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Sample records for safety analysis addendum

  1. Addendum to the Safety Analysis Report for the Steel Waste Packaging. Revision 1

    International Nuclear Information System (INIS)

    Crow, S.R.

    1996-01-01

    The Battelle Pacific Northwest National Laboratory Safety Analysis Report (SAR) for the Steel Waste Package requires additional analyses to support the shipment of remote-handled radioactive waste and special-case waste from the 324 building hot cells to PUREX for interim storage. This addendum provides the analyses required to show that this waste can be safely shipped onsite in the configuration shown

  2. PAT-1 safety analysis report addendum.

    Energy Technology Data Exchange (ETDEWEB)

    Weiner, Ruth F.; Schmale, David T.; Kalan, Robert J.; Akin, Lili A.; Miller, David Russell; Knorovsky, Gerald Albert; Yoshimura, Richard Hiroyuki; Lopez, Carlos; Harding, David Cameron; Jones, Perry L.; Morrow, Charles W.

    2010-09-01

    The Plutonium Air Transportable Package, Model PAT-1, is certified under Title 10, Code of Federal Regulations Part 71 by the U.S. Nuclear Regulatory Commission (NRC) per Certificate of Compliance (CoC) USA/0361B(U)F-96 (currently Revision 9). The purpose of this SAR Addendum is to incorporate plutonium (Pu) metal as a new payload for the PAT-1 package. The Pu metal is packed in an inner container (designated the T-Ampoule) that replaces the PC-1 inner container. The documentation and results from analysis contained in this addendum demonstrate that the replacement of the PC-1 and associated packaging material with the T-Ampoule and associated packaging with the addition of the plutonium metal content are not significant with respect to the design, operating characteristics, or safe performance of the containment system and prevention of criticality when the package is subjected to the tests specified in 10 CFR 71.71, 71.73 and 71.74.

  3. Addendum to Composite Analysis for Low-Level Waste Disposal in the 200 Area Plateau of the Hanford Site

    Energy Technology Data Exchange (ETDEWEB)

    Bergeron, Marcel P.; Freeman, Eugene J.; Wurstner, Signe K.; Kincaid, Charles T.; Coony, Mike M.; Strenge, Dennis L.; Aaberg, Rosanne L.; Eslinger, Paul W.

    2001-09-28

    This report summarizes efforts to complete an addendum analysis to the first iteration of the Composite Analysis for Low-Level Waste Disposal in the 200 Area Plateau of the Hanford Site (Composite Analysis). This document describes the background and performance objectives of the Composite Analysis and this addendum analysis. The methods used, results, and conclusions for this Addendum analysis are summarized, and recommendations are made for work to be undertaken in anticipation of a second analysis.

  4. Addendum to the 2015 Eastern Interconnect Baselining and Analysis Report

    Energy Technology Data Exchange (ETDEWEB)

    Amidan, Brett G. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Follum, James D. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2016-06-30

    This report serves as an addendum to the report 2015 Eastern Interconnect Baselining and Analysis Report (Amidan, Follum, and Freeman, 2015). This addendum report investigates the following: the impact of shorter record lengths and of adding a daily regularization term to the date/time models for angle pair measurements, additional development of a method to monitor the trend in phase angle pairs, the effect of changing the length of time to determine a baseline, when calculating atypical events, and a comparison between quantitatively discovered atypical events and actual events.

  5. Addendum to the health and safety plan for Waste Area Grouping 6 at Oak Ridge National Laboratory, Oak Ridge, Tennessee. Environmental Restoration Program

    International Nuclear Information System (INIS)

    Clark, C. Jr.; Burman, S.N.; Wilson, K.A.

    1995-08-01

    There are three purposes for this addendum to the health and safety plan for Waste Area Grouping 6. The first purpose is to provide record of a corrective action response concerning an occurrence on WAG 6 in October 1994 (ORO-MMES-ENVRES-1994-0016.) This occurrence involved a precautionary evacuation of subcontractor field crews due to malfunctioning monitor alarms for organic vapors. The corrective action is to revise the WAG 6 Site health and safety plan to improve communications during emergency events. The second purpose is to incorporate any outstanding health and safety issues not addressed in the original health and safety plan for WAG 6 document (ORNL/ER-183). The only variance of note is tritium air monitoring in the Tumulus building. The tritium air monitor is added in this addendum as monitoring equipment for WAG 6 with description of action level and calibration. The third purpose of this addendum is to satisfy a condition of approval for the pending Nuclear Criticality Safety Assessment (NCSA) pertaining to KEMA fuel storage at WAG 6. This approval condition requires the following: ''The location of the KEMA burial shall be recorded and maintained in a controlled document that identifies the quantity and the general physical conditions at the time of the entombment with an admonishment to obtain nuclear criticality safety guidance before altering the burial condition.'' In order to satisfy the approval, this document must be controlled. The predecessor to the pending NCSA is NSR No. 0002WM22001

  6. Safe Handling of Radioisotopes. Medical Addendum

    International Nuclear Information System (INIS)

    Hercik, F.; Jammet, H.

    1960-01-01

    The International Atomic Energy Agency published in 1958 a Manual entitled ''Safe Handling of Radioisotopes'' (Safety Series No. 1 - STI/PUB/1), based on the work of an international panel convened by the Agency. As recommended by that panel and approved by the Agency's Board of Governors, this Addendum has now been prepared, primarily as a supplement to the Manual. It contains information necessary to medical officers concerned with the implementation of the controls given in the Manual. In addition, it is intended to serve as a brief introduction to the medical problems encountered in radiological protection work and to the methods of resolving them. As in the case of the Manual itself, the information given in this Addendum is particularly relevant to the problems encountered by the small user of radioisotopes. Although the basic principles set forth in it apply to all work with radiation sources, the Addendum is not intended to serve as a radiological protection manual for use in reactor installations or large-scale nuclear industry, where more specialized techniques and information are required.

  7. Safe Handling of Radioisotopes. Health Physics Addendum

    Energy Technology Data Exchange (ETDEWEB)

    Appleton, G J; Krishnamoorthy, P N

    1960-07-15

    The International Atomic Energy Agency published in 1958 a Manual entitled ''Safe Handling of Radioisotopes'' (Safety Series No. 1 - STI/PUB/1), based on the work of an international panel convened by the Agency. As recommended by that panel and approved by the Agency's Board of Governors, this Addendum has now been prepared, primarily as a supplement to the Manual. It contains technical information necessary for the implementation of the controls given in the Manual. In addition, it is intended to serve as a brief introduction to the technical problems encountered in radiological protection work and to the methods of resolving them. As in the case of the Manual itself, the information given in this Addendum is particularly relevant to the problems encountered by the small user of radioisotopes. Although the basic principles set forth in it apply to all work with radiation sources, the Addendum is not intended to serve as a radiological protection manual for use in reactor installations or large-scale nuclear industry, where more specialized techniques and information are required.

  8. Safe Handling of Radioisotopes. Health Physics Addendum

    International Nuclear Information System (INIS)

    Appleton, G.J.; Krishnamoorthy, P.N.

    1960-01-01

    The International Atomic Energy Agency published in 1958 a Manual entitled ''Safe Handling of Radioisotopes'' (Safety Series No. 1 - STI/PUB/1), based on the work of an international panel convened by the Agency. As recommended by that panel and approved by the Agency's Board of Governors, this Addendum has now been prepared, primarily as a supplement to the Manual. It contains technical information necessary for the implementation of the controls given in the Manual. In addition, it is intended to serve as a brief introduction to the technical problems encountered in radiological protection work and to the methods of resolving them. As in the case of the Manual itself, the information given in this Addendum is particularly relevant to the problems encountered by the small user of radioisotopes. Although the basic principles set forth in it apply to all work with radiation sources, the Addendum is not intended to serve as a radiological protection manual for use in reactor installations or large-scale nuclear industry, where more specialized techniques and information are required.

  9. Addendum to the Composite Analysis for the E-Area Vaults and Saltstone Disposal Facilities

    International Nuclear Information System (INIS)

    Cook, J.R.

    2002-01-01

    Revision 1 of the Composite Analysis (CA) Addendum has been prepared to respond to the U.S. Department of Energy (DOE) Low-Level Waste Disposal Facilities Federal Review Group review of the CA. This addendum to the composite analysis responds to the conditions of approval. The composite analysis was performed on the two active SRS low-level radioactive waste disposal facilities. The facilities are the Z-Area Saltstone Disposal Facility and the E-Area Vaults Disposal Facility. The analysis calculated potential releases to the environment from all sources of residual radioactive material expected to remain in the General Separations Area (GSA). The GSA is the central part of the Savannah River Site and contains all of the waste disposal facilities, the chemical separation facilities and associated high-level waste storage facilities, as well as numerous other sources of radioactive material

  10. The storage of liquid high level waste at BNFL, Sellafield. Addendum to February 2000 report

    International Nuclear Information System (INIS)

    2001-08-01

    On 18 February 2000 the Health and Safety Executive (HSE) published a report on the work of its Nuclear Installations Inspectorate (NIl) in regulating the storage of liquid high level waste at the BNFL Sellafield site. Within the report NIl gave two undertakings. One was to publish an addendum around 1 year later covering its assessment of the new safety case for the storage plant and the second was to publish a further addendum when progress had been made with options studies for reducing the stocks of liquid high level waste (HLW), also referred to as highly active liquor (HAL), to a buffer level. A progress report was published in February 2001 which included a summary of the assessment of the new safety case and NIl's regulatory action to enforce liquid HLW stock reductions. This addendum provides a more detailed update on the position reached based on consideration of BNFL's responses to the recommendations from the February 2000 HLW report since its publication. It embodies the two addenda referred to above integrated into a single document for publication

  11. Fuel Receiving and Storage Station. License application, amendment 5, addendum

    International Nuclear Information System (INIS)

    1975-06-01

    This Addendum to AG-L105 addresses the utilization of the Service Concentrator for evaporation of low level wastes generated during fuel receiving and storage operations. The Service Concentrator is described from various viewpoints and necessary relevant data are included for adequate assessment of safety. (U.S.)

  12. Addendum to the Phase 2 Sampling and Analysis Plan for the Clinch River Remedial Investigation

    International Nuclear Information System (INIS)

    1994-03-01

    This document is an addendum to the Phase 2 Sampling and Analysis Plan for the Clinch River Remedial Investigation (DOE 1993). The Department of Energy--Oak Ridge Operations (DOE-ORO) is proposing this addendum to the US Envianmental Protection Agency, Region IV (EPA-IV), and the Tennessee Department of Environment and Conservation (TDEC) as a reduced sampling program on the Clinch River arm of Watts Bar Reservoir and on Poplar Creek. DOE-ORO is proposing to maximize the use of existing data and minimize the collection of new data for water, sediment, and biota during Phase 2 of the Clinch River Remedial Investigation. The existing data along with the additional data collected in Phase 2 would be used to perform a baseline risk assessment and make remedial decisions. DOE-ORO considers that the existing data, the additional data collected in Phase 2, and on-site remedial investigation data would be sufficient to understand the nature and extent of the contamination problem in the Clinch River, perform a baseline risk assessment,and make remedial decisions. This addendum is organized in three sections. The first section provides background information and describes a rationale for modifying the Phase 2 Sampling and Analysis Plan. Section 2 presents a summary of the existing data for the Clinch River arm of Watts Bar Reservoir and an evaluation of the sufficiency of this data for a baseline human health and ecological risk assessment. Section 3 describes the revised Phase 2 Sampling and Analysis Plan for surface water, sediment, and biota in the Clinch River OU and in the Poplar Creek OU

  13. CSER 90-006, addendum 1: Criticality safety control for source term reduction project in the scrubber glovebox of Building 232-Z. Revision 1

    International Nuclear Information System (INIS)

    Hess, A.L.

    1995-01-01

    This Criticality Safety Evaluation Report addendum extends the coverage of the original CSER (90-006) about dismantling the ductwork in 232-Z to include cleanout of the Scrubber Glovebox, with an estimated residual Pu holdup of less than 200 grams. For conservatism and containment considerations, the provisions about waste packaging and water exclusion from the original work are retained, even though it is not credible for the Scrubber Pu content to be made critical with water added (NDA gives about 1/3 a minimum critical mass)

  14. Addendum 3 to CSAR 80-027, Use of calorimeter 109B for fissile material measurement

    International Nuclear Information System (INIS)

    Chiao, T.

    1994-01-01

    This modification to the Plutonium Finishing Plant (PFP) calorimeter system involves removing current calorimeter No. 3 from the water bath and replacing it with a calorimeter that can accommodate larger diameter items (an oversize can). The inside diameters of both the sample and the reference cells will be increased to 5.835 inches at the top opening and to 5.22 inches at the bottom, the 8 inch high measurement zone. This Addendum 3 to Criticality Safety Analysis Report 80-027 examines criticality safety during the use of the modified calorimeter (Calorimeter 109B) with enlarged cell tube diameters to assure that an adequate margin of subcriticality is maintained for all normal and contingency conditions

  15. 48 CFR 1352.239-70 - Software license addendum.

    Science.gov (United States)

    2010-10-01

    ... 48 Federal Acquisition Regulations System 5 2010-10-01 2010-10-01 false Software license addendum... Software license addendum. As prescribed in 48 CFR 1339.107, insert the following clause: Software License....S.C. Section 516). (h) Renewal of Support after Expiration of this Award: Service will not...

  16. Technical Review Report for the Model 9977 Safety Analysis Report for Packaging Addendum 1 Justification for DNDO Contents

    Energy Technology Data Exchange (ETDEWEB)

    West, M H

    2008-12-17

    The Model 9977 Package is currently certified for Content Envelope C.1, {sup 238}Pu Heat Sources, either in Radioisotope Thermoelectric Generator (RTG), or in Food-Pack Can configurations, under Certificate of Compliance (CoC) Certificate Number 9977 and Package Identification Number USA/9977/B(M)F-96 (DOE). Addendum 1, Justification for DNDO Contents,--the Submittal--supplements Revision 2 of the Safety Analysis Report for Packaging for the Model 9977 Package. The Submittal adds five new contents to the Model 9977 Package, Content Envelopes, AC.1 through AC.5. The Content Envelopes are neptunium metal, the beryllium-reflected plutonium ball (BeRP Ball), plutonium/uranium metal, plutonium/uranium metal with enhanced wt% {sup 240}Pu (to 50 wt%), and uranium metal. The last three Content Envelopes are stabilized to DOE-STD-3013. These Content Envelopes will be shipped to the Device Assembly Facility (DAF) at the Nevada Test Site (NTS), where they will reside, and, hence, to off-site locations in support of the Department of Homeland Security (DHS) Domestic Nuclear Detection Office (DNDO). The new certificate will apply to a limited number of Model 9977 Packages. At the same time, the Submittal requests an extension of the periodic maintenance requirements from one (1) year to up to five (5) years using Radio-Frequency Identification (RFID) temperature-monitoring systems to measure the ambient storage temperature in order to ensure that the temperature of the Viton{reg_sign} O-rings for the 6-inch Containment Vessel (6CV) remain less than 200 F. The RFIDs have been developed by Argonne National Laboratory. An on-going surveillance program at the K-Area Materials Storage (KAMS) facility at the Savannah River Site, and an on-going examination of Viton{reg_sign} O-rings from mock Primary Containment Vessels (PCVs) at Savannah River National Laboratory (SRNL) provide the technical justification for the extension of the periodic maintenance interval. Where extended

  17. Integrated thermal treatment system study. Phase 2. Addendum system A8

    International Nuclear Information System (INIS)

    Biagi, C.; Teheranian, B.; Quapp, W.J.; Schwinkendorf, W.E.

    1996-05-01

    This is an addendum to the Integrated Treatment System Study - Phase 2 Results report. This addendum describes the technology and the operation of System A-8, Rotary Kiln, Air Combustion Gas, Dry-Wet APC, and Grout Stabilization. A process flow diagram, functional allocation diagrams, and plan views and perspective views for this system are attached. Detailed cost information for this subsystem is reported in Appendix A of this addendum

  18. Upgraded safety analysis document including operations policies, operational safety limits and policy changes. Revision 2

    International Nuclear Information System (INIS)

    Batchelor, K.

    1996-03-01

    The National Synchrotron Light Source Safety Analysis Reports (1), (2), (3), BNL reports number-sign 51584, number-sign 52205 and number-sign 52205 (addendum) describe the basic Environmental Safety and Health issues associated with the department's operations. They include the operating envelope for the Storage Rings and also the rest of the facility. These documents contain the operational limits as perceived prior or during construction of the facility, much of which still are appropriate for current operations. However, as the machine has matured, the experimental program has grown in size, requiring more supervision in that area. Also, machine studies have either verified or modified knowledge of beam loss modes and/or radiation loss patterns around the facility. This document is written to allow for these changes in procedure or standards resulting from their current mode of operation and shall be used in conjunction with the above reports. These changes have been reviewed by NSLS and BNL ES and H committee and approved by BNL management

  19. Grøstl Addendum

    DEFF Research Database (Denmark)

    Gauravaram, Praveen; Knudsen, Lars R.; Matusiewicz, Krystian

    2009-01-01

    This document is an addendum to the submission document of Grøstl, which was selected for the second round of NIST’s SHA-3 competition [18]. We stress that we do not change the specification of Grøstl. In other words, Grøstl is defined exactly as specified in the original submission document [8]....

  20. Health effects models for nuclear power plant accident consequence analysis. Modification of models resulting from addition of effects of exposure to alpha-emitting radionuclides: Revision 1, Part 2, Scientific bases for health effects models, Addendum 2

    Energy Technology Data Exchange (ETDEWEB)

    Abrahamson, S. [Wisconsin Univ., Madison, WI (United States); Bender, M.A. [Brookhaven National Lab., Upton, NY (United States); Boecker, B.B.; Scott, B.R. [Lovelace Biomedical and Environmental Research Inst., Albuquerque, NM (United States). Inhalation Toxicology Research Inst.; Gilbert, E.S. [Pacific Northwest Lab., Richland, WA (United States)

    1993-05-01

    The Nuclear Regulatory Commission (NRC) has sponsored several studies to identify and quantify, through the use of models, the potential health effects of accidental releases of radionuclides from nuclear power plants. The Reactor Safety Study provided the basis for most of the earlier estimates related to these health effects. Subsequent efforts by NRC-supported groups resulted in improved health effects models that were published in the report entitled {open_quotes}Health Effects Models for Nuclear Power Plant Consequence Analysis{close_quotes}, NUREG/CR-4214, 1985 and revised further in the 1989 report NUREG/CR-4214, Rev. 1, Part 2. The health effects models presented in the 1989 NUREG/CR-4214 report were developed for exposure to low-linear energy transfer (LET) (beta and gamma) radiation based on the best scientific information available at that time. Since the 1989 report was published, two addenda to that report have been prepared to (1) incorporate other scientific information related to low-LET health effects models and (2) extend the models to consider the possible health consequences of the addition of alpha-emitting radionuclides to the exposure source term. The first addendum report, entitled {open_quotes}Health Effects Models for Nuclear Power Plant Accident Consequence Analysis, Modifications of Models Resulting from Recent Reports on Health Effects of Ionizing Radiation, Low LET Radiation, Part 2: Scientific Bases for Health Effects Models,{close_quotes} was published in 1991 as NUREG/CR-4214, Rev. 1, Part 2, Addendum 1. This second addendum addresses the possibility that some fraction of the accident source term from an operating nuclear power plant comprises alpha-emitting radionuclides. Consideration of chronic high-LET exposure from alpha radiation as well as acute and chronic exposure to low-LET beta and gamma radiations is a reasonable extension of the health effects model.

  1. Addendum to the 1996 Gunnison Monitoring Report for the Gunnison, Colorado Wetlands Mitigation Plan

    International Nuclear Information System (INIS)

    1997-10-01

    This document is an addendum to the 1996 Gunnison Monitoring Report for the Gunnison, Colorado, Wetlands Mitigation Report, dated July 1997. The purpose of this addendum is to: (1) modify how information on plant height and plant species criteria are presented; and (2) provide more detailed information regarding the evaluation of the bare ground criteria at the Camp Ketle site. The information in this addendum is provided at the request of the Bureau of Land Management to aid in future monitoring and evaluation of the wetland mitigation sites

  2. Trainees May Add Value to Patient Care by Decreasing Addendum Utilization in Radiology Reports.

    Science.gov (United States)

    Balthazar, Patricia; Konstantopoulos, Christina; Wick, Carson A; DeSimone, Ariadne K; Tridandapani, Srini; Simoneaux, Stephen; Applegate, Kimberly E

    2017-11-01

    The purpose of this study was to evaluate the impact of trainee involvement and other factors on addendum rates in radiology reports. This retrospective study was performed in a tertiary care pediatric hospital. From the institutional radiology data repository, we extracted all radiology reports from January 1 to June 30, 2016, as well as trainee (resident or fellow) involvement, imaging modality, patient setting (emergency, inpatient, or outpatient), order status (routine vs immediate), time of interpretation (regular work hours vs off-hours), radiologist's years of experience, and sex. We grouped imaging modalities as advanced (CT, MRI, and PET) or nonadvanced (any modality that was not CT, MRI, or PET) and radiologist experience level as ≤ 20 years or > 20 years. Our outcome measure was the rate of addenda in radiology reports. Statistical analysis was performed using multivariate logistic regression. From 129,033 reports finalized during the study period, 418 (0.3%) had addenda. Reports generated without trainees were 12 times more likely than reports with trainee involvement to have addenda (odds ratio [OR] = 12.2, p use (OR = 4.7, p use than those in an inpatient setting (OR = 1.5, p = 0.04; and OR = 1.3, p = 0.04, respectively). Routine orders had a slightly higher likelihood of addendum use compared with immediate orders (OR = 1.3, p = 0.01). We found no difference in addendum use by radiologist's sex, radiologist's years of experience, emergency versus outpatient setting, or time of interpretation. Trainees may add value to patient care by decreasing addendum rates in radiology reports.

  3. Health and safety plan for operations performed for the Environmental Restoration Program

    Energy Technology Data Exchange (ETDEWEB)

    Trippet, W.A. II (IT Corp., (United States)); Reneau, M.; Morton, S.L. (EG and G Idaho, Inc., Idaho Falls, ID (United States))

    1992-04-01

    This document constitutes the generic health and safety plan for the Environmental Restoration Program (ERP). It addresses the health and safety requirements of the Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA); Occupational Safety and Health Administration (OSHA) 29 CFR 1910.120 standard; and EG G Idaho, Inc. This plan is a guide to individuals who must complete a health and safety plan for a task performed for the EPR. It contains a task specific addendum that, when completed, specifically addresses task specific health and safety issues. This health and safety plan reduces the time it takes to write a task specific health and safety plan by providing discussions of requirements, guidance on where specific information is located, and specific topics in the Addendum that must be discussed at a task level. This format encourages a complete task specific health and safety plan and a standard for all health and safety plans written for ERP.

  4. Health and safety plan for operations performed for the Environmental Restoration Program

    International Nuclear Information System (INIS)

    Trippet, W.A. II; Reneau, M.; Morton, S.L.

    1992-04-01

    This document constitutes the generic health and safety plan for the Environmental Restoration Program (ERP). It addresses the health and safety requirements of the Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA); Occupational Safety and Health Administration (OSHA) 29 CFR 1910.120 standard; and EG ampersand G Idaho, Inc. This plan is a guide to individuals who must complete a health and safety plan for a task performed for the EPR. It contains a task specific addendum that, when completed, specifically addresses task specific health and safety issues. This health and safety plan reduces the time it takes to write a task specific health and safety plan by providing discussions of requirements, guidance on where specific information is located, and specific topics in the Addendum that must be discussed at a task level. This format encourages a complete task specific health and safety plan and a standard for all health and safety plans written for ERP

  5. Reference Mission Version 3.0 Addendum to the Human Exploration of Mars: The Reference Mission of the NASA Mars Exploration Study Team. Addendum; 3.0

    Science.gov (United States)

    Drake, Bret G. (Editor)

    1998-01-01

    This Addendum to the Mars Reference Mission was developed as a companion document to the NASA Special Publication 6107, "Human Exploration of Mars: The Reference Mission of the NASA Mars Exploration Study Team." It summarizes changes and updates to the Mars Reference Missions that were developed by the Exploration Office since the final draft of SP 6107 was printed in early 1999. The Reference Mission is a tool used by the exploration community to compare and evaluate approaches to mission and system concepts that could be used for human missions to Mars. It is intended to identify and clarify system drivers, significant sources of cost, performance, risk, and schedule variation. Several alternative scenarios, employing different technical approaches to solving mission and technology challenges, are discussed in this Addendum. Comparing alternative approaches provides the basis for continual improvement to technology investment plan and a general understanding of future human missions to Mars. The Addendum represents a snapshot of work in progress in support of planning for future human exploration missions through May 1998.

  6. Addendum to Environmental Monitoring Plan, Nevada Test Site and Support Facilities; Addendum 2

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1993-11-01

    This 1993 Addendum to the ``Environmental Monitoring Plan Nevada Test Site and Support Facilities -- 1991,`` Report No. DOE/NV/10630-28 (EMP) applies to the US Department of Energy`s (DOE`s) operations on the Continental US (including Amchitka Island, Alaska) that are under the purview of the DOE Nevada Operations Office (DOE/NV). The primary purpose of these operations is the conduct of the nuclear weapons testing program for the DOE and the Department of Defense. Since 1951, these tests have been conducted principally at the Nevada Test Site (NTS), which is located approximately 100 miles northwest of Las Vegas, Nevada. In accordance with DOE Order 5400.1, this 1993 Addendum to the EMP brings together, in one document, updated information and/or new sections to the description of the environmental activities conducted at the NTS by user organizations, operations support contractors, and the US Environmental Protection Agency (EPA) originally published in the EMP. The EPA conducts both the offsite environmental monitoring program around the NTS and post-operational monitoring efforts at non-NTS test locations used between 1961 and 1973 in other parts of the continental US. All of these monitoring activities are conducted under the auspices of the DOE/NV, which has the stated policy of conducting its operations in compliance with both the letter and the spirit of applicable environmental statutes, regulations, and standards.

  7. Engineering task plan for the annual revision of the rotary mode core sampling system safety equipment list

    International Nuclear Information System (INIS)

    BOGER, R.M.

    1999-01-01

    This Engineering Task Plan addresses an effort to provide an update to the RMCS Systems 3 and 4 SEL and DCM in order to incorporate the changes to the authorization basis implemented by HNF-SD-WM-BIO-001, Rev. 0 (Draft), Addendum 5 , Safety Analysis for Rotary Mode Core Sampling. Responsibilities, task description, cost estimate, and schedule are presented

  8. Advancing Geographic Information Science—Proposal to NSF, with Addendum

    OpenAIRE

    National Center for Geographic Information and Analysis

    1995-01-01

    This document from 1995 and the addendum from 1996 represent NCGIA's proposal for Varenius—a program of research initiatives to advance geographic information science in the context of digitial worlds and the information society.

  9. 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

  10. The management of radioactive wastes produced by radioisotope users. Technical addendum

    International Nuclear Information System (INIS)

    1966-01-01

    This Addendum contains detailed technical information on processes and procedures that are outlined in more general terms in the Code of Practice. The information is particularly relevant to the problem of handling the relatively small quantities of waste arising from the use of radionuclides in laboratories, hospitals and industry when no special facilities for radioactive waste treatment are available on site. The Addendum is directed toward providing the radioisotope user with the type and amount of information required for him to be able to assess the alternatives available to him in terms of his particular needs, develop a preliminary design of an optimum waste-handling system and get help and guidance in his search for more detailed information.

  11. 78 FR 57873 - 60-Day Notice of Proposed Information Collection: HUD-Owned Real Estate-Sales Contract and Addendums

    Science.gov (United States)

    2013-09-20

    ... DEPARTMENT OF HOUSING AND URBAN DEVELOPMENT [Docket No. FR-5687-N-36] 60-Day Notice of Proposed Information Collection: HUD-Owned Real Estate--Sales Contract and Addendums AGENCY: Office of the Assistant... Information Collection Title of Information Collection: HUD-Owned Real Estate--Sales Contract and Addendums...

  12. Addendum to ''Test Plan: Small-Scale Seal Performance Tests (SSSPT)''

    International Nuclear Information System (INIS)

    Finley, R.E.

    1992-01-01

    This document describes activities that are intended to update the data base of fluid flow measurements made on expansive salt concrete (ESC) seals as part of the Small-Scale Seal Performance Tests (SSSPT). The original plans for the SSSPT experiments are described by Stormont (1985a and 1985b). These seals have previously been tested with brine and gas during the early stages (less than 450 days) after emplacement. The purpose of this Test Plan Addendum is to detail the activities necessary to, update the gas and brine measurements previously performed on the SSSPT Series A and Series B seals and to identify the key personnel responsible for implementing these activities. This addendum describes a limited undertaking and does NOT change the scope of the original test plan

  13. Alcator C-MOD proposal addendum

    International Nuclear Information System (INIS)

    Bonoli, P.; Greenwald, M.; Gwinn, D.

    1986-04-01

    Since the design concept and overall purpose of the Alcator C-MOD device are similar to that proposed in October 1985, we have chosen in this document only to highlight areas where changes or additions have been made. Chapters in the Addendum correspond to those in the Proposal, except Chapter 9 which describes a number of toroidal improvement concepts which are being considered for inclusion in the Alcator C-MOD experimental program. A description of the redesign and a discussion of the objectives of the experimental program are given

  14. 78 FR 39763 - Recovery Plan Addendum; Thick-Billed Parrot

    Science.gov (United States)

    2013-07-02

    ... particular species. Species' History Historically the thick-billed parrot's range extended from Mexico into... and the Secretary of the Interior. This species is currently found in Mexico but has not been detected... included southern Arizona and possibly southwestern New Mexico. The recovery plan addendum includes...

  15. Addendum: First injection technique recommendations for patients with diabetes, Forum for Injection Techniques India

    Directory of Open Access Journals (Sweden)

    Sanjay Kalra

    2013-01-01

    Full Text Available The forum for injection techniques, India recommendation, the first ever in the country on insulin injcetion techniques, have covered the science and the art of insulin injection technique in an exhaustive manner. However, a few gaps were identified in the document, which are addressed in the current addendum. This article focuses on insulin injection technique in special clinical situations, including geriatric people, women in pregnancy and those with dermatological or surgical disease who live with diabetes. The addendum also covers salient features of administration of insulin using the insulin pump.

  16. Final environmental statement. Final addendum to Part II: Manufacture of floating nuclear power plants by Offshore Power Systems. DOCKET-STN--50-437

    International Nuclear Information System (INIS)

    1978-06-01

    This Addendum to Part II of the Final Environmental Statement related to manufacture of floating nuclear power plants by Offshore Power Systems (OPS), NUREG-0056, issued September 1976, was prepared by the U.S. Nuclear Regulatory Commission (NRC), Office of Nuclear Reactor Regulation. The staff's basic evaluation is presented in NUREG-0056. The current Addendum provides further consideration of a number of topics discussed in NUREG-0056, particularly additional consideration of shore zone siting at estuarine and ocean regions. This Summary and Conclusions recapitulates and is cumulative for Part II of the FES and the current Addendum. Augmentations to the Summary and Conclusions presented in Part II of the FES and arising from the evaluations contained in this Addendum are italicized

  17. Health and safety plan for operations performed for the Environmental Restoration Program. Task, OU 1-03 and OU 4-10 Track 2 investigations

    Energy Technology Data Exchange (ETDEWEB)

    Trippet, W.A. II [IT Corp., (United States); Reneau, M.; Morton, S.L. [EG and G Idaho, Inc., Idaho Falls, ID (United States)

    1992-04-01

    This document constitutes the generic health and safety plan for the Environmental Restoration Program (ERP). It addresses the health and safety requirements of the Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA); Occupational Safety and Health Administration (OSHA) 29 CFR 1910.120 standard; and EG&G Idaho, Inc. This plan is a guide to individuals who must complete a health and safety plan for a task performed for the EPR. It contains a task specific addendum that, when completed, specifically addresses task specific health and safety issues. This health and safety plan reduces the time it takes to write a task specific health and safety plan by providing discussions of requirements, guidance on where specific information is located, and specific topics in the Addendum that must be discussed at a task level. This format encourages a complete task specific health and safety plan and a standard for all health and safety plans written for ERP.

  18. Systems engineering functions and requirements for the Hanford cleanup mission. First issue, Addendum 2

    Energy Technology Data Exchange (ETDEWEB)

    Holmes, J.J.

    1994-01-01

    This addendum provides the technical detail of a systems engineering functional analysis for the Hanford cleanup mission. Details of the mission analysis including mission statement, scope, problem statement, initial state definition, and final state definition are provided in the parent document. The functional analysis consists of Input Computer Automated Manufacturing Definition (IDEFO) diagrams an definitions, which will be understood by systems engineers, but which may be difficult for others to comprehend. For a more complete explanation of this work, refer to the parent document. The analysis covers the total Hanford cleanup mission including the decomposition levels at which the various Hanford programs or integrated activities are encountered.

  19. HFBR restart activity A2.6: Review of FSAR and 60 MW addendum to assure consistency of operation at 40 MW

    International Nuclear Information System (INIS)

    Rao, D.V.; Ross, S.B.; Darby, J.L.; Clark, R.A.

    1990-01-01

    The purpose of this task (HFBR Restart Activity A2.6) is to perform a review of the design basis accident (DBA) analyses sections of the 1964 HFBR-Final Safety Analysis Report; Volumes I and II, and the 1982 Addendum to the HFBR-FSAR for 60 MW operation to assure that operation at 40 MW will be consistent with these analyses. Additional documents utilized in the review included the Level 1 PRA for HFBR, HFBR-PDMs and HFBR-OPMs. The review indicates that the 1964 FSAR-DBA analysis in incomplete in the sense that it did not analyze some of the important initiators for 1-loop operation that include: Accidental throttling of primary flow control valves; seizure of primary pump; loss of secondary pump; accidental throttling of secondary flow control valves; rupture of secondary piping. The first three initiators were later studied in the 1982 addendum. The other two initiators have not been examined to-date for 1-loop operation. It is recommended that the impact of these initiators be assessed prior to the restart, if 1-loop operation is chosen for the restart. The review demonstrated that at 40 MW operation there are only a few accident initiators that will culminate in core damage (fuel melting and /or cladding failure) regardless of the availability of mitigating systems. For 1-loop Operation these accidents include: Fuel channel blockage, primary pump seizure, and large-large LOCA (a LOCA with effective break diameter > 2.81 in. is referred to as a large-large LOCA in this document as well as in PRA). Although all these accidents listed above could lead to core damage for 1-loop operation as well, the probability is expected be very low

  20. Operation and maintenance manual addendum to Westinghouse OMM-051-00-005 for intermediate-size inducer pump (ISIP) Model 266

    International Nuclear Information System (INIS)

    1979-01-01

    This Addendum A to the Westinghouse Operation and Maintenance Manual OMM-051-00-005 contains additions, changes and deletions that modify data for the Liquid Metal Coolant Pump, Model LMP-1 to data applicable to the Rockwell International Intermediate-Size Inducer Pump (ISIP), Model 266. The major modifications on the ISIP are the new impeller/inducer assembly, diffuser vanes and new securing hardware. The paragraphs affected by Addendum A use the numbering system used in OMM-051-00-005 except each paragraph number is prefixed with an A preceding the paragraph number signifying that the paragraph is changed to conform with the ISIP configuration. Paragraphs within OMM-051-00-005 not identified in Addendum A remain unchanged with the data still valid and useable

  1. 2011 Addendum to the SNL/NM SWEIS Supplemental Information Source Documents

    Energy Technology Data Exchange (ETDEWEB)

    Dimmick, Ross [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2014-12-01

    This document contains updates to the Supplemental Information Sandia National Laboratories/New Mexico Site-Wide Environmental Impact Statement Source Documents that were developed in 2010. In general, this addendum provides calendar year 2010 data, along with changes or additions to text in the original documents.

  2. 77 FR 7131 - Addendum to Environmental Technologies Trade Advisory Committee Public Meeting

    Science.gov (United States)

    2012-02-10

    ... DEPARTMENT OF COMMERCE International Trade Administration Addendum to Environmental Technologies... agenda of a meeting of the Environmental Technologies Trade Advisory Committee (ETTAC) will be changed to... & Environmental Industries, International Trade Administration, Room 4053, 1401 Constitution Avenue NW...

  3. Addendum Guidelines for the Prevention of Peanut Allergy in the United States

    Science.gov (United States)

    Togias, Alkis; Cooper, Susan F.; Acebal, Maria L.; Assa’ad, Amal; Baker, James R.; Beck, Lisa A.; Block, Julie; Byrd-Bredbenner, Carol; Chan, Edmond S.; Eichenfield, Lawrence F.; Fleischer, David M.; Fuchs, George J.; Furuta, Glenn T.; Greenhawt, Matthew J.; Gupta, Ruchi S.; Habich, Michele; Jones, Stacie M.; Keaton, Kari; Muraro, Antonella; Plaut, Marshall; Rosenwasser, Lanny J.; Rotrosen, Daniel; Sampson, Hugh A.; Schneider, Lynda C.; Sicherer, Scott H.; Sidbury, Robert; Spergel, Jonathan; Stukus, David R.; Venter, Carina; Boyce, Joshua A.

    2016-01-01

    Background Food allergy is an important public health problem because it affects children and adults, it may be severe and even life-threatening, and it may be increasing in prevalence. Beginning in 2008, the National Institute of Allergy and Infectious Diseases (NIAID), working with other organizations and advocacy groups, led the development of the first clinical guidelines for the diagnosis and management of food allergy. A recent landmark clinical trial and other emerging data suggest that peanut allergy can be prevented through introduction of peanut-containing foods beginning in infancy. Objectives Prompted by these findings, along with 25 professional organizations, federal agencies, and patient advocacy groups, NIAID facilitated development of Addendum Guidelines to specifically address the prevention of peanut allergy. Results The Addendum provides three separate guidelines for infants at various risk levels for development of peanut allergy and is intended for use by a wide variety of health care providers. Topics addressed include the definition of risk categories, appropriate use of testing (specific IgE, skin prick testing, and oral food challenge) and the timing and approaches for introduction of peanut-containing foods in the health care provider’s office or at home. The Addendum Guidelines provide the background, rationale, and strength of evidence for each recommendation. Conclusions Guidelines have been developed for early introduction of peanut-containing foods into the diets of infants at various risk levels for peanut allergy. PMID:28065278

  4. Health and Safety Plan for Operations Performed for the Environmental Restoration Program: Task, Characterization of Potential Waste Sources at Auxiliary Reactor Area-1 Operable Unit 5--07 site ARA-02

    International Nuclear Information System (INIS)

    Pickett, S.L.; Morton, S.L.

    1992-06-01

    This document constitutes the generic health and safety plan for the Environmental Restoration Program (ERP). It addresses the health and safety requirements of the Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA); Occupational Safety and Health Administration (OSHA) 29 CFR 1910.120 standard; and EG ampersand G Idaho, Inc. This plan is a guide to individuals who must complete a health and safety plan for a task performed for the ERP. It contains a task specific addendum that, when completed, specifically addresses task specific health and safety issues. This health and safety plan reduces the time it takes to write a task specific health and safety plan by providing discussions of requirements, guidance on where specific information is located, and specific topics in the Addendum that must be discussed at a task level. This format encourages a complete task specific health and safety plan and a standard for all health and safety plans written for ERP

  5. Health and Safety Plan for Operations Performed for the Environmental Restoration Program: Task, Characterization of Potential Waste Sources at Auxiliary Reactor Area-1 Operable Unit 5--07 site ARA-02

    Energy Technology Data Exchange (ETDEWEB)

    Pickett, S.L.; Morton, S.L.

    1992-06-01

    This document constitutes the generic health and safety plan for the Environmental Restoration Program (ERP). It addresses the health and safety requirements of the Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA); Occupational Safety and Health Administration (OSHA) 29 CFR 1910.120 standard; and EG G Idaho, Inc. This plan is a guide to individuals who must complete a health and safety plan for a task performed for the ERP. It contains a task specific addendum that, when completed, specifically addresses task specific health and safety issues. This health and safety plan reduces the time it takes to write a task specific health and safety plan by providing discussions of requirements, guidance on where specific information is located, and specific topics in the Addendum that must be discussed at a task level. This format encourages a complete task specific health and safety plan and a standard for all health and safety plans written for ERP.

  6. Health and Safety Plan for Operations Performed for the Environmental Restoration Program: Task, Characterization of Potential Waste Sources at Auxiliary Reactor Area-1 Operable Unit 5--07 site ARA-02

    Energy Technology Data Exchange (ETDEWEB)

    Pickett, S.L.; Morton, S.L.

    1992-06-01

    This document constitutes the generic health and safety plan for the Environmental Restoration Program (ERP). It addresses the health and safety requirements of the Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA); Occupational Safety and Health Administration (OSHA) 29 CFR 1910.120 standard; and EG&G Idaho, Inc. This plan is a guide to individuals who must complete a health and safety plan for a task performed for the ERP. It contains a task specific addendum that, when completed, specifically addresses task specific health and safety issues. This health and safety plan reduces the time it takes to write a task specific health and safety plan by providing discussions of requirements, guidance on where specific information is located, and specific topics in the Addendum that must be discussed at a task level. This format encourages a complete task specific health and safety plan and a standard for all health and safety plans written for ERP.

  7. Phase 2 sampling and analysis plan, Quality Assurance Project Plan, and environmental health and safety plan for the Clinch River Remedial Investigation: An addendum to the Clinch River RCRA Facility Investigation plan

    International Nuclear Information System (INIS)

    Cook, R.B.; Adams, S.M.; Beauchamp, J.J.; Bevelhimer, M.S.; Blaylock, B.G.; Brandt, C.C.; Etnier, E.L.; Ford, C.J.; Frank, M.L.; Gentry, M.J.; Greeley, M.S.; Halbrook, R.S.; Harris, R.A.; Holladay, S.K.; Hook, L.A.; Howell, P.L.; Kszos, L.A.; Levine, D.A.; Skiles, J.L.; Suter, G.W.

    1992-12-01

    This document contains a three-part addendum to the Clinch River Resource Conservation and Recovery Act (RCRA) Facility Investigation Plan. The Clinch River RCRA Facility Investigation began in 1989, as part of the comprehensive remediation of facilities on the US Department of Energy Oak Ridge Reservation (ORR). The ORR was added to the National Priorities List in December 1989. The regulatory agencies have encouraged the adoption of Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA) terminology; therefore, the Clinch River activity is now referred to as the Clinch River Remedial Investigation (CRRI), not the Clinch River RCRA Facility Investigation. Part 1 of this document is the plan for sampling and analysis (S ampersand A) during Phase 2 of the CRRI. Part 2 is a revision of the Quality Assurance Project Plan for the CRRI, and Part 3 is a revision of the Environmental Health and Safety Plan for the CRRI. The Clinch River RI (CRRI) is designed to address the transport, fate, and distribution of waterborne contaminants (radionuclides, metals, and organic compounds) released from the DOE Oak Ridge Reservation (ORR) and to assess potential risks to human health and the environment associated with these contaminants. Primary areas of investigation are Melton Hill Reservoir, the Clinch River from Melton Hill Dam to its confluence with the Tennessee River, Poplar Creek, and Watts Bar Reservoir. The contaminants identified in the Clinch River/Watts Bar Reservoir (CR/WBR) downstream of the ORR are those associated with the water, suspended particles, deposited sediments, aquatic organisms, and wildlife feeding on aquatic organisms. The purpose of the Phase 2 S ampersand A Plan is to describe the proposed tasks and subtasks developed to meet the primary objectives of the CRRI

  8. Phase 2 sampling and analysis plan, Quality Assurance Project Plan, and environmental health and safety plan for the Clinch River Remedial Investigation: An addendum to the Clinch River RCRA Facility Investigation plan

    Energy Technology Data Exchange (ETDEWEB)

    Cook, R.B.; Adams, S.M.; Beauchamp, J.J.; Bevelhimer, M.S.; Blaylock, B.G.; Brandt, C.C.; Etnier, E.L.; Ford, C.J.; Frank, M.L.; Gentry, M.J.; Greeley, M.S.; Halbrook, R.S.; Harris, R.A.; Holladay, S.K.; Hook, L.A.; Howell, P.L.; Kszos, L.A.; Levine, D.A.; Skiles, J.L.; Suter, G.W.

    1992-12-01

    This document contains a three-part addendum to the Clinch River Resource Conservation and Recovery Act (RCRA) Facility Investigation Plan. The Clinch River RCRA Facility Investigation began in 1989, as part of the comprehensive remediation of facilities on the US Department of Energy Oak Ridge Reservation (ORR). The ORR was added to the National Priorities List in December 1989. The regulatory agencies have encouraged the adoption of Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA) terminology; therefore, the Clinch River activity is now referred to as the Clinch River Remedial Investigation (CRRI), not the Clinch River RCRA Facility Investigation. Part 1 of this document is the plan for sampling and analysis (S A) during Phase 2 of the CRRI. Part 2 is a revision of the Quality Assurance Project Plan for the CRRI, and Part 3 is a revision of the Environmental Health and Safety Plan for the CRRI. The Clinch River RI (CRRI) is designed to address the transport, fate, and distribution of waterborne contaminants (radionuclides, metals, and organic compounds) released from the DOE Oak Ridge Reservation (ORR) and to assess potential risks to human health and the environment associated with these contaminants. Primary areas of investigation are Melton Hill Reservoir, the Clinch River from Melton Hill Dam to its confluence with the Tennessee River, Poplar Creek, and Watts Bar Reservoir. The contaminants identified in the Clinch River/Watts Bar Reservoir (CR/WBR) downstream of the ORR are those associated with the water, suspended particles, deposited sediments, aquatic organisms, and wildlife feeding on aquatic organisms. The purpose of the Phase 2 S A Plan is to describe the proposed tasks and subtasks developed to meet the primary objectives of the CRRI.

  9. Nuclear safety in France

    International Nuclear Information System (INIS)

    Laverie, M.

    1981-02-01

    The principles and rules governing the safety of nuclear installations are defined as from three fundamental principles and three practical rules as follows: First principle: the operator is responsible and of the highest order. Second principle: the public authorities exercise their control responsibility with respect to the design, construction and running of the installations. Third principle: nuclear safety, this is to accept that man and his technique are not infallible and that one must be prepared to control the unpredictable. First rule: the installations must include several 'lines of defence' in succession and to the extent where this is possible these must be independent of each other. Second rule: procedures are required and supervised by the Government Departments. Third rule: nuclear safety requires that any incident or anomaly must undergo an analysis in depth and is also based on a standing 'clinical' examination of the installations. The definition is given as to how the public authorities exercise their intervention: terms and conditions of the intervention by the safety authorities, authorization procedures, surveillance of the installations, general technical regulations. Two specific subjects are presented in the addendum, (a) the choice of nuclear power station sites in France and (b) the storage of radioactive wastes [fr

  10. Sierra/SolidMechanics 4.48 User's Guide: Addendum for Shock Capabilities.

    Energy Technology Data Exchange (ETDEWEB)

    Plews, Julia A.; Crane, Nathan K; de Frias, Gabriel Jose; Le, San; Littlewood, David John; Merewether, Mark Thomas; Mosby, Matthew David; Pierson, Kendall H.; Porter, Vicki L.; Shelton, Timothy; Thomas, Jesse David; Tupek, Michael R.; Veilleux, Michael; Xavier, Patrick G.

    2018-03-01

    This is an addendum to the Sierra/SolidMechanics 4.48 User's Guide that documents additional capabilities available only in alternate versions of the Sierra/SolidMechanics (Sierra/SM) code. These alternate versions are enhanced to provide capabilities that are regulated under the U.S. Department of State's International Traffic in Arms Regulations (ITAR) export-control rules. The ITAR regulated codes are only distributed to entities that comply with the ITAR export-control requirements. The ITAR enhancements to Sierra/SM in- clude material models with an energy-dependent pressure response (appropriate for very large deformations and strain rates) and capabilities for blast modeling. Since this is an addendum to the standard Sierra/SM user's guide, please refer to that document first for general descriptions of code capability and use.

  11. Addendum: Observables for general relativity related to geometry

    International Nuclear Information System (INIS)

    Duch, Paweł; Kamiński, Wojciech; Lewandowski, Jerzy; Świeżewski, Jędrzej

    2015-01-01

    In this addendum we clarify a point which strengthens one of the results from http://dx.doi.org/10.1007/JHEP05(2014)077. Namely, we show that the algebra of the observables F(r,θ) is yet simpler then it was described in http://dx.doi.org/10.1007/JHEP05(2014)077. This is an important point, because with this simplification an important subalgebra becomes canonical, allowing for a natural reduction of the phase space.

  12. Addendum: Observables for general relativity related to geometry

    Energy Technology Data Exchange (ETDEWEB)

    Duch, Paweł [Institute of Physics, Jagiellonian University,Reymonta 4, 30-059 Kraków (Poland); Kamiński, Wojciech [Faculty of Physics, University of Warsaw,Hoża 69, 00-681 Warszawa (Poland); Lewandowski, Jerzy [Faculty of Physics, University of Warsaw,Hoża 69, 00-681 Warszawa (Poland); Institute for Quantum Gravity (IQG), FAU Erlangen - Nurnberg,Staudtstr. 7, 91058 Erlangen (Germany); Świeżewski, Jędrzej [Faculty of Physics, University of Warsaw,Hoża 69, 00-681 Warszawa (Poland)

    2015-04-15

    In this addendum we clarify a point which strengthens one of the results from http://dx.doi.org/10.1007/JHEP05(2014)077. Namely, we show that the algebra of the observables F(r,θ) is yet simpler then it was described in http://dx.doi.org/10.1007/JHEP05(2014)077. This is an important point, because with this simplification an important subalgebra becomes canonical, allowing for a natural reduction of the phase space.

  13. Hanford Tank Farms Vadose Zone, Addendum to the TX Tank Farm Report

    International Nuclear Information System (INIS)

    Spatz, R.

    2000-01-01

    This addendum to the TX Tank Farm Report (GJO-97-13-TAR, GJO-HAN-11) published in September 1997 incorporates the results of high-rate and repeat logging activities along with shape factor analysis of the logging data. A high-rate logging system was developed and deployed in the TX Tank Farm to measure cesium-137 concentration levels in high gamma flux zones where the spectral gamma logging system was unable to collect usable data because of high dead times and detector saturation. This report presents additional data and revised visualizations of subsurface contaminant distribution in the TX Tank Farm at the DOE Hanford Site in the state of Washington

  14. Hanford Tank Farms Vadose Zone, Addendum to the T Tank Farm Report

    Energy Technology Data Exchange (ETDEWEB)

    Spatz, Robert

    2000-07-01

    This addendum to the T Tank Farm Report (GJO-99-101-TARA, GJO-HAN-27) published in September 1999 incorporates the results of high-rate and repeat logging activities along with shape factor analysis of the logging incorporates the results of high-rate and repeat logging activities along with shape factor analysis of the logging data. A high-rate logging system was developed and deployed in the T Tank Farm to measure cesium-137 concentration levels in high gamma flux zones where the spectral gamma logging system was unable to collect usable data because of high dead times and detector saturation. This report presents additional data and revised visualizations of subsurface contaminant distribution in the T Tank Farm at the DOE Hanford Site in the state of Washington.

  15. Dictionary of minor planet names addendum to 6th edition 2012-2014

    CERN Document Server

    Schmadel, Lutz D

    2015-01-01

    The quantity of numbered minor planets is now approaching half a million. Together with this Addendum, the sixth edition of the Dictionary of Minor Planet Names, which is the IAU's official reference for the field, now covers more than 19,000 named minor planets. In addition to being of practical value for identification purposes, the Dictionary of Minor Planet Names provides authoritative information about the basis for the rich and colorful variety of ingenious names, from heavenly goddesses to artists, from scientists to Nobel laureates, from historical or political figures to ordinary women and men, from mountains to buildings, as well as a variety of compound terms and curiosities. This Addendum to the 6th edition of the Dictionary of Minor Planet Names adds approximately 2200 entries. It also contains many corrections, revisions and updates to the entries published in earlier editions. This work is an abundant source of information for anyone interested in minor planets and who enjoys reading about the ...

  16. Statement of Agreements Registered with the Agency. Addendum

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1961-10-24

    This addendum is divided into two parts. Part I contains supplementary information on certain agreements which were registered with the Agency up to 30 June 1960; part II is a statement of all agreements so registered between July 1960 and 30 June 1961. In compliance with Article VI of the Regulations for the Registration of Agreements this document is hereby transmitted to all Members of the Agency for their information. A copy is also being sent to the Secretary-General of the United Nations.

  17. Statement of Agreements Registered with the Agency. Addendum

    International Nuclear Information System (INIS)

    1961-01-01

    This addendum is divided into two parts. Part I contains supplementary information on certain agreements which were registered with the Agency up to 30 June 1960; part II is a statement of all agreements so registered between July 1960 and 30 June 1961. In compliance with Article VI of the Regulations for the Registration of Agreements this document is hereby transmitted to all Members of the Agency for their information. A copy is also being sent to the Secretary-General of the United Nations

  18. Addendum to 'Half coalescence of the m=1, n=1 magnetic island in tokamaks'

    International Nuclear Information System (INIS)

    Bussac, M.N.; Pellat, R.

    1985-01-01

    As an addendum to our previous work concerning the half-coalescence instability of an m=1, n=1 magnetic island in tokamaks, the potential energy is given for an arbitrary shape of the separatrix. (orig.)

  19. Safety balance: Analysis of safety systems

    International Nuclear Information System (INIS)

    Delage, M.; Giroux, C.

    1990-12-01

    Safety analysis, and particularly analysis of exploitation of NPPs is constantly affected by EDF and by the safety authorities and their methodologies. Periodic safety reports ensure that important issues are not missed on daily basis, that incidents are identified and that relevant actions are undertaken. French safety analysis method consists of three principal steps. First type of safety balance is analyzed at the normal start-up phase for each unit including the final safety report. This enables analysis of behaviour of units ten years after their licensing. Second type is periodic operational safety analysis performed during a few years. Finally, the third step consists of safety analysis of the oldest units with the aim to improve the safety standards. The three steps of safety analysis are described in this presentation in detail with the aim to present the objectives and principles. Examples of most recent exercises are included in order to illustrate the importance of such analyses

  20. Addendum to 'Reconciling grand unification with strings by anisotropic compactifications'

    International Nuclear Information System (INIS)

    Dundee, Ben; Raby, Stuart; Wingerter, Akin

    2009-01-01

    In this brief addendum, we clarify a point which we left unaddressed in a previous publication [Phys. Rev. D 78, 066006 (2008)]. In particular, we show that a specific vacuum configuration constructed in one of our models satisfies the condition D=0. In the previous publication, we only showed F=0. Both D=0 and F=0 are necessary to ensure that supersymmetry survives to the weak scale.

  1. Safety margins in deterministic safety analysis

    International Nuclear Information System (INIS)

    Viktorov, A.

    2011-01-01

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

  2. Safety analysis fundamentals

    International Nuclear Information System (INIS)

    Wright, A.C.D.

    2002-01-01

    This paper discusses the safety analysis fundamentals in reactor design. This study includes safety analysis done to show consequences of postulated accidents are acceptable. Safety analysis is also used to set design of special safety systems and includes design assist analysis to support conceptual design. safety analysis is necessary for licensing a reactor, to maintain an operating license, support changes in plant operations

  3. Addendum to ‘Understanding risks in the light of uncertainty: low-probability, high-impact coastal events in cities’

    Science.gov (United States)

    Galarraga, Ibon; Sainz de Murieta, Elisa; Markandya, Anil; María Abadie, Luis

    2018-02-01

    This addendum adds to the analysis presented in ‘Understanding risks in the light of uncertainty: low-probability, high-impact coastal events in cities’ Abadie et al (2017 Environ. Res. Lett. 12 014017). We propose to use the framework developed earlier to enhance communication and understanding of risks, with the aim of bridging the gap between highly technical risk management discussion to the public risk aversion debate. We also propose that the framework could be used for stress-testing resilience.

  4. 77 FR 21782 - International Conference on Harmonisation; Draft Guidance for Industry on E2C(R2) Periodic...

    Science.gov (United States)

    2012-04-11

    ... Safety Data Management: Periodic Safety Update Reports for Marketed Drugs'' (E2C guidance) and ``Addendum to E2C Clinical Safety Data Management: Periodic Safety Update Reports for Marketed Drugs'' (addendum....regulations.gov . Submit written comments to the Division of Dockets Management (HFA-305), Food and Drug...

  5. Transportation Safety Excellence in Operations Through Improved Transportation Safety Document

    International Nuclear Information System (INIS)

    Dr. Michael A. Lehto; MAL

    2007-01-01

    requirements with waste requirements. MFC's efforts illustrate that utilizing the requirements of other disciplines, beyond nuclear safety, can provide an efficient process. Analyzing current processes to find better ways of meeting the requirements of multiple disciplines within a safety basis can lead to a more cost-effective, streamlined process. (2) Incorporating the DOT Type A drums into the MFC TSD was efficient because safety analysts utilized a transportation plan that provided analysis that could also be used for the change to the TSD addendum. In addition, because the plan they used had already been approved and was in use by the Idaho Cleanup Project (ICP) at the INL, justification for the change to the TSD was more compelling. MFC safety analysts proved that streamlining a process can be made more feasible by drawing from analysis that has already been completed

  6. Dictionary of Minor Planet Names Addendum to Fifth Edition: 2006 - 2008

    CERN Document Server

    Schmadel, Lutz

    2009-01-01

    The second Addendum to the Dictionary of Minor Planet Names, fifth edition, which is the IAU's official reference for the field, contains all newly published names from the period 2006-2008 as well as corrections and amendments to earlier editions. In total the Dictionary of Minor Planet Names now covers some 15000 named minor planets. It provides authoritative information about the basis for the rich and colorful variety of ingenious names, from heavenly goddesses to more prosaic constructions.

  7. Addendum to environmental monitoring plan Nevada Test Site and support facilities

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1992-11-01

    This 1992 Addendum to the ``Environmental Monitoring Plan Nevada Test Site and Support Facilities -- 1991,`` Report No. DOE/NV/1 0630-28 (EMP) applies to the US Department of Energy`s (DOE`s) operations on the Continental US (including Amchitka Island, Alaska) that are under the purview of the DOE Nevada Field Office (DOE/NV). The primary purpose of these operations is the conduct of the nuclear weapons testing program for the DOE and the Department of Defense. Since 1951, these tests have been conducted principally at the Nevada Test Site (NTS), which is located approximately 100 miles northwest of Las Vegas, Nevada. In accordance with DOE Order 5400.1, this 1992 Addendum to the EMP brings together, in one document, updated information and/or new sections to the description of the environmental activities conducted at the NTS by user organizations, operations support contractors, and the US Environmental Protection Agency (EPA) originally published in the EMP. The EPA conducts both the offsite environmental monitoring program around the NTS and post-operational monitoring efforts at non-NTS test locations used between 1961 and 1973 in other parts of the continental US All of these monitoring activities are conducted under the auspices of the DOE/NV, which has the stated policy of conducting its operations in compliance with both the letter and the spirit of applicable environmental statutes, regulations, and standards.

  8. Addendum to Environmental Monitoring Plan, Nevada Test Site and Support Facilities

    International Nuclear Information System (INIS)

    1993-11-01

    This 1993 Addendum to the ''Environmental Monitoring Plan Nevada Test Site and Support Facilities -- 1991,'' Report No. DOE/NV/10630-28 (EMP) applies to the US Department of Energy's (DOE's) operations on the Continental US (including Amchitka Island, Alaska) that are under the purview of the DOE Nevada Operations Office (DOE/NV). The primary purpose of these operations is the conduct of the nuclear weapons testing program for the DOE and the Department of Defense. Since 1951, these tests have been conducted principally at the Nevada Test Site (NTS), which is located approximately 100 miles northwest of Las Vegas, Nevada. In accordance with DOE Order 5400.1, this 1993 Addendum to the EMP brings together, in one document, updated information and/or new sections to the description of the environmental activities conducted at the NTS by user organizations, operations support contractors, and the US Environmental Protection Agency (EPA) originally published in the EMP. The EPA conducts both the offsite environmental monitoring program around the NTS and post-operational monitoring efforts at non-NTS test locations used between 1961 and 1973 in other parts of the continental US. All of these monitoring activities are conducted under the auspices of the DOE/NV, which has the stated policy of conducting its operations in compliance with both the letter and the spirit of applicable environmental statutes, regulations, and standards

  9. Decommissioning of eight surplus production reactors at the Hanford Site, Richland, Washington. Addendum (Final Environmental Impact Statement)

    Energy Technology Data Exchange (ETDEWEB)

    1992-12-01

    The first section of this volume summarizes the content of the draft environmental impact statement (DEIS) and this Addendum, which together constitute the final environmental impact statement (FEIS) prepared on the decommissioning of eight surplus plutonium production reactors at Hanford. The FEIS consists of two volumes. The first volume is the DEIS as written. The second volume (this Addendum) consists of a summary; Chapter 9, which contains comments on the DEIS and provides DOE`s responses to the comments; Appendix F, which provides additional health effects information; Appendix K, which contains costs of decommissioning in 1990 dollars; Appendix L, which contains additional graphite leaching data; Appendix M, which contains a discussion of accident scenarios; Appendix N, which contains errata; and Appendix 0, which contains reproductions of the letters, transcripts, and exhibits that constitute the record for the public comment period.

  10. Linking Safety Analysis to Safety Requirements

    DEFF Research Database (Denmark)

    Hansen, Kirsten Mark

    Software for safety critical systems must deal with the hazards identified by safety analysistechniques: Fault trees, event trees,and cause consequence diagrams can be interpreted as safety requirements and used in the design activity. We propose that the safety analysis and the system design use...

  11. In Vitro Osteoblast Model for Bone Wound Infections and Antimicrobial Therapy (Addendum)

    Science.gov (United States)

    2012-01-01

    n=2). Previous investigations on bacterial arthritis induced by S . aureus and S . agalactiae have suggested that TNFα, interleukin1β, and IL6, are...REPOR PREPARED FOR DISTRIBUTION S Approved fo The views, opinio should not be con unless so designa January 2012 T: Addendum to Final : U.S...contained in this report are those of the author( s ) and strued as an official Department of the Army position, policy or decision ted by other

  12. Addendum 6 to CSAR 79-038 out-of-hood plutonium storage (burial box)

    International Nuclear Information System (INIS)

    Chiao, T.

    1995-01-01

    The Addendum considered an increase in the limit of fissile material in a stacked container array to 500 grams. In other words, the sum of fissile material in an array of containers is limited to 500 grams, regardless of whether the containers are stacked or not. The results of this evaluation indicates that with the modification of the fissile limits described, the system of a container array will stay sub-critical

  13. ARIADNE: Beam Time and Resources Request - Addendum

    CERN Document Server

    Hollywood, D; Mavrokoridis, K; McCormick, KJ; Philippou, B; Roberts, A; Stavrakis, G; Touramanis, C; Vann, J

    2017-01-01

    This addendum provides additional information and clarification requested by the SPS Committee in response to our previously submitted Beam Time Request (CERN-SPSC-2017-031 [SPSC-P-356]). The ARIADNE research programme is based around developing an optical readout system for use in LAr TPC detectors, which could potentially have a number of benefits over the currently used segmented anode planes - particularly in matters of scalability for future kilo-tonne scale detectors, ease of installation, use and maintenance, and reconstruction resolution. As detailed in our previously submitted Letter of Intent (CERN-SPSC-2016-008 [SPSC-I-244]), ARIADNE itself is a 1-tonne, two-phase LAr TPC detector, designed and built at the University of Liverpool, which uses an array of EMCCD cameras as its optical readout system, alongside PMTs for scintillation light observation and a segmented THGEM for charge readout.

  14. Safety analysis for research reactors

    International Nuclear Information System (INIS)

    2008-01-01

    The aim of safety analysis for research reactors is to establish and confirm the design basis for items important to safety using appropriate analytical tools. The design, manufacture, construction and commissioning should be integrated with the safety analysis to ensure that the design intent has been incorporated into the as-built reactor. Safety analysis assesses the performance of the reactor against a broad range of operating conditions, postulated initiating events and other circumstances, in order to obtain a complete understanding of how the reactor is expected to perform in these situations. Safety analysis demonstrates that the reactor can be kept within the safety operating regimes established by the designer and approved by the regulatory body. This analysis can also be used as appropriate in the development of operating procedures, periodic testing and inspection programmes, proposals for modifications and experiments and emergency planning. The IAEA Safety Requirements publication on the Safety of Research Reactors states that the scope of safety analysis is required to include analysis of event sequences and evaluation of the consequences of the postulated initiating events and comparison of the results of the analysis with radiological acceptance criteria and design limits. This Safety Report elaborates on the requirements established in IAEA Safety Standards Series No. NS-R-4 on the Safety of Research Reactors, and the guidance given in IAEA Safety Series No. 35-G1, Safety Assessment of Research Reactors and Preparation of the Safety Analysis Report, providing detailed discussion and examples of related topics. Guidance is given in this report for carrying out safety analyses of research reactors, based on current international good practices. The report covers all the various steps required for a safety analysis; that is, selection of initiating events and acceptance criteria, rules and conventions, types of safety analysis, selection of

  15. Addendum to the post-closure permit application for the Bear Creek hydrogeologic regime at the Y-12 plant: Walk-in pits

    International Nuclear Information System (INIS)

    1995-04-01

    In June 1987, the Resource Conservation and Recovery Act (RCRA) Closure/Post-Closure Plan for the Bear Creek Burial Grounds (BCBG) located at the Y-12 Plant on the Oak Ridge Reservation in Oak Ridge, Tennessee was submitted to the Tennessee Department of Environment and Conservation (TDEC) for review and approval.The Closure Plan has been modified and revised several times. This document is an addendum to the Post-Closure Permit Application submitted to TDEC in June, 1994. This addendum contains information on the Walk-In Pits of the BCBG which is meant to supplement the information provided in the Post-Closure Permit Application submitted for the BCBG. This document is not intended to be a stand-alone document.

  16. 78 FR 76889 - Proposed Addendum to the Interagency Policy Statement on Income Tax Allocation in a Holding...

    Science.gov (United States)

    2013-12-19

    .... Please use the title ``Proposed Addendum to the Interagency Policy Statement on Income Tax Allocation in... to the Interagency Policy Statement on Income Tax Allocation in a Holding Company Structure'' to... for the Institution. Any tax refund attributable to income earned, taxes paid, and losses incurred by...

  17. 75 FR 35786 - White River Minimum Flows-Addendum to Final Determination of Federal and Non-Federal Hydropower...

    Science.gov (United States)

    2010-06-23

    ... a revised discount rate selection for calculation of the present value of the losses for both the... determination rather than the discount rate selection proposed in the June 2009 draft addendum. Therefore, no change is required to the final determination related to the discount rate. The discount rate change...

  18. Time Based Workload Analysis Method for Safety-Related Operator Actions in Safety Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Yun Goo; Oh, Eung Se [Korea Hydro and Nuclear Power Co., Daejeon (Korea, Republic of)

    2016-05-15

    During the design basis event, the safety system performs safety functions to mitigate the event. The most of safety system is actuated by automatic system however, there are operator manual actions that are needed for the plant safety. These operator actions are classified as important human actions in human factors engineering design. The human factors engineering analysis and evaluation is needed for these important human actions to assure that operator successfully perform their tasks for plant safety and operational goals. The work load analysis is one of the required analysis for the important human actions.

  19. Time Based Workload Analysis Method for Safety-Related Operator Actions in Safety Analysis

    International Nuclear Information System (INIS)

    Kim, Yun Goo; Oh, Eung Se

    2016-01-01

    During the design basis event, the safety system performs safety functions to mitigate the event. The most of safety system is actuated by automatic system however, there are operator manual actions that are needed for the plant safety. These operator actions are classified as important human actions in human factors engineering design. The human factors engineering analysis and evaluation is needed for these important human actions to assure that operator successfully perform their tasks for plant safety and operational goals. The work load analysis is one of the required analysis for the important human actions.

  20. Reply to the SPSC questions on Addendum CERN-SPSC-2018-008 entitled Study of Hadron-Nucleus and Nucleus-Nucleus Collisions at the CERN SPS: Early Post-LS2 Measurements and Future Plans

    CERN Document Server

    Aduszkiewicz, A

    2018-01-01

    This document presents answers of the NA61/SHINE Collaborations to the SPSC questions on the addendum CERN-SPSC-2018-008 (referred to as ''Addendum'') to the NA61/SHINE proposal. Addendum requests an extension of the NA61/SHINE measurements beyond the Long Shutdown 2 and approval of the first physics data taking in 2022. The SPSC requested NA61/SHINE to consider a possibility to start the first physics data taking in 2021. NA61/SHINE concludes that this is possible and requests beam time in 2021 for detector commissioning and tests as well as for data taking with hadron (for neutrino physics) and Pb (for open charm measurements in Pb+Pb collisions) beams. The critical issue is timely flow of financial resources needed for the hardware of the detector upgrade.

  1. 78 FR 66040 - 30-Day Notice of Proposed Information Collection: HUD-Owned Real Estate-Sales Contract and Addendums

    Science.gov (United States)

    2013-11-04

    ... DEPARTMENT OF HOUSING AND URBAN DEVELOPMENT [Docket No. FR-5683-N-91] 30-Day Notice of Proposed Information Collection: HUD-Owned Real Estate--Sales Contract and Addendums AGENCY: Office of the Chief... published a 30 day notice of proposed information collection entitled HUD-Owned Real Estate-Sales Contract...

  2. 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

  3. Addendum to: Derivation of in situ opalinus clay porewater compositions from experimental and geochemical modelling studies

    International Nuclear Information System (INIS)

    Bradbury, M.H.; Baeyens; Pearson, F.J.; Berner, U.

    1998-01-01

    As part of the synthesis of water chemistry studies within the hydrochemical program at Mont Terri (Switzerland), a reexamination of the modelling method showed that it should lead to a range of water compositions rather than to a single composition. The single composition resulted from two compensating oversights, a theoretical one and a modelling one. These are discussed in this Addendum. (author)

  4. K Basin safety analysis

    International Nuclear Information System (INIS)

    Porten, D.R.; Crowe, R.D.

    1994-01-01

    The purpose of this accident safety analysis is to document in detail, analyses whose results were reported in summary form in the K Basins Safety Analysis Report WHC-SD-SNF-SAR-001. The safety analysis addressed the potential for release of radioactive and non-radioactive hazardous material located in the K Basins and their supporting facilities. The safety analysis covers the hazards associated with normal K Basin fuel storage and handling operations, fuel encapsulation, sludge encapsulation, and canister clean-up and disposal. After a review of the Criticality Safety Evaluation of the K Basin activities, the following postulated events were evaluated: Crane failure and casks dropped into loadout pit; Design basis earthquake; Hypothetical loss of basin water accident analysis; Combustion of uranium fuel following dryout; Crane failure and cask dropped onto floor of transfer area; Spent ion exchange shipment for burial; Hydrogen deflagration in ion exchange modules and filters; Release of Chlorine; Power availability and reliability; and Ashfall

  5. Safety analysis for 'Fugen'

    International Nuclear Information System (INIS)

    1997-10-01

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

  6. An eight pushbutton control system for the PSU ultra-safe nuclear plant: Addendum to annual report

    International Nuclear Information System (INIS)

    Schultz, M.A.

    1988-10-01

    The control system described in this report was developed for a ''Light Water Ultra-Safe Plant Concept'' studied at Penn State during a 2 year program funded by the Department of Energy. A reconfigured pressurized water reactor was designed which eliminated the conventional PWR pressurizer and added an active pumping system with a large dedicated water storage tank to control pressure. The thermal hydraulics aspects of the design were studied and reported in several master's theses and other addendum reports and first annual report. In addition to improving the inherent safety of an advanced plant by changes in the thermal hydraulic aspects of the design, a new approach to control system organization and design should also be included in an advanced plant. The study presented in this report modifies a conventional PWR plant control system to the needs of the Ultra-Safe Plant Concept and then describes implementation of the control system in a distributed network of control computers. The end result is a control system that is much simpler at the operator level. This simplicity should eliminate the possibility of operator error in the Ultra-Safe plant

  7. Reactor Safety Analysis

    International Nuclear Information System (INIS)

    Arien, B.

    2000-01-01

    The objective of SCK-CEN's programme on reactor safety is to develop expertise in probabilistic and deterministic reactor safety analysis. The research programme consists of two main activities, in particular the development of software for reliability analysis of large systems and participation in the international PHEBUS-FP programme for severe accidents. Main achievements in 1999 are reported

  8. Annual Status Report (FY2008) Composite Analysis of Low-Level Waste Disposal in the Central Plateau at the Hanford Site

    Energy Technology Data Exchange (ETDEWEB)

    Nichols, W. E. [Hanford Site (HNF), Richland, WA (United States)

    2009-12-18

    In accordance with the U.S. Department of Energy (DOE) requirements in DOE 0 435.1, Radioactive to be considered or purposes of Waste Management, and implemented by DOE/RL-2000-292, Maintenance Plan for the Composite Analysis of the Hanford Site, Southeast Washington, the DOE Richland Operations Office has prepared this annual report for fiscal year 2008 of PNNL-1 1800, Composite Analysis for the Low-Level Waste Disposal in the 200-Area Plateau of the Hanford Site, hereafter referred to as the Composite Analysis. The main emphasis of DOE/RL-2000-29 Is to identify additional data and information to enhance the Composite Analysis and the subsequent PNNL- 11800 Addendum, Addendum to Composite Analysis for Low-Level Waste Disposal in the 200 Area Plateau of the Hanford Site, hereafter referred to as the Addendum, and to address secondary issues identified during the review of the Composite Analysis.

  9. Reliability analysis of PLC safety equipment

    Energy Technology Data Exchange (ETDEWEB)

    Yu, J.; Kim, J. Y. [Chungnam Nat. Univ., Daejeon (Korea, Republic of)

    2006-06-15

    FMEA analysis for Nuclear Safety Grade PLC, failure rate prediction for nuclear safety grade PLC, sensitivity analysis for components failure rate of nuclear safety grade PLC, unavailability analysis support for nuclear safety system.

  10. Reliability analysis of PLC safety equipment

    International Nuclear Information System (INIS)

    Yu, J.; Kim, J. Y.

    2006-06-01

    FMEA analysis for Nuclear Safety Grade PLC, failure rate prediction for nuclear safety grade PLC, sensitivity analysis for components failure rate of nuclear safety grade PLC, unavailability analysis support for nuclear safety system

  11. Addendum report of the JHPS expert committee on radiation protection of the lens of the eye (3). Recent related overseas activities

    International Nuclear Information System (INIS)

    Akahane, Keiichi; Tatsuzaki, Hideo; Iimoto, Takeshi; Ichiji, Takeshi; Hamada, Nobuyuki; Iwai, Satoshi; Ohguchi, Hiroyuki; Ohno, Kazuko; Kawaura, Chiyo; Tsujimura, Norio; Hotta, Yutaka; Yamasaki, Tadashi; Yokoyama, Sumi

    2015-01-01

    Following the Statement on Tissue Reaction issued by the International Commission on Radiological Protection (ICRP), revisions and discussions are underway in various international organizations about the standards, guidelines and methods for the eye dosimetry. We have previously discussed these issues in six interim reports, followed by two addendum reports on lens dosimetry. This addendum report reviews ongoing overseas discussion and work in relation to research on radiogenic cataracts and implementation of the new eye lens dose limit, such as those in the European Commission, Low Dose Research towards Multidisciplinary Integration (DoReMi), Open Project for European Radiation Research Area (OPERRA), and the Unite Sates National Council on Radiation Protection and Measurements (NCRP). This report serves as the last part of a series of reports by the Japan Health Physics Society Expert Committee on Radiation Protection of the Lens of the Eye. (author)

  12. Reactor safety analysis

    International Nuclear Information System (INIS)

    Arien, B.

    1998-01-01

    Risk assessments of nuclear installations require accurate safety and reliability analyses to estimate the consequences of accidental events and their probability of occurrence. The objective of the work performed in this field at the Belgian Nuclear Research Centre SCK-CEN is to develop expertise in probabilistic and deterministic reactor safety analysis. The four main activities of the research project on reactor safety analysis are: (1) the development of software for the reliable analysis of large systems; (2) the development of an expert system for the aid to diagnosis; (3) the development and the application of a probabilistic reactor-dynamics method, and (4) to participate in the international PHEBUS-FP programme for severe accidents. Progress in research during 1997 is described

  13. TU-D-201-03: Results of a Survey On the Implementation of the TG-51 Protocol and Associated Addendum On Reference Dosimetry of External Beams

    Energy Technology Data Exchange (ETDEWEB)

    Kim, G [University of California, San Diego, La Jolla, CA (United States); Muir, B [National Research Council, Ottawa, AB (Canada); Culberson, W [University of Wisconsin Madison, Madison, WI (United States); Davis, S [McGill University Health Center, Montreal, QC (Canada); Huang, Y [Henry Ford Health System, West Bloomfield, MI (United States); Lee, S [University of Maryland School of Medicine, Columbia, MD (United States); Lowenstein, J [UT MD Anderson Cancer Center, Houston, TX (United States); Sarfehnia, A [Sunnybrook Health Science Center, Toronto, ON (Canada); Tolani, N [Michael E. DeBakey VA Medical Center, Sugarland (United States); Siebers, J [University of Virginia Health System, Charlottesville, VA (United States)

    2016-06-15

    Purpose: The working group on the review and extension of the TG-51 protocol (WGTG51) collected data from American Association of Physicists in Medicine (AAPM) members with respect to their current TG-51 and associated addendum usage in the interest of considering future protocol addenda and guidance on reference dosimetry best practices. This study reports an overview of this survey on dosimetry of external beams. Methods: Fourteen survey questions were developed by WGTG51 and released in November 2015. The questions collected information on reference dosimetry, beam quality specification, and ancillary calibration equipment. Results: Of the 190 submissions completed worldwide (U.S. 70%), 83% were AAPM members. Of the respondents, 33.5% implemented the TG-51 addendum, with the maximum calibration difference for any photon beam, with respect to the original TG-51 protocol, being <1% for 97.4% of responses. One major finding is that 81.8% of respondents used the same cylindrical ionization chamber for photon and electron dosimetry, implying that many clinics are foregoing the use of parallel-plate chambers. Other evidence suggests equivalent dosimetric results can be obtained with both cylindrical and parallel-plate chambers in electron beams. This, combined with users comfort with cylindrical chambers for electrons will likely impact recommendations put forward in an upcoming electron beam addendum to the TG-51 protocol. Data collected on ancillary equipment showed 58.2% (45.0%) of the thermometers (barometers) in use for beam calibration had NIST traceable calibration certificates, but 48.4% (42.7%) were never recalibrated. Conclusion: This survey provides a snapshot of TG-51 external beam reference dosimetry practice in radiotherapy centers. Findings demonstrate the rapid take-up of the TG-51 photon beam addendum and raise issues for the WGTG51 to focus on going forward, including guidelines on ancillary equipment and the choice of chamber for electron beam

  14. SU-G-TeP2-03: Comparison of Standard Dosimetry Protocol in Japan and AAPM TG-51 Addendum in Order to Establish Optimal Dosimetry for FFF Beam

    Energy Technology Data Exchange (ETDEWEB)

    Matsunaga, T; Adachi, Y [Department of Radiology, Seirei Hamamatsu General Hospital, Hamamatsu, Shizuoka (Japan); Hayashi, N [Graduate School of Health Sciences, Fujita Health University, Tayoake, Aichi (Japan); Nozue, M [Department of Radiation Oncology, Seirei Hamamtsu General Hospital, Hamamatsu, Shizuoka (Japan)

    2016-06-15

    Purpose: Japan Standard Dosimetry of Absorbed dose to water in external beam radiotherapy (JSDP12) is widely used to measure radiation dose in radiotherapy. However, JSDP12 does not take flattening-filter-free (FFF) beam into consideration. In addition, JSDP12 applied TPR20,10 for dose quality index for photon beam. The purpose of this study is to compare JSDP12 with AAPM TG-51 addendum in order to establish optimal dosimetry procedure for FFF beam. Method: We evaluated the ion-recombination factor (ks) and the correction factor of radial beam profile (Prp) in FFF beam dosimetry. The ks was introduced by 2 voltages method and verified by Jaffe’s plot. The Prp was given by both film measurement and calculation of treatment planning system, and compared them. Next, we compared the dose quality indexes (kQ) between TPR20,10 method and PDD(10)x method. Finally we considered optimal dosimetry protocol for FFF photon beam using JSDP12 with referring TG-51 addendum protocols. The FFF photon beams of 6 MV (6X-FFF) and 10 MV (10X-FFF) from TrueBeam were investigated in this study. Results: The ks for 6X-FFF and 10X-FFF beams were 1.005 and 1.010, respectively. The Prp of 0.6 cc ionization chamber for 6X-FFF and 10X-FFF beams (Film, TPS) were (1.004, 1.008) and (1.005, 1.008), respectively. The kQ for 6X-FFF and 10X-FFF beams (JSDP12, TG-51 addendum) were (0.9950, 0.9947) and (0.9851, 0.9845), respectively. The most effective factor for uncertainty in FFF photon beam measurement was Prp for JSDP12 formalism. Total dosimetric differences between JSDP12 and TG-51 addendum for 6X-FFF and 10X-FFF were -0.47% and -0.73%, respectively. Conclusion: The total dosimetric difference between JSDP12 and TG-51 addendum was within 1%. The introduction of kQ given by JSDP is feasible for FFF photon beam dosimetry. However, we think Prp should be considered for optimal dosimetry procedure even if JSDP12 is used for FFF photon beam dosimetry.

  15. Periodic safety review of the HTR-10 safety analysis

    International Nuclear Information System (INIS)

    Chen Fubing; Zheng Yanhua; Shi Lei; Li Fu

    2015-01-01

    Designed by the Institute of Nuclear and New Energy Technology (INET) of Tsinghua University, the 10 MW High Temperature Gas-cooled Reactor-Test Module (HTR-10) is the first modular High Temperature Gas-cooled Reactor (HTGR) in China. According to the nuclear safety regulations of China, the periodic safety review (PSR) of the HTR-10 was initiated by INET after approved by the National Nuclear Safety Administration (NNSA) of China. Safety analysis of the HTR-10 is one of the key safety factors of the PSR. In this paper, the main contents in the review of safety analysis are summarized; meanwhile, the internal evaluation on the review results is presented by INET. (authors)

  16. 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

  17. 10 CFR 70.62 - Safety program and integrated safety analysis.

    Science.gov (United States)

    2010-01-01

    ...; (iv) Potential accident sequences caused by process deviations or other events internal to the... have experience in nuclear criticality safety, radiation safety, fire safety, and chemical process... this safety program; namely, process safety information, integrated safety analysis, and management...

  18. The practical implementation of integrated safety management for nuclear safety analysis and fire hazards analysis documentation

    International Nuclear Information System (INIS)

    COLLOPY, M.T.

    1999-01-01

    In 1995 Mr. Joseph DiNunno of the Defense Nuclear Facilities Safety Board issued an approach to describe the concept of an integrated safety management program which incorporates hazard and safety analysis to address a multitude of hazards affecting the public, worker, property, and the environment. Since then the U S . Department of Energy (DOE) has adopted a policy to systematically integrate safety into management and work practices at all levels so that missions can be completed while protecting the public, worker, and the environment. While the DOE and its contractors possessed a variety of processes for analyzing fire hazards at a facility, activity, and job; the outcome and assumptions of these processes have not always been consistent for similar types of hazards within the safety analysis and the fire hazard analysis. Although the safety analysis and the fire hazard analysis are driven by different DOE Orders and requirements, these analyses should not be entirely independent and their preparation should be integrated to ensure consistency of assumptions, consequences, design considerations, and other controls. Under the DOE policy to implement an integrated safety management system, identification of hazards must be evaluated and agreed upon to ensure that the public. the workers. and the environment are protected from adverse consequences. The DOE program and contractor management need a uniform, up-to-date reference with which to plan. budget, and manage nuclear programs. It is crucial that DOE understand the hazards and risks necessarily to authorize the work needed to be performed. If integrated safety management is not incorporated into the preparation of the safety analysis and the fire hazard analysis, inconsistencies between assumptions, consequences, design considerations, and controls may occur that affect safety. Furthermore, confusion created by inconsistencies may occur in the DOE process to grant authorization of the work. In accordance with

  19. The detection and settlement of VAT fraud in four countries : Addendum to the report value-added tax fraud in the European Union

    NARCIS (Netherlands)

    Aronowitz, A.A.; Laagland, D.C.G.; Paulides, G.

    1996-01-01

    This addendum to the report 'Value-added tax fraud in the European Union' gives additional information on the detection and settlement of VAT fraud in Belgium, Germany, the United Kingdom and the Netherlands. See link to the report at: More information.

  20. IAEA Review for Gap Analysis of Safety Analysis Capability

    International Nuclear Information System (INIS)

    Basic, Ivica; Kim, Manwoong; Huges, Peter; Lim, B-K; D'Auria, Francesco; Louis, Vidard Michael

    2014-01-01

    The IAEA Asian Nuclear Safety Network (ANSN) was launched in 2002 in the framework of the Extra Budgetary Programme (EBP) on the Safety of Nuclear Installations in the South East Asia, Pacific and Far East Countries. The main objective is to strengthen and expand human and advanced Information Technology (IT) network to pool, analyse and share nuclear safety knowledge and practical experience for peaceful uses in this region. Under the ANSN framework, a technical group on Safety Analysis (SATG) was established in 2004 aimed to providing a forum for the exchange of experience in the following areas of safety analysis: · To provide a forum for an exchange of experience in the area of safety analysis, · To maintain and improve the knowledge on safety analysis method, · To enhance the utilization of computer codes, · To pool and analyse the issues related with safety analysis of research reactor, and · To facilitate mutual interested on safety analysis among member countries. A sustainable and successful nuclear energy programme requires a strong technical infrastructure, including a workforce made up of highly specialized and well-educated professionals. A significant portion of this technical capacity must be dedicated to safety- especially to safety analysis- as only then can it serve as the basis for making the right decisions during the planning, licensing, construction and operation of new nuclear facilities. In this regard, the IAEA has provided ANSN member countries with comprehensive training opportunities for capacity building in safety analysis. Nevertheless, the SATG recognizes that it is difficult to achieve harmonization in this area among all member countries because of their different competency levels. Therefore, it is necessary to quickly identify the most obvious gaps in safety analysis capability and then to use existing resources to begin to fill those gaps. The goal of this Expert Mission (EM) for gap finding service is to facilitate

  1. Addendum to foundations of multidimensional wave field signal theory: Gaussian source function

    Directory of Open Access Journals (Sweden)

    Natalie Baddour

    2018-02-01

    Full Text Available Many important physical phenomena are described by wave or diffusion-wave type equations. Recent work has shown that a transform domain signal description from linear system theory can give meaningful insight to multi-dimensional wave fields. In N. Baddour [AIP Adv. 1, 022120 (2011], certain results were derived that are mathematically useful for the inversion of multi-dimensional Fourier transforms, but more importantly provide useful insight into how source functions are related to the resulting wave field. In this short addendum to that work, it is shown that these results can be applied with a Gaussian source function, which is often useful for modelling various physical phenomena.

  2. Addendum to foundations of multidimensional wave field signal theory: Gaussian source function

    Science.gov (United States)

    Baddour, Natalie

    2018-02-01

    Many important physical phenomena are described by wave or diffusion-wave type equations. Recent work has shown that a transform domain signal description from linear system theory can give meaningful insight to multi-dimensional wave fields. In N. Baddour [AIP Adv. 1, 022120 (2011)], certain results were derived that are mathematically useful for the inversion of multi-dimensional Fourier transforms, but more importantly provide useful insight into how source functions are related to the resulting wave field. In this short addendum to that work, it is shown that these results can be applied with a Gaussian source function, which is often useful for modelling various physical phenomena.

  3. Test Plan Addendum No. 1: Waste Isolation Pilot Plant bin-scale CH TRU waste tests

    International Nuclear Information System (INIS)

    Molecke, M.A.; Lappin, A.R.

    1990-12-01

    This document is the first major revision to the Test Plan: WIPP Bin-Scale CH TRU Waste Tests. Factors that make this revision necessary are described and justified in Section 1, and elaborated upon in Section 4. This addendum contains recommended estimates of, and details for: (1) The total separation of waste leaching/solubility tests from bin-scale gas tests, including preliminary details and quantities of leaching tests required for testing of Levels 1, 2, and 3 WIPP CH TRU wastes; (2) An initial description and quantification of bin-scale gas test Phase 0, added to provide a crucial tie to pretest waste characterization representatives and overall test statistical validation; (3) A revision to the number of test bins required for Phases 1 and 2 of the bin gas test program, and specification of the numbers of additional bin tests required for incorporating gas testing of Level 2 wastes into test Phase 3. Contingencies are stated for the total number of test bins required, both positive and negative, including the supporting assumptions, logic, and decision points. (4) Several other general test detail updates occurring since the Test Plan was approved and published in January, 1990. Possible impacts of recommended revisions included in this Addendum on WIPP site operations are called out and described. 56 refs., 12 tabs

  4. Reactor Safety Analysis

    International Nuclear Information System (INIS)

    Arien, B.

    1998-01-01

    The objective of SCK-CEN's programme on reactor safety is to develop expertise in probabilistic and deterministic reactor safety analysis. The research programme consists of four main activities, in particular the development of software for reliability analysis of large systems and participation in the international PHEBUS-FP programme for severe accidents, the development of an expert system for the aid to diagnosis; the development and application of a probabilistic reactor dynamics method. Main achievements in 1999 are reported

  5. The Economic and Workforce Development Program (ED>Net) Annual Report, 2001-02 [and] Addendum to FY 01-02 Annual Report.

    Science.gov (United States)

    California Community Colleges, Sacramento. Economic Development Coordination Network (EDNet).

    This document contains an annual report and its addendum from the Economic and Workforce Development Program of California Community Colleges. The annual report provides an overview of the Program's evaluation processes, regional centers, short-term projects, legislation, strategic plan, etc. It also provides vital facts about the program such as…

  6. Systems engineered health and safety criteria for safety analysis reports

    International Nuclear Information System (INIS)

    Beitel, G.A.; Morcos, N.

    1993-01-01

    The world of safety analysis is filled with ambiguous words: codes and standards, consequences and risks, hazard and accident, and health and safety. These words have been subject to disparate interpretations by safety analysis report (SAR) writers, readers, and users. open-quotes Principal health and safety criteriaclose quotes has been one of the most frequently misused phrases; rarely is it used consistently or effectively. This paper offers an easily understood definition for open-quotes principal health and safety criteriaclose quotes and uses systems engineering to convert an otherwise mysterious topic into the primary means of producing an integrated SAR. This paper is based on SARs being written for environmental restoration and waste management activities for the U.S. Department of Energy (DOE). Requirements for these SARs are prescribed in DOE Order 5480-23, open-quotes Nuclear Safety Analysis Reports.close quotes

  7. Ignalina Safety Analysis Group

    International Nuclear Information System (INIS)

    Ushpuras, E.

    1995-01-01

    The article describes the fields of activities of Ignalina NPP Safety Analysis Group (ISAG) in the Lithuanian Energy Institute and overview the main achievements gained since the group establishment in 1992. The group is working under the following guidelines: in-depth analysis of the fundamental physical processes of RBMK-1500 reactors; collection, systematization and verification of the design and operational data; simulation and analysis of potential accident consequences; analysis of thermohydraulic and neutronic characteristics of the plant; provision of technical and scientific consultations to VATESI, Governmental authorities, and also international institutions, participating in various projects aiming at Ignalina NPP safety enhancement. The ISAG is performing broad scientific co-operation programs with both Eastern and Western scientific groups, supplying engineering assistance for Ignalina NPP. ISAG is also participating in the joint Lithuanian - Swedish - Russian project - Barselina, the first Probabilistic Safety Assessment (PSA) study of Ignalina NPP. The work is underway together with Maryland University (USA) for assessment of the accident confinement system for a range of breaks in the primary circuit. At present the ISAG personnel is also involved in the project under the grant from the Nuclear Safety Account, administered by the European Bank for reconstruction and development for the preparation and review of an in-depth safety assessment of the Ignalina plant

  8. Health effects models for nuclear power plant accident consequence analysis

    International Nuclear Information System (INIS)

    Abrahamson, S.; Bender, M.A.; Boecker, B.B.; Scott, B.R.

    1993-05-01

    The Nuclear Regulatory Commission (NRC) has sponsored several studies to identify and quantify, through the use of models, the potential health effects of accidental releases of radionuclides from nuclear power plants. The Reactor Safety Study provided the basis for most of the earlier estimates related to these health effects. Subsequent efforts by NRC-supported groups resulted in improved health effects models that were published in the report entitled open-quotes Health Effects Models for Nuclear Power Plant Consequence Analysisclose quotes, NUREG/CR-4214, 1985 and revised further in the 1989 report NUREG/CR-4214, Rev. 1, Part 2. The health effects models presented in the 1989 NUREG/CR-4214 report were developed for exposure to low-linear energy transfer (LET) (beta and gamma) radiation based on the best scientific information available at that time. Since the 1989 report was published, two addenda to that report have been prepared to (1) incorporate other scientific information related to low-LET health effects models and (2) extend the models to consider the possible health consequences of the addition of alpha-emitting radionuclides to the exposure source term. The first addendum report, entitled open-quotes Health Effects Models for Nuclear Power Plant Accident Consequence Analysis, Modifications of Models Resulting from Recent Reports on Health Effects of Ionizing Radiation, Low LET Radiation, Part 2: Scientific Bases for Health Effects Models,close quotes was published in 1991 as NUREG/CR-4214, Rev. 1, Part 2, Addendum 1. This second addendum addresses the possibility that some fraction of the accident source term from an operating nuclear power plant comprises alpha-emitting radionuclides. Consideration of chronic high-LET exposure from alpha radiation as well as acute and chronic exposure to low-LET beta and gamma radiations is a reasonable extension of the health effects model

  9. Preliminary safety analysis methodology for the SMART

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Kyoo Hwan; Chung, Y. J.; Kim, H. C.; Sim, S. K.; Lee, W. J.; Chung, B. D.; Song, J. H. [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2000-03-01

    This technical report was prepared for a preliminary safety analysis methodology of the 330MWt SMART (System-integrated Modular Advanced ReacTor) which has been developed by Korea Atomic Energy Research Institute (KAERI) and funded by the Ministry of Science and Technology (MOST) since July 1996. This preliminary safety analysis methodology has been used to identify an envelope for the safety of the SMART conceptual design. As the SMART design evolves, further validated final safety analysis methodology will be developed. Current licensing safety analysis methodology of the Westinghouse and KSNPP PWRs operating and under development in Korea as well as the Russian licensing safety analysis methodology for the integral reactors have been reviewed and compared to develop the preliminary SMART safety analysis methodology. SMART design characteristics and safety systems have been reviewed against licensing practices of the PWRs operating or KNGR (Korean Next Generation Reactor) under construction in Korea. Detailed safety analysis methodology has been developed for the potential SMART limiting events of main steam line break, main feedwater pipe break, loss of reactor coolant flow, CEA withdrawal, primary to secondary pipe break and the small break loss of coolant accident. SMART preliminary safety analysis methodology will be further developed and validated in parallel with the safety analysis codes as the SMART design further evolves. Validated safety analysis methodology will be submitted to MOST as a Topical Report for a review of the SMART licensing safety analysis methodology. Thus, it is recommended for the nuclear regulatory authority to establish regulatory guides and criteria for the integral reactor. 22 refs., 18 figs., 16 tabs. (Author)

  10. Transient analysis for resolving safety issues

    International Nuclear Information System (INIS)

    Chao, J.; Layman, W.

    1987-01-01

    The Nuclear Safety Analysis Center (NSAC) has a Generic Safety Analysis Program to help resolve high priority generic safety issues. This paper describes several high priority safety issues considered at NSAC and how they were resolved by transient analysis using thermal hydraulics and neutronics codes. These issues are pressurized thermal shock (PTS), anticipated transients without scram (ATWS), steam generator tube rupture (SGTR), and reactivity transients in light of the Chernobyl accident

  11. Addendum to "Increasing Early Childhood Educators' Use of Communication-Facilitating and Language-Modelling Strategies: Brief Speech and Language Therapy Training"

    Science.gov (United States)

    McDonald, David; Proctor, Penny; Gill, Wendy; Heaven, Sue; Marr, Jane; Young, Jane

    2015-01-01

    The purpose of this addendum is to include within our discussion the findings of Girolametto et al. (2007), a randomized controlled trial of Teacher Talk training, an adapted version of Learning Language and Loving It (LLLI) (Weitzman and Greenberg, 2002). Teacher Talk does not include the coaching and video feedback elements of LLLI. Girolametto…

  12. Super-GCA connection with tensionless strings: Addendum to “Super-GCA from N=(2,2 super-Virasoro” [Phys. Lett. B 754 (2016 195–200

    Directory of Open Access Journals (Sweden)

    Ipsita Mandal

    2016-09-01

    Full Text Available In this addendum, we consider the connection between certain 2d super-GCA, obtained from the parametric contractions of 2d SCFTs, which can describe the constraint algebra of null spinning strings.

  13. A study of software safety analysis system for safety-critical software

    International Nuclear Information System (INIS)

    Chang, H. S.; Shin, H. K.; Chang, Y. W.; Jung, J. C.; Kim, J. H.; Han, H. H.; Son, H. S.

    2004-01-01

    The core factors and requirements for the safety-critical software traced and the methodology adopted in each stage of software life cycle are presented. In concept phase, Failure Modes and Effects Analysis (FMEA) for the system has been performed. The feasibility evaluation of selected safety parameter was performed and Preliminary Hazards Analysis list was prepared using HAZOP(Hazard and Operability) technique. And the check list for management control has been produced via walk-through technique. Based on the evaluation of the check list, activities to be performed in requirement phase have been determined. In the design phase, hazard analysis has been performed to check the safety capability of the system with regard to safety software algorithm using Fault Tree Analysis (FTA). In the test phase, the test items based on FMEA have been checked for fitness guided by an accident scenario. The pressurizer low pressure trip algorithm has been selected to apply FTA method to software safety analysis as a sample. By applying CASE tool, the requirements traceability of safety critical system has been enhanced during all of software life cycle phases

  14. CSER-00-007 Addendum 1 Criticality Safety Evaluation of Shippingport PWR Core 2 Blanket Fuel Assemblies at Lower Exposures

    International Nuclear Information System (INIS)

    WITTEKIND, W.D.

    2001-01-01

    This analysis meets the requirements of HNF-7098, Criticality Safety Program, (FH 2001a). HNF-7098 states that before starting a new operation with fissile material or before an existing operation is changed, it shall be determined that the entire process will be subcritical under both normal and credible abnormal conditions. To demonstrate the Incredibility Principle is satisfied, this Criticality Safety Evaluation Report (CSER) shows that the form or distribution is such that criticality is impossible. This evaluation demonstrated, that on the basis of effective 235 U enrichment, criticality is not possible. The minimum blanket assembly exposure is 4,375 MW t d/MTU for fissile material that is shown to fulfill the Incredibility Principle safety criterion on the basis of enrichment

  15. Safety analysis of autonomous excavator functionality

    International Nuclear Information System (INIS)

    Seward, D.; Pace, C.; Morrey, R.; Sommerville, I.

    2000-01-01

    This paper presents an account of carrying out a hazard analysis to define the safety requirements for an autonomous robotic excavator. The work is also relevant to the growing generic class of heavy automated mobile machinery. An overview of the excavator design is provided and the concept of a safety manager is introduced. The safety manager is an autonomous module responsible for all aspects of system operational safety, and is central to the control system's architecture. Each stage of the hazard analysis is described, i.e. system model creation, hazard definition and hazard analysis. Analysis at an early stage of the design process, and on a system that interfaces directly to an unstructured environment, exposes certain issues relevant to the application of current hazard analysis methods. The approach taken in the analysis is described. Finally, it is explained how the results of the hazard analysis have influenced system design, in particular, safety manager specifications. Conclusions are then drawn about the applicability of hazard analysis of requirements in general, and suggestions are made as to how the approach can be taken further

  16. Solid waste burial grounds interim safety analysis

    International Nuclear Information System (INIS)

    Saito, G.H.

    1994-01-01

    This Interim Safety Analysis document supports the authorization basis for the interim operation and restrictions on interim operations for the near-surface land disposal of solid waste in the Solid Waste Burial Grounds. The Solid Waste Burial Grounds Interim Safety Basis supports the upgrade progress for the safety analysis report and the technical safety requirements for the operations in the Solid Waste Burial Grounds. Accident safety analysis scenarios have been analyzed based on the significant events identified in the preliminary hazards analysis. The interim safety analysis provides an evaluation of the operations in the Solid Waste Burial Grounds to determine if the radiological and hazardous material exposures will be acceptable from an overall health and safety standpoint to the worker, the onsite personnel, the public, and the environment

  17. Solid waste burial grounds interim safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Saito, G.H.

    1994-10-01

    This Interim Safety Analysis document supports the authorization basis for the interim operation and restrictions on interim operations for the near-surface land disposal of solid waste in the Solid Waste Burial Grounds. The Solid Waste Burial Grounds Interim Safety Basis supports the upgrade progress for the safety analysis report and the technical safety requirements for the operations in the Solid Waste Burial Grounds. Accident safety analysis scenarios have been analyzed based on the significant events identified in the preliminary hazards analysis. The interim safety analysis provides an evaluation of the operations in the Solid Waste Burial Grounds to determine if the radiological and hazardous material exposures will be acceptable from an overall health and safety standpoint to the worker, the onsite personnel, the public, and the environment.

  18. Addendum Guidelines for the Prevention of Peanut Allergy in the United States: Report of the National Institute of Allergy and Infectious Diseases-Sponsored Expert Panel.

    Science.gov (United States)

    Togias, Alkis; Cooper, Susan F; Acebal, Maria L; Assa'ad, Amal; Baker, James R; Beck, Lisa A; Block, Julie; Byrd-Bredbenner, Carol; Chan, Edmond S; Eichenfield, Lawrence F; Fleischer, David M; Fuchs, George J; Furuta, Glenn T; Greenhawt, Matthew J; Gupta, Ruchi S; Habich, Michele; Jones, Stacie M; Keaton, Kari; Muraro, Antonella; Plaut, Marshall; Rosenwasser, Lanny J; Rotrosen, Daniel; Sampson, Hugh A; Schneider, Lynda C; Sicherer, Scott H; Sidbury, Robert; Spergel, Jonathan; Stukus, David R; Venter, Carina; Boyce, Joshua A

    Food allergy is an important public health problem because it affects children and adults, can be severe and even life-threatening, and may be increasing in prevalence. Beginning in 2008, the National Institute of Allergy and Infectious Diseases, working with other organizations and advocacy groups, led the development of the first clinical guidelines for the diagnosis and management of food allergy. A recent landmark clinical trial and other emerging data suggest that peanut allergy can be prevented through introduction of peanut-containing foods beginning in infancy. Prompted by these findings, along with 25 professional organizations, federal agencies, and patient advocacy groups, the National Institute of Allergy and Infectious Diseases facilitated development of addendum guidelines to specifically address the prevention of peanut allergy. The addendum provides 3 separate guidelines for infants at various risk levels for the development of peanut allergy and is intended for use by a wide variety of health care providers. Topics addressed include the definition of risk categories, appropriate use of testing (specific IgE measurement, skin prick tests, and oral food challenges), and the timing and approaches for introduction of peanut-containing foods in the health care provider's office or at home. The addendum guidelines provide the background, rationale, and strength of evidence for each recommendation. Guidelines have been developed for early introduction of peanut-containing foods into the diets of infants at various risk levels for peanut allergy. Published by Elsevier Inc.

  19. Safety analysis procedures for PHWR

    International Nuclear Information System (INIS)

    Min, Byung Joo; Kim, Hyoung Tae; Yoo, Kun Joong

    2004-03-01

    The methodology of safety analyses for CANDU reactors in Canada, a vendor country, uses a combination of best-estimate physical models and conservative input parameters so as to minimize the uncertainty of the plant behavior predictions. As using the conservative input parameters, the results of the safety analyses are assured the regulatory requirements such as the public dose, the integrity of fuel and fuel channel, the integrity of containment and reactor structures, etc. However, there is not the comprehensive and systematic procedures for safety analyses for CANDU reactors in Korea. In this regard, the development of the safety analyses procedures for CANDU reactors is being conducted not only to establish the safety analyses system, but also to enhance the quality assurance of the safety assessment. In the first phase of this study, the general procedures of the deterministic safety analyses are developed. The general safety procedures are covered the specification of the initial event, selection of the methodology and accident sequences, computer codes, safety analysis procedures, verification of errors and uncertainties, etc. Finally, These general procedures of the safety analyses are applied to the Large Break Loss Of Coolant Accident (LBLOCA) in Final Safety Analysis Report (FSAR) for Wolsong units 2, 3, 4

  20. Addendum to Technical Proposal: A Facility to Search for Hidden Particles (SHiP) at the CERN SPS

    CERN Document Server

    SHiP Collaboration

    2015-01-01

    With the Technical Proposal submitted to the SPSC committee in April 2015, the SHiP collaboration declared its interest in proceeding towards a Comprehensive Design Study phase with the aim of preparing for the Technical Design Reports pending an approval by the CERN committees. Following the recommendation by the SPSC, it has been decided to complement the TP with this addendum that provides an update of the key aspects for the review of the SHiP project.

  1. SEISMIC ANALYSIS FOR PRECLOSURE SAFETY

    Energy Technology Data Exchange (ETDEWEB)

    E.N. Lindner

    2004-12-03

    The purpose of this seismic preclosure safety analysis is to identify the potential seismically-initiated event sequences associated with preclosure operations of the repository at Yucca Mountain and assign appropriate design bases to provide assurance of achieving the performance objectives specified in the Code of Federal Regulations (CFR) 10 CFR Part 63 for radiological consequences. This seismic preclosure safety analysis is performed in support of the License Application for the Yucca Mountain Project. In more detail, this analysis identifies the systems, structures, and components (SSCs) that are subject to seismic design bases. This analysis assigns one of two design basis ground motion (DBGM) levels, DBGM-1 or DBGM-2, to SSCs important to safety (ITS) that are credited in the prevention or mitigation of seismically-initiated event sequences. An application of seismic margins approach is also demonstrated for SSCs assigned to DBGM-2 by showing a high confidence of a low probability of failure at a higher ground acceleration value, termed a beyond-design basis ground motion (BDBGM) level. The objective of this analysis is to meet the performance requirements of 10 CFR 63.111(a) and 10 CFR 63.111(b) for offsite and worker doses. The results of this calculation are used as inputs to the following: (1) A classification analysis of SSCs ITS by identifying potential seismically-initiated failures (loss of safety function) that could lead to undesired consequences; (2) An assignment of either DBGM-1 or DBGM-2 to each SSC ITS credited in the prevention or mitigation of a seismically-initiated event sequence; and (3) A nuclear safety design basis report that will state the seismic design requirements that are credited in this analysis. The present analysis reflects the design information available as of October 2004 and is considered preliminary. The evolving design of the repository will be re-evaluated periodically to ensure that seismic hazards are properly

  2. SEISMIC ANALYSIS FOR PRECLOSURE SAFETY

    International Nuclear Information System (INIS)

    E.N. Lindner

    2004-01-01

    The purpose of this seismic preclosure safety analysis is to identify the potential seismically-initiated event sequences associated with preclosure operations of the repository at Yucca Mountain and assign appropriate design bases to provide assurance of achieving the performance objectives specified in the Code of Federal Regulations (CFR) 10 CFR Part 63 for radiological consequences. This seismic preclosure safety analysis is performed in support of the License Application for the Yucca Mountain Project. In more detail, this analysis identifies the systems, structures, and components (SSCs) that are subject to seismic design bases. This analysis assigns one of two design basis ground motion (DBGM) levels, DBGM-1 or DBGM-2, to SSCs important to safety (ITS) that are credited in the prevention or mitigation of seismically-initiated event sequences. An application of seismic margins approach is also demonstrated for SSCs assigned to DBGM-2 by showing a high confidence of a low probability of failure at a higher ground acceleration value, termed a beyond-design basis ground motion (BDBGM) level. The objective of this analysis is to meet the performance requirements of 10 CFR 63.111(a) and 10 CFR 63.111(b) for offsite and worker doses. The results of this calculation are used as inputs to the following: (1) A classification analysis of SSCs ITS by identifying potential seismically-initiated failures (loss of safety function) that could lead to undesired consequences; (2) An assignment of either DBGM-1 or DBGM-2 to each SSC ITS credited in the prevention or mitigation of a seismically-initiated event sequence; and (3) A nuclear safety design basis report that will state the seismic design requirements that are credited in this analysis. The present analysis reflects the design information available as of October 2004 and is considered preliminary. The evolving design of the repository will be re-evaluated periodically to ensure that seismic hazards are properly

  3. Safety balance: Analysis of safety systems; Bilans de surete: analyse par les organismes de surete

    Energy Technology Data Exchange (ETDEWEB)

    Delage, M; Giroux, C

    1990-12-01

    Safety analysis, and particularly analysis of exploitation of NPPs is constantly affected by EDF and by the safety authorities and their methodologies. Periodic safety reports ensure that important issues are not missed on daily basis, that incidents are identified and that relevant actions are undertaken. French safety analysis method consists of three principal steps. First type of safety balance is analyzed at the normal start-up phase for each unit including the final safety report. This enables analysis of behaviour of units ten years after their licensing. Second type is periodic operational safety analysis performed during a few years. Finally, the third step consists of safety analysis of the oldest units with the aim to improve the safety standards. The three steps of safety analysis are described in this presentation in detail with the aim to present the objectives and principles. Examples of most recent exercises are included in order to illustrate the importance of such analyses.

  4. Advanced Photon Source experimental beamline Safety Assessment Document: Addendum to the Advanced Photon Source Accelerator Systems Safety Assessment Document (APS-3.2.2.1.0)

    International Nuclear Information System (INIS)

    1995-01-01

    This Safety Assessment Document (SAD) addresses commissioning and operation of the experimental beamlines at the Advanced Photon Source (APS). Purpose of this document is to identify and describe the hazards associated with commissioning and operation of these beamlines and to document the measures taken to minimize these hazards and mitigate the hazard consequences. The potential hazards associated with the commissioning and operation of the APS facility have been identified and analyzed. Physical and administrative controls mitigate identified hazards. No hazard exists in this facility that has not been previously encountered and successfully mitigated in other accelerator and synchrotron radiation research facilities. This document is an updated version of the APS Preliminary Safety Analysis Report (PSAR). During the review of the PSAR in February 1990, the APS was determined to be a Low Hazard Facility. On June 14, 1993, the Acting Director of the Office of Energy Research endorsed the designation of the APS as a Low Hazard Facility, and this Safety Assessment Document supports that designation

  5. SIMMER as a safety analysis tool

    International Nuclear Information System (INIS)

    Smith, L.L.; Bell, C.R.; Bohl, W.R.; Bott, T.F.; Dearing, J.F.; Luck, L.B.

    1982-01-01

    SIMMER has been used for numerous applications in fast reactor safety, encompassing both accident and experiment analysis. Recent analyses of transition-phase behavior in potential core disruptive accidents have integrated SIMMER testing with the accident analysis. Results of both the accident analysis and the verification effort are presented as a comprehensive safety analysis program

  6. The Role of the Inspector during an Emergency. Addendum to the Inspection of Emergency Arrangements [NEA/CNRA/R(2013)2

    International Nuclear Information System (INIS)

    Veyret, Olivier; Kupila, Jukka

    2015-06-01

    A fundamental goal in a regulatory authority's oversight of nuclear facilities is to establish confidence that adequate arrangements are in place for emergency preparedness and response. The CNRA published a report in 1998 on the inspection of licensee activities in emergency planning [NEA/CNRA/R(98)2]. Significant changes have occurred in the interim period, including early lessons learnt from the Fukushima accident. Therefore, the WGIP determined that it would be prudent for member countries to share good inspection practices associated with emergency arrangements (preparedness and response) in an updated report. In December 2013, the WGIP therefore presented the report entitled Inspection of Emergency Arrangements [NEA/CNRA/R(2013)2] for CNRA approval. While the report was approved at the December 2013 CNRA meeting, members requested that a short addendum be developed to compile member countries experiences of the role of the inspector during an emergency. The present addendum to NEA/CNRA/R(2013)2 identifies six additional commendable practices. The WGIP generated a questionnaire to address aspects of the role of the inspector during an emergency. Details of the questions posed to WGIP members are set out in Section 3 of this report. The following twenty countries responded to the questionnaire: Belgium, Canada, Czech Republic, Finland, France, Germany, India, Italy, Japan, Korea, Mexico, the Netherlands, Russia, Slovak Republic, Slovenia, Spain, Sweden, Switzerland, United Kingdom and United States. The WGIP members were requested to response concisely to the two main questions in the questionnaire. Appendix A provides the detailed responses to the questions. Appendix B includes supplementary information provided by several countries. An analysis of WGIP members' responses is presented in Section 4 of this report and conclusions are presented in Section 5

  7. Development of safety analysis technology for integral reactor

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-03-01

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

  8. Computer aided safety analysis 1989

    International Nuclear Information System (INIS)

    1990-04-01

    The meeting was conducted in a workshop style, to encourage involvement of all participants during the discussions. Forty-five (45) experts from 19 countries, plus 22 experts from the GDR participated in the meeting. A list of participants can be found at the end of this volume. Forty-two (42) papers were presented and discussed during the meeting. Additionally an open discussion was held on the possible directions of the IAEA programme on Computer Aided Safety Analysis. A summary of the conclusions of these discussions is presented in the publication. The remainder of this proceedings volume comprises the transcript of selected technical papers (22) presented in the meeting. It is the intention of the IAEA that the publication of these proceedings will extend the benefits of the discussions held during the meeting to a larger audience throughout the world. The Technical Committee/Workshop on Computer Aided Safety Analysis was organized by the IAEA in cooperation with the National Board for Safety and Radiological Protection (SAAS) of the German Democratic Republic in Berlin. The purpose of the meeting was to provide an opportunity for discussions on experiences in the use of computer codes used for safety analysis of nuclear power plants. In particular it was intended to provide a forum for exchange of information among experts using computer codes for safety analysis under the Technical Cooperation Programme on Safety of WWER Type Reactors (RER/9/004) and other experts throughout the world. A separate abstract was prepared for each of the 22 selected papers. Refs, figs tabs and pictures

  9. Project W-320, waste retrieval sluicing system: BIO/SER implementation matrices

    International Nuclear Information System (INIS)

    Bailey, J.W.

    1998-01-01

    This document provides verification that the safety related commitments specified in HNF-SD-WM-810-001, Addendum 1 for the Waste Retrieval Sluicing System, Project W-320 and Project W-320 Safety Evaluation Report (SER), have been implemented in the project hardware, procedures and administrative controls. Four appendices include matrices which show where the 810 commitments are implemented for limiting conditions of operation and surveillance requirements controls, administrative controls, defense-in-depth controls and controls discussed in 810 Addendum 1. A fifth appendix includes the implementation of Project W-320 SER issues and provisions

  10. Subseabed disposal safety analysis

    International Nuclear Information System (INIS)

    Koplick, C.M.; Kabele, T.J.

    1982-01-01

    This report summarizes the status of work performed by Analytic Sciences Corporation (TASC) in FY'81 on subseabed disposal safety analysis. Safety analysis for subseabed disposal is divided into two phases: pre-emplacement which includes all transportation, handling, and emplacement activities; and long-term (post-emplacement), which is concerned with the potential hazard after waste is safely emplaced. Details of TASC work in these two areas are provided in two technical reports. The work to date, while preliminary, supports the technical and environmental feasibility of subseabed disposal of HLW

  11. 14 CFR 33.75 - Safety analysis.

    Science.gov (United States)

    2010-01-01

    ... 14 Aeronautics and Space 1 2010-01-01 2010-01-01 false Safety analysis. 33.75 Section 33.75... STANDARDS: AIRCRAFT ENGINES Design and Construction; Turbine Aircraft Engines § 33.75 Safety analysis. (a... consequences of all failures that can reasonably be expected to occur. This analysis will take into account, if...

  12. 14 CFR 35.15 - Safety analysis.

    Science.gov (United States)

    2010-01-01

    ... 14 Aeronautics and Space 1 2010-01-01 2010-01-01 false Safety analysis. 35.15 Section 35.15... STANDARDS: PROPELLERS Design and Construction § 35.15 Safety analysis. (a)(1) The applicant must analyze the.... This analysis will take into account, if applicable: (i) The propeller system in a typical installation...

  13. Health and Safety Plan for Waste Area Grouping 6 at Oak Ridge National Laboratory, Oak Ridge, Tennessee

    International Nuclear Information System (INIS)

    Van Hoesen, S.D.; Clark, C. Jr.; Burman, S.N.; Manis, L.W.; Barre, W.L.

    1993-12-01

    The Martin Marietta Energy Systems, Inc. (Energy Systems), policy is to provide a safe and healthful workplace for all employees and subcontractors. The accomplishment of this policy requires that operations at Waste Area Grouping (WAG) 6 at the Department of Energy (DOE) Oak Ridge National Laboratory are guided by an overall plan and consistent proactive approach to safety and health (S ampersand H) issues. The plan is written to utilize past experience and best management practices to minimize hazards to human health or the environment from events such as fires, explosions, falls, mechanical hazards, or any unplanned release of hazardous or radioactive materials to air, soil, or surface water This plan explains additional site-specific health and safety requirements such as Site Specific Hazards Evaluation Addendums (SSHEAs) to the Site Safety and Health Plan which should be used in concert with this plan and existing established procedures

  14. Safety analysis of spent fuel packaging

    International Nuclear Information System (INIS)

    Akamatsu, Hiroshi; Taniuchi, Hiroaki; Tai, Hideto

    1987-01-01

    Many types of spent fuel packagings have been manufactured and been used for transport of spent fuels discharged from nuclear power plant. These spent fuel packagings need to be assesed thoroughly about safety transportation because spent fuels loaded into the packaging have high radioactivity and generation of heat. This paper explains the outline of safety analysis of a packaging, Safety analysis is performed for structural, thermal, containment, shielding and criticality factors, and MARC-CDC, TRUMP, ORIGEN, QAD, ANISN, KENO, etc computer codes are used for such analysis. (author)

  15. Statistical considerations on safety analysis

    International Nuclear Information System (INIS)

    Pal, L.; Makai, M.

    2004-01-01

    The authors have investigated the statistical methods applied to safety analysis of nuclear reactors and arrived at alarming conclusions: a series of calculations with the generally appreciated safety code ATHLET were carried out to ascertain the stability of the results against input uncertainties in a simple experimental situation. Scrutinizing those calculations, we came to the conclusion that the ATHLET results may exhibit chaotic behavior. A further conclusion is that the technological limits are incorrectly set when the output variables are correlated. Another formerly unnoticed conclusion of the previous ATHLET calculations that certain innocent looking parameters (like wall roughness factor, the number of bubbles per unit volume, the number of droplets per unit volume) can influence considerably such output parameters as water levels. The authors are concerned with the statistical foundation of present day safety analysis practices and can only hope that their own misjudgment will be dispelled. Until then, the authors suggest applying correct statistical methods in safety analysis even if it makes the analysis more expensive. It would be desirable to continue exploring the role of internal parameters (wall roughness factor, steam-water surface in thermal hydraulics codes, homogenization methods in neutronics codes) in system safety codes and to study their effects on the analysis. In the validation and verification process of a code one carries out a series of computations. The input data are not precisely determined because measured data have an error, calculated data are often obtained from a more or less accurate model. Some users of large codes are content with comparing the nominal output obtained from the nominal input, whereas all the possible inputs should be taken into account when judging safety. At the same time, any statement concerning safety must be aleatory, and its merit can be judged only when the probability is known with which the

  16. Addendum to engineering evaluation/cost analysis for the proposed management of 15 nonprocess buildings (15 Series) at the Weldon Spring site chemical plant, Weldon Spring, Missouri

    International Nuclear Information System (INIS)

    MacDonell, M.M.; Peterson, J.M.

    1990-08-01

    An engineering evaluation/cost analysis (EE/CA) report was prepared in May 1989 to analyze alternatives for a proposed removal action to manage 15 nonprocess buildings, designated as the 15 Series buildings, at the chemical plant area of the Weldon Spring site. The alternative selected as a result of the analyses was to dismantle the buildings and to salvage or transport off-site for treatment or disposal all nonradioactively contaminated materials and to store on-site in a material staging area (MSA) all radioactively contaminated materials, pending a decision for disposal of all wastes resulting from remediation of the Weldon Spring site. Region VII of the US Environmental Protection Agency (EPA) and the state of Missouri concurred with the selection of this alternative and provided comments on the EE/CA report. The proposed removal action was not initiated at that time due to funding constraints. This addendum has been prepared to update information provided in the EE/CA report, provide additional information on the MSA, and respond to EPA Region VII and state of Missouri comments on the EE/CA. 5 refs., 1 tab

  17. Software FMEA analysis for safety-related application software

    International Nuclear Information System (INIS)

    Park, Gee-Yong; Kim, Dong Hoon; Lee, Dong Young

    2014-01-01

    Highlights: • We develop a modified FMEA analysis suited for applying to software architecture. • A template for failure modes on a specific software language is established. • A detailed-level software FMEA analysis on nuclear safety software is presented. - Abstract: A method of a software safety analysis is described in this paper for safety-related application software. The target software system is a software code installed at an Automatic Test and Interface Processor (ATIP) in a digital reactor protection system (DRPS). For the ATIP software safety analysis, at first, an overall safety or hazard analysis is performed over the software architecture and modules, and then a detailed safety analysis based on the software FMEA (Failure Modes and Effect Analysis) method is applied to the ATIP program. For an efficient analysis, the software FMEA analysis is carried out based on the so-called failure-mode template extracted from the function blocks used in the function block diagram (FBD) for the ATIP software. The software safety analysis by the software FMEA analysis, being applied to the ATIP software code, which has been integrated and passed through a very rigorous system test procedure, is proven to be able to provide very valuable results (i.e., software defects) that could not be identified during various system tests

  18. Work plan addendum for David Witherspoon, Inc., 901 Site Building Characterization, Knoxville, Tennessee

    International Nuclear Information System (INIS)

    1997-01-01

    This building characterization plan was developed as an addendum to the existing site characterization work plan documents, which are in Appendix B of the David Witherspoon, Inc., (DWI) preliminary remedial investigation (RI)/feasibility study (FS). All building characterization activities will be conducted in accordance with the rules of the Hazardous Substance Remedial Action Program under the direction of the Tennessee Department of Environment and Conservation, Division of Superfund (TN Rules 1200-1-3) and its implementing regulations. Additional rules of the state of Tennessee, Comprehensive Environmental Response, Compensation, and Liability Act of 1980, and the U.S. Environmental Protection Agency guidance were consulted during development of this plan. Activities at the DWI site were concerned with scrap metal processing and scrap metal resale

  19. Gas-cooled reactor safety and accident analysis

    International Nuclear Information System (INIS)

    1985-12-01

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

  20. Safety analysis - current and future regulatory challenges

    Energy Technology Data Exchange (ETDEWEB)

    Jamieson, T., E-mail: Terry.Jamieson@cnsc-ccsn.gc.ca [Canadian Nuclear Safety Commission, Ottawa, ON (Canada)

    2015-07-01

    'Full text:' The current and future regulatory challenges associated with deterministic safety analysis are reviewed, including: 1. The CNSC's and safety control areas. 2. Traditional safety analysis approach. 3. Experience gained and impact. 4. Current analysis and regulatory approaches. 5. Current status. 6. Complexity and challenges In particular, the technical, regulatory and strategic aspects of these challenges are discussed. (author)

  1. Safety analysis - current and future regulatory challenges

    International Nuclear Information System (INIS)

    Jamieson, T.

    2015-01-01

    'Full text:' The current and future regulatory challenges associated with deterministic safety analysis are reviewed, including: 1. The CNSC's and safety control areas. 2. Traditional safety analysis approach. 3. Experience gained and impact. 4. Current analysis and regulatory approaches. 5. Current status. 6. Complexity and challenges In particular, the technical, regulatory and strategic aspects of these challenges are discussed. (author)

  2. Development of safety analysis technology for LMR

    International Nuclear Information System (INIS)

    Hahn, Do Hee; Kwon, Y. M.; Kim, K. D.

    2000-05-01

    The analysis methodologies as well as the analysis computer code system for the transient, HCDA, and containment performance analyses, which are required for KALIMER safety analyses, have been developed. The SSC-K code has been developed based on SSC-L which is an analysis code for loop type LMR, by improving models necessary for the KALIMER system analysis, and additional models have been added to the code. In addition, HCDA analysis model has been developed and the containment performance analysis code has been also improved. The preliminary basis for the safety analysis has been established, and the preliminary safety analyses for the key design features have been performed. In addition, a state-of-art analysis for LMR PSA and overseas safety and licensing requirements have been reviewed. The design database for the systematic management of the design documents as well as design processes has been established as well

  3. Development of safety analysis technology for LMR

    Energy Technology Data Exchange (ETDEWEB)

    Hahn, Do Hee; Kwon, Y. M.; Kim, K. D. [and others

    2000-05-01

    The analysis methodologies as well as the analysis computer code system for the transient, HCDA, and containment performance analyses, which are required for KALIMER safety analyses, have been developed. The SSC-K code has been developed based on SSC-L which is an analysis code for loop type LMR, by improving models necessary for the KALIMER system analysis, and additional models have been added to the code. In addition, HCDA analysis model has been developed and the containment performance analysis code has been also improved. The preliminary basis for the safety analysis has been established, and the preliminary safety analyses for the key design features have been performed. In addition, a state-of-art analysis for LMR PSA and overseas safety and licensing requirements have been reviewed. The design database for the systematic management of the design documents as well as design processes has been established as well.

  4. Safety analysis reports - new strategies

    International Nuclear Information System (INIS)

    Booth, J.A.

    1994-01-01

    Within the past year there have been many external changes in the requirements of safety analysis reports. Now there is emphasis on open-quotes graded approachesclose quotes depending on the Hazard Classification of the project. The Energy Facility Contractors Group (EFCOG) has a Safety Analysis Working Group. The results of this group for the past year are discussed as well as the implications for EG ampersand G. New strategies include ideas for incorporating the graded approach, auditable safety documents, additional guidance for Hazard Classification per DOE-STD-1027-92. The emphasis in the paper is on those projects whose hazard classification is category three or less

  5. NASA satellite communications application research, phase 2 addendum. Efficient high power, solid state amplifier for EHF communications

    Science.gov (United States)

    Benet, James

    1994-01-01

    This document is an addendum to the NASA Satellite Communications Application Research (SCAR) Phase 2 Final Report, 'Efficient High Power, Solid State Amplifier for EHF Communications.' This report describes the work performed from 1 August 1993 to 11 March 1994, under contract number NASW-4513. During this reporting period an array of transistor amplifiers was repaired by replacing all MMIC amplifier chips. The amplifier array was then tested using three different feedhorn configurations. Descriptions, procedures, and results of this testing are presented in this report, and conclusions are drawn based on the test results obtained.

  6. Deep Borehole Disposal Safety Analysis.

    Energy Technology Data Exchange (ETDEWEB)

    Freeze, Geoffrey A. [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States); Stein, Emily [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States); Price, Laura L. [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States); MacKinnon, Robert J. [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States); Tillman, Jack Bruce [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States)

    2016-10-01

    This report presents a preliminary safety analysis for the deep borehole disposal (DBD) concept, using a safety case framework. A safety case is an integrated collection of qualitative and quantitative arguments, evidence, and analyses that substantiate the safety, and the level of confidence in the safety, of a geologic repository. This safety case framework for DBD follows the outline of the elements of a safety case, and identifies the types of information that will be required to satisfy these elements. At this very preliminary phase of development, the DBD safety case focuses on the generic feasibility of the DBD concept. It is based on potential system designs, waste forms, engineering, and geologic conditions; however, no specific site or regulatory framework exists. It will progress to a site-specific safety case as the DBD concept advances into a site-specific phase, progressing through consent-based site selection and site investigation and characterization.

  7. National Waste Repository Novi Han operational safety analysis report. Safety assessment methodology

    International Nuclear Information System (INIS)

    2003-01-01

    The scope of the safety assessment (SA), presented includes: waste management functions (acceptance, conditioning, storage, disposal), inventory (current and expected in the future), hazards (radiological and non-radiological) and normal and accidental modes. The stages in the development of the SA are: criteria selection, information collection, safety analysis and safety assessment documentation. After the review the facilities functions and the national and international requirements, the criteria for safety level assessment are set. As a result from the 2nd stage actual parameters of the facility, necessary for safety analysis are obtained.The methodology is selected on the base of the comparability of the results with the results of previous safety assessments and existing standards and requirements. The procedure and requirements for scenarios selection are described. A radiological hazard categorisation of the facilities is presented. Qualitative hazards and operability analysis is applied. The resulting list of events are subjected to procedure for prioritization by method of 'criticality analysis', so the estimation of the risk is given for each event. The events that fall into category of risk on the boundary of acceptability or are unacceptable are subjected to the next steps of the analysis. As a result the lists with scenarios for PSA and possible design scenarios are established. PSA logical modeling and quantitative calculations of accident sequences are presented

  8. GPHS-RTGs in support of the Cassini RTG Program. Addendum to the final technical report, May 1--December 31, 1998

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-12-01

    This Addendum to the Cassini GPHS-RTG Program Final Technical Progress Report describes activities performed during the period 1 May 1998 through 31 December 1998, including effort reflecting contract modification M058. These activities include Earth Gravity Assist (EGA) reentry and related analyses which are detailed in Part A, and effort related to the installation of CAGO equipment within Lockheed Martin`s Building 100 facility in Valley Forge, PA, which is detailed in Part B.

  9. GPHS-RTGs in support of the Cassini RTG Program. Addendum to the final technical report, May 1-December 31, 1998

    International Nuclear Information System (INIS)

    1998-12-01

    This Addendum to the Cassini GPHS-RTG Program Final Technical Progress Report describes activities performed during the period 1 May 1998 through 31 December 1998, including effort reflecting contract modification M058. These activities include Earth Gravity Assist (EGA) reentry and related analyses which are detailed in Part A, and effort related to the installation of CAGO equipment within Lockheed Martin's Building 100 facility in Valley Forge, PA, which is detailed in Part B

  10. NPP Temelin safety analysis reports and PSA status

    International Nuclear Information System (INIS)

    Mlady, O.

    1999-01-01

    To enhance the safety level of Temelin NPP, recommendations of the international reviews were implemented into the design as well as into organization of the plant construction and preparation for operation. The safety assessment of these design changes has been integrated and reflected in the Safety Analysis Reports, which follow the internationally accepted guidelines. All safety analyses within Safety Analysis Reports were repeated carefully considering technical improvements and replacements to complement preliminary safety documentation. These analyses were performed by advanced western computer codes to the depth and in the structure required by western standards. The Temelin NPP followed a systematic approach in the functional design of the Reactor Protection System and related safety analyses. Modifications of reactor protection system increase defense in depth and facilitate demonstrating that LOCA and radiological limits are met for non-LOCA events. The rigorous safety analysis methodology provides assurance that LOCA and radiological limits are met. Established and accepted safety analysis methodology and accepted criteria were applied to Temelin NPP meeting US NRC and Czech Republic requirements. IAEA guidelines and recommendations

  11. System safety engineering analysis handbook

    Science.gov (United States)

    Ijams, T. E.

    1972-01-01

    The basic requirements and guidelines for the preparation of System Safety Engineering Analysis are presented. The philosophy of System Safety and the various analytic methods available to the engineering profession are discussed. A text-book description of each of the methods is included.

  12. Current status of safety analysis report for ANPP

    International Nuclear Information System (INIS)

    Amirjanyan, A.

    1999-01-01

    Current situation concerning Armenian NPP safety analysis report is considered within the frame of accepted safety practice. Licensing procedure is being developed. Technical support group was established in the Armenian Nuclear Regulatory Authority (ANRA). The task of the group is to study modern methods of NPP in depth safety analysis for technical assistance for the ANRA, and perform independent safety assessments. ANRA will be obliged to demand assistance from various foreign organisations for preparation of different parts of the Safety Analysis Report like determination though certain parts can be prepared in Armenia

  13. NKS/SOS-1 seminar on safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Lauridsen, K. [Risoe National Lab., Roskilde (Denmark); Anderson, K. [Karinta-Konsult (Sweden); Pulkkinen, U. [VTT Automation (Finland)

    2001-05-01

    The report describes presentations and discussions at a seminar held at Risoe on March 22-23, 2000. The title of the seminar was NKS/SOS-1 - Safety Analysis. It dealt with issues of relevance for the safety analysis for the entire nuclear safety field (notably reactors and nuclear waste repositories). Such issues were: objectives of safety analysis, risk criteria, decision analysis, expert judgement and risk communication. In addition, one talk dealt with criteria for chemical industries in Europe. The seminar clearly showed that the concept of risk is multidimensional, which makes clarity and transparency essential elements in risk communication, and that there are issues of common concern between different applications, such as how to deal with different kinds of uncertainty and expert judgement. (au)

  14. HANFORD SAFETY ANALYSIS & RISK ASSESSMENT HANDBOOK (SARAH)

    Energy Technology Data Exchange (ETDEWEB)

    EVANS, C B

    2004-12-21

    The purpose of the Hanford Safety Analysis and Risk Assessment Handbook (SARAH) is to support the development of safety basis documentation for Hazard Category 2 and 3 (HC-2 and 3) U.S. Department of Energy (DOE) nuclear facilities to meet the requirements of 10 CFR 830, ''Nuclear Safety Management''. Subpart B, ''Safety Basis Requirements.'' Consistent with DOE-STD-3009-94, Change Notice 2, ''Preparation Guide for U.S. Department of Energy Nonreactor Nuclear Facility Documented Safety Analyses'' (STD-3009), and DOE-STD-3011-2002, ''Guidance for Preparation of Basis for Interim Operation (BIO) Documents'' (STD-3011), the Hanford SARAH describes methodology for performing a safety analysis leading to development of a Documented Safety Analysis (DSA) and derivation of Technical Safety Requirements (TSR), and provides the information necessary to ensure a consistently rigorous approach that meets DOE expectations. The DSA and TSR documents, together with the DOE-issued Safety Evaluation Report (SER), are the basic components of facility safety basis documentation. For HC-2 or 3 nuclear facilities in long-term surveillance and maintenance (S&M), for decommissioning activities, where source term has been eliminated to the point that only low-level, residual fixed contamination is present, or for environmental remediation activities outside of a facility structure, DOE-STD-1120-98, ''Integration of Environment, Safety, and Health into Facility Disposition Activities'' (STD-1120), may serve as the basis for the DSA. HC-2 and 3 environmental remediation sites also are subject to the hazard analysis methodologies of this standard.

  15. Safety analysis reports. Current status (third key report)

    International Nuclear Information System (INIS)

    1999-01-01

    A review of Ukrainian regulations and laws concerned with Nuclear power and radiation safety is presented with an overview of the requirements for the Safety Analysis Report Contents. Status of Safety Analysis Reports (SAR) is listed for each particular Ukrainian NPP including SAR development schedules. Organisational scheme of SAR development works includes: general technical co-ordination on Safety Analysis Report development; list of leading organisations and utilization of technical support within international projects

  16. Status of Ignalina's safety analysis reports

    International Nuclear Information System (INIS)

    Uspuras, E.

    1999-01-01

    Ignalina NPP is unique among RBMK type reactors in the scope and comprehensiveness of international studies which have been performed to verify its design parameters and analyze risk levels. International assistance took several forms, a very valuable mod of assistance utilized the knowledge of international experts in extensive international studies whose purpose was: collection, systematization and verification of plant design data; analysis of risk levels; recommendations leading to improvements in the safety lave; transfer of state of the art analytical methodology to Lithuanian specialists. The major large scale international studies include: probabilistic risk analysis; extensive international study meant to provide comprehensive overview of plant status with special emphasis on safety aspects; an extensive review of the Safety Analysis Report by an independent group of international experts. In spite of the safety improvements and analyses which have been performed at the Ignalina NPP, much remains to be done in the nearest future

  17. Hot Cell Facility (HCF) Safety Analysis Report

    Energy Technology Data Exchange (ETDEWEB)

    MITCHELL,GERRY W.; LONGLEY,SUSAN W.; PHILBIN,JEFFREY S.; MAHN,JEFFREY A.; BERRY,DONALD T.; SCHWERS,NORMAN F.; VANDERBEEK,THOMAS E.; NAEGELI,ROBERT E.

    2000-11-01

    This Safety Analysis Report (SAR) is prepared in compliance with the requirements of DOE Order 5480.23, Nuclear Safety Analysis Reports, and has been written to the format and content guide of DOE-STD-3009-94 Preparation Guide for U. S. Department of Energy Nonreactor Nuclear Safety Analysis Reports. The Hot Cell Facility is a Hazard Category 2 nonreactor nuclear facility, and is operated by Sandia National Laboratories for the Department of Energy. This SAR provides a description of the HCF and its operations, an assessment of the hazards and potential accidents which may occur in the facility. The potential consequences and likelihood of these accidents are analyzed and described. Using the process and criteria described in DOE-STD-3009-94, safety-related structures, systems and components are identified, and the important safety functions of each SSC are described. Additionally, information which describes the safety management programs at SNL are described in ancillary chapters of the SAR.

  18. Approach to uncertainty evaluation for safety analysis

    International Nuclear Information System (INIS)

    Ogura, Katsunori

    2005-01-01

    Nuclear power plant safety used to be verified and confirmed through accident simulations using computer codes generally because it is very difficult to perform integrated experiments or tests for the verification and validation of the plant safety due to radioactive consequence, cost, and scaling to the actual plant. Traditionally the plant safety had been secured owing to the sufficient safety margin through the conservative assumptions and models to be applied to those simulations. Meanwhile the best-estimate analysis based on the realistic assumptions and models in support of the accumulated insights could be performed recently, inducing the reduction of safety margin in the analysis results and the increase of necessity to evaluate the reliability or uncertainty of the analysis results. This paper introduces an approach to evaluate the uncertainty of accident simulation and its results. (Note: This research had been done not in the Japan Nuclear Energy Safety Organization but in the Tokyo Institute of Technology.) (author)

  19. Hot Cell Facility (HCF) Safety Analysis Report

    International Nuclear Information System (INIS)

    MITCHELL, GERRY W.; LONGLEY, SUSAN W.; PHILBIN, JEFFREY S.; MAHN, JEFFREY A.; BERRY, DONALD T.; SCHWERS, NORMAN F.; VANDERBEEK, THOMAS E.; NAEGELI, ROBERT E.

    2000-01-01

    This Safety Analysis Report (SAR) is prepared in compliance with the requirements of DOE Order 5480.23, Nuclear Safety Analysis Reports, and has been written to the format and content guide of DOE-STD-3009-94 Preparation Guide for U. S. Department of Energy Nonreactor Nuclear Safety Analysis Reports. The Hot Cell Facility is a Hazard Category 2 nonreactor nuclear facility, and is operated by Sandia National Laboratories for the Department of Energy. This SAR provides a description of the HCF and its operations, an assessment of the hazards and potential accidents which may occur in the facility. The potential consequences and likelihood of these accidents are analyzed and described. Using the process and criteria described in DOE-STD-3009-94, safety-related structures, systems and components are identified, and the important safety functions of each SSC are described. Additionally, information which describes the safety management programs at SNL are described in ancillary chapters of the SAR

  20. Manpower analysis in transportation safety. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Bauer, C.S.; Bowden, H.M.; Colford, C.A.; DeFilipps, P.J.; Dennis, J.D.; Ehlert, A.K.; Popkin, H.A.; Schrader, G.F.; Smith, Q.N.

    1977-05-01

    The project described provides a manpower review of national, state and local needs for safety skills, and projects future manning levels for transportation safety personnel in both the public and private sectors. Survey information revealed that there are currently approximately 121,000 persons employed directly in transportation safety occupations within the air carrier, highway and traffic safety, motor carrier, pipeline, rail carrier, and marine carrier transportation industry groups. The projected need for 1980 is over 145,000 of which over 80 percent will be in highway safety. An analysis of transportation tasks is included, and shows ten general categories about which the majority of safety activities are focused. A skills analysis shows a generally high level of educational background and several years of experience are required for most transportation safety jobs. An overall review of safety programs in the transportation industry is included, together with chapters on the individual transportation modes.

  1. Safety analysis of a high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Shimazu, Akira; Morimoto, Toshio

    1975-01-01

    In recent years, in order to satisfy the social requirements of environment and safety and also to cope with the current energy stringency, the installation of safe nuclear power plants is indispensable. Herein, safety analysis and evaluation to confirm quantitatively the safety design of a nuclear power plant become more and more important. The safety analysis and its methods for a high temperature gas-cooled reactor are described, with emphasis placed on the practices by Fuji Electric Manufacturing Co. Fundamental rule of securing plant safety ; safety analysis in normal operation regarding plant dynamic characteristics and radioactivity evaluation ; and safety analysis at the time of accidents regarding plant response to the accidents and radioactivity evaluation are explained. (Mori, K.)

  2. Safety analysis SFR 1. Long-term safety

    Energy Technology Data Exchange (ETDEWEB)

    2008-12-15

    An updated assessment of the long-term safety of SKB's final repository for radioactive operational waste, SFR 1, is presented in this report. The report is included in the safety analysis report for SFR 1. The most recent account of long-term safety was submitted to the regulatory authorities in 2001. The present report has been compiled on SKB's initiative to address the regulatory authorities' viewpoints regarding the preceding account of long-term safety. Besides the new mode of working with safety functions there is another important difference between the 2001 safety assessment and the current assessment: The time horizon in the current assessment has been extended to 100,000 years in order to include the effect of future climate changes. The purpose of this renewed assessment of the long-term safety of SFR 1 is to show with improved data that the repository is capable of protecting human health and the environment against ionizing radiation in a long-term perspective. This is done by showing that calculated risks lie below the risk criteria stipulated by the regulatory authorities. SFR 1 is built to receive, and after closure serve as a passive repository for, low. and intermediate-level radioactive waste. The disposal chambers are situated in rock beneath the sea floor, covered by about 60 metres of rock. The underground part of the facility is reached via two tunnels whose entrances are near the harbour. The repository has been designed so that it can be abandoned after closure without further measures needing to be taken to maintain its function. The waste in SFR 1 is short-lived low- and intermediate-level waste. After 100 years the activity is less than half, and after 1,000 years only about 2% of the original activity remains. The report on long-term safety comprises eleven chapters. Chapter 1 Introduction. The chapter describes the purpose, background, format and contents of SAR-08, applicable regulations and injunctions, and the regulatory

  3. Safety analysis SFR 1. Long-term safety

    International Nuclear Information System (INIS)

    2008-12-01

    An updated assessment of the long-term safety of SKB's final repository for radioactive operational waste, SFR 1, is presented in this report. The report is included in the safety analysis report for SFR 1. The most recent account of long-term safety was submitted to the regulatory authorities in 2001. The present report has been compiled on SKB's initiative to address the regulatory authorities' viewpoints regarding the preceding account of long-term safety. Besides the new mode of working with safety functions there is another important difference between the 2001 safety assessment and the current assessment: The time horizon in the current assessment has been extended to 100,000 years in order to include the effect of future climate changes. The purpose of this renewed assessment of the long-term safety of SFR 1 is to show with improved data that the repository is capable of protecting human health and the environment against ionizing radiation in a long-term perspective. This is done by showing that calculated risks lie below the risk criteria stipulated by the regulatory authorities. SFR 1 is built to receive, and after closure serve as a passive repository for, low. and intermediate-level radioactive waste. The disposal chambers are situated in rock beneath the sea floor, covered by about 60 metres of rock. The underground part of the facility is reached via two tunnels whose entrances are near the harbour. The repository has been designed so that it can be abandoned after closure without further measures needing to be taken to maintain its function. The waste in SFR 1 is short-lived low- and intermediate-level waste. After 100 years the activity is less than half, and after 1,000 years only about 2% of the original activity remains. The report on long-term safety comprises eleven chapters. Chapter 1 Introduction. The chapter describes the purpose, background, format and contents of SAR-08, applicable regulations and injunctions, and the regulatory

  4. Safety analysis SFR 1. Long-term safety

    Energy Technology Data Exchange (ETDEWEB)

    2008-12-15

    An updated assessment of the long-term safety of SKB's final repository for radioactive operational waste, SFR 1, is presented in this report. The report is included in the safety analysis report for SFR 1. The most recent account of long-term safety was submitted to the regulatory authorities in 2001. The present report has been compiled on SKB's initiative to address the regulatory authorities' viewpoints regarding the preceding account of long-term safety. Besides the new mode of working with safety functions there is another important difference between the 2001 safety assessment and the current assessment: The time horizon in the current assessment has been extended to 100,000 years in order to include the effect of future climate changes. The purpose of this renewed assessment of the long-term safety of SFR 1 is to show with improved data that the repository is capable of protecting human health and the environment against ionizing radiation in a long-term perspective. This is done by showing that calculated risks lie below the risk criteria stipulated by the regulatory authorities. SFR 1 is built to receive, and after closure serve as a passive repository for, low. and intermediate-level radioactive waste. The disposal chambers are situated in rock beneath the sea floor, covered by about 60 metres of rock. The underground part of the facility is reached via two tunnels whose entrances are near the harbour. The repository has been designed so that it can be abandoned after closure without further measures needing to be taken to maintain its function. The waste in SFR 1 is short-lived low- and intermediate-level waste. After 100 years the activity is less than half, and after 1,000 years only about 2% of the original activity remains. The report on long-term safety comprises eleven chapters. Chapter 1 Introduction. The chapter describes the purpose, background, format and contents of SAR-08, applicable regulations and injunctions, and the

  5. Health and Safety Plan for Waste Area Grouping 6 at Oak Ridge National Laboratory, Oak Ridge, Tennessee. Environmental Restoration Program

    Energy Technology Data Exchange (ETDEWEB)

    Van Hoesen, S.D.; Clark, C. Jr.; Burman, S.N. [Oak Ridge National Lab., TN (United States); Manis, L.W.; Barre, W.L. [Analysas Corp., Oak Ridge, TN (United States)

    1993-12-01

    The Martin Marietta Energy Systems, Inc. (Energy Systems), policy is to provide a safe and healthful workplace for all employees and subcontractors. The accomplishment of this policy requires that operations at Waste Area Grouping (WAG) 6 at the Department of Energy (DOE) Oak Ridge National Laboratory are guided by an overall plan and consistent proactive approach to safety and health (S&H) issues. The plan is written to utilize past experience and best management practices to minimize hazards to human health or the environment from events such as fires, explosions, falls, mechanical hazards, or any unplanned release of hazardous or radioactive materials to air, soil, or surface water This plan explains additional site-specific health and safety requirements such as Site Specific Hazards Evaluation Addendums (SSHEAs) to the Site Safety and Health Plan which should be used in concert with this plan and existing established procedures.

  6. Applications of probabilistic risk analysis in nuclear criticality safety design

    International Nuclear Information System (INIS)

    Chang, J.K.

    1992-01-01

    Many documents have been prepared that try to define the scope of the criticality analysis and that suggest adding probabilistic risk analysis (PRA) to the deterministic safety analysis. The report of the US Department of Energy (DOE) AL 5481.1B suggested that an accident is credible if the occurrence probability is >1 x 10 -6 /yr. The draft DOE 5480 safety analysis report suggested that safety analyses should include the application of methods such as deterministic safety analysis, risk assessment, reliability engineering, common-cause failure analysis, human reliability analysis, and human factor safety analysis techniques. The US Nuclear Regulatory Commission (NRC) report NRC SG830.110 suggested that major safety analysis methods should include but not be limited to risk assessment, reliability engineering, and human factor safety analysis. All of these suggestions have recommended including PRA in the traditional criticality analysis

  7. Software safety hazard analysis

    International Nuclear Information System (INIS)

    Lawrence, J.D.

    1996-02-01

    Techniques for analyzing the safety and reliability of analog-based electronic protection systems that serve to mitigate hazards in process control systems have been developed over many years, and are reasonably well understood. An example is the protection system in a nuclear power plant. The extension of these techniques to systems which include digital computers is not well developed, and there is little consensus among software engineering experts and safety experts on how to analyze such systems. One possible technique is to extend hazard analysis to include digital computer-based systems. Software is frequently overlooked during system hazard analyses, but this is unacceptable when the software is in control of a potentially hazardous operation. In such cases, hazard analysis should be extended to fully cover the software. A method for performing software hazard analysis is proposed in this paper

  8. Infusing Reliability Techniques into Software Safety Analysis

    Science.gov (United States)

    Shi, Ying

    2015-01-01

    Software safety analysis for a large software intensive system is always a challenge. Software safety practitioners need to ensure that software related hazards are completely identified, controlled, and tracked. This paper discusses in detail how to incorporate the traditional reliability techniques into the entire software safety analysis process. In addition, this paper addresses how information can be effectively shared between the various practitioners involved in the software safety analyses. The author has successfully applied the approach to several aerospace applications. Examples are provided to illustrate the key steps of the proposed approach.

  9. Incorporation of advanced accident analysis methodology into safety analysis reports

    International Nuclear Information System (INIS)

    2003-05-01

    The IAEA Safety Guide on Safety Assessment and Verification defines that the aim of the safety analysis should be by means of appropriate analytical tools to establish and confirm the design basis for the items important to safety, and to ensure that the overall plant design is capable of meeting the prescribed and acceptable limits for radiation doses and releases for each plant condition category. Practical guidance on how to perform accident analyses of nuclear power plants (NPPs) is provided by the IAEA Safety Report on Accident Analysis for Nuclear Power Plants. The safety analyses are performed both in the form of deterministic and probabilistic analyses for NPPs. It is customary to refer to deterministic safety analyses as accident analyses. This report discusses the aspects of using the advanced accident analysis methods to carry out accident analyses in order to introduce them into the Safety Analysis Reports (SARs). In relation to the SAR, purposes of deterministic safety analysis can be further specified as (1) to demonstrate compliance with specific regulatory acceptance criteria; (2) to complement other analyses and evaluations in defining a complete set of design and operating requirements; (3) to identify and quantify limiting safety system set points and limiting conditions for operation to be used in the NPP limits and conditions; (4) to justify appropriateness of the technical solutions employed in the fulfillment of predetermined safety requirements. The essential parts of accident analyses are performed by applying sophisticated computer code packages, which have been specifically developed for this purpose. These code packages include mainly thermal-hydraulic system codes and reactor dynamics codes meant for the transient and accident analyses. There are also specific codes such as those for the containment thermal-hydraulics, for the radiological consequences and for severe accident analyses. In some cases, codes of a more general nature such

  10. Software safety analysis practice in installation phase

    Energy Technology Data Exchange (ETDEWEB)

    Huang, H. W.; Chen, M. H.; Shyu, S. S., E-mail: hwhwang@iner.gov.t [Institute of Nuclear Energy Research, No. 1000 Wenhua Road, Chiaan Village, Longtan Township, 32546 Taoyuan County, Taiwan (China)

    2010-10-15

    This work performed a software safety analysis in the installation phase of the Lung men nuclear power plant in Taiwan, under the cooperation of Institute of Nuclear Energy Research and Tpc. The US Nuclear Regulatory Commission requests licensee to perform software safety analysis and software verification and validation in each phase of software development life cycle with Branch Technical Position 7-14. In this work, 37 safety grade digital instrumentation and control systems were analyzed by failure mode and effects analysis, which is suggested by IEEE standard 7-4.3.2-2003. During the installation phase, skew tests for safety grade network and point to point tests were performed. The failure mode and effects analysis showed all the single failure modes can be resolved by the redundant means. Most of the common mode failures can be resolved by operator manual actions. (Author)

  11. Software safety analysis practice in installation phase

    International Nuclear Information System (INIS)

    Huang, H. W.; Chen, M. H.; Shyu, S. S.

    2010-10-01

    This work performed a software safety analysis in the installation phase of the Lung men nuclear power plant in Taiwan, under the cooperation of Institute of Nuclear Energy Research and Tpc. The US Nuclear Regulatory Commission requests licensee to perform software safety analysis and software verification and validation in each phase of software development life cycle with Branch Technical Position 7-14. In this work, 37 safety grade digital instrumentation and control systems were analyzed by failure mode and effects analysis, which is suggested by IEEE standard 7-4.3.2-2003. During the installation phase, skew tests for safety grade network and point to point tests were performed. The failure mode and effects analysis showed all the single failure modes can be resolved by the redundant means. Most of the common mode failures can be resolved by operator manual actions. (Author)

  12. Guidelines for nuclear reactor equipments safety-analysis

    International Nuclear Information System (INIS)

    1978-01-01

    The safety analysis in approving the applications for nuclear reactor constructions (or alterations) is performed by the Committee on Examination of Reactor Safety in accordance with various guidelines prescribed by the Atomic Energy Commission. In addition, the above Committee set forth its own regulations for the safety analysis on common problems among various types of nuclear reactors. This book has collected and edited those guidelines and regulations. It has two parts: Part I includes the guidelines issued to date by the Atomic Energy Commission: and Part II - regulations of the Committee. Part I has collected 8 categories of guidelines which relate to following matters: nuclear reactor sites analysis guidelines and standards for their applications; standard exposure dose of plutonium; nuclear ship operation guidelines; safety design analysis guidelines for light-water type, electricity generating nuclear reactor equipments; safety evaluation guidelines for emergency reactor core cooling system of light-water type power reactors; guidelines for exposure dose target values around light-water type electricity generating nuclear reactor equipments, and guidelines for evaluation of above target values; and meteorological guidelines for the safety analysis of electricity generating nuclear reactor equipments. Part II includes regulations of the Committee concerning - the fuel assembly used in boiling-water type and in pressurized-water type reactors; techniques of reactor core heat designs, etc. in boiling-water reactors; and others

  13. Preliminary Integrated Safety Analysis Status Report

    International Nuclear Information System (INIS)

    Gwyn, D.

    2001-01-01

    This report provides the status of the potential Monitored Geologic Repository (MGR) Integrated Safety Analysis (EA) by identifying the initial work scope scheduled for completion during the ISA development period, the schedules associated with the tasks identified, safety analysis issues encountered, and a summary of accomplishments during the reporting period. This status covers the period from October 1, 2000 through March 30, 2001

  14. A formal safety analysis for PLC software-based safety critical system using Z

    International Nuclear Information System (INIS)

    Koh, Jung Soo; Seong, Poong Hyun

    1997-01-01

    This paper describes a formal safety analysis technique which is demonstrated by performing empirical formal safety analysis with the case study of beamline hutch door Interlock system that is developed by using PLC (Programmable Logic Controller) systems at the Pohang Accelerator Laboratory. In order to perform formed safety analysis, we have built the Z formal specifications representation from user requirement written in ambiguous natural language and target PLC ladder logic, respectively. We have also studied the effective method to express typical PLC timer component by using specific Z formal notation which is supported by temporal history. We present a formal proof technique specifying and verifying that the hazardous states are not introduced into ladder logic in the PLC-based safety critical system

  15. A formal safety analysis for PLC software-based safety critical system using Z

    International Nuclear Information System (INIS)

    Koh, Jung Soo

    1997-02-01

    This paper describes a formal safety analysis technique which is demonstrated by performing empirical formal safety analysis with the case study of beamline hutch door Interlock system that is developed by using PLC (Programmable Logic Controller) systems at the Pohang Accelerator Laboratory. In order to perform formal safety analysis, we have built the Z formal specifications representation from user requirement written in ambiguous natural language and target PLC ladder logic, respectively. We have also studied the effective method to express typical PLC timer component by using specific Z formal notation which is supported by temporal history. We present a formal proof technique specifying and verifying that the hazardous states are not introduced into ladder logic in the PLC-based safety critical system. And also, we have found that some errors or mismatches in user requirement and final implemented PLC ladder logic while analyzing the process of the consistency and completeness of Z translated formal specifications. In the case of relatively small systems like Beamline hutch door interlock system, a formal safety analysis including explicit proof is highly recommended so that the safety of PLC-based critical system may be enhanced and guaranteed. It also provides a helpful benefits enough to comprehend user requirement expressed by ambiguous natural language

  16. An overview-probabilistic safety analysis for research reactors

    International Nuclear Information System (INIS)

    Liu Jinlin; Peng Changhong

    2015-01-01

    For long-term application, Probabilistic Safety Analysis (PSA) has proved to be a valuable tool for improving the safety and reliability of power reactors. In China, 'Nuclear safety and radioactive pollution prevention 'Twelfth Five Year Plan' and the 2020 vision' raises clearly that: to develop probabilistic safety analysis and aging evaluation for research reactors. Comparing with the power reactors, it reveals some specific features in research reactors: lower operating power, lower coolant temperature and pressure, etc. However, the core configurations may be changed very often and human actions play an important safety role in research reactors due to its specific experimental requirement. As a result, there is a necessary to conduct the PSA analysis of research reactors. This paper discusses the special characteristics related to the structure and operation and the methods to develop the PSA of research reactors, including initiating event analysis, event tree analysis, fault tree analysis, dependent failure analysis, human reliability analysis and quantification as well as the experimental and external event evaluation through the investigation of various research reactors and their PSAs home and abroad, to provide the current situation and features of research reactors PSAs. (author)

  17. Ignalina Safety Analysis Group's report for the year 1998

    International Nuclear Information System (INIS)

    Uspuras, E.; Augutis, J.; Bubelis, E.; Cesna, B.; Kaliatka, A.

    1999-02-01

    Results of Ignalina NPP Safety Analysis Group's research are presented. The main fields of group's activities in 1998 were following: safety analysis of reactor's cooling system, safety analysis of accident localization system, investigation of the problem graphite - fuel channel, reactor core modelling, assistance to the regulatory body VATESI in drafting regulations and reviewing safety reports presented by Ignalina NPP during the process of licensing of unit 1

  18. 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

  19. Computer codes for safety analysis

    International Nuclear Information System (INIS)

    Holland, D.F.

    1986-11-01

    Computer codes for fusion safety analysis have been under development in the United States for about a decade. This paper will discuss five codes that are currently under development by the Fusion Safety Program. The purpose and capability of each code will be presented, a sample given, followed by a discussion of the present status and future development plans

  20. Removing unreasonable conservatisms in DOE safety analysis

    International Nuclear Information System (INIS)

    BISHOP, G.E.

    1999-01-01

    While nuclear safety analyses must always be conservative, invoking excessive conservatisms does not provide additional margins of safety. Rather, beyond a fairly narrow point, conservatisms skew a facility's true safety envelope by exaggerating risks and creating unreasonable bounds on what is required for safety. The conservatism has itself become unreasonable. A thorough review of the assumptions and methodologies contained in a facility's safety analysis can provide substantial reward, reducing both construction and operational costs without compromising actual safety

  1. HANFORD SAFETY ANALYSIS and RISK ASSESSMENT HANDBOOK (SARAH)

    International Nuclear Information System (INIS)

    EVANS, C.B.

    2004-01-01

    The purpose of the Hanford Safety Analysis and Risk Assessment Handbook (SARAH) is to support the development of safety basis documentation for Hazard Category 2 and 3 (HC-2 and 3) U.S. Department of Energy (DOE) nuclear facilities to meet the requirements of 10 CFR 830, ''Nuclear Safety Management''. Subpart B, ''Safety Basis Requirements.'' Consistent with DOE-STD-3009-94, Change Notice 2, ''Preparation Guide for U.S. Department of Energy Nonreactor Nuclear Facility Documented Safety Analyses'' (STD-3009), and DOE-STD-3011-2002, ''Guidance for Preparation of Basis for Interim Operation (BIO) Documents'' (STD-3011), the Hanford SARAH describes methodology for performing a safety analysis leading to development of a Documented Safety Analysis (DSA) and derivation of Technical Safety Requirements (TSR), and provides the information necessary to ensure a consistently rigorous approach that meets DOE expectations. The DSA and TSR documents, together with the DOE-issued Safety Evaluation Report (SER), are the basic components of facility safety basis documentation. For HC-2 or 3 nuclear facilities in long-term surveillance and maintenance (S and M), for decommissioning activities, where source term has been eliminated to the point that only low-level, residual fixed contamination is present, or for environmental remediation activities outside of a facility structure, DOE-STD-1120-98, ''Integration of Environment, Safety, and Health into Facility Disposition Activities'' (STD-1120), may serve as the basis for the DSA. HC-2 and 3 environmental remediation sites also are subject to the hazard analysis methodologies of this standard

  2. Safety assessment of research reactors and preparation of the safety analysis report

    International Nuclear Information System (INIS)

    1994-01-01

    This Safety Guide presents guidelines, approved by international consensus, for the preparation, review and assessment of safety documentation for research reactors such as the Safety Analysis Report. While the Guide is most applicable to research reactors in the design and construction stage, it is also recommended for use during relicensing or reassessment of existing reactors

  3. Site Safety Plan for Lawrence Livermore National Laboratory CERCLA investigations

    Energy Technology Data Exchange (ETDEWEB)

    Bainer, R.; Duarte, J.

    1993-07-01

    The safety policy of LLNL is to take every reasonable precaution in the performance of work to protect the environment and the health and safety of employees and the public, and to prevent property damage. With respect to hazardous agents, this protection is provided by limiting human exposures, releases to the environment, and contamination of property to levels that are as low as reasonably achievable (ALARA). It is the intent of this Plan to supply the broad outline for completing environmental investigations within ALARA guidelines. It may not be possible to determine actual working conditions in advance of the work; therefore, planning must allow the opportunity to provide a range of protection based upon actual working conditions. Requirements will be the least restrictive possible for a given set of circumstances, such that work can be completed in an efficient and timely fashion. Due to the relatively large size of the LLNL Site and the different types of activities underway, site-specific Operational Safety Procedures (OSPs) will be prepared to supplement activities not covered by this Plan. These site-specific OSPs provide the detailed information for each specific activity and act as an addendum to this Plan, which provides the general plan for LLNL Main Site operation.

  4. Application of Software Safety Analysis Methods

    International Nuclear Information System (INIS)

    Park, G. Y.; Hur, S.; Cheon, S. W.; Kim, D. H.; Lee, D. Y.; Kwon, K. C.; Lee, S. J.; Koo, Y. H.

    2009-01-01

    A fully digitalized reactor protection system, which is called the IDiPS-RPS, was developed through the KNICS project. The IDiPS-RPS has four redundant and separated channels. Each channel is mainly composed of a group of bistable processors which redundantly compare process variables with their corresponding setpoints and a group of coincidence processors that generate a final trip signal when a trip condition is satisfied. Each channel also contains a test processor called the ATIP and a display and command processor called the COM. All the functions were implemented in software. During the development of the safety software, various software safety analysis methods were applied, in parallel to the verification and validation (V and V) activities, along the software development life cycle. The software safety analysis methods employed were the software hazard and operability (Software HAZOP) study, the software fault tree analysis (Software FTA), and the software failure modes and effects analysis (Software FMEA)

  5. Nuclear safety in Slovak Republic. Safety analysis reports for WWER 440 reactors

    International Nuclear Information System (INIS)

    Rohar, S.

    1999-01-01

    Implementation of nuclear power program is connected to establishment of regulatory body for safe regulation of siting, construction, operation and decommissioning of nuclear installations. Licensing being one of the most important regulatory surveillance activity is based on independent regulatory review and assessment of information on nuclear safety for particular nuclear facility. Documents required to be submitted to the regulatory body by the licensee in Slovakia for the review and assessment usually named Safety Analysis Report (SAR) are presented in detail in this paper. Current status of Safety Analysis Reports for Bohunice V-1, Bohunice V-2 and Mochovce NPP is shown

  6. From Safety Analysis to Formal Specification

    DEFF Research Database (Denmark)

    Hansen, Kirsten Mark; Ravn, Anders P.; Stavridou, Victoria

    1998-01-01

    Software for safety critical systems must deal with the hazards identified bysafety analysis. This paper investigates, how the results of onesafety analysis technique, fault trees, are interpreted as software safetyrequirements to be used in the program design process. We propose thatfault tree...... analysis and program development use the samesystem model. This model is formalized in areal-time, interval logic, based on a conventional dynamic systems modelwith state evolving over time. Fault trees are interpreted astemporal formulas, and it is shown how such formulas can be usedfor deriving safety...

  7. Coeficiente de corrección en engranajes cilíndricos como factor de conversión entre sistemas AGMA e ISO. // Addendum modification coefficient as conversion factor of cylindrical gears in AGMA system to ISO standards.

    Directory of Open Access Journals (Sweden)

    G. González Rey

    2007-09-01

    Full Text Available A partir de la experiencia de los autores, en trabajos asociados con el análisis para la reconversión de engranajes ensistemas AGMA a partir de herramientas con referencia a normas ISO, en el artículo se pretende exponer las principalesdefiniciones y recomendaciones asociadas al coeficiente de corrección relativo al módulo en engranajes cilíndricos de ejesparalelos con dientes de flanco de evolvente que permitan un racional cálculo de la geometría del engranaje y apreciarmediante ejemplos la ventaja del coeficiente de corrección en la solución de problemas prácticos de conversión deengranajes en sistemas AGMA a ISO.Palabras claves: geometría de engranaje cilíndrico, coeficiente de corrección, conversión AGMA a ISO________________________________________________________________________________AbstractDefinitions and recommendations associated to the addendum modification coefficient are presented. Examples, based onauthor’s experiences in the analysis, recovery and conversion of helical and spur gears in AGMA system to ISO standards showthe advantage of application of the addendum modification coefficient in the solution of practical problems and rationalcalculation of gears.Key words: external parallel-axis cylindrical involute gears, gear geometry, , ISO gear standards, rack shift coefficient x

  8. Business of Nuclear Safety Analysis Office, Nuclear Technology Test Center

    International Nuclear Information System (INIS)

    Hayakawa, Masahiko

    1981-01-01

    The Nuclear Technology Test Center established the Nuclear Safety Analysis Office to execute newly the works concerning nuclear safety analysis in addition to the works related to the proving tests of nuclear machinery and equipments. The regulations for the Nuclear Safety Analysis Office concerning its organization, business and others were specially decided, and it started the business formally in August, 1980. It is a most important subject to secure the safety of nuclear facilities in nuclear fuel cycle as the premise of developing atomic energy. In Japan, the strict regulation of safety is executed by the government at each stage of the installation, construction, operation and maintenance of nuclear facilities, based on the responsibility for the security of installers themselves. The Nuclear Safety Analysis Office was established as the special organ to help the safety examination related to the installation of nuclear power stations and others by the government. It improves and puts in order the safety analysis codes required for the cross checking in the safety examination, and carries out safety analysis calculation. It is operated by the cooperation of the Science and Technology Agency and the Agency of Natural Resources and Energy. The purpose of establishment, the operation and the business of the Nuclear Safety Analysis Office, the plan of improving and putting in order of analysis codes, and the state of the similar organs in foreign countries are described. (Kako, I.)

  9. Establishment of Safety Analysis System and Technology for CANDU Reactors

    International Nuclear Information System (INIS)

    Park, Joo Hwan; Rhee, B. W.; Min, B. J.; Kim, H. T.; Kim, W. Y.; Yoon, C.; Chun, J. S.; Cho, M. S.; Jeong, J. Y.; Kang, H. S.

    2007-06-01

    The following 4 research items have been studied to establish a CANDU safety analysis system and to develop the relevant elementary technology for CANDU reactors. First, to improve and validate the CANDU design and operational safety analysis codes, the CANDU physics cell code WIMS-CANDU was improved, and validated, and an analysis of the moderator subcooling and pressure tube integrity has been performed for the large break LOCAs without ECCS. Also a CATHENA model and a CFD model for a post-blowdown fuel channel analysis have been developed and validated against two high temperature thermal-chemical experiments, CS28-1 and 2. Second, to improve the integrated operating system of the CANDU safety analysis codes, an extension has been made to them to include the core and fuel accident analyses, and a web-based CANDU database, CANTHIS version 2.0 was completed. Third, to assess the applicability of the ACR-7 safety analysis methodology to CANDU-6 the ACR-7 safety analysis methods were reviewed and the safety analysis methods of ACR-7 applicable to CANDU-6 were recommended. Last, to supplement and improve the existing CANDU safety analysis procedures, detailed analysis procedures have been prepared for individual accident scenarios. The results of this study can be used to resolve the CANDU safety issues, to improve the current design and operational safety analysis codes, and to technically support the Wolsong site to resolve their problems

  10. Establishment of Safety Analysis System and Technology for CANDU Reactors

    International Nuclear Information System (INIS)

    Min, Byung Joo; Kim, W. Y.; Kim, H. T.; Rhee, B. W.; Yoon, C.; Kang, H. S.; Yoo, K. J.

    2005-03-01

    To improve the CANDU design/operation safety analysis codes and the CANDU safety analysis methodology, the following works have been done. From the development of the lattice codes (WIMS/CANDU), the lattice model simulates the real core lattice geometry and the effect of the pressure tube creep to the core lattice parameter has been evaluated. From the development of the 3-dimensional thermal-hydraulic analysis model of the moderator behavior (CFX4-CAMO), validation of the model against STERN Lab experiment has been executed. The butterfly-shaped grid structure and the 3-dimensional flow resistance model for porous media were developed and applied to the moderator analysis for Wolsong units 2/3/4. The single fuel channel analysis codes for blowdown and post-blowdown were unified by CATHENA. The 3-dimensional fuel channel analysis model (CFX-CACH) has been developed for validation of CATHENA fuel channel analysis model. The interlinking analysis system (CANVAS) of the thermal-hydraulic safety analysis codes for the primary heat transport system and containment system has been executed. The database system of core physics and thermal-hydraulics experimental data for safety analysis has been established on the URL: http://CANTHIS.kaeri.re.kr. For documentation and Standardization of the general safety analysis procedure, the general safety analysis procedure is developed and applied to a large break LOCA. The present research results can be utilized for establishment of the independent safety analysis technology and acquisition of the optimal safety analysis technology

  11. Ignalina NPP Safety Analysis: Models and Results

    International Nuclear Information System (INIS)

    Uspuras, E.

    1999-01-01

    Research directions, linked to safety assessment of the Ignalina NPP, of the scientific safety analysis group are presented: Thermal-hydraulic analysis of accidents and operational transients; Thermal-hydraulic assessment of Ignalina NPP Accident Localization System and other compartments; Structural analysis of plant components, piping and other parts of Main Circulation Circuit; Assessment of RBMK-1500 reactor core and other. Models and main works carried out last year are described. (author)

  12. Reload safety analysis automation tools

    International Nuclear Information System (INIS)

    Havlůj, F.; Hejzlar, J.; Vočka, R.

    2013-01-01

    Performing core physics calculations for the sake of reload safety analysis is a very demanding and time consuming process. This process generally begins with the preparation of libraries for the core physics code using a lattice code. The next step involves creating a very large set of calculations with the core physics code. Lastly, the results of the calculations must be interpreted, correctly applying uncertainties and checking whether applicable limits are satisfied. Such a procedure requires three specialized experts. One must understand the lattice code in order to correctly calculate and interpret its results. The next expert must have a good understanding of the physics code in order to create libraries from the lattice code results and to correctly define all the calculations involved. The third expert must have a deep knowledge of the power plant and the reload safety analysis procedure in order to verify, that all the necessary calculations were performed. Such a procedure involves many steps and is very time consuming. At ÚJV Řež, a.s., we have developed a set of tools which can be used to automate and simplify the whole process of performing reload safety analysis. Our application QUADRIGA automates lattice code calculations for library preparation. It removes user interaction with the lattice code and reduces his task to defining fuel pin types, enrichments, assembly maps and operational parameters all through a very nice and user-friendly GUI. The second part in reload safety analysis calculations is done by CycleKit, a code which is linked with our core physics code ANDREA. Through CycleKit large sets of calculations with complicated interdependencies can be performed using simple and convenient notation. CycleKit automates the interaction with ANDREA, organizes all the calculations, collects the results, performs limit verification and displays the output in clickable html format. Using this set of tools for reload safety analysis simplifies

  13. Software safety analysis application in installation phase

    International Nuclear Information System (INIS)

    Huang, H. W.; Yih, S.; Wang, L. H.; Liao, B. C.; Lin, J. M.; Kao, T. M.

    2010-01-01

    This work performed a software safety analysis (SSA) in the installation phase of the Lungmen nuclear power plant (LMNPP) in Taiwan, under the cooperation of INER and TPC. The US Nuclear Regulatory Commission (USNRC) requests licensee to perform software safety analysis (SSA) and software verification and validation (SV and V) in each phase of software development life cycle with Branch Technical Position (BTP) 7-14. In this work, 37 safety grade digital instrumentation and control (I and C) systems were analyzed by Failure Mode and Effects Analysis (FMEA), which is suggested by IEEE Standard 7-4.3.2-2003. During the installation phase, skew tests for safety grade network and point to point tests were performed. The FMEA showed all the single failure modes can be resolved by the redundant means. Most of the common mode failures can be resolved by operator manual actions. (authors)

  14. Software safety analysis techniques for developing safety critical software in the digital protection system of the LMR

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jang Soo; Cheon, Se Woo; Kim, Chang Hoi; Sim, Yun Sub

    2001-02-01

    This report has described the software safety analysis techniques and the engineering guidelines for developing safety critical software to identify the state of the art in this field and to give the software safety engineer a trail map between the code and standards layer and the design methodology and documents layer. We have surveyed the management aspects of software safety activities during the software lifecycle in order to improve the safety. After identifying the conventional safety analysis techniques for systems, we have surveyed in details the software safety analysis techniques, software FMEA(Failure Mode and Effects Analysis), software HAZOP(Hazard and Operability Analysis), and software FTA(Fault Tree Analysis). We have also surveyed the state of the art in the software reliability assessment techniques. The most important results from the reliability techniques are not the specific probability numbers generated, but the insights into the risk importance of software features. To defend against potential common-mode failures, high quality, defense-in-depth, and diversity are considered to be key elements in digital I and C system design. To minimize the possibility of CMFs and thus increase the plant reliability, we have provided D-in-D and D analysis guidelines.

  15. Software safety analysis techniques for developing safety critical software in the digital protection system of the LMR

    International Nuclear Information System (INIS)

    Lee, Jang Soo; Cheon, Se Woo; Kim, Chang Hoi; Sim, Yun Sub

    2001-02-01

    This report has described the software safety analysis techniques and the engineering guidelines for developing safety critical software to identify the state of the art in this field and to give the software safety engineer a trail map between the code and standards layer and the design methodology and documents layer. We have surveyed the management aspects of software safety activities during the software lifecycle in order to improve the safety. After identifying the conventional safety analysis techniques for systems, we have surveyed in details the software safety analysis techniques, software FMEA(Failure Mode and Effects Analysis), software HAZOP(Hazard and Operability Analysis), and software FTA(Fault Tree Analysis). We have also surveyed the state of the art in the software reliability assessment techniques. The most important results from the reliability techniques are not the specific probability numbers generated, but the insights into the risk importance of software features. To defend against potential common-mode failures, high quality, defense-in-depth, and diversity are considered to be key elements in digital I and C system design. To minimize the possibility of CMFs and thus increase the plant reliability, we have provided D-in-D and D analysis guidelines

  16. Status of SPACE Safety Analysis Code Development

    International Nuclear Information System (INIS)

    Lee, Dong Hyuk; Yang, Chang Keun; Kim, Se Yun; Ha, Sang Jun

    2009-01-01

    In 2006, the Korean the Korean nuclear industry started developing a thermal-hydraulic analysis code for safety analysis of PWR(Pressurized Water Reactor). The new code is named as SPACE(Safety and Performance Analysis Code for Nuclear Power Plant). The SPACE code can solve two-fluid, three-field governing equations in one dimensional or three dimensional geometry. The SPACE code has many component models required for modeling a PWR, such as reactor coolant pump, safety injection tank, etc. The programming language used in the new code is C++, for new generation of engineers who are more comfortable with C/C++ than old FORTRAN language. This paper describes general characteristics of SPACE code and current status of SPACE code development

  17. Systems Analysis of NASA Aviation Safety Program: Final Report

    Science.gov (United States)

    Jones, Sharon M.; Reveley, Mary S.; Withrow, Colleen A.; Evans, Joni K.; Barr, Lawrence; Leone, Karen

    2013-01-01

    A three-month study (February to April 2010) of the NASA Aviation Safety (AvSafe) program was conducted. This study comprised three components: (1) a statistical analysis of currently available civilian subsonic aircraft data from the National Transportation Safety Board (NTSB), the Federal Aviation Administration (FAA), and the Aviation Safety Information Analysis and Sharing (ASIAS) system to identify any significant or overlooked aviation safety issues; (2) a high-level qualitative identification of future safety risks, with an assessment of the potential impact of the NASA AvSafe research on the National Airspace System (NAS) based on these risks; and (3) a detailed, top-down analysis of the NASA AvSafe program using an established and peer-reviewed systems analysis methodology. The statistical analysis identified the top aviation "tall poles" based on NTSB accident and FAA incident data from 1997 to 2006. A separate examination of medical helicopter accidents in the United States was also conducted. Multiple external sources were used to develop a compilation of ten "tall poles" in future safety issues/risks. The top-down analysis of the AvSafe was conducted by using a modification of the Gibson methodology. Of the 17 challenging safety issues that were identified, 11 were directly addressed by the AvSafe program research portfolio.

  18. Status of generic actions items and safety analysis system of PHWR

    Energy Technology Data Exchange (ETDEWEB)

    Park, Joo Hwan; Min, Byung Joo

    2001-05-01

    This report described the review results of a GAIs(Generic Action Item) currently issued on safety analysis of PHWR(Pressurized Heavy Water Reactor) and the research activities and positions to solve the GAIs in each country which possess PHWRs. eviewing the Final Safety Analysis Report for Wolsong-2/3/4 Units, the safety analysis methodology, classification for accident scenarios, safety analysis codes, their interface, etc.. were described. From the present review report, it is intended to establish the CANDU safety analysis system by providing the better understandings and development plans for the safety analysis of PHWR. esults.

  19. Addendum to the Closure Report for Corrective Action Unit 411: Double Tracks Plutonium Dispersion (Nellis), Nevada Test and Training Range, Nevada, Revision 0

    Energy Technology Data Exchange (ETDEWEB)

    Burmeister, Mark [Navarro, Las Vegas, NV (United States)

    2016-11-01

    The Corrective Action Unit (CAU) 411 Closure Report (CR) was published in June 2016 (NNSA/NFO, 2016). The purpose of this addendum is to clarify language in the CR relating to the field instrument for the detection of low-energy radiation (FIDLER), provide the waste disposal documentation for waste generated during the corrective action investigation (CAI), and reference a letter from the U.S. Air Force (USAF) regarding the closure of CAU 411.

  20. 2005 dossier: granite. Tome: safety analysis of the geologic disposal

    International Nuclear Information System (INIS)

    2005-01-01

    This document makes a status of the researches carried out by the French national agency of radioactive wastes (ANDRA) about the safety aspects of the geologic disposal of high-level and long-lived (HLLL) radioactive wastes in granite formations. Content: 1 - safety approach: context and general goal, references, design approach by safety functions, safety approach during the construction-exploitation-observation-closure phase, safety analysis during the post-closure phase; 2 - general description: HLLL wastes, granitic environment, general structure of the architecture of a disposal facility; 3 - safety functions and disposal design: general context, safety functions of the long-term disposal, design dispositions retained to answer the functions; 4 - operational safety: people's protection, radiological risks during exploitation, risk analysis in accident situation; 5 - qualitative safety analysis: methodology, main results of the analysis of the features, events and processes (FEP) database; 6 - disposal efficiency evaluation during post-closure phase: calculation models, calculation tools used for the modeling of radionuclides transport, calculation results and main lessons. (J.S.)

  1. Integrated framework for dynamic safety analysis

    International Nuclear Information System (INIS)

    Kim, Tae Wan; Karanki, Durga R.

    2012-01-01

    In the conventional PSA (Probabilistic Safety Assessment), detailed plant simulations by independent thermal hydraulic (TH) codes are used in the development of accident sequence models. Typical accidents in a NPP involve complex interactions among process, safety systems, and operator actions. As independent TH codes do not have the models of operator actions and full safety systems, they cannot literally simulate the integrated and dynamic interactions of process, safety systems, and operator responses. Offline simulation with pre decided states and time delays may not model the accident sequences properly. Moreover, when stochastic variability in responses of accident models is considered, defining all the combinations for simulations will be cumbersome task. To overcome some of these limitations of conventional safety analysis approach, TH models are coupled with the stochastic models in the dynamic event tree (DET) framework, which provides flexibility to model the integrated response due to better communication as all the accident elements are in the same model. The advantages of this framework also include: Realistic modeling in dynamic scenarios, comprehensive results, integrated approach (both deterministic and probabilistic models), and support for HRA (Human Reliability Analysis)

  2. OASIS: An automotive analysis and safety engineering instrument

    International Nuclear Information System (INIS)

    Mader, Roland; Armengaud, Eric; Grießnig, Gerhard; Kreiner, Christian; Steger, Christian; Weiß, Reinhold

    2013-01-01

    In this paper, we describe a novel software tool named OASIS (AutOmotive Analysis and Safety EngIneering InStrument). OASIS supports automotive safety engineering with features allowing the creation of consistent and complete work products and to simplify and automate workflow steps from early analysis through system development to software development. More precisely, it provides support for (a) model creation and reuse, (b) analysis and documentation and (c) configuration and code generation. We present OASIS as a part of a tool chain supporting the application of a safety engineering workflow aligned with the automotive safety standard ISO 26262. In particular, we focus on OASIS' (1) support for property checking and model correction as well as its (2) support for fault tree generation and FMEA (Failure Modes and Effects Analysis) table generation. Finally, based on the case study of hybrid electric vehicle development, we demonstrate that (1) and (2) are able to strongly support FTA (Fault Tree Analysis) and FMEA

  3. PWR core safety analysis with 3-dimensional methods

    International Nuclear Information System (INIS)

    Gensler, A.; Kühnel, K.; Kuch, S.

    2015-01-01

    Highlights: • An overview of AREVA’s safety analysis codes their coupling is provided. • The validation base and licensing applications of these codes are summarized. • Coupled codes and methods provide improved margins and non-conservative results. • Examples for REA and inadvertent opening of the pressurizer safety valve are given. - Abstract: The main focus of safety analysis is to demonstrate the required safety level of the reactor core. Because of the demanding requirements, the quality of the safety analysis strongly affects the confidence in the operational safety of a reactor. To ensure the highest quality, it is essential that the methodology consists of appropriate analysis tools, an extensive validation base, and last but not least highly educated engineers applying the methodology. The sophisticated 3-dimensional core models applied by AREVA ensure that all physical effects relevant for safety are treated and the results are reliable and conservative. Presently AREVA employs SCIENCE, CASMO/NEMO and CASCADE-3D for pressurized water reactors. These codes are currently being consolidated into the next generation 3D code system ARCADIA®. AREVA continuously extends the validation base, including measurement campaigns in test facilities and comparisons of the predictions of steady state and transient measured data gathered from plants during many years of operation. Thus, the core models provide reliable and comprehensive results for a wide range of applications. For the application of these powerful tools, AREVA is taking benefit of its interdisciplinary know-how and international teamwork. Experienced engineers of different technical backgrounds are working together to ensure an appropriate interpretation of the calculation results, uncertainty analysis, along with continuously maintaining and enhancing the quality of the analysis methodologies. In this paper, an overview of AREVA’s broad application experience as well as the broad validation

  4. Safety analysis of the UTSI-CFFF superconducting magnet

    International Nuclear Information System (INIS)

    Turner, L.R.; Wang, S.T.; Smith, R.P.; VanderArend, P.C.; Hsu, Y.H.

    1979-01-01

    In designing a large superconducting magnet such as the UTSI-CFFF dipole, great attention must be devoted to the safety of the magnet and personnel. The conductor for the UTSI-CFFF magnet incorporates much copper stabilizer, which both insures its cryostability, and contributes to the magnet safety. The quench analysis and the cryostat fault condition analysis are presented. Two analyses of exposed turns follow; the first shows that gas cooling protects uncovered turns; the second, that the cryostat pressure relief system protects them. Finally the failure mode and safety analysis is presented

  5. Health and Safety Management Plan for the Plutonium Stabilization and Packaging System

    International Nuclear Information System (INIS)

    1996-01-01

    This Health and Safety Management Plan (HSMP) presents safety and health policies and a project health and safety organizational structure designed to minimize potential risks of harm to personnel performing activities associated with Plutonium Stabilization and Packaging System (Pu SPS). The objectives of the Pu SPS are to design, fabricate, install, and startup of a glovebox system for the safe repackaging of plutonium oxides and metals, with a requirement of a 50-year storage period. This HSMP is intended as an initial project health and safety submittal as part of a three phase effort to address health and safety issues related to personnel working the Pu SPS project. Phase 1 includes this HSMP and sets up the basic approach to health and safety on the project and addresses health and safety issues related to the engineering and design effort. Phase 2 will include the Site Specific Construction health and Safety Plan (SSCHSP). Phase 3 will include an additional addendum to this HSMP and address health and safety issues associated with the start up and on-site test phase of the project. This initial submittal of the HSMP is intended to address those activities anticipated to be performed during phase 1 of the project. This HSMP is intended to be a living document which shall be modified as information regarding the individual tasks associated with the project becomes available. These modifications will be in the form of addenda to be submitted prior to the initiation of each phase of the project. For additional work authorized under this project this HSMP will be modified as described in section 1.4

  6. Reliability Analysis for Safety Grade PLC(POSAFE-Q)

    International Nuclear Information System (INIS)

    Choi, Kyung Chul; Song, Seung Whan; Park, Gang Min; Hwang, Sung Jae

    2012-01-01

    Safety Grade PLC(Programmable Logic Controller), POSAFE-Q, was developed recently in accordance with nuclear regulatory and requirements. In this paper, describe reliability analysis for digital safety grade PLC (especially POSAFE-Q). Reliability analysis scope is Prediction, Calculation of MTBF (Mean Time Between Failure), FMEA (Failure Mode Effect Analysis), PFD (Probability of Failure on Demand). (author)

  7. Status of safety analysis reports

    Energy Technology Data Exchange (ETDEWEB)

    Cserhati, A

    1999-06-01

    The safety regulation connected to both of the Atomic Acts from 1980 and 1996 requires preparation of the Preliminary Safety Analysis Report (PSAR) as well as Final SAR (FSAR). In this respect the licensing procedure for the construction and commissioning of Paks NPP did not formally deviate from the standards applied in developed countries; this is particularly true if comparison is made with the standards applied for commissioning NPPs in the second half of the seventies. By the time the overall development of internationally accepted safety standards and some existing deficiencies of earlier SAR made necessary a general reassessment of the plant safety (AGNES project). The carried out PSR for Paks-1 and 2 also added a valuable contribution to the SAR content, however a formal update of SAR is not made yet. A Hungarian nuclear authority decree from 1997 obligates the licensee to prepare and submit a major upgrade of FSAR until the mid of 2000, after finishing the PSR for Paks-3 and 4. From this date a periodic update of FSAR is required every year. The operational license renewal affects only the PSR but not the FSAR updating. The new Nuclear Safety Code outlines the contents of PSAR and FSAR, based on US NRC Reg. Guide 1. 70. Rev. 3. Hungary by now can fulfill the upgrading of SAR without major external technical or financial help. The AGNES project covered the safety analysis chapters of SAR. It was financed mainly by the country. In the project there have been involved in limited cases as performers the VTT (Finland), Belgatom (Belgium), GRS (Germany), etc., the IVO (Finland) fulfilled tasks of an independent reviewer for safety analysis. The AGNES had certain interconnection with the similar IAEA RER safety reassessment project for WWER-440/213. The PSR for Paks-1 and 2 have been carried out by the Paks staff from the resources of the plant. During the evaluation of several parts of Paks-3 and 4 PSR documentation the authority intends to use certain

  8. Status of safety analysis reports

    International Nuclear Information System (INIS)

    Cserhati, A.

    1999-01-01

    The safety regulation connected to both of the Atomic Acts from 1980 and 1996 requires preparation of the Preliminary Safety Analysis Report (PSAR) as well as Final SAR (FSAR). In this respect the licensing procedure for the construction and commissioning of Paks NPP did not formally deviate from the standards applied in developed countries; this is particularly true if comparison is made with the standards applied for commissioning NPPs in the second half of the seventies. By the time the overall development of internationally accepted safety standards and some existing deficiencies of earlier SAR made necessary a general reassessment of the plant safety (AGNES project). The carried out PSR for Paks-1 and 2 also added a valuable contribution to the SAR content, however a formal update of SAR is not made yet. A Hungarian nuclear authority decree from 1997 obligates the licensee to prepare and submit a major upgrade of FSAR until the mid of 2000, after finishing the PSR for Paks-3 and 4. From this date a periodic update of FSAR is required every year. The operational license renewal affects only the PSR but not the FSAR updating. The new Nuclear Safety Code outlines the contents of PSAR and FSAR, based on US NRC Reg. Guide 1. 70. Rev. 3. Hungary by now can fulfill the upgrading of SAR without major external technical or financial help. The AGNES project covered the safety analysis chapters of SAR. It was financed mainly by the country. In the project there have been involved in limited cases as performers the VTT (Finland), Belgatom (Belgium), GRS (Germany), etc., the IVO (Finland) fulfilled tasks of an independent reviewer for safety analysis. The AGNES had certain interconnection with the similar IAEA RER safety reassessment project for WWER-440/213. The PSR for Paks-1 and 2 have been carried out by the Paks staff from the resources of the plant. During the evaluation of several parts of Paks-3 and 4 PSR documentation the authority intends to use certain

  9. Safety analysis of the nuclear chemistry Building 151

    International Nuclear Information System (INIS)

    Kvam, D.

    1984-01-01

    This report summarizes the results of a safety analysis that was done on Building 151. The report outlines the methodology, the analysis, and the findings that led to the low hazard classification. No further safety evaluation is indicated at this time. 5 tables

  10. DESIGN PACKAGE 1E SYSTEM SAFETY ANALYSIS

    Energy Technology Data Exchange (ETDEWEB)

    M. Salem

    1995-06-23

    The purpose of this analysis is to systematically identify and evaluate hazards related to the Yucca Mountain Project Exploratory Studies Facility (ESF) Design Package 1E, Surface Facilities, (for a list of design items included in the package 1E system safety analysis see section 3). This process is an integral part of the systems engineering process; whereby safety is considered during planning, design, testing, and construction. A largely qualitative approach was used since a radiological System Safety Analysis is not required. The risk assessment in this analysis characterizes the accident scenarios associated with the Design Package 1E structures/systems/components(S/S/Cs) in terms of relative risk and includes recommendations for mitigating all identified risks. The priority for recommending and implementing mitigation control features is: (1) Incorporate measures to reduce risks and hazards into the structure/system/component design, (2) add safety devices and capabilities to the designs that reduce risk, (3) provide devices that detect and warn personnel of hazardous conditions, and (4) develop procedures and conduct training to increase worker awareness of potential hazards, on methods to reduce exposure to hazards, and on the actions required to avoid accidents or correct hazardous conditions.

  11. The influence of sodium fires on LMFBRs safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Justin, F [DSN/Centre de Fontenay-aux-Roses, Fontenay-aux-Roses (France)

    1979-03-01

    In a sodium cooled reactor, sodium fires are accidental conditions to be taken into account in safety analysis. For the various sodium categories, fire conditions, associated risks, safety analysis objectives and detailed corresponding issues are indicated, An experimental research program can be deduced from these considerations. This report covers the following: safety analysis methodology; primary sodium fires; secondary sodium fires; auxiliary sodium fires, and related experimental research programs.

  12. The influence of sodium fires on LMFBRs safety analysis

    International Nuclear Information System (INIS)

    Justin, F.

    1979-01-01

    In a sodium cooled reactor, sodium fires are accidental conditions to be taken into account in safety analysis. For the various sodium categories, fire conditions, associated risks, safety analysis objectives and detailed corresponding issues are indicated, An experimental research program can be deduced from these considerations. This report covers the following: safety analysis methodology; primary sodium fires; secondary sodium fires; auxiliary sodium fires, and related experimental research programs

  13. Meta-analysis of surgical safety checklist effects on teamwork, communication, morbidity, mortality, and safety.

    Science.gov (United States)

    Lyons, Vanessa E; Popejoy, Lori L

    2014-02-01

    The purpose of this study is to examine the effectiveness of surgical safety checklists on teamwork, communication, morbidity, mortality, and compliance with safety measures through meta-analysis. Four meta-analyses were conducted on 19 studies that met the inclusion criteria. The effect size of checklists on teamwork and communication was 1.180 (p = .003), on morbidity and mortality was 0.123 (p = .003) and 0.088 (p = .001), respectively, and on compliance with safety measures was 0.268 (p teamwork and communication, reduce morbidity and mortality, and improve compliance with safety measures. This meta-analysis is limited in its generalizability based on the limited number of studies and the inclusion of only published research. Future research is needed to examine possible moderating variables for the effects of surgical safety checklists.

  14. Reliability analysis of software based safety functions

    International Nuclear Information System (INIS)

    Pulkkinen, U.

    1993-05-01

    The methods applicable in the reliability analysis of software based safety functions are described in the report. Although the safety functions also include other components, the main emphasis in the report is on the reliability analysis of software. The check list type qualitative reliability analysis methods, such as failure mode and effects analysis (FMEA), are described, as well as the software fault tree analysis. The safety analysis based on the Petri nets is discussed. The most essential concepts and models of quantitative software reliability analysis are described. The most common software metrics and their combined use with software reliability models are discussed. The application of software reliability models in PSA is evaluated; it is observed that the recent software reliability models do not produce the estimates needed in PSA directly. As a result from the study some recommendations and conclusions are drawn. The need of formal methods in the analysis and development of software based systems, the applicability of qualitative reliability engineering methods in connection to PSA and the need to make more precise the requirements for software based systems and their analyses in the regulatory guides should be mentioned. (orig.). (46 refs., 13 figs., 1 tab.)

  15. Transmittal of the Calculation Package that Supports the Analysis of Performance of the Environmental Management Waste Management Facility Oak Ridge, Tennessee (Based 5-Cell Design Issued 8/14/09)

    Energy Technology Data Exchange (ETDEWEB)

    Williams M.J.

    2009-09-14

    This document presents the results of an assessment of the performance of a build-out of the Environmental Management Waste Management Facility (EMWMF). The EMWMF configuration that was assessed includes the as-constructed Cells 1 through 4, with a groundwater underdrain that was installed beneath Cell 3 during the winter of 2003-2004, and Cell 5, whose proposed design is an Addendum to Remedial Design Report for the Disposal of Oak Ridge Reservation Comprehensive Environmental Response, Compensation, and Liability Act of 1980 Waste, Oak Ridge, Tennessee, DOE/OR/01-1873&D2/A5/R1. The total capacity of the EMWMF with 5 cells is about 1.7 million cubic yards. This assessment was conducted to determine the conditions under which the approved Waste Acceptance Criteria (WAC) for the EMWMF found in the Attainment Plan for Risk/Toxicity-Based Waste Acceptance Criteria at the Oak Ridge Reservation, Oak Ridge, Tennessee [U.S. Department of Energy (DOE) 2001a], as revised for constituents added up to October 2008, would remain protective of public health and safety for a five-cell disposal facility. For consistency, the methods of analyses and the exposure scenario used to predict the performance of a five-cell disposal facility were identical to those used in the Remedial Investigation and Feasibility Study (RI/FS) and its addendum (DOE 1998a, DOE 1998b) to develop the approved WAC. To take advantage of new information and design changes departing from the conceptual design, the modeling domain and model calibration were upaded from those used in the RI/FS and its addendum. It should be noted that this analysis is not intended to justify or propose a change in the approved WAC.

  16. Project health and safety plan for the Gunite and Associated Tanks at Oak Ridge National Laboratory, Oak Ridge, Tennessee

    International Nuclear Information System (INIS)

    Abston, J.P.

    1997-04-01

    The Lockheed Martin Energy Systems, Inc. (Energy Systems) policy is to provide a safe and healthful workplace for all employees and subcontractors. The accomplishment of this policy requires that operations at the Gunite and Associated Tanks (GAAT) in the North and South Tank Farms (NTF and STF) at the Department of Energy (DOE) Oak Ridge National Laboratory are guided by an overall plan and consistent proactive approach to health and safety (H and S) issues. The policy and procedures in this plan apply to all GAAT operations in the NTF and STF. The provisions of this plan are to be carried out whenever activities identifies s part of the GAAT are initiated that could be a threat to human health or the environment. This plan implements a policy and establishes criteria for the development of procedures for day-to-day operations to prevent or minimize any adverse impact to the environment and personnel safety and health and to meet standards that define acceptable management of hazardous and radioactive materials and wastes. The plan is written to utilize past experience and best management practices in order to minimize hazards to human health or the environment from events such as fires, explosions, falls, mechanical hazards, or any unplanned release of hazardous or radioactive materials to the air. This plan explains additional task-specific health and safety requirements such as the Site Safety and health Addendum and Activity Hazard Analysis, which should be used in concert with this plan and existing established procedures

  17. Project health and safety plan for the Gunite and Associated Tanks at Oak Ridge National Laboratory, Oak Ridge, Tennessee

    Energy Technology Data Exchange (ETDEWEB)

    Abston, J.P.

    1997-04-01

    The Lockheed Martin Energy Systems, Inc. (Energy Systems) policy is to provide a safe and healthful workplace for all employees and subcontractors. The accomplishment of this policy requires that operations at the Gunite and Associated Tanks (GAAT) in the North and South Tank Farms (NTF and STF) at the Department of Energy (DOE) Oak Ridge National Laboratory are guided by an overall plan and consistent proactive approach to health and safety (H and S) issues. The policy and procedures in this plan apply to all GAAT operations in the NTF and STF. The provisions of this plan are to be carried out whenever activities identifies s part of the GAAT are initiated that could be a threat to human health or the environment. This plan implements a policy and establishes criteria for the development of procedures for day-to-day operations to prevent or minimize any adverse impact to the environment and personnel safety and health and to meet standards that define acceptable management of hazardous and radioactive materials and wastes. The plan is written to utilize past experience and best management practices in order to minimize hazards to human health or the environment from events such as fires, explosions, falls, mechanical hazards, or any unplanned release of hazardous or radioactive materials to the air. This plan explains additional task-specific health and safety requirements such as the Site Safety and health Addendum and Activity Hazard Analysis, which should be used in concert with this plan and existing established procedures.

  18. Issues affecting advanced passive light-water reactor safety analysis

    International Nuclear Information System (INIS)

    Beelman, R.J.; Fletcher, C.D.; Modro, S.M.

    1992-01-01

    Next generation commercial reactor designs emphasize enhanced safety through improved safety system reliability and performance by means of system simplification and reliance on immutable natural forces for system operation. Simulating the performance of these safety systems will be central to analytical safety evaluation of advanced passive reactor designs. Yet the characteristically small driving forces of these safety systems pose challenging computational problems to current thermal-hydraulic systems analysis codes. Additionally, the safety systems generally interact closely with one another, requiring accurate, integrated simulation of the nuclear steam supply system, engineered safeguards and containment. Furthermore, numerical safety analysis of these advanced passive reactor designs wig necessitate simulation of long-duration, slowly-developing transients compared with current reactor designs. The composite effects of small computational inaccuracies on induced system interactions and perturbations over long periods may well lead to predicted results which are significantly different than would otherwise be expected or might actually occur. Comparisons between the engineered safety features of competing US advanced light water reactor designs and analogous present day reactor designs are examined relative to the adequacy of existing thermal-hydraulic safety codes in predicting the mechanisms of passive safety. Areas where existing codes might require modification, extension or assessment relative to passive safety designs are identified. Conclusions concerning the applicability of these codes to advanced passive light water reactor safety analysis are presented

  19. STARS software tool for analysis of reliability and safety

    International Nuclear Information System (INIS)

    Poucet, A.; Guagnini, E.

    1989-01-01

    This paper reports on the STARS (Software Tool for the Analysis of Reliability and Safety) project aims at developing an integrated set of Computer Aided Reliability Analysis tools for the various tasks involved in systems safety and reliability analysis including hazard identification, qualitative analysis, logic model construction and evaluation. The expert system technology offers the most promising perspective for developing a Computer Aided Reliability Analysis tool. Combined with graphics and analysis capabilities, it can provide a natural engineering oriented environment for computer assisted reliability and safety modelling and analysis. For hazard identification and fault tree construction, a frame/rule based expert system is used, in which the deductive (goal driven) reasoning and the heuristic, applied during manual fault tree construction, is modelled. Expert system can explain their reasoning so that the analyst can become aware of the why and the how results are being obtained. Hence, the learning aspect involved in manual reliability and safety analysis can be maintained and improved

  20. Safety analysis methodology for OPR 1000

    International Nuclear Information System (INIS)

    Hwang-Yong, Jun

    2005-01-01

    Full text: Korea Electric Power Research Institute (KEPRI) has been developing inhouse safety analysis methodology based on the delicate codes available to KEPRI to overcome the problems arising from currently used vendor oriented methodologies. For the Loss of Coolant Accident (LOCA) analysis, the KREM (KEPRI Realistic Evaluation Methodology) has been developed based on the RELAP-5 code. The methodology was approved for the Westinghouse 3-loop plants by the Korean regulatory organization and the project to extent the methodology to the Optimized Power Reactor 1000 (OPR1000) has been ongoing since 2001. Also, for the Non-LOCA analysis, the KNAP (Korea Non-LOCA Analysis Package) has been developed using the UNICORN-TM code system. To demonstrate the feasibility of these codes systems and methodologies, some typical cases of the design basis accidents mentioned in the final safety analysis report (FSAR) were analyzed. (author)

  1. Swimming pool reactor reliability and safety analysis

    International Nuclear Information System (INIS)

    Li Zhaohuan

    1997-01-01

    A reliability and safety analysis of Swimming Pool Reactor in China Institute of Atomic Energy is done by use of event/fault tree technique. The paper briefly describes the analysis model, analysis code and main results. Meanwhile it also describes the impact of unassigned operation status on safety, the estimation of effectiveness of defense tactics in maintenance against common cause failure, the effectiveness of recovering actions on the system reliability, the comparison of occurrence frequencies of the core damage by use of generic and specific data

  2. Analysis on safety production in coal mines Henan Province

    Institute of Scientific and Technical Information of China (English)

    KONG Liu-an; ZHANG Wen-yong

    2006-01-01

    Based on the rigorous situation of safety production in coal mines, the paper analyzed the statistical data of recent accidents indexes in Henan's coal mines. Using investigation and comparison analysis methods, a specified analysis on mining conditions, technical facility level, safety input and vocational quality of workers in Henan's coal mines was conducted. The result indicates that there have been existing such main safety production problems as weak safety management, low-level facilities, inadequate safety input and poor vocational quality and so on. Finally it proposes such reference solutions as to establish and perfect coal mining supervision and management system, to increase safety investment into techniques and facilities and to strengthen workers' safety education and introduction of more high-level professional talents.

  3. Preliminary safety analysis of unscrammed events for KLFR

    International Nuclear Information System (INIS)

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

    2005-01-01

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

  4. SACS2: Dynamic and Formal Safety Analysis Method for Complex Safety Critical System

    International Nuclear Information System (INIS)

    Koh, Kwang Yong; Seong, Poong Hyun

    2009-01-01

    Fault tree analysis (FTA) is one of the most widely used safety analysis technique in the development of safety critical systems. However, over the years, several drawbacks of the conventional FTA have become apparent. One major drawback is that conventional FTA uses only static gates and hence can not capture dynamic behaviors of the complex system precisely. Although several attempts such as dynamic fault tree (DFT), PANDORA, formal fault tree (FFT) and so on, have been made to overcome this problem, they can not still do absolute or actual time modeling because they adapt relative time concept and can capture only sequential behaviors of the system. Second drawback of conventional FTA is its lack of rigorous semantics. Because it is informal in nature, safety analysis results heavily depend on an analyst's ability and are error-prone. Finally reasoning process which is to check whether basic events really cause top events is done manually and hence very labor-intensive and timeconsuming for the complex systems. In this paper, we propose a new safety analysis method for complex safety critical system in qualitative manner. We introduce several temporal gates based on timed computational tree logic (TCTL) which can represent quantitative notion of time. Then, we translate the information of the fault trees into UPPAAL query language and the reasoning process is automatically done by UPPAAL which is the model checker for time critical system

  5. MSSV Modeling for Wolsong-1 Safety Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Moon, Bok Ja; Choi, Chul Jin; Kim, Seoung Rae [KEPCO EandC, Daejeon (Korea, Republic of)

    2010-10-15

    The main steam safety valves (MSSVs) are installed on the main steam line to prevent the overpressurization of the system. MSSVs are held in closed position by spring force and the valves pop open by internal force when the main steam pressure increases to open set pressure. If the overpressure condition is relieved, the valves begin to close. For the safety analysis of anticipated accident condition, the safety systems are modeled conservatively to simulate the accident condition more severe. MSSVs are also modeled conservatively for the analysis of over-pressurization accidents. In this paper, the pressure transient is analyzed at over-pressurization condition to evaluate the conservatism for MSSV models

  6. Economic consideration of nuclear safety and cost benefit analysis in nuclear safety regulation

    International Nuclear Information System (INIS)

    Choi, Y. S.; Choi, K. S.; Choi, K. W.; Song, I. J.; Park, D. K.

    2001-01-01

    For the optimization of nuclear safety regulation, understanding of economic aspects of it becomes increasingly important together with the technical approach used so far to secure nuclear safety. Relevant economic theories on private and public goods were reviewed to re-illuminate nuclear safety from the economic perspective. The characteristics of nuclear safety as a public good was reviewed and discussed in comparison with the car safety as a private safety good. It was shown that the change of social welfare resulted from the policy change induced can be calculated by the summation of compensating variation(CV) of individuals. It was shown that the value of nuclear safety could be determined in monetary term by this approach. The theoretical background and history of cost benefit analysis of nuclear safety regulation were presented and topics for future study were suggested

  7. Safety culture and accident analysis-A socio-management approach based on organizational safety social capital

    International Nuclear Information System (INIS)

    Rao, Suman

    2007-01-01

    One of the biggest challenges for organizations in today's competitive business environment is to create and preserve a self-sustaining safety culture. Typically, Key drivers of safety culture in many organizations are regulation, audits, safety training, various types of employee exhortations to comply with safety norms, etc. However, less evident factors like networking relationships and social trust amongst employees, as also extended networking relationships and social trust of organizations with external stakeholders like government, suppliers, regulators, etc., which constitute the safety social capital in the Organization-seem to also influence the sustenance of organizational safety culture. Can erosion in safety social capital cause deterioration in safety culture and contribute to accidents? If so, how does it contribute? As existing accident analysis models do not provide answers to these questions, CAMSoC (Curtailing Accidents by Managing Social Capital), an accident analysis model, is proposed. As an illustration, five accidents: Bhopal (India), Hyatt Regency (USA), Tenerife (Canary Islands), Westray (Canada) and Exxon Valdez (USA) have been analyzed using CAMSoC. This limited cross-industry analysis provides two key socio-management insights: the biggest source of motivation that causes deviant behavior leading to accidents is 'Faulty Value Systems'. The second biggest source is 'Enforceable Trust'. From a management control perspective, deterioration in safety culture and resultant accidents is more due to the 'action controls' rather than explicit 'cultural controls'. Future research directions to enhance the model's utility through layering are addressed briefly

  8. The PEC reactor. Safety analysis: Detailed reports

    Energy Technology Data Exchange (ETDEWEB)

    1988-01-01

    In the safety-analysis of the PEC Brasimone reactor (Italy), attention was focused on the role of plant-incident analysis during the design stage and the conclusions reached. The analysis regarded the following: thermohydraulic incidents at full power; incidents with the reactor shut down; reactivity incidents; core local faults; analysis of fuel-handling incidents; engineered safeguards and passive safety features; coolant leakage and sodium fires; research and development studies on the seismic behaviour of the PEC fast reactor; generalized sodium fire; severe accidents, accident sequences with shudown; reference accident. Both the theoretical and experimental analyses demonstrated the adequacy of the design of the PEC fast reactor, aimed at minimizing the consequences of a hypothetical disruptive core accident with mechanical energy release. It was shown that the containment barriers were sized correctly and that the residual heat from a disassembled core would be removed. The re-evaluation of the source term emphasized the conservative nature of the hypotheses assumed in the preliminary safety analysis for calculating the risk to the public.

  9. Rad waste disposal safety analysis / Integrated safety assessment of a waste repository

    International Nuclear Information System (INIS)

    Jeong, Jongtae; Choi, Jongwon; Kang, Chulhyung

    2012-04-01

    We developed CYPRUS+and adopted PID and RES method for the development of scenario. Safety performance assessment program was developed using GoldSim for the safety assessment of disposal system for the disposal of spnet fuels and wastes resulting from the pyrpoprocessing. Biosphere model was developed and verified in cooperation with JAEA. The capability to evaluate post-closure performance and safety was added to the previously developed program. And, nuclide migration and release to the biosphere considering site characteristics was evaluated by using deterministic and probabilistic approach. Operational safety assessment for drop, fire, and earthquake was also statistically evaluated considering well-established input parameter distribution. Conservative assessment showed that dose rate is below the limit value of low- and intermediate-level repository. Gas generation mechanism within engineered barrier was defined and its influence on safety was evaluated. We made probabilistic safety assessment by obtaining the probability distribution functions of important input variables and also made a sensitivity analysis. The maximum annual dose rate was shown to be below the safety limit value of 10 mSv/yr. The structure and element of safety case was developed to increase reliability of safety assessment methodology for a deep geological repository. Finally, milestone for safety case development and implementation strategy for each safety case element was also proposed

  10. Cost benefit analysis of reactor safety systems

    International Nuclear Information System (INIS)

    Maurer, H.A.

    1984-01-01

    Cost/benefit analysis of reactor safety systems is a possibility appropriate to deal with reactor safety. The Commission of the European Communities supported a study on the cost-benefit or cost effectiveness of safety systems installed in modern PWR nuclear power plants. The following systems and their cooperation in emergency cases were in particular investigated in this study: the containment system (double containment), the leakage exhaust and control system, the annulus release exhaust system and the containment spray system. The benefit of a safety system is defined according to its contribution to the reduction of the radiological consequences for the environment after a LOCA. The analysis is so far performed in two different steps: the emergency core cooling system is considered to function properly, failure of the emergency core cooling system is assumed (with the possible consequence of core melt-down) and the results may demonstrate the evidence that striving for cost-effectiveness can produce a safer end result than the philosophy of safety at any cost. (orig.)

  11. Addendum to the Closure Report for Corrective Action Unit 113: Area 25 R-MAD Facility, Nevada National Security Site, Nevada

    International Nuclear Information System (INIS)

    2011-01-01

    This addendum to the Closure Report for Corrective Action Unit 113: Area 25, Reactor Maintenance, Assembly, and Disassembly Facility, Building 3110, Nevada Test Site, Nevada, DOE/NV--891-VOL I-Rev. 1, dated July 2003, provides details of demolition, waste disposal, and use restriction (UR) modification for Corrective Action Unit 113, Area 25 R-MAD Facility. Demolition was completed on July 15, 2010, when the last of the building debris was disposed. Final field activities were concluded on August 30, 2010, after all equipment was demobilized and UR signs were posted. This work was funded by the American Recovery and Reinvestment Act.

  12. Safety systems and safety analysis of the Qinshan phase III CANDU nuclear power plant

    International Nuclear Information System (INIS)

    Cai Jianping; Shen Sen; Barkman, N.

    1999-01-01

    The author introduces the Canadian nuclear reactor safety philosophy and the Qinshan Phase III CANDU NPP safety systems and safety analysis, which are designed and performed according to this philosophy. The concept of 'defence-in-depth' is a key element of the Canadian nuclear reactor safety philosophy. The design concepts of redundancy, diversity, separation, equipment qualification, quality assurance, and use of appropriate design codes and standards are adopted in the design. Four special safety systems as well as a set of reliable safety support systems are incorporated in the design of Qinshan phase III CANDU for accident mitigation. The assessment results for safety systems performance show that the fundamental safety criteria for public dose, and integrity of fuel, channels and the reactor building, are satisfied

  13. Challenges on innovations of newly-developed safety analysis codes

    International Nuclear Information System (INIS)

    Yang, Yanhua; Zhang, Hao

    2016-01-01

    With the development of safety analysis method, the safety analysis codes meet more challenges. Three challenges are presented in this paper, which are mathematic model, code design and user interface. Combined with the self-reliance safety analysis code named COSINE, the ways of meeting these requirements are suggested, that is to develop multi-phases, multi-fields and multi-dimension models, to adopt object-oriented code design ideal and to improve the way of modeling, calculation control and data post-processing in the user interface.

  14. Challenges on innovations of newly-developed safety analysis codes

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Yanhua [Shanghai Jiao Tong Univ. (China). School of Nuclear Science and Engineering; Zhang, Hao [State Nuclear Power Software Development Center, Beijing (China). Beijing Future Science and Technology City

    2016-05-15

    With the development of safety analysis method, the safety analysis codes meet more challenges. Three challenges are presented in this paper, which are mathematic model, code design and user interface. Combined with the self-reliance safety analysis code named COSINE, the ways of meeting these requirements are suggested, that is to develop multi-phases, multi-fields and multi-dimension models, to adopt object-oriented code design ideal and to improve the way of modeling, calculation control and data post-processing in the user interface.

  15. Safety Analysis Report for Ignalina NPP

    International Nuclear Information System (INIS)

    Negrivoda, G.

    1997-01-01

    In December 1994 an agreement was signed between the European Bank for Reconstruction and Development and the Republic of Lithuania for the grant of 32.86 MECU for the safety Improvement at Ignalina NPP. One of the conditions for the provision of the grant, was a requirement for an in-depth analysis of the safety level at Ignalina NPP in the scope and according to the standards acceptable for a western nuclear power plant, and to publish a Safety Analysis Report (SAR). The report should investigate and analyze any factor that could limit a safe operation of the plant, and provide recommendations for actual safety improvements. According to the agreement, Lithuania had to finalize the SAR until 31 December, 1995. The bank has also organized and financed investigation of safety at Ignalina NPP and preparation of the SAR. EBRD made an agreement with Sweden's Vattenfall, which subcontracted well-known companies from Canada, USA, Germany, etc., and also the Russian Research and Development Institute of Power Engineering (NIKIET), reactor designer of Ignalina NPP. The SAR is a very comprehensive document and contains about 8000 pages of text, diagrams and tables. The main findings of the SAR are provided in the article. A large number of discrepancies with modern rules and western practices was detected, but they were not proved to be serious enough to require reactors shutdown. Based on the recommendations of the SAR Ignalina NPP has worked out Safety Improvement Program No. 2 (SIP-2), which is planned for three years and will cost 486 MLT. (author)

  16. Safety culture and accident analysis-A socio-management approach based on organizational safety social capital

    Energy Technology Data Exchange (ETDEWEB)

    Rao, Suman [Risk Analyst (India)]. E-mail: sumanashokrao@yahoo.co.in

    2007-04-11

    One of the biggest challenges for organizations in today's competitive business environment is to create and preserve a self-sustaining safety culture. Typically, Key drivers of safety culture in many organizations are regulation, audits, safety training, various types of employee exhortations to comply with safety norms, etc. However, less evident factors like networking relationships and social trust amongst employees, as also extended networking relationships and social trust of organizations with external stakeholders like government, suppliers, regulators, etc., which constitute the safety social capital in the Organization-seem to also influence the sustenance of organizational safety culture. Can erosion in safety social capital cause deterioration in safety culture and contribute to accidents? If so, how does it contribute? As existing accident analysis models do not provide answers to these questions, CAMSoC (Curtailing Accidents by Managing Social Capital), an accident analysis model, is proposed. As an illustration, five accidents: Bhopal (India), Hyatt Regency (USA), Tenerife (Canary Islands), Westray (Canada) and Exxon Valdez (USA) have been analyzed using CAMSoC. This limited cross-industry analysis provides two key socio-management insights: the biggest source of motivation that causes deviant behavior leading to accidents is 'Faulty Value Systems'. The second biggest source is 'Enforceable Trust'. From a management control perspective, deterioration in safety culture and resultant accidents is more due to the 'action controls' rather than explicit 'cultural controls'. Future research directions to enhance the model's utility through layering are addressed briefly.

  17. Construction safety and waste management an economic analysis

    CERN Document Server

    Li, Rita Yi Man

    2015-01-01

    This monograph presents an analysis of construction safety problems and on-site safety measures from an economist’s point of view. The book includes examples from both emerging countries, e.g. China and India, and developed countries, e.g. Australia and Hong Kong. Moreover, the author covers an analysis on construction safety knowledge sharing by means of updatable mobile technology such as apps in Androids and iOS platform mobile devices. The target audience comprises primarily researchers and experts in the field but the book may also be beneficial for graduate students.

  18. Using Addenda in Documented Safety Analysis Reports

    International Nuclear Information System (INIS)

    Swanson, D.S.; Thieme, M.A.

    2003-01-01

    This paper discusses the use of addenda to the Radioactive Waste Management Complex (RWMC) Documented Safety Analysis (DSA) located at the Idaho National Engineering and Environmental Laboratory (INEEL). Addenda were prepared for several systems and processes at the facility that lacked adequate descriptive information and hazard analysis in the DSA. They were also prepared for several new activities involving unreviewed safety questions (USQs). Ten addenda to the RWMC DSA have been prepared since the last annual update

  19. System safety analysis of an autonomous mobile robot

    International Nuclear Information System (INIS)

    Bartos, R.J.

    1994-01-01

    Analysis of the safety of operating and maintaining the Stored Waste Autonomous Mobile Inspector (SWAMI) II in a hazardous environment at the Fernald Environmental Management Project (FEMP) was completed. The SWAMI II is a version of a commercial robot, the HelpMate trademark robot produced by the Transitions Research Corporation, which is being updated to incorporate the systems required for inspecting mixed toxic chemical and radioactive waste drums at the FEMP. It also has modified obstacle detection and collision avoidance subsystems. The robot will autonomously travel down the aisles in storage warehouses to record images of containers and collect other data which are transmitted to an inspector at a remote computer terminal. A previous study showed the SWAMI II has economic feasibility. The SWAMI II will more accurately locate radioactive contamination than human inspectors. This thesis includes a System Safety Hazard Analysis and a quantitative Fault Tree Analysis (FTA). The objectives of the analyses are to prevent potentially serious events and to derive a comprehensive set of safety requirements from which the safety of the SWAMI II and other autonomous mobile robots can be evaluated. The Computer-Aided Fault Tree Analysis (CAFTA copyright) software is utilized for the FTA. The FTA shows that more than 99% of the safety risk occurs during maintenance, and that when the derived safety requirements are implemented the rate of serious events is reduced to below one event per million operating hours. Training and procedures in SWAMI II operation and maintenance provide an added safety margin. This study will promote the safe use of the SWAMI II and other autonomous mobile robots in the emerging technology of mobile robotic inspection

  20. Preliminary Safety Analysis Report for the Tokamak Physics Experiment

    International Nuclear Information System (INIS)

    Motloch, C.G.; Bonney, R.F.; Levine, J.D.; Masson, L.S.; Commander, J.C.

    1995-04-01

    This Preliminary Safety Analysis Report (PSAR), includes an indication of the magnitude of facility hazards, complexity of facility operations, and the stage of the facility life-cycle. It presents the results of safety analyses, safety assurance programs, identified vulnerabilities, compensatory measures, and, in general, the rationale describing why the Tokamak Physics Experiment (TPX) can be safely operated. It discusses application of the graded approach to the TPX safety analysis, including the basis for using Department of Energy (DOE) Order 5480.23 and DOE-STD-3009-94 in the development of the PSAR

  1. Reliability analysis of diverse safety logic systems of fast breeder reactor

    International Nuclear Information System (INIS)

    Ravi Kumar, Bh.; Apte, P.R.; Srivani, L.; Ilango Sambasivan, S.; Swaminathan, P.

    2006-01-01

    Safety Logic for Fast Breeder Reactor (FBR) is designed to initiate safety action against Design Basis Events. Based on the outputs of various processing circuits, Safety logic system drives the control rods of the shutdown system. So, Safety Logic system is classified as safety critical system. Therefore, reliability analysis has to be performed. This paper discusses the Reliability analysis of Diverse Safety logic systems of FBRs. For this literature survey on safety critical systems, system reliability approach and standards to be followed like IEC-61508 are discussed in detail. For Programmable Logic device based systems, Hardware Description Languages (HDL) are used. So this paper also discusses the Verification and Validation for HDLs. Finally a case study for the Reliability analysis of Safety logic is discussed. (author)

  2. Uncertainty analysis for Ulysses safety evaluation report

    International Nuclear Information System (INIS)

    Frank, M.V.

    1991-01-01

    As part of the effort to review the Ulysses Final Safety Analysis Report and to understand the risk of plutonium release from the Ulysses spacecraft General Purpose Heat Source---Radioisotope Thermal Generator (GPHS-RTG), the Interagency Nuclear Safety Review Panel (INSRP) and the author performed an integrated, quantitative analysis of the uncertainties of the calculated risk of plutonium release from Ulysses. Using state-of-art probabilistic risk assessment technology, the uncertainty analysis accounted for both variability and uncertainty of the key parameters of the risk analysis. The results show that INSRP had high confidence that risk of fatal cancers from potential plutonium release associated with calculated launch and deployment accident scenarios is low

  3. COLD-SAT feasibility study safety analysis

    Science.gov (United States)

    Mchenry, Steven T.; Yost, James M.

    1991-01-01

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

  4. Guidance for preparation of safety analysis reports for nonreactor facilities and operations

    International Nuclear Information System (INIS)

    1992-01-01

    Department of Energy (DOE) Orders 5480.23, ''Nuclear Safety Analysis Reports,'' and 5481.1B, ''Safety Analysis and Review System'' require the preparation of appropriate safety analyses for each DOE operation and subsequent significant modifications including decommissioning, and independent review of each safety analysis. The purpose of this guide is to assist in the preparation and review of safety documentation for Oak Ridge Field Office (OR) nonreactor facilities and operation. Appendix A lists DOE Orders, NRC Regulatory Guides and other documents applicable to the preparation of safety analysis reports

  5. Safety analysis and risk assessment handbook

    International Nuclear Information System (INIS)

    Peterson, V.L.; Colwell, R.G.; Dickey, R.L.

    1997-01-01

    This Safety Analysis and Risk Assessment Handbook (SARAH) provides guidance to the safety analyst at the Rocky Flats Environmental Technology Site (RFETS) in the preparation of safety analyses and risk assessments. Although the older guidance (the Rocky Flats Risk Assessment Guide) continues to be used for updating the Final Safety Analysis Reports developed in the mid-1980s, this new guidance is used with all new authorization basis documents. With the mission change at RFETS came the need to establish new authorization basis documents for its facilities, whose functions had changed. The methodology and databases for performing the evaluations that support the new authorization basis documents had to be standardized, to avoid the use of different approaches and/or databases for similar accidents in different facilities. This handbook presents this new standardized approach. The handbook begins with a discussion of the requirements of the different types of authorization basis documents and how to choose the one appropriate for the facility to be evaluated. It then walks the analyst through the process of identifying all the potential hazards in the facility, classifying them, and choosing the ones that need to be analyzed further. It then discusses the methods for evaluating accident initiation and progression and covers the basic steps in a safety analysis, including consequence and frequency binning and risk ranking. The handbook lays out standardized approaches for determining the source terms of the various accidents (including airborne release fractions, leakpath factors, etc.), the atmospheric dispersion factors appropriate for Rocky Flats, and the methods for radiological and chemical consequence assessments. The radiological assessments use a radiological open-quotes templateclose quotes, a spreadsheet that incorporates the standard values of parameters, whereas the chemical assessments use the standard codes ARCHIE and ALOHA

  6. Multivariate time series analysis of SafetyNet data. SafetyNet, Building the European Road Safety Observatory, Workpackage 7, Deliverable 7.7.

    NARCIS (Netherlands)

    Commandeur, J.J.F. Bijleveld, F.D. & Bergel, R.

    2009-01-01

    This deliverable provides an application of theories and methods documented in Deliverables 7.4 and 7.5 of work package 7 of the SafetyNet project. In this deliverable, use of select analysis techniques is demonstrated through real world road safety analysis problems, using aggregate data which may

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

    Energy Technology Data Exchange (ETDEWEB)

    1988-01-15

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

  8. Computer aided safety analysis

    International Nuclear Information System (INIS)

    1988-05-01

    The document reproduces 20 selected papers from the 38 papers presented at the Technical Committee/Workshop on Computer Aided Safety Analysis organized by the IAEA in co-operation with the Institute of Atomic Energy in Otwock-Swierk, Poland on 25-29 May 1987. A separate abstract was prepared for each of these 20 technical papers. Refs, figs and tabs

  9. 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)

  10. Addendum to the remedial investigation report on Bear Creek Valley Operable Unit 2 (Rust Spoil Area, Spoil Area 1, and SY-200 Yard) at the Oak Ridge Y-12 Plant Oak Ridge, Tennessee. Volume 1: Main text

    International Nuclear Information System (INIS)

    1995-04-01

    This addendum to the Remedial Investigation (RI) Report on Bear Creek Valley Operable Unit (OU) 2 at the Oak Ridge Y-12 Plant was prepared in accordance with requirements under the Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA) for reporting the results of a site characterization for public review. This addendum is a supplement to a document that was previously issued in January 1995 and that provided the Environmental Restoration Program with information about the results of the 1993 investigation performed at OU 2. The January 1995 D2 version of the RI Report on Bear Creek Valley OU 2 included information on risk assessments that have evaluated impacts to human health and the environment. Information provided in the document formed the basis for the development of the Feasibility Study Report. This addendum includes revisions to four chapters of information that were a part of the document issued in January 1995. Specifically, it includes revisions to Chaps. 2, 3, 4, and 9. Volume 1 of this document is not being reissued in its entirety as a D3 version because only the four chapters just mentioned have been affected by requested changes. Note also that Volume 2 of this RI Report on Bear Creek Valley OU 2 is not being reissued in conjunction with Volume 1 of this document because there have been no changes requested or made to the previously issued version of Volume 2 of this document.

  11. Special characteristics of the safety analysis of HWRs

    International Nuclear Information System (INIS)

    Kugler, G.

    1980-01-01

    Two lectures are presented in this report. The CANDU-PHW reactor is used as a model for discussion. The first lecture describes the distinctive features of the CANDU reactor, and how they impact on reactor safety. In the second lecture the Canadian safety philosophy, the safety design objective, and other selected topics on reactor safety analysis are discussed. The material in this report was selected with a view to assisting those not familiar with the CANDU heavy water reactor design in evaluating the distinctive safety aspects of these reactors. (orig./RW)

  12. SCALE 5: Powerful new criticality safety analysis tools

    International Nuclear Information System (INIS)

    Bowman, Stephen M.; Hollenbach, Daniel F.; Dehart, Mark D.; Rearden, Bradley T.; Gauld, Ian C.; Goluoglu, Sedat

    2003-01-01

    Version 5 of the SCALE computer software system developed at Oak Ridge National Laboratory, scheduled for release in December 2003, contains several significant new modules and sequences for criticality safety analysis and marks the most important update to SCALE in more than a decade. This paper highlights the capabilities of these new modules and sequences, including continuous energy flux spectra for processing multigroup problem-dependent cross sections; one- and three-dimensional sensitivity and uncertainty analyses for criticality safety evaluations; two-dimensional flexible mesh discrete ordinates code; automated burnup-credit analysis sequence; and one-dimensional material distribution optimization for criticality safety. (author)

  13. Support analysis for safety analysis development for CANDU nuclear power plant

    International Nuclear Information System (INIS)

    Bedreaga, L.; Florescu, Gh.; Apostol, M.; Nitoi, M.

    2004-01-01

    Probabilistic Safety Assessment analysis (PSA) is a technique used to assess the safety of a nuclear power plant. Assessments of the nuclear plant systems/components from safety point of view consist in accomplishment of a lot of support analyses that are the base for the main analysis, in order to evaluate the impact of occurrences of abnormal states for these systems. Evaluation of initiating events frequency and components failure rate is based on underlying probabilistic theory and mathematic statistics. Some of these analyses are detailed analyses and are known very well in PSA. There are also some analyses, named support analyses for PSA, which are very important but less applicable because they involve a huge human effort and hardware facilities to accomplish. The usual methods applicable in PSA such as input data extracted from the specific documentation (operation procedures, testing procedures, maintenance procedures and so on) or conservative evaluation provide a high level of uncertainty for both input and output data. The paper describes support analysis required to improve the certainty level in evaluation of reliability parameters and also in the final results (either risk, reliability or safety assessment). (author)

  14. Preparing a Safety Analysis Report using the building block approach

    International Nuclear Information System (INIS)

    Herrington, C.C.

    1990-01-01

    The credibility of the applicant in a licensing proceeding is severely impacted by the quality of the license application, particularly the Safety Analysis Report. To ensure the highest possible credibility, the building block approach was devised to support the development of a quality Safety Analysis Report. The approach incorporates a comprehensive planning scheme that logically ties together all levels of the investigation and provides the direction necessary to prepare a superior Safety Analysis Report

  15. Utilization of the MCNP-3A code for criticality safety analysis

    International Nuclear Information System (INIS)

    Maragni, M.G.; Moreira, J.M.L.

    1996-01-01

    In the last decade, Brazil started to operate facilities for processing and storing uranium in different forms. The necessity of criticality safety analysis appeared in the design phase of the uranium pilot process plants and also in the licensing of transportation and storage of fissile materials. The 2-MW research reactor and the Angra I power plant also required criticality safety assessments because their spent-fuel storage was approaching full-capacity utilization. The criticality safety analysis in Brazil has been based on KENO IV code calculations, which present some difficulties for correct geometry representation. The MCNP-3A code is not reported to be used frequently for criticality safety analysis in Brazil, but its good geometry representation makes it a possible tool for treating problems of complex geometry. A set of benchmark tests was performed to verify its applicability for criticality safety analysis in Brazil. This paper presents several benchmark tests aimed at selecting a set of options available in the MCNP-3A code that would be adequate for criticality safety analysis. The MCNP-3A code is also compared with the KENO-IV code regarding its performance for criticality safety analysis

  16. Safety analysis of tritium processing system based on PHA

    International Nuclear Information System (INIS)

    Fu Wanfa; Luo Deli; Tang Tao

    2012-01-01

    Safety analysis on primary confinement of tritium processing system for TBM was carried out with Preliminary Hazard Analysis. Firstly, the basic PHA process was given. Then the function and safe measures with multiple confinements about tritium system were described and analyzed briefly, dividing the two kinds of boundaries of tritium transferring through, that are multiple confinement systems division and fluid loops division. Analysis on tritium releasing is the key of PHA. Besides, PHA table about tritium releasing was put forward, the causes and harmful results being analyzed, and the safety measures were put forward also. On the basis of PHA, several kinds of typical accidents were supposed to be further analyzed. And 8 factors influencing the tritium safety were analyzed, laying the foundation of evaluating quantitatively the safety grade of various nuclear facilities. (authors)

  17. Supplement to proceedings of the second analysis meeting on JUPITER Program

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1981-10-01

    This document provides a supplement to the proceedings which provided an analysis of the JUPITER Experiment in ZPPR-10 (including ZPPR-9). The Addendum and Comments discusses Reaction Rate Ratio and Distribution. Topics provided are: Some Additional Corrections for ZPPR-9 Reaction Rate Distribution; An Analysis of the Sample Worth Experiments in ZPPR-10A; Homogenization Method of Pin Rods in ZPPR- 10A Assembly; and Control Rod Analysis in ZPPR-3.

  18. A Study of Time Response for Safety-Related Operator Actions in Non-LOCA Safety Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Min Seok; Lee, Sang Seob; Park, Min Soo; Lee, Gyu Cheon; Kim, Shin Whan [KEPCO E and C Company, Daejeon (Korea, Republic of)

    2014-10-15

    The classification of initiating events for safety analysis report (SAR) chapter 15 is categorized into moderate frequency events (MF), infrequent events (IF), and limiting faults (LF) depending on the frequency of its occurrence. For the non-LOCA safety analysis with the purpose to get construction or operation license, however, it is assumed that the operator response action to mitigate the events starts at 30 minutes after the initiation of the transient regardless of the event categorization. Such an assumption of corresponding operator response time may have over conservatism with the MF and IF events and results in a decrease in the safety margin compared to its acceptance criteria. In this paper, the plant conditions (PC) are categorized with the definitions in SAR 15 and ANS 51.1. Then, the consequence of response for safety-related operator action time is determined based on the PC in ANSI 58.8. The operator response time for safety analysis regarding PC are reviewed and suggested. The clarifying alarm response procedure would be required for the guideline to reduce the operator response time when the alarms indicate the occurrence of the transient.

  19. Development of safety analysis technology for LMR

    International Nuclear Information System (INIS)

    Hahn, Do Hee; Kwon, Y. M.; Suk, S. D.

    2002-05-01

    In the present study, the KALIMER safety analysis has been made for the transients considered in the design concept, hypothetical core disruptive accident (HCDA), and containment performance with the establishment of the design basis. Such analyses have not been possible without the computer code improvement, and the experience attained during this research period must have greatly contributed to the achievement of the self reliance in the domestic technology establishment on the safety analysis areas of the conceptual design. The safety analysis codes have been improved to extend their applicable ranges for detailed conceptual design, and a basic computer code system has been established for HCDA analysis. A code-to-code comparison analysis has been performed as a part of code verification attempt, and the leading edge technology of JNC also has been brought for the technology upgrade. In addition, the research and development on the area of the database establishment has been made for the efficient and systematic project implementation of the conceptual design, through performances on the development of a project scheduling management, integration of the individually developed technology, establishment of the product database, and so on, taking into account coupling of the activities conducted in each specific area

  20. Preclosure Safety Analysis Guide

    International Nuclear Information System (INIS)

    D.D. Orvis

    2003-01-01

    A preclosure safety analysis (PSA) is a required element of the License Application (LA) for the high- level radioactive waste repository at Yucca Mountain. This guide provides analysts and other Yucca Mountain Repository Project (the Project) personnel with standardized methods for developing and documenting the PSA. The definition of the PSA is provided in 10 CFR 63.2, while more specific requirements for the PSA are provided in 10 CFR 63.112, as described in Sections 1.2 and 2. The PSA requirements described in 10 CFR Part 63 were developed as risk-informed performance-based regulations. These requirements must be met for the LA. The PSA addresses the safety of the Geologic Repository Operations Area (GROA) for the preclosure period (the time up to permanent closure) in accordance with the radiological performance objectives of 10 CFR 63.111. Performance objectives for the repository after permanent closure (described in 10 CFR 63.113) are not mentioned in the requirements for the PSA and they are not considered in this guide. The LA will be comprised of two phases: the LA for construction authorization (CA) and the LA amendment to receive and possess (R and P) high-level radioactive waste (HLW). PSA methods must support the safety analyses that will be based on the differing degrees of design detail in the two phases. The methods described herein combine elements of probabilistic risk assessment (PRA) and deterministic analyses that comprise a risk-informed performance-based safety analysis. This revision to the PSA guide was prepared for the following objectives: (1) To correct factual and typographical errors. (2) To provide additional material suggested from reviews by the Project, the U.S. Department of Energy (DOE), and U.S. Nuclear Regulatory Commission (NRC) Staffs. (3) To update material in accordance with approaches and/or strategies adopted by the Project. In addition, a principal objective for the planned revision was to ensure that the methods and

  1. Analysis of high burnup fuel safety issues

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Chan Bock; Kim, D. H.; Bang, J. G.; Kim, Y. M.; Yang, Y. S.; Jung, Y. H.; Jeong, Y. H.; Nam, C.; Baik, J. H.; Song, K. W.; Kim, K. S

    2000-12-01

    Safety issues in steady state and transient behavior of high burnup LWR fuel above 50 - 60 MWD/kgU were analyzed. Effects of burnup extension upon fuel performance parameters was reviewed, and validity of both the fuel safety criteria and the performance analysis models which were based upon the lower burnup fuel test results was analyzed. It was found that further tests would be necessary in such areas as fuel failure and dispersion for RIA, and high temperature cladding corrosion and mechanical deformation for LOCA. Since domestic fuels have been irradiated in PWR up to burnup higher than 55 MWD/kgU-rod. avg., it can be said that Korea is in the same situation as the other countries in the high burnup fuel safety issues. Therefore, necessary research areas to be performed in Korea were derived. Considering that post-irradiation examination(PIE) for the domestic fuel of burnup higher than 30 MWD/kgU has not been done so far at all, it is primarily necessary to perform PIE for high burnup fuel, and then simulation tests for RIA and LOCA could be performed by using high burnup fuel specimens. For the areas which can not be performed in Korea, international cooperation will be helpful to obtain the test results. With those data base, safety of high burnup domestic fuels will be confirmed, current fuel safety criteria will be re-evaluated, and finally transient high burnup fuel behavior analysis technology will be developed through the fuel performance analysis code development.

  2. Analysis of high burnup fuel safety issues

    International Nuclear Information System (INIS)

    Lee, Chan Bock; Kim, D. H.; Bang, J. G.; Kim, Y. M.; Yang, Y. S.; Jung, Y. H.; Jeong, Y. H.; Nam, C.; Baik, J. H.; Song, K. W.; Kim, K. S

    2000-12-01

    Safety issues in steady state and transient behavior of high burnup LWR fuel above 50 - 60 MWD/kgU were analyzed. Effects of burnup extension upon fuel performance parameters was reviewed, and validity of both the fuel safety criteria and the performance analysis models which were based upon the lower burnup fuel test results was analyzed. It was found that further tests would be necessary in such areas as fuel failure and dispersion for RIA, and high temperature cladding corrosion and mechanical deformation for LOCA. Since domestic fuels have been irradiated in PWR up to burnup higher than 55 MWD/kgU-rod. avg., it can be said that Korea is in the same situation as the other countries in the high burnup fuel safety issues. Therefore, necessary research areas to be performed in Korea were derived. Considering that post-irradiation examination(PIE) for the domestic fuel of burnup higher than 30 MWD/kgU has not been done so far at all, it is primarily necessary to perform PIE for high burnup fuel, and then simulation tests for RIA and LOCA could be performed by using high burnup fuel specimens. For the areas which can not be performed in Korea, international cooperation will be helpful to obtain the test results. With those data base, safety of high burnup domestic fuels will be confirmed, current fuel safety criteria will be re-evaluated, and finally transient high burnup fuel behavior analysis technology will be developed through the fuel performance analysis code development

  3. LOCA analysis of SCWR-M with passive safety system

    Energy Technology Data Exchange (ETDEWEB)

    Liu, X.J., E-mail: xiaojingliu@sjtu.edu.cn [School of Nuclear Science and Engineering, Shanghai Jiao Tong University, 800 Dong Chuan Road, Shanghai 200240 (China); Fu, S.W. [Navy University of Engineering, Wuhan, Hubei (China); Xu, Z.H. [Shanghai Nuclear Engineering Research and Design Institute, Shanghai (China); Yang, Y.H. [School of Nuclear Science and Engineering, Shanghai Jiao Tong University, 800 Dong Chuan Road, Shanghai 200240 (China); Cheng, X. [Institute of Fusion and Nuclear Technology, Karlsruhe Institute of Technology (KIT), Kaiserstr. 12, 76131 Karlsruhe (Germany)

    2013-06-15

    Highlights: • Application of the ATHLET-SC code to the trans-critical analysis for SCWR. • Development of a passive safety system for SCWR-M. • Analysis of hot/cold leg LOCA behaviour with different break size. • Introduction of some mitigation measures for SCWR-M -- Abstract: A new SCWR conceptual design (mixed spectrum supercritical water cooled reactor: SCWR-M) is proposed by Shanghai Jiao Tong University (SJTU). R and D activities covering core design, safety system design and code development of SCWR-M are launched at SJTU. Safety system design and analysis is one of the key tasks during the development of SCWR-M. Considering the current advanced reactor design, a new passive safety system for SCWR-M including isolation cooling system (ICS), accumulator injection system (ACC), gravity driven cooling system (GDCS) and automatic depressurization system (ADS) is proposed. Based on the modified and preliminarily assessed system code ATHLET-SC, loss of coolant accident (LOCA) analysis for hot and cold leg is performed in this paper. Three different break sizes are analyzed to clarify the hot and cold LOCA characteristics of the SCWR-M. The influence of the break location and break size on the safety performance of SCWR-M is also concluded. Several measures to induce the core coolant flow and to mitigate core heating up are also discussed. The results achieved so far demonstrate the feasibility of the proposed passive safety system to keep the SCWR-M core at safety condition during loss of coolant accident.

  4. 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

  5. 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

  6. 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.

  7. Posttest analysis of the FFTF inherent safety tests

    International Nuclear Information System (INIS)

    Padilla, A. Jr.; Claybrook, S.W.

    1987-01-01

    Inherent safety tests were performed during 1986 in the 400-MW (thermal) Fast Flux Test Facility (FFTF) reactor to demonstrate the effectiveness of an inherent shutdown device called the gas expansion module (GEM). The GEM device provided a strong negative reactivity feedback during loss-of-flow conditions by increasing the neutron leakage as a result of an expanding gas bubble. The best-estimate pretest calculations for these tests were performed using the IANUS plant analysis code (Westinghouse Electric Corporation proprietary code) and the MELT/SIEX3 core analysis code. These two codes were also used to perform the required operational safety analyses for the FFTF reactor and plant. Although it was intended to also use the SASSYS systems (core and plant) analysis code, the calibration of the SASSYS code for FFTF core and plant analysis was not completed in time to perform pretest analyses. The purpose of this paper is to present the results of the posttest analysis of the 1986 FFTF inherent safety tests using the SASSYS code

  8. SNF fuel retrieval sub project safety analysis document

    International Nuclear Information System (INIS)

    BERGMANN, D.W.

    1999-01-01

    This safety analysis is for the SNF Fuel Retrieval (FRS) Sub Project. The FRS equipment will be added to K West and K East Basins to facilitate retrieval, cleaning and repackaging the spent nuclear fuel into Multi-Canister Overpack baskets. The document includes a hazard evaluation, identifies bounding accidents, documents analyses of the accidents and establishes safety class or safety significant equipment to mitigate accidents as needed

  9. SNF fuel retrieval sub project safety analysis document

    Energy Technology Data Exchange (ETDEWEB)

    BERGMANN, D.W.

    1999-02-24

    This safety analysis is for the SNF Fuel Retrieval (FRS) Sub Project. The FRS equipment will be added to K West and K East Basins to facilitate retrieval, cleaning and repackaging the spent nuclear fuel into Multi-Canister Overpack baskets. The document includes a hazard evaluation, identifies bounding accidents, documents analyses of the accidents and establishes safety class or safety significant equipment to mitigate accidents as needed.

  10. Development and improvement of safety analysis code for geological disposal

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-08-15

    In order to confirm the long-term safety concerning geological disposal, probabilistic safety assessment code and other analysis codes, which can evaluate possibility of each event and influence on engineered barrier and natural barrier by the event, were introduced. We confirmed basic functions of those codes and studied the relation between those functions and FEP/PID which should be taken into consideration in safety assessment. We are planning to develop 'Nuclide Migration Assessment System' for the purpose of realizing improvement in efficiency of assessment work, human error prevention for analysis, and quality assurance of the analysis environment and analysis work for safety assessment by using it. As the first step, we defined the system requirements and decided the system composition and functions which should be mounted in them based on those requirements. (author)

  11. Safety analysis of control rod drive computers

    International Nuclear Information System (INIS)

    Ehrenberger, W.; Rauch, G.; Schmeil, U.; Maertz, J.; Mainka, E.U.; Nordland, O.; Gloee, G.

    1985-01-01

    The analysis of the most significant user programmes revealed no errors in these programmes. The evaluation of approximately 82 cumulated years of operation demonstrated that the operating system of the control rod positioning processor has a reliability that is sufficiently good for the tasks this computer has to fulfil. Computers can be used for safety relevant tasks. The experience gained with the control rod positioning processor confirms that computers are not less reliable than conventional instrumentation and control system for comparable tasks. The examination and evaluation of computers for safety relevant tasks can be done with programme analysis or statistical evaluation of the operating experience. Programme analysis is recommended for seldom used and well structured programmes. For programmes with a long, cumulated operating time a statistical evaluation is more advisable. The effort for examination and evaluation is not greater than the corresponding effort for conventional instrumentation and control systems. This project has also revealed that, where it is technologically sensible, process controlling computers or microprocessors can be qualified for safety relevant tasks without undue effort. (orig./HP) [de

  12. Safety evaluation report of the Waste Isolation Pilot Plant safety analysis report: Contact-handled transuranic waste disposal operations

    International Nuclear Information System (INIS)

    1997-02-01

    DOE 5480.23, Nuclear Safety Analysis Reports, requires that the US Department of Energy conduct an independent, defensible, review in order to approve a Safety Analysis Report (SAR). That review and the SAR approval basis is documented in this formal Safety Evaluation Report (SER). This SER documents the DOE's review of the Waste Isolation Pilot Plant SAR and provides the Carlsbad Area Office Manager, the WIPP SAR approval authority, with the basis for approving the safety document. It concludes that the safety basis documented in the WIPP SAR is comprehensive, correct, and commensurate with hazards associated with planned waste disposal operations

  13. Safety analysis of the existing 850 Firing Facility

    International Nuclear Information System (INIS)

    Odell, B.N.

    1986-01-01

    A safety analysis was performed to determine if normal operations and/or potential accidents at the 850 Firing Facility at Site 300 could present undue hazards to the general public, personnel at Site 300, or have an adverse effect on the environment. The normal operations and credible accidents that might have an effect on these facilities or have off-site consequences were considered. It was determined by this analysis that all but one of the hazards were either low or of the type or magnitude routinely encountered and/or accepted by the public. The exception was explosives, which was classified as a moderate hazard per the requirements given in DOE Order 5481.1A. This safety analysis concluded that the operation at this facility will present no undue risk to the health and safety of LLNL employees or the public

  14. Safety analysis of the existing 851 Firing Facility

    International Nuclear Information System (INIS)

    Odell, B.N.

    1986-01-01

    A safety analysis was performed to determine if normal operations and/or potential accidents at the 851 Firing Facility at Site 300 could present undue hazards to the general public, personnel at Site 300, or have an adverse effect on the environment. The normal operations and credible accidents that might have an effect on these facilities or have off-site consequences were considered. It was determined by this analysis that all but two of the hazards were either low or of the type or magnitude routinely encountered and/or accepted by the public. The exceptions were the linear accelerator and explosives, which were classified as moderate hazards per the requirements given in DOE Order 5481.1A. This safety analysis concluded that the operation at this facility will present no undue risk to the health and safety of LLNL employees or the public

  15. Statistical margin to DNB safety analysis approach for LOFT

    International Nuclear Information System (INIS)

    Atkinson, S.A.

    1982-01-01

    A method was developed and used for LOFT thermal safety analysis to estimate the statistical margin to DNB for the hot rod, and to base safety analysis on desired DNB probability limits. This method is an advanced approach using response surface analysis methods, a very efficient experimental design, and a 2nd-order response surface equation with a 2nd-order error propagation analysis to define the MDNBR probability density function. Calculations for limiting transients were used in the response surface analysis thereby including transient interactions and trip uncertainties in the MDNBR probability density

  16. Tolerability of risk, safety assessment principles and their implications for probabilistic safety analysis

    International Nuclear Information System (INIS)

    Ewing, D.J.F.; Campbell, J.F.

    1994-01-01

    This paper gives a regulatory view of probabilistic safety assessment as seen by the Nuclear Installations Inspectorate (NII) and in the light of the general regulatory risk aims set out in the Health and Safety Executive's (HSE) The tolerability of risk from nuclear power stations (TOR) and in Safety assessment principles for nuclear plants (SAPs), prepared by NII on behalf of the HSE. Both of these publications were revised and republished in 1992. This paper describes the SAPs, together with the historical background, the motivation for review, the effects of the Sizewell and Hinkley Point C public inquiries, changes since the original versions, comparison with international standards and use in assessment. For new plant, probabilistic safety analysis (PSA) is seen as an essential tool in balancing the safety of the design and in demonstrating compliance with TOR and the SAPs. (Author)

  17. Development of regulatory technology for thermal-hydraulic safety analysis

    International Nuclear Information System (INIS)

    Bang, Young Seok; Lee, S. H.; Ryu, Y. H.

    2001-02-01

    The present study aims to develop the regulation capability in thermal-hydraulic safety analysis which was required for the reasonable safety regulation in the current NPP, the next generation reactors, and the future-type reactors. The fourth fiscal year of the first phase of the research was focused on the following research topics: Investigation on the current status of the thermal-hydraulic safety analysis technology outside and inside of the country; Review on the improved features of the thermal-hydraulic safety analysis regulatory audit code, RELAP5/MOD3; Assessments of code with LOFT L9-3 ATWS experiment and LSTF SB-SG-10 multiple SGTR experiment; Application of the RELAP5/CANDU code to analyses of SLB and LBLOCA and evaluation of its effect on safety; Application of the code to IAEA PHWR ISP analysis; Assessments of RELAP5 and TRAC with UPTF downcomer injection test and Analysis of LBLOCA with RELAP5 for the performance evaluation of KNGR DVI; Setup of a coupled 3-D kinetics and thermal-hydraulics and application it to a reactivity accident analysis; and Extension of database and improvement of plant input decks. For supporting the resolution of safety issues, loss of RHR event during midloop operation was analyzed for Kori Unit 3, issues on high burnup fuel were reviewed and performance of FRAPCON-3 assessed. Also MSLB was analyzed to figure out the sensitivity of downcomer temperature supporting the PTS risk evaluation of Kori Unit 1. Thermal stratification in pipe was analyzed using the method proposed. And a method predicting the thermal-hydraulic performance of IRWST of KNGR was explored. The PWR ECCS performance criteria was issued as a MOST Article 200-19.and a regulatory guide on evaluation methodology was improved to cover concerns raised from the related licensing review process

  18. Safety analysis in support of regulatory decision marking

    International Nuclear Information System (INIS)

    Pomier Baez, L.; Troncoso Fleitas, M.; Valhuerdi Debesa, C.; Valle Cepero, R.; Hernandez, J.L.

    1996-01-01

    Features of different safety analysis techniques by means of calculation thermohydraulic a probabilistic and severe accidents used in the safety assessment, as well as the development of these techniques in Cuba and their use in support of regulatory decision making are presented

  19. Operational safety analysis status of Novi Han repository

    International Nuclear Information System (INIS)

    Boiadjiev, A.

    2000-01-01

    This article presents the status of the safety studies and activities related to Novi Han repository. The case of this facility is such that no clear boundary exists between post-closure safety assessment and operational safety assessment. The major findings of these activities are given. The Safety Analysis Report (SAR) for Novi Han repository is developed by Risk Engineering Ltd. under a contract with the Committee on the Use of Atomic Energy for Peaceful Purposes. The general structure and main conclusions and recommendations of the SAR are presented. (author)

  20. Safety analysis and evaluation methodology for fusion systems

    International Nuclear Information System (INIS)

    Fujii-e, Y.; Kozawa, Y.; Namba, C.

    1987-03-01

    Fusion systems which are under development as future energy systems have reached a stage that the break even is expected to be realized in the near future. It is desirable to demonstrate that fusion systems are well acceptable to the societal environment. There are three crucial viewpoints to measure the acceptability, that is, technological feasibility, economy and safety. These three points have close interrelation. The safety problem is more important since three large scale tokamaks, JET, TFTR and JT-60, start experiment, and tritium will be introduced into some of them as the fusion fuel. It is desirable to establish a methodology to resolve the safety-related issues in harmony with the technological evolution. The promising fusion system toward reactors is not yet settled. This study has the objective to develop and adequate methodology which promotes the safety design of general fusion systems and to present a basis for proposing the R and D themes and establishing the data base. A framework of the methodology, the understanding and modeling of fusion systems, the principle of ensuring safety, the safety analysis based on the function and the application of the methodology are discussed. As the result of this study, the methodology for the safety analysis and evaluation of fusion systems was developed. New idea and approach were presented in the course of the methodology development. (Kako, I.)

  1. Gap Analysis Approach for Construction Safety Program Improvement

    Directory of Open Access Journals (Sweden)

    Thanet Aksorn

    2007-06-01

    Full Text Available To improve construction site safety, emphasis has been placed on the implementation of safety programs. In order to successfully gain from safety programs, factors that affect their improvement need to be studied. Sixteen critical success factors of safety programs were identified from safety literature, and these were validated by safety experts. This study was undertaken by surveying 70 respondents from medium- and large-scale construction projects. It explored the importance and the actual status of critical success factors (CSFs. Gap analysis was used to examine the differences between the importance of these CSFs and their actual status. This study found that the most critical problems characterized by the largest gaps were management support, appropriate supervision, sufficient resource allocation, teamwork, and effective enforcement. Raising these priority factors to satisfactory levels would lead to successful safety programs, thereby minimizing accidents.

  2. Incorporating Traffic Control and Safety Hardware Performance Functions into Risk-based Highway Safety Analysis

    Directory of Open Access Journals (Sweden)

    Zongzhi Li

    2017-04-01

    Full Text Available Traffic control and safety hardware such as traffic signs, lighting, signals, pavement markings, guardrails, barriers, and crash cushions form an important and inseparable part of highway infrastructure affecting safety performance. Significant progress has been made in recent decades to develop safety performance functions and crash modification factors for site-specific crash predictions. However, the existing models and methods lack rigorous treatments of safety impacts of time-deteriorating conditions of traffic control and safety hardware. This study introduces a refined method for computing the Safety Index (SI as a means of crash predictions for a highway segment that incorporates traffic control and safety hardware performance functions into the analysis. The proposed method is applied in a computation experiment using five-year data on nearly two hundred rural and urban highway segments. The root-mean square error (RMSE, Chi-square, Spearman’s rank correlation, and Mann-Whitney U tests are employed for validation.

  3. Quantitative Safety and Security Analysis from a Communication Perspective

    Directory of Open Access Journals (Sweden)

    Boris Malinowsky

    2015-12-01

    Full Text Available This paper introduces and exemplifies a trade-off analysis of safety and security properties in distributed systems. The aim is to support analysis for real-time communication and authentication building blocks in a wireless communication scenario. By embedding an authentication scheme into a real-time communication protocol for safety-critical scenarios, we can rely on the protocol’s individual safety and security properties. The resulting communication protocol satisfies selected safety and security properties for deployment in safety-critical use-case scenarios with security requirements. We look at handover situations in a IEEE 802.11 wireless setup between mobile nodes and access points. The trade-offs involve application-layer data goodput, probability of completed handovers, and effect on usable protocol slots, to quantify the impact of security from a lower-layer communication perspective on the communication protocols. The results are obtained using the network simulator ns-3.

  4. Safety analysis and review system: a Department of Energy safety assurance tool

    International Nuclear Information System (INIS)

    Rosenthal, H.B.

    1981-01-01

    The concept of the Safety Analysis and Review System is not new. It has been used within the Department and its predecessor agencies, Atomic Energy Commission (AEC) and Energy Research and Development Administration (ERDA), for over 20 years. To minimize the risks from nuclear reactor and power plants, the AEC developed a process to support management authorization of each operation through identification and analysis of potential hazards and the measures taken to control them. As the agency evolved from AEC through ERDA to the Department of Energy, its responsibilities were broadened to cover a diversity of technologies, including those associated with the development of fossil, solar, and geothermal energy. Because the safety analysis process had proved effective in a technology of high potential hazard, the Department investigated the applicability of the process to the other technologies. This paper describes the system and discusses how it is implemented within the Department

  5. Upgrading the safety toolkit: Initiatives of the accident analysis subgroup

    International Nuclear Information System (INIS)

    O'Kula, K.R.; Chung, D.Y.

    1999-01-01

    Since its inception, the Accident Analysis Subgroup (AAS) of the Energy Facility Contractors Group (EFCOG) has been a leading organization promoting development and application of appropriate methodologies for safety analysis of US Department of Energy (DOE) installations. The AAS, one of seven chartered by the EFCOG Safety Analysis Working Group, has performed an oversight function and provided direction to several technical groups. These efforts have been instrumental toward formal evaluation of computer models, improving the pedigree on high-use computer models, and development of the user-friendly Accident Analysis Guidebook (AAG). All of these improvements have improved the analytical toolkit for best complying with DOE orders and standards shaping safety analysis reports (SARs) and related documentation. Major support for these objectives has been through DOE/DP-45

  6. 3D analysis methods - Study and seminar[BWR safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Daaviittila, A [Valtion Teknillinen Tutkimuskeskus (Finland)

    2003-10-01

    The first part of the report results from a study that was performed as a Nordic co-operation activity with active participation from Studsvik Scandpower and Westinghouse Atom in Sweden, and VTT in Finland. The purpose of the study was to identify and investigate the effects rising from using the 3D transient com-puter codes in BWR safety analysis, and their influence on the transient analysis methodology. One of the main questions involves the critical power ratio (CPR) calculation methodology. The present way, where the CPR calculation is per-formed with a separate hot channel calculation, can be artificially conservative. In the investigated cases, no dramatic minimum CPR effect coming from the 3D calculation is apparent. Some cases show some decrease in the transient change of minimum CPR with the 3D calculation, which confirms the general thinking that the 1D calculation is conservative. On the other hand, the observed effect on neutron flux behaviour is quite large. In a slower transient the 3D effect might be stronger. The second part of the report is a summary of a related seminar that was held on the 3D analysis methods. The seminar was sponsored by the Reactor Safety part (NKS-R) of the Nordic Nuclear Safety Research Programme (NKS). (au)

  7. Analysis approach for common cause failure on non-safety digital control system

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Yun Goo; Oh, Eungse [Korea Hydro and Nuclear Power Co. Ltd., Daejeon (Korea, Republic of)

    2014-05-15

    The effects of common cause failure (CCF) on safety digital instrumentation and control (I and C) system had been considered in defense in depth and diversity coping analysis with safety analysis method. For the non-safety system, single failure had been considered for safety analysis. IEEE Std. 603-1991, Clause 5.6.3.1(2), 'Isolation' states that no credible failure on the non-safety side of an isolation device shall prevent any portion of a safety system from meeting its minimum performance requirements during and following any design basis event requiring that safety function. The software CCF is one of the credible failure on the non-safety side. In advanced digital I and C system, same hardware component is used for different control system and the defect in manufacture or common external event can generate CCF. Moreover, the non-safety I and C system uses complex software for its various function and software quality assurance for the development process is less severe than safety software for the cost effective design. Therefore the potential defects in software cannot be ignored and the effect of software CCF on non-safety I and C system is needed to be evaluated. This paper proposes the general process and considerations for the analysis of CCF on non-safety I and C system.

  8. Safety analysis and synthesis using fuzzy sets and evidential reasoning

    International Nuclear Information System (INIS)

    Wang, J.; Yang, J.B.; Sen, P.

    1995-01-01

    This paper presents a new methodology for safety analysis and synthesis of a complex engineering system with a structure that is capable of being decomposed into a hierarchy of levels. In this methodology, fuzzy set theory is used to describe each failure event and an evidential reasoning approach is then employed to synthesise the information thus produced to assess the safety of the whole system. Three basic parameters--failure likelihood, consequence severity and failure consequence probability, are used to analyse a failure event. These three parameters are described by linguistic variables which are characterised by a membership function to the defined categories. As safety can also be clearly described by linguistic variables referred to as the safety expressions, the obtained fuzzy safety score can be mapped back to the safety expressions which are characterised by membership functions over the same categories. This mapping results in the identification of the safety of each failure event in terms of the degree to which the fuzzy safety score belongs to each of the safety expressions. Such degrees represent the uncertainty in safety evaluations and can be synthesised using an evidential reasoning approach so that the safety of the whole system can be evaluated in terms of these safety expressions. Finally, a practical engineering example is presented to demonstrate the proposed safety analysis and synthesis methodology

  9. Bayesian-network-based safety risk analysis in construction projects

    International Nuclear Information System (INIS)

    Zhang, Limao; Wu, Xianguo; Skibniewski, Miroslaw J.; Zhong, Jingbing; Lu, Yujie

    2014-01-01

    This paper presents a systemic decision support approach for safety risk analysis under uncertainty in tunnel construction. Fuzzy Bayesian Networks (FBN) is used to investigate causal relationships between tunnel-induced damage and its influential variables based upon the risk/hazard mechanism analysis. Aiming to overcome limitations on the current probability estimation, an expert confidence indicator is proposed to ensure the reliability of the surveyed data for fuzzy probability assessment of basic risk factors. A detailed fuzzy-based inference procedure is developed, which has a capacity of implementing deductive reasoning, sensitivity analysis and abductive reasoning. The “3σ criterion” is adopted to calculate the characteristic values of a triangular fuzzy number in the probability fuzzification process, and the α-weighted valuation method is adopted for defuzzification. The construction safety analysis progress is extended to the entire life cycle of risk-prone events, including the pre-accident, during-construction continuous and post-accident control. A typical hazard concerning the tunnel leakage in the construction of Wuhan Yangtze Metro Tunnel in China is presented as a case study, in order to verify the applicability of the proposed approach. The results demonstrate the feasibility of the proposed approach and its application potential. A comparison of advantages and disadvantages between FBN and fuzzy fault tree analysis (FFTA) as risk analysis tools is also conducted. The proposed approach can be used to provide guidelines for safety analysis and management in construction projects, and thus increase the likelihood of a successful project in a complex environment. - Highlights: • A systemic Bayesian network based approach for safety risk analysis is developed. • An expert confidence indicator for probability fuzzification is proposed. • Safety risk analysis progress is extended to entire life cycle of risk-prone events. • A typical

  10. Galileo and Ulysses missions safety analysis and launch readiness status

    International Nuclear Information System (INIS)

    Cork, M.J.; Turi, J.A.

    1989-01-01

    The Galileo spacecraft will explore the Jupiter system and Ulysses will fly by Jupiter en route to a polar orbit of the sun. Both spacecraft are powered by general purpose heat source radioisotope thermoelectric generators (RTGs). As a result of the Challenger accident and subsequent mission reprogramming, the Galileo and Ulysses missions' safety analysis had to be repeated. In addition to presenting an overview of the safety analysis status for the missions, this paper presents a brief review of the missions' objectives and design approaches, RTG design characteristics and development history, and a description of the safety analysis process. (author)

  11. Addendum 1 Composite Analysis for the Area 5 Radioactive Waste Management Site at the Nevada Test Site, Nye County, Nevada

    International Nuclear Information System (INIS)

    Vefa Yucel

    2001-01-01

    A disposal authorization statement (DAS) was issued by the U.S. Department of Energy/Headquarters (DOE/HQ) on December 5, 2000, authorizing the DOE's National Nuclear Security Administration Nevada Operations Office to continue the operation of the Area 5 Radioactive Waste Management Site (RWMS) at the Nevada Test Site for the disposal of low-level waste and mixed low-level waste. Prior to the issuance of the DAS, the Low-Level Waste Disposal Facility Federal Review Group (LFRG) had conducted reviews of the performance assessment (PA) and the composite analysis (CA) for the Area 5 RWMS, in accordance with the requirements of the DOE Radioactive Waste Management Order DOE O 435.1. A brief history of the reviews is as follows. (The reviews were conducted by independent review teams chartered by the LFRG; the review findings and recommendations were issued in review team reports to the LFRG.) The LFRG accepted the initial PA, with conditions, on August 30, 1996. Revision 2.1 to the PA was issued in January 1998, implementing the conditions of acceptance of the 1996 PA. The LFRG reviewed Revision 2.1 as part of the Area 5 RWMS CA review during 2000, and found it acceptable. The CA and the Supplemental Information provided in response to issues identified during the initial review of the CA were accepted by the LFRG. The Supplemental Information (including the responses to four key issues) is included in the Review Team Report to the LFRG, which recommends that it be incorporated into the CA and issued to all known holders of the CA. The Area 5 RWMS DAS requires that the Supplemental Information generated during the DOE/HQ review of the CA be incorporated into the CA within one year of the date of issuance of the DAS. This report, the first addendum to the Area 5 CA, is prepared to fulfill that requirement. The Supplemental Information includes the following: Issues Identified in the Review Team Report; Crosswalk Presentation; and Maintaining Doses As Low As Reasonably

  12. A root cause analysis project in a medication safety course.

    Science.gov (United States)

    Schafer, Jason J

    2012-08-10

    To develop, implement, and evaluate team-based root cause analysis projects as part of a required medication safety course for second-year pharmacy students. Lectures, in-class activities, and out-of-class reading assignments were used to develop students' medication safety skills and introduce them to the culture of medication safety. Students applied these skills within teams by evaluating cases of medication errors using root cause analyses. Teams also developed error prevention strategies and formally presented their findings. Student performance was assessed using a medication errors evaluation rubric. Of the 211 students who completed the course, the majority performed well on root cause analysis assignments and rated them favorably on course evaluations. Medication error evaluation and prevention was successfully introduced in a medication safety course using team-based root cause analysis projects.

  13. Establishment of joint application system of safety analysis codes between Korea and Vietnam

    International Nuclear Information System (INIS)

    Chung, Bub Dong; Kim, Kyung Doo; Park, Cheol; Bae, Sung Won; Baek, Won Pil; Song, Cheol hwa; Jeong, Jae Jun; Lee, Seung Wook; Hwang, Moon Kyu; Lee, Chang Sup

    2011-04-01

    The following KAERI-VAEI collaboration works have been performed during the 2 year project ('09.4∼'11.4). 1) On the job training of Vietnam code users(1st training for 4 VAEI staff-3 months. 2nd training for 3 VAEI staff- 3 month), 2) Lecture of nuclear safety analysis (30 hrs basic course and 30 hrs advanced course), 3) Review of safety analysis method (IAEA safety concept and requirements), 4) Collaborative assessment of safety analysis code MARS (13 conceptual problem, 2 separate effect test problem, 1 integral effect test problem), 5) Input deck preparation of standard PWR (Preparation of APR1400 input deck and safety analysis of DBA). VAEI staffs have been familiarized to Korean PWR safety assessment technology through the collaboration assessment work using a computer code developed in Korea. The lectures for Vietnamese research will be contributed to the utilization and cultivation of Korean safety technology. The collaborated assessment works will be used for the establishment of MARS based safety analysis system which is independent from US safety assessment system

  14. Establishment of joint application system of safety analysis codes between Korea and Vietnam

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Bub Dong; Kim, Kyung Doo; Park, Cheol; Bae, Sung Won; Baek, Won Pil; Song, Cheol hwa; Jeong, Jae Jun; Lee, Seung Wook; Hwang, Moon Kyu; Lee, Chang Sup [KAERI, Daejeon (Korea, Republic of)

    2011-04-15

    The following KAERI-VAEI collaboration works have been performed during the 2 year project ('09.4{approx}'11.4). 1) On the job training of Vietnam code users(1st training for 4 VAEI staff-3 months. 2nd training for 3 VAEI staff- 3 month), 2) Lecture of nuclear safety analysis (30 hrs basic course and 30 hrs advanced course), 3) Review of safety analysis method (IAEA safety concept and requirements), 4) Collaborative assessment of safety analysis code MARS (13 conceptual problem, 2 separate effect test problem, 1 integral effect test problem), 5) Input deck preparation of standard PWR (Preparation of APR1400 input deck and safety analysis of DBA). VAEI staffs have been familiarized to Korean PWR safety assessment technology through the collaboration assessment work using a computer code developed in Korea. The lectures for Vietnamese research will be contributed to the utilization and cultivation of Korean safety technology. The collaborated assessment works will be used for the establishment of MARS based safety analysis system which is independent from US safety assessment system

  15. Reliability analysis of Angra I safety systems

    International Nuclear Information System (INIS)

    Oliveira, L.F.S. de; Soto, J.B.; Maciel, C.C.; Gibelli, S.M.O.; Fleming, P.V.; Arrieta, L.A.

    1980-07-01

    An extensive reliability analysis of some safety systems of Angra I, are presented. The fault tree technique, which has been successfully used in most reliability studies of nuclear safety systems performed to date is employed. Results of a quantitative determination of the unvailability of the accumulator and the containment spray injection systems are presented. These results are also compared to those reported in WASH-1400. (E.G.) [pt

  16. Safety analysis methodologies for radioactive waste repositories in shallow ground

    International Nuclear Information System (INIS)

    1984-01-01

    The report is part of the IAEA Safety Series and is addressed to authorities and specialists responsible for or involved in planning, performing and/or reviewing safety assessments of shallow ground radioactive waste repositories. It discusses approaches that are applicable for safety analysis of a shallow ground repository. The methodologies, analysis techniques and models described are pertinent to the task of predicting the long-term performance of a shallow ground disposal system. They may be used during the processes of selection, confirmation and licensing of new sites and disposal systems or to evaluate the long-term consequences in the post-sealing phase of existing operating or inactive sites. The analysis may point out need for remedial action, or provide information to be used in deciding on the duration of surveillance. Safety analysis both general in nature and specific to a certain repository, site or design concept, are discussed, with emphasis on deterministic and probabilistic studies

  17. A risk-informed perspective on deterministic safety analysis of nuclear power plants

    International Nuclear Information System (INIS)

    Wan, P.T.

    2009-01-01

    In this work, the deterministic safety analysis (DSA) approach to nuclear safety is examined from a risk-informed perspective. One objective of safety analysis of a nuclear power plant is to demonstrate via analysis that the risks to the public from events or accidents that are within the design basis of the power plant are within acceptable levels with a high degree of assurance. This nuclear safety analysis objective can be translated into two requirements on the risk estimates of design basis events or accidents: the nominal risk estimate to the public must be shown to be within acceptable levels, and the uncertainty in the risk estimates must be shown to be small on an absolute or relative basis. The DSA approach combined with the defense-in-depth (DID) principle is a simplified safety analysis approach that attempts to achieve the above safety analysis objective in the face of potentially large uncertainties in the risk estimates of a nuclear power plant by treating the various uncertainty contributors using a stylized conservative binary (yes-no) approach, and applying multiple overlapping physical barriers and defense levels to protect against the release of radioactivity from the reactor. It is shown that by focusing on the consequence aspect of risk, the previous two nuclear safety analysis requirements on risk can be satisfied with the DSA-DID approach to nuclear safety. It is also shown the use of multiple overlapping physical barriers and defense levels in the traditional DSA-DID approach to nuclear safety is risk-informed in the sense that it provides a consistently high level of confidence in the validity of the safety analysis results for various design basis events or accidents with a wide range of frequency of occurrence. It is hoped that by providing a linkage between the consequence analysis approach in DSA with a risk-informed perspective, greater understanding of the limitation and capability of the DSA approach is obtained. (author)

  18. GAMBIT. The global and modular beyond-the-standard-model inference tool. Addendum for GAMBIT 1.1. Mathematica backends, SUSYHD interface and updated likelihoods

    Energy Technology Data Exchange (ETDEWEB)

    Athron, Peter; Balazs, Csaba [Monash University, School of Physics and Astronomy, Melbourne, VIC (Australia); Australian Research Council Centre of Excellence for Particle Physics at the Tera-scale (Australia); Bringmann, Torsten; Dal, Lars A.; Gonzalo, Tomas E.; Krislock, Abram; Raklev, Are [University of Oslo, Department of Physics, Oslo (Norway); Buckley, Andy [University of Glasgow, SUPA, School of Physics and Astronomy, Glasgow (United Kingdom); Chrzaszcz, Marcin [Universitaet Zuerich, Physik-Institut, Zurich (Switzerland); Polish Academy of Sciences, H. Niewodniczanski Institute of Nuclear Physics, Krakow (Poland); Conrad, Jan; Edsjoe, Joakim; Farmer, Ben; Lundberg, Johan [AlbaNova University Centre, Oskar Klein Centre for Cosmoparticle Physics, Stockholm (Sweden); Stockholm University, Department of Physics, Stockholm (Sweden); Cornell, Jonathan M. [McGill University, Department of Physics, Montreal, QC (Canada); Dickinson, Hugh [University of Minnesota, Minnesota Institute for Astrophysics, Minneapolis, MN (United States); Jackson, Paul; White, Martin [Australian Research Council Centre of Excellence for Particle Physics at the Tera-scale (Australia); University of Adelaide, Department of Physics, Adelaide, SA (Australia); Kvellestad, Anders; Savage, Christopher [NORDITA, Stockholm (Sweden); McKay, James [Imperial College London, Department of Physics, Blackett Laboratory, London (United Kingdom); Mahmoudi, Farvah [Lyon 1 Univ., ENS de Lyon, CNRS, Centre de Recherche Astrophysique de Lyon UMR5574, Saint-Genis-Laval (France); CERN, Theoretical Physics Department, Geneva (Switzerland); Institut Universitaire de France, Paris (France); Martinez, Gregory D. [University of California, Physics and Astronomy Department, Los Angeles, CA (United States); Putze, Antje [LAPTh, Universite de Savoie, CNRS, Annecy-le-Vieux (France); Ripken, Joachim [Max Planck Institute for Solar System Research, Goettingen (Germany); Rogan, Christopher [Harvard University, Department of Physics, Cambridge, MA (United States); Saavedra, Aldo [Australian Research Council Centre of Excellence for Particle Physics at the Tera-scale (Australia); University of Sydney, Centre for Translational Data Science, Faculty of Engineering and Information Technologies, School of Physics, Sydney, NSW (Australia); Scott, Pat [Imperial College London, Department of Physics, Blackett Laboratory, London (United Kingdom); Seo, Seon-Hee [Seoul National University, Department of Physics and Astronomy, Seoul (Korea, Republic of); Serra, Nicola [Universitaet Zuerich, Physik-Institut, Zurich (Switzerland); Weniger, Christoph [University of Amsterdam, GRAPPA, Institute of Physics, Amsterdam (Netherlands); Wild, Sebastian [DESY, Hamburg (Germany); Collaboration: The GAMBIT Collaboration

    2018-02-15

    In Ref. (GAMBIT Collaboration: Athron et. al., Eur. Phys. J. C.arXiv:1705.07908, 2017) we introduced the global-fitting framework GAMBIT. In this addendum, we describe a new minor version increment of this package. GAMBIT 1.1 includes full support for Mathematica backends, which we describe in some detail here. As an example, we backend SUSYHD (Vega and Villadoro, JHEP 07:159, 2015), which calculates the mass of the Higgs boson in the MSSM from effective field theory. We also describe updated likelihoods in PrecisionBit and DarkBit, and updated decay data included in DecayBit. (orig.)

  19. GAMBIT: the global and modular beyond-the-standard-model inference tool. Addendum for GAMBIT 1.1: Mathematica backends, SUSYHD interface and updated likelihoods

    Science.gov (United States)

    Athron, Peter; Balazs, Csaba; Bringmann, Torsten; Buckley, Andy; Chrząszcz, Marcin; Conrad, Jan; Cornell, Jonathan M.; Dal, Lars A.; Dickinson, Hugh; Edsjö, Joakim; Farmer, Ben; Gonzalo, Tomás E.; Jackson, Paul; Krislock, Abram; Kvellestad, Anders; Lundberg, Johan; McKay, James; Mahmoudi, Farvah; Martinez, Gregory D.; Putze, Antje; Raklev, Are; Ripken, Joachim; Rogan, Christopher; Saavedra, Aldo; Savage, Christopher; Scott, Pat; Seo, Seon-Hee; Serra, Nicola; Weniger, Christoph; White, Martin; Wild, Sebastian

    2018-02-01

    In Ref. (GAMBIT Collaboration: Athron et. al., Eur. Phys. J. C. arXiv:1705.07908, 2017) we introduced the global-fitting framework GAMBIT. In this addendum, we describe a new minor version increment of this package. GAMBIT 1.1 includes full support for Mathematica backends, which we describe in some detail here. As an example, we backend SUSYHD (Vega and Villadoro, JHEP 07:159, 2015), which calculates the mass of the Higgs boson in the MSSM from effective field theory. We also describe updated likelihoods in PrecisionBit and DarkBit, and updated decay data included in DecayBit.

  20. GAMBIT. The global and modular beyond-the-standard-model inference tool. Addendum for GAMBIT 1.1. Mathematica backends, SUSYHD interface and updated likelihoods

    International Nuclear Information System (INIS)

    Athron, Peter; Balazs, Csaba; Bringmann, Torsten; Dal, Lars A.; Gonzalo, Tomas E.; Krislock, Abram; Raklev, Are; Buckley, Andy; Chrzaszcz, Marcin; Conrad, Jan; Edsjoe, Joakim; Farmer, Ben; Lundberg, Johan; Cornell, Jonathan M.; Dickinson, Hugh; Jackson, Paul; White, Martin; Kvellestad, Anders; Savage, Christopher; McKay, James; Mahmoudi, Farvah; Martinez, Gregory D.; Putze, Antje; Ripken, Joachim; Rogan, Christopher; Saavedra, Aldo; Scott, Pat; Seo, Seon-Hee; Serra, Nicola; Weniger, Christoph; Wild, Sebastian

    2018-01-01

    In Ref. (GAMBIT Collaboration: Athron et. al., Eur. Phys. J. C.arXiv:1705.07908, 2017) we introduced the global-fitting framework GAMBIT. In this addendum, we describe a new minor version increment of this package. GAMBIT 1.1 includes full support for Mathematica backends, which we describe in some detail here. As an example, we backend SUSYHD (Vega and Villadoro, JHEP 07:159, 2015), which calculates the mass of the Higgs boson in the MSSM from effective field theory. We also describe updated likelihoods in PrecisionBit and DarkBit, and updated decay data included in DecayBit. (orig.)

  1. Safety of GM crops: compositional analysis.

    Science.gov (United States)

    Brune, Philip D; Culler, Angela Hendrickson; Ridley, William P; Walker, Kate

    2013-09-04

    The compositional analysis of genetically modified (GM) crops has continued to be an important part of the overall evaluation in the safety assessment program for these materials. The variety and complexity of genetically engineered traits and modes of action that will be used in GM crops in the near future, as well as our expanded knowledge of compositional variability and factors that can affect composition, raise questions about compositional analysis and how it should be applied to evaluate the safety of traits. The International Life Sciences Institute (ILSI), a nonprofit foundation whose mission is to provide science that improves public health and well-being by fostering collaboration among experts from academia, government, and industry, convened a workshop in September 2012 to examine these and related questions, and a series of papers has been assembled to describe the outcomes of that meeting.

  2. Multi-dimensional Code Development for Safety Analysis of LMR

    International Nuclear Information System (INIS)

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

    2006-08-01

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

  3. Job safety and awareness analysis of safety implementation among electrical workers in airport service company

    Directory of Open Access Journals (Sweden)

    Putra Perdana Suteja

    2018-01-01

    Full Text Available Electrical is a fundamental process in the company that has high risk and responsibility especially in public service company such as an airport. Hence, the company that operates activities in the airport has to identify and control the safety activities of workers. On the safety implementation, the lack of workers’ awareness is fundamental aspects to the safety failure. Therefore, this study aimed to analyse the safety awareness and identify risk in the electrical workplace. Safety awareness questionnaires are distributed to ten workers in order to analyse their awareness. Job safety analysis method used to identify the risk in the electrical workplace. The preliminary study stated that workers were not aware of personal protective equipment usage so that the awareness and behavioural need to be analysed. The result is the hazard was found such as electrical shock and noise for various intensity in the workplace. While electrical workers were aware of safety implementation but less of safety behaviour. Furthermore, the recommendation can be implemented are the implementation of behaviour-based safety (BBS, 5S implementation and accident report list.

  4. Analysis Method of Common Cause Failure on Non-safety Digital Control System

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Yun Goo; Oh, Eun Gse [KHNP, Daejeon (Korea, Republic of)

    2014-08-15

    The effects of common cause failure on safety digital instrumentation and control system had been considered in defense in depth analysis with safety analysis method. However, the effects of common cause failure on non-safety digital instrumentation and control system also should be evaluated. The common cause failure can be included in credible failure on the non-safety system. In the I and C architecture of nuclear power plant, many design feature has been applied for the functional integrity of control system. One of that is segmentation. Segmentation defenses the propagation of faults in the I and C architecture. Some of effects from common cause failure also can be limited by segmentation. Therefore, in this paper there are two type of failure mode, one is failures in one control group which is segmented, and the other is failures in multiple control group because that the segmentation cannot defense all effects from common cause failure. For each type, the worst failure scenario is needed to be determined, so the analysis method has been proposed in this paper. The evaluation can be qualitative when there is sufficient justification that the effects are bounded in previous safety analysis. When it is not bounded in previous safety analysis, additional analysis should be done with conservative assumptions method of previous safety analysis or best estimation method with realistic assumptions.

  5. Applicability of trends in nuclear safety analysis to space nuclear power systems

    International Nuclear Information System (INIS)

    Bari, R.A.

    1992-01-01

    A survey is presented of some current trends in nuclear safety analysis that may be relevant to space nuclear power systems. This includes: lessons learned from operating power reactor safety and licensing; approaches to the safety design of advanced and novel reactors and facilities; the roles of risk assessment, extremely unlikely accidents, safety goals/targets; and risk-benefit analysis and communication

  6. Human reliability analysis methods for probabilistic safety assessment

    International Nuclear Information System (INIS)

    Pyy, P.

    2000-11-01

    Human reliability analysis (HRA) of a probabilistic safety assessment (PSA) includes identifying human actions from safety point of view, modelling the most important of them in PSA models, and assessing their probabilities. As manifested by many incidents and studies, human actions may have both positive and negative effect on safety and economy. Human reliability analysis is one of the areas of probabilistic safety assessment (PSA) that has direct applications outside the nuclear industry. The thesis focuses upon developments in human reliability analysis methods and data. The aim is to support PSA by extending the applicability of HRA. The thesis consists of six publications and a summary. The summary includes general considerations and a discussion about human actions in the nuclear power plant (NPP) environment. A condensed discussion about the results of the attached publications is then given, including new development in methods and data. At the end of the summary part, the contribution of the publications to good practice in HRA is presented. In the publications, studies based on the collection of data on maintenance-related failures, simulator runs and expert judgement are presented in order to extend the human reliability analysis database. Furthermore, methodological frameworks are presented to perform a comprehensive HRA, including shutdown conditions, to study reliability of decision making, and to study the effects of wrong human actions. In the last publication, an interdisciplinary approach to analysing human decision making is presented. The publications also include practical applications of the presented methodological frameworks. (orig.)

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

    Directory of Open Access Journals (Sweden)

    Krzysiak Zbigniew

    2017-12-01

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

  8. Westinghouse Hanford Company safety analysis reports and technical safety requirements upgrade program

    International Nuclear Information System (INIS)

    Busche, D.M.

    1995-09-01

    During Fiscal Year 1992, the US Department of Energy, Richland Operations Office (RL) separately transmitted the following US Department of Energy (DOE) Orders to Westinghouse Hanford Company (WHC) for compliance: DOE 5480.21, ''Unreviewed Safety Questions,'' DOE 5480.22, ''Technical Safety Requirements,'' and DOE 5480.23, ''Nuclear Safety Analysis Reports.'' WHC has proceeded with its impact assessment and implementation process for the Orders. The Orders are closely-related and contain some requirements that are either identical, similar, or logically-related. Consequently, WHC has developed a strategy calling for an integrated implementation of the three Orders. The strategy is comprised of three primary objectives, namely: Obtain DOE approval of a single list of DOE-owned and WHC-managed Nuclear Facilities, Establish and/or upgrade the ''Safety Basis'' for each Nuclear Facility, and Establish a functional Unreviewed Safety Question (USQ) process to govern the management and preservation of the Safety Basis for each Nuclear Facility. WHC has developed policy-revision and facility-specific implementation plans to accomplish near-term tasks associated with the above strategic objectives. This plan, which as originally submitted in August 1993 and approved, provided an interpretation of the new DOE Nuclear Facility definition and an initial list of WHC-managed Nuclear Facilities. For each current existing Nuclear Facility, existing Safety Basis documents are identified and the plan/status is provided for the ISB. Plans for upgrading SARs and developing TSRs will be provided after issuance of the corresponding Rules

  9. Hazard Analysis and Safety Requirements for Small Drone Operations: To What Extent Do Popular Drones Embed Safety?

    Science.gov (United States)

    Plioutsias, Anastasios; Karanikas, Nektarios; Chatzimihailidou, Maria Mikela

    2018-03-01

    Currently, published risk analyses for drones refer mainly to commercial systems, use data from civil aviation, and are based on probabilistic approaches without suggesting an inclusive list of hazards and respective requirements. Within this context, this article presents: (1) a set of safety requirements generated from the application of the systems theoretic process analysis (STPA) technique on a generic small drone system; (2) a gap analysis between the set of safety requirements and the ones met by 19 popular drone models; (3) the extent of the differences between those models, their manufacturers, and the countries of origin; and (4) the association of drone prices with the extent they meet the requirements derived by STPA. The application of STPA resulted in 70 safety requirements distributed across the authority, manufacturer, end user, or drone automation levels. A gap analysis showed high dissimilarities regarding the extent to which the 19 drones meet the same safety requirements. Statistical results suggested a positive correlation between drone prices and the extent that the 19 drones studied herein met the safety requirements generated by STPA, and significant differences were identified among the manufacturers. This work complements the existing risk assessment frameworks for small drones, and contributes to the establishment of a commonly endorsed international risk analysis framework. Such a framework will support the development of a holistic and methodologically justified standardization scheme for small drone flights. © 2017 Society for Risk Analysis.

  10. Qualitative safety analysis in accelerator based systems

    International Nuclear Information System (INIS)

    Sarkar, P.K.; Chowdhury, Lekha M.

    2006-01-01

    In recent developments connected to high energy and high current accelerators, the accelerator driven systems (ADS) and the Radioactive Ion Beam (RIB) facilities come in the forefront of application. For medical and industrial applications high current accelerators often need to be located in populated areas. These facilities pose significant radiological hazard during their operation and accidental situations. We have done a qualitative evaluation of radiological safety analysis using the probabilistic safety analysis (PSA) methods for accelerator-based systems. The major contribution to hazard comes from a target rupture scenario in both ADS and RIB facilities. Other significant contributors to hazard in the facilities are also discussed using fault tree and event tree methodologies. (author)

  11. Human Resources Readiness as TSO for Deterministic Safety Analysis on the First NPP in Indonesia

    International Nuclear Information System (INIS)

    Sony Tjahyani, D. T.

    2010-01-01

    In government regulation no. 43 year 2006 it is mentioned that preliminary safety analysis report and final safety analysis report are one of requirements which should be applied in construction and operation licensing for commercial power reactor (NPPs). The purpose of safety analysis report is to confirm the adequacy and efficiency of provisions within the defence in depth of nuclear reactor. Deterministic analysis is used on the safety analysis report. One of the TSO task is to evaluate this report based on request of operator or regulatory body. This paper discusses about human resources readiness as TSO for deterministic safety analysis on the first NPP in Indonesia. The assessment is done by comparing the analysis step on SS-23 and SS-30 with human resources status of BATAN currently. The assessment results showed that human resources for deterministic safety analysis are ready as TSO especially to review preliminary safety analysis report and to revise final safety analysis report in licensing on the first NPP in Indonesia. Otherwise, to prepare the safety analysis report is still needed many competency human resources. (author)

  12. Safety Analysis Of Actinide Recycled Fast Power Reactor

    International Nuclear Information System (INIS)

    Taufik, Mohammad

    2001-01-01

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

  13. Development of evaluation method for software safety analysis techniques

    International Nuclear Information System (INIS)

    Huang, H.; Tu, W.; Shih, C.; Chen, C.; Yang, W.; Yih, S.; Kuo, C.; Chen, M.

    2006-01-01

    Full text: Full text: Following the massive adoption of digital Instrumentation and Control (I and C) system for nuclear power plant (NPP), various Software Safety Analysis (SSA) techniques are used to evaluate the NPP safety for adopting appropriate digital I and C system, and then to reduce risk to acceptable level. However, each technique has its specific advantage and disadvantage. If the two or more techniques can be complementarily incorporated, the SSA combination would be more acceptable. As a result, if proper evaluation criteria are available, the analyst can then choose appropriate technique combination to perform analysis on the basis of resources. This research evaluated the applicable software safety analysis techniques nowadays, such as, Preliminary Hazard Analysis (PHA), Failure Modes and Effects Analysis (FMEA), Fault Tree Analysis (FTA), Markov chain modeling, Dynamic Flowgraph Methodology (DFM), and simulation-based model analysis; and then determined indexes in view of their characteristics, which include dynamic capability, completeness, achievability, detail, signal/ noise ratio, complexity, and implementation cost. These indexes may help the decision makers and the software safety analysts to choose the best SSA combination arrange their own software safety plan. By this proposed method, the analysts can evaluate various SSA combinations for specific purpose. According to the case study results, the traditional PHA + FMEA + FTA (with failure rate) + Markov chain modeling (without transfer rate) combination is not competitive due to the dilemma for obtaining acceptable software failure rates. However, the systematic architecture of FTA and Markov chain modeling is still valuable for realizing the software fault structure. The system centric techniques, such as DFM and Simulation-based model analysis, show the advantage on dynamic capability, achievability, detail, signal/noise ratio. However, their disadvantage are the completeness complexity

  14. Probabilistic safety analysis procedures guide

    International Nuclear Information System (INIS)

    Papazoglou, I.A.; Bari, R.A.; Buslik, A.J.

    1984-01-01

    A procedures guide for the performance of probabilistic safety assessment has been prepared for interim use in the Nuclear Regulatory Commission programs. The probabilistic safety assessment studies performed are intended to produce probabilistic predictive models that can be used and extended by the utilities and by NRC to sharpen the focus of inquiries into a range of tissues affecting reactor safety. This guide addresses the determination of the probability (per year) of core damage resulting from accident initiators internal to the plant and from loss of offsite electric power. The scope includes analyses of problem-solving (cognitive) human errors, a determination of importance of the various core damage accident sequences, and an explicit treatment and display of uncertainties for the key accident sequences. Ultimately, the guide will be augmented to include the plant-specific analysis of in-plant processes (i.e., containment performance) and the risk associated with external accident initiators, as consensus is developed regarding suitable methodologies in these areas. This guide provides the structure of a probabilistic safety study to be performed, and indicates what products of the study are essential for regulatory decision making. Methodology is treated in the guide only to the extent necessary to indicate the range of methods which is acceptable; ample reference is given to alternative methodologies which may be utilized in the performance of the study

  15. Random safety auditing, root cause analysis, failure mode and effects analysis.

    Science.gov (United States)

    Ursprung, Robert; Gray, James

    2010-03-01

    Improving quality and safety in health care is a major concern for health care providers, the general public, and policy makers. Errors and quality issues are leading causes of morbidity and mortality across the health care industry. There is evidence that patients in the neonatal intensive care unit (NICU) are at high risk for serious medical errors. To facilitate compliance with safe practices, many institutions have established quality-assurance monitoring procedures. Three techniques that have been found useful in the health care setting are failure mode and effects analysis, root cause analysis, and random safety auditing. When used together, these techniques are effective tools for system analysis and redesign focused on providing safe delivery of care in the complex NICU system. Copyright 2010 Elsevier Inc. All rights reserved.

  16. Evaluation of safety assessment methodologies in Rocky Flats Risk Assessment Guide (1985) and Building 707 Final Safety Analysis Report (1987)

    International Nuclear Information System (INIS)

    Walsh, B.; Fisher, C.; Zigler, G.; Clark, R.A.

    1990-01-01

    FSARs. Rockwell International, as operating contractor at the Rocky Flats plant, conducted a safety analysis program during the 1980s. That effort resulted in Final Safety Analysis Reports (FSARs) for several buildings, one of them being the Building 707 Final Safety Analysis Report, June 87 (707FSAR) and a Plant Safety Analysis Report. Rocky Flats Risk Assessment Guide, March 1985 (RFRAG85) documents the methodologies that were used for those FSARs. Resources available for preparation of those Rocky Flats FSARs were very limited. After addressing the more pressing safety issues, some of which are described below, the present contractor (EG ampersand G) intends to conduct a program of upgrading the FSARs. This report presents the results of a review of the methodologies described in RFRAG85 and 707FSAR and contains suggestions that might be incorporated into the methodology for the FSAR upgrade effort

  17. Survey of systems safety analysis methods and their application to nuclear waste management systems

    International Nuclear Information System (INIS)

    Pelto, P.J.; Winegardner, W.K.; Gallucci, R.H.V.

    1981-11-01

    This report reviews system safety analysis methods and examines their application to nuclear waste management systems. The safety analysis methods examined include expert opinion, maximum credible accident approach, design basis accidents approach, hazard indices, preliminary hazards analysis, failure modes and effects analysis, fault trees, event trees, cause-consequence diagrams, G0 methodology, Markov modeling, and a general category of consequence analysis models. Previous and ongoing studies on the safety of waste management systems are discussed along with their limitations and potential improvements. The major safety methods and waste management safety related studies are surveyed. This survey provides information on what safety methods are available, what waste management safety areas have been analyzed, and what are potential areas for future study

  18. Survey of systems safety analysis methods and their application to nuclear waste management systems

    Energy Technology Data Exchange (ETDEWEB)

    Pelto, P.J.; Winegardner, W.K.; Gallucci, R.H.V.

    1981-11-01

    This report reviews system safety analysis methods and examines their application to nuclear waste management systems. The safety analysis methods examined include expert opinion, maximum credible accident approach, design basis accidents approach, hazard indices, preliminary hazards analysis, failure modes and effects analysis, fault trees, event trees, cause-consequence diagrams, G0 methodology, Markov modeling, and a general category of consequence analysis models. Previous and ongoing studies on the safety of waste management systems are discussed along with their limitations and potential improvements. The major safety methods and waste management safety related studies are surveyed. This survey provides information on what safety methods are available, what waste management safety areas have been analyzed, and what are potential areas for future study.

  19. Criticality safety analysis for plutonium dissolver using silver mediated electrolytic oxidation method

    International Nuclear Information System (INIS)

    Umeda, Miki; Sugikawa, Susumu; Nakamura, Kazuhito; Egashira, Tetsurou

    1998-08-01

    Design and construction of a plutonium dissolver using silver mediated electrolytic oxidation method are promoted in NUCEF. Criticality safety analysis for the plutonium dissolver is described in this report. The electrolytic plutonium dissolver consists of connection pipes and three pots for MOX powder supply, circulation and electrolysis. The criticality control for the dissolver is made by geometrically safe shape with mass limitation. Monte Carlo code KENO-IV using MGCL-137 library based on ENDF/B-IV was used for the criticality safety analysis for the plutonium dissolver. Considering the required size for construction and criticality safety, diameter of pot and distance between two pots were determined. On this condition, the criticality safety analysis for the plutonium dissolver with connection pipes was carried out. As the result of the criticality safety analysis, an effective neutron multiplication factor keff of 0.91 was obtained and the criticality safety of the plutonium dissolver was confirmed on the basis of criteria of ≤0.95. (author)

  20. K West integrated water treatment system subproject safety analysis document

    International Nuclear Information System (INIS)

    SEMMENS, L.S.

    1999-01-01

    This Accident Analysis evaluates unmitigated accident scenarios, and identifies Safety Significant and Safety Class structures, systems, and components for the K West Integrated Water Treatment System

  1. K West integrated water treatment system subproject safety analysis document

    Energy Technology Data Exchange (ETDEWEB)

    SEMMENS, L.S.

    1999-02-24

    This Accident Analysis evaluates unmitigated accident scenarios, and identifies Safety Significant and Safety Class structures, systems, and components for the K West Integrated Water Treatment System.

  2. Safety analysis of disposal of spent nuclear fuel

    International Nuclear Information System (INIS)

    Vieno, T.

    1994-04-01

    The spent fuel from the Olkiluoto NPP (TVO I and II) is planned to be disposed of in a repository to be constructed at a depth of about 500 meters in the crystalline bedrock. The thesis is dealing with the safety analysis of the disposal. The main topics presented in the thesis are: (1) The amount of radioactive properties of the spent fuel, (2) The canister design and the planned disposal concept, (3) The results of the preliminary site investigations, (4) Discussion of the multi-barrier principle, (5) The general principles and methodology of the TVO-92 safety analysis, (6) Groundwater flow analysis, (7) Durability and behaviour of the canister, (8) Biosphere analysis and reference scenario, and (9) The sensitivity and uncertainty analyses. (246 refs., 75 figs., 44 tabs.)

  3. Fault tree synthesis for software design analysis of PLC based safety-critical systems

    International Nuclear Information System (INIS)

    Koo, S. R.; Cho, C. H.; Seong, P. H.

    2006-01-01

    As a software verification and validation should be performed for the development of PLC based safety-critical systems, a software safety analysis is also considered in line with entire software life cycle. In this paper, we propose a technique of software safety analysis in the design phase. Among various software hazard analysis techniques, fault tree analysis is most widely used for the safety analysis of nuclear power plant systems. Fault tree analysis also has the most intuitive notation and makes both qualitative and quantitative analyses possible. To analyze the design phase more effectively, we propose a technique of fault tree synthesis, along with a universal fault tree template for the architecture modules of nuclear software. Consequently, we can analyze the safety of software on the basis of fault tree synthesis. (authors)

  4. Safety margins of operating reactors. Analysis of uncertainties and implications for decision making

    International Nuclear Information System (INIS)

    2003-01-01

    Maintaining safety in the design and operation of nuclear power plants (NPPs) is a very important task under the conditions of a challenging environment, affected by the deregulated electricity market and implementation of risk informed regulations. In Member States, advanced computer codes are widely used as safety analysis tools in the framework of licensing of new NPP projects, safety upgrading programmes of existing NPPs, periodic safety reviews, renewal of operating licences, use of the safety margins for reactor power uprating, better utilization of nuclear fuel and higher operational flexibility, for justification of lifetime extensions, development of new emergency operating procedures, analysis of operational events, and development of accident management programmes. The issue of inadequate quality of safety analysis is becoming important due to a general tendency to use advanced tools for better establishment and utilization of safety margins, while the existence of such margins assure that NPPs operate safely in all modes of operation and at all times. The most important safety margins relate to physical barriers against release of radioactive material, such as fuel matrix and fuel cladding, reactor coolant system boundary, and the containment. Typically, safety margins are determined with use of computational tools for safety analysis. Advanced best estimate computer codes are suggested e.g. in the IAEA Safety Guide on Safety Assessment and Verification for Nuclear Power Plants to be used for current safety analysis. Such computer codes require their careful application to avoid unjustified reduction in robustness of the reactor safety. The issue of uncertainties in safety analyses and their impact on evaluation of safety margins is addressed in a number of IAEA guidance documents, in particular in the Safety Report on Accident Analysis for Nuclear Power Plants. It is also discussed in various technical meetings and workshops devoted to this area. The

  5. Prototype application of best estimate and uncertainty safety analysis methodology to large LOCA analysis

    International Nuclear Information System (INIS)

    Luxat, J.C.; Huget, R.G.

    2001-01-01

    Development of a methodology to perform best estimate and uncertainty nuclear safety analysis has been underway at Ontario Power Generation for the past two and one half years. A key driver for the methodology development, and one of the major challenges faced, is the need to re-establish demonstrated safety margins that have progressively been undermined through excessive and compounding conservatism in deterministic analyses. The major focus of the prototyping applications was to quantify the safety margins that exist at the probable range of high power operating conditions, rather than the highly improbable operating states associated with Limit of the Envelope (LOE) assumptions. In LOE, all parameters of significance to the consequences of a postulated accident are assumed to simultaneously deviate to their limiting values. Another equally important objective of the prototyping was to demonstrate the feasibility of conducting safety analysis as an incremental analysis activity, as opposed to a major re-analysis activity. The prototype analysis solely employed prior analyses of Bruce B large break LOCA events - no new computer simulations were undertaken. This is a significant and novel feature of the prototyping work. This methodology framework has been applied to a postulated large break LOCA in a Bruce generating unit on a prototype basis. This paper presents results of the application. (author)

  6. Lithium-thionyl chloride cell system safety hazard analysis

    Science.gov (United States)

    Dampier, F. W.

    1985-03-01

    This system safety analysis for the lithium thionyl chloride cell is a critical review of the technical literature pertaining to cell safety and draws conclusions and makes recommendations based on this data. The thermodynamics and kinetics of the electrochemical reactions occurring during discharge are discussed with particular attention given to unstable SOCl2 reduction intermediates. Potentially hazardous reactions between the various cell components and discharge products or impurities that could occur during electrical or thermal abuse are described and the most hazardous conditions and reactions identified. Design factors influencing the safety of Li/SOCl2 cells, shipping and disposal methods and the toxicity of Li/SOCl2 battery components are additional safety issues that are also addressed.

  7. An intelligent hybrid system for surface coal mine safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Lilic, N.; Obradovic, I.; Cvjetic, A. [University of Belgrade, Belgrade (Serbia)

    2010-06-15

    Analysis of safety in surface coal mines represents a very complex process. Published studies on mine safety analysis are usually based on research related to accidents statistics and hazard identification with risk assessment within the mining industry. Discussion in this paper is focused on the application of AI methods in the analysis of safety in mining environment. Complexity of the subject matter requires a high level of expert knowledge and great experience. The solution was found in the creation of a hybrid system PROTECTOR, whose knowledge base represents a formalization of the expert knowledge in the mine safety field. The main goal of the system is the estimation of mining environment as one of the significant components of general safety state in a mine. This global goal is subdivided into a hierarchical structure of subgoals where each subgoal can be viewed as the estimation of a set of parameters (gas, dust, climate, noise, vibration, illumination, geotechnical hazard) which determine the general mine safety state and category of hazard in mining environment. Both the hybrid nature of the system and the possibilities it offers are illustrated through a case study using field data related to an existing Serbian surface coal mine.

  8. Presurized water reactor safety approach and analysis. From conception to experience feedback

    International Nuclear Information System (INIS)

    Libmann, J.

    1987-04-01

    This report deals in ten chapters, with the following subjects: 1. Safety approach methods; 2. Study of accidents; 3. Safety analysis; 4. Study of internal aggressions or those involved by the site; 5. Consideration of complementary situations; 6. Three Mile Island accident; 7. Safety during operation and experience feedback; 8. An example of analysis: steam generator closure plug; 9. Probabilistic safety evaluation; 10. Chernobyl accident. 30 refs [fr

  9. Safety evaluation status report for the prototype license application safety analysis report

    International Nuclear Information System (INIS)

    1989-07-01

    The US Nuclear Regulatory Commission (NRC) staff and consultants reviewed a Prototype License Application Safety Analysis Report (PLASAR) submitted by the US Department of Energy (DOE) for the earth-mounded concrete bunker (EMCB) alternative method of low-level radioactive waste disposal. The NRC reviewers relied extensively on the Standard Review Plan (SRP), Rev.1 (NUREG-1200), to evaluate the acceptability of the information provided in the EMCB PLASAR. The NRC staff selected certain review areas in the PLASAR for development of safety evaluation report input to provide examples of safety assessments that are necessary as part of a licensing review. Because of the fictitious nature of the assumed disposal site, and the decision to limit the review to essentially first-round review status, the NRC staff report is labeled a ''Safety Evaluation Status Report'' (SESR). Appendix A comprises the NRC review comments and questions on the information that DOE submitted in the PLASAR. The NRC concentrated its review on the design and operations-related portions of the EMCB PLASAR

  10. Safety and reliability analysis based on nonprobabilistic methods

    International Nuclear Information System (INIS)

    Kozin, I.O.; Petersen, K.E.

    1996-01-01

    Imprecise probabilities, being developed during the last two decades, offer a considerably more general theory having many advantages which make it very promising for reliability and safety analysis. The objective of the paper is to argue that imprecise probabilities are more appropriate tool for reliability and safety analysis, that they allow to model the behavior of nuclear industry objects more comprehensively and give a possibility to solve some problems unsolved in the framework of conventional approach. Furthermore, some specific examples are given from which we can see the usefulness of the tool for solving some reliability tasks

  11. Use of safety analysis results to support process operation

    International Nuclear Information System (INIS)

    Karvonen, I.; Heino, P.

    1990-01-01

    Safety and risk analysis carried out during the design phase of a process plant produces useful knowledge about the behavior and the disturbances of the system. This knowledge, however, often remains to the designer though it would be of benefit to the operators and supervisors of the process plant, too. In Technical Research Centre of Finland a project has been started to plan and construct a prototype of an information system to make use of the analysis knowledge during the operation phase. The project belongs to a Nordic KRM project (Knowledge Based Risk Management System). The information system is planned to base on safety and risk analysis carried out during the design phase and completed with operational experience. The safety analysis includes knowledge about potential disturbances, their causes and consequences in the form of Hazard and Operability Study, faut trees and/or event trees. During the operation disturbances can however, occur, which are not included in the safety analysis, or the causes or consequences of which have been incompletely identified. Thus the information system must also have an interface for the documentation of the operational knowledge missing from the analysis results. The main tasks off the system when supporting the management of a disturbance are to identify it (or the most important of the coexistent ones) from the stored knowledge and to present it in a proper form (for example as a deviation graph). The information system may also be used to transfer knowledge from one shift to another and to train process personnel

  12. Performance and safety analysis of WP-cave concept

    International Nuclear Information System (INIS)

    Skagius, K.; Svemar, C.

    1989-08-01

    The report presents a performance safety, and cost analysis of the WP-cave, WPC, concept. In the performance analysis, questions specific to the WPC have been addressed which have been identified to require more detailed studies. Based on the outcome of this analysis, a safety analysis has been made which comprises of the modeling and calculation of radionuclide transport from the repository to the biosphere and the resulting dose exposure to man. The result of the safety analysis indicates that the present design of a WPC repository may give unacceptably high doses. By improving the properties of the bentonite/sand barrier such that the hydraulic conductivity is reduced, or by changing the short-lived steel canisters to more long-lived canisters, e.g. copper canisters, it is judged possible to achieve a sufficiently low level of dose exposure rates to man. The cost for a WPC repository of the studied design is significantly higher than for a KBS-3 repository considering the Swedish conditions and the Swedish amount of spent fuel. The major costs are connected to the excavation and backfilling of the bentonite/sand barrier. The potential for cost savings is high but it is not judged possible to account for savings in such a way that the WPC concept shows lower cost than the KBS-3 concept. (34 figs., 33 tabs., 29 refs.)

  13. Probabilistic safety analysis of earth retaining structures during earthquakes

    Science.gov (United States)

    Grivas, D. A.; Souflis, C.

    1982-07-01

    A procedure is presented for determining the probability of failure of Earth retaining structures under static or seismic conditions. Four possible modes of failure (overturning, base sliding, bearing capacity, and overall sliding) are examined and their combined effect is evaluated with the aid of combinatorial analysis. The probability of failure is shown to be a more adequate measure of safety than the customary factor of safety. As Earth retaining structures may fail in four distinct modes, a system analysis can provide a single estimate for the possibility of failure. A Bayesian formulation of the safety retaining walls is found to provide an improved measure for the predicted probability of failure under seismic loading. The presented Bayesian analysis can account for the damage incurred to a retaining wall during an earthquake to provide an improved estimate for its probability of failure during future seismic events.

  14. Oak Ridge National Laboratory site data for safety-analysis report

    International Nuclear Information System (INIS)

    Fitzpatrick, F.C.

    1982-12-01

    The Oak Ridge National Laboratory site data contained herein were compiled in support of the United States Department of Energy (USDOE) Oak Ridge Operations Office Order OR 5481.1. That order sets forth assignment of responsibilities for safety analysis and review responsibilities and provides guidance relative to the content and format of safety analysis reports. The information presented in this document is intended for use by reference in individual safety analysis reports where applicable to support accident analyses or the establishment of design bases of significance to safety, and it is applicable only to Oak Ridge National Laboratory facilities in Bethel and Melton Valleys. This information includes broad descriptions of the site characteristics, radioactive waste handling and monitoring practices, and the organization and operating policies at Oak Ridge National Laboratory. The historical background of the Laboratory is discussed briefly and the overall physical situation of the facilities is described in the following paragraphs

  15. Oak Ridge National Laboratory site data for safety-analysis report

    Energy Technology Data Exchange (ETDEWEB)

    Fitzpatrick, F.C.

    1982-12-01

    The Oak Ridge National Laboratory site data contained herein were compiled in support of the United States Department of Energy (USDOE) Oak Ridge Operations Office Order OR 5481.1. That order sets forth assignment of responsibilities for safety analysis and review responsibilities and provides guidance relative to the content and format of safety analysis reports. The information presented in this document is intended for use by reference in individual safety analysis reports where applicable to support accident analyses or the establishment of design bases of significance to safety, and it is applicable only to Oak Ridge National Laboratory facilities in Bethel and Melton Valleys. This information includes broad descriptions of the site characteristics, radioactive waste handling and monitoring practices, and the organization and operating policies at Oak Ridge National Laboratory. The historical background of the Laboratory is discussed briefly and the overall physical situation of the facilities is described in the following paragraphs.

  16. 340 Waste handling Facility Hazard Categorization and Safety Analysis

    International Nuclear Information System (INIS)

    Rodovsky, T.J.

    2010-01-01

    The analysis presented in this document provides the basis for categorizing the facility as less than Hazard Category 3. The final hazard categorization for the deactivated 340 Waste Handling Facility (340 Facility) is presented in this document. This hazard categorization was prepared in accordance with DOE-STD-1 027-92, Change Notice 1, Hazard Categorization and Accident Analysis Techniques for Compliance with Doe Order 5480.23, Nuclear Safety Analysis Reports. The analysis presented in this document provides the basis for categorizing the facility as less than Hazard Category (HC) 3. Routine nuclear waste receiving, storage, handling, and shipping operations at the 340 Facility have been deactivated, however, the facility contains a small amount of radioactive liquid and/or dry saltcake in two underground vault tanks. A seismic event and hydrogen deflagration were selected as bounding accidents. The generation of hydrogen in the vault tanks without active ventilation was determined to achieve a steady state volume of 0.33%, which is significantly less than the lower flammability limit of 4%. Therefore, a hydrogen deflagration is not possible in these tanks. The unmitigated release from a seismic event was used to categorize the facility consistent with the process defined in Nuclear Safety Technical Position (NSTP) 2002-2. The final sum-of-fractions calculation concluded that the facility is less than HC 3. The analysis did not identify any required engineered controls or design features. The Administrative Controls that were derived from the analysis are: (1) radiological inventory control, (2) facility change control, and (3) Safety Management Programs (SMPs). The facility configuration and radiological inventory shall be controlled to ensure that the assumptions in the analysis remain valid. The facility commitment to SMPs protects the integrity of the facility and environment by ensuring training, emergency response, and radiation protection. The full scale

  17. Fuel reprocessing: safety analysis of extraction cycles

    International Nuclear Information System (INIS)

    Dinh, B.; Mauborgne, B.; Baron, P.; Mercier, J.P.

    1991-01-01

    An essential part of the safety analysis related to the extraction cycles of reprocessing plants, is the analysis of their behaviour during steady-state and transient operations, by means of simulation codes. These codes are based on the chemical properties of the main species involved (distribution coefficient and kinetics) and the hydrodynamics inside the contactors (mixer-settlers and pulsed columns). These codes have been consolidated by comparison of calculations with experimental results. The safety analysis is essentially performed in two steps. The first step is a parametric sensitivity analysis of the chemical flowsheet operated: the effect of a misadjustment (flowrate of feed, solvent, etc) is evaluated by successive steady-state calculations. These calculations help the identification of the sensitive parameters for the risk of plutonium accumulation, while indicating the permissible level of misadjustment. These calculations also serve to identify the parameters which should be measured during plant operation. The second step is the study of transient regimes, for the most sensitive parameters related to plutonium accumulation risk. The aim is to confirm the conclusions of the first step and to check that the characteristic process parameters chosen effectively allow, the early and reliable detection of any drift towards a plutonium accumulating regime. The procedures to drive the process backwards to a specified convenient steady-state regime from a drifting-state are also verified. The identification of the sensitive parameters, the process status parameters and the process transient analysis, allow a good control of process operation. This procedure, applied to the first purification cycle of COGEMA's UP3-A La Hague plant has demonstrated the total safety of facility operations

  18. SAFETY ANALYSIS METHODOLOGY FOR AGED CANDU® 6 NUCLEAR REACTORS

    Directory of Open Access Journals (Sweden)

    WOLFGANG HARTMANN

    2013-10-01

    Full Text Available This paper deals with the Safety Analysis for CANDU® 6 nuclear reactors as affected by main Heat Transport System (HTS aging. Operational and aging related changes of the HTS throughout its lifetime may lead to restrictions in certain safety system settings and hence some restriction in performance under certain conditions. A step in confirming safe reactor operation is the tracking of relevant data and their corresponding interpretation by the use of appropriate thermalhydraulic analytic models. Safety analyses ranging from the assessment of safety limits associated with the prevention of intermittent fuel sheath dryout for a slow Loss of Regulation (LOR analysis and fission gas release after a fuel failure are summarized. Specifically for fission gas release, the thermalhydraulic analysis for a fresh core and an 11 Effective Full Power Years (EFPY aged core was summarized, leading to the most severe stagnation break sizes for the inlet feeder break and the channel failure time. Associated coolant conditions provide the input data for fuel analyses. Based on the thermalhydraulic data, the fission product inventory under normal operating conditions may be calculated for both fresh and aged cores, and the fission gas release may be evaluated during the transient. This analysis plays a major role in determining possible radiation doses to the public after postulated accidents have occurred.

  19. Development of an auditable safety analysis in support of a radiological facility classification

    International Nuclear Information System (INIS)

    Kinney, M.D.; Young, B.

    1995-01-01

    In recent years, U.S. Department of Energy (DOE) facilities commonly have been classified as reactor, non-reactor nuclear, or nuclear facilities. Safety analysis documentation was prepared for these facilities, with few exceptions, using the requirements in either DOE Order 5481.1B, Safety Analysis and Review System; or DOE Order 5480.23, Nuclear Safety Analysis Reports. Traditionally, this has been accomplished by development of an extensive Safety Analysis Report (SAR), which identifies hazards, assesses risks of facility operation, describes and analyzes adequacy of measures taken to control hazards, and evaluates potential accidents and their associated risks. This process is complicated by analysis of secondary hazards and adequacy of backup (redundant) systems. The traditional SAR process is advantageous for DOE facilities with appreciable hazards or operational risks. SAR preparation for a low-risk facility or process can be cost-prohibitive and quite challenging because conventional safety analysis protocols may not readily be applied to a low-risk facility. The DOE Office of Environmental Restoration and Waste Management recognized this potential disadvantage and issued an EM limited technical standard, No. 5502-94, Hazard Baseline Documentation. This standard can be used for developing documentation for a facility classified as radiological, including preparation of an auditable (defensible) safety analysis. In support of the radiological facility classification process, the Uranium Mill Tailings Remedial Action (UMTRA) Project has developed an auditable safety analysis document based upon the postulation criteria and hazards analysis techniques defined in DOE Order 5480.23

  20. Valuation of road safety effects in cost-benefit analysis.

    Science.gov (United States)

    Wijnen, Wim; Wesemann, Paul; de Blaeij, Arianne

    2009-11-01

    Cost-benefit analysis is a common method for evaluating the social economic impact of transport projects, and in many of these projects the saving of human lives is an issue. This implies, within the framework of cost-benefit analysis, that a monetary value should be attached to saving human lives. This paper discusses the 'Value of a Statistical Life' (VoSL), a concept that is often used for monetising safety effects, in the context of road safety. Firstly, the concept of 'willingness to pay' for road safety and its relation to the VoSL are explained. The VoSL approach will be compared to other approaches to monetise safety effects, in particular the human capital approach and 'quality adjusted life years'. Secondly, methods to estimate the VoSL and their applicability to road safety will be discussed. Thirdly, the paper reviews the VoSL estimates that have been found in scientific research and compares them with the values that are used in policy evaluations. Finally, a VoSL study in the Netherlands will be presented as a case study, and its applicability in policy evaluation will be illustrated.

  1. The use of case tools in OPG safety analysis code qualification

    International Nuclear Information System (INIS)

    Pascoe, J.; Cheung, A.; Westbye, C.

    2001-01-01

    Ontario Power Generation (OPG) is currently qualifying its critical safety analysis software. The software quality assurance (SQA) framework is described. Given the legacy nature of much of the safety analysis software the reverse engineering methodology has been adopted. The safety analysis suite of codes was developed over a period of many years to differing standards of quality and had sparse or incomplete documentation. Key elements of the reverse engineering process require recovery of design information from existing coding. This recovery, if performed manually, could represent an enormous effort. Driven by a need to maximize productivity and enhance the repeatability and objectivity of software qualification activities the decision was made to acquire or develop and implement Computer Aided Software Engineering (CASE) tools. This paper presents relevant background information on CASE tools and discusses how the OPG SQA requirements were used to assess the suitability of available CASE tools. Key findings from the application of CASE tools to the qualification of the OPG safety analysis software are discussed. (author)

  2. Idaho National Engineering Laboratory (INEL) Environmental Restoration (ER) Program Baseline Safety Analysis File (BSAF)

    International Nuclear Information System (INIS)

    1995-09-01

    The Baseline Safety Analysis File (BSAF) is a facility safety reference document for the Idaho National Engineering Laboratory (INEL) environmental restoration activities. The BSAF contains information and guidance for safety analysis documentation required by the U.S. Department of Energy (DOE) for environmental restoration (ER) activities, including: Characterization of potentially contaminated sites. Remedial investigations to identify and remedial actions to clean up existing and potential releases from inactive waste sites Decontamination and dismantlement of surplus facilities. The information is INEL-specific and is in the format required by DOE-EM-STD-3009-94, Preparation Guide for U.S. Department of Energy Nonreactor Nuclear Facility Safety Analysis Reports. An author of safety analysis documentation need only write information concerning that activity and refer to BSAF for further information or copy applicable chapters and sections. The information and guidance provided are suitable for: sm-bullet Nuclear facilities (DOE Order 5480-23, Nuclear Safety Analysis Reports) with hazards that meet the Category 3 threshold (DOE-STD-1027-92, Hazard Categorization and Accident Analysis Techniques for Compliance with DOE Order 5480.23, Nuclear Safety Analysis Reports) sm-bullet Radiological facilities (DOE-EM-STD-5502-94, Hazard Baseline Documentation) Nonnuclear facilities (DOE-EM-STD-5502-94) that are classified as open-quotes lowclose quotes hazard facilities (DOE Order 5481.1B, Safety Analysis and Review System). Additionally, the BSAF could be used as an information source for Health and Safety Plans and for Safety Analysis Reports (SARs) for nuclear facilities with hazards equal to or greater than the Category 2 thresholds, or for nonnuclear facilities with open-quotes moderateclose quotes or open-quotes highclose quotes hazard classifications

  3. Idaho National Engineering Laboratory (INEL) Environmental Restoration (ER) Program Baseline Safety Analysis File (BSAF)

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-09-01

    The Baseline Safety Analysis File (BSAF) is a facility safety reference document for the Idaho National Engineering Laboratory (INEL) environmental restoration activities. The BSAF contains information and guidance for safety analysis documentation required by the U.S. Department of Energy (DOE) for environmental restoration (ER) activities, including: Characterization of potentially contaminated sites. Remedial investigations to identify and remedial actions to clean up existing and potential releases from inactive waste sites Decontamination and dismantlement of surplus facilities. The information is INEL-specific and is in the format required by DOE-EM-STD-3009-94, Preparation Guide for U.S. Department of Energy Nonreactor Nuclear Facility Safety Analysis Reports. An author of safety analysis documentation need only write information concerning that activity and refer to BSAF for further information or copy applicable chapters and sections. The information and guidance provided are suitable for: {sm_bullet} Nuclear facilities (DOE Order 5480-23, Nuclear Safety Analysis Reports) with hazards that meet the Category 3 threshold (DOE-STD-1027-92, Hazard Categorization and Accident Analysis Techniques for Compliance with DOE Order 5480.23, Nuclear Safety Analysis Reports) {sm_bullet} Radiological facilities (DOE-EM-STD-5502-94, Hazard Baseline Documentation) Nonnuclear facilities (DOE-EM-STD-5502-94) that are classified as {open_quotes}low{close_quotes} hazard facilities (DOE Order 5481.1B, Safety Analysis and Review System). Additionally, the BSAF could be used as an information source for Health and Safety Plans and for Safety Analysis Reports (SARs) for nuclear facilities with hazards equal to or greater than the Category 2 thresholds, or for nonnuclear facilities with {open_quotes}moderate{close_quotes} or {open_quotes}high{close_quotes} hazard classifications.

  4. Preliminary safety analysis for key design features of KALIMER

    Energy Technology Data Exchange (ETDEWEB)

    Hahn, D. H.; Kwon, Y. M.; Chang, W. P.; Suk, S. D.; Lee, S. O.; Lee, Y. B.; Jeong, K. S

    2000-07-01

    KAERI is currently developing the conceptual design of a liquid metal reactor, KALIMER(Korea Advanced Liquid Metal Reactor) under the long-term nuclear R and D program. In this report, descriptions of the KALIMER safety design features and safety analyses results for selected ATWS accidents are presented. First, the basic approach to achieve the safety goal is introduced in chapter 1, and the safety evaluation procedure for the KALIMER design is described in chapter 2. It includes event selection, event categorization, description of design basis events, and beyond design basis events. In chapter 3, results of inherent safety evaluations for the KALIMER conceptual design are presented. The KALIMER core and plant system are designed to assure design performance during a selected set of events without either reactor control or protection system intervention. Safety analyses for the postulated anticipated transient without scram(ATWS) have been performed to investigate the KALIMER system response to the events. They are categorized as bounding events(BEs) because of their low probability of occurrence. In chapter 4, the design of the KALIMER containment dome and the results of its performance analysis are presented. The designs of the existing LMR containment and the KALIMER containment dome have been compared in this chapter. Procedure of the containment performance analysis and the analysis results are described along with the accident scenario and source terms. Finally, a simple methodology is introduced to investigate the core kinetics and hydraulic behavior during HCDA in chapter 5. Mathematical formulations have been developed in the framework of the modified bethe-tait method, and scoping analyses have been performed for the KALIMER core behavior during super-prompt critical excursions.

  5. Introduction to Safety Analysis Approach for Research Reactors

    International Nuclear Information System (INIS)

    Park, Suki

    2016-01-01

    The research reactors have a wide variety in terms of thermal powers, coolants, moderators, reflectors, fuels, reactor tanks and pools, flow direction in the core, and the operating pressure and temperature of the cooling system. Around 110 research reactors have a thermal power greater than 1 MW. This paper introduces a general approach to safety analysis for research reactors and deals with the experience of safety analysis on a 10 MW research reactor with an open-pool and open-tank reactor and a downward flow in the reactor core during normal operation. The general approach to safety analysis for research reactors is described and the design features of a typical open-pool and open-tank type reactor are discussed. The representative events expected in research reactors are investigated. The reactor responses and the thermal hydraulic behavior to the events are presented and discussed. From the minimum CHFR and the maximum fuel temperature calculated, it is ensured that the fuel is not damaged in the step insertion of reactivity by 1.8 mk and the failure of all primary pumps for the reactor with a 10 MW thermal power and downward core flow

  6. Computational methods for nuclear criticality safety analysis

    International Nuclear Information System (INIS)

    Maragni, M.G.

    1992-01-01

    Nuclear criticality safety analyses require the utilization of methods which have been tested and verified against benchmarks results. In this work, criticality calculations based on the KENO-IV and MCNP codes are studied aiming the qualification of these methods at the IPEN-CNEN/SP and COPESP. The utilization of variance reduction techniques is important to reduce the computer execution time, and several of them are analysed. As practical example of the above methods, a criticality safety analysis for the storage tubes for irradiated fuel elements from the IEA-R1 research has been carried out. This analysis showed that the MCNP code is more adequate for problems with complex geometries, and the KENO-IV code shows conservative results when it is not used the generalized geometry option. (author)

  7. Advances in methods and applications of reliability and safety analysis

    International Nuclear Information System (INIS)

    Fieandt, J.; Hossi, H.; Laakso, K.; Lyytikaeinen, A.; Niemelae, I.; Pulkkinen, U.; Pulli, T.

    1986-01-01

    The know-how of the reliability and safety design and analysis techniques of Vtt has been established over several years in analyzing the reliability in the Finnish nuclear power plants Loviisa and Olkiluoto. This experience has been later on applied and developed to be used in the process industry, conventional power industry, automation and electronics. VTT develops and transfers methods and tools for reliability and safety analysis to the private and public sectors. The technology transfer takes place in joint development projects with potential users. Several computer-aided methods, such as RELVEC for reliability modelling and analysis, have been developed. The tool developed are today used by major Finnish companies in the fields of automation, nuclear power, shipbuilding and electronics. Development of computer-aided and other methods needed in analysis of operating experience, reliability or safety is further going on in a number of research and development projects

  8. Safety Analysis for Enlargement of Allowance Band of Main Steam Safety Valve Opening Setpoint of Wolsong Unit 1

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sungmin [Korea Hydro and Nuclear Power Co., Ltd., Daejeon (Korea, Republic of); Kim, Jonghyun; Cho, Cheonhwey [Atomic Creative Technology Co., Ltd., Daejeon (Korea, Republic of)

    2013-05-15

    The target events were selected to be the two most secondary system pressurization events - Loss of Class IV Power (LOCL4) and Loss of Condenser Vacuum (LOCV). In the actual analysis, an uncertainty of 1% was added to be conservative, so an allowance band of ±4% was used. A safety analysis was performed with CATHENA code to evaluate the safety of increasing the opening setpoint allowance band of MSSVs in WSNPP-1 The analysis results for both LOCL4 and LOCV confirm that the enlarged allowance would bring no harm to the safety of the plant from the viewpoint of fuel integrity and pressure boundary integrity. Therefore, the new allowance band of MSSVs will be incorporated into the Technical Specifications of WSNPP-1.

  9. Applying importance-performance analysis to patient safety culture.

    Science.gov (United States)

    Lee, Yii-Ching; Wu, Hsin-Hung; Hsieh, Wan-Lin; Weng, Shao-Jen; Hsieh, Liang-Po; Huang, Chih-Hsuan

    2015-01-01

    The Sexton et al.'s (2006) safety attitudes questionnaire (SAQ) has been widely used to assess staff's attitudes towards patient safety in healthcare organizations. However, to date there have been few studies that discuss the perceptions of patient safety both from hospital staff and upper management. The purpose of this paper is to improve and to develop better strategies regarding patient safety in healthcare organizations. The Chinese version of SAQ based on the Taiwan Joint Commission on Hospital Accreditation is used to evaluate the perceptions of hospital staff. The current study then lies in applying importance-performance analysis technique to identify the major strengths and weaknesses of the safety culture. The results show that teamwork climate, safety climate, job satisfaction, stress recognition and working conditions are major strengths and should be maintained in order to provide a better patient safety culture. On the contrary, perceptions of management and hospital handoffs and transitions are important weaknesses and should be improved immediately. Research limitations/implications - The research is restricted in generalizability. The assessment of hospital staff in patient safety culture is physicians and registered nurses. It would be interesting to further evaluate other staff's (e.g. technicians, pharmacists and others) opinions regarding patient safety culture in the hospital. Few studies have clearly evaluated the perceptions of healthcare organization management regarding patient safety culture. Healthcare managers enable to take more effective actions to improve the level of patient safety by investigating key characteristics (either strengths or weaknesses) that healthcare organizations should focus on.

  10. Discussion on safety analysis approach for sodium fast reactors

    International Nuclear Information System (INIS)

    Hong, Soon Joon; Choo, Yeon Joon; Suh, Nam Duk; Shin, Ahn Dong; Bae, Moo Hoon

    2012-01-01

    Utilization of nuclear energy is increasingly necessary not only because of the increasing energy consumption but also because of the controls on greenhouse emissions against global warming. To keep step with such demands, advanced reactors are now world widely under development with the aims of highly economical advances, and enhanced safety. Recently, further elaborating is encouraged on the research and development program for Generation IV (GEN IV) reactors, and in collaboration with other interested countries through the Generation IV International Forum (GIF). Sodium cooled Fast Reactor (SFR) is a strong contender amongst the GEN IV reactor concepts. Korea also takes part in that program and plans to construct demonstration reactor of SFR. SFR is under the development for a candidate of small modular reactors, for example, PRISM (Power Reactor Innovative Small Module). Understanding of safety analysis approach has also advanced by the demand of increasing comprehensive safety requirement. Reviewing the past development of the licensing and safety basis in the advanced reactors, such approaches seemed primarily not so satisfactory because the reference framework of licensing and safety analysis approach in the advanced reactors was always the one in water reactors. And, the framework is very plant specific one and thereby the advanced reactors and their frameworks don't look like a well assorted couple. Recently as a result of considerable advances in probabilistic safety assessment (PSA), risk informed approaches are increasingly applied together with some of the deterministic approaches like as the ones in water reactors. Technology neutral framework (TNF) can be said to be the utmost works of such risk informed approaches, even though an intensive assessment of the applicability has not been sufficiently accomplished. This study discusses the viable safety analysis approaches for the urgent application to the construction of pool type SFR. As discussed in

  11. Process hazards analysis (PrHA) program, bridging accident analyses and operational safety

    International Nuclear Information System (INIS)

    Richardson, J.A.; McKernan, S.A.; Vigil, M.J.

    2003-01-01

    Recently the Final Safety Analysis Report (FSAR) for the Plutonium Facility at Los Alamos National Laboratory, Technical Area 55 (TA-55) was revised and submitted to the US. Department of Energy (DOE). As a part of this effort, over seventy Process Hazards Analyses (PrHAs) were written and/or revised over the six years prior to the FSAR revision. TA-55 is a research, development, and production nuclear facility that primarily supports US. defense and space programs. Nuclear fuels and material research; material recovery, refining and analyses; and the casting, machining and fabrication of plutonium components are some of the activities conducted at TA-35. These operations involve a wide variety of industrial, chemical and nuclear hazards. Operational personnel along with safety analysts work as a team to prepare the PrHA. PrHAs describe the process; identi fy the hazards; and analyze hazards including determining hazard scenarios, their likelihood, and consequences. In addition, the interaction of the process to facility systems, structures and operational specific protective features are part of the PrHA. This information is rolled-up to determine bounding accidents and mitigating systems and structures. Further detailed accident analysis is performed for the bounding accidents and included in the FSAR. The FSAR is part of the Documented Safety Analysis (DSA) that defines the safety envelope for all facility operations in order to protect the worker, the public, and the environment. The DSA is in compliance with the US. Code of Federal Regulations, 10 CFR 830, Nuclear Safety Management and is approved by DOE. The DSA sets forth the bounding conditions necessary for the safe operation for the facility and is essentially a 'license to operate.' Safely of day-to-day operations is based on Hazard Control Plans (HCPs). Hazards are initially identified in the PrI-IA for the specific operation and act as input to the HCP. Specific protective features important to worker

  12. Safety analysis of the proposed Canadian geologic nuclear waste repository

    International Nuclear Information System (INIS)

    Prowse, D.R.

    1977-01-01

    The Canadian program for development and qualification of a geologic repository for emplacement of high-level and long-lived, alpha-emitting waste from irradiated nuclear fuel has been inititiated and is in its initial development stage. Fieldwork programs to locate candidate sites with suitable geological characteristics have begun. Laboratory studies and development of models for use in safety analysis of the emplaced nuclear waste have been initiated. The immediate objective is to complete a simplified safety analysis of a model geologic repository by mid-1978. This analysis will be progressively updated and will form part of an environmental Assessment Report of a Model Fuel Center which will be issued in mid-1979. The long-term objectives are to develop advanced safety assessment models of a geologic repository which will be available by 1980

  13. Safety analysis report for packaging (onsite) steel drum

    International Nuclear Information System (INIS)

    McCormick, W.A.

    1998-01-01

    This Safety Analysis Report for Packaging (SARP) provides the analyses and evaluations necessary to demonstrate that the steel drum packaging system meets the transportation safety requirements of HNF-PRO-154, Responsibilities and Procedures for all Hazardous Material Shipments, for an onsite packaging containing Type B quantities of solid and liquid radioactive materials. The basic component of the steel drum packaging system is the 208 L (55-gal) steel drum

  14. Use of safety analysis to site comfirmation procedure in case of hard rock repository

    International Nuclear Information System (INIS)

    Peltonen, E.K.

    1984-02-01

    The role of safety analysis in a confirmation procedure of a candidate disposal site of radioactive wastes is discussed. Items dealt with include principle reasons and practical goals of the use of safety analysis, methodology of safety analysis and assessment, as well as usefulness and adequacy of the present safety analysis. Safety analysis is a tool, which enables one to estimate quantitatively the possible radiological impacts from the disposal. The results can be compared with the criteria and the suitability conclusions drawn. Because of its systems analytical nature safety analysis is an effective method to reveal, what are the most important factors of the disposal system and the most critical site characteristics inside the lumped parameters often provided by the experimental site investigation methods. Furthermore it gives information on the accuracy needs of different site properties. This can be utilized to judge whether the quality and quantity of the measurements for the characterization are sufficient as well as to guide the further site investigations. A more practical discussion regarding the applicability of the use of safety analysis is presented by an example concerning the assessment of a Finnish candidate site for low- and intermediate-level radioactive waste repository. (author)

  15. LFR safety approach and main ELFR safety analysis results

    International Nuclear Information System (INIS)

    Bubelis, E.; Schikorr, M.; Frogheri, M.; Mansani, L.; Bandini, G.; Burgazzi, L.; Mikityuk, K.; Zhang, Y.; Lo Frano, R.; Forgione, N.

    2013-01-01

    LFR safety approach: → A global safety approach for the LFR reference plant has been assessed and the safety analyses methodology has been developed. → LFR follows the general guidelines of the Generation IV safety concept recommendations. Thus, improved safety and higher reliability are recognized as an essential priority. → The fundamental safety objectives and the Defence-in-Depth (DiD) approach, as described by IAEA Safety Guides, have been preserved. → The recommendations of the Risk and Safety Working Group (RSWG) of GEN-IV IF has been taken into account: • safety is to be “built-in” in the fundamental design rather than “added on”; • full implementation of the Defence-in-Depth principles in a manner that is demonstrably exhaustive, progressive, tolerant, forgiving and well-balanced; • “risk-informed” approach - deterministic approach complemented with a probabilistic one; • adoption of an integrated methodology that can be used to evaluate and document the safety of Gen IV nuclear systems - ISAM. In particular the OPT tool is the fundamental methodology used throughout the design process

  16. Idaho National Engineering Laboratory (INEL) Environmental Restoration Program (ERP), Baseline Safety Analysis File (BSAF). Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    1994-06-20

    This document was prepared to take the place of a Safety Evaluation Report since the Baseline Safety Analysis File (BSAF)and associated Baseline Technical Safety Requirements (TSR) File do not meet the requirements of a complete safety analysis documentation. Its purpose is to present in summary form the background of how the BSAF and Baseline TSR originated and a description of the process by which it was produced and approved for use in the Environmental Restoration Program.The BSAF is a facility safety reference document for INEL environmental restoration activities including environmental remediation of inactive waste sites and decontamination and decommissioning (D&D) of surplus facilities. The BSAF contains safety bases common to environmental restoration activities and guidelines for performing and documenting safety analysis. The common safety bases can be incorporated by reference into the safety analysis documentation prepared for individual environmental restoration activities with justification and any necessary revisions. The safety analysis guidelines in BSAF provide an accepted method for hazard analysis; analysis of normal, abnormal, and accident conditions; human factors analysis; and derivation of TSRS. The BSAF safety bases and guidelines are graded for environmental restoration activities.

  17. Idaho National Engineering Laboratory (INEL) Environmental Restoration Program (ERP), Baseline Safety Analysis File (BSAF). Revision 1

    International Nuclear Information System (INIS)

    1994-01-01

    This document was prepared to take the place of a Safety Evaluation Report since the Baseline Safety Analysis File (BSAF)and associated Baseline Technical Safety Requirements (TSR) File do not meet the requirements of a complete safety analysis documentation. Its purpose is to present in summary form the background of how the BSAF and Baseline TSR originated and a description of the process by which it was produced and approved for use in the Environmental Restoration Program.The BSAF is a facility safety reference document for INEL environmental restoration activities including environmental remediation of inactive waste sites and decontamination and decommissioning (D ampersand D) of surplus facilities. The BSAF contains safety bases common to environmental restoration activities and guidelines for performing and documenting safety analysis. The common safety bases can be incorporated by reference into the safety analysis documentation prepared for individual environmental restoration activities with justification and any necessary revisions. The safety analysis guidelines in BSAF provide an accepted method for hazard analysis; analysis of normal, abnormal, and accident conditions; human factors analysis; and derivation of TSRS. The BSAF safety bases and guidelines are graded for environmental restoration activities

  18. Probabilistic safety analysis and human reliability analysis. Proceedings. Working material

    International Nuclear Information System (INIS)

    1996-01-01

    An international meeting on Probabilistic Safety Assessment (PSA) and Human Reliability Analysis (HRA) was jointly organized by Electricite de France - Research and Development (EDF DER) and SRI International in co-ordination with the International Atomic Energy Agency. The meeting was held in Paris 21-23 November 1994. A group of international and French specialists in PSA and HRA participated at the meeting and discussed the state of the art and current trends in the following six topics: PSA Methodology; PSA Applications; From PSA to Dependability; Incident Analysis; Safety Indicators; Human Reliability. For each topic a background paper was prepared by EDF/DER and reviewed by the international group of specialists who attended the meeting. The results of this meeting provide a comprehensive overview of the most important questions related to the readiness of PSA for specific uses and areas where further research and development is required. Refs, figs, tabs

  19. Probabilistic safety analysis and human reliability analysis. Proceedings. Working material

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-12-31

    An international meeting on Probabilistic Safety Assessment (PSA) and Human Reliability Analysis (HRA) was jointly organized by Electricite de France - Research and Development (EDF DER) and SRI International in co-ordination with the International Atomic Energy Agency. The meeting was held in Paris 21-23 November 1994. A group of international and French specialists in PSA and HRA participated at the meeting and discussed the state of the art and current trends in the following six topics: PSA Methodology; PSA Applications; From PSA to Dependability; Incident Analysis; Safety Indicators; Human Reliability. For each topic a background paper was prepared by EDF/DER and reviewed by the international group of specialists who attended the meeting. The results of this meeting provide a comprehensive overview of the most important questions related to the readiness of PSA for specific uses and areas where further research and development is required. Refs, figs, tabs.

  20. Conception of a PWR simulator as a tool for safety analysis

    International Nuclear Information System (INIS)

    Lanore, J.M.; Bernard, P.; Romeyer Dherbey, J.; Bonnet, C.; Quilchini, P.

    1982-09-01

    A simulator can be a very useful tool for safety analysis to study accident sequences involving malfunctions of the systems and operator interventions. The main characteristics of the simulator SALAMANDRE (description of the systems, physical models, programming organization, control desk) have then been selected according tot he objectives of safety analysis

  1. 10 CFR 52.157 - Contents of applications; technical information in final safety analysis report.

    Science.gov (United States)

    2010-01-01

    ...; technical information in final safety analysis report. The application must contain a final safety analysis... 10 Energy 2 2010-01-01 2010-01-01 false Contents of applications; technical information in final safety analysis report. 52.157 Section 52.157 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) LICENSES...

  2. Deterministic and probabilistic approach to safety analysis

    International Nuclear Information System (INIS)

    Heuser, F.W.

    1980-01-01

    The examples discussed in this paper show that reliability analysis methods fairly well can be applied in order to interpret deterministic safety criteria in quantitative terms. For further improved extension of applied reliability analysis it has turned out that the influence of operational and control systems and of component protection devices should be considered with the aid of reliability analysis methods in detail. Of course, an extension of probabilistic analysis must be accompanied by further development of the methods and a broadening of the data base. (orig.)

  3. Safety Analysis Report - Packages, 9965, 9968, 9972-9975 Packages

    International Nuclear Information System (INIS)

    Van Alstine, M.N.

    1999-01-01

    This Safety Analysis Report for Packaging (SARP) documents the performance of the 9965 B, 9968 B, 9972 B(U), 9973 B(U), 9974 B(U), and 9975 B(U) packages in satisfying the regulatory safety requirements of the Code of Federal Regulations (CFR) 711 and the International Atomic Energy Agency (IAEA) Safety Series No. 6, Regulations for the Safe Transport of Radioactive Material, 1985 edition2. Results of the analysis and testing performed on the 9965 B, 9968 B, 9972 B(U), 9973 B(U), 9974 B(U), and 9975 B(U) packages are presented in this SARP, which was prepared in accordance with U.S. Department of energy (DOE) Order 5480.33 and in the format specified in the Nuclear Regulatory Commission (NRC) Regulatory Guides 7.94 and 7.10.5

  4. Safety analysis report - packages 9965, 9968, 9972-9975 packages

    International Nuclear Information System (INIS)

    Van Alstine, M.N.

    1997-10-01

    This Safety Analysis Report for Packaging (SARP) documents the performance of the 9965 B( ), 9968 B( ), 9972 B(U), 9973 B(U), 9974 B(U), and 9975 B(U) packages in satisfying the regulatory safety requirements of the Code of Federal Regulations (CFR) 10 CFR 71 and the International Atomic Energy Agency (IAEA) Safety Series No. 6, Regulations for the Safe Transport of Radioactive Material, 1985 edition. Results of the analysis and testing performed on the 9965 B(), 9968 B(), 9972 B(U), 9973 B(U), and 9975 B(U) packages are presented in this SARP, which was prepared in accordance with U.S. Department of Energy (DOE) Order 5480.3 and in the format specified in the Nuclear Regulatory Commission (NRC) Regulatory Guides 7.9 and 7.10

  5. Computational methods for criticality safety analysis within the scale system

    International Nuclear Information System (INIS)

    Parks, C.V.; Petrie, L.M.; Landers, N.F.; Bucholz, J.A.

    1986-01-01

    The criticality safety analysis capabilities within the SCALE system are centered around the Monte Carlo codes KENO IV and KENO V.a, which are both included in SCALE as functional modules. The XSDRNPM-S module is also an important tool within SCALE for obtaining multiplication factors for one-dimensional system models. This paper reviews the features and modeling capabilities of these codes along with their implementation within the Criticality Safety Analysis Sequences (CSAS) of SCALE. The CSAS modules provide automated cross-section processing and user-friendly input that allow criticality safety analyses to be done in an efficient and accurate manner. 14 refs., 2 figs., 3 tabs

  6. Standard model for the safety analysis report of nuclear fuel reprocessing plants

    International Nuclear Information System (INIS)

    1980-02-01

    This norm establishes the Standard Model for the Safety Analysis Report of Nuclear Fuel Reprocessing Plants, comprehending the presentation format, the detailing level of the minimum information required by the CNEN for evaluation the requests of Construction License or Operation Authorization, in accordance with the legislation in force. This regulation applies to the following basic reports: Preliminary Safety Analysis Report - PSAR, integrating part of the requirement of Construction License; and Final Safety Analysis Report (FSAR) which is the integrating part of the requirement for Operation Authorization

  7. Sources of Safety Data and Statistical Strategies for Design and Analysis: Clinical Trials.

    Science.gov (United States)

    Zink, Richard C; Marchenko, Olga; Sanchez-Kam, Matilde; Ma, Haijun; Jiang, Qi

    2018-03-01

    There has been an increased emphasis on the proactive and comprehensive evaluation of safety endpoints to ensure patient well-being throughout the medical product life cycle. In fact, depending on the severity of the underlying disease, it is important to plan for a comprehensive safety evaluation at the start of any development program. Statisticians should be intimately involved in this process and contribute their expertise to study design, safety data collection, analysis, reporting (including data visualization), and interpretation. In this manuscript, we review the challenges associated with the analysis of safety endpoints and describe the safety data that are available to influence the design and analysis of premarket clinical trials. We share our recommendations for the statistical and graphical methodologies necessary to appropriately analyze, report, and interpret safety outcomes, and we discuss the advantages and disadvantages of safety data obtained from clinical trials compared to other sources. Clinical trials are an important source of safety data that contribute to the totality of safety information available to generate evidence for regulators, sponsors, payers, physicians, and patients. This work is a result of the efforts of the American Statistical Association Biopharmaceutical Section Safety Working Group.

  8. A Technique of Software Safety Analysis in the Design Phase for PLC Based Safety-Critical Systems

    International Nuclear Information System (INIS)

    Koo, Seo-Ryong; Kim, Chang-Hwoi

    2017-01-01

    The purpose of safety analysis, which is a method of identifying portions of a system that have the potential for unacceptable hazards, is firstly to encourage design changes that will reduce or eliminate hazards and, secondly, to conduct special analyses and tests that can provide increased confidence in especially vulnerable portions of the system. For the design and implementation phase of the PLC based systems, we proposed a technique for software design specification and analysis, and this technique enables us to generate software design specifications (SDSs) in nuclear fields. For the safety analysis in the design phase, we used architecture design blocks of NuFDS to represent the architecture of the software. On the basis of the architecture design specification, we can directly generate the fault tree and then use the fault tree for qualitative analysis. Therefore, we proposed a technique of fault tree synthesis, along with a universal fault tree template for the architecture modules of nuclear software. Through our proposed fault tree synthesis in this work, users can use the architecture specification of the NuFDS approach to intuitively compose fault trees that help analyze the safety design features of software.

  9. Final safety analysis report (FSAR) for waste receiving and processing (WRAP) facility

    International Nuclear Information System (INIS)

    Weidert, J.R.

    1997-01-01

    This safety analysis report provides a summary description of the WRAP Facility, focusing on significant safety-related characteristics of the location and facility design. This report demonstrates that adherence to the safety basis wi11 ensure necessary operational safety considerations have been addressed sufficiently and justifies the adequacy of the safety basis in protecting the health and safety of the public, workers, and the environment

  10. Safety Analysis for Key Design Features of KALIMER-600 Design Concept

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Yong Bum; Kwon, Y. M.; Kim, E. K.; Suk, S. D.; Chang, W. P.; Jeong, H. Y.; Ha, K. S

    2007-02-15

    This report contains the safety analyses of the KALIMER-600 conceptual design which KAERI has been developing under the Long-term Nuclear R and D Program. The analyses have been performed reflecting the design developments during the second year of the 4th design phase in the program. The specific presentations are the key design features with the safety principles for achieving the safety objectives, the event categorization and safety criteria, and results on the safety analyses for the DBAs and ATWS events, the containment performance, and the channel blockages. The safety analyses for both the DBAs and ATWS events have been performed using SSC-K version 1.3., and the results have shown the fulfillment of the safety criteria for DBAs with conservative assumptions. The safety margins as well as the inherent safety also have been confirmed for the ATWS events. For the containment performance analysis, ORIGEN-2.1 and CONTAIN-LMR have been used. In results, the structural integrity has been acceptable and the evaluated exposure dose rate has been complied with 10 CFR 100 and PAG limits. The analysis results for flow blockages of 6-subchannels, 24-subchannels, and 54- subchannels with the MATRA-LMR-FB code, have assured the integrity of subassemblies.

  11. Organizational Culture and Safety Performance in the Manufacturing Companies in Malaysia: A Conceptual Analysis

    OpenAIRE

    Ong Choon Hee; Lim Lee Ping

    2014-01-01

    The purpose of this paper is to provide a conceptual analysis of organizational culture and safety performance in the manufacturing companies in Malaysia. Our conceptual analysis suggests that manufacturing companies that adopt group culture or hierarchical culture are more likely to demonstrate safety compliance and safety participation. Manufacturing companies that adopt rational culture or developmental culture are less likely to demonstrate safety compliance and safety participation. Give...

  12. Development of safety evaluation methods and analysis codes applied to the safety regulations for the design and construction stage of fast breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-08-15

    The purposes of this study are to develop the safety evaluation methods and analysis codes needed in the design and construction stage of fast breeder reactor (FBR). In JFY 2012, the following results are obtained. As for the development of safety evaluation methods needed in the safety examination conducted for the reactor establishment permission, development of the analysis codes, such as core damage analysis code, were carried out following the planned schedule. As for the development of the safety evaluation method needed for the risk informed safety regulation, the quantification technique of the event tree using the Continuous Markov chain Monte Carlo method (CMMC method) were studied. (author)

  13. Safety analysis of accident localization system

    International Nuclear Information System (INIS)

    1999-01-01

    A complex safety analysis of accident localization system of Ignalina NPP was performed. Calculation results obtained, results of non-destruct ing testing and experimental data of reinforced concrete testing of buildings does not revealed deficiencies of buildings of accident localization system at unit 1 of Ignalina NPP. Calculations were performed using codes NEPTUNE, ALGOR, CONTAIN

  14. Comparative analysis of safety related site characteristics

    International Nuclear Information System (INIS)

    Andersson, Johan

    2010-12-01

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

  15. Comparative analysis of safety related site characteristics

    Energy Technology Data Exchange (ETDEWEB)

    Andersson, Johan (ed.)

    2010-12-15

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

  16. Safety analysis report upgrade program at the Plutonium Facility, Los Alamos National Laboratory

    International Nuclear Information System (INIS)

    Pan, P.Y.

    1993-01-01

    Plutonium research and development activities have resided at the Los Alamos National Laboratory (LANL) since 1943. The function of the Plutonium Facility (PF-4) has been to perform basic special nuclear materials research and development and to support national defense and energy programs. The original Final Safety Analysis Report (FSAR) for PF-4 was approved by DOE in 1978. This FSAR analyzed design-basis and bounding accidents. In 1986, DOE/AL published DOE/AL Order 5481.1B, ''Safety Analysis and Review System'', as a requirement for preparation and review of safety analyses. To meet the new DOE requirements, the Facilities Management Group of the Nuclear Material Technology Division submitted a draft FSAR to DOE for approval in April 1991. This draft FSAR analyzed the new configurations and used a limited-scope probabilistic risk analysis for accident analysis. During the DOE review of the draft FSAR, DOE Order 5480.23 ''Nuclear Safety Analysis Reports'', was promulgated and was later officially released in April 1992. The new order significantly expands the scope, preparation, and maintenance efforts beyond those required in DOE/AL Order 5481.1B by requiring: description of institutional and human-factor safety programs; clear definitions of all facility-specific safety commitments; more comprehensive and detailed hazard assessment; use of new safety analysis methods; and annual updates of FSARs. This paper describes the safety analysis report (SAR) upgrade program at the Plutonium Facility in LANL. The SAR upgrade program is established to meet the requirements in DOE Order 5480.23. Described in this paper are the SAR background, authorization basis for operations, hazard classification, and technical program elements

  17. 242-A evaporator safety analysis report

    International Nuclear Information System (INIS)

    CAMPBELL, T.A.

    1999-01-01

    This report provides a revised safety analysis for the upgraded 242-A Evaporator (the Evaporator). This safety analysis report (SAR) supports the operation of the Evaporator following life extension upgrades and other facility and operations upgrades (e.g., Project B-534) that were undertaken to enhance the capabilities of the Evaporator. The Evaporator has been classified as a moderate-hazard facility (Johnson 1990). The information contained in this SAR is based on information provided by 242-A Evaporator Operations, Westinghouse Hanford Company, site maintenance and operations contractor from June 1987 to October 1996, and the existing operating contractor, Waste Management Hanford (WMH) policies. Where appropriate, a discussion address the US Department of Energy (DOE) Orders applicable to a topic is provided. Operation of the facility will be compared to the operating contractor procedures using appropriate audits and appraisals. The following subsections provide introductory and background information, including a general description of the Evaporator facility and process, a description of the scope of this SAR revision,a nd a description of the basic changes made to the original SAR

  18. 242-A evaporator safety analysis report

    Energy Technology Data Exchange (ETDEWEB)

    CAMPBELL, T.A.

    1999-05-17

    This report provides a revised safety analysis for the upgraded 242-A Evaporator (the Evaporator). This safety analysis report (SAR) supports the operation of the Evaporator following life extension upgrades and other facility and operations upgrades (e.g., Project B-534) that were undertaken to enhance the capabilities of the Evaporator. The Evaporator has been classified as a moderate-hazard facility (Johnson 1990). The information contained in this SAR is based on information provided by 242-A Evaporator Operations, Westinghouse Hanford Company, site maintenance and operations contractor from June 1987 to October 1996, and the existing operating contractor, Waste Management Hanford (WMH) policies. Where appropriate, a discussion address the US Department of Energy (DOE) Orders applicable to a topic is provided. Operation of the facility will be compared to the operating contractor procedures using appropriate audits and appraisals. The following subsections provide introductory and background information, including a general description of the Evaporator facility and process, a description of the scope of this SAR revision,a nd a description of the basic changes made to the original SAR.

  19. Integrated program of using of Probabilistic Safety Analysis in Spain

    International Nuclear Information System (INIS)

    1998-01-01

    Since 25 June 1986, when the CSN (Nuclear Safety Conseil) approve the Integrated Program of Probabilistic Safety Analysis, this program has articulated the main activities of CSN. This document summarize the activities developed during these years and reviews the Integrated programme

  20. 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'

  1. Effect of the Operation of Kerr and Hungry Horse Dams on the Reproductive Success of Kokanee in the Flathead System; Technical Addendum to the Final Report.

    Energy Technology Data Exchange (ETDEWEB)

    Beattie, Will; Tohtz, Joel

    1990-03-01

    This addendum to the Final Report presents results of research on the zooplankton and fish communities of Flathead Lade. The intent of the Study has been to identify the impacts of hydroelectric operations at Kerr and Hungry Horse Dam on the reproductive success of kokanee an to propose mitigation for these impacts. Recent changes in the trophic ecology of the lake, have reduced the survival of kokanee. In the last three year the Study has been redirected to identify, if possible, the biological mechanisms which now limit kokanee survival, and to test methods of enhancing the kokanee fishery by artificial supplementation. These studies were necessary to the formulation of mitigation plans. The possibility of successfully rehabilitating the kokanee population, is the doubt because of change in the trophic ecology of the system. This report first presents the results of studies of the population dynamics of crustacean zooplankton, upon which planktivorous fish depend. A modest effort was directed to measuring the spawning escapement of kokanee in 1988. Because of its relevance to the study, we also report assessments of 1989 kokanee spawning escapement. Hydroacoustic assessment of the abundance of all fish species in Flathead Lake was conducted in November, 1988. Summary of the continued efforts to document the growth rates and food habits of kokanee and lake whitefish are included in this report. Revised kokanee spawning and harvest estimates, and management implications of the altered ecology of Flathead Lake comprise the final sections of this addendum. 83 refs., 20 figs., 25 tabs.

  2. Safety analysis code SCTRAN development for SCWR and its application to CGNPC SCWR

    International Nuclear Information System (INIS)

    Wu, Pan; Gou, Junli; Shan, Jianqiang; Jiang, Yang; Yang, Jue; Zhang, Bo

    2013-01-01

    Highlights: ► A new safety analysis code named SCTRAN is developed for SCWRs. ► Capability of SCTRAN is verified by comparing with code APROS and RELAP5-3D. ► A new passive safety system is proposed for CGNPC SCWR and analyzed with SCTRAN. ► CGNPC SCWR is able to cope with two critical accidents for SCWRs, LOFA and LOCA. - Abstract: Design analysis is one of the main difficulties during the research and design of SCWRs. Currently, the development of safety analysis code for SCWR is still in its infancy all around the world, and very few computer codes could carry out the trans-critical calculations where significant changes in water properties would take place. In this paper, a safety analysis code SCTRAN for SCWRs has been developed based on code RETRAN-02, the best estimate code used for safety analysis of light water reactors. The ability of SCTRAN code to simulate transients where both supercritical and subcritical regimes are encountered has been verified by comparing with APROS and RELAP5-3D codes. Furthermore, the LOFA and LOCA transients for the CGNPC SCWR design were analyzed with SCTRAN code. The characteristics and performance of the passive safety systems applied to CGNPC SCWR were evaluated. The results show that: (1) The SCTRAN computer code developed in this study is capable to perform design analysis for SCWRs; (2) During LOFA and LOCA accidents in a CGNPC SCWR, the passive safety systems would significantly mitigate the consequences of these transients and enhance the inherent safety

  3. Annual activity report of Ignalina NPP Safety Analysis Group for 1996 year

    International Nuclear Information System (INIS)

    Ushpuras, E.; Augutis, J.; Bubelis, E.

    1997-03-01

    The main results of Ignalina NPP Safety Analysis Group (ISAG) investigations for 1996 are presented. ISAG is concentrating its research activities into four areas: the neutrons dynamics modelling, simulation of transient processes during loss of coolant accident, the reactor cooling systems modelling and the probabilistic safety assessment of accident confinement system. Ignalina Safety Analysis Report was prepared on the basis of these results. 37 refs., 9 tabs., 96 figs

  4. Development of safety analysis technology for integral reactor

    International Nuclear Information System (INIS)

    Kim, Hee Cheol; Kim, K. K.; Kim, S. H.

    2002-04-01

    The state-of-the-arts for the integral reactor was performed to investigate the safety features. The safety and performance of SMART were assessed using the technologies developed during the study. For this purpose, the computer code system and the analysis methodology were developed and the safety and performance analyses on SMART basic design were carried out for the design basis event and accident. The experimental facilities were designed for the core flow distribution test and the self-pressurizing pressurizer performance test. The tests on the 2-phase critical flow with non-condensable gas were completed and the results were used to assess the critical flow model. Probabilistic Safety Assessment(PSA) was carried out to evaluate the safety level and to optimize the design by identifying and remedying any weakness in the design. A joint study with KINS was carried out to promote licensing environment. The generic safety issues of integral reactors were identified and the solutions were formulated. The economic evaluation of the SMART desalination plant and the activities related to the process control were carried out in the scope of the study

  5. Development of vendor independent safety analysis capability for nuclear power plants in Taiwan

    International Nuclear Information System (INIS)

    Tang, J.-R.

    2001-01-01

    The Institute of Nuclear Energy Research (INER) and the Taiwan Power Company (TPC) have long-term cooperation to develop vendor independent safety analysis capability to provide support to nuclear power plants in Taiwan in many aspects. This paper presents some applications of this analysis capability, introduces the analysis methodology, and discusses the significance of vendor independent analysis capability now and future. The applications include a safety analysis of core shroud crack for Chinshan BWR/4 Unit 2, a parallel reload safety analysis of the first 18-month extended fuel cycle for Kuosheng BWR/6 Unit 2 Cycle 13, an analysis to support Technical Specification change for Maanshan three-loop PWR, and a design analysis to support the review of Preliminary Safety Analysis Report of Lungmen ABWR. In addition, some recent applications such as an analysis to support the review of BWR fuel bid for Chinshan and Kuosheng demonstrates the needs of further development of the analysis capability to support nuclear power plants in the 21 st century. (authors)

  6. WHO position on the use of fractional doses - June 2017, addendum to vaccines and vaccination against yellow fever WHO: Position paper - June 2013.

    Science.gov (United States)

    World Health Organization

    2017-10-13

    This article presents the World Health Organization's (WHO) recommendations on the use of fractional doses of yellow fever vaccines excerpted from the "Yellow fever vaccine: WHO position on the use of fractional doses - June 2017, Addendum to Vaccines and vaccination against yellow fever WHO: Position Paper - June 2013″, published in the Weekly Epidemiological Record [1,2]. This addendum to the 2013 position paper pertains specifically to use of fractional dose YF (fYF) vaccination (fractional dose yellow fever vaccination refers to administration of a reduced volume of vaccine dose, which has been reconstituted as usual per manufacturer recommendations) in the context of YF vaccine supply shortages beyond the capacity of the global stockpile. The current WHO position on the use of yellow fever (YF) vaccine is set out in the 2013 WHO position paper on vaccines and vaccination against YF and those recommendations are unchanged. Footnotes to this paper provide a number of core references including references to grading tables that assess the quality of the scientific evidence, and to the evidence-to-recommendation table. In accordance with its mandate to provide guidance to Member States on health policy matters, WHO issues a series of regularly updated position papers on vaccines and combinations of vaccines against diseases that have an international public health impact. These papers are concerned primarily with the use of vaccines in large-scale immunization programmes; they summarize essential background information on diseases and vaccines, and conclude with WHO's current position on the use of vaccines in the global context. Recommendations on the use of Yellow Fever vaccines were discussed by SAGE in October 2016; evidence presented at these meetings can be accessed at: www.who.int/immunization/sage/meetings/2016/October/presentations_background_docs/en/. Copyright © 2017. Published by Elsevier Ltd.

  7. Safety analysis of passing maneuvers using extreme value theory

    Directory of Open Access Journals (Sweden)

    Haneen Farah

    2017-04-01

    The results indicate that this is a promising approach for safety evaluation. On-going work of the authors will attempt to generalize this method to other safety measures related to passing maneuvers, test it for the detailed analysis of the effect of demographic factors on passing maneuvers' crash probability and for its usefulness in a traffic simulation environment.

  8. ACRR fuel storage racks criticality safety analysis

    International Nuclear Information System (INIS)

    Bodette, D.E.; Naegeli, R.E.

    1997-10-01

    This document presents the criticality safety analysis for a new fuel storage rack to support modification of the Annular Core Research Reactor for production of molybdenum-99 at Sandia National Laboratories, Technical Area V facilities. Criticality calculations with the MCNP code investigated various contingencies for the criticality control parameters. Important contingencies included mix of fuel element types stored, water density due to air bubbles or water level for the over-moderated racks, interaction with existing fuel storage racks and fuel storage holsters in the fuel storage pool, neutron absorption of planned rack design and materials, and criticality changes due to manufacturing tolerances or damage. Some limitations or restrictions on use of the new fuel storage rack for storage operations were developed through the criticality analysis and are required to meet the double contingency requirements of criticality safety. As shown in the analysis, this system will remain subcritical under all credible upset conditions. Administrative controls are necessary for loading, moving, and handling the storage rack as well as for control of operations around it. 21 refs., 16 figs., 4 tabs

  9. CCF analysis of high redundancy systems safety/relief valve data analysis and reference BWR application

    International Nuclear Information System (INIS)

    Mankamo, T.; Bjoere, S.; Olsson, Lena

    1992-12-01

    Dependent failure analysis and modeling were developed for high redundancy systems. The study included a comprehensive data analysis of safety and relief valves at the Finnish and Swedish BWR plants, resulting in improved understanding of Common Cause Failure mechanisms in these components. The reference application on the Forsmark 1/2 reactor relief system, constituting of twelve safety/relief lines and two regulating relief lines, covered different safety criteria cases of reactor depressurization and overpressure protection function, and failure to re close sequences. For the quantification of dependencies, the Alpha Factor Model, the Binomial Probability Model and the Common Load Model were compared for applicability in high redundancy systems

  10. System analysis of vehicle active safety problem

    Science.gov (United States)

    Buznikov, S. E.

    2018-02-01

    The problem of the road transport safety affects the vital interests of the most of the population and is characterized by a global level of significance. The system analysis of problem of creation of competitive active vehicle safety systems is presented as an interrelated complex of tasks of multi-criterion optimization and dynamic stabilization of the state variables of a controlled object. Solving them requires generation of all possible variants of technical solutions within the software and hardware domains and synthesis of the control, which is close to optimum. For implementing the task of the system analysis the Zwicky “morphological box” method is used. Creation of comprehensive active safety systems involves solution of the problem of preventing typical collisions. For solving it, a structured set of collisions is introduced with its elements being generated also using the Zwicky “morphological box” method. The obstacle speed, the longitudinal acceleration of the controlled object and the unpredictable changes in its movement direction due to certain faults, the road surface condition and the control errors are taken as structure variables that characterize the conditions of collisions. The conditions for preventing typical collisions are presented as inequalities for physical variables that define the state vector of the object and its dynamic limits.

  11. Final Safety Analysis Report (FSAR) for Building 332, Increment III

    Energy Technology Data Exchange (ETDEWEB)

    Odell, B. N.; Toy, Jr., A. J.

    1977-08-31

    This Final Safety Analysis Report (FSAR) supplements the Preliminary Safety Analysis Report (PSAR), dated January 18, 1974, for Building 332, Increment III of the Plutonium Materials Engineering Facility located at the Lawrence Livermore Laboratory (LLL). The FSAR, in conjunction with the PSAR, shows that the completed increment provides facilities for safely conducting the operations as described. These documents satisfy the requirements of ERDA Manual Appendix 6101, Annex C, dated April 8, 1971. The format and content of this FSAR complies with the basic requirements of the letter of request from ERDA San to LLL, dated March 10, 1972. Included as appendices in support of th FSAR are the Building 332 Operational Safety Procedure and the LLL Disaster Control Plan.

  12. Preliminary safety analysis report for the Waste Characterization Facility

    International Nuclear Information System (INIS)

    1994-10-01

    This safety analysis report outlines the safety concerns associated with the Waste Characterization Facility located in the Radioactive Waste Management Complex at the Idaho National Engineering Laboratory. The three main objectives of the report are to: define and document a safety basis for the Waste Characterization Facility activities; demonstrate how the activities will be carried out to adequately protect the workers, public, and environment; and provide a basis for review and acceptance of the identified risk that the managers, operators, and owners will assume. 142 refs., 38 figs., 39 tabs

  13. TVO-92 safety analysis of spent fuel disposal

    International Nuclear Information System (INIS)

    Vieno, T.; Hautojaervi, A.; Koskinen, L.; Nordman, H.

    1993-08-01

    The spent fuel from the TVO I and TVO II reactors at the Olkiluoto nuclear power plant is planned to be disposed in a repository constructed at a depth of about 500 meters in crystalline bedrock. Teollisuuden Voima Oy (TVO) has carried out preliminary site investigations for spent fuel disposal between 1987 and 1992 at five areas in Finland (Olkiluoto, Kivetty, Romuvaara, Syyry and Veitsivaara). The Safety analysis of the disposal system is presented in the report. Spent fuel will be encapsulated in composite copper-steel canisters. The canister design (ACP canister) consists of an inner container of steel as a load-bearing element and an outer container of oxygen-free copper to provide a shield against corrosion. In the repository the canisters will be emplaced in vertical holes drilled in the floors of horizontal deposition tunnels. The annulus between the canister and the rock is filled with compacted bentonite. The results of the safety analysis attest that the planned disposal system fulfils the safety requirements. Suitable places for the repository can be found at each of the five investigation sites

  14. The Oak Ridge Research Reactor: safety analysis: Volume 2, supplement 2

    International Nuclear Information System (INIS)

    Hurt, S.S.

    1986-11-01

    The Oak Ridge Research Reactor Safety Analysis was last updated via ORNL-4169, Vol. 2, Supplement 1, in May of 1978. Since that date, several changes have been effected through the change-memo system described below. While these changes have involved the cooling system, the electrical system, and the reactor instrumentation and controls, they have not, for the most part, presented new or unreviewed safety questions. However, some of the changes have been based on questions or recommendations stemming from safety reviews or from reactor events at other sites. This paper discusses those changes which were judged to be safety related and which include revisions to the syphon-break system and changes related to seismic considerations which were very recently completed. The maximum hypothetical accident postulated in the original safety analysis requires dynamic containment and filtered flow for compliance with 10CFR100 limits at the site boundary

  15. A survey on reliability and safety analysis techniques of robot systems in nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Eom, H S; Kim, J H; Lee, J C; Choi, Y R; Moon, S S

    2000-12-01

    The reliability and safety analysis techniques was surveyed for the purpose of overall quality improvement of reactor inspection system which is under development in our current project. The contents of this report are : 1. Reliability and safety analysis techniques suvey - Reviewed reliability and safety analysis techniques are generally accepted techniques in many industries including nuclear industry. And we selected a few techniques which are suitable for our robot system. They are falut tree analysis, failure mode and effect analysis, reliability block diagram, markov model, combinational method, and simulation method. 2. Survey on the characteristics of robot systems which are distinguished from other systems and which are important to the analysis. 3. Survey on the nuclear environmental factors which affect the reliability and safety analysis of robot system 4. Collection of the case studies of robot reliability and safety analysis which are performed in foreign countries. The analysis results of this survey will be applied to the improvement of reliability and safety of our robot system and also will be used for the formal qualification and certification of our reactor inspection system.

  16. A survey on reliability and safety analysis techniques of robot systems in nuclear power plants

    International Nuclear Information System (INIS)

    Eom, H.S.; Kim, J.H.; Lee, J.C.; Choi, Y.R.; Moon, S.S.

    2000-12-01

    The reliability and safety analysis techniques was surveyed for the purpose of overall quality improvement of reactor inspection system which is under development in our current project. The contents of this report are : 1. Reliability and safety analysis techniques suvey - Reviewed reliability and safety analysis techniques are generally accepted techniques in many industries including nuclear industry. And we selected a few techniques which are suitable for our robot system. They are falut tree analysis, failure mode and effect analysis, reliability block diagram, markov model, combinational method, and simulation method. 2. Survey on the characteristics of robot systems which are distinguished from other systems and which are important to the analysis. 3. Survey on the nuclear environmental factors which affect the reliability and safety analysis of robot system 4. Collection of the case studies of robot reliability and safety analysis which are performed in foreign countries. The analysis results of this survey will be applied to the improvement of reliability and safety of our robot system and also will be used for the formal qualification and certification of our reactor inspection system

  17. Enhancing Safety of Artificially Ventilated Patients Using Ambient Process Analysis.

    Science.gov (United States)

    Lins, Christian; Gerka, Alexander; Lüpkes, Christian; Röhrig, Rainer; Hein, Andreas

    2018-01-01

    In this paper, we present an approach for enhancing the safety of artificially ventilated patients using ambient process analysis. We propose to use an analysis system consisting of low-cost ambient sensors such as power sensor, RGB-D sensor, passage detector, and matrix infrared temperature sensor to reduce risks for artificially ventilated patients in both home and clinical environments. We describe the system concept and our implementation and show how the system can contribute to patient safety.

  18. Safety analysis of Ignalina NPP during shutdown conditions

    International Nuclear Information System (INIS)

    Kaliatka, A.; Uspuras, E.

    2000-01-01

    The accident analysis for the Ignalina NPP with RBMK-1500 reactors at normal operating conditions and at minimum controlled power level (during startup of the reactor) has been performed in the frame of the project I n-Depth Safety Assessment of the Ignalina NPP , which was completed in 1996. However, the plant conditions during the reactor shutdown differ from conditions during reactor operation at full power (equipment status in protection systems, set points for actuation of safety and protection systems, etc.). Results of RELAP5 simulation of two worst initiating events during reactor shutdown - Pressure Header rupture in case of steam reactor cooldown as well as Pressure Header rupture in case of water reactor cooldown are discussed in the paper. Results of analysis shown that reactor are reliably cooled in both cases. Further analysis for all range of initial events during reactor shutdown and at shutdown conditions is recommended. (author)

  19. Safety analysis report for packaging (onsite) Castor GSF cask

    International Nuclear Information System (INIS)

    Clements, E.P.

    1997-01-01

    The CASTOR GSF packaging was designed and fabricated to be a certified Type B(U) packaging and comply with the requirements of the International Atomic Energy Agency (IAEA) for transport of up to five sealed canisters of vitrified radioactive materials. This onsite Safety Analysis Report for Packaging (SARP) provides the analysis and evaluations necessary to demonstrate that the casks, with the canister payload, meet the intent of the Type B packaging regulations set forth in 10 CFR 71 and therefore meet the onsite transportation safety requirements of WHC-CM-2-14, Hazardous Material Packaging and Shipping

  20. Atlantic Richfield Hanford Company californium multiplier/delayed neutron counter safety analysis

    International Nuclear Information System (INIS)

    Zimmer, W.H.

    1976-08-01

    The Californium Multiplier (CFX) is a subcritical assembly of uranium surrounding 252 Cf spontaneously fissioning neutron sources; its function is to multiply the neutron flux to a level useful for activation analysis. This document summarizes the safety analysis aspects of the CFX, DNC, pneumatic transfer system, and instrumentation and to detail all the aspects of the total facility as a starting point for the ARHCO Safety Analysis Review. Recognized hazards and steps already taken to neutralize them are itemized

  1. Lessons learned - development of the tritium facilities 5480.23 safety analysis report and technical safety requirements

    International Nuclear Information System (INIS)

    Cappucci, A.J. Jr.; Bowman, M.E.; Goff, L.

    1997-01-01

    A review was performed which identified open-quotes Lessons Learnedclose quotes from the development of the 5480.23 Tritium Safety Analysis Report (SAR) and the Technical Safety Requirements (TSR) for the Tritium Facilities (TF). The open-quotes Lessons Learnedclose quotes were based on an evaluation of the use of the SRS procedures, processes, and work practices which contributed to the success or lack thereof. This review also identified recommendations and suggestions for improving the development of SARs and TSRs at SRS. The 5480.23 SAR describes the site for the TF, the various process systems in the process buildings, a complete hazards and accident analysis of the most significant hazards affecting the nearby offsite population, and the selection of safety systems, structures, and components to protect both the public and site workers. It also provides descriptions of important programs and processes which add defense in depth to public and worker protection

  2. An analysis of electronic health record-related patient safety concerns

    Science.gov (United States)

    Meeks, Derek W; Smith, Michael W; Taylor, Lesley; Sittig, Dean F; Scott, Jean M; Singh, Hardeep

    2014-01-01

    Objective A recent Institute of Medicine report called for attention to safety issues related to electronic health records (EHRs). We analyzed EHR-related safety concerns reported within a large, integrated healthcare system. Methods The Informatics Patient Safety Office of the Veterans Health Administration (VA) maintains a non-punitive, voluntary reporting system to collect and investigate EHR-related safety concerns (ie, adverse events, potential events, and near misses). We analyzed completed investigations using an eight-dimension sociotechnical conceptual model that accounted for both technical and non-technical dimensions of safety. Using the framework analysis approach to qualitative data, we identified emergent and recurring safety concerns common to multiple reports. Results We extracted 100 consecutive, unique, closed investigations between August 2009 and May 2013 from 344 reported incidents. Seventy-four involved unsafe technology and 25 involved unsafe use of technology. A majority (70%) involved two or more model dimensions. Most often, non-technical dimensions such as workflow, policies, and personnel interacted in a complex fashion with technical dimensions such as software/hardware, content, and user interface to produce safety concerns. Most (94%) safety concerns related to either unmet data-display needs in the EHR (ie, displayed information available to the end user failed to reduce uncertainty or led to increased potential for patient harm), software upgrades or modifications, data transmission between components of the EHR, or ‘hidden dependencies’ within the EHR. Discussion EHR-related safety concerns involving both unsafe technology and unsafe use of technology persist long after ‘go-live’ and despite the sophisticated EHR infrastructure represented in our data source. Currently, few healthcare institutions have reporting and analysis capabilities similar to the VA. Conclusions Because EHR-related safety concerns have complex

  3. Preliminary safety analysis report for the TFTR

    International Nuclear Information System (INIS)

    Lind, K.E.; Levine, J.D.; Howe, H.J.

    A Preliminary Safety Analysis Report has been prepared for the Tokamak Fusion Test Reactor. No accident scenarios have been identified which would result in exposures to on-site personnel or the general public in excess of the guidelines defined for the project by DOE

  4. Safety design concept and analysis for the upgrading JRR-3

    International Nuclear Information System (INIS)

    Onishi, N.; Isshiki, M.; Takahashi, H.; Takayanagi, M.

    1990-01-01

    The Research Reactor No.3 (JRR-3) is under reconstruction for upgrading. This paper describes the safety design concepts of the architectural and engineering design, anticipated operational transients and accident conditions which are the postulated initiating events for the safety evaluation, and the safety criteria of the upgraded JRR-3. The safety criteria are defined taking into account those of Light Water Reactors and the characteristics of the research reactor. Using the example of the safety analysis, this paper describes analytical results of a reactivity insertion by removal of in-core irradiation samples, a pipeline break at the primary coolant loop and flow blockage to a coolant channel, which are the severest postulated initiating events of the JRR-3

  5. Qualitative uncertainty analysis in probabilistic safety assessment context

    International Nuclear Information System (INIS)

    Apostol, M.; Constantin, M; Turcu, I.

    2007-01-01

    In Probabilistic Safety Assessment (PSA) context, an uncertainty analysis is performed either to estimate the uncertainty in the final results (the risk to public health and safety) or to estimate the uncertainty in some intermediate quantities (the core damage frequency, the radionuclide release frequency or fatality frequency). The identification and evaluation of uncertainty are important tasks because they afford credit to the results and help in the decision-making process. Uncertainty analysis can be performed qualitatively or quantitatively. This paper performs a preliminary qualitative uncertainty analysis, by identification of major uncertainty in PSA level 1- level 2 interface and in the other two major procedural steps of a level 2 PSA i.e. the analysis of accident progression and of the containment and analysis of source term for severe accidents. One should mention that a level 2 PSA for a Nuclear Power Plant (NPP) involves the evaluation and quantification of the mechanisms, amount and probabilities of subsequent radioactive material releases from the containment. According to NUREG 1150, an important task in source term analysis is fission products transport analysis. The uncertainties related to the isotopes distribution in CANDU NPP primary circuit and isotopes' masses transferred in the containment, using SOPHAEROS module from ASTEC computer code will be also presented. (authors)

  6. Compendium of computer codes for the safety analysis of LMFBR's

    International Nuclear Information System (INIS)

    1975-06-01

    A high level of mathematical sophistication is required in the safety analysis of LMFBR's to adequately meet the demands for realism and confidence in all areas of accident consequence evaluation. The numerical solution procedures associated with these analyses are generally so complex and time consuming as to necessitate their programming into computer codes. These computer codes have become extremely powerful tools for safety analysis, combining unique advantages in accuracy, speed and cost. The number, diversity and complexity of LMFBR safety codes in the U. S. has grown rapidly in recent years. It is estimated that over 100 such codes exist in various stages of development throughout the country. It is inevitable that such a large assortment of codes will require rigorous cataloguing and abstracting to aid individuals in identifying what is available. It is the purpose of this compendium to provide such a service through the compilation of code summaries which describe and clarify the status of domestic LMFBR safety codes. (U.S.)

  7. Statistical analysis applied to safety culture self-assessment

    International Nuclear Information System (INIS)

    Macedo Soares, P.P.

    2002-01-01

    Interviews and opinion surveys are instruments used to assess the safety culture in an organization as part of the Safety Culture Enhancement Programme. Specific statistical tools are used to analyse the survey results. This paper presents an example of an opinion survey with the corresponding application of the statistical analysis and the conclusions obtained. Survey validation, Frequency statistics, Kolmogorov-Smirnov non-parametric test, Student (T-test) and ANOVA means comparison tests and LSD post-hoc multiple comparison test, are discussed. (author)

  8. Review of design criteria and safety analysis of safety class electric building for fuel test loop

    Energy Technology Data Exchange (ETDEWEB)

    Kim, J. Y.

    1998-02-01

    Steady state fuel test loop will be equipped in HANARO to obtain the development and betterment of advanced fuel and materials through the irradiation tests. HANARO fuel test loop was designed for CANDU and PWR fuel testing. Safety related system of Fuel Test Loop such as emergency cooling water system, component cooling water system, safety ventilation system, high energy line break mitigation system and remote control room was required 1E class electric supply to meet the safety operation in accordance with related code. Therefore, FTL electric building was designed to construction and install the related equipment based on seismic category I. The objective of this study is to review the design criteria and analysis the safety function of safety class electric building for fuel test loop, and this results will become guidance for the irradiation testing in future. (author). 10 refs., 6 tabs., 30 figs.

  9. SYSTEMS SAFETY ANALYSIS FOR FIRE EVENTS ASSOCIATED WITH THE ECRB CROSS DRIFT

    International Nuclear Information System (INIS)

    R. J. Garrett

    2001-01-01

    The purpose of this analysis is to systematically identify and evaluate fire hazards related to the Yucca Mountain Site Characterization Project (YMP) Enhanced Characterization of the Repository Block (ECRB) East-West Cross Drift (commonly referred to as the ECRB Cross-Drift). This analysis builds upon prior Exploratory Studies Facility (ESF) System Safety Analyses and incorporates Topopah Springs (TS) Main Drift fire scenarios and ECRB Cross-Drift fire scenarios. Accident scenarios involving the fires in the Main Drift and the ECRB Cross-Drift were previously evaluated in ''Topopah Springs Main Drift System Safety Analysis'' (CRWMS M and O 1995) and the ''Yucca Mountain Site Characterization Project East-West Drift System Safety Analysis'' (CRWMS M and O 1998). In addition to listing required mitigation/control features, this analysis identifies the potential need for procedures and training as part of defense-in-depth mitigation/control features. The inclusion of this information in the System Safety Analysis (SSA) is intended to assist the organization(s) (e.g., Construction, Environmental Safety and Health, Design) responsible for these aspects of the ECRB Cross-Drift in developing mitigation/control features for fire events, including Emergency Refuge Station(s). This SSA was prepared, in part, in response to Condition/Issue Identification and Reporting/Resolution System (CIRS) item 1966. The SSA is an integral part of the systems engineering process, whereby safety is considered during planning, design, testing, and construction. A largely qualitative approach is used which incorporates operating experiences and recommendations from vendors, the constructor and the operating contractor. The risk assessment in this analysis characterizes the scenarios associated with fires in terms of relative risk and includes recommendations for mitigating all identified hazards. The priority for recommending and implementing mitigation control features is: (1) Incorporate

  10. A 'Toolbox' Equivalent Process for Safety Analysis Software

    International Nuclear Information System (INIS)

    O'Kula, K.R.; Eng, Tony

    2004-01-01

    Defense Nuclear Facilities Safety Board (DNFSB) Recommendation 2002-1 (Quality Assurance for Safety-Related Software) identified a number of quality assurance issues on the use of software in Department of Energy (DOE) facilities for analyzing hazards, and designing and operating controls that prevent or mitigate potential accidents. The development and maintenance of a collection, or 'toolbox', of multiple-site use, standard solution, Software Quality Assurance (SQA)-compliant safety software is one of the major improvements identified in the associated DOE Implementation Plan (IP). The DOE safety analysis toolbox will contain a set of appropriately quality-assured, configuration-controlled, safety analysis codes, recognized for DOE-broad, safety basis applications. Currently, six widely applied safety analysis computer codes have been designated for toolbox consideration. While the toolbox concept considerably reduces SQA burdens among DOE users of these codes, many users of unique, single-purpose, or single-site software may still have sufficient technical justification to continue use of their computer code of choice, but are thwarted by the multiple-site condition on toolbox candidate software. The process discussed here provides a roadmap for an equivalency argument, i.e., establishing satisfactory SQA credentials for single-site software that can be deemed ''toolbox-equivalent''. The process is based on the model established to meet IP Commitment 4.2.1.2: Establish SQA criteria for the safety analysis ''toolbox'' codes. Implementing criteria that establish the set of prescriptive SQA requirements are based on implementation plan/procedures from the Savannah River Site, also incorporating aspects of those from the Waste Isolation Pilot Plant (SNL component) and the Yucca Mountain Project. The major requirements are met with evidence of a software quality assurance plan, software requirements and design documentation, user's instructions, test report, a

  11. Hazard Identification and Risk Assessment of Health and Safety Approach JSA (Job Safety Analysis) in Plantation Company

    Science.gov (United States)

    Sugarindra, Muchamad; Ragil Suryoputro, Muhammad; Tiya Novitasari, Adi

    2017-06-01

    Plantation company needed to identify hazard and perform risk assessment as an Identification of Hazard and Risk Assessment Crime and Safety which was approached by using JSA (Job Safety Analysis). The identification was aimed to identify the potential hazards that might be the risk of workplace accidents so that preventive action could be taken to minimize the accidents. The data was collected by direct observation to the workers concerned and the results were recorded on a Job Safety Analysis form. The data were as forklift operator, macerator worker, worker’s creeper, shredder worker, workers’ workshop, mechanical line worker, trolley cleaning workers and workers’ crepe decline. The result showed that shredder worker value was 30 and had the working level with extreme risk with the risk value range was above 20. So to minimize the accidents could provide Personal Protective Equipment (PPE) which were appropriate, information about health and safety, the company should have watched the activities of workers, and rewards for the workers who obey the rules that applied in the plantation.

  12. YUCCA MOUNTAIN SITE CHARACTERIZATION PROJECT EAST-WEST DRIFT SYSTEM SAFETY ANALYSIS

    Energy Technology Data Exchange (ETDEWEB)

    NA

    1999-06-08

    The purpose of this analysis is to systematically identify and evaluate hazards related to the design of the Yucca Mountain Project Exploratory Studies Facility (ESF) East-West Cross Drift. This analysis builds upon prior ESF System Safety Analyses and incorporates TS Main Drift scenarios, where applicable, into the East-West Drift scenarios. This System Safety Analysis (SSA) focuses on the personnel safety and health hazards associated with the engineered design of the East-West Drift. The analysis also evaluates other aspects of the East-West Drift, including purchased equipment (e.g., scientific mapping platform) or Systems/Structures/Components (SSCs) and out-of-tolerance conditions. In addition to recommending design mitigation features, the analysis identifies the potential need for procedures, training, or Job Safety Analyses (JSAs). The inclusion of this information in the SSA is intended to assist the organization(s) (e.g., constructor, Safety and Health, design) responsible for these aspects of the East-West Drift in evaluating personnel hazards and augment the information developed by these organizations. The SSA is an integral part of the systems engineering process, whereby safety is considered during planning, design, testing, and construction. A largely qualitative approach is used which incorporates operating experiences and recommendations from vendors, the constructor and the operating contractor. The risk assessment in this analysis characterizes the scenarios associated with East-West Drift SSCs in terms of relative risk and includes recommendations for mitigating all identified hazards. The priority for recommending and implementing mitigation control features is: (1) Incorporate measures to reduce risks and hazards into SSC designs. (2) Add safety features and capabilities to existing designs. (3) Develop procedures and conduct training to increase worker awareness of potential hazards, reduce exposure to hazards, and inform personnel of the

  13. 10 CFR 52.79 - Contents of applications; technical information in final safety analysis report.

    Science.gov (United States)

    2010-01-01

    ...; technical information in final safety analysis report. (a) The application must contain a final safety... 10 Energy 2 2010-01-01 2010-01-01 false Contents of applications; technical information in final safety analysis report. 52.79 Section 52.79 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) LICENSES...

  14. Probabilistic safety analysis : a new nuclear power plants licensing method

    International Nuclear Information System (INIS)

    Oliveira, L.F.S. de.

    1982-04-01

    After a brief retrospect of the application of Probabilistic Safety Analysis in the nuclear field, the basic differences between the deterministic licensing method, currently in use, and the probabilistic method are explained. Next, the two main proposals (by the AIF and the ACRS) concerning the establishment of the so-called quantitative safety goals (or simply 'safety goals') are separately presented and afterwards compared in their most fundamental aspects. Finally, some recent applications and future possibilities are discussed. (Author) [pt

  15. Safety analysis of the existing 804 and 845 firing facilities

    International Nuclear Information System (INIS)

    Odell, B.N.

    1986-01-01

    A safety analysis was performed to determine if normal operations and/or potential accidents at the 804 and 845 Firing Facilities at Site 300 could present undue hazards to the general public, peronnel at Site 300, or have an adverse effect on the environment. The normal operation and credible accident that might have an effect on these facilities or have off-site consequence were considered. It was determined by this analysis that all but one of the hazards were either low or of the type or magnitude routinely encountered and/or accepted by the public. The exception was explosives. Since this hazard has the potential for causing significant on-site and minimum off-site consequences, Bunkers 804 and 845 have been classified as moderate hazard facilties per DOE Order 5481.1A. This safety analysis concluded that the operation at these facilities will present no undue risk to the health and safety of LLNL employees or the public

  16. Short course on system safety analysis

    International Nuclear Information System (INIS)

    Sudmann, R.H.

    1992-01-01

    This course provides and introduction to methods generally used in safety analysis and accident investigation. It is a non-mathematical approach, directed toward a casual user. The participant will learn techniques allowing them to dissect a system or incident in order identify real or potential safety problems. These techniques will be applied to analyze events which have occurred within DOE facilities. As a manager or staff person with general oversight responsibilities, the participant should gain an awareness of the big picture and not just ''dig for facts.'' This can be accomplished by being alert and responsive to the atmosphere and condition of the plant; mood and impression of the worker and the behavioral climate. The techniques taught in the course can be used to identify critical areas or indicators. These indicators will signal problems before the ''facts'' will. Analysis techniques taught are used to gauge the breadth of the ''forest'' and not necessarily to identify the trees. For this course includes a technical background with experience in a chemical processing operations and a knowledge of basic chemistry and engineering is desirable. The course should help in a present or future assignment in an oversight role

  17. Comparison of a Traditional Probabilistic Risk Assessment Approach with Advanced Safety Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Smith, Curtis L; Mandelli, Diego; Zhegang Ma

    2014-11-01

    As part of the Light Water Sustainability Program (LWRS) [1], the purpose of the Risk Informed Safety Margin Characterization (RISMC) [2] Pathway research and development (R&D) is to support plant decisions for risk-informed margin management with the aim to improve economics, reliability, and sustain safety of current NPPs. In this paper, we describe the RISMC analysis process illustrating how mechanistic and probabilistic approaches are combined in order to estimate a safety margin. We use the scenario of a “station blackout” (SBO) wherein offsite power and onsite power is lost, thereby causing a challenge to plant safety systems. We describe the RISMC approach, illustrate the station blackout modeling, and contrast this with traditional risk analysis modeling for this type of accident scenario. We also describe our approach we are using to represent advanced flooding analysis.

  18. A proposal for performing software safety hazard analysis

    International Nuclear Information System (INIS)

    Lawrence, J.D.; Gallagher, J.M.

    1997-01-01

    Techniques for analyzing the safety and reliability of analog-based electronic protection systems that serve to mitigate hazards in process control systems have been developed over many years, and are reasonably understood. An example is the protection system in a nuclear power plant. The extension of these techniques to systems which include digital computers is not well developed, and there is little consensus among software engineering experts and safety experts on how to analyze such systems. One possible technique is to extend hazard analysis to include digital computer-based systems. Software is frequently overlooked during system hazard analyses, but this is unacceptable when the software is in control of a potentially hazardous operation. In such cases, hazard analysis should be extended to fully cover the software. A method for performing software hazard analysis is proposed in this paper. The method concentrates on finding hazards during the early stages of the software life cycle, using an extension of HAZOP

  19. Development of design and safety analysis supporting system for casks

    International Nuclear Information System (INIS)

    Ohsono, Katsunari; Higashino, Akira; Endoh, Shuji

    1993-01-01

    Mitsubishi heavy Industries has developed a design and safety analysis supporting system 'CADDIE' (Cask Computer Aided Design, Drawing and Integrated Evaluation System), with the following objectives: (1) Enhancement of efficiency of the design and safety analysis (2) Further advancement of design quality (3) Response to the diversification of design requirements. The features of this system are as follows: (1) The analysis model data common to analyses is established, and it is prepared automatically from the model made by CAD. (2) The input data for the analysis code is available by simple operation of conversation type from the analysis model data. (3) The analysis results are drawn out in diagrams by output generator, so as to facilitate easy observation. (4) The data of material properties, fuel assembly data, etc. required for the analyses are made available as a data base. (J.P.N.)

  20. Safety analysis, risk assessment, and risk acceptance criteria

    International Nuclear Information System (INIS)

    Jamali, K.

    1997-01-01

    This paper discusses a number of topics that relate safety analysis as documented in the Department of Energy (DOE) safety analysis reports (SARs), probabilistic risk assessments (PRA) as characterized primarily in the context of the techniques that have assumed some level of formality in commercial nuclear power plant applications, and risk acceptance criteria as an outgrowth of PRA applications. DOE SARs of interest are those that are prepared for DOE facilities under DOE Order 5480.23 and the implementing guidance in DOE STD-3009-94. It must be noted that the primary area of application for DOE STD-3009 is existing DOE facilities and that certain modifications of the STD-3009 approach are necessary in SARs for new facilities. Moreover, it is the hazard analysis (HA) and accident analysis (AA) portions of these SARs that are relevant to the present discussions. Although PRAs can be qualitative in nature, PRA as used in this paper refers more generally to all quantitative risk assessments and their underlying methods. HA as used in this paper refers more generally to all qualitative risk assessments and their underlying methods that have been in use in hazardous facilities other than nuclear power plants. This discussion includes both quantitative and qualitative risk assessment methods. PRA has been used, improved, developed, and refined since the Reactor Safety Study (WASH-1400) was published in 1975 by the Nuclear Regulatory Commission (NRC). Much debate has ensued since WASH-1400 on exactly what the role of PRA should be in plant design, reactor licensing, 'ensuring' plant and process safety, and a large number of other decisions that must be made for potentially hazardous activities. Of particular interest in this area is whether the risks quantified using PRA should be compared with numerical risk acceptance criteria (RACs) to determine whether a facility is 'safe.' Use of RACs requires quantitative estimates of consequence frequency and magnitude

  1. Management system of health and safety work (SMK3) with job safety analysis (JSA) in PT. Nira Murni construction

    Science.gov (United States)

    Melliana, Armen, Yusrizal, Akmal, Syarifah

    2017-11-01

    PT Nira Murni construction is a contractor of PT Chevron Pacific Indonesia which engaged in contractor, fabrication, maintenance construction suppliers, and labor services. The high of accident rate in this company is caused the lack of awareness of workplace safety. Therefore, it requires an effort to reduce the accident rate on the company so that the financial losses can be minimized. In this study, Safe T-Score method is used to analyze the accident rate by measuring the level of frequency. Analysis is continued using risk management methods which identify hazards, risk measurement and risk management. The last analysis uses Job safety analysis (JSA) which will identify the effect of accidents. From the result of this study can be concluded that Job Safety Analysis (JSA) methods has not been implemented properly. Therefore, JSA method needs to follow-up in the next study, so that can be well applied as prevention of occupational accidents.

  2. Reliability Analysis and Calibration of Partial Safety Factors for Redundant Structures

    DEFF Research Database (Denmark)

    Sørensen, John Dalsgaard

    1998-01-01

    Redundancy is important to include in the design and analysis of structural systems. In most codes of practice redundancy is not directly taken into account. In the paper various definitions of a deterministic and reliability based redundancy measure are reviewed. It is described how reundancy can...... be included in the safety system and how partial safety factors can be calibrated. An example is presented illustrating how redundancy is taken into account in the safety system in e.g. the Danish codes. The example shows how partial safety factors can be calibrated to comply with the safety level...

  3. Documentation of Hanford Site independent review of the Hanford Waste Vitrification Plant Preliminary Safety Analysis Report

    International Nuclear Information System (INIS)

    Herborn, D.I.

    1993-11-01

    Westinghouse Hanford Company (WHC) is the Integrating Contractor for the Hanford Waste Vitrification Plant (HWVP) Project, and as such is responsible for preparation of the HWVP Preliminary Safety Analysis Report (PSAR). The HWVP PSAR was prepared pursuant to the requirements for safety analyses contained in US Department of Energy (DOE) Orders 4700.1, Project Management System (DOE 1987); 5480.5, Safety of Nuclear Facilities (DOE 1986a); 5481.lB, Safety Analysis and Review System (DOE 1986b) which was superseded by DOE order 5480-23, Nuclear Safety Analysis Reports, for nuclear facilities effective April 30, 1992 (DOE 1992); and 6430.lA, General Design Criteria (DOE 1989). The WHC procedures that, in large part, implement these DOE requirements are contained in WHC-CM-4-46, Nonreactor Facility Safety Analysis Manual. This manual describes the overall WHC safety analysis process in terms of requirements for safety analyses, responsibilities of the various contributing organizations, and required reviews and approvals

  4. Safety analysis of the VLJ repository

    International Nuclear Information System (INIS)

    Vieno, T.; Nordman, H.

    1991-05-01

    The VLJ repository is an underground disposal facility for the low and medium level waste generated at the Olkiluoto nuclear power plant. The repository is located within 1 km from TVO I and TVO II (2 x 710 MWe) BWR's on the Olkiluoto island at the west coast of Finland. It contains two rock silos excavated at the depth of 60...100 meters in the bedrock. Low level waste will be disposed of in a shotcreted rock silo. For bituminized medium level waste, a separate silo of reinforced concrete has been built inside the shotcreted rock silo. The post-closure safety analysis has been done for the Final Safety Analysis Report (FSAR) of the VLJ repository. In addition to the normal evolution scenario, several disturbed evolution and accident scenarios have been analysed. In the reference scenario, radio-nuclides are assumed to be released from the bituminized waste within 500 years, the concrete silo is assumed to gradually disintegrate and finally to collapse at 5 000 years, all concrete in the silo is assumed to be also chemically depleted within 6 000 years, and all the seals of the repository are assumed to deteriorate within 12 000 years. The ability of alone natural barriers to restrict the release of radionuclides into the biosphere has been evaluated by means of scenarios where the degradation of engineered barriers has been assumed to take place at a still faster rate. In one of the disturbed evolution scenarios it has been assumed that the concrete silo for medium level waste is severely impaired immediately after sealing of the repository. Effects of gas generation and consequences of human intrusion have been evaluated, too. The results of the safety analysis show that radiation doses of any significance are caused only if a well is bored in the vicinity of the repository or if the groundwater discharge spot is inhabited and used for cultivation. In the reference scenario the maximum expectation value of the individual dose rate is 0.3 mSv/a

  5. Development of safety analysis technology for LMR

    International Nuclear Information System (INIS)

    Lee, Y. B.; Kwon, Y. M.; Suk, S. D.

    2005-03-01

    The MATRA-LMR-FB has been developed internally for the damage prevention as well as the safety assessment during a channel blockage accident and, as a the result, the quality of the code becomes comparable to that developed in the leading countries. For a code-to-code comparison, KAERI could have access to the SASSYS-1 through a bilateral collaboration between KAERI and ANL. The study could bring into the reliability improvements both on the reactivity models in the SSC-K and on the SSC-K prediction capability. It finally leads to the completion of the SSC-K version 1.3 resulting from the qualitative and quantitative code-to-code comparison. The preliminary analysis for a metal fueled LMR could also become possible with the MELT-III and the VENUS-II, which had originally been developed for the HCDA analysis with an oxidized fuel, by developing the relevant models For the development of the safety evaluation technology, the safety limits have been set up, and the analyses of the internal and external channel blockages in an assembly have also been performed. Besides, the more reliable analysis results on the key design concepts could be obtained by way of the methodology improvement resulting from the qualitative and quantitative comparison study. For an efficient and systematic control of the main project, the integration of the developed technologies and the establishment of their data base have been pursued. It has gone through the development of the process control with taking account of interfaces among the sub-projects, the overall coordination of the developed technologies, the data base for the design products, and so on

  6. Probabilistic analysis of safety in industrial irradiation plants

    International Nuclear Information System (INIS)

    Alderete, F.; Elechosa, C.

    2006-01-01

    The Argentinean Nuclear Regulatory Authority is carrying out the Probabilistic Safety Analysis (PSA) of the two industrial irradiation plants existent in the country. The objective of this presentation is to show from the regulatory point of view, the advantages of applying this tool, as well as the appeared difficulties; for it will be made a brief description of the facilities, of the method and of the normative one. Both plants are multipurpose facilities classified as 'industrial irradiator category IV' (panoramic irradiator with source deposited in pool). Basically, the execution of an APS consists of the following stages: 1. Identification of initiating events. 2. Modeling of Accidental Sequences (Event Trees). 3. Analysis of Systems (Fault trees). 4. Quantification of Accidental Sequences. The argentine normative doesn't demand to these facilities the realization of an APS, however the basic standard of Radiological Safety establishes that in the design of this type of facilities in the cases that is justified, should make sure that the annual probability of occurrence of an accidental sequence and the resulting dose in a person gives as result an radiological risk inferior to the risk limit adopted as acceptance criteria. On the other hand the design standard specifies for these irradiators it demands a maximum fault rate of 10 -2 for the related components with the systems of radiological safety. In our case, the possible initiating events have been identified that carried out to not wanted situations (about people exposure, radioactive contamination). Then, for each one of the significant initiating events, the corresponding accidental sequences were modeled and the safety systems that intervene in this sequences by means of fault trees were analyzed, for then to determine the fault probabilities of the same ones. At the moment they are completing these fault trees, but the difficulty resides in the impossibility of obtaining real data of the reliability

  7. Safety analysis report for the Waste Storage Facility. Revision 2

    Energy Technology Data Exchange (ETDEWEB)

    Bengston, S.J.

    1994-05-01

    This safety analysis report outlines the safety concerns associated with the Waste Storage Facility located in the Radioactive Waste Management Complex at the Idaho National Engineering Laboratory. The three main objectives of the report are: define and document a safety basis for the Waste Storage Facility activities; demonstrate how the activities will be carried out to adequately protect the workers, public, and environment; and provide a basis for review and acceptance of the identified risk that the managers, operators, and owners will assume.

  8. Role of in-house safety analysis and research activities in regulatory decision making

    International Nuclear Information System (INIS)

    Pradhan, Santosh K.; Nagrale, Dhanesh B.; Gaikwad, Avinash J.

    2015-01-01

    Achievement of an acceptable level of nuclear safety is an essential requirement for the peaceful utilization of nuclear energy. The success of Global Nuclear Safety Regime is built upon a foundation of research. Such research has been sponsored by Governments and industry and has led to improved designs, safer and more reliable plant operation, and improvements in operating plant efficiency. A key element of this research has been the nuclear safety research performed or sponsored by regulatory organizations. In part, it has been the safety research performed or sponsored by regulatory organizations that has contributed to improved safety and has laid the foundation for activities such as risk-informed regulation, plant life extension, improved plant performance (e.g. power uprates) and new plant designs. The regulatory research program is meant to improve the regulatory authority’s knowledge where uncertainty exists, where safety margins are not well-characterized, and where regulatory decisions need to be confirmed in existing or new designs and technologies. The regulatory body get research initiated either in-house or by the licensee or through technical support organizations (TSOs). Research and analysis carried out within the regulatory body is of immense value in this context. This could be in the form of analysis of safety significant events, analysis of severe accidents, review of operating experience, independent checks of critical designs and even review of operator responses under different situations towards arriving at modifications to training programmes and licensing procedures for operating personnel. A latent benefit of regulatory research carried out by the regulators themselves is that it improves their technical competence considerably which in turn leads to high quality safety reviews and improved regulation in general. The aim of the present paper is to provide an overview of role of regulatory research and the in-house regulatory safety

  9. Influence of probabilistic safety analysis on design and operation of PWR plants

    International Nuclear Information System (INIS)

    Bastl, W.; Hoertner, H.; Kafka, P.

    1978-01-01

    This paper gives a comprehensive presentation of the connections and influences of probabilistic safety analysis on design and operation of PWR plants. In this context a short historical retrospective view concerning probabilistic reliability analysis is given. In the main part of this paper some examples are presented in detail, showing special outcomes of such probabilistic investigations. Additional paragraphs illustrate some activities and issues in the field of probabilistic safety analysis

  10. Analysis of Critical Characteristics for Safety Graded Personnel Computers in the KNICS Architecture

    International Nuclear Information System (INIS)

    Lee, Hyun Chul; Lee, Dong Young

    2009-01-01

    Critical characteristics analysis of a safety related item is to identify characteristics to be verified to replace an original item with the dedicated item. It is sure that the dedicated item meeting critical characteristics would perform its intended safety function instead of the specified item. KNICS project developed two safety systems: IDiPS RPS (Reactor Protection System) and IDiPS ESF-CCS (Engineered Safety Features-Component Control System). Two safety systems of IDiPS are equipped with personnel computers, so-called COMs (Cabinet Operator Modules), in their cabinets. The personnel computers, COMs, are responsible for safety system monitoring, testing, and maintaining. Even though two safety systems are safety critical system, the personnel computers of two systems, i.e. COMs, are not graded as safety-graded items. Regulation requirements are expected to be strengthened, and the functions of the personnel computer may be enhanced to include safety-related functions and safety functions, it would be necessary that the grade of the personnel computers is adjusted to a higher level, the safety grade. To try to upgrade a non safety system, i.e. COMs, to a safety system, its safety functions and requirements, i.e. critical characteristics, must be identified and verified. This paper describes the process of the identification of critical characteristics and the results of analysis

  11. Safety analysis and risk assessment of the National Ignition Facility

    International Nuclear Information System (INIS)

    Brereton, S.; McLouth, L.; Odell, B.

    1996-01-01

    The National Ignition Facility (NIF) is a proposed U.S. Department of Energy inertial confinement laser fusion facility. The candidate sites for locating the NIF are: Los Alamos National Laboratory, Sandia National Laboratory, the Nevada Test Site, and Lawrence Livermore National Laboratory (LLNL), the preferred site. The NIF will operate by focusing 192 laser beams onto a tiny deuterium-tritium target located at the center of a spherical target chamber. The NIF mission is to achieve inertial confinement fusion (ICF) ignition, access physical conditions in matter of interest to nuclear weapons physics, provide an above ground simulation capability for nuclear weapons effects testing, and contribute to the development of inertial fusion for electrical power production. The NIF has been classified as a radiological, low hazard facility on the basis of a preliminary hazards analysis and according to the DOE methodology for facility classification. This requires that a safety analysis be prepared under DOE Order 5481.1B, Safety Analysis and Review System. A draft Preliminary Safety Analysis Report (PSAR) has been written, and this will be finalized later in 1996. This paper summarizes the safety issues associated with the operation of the NIF and the methodology used to study them. It provides a summary of the methodology, an overview of the hazards, estimates maximum routine and accidental exposures for the preferred site of LLNL, and concludes that the risks from NIF operations are low

  12. Representation of human behaviour in probabilistic safety analysis

    International Nuclear Information System (INIS)

    Whittingham, R.B.

    1991-01-01

    This paper provides an overview of the representation of human behaviour in probabilistic safety assessment. Human performance problems which may result in errors leading to accidents are considered in terms of methods of identification using task analysis, screening analysis of critical errors, representation and quantification of human errors in fault trees and event trees and error reduction measures. (author) figs., tabs., 43 refs

  13. Analysis on Dangerous Source of Large Safety Accident in Storage Tank Area

    Science.gov (United States)

    Wang, Tong; Li, Ying; Xie, Tiansheng; Liu, Yu; Zhu, Xueyuan

    2018-01-01

    The difference between a large safety accident and a general accident is that the consequences of a large safety accident are particularly serious. To study the tank area which factors directly or indirectly lead to the occurrence of large-sized safety accidents. According to the three kinds of hazard source theory and the consequence cause analysis of the super safety accident, this paper analyzes the dangerous source of the super safety accident in the tank area from four aspects, such as energy source, large-sized safety accident reason, management missing, environmental impact Based on the analysis of three kinds of hazard sources and environmental analysis to derive the main risk factors and the AHP evaluation model is established, and after rigorous and scientific calculation, the weights of the related factors in four kinds of risk factors and each type of risk factors are obtained. The result of analytic hierarchy process shows that management reasons is the most important one, and then the environmental factors and the direct cause and Energy source. It should be noted that although the direct cause is relatively low overall importance, the direct cause of Failure of emergency measures and Failure of prevention and control facilities in greater weight.

  14. Criticality safety analysis of Hanford Waste Tank 241-101-SY

    International Nuclear Information System (INIS)

    Perry, R.T.; Sapir, J.L.; Krohn, B.J.

    1993-01-01

    As part of a safety assessment for proposed pump mixing operations to mitigate episodic gas releases in Tank 241-101-SY at the Hanford Site, Richland, Washington, a criticality safety analysis was made using the Sn transport code ONEDANT. The tank contains approximately one million gallons of waste and an estimated 910 G of plutonium. the criticality analysis considers reconfiguration and underestimation of plutonium content. The results indicate that Tank SY-101 does not present a criticality hazard. These methods are also used in criticality analyses of other Hanford tanks

  15. IRSN global process for leading a comprehensive fire safety analysis for nuclear installations

    International Nuclear Information System (INIS)

    Ormieres, Yannick; Lacoue, Jocelyne

    2013-01-01

    A fire safety analysis (FSA) is requested to justify the adequacy of fire protection measures set by the operator. A recent document written by IRSN outlines a global process for such a comprehensive fire safety analysis. Thanks to the French nuclear fire safety regulation evolutions, from prescriptive requirements to objective requirements, the proposed fire safety justification process focuses on compliance with performance criteria for fire protection measures. These performance criteria are related to the vulnerability of targets to effects of fire, and not only based upon radiological consequences out side the installation caused by a fire. In his FSA, the operator has to define the safety functions that should continue to ensure its mission even in the case of fire in order to be in compliance with nuclear safety objectives. Then, in order to maintain these safety functions, the operator has to justify the adequacy of fire protection measures, defined according to defence in depth principles. To reach the objective, the analysis process is based on the identification of targets to be protected in order to maintain safety functions, taken into account facility characteristics. These targets include structures, systems, components and personal important to safety. Facility characteristics include, for all operating conditions, potential ignition sources and fire protections systems. One of the key points of the fire analysis is the assessment of possible fire scenarios in the facility. Given the large number of possible fire scenarios, it is then necessary to evaluate 'reference fires' which are the worst case scenarios of all possible fire scenarios and which are used by the operator for the design of fire protection measures. (authors)

  16. Addendum to: Corrective Action Decision Document/Corrective Action Plan (CADD/CAP) for Corrective Action Unit (CAU) 443: Central Nevada Test Area (CNTA)-Subsurface Central Nevada Test Area, DOE/NV-977

    International Nuclear Information System (INIS)

    2008-01-01

    The environmental remediation closure process for the nuclear test at the Central Nevada Test Area (CNTA) has progressed from the approved Corrective Action Decision Document/Corrective Action Plan (CADD/CAP) to this addendum. The closure process required the installation of three monitoring/validation (MV) wells and validation analysis of the flow and transport model. The model validation analysis led to the conclusion that the hydraulic heads simulated by the flow model did not adequately predict observed heads at the MV-1, MV-2, and MV-3 validation points (wells and piezometers). The observed heads from screened intervals near the test horizon were higher than the model predicted and are believed to be the result of detonation-related effects that have persisted since the nuclear test. These effects, which include elevated heads out from the detonation zone and lower heads in the immediate vicinity of the detonation, are seen at other nuclear tests and typically dissipate within a few years. These effects were not included in the initial head distribution of the model. The head variations at CNTA are believed to have persisted due to the very low permeability of the material at the detonation level.

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

    International Nuclear Information System (INIS)

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

    1982-01-01

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

  18. Transit safety & security statistics & analysis 2002 annual report (formerly SAMIS)

    Science.gov (United States)

    2004-12-01

    The Transit Safety & Security Statistics & Analysis 2002 Annual Report (formerly SAMIS) is a compilation and analysis of mass transit accident, casualty, and crime statistics reported under the Federal Transit Administrations (FTAs) National Tr...

  19. Transit safety & security statistics & analysis 2003 annual report (formerly SAMIS)

    Science.gov (United States)

    2005-12-01

    The Transit Safety & Security Statistics & Analysis 2003 Annual Report (formerly SAMIS) is a compilation and analysis of mass transit accident, casualty, and crime statistics reported under the Federal Transit Administrations (FTAs) National Tr...

  20. Light-water reactor safety analysis codes

    International Nuclear Information System (INIS)

    Jackson, J.F.; Ransom, V.H.; Ybarrondo, L.J.; Liles, D.R.

    1980-01-01

    A brief review of the evolution of light-water reactor safety analysis codes is presented. Included is a summary comparison of the technical capabilities of major system codes. Three recent codes are described in more detail to serve as examples of currently used techniques. Example comparisons between calculated results using these codes and experimental data are given. Finally, a brief evaluation of current code capability and future development trends is presented

  1. Development of Realistic Safety Analysis Technology for CANDU Reactors

    International Nuclear Information System (INIS)

    Park, Joo Hwan; Rhee, B. W.; Rho, G. H.

    2010-04-01

    The following 3 research items have been studied to develop and establish the realistic safety analysis and the associated technologies for a CANDU reactor. At the first, WIMS-CANDU which is physics cell code for a CANDU has been improved and validated against the physics criticality experiment data transferred through the international cooperation programs. Also an improved physics model to take into account the pressure tube creep was developed and utilized to assess the effects of the pressure tube creep of 0%, 2.5% and 5% diametral increase of pressure tube on core physics parameters. Secondly, the interfacing module between physics and thermal-hydraulics codes has been developed to provide the enhancement of reliability and convenience of the calculation results of the physics parameters such as power coefficient which was calculated by independent code systems. Finally, the important parameters related to the complex heat transfer mechanisms in the crept pressure tubes were identified to find how to improve the existing fuel channel models. One of the important parameters such as the oxidation model of Zr-steam reaction was identified, implemented and verified with the experimental data of the high pressure and temperature fuel channel and its model was utilized for CFD analysis of the crept pressure tube effect on the reactor safety. The results were also provided to validate the CATNENA models of the crept pressure tube and the effects of the pressure tube creep on the blowdown and post-blowdown phase during LOCA was assessed. The results of this study can be used to assess the uncertainty analysis of coolant void reactivity and the effects of the creep deformed pressure tubes on physics/TH/safety issues. Also, those results will be used to improve the current design and operational safety analysis codes, and to technically support the related issues to resolve their problems

  2. STAMP model and its application prospect in DCS safety analysis of nuclear power plant

    International Nuclear Information System (INIS)

    Yang Xiaohua; Liu Jie; Liu Zhaohui; Liu Hua; Yu Tonglan

    2013-01-01

    The application of DCS (Digit Control System) is a certain trend for the development of nuclear power. DCS not only improves the control capability of nuclear power system, but also increases the complexity of the system. Traditional safety analysis techniques based on event-chain model are facing challenges. In order to improve the safety performance of nuclear power DCS, the latest research achievement in the field of safety engineering should be focused, studied and applied into nuclear power safety. This paper introduces a new safety analysis model named STAMP (Systems-Theoretic Accident Modeling and Processes) based on the system theory, analyzes its advantages and disadvantages compared with the traditional ones, and explains the basic steps of STPA (STAMP-Based Hazard Analysis) technology. Finally, according to the application status of STAMP at home and abroad, it prospects the development of STAMP in China's nuclear power safety. (authors)

  3. Preliminary safety analysis of the HTTR-IS nuclear hydrogen production system

    International Nuclear Information System (INIS)

    Sato, Hiroyuki; Ohashi, Hirofumi; Tazawa, Yujiro; Tachibana, Yukio; Sakaba, Nariaki

    2010-06-01

    Japan Atomic Energy Agency is planning to demonstrate hydrogen production by thermochemical water-splitting IS process utilizing heat from the high-temperature gas-cooled reactor HTTR (HTTR-IS system). The previous study identified that the HTTR modification due to the coupling of hydrogen production plant requires an additional safety review since the scenario and quantitative values of the evaluation items would be altered from the original HTTR safety review. Hence, preliminary safety analyses are conducted by using the system analysis code. Calculation results showed that evaluation items such as a coolant pressure, temperatures of heat transfer tubes at the pressure boundary, etc., did not exceed allowable values. Also, the peak fuel temperature did not exceed allowable value and therefore the reactor core was not damaged and cooled sufficiently. This report compiles calculation conditions, event scenarios and the calculation results of the preliminary safety analysis. (author)

  4. The safety analysis report for nuclear power plants in the Federal Republic of Germany

    International Nuclear Information System (INIS)

    Kohler, H.A.G.

    1976-01-01

    On the basis of the standard format for drawing up safety analysis reports for stationary fission reactors - this format was published in 1959 by the competent Federal Ministry for Atomic Energy - a report is made on the format and scope of German safety analysis reports. The reasons for drawing up a standard safety analysis report for nuclear power plants with pressurized water reactor or boiling water reactor and a KWU draft are discussed. (orig./RW) [de

  5. System and software safety analysis for the ERA control computer

    International Nuclear Information System (INIS)

    Beerthuizen, P.G.; Kruidhof, W.

    2001-01-01

    The European Robotic Arm (ERA) is a seven degrees of freedom relocatable anthropomorphic robotic manipulator system, to be used in manned space operation on the International Space Station, supporting the assembly and external servicing of the Russian segment. The safety design concept and implementation of the ERA is described, in particular with respect to the central computer's software design. A top-down analysis and specification process is used to down flow the safety aspects of the ERA system towards the subsystems, which are produced by a consortium of companies in many countries. The user requirements documents and the critical function list are the key documents in this process. Bottom-up analysis (FMECA) and test, on both subsystem and system level, are the basis for safety verification. A number of examples show the use of the approach and methods used

  6. Risk prediction, safety analysis and quantitative probability methods - a caveat

    International Nuclear Information System (INIS)

    Critchley, O.H.

    1976-01-01

    Views are expressed on the use of quantitative techniques for the determination of value judgements in nuclear safety assessments, hazard evaluation, and risk prediction. Caution is urged when attempts are made to quantify value judgements in the field of nuclear safety. Criteria are given the meaningful application of reliability methods but doubts are expressed about their application to safety analysis, risk prediction and design guidances for experimental or prototype plant. Doubts are also expressed about some concomitant methods of population dose evaluation. The complexities of new designs of nuclear power plants make the problem of safety assessment more difficult but some possible approaches are suggested as alternatives to the quantitative techniques criticized. (U.K.)

  7. Design and safety analysis of the helium cooled solid breeder blanket for CFETR

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Shuai; Zhou, Guangming; Lv, Zhongliang; Jin, Cheng; Chen, Hongli [University of Science and Technology of China, Anhui (China). School of Nuclear Science and Technology

    2016-05-15

    This paper reports the design and safety analysis results of the helium cooled solid breeder blanket of the Chinese Fusion Engineering Test Reactor (CFETR). Materials selection and basic structure of the blanket have been presented. Performance analysis including neutronics analysis and thermo-mechanical analysis has shown good results. And the safety analysis of the blanket under Loss Of Coolant Accident (LOCA) conditions has been described. Results showed the current design can deal well with the selected accident scenarios.

  8. An Integrated Approach of Model checking and Temporal Fault Tree for System Safety Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Koh, Kwang Yong; Seong, Poong Hyun [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2009-10-15

    Digitalization of instruments and control systems in nuclear power plants offers the potential to improve plant safety and reliability through features such as increased hardware reliability and stability, and improved failure detection capability. It however makes the systems and their safety analysis more complex. Originally, safety analysis was applied to hardware system components and formal methods mainly to software. For software-controlled or digitalized systems, it is necessary to integrate both. Fault tree analysis (FTA) which has been one of the most widely used safety analysis technique in nuclear industry suffers from several drawbacks as described in. In this work, to resolve the problems, FTA and model checking are integrated to provide formal, automated and qualitative assistance to informal and/or quantitative safety analysis. Our approach proposes to build a formal model of the system together with fault trees. We introduce several temporal gates based on timed computational tree logic (TCTL) to capture absolute time behaviors of the system and to give concrete semantics to fault tree gates to reduce errors during the analysis, and use model checking technique to automate the reasoning process of FTA.

  9. Safety evaluation review of the prototype license application safety analysis report

    International Nuclear Information System (INIS)

    1991-08-01

    The US Nuclear Regulatory Commission (NRC) staff and consultants reviewed a Prototype License Application Safety Analysis Report (PLASAR) submitted by the US Department of Energy (DOE) for the belowground vault (BGV) alternative method of low-level radioactive waste disposal. In Volume 1 of NUREG-1375, the NRC staff provided the safety review results for an earth-mounded concrete bunker PLASAR. In the current report, the staff focused its review on the design, construction, and operational aspects of the BGV PLASAR. The staff developed review comments and questions using the Standard Review Plan (SRP), Rev. 1 (NUREG-1200) as the basis for evaluating the acceptability of the information provided in the BGV PLASAR. The detailed review comments provided in this report are intended to be useful guidance to facility developers and State regulators in addressing issues likely to be encountered in the review of a license application for a low-level-waste disposal facility. 44 refs

  10. RISMC Advanced Safety Analysis Project Plan – FY 2015 - FY 2019

    Energy Technology Data Exchange (ETDEWEB)

    Szilard, Ronaldo H. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Smith, Curtis L. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Youngblood, Robert [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2014-09-01

    In this report, a project plan is developed, focused on industry applications, using Risk-Informed Safety Margin Characterization (RISMC) tools and methods applied to realistic, relevant, and current interest issues to the operating nuclear fleet. RISMC focuses on modernization of nuclear power safety analysis (tools, methods and data); implementing state-of-the-art modeling techniques (which include, for example, enabling incorporation of more detailed physics as they become available); taking advantage of modern computing hardware; and combining probabilistic and mechanistic analyses to enable a risk informed safety analysis process. The modernized tools will maintain the current high level of safety in our nuclear power plant fleet, while providing an improved understanding of safety margins and the critical parameters that affect them. Thus, the set of tools will provide information to inform decisions on plant modifications, refurbishments, and surveillance programs, while improving economics. This set of tools will also benefit the design of new reactors, enhancing safety per unit cost of a nuclear plant. The proposed plan will focus on application of the RISMC toolkit, in particular, solving realistic problems of important current issues to the nuclear industry, in collaboration with plant owners and operators to demonstrate the usefulness of these tools in decision making.

  11. Calculation of partial derivatives of thermophysical properties of sodium for safety analysis

    International Nuclear Information System (INIS)

    Shan Jianqiang; Qiu Suizhang; Zhu Jizhou; Zhang Guiqin

    1997-01-01

    According to the characters of safety analysis of LMFBR, the partial derivatives formula of some special thermophysical properties of sodium, including single-and two-phase properties, are calculated based on the basic Maxwell equations, and on the formulae of basic thermophysical properties of sodium which were verified abroad. The present study can provide theoretical base for safety analysis of LMFBR

  12. Improvement of safety by analysis of costs and benefits of the system

    OpenAIRE

    T. Karkoszka; M. Andraczke

    2011-01-01

    Purpose: of the paper has been the assessment of the dependence between improvement of the implemented occupational health and safety management system and both minimization of costs connected with occupational health and safety assurance and optimization of real work conditions.Design/methodology/approach: used for the analysis has included definition of the occupational health and safety system with regard to the rules and tool allowing for occupational safety assurance in the organisationa...

  13. Software Safety Analysis of Digital Protection System Requirements Using a Qualitative Formal Method

    International Nuclear Information System (INIS)

    Lee, Jang-Soo; Kwon, Kee-Choon; Cha, Sung-Deok

    2004-01-01

    The safety analysis of requirements is a key problem area in the development of software for the digital protection systems of a nuclear power plant. When specifying requirements for software of the digital protection systems and conducting safety analysis, engineers find that requirements are often known only in qualitative terms and that existing fault-tree analysis techniques provide little guidance on formulating and evaluating potential failure modes. A framework for the requirements engineering process is proposed that consists of a qualitative method for requirements specification, called the qualitative formal method (QFM), and a safety analysis method for the requirements based on causality information, called the causal requirements safety analysis (CRSA). CRSA is a technique that qualitatively evaluates causal relationships between software faults and physical hazards. This technique, extending the qualitative formal method process and utilizing information captured in the state trajectory, provides specific guidelines on how to identify failure modes and the relationship among them. The QFM and CRSA processes are described using shutdown system 2 of the Wolsong nuclear power plants as the digital protection system example

  14. Computer graphics in reactor safety analysis

    International Nuclear Information System (INIS)

    Fiala, C.; Kulak, R.F.

    1989-01-01

    This paper describes a family of three computer graphics codes designed to assist the analyst in three areas: the modelling of complex three-dimensional finite element models of reactor structures; the interpretation of computational results; and the reporting of the results of numerical simulations. The purpose and key features of each code are presented. The graphics output used in actual safety analysis are used to illustrate the capabilities of each code. 5 refs., 10 figs

  15. Operator reliability study for Probabilistic Safety Analysis of an operating research reactor

    International Nuclear Information System (INIS)

    Mohamed, F.; Hassan, A.; Yahaya, R.; Rahman, I.; Maskin, M.; Praktom, P.; Charlie, F.

    2015-01-01

    Highlights: • Human Reliability Analysis (HRA) for Level 1 Probabilistic Safety Analysis (PSA) is performed on research nuclear reactor. • Implemented qualitative HRA framework is addressed. • Human Failure Events of significant impact to the reactor safety are derived. - Abstract: A Level 1 Probabilistic Safety Analysis (PSA) for the TRIGA Mark II research reactor of Malaysian Nuclear Agency has been developed to evaluate the potential risk in its operation. In conjunction to this PSA development, Human Reliability Analysis (HRA) is performed in order to determine human contribution to the risk. The aim of this study is to qualitatively analyze human actions (HAs) involved in the operation of this reactor according to the qualitative part of the HRA framework for PSA which is namely the identification, qualitative screening and modeling of HAs. By performing this framework, Human Failure Events (HFEs) of significant impact to the reactor safety are systematically analyzed and incorporated into the PSA structure. A part of the findings in this study will become the input for the subsequent quantitative part of the HRA framework, i.e. the Human Error Probability (HEP) quantification

  16. Uncertainty and sensitivity analysis in a Probabilistic Safety Analysis level-1

    International Nuclear Information System (INIS)

    Nunez Mc Leod, Jorge E.; Rivera, Selva S.

    1996-01-01

    A methodology for sensitivity and uncertainty analysis, applicable to a Probabilistic Safety Assessment Level I has been presented. The work contents are: correct association of distributions to parameters, importance and qualification of expert opinions, generations of samples according to sample sizes, and study of the relationships among system variables and systems response. A series of statistical-mathematical techniques are recommended along the development of the analysis methodology, as well as different graphical visualization for the control of the study. (author)

  17. Implementation of a patient safety program at a tertiary health system: A longitudinal analysis of interventions and serious safety events.

    Science.gov (United States)

    Cropper, Douglas P; Harb, Nidal H; Said, Patricia A; Lemke, Jon H; Shammas, Nicolas W

    2018-04-01

    We hypothesize that implementation of a safety program based on high reliability organization principles will reduce serious safety events (SSE). The safety program focused on 7 essential elements: (a) safety rounding, (b) safety oversight teams, (c) safety huddles, (d) safety coaches, (e) good catches/safety heroes, (f) safety education, and (g) red rule. An educational curriculum was implemented focusing on changing high-risk behaviors and implementing critical safety policies. All unusual occurrences were captured in the Midas system and investigated by risk specialists, the safety officer, and the chief medical officer. A multidepartmental committee evaluated these events, and a root cause analysis (RCA) was performed. Events were tabulated and serious safety event (SSE) recorded and plotted over time. Safety success stories (SSSs) were also evaluated over time. A steady drop in SSEs was seen over 9 years. Also a rise in SSSs was evident, reflecting on staff engagement in the program. The parallel change in SSEs, SSSs, and the implementation of various safety interventions highly suggest that the program was successful in achieving its goals. A safety program based on high-reliability organization principles and made a core value of the institution can have a significant positive impact on reducing SSEs. © 2018 American Society for Healthcare Risk Management of the American Hospital Association.

  18. Safety analysis on Non-LOCA events for the revision of Wolsong NPP unit 2,3,4 sar

    International Nuclear Information System (INIS)

    Kim, Jong Hyun; Jin, Dong Sik; Ryu, Eui Seung; Kho, Dong Wook; Kim, Sung Min

    2015-01-01

    Korean Wolsong Nuclear Power Plant Units 2,3,4 (CANDU-6 Type) has prepared the revision of safety analysis report (Final Safety Analysis Report (FSAR) chapter 15) from the original performed in the year of 1990s, using the updated and state-of-the-art methodology and tools including IST safety analysis codes and more detail modelling. Compared with the original FSAR15, the revised FSAR15 has significant improvement in both the scope and the depth of safety analysis, which has demonstrated the safety analysis results have complied with the safety requirements(acceptance criteria). This paper will present the analysis scope for Non-LOCA events re-analyzed or added for the FSAR15 revision, methodologies applied such as codes and modelling and some important analysis results will be demonstrated with comparison to acceptance criteria. Application of more detail and near-realistic assumptions and method including Dev-PDO options and uncertainty related to the CHF correlations has altogether brought about more safety margin compared with the original FSAR15 with respect to SDS trip effectiveness etc. (author)

  19. Safety analysis report for packaging (onsite) multicanister overpack cask

    International Nuclear Information System (INIS)

    Edwards, W.S.

    1997-01-01

    This safety analysis report for packaging (SARP) documents the safety of shipments of irradiated fuel elements in the MUlticanister Overpack (MCO) and MCO Cask for a highway route controlled quantity, Type B fissile package. This SARP evaluates the package during transfers of (1) water-filled MCOs from the K Basins to the Cold Vacuum Drying Facility (CVDF) and (2) sealed and cold vacuum dried MCOs from the CVDF in the 100 K Area to the Canister Storage Building in the 200 East Area

  20. Safety analysis report for packaging (onsite) multicanister overpack cask

    Energy Technology Data Exchange (ETDEWEB)

    Edwards, W.S.

    1997-07-14

    This safety analysis report for packaging (SARP) documents the safety of shipments of irradiated fuel elements in the MUlticanister Overpack (MCO) and MCO Cask for a highway route controlled quantity, Type B fissile package. This SARP evaluates the package during transfers of (1) water-filled MCOs from the K Basins to the Cold Vacuum Drying Facility (CVDF) and (2) sealed and cold vacuum dried MCOs from the CVDF in the 100 K Area to the Canister Storage Building in the 200 East Area.

  1. Mixed Waste Management Facility Preliminary Safety Analysis Report. Chapters 1 to 20

    Energy Technology Data Exchange (ETDEWEB)

    1994-09-01

    This document provides information on waste management practices, occupational safety, and a site characterization of the Lawrence Livermore National Laboratory. A facility description, safety engineering analysis, mixed waste processing techniques, and auxiliary support systems are included.

  2. Mixed Waste Management Facility Preliminary Safety Analysis Report. Chapters 1 to 20

    International Nuclear Information System (INIS)

    1994-09-01

    This document provides information on waste management practices, occupational safety, and a site characterization of the Lawrence Livermore National Laboratory. A facility description, safety engineering analysis, mixed waste processing techniques, and auxiliary support systems are included

  3. Screening of Gas-Cooled Reactor Thermal-Hydraulic and Safety Analysis Tools and Experimental Database

    International Nuclear Information System (INIS)

    Lee, Won Jae; Kim, Min Hwan; Lee, Seung Wook

    2007-08-01

    This report is a final report of I-NERI Project, 'Screening of Gas-cooled Reactor Thermal Hydraulic and Safety Analysis Tools and Experimental Database 'jointly carried out by KAERI, ANL and INL. In this study, we developed the basic technologies required to develop and validate the VHTR TH/safety analysis tools and evaluated the TH/safety database information. The research tasks consist of; 1) code qualification methodology (INL), 2) high-level PIRTs for major nucleus set of events (KAERI, ANL, INL), 3) initial scaling and scoping analysis (ANL, KAERI, INL), 4) filtering of TH/safety tools (KAERI, INL), 5) evaluation of TH/safety database information (KAERI, INL, ANL) and 6) key scoping analysis (KAERI). The code qualification methodology identifies the role of PIRTs in the R and D process and the bottom-up and top-down code validation methods. Since the design of VHTR is still evolving, we generated the high-level PIRTs referencing 600MWth block-type GT-MHR and 400MWth pebble-type PBMR. Nucleus set of events that represents the VHTR safety and operational transients consists of the enveloping scenarios of HPCC (high pressure conduction cooling: loss of primary flow), LPCC/Air-Ingress (low pressure conduction cooling: loss of coolant), LC (load changes: power maneuvering), ATWS (anticipated transients without scram: reactivity insertion), WS (water ingress: water-interfacing system break) and HU (hydrogen-side upset: loss of heat sink). The initial scaling analysis defines dimensionless parameters that need to be reflected in mixed convection modeling and the initial scoping analysis provided the reference system transients used in the PIRTs generation. For the PIRTs phenomena, we evaluated the modeling capability of the candidate TH/safety tools and derived a model improvement need. By surveying and evaluating the TH/safety database information, a tools V and V matrix has been developed. Through the key scoping analysis using available database, the modeling

  4. Screening of Gas-Cooled Reactor Thermal-Hydraulic and Safety Analysis Tools and Experimental Database

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Won Jae; Kim, Min Hwan; Lee, Seung Wook (and others)

    2007-08-15

    This report is a final report of I-NERI Project, 'Screening of Gas-cooled Reactor Thermal Hydraulic and Safety Analysis Tools and Experimental Database 'jointly carried out by KAERI, ANL and INL. In this study, we developed the basic technologies required to develop and validate the VHTR TH/safety analysis tools and evaluated the TH/safety database information. The research tasks consist of; 1) code qualification methodology (INL), 2) high-level PIRTs for major nucleus set of events (KAERI, ANL, INL), 3) initial scaling and scoping analysis (ANL, KAERI, INL), 4) filtering of TH/safety tools (KAERI, INL), 5) evaluation of TH/safety database information (KAERI, INL, ANL) and 6) key scoping analysis (KAERI). The code qualification methodology identifies the role of PIRTs in the R and D process and the bottom-up and top-down code validation methods. Since the design of VHTR is still evolving, we generated the high-level PIRTs referencing 600MWth block-type GT-MHR and 400MWth pebble-type PBMR. Nucleus set of events that represents the VHTR safety and operational transients consists of the enveloping scenarios of HPCC (high pressure conduction cooling: loss of primary flow), LPCC/Air-Ingress (low pressure conduction cooling: loss of coolant), LC (load changes: power maneuvering), ATWS (anticipated transients without scram: reactivity insertion), WS (water ingress: water-interfacing system break) and HU (hydrogen-side upset: loss of heat sink). The initial scaling analysis defines dimensionless parameters that need to be reflected in mixed convection modeling and the initial scoping analysis provided the reference system transients used in the PIRTs generation. For the PIRTs phenomena, we evaluated the modeling capability of the candidate TH/safety tools and derived a model improvement need. By surveying and evaluating the TH/safety database information, a tools V and V matrix has been developed. Through the key scoping analysis using available database, the

  5. Modern design and safety analysis of the University of Florida Training Reactor

    International Nuclear Information System (INIS)

    Jordan, K.A.; Springfels, D.; Schubring, D.

    2015-01-01

    Highlights: • A new safety analysis of the University of Florida Training Reactor is presented. • This analysis uses modern codes and replaces the NRC approved analysis from 1982. • Reduction in engineering margin confirms that the UFTR is a negligible risk reactor. • Safety systems are not required to ensure that safety limits are not breached. • Negligible risk reactors are ideal for testing digital I&C equipment. - Abstract: A comprehensive series of neutronics and thermal hydraulics analyses were conducted to demonstrate the University of Florida Training Reactor (UFTR), an ARGONAUT type research reactor, as a negligible risk reactor that does not require safety-related systems or components to prevent breach of a safety limit. These analyses show that there is no credible UFTR accident that would result in major fuel damage or risk to public health and safety. The analysis was based on two limiting scenarios, whose extremity bound all other accidents of consequence: (1) the large step insertion of positive reactivity and (2) the release of fission products due to mechanical damage to a spent fuel plate. The maximum step insertion of positive reactivity was modeled using PARET/ANL software and shows a maximum peak fuel temperature of 283.2 °C, which is significantly below the failure limit of 530 °C. The exposure to the staff and general public was calculated for the worst-case fission product release scenario using the ORIGEN-S and COMPLY codes and was shown to be 6.5% of the annual limit. Impacts on reactor operations and an Instrumentation & Control System (I&C) upgrade are discussed

  6. Modern design and safety analysis of the University of Florida Training Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jordan, K.A., E-mail: kjordan@ufl.edu [University of Florida, 106 UFTR Bldg., PO Box 116400, Gainesville, FL 32611-6400 (United States); Springfels, D., E-mail: dspringfels@ufl.edu [University of Florida, 106 UFTR Bldg., PO Box 116400, Gainesville, FL 32611-6400 (United States); Schubring, D., E-mail: dlschubring@ufl.edu [University of Florida, 202 Nuclear Science Building, PO Box 118300, Gainesville, FL 32611-8300 (United States)

    2015-05-15

    Highlights: • A new safety analysis of the University of Florida Training Reactor is presented. • This analysis uses modern codes and replaces the NRC approved analysis from 1982. • Reduction in engineering margin confirms that the UFTR is a negligible risk reactor. • Safety systems are not required to ensure that safety limits are not breached. • Negligible risk reactors are ideal for testing digital I&C equipment. - Abstract: A comprehensive series of neutronics and thermal hydraulics analyses were conducted to demonstrate the University of Florida Training Reactor (UFTR), an ARGONAUT type research reactor, as a negligible risk reactor that does not require safety-related systems or components to prevent breach of a safety limit. These analyses show that there is no credible UFTR accident that would result in major fuel damage or risk to public health and safety. The analysis was based on two limiting scenarios, whose extremity bound all other accidents of consequence: (1) the large step insertion of positive reactivity and (2) the release of fission products due to mechanical damage to a spent fuel plate. The maximum step insertion of positive reactivity was modeled using PARET/ANL software and shows a maximum peak fuel temperature of 283.2 °C, which is significantly below the failure limit of 530 °C. The exposure to the staff and general public was calculated for the worst-case fission product release scenario using the ORIGEN-S and COMPLY codes and was shown to be 6.5% of the annual limit. Impacts on reactor operations and an Instrumentation & Control System (I&C) upgrade are discussed.

  7. Improving patient safety in radiotherapy through error reporting and analysis

    International Nuclear Information System (INIS)

    Findlay, Ú.; Best, H.; Ottrey, M.

    2016-01-01

    Aim: To improve patient safety in radiotherapy (RT) through the analysis and publication of radiotherapy errors and near misses (RTE). Materials and methods: RTE are submitted on a voluntary basis by NHS RT departments throughout the UK to the National Reporting and Learning System (NRLS) or directly to Public Health England (PHE). RTE are analysed by PHE staff using frequency trend analysis based on the classification and pathway coding from Towards Safer Radiotherapy (TSRT). PHE in conjunction with the Patient Safety in Radiotherapy Steering Group publish learning from these events, on a triannual and summarised on a biennial basis, so their occurrence might be mitigated. Results: Since the introduction of this initiative in 2010, over 30,000 (RTE) reports have been submitted. The number of RTE reported in each biennial cycle has grown, ranging from 680 (2010) to 12,691 (2016) RTE. The vast majority of the RTE reported are lower level events, thus not affecting the outcome of patient care. Of the level 1 and 2 incidents reported, it is known the majority of them affected only one fraction of a course of treatment. This means that corrective action could be taken over the remaining treatment fractions so the incident did not have a significant impact on the patient or the outcome of their treatment. Analysis of the RTE reports demonstrates that generation of error is not confined to one professional group or to any particular point in the pathway. It also indicates that the pattern of errors is replicated across service providers in the UK. Conclusion: Use of the terminology, classification and coding of TSRT, together with implementation of the national voluntary reporting system described within this report, allows clinical departments to compare their local analysis to the national picture. Further opportunities to improve learning from this dataset must be exploited through development of the analysis and development of proactive risk management strategies

  8. Final safety analysis report for the Ground Test Accelerator (GTA), Phase 2

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1994-10-01

    This document is the second volume of a 3 volume safety analysis report on the Ground Test Accelerator (GTA). The GTA program at the Los Alamos National Laboratory (LANL) is the major element of the national Neutral Particle Beam (NPB) program, which is supported by the Strategic Defense Initiative Office (SDIO). A principal goal of the national NPB program is to assess the feasibility of using hydrogen and deuterium neutral particle beams outside the Earth`s atmosphere. The main effort of the NPB program at Los Alamos concentrates on developing the GTA. The GTA is classified as a low-hazard facility, except for the cryogenic-cooling system, which is classified as a moderate-hazard facility. This volume consists of failure modes and effects analysis; accident analysis; operational safety requirements; quality assurance program; ES&H management program; environmental, safety, and health systems critical to safety; summary of waste-management program; environmental monitoring program; facility expansion, decontamination, and decommissioning; summary of emergency response plan; summary plan for employee training; summary plan for operating procedures; glossary; and appendices A and B.

  9. Study on mixed analysis method for fatigue analysis of oblique safety injection nozzle on main piping

    International Nuclear Information System (INIS)

    Lu Xifeng; Zhang Yixiong; Ai Honglei; Wang Xinjun; He Feng

    2014-01-01

    The simplified analysis method and the detailed analysis method were used for the fatigue analysis of the nozzle on the main piping. Because the structure of the oblique safety injection nozzle is complex and some more severe transients are subjected. The results obtained are more penalized and cannot be validate when the simplified analysis method used for the fatigue analysis. It will be little conservative when the detailed analysis method used, but it is more complex and time-consuming and boring labor. To reduce the conservatism and save time, the mixed analysis method which combining the simplified analysis method with the detailed analysis method is used for the fatigue analysis. The heat transfer parameters between the fluid and the structure which used for analysis were obtained by heat transfer property experiment. The results show that the mixed analysis which heat transfer property is considered can reduce the conservatism effectively, and the mixed analysis method is a more effective and practical method used for the fatigue analysis of the oblique safety injection nozzle. (authors)

  10. Evolution of Safety Analysis to Support New Exploration Missions

    Science.gov (United States)

    Thrasher, Chard W.

    2008-01-01

    NASA is currently developing the Ares I launch vehicle as a key component of the Constellation program which will provide safe and reliable transportation to the International Space Station, back to the moon, and later to Mars. The risks and costs of the Ares I must be significantly lowered, as compared to other manned launch vehicles, to enable the continuation of space exploration. It is essential that safety be significantly improved, and cost-effectively incorporated into the design process. This paper justifies early and effective safety analysis of complex space systems. Interactions and dependences between design, logistics, modeling, reliability, and safety engineers will be discussed to illustrate methods to lower cost, reduce design cycles and lessen the likelihood of catastrophic events.

  11. Probabilistic safety analysis for control rod drive system of ET-RR-1

    International Nuclear Information System (INIS)

    Nasr, M.; Nasser, O.

    1988-01-01

    The International Atomic Energy Agency (IAEA) co-ordinated a Research programme on Probabilistic Safety Analysis (PSA) for research reactors; with the participation of several countries. In the framework of this project (Project Int. 9/063) the Egyptian Atomic Energy Authority decided to perform a PSA study on the ET-RR-1 (Egypt Thermal Research Reactor). The study is conducted in collaboration between the nuclear regulatory and safety centre (NRSC) and the reactor department of the nuclear research centre at Inchass. The present work is a part of the PSA study on ET-RR- it is concerning a probabilistic safety analysis of the control rod drive mechanism

  12. Safety Assessment for Research Reactors and Preparation of the Safety Analysis Report. Specific Safety Guide

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2011-11-15

    The IAEA's Statute authorizes the Agency to 'establish or adopt' standards of safety for protection of health and minimization of danger to life and property' - standards that the IAEA must use in its own operations, and which States can apply by means of their regulatory provisions for nuclear and radiation safety. The IAEA does this in consultation with the competent organs of the United Nations and with the specialized agencies concerned. A comprehensive set of high quality standards under regular review is a key element of a stable and sustainable global safety regime, as is the IAEA's assistance in their application. The IAEA commenced its safety standards programme in 1958. The emphasis placed on quality, fitness for purpose and continuous improvement has led to the widespread use of the IAEA standards throughout the world. The Safety Standards Series now includes unified Fundamental Safety Principles, which represent an international consensus on what must constitute a high level of protection and safety. With the strong support of the Commission on Safety Standards, the IAEA is working to promote the global acceptance and use of its standards. Standards are only effective if they are properly applied in practice. The IAEA's safety services encompass design, siting and engineering safety, operational safety, radiation safety, safe transport of radioactive material and safe management of radioactive waste, as well as governmental organization, regulatory matters and safety culture in organizations. These safety services assist Member States in the application of the standards and enable valuable experience and insights to be shared. Regulating safety is a national responsibility, and many States have decided to adopt the IAEA's standards for use in their national regulations. For parties to the various international safety conventions, IAEA standards provide a consistent, reliable means of ensuring the effective fulfilment of obligations under the conventions

  13. Safety Assessment for Research Reactors and Preparation of the Safety Analysis Report. Specific Safety Guide

    International Nuclear Information System (INIS)

    2011-01-01

    The IAEA's Statute authorizes the Agency to 'establish or adopt' standards of safety for protection of health and minimization of danger to life and property' - standards that the IAEA must use in its own operations, and which States can apply by means of their regulatory provisions for nuclear and radiation safety. The IAEA does this in consultation with the competent organs of the United Nations and with the specialized agencies concerned. A comprehensive set of high quality standards under regular review is a key element of a stable and sustainable global safety regime, as is the IAEA's assistance in their application. The IAEA commenced its safety standards programme in 1958. The emphasis placed on quality, fitness for purpose and continuous improvement has led to the widespread use of the IAEA standards throughout the world. The Safety Standards Series now includes unified Fundamental Safety Principles, which represent an international consensus on what must constitute a high level of protection and safety. With the strong support of the Commission on Safety Standards, the IAEA is working to promote the global acceptance and use of its standards. Standards are only effective if they are properly applied in practice. The IAEA's safety services encompass design, siting and engineering safety, operational safety, radiation safety, safe transport of radioactive material and safe management of radioactive waste, as well as governmental organization, regulatory matters and safety culture in organizations. These safety services assist Member States in the application of the standards and enable valuable experience and insights to be shared. Regulating safety is a national responsibility, and many States have decided to adopt the IAEA's standards for use in their national regulations. For parties to the various international safety conventions, IAEA standards provide a consistent, reliable means of ensuring the effective fulfilment of obligations under the conventions

  14. An Empirical Analysis of Human Performance and Nuclear Safety Culture

    International Nuclear Information System (INIS)

    Jeffrey Joe; Larry G. Blackwood

    2006-01-01

    The purpose of this analysis, which was conducted for the US Nuclear Regulatory Commission (NRC), was to test whether an empirical connection exists between human performance and nuclear power plant safety culture. This was accomplished through analyzing the relationship between a measure of human performance and a plant's Safety Conscious Work Environment (SCWE). SCWE is an important component of safety culture the NRC has developed, but it is not synonymous with it. SCWE is an environment in which employees are encouraged to raise safety concerns both to their own management and to the NRC without fear of harassment, intimidation, retaliation, or discrimination. Because the relationship between human performance and allegations is intuitively reciprocal and both relationship directions need exploration, two series of analyses were performed. First, human performance data could be indicative of safety culture, so regression analyses were performed using human performance data to predict SCWE. It also is likely that safety culture contributes to human performance issues at a plant, so a second set of regressions were performed using allegations to predict HFIS results

  15. ARIES-AT safety design and analysis

    Energy Technology Data Exchange (ETDEWEB)

    Petti, D.A. [Idaho National Engineering and Environmental Laboratory, Fusion Safety Program, P.O. Box 1625, Idaho Falls, ID 83415 (United States)]. E-mail: David.Petti@inl.gov; Merrill, B.J. [Idaho National Engineering and Environmental Laboratory, Fusion Safety Program, P.O. Box 1625, Idaho Falls, ID 83415 (United States); Moore, R.L. [Idaho National Engineering and Environmental Laboratory, Fusion Safety Program, P.O. Box 1625, Idaho Falls, ID 83415 (United States); Longhurst, G.R. [Idaho National Engineering and Environmental Laboratory, Fusion Safety Program, P.O. Box 1625, Idaho Falls, ID 83415 (United States); El-Guebaly, L. [Fusion Technology Institute, 1500 Engineering Drive, University of Wisconsin-Madison, Madison, WI 53706 (United States); Mogahed, E. [Fusion Technology Institute, 1500 Engineering Drive, University of Wisconsin-Madison, Madison, WI 53706 (United States); Henderson, D. [Fusion Technology Institute, 1500 Engineering Drive, University of Wisconsin-Madison, Madison, WI 53706 (United States); Wilson, P. [Fusion Technology Institute, 1500 Engineering Drive, University of Wisconsin-Madison, Madison, WI 53706 (United States); Abdou, A. [Fusion Technology Institute, 1500 Engineering Drive, University of Wisconsin-Madison, Madison, WI 53706 (United States)

    2006-01-15

    ARIES-AT is a 1000 MWe conceptual fusion power plant design with a very low projected cost of electricity. The design contains many innovative features to improve both the physics and engineering performance of the system. From the safety and environmental perspective, there is greater depth to the overall analysis than in past ARIES studies. For ARIES-AT, the overall spectrum of off-normal events to be examined has been broadened. They include conventional loss of coolant and loss of flow events, an ex-vessel loss of coolant, and in-vessel off-normal events that mobilize in-vessel inventories (e.g., tritium and tokamak dust) and bypass primary confinement such as a loss of vacuum and an in-vessel loss of coolant with bypass. This broader examination of accidents improves the robustness of the design from the safety perspective and gives additional confidence that the facility can meet the no-evacuation requirement under average weather conditions. We also provide a systematic assessment of the design to address key safety functions such as confinement, decay heat removal, and chemical energy control. In the area of waste management, both the volume of the component and its hazard are used to classify the waste. In comparison to previous ARIES designs, the overall waste volume is less because of the compact design.

  16. Automation for System Safety Analysis

    Science.gov (United States)

    Malin, Jane T.; Fleming, Land; Throop, David; Thronesbery, Carroll; Flores, Joshua; Bennett, Ted; Wennberg, Paul

    2009-01-01

    This presentation describes work to integrate a set of tools to support early model-based analysis of failures and hazards due to system-software interactions. The tools perform and assist analysts in the following tasks: 1) extract model parts from text for architecture and safety/hazard models; 2) combine the parts with library information to develop the models for visualization and analysis; 3) perform graph analysis and simulation to identify and evaluate possible paths from hazard sources to vulnerable entities and functions, in nominal and anomalous system-software configurations and scenarios; and 4) identify resulting candidate scenarios for software integration testing. There has been significant technical progress in model extraction from Orion program text sources, architecture model derivation (components and connections) and documentation of extraction sources. Models have been derived from Internal Interface Requirements Documents (IIRDs) and FMEA documents. Linguistic text processing is used to extract model parts and relationships, and the Aerospace Ontology also aids automated model development from the extracted information. Visualizations of these models assist analysts in requirements overview and in checking consistency and completeness.

  17. YUCCA MOUNTAIN SITE CHARACTERIZATIONS PROJECT TUNNEL BORING MACHINE (TBM) SYSTEM SAFETY ANALYSIS

    International Nuclear Information System (INIS)

    1997-01-01

    The purpose of this analysis is to systematically identify and evaluate hazards related to the tunnel boring machine (TBM) used in the Exploratory Studies Facility (ESF) at the Yucca Mountain Site Characterization Project. This process is an integral part of the systems engineering process; whereby safety is considered during planning, design, testing, and construction. Since the TBM is an ''as built'' system, the MandO is conducting the System Safety Analysis during the construction or assembly phase of the TBM. A largely qualitative approach was used since a radiological System Safety Analysis is not required. The risk assessment in this analysis characterizes the accident scenarios associated with the TBM in terms of relative risk and includes recommendations for mitigating all identified risks. The priority for recommending and implementing mitigation control features is: (1) Incorporate measures to reduce risks and hazards into the system/subsystem/component design, (2) add safety features and capabilities to existing designs, and (3) develop procedures and conduct training to increase worker awareness of potential hazards, on methods to reduce exposure to hazards, and on the actions required to avoid accidents or correct hazardous conditions. The scope of this analysis is limited to the TBM during normal operations, excluding hazards occurring during assembly and test of the TBM or maintenance of the TBM equipment

  18. YUCCA MOUNTAIN SITE CHARACTERIZATIONS PROJECT TUNNEL BORING MACHINE (TBM) SYSTEM SAFETY ANALYSIS

    Energy Technology Data Exchange (ETDEWEB)

    N/A

    1997-02-19

    The purpose of this analysis is to systematically identify and evaluate hazards related to the tunnel boring machine (TBM) used in the Exploratory Studies Facility (ESF) at the Yucca Mountain Site Characterization Project. This process is an integral part of the systems engineering process; whereby safety is considered during planning, design, testing, and construction. Since the TBM is an ''as built'' system, the M&O is conducting the System Safety Analysis during the construction or assembly phase of the TBM. A largely qualitative approach was used since a radiological System Safety Analysis is not required. The risk assessment in this analysis characterizes the accident scenarios associated with the TBM in terms of relative risk and includes recommendations for mitigating all identified risks. The priority for recommending and implementing mitigation control features is: (1) Incorporate measures to reduce risks and hazards into the system/subsystem/component design, (2) add safety features and capabilities to existing designs, and (3) develop procedures and conduct training to increase worker awareness of potential hazards, on methods to reduce exposure to hazards, and on the actions required to avoid accidents or correct hazardous conditions. The scope of this analysis is limited to the TBM during normal operations, excluding hazards occurring during assembly and test of the TBM or maintenance of the TBM equipment.

  19. Recent progress in safety-related applications of reactor noise analysis

    International Nuclear Information System (INIS)

    Hirota, Jitsuya; Shinohara, Yoshikuni; Saito, Keiichi

    1982-01-01

    Recent progress in safety-related applications of reactor noise analysis is reviewed, mainly referring to various papers presented at the Third Specialists' Meeting on Reactor Noise (SMORN-III) held in Tokyo in 1981. Advances in application of autoregressive model, coherence analysis and pattern recognition technique are significant since SMORN-II in 1977. Development of reactor diagnosis systems based on noise analysis is in progress. Practical experiences in the safety-related applications to power plants are being accumulated. Advances in quantitative monitoring of vibration of internal structures in PWR and diagnosis of core stability and control system characteristics in BWR are notable. Acoustic methods are also improved to detect sodium boiling in LMFBR. The Reactor Noise Analysis Benchmark Test performed by Japan in connection with SMORN-III is successful so that it is possible to proceed to the second stage of the benchmark test. (author)

  20. Aging effect on the fuel behaviors for CANDU fuel safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Jung, J.Y.; Bae, J.H.; Park, J.H.; Song, Y.M., E-mail: agahee@kaeri.re.kr [Korea Atomic Energy Research Inst., Yuseong-gu, Daejeon (Korea, Republic of)

    2013-07-01

    Because of the aging of heat transport system components, the reactor thermalhydraulic conditions can vary, which may affect the safety response. In a recent safety analysis for the refurbished Wolsong 1 NPP, various aging effects were incorporated into the hydraulic models of the components in the primary heat transport system (PHTS) for conservatism. The aging data of the thermal-hydraulic components for an 11 EFPY of Wolsong 1 were derived based on the site operation data and were modified to the appropriate input data for the thermal-hydraulic code for a safety analysis of a postulated accident. This paper deals with the aging effect of the PHTS of the CANDU reactor on the fuel performance during normal operation and transient period following a postulated accident such as a feeder stagnation break. (author)

  1. Core conversion effects on the safety analysis of research reactors

    International Nuclear Information System (INIS)

    Anoussis, J.N.; Chrysochoides, N.G.; Papastergiou, C.N.

    1982-07-01

    The safety related parameters of the 5 MW Democritus research reactor that will be affected by the scheduled core conversion to use LEU instead of HEU are considered. The analysis of the safety related items involved in such a core conversion, mainly the consequences due to MCA, DBA, etc., is of a general nature and can, therefore, be applied to other similar pool type reactors as well. (T.A.)

  2. Analysis of factors influencing safety management for metro construction in China.

    Science.gov (United States)

    Yu, Q Z; Ding, L Y; Zhou, C; Luo, H B

    2014-07-01

    With the rapid development of urbanization in China, the number and size of metro construction projects are increasing quickly. At the same time, and increasing number of accidents in metro construction make it a disturbing focus of social attention. In order to improve safety management in metro construction, an investigation of the participants' perspectives on safety factors in China metro construction has been conducted to identify the key safety factors, and their ranking consistency among the main participants, including clients, consultants, designers, contractors and supervisors. The result of factor analysis indicates that there are five key factors which influence the safety of metro construction including safety attitude, construction site safety, government supervision, market restrictions and task unpredictability. In addition, ANOVA and Spearman rank correlation coefficients were performed to test the consistency of the means rating and the ranking of safety factors. The results indicated that the main participants have significant disagreement about the importance of safety factors on more than half of the items. Suggestions and recommendations on practical countermeasures to improve metro construction safety management in China are proposed. Copyright © 2013 Elsevier Ltd. All rights reserved.

  3. 49 CFR 209.501 - Review of rail transportation safety and security route analysis.

    Science.gov (United States)

    2010-10-01

    ... 49 Transportation 4 2010-10-01 2010-10-01 false Review of rail transportation safety and security....820 § 209.501 Review of rail transportation safety and security route analysis. (a) Review of route... establish that the route chosen by the carrier poses the least overall safety and security risk, the...

  4. Analysis of the criticality safety of a nuclear fuel deposit

    International Nuclear Information System (INIS)

    Landeyro, P.A.; Mincarini, M.

    1987-01-01

    In the present work a safety analysis from criticality accidents of nuclear fuel deposits is performed. The analysis is performed utilizing two methods derived from different physical principes: 1) superficial density method, obtained from experimental research; 2) solid angle method, derived from transport theory

  5. 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

  6. Annual Status Report (Fiscal Year 2012) Composite Analysis of Low-Level Waste Disposal in the Central Plateau at the Hanford Site

    Energy Technology Data Exchange (ETDEWEB)

    Nichols, W. E. [CH2M HILL Plateau Remediation Company, Richland, WA (United States)

    2012-12-27

    In accordance with U.S. Department of Energy (DOE) requirements in DOE O 435.1, Chg. 1,1 and as implemented by DOE/RL-2000-29, Rev. 2,2 the DOE Richland Operations Office (DOE-RL) has prepared this annual summary of the Hanford Site Composite Analysis for fiscal year (FY) 2012, as originally reported in PNNL-118003 and PNNL-11800, Addendum 14 (hereafter these reports are referred to collectively as the Composite Analysis), and to address secondary issues identified during the review of the Composite Analysis.

  7. Model-based safety analysis of a control system using Simulink and Simscape extended models

    Directory of Open Access Journals (Sweden)

    Shao Nian

    2017-01-01

    Full Text Available The aircraft or system safety assessment process is an integral part of the overall aircraft development cycle. It is usually characterized by a very high timely and financial effort and can become a critical design driver in certain cases. Therefore, an increasing demand of effective methods to assist the safety assessment process arises within the aerospace community. One approach is the utilization of model-based technology, which is already well-established in the system development, for safety assessment purposes. This paper mainly describes a new tool for Model-Based Safety Analysis. A formal model for an example system is generated and enriched with extended models. Then, system safety analyses are performed on the model with the assistance of automation tools and compared to the results of a manual analysis. The objective of this paper is to improve the increasingly complex aircraft systems development process. This paper develops a new model-based analysis tool in Simulink/Simscape environment.

  8. Supplement to safety analysis report. 306-W building operations safety requirement

    International Nuclear Information System (INIS)

    Richey, C.R.

    1979-08-01

    The operations safety requirements (OSRs) presented in this report define the conditions, safe boundaries, and management control needed for safely conducting operations with radioactive materials in the Pacific Northwest Laboratory (PNL) 306-W building. The safety requirements are organized in five sections. Safety limits are safety-related process variables that are observable and measurable. Limiting conditions cover: equipment and technical conditions and characteristics of the facility and operations necessary for continued safe operation. Surveillance requirements prescribe the requirements for checking systems and components that are essential to safety. Equipment design controls require that changes to process equipment and systems be independently checked and approved to assure that the changes will have no adverse effect on safety. Administrative controls describe and discuss the organization and administrative systems and procedures to be used for safe operation of the facility. Details of the implementation of the operations safety requirements are prescribed by internal PNL documents such as criticality safety specifications and radiation work procedures

  9. Technology development of maintenance optimization and reliability analysis for safety features in nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Tae Woon; Choi, Seong Soo; Lee, Dong Gue; Kim, Young Il

    1999-12-01

    The reliability data management system (RDMS) for safety systems of PHWR type plants has been developed and utilized in the reliability analysis of the special safety systems of Wolsong Unit 1,2 with plant overhaul period lengthened. The RDMS is developed for the periodic efficient reliability analysis of the safety systems of Wolsong Unit 1,2. In addition, this system provides the function of analyzing the effects on safety system unavailability if the test period of a test procedure changes as well as the function of optimizing the test periods of safety-related test procedures. The RDMS can be utilized in handling the requests of the regulatory institute actively with regard to the reliability validation of safety systems. (author)

  10. Development of time dependent safety analysis code for plasma anomaly events in fusion reactors

    International Nuclear Information System (INIS)

    Honda, Takuro; Okazaki, Takashi; Bartels, H.W.; Uckan, N.A.; Seki, Yasushi.

    1997-01-01

    A safety analysis code SAFALY has been developed to analyze plasma anomaly events in fusion reactors, e.g., a loss of plasma control. The code is a hybrid code comprising a zero-dimensional plasma dynamics and a one-dimensional thermal analysis of in-vessel components. The code evaluates the time evolution of plasma parameters and temperature distributions of in-vessel components. As the plasma-safety interface model, we proposed a robust plasma physics model taking into account updated data for safety assessment. For example, physics safety guidelines for beta limit, density limit and H-L mode confinement transition threshold power, etc. are provided in the model. The model of the in-vessel components are divided into twenty temperature regions in the poloidal direction taking account of radiative heat transfer between each surface of each region. This code can also describe the coolant behavior under hydraulic accidents with the results by hydraulics code and treat vaporization (sublimation) from plasma facing components (PFCs). Furthermore, the code includes the model of impurity transport form PFCs by using a transport probability and a time delay. Quantitative analysis based on the model is possible for a scenario of plasma passive shutdown. We examined the possibility of the code as a safety analysis code for plasma anomaly events in fusion reactors and had a prospect that it would contribute to the safety analysis of the International Thermonuclear Experimental Reactor (ITER). (author)

  11. Dependability Assessment by Static Analysis of Software Important to Nuclear Power Plant Safety

    Energy Technology Data Exchange (ETDEWEB)

    Ourghanlian, Alain [EDF Lab, Chatou (France)

    2014-08-15

    We describe a practical experimentation of safety assessment of safety-critical software used in Nuclear Power Plants. To enhance the credibility of safety assessments and to optimize safety justification costs, Electricite de France (EDF) investigates the use of methods and tools for source code semantic analysis, to obtain indisputable evidence and help assessors focus on the most critical issues. EDF has been using the PolySpace tool for more than 10 years. Today, new industrial tools, based on the same formal approach, Abstract Interpretation, are available. Practical experimentation with these new tools shows that the precision obtained on one of our shutdown systems software is very significantly improved. In a first part, we present the analysis principles of the tools used in our experimentation. In a second part, we present the main characteristics of protection-system software, and why these characteristics are well adapted for the new analysis tools. In the last part, we present an overview of the results and the limitation of the tools.

  12. ACCIDENT ANALYSES & CONTROL OPTIONS IN SUPPORT OF THE SLUDGE WATER SYSTEM SAFETY ANALYSIS

    Energy Technology Data Exchange (ETDEWEB)

    WILLIAMS, J.C.

    2003-11-15

    This report documents the accident analyses and nuclear safety control options for use in Revision 7 of HNF-SD-WM-SAR-062, ''K Basins Safety Analysis Report'' and Revision 4 of HNF-SD-SNF-TSR-001, ''Technical Safety Requirements - 100 KE and 100 KW Fuel Storage Basins''. These documents will define the authorization basis for Sludge Water System (SWS) operations. This report follows the guidance of DOE-STD-3009-94, ''Preparation Guide for US. Department of Energy Nonreactor Nuclear Facility Safety Analysis Reports'', for calculating onsite and offsite consequences. The accident analysis summary is shown in Table ES-1 below. While this document describes and discusses potential control options to either mitigate or prevent the accidents discussed herein, it should be made clear that the final control selection for any accident is determined and presented in HNF-SD-WM-SAR-062.

  13. The application of new mathematical structures to safety analysis

    International Nuclear Information System (INIS)

    Cooper, J.A.; Ross, T.J.

    1997-10-01

    Probabilistic safety analyses (PSAs) often depend on significant subjectivity. The recent successes of fuzzy logic and fuzzy and hybrid mathematics in portraying subjectivity is a reminder that a selection made from the most applicable mathematical tools is more important than forced adaptation of conventional tools. In this paper, the authors consider new approaches that enhance conventional and fuzzy PSA by improved handling of subjectivity. The most significant of the mathematical structures were have investigated (from a standpoint of safety analysis applications) will be described, and the general types of applications will be outlined

  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. Potential safety features and safety analysis aspects for high performance light water reactor (HPLWR)

    International Nuclear Information System (INIS)

    Aksan, N.; Schulenberg, T.; Squarer, D.

    2003-01-01

    Research Activities are ongoing worldwide to develop advanced nuclear power plants with high thermal efficiency for the purpose to improve their economical competitiveness. Within the 5th Framework Programme of the European Commission, a project has been launched with the main objective to assess the technical and economical feasibility of a high efficiency LWR operating at super critical pressure conditions. Several European research institutions, industrial partners and the University of Tokyo participated and worked in this common research project. Within the aims of the development of the HPLWR is to use both passive and active safety systems for performing safety related functions in the event of transients or accidents. Consequently substantial effort has been invested in order to define the safety features of the plant in a European environment, as well as to incorporate passive safety features into the design. Throughout this process, the European Utility Requirements (EUR) and requirements known from Generation IV initiative were considered as a guideline in general terms in order to include further advanced ideas. The HPLWR general features were compared to both requirements, indicating a potential to meet these. Since, the supercritical HPLWR represents a challenge for best-estimate safety codes like RELAP5, CATHARE and TRAB due to the fact that these codes were developed for two-phase or single-phase coolant at pressures far below critical point, work on the preliminary assessment of the appropriateness of these codes have been performed for selected relevant phenomena, and application of the codes to the selected transients on the basis of defined 'reference design'. An overview on their successful upgrade to supercritical pressures and application to some plant safety analysis are provided in the paper. Further elaborations in relation to future needs are also discussed. (author)

  16. Analysis of the survey results for the safety culture attitudes of HANARO management division staff

    Energy Technology Data Exchange (ETDEWEB)

    Lee, M.; Hwang, S. Y.; Lim, I. C.; Kang, T. J.; Choi, H. Y

    2003-12-01

    This report provides the that analysis results of the surveys on the safety culture attitude and related topics for the HANARO Management Division staff. The survey form consisted of the questionnaire on the safety culture attitude, happiness and life style, drinking habit and traffic safety attitude. The key information drawn from the analysis are as follow; First, the level of safety culture attitude of the staff was confirmed and it was found that the attitude improved when the present analysis results were compared with them for the survey in 1998. Second, the degree of happiness that the staff feel is higher than the average Korean and they enjoy more leisure time than the average Korean. Third, the staff drink more alcohol than the average Korean. Thus, the drinking habit must be improved for the health of the staff. Fourth, the questionnaire of the happiness and the traffic safety attitude should be improved to verify the hypotheses on the relation between them and the safety culture attitude.

  17. Analysis of the survey results for the safety culture attitudes of HANARO management division staff

    International Nuclear Information System (INIS)

    Lee, M.; Hwang, S. Y.; Lim, I. C.; Kang, T. J.; Choi, H. Y.

    2003-12-01

    This report provides the that analysis results of the surveys on the safety culture attitude and related topics for the HANARO Management Division staff. The survey form consisted of the questionnaire on the safety culture attitude, happiness and life style, drinking habit and traffic safety attitude. The key information drawn from the analysis are as follow; First, the level of safety culture attitude of the staff was confirmed and it was found that the attitude improved when the present analysis results were compared with them for the survey in 1998. Second, the degree of happiness that the staff feel is higher than the average Korean and they enjoy more leisure time than the average Korean. Third, the staff drink more alcohol than the average Korean. Thus, the drinking habit must be improved for the health of the staff. Fourth, the questionnaire of the happiness and the traffic safety attitude should be improved to verify the hypotheses on the relation between them and the safety culture attitude

  18. ETR fuel element shipping container addendum to PR-T-79-011 (TR-466). Internal technical report

    International Nuclear Information System (INIS)

    Smith, M.C.

    1979-01-01

    In July, 1979, EG and G Idaho, Inc. was requested to evaluate the ETR Fuel Element Shipping Container for compliance with existing transport regulations, in order to ship GETR fuel elements from Vallecitos, California to the INEL. Technical report PR-T-79-011 (TR-466), ATR Fuel Element Shipping Container Safety Analysis, was used as a basis for this evaluation. The safety analysis contained in technical report PR-T-79-011 (TR-466) was performed utilizing the ATR, ETR, MTR, and SPERT shipping containers. The report determined the ETR Fuel Element Shipping Container does comply with the existing transport regulations for a Type A quantity, Fissile Class I shipping container. The ETR and GETR fuel elements are essentially identical in physical size, construction, and fissile material content, the analysis documented in this report has determined the shipment of GETR fuel elements in the ETR shipping container to be safe and pose no threat to the public health and safety

  19. Fault Tree Analysis for Safety/Security Verification in Aviation Software

    Directory of Open Access Journals (Sweden)

    Andrew J. Kornecki

    2013-01-01

    Full Text Available The Next Generation Air Traffic Management system (NextGen is a blueprint of the future National Airspace System. Supporting NextGen is a nation-wide Aviation Simulation Network (ASN, which allows integration of a variety of real-time simulations to facilitate development and validation of the NextGen software by simulating a wide range of operational scenarios. The ASN system is an environment, including both simulated and human-in-the-loop real-life components (pilots and air traffic controllers. Real Time Distributed Simulation (RTDS developed at Embry Riddle Aeronautical University, a suite of applications providing low and medium fidelity en-route simulation capabilities, is one of the simulations contributing to the ASN. To support the interconnectivity with the ASN, we designed and implemented a dedicated gateway acting as an intermediary, providing logic for two-way communication and transfer messages between RTDS and ASN and storage for the exchanged data. It has been necessary to develop and analyze safety/security requirements for the gateway software based on analysis of system assets, hazards, threats and attacks related to ultimate real-life future implementation. Due to the nature of the system, the focus was placed on communication security and the related safety of the impacted aircraft in the simulation scenario. To support development of safety/security requirements, a well-established fault tree analysis technique was used. This fault tree model-based analysis, supported by a commercial tool, was a foundation to propose mitigations assuring the gateway system safety and security. 

  20. System safety education focused on flight safety

    Science.gov (United States)

    Holt, E.

    1971-01-01

    The measures necessary for achieving higher levels of system safety are analyzed with an eye toward maintaining the combat capability of the Air Force. Several education courses were provided for personnel involved in safety management. Data include: (1) Flight Safety Officer Course, (2) Advanced Safety Program Management, (3) Fundamentals of System Safety, and (4) Quantitative Methods of Safety Analysis.