Subsurface Interim Measures/Interim Remedial Action Plan/Environmental Assessment and Decision Document, Operable Unit No. 2

International Nuclear Information System (INIS)

1992-01-01

The subject Interim Measures/Interim Remedial Action plan/Environmental Assessment (IM/IRAP/EA) addresses residual free-phase volatile organic compound (VOC) contamination suspected in the subsurface within an area identified as Operable Unit No. 2 (OU2). This IM/IRAP/EA also addresses radionuclide contamination beneath the 903 Pad at OU2. Although subsurface VOC and radionuclide contamination on represent a source of OU2 ground-water contamination, they pose no immediate threat to public health or the environment. This IM/IRAP/EA identifies and evaluates interim remedial actions for removal of residual free-phase VOC contamination from three different subsurface environments at OU2. The term ''residual'' refers to the non-aqueous phase contamination remaining in the soil matrix (by capillary force) subsequent to the passage of non-aqueous or free-phase liquid through the subsurface. In addition to the proposed actions, this IM/IRAP/EA presents an assessment of the No Action Alternative. This document also considers an interim remedial action for the removal of radionuclides from beneath the 903 Pad

Review of SKB's interim report of SR-Can: SKI's and SSI's evaluation of SKB's up-dated methodology for safety assessment

International Nuclear Information System (INIS)

Dverstorp, Bjoern; Moberg, Leif; Wiebert, Anders; Xu Shulan; Stroemberg, Bo; Kautsky, Fritz; Lilja, Christina; Simic, Eva; Sundstroem, Benny; Toverud, Oeivind

2005-07-01

This report presents the findings of a review of the Swedish Nuclear Fuel and Waste Management Co.'s (SKB) interim report of the safety assessment SR-Can (SKB TR 04-11), conducted by the Swedish Radiation Protection Authority (SSI) and the Swedish Nuclear Power Inspectorate (SKI). SKB's interim report describes and exemplifies the safety assessment methodology that SKB plans to use in the oncoming licence applications for an encapsulation plant and a final repository for spent nuclear fuel. The authorities' review takes into account the findings of an international peer review of SKB's interim report. The authorities conclude that SKB has improved its safety assessment methodology in several aspects compared to earlier safety reports. Among other things the authorities commend SKB for giving a comprehensive account of relevant regulations and guidance, and for the systematic approach to identification and documentation of features, events and processes that need to be considered in the safety assessment. However, the authorities also conclude that important parts of SKB's method need to be further developed before they are mature enough to be used as a basis for a license application. The authorities' overall assessment is summarised in chapter 8 of this report

Fuel Supply Shutdown Facility Interim Operational Safety Requirements

International Nuclear Information System (INIS)

BENECKE, M.W.

2000-01-01

The Interim Operational Safety Requirements for the Fuel Supply Shutdown (FSS) Facility define acceptable conditions, safe boundaries, bases thereof, and management of administrative controls to ensure safe operation of the facility

K basins interim remedial action health and safety plan

Energy Technology Data Exchange (ETDEWEB)

DAY, P.T.

1999-09-14

The K Basins Interim Remedial Action Health and Safety Plan addresses the requirements of the Comprehensive Environmental Response, Compensation and Liability Act (CERCLA), as they apply to the CERCLA work that will take place at the K East and K West Basins. The provisions of this plan become effective on the date the US Environmental Protection Agency issues the Record of Decision for the K Basins Interim Remedial Action, currently planned in late August 1999.

An Assessment of SKB's Performance Assessment Calculations in the Interim Main Report for the Safety Assessment SR-Can

International Nuclear Information System (INIS)

Maul, Philip; Robinson, Peter

2005-03-01

SKB have published their Interim Main Report of the safety assessment SR-Can, which is intended to establish the framework for what will be submitted in 2006 in support of a licence application for construction of the spent fuel encapsulation plant. This follows on from the SR-Can Planning Document published in 2003. The purpose of the Interim Report is stated to be to demonstrate the methodology that will be used for safety assessment. The present report evaluates the information provided in the Interim SR-Can Report that is relevant to the Performance Assessment (PA) calculations that SKB intend to undertake, using independent calculations to facilitate this process. SKB consider that the primary safety function is to isolate completely the fuel within the canisters over the entire assessment period. Should a canister be damaged, the secondary safety function is to ensure that any release is retarded and dispersed sufficiently to ensure that concentrations levels in the accessible environment cannot cause unacceptable consequences. In this report PA calculations are considered to include both a high-level representation of the evolution of the system (relevant to the primary safety function), and any subsequent radionuclide transport (relevant to the secondary safety function). The main conclusions drawn are: 1. The effects of climate evolution on engineered barriers have not been analysed in detail in the Interim Report, and this limits the usefulness of the preliminary calculations that have been undertaken. 2. A key aspect of SKB's approach is the use of an integrated near-field evolution model. The information provided on this model demonstrates its capability efficiently to reproduce calculations from individual process models, but insufficient information is given at the present time to justify statements about interactions between processes. In particular it is assumed that relatively short term thermal and resaturation processes do not affect the

Safety research activities for Japanese regulations of spent fuel interim storage facilities

Energy Technology Data Exchange (ETDEWEB)

NONE

2013-08-15

Japan Nuclear Energy Safety Organization (JNES) carries out (a) preparation of technical documents, (b) technical evaluations of standards (prepared by academic societies), etc. and (c) other R and D activities, to support Nuclear Regulation Authority (NRA: which controls the regulations for Spent Fuel Interim Storage Facilities). In 2012 fiscal year, JNES carried out dynamic test of spent fuel to examine the integrity of spent fuel under cask drop accidents, and preparation for PWR spent fuel storage test to prove long term integrity of spent fuel and cask itself. Some of these tests will be also carried out in 2013 fiscal year and after. (author)

Fuel supply shutdown facility interim operational safety requirements

International Nuclear Information System (INIS)

Besser, R.L.; Brehm, J.R.; Benecke, M.W.; Remaize, J.A.

1995-01-01

These Interim Operational Safety Requirements (IOSR) for the Fuel Supply Shutdown (FSS) facility define acceptable conditions, safe boundaries, bases thereof, and management or administrative controls to ensure safe operation. The IOSRs apply to the fuel material storage buildings in various modes (operation, storage, surveillance)

Generic safety documentation model

International Nuclear Information System (INIS)

Mahn, J.A.

1994-04-01

This document is intended to be a resource for preparers of safety documentation for Sandia National Laboratories, New Mexico facilities. It provides standardized discussions of some topics that are generic to most, if not all, Sandia/NM facilities safety documents. The material provides a ''core'' upon which to develop facility-specific safety documentation. The use of the information in this document will reduce the cost of safety document preparation and improve consistency of information

Scientific criteria document for the development of an interim provincial water quality objective for aniline

Energy Technology Data Exchange (ETDEWEB)

Angelow, R.V.; Bazinet, N.

1996-11-01

The purpose of this document is to develop an interim provincial water quality objective for aniline for the protection of aquatic life in Ontario. It reviews the sources of aniline in the environment, its environmental fate and properties, acute and chronic toxicity as determined from results reported in the literature on toxicity tests using vertebrates and invertebrates, the bioaccumulation of aniline in the environment, mutagenic effects, and threshold aniline concentrations affecting fish odour and taste. The document then explains the derivation of the interim water quality objective. Water quality criteria for aniline developed in other jurisdictions are noted.

Safety of Long-term Interim Storage Facilities - Workshop Proceedings

International Nuclear Information System (INIS)

2014-01-01

The objective of this workshop was to discuss and review current national activities, plans and regulatory approaches for the safety of long term interim storage facilities dedicated to spent nuclear fuel (SF), high level waste (HLW) and other radioactive materials with prolonged storage regimes. It was also intended to discuss results of experiments and to identify necessary R and D to confirm safety of fuel and cask during the long-term storage. Safety authorities and their Technical Support Organisation (TSO), Fuel Cycle Facilities (FCF) operating organisations and international organisations were invited to share information on their approaches, practices and current developments. The workshop was organised in an opening session, three technical sessions, and a conclusion session. The technical sessions were focused on: - National approaches for long term interim storage facilities; - Safety requirements, regulatory framework and implementation issues; - Technical issues and operational experience, needs for R and D. Each session consisted of a number of presentations followed by a panel discussion moderated by the session Chairs. A summary of each session and subsequent discussion that ensued are provided as well as a summary of the results of the workshop with the text of the papers given and presentations made

Waste Encapsulation and Storage Facility interim operational safety requirements

CERN Document Server

Covey, L I

2000-01-01

The Interim Operational Safety Requirements (IOSRs) for the Waste Encapsulation and Storage Facility (WESF) define acceptable conditions, safe boundaries, bases thereof, and management or administrative controls required to ensure safe operation during receipt and inspection of cesium and strontium capsules from private irradiators; decontamination of the capsules and equipment; surveillance of the stored capsules; and maintenance activities. Controls required for public safety, significant defense-in-depth, significant worker safety, and for maintaining radiological consequences below risk evaluation guidelines (EGs) are included.

Safety evaluation for the interim stabilization of Tank 241-C-103

Energy Technology Data Exchange (ETDEWEB)

Geschke, G.R.

1995-03-01

This document provides the basis for interim stabilization of tank 241-C-103. The document covers the removal of the organic liquid layer and the aqueous supernatant from tank 241-C-103. Hazards are identified, consequences are calculated and controls to mitigate or prevent potential accidents are developed.

Safety evaluation for the interim stabilization of Tank 241-C-103

International Nuclear Information System (INIS)

Geschke, G.R.

1995-03-01

This document provides the basis for interim stabilization of tank 241-C-103. The document covers the removal of the organic liquid layer and the aqueous supernatant from tank 241-C-103. Hazards are identified, consequences are calculated and controls to mitigate or prevent potential accidents are developed

Safety evaluation of interim stabilization of non-stabilized single-shell watch list tanks

Energy Technology Data Exchange (ETDEWEB)

Stahl, S.M.

1994-12-30

The report provides a summation of the status of safety issues associated with interim stabilization of Watch List SSTs (organic, ferrocyanide, and flammable gas), as extracted from recent safety analyses, including the Tank Farms Accelerated Safety Analysis efforts.

Safety evaluation of interim stabilization of non-stabilized single-shell watch list tanks

International Nuclear Information System (INIS)

Stahl, S.M.

1994-01-01

The report provides a summation of the status of safety issues associated with interim stabilization of Watch List SSTs (organic, ferrocyanide, and flammable gas), as extracted from recent safety analyses, including the Tank Farms Accelerated Safety Analysis efforts

Interim main report of the safety assessment SR-Can

Energy Technology Data Exchange (ETDEWEB)

Hedin, Allan [and others

2004-08-01

This document is an interim report on the safety assessment SR-Can (SR in the acronym stands for Safety Report and Can is short for canister). The final SR-Can report will support SKB's application to build an Encapsulation plant for spent nuclear fuel and is to be produced in 2006. The purpose of the present interim report is to demonstrate the methodology for safety assessment so that it can be reviewed before it is used in a license application. The assessment relates to the KBS-3 disposal concept in which copper canisters with a cast iron insert containing spent nuclear fuel are surrounded by bentonite clay and deposited at approximately 500 m depth in saturated, granitic rock. Preliminary data from the Forsmark site, presently being investigated by SKB as one of the candidate for a KBS-3 repository are used to some extent as examples. However, the collected data are yet too sparse to allow an evaluation of safety for this site. An important aim of this report is to demonstrate the proper handling of requirements on the safety assessment in applicable regulations. Therefore, regulations issued by the Swedish Nuclear Power Inspectorate and the Swedish Radiation Protection Authority are duplicated in an Appendix. The principal acceptance criterion requires that 'the annual risk of harmful effects after closure does not exceed 10{sup -6} for a representative individual in the group exposed to the greatest risk'. 'Harmful effects' refer to cancer and hereditary effects. Following the introductory chapter 1, this report outlines the methodology for the SR-Can assessment in chapter 2, and presents in chapters 3, 4 and 5 the initial state of the system and the plans and methods for handling external influences and internal processes, respectively. Function indicators are introduced in chapter 6 and a preliminary evaluation of these is given in chapter 7. The material presented in the first seven chapters is utilised in the scenario selection in chapter 8

Interim main report of the safety assessment SR-Can

International Nuclear Information System (INIS)

Hedin, Allan

2004-08-01

This document is an interim report on the safety assessment SR-Can (SR in the acronym stands for Safety Report and Can is short for canister). The final SR-Can report will support SKB's application to build an Encapsulation plant for spent nuclear fuel and is to be produced in 2006. The purpose of the present interim report is to demonstrate the methodology for safety assessment so that it can be reviewed before it is used in a license application. The assessment relates to the KBS-3 disposal concept in which copper canisters with a cast iron insert containing spent nuclear fuel are surrounded by bentonite clay and deposited at approximately 500 m depth in saturated, granitic rock. Preliminary data from the Forsmark site, presently being investigated by SKB as one of the candidate for a KBS-3 repository are used to some extent as examples. However, the collected data are yet too sparse to allow an evaluation of safety for this site. An important aim of this report is to demonstrate the proper handling of requirements on the safety assessment in applicable regulations. Therefore, regulations issued by the Swedish Nuclear Power Inspectorate and the Swedish Radiation Protection Authority are duplicated in an Appendix. The principal acceptance criterion requires that 'the annual risk of harmful effects after closure does not exceed 10 -6 for a representative individual in the group exposed to the greatest risk'. 'Harmful effects' refer to cancer and hereditary effects. Following the introductory chapter 1, this report outlines the methodology for the SR-Can assessment in chapter 2, and presents in chapters 3, 4 and 5 the initial state of the system and the plans and methods for handling external influences and internal processes, respectively. Function indicators are introduced in chapter 6 and a preliminary evaluation of these is given in chapter 7. The material presented in the first seven chapters is utilised in the scenario selection in chapter 8. Hydrogeological

Toward introduction of risk informed safety regulation. Nuclear Safety Commission taskforce's interim report

International Nuclear Information System (INIS)

2006-01-01

Nuclear Safety Commission's taskforce on 'Introduction of Safety Regulation Utilizing Risk Information' completed the interim report on its future subjects and directions in December 2005. Although current safety regulatory activities have been based on deterministic approach, this report shows the risk informed approach is expected to be very useful for making nuclear safety regulation and assurance activities reasonable and also for appropriate allocation of regulatory resources. For introduction of risk informed regulation, it also recommends pileups of experiences with gradual introduction and trial of the risk informed approach, improvement of plant maintenance rules and regulatory requirements utilizing risk information, and establishment of framework to assure quality of risk evaluation. (T. Tanaka)

Central waste complex interim safety basis

International Nuclear Information System (INIS)

Cain, F.G.

1995-01-01

This interim safety basis provides the necessary information to conclude that hazards at the Central Waste Complex are controlled and that current and planned activities at the CWC can be conducted safely. CWC is a multi-facility complex within the Solid Waste Management Complex that receives and stores most of the solid wastes generated and received at the Hanford Site. The solid wastes that will be handled at CWC include both currently stored and newly generated low-level waste, low-level mixed waste, contact-handled transuranic, and contact-handled TRU mixed waste

Interim main report of the safety assessment SR-Can

Energy Technology Data Exchange (ETDEWEB)

Hedin, Allan (ed.) [and others

2004-08-01

This document is an interim report on the safety assessment SR-Can (SR in the acronym stands for Safety Report and Can is short for canister). The final SR-Can report will support SKB's application to build an Encapsulation plant for spent nuclear fuel and is to be produced in 2006. The purpose of the present interim report is to demonstrate the methodology for safety assessment so that it can be reviewed before it is used in a license application. The assessment relates to the KBS-3 disposal concept in which copper canisters with a cast iron insert containing spent nuclear fuel are surrounded by bentonite clay and deposited at approximately 500 m depth in saturated, granitic rock. Preliminary data from the Forsmark site, presently being investigated by SKB as one of the candidate for a KBS-3 repository are used to some extent as examples. However, the collected data are yet too sparse to allow an evaluation of safety for this site. An important aim of this report is to demonstrate the proper handling of requirements on the safety assessment in applicable regulations. Therefore, regulations issued by the Swedish Nuclear Power Inspectorate and the Swedish Radiation Protection Authority are duplicated in an Appendix. The principal acceptance criterion requires that 'the annual risk of harmful effects after closure does not exceed 10{sup -6} for a representative individual in the group exposed to the greatest risk'. 'Harmful effects' refer to cancer and hereditary effects. Following the introductory chapter 1, this report outlines the methodology for the SR-Can assessment in chapter 2, and presents in chapters 3, 4 and 5 the initial state of the system and the plans and methods for handling external influences and internal processes, respectively. Function indicators are introduced in chapter 6 and a preliminary evaluation of these is given in chapter 7. The material presented in the first seven chapters is utilised in the scenario selection

33 CFR 96.360 - Interim Safety Management Certificate: what is it and when can it be used?

Science.gov (United States)

2010-07-01

...? § 96.360 Interim Safety Management Certificate: what is it and when can it be used? (a) A responsible... 33 Navigation and Navigable Waters 1 2010-07-01 2010-07-01 false Interim Safety Management Certificate: what is it and when can it be used? 96.360 Section 96.360 Navigation and Navigable Waters COAST...

Ferrocyanide Safety Program rationale for removing six tanks from the safety watch list

International Nuclear Information System (INIS)

Borsheim, G.L.

1993-09-01

This report documents an in-depth study of single-shell tanks containing ferrocyanide wastes. Topics include: safety assessments, tank histories, supportive documentation about interim stabilization and planned remedial activities

Subsurface Interim Measures/Interim Remedial Action Plan and Decision Document for the 903 Pad, Mound, and East Trenches Areas (Operable Unit No. 2)

International Nuclear Information System (INIS)

1992-01-01

The Department of Energy (DOE) is pursuing an Interim Measure/Interim Remedial Action (IM/IRA) at the 903 Pad, Mound, and East Trenches Areas (Operable Unit No. 2) at the Rocky Flats Plant (RFP). This MIRA is to be conducted to provide information that will aid in the selection and design of final remedial actions at OU2 that will address removal of suspected free-phase volatile organic compound (VOC) contamination. The Plan involves investigating the removal of residual free-phase VOCs by in situ vacuum-enhanced vapor extraction technology at 3 suspected VOC source areas within OU2. VOC-contaminated vapors extracted from the subsurface would be treated by granular activated carbon (GAC) adsorption and discharged. The Plan also includes water table depression, when applicable at the test sites, to investigate the performance of vapor extraction technology in the saturated zone. The Plan provides for treatment of any contaminated ground water recovered during the IM/IRA at existing RFP treatment facilities. The proposed MVIRA Plan is presented in the document entitled ''Proposed Subsurface Interim Measures/Interim Remedial Action Plan/Environmental Assessment and Decision Document, 903 Pad, Mound, and East Trenches Areas, Operable Unit No. 2, '' dated 20 March 1992. Information concerning the proposed Subsurface IM/IRA was presented during a DOE Quarterly Review meeting held on 07 April 1992 and a public meeting held on 07 May 1992, at the Marriott Hotel in Golden, Colorado. The Responsiveness Summary presents DOE's response to all comments received at the public meeting, as well as those mailed to date to DOE during the public comment period

Review of SKB's interim report of SR-Can: SKI's and SSI's evaluation of SKB's up-dated methodology for safety assessment

Energy Technology Data Exchange (ETDEWEB)

Dverstorp, Bjoern; Moberg, Leif; Wiebert, Anders; Xu Shulan [Swedish Radiation Protection Authority, Stockholm (Sweden); Stroemberg, Bo; Kautsky, Fritz; Lilja, Christina; Simic, Eva; Sundstroem, Benny; Toverud, Oeivind [Swedish Nuclear Power Inspectorate, Stockholm (Sweden)

2005-07-01

This report presents the findings of a review of the Swedish Nuclear Fuel and Waste Management Co.'s (SKB) interim report of the safety assessment SR-Can (SKB TR 04-11), conducted by the Swedish Radiation Protection Authority (SSI) and the Swedish Nuclear Power Inspectorate (SKI). SKB's interim report describes and exemplifies the safety assessment methodology that SKB plans to use in the oncoming licence applications for an encapsulation plant and a final repository for spent nuclear fuel. The authorities' review takes into account the findings of an international peer review of SKB's interim report. The authorities conclude that SKB has improved its safety assessment methodology in several aspects compared to earlier safety reports. Among other things the authorities commend SKB for giving a comprehensive account of relevant regulations and guidance, and for the systematic approach to identification and documentation of features, events and processes that need to be considered in the safety assessment. However, the authorities also conclude that important parts of SKB's method need to be further developed before they are mature enough to be used as a basis for a license application. The authorities' overall assessment is summarised in chapter 8 of this report.

Design requirements document for Project W-465, immobilized low-activity waste interim storage

International Nuclear Information System (INIS)

Burbank, D.A.

1998-01-01

The scope of this Design Requirements Document (DRD) is to identify the functions and associated requirements that must be performed to accept, transport, handle, and store immobilized low-activity waste (ILAW) produced by the privatized Tank Waste Remediation System (TWRS) treatment contractors. The functional and performance requirements in this document provide the basis for the conceptual design of the TWRS ILAW Interim Storage facility project and provides traceability from the program level requirements to the project design activity. Technical and programmatic risk associated with the TWRS planning basis are discussed in the Tank Waste Remediation System Decisions and Risk Assessment (Johnson 1994). The design requirements provided in this document will be augmented by additional detailed design data documented by the project

Safety evaluation of interim stabilization of non-stabilized single-shell watch list tanks

International Nuclear Information System (INIS)

Stahl, S.M.

1994-01-01

This report provides results of a review of recently completed safety analyses related to hazards associated with Interim Stabilization of Single analyses related to hazards included oh the Hanford Site Waste Tank-Watch Shell Tanks (SSTs) that are included on the Hanford List. The purpose of the review was to identify and summarize conclusions regarding the safety of interim stabilization of Watch List SSTs, and to highlight applicable limitations, restrictions, and controls. The scope of this review was restricted to SSTs identified List in the categories of flammable gas ferrocyanide, and organic salts. High heat tanks were not included in the scope. A Watch List tank is defined as an underground storage tank containing waste that requires special safety precautions because it may have a serious potential for release of high level radioactive waste because of uncontrolled increases in temperature or pressure. Special restrictions have been placed on these tanks

Interim data report for the safety assessment SR-Can

International Nuclear Information System (INIS)

Vahlund, Fredrik; Andersson, Johan

2004-08-01

This document is the interim data report in the project SR-Can. The purpose of the data report is to present input data, with uncertainty estimates, for the SR-Can assessment calculations. Besides input data, the report also describes the standardised procedures used when deriving the input data and the corresponding uncertainty estimates. However, in the present interim version of the report (written in the initial stage of the project when site characterisation has yet not been completed) the standardised procedures have not been possible to apply for most of the data and, in order to present a compilation of the data used in the assessment, much of the input data is presented without following the standardised procedures. This will however be changed for the final version of the SR-Can data report, in order to show the methodology that will be used in the final version one example of how input data will be presented is included (migration data for buffer) . The recommended input data for the assessment calculations are, for the interim version, mainly based on SR 97 Beberg data, these are merely presented without any background or uncertainty discussion (this is presented in the SR 97 data report)

Interim safety evaluation report related to operation of Enrico Fermi Atomic Power Plant, Unit 2, Detroit Edison Company

International Nuclear Information System (INIS)

1977-09-01

This interim report summarizes the scope and results of the radiological safety review performed to date by the NRC staff with respect to the operating license phase for the Enrico Fermi Atomic Power Plant, Unit 2. The major effort was the review of the facility design and proposed operating procedures described in applicant's Final Safety Analysis Report. In the course of the review, several meetings were held with representatives of the applicant to discuss plant design, construction and proposed operation. Additional information was requested, which the applicant provided through Amendment 7 to the Final Safety Analysis Report. A chronology of the principal actions relating to the review of the application is attached as Appendix A to the report. The Final Safety Analysis Report and amendments thereto are available for public inspection at the Nuclear Regulatory Commission Public Document Room, 1717 H Street, N. W., Washington, D.C. and at Monroe County Library System, 3700 South Custer Road, Monroe, Michigan 48161

Surface Water Interim Measures/Interim Remedial Action Plan/Environmental Assessment and Decision Document for South Walnut Creek Basin (Operable Unit No. 2)

International Nuclear Information System (INIS)

1991-01-01

The Department of Energy (DOE) is pursuing an Interim Measure/Interim Remedial Action (IM/IRA) at the 903 Pad, Mound, and East Trenches Areas (Operable Unit No. 2) at the Rocky Flats Plant (RFP). This IM/IRA is to be conducted to minimize the release from these areas of hazardous substances that pose a potential threat to the public health and environment. The Plan involved the collection of contaminated surface water at specific locations, treatment by chemical precipitation, cross-flow membrane filtration and granular activated carbon (GAC) adsorption, and surface discharge of treated water. Information for the initial configuration of the Plan is presented in the document entitled ''Proposed Interim Measures/Interim Remedial Action Plan and Decision Document, 903 Pad, Mound, and East Trenches Areas, Operable Unit No. 2'' (IM/IRAP) dated 26 September 1990. Information concerning the proposed Surface Water IM/IRA was presented during a public meeting held from 7 to 10 p.m., Tuesday, 23 October 1990, at the Westminster City Park Recreation Center in Westminster, Colorado. This Responsiveness Summary presents DOE's response to all comments received at the public meeting, as well as those mailed to DOE during the public comment period which ended 24 November 1990. There were a number of technical comments on the plan that DOE has addressed herein. It is noted that several major issues were raised by the comments. Regardless of the estimated low risk to the public from construction and water transport activities, the popular sentiment of the public, based on comments received, is strong concern over worker and public health risks from these activities. In the light of public and municipal concerns, DOE proposes to eliminate from this IM/IRA the interbasin transfer of Woman Creek seepage to the South Walnut Creek drainage and to address collection and treatment of contaminated South Walnut Creek and Woman Creek surface water under two separate IM/IRAs

Safety regulations for radioisotopes, etc. (interim report)

International Nuclear Information System (INIS)

1980-01-01

An (interim) report by an ad hoc expert committee to the Nuclear Safety Commission, on the safety regulations for radioisotopes, etc., was presented. For the utilization of radioisotopes, etc., there is the Law Concerning Prevention of Radiation Injury Due to Radioisotopes, etc. with the advances in this field and the improvement in international standards, the regulations by the law have been examined. After explaining the basic ideas of the regulations, the problems and countermeasures in the current regulations are described: legal system, rationalization in permission procedures and others, inspection on RI management, the system of the persons in charge of radiation handling, RI transport, low-level radioactive wastes, consumer goods, definitions of RIs, radiation and sealed sources, regulations by group partitioning, RI facilities, system of personnel exposure registration, entrusting of inspection, etc. to private firms, and reduction in the works for permission among governmental offices. (author)

Safety Consideration for a Wet Interim Spent Fuel Store at Conceptual Design Stage

International Nuclear Information System (INIS)

Astoux, Marion

2014-01-01

EDF Energy plans to build and operate two UK EPRs at the Hinkley Point C (HPC) site in Somerset, England. Spent fuel from the UK EPRs will need to be managed from the time it is discharged from the reactor until it is ultimately disposed of and this will involve storing the spent fuel for a period in the fuel building and thereafter in a dedicated interim facility until it can be emplaced within the UK Geological Disposal Facility. EDF Energy has proposed that this interim store should be located on the Hinkley Point site which is consistent with UK policy. This Interim Spent Fuel Store (ISFS) will have the capability to store for at least one hundred years the spent fuel arising from the operation of the two EPR units (sixty years operation). Therefore, specificities regarding the lifetime of the facility have to be accounted for its design. The choice of interim storage technology was considered in some depth for the HPC project and wet storage (pool) was selected. The facility is currently at conceptual design stage, although its construction will be part of main site construction phase. Safety functions and safety requirements for this storage facility have been defined, in compliance with WENRA 'Waste and Spent Fuel Storage - Safety Reference Level Report' and IAEA Specific Safety Guide no. 15 'Storage of Spent Nuclear Fuel'. EDF technical know-how, operational feedback on existing storage pools, UK regulatory context and Fukushima experience feedback have also been accounted for. Achievement of the safety functions as passively as reasonably practicable is a key issue for the design, especially in accident situations. Regarding lifetime aspects, ageing management of equipments, optimisation of the refurbishment, climate change, passivity of the facility, and long-term achievement of the safety functions are among the subjects to consider. Adequate Operational Limits and Conditions will also have to be defined, to enable the long-term achievement of the safety

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

International Nuclear Information System (INIS)

SCHWEHR, B.A.

1999-01-01

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

Nuclear criticality safety evaluation of the passage of decontaminated salt solution from the ITP filters into tank 50H for interim storage

International Nuclear Information System (INIS)

Hobbs, D.T.; Davis, J.R.

1994-01-01

This report assesses the nuclear criticality safety associated with the decontaminated salt solution after passing through the In-Tank Precipitation (ITP) filters, through the stripper columns and into Tank 50H for interim storage until transfer to the Saltstone facility. The criticality safety basis for the ITP process is documented. Criticality safety in the ITP filtrate has been analyzed under normal and process upset conditions. This report evaluates the potential for criticality due to the precipitation or crystallization of fissionable material from solution and an ITP process filter failure in which insoluble material carryover from salt dissolution is present. It is concluded that no single inadvertent error will cause criticality and that the process will remain subcritical under normal and credible abnormal conditions

Hanford Generic Interim Safety Basis

International Nuclear Information System (INIS)

Lavender, J.C.

1994-01-01

The purpose of this document is to identify WHC programs and requirements that are an integral part of the authorization basis for nuclear facilities that are generic to all WHC-managed facilities. The purpose of these programs is to implement the DOE Orders, as WHC becomes contractually obligated to implement them. The Hanford Generic ISB focuses on the institutional controls and safety requirements identified in DOE Order 5480.23, Nuclear Safety Analysis Reports

Hanford Generic Interim Safety Basis

Energy Technology Data Exchange (ETDEWEB)

Lavender, J.C.

1994-09-09

The purpose of this document is to identify WHC programs and requirements that are an integral part of the authorization basis for nuclear facilities that are generic to all WHC-managed facilities. The purpose of these programs is to implement the DOE Orders, as WHC becomes contractually obligated to implement them. The Hanford Generic ISB focuses on the institutional controls and safety requirements identified in DOE Order 5480.23, Nuclear Safety Analysis Reports.

A safety equipment list for rotary mode core sampling systems operation in single shell flammable gas tanks

International Nuclear Information System (INIS)

SMALLEY, J.L.

1999-01-01

This document identifies all interim safety equipment to be used for rotary mode core sampling of single-shell flammable gas tanks utilizing Rotary Mode Core Sampling systems (RMCS). This document provides the safety equipment for RMCS trucks HO-68K-4600, HO-68K-4647, trucks three and four respectively, and associated equipment. It is not intended to replace or supersede WHC-SD-WM-SEL-023, (Kelly 1991), or WHC-SD-WM-SEL-032, (Corbett 1994), which classifies 80-68K-4344 and HO-68K-4345 respectively. The term ''safety equipment'' refers to safety class (SC) and safety significant (SS) equipment, where equipment refers to structures, systems and components (SSC's). The identification of safety equipment in this document is based on the credited design safety features and analysis contained in the Authorization Basis (AB) for rotary mode core sampling operations in single-shell flammable gas tanks. This is an interim safety classification since the AB is interim. This document will be updated to reflect the final RMCS equipment safety classification designations upon completion of a final AB which will be implemented with the release of the Final Safety Analysis Report (FSAR)

A safety equipment list for rotary mode core sampling systems operation in single shell flammable gas tanks; TOPICAL

International Nuclear Information System (INIS)

SMALLEY, J.L.

1999-01-01

This document identifies all interim safety equipment to be used for rotary mode core sampling of single-shell flammable gas tanks utilizing Rotary Mode Core Sampling systems (RMCS). This document provides the safety equipment for RMCS trucks HO-68K-4600, HO-68K-4647, trucks three and four respectively, and associated equipment. It is not intended to replace or supersede WHC-SD-WM-SEL-023, (Kelly 1991), or WHC-SD-WM-SEL-032, (Corbett 1994), which classifies 80-68K-4344 and HO-68K-4345 respectively. The term ''safety equipment'' refers to safety class (SC) and safety significant (SS) equipment, where equipment refers to structures, systems and components (SSC's). The identification of safety equipment in this document is based on the credited design safety features and analysis contained in the Authorization Basis (AB) for rotary mode core sampling operations in single-shell flammable gas tanks. This is an interim safety classification since the AB is interim. This document will be updated to reflect the final RMCS equipment safety classification designations upon completion of a final AB which will be implemented with the release of the Final Safety Analysis Report (FSAR)

Final Hazard Classification and Auditable Safety Analysis for the 105-F Building Interim Safe Storage Project

International Nuclear Information System (INIS)

Rodovsky, T.J.; Bond, S.L.

1998-07-01

The auditable safety analysis (ASA) documents the authorization basis for the partial decommissioning and facility modifications to place the 105-F Building into interim safe storage (ISS). Placement into the ISS is consistent with the preferred alternative identified in the Record of Decision (58 FR). Modifications will reduce the potential for release and worker exposure to hazardous and radioactive materials, as well as lower surveillance and maintenance (S ampersand M) costs. This analysis includes the following: A description of the activities to be performed in the course of the 105-F Building ISS Project. An assessment of the inventory of radioactive and other hazardous materials within the 105-F Building. Identification of the hazards associated with the activities of the 105-F Building ISS Project. Identification of internally and externally initiated accident scenarios with the potential to produce significant local or offsite consequences during the 105-F Building ISS Project. Bounding evaluation of the consequences of the potentially significant accident scenarios. Hazard classification based on the bounding consequence evaluation. Associated safety function and controls, including commitments. Radiological and other employee safety and health considerations

Simplifying documentation while approaching site closure: integrated health and safety plans as documented safety analysis

International Nuclear Information System (INIS)

Brown, Tulanda

2003-01-01

At the Fernald Closure Project (FCP) near Cincinnati, Ohio, environmental restoration activities are supported by Documented Safety Analyses (DSAs) that combine the required project-specific Health and Safety Plans, Safety Basis Requirements (SBRs), and Process Requirements (PRs) into single Integrated Health and Safety Plans (I-HASPs). By isolating any remediation activities that deal with Enriched Restricted Materials, the SBRs and PRs assure that the hazard categories of former nuclear facilities undergoing remediation remain less than Nuclear. These integrated DSAs employ Integrated Safety Management methodology in support of simplified restoration and remediation activities that, so far, have resulted in the decontamination and demolition (D and D) of over 150 structures, including six major nuclear production plants. This paper presents the FCP method for maintaining safety basis documentation, using the D and D I-HASP as an example

Radioactive waste interim storage in Germany

International Nuclear Information System (INIS)

2015-12-01

The short summary on the radioactive waste interim storage in Germany covers the following issues: importance of interim storage in the frame of radioactive waste management, responsibilities and regulations, waste forms, storage containers, transport of vitrified high-level radioactive wastes from the reprocessing plants, central interim storage facilities (Gorleben, Ahaus, Nord/Lubmin), local interim storage facilities at nuclear power plant sites, federal state collecting facilities, safety, radiation exposure in Germany.

Periodic Safety Review in Interim Storage Facilities - Current Regulation and Experiences in Germany

International Nuclear Information System (INIS)

Neles, Julia Mareike; Schmidt, Gerhard

2014-01-01

Periodic safety reviews in nuclear power plants in Germany have been performed since the end of the 1980's as an indirect follow-up of the accident in Chernobyl and, in the meantime, are formally required by law. During this process the guidelines governing this review were developed in stages and reached their final form in 1996. Interim storage facilities and other nuclear facilities at that time were not included, so the guidelines were solely focused on the specific safety issues of nuclear power plants. Following IAEA's recommendations, the Western European Nuclear Regulator Association (WENRA) introduced PSRs in its safety reference levels for storage facilities (current version in WGWD report 2.1 as of Feb 2011: SRLs 59 - 61). Based on these formulations, Germany improved its regulation in 2010 with a recommendation of the Nuclear Waste Management Commission (Entsorgungskommission, ESK), an expert advisory commission for the federal regulatory body BMU. The ESK formulated these detailed requirements in the 'ESK recommendation for guides to the performance of periodic safety reviews for interim storage facilities for irradiated fuel elements and heat-generating radioactive waste'. Before finalization of the guideline a test phase was introduced, aimed to test the new regulation in practice and to later include the lessons learned in the final formulation of the guideline. The two-year test phase started in October 2011 in which the performance of a PSR will be tested at two selected interim storage facilities. Currently these recommendations are discussed with interested/concerned institutions. The results of the test phase shall be considered for improvements of the draft and during the final preparation of guidelines. Currently the PSR for the first ISF is in an advanced stage, the second facility just started the process. Preliminary conclusions from the test phase show that the implementation of the draft guideline requires interpretation. The aim of a

International Peer Review of Swedish Nuclear Fuel and Waste Management Company's SR-Can interim report

International Nuclear Information System (INIS)

Sagar, Budhi; Bailey, Lucy; Bennett, David G.; Egan, Mike; Roehlig, Klaus

2004-12-01

SKB has produced an interim safety assessment report as part of its work to develop a licence application for the construction of a spent nuclear fuel encapsulation plant. The purpose of the interim report is to set out and demonstrate SKB's proposed methodology for long-term safety assessment. The aim of producing an interim report is to allow the Swedish regulatory authorities (SKI and SSI) to review and comment on SKB's proposed methodology before it is used in support of a formal licence application. To help inform their review of SKB's proposed methodology, the authorities appointed an international review team (IRT) to carry out a review of SKB's interim safety assessment report. Comments from the IRT are presented in this document and will be considered by the regulatory authorities in developing their own view of SKB's proposed methodology. The IRT's review included examination of SKB's documentation (the 'Interim Main Report of the Safety Assessment SR-Can' and four supporting documents) and hearings with SKB staff and contractors. The hearings provided an opportunity for the IRT to discuss the SR-Can safety assessment with the authors and contributors to SKB's work. As directed by SKI and SSI, the IRT's review focused on methodological aspects and sought to determine whether SKB's proposed safety assessment methodology: (i) is fit for the purpose of supporting a licence application; (ii) has a reasonable prospect of leading to a safety assessment that is sufficiently comprehensive, reproducible, traceable and transparent; (iii) is compatible with the authorities' regulations and guidance. No evaluation of long term safety or site acceptability was attempted by the IRT. At the request of SKI and SSI, the IRT's review considered and made recommendations on the following issues: Description of the initial state of the repository and its components; Description of features, events and processes (FEPs) relevant to repository evolution; Strategy for safety

Subsurface Interim Measures/Interim Remedial Action Plan/ Environmental Assessment and Decision Document, Operable Unit No. 2

International Nuclear Information System (INIS)

1992-01-01

The subject Interim Measures/Interim Remedial Action plan/Environmental Assessment (IM/IRAP/EA) addresses residual free-phase volatile organic compound (VOC) contamination suspected in the subsurface within an area identified as Operable Unit No. 2 (OU2). This IM/IRAP/EA also addresses radionuclide contamination beneath the 903 Pad at OU2. Although subsurface VOC and radionuclide contamination on represent a source of OU2 ground-water contamination, they pose no immediate threat to public health or the environment. This volume contains five appendices

Flammable gas deflagration consequence calculations for the tank waste remediation system basis for interim operation

Energy Technology Data Exchange (ETDEWEB)

Van Vleet, R.J., Westinghouse Hanford

1996-08-13

This paper calculates the radiological dose consequences and the toxic exposures for deflagration accidents at various Tank Waste Remediation System facilities. These will be used in support of the Tank Waste Remediation System Basis for Interim Operation.The attached SD documents the originator`s analysis only. It shall not be used as the final or sole document for effecting changes to an authorization basis or safety basis for a facility or activity.

Criticality safety evaluation for long term storage of FFTF fuel in interim storage casks

International Nuclear Information System (INIS)

Richard, R.F.

1995-01-01

It has been postulated that a degradation phenomenon, referred to as ''hot cell rot'', may affect irradiated FFTF mixed plutonium-uranium oxide (MOX) fuel during dry interim storage. ''Hot cell rot'' refers to a variety of phenomena that degrade fuel pin cladding during exposure to air and inert gas environments. It is thought to be a form of caustic stress corrosion cracking or environmentally assisted cracking. Here, a criticality safety analysis was performed to address the effect of the ''hot cell rot'' phenomenon on the long term storage of irradiated FFTF fuel in core component containers. The results show that seven FFTF fuel assemblies or six Ident-69 pin containers stored in core component containers within interim storage casks will remain safely subcritical

Safety of interim storage solutions of used nuclear fuel during extended term

Energy Technology Data Exchange (ETDEWEB)

Shelton, C.; Bader, S.; Issard, H.; Arslan, M. [AREVA, 7135 Minstrel Way, Suite 300 Columbia, MD 21045 (United States)

2013-07-01

In 2013, the total amount of stored used nuclear fuel (UNF) in the world will reach 225,000 T HM. The UNF inventory in wet storage will take up over 80% of the available total spent fuel pool (SFP) capacity. Interim storage solutions are needed. They give flexibility to the nuclear operators and ensure that nuclear reactors continue to operate. However, we need to keep in mind that they are also an easy way to differ final decision and implementation of a UNF management approach (recycling or final disposal). In term of public perception, they can have a negative impact overtime as it may appear that nuclear industry may have significant issues to resolve. In countries lacking an integrated UNF management approach, the UNF are being discharged from the SFPs to interim storage (mostly to dry storage) at the same rate as UNF is being discharged from reactors, as the SFPs at the reactor sites are becoming full. This is now the case in USA, Taiwan, Switzerland, Spain, South Africa and Germany. For interim storage, AREVA has developed different solutions in order to allow the continued operation of reactors while meeting the current requirements of Safety Authorities: -) Dry storage canisters on pads, -) Dual-purpose casks (dry storage and transportation), -) Vault dry storage, and -) Centralized pool storage.

Pre-conceptual study on the review framework for the radiation shielding safety of the PWR spent fuel cask interim storage in Korea

International Nuclear Information System (INIS)

Kim, Byeong-Soo; Jeong, Jae-Hak; Jeong, Chan-Woo

2006-01-01

In Korea, 20 nuclear power plants are in operation and lots of spent fuels are on the onsite storage. The onsite storage capacity in Korea is supposed to be full around at the year of 2016 and interim storage facilities could be considered to be constructed before 2016. A review framework to evaluate the radiation shielding safety of the interim storage facilities is developed in this study. It includes acceptance criteria, review procedures and activities of independent analyses. A case study is performed to apply the review framework. Modeling the review reference storage, evaluating the source terms and calculating the photon fluxes are performed. It is shown that the application of the review framework could satisfy the regulatory demand that would arise in the near future in the review area of the radiation shielding safety of the interim storage in Korea. (author)

Environment, Health, and Safety - Construction Subcontractors Documents |

Science.gov (United States)

NREL Environment, Health, and Safety - Construction Subcontractors Documents Environment Environment, Health and Safety (EH&S) requirements are understood by construction subcontractors and with these requirements before submitting proposals and/or environment, health and safety plans for the

PUREX Deactivation Health and Safety documentation

Energy Technology Data Exchange (ETDEWEB)

Dodd, E.N. III

1995-01-01

The purpose of the PUREX Deactivation Project is to establish a passively safe and environmentally secure configuration of PUREX at the Hanford Site, and to preserve that configuration for a 10-year horizon. The 10-year horizon is used to predict future maintenance requirements and represents they typical time duration expended to define, authorize, and initiate the follow-on Decontamination and Decommissioning (D&D) activities. This document was prepared to increase attention to worker safety issues during the deactivation project and, as such, identifies the documentation and programs associated with PUREX Deactivation Health and Safety.

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

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.

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

International Nuclear Information System (INIS)

1985-04-01

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

Interim Safety Basis for Fuel Supply Shutdown Facility

International Nuclear Information System (INIS)

BENECKE, M.W.

2000-01-01

This ISB, in conjunction with the IOSR, provides the required basis for interim operation or restrictions on interim operations and administrative controls for the facility until a SAR is prepared in accordance with the new requirements or the facility is shut down. It is concluded that the risks associated with tha current and anticipated mode of the facility, uranium disposition, clean up, and transition activities required for permanent closure, are within risk guidelines

PUREX Deactivation Health and Safety documentation

International Nuclear Information System (INIS)

Dodd, E.N. III.

1995-01-01

The purpose of the PUREX Deactivation Project is to establish a passively safe and environmentally secure configuration of PUREX at the Hanford Site, and to preserve that configuration for a 10-year horizon. The 10-year horizon is used to predict future maintenance requirements and represents they typical time duration expended to define, authorize, and initiate the follow-on Decontamination and Decommissioning (D ampersand D) activities. This document was prepared to increase attention to worker safety issues during the deactivation project and, as such, identifies the documentation and programs associated with PUREX Deactivation Health and Safety

Environmental restoration and decontamination and decommissioning safety documentation

International Nuclear Information System (INIS)

Hansen, J.L.; Frauenholz, L.H.; Kerr, N.R.

1993-01-01

This document presents recommendations of a working group designated by the Environmental Restoration and Remediation (ER) and Decontamination and Decommissioning (D ampersand D) subcommittees of the Westinghouse M ampersand O (Management and Operation) Nuclear Facility Safety Committee. A commonalty of approach to safety documentation specific to ER and D ampersand D activities was developed and is summarized below. Allowance for interpretative tolerance and documentation flexibility appropriate to the activity, graded for hazard category, duration, and complexity, was a primary consideration in development of this guidance

Hydrazine blending and storage facility, interim response action, draft implementation document for rinsewater transfer, phase 2

Energy Technology Data Exchange (ETDEWEB)

1991-08-09

This Draft Implementation Document (ID) for Rinsewater Transfer has been prepared as a requirement for conducting and completing the Interim Response Action (IRA) at the Hydrazine Blending and Storage Facility (HBSF) located at Rocky Mountain Arsenal (RMA) in Commerce City, Colorado. This document has been prepared in accordance with requirements set forth in the October 1988 Final Decision Document for the HBSF IRA (Peer, 1988) and the Amendment to the Final Decision Document (HLA, 1991). The HBSF IRA task was separated into two phases that comprise complete decommissioning of the HBSF as cited in the Federal Facility Agreement. The design portion of Phase I of the HBSF IRA included analytical methods development and laboratory certification for analysis of hydrazine fuel compounds (hydrazine, monomethyl hydrazine) (MMH), and unsymmetrical dimethyl hydrazine (UDMH) and n-nitrosodimethylamine (NDMA) in HBSF rinsewater, chemical characterization of hydrazine rinsewater, bench- and pilot-scale testing of ultraviolet (UV) light/chemical oxidation treatment systems for treatment of hydrazine rinsewater, full-scale startup testing of a UV light/chemical oxidation treatment system, and air monitoring during startup testing as described in the Draft Final Treatment Report (HLA, 1991).

Interim safety basis for fuel supply shutdown facility

International Nuclear Information System (INIS)

Brehm, J.R.; Deobald, T.L.; Benecke, M.W.; Remaize, J.A.

1995-01-01

This ISB in conjunction with the new TSRs, will provide the required basis for interim operation or restrictions on interim operations and administrative controls for the Facility until a SAR is prepared in accordance with the new requirements. It is concluded that the risk associated with the current operational mode of the Facility, uranium closure, clean up, and transition activities required for permanent closure, are within Risk Acceptance Guidelines. The Facility is classified as a Moderate Hazard Facility because of the potential for an unmitigated fire associated with the uranium storage buildings

Reference values on safety regulation of land disposal of low level radioactive solid waste (the second interim report) and its incorporation into legal regulations

International Nuclear Information System (INIS)

Aoki, Terumi

1994-01-01

Safety regulation of land disposal of low level radioactive solid waste in Japan is based on 'the basic philosophy on the safety regulation of land disposal of low level radioactive solid waste' determined by the Nuclear safety Committee (October 1985). The basic philosophy on the upper limit of radioactivity of disposed wastes was published as the reference values in the interim report (February 1987) and in the second interim report (June 1992). In the second interim report, the upper limits of radioactivity are established for three types of solid radioactive wastes: 1) metals, incombustible or flame resistant wastes generated nuclear reactor facilities and solidified in vessels, 2) large metallic structures generated from decommissioning of reactor facilities and difficult to solidify in vessels, and 3) radioactive concrete waste generated from decommissioning of reactor facilities. The upper limits of radioactivity are presented for C-14, Co-60, Ni-63, Sr-90, Cs-137, alfa-emmitters, Ca-41 (for concrete) and Eu-152 (for concrete). Related laws and regulations in Japan on safe disposal of low level wastes are explained. (T.H.)

Environmental Restoration Disposal Facility (Project W-296) Safety Assessment

International Nuclear Information System (INIS)

Armstrong, D.L.

1994-08-01

This Safety Assessment is based on information derived from the Conceptual Design Report for the Environmental Restoration Disposal Facility (DOE/RL 1994) and ancillary documentation developed during the conceptual design phase of Project W-296. The Safety Assessment has been prepared to support the Solid Waste Burial Ground Interim Safety Basis document. The purpose of the Safety Assessment is to provide an evaluation of the design to determine if the process, as proposed, will comply with US Department of Energy (DOE) Limits for radioactive and hazardous material exposures and be acceptable from an overall health and safety standpoint. The evaluation considered affects on the worker, onsite personnel, the public, and the environment

Environmental Restoration Disposal Facility (Project W-296) Safety Assessment

Energy Technology Data Exchange (ETDEWEB)

Armstrong, D.L.

1994-08-01

This Safety Assessment is based on information derived from the Conceptual Design Report for the Environmental Restoration Disposal Facility (DOE/RL 1994) and ancillary documentation developed during the conceptual design phase of Project W-296. The Safety Assessment has been prepared to support the Solid Waste Burial Ground Interim Safety Basis document. The purpose of the Safety Assessment is to provide an evaluation of the design to determine if the process, as proposed, will comply with US Department of Energy (DOE) Limits for radioactive and hazardous material exposures and be acceptable from an overall health and safety standpoint. The evaluation considered affects on the worker, onsite personnel, the public, and the environment.

Upgrading safety documentation for exported nuclear power plants

International Nuclear Information System (INIS)

Rosen, M.

1978-01-01

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

Annex D-200 Area Interim Storage Area Final Safety Analysis Report [FSAR] [Section 1 & 2

Energy Technology Data Exchange (ETDEWEB)

CARRELL, R D

2002-07-16

The 200 Area Interim Storage Area (200 Area ISA) at the Hanford Site provides for the interim storage of non-defense reactor spent nuclear fuel (SNF) housed in aboveground dry cask storage systems. The 200 Area ISA is a relatively simple facility consisting of a boundary fence with gates, perimeter lighting, and concrete and gravel pads on which to place the dry storage casks. The fence supports safeguards and security and establishes a radiation protection buffer zone. The 200 Area ISA is nominally 200,000 ft{sup 2} and is located west of the Canister Storage Building (CSB). Interim storage at the 200 Area ISA is intended for a period of up to 40 years until the materials are shipped off-site to a disposal facility. This Final Safety Analysis Report (FSAR) does not address removal from storage or shipment from the 200 Area ISA. Three different SNF types contained in three different dry cask storage systems are to be stored at the 200 Area ISA, as follows: (1) Fast Flux Test Facility Fuel--Fifty-three interim storage casks (ISC), each holding a core component container (CCC), will be used to store the Fast Flux Test Facility (FFTF) SNF currently in the 400 Area. (2) Neutron Radiography Facility (NRF) TRIGA'--One Rad-Vault' container will store two DOT-6M3 containers and six NRF TRIGA casks currently stored in the 400 Area. (3) Commercial Light Water Reactor Fuel--Six International Standards Organization (ISO) containers, each holding a NAC-I cask4 with an inner commercial light water reactor (LWR) canister, will be used for commercial LWR SNF from the 300 Area. An aboveground dry cask storage location is necessary for the spent fuel because the current storage facilities are being shut down and deactivated. The spent fuel is being transferred to interim storage because there is no permanent repository storage currently available.

Interim summary report of the safety case 2009

International Nuclear Information System (INIS)

2010-03-01

Following the guidelines set forth by the Ministry of Trade and Industry (now Ministry of Employment and Economy), Posiva is preparing to submit a construction license application for the final disposal spent nuclear fuel at the Olkiluoto site, Finland, by the end of the year 2012. Disposal will take place in a geological repository implemented according to the KBS-3 method. The long-term safety section supporting the license application will be based on a safety case that, according to the internationally adopted definition, will be a compilation of the evidence, analyses and arguments that quantify and substantiate the safety and the level of expert confidence in the safety of the planned repository. The present Interim Summary Report represents a major contribution to the development of this safety case. The report has been compiled in accordance with Posiva's current plan for preparing this safety case. A full safety case for the KBS-3V variant will be developed to support the Preliminary Safety Assessment Report (PSAR) in 2012. The report outlines the current design and safety concept for the planned repository. It summarises the approach used to formulate scenarios for the evolution of the disposal system over time, describes these scenarios and presents the main models and computer codes used to analyse them. It also discusses compliance with Finnish regulatory requirements for long-term safety of a geological repository and gives the main evidence, arguments and analyses that lead to confidence, on the part of Posiva, in the long-term safety of the planned repository. Current understanding of the evolution of the disposal system indicates that, except a few unlikely circumstances affecting a small number of canisters, spent fuel will remain isolated, and the radionuclides contained within the canisters, for hundreds of thousands of years or more, in accordance with the base scenario. Confidence in this base scenario derives, in the first place, from the

Interim performance criteria for photovoltaic energy systems. [Glossary included

Energy Technology Data Exchange (ETDEWEB)

DeBlasio, R.; Forman, S.; Hogan, S.; Nuss, G.; Post, H.; Ross, R.; Schafft, H.

1980-12-01

This document is a response to the Photovoltaic Research, Development, and Demonstration Act of 1978 (P.L. 95-590) which required the generation of performance criteria for photovoltaic energy systems. Since the document is evolutionary and will be updated, the term interim is used. More than 50 experts in the photovoltaic field have contributed in the writing and review of the 179 performance criteria listed in this document. The performance criteria address characteristics of present-day photovoltaic systems that are of interest to manufacturers, government agencies, purchasers, and all others interested in various aspects of photovoltaic system performance and safety. The performance criteria apply to the system as a whole and to its possible subsystems: array, power conditioning, monitor and control, storage, cabling, and power distribution. They are further categorized according to the following performance attributes: electrical, thermal, mechanical/structural, safety, durability/reliability, installation/operation/maintenance, and building/site. Each criterion contains a statement of expected performance (nonprescriptive), a method of evaluation, and a commentary with further information or justification. Over 50 references for background information are also given. A glossary with definitions relevant to photovoltaic systems and a section on test methods are presented in the appendices. Twenty test methods are included to measure performance characteristics of the subsystem elements. These test methods and other parts of the document will be expanded or revised as future experience and needs dictate.

A graded approach to safety documentation at processing facilities

International Nuclear Information System (INIS)

Cowen, M.L.

1992-01-01

Westinghouse Savannah River Company (WSRC) has over 40 major Safety Analysis Reports (SARs) in preparation for non-reactor facilities. These facilities include nuclear material production facilities, waste management facilities, support laboratories and environmental remediation facilities. The SARs for these various projects encompass hazard levels from High to Low, and mission times from startup, through operation, to shutdown. All of these efforts are competing for scarce resources, and therefore some mechanism is required for balancing the documentation requirements. Three of the key variables useful for the decision making process are Depth of Safety Analysis, Urgency of Safety Analysis, and Resource Availability. This report discusses safety documentation at processing facilities

Using resources for scientific-driven pharmacovigilance: from many product safety documents to one product safety master file.

Science.gov (United States)

Furlan, Giovanni

2012-08-01

Current regulations require a description of the overall safety profile or the specific risks of a drug in multiple documents such as the Periodic and Development Safety Update Reports, Risk Management Plans (RMPs) and Signal Detection Reports. In a resource-constrained world, the need for preparing multiple documents reporting the same information results in shifting the focus from a thorough scientific and medical evaluation of the available data to maintaining compliance with regulatory timelines. Since the aim of drug safety is to understand and characterize product issues to take adequate risk minimization measures rather than to comply with bureaucratic requirements, there is the need to avoid redundancy. In order to identify core drug safety activities that need to be undertaken to protect patient safety and reduce the number of documents reporting the results of these activities, the author has reviewed the main topics included in the drug safety guidelines and templates. The topics and sources that need to be taken into account in the main regulatory documents have been found to greatly overlap and, in the future, as a result of the new Periodic Safety Update Report structure and requirements, in the author's opinion this overlap is likely to further increase. Many of the identified inter-document differences seemed to be substantially formal. The Development Safety Update Report, for example, requires separate presentation of the safety issues emerging from different sources followed by an overall evaluation of each safety issue. The RMP, instead, requires a detailed description of the safety issues without separate presentation of the evidence derived from each source. To some extent, however, the individual documents require an in-depth analysis of different aspects; the RMP, for example, requires an epidemiological description of the indication for which the drug is used and its risks. At the time of writing this article, this is not specifically

Documents pertaining to safety control of nuclear facilities

International Nuclear Information System (INIS)

1998-01-01

The Finnish Radiation and Nuclear Safety Authority (STUK) controls the safety of nuclear facilities in Finland. This control encompasses on one hand the evaluation of plant safety on the basis of plans and analyses pertaining to the plant and on the other hand the inspection of plant structures, systems and components as well as of operational activity. STUK also monitors plants operational experience feedback and technical developments in the field, as well as the development of safety research and takes the necessary measures on their basis. Guide YVL 1.1 describes how STUK controls the design, construction and operation of nuclear power plants. The documents to be submitted to STUK are described in the nuclear energy legislation and YVL guides. This guide presents the mode of delivery, quality, contents and number of documents to be submitted to STUK

Interim report on construction of data base for atomic energy science documents (concerning Kyoto University Reactor)

International Nuclear Information System (INIS)

Takeuchi, Takayuki

1984-01-01

The Kyoto University Research Reactor Institute was established in 1963 as a research institute for all universities in Japan utilizing the facilities in common. The construction of a document data base has been undertaken in commemoration of the 20th anniversary of the institute. The data base concerns the research works performed at the institute and also the publications and reports on the research made by the personnel belonging to the institute. Input data are gathered from concerned researchers. In this interim report, the structure and contents of this data base are shortly described. One of the features of this data base is that it handles data with both Japanese and English at the same time. (Aoki, K.)

Physical protection of shipments of irradiated reactor fuel; Interim guidance. Regulatory report

International Nuclear Information System (INIS)

1980-06-01

During May, 1979, the U.S. Nuclear Regulatory Commission approved for issuance in effective form new interim regulations for strengthening the protection of spent fuel shipments against sabotage and diversion. The new regulations were issued without benefit of public comment, but comments from the public were solicited after the effective date. Based upon the public comments received, the interim regulations were amended and reissued in effective form as a final interim rule in May, 1980. The present document supersedes a previously issued interim guidance document, NUREG-0561 (June, 1979) which accompanied the original rule. This report has been revised to conform to the new interim regulations on the physical protection of shipments of irradiated reactor fuel which are likely to remain in effect until the completion of an ongoing research program concerning the response of spent fuel to certain forms of sabotage, at which time the regulations may be rescinded, modified or made permanent, as appropriate. This report discusses the amended regulations and provides a basis on which licensees can develop an acceptable interim program for the protection of spent fuel shipments

75 FR 69648 - Safety Analysis Requirements for Defining Adequate Protection for the Public and the Workers

Science.gov (United States)

2010-11-15

... interpretative posture weakens the safety structure the rule is designed to hold firmly in place. 10 CFR Part 830... Basis Documents, and notes that the Safety Basis Approval Authority may prescribe interim controls and... managers ``are expected to carefully evaluate situations that fall short of expectations and only provide...

75 FR 74022 - Safety Analysis Requirements for Defining Adequate Protection for the Public and the Workers

Science.gov (United States)

2010-11-30

... posture weakens the safety structure the rule is designed to hold firmly in place. 10 CFR Part 830 imposes... Basis Documents, and notes that the Safety Basis Approval Authority may prescribe interim controls and... managers ``are expected to carefully evaluate situations that fall short of expectations and only provide...

Intranet-based safety documentation in management of major hazards and occupational health and safety.

Science.gov (United States)

Leino, Antti

2002-01-01

In the European Union, Council Directive 96/82/EC requires operators producing, using, or handling significant amounts of dangerous substances to improve their safety management systems in order to better manage the major accident potentials deriving from human error. A new safety management system for the Viikinmäki wastewater treatment plant in Helsinki, Finland, was implemented in this study. The system was designed to comply with both the new safety liabilities and the requirements of OHSAS 18001 (British Standards Institute, 1999). During the implementation phase experiences were gathered from the development processes in this small organisation. The complete documentation was placed in the intranet of the plant. Hyperlinks between documents were created to ensure convenience of use. Documentation was made accessible for all workers from every workstation.

Annex D-200 Area Interim Storage Area Final Safety Analysis Report [FSAR] [Section 1 and 2

International Nuclear Information System (INIS)

CARRELL, R.D.

2002-01-01

The 200 Area Interim Storage Area (200 Area ISA) at the Hanford Site provides for the interim storage of non-defense reactor spent nuclear fuel (SNF) housed in aboveground dry cask storage systems. The 200 Area ISA is a relatively simple facility consisting of a boundary fence with gates, perimeter lighting, and concrete and gravel pads on which to place the dry storage casks. The fence supports safeguards and security and establishes a radiation protection buffer zone. The 200 Area ISA is nominally 200,000 ft 2 and is located west of the Canister Storage Building (CSB). Interim storage at the 200 Area ISA is intended for a period of up to 40 years until the materials are shipped off-site to a disposal facility. This Final Safety Analysis Report (FSAR) does not address removal from storage or shipment from the 200 Area ISA. Three different SNF types contained in three different dry cask storage systems are to be stored at the 200 Area ISA, as follows: (1) Fast Flux Test Facility Fuel--Fifty-three interim storage casks (ISC), each holding a core component container (CCC), will be used to store the Fast Flux Test Facility (FFTF) SNF currently in the 400 Area. (2) Neutron Radiography Facility (NRF) TRIGA'--One Rad-Vault' container will store two DOT-6M3 containers and six NRF TRIGA casks currently stored in the 400 Area. (3) Commercial Light Water Reactor Fuel--Six International Standards Organization (ISO) containers, each holding a NAC-I cask4 with an inner commercial light water reactor (LWR) canister, will be used for commercial LWR SNF from the 300 Area. An aboveground dry cask storage location is necessary for the spent fuel because the current storage facilities are being shut down and deactivated. The spent fuel is being transferred to interim storage because there is no permanent repository storage currently available

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

Synthesis of the IRSN report on its analysis of the safety guidance package (DOrS) of the ASTRID reactor project. Safety guidance document for the ASTRID prototype: Referral to the GPR. Opinion related to the safety guidance document of the ASTRID reactor project. ASTRID prototype: Safety guidance document for the ASTRID prototype

International Nuclear Information System (INIS)

Lachaume, Jean-Luc; Niel, Jean-Christophe

2013-01-01

A first document indicates the improvement guidelines for the ASTRID project based on the French experience in the field of sodium-cooled fast neutron reactors, addresses the safety objectives as they are presented for the ASTRID project, discusses how the project includes a regulation and design referential, and how it addresses various aspects of the design approach (ranking and analysis of operation situations, defence in depth, use of probabilistic studies, safety classification and qualification to accidental situations, taking internal and external aggressions into account and taking severe accidents into account at the design level). It comments the guidelines related to the first two barriers, to main safety functions (control of reactivity and of reactor cooling, containment of radioactive and toxic materials), to dismantling, to R and D for safety support. A second document is a letter sent by the ASN to the GPR (permanent group of experts in charge of nuclear reactors) about the safety guidance document for the ASTRID prototype. The third document is the answer and contains comments and recommendations by this group about the content of this document, and therefore addresses the same topics as the first document. The last document defines the framework of the approach to this document

Annex D 200 Area Interim Storage Area Final Safety Analysis Report Volume 5 (FSAR) (Section 1 and 2)

International Nuclear Information System (INIS)

CARRELL, R.D.

2003-01-01

The 200 Area Interim Storage Area (200 Area ISA) at the Hanford Site provides for the interim storage of non-defense reactor spent nuclear fuel (SNF) housed in aboveground dry cask storage systems. The 200 Area ISA is a relatively simple facility consisting of a boundary fence with gates, perimeter lighting, and concrete and gravel pads on which to place the dry storage casks. The fence supports safeguards and security and establishes a radiation protection buffer zone. The 200 Area ISA is nominally 200,000 ft 2 and is located west of the Canister Storage Building (CSB). Interim storage at the 200 Area ISA is intended for a period of up to 40 years until the materials are shipped offsite to a disposal facility. This Final Safety Analysis Report (FSAR) does not address removal from storage or shipment from the 200 Area ISA. Three different SNF types contained in three different dry cask storage systems are to be stored at the 200 Area ISA, as follows: (1) Fast Flux Test Facility (FFTF) Fuel--Fifty-three interim storage casks (ISC), each holding a core component container (CCC), will be used to store the FFTF SNF currently in the 400 Area. (2) Neutron Radiography Facility (NRF)TRIGA--One Rad-Vault container stores two DOT-6M 3 containers and six NRF TRIGA casks. (3) Commercial Light Water Reactor Fuel--Six International Standards Organization (ISO) containers, each holding a NAC-1 cask with an inner commercial light water reactor (LWR) canister, are used for storing commercial LWR SNF from the 300 Area. An aboveground dry cask storage location is necessary for the spent fuel because the current storage facilities are being shut down and deactivated. The spent fuel is being transferred to interim storage because there is no permanent repository storage currently available

Interim performance specifications for conceptual waste-package designs for geologic isolation in salt repositories

International Nuclear Information System (INIS)

1983-06-01

The interim performance specifications and data requirements presented apply to conceptual waste package designs for all waste forms which will be isolated in salt geologic repositories. The waste package performance specifications and data requirements respond to the waste package performance criteria. Subject areas treated include: containment and controlled release, operational period safety, criticality control, identification, and waste package performance testing requirements. This document was generated for use in the development of conceptual waste package designs in salt. It will be revised as additional data, analyses, and regulatory requirements become available

DOE UST interim subsurface barrier technologies workshop

International Nuclear Information System (INIS)

1992-09-01

This document contains information which was presented at a workshop regarding interim subsurface barrier technologies that could be used for underground storage tanks, particularly the tank 241-C-106 at the Hanford Reservation

Preparation of safety and regulatory document for BARC Facilities

International Nuclear Information System (INIS)

Prasad, S.S.; Jayarajan, K.

2017-01-01

In India, the necessary codes and safety guidelines for achieving the safety objectives are provided by the Atomic Energy Regulatory Board (AERB), which are in conformity with the principles of radiation protection as formulated by the International Council of Radiation Protection (ICRP) and International Atomic Energy Agency (IAEA). The same is followed by BARC Safety Council (BSC), which is the regulatory body for the BARC facilities. In addition to all types of fuel cycle facilities, BSC regulates safety of many types of conventional facilities. Many such types of facilities and projects are not under the regulatory purview of AERB. Therefore, the Council has also initiated a programme for development and publication of safety documents for installations in BARC in the fields/ topics yet not addressed by IAEA or AERB. This makes the task pioneering, as some of the areas taken up for defining the regulatory requirements are new, where standard regulatory documents are not available

Interim storage of radioactive waste packages

International Nuclear Information System (INIS)

1998-01-01

This report covers all the principal aspects of production and interim storage of radioactive waste packages. The latest design solutions of waste storage facilities and the operational experiences of developed countries are described and evaluated in order to assist developing Member States in decision making and design and construction of their own storage facilities. This report is applicable to any category of radioactive waste package prepared for interim storage, including conditioned spent fuel, high level waste and sealed radiation sources. This report addresses the following issues: safety principles and requirements for storage of waste packages; treatment and conditioning methods for the main categories of radioactive waste; examples of existing interim storage facilities for LILW, spent fuel and high level waste; operational experience of Member States in waste storage operations including control of storage conditions, surveillance of waste packages and observation of the behaviour of waste packages during storage; retrieval of waste packages from storage facilities; technical and administrative measures that will ensure optimal performance of waste packages subject to various periods of interim storage

Remedial design report and remedial action work plan for the 100-HR-3 and 100-KR-4 groundwater operable units' interim action

International Nuclear Information System (INIS)

1996-09-01

This document is a combination remedial design report and remedial action work plan for the 100-HR-3 and 100-KR-4 Operable Units (located on the Hanford Site in Richland, Washington) interim action. The interim actions described in this document represent the first of an ongoing program to address groundwater contamination in each operable unit. This document describes the design basis, provides a description of the interim action, and identifies how they will meet the requirements set forth in the interim action Record of Decision

Interim data quality objectives for waste pretreatment and vitrification. Revision 1

International Nuclear Information System (INIS)

Kupfer, M.J.; Conner, J.M.; Kirkbride, R.A.; Mobley, J.R.

1994-01-01

The Tank Waste Remediation System (TWRS) is responsible for storing, processing, and immobilizing the Hanford Site tank wastes. Characterization information on the tank wastes is needed so that safety concerns can be addressed, and retrieval, pretreatment, and immobilization processes can be designed, permitted, and implemented. This document describes the near-term tank waste sampling and characterization needs of the Pretreatment, High-Level Waste (HLW) Disposal, and Low-Level Waste (LLW) Disposal Programs to support the TWRS disposal mission. The final DQO (Data Quality Objective) will define specific waste tanks to be sampled, sample timing requirements, an appropriate analytical scheme, and a list of required analytes. This interim DQO, however, focuses primarily on the required analytes since the tanks to be sampled in FY 1994 and early FY 1995 are being driven most heavily by other considerations, particularly safety. The major objective of this Interim DQO is to provide guidance for tank waste characterization requirements for samples taken before completion of the final DQO. The characterization data needs defined herein will support the final DQO to help perform the following: Support the TWRS technical strategy by identification of the chemical and physical composition of the waste in the tanks and Guide development efforts to define waste pretreatment processes, which will in turn define HLW and LLW feed to vitrification processes

IAEA activities in preparation of reglamentary documents on nuclear power plant safety

International Nuclear Information System (INIS)

Konstantinov, L.V.

1976-01-01

The activities of the IAEA in the field of working out practical rules and recommendations ensuring the nuclear power plant safety are discussed. The practical rules will establish the aims and the minimum of requirements, that must be carried out to ensure the necessary safety of systems, components and equipment of the nuclear power plant throughout the whole period of its exploitation. Described is the procedure of the document preparation, consisting of the collection of documents, edited in different countries, the integration of documents by the IAEA Secretariat, the consideratiom of documents by the Group of senior advisers, the preparation of the draft document, the additional wort at the document in accordaqce with the remarks of the IAEA member-countries, the edition and dissemination of documents. The necessity for the active participation of the CMEA member-countries in the development and discussion of documents concerning the nuclear power plant safety is stated [ru

Accident consequence calculations for project W-058 safety analysis

International Nuclear Information System (INIS)

Van Keuren, J.C.

1997-01-01

This document describes the calculations performed to determine the accident consequences for the W-058 safety analysis. Project W-058 is the replacement cross site transfer system (RCSTS), which is designed to transort liquid waste between the 200 W and 200 E areas. Calculations for RCSTS safety analyses used the same methods as the calculations for the Tank Waste Remediation System (TWRS) Basis for Interim Operation (BIO) and its supporting calculation notes. Revised analyses were performed for the spray and pool leak accidents since the RCSTS flows and pressures differ from those assumed in the TWRS BIO. Revision 1 of the document incorporates review comments

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

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

Site safety progress review of spent fuel central interim storage facility. Final report

International Nuclear Information System (INIS)

Gurpinar, A.; Serva, L.; Giuliani

1995-01-01

Following the request of the Czech Power Board (CEZ) and within the scope of the Technical Cooperation Project CZR/9/003, a progress review of the site safety of the Spent Fuel Central Interim Storage Facility (SFCISF) was performed. The review involved the first two stages of the works comprising the regional survey and identification of candidate sites for the underground and surface storage options. Five sites have been identified as a result of the previous works. The following two stages will involved the identification of the preferred candidate sites for the two options and the final site qualification. The present review had the purpose of assessing the work already performed and making recommendations for the next two stages of works

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.

Interim format and content for a physical security plan for nuclear power plants

International Nuclear Information System (INIS)

1977-02-01

The document serves as interim guidance to assist the licensee or applicant in the preparation of a physical security plan. It is to be used in conjunction with interim acceptance criteria for physical security programs, which will be distributed at a later date

Immobilized high-level waste interim storage alternatives generation and analysis and decision report

International Nuclear Information System (INIS)

CALMUS, R.B.

1999-01-01

This report presents a study of alternative system architectures to provide onsite interim storage for the immobilized high-level waste produced by the Tank Waste Remediation System (TWRS) privatization vendor. It examines the contract and program changes that have occurred and evaluates their impacts on the baseline immobilized high-level waste (IHLW) interim storage strategy. In addition, this report documents the recommended initial interim storage architecture and implementation path forward

German Approach for the Transport of Spent Fuel Packages after Interim Storage

International Nuclear Information System (INIS)

Wille, Frank; Wolff, Dietmar; Droste, Bernhard; Voelzke, Holger

2014-01-01

In Germany the concept of dry interim storage of spent nuclear fuel in dual purpose metal casks is implemented, currently for periods of up to 40 years. The casks being used have an approved package design in accordance with the international transport regulations. The license for dry storage is granted on the German Atomic Energy Act with respect to the recently (in 2012) revised 'Guidelines for dry cask storage of spent nuclear fuel and heat-generating waste' by the German Waste management Commission (ESK) which are very similar to the former RSK (reactor safety commission) guidelines. For transport on public routes between or after long term interim storage periods, it has to be ensured that the transport and storage casks fulfil the specifications of the transport approval or other sufficient properties which satisfy the proofs for the compliance of the safety objectives at that time. In recent years the validation period of transport approval certificates for manufactured, loaded and stored packages were discussed among authorities and applicants. A case dependent system of 3, 5 and 10 years was established. There are consequences for the safety cases in the Package Design Safety Report including evaluation of long term behavior of components and specific operating procedures of the package. Present research and knowledge concerning the long term behavior of transport and storage cask components have to be consulted as well as experiences from interim cask storage operations. Challenges in the safety assessment are e.g. the behavior of aged metal and elastomeric seals under IAEA test conditions to ensure that the results of drop tests can be transferred to the compliance of the safety objectives at the time of transport after the interim storage period (aged package). Assessment methods for the material compatibility, the behavior of fuel assemblies and the aging behavior of shielding parts are issues as well. This paper describes the state

Interim criteria for Organic Watch List tanks at the Hanford Site

International Nuclear Information System (INIS)

Babad, S.; Turner, D.A.

1993-09-01

This document establishes interim criteria for identifying single-shell radioactive waste storage tanks at the Hanford Site that contain organic chemicals mixed with nitrate/nitrite salts in potentially hazardous concentrations. These tanks are designated as ''organic Watch List tanks.'' Watch List tanks are radioactive waste storage tanks that have the potential for release of high-level waste as a result of uncontrolled increases in temperature or pressure. Organic Watch List tanks are those Watch List tanks that contain relatively high concentrations of organic chemicals. Because of the potential for release of high-level waste resulting from uncontrolled increases in temperature or pressure, the organic Watch List tanks (collectively) constitute a Hanford Site radioactive waste storage tank ''safety issue.''

Documents and legal texts

International Nuclear Information System (INIS)

2013-01-01

This section reprints a selection of recently published legislative texts and documents: - Russian Federation: Federal Law No.170 of 21 November 1995 on the use of atomic energy, Adopted by the State Duma on 20 October 1995; - Uruguay: Law No.19.056 On the Radiological Protection and Safety of Persons, Property and the Environment (4 January 2013); - Japan: Third Supplement to Interim Guidelines on Determination of the Scope of Nuclear Damage resulting from the Accident at the Tokyo Electric Power Company Fukushima Daiichi and Daini Nuclear Power Plants (concerning Damages related to Rumour-Related Damage in the Agriculture, Forestry, Fishery and Food Industries), 30 January 2013; - France and the United States: Joint Statement on Liability for Nuclear Damage (Aug 2013); - Franco-Russian Nuclear Power Declaration (1 November 2013)

Hanford Site Wide Transportation Safety Document [SEC 1 Thru 3

Energy Technology Data Exchange (ETDEWEB)

MCCALL, D L

2002-06-01

This safety evaluation report (SER) documents the basis for the US Department of Energy (DOE), Richland Operations Office (RL) to approve the Hanford Sitewide Transportation Safety Document (TSD) for onsite Transportation and Packaging (T&P) at Hanford. Hanford contractors, on behalf of DOE-RL, prepared and submitted the Hanford Sitewide Transportation Safety Document, DOE/RL-2001-0036, Revision 0, (DOE/RL 2001), dated October 4, 2001, which is referred to throughout this report as the TSD. In the context of the TSD, Hanford onsite shipments are the activities of moving hazardous materials, substances, and wastes between DOE facilities and over roadways where public access is controlled or restricted and includes intra-area and inter-area movements. The TSD sets forth requirements and standards for onsite shipment of radioactive and hazardous materials and wastes within the confines of the Hanford Site on roadways where public access is restricted by signs, barricades, fences, or other means including road closures and moving convoys controlled by Hanford Site security forces.

Permitting plan for the high-level waste interim storage

International Nuclear Information System (INIS)

Deffenbaugh, M.L.

1997-01-01

This document addresses the environmental permitting requirements for the transportation and interim storage of solidified high-level waste (HLW) produced during Phase 1 of the Hanford Site privatization effort. Solidified HLW consists of canisters containing vitrified HLW (glass) and containers that hold cesium separated during low-level waste pretreatment. The glass canisters and cesium containers will be transported to the Canister Storage Building (CSB) in a U.S. Department of Energy (DOE)-provided transportation cask via diesel-powered tractor trailer. Tri-Party Agreement (TPA) Milestone M-90 establishes a new major milestone, and associated interim milestones and target dates, governing acquisition and/or modification of facilities necessary for: (1) interim storage of Tank Waste Remediation Systems (TWRS) immobilized HLW (IHLW) and other canistered high-level waste forms; and (2) interim storage and disposal of TWRS immobilized low-activity tank waste (ILAW). An environmental requirements checklist and narrative was developed to identify the permitting path forward for the HLW interim storage (HLWIS) project (See Appendix B). This permitting plan will follow the permitting logic developed in that checklist

Insights from the interim reliability evaluation program pertinent to reactor safety issues

International Nuclear Information System (INIS)

Carlson, D.D.

1983-01-01

The Interim Reliability Evaluation Program (IREP) consisted of concurrent probabilistic analyses of four operating nuclear power plants. This paper presents and integrated view of the results of the analyses drawing insights pertinent to reactor safety. The importance to risk of accident sequences initiated by transients and small loss-of-coolant accidents was confirmed. Support systems were found to contribute significantly to the sets of dominant accident sequences, either due to single failures which could disable one or more mitigating systems or due to their initiating plant transients. Human errors in response to accidents also were important risk contributors. Consideration of operator recovery actions influences accident sequence frequency estimates, the list of accident sequences dominating core melt, and the set of dominant risk contributors. Accidents involving station blackout, reactor coolant pump seal leaks and ruptures, and loss-of-coolant accidents requiring manual initiation of coolant injection were found to be risk significant

Electronic Medical Record Documentation of Driving Safety for Veterans with Diagnosed Dementia.

Science.gov (United States)

Vair, Christina L; King, Paul R; Gass, Julie; Eaker, April; Kusche, Anna; Wray, Laura O

2018-01-01

Many older adults continue to drive following dementia diagnosis, with medical providers increasingly likely to be involved in addressing such safety concerns. This study examined electronic medical record (EMR) documentation of driving safety for veterans with dementia (N = 118) seen in Veterans Affairs primary care and interdisciplinary geriatrics clinics in one geographic region over a 10-year period. Qualitative directed content analysis of retrospective EMR data. Assessment of known risk factors or subjective concerns for unsafe driving were documented in fewer than half of observed cases; specific recommendations for driving safety were evident for a minority of patients, with formal driving evaluation the most frequently documented recommendation by providers. Utilizing data from actual clinical encounters provides a unique snapshot of how driving risk and safety concerns are addressed for veterans with dementia. This information provides a meaningful frame of reference for understanding potential strengths and possible gaps in how this important topic area is being addressed in the course of clinical care. The EMR is an important forum for interprofessional communication, with documentation of driving risk and safety concerns an essential element for continuity of care and ensuring consistency of information delivered to patients and caregivers.

Fire Hazards Analysis for the 200 Area Interim Storage Area

International Nuclear Information System (INIS)

JOHNSON, D.M.

2000-01-01

This documents the Fire Hazards Analysis (FHA) for the 200 Area Interim Storage Area. The Interim Storage Cask, Rad-Vault, and NAC-1 Cask are analyzed for fire hazards and the 200 Area Interim Storage Area is assessed according to HNF-PRO-350 and the objectives of DOE Order 5480 7A. This FHA addresses the potential fire hazards associated with the Interim Storage Area (ISA) facility in accordance with the requirements of DOE Order 5480 7A. It is intended to assess the risk from fire to ensure there are no undue fire hazards to site personnel and the public and to ensure property damage potential from fire is within acceptable limits. This FHA will be in the form of a graded approach commensurate with the complexity of the structure or area and the associated fire hazards

The Safety Assessment of Long term Interim Storage at Sellafield

International Nuclear Information System (INIS)

Buchan, Andrew B.

2014-01-01

that are most significant in terms of frequency and unmitigated potential consequences PSA looks at the full range of fault sequences and allows full incorporation of the reliability and failure probability of the safety measures and other features of the design and operations SAA considers significant but unlikely accidents where off-site consequences are likely to significantly affect the critical group and provides information on their progression, within the facility and also beyond the site boundary. The paper will illustrate how these techniques have been utilised to facilitate design, operation, resilience evaluation and accident management of facilities supporting long term interim storage at Sellafield. (author)

Interim safety equipment list for 241-C-106 waste retrieval, project W-320

International Nuclear Information System (INIS)

Conner, J.C.

1996-01-01

The purpose of this supporting document is to provide safety classifications for systems, structures, and components of the Tank 241-C-106 Waste Retrieval Sluicing System (WRSS) and to document the methodology used to develop these safety classifications. The WRSS requires two transfer lines, one to carry sluiced waste slurry to tank 241-AY-102 and the other to return supernatant to tank 241-C-106; pumps in each tank; sluicers to direct the supernatant stream inside tank 241-C-106; a slurry distributor in tank 241-AY-102; heating, ventilation, and air conditioning for tank 241-C-106; and instrumentation and control devices

Interim safety equipment list for 241-C-106 waste retrieval, project W-320

Energy Technology Data Exchange (ETDEWEB)

Conner, J.C.

1996-01-25

The purpose of this supporting document is to provide safety classifications for systems, structures, and components of the Tank 241-C-106 Waste Retrieval Sluicing System (WRSS) and to document the methodology used to develop these safety classifications. The WRSS requires two transfer lines, one to carry sluiced waste slurry to tank 241-AY-102 and the other to return supernatant to tank 241-C-106; pumps in each tank; sluicers to direct the supernatant stream inside tank 241-C-106; a slurry distributor in tank 241-AY-102; heating, ventilation, and air conditioning for tank 241-C-106; and instrumentation and control devices.

Interim Reliability Evaluation Program procedures guide

International Nuclear Information System (INIS)

Carlson, D.D.; Gallup, D.R.; Kolaczkowski, A.M.; Kolb, G.J.; Stack, D.W.; Lofgren, E.; Horton, W.H.; Lobner, P.R.

1983-01-01

This document presents procedures for conducting analyses of a scope similar to those performed in Phase II of the Interim Reliability Evaluation Program (IREP). It documents the current state of the art in performing the plant systems analysis portion of a probabilistic risk assessment. Insights gained into managing such an analysis are discussed. Step-by-step procedures and methodological guidance constitute the major portion of the document. While not to be viewed as a cookbook, the procedures set forth the principal steps in performing an IREP analysis. Guidance for resolving the problems encountered in previous analyses is offered. Numerous examples and representative products from previous analyses clarify the discussion

Assessment by peer review of the effectiveness of a regulatory programme for radiation safety. Interim report for comment

International Nuclear Information System (INIS)

2002-06-01

This document covers assessment of those aspects of a radiation protection and safety infrastructure that are implemented by the Regulatory Authority for radiation sources and practices using such sources and necessarily includes those ancillary technical services, such as dosimetry services, which directly affect the ability of the Regulatory Authority to discharge its responsibilities. The focus of the guidance in this TECDOC is on assessment of a regulatory programme intended to implement the BSS. The BSS address transportation and waste safety mainly by reference to other IAEA documents. When conducting an assessment, the Review Team members should be aware of the latest IAEA documents (or similar national documents) concerning transportation and waste safety and, if appropriate, nuclear safety, and take them into account to the extent applicable when assessing the effectiveness of the regulatory programme governing radiation protection and safety of radiation source practices in a particular State

Assessment by peer review of the effectiveness of a regulatory programme for radiation safety. Interim report for comment

International Nuclear Information System (INIS)

2001-05-01

This document covers assessment of those aspects of a radiation protection and safety infrastructure that are implemented by the Regulatory Authority for radiation sources and practices using such sources and necessarily includes those ancillary technical services, such as dosimetry services, which directly affect the ability of the Regulatory Authority to discharge its responsibilities. The focus of the guidance in this TECDOC is on assessment of a regulatory programme intended to implement the BSS. The BSS address transportation and waste safety mainly by reference to other IAEA documents. When conducting an assessment, the Review Team members should be aware of the latest IAEA documents (or similar national documents) concerning transportation and waste safety and, if appropriate, nuclear safety, and take them into account to the extent applicable when assessing the effectiveness of the regulatory programme governing radiation protection and safety of radiation source practices in a particular State

Interim storage report

International Nuclear Information System (INIS)

Rawlins, J.K.

1998-02-01

High-level radioactive waste (HLW) stored at the Idaho Chemical Processing Plant (ICPP) in the form of calcine and liquid and liquid sodium-bearing waste (SBW) will be processed to provide a stable waste form and prepare the waste to be transported to a permanent repository. Because a permanent repository will not be available when the waste is processed, the waste must be stored at ICPP in an Interim Storage Facility (ISF). This report documents consideration of an ISF for each of the waste processing options under consideration

CMM Interim Check (U)

Energy Technology Data Exchange (ETDEWEB)

Montano, Joshua Daniel [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2015-03-23

Coordinate Measuring Machines (CMM) are widely used in industry, throughout the Nuclear Weapons Complex and at Los Alamos National Laboratory (LANL) to verify part conformance to design definition. Calibration cycles for CMMs at LANL are predominantly one year in length. Unfortunately, several nonconformance reports have been generated to document the discovery of a certified machine found out of tolerance during a calibration closeout. In an effort to reduce risk to product quality two solutions were proposed – shorten the calibration cycle which could be costly, or perform an interim check to monitor the machine’s performance between cycles. The CMM interim check discussed makes use of Renishaw’s Machine Checking Gauge. This off-the-shelf product simulates a large sphere within a CMM’s measurement volume and allows for error estimation. Data was gathered, analyzed, and simulated from seven machines in seventeen different configurations to create statistical process control run charts for on-the-floor monitoring.

105-C Reactor interim safe storage project technology integration plan

International Nuclear Information System (INIS)

Pulsford, S.K.

1997-01-01

The 105-C Reactor Interim Safe Storage Project Technology Integration Plan involves the decontamination, dismantlement, and interim safe storage of a surplus production reactor. A major goal is to identify and demonstrate new and innovative D and D technologies that will reduce costs, shorten schedules, enhance safety, and have the potential for general use across the RL complex. Innovative technologies are to be demonstrated in the following areas: Characterization; Decontamination; Waste Disposition; Dismantlement, Segmentation, and Demolition; Facility Stabilization; and Health and Safety. The evaluation and ranking of innovative technologies has been completed. Demonstrations will be selected from the ranked technologies according to priority. The contractor team members will review and evaluate the demonstration performances and make final recommendations to DOE

PROJECT W-551 INTERIM PRETREATMENT SYSTEM TECHNOLOGY SELECTION SUMMARY DECISION REPORT AND RECOMMENDATION

International Nuclear Information System (INIS)

CONRAD EA

2008-01-01

This report provides the conclusions of the tank farm interim pretreatment technology decision process. It documents the methodology, data, and results of the selection of cross-flow filtration and ion exchange technologies for implementation in project W-551, Interim Pretreatment System. This selection resulted from the evaluation of specific scope criteria using quantitative and qualitative analyses, group workshops, and technical expert personnel

Sustainable Solutions for Nuclear used Fuels Interim Storage

International Nuclear Information System (INIS)

Arslan, Marc; Favet, Dominique; Issard, Herve; Le Jemtel, Amaury; Drevon, Caroline

2014-01-01

AREVA has a unique experience in providing sustainable solutions for used fuel management, fitted with the needs of different customers in the world and with regulation in different countries. These solutions entail both recycling and interim storage technologies. In a first part, we will describe the various types of solutions for Interim Storage of UNF that have been implemented around the world for interim storage at reactor or centralized Pad solution in canisters dry storage, vault type storages for dry storage, dry storage of transportation casks (dual purpose) pools for wet storage, The experience for all these different families of interim storages in which AREVA is involved is extensive and will be discussed with respect to the new challenges: increase of the duration of the interim storage (long term interim storage) increase of burn up of the fuels In a second part of the presentation, special recycling features will be presented. In that case, interim storage of the used fuels is ensured in pools. This provides in the long term good conditions for the behaviour of the fuel and its retrievability. With recycling, the final waste (Universal Canister of vitrified fission products and compacted hulls and end pieces): is stable and licensed in many countries for the final disposal (France, UK, Belgium, NL, Switzerland, Germany, Japan, upcoming: Spain, Australia, Italy). Presents neither safety criticality risks nor proliferation risks (AREVA conditioned HLW and LL-ILW are free of IAEA safeguard constraints thanks to AREVA process high recovery and purification yields). It can therefore be safely stored in interim storage for more than 100 years before final disposal. Some economic considerations will also be discussed. In particular, in the case of long term interim storage of used fuels, there are growing uncertainties regarding the future needs of repackaging and transportation, which can result in future cost overruns. Meanwhile, in the recycling policy

Interim restorations.

Science.gov (United States)

Gratton, David G; Aquilino, Steven A

2004-04-01

Interim restorations are a critical component of fixed prosthodontic treatment, biologically and biomechanically. Interim restoration serves an important diagnostic role as a functional and esthetic try-in and as a blueprint for the design of the definitive prosthesis. When selecting materials for any interim restoration, clinicians must consider physical properties, handling properties, patient acceptance, and material cost. Although no single material meets all the requirements and material classification alone of a given product is not a predictor of clinical performance, bis-acryl materials are typically best suited to single-unit restorations, and poly(methylmethacrylate) interim materials are generally ideal for multi-unit, complex, long-term, interim fixed prostheses. As with most dental procedures, the technique used for fabrication has a greater effect on the final result than the specific material chosen.

75 FR 35510 - License Renewal Interim Staff Guidance Process, Revision 2 Notice of Availability

Science.gov (United States)

2010-06-22

... Related Regulatory Functions.'' An electronic copy of the revised LR-ISG process is available in the NRC's Agencywide Documents Access and Management System (ADAMS) under Accession No. ML100920158. The revised LR-ISG... interim changes to certain NRC license renewal guidance documents. These guidance documents facilitate the...

Safety aspects of spent nuclear fuel interim storage installations

Energy Technology Data Exchange (ETDEWEB)

Romanato, Luiz Sergio [Centro Tecnologico da Marinha em Sao Paulo (CTMSP), Sao Paulo, SP (Brazil). Dept. da Qualidade. Div. de Sistemas da Qualidade]. E-mail: romanato@ctmsp.mar.mil.br; Rzyski, Barbara Maria [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil). Div. de Ensino]. E-mail: bmrzyski@ipen.br

2007-07-01

Nowadays safety and security of spent nuclear fuel (SNF) interim storage installations are very important, due to a great concentration of fission products, actinides and activation products. In this kind of storage it is necessary to consider the physical security. Nuclear installations have become more vulnerable. New types of accidents must be considered in the design of these installations, which in the early days were not considered like: fissile material stolen, terrorists' acts and war conflicts, and traditional accidents concerning the transport of the spent fuel from the reactor to the storage location, earthquakes occurrence, airplanes crash, etc. Studies related to airplane falling had showed that a collision of big commercials airplanes at velocity of 800 km/h against SNF storage and specially designed concrete casks, do not result in serious structural injury to the casks, and not even radionuclides liberation to the environment. However, it was demonstrated that attacks with modern military ammunitions, against metallic casks, are calamitous. The casks could not support a direct impact of this ammo and the released radioactive materials can expose the workers and public as well the local environment to harmful radiation. This paper deals about the main basic aspects of a dry SNF storage installation, that must be physically well protected, getting barriers that difficult the access of unauthorized persons or vehicles, as well as, must structurally resist to incidents or accidents caused by unauthorized intrusion. (author)

Safety aspects of spent nuclear fuel interim storage installations

International Nuclear Information System (INIS)

Romanato, Luiz Sergio

2007-01-01

Nowadays safety and security of spent nuclear fuel (SNF) interim storage installations are very important, due to a great concentration of fission products, actinides and activation products. In this kind of storage it is necessary to consider the physical security. Nuclear installations have become more vulnerable. New types of accidents must be considered in the design of these installations, which in the early days were not considered like: fissile material stolen, terrorists' acts and war conflicts, and traditional accidents concerning the transport of the spent fuel from the reactor to the storage location, earthquakes occurrence, airplanes crash, etc. Studies related to airplane falling had showed that a collision of big commercials airplanes at velocity of 800 km/h against SNF storage and specially designed concrete casks, do not result in serious structural injury to the casks, and not even radionuclides liberation to the environment. However, it was demonstrated that attacks with modern military ammunitions, against metallic casks, are calamitous. The casks could not support a direct impact of this ammo and the released radioactive materials can expose the workers and public as well the local environment to harmful radiation. This paper deals about the main basic aspects of a dry SNF storage installation, that must be physically well protected, getting barriers that difficult the access of unauthorized persons or vehicles, as well as, must structurally resist to incidents or accidents caused by unauthorized intrusion. (author)

76 FR 74834 - Interim Staff Guidance on Aging Management Program for Steam Generators

Science.gov (United States)

2011-12-01

... for Steam Generators AGENCY: Nuclear Regulatory Commission. ACTION: Interim staff guidance; issuance... (LR-ISG), LR-ISG-2011-02, ``Aging Management Program for Steam Generators.'' This LR-ISG provides the...) document, NEI 97-06, ``Steam Generator Program Guidelines,'' (NRC's Agencywide Documents Access and...

Interim storage study report

Energy Technology Data Exchange (ETDEWEB)

Rawlins, J.K.

1998-02-01

High-level radioactive waste (HLW) stored at the Idaho Chemical Processing Plant (ICPP) in the form of calcine and liquid and liquid sodium-bearing waste (SBW) will be processed to provide a stable waste form and prepare the waste to be transported to a permanent repository. Because a permanent repository will not be available when the waste is processed, the waste must be stored at ICPP in an Interim Storage Facility (ISF). This report documents consideration of an ISF for each of the waste processing options under consideration.

Surface Water Interim Measures/Interim Remedial Action Plan/ Environmental and Decision Document, South Walnut Creek Basin, Operable Unit No.2

International Nuclear Information System (INIS)

1991-01-01

Water quality investigations have identified the presence of volatile organic compound (VOC) and radionuclide contamination of surface water at the Rocky Flats Plant (RFP). The subject interim Measures/Interim Remedial Action Plan/Environmental Assessment (IM/IRAP/EA) addresses contaminated surface water in a portion of the South Walnut Creek drainage basin located within an area identified as Operable Unit No. 2 (OU 2). There is no immediate threat to public health and the environment posed by this surface water contamination. The affected surface water is contained within the plant boundary by existing detention ponds, and is treated prior to discharge for removal of volatile contaminants and suspended particulates to which radionuclides, if present, are likely to absorb. However, there is a potential threat and the Department of Energy (DOE) is implementing this Surface Water IM/IRAP at the request of the US Environmental Protection Agency (EPA) and Colorado Department of Health (CDH). Implementation of the Surface Water IM/IRA will enhance the DOE's efforts towards containing and managing contaminated surface water, and will mitigate downgradient migration of contaminants. Another factor in implementing this IM/IRA is the length of time it will take to complete the investigations and engineering studies necessary to determine the final remedy for OU 2. 44 refs., 23 figs., 14 tabs

Transportation Safety Excellence in Operations Through Improved Transportation Safety Document

International Nuclear Information System (INIS)

Dr. Michael A. Lehto; MAL

2007-01-01

A recent accomplishment of the Idaho National Laboratory (INL) Materials and Fuels Complex (MFC) Nuclear Safety analysis group was to obtain DOE-ID approval for the inter-facility transfer of greater-than-Hazard-Category-3 quantity radioactive/fissionable waste in Department of Transportation (DOT) Type A drums at MFC. This accomplishment supported excellence in operations through safety analysis by better integrating nuclear safety requirements with waste requirements in the Transportation Safety Document (TSD); reducing container and transport costs; and making facility operations more efficient. The MFC TSD governs and controls the inter-facility transfer of greater-than-Hazard-Category-3 radioactive and/or fissionable materials in non-DOT approved containers. Previously, the TSD did not include the capability to transfer payloads of greater-than-Hazard-Category-3 radioactive and/or fissionable materials using DOT Type A drums. Previous practice was to package the waste materials to less-than-Hazard-Category-3 quantities when loading DOT Type A drums for transfer out of facilities to reduce facility waste accumulations. This practice allowed operations to proceed, but resulted in drums being loaded to less than the Waste Isolation Pilot Plant (WIPP) waste acceptance criteria (WAC) waste limits, which was not cost effective or operations friendly. An improved and revised safety analysis was used to gain DOE-ID approval for adding this container configuration to the MFC TSD safety basis. In the process of obtaining approval of the revised safety basis, safety analysis practices were used effectively to directly support excellence in operations. Several factors contributed to the success of MFC's effort to obtain approval for the use of DOT Type A drums, including two practices that could help in future safety basis changes at other facilities. (1) The process of incorporating the DOT Type A drums into the TSD at MFC helped to better integrate nuclear safety

Basis for Interim Operation for the K-Reactor in Cold Standby

Energy Technology Data Exchange (ETDEWEB)

Shedrow, B.

1998-10-19

The Basis for Interim Operation (BIO) document for K Reactor in Cold Standby and the L- and P-Reactor Disassembly Basins was prepared in accordance with the draft DOE standard for BIO preparation (dated October 26, 1993).

Basis for Interim Operation for the K-Reactor in Cold Standby

International Nuclear Information System (INIS)

Shedrow, B.

1998-01-01

The Basis for Interim Operation (BIO) document for K Reactor in Cold Standby and the L- and P-Reactor Disassembly Basins was prepared in accordance with the draft DOE standard for BIO preparation (dated October 26, 1993)

Implementing 10 CFR 830 at the FEMP Silos: Nuclear Health and Safety Plans as Documented Safety Analysis

International Nuclear Information System (INIS)

Fisk, Patricia; Rutherford, Lavon

2003-01-01

The objective of the Silos Project at the Fernald Closure Project (FCP) is to safely remediate high-grade uranium ore residues (Silos 1 and 2) and metal oxide residues (Silo 3). The evolution of Documented Safety Analyses (DSAs) for these facilities has reflected the changes in remediation processes. The final stage in silos DSAs is an interpretation of 10 CFR 830 Safe Harbor Requirements that combines a Health and Safety Plan with nuclear safety requirements. This paper will address the development of a Nuclear Health and Safety Plan, or N-HASP

Plutonium Finishing Plant. Interim plutonium stabilization engineering study

Energy Technology Data Exchange (ETDEWEB)

Sevigny, G.J.; Gallucci, R.H.; Garrett, S.M.K.; Geeting, J.G.H.; Goheen, R.S.; Molton, P.M.; Templeton, K.J.; Villegas, A.J. [Pacific Northwest Lab., Richland, WA (United States); Nass, R. [Nuclear Fuel Services, Inc. (United States)

1995-08-01

This report provides the results of an engineering study that evaluated the available technologies for stabilizing the plutonium stored at the Plutonium Finishing Plant located at the hanford Site in southeastern Washington. Further processing of the plutonium may be required to prepare the plutonium for interim (<50 years) storage. Specifically this document provides the current plutonium inventory and characterization, the initial screening process, and the process descriptions and flowsheets of the technologies that passed the initial screening. The conclusions and recommendations also are provided. The information contained in this report will be used to assist in the preparation of the environmental impact statement and to help decision makers determine which is the preferred technology to process the plutonium for interim storage.

Plutonium Finishing Plant. Interim plutonium stabilization engineering study

International Nuclear Information System (INIS)

Sevigny, G.J.; Gallucci, R.H.; Garrett, S.M.K.; Geeting, J.G.H.; Goheen, R.S.; Molton, P.M.; Templeton, K.J.; Villegas, A.J.; Nass, R.

1995-08-01

This report provides the results of an engineering study that evaluated the available technologies for stabilizing the plutonium stored at the Plutonium Finishing Plant located at the hanford Site in southeastern Washington. Further processing of the plutonium may be required to prepare the plutonium for interim (<50 years) storage. Specifically this document provides the current plutonium inventory and characterization, the initial screening process, and the process descriptions and flowsheets of the technologies that passed the initial screening. The conclusions and recommendations also are provided. The information contained in this report will be used to assist in the preparation of the environmental impact statement and to help decision makers determine which is the preferred technology to process the plutonium for interim storage

Public School Finance Problems in Texas. An Interim Report.

Science.gov (United States)

Texas Research League, Austin.

The U.S. District Court ruling in Rodriguez vs San Antonio Independent School District, which struck down Texas' school finance system as inequitable and unconstitutional, provided the impetus for publishing this interim report. The report documents the growing cost of State-supported public school programs--the primary concern prior to the…

Interim Storage of Spent Nuclear Fuel before Final Disposal in Germany - Regulator's view

International Nuclear Information System (INIS)

Arens, G.; Goetz, Ch.; Geupel, Sandra; Gmal, B.; Mester, W.

2014-01-01

For spent nuclear fuel management in Germany the concept of dry interim storage in dual purpose casks before direct disposal is applied. The Federal Office for Radiation Protection (BfS) is the competent authority for licensing of interim storage facilities. The competent authority for surveillance of operation is the responsible authority of the respective federal state (Land). Currently operation licenses for storage facilities have been granted for a storage time of 40 years and are based on safety demonstrations for all safety issues as safe enclosure, shielding, sub-criticality and decay heat removal under consideration of operation conditions. In addition, transportability of the casks for the whole storage period has to be provided. Due to current delay in site selection and exploration of a disposal site, an extension of the storage time beyond 40 years could be needed. This will cause appropriate actions by the licensee and the competent authorities as well. A brief description of the regulatory base of licensing and surveillance of interim storage is given from the regulators view. Furthermore the current planning for final disposal of spent nuclear fuel and high level waste and its interconnections between storage and disposal concepts are shortly explained. Finally the relevant aspects for licensing of extended storage time beyond 40 years will be discussed. Current activities on this issue, which have been initiated by the Federal Government, will be addressed. On the regulatory side a review and amendment of the safety guideline for interim storage of spent fuel has been performed and the procedure of periodic safety review is being implemented. A guideline for implementing an ageing management programme is available in a draft version. Regarding safety of long term storage a study focussing on the identification and evaluation of long term effects as well as gaps of knowledge has been finished in 2010. A continuation and update is currently underway

LESSONS LEARNED IN DEVELOPMENT OF THE HANFORD SWOC MASTER DOCUMENTED SAFETY ANALYSIS (MDSA) and IMPLEMENTATION VALIDATION REVIEW (IVR)

International Nuclear Information System (INIS)

MORENO, M.R.

2004-01-01

DOE set clear expectations on a cost-effective approach for achieving compliance with the Nuclear Safety Management requirements (20 CFR 830, Nuclear Safety Rule), which ensured long-term benefit to Hanford, via issuance of a nuclear safety strategy in February 2003. To facilitate implementation of these expectations, tools were developed to streamline and standardize safety analysis and safety document development with the goal of a shorter and more predictable DOE approval cycle. A Hanford Safety Analysis and Risk Assessment Handbook (SARAH) was approved to standardize methodologies for development of safety analyses. A Microsoft Excel spreadsheet (RADIDOSE) was approved for the evaluation of radiological consequences for accident scenarios often postulated at Hanford. Standard safety management program chapters were approved for use as a means of compliance with the programmatic chapters of DOE-STD-3009, ''Preparation Guide for U.S. Department of Energy Nonreactor Nuclear Facility Safety Analysis Reports''. An in-process review was developed between DOE and the Contractor to facilitate DOE approval and provide early course correction. The new Documented Safety Analysis (DSA) developed to address the operations of four facilities within the Solid Waste Operations Complex (SWOC) necessitated development of an Implementation Validation Review (IVR) process. The IVR process encompasses the following objectives: safety basis controls and requirements are adequately incorporated into appropriate facility documents and work instructions, facility personnel are knowledgeable of controls and requirements, and the DSA/TSR controls have been implemented. Based on DOE direction and safety analysis tools, four waste management nuclear facilities were integrated into one safety basis document. With successful completion of implementation of this safety document, lessons-learned from the in-process review, safety analysis tools and IVR process were documented for future action

Criticality safety evaluation report for FFTF 42% fuel assemblies

International Nuclear Information System (INIS)

Richard, R.F.

1997-01-01

An FFTF tritium/isotope production mission will require a new fuel supply. The reference design core will use a mixed oxide fuel nominally enriched to 40 wt% Pu. This enrichment is significantly higher than that of the standard Driver Fuel Assemblies used in past operations. Consequently, criticality safety for handling and storage of this fuel must be addressed. The purpose of this document is to begin the process by determining the minimum critical number for these new fuel assemblies in water, sodium and air. This analysis is preliminary and further work can be done to refine the results reported here. Analysis was initially done using 45 wt 5 PuO. Additionally, a preliminary assessment is done concerning storage of these fuel assemblies in Interim Decay Storage (IDS), Fuel Storage Facility (FSF), and Core Component Containers/Interim Storage Casks (CCC/ISC)

Taking interim actions: Integrating CERCLA and NEPA to move ahead with site cleanup

International Nuclear Information System (INIS)

MacDonell, M.M.; Peterson, J.M.; Valett, G.L.; McCracken, S.H.

1991-01-01

The cleanup of contaminated sites can be expedited by using interim response actions in accordance with the Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA), as amended, and the National Oil and Hazardous Substances Pollution Contingency Plan (NCP). In fact, a major portion of some Superfund sites can be cleaned up using interim actions. For CERCLA sites being remediated by the US Department of Energy (DOE), such actions must also comply with the National Environmental Policy Act (NEPA) because the DOE has established a policy for integrating CERCLA and NEPA requirements. A strategy for the integrated documentation with implementation of interim actions has been applied successfully at the Weldon Spring site, and major cleanup projects are currently underway. This paper discusses some of the issues associated with integrating CERCLA and NEPA for interim actions and summarizes those actions that have been identified for the Weldon Spring site

Taking interim actions: Integrating CERCLA and NEPA to move ahead with site cleanup

International Nuclear Information System (INIS)

MacDonell, M.M.; Peterson, J.M.; Valett, G.L.; McCracken, S.H.

1991-01-01

The cleanup of contaminated sites can be expedited by using interim response actions in accordance with the Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA), as amended, and the National Oil and Hazardous Substances Pollution Contingency Plan (NCP). In fact, a major portion of some Superfund sites can be cleaned up using interim actions. For CERCLA sites being remediated by the US Department of Energy (DOE), such actions must also comply with the National Environmental Policy Act (NEPA) because the DOE has established a policy for integrating CERCLA and NEPA requirements. A strategy for the integrated documentation and implementation of interim actions has been applied successfully at the Weldon Spring site, and major cleanup projects are currently underway. This paper discusses some of the issues associated with integrating CERCLA and NEPA for interim actions and summarizes those actions that have been identified for the Weldon Spring site

Review of Policy Documents for Nuclear Safety and Regulation

International Nuclear Information System (INIS)

Kim, Woong Sik; Choi, Kwang Sik; Choi, Young Sung; Kim, Hho Jung; Kim, Ho Ki

2006-01-01

The goal of regulation is to protect public health and safety as well as environment from radiological hazards that may occur as a result of the use of atomic energy. In September 1994, the Korean government issued the Nuclear Safety Policy Statement (NSPS) to establish policy goals of maintaining and achieving high-level of nuclear safety and also help the public understand the national policy and a strong will of the government toward nuclear safety. It declares the importance of establishing safety culture in nuclear community and also specifies five nuclear regulatory principles (Independence, Openness, Clarity, Efficiency and Reliability) and provides the eleven regulatory policy directions. In 2001, the Nuclear Safety Charter was declared to make the highest goal of safety in driving nuclear business clearer; to encourage atomic energy- related institutions and workers to keep in mind the mission and responsibility for assuring safety; to guarantee public confidence in related organizations. The Ministry of Science and Technology (MOST) also issues Yearly Regulatory Policy Directions at the beginning of every year. Recently, the third Atomic Energy Promotion Plan (2007-2011) has been established. It becomes necessary for the relevant organizations to prepare the detailed plans on such areas as nuclear development, safety management, regulation, etc. This paper introduces a multi-level structure of nuclear safety and regulation policy documents in Korea and presents some improvements necessary for better application of the policies

Review of Policy Documents for Nuclear Safety and Regulation

Energy Technology Data Exchange (ETDEWEB)

Kim, Woong Sik; Choi, Kwang Sik; Choi, Young Sung; Kim, Hho Jung; Kim, Ho Ki [Korea Institute of Nuclear Safety, Taejon (Korea, Republic of)

2006-07-01

The goal of regulation is to protect public health and safety as well as environment from radiological hazards that may occur as a result of the use of atomic energy. In September 1994, the Korean government issued the Nuclear Safety Policy Statement (NSPS) to establish policy goals of maintaining and achieving high-level of nuclear safety and also help the public understand the national policy and a strong will of the government toward nuclear safety. It declares the importance of establishing safety culture in nuclear community and also specifies five nuclear regulatory principles (Independence, Openness, Clarity, Efficiency and Reliability) and provides the eleven regulatory policy directions. In 2001, the Nuclear Safety Charter was declared to make the highest goal of safety in driving nuclear business clearer; to encourage atomic energy- related institutions and workers to keep in mind the mission and responsibility for assuring safety; to guarantee public confidence in related organizations. The Ministry of Science and Technology (MOST) also issues Yearly Regulatory Policy Directions at the beginning of every year. Recently, the third Atomic Energy Promotion Plan (2007-2011) has been established. It becomes necessary for the relevant organizations to prepare the detailed plans on such areas as nuclear development, safety management, regulation, etc. This paper introduces a multi-level structure of nuclear safety and regulation policy documents in Korea and presents some improvements necessary for better application of the policies.

A refined safety analysis approach for closure of the Hanford Site flammable gas unreviewed safety question

International Nuclear Information System (INIS)

Bratzel, D.R.

1997-01-01

Following a 1990 investigation into flammable gas generation, retention, and release mechanisms within the Hanford Site high-level waste tanks, personnel concluded that the existing Authorization Basis documentation did not adequately evaluate flammable gas hazards. This declaration was based primarily on the fact that personnel did not adequately consider hydrogen and nitrous oxide evolution within the material in certain waste tanks and subsequent hypothetical ignition in the development of safety documentation for the waste tanks. The US Department of Energy-Headquarters subsequently declared an Unreviewed Safety Question (USQ). Although work scope has been focused on closure of the USQ since 1990, the DOE has yet to close the USQ because of considerable uncertainty regarding essential technical parameters and associated risk. The DOE recently approved a Basis for Interim Operation to revise the Authorization Basis for managing the tank farms, however, the USQ remains open. The two fundamental requirements for closure of the flammable gas USQ are as follows: development of a defensible technical basis for existing controls; development of a process to assess the adequacy of controls as the waste tank mission progresses

33 CFR 96.250 - What documents and reports must a safety management system have?

Science.gov (United States)

2010-07-01

... safety management system have? 96.250 Section 96.250 Navigation and Navigable Waters COAST GUARD, DEPARTMENT OF HOMELAND SECURITY VESSEL OPERATING REGULATIONS RULES FOR THE SAFE OPERATION OF VESSELS AND SAFETY MANAGEMENT SYSTEMS Company and Vessel Safety Management Systems § 96.250 What documents and...

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

CP-50 calibration facility radiological safety assessment document

International Nuclear Information System (INIS)

Chilton, M.W.; Hill, R.L.; Eubank, B.F.

1980-03-01

The CP-50 Calibration Facility Radiological Safety Assessment document, prepared at the request of the Nevada Operations Office of the US Department of Energy to satisfy provisions of ERDA Manual Chapter 0531, presents design features, systems controls, and procedures used in the operation of the calibration facility. Site and facility characteristics and routine and non-routine operations, including hypothetical incidents or accidents are discussed and design factors, source control systems, and radiation monitoring considerations are described

AGR-1 Data Qualification Interim Report

International Nuclear Information System (INIS)

Abbott, Machael

2009-01-01

Projects for the very-high-temperature reactor (VHTR) program provide data in support of Nuclear Regulatory Commission licensing of the VHTR. Fuel and materials to be used in the reactor are tested and characterized to quantify performance in high temperature and high fluence environments. The VHTR Program has established the NGNP Data Management and Analysis System (NDMAS) to ensure that VHTR data are (1) qualified for use, (2) stored in a readily accessible electronic form, and (3) analyzed to extract useful results. This document focuses on the first NDMAS objective. It describes the data streams associated with the first Advanced Gas Reactor (AGR-1) experiment, the processing of these data within NDMAS, and reports the interim FY09 qualification status of the AGR-1 data to date. Data qualification activities within NDMAS for specific types of data are determined by the data qualification category, which is assigned by the data generator, and include: (1) capture testing, to confirm that the data stored within NDMAS are identical to the raw data supplied, (2) accuracy testing, to confirm that the data are an accurate representation of the system or object being measured, and (3) documentation that the data were collected under an NQA-1 or equivalent QA program. The interim qualification status of the following four data streams is reported in this document: (1) fuel fabrication data, (2) fuel irradiation data, (3) fission product monitoring system (FPMS) data, and (4) Advanced Test Reactor (ATR) operating conditions data. A final report giving the NDMAS qualification status of all AGR-1 data (including cycle 145A) is planned for February 2010

Interim Stabilization Equipment Essential and Support Drawing Plan

International Nuclear Information System (INIS)

KOCH, M.R.

1999-01-01

The purpose of this document is to list the Interim Stabilization equipment drawings that are classified as Essential or Support drawings. Essential Drawings: Those drawings identified by the facility staff as necessary to directly support the safe operation of the facility or equipment. Support Drawings: Those drawings identified by the facility staff that further describe the design details of structures, systems or components shown on essential drawings

Safety and Effectiveness of Natalizumab: First Report of Interim Results of Post-Marketing Surveillance in Japan.

Science.gov (United States)

Saida, Takahiko; Yokoyama, Kazumasa; Sato, Ryusuke; Makioka, Haruki; Iizuka, Yukihiko; Hase, Masakazu; Ling, Yan; Torii, Shinichi

2017-12-01

Natalizumab, a humanized anti-α4 integrin monoclonal antibody, received marketing approval in Japan in 2014 for the treatment of multiple sclerosis (MS). Because the previous large-scale clinical trials of natalizumab were mainly conducted in Europe and North American countries, and data in patients with MS from Japan were limited, we conducted an all-case post-marketing surveillance of natalizumab-treated MS patients from Japan to investigate the safety and effectiveness of natalizumab in a real-world clinical setting in Japan. Here, we report the results of an interim analysis. During the observation period of 2 years, all patients who were treated with natalizumab subsequent to its approval in Japan were followed. The effectiveness of natalizumab was assessed by examining the changes in expanded disability status scale (EDSS) score and annualized relapse rate (ARR) from baseline. Safety was assessed by analyzing the incidence of adverse drug reactions (ADRs). The safety analysis included 106 patients (mean age 39.3 years; women 62.3%) whose data were collected until the data lock point (February 7, 2016). The effectiveness analysis included 75 patients. The majority of patients had relapsing-remitting MS (93/106 patients; 87.7%). The mean length of treatment exposure in the present study was 6.6 months. During the 2-year observation period, no significant change in the EDSS was observed, while the ARR decreased significantly from baseline (72.9% reduction, p = 0.001). ADRs and serious ADRs were observed in 11.3% and 3.8% of patients, respectively; however, no new safety concerns were detected. No patient had progressive multifocal leukoencephalopathy (PML) during the present study period. The safety and effectiveness of natalizumab were confirmed in Japanese patients with MS in clinical practice. Nevertheless, potential risks including PML require continuous, careful observation. Biogen Japan Ltd (Tokyo, Japan).

Environmental assessment for 881 Hillside (High Priority Sites) interim remedial action

International Nuclear Information System (INIS)

1990-01-01

This Environmental Assessment evaluates the impact of an interim remedial action proposed for the High Priority Sites (881 Hillside Area) at the Rocky Flats Plant (RFP). This interim action is to be conducted to minimize the release of hazardous substances from the 881 Hillside Area that pose a potential long-term threat to public health and the environment. This document integrates current site characterization data and environmental analyses required by the Comprehensive Environmental Response, Compensation and Liability Act (CERCLA) or ''Superfund'' process, into an environmental assessment pursuant to the National Environmental Policy Act (NEPA). Characterization of the 881 Hillside Area is continuing. Consequently, a final remedial action has not yet been proposed. Environmental impacts associated with the proposed interim remedial action and reasonable alternatives designed to remove organic and inorganic contaminants, including radionuclides, from alluvial groundwater in the 881 Hillside Area are addressed. 24 refs., 5 figs., 23 tabs

Immobilized High-Level Waste (HLW) Interim Storage Alternative Generation and analysis and Decision Report - second Generation Implementing Architecture

International Nuclear Information System (INIS)

CALMUS, R.B.

2000-01-01

Two alternative approaches were previously identified to provide second-generation interim storage of Immobilized High-Level Waste (IHLW). One approach was retrofit modification of the Fuel and Materials Examination Facility (FMEF) to accommodate IHLW. The results of the evaluation of the FMEF as the second-generation IHLW interim storage facility and subsequent decision process are provided in this document

Interim Design Report

CERN Document Server

Choubey, S.; Goswami, S.; Berg, J.S.; Fernow, R.; Gallardo, J.C.; Gupta, R.; Kirk, H.; Simos, N.; Souchlas, N.; Ellis, M.; Kyberd, P.; Benedetto, E.; Fernandez-Martinez, E.; Efthymiopoulos, I.; Garoby, R.; Gilardoni, S.; Martini, M.; Prior, G.; Ballett, P.; Pascoli, S.; Bross, A.; Geer, S.; Johnstone, C.; Kopp, J.; Mokhov, N.; Morfin, J.; Neuffer, D.; Parke, S.; Popovic, M.; Strait, J.; Striganov, S.; Blondel, A.; Dufour, F.; Laing, A.; Soler, F.J.P; Lindner, M.; Schwetz, T.; Alekou, A.; Apollonio, M.; Aslaninejad, M.; Bontoiu, C.; Dornan, P.; Eccleston, R.; Kurup, A.; Long, K.; Pasternak, J.; Pozimski, J.; Bogacz, A.; Morozov, V.; Roblin, Y.; Bhattacharya, S.; Majumdar, D.; Mori, Y.; Planche, T.; Zisman, M.; Cline, D.; Stratakis, D.; Ding, X.; Coloma, P.; Donini, A.; Gavela, B.; Lopez Pavon, J.; Maltoni, M.; Bromberg, C.; Bonesini, M.; Hart, T.; Kudenko, Y.; Mondal, N.; Antusch, S.; Blennow, M.; Ota, T.; Abrams, R.J.; Ankenbrandt, C.M.; Beard, K.B.; Cummings, M.A.C.; Flanagan, G.; Johnson, R.P.; Roberts, T.J.; Yoshikawa, C.Y.; Migliozzi, P.; Palladino, V.; de Gouvea, A.; Graves, V.B.; Kuno, Y.; Peltoniemi, J.; Blackmore, V.; Cobb, J.; Witte, H.; Mezzetto, M.; Rigolin, S.; McDonald, K.T.; Coney, L.; Hanson, G.; Snopok, P.; Tortora, L.; Andreopoulos, C.; Bennett, J.R.J.; Brooks, S.; Caretta, O.; Davenne, T.; Densham, C.; Edgecock, R.; Kelliher, D.; Loveridge, P.; McFarland, A.; Machida, S.; Prior, C.; Rees, G.; Rogers, C.; Thomason, J.W.G.; Booth, C.; Skoro, G.; Karadzhov, Y.; Matev, R.; Tsenov, R.; Samulyak, R.; Mishra, S.R.; Petti, R.; Dracos, M.; Yasuda, O.; Agarwalla, S.K.; Cervera-Villanueva, A.; Gomez-Cadenas, J.J.; Hernandez, P.; Li, T.; Martin-Albo, J.; Huber, P.; Back, J.; Barker, G.; Harrison, P.; Meloni, D.; Tang, J.; Winter, W.

2011-01-01

The International Design Study for the Neutrino Factory (the IDS-NF) was established by the community at the ninth "International Workshop on Neutrino Factories, super-beams, and beta- beams" which was held in Okayama in August 2007. The IDS-NF mandate is to deliver the Reference Design Report (RDR) for the facility on the timescale of 2012/13. In addition, the mandate for the study [3] requires an Interim Design Report to be delivered midway through the project as a step on the way to the RDR. This document, the IDR, has two functions: it marks the point in the IDS-NF at which the emphasis turns to the engineering studies required to deliver the RDR and it documents baseline concepts for the accelerator complex, the neutrino detectors, and the instrumentation systems. The IDS-NF is, in essence, a site-independent study. Example sites, CERN, FNAL, and RAL, have been identified to allow site-specific issues to be addressed in the cost analysis that will be presented in the RDR. The choice of example sites shou...

Interim-status groundwater monitoring plan for the 216-B-63 trench

Energy Technology Data Exchange (ETDEWEB)

Sweeney, M.D.

1995-02-09

This document outlines the groundwater monitoring plan, under RCRA regulations in 40 CFR 265 Subpart F and WAC173-300-400, for the 216-B-63 Trench. This interim status facility is being sampled under detection monitoring criteria and this plan provides current program conditions and requirements.

Methods for assessing environmental impacts of a FUSRAP property-cleanup/interim-storage remedial action

International Nuclear Information System (INIS)

Wyman, D.J.

1982-12-01

This document provides a description of a property-cleanup/interim-storage action, explanation of how environmental impacts might occur, comprehensive treatment of most potential impacts that might occur as a result of this type of action, discussion of existing methodologies for estimating and assessing impacts, justification of the choice of specific methodologies for use in FUSRAP environmental reviews, assessments of representative impacts (or expected ranges of impacts where possible), suggested mitigation measures, and some key sources of information. The major topical areas covered are physical and biological impacts, radiological impacts, and socioeconomic impacts. Some project-related issues were beyond the scope of this document, including dollar costs, specific accident scenarios, project funding and changes in Congressional mandates, and project management (contracts, labor relations, quality assurance, liability, emergency preparedness, etc.). These issues will be covered in other documents supporting the decision-making process. Although the scope of this document covers property-cleanup and interim-storage actions, it is applicable to other similar remedial actions. For example, the analyses discussed herein for cleanup activities are applicable to any FUSRAP action that includes site cleanup

Waste Management System Description Document (WMSD)

International Nuclear Information System (INIS)

1992-02-01

This report is an appendix of the ''Waste Management Description Project, Revision 1''. This appendix is about the interim approach for the technical baseline of the waste management system. It describes the documentation and regulations of the waste management system requirements and description. (MB)

Health Information Technology, Patient Safety, and Professional Nursing Care Documentation in Acute Care Settings.

Science.gov (United States)

Lavin, Mary Ann; Harper, Ellen; Barr, Nancy

2015-04-14

The electronic health record (EHR) is a documentation tool that yields data useful in enhancing patient safety, evaluating care quality, maximizing efficiency, and measuring staffing needs. Although nurses applaud the EHR, they also indicate dissatisfaction with its design and cumbersome electronic processes. This article describes the views of nurses shared by members of the Nursing Practice Committee of the Missouri Nurses Association; it encourages nurses to share their EHR concerns with Information Technology (IT) staff and vendors and to take their place at the table when nursing-related IT decisions are made. In this article, we describe the experiential-reflective reasoning and action model used to understand staff nurses' perspectives, share committee reflections and recommendations for improving both documentation and documentation technology, and conclude by encouraging nurses to develop their documentation and informatics skills. Nursing issues include medication safety, documentation and standards of practice, and EHR efficiency. IT concerns include interoperability, vendors, innovation, nursing voice, education, and collaboration.

List of documents received by the INDC Secretariat

International Nuclear Information System (INIS)

1980-06-01

This list is produced directly from computer printout in two sorts: one ordered by accession number, and the other ordered by document number within each origin series (e.g. listing all INDC(SEC)-documents in one block). Reference to earlier INDSWG and ''interim'' INDC reports received between 1962 and 1967 are listed in report INDC/199 (dated November 1967)

Spent fuel interim storage

International Nuclear Information System (INIS)

Bilegan, Iosif C.

2003-01-01

The official inauguration of the spent fuel interim storage took place on Monday July 28, 2003 at Cernavoda NNP. The inaugural event was attended by local and central public authority representatives, a Canadian Government delegation as well as newsmen from local and central mass media and numerous specialists from Cernavoda NPP compound. Mr Andrei Grigorescu, State Secretary with the Economy and Commerce Ministry, underlined in his talk the importance of this objective for the continuous development of nuclear power in Romania as well as for Romania's complying with the EU practice in this field. Also the excellent collaboration between the Canadian contractor AECL and the Romanian partners Nuclear Montaj, CITON, UTI, General Concret in the accomplishment of this unit at the planned terms and costs. On behalf of Canadian delegation, spoke Minister Don Boudria. He underlined the importance which the Canadian Government affords to the cooperation with Romania aiming at specific objectives in the field of nuclear power such as the Cernavoda NPP Unit 2 and spent fuel interim storage. After traditional cutting of the inaugural ribbon by the two Ministers the festivities continued on the Cernavoda NPP Compound with undersigning the documents regarding the project completion and a press conference

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

International Nuclear Information System (INIS)

2000-01-01

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

1987 Federal interim storage fee study: A technical and economic analysis

International Nuclear Information System (INIS)

1987-09-01

This document is the latest in a series of reports that are published annually by Pacific Northwest Laboratory (PNL) for the US Department of Energy (DOE). This information in the report, which was prepared by E.R. Johnson Associates under subcontract to PNL, will be used by the DOE to establish a payment schedule for interim storage of spent nuclear fuel under the Federal Interim Storage (FIS) Program, which was mandated by the Nuclear Waste Policy Act of 1982. The information in this report will be used to establish the schedule of charges for FIS services for the year commencing January 1, 1988. 13 tabs

1987 Federal interim storage fee study: A technical and economic analysis

Energy Technology Data Exchange (ETDEWEB)

1987-09-01

This document is the latest in a series of reports that are published annually by Pacific Northwest Laboratory (PNL) for the US Department of Energy (DOE). This information in the report, which was prepared by E.R. Johnson Associates under subcontract to PNL, will be used by the DOE to establish a payment schedule for interim storage of spent nuclear fuel under the Federal Interim Storage (FIS) Program, which was mandated by the Nuclear Waste Policy Act of 1982. The information in this report will be used to establish the schedule of charges for FIS services for the year commencing January 1, 1988. 13 tabs.

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'

Safety and Efficacy of Teneligliptin in Patients with Type 2 Diabetes Mellitus and Impaired Renal Function: Interim Report from Post-marketing Surveillance.

Science.gov (United States)

Haneda, Masakazu; Kadowaki, Takashi; Ito, Hiroshi; Sasaki, Kazuyo; Hiraide, Sonoe; Ishii, Manabu; Matsukawa, Miyuki; Ueno, Makoto

2018-06-01

Teneligliptin is a novel oral dipeptidyl peptidase-4 inhibitor for the treatment of type 2 diabetes mellitus (T2DM). Safety and efficacy of teneligliptin have been demonstrated in clinical studies; however, data supporting its use in patients with moderate or severe renal impairment are limited. This interim analysis of a post-marketing surveillance of teneligliptin, exploRing the long-term efficacy and safety included cardiovascUlar events in patients with type 2 diaBetes treated bY teneligliptin in the real-world (RUBY), aims to verify the long-term safety and efficacy of teneligliptin in Japanese patients with T2DM and impaired renal function. For this analysis, we used the data from case report forms of the RUBY surveillance between May 2013 and June 2017. The patients were classified into G1-G5 stages of chronic kidney disease according to estimated glomerular filtration rate (eGFR) at initiation of teneligliptin treatment. Safety and efficacy were evaluated in these subgroups. Patients on dialysis were also assessed. Safety was assessed from adverse drug reactions (ADRs). Glycemic control was evaluated up to 2 years after teneligliptin initiation. A total of 11,677 patients were enrolled in the surveillance and 11,425 patient case-report forms were collected for the interim analysis. The incidence of ADRs in each subgroup was 2.98-6.98% of patients, with no differences in the ADR profile (including hypoglycemia and renal function ADRs) between subgroups. At 1 and 2 years after starting teneligliptin, the least-squares mean change in HbA1c adjusted to the baseline was - 0.68 to - 0.85% and - 0.71 to - 0.85% across the eGFR groups, respectively. Treatment with teneligliptin in patients on dialysis reduced or tended to reduce glycated albumin levels [- 2.29%, (p < 0.001) after 1 year; - 1.64%, (p = 0.064) after 2 years]. During long-term treatment, teneligliptin was generally well tolerated in patients with any stage of renal impairment from

National Ignition Facility Cryogenic Target Systems Interim Management Plan

International Nuclear Information System (INIS)

Warner, B

2002-01-01

Restricted availability of funding has had an adverse impact, unforeseen at the time of the original decision to projectize the National Ignition Facility (NIF) Cryogenic Target Handling Systems (NCTS) Program, on the planning and initiation of these efforts. The purpose of this document is to provide an interim project management plan describing the organizational structure and management processes currently in place for NCTS. Preparation of a Program Execution Plan (PEP) for NCTS has been initiated, and a current draft is provided as Attachment 1 to this document. The National Ignition Facility is a multi-megajoule laser facility being constructed at Lawrence Livermore National Laboratory (LLNL) by the National Nuclear Security Administration (NNSA) in the Department of Energy (DOE). Its primary mission is to support the Stockpile Stewardship Program (SSP) by performing experiments studying weapons physics, including fusion ignition. NIF also supports the missions of weapons effects, inertial fusion energy, and basic science in high-energy-density physics. NIF will be operated by LLNL under contract to the University of California (UC) as a national user facility. NIF is a low-hazard, radiological facility, and its operation will meet all applicable federal, state, and local Environmental Safety and Health (ES and H) requirements. The NCTS Interim Management Plan provides a summary of primary design criteria and functional requirements, current organizational structure, tracking and reporting procedures, and current planning estimates of project scope, cost, and schedule. The NIF Director controls the NIF Cryogenic Target Systems Interim Management Plan. Overall scope content and execution schedules for the High Energy Density Physics Campaign (SSP Campaign 10) are currently undergoing rebaselining and will be brought into alignment with resources expected to be available throughout the NNSA Future Years National Security Plan (FYNSP). The revised schedule for

Documented Safety Analysis for the B695 Segment

Energy Technology Data Exchange (ETDEWEB)

Laycak, D

2008-09-11

This Documented Safety Analysis (DSA) was prepared for the Lawrence Livermore National Laboratory (LLNL) Building 695 (B695) Segment of the Decontamination and Waste Treatment Facility (DWTF). The report provides comprehensive information on design and operations, including safety programs and safety structures, systems and components to address the potential process-related hazards, natural phenomena, and external hazards that can affect the public, facility workers, and the environment. Consideration is given to all modes of operation, including the potential for both equipment failure and human error. The facilities known collectively as the DWTF are used by LLNL's Radioactive and Hazardous Waste Management (RHWM) Division to store and treat regulated wastes generated at LLNL. RHWM generally processes low-level radioactive waste with no, or extremely low, concentrations of transuranics (e.g., much less than 100 nCi/g). Wastes processed often contain only depleted uranium and beta- and gamma-emitting nuclides, e.g., {sup 90}Sr, {sup 137}Cs, or {sup 3}H. The mission of the B695 Segment centers on container storage, lab-packing, repacking, overpacking, bulking, sampling, waste transfer, and waste treatment. The B695 Segment is used for storage of radioactive waste (including transuranic and low-level), hazardous, nonhazardous, mixed, and other waste. Storage of hazardous and mixed waste in B695 Segment facilities is in compliance with the Resource Conservation and Recovery Act (RCRA). LLNL is operated by the Lawrence Livermore National Security, LLC, for the Department of Energy (DOE). The B695 Segment is operated by the RHWM Division of LLNL. Many operations in the B695 Segment are performed under a Resource Conservation and Recovery Act (RCRA) operation plan, similar to commercial treatment operations with best demonstrated available technologies. The buildings of the B695 Segment were designed and built considering such operations, using proven building

Documented Safety Analysis for the B695 Segment

International Nuclear Information System (INIS)

Laycak, D.

2008-01-01

This Documented Safety Analysis (DSA) was prepared for the Lawrence Livermore National Laboratory (LLNL) Building 695 (B695) Segment of the Decontamination and Waste Treatment Facility (DWTF). The report provides comprehensive information on design and operations, including safety programs and safety structures, systems and components to address the potential process-related hazards, natural phenomena, and external hazards that can affect the public, facility workers, and the environment. Consideration is given to all modes of operation, including the potential for both equipment failure and human error. The facilities known collectively as the DWTF are used by LLNL's Radioactive and Hazardous Waste Management (RHWM) Division to store and treat regulated wastes generated at LLNL. RHWM generally processes low-level radioactive waste with no, or extremely low, concentrations of transuranics (e.g., much less than 100 nCi/g). Wastes processed often contain only depleted uranium and beta- and gamma-emitting nuclides, e.g., 90 Sr, 137 Cs, or 3 H. The mission of the B695 Segment centers on container storage, lab-packing, repacking, overpacking, bulking, sampling, waste transfer, and waste treatment. The B695 Segment is used for storage of radioactive waste (including transuranic and low-level), hazardous, nonhazardous, mixed, and other waste. Storage of hazardous and mixed waste in B695 Segment facilities is in compliance with the Resource Conservation and Recovery Act (RCRA). LLNL is operated by the Lawrence Livermore National Security, LLC, for the Department of Energy (DOE). The B695 Segment is operated by the RHWM Division of LLNL. Many operations in the B695 Segment are performed under a Resource Conservation and Recovery Act (RCRA) operation plan, similar to commercial treatment operations with best demonstrated available technologies. The buildings of the B695 Segment were designed and built considering such operations, using proven building systems, and keeping

Czech interim spent fuel storage facility: operation experience, inspections and future plans

International Nuclear Information System (INIS)

Fajman, V.; Bartak, L.; Coufal, J.; Brzobohaty, K.; Kuba, S.

1999-01-01

The paper describes the situation in the spent fuel management in the Czech Republic. The interim Spent Fuel Storage Facility (ISFSF) at Dukovany, which was commissioned in January 1997 and is using dual transport and storage CASTOR - 440/84 casks, is briefly described. The authors deal with their experience in operating and inspecting the ISFSF Dukovany. The structure of the basic safety document 'Limits and Conditions of Normal Operation' is also mentioned, including the experience of the performance. The inspection activities focused on permanent checking of the leak tightness of the CASTOR 440/84 casks, the maximum cask temperature and inspections monitoring both the neutron and gamma dose rate as well as the surface contamination. The results of the inspections are mentioned in the presentation as well. The operator's experience with re-opening partly loaded and already dried CASTOR-440/84 cask, after its transport from NPP Jaslovske Bohunice to the NPP Dukovany is also described. The paper introduces briefly the concept of future spent fuel storage both from the NPP Dukovany and the NPP Temelin, as prepared by the CEZ. The preparatory work for the Central Interim Spent Nuclear Fuel Storage Facility (CISFSF) in the Czech Republic and the information concerning the planned storage technology for this facility is discussed in the paper as well. The authors describe the site selection process and the preparatory steps concerning new spent fuel facility construction including the Environmental Impact Assessment studies. (author)

Relative risk measure suitable for comparison of design alternatives of interim spent nuclear fuel storage facility

International Nuclear Information System (INIS)

Ferjencik, M.

1997-01-01

Accessible reports on risk assessment of interim spent nuclear fuel storage facilities presume that only releases of radioactive substances represent undesired consequences. However, only certain part of the undesired consequences is represented by them. Many other events are connected with safety and are able to cause losses to the operating company. The following two presumptions are pronounced based on this. 1. Any event causing a disturbance of a safety function of the storage facility is an incident event. 2. Any disturbance of a safety function is an undesired consequence. If the facility safety functions are identified and if the severity of their disturbances is quantified, then it is possible to combine consequence severity quantifications and event frequencies into a risk measure. Construction and application of such a risk measure is described in this paper. The measure is shown to be a tool suitable for comparison of interim storage technology design alternatives. (author)

Waste Management System Requirements Document

International Nuclear Information System (INIS)

1992-02-01

This DCP establishes an interim plan for the Office of Civilian Radioactive Waste Management (OCRWM) technical baseline until the results of the OCRWM Document Hierarchy Task Force can be implemented. This plan is needed to maintain continuity in the Program for ongoing work in the areas of Waste Acceptance, Transportation, Monitored Retrievable Storage (MRS) and Yucca Mountain Site Characterization

Interim Stabilization Equipment Essential and Support Drawing Plan

International Nuclear Information System (INIS)

HORNER, T.M.

2000-01-01

The purpose of this document is to list the Interim Stabilization equipment drawings that are classified as Essential or Support drawings. Essential Drawings are those drawings identified by the facility staff as necessary to directly support the safe operation of the facility or equipment. [CHG 2000a]. Support Drawings are those drawings identified by the facility staff that further describe the design details of structures, systems or components shown on essential drawings. [CHG 2000a

Status of High Flux Isotope Reactor (HFIR) post-restart safety analysis and documentation upgrades

International Nuclear Information System (INIS)

Cook, D.H.; Radcliff, T.D.; Rothrock, R.B.; Schreiber, R.E.

1990-01-01

The High Flux Isotope Reactor (HFIR), an experimental reactor located at the Oak Ridge National Laboratory (ORNL) and operated for the US Department of Energy by Martin Marietta Energy Systems, was shut down in November, 1986 after the discovery of unexpected neutron embrittlement of the reactor vessel. The reactor was restarted in April, 1989, following an extensive review by DOE and ORNL of the HFIR design, safety, operation, maintenance and management, and the implementation of several upgrades to HFIR safety-related hardware, analyses, documents and procedures. This included establishing new operating conditions to provide added margin against pressure vessel failure, as well as the addition, or upgrading, of specific safety-related hardware. This paper summarizes the status of some of the follow-on (post-restart) activities which are currently in progress, and which will result in a comprehensive set of safety analyses and documentation for the HFIR, comparable with current practice in commercial nuclear power plants. 8 refs

Analysis of compatibility of current Czech initial documentation in the area of technical assurance of nuclear safety with the requirements of the EUR document

International Nuclear Information System (INIS)

Zdebor, J.; Zdebor, R.; Kratochvil, L.

2001-11-01

The publication is structured as follows: Description of existing documentation. General requirements, goals, principles and design principles: Documents being compared; Method of comparison; Results and partial evaluation of comparison of requirements between EUR and Czech regulations (basic goals and safety philosophy; quantitative safety objectives; basic design requirements; extended design requirements; external and internal threats; technical requirements; site conditions); Summary of the comparison of safety requirements. Comparison of requirements for the systems: Requirements for the nuclear reactor unit systems; Barrier systems (fuel system; reactor cooling system; containment system); Remaining systems (control systems; protection systems; coolant makeup and purification system; residual heat removal system; emergency cooling system; power systems); Common technical requirements for systems (technical requirements for systems; internal and external events). (P.A.)

THE FORMATION OF THE CONTOUR OF THE DOCUMENTED AND REAL FLIGHT SAFETY IN THE SYSTEM OF THE INFORMATION PROVISION OF SAFETY OF FLIGHTS

Directory of Open Access Journals (Sweden)

B. I. Bachkalo

2015-01-01

Full Text Available The article discusses the principles and mechanisms of formation of the contour of the real safety of flights and contour of the documented safety, allowing us to obtain information to control fligh safety. The proposed approach can be used in the algorithms of active on-board flight safety management system for the implementation of information support to the crew in flight and automatic control of flight safety.

Conceptual waste package interim product specifications and data requirements for disposal of borosilicate glass defense high-level waste forms in salt geologic repositories

International Nuclear Information System (INIS)

1983-06-01

The conceptual waste package interim product specifications and data requirements presented are applicable specifically to the normal borosilicate glass product of the Defense Waste Processing Facility (DWPF). They provide preliminary numerical values for the defense high-level waste form parameters and properties identified in the waste form performance specification for geologic isolation in salt repositories. Subject areas treated include containment and isolation, operational period safety, criticality control, waste form/production canister identification, and waste package performance testing requirements. This document was generated for use in the development of conceptual waste package designs in salt. It will be revised as additional data, analyses, and regulatory requirements become available

Planning Document for an NBSR Conversion Safety Analysis Report

Energy Technology Data Exchange (ETDEWEB)

Diamond D. J.; Baek J.; Hanson, A.L.; Cheng, L-Y.; Brown, N.; Cuadra, A.

2013-09-25

The NIST Center for Neutron Research (NCNR) is a reactor-laboratory complex providing the National Institute of Standards and Technology (NIST) and the nation with a world-class facility for the performance of neutron-based research. The heart of this facility is the National Bureau of Standards Reactor (NBSR). The NBSR is a heavy water moderated and cooled reactor operating at 20 MW. It is fueled with high-enriched uranium (HEU) fuel elements. A Global Threat Reduction Initiative (GTRI) program is underway to convert the reactor to low-enriched uranium (LEU) fuel. This program includes the qualification of the proposed fuel, uranium and molybdenum alloy foil clad in an aluminum alloy, and the development of the fabrication techniques. This report is a planning document for the conversion Safety Analysis Report (SAR) that would be submitted to, and approved by, the Nuclear Regulatory Commission (NRC) before the reactor could be converted.This report follows the recommended format and content from the NRC codified in NUREG-1537, “Guidelines for Preparing and Reviewing Applications for the Licensing of Non-power Reactors,” Chapter 18, “Highly Enriched to Low-Enriched Uranium Conversions.” The emphasis herein is on the SAR chapters that require significant changes as a result of conversion, primarily Chapter 4, Reactor Description, and Chapter 13, Safety Analysis. The document provides information on the proposed design for the LEU fuel elements and identifies what information is still missing. This document is intended to assist ongoing fuel development efforts, and to provide a platform for the development of the final conversion SAR. This report contributes directly to the reactor conversion pillar of the GTRI program, but also acts as a boundary condition for the fuel development and fuel fabrication pillars.

Second interim assessment of the Canadian concept for nuclear fuel waste disposal. Volume 3

International Nuclear Information System (INIS)

Johansen, K.; Donnelly, K.J.; Gee, J.H.; Green, B.J.; Nathwani, J.S.; Quinn, A.M.; Rogers, B.G.; Stevenson, M.A.; Dunford, W.E.; Tamm, J.A.

1985-12-01

The nuclear fuel waste disposal concept chosen for development and assessment in Canada involves the isolation of corrosion-resistant containers of waste in a vault located deep in plutonic rock. As the concept and the assessment tools are developed, periodic assessments are performed to permit evaluation of the methodology and provide feedback to those developing the concept. The ultimate goal of these assessments is to predict what impact the disposal system would have on man and the environment if the concept were implemented. The second such assessment was completed in 1984 and is documented in the Second Interim Assessment of the Canadian Concept for Nuclear Fuel Waste Disposal - Volumes 1-4. This, the third volume of the report, summarizes the pre-closure environmental and safety assessments completed by Ontario Hydro for Atomic Energy of Canada Limited. The preliminary results and their sigificance are discussed. 85 refs

Computerising documentation

International Nuclear Information System (INIS)

Anon.

1992-01-01

The nuclear power generation industry is faced with public concern and government pressures over safety, efficiency and risk. Operators throughout the industry are addressing these issues with the aid of a new technology - technical document management systems (TDMS). Used for strategic and tactical advantage, the systems enable users to scan, archive, retrieve, store, edit, distribute worldwide and manage the huge volume of documentation (paper drawings, CAD data and film-based information) generated in building, maintaining and ensuring safety in the UK's power plants. The power generation industry has recognized that the management and modification of operation critical information is vital to the safety and efficiency of its power plants. Regulatory pressure from the Nuclear Installations Inspectorate (NII) to operate within strict safety margins or lose Site Licences has prompted the need for accurate, up-to-data documentation. A document capture and management retrieval system provides a powerful cost-effective solution, giving rapid access to documentation in a tightly controlled environment. The computerisation of documents and plans is discussed in this article. (Author)

Interim overdentures.

Science.gov (United States)

Fenton, A H

1976-07-01

The construction of an interim overdenture using existing removable partial dentures with natural tooth crowns and artificial teeth can be a simple and economical method of providing patients with dentures while tissues heal and teeth are prepared and restored. A more definite prognosis for both the patient and his remaining dentition can be established before the final overdenture is completed. The procedures necessary to provide three types of interim overdentures have been outlined. Patients tolerate this method of changing their dentitions extremely well.

Development of Accident Scenario for Interim Spent Fuel Storage Facility Based on Fukushima Accident

Energy Technology Data Exchange (ETDEWEB)

Lee, Dongjin; Choi, Kwangsoon; Yoon, Hyungjoon; Park, Jungsu [KEPCO-E and C, Yongin (Korea, Republic of)

2014-05-15

700 MTU of spent nuclear fuel is discharged from nuclear fleet every year and spent fuel storage is currently 70.9% full. The on-site wet type spent fuel storage pool of each NPP(nuclear power plants) in Korea will shortly exceed its storage limit. Backdrop, the Korean government has rolled out a plan to construct an interim spent fuel storage facility by 2024. However, the type of interim spent fuel storage facility has not been decided yet in detail. The Fukushima accident has resulted in more stringent requirements for nuclear facilities in case of beyond design basis accidents. Therefore, there has been growing demand for developing scenario on interim storage facility to prepare for beyond design basis accidents and conducting dose assessment based on the scenario to verify the safety of each type of storage.

Savannah River Interim Waste Management Program Plan - FY 1986

International Nuclear Information System (INIS)

1985-09-01

This document provides the program plan as requested by the Savannah River Operations Office of the Department of Energy. The plan was developed to provide a working knowledge of the nature and extent of the interim waste management programs being undertaken by Savannah River (SR) contractors for the Fiscal Year 1986. In addition, the document projects activities for several years beyond 1986 to adequately plan for safe handling and storage of radioactive wastes generated at Savannah River and for developing technology for improved management of low-level solid wastes. A revised plan will be issued prior to the beginning of the first quarter of each fiscal year. In this document, work descriptions and milestone schedules are current as of the date of publication. Budgets are based on available information as of May 1985

Savannah River Interim Waste Management Program plan, FY-1987

International Nuclear Information System (INIS)

1986-09-01

This document provides the program plan as requested by the Savannah River Operations office of the Department of Energy. The plan was developed to provide a working knowledge of the nature and extent of the interim waste management programs being undertaken by Savannah River (SR) contractors for the Fiscal Year 1987. In addition, the document projects activities for several years beyond 1987 to adequately plan for safe handling and storage of radioactive wastes generated at Savannah River and for developing technology for improved management of low-level solid wastes. A revised plan will be issued prior to the beginning of the first quarter of each fiscal year. In this document, work descriptions and milestone schedules are current as of the date of publication. Budgets are based on available information as of June 1986

Documented Safety Analysis for the Waste Storage Facilities

Energy Technology Data Exchange (ETDEWEB)

Laycak, D

2008-06-16

This documented safety analysis (DSA) for the Waste Storage Facilities was developed in accordance with 10 CFR 830, Subpart B, 'Safety Basis Requirements', and utilizes the methodology outlined in DOE-STD-3009-94, Change Notice 3. The Waste Storage Facilities consist of Area 625 (A625) and the Decontamination and Waste Treatment Facility (DWTF) Storage Area portion of the DWTF complex. These two areas are combined into a single DSA, as their functions as storage for radioactive and hazardous waste are essentially identical. The B695 Segment of DWTF is addressed under a separate DSA. This DSA provides a description of the Waste Storage Facilities and the operations conducted therein; identification of hazards; analyses of the hazards, including inventories, bounding releases, consequences, and conclusions; and programmatic elements that describe the current capacity for safe operations. The mission of the Waste Storage Facilities is to safely handle, store, and treat hazardous waste, transuranic (TRU) waste, low-level waste (LLW), mixed waste, combined waste, nonhazardous industrial waste, and conditionally accepted waste generated at LLNL (as well as small amounts from other DOE facilities).

Technical basis for the ITER detailed design report, cost review and safety analysis (DDR)

International Nuclear Information System (INIS)

1997-01-01

The ITER Detailed Design Report (DDR), Cost Review and Safety Analysis is the 3rd major milestone representing the progress made in the ITER Engineering Design Activities. With the approval of the Interim Design Report (IDR), it has been possible to freeze the main concepts and system approaches for ITER and to develop the design in more detail for the individual components and sub-systems. This report, although designed to be fully understandable as a separate document, focusses particularly on the main changes since the IDR

Sample results from the interim salt disposition program macrobatch 9 tank 21H qualification samples

Energy Technology Data Exchange (ETDEWEB)

Peters, T. B. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2015-11-01

Savannah River National Laboratory (SRNL) analyzed samples from Tank 21H in support of qualification of Macrobatch (Salt Batch) 9 for the Interim Salt Disposition Program (ISDP). This document reports characterization data on the samples of Tank 21H.

Engineering Task Plan for Hose-In-Hose Transfer Lines for the Interim Stabilization Program

International Nuclear Information System (INIS)

TORRES, T.D.

2000-01-01

The document is the Engineering Task Plan for the engineering, design services, planning, project integration and management support for the design, modification, installation and testing of an over ground transfer (OGT) system to support the interim stabilization of S/SX and U Tank Farms

Conceptual waste package interim product specifications and data requirements for disposal of glass commercial high-level waste forms in salt geologic repositories

International Nuclear Information System (INIS)

1983-10-01

The conceptual waste package interim product specifications and data requirements presented are applicable to the reference glass composition described in PNL-3838 and carbon steel canister described in ONWI-438. They provide preliminary numerical values for the commercial high-level waste form parameters and properties identified in the waste form performance specification for geologic isolation in salt repositories. Subject areas treated include containment and isolation, operational period safety, criticality control, waste form/production canister identification, and waste package performance testing requirements. This document was generated for use in the development of conceptual waste package designs in salt. It will be revised as additional data, analyses and regulatory requirements become available. 13 references, 1 figure

National NIF Diagnostic Program Interim Management Plan

International Nuclear Information System (INIS)

Warner, B

2002-01-01

The National Ignition Facility (NIF) has the mission of supporting Stockpile Stewardship and Basic Science research in high-energy-density plasmas. To execute those missions, the facility must provide diagnostic instrumentation capable of observing and resolving in time events and radiation emissions characteristic of the plasmas of interest. The diagnostic instrumentation must conform to high standards of operability and reliability within the NIF environment. These exacting standards, together with the facility mission of supporting a diverse user base, has led to the need for a central organization charged with delivering diagnostic capability to the NIF. The National NIF Diagnostics Program (NNDP) has been set up under the aegis of the NIF Director to provide that organization authority and accountability to the wide user community for NIF. The funds necessary to perform the work of developing diagnostics for NIF will be allocated from the National NIF Diagnostics Program to the participating laboratories and organizations. The participating laboratories and organizations will design, build, and commission the diagnostics for NIF. Restricted availability of funding has had an adverse impact, unforeseen at the time of the original decision to projectize NIF Core Diagnostics Systems and Cryogenic Target Handing Systems, on the planning and initiation of these efforts. The purpose of this document is to provide an interim project management plan describing the organizational structure and management processes currently in place for NIF Core Diagnostics Systems. Preparation of a Program Execution Plan for NIF Core Diagnostics Systems has been initiated and a current draft is provided as Attachment 1 to this document. The National NIF Diagnostics Program Interim Management Plan provides a summary of primary design criteria and functional requirements, current organizational structure, tracking and reporting procedures, and current planning estimates of project scope

Security in the transport of radioactive material - interim guidance for comment

International Nuclear Information System (INIS)

Legoux, P.; Wangler, M.

2004-01-01

While the IAEA has provided specific guidance for physical protection in the transport of nuclear material, its previous publications have only provided some general guidelines for security of non-nuclear radioactive material in transport. Some basic practical advice has been provided in the requirements of the International Basic Safety Standards for Protection against Ionising Radiation and for the Safety of Radiation Sources (BSS) [1]. These guidelines were primarily directed toward such issues as unintentional exposure to radiation, negligence and inadvertent loss. Recently, the IAEA published a document on the security of sources, which included some general guidance on providing security during transport of the sources. However, it is clear that more guidance is needed for security during the transport of radioactive material in addition to those already existing for nuclear material. Member States have requested guidance on the type and nature of security measures that might be put in place for radioactive material in general during its transport and on the methodology to be used in choosing and implementing such measures. The purpose of the TECDOC on Security in the Transport of Radioactive Material being developed by the IAEA is to provide an initial response to that request. This interim guidance is being developed with a view to harmonizing the security guidance - as much as possible - with existing guidance from the IAEA for the transport of radioactive sources and nuclear material. It is also intended to harmonize with model requirements developed in 2002-2003 by the United Nations Economic and Social Council's Committee of Experts on the Transport of Dangerous Goods and on the Globally Harmonised System of Classification and Labelling of Chemicals which was issued as general security guidelines for all dangerous goods, including radioactive material, and that will shortly be implemented as binding regulations by the international modal authorities

Security in the transport of radioactive material - interim guidance for comment

Energy Technology Data Exchange (ETDEWEB)

Legoux, P.; Wangler, M. [International Atomic Energy Agency, Vienna (Austria)

2004-07-01

While the IAEA has provided specific guidance for physical protection in the transport of nuclear material, its previous publications have only provided some general guidelines for security of non-nuclear radioactive material in transport. Some basic practical advice has been provided in the requirements of the International Basic Safety Standards for Protection against Ionising Radiation and for the Safety of Radiation Sources (BSS) [1]. These guidelines were primarily directed toward such issues as unintentional exposure to radiation, negligence and inadvertent loss. Recently, the IAEA published a document on the security of sources, which included some general guidance on providing security during transport of the sources. However, it is clear that more guidance is needed for security during the transport of radioactive material in addition to those already existing for nuclear material. Member States have requested guidance on the type and nature of security measures that might be put in place for radioactive material in general during its transport and on the methodology to be used in choosing and implementing such measures. The purpose of the TECDOC on Security in the Transport of Radioactive Material being developed by the IAEA is to provide an initial response to that request. This interim guidance is being developed with a view to harmonizing the security guidance - as much as possible - with existing guidance from the IAEA for the transport of radioactive sources and nuclear material. It is also intended to harmonize with model requirements developed in 2002-2003 by the United Nations Economic and Social Council's Committee of Experts on the Transport of Dangerous Goods and on the Globally Harmonised System of Classification and Labelling of Chemicals which was issued as general security guidelines for all dangerous goods, including radioactive material, and that will shortly be implemented as binding regulations by the international modal

Safety aspects of long-term dry interim storage of Type B spent fuel and high-level transport casks

International Nuclear Information System (INIS)

Wolff, D.; Probst, U.; Voelzke, H.; Droste, B.; Roedel, R.

2004-01-01

Based on the German decision to minimise transport of spent fuel casks between nuclear power plants, reprocessing plants and central storage facilities several on-site storage facilities were licensed until the end of 2003. Because of the large amount of Type B(U) transport casks which are going to be used for long-term interim storage the question of time-limited Type B(U) licence maintenance during the storage period of up to 40 years has been discussed under different aspects. This paper describes present technical aspects of the discussion. A main aspect of qualification of transport casks for interim storage is the long-term behaviour of the metallic seal-lid system. Here we present results from current long-term experimental tests with metallic 'Helicoflex' seals in which pool water is enclosed. This series of tests has been performed by the Federal Institute for Materials Research and Testing (BAM) on behalf of the Federal Office for Radiation Protection (BfS) since 2001. Finally, the paper presents a German concept for an exchange of experience, know-how and state-of-the-art between authorities and technical experts with regard to cask dispatch in nuclear facilities. BAM has taken over a central role in this so-called 'coordinating institution for cask dispatching information' ('KOBAF') which entails management of an online database of cask-specific documents and a technical working group meeting twice a year. The goal is to keep comparable technical standards for all nuclear sites and storage facilities which are going to load and dispatch casks of the same or similar types under the responsibility of different German state governments for the coming decades. (author)

Evaluation of Hose in Hose Transfer Line Service Life for Hanfords Interim Stabilization Program

International Nuclear Information System (INIS)

TORRES, T.D.

2001-01-01

RPP-6153, Engineering Task Plan for Hose-in-Hose Transfer System for the Interim Stabilization Program (Torres, 2000a), defines the programmatic goals, functional requirements, and technical criteria for the development and subsequent installation of waste transfer line equipment to support Hanford's Interim Stabilization Program. RPP-6028, Specification for Hose in Hose Transfer Lines for Hanford's Interim Stabilization Program (Torres, 2000b), has been issued to define the specific requirements for the design, manufacture, and verification of transfer line assemblies for specific waste transfer applications associated with Interim Stabilization. Included in RPP-6028 are tables defining the chemical constituents of concern to which transfer lines will be exposed. Current Interim Stabilization Program planning forecasts that the at-grade transfer lines will be required to convey pumpable waste for as much as three years after commissioning, RPP-6028 Section 3.2.7. Performance Incentive Number ORP-05 requires that all the Single Shell Tanks be Interim Stabilized by September 30, 2003. The Tri-Party Agreement (TPA) milestone M-41-00, enforced by a federal consent decree, requires all the Single Shell Tanks to be Interim stabilized by September 30, 2004. By meeting the Performance Incentive the TPA milestone is met. Prudent engineering dictates that the equipment used to transfer waste have a life in excess of the forecasted operational time period, with some margin to allow for future adjustments to the planned schedule. This document evaluates the effective service life of the Hose-in-Hose Transfer Lines, based on information submitted by the manufacturer, published literature and calculations. The effective service life of transfer line assemblies is a function of several factors. Foremost among these are the hose material's resistance to the harmful effects of process fluid characteristics, ambient environmental conditions, exposure to ionizing radiation and the

Safety and effectiveness of certolizumab pegol in patients with rheumatoid arthritis: Interim analysis of post-marketing surveillance.

Science.gov (United States)

Kameda, Hideto; Nishida, Keiichiro; Nannki, Toshihiro; Watanabe, Akira; Oshima, Yukiya; Momohara, Shigaki

2017-01-01

Objective: To evaluate the safety and effectiveness of certolizumab pegol (CZP) in a real-world setting among Japanese patients with rheumatoid arthritis. Post-marketing surveillance data from 2,579 patients treated with CZP were analyzed. Adverse events (AEs) observed during the 24-week CZP treatment period were recorded. Disease activity was evaluated using DAS28-ESR and DAS28-CRP at baseline, Week 12, Week 24, or at withdrawal. The total period of exposure to CZP was 1313.8 patient-years (PY). AEs were reported in 658 (25.5%) patients, at an event rate (ER) of 73.68/100 PY. The most frequent serious AEs were pneumonia, herpes zoster, and interstitial lung disease, at ER per 100 PY of 2.06, 1.29, and 1.22, respectively. Mean disease activity scores at baseline, as measured by DAS28-ESR and DAS28-CRP, were 4.77 ± 1.34 and 4.21 ± 1.27, respectively. Mean changes from baseline at the last observation were -1.29 ± 1.46 and -1.30 ± 1.42, respectively. EULAR good or moderate responses were achieved in 65% of patients. Longer disease duration, prior biologics use, and treatment without MTX co-therapy were associated with EULAR no response. In this interim analysis, no new safety signals were observed. Clinical response to CZP was observed in approximately two thirds of patients.

Engineering Task Plan for Hose-In-Hose Transfer Lines for the Interim Stabilization Program

International Nuclear Information System (INIS)

RUNG, M.P.

2000-01-01

This document is the Engineering Task Plan for the engineering, design services, planning, project integration and management support for the design, modification, installation and testing of an over ground transfer (OGT) system to support the interim stabilization of nine tanks in the 241-S/SX Tank Farms

7 CFR 1738.21 - Interim financing.

Science.gov (United States)

2010-01-01

... 7 Agriculture 11 2010-01-01 2010-01-01 false Interim financing. 1738.21 Section 1738.21... Interim financing. (a) Upon notification by RUS that an applicant's application is considered complete, the applicant may enter into an interim financing agreement with a lender other than RUS or use its...

Technical basis for the ITER detailed design report, cost review and safety analysis (DDR)

Energy Technology Data Exchange (ETDEWEB)

NONE

1997-12-01

The ITER Detailed Design Report (DDR), Cost Review and Safety Analysis is the 3rd major milestone representing the progress made in the ITER Engineering Design Activities. With the approval of the Interim Design Report (IDR), it has been possible to freeze the main concepts and system approaches for ITER and to develop the design in more detail for the individual components and sub-systems. This report, although designed to be fully understandable as a separate document, focusses particularly on the main changes since the IDR. Refs, figs, tabs

List of documents received by the INDC Secretariat

International Nuclear Information System (INIS)

1979-05-01

The Nuclear Data Section of the International Atomic Energy Agency receives documents originated by or for the International Nuclear Data Committee for distribution. This list includes all INDC documents received and distributed between January 1968 and May 1979, and supersedes INDC(SEC)-66/UN. This list is produced directly from computer printout in two sorts: one ordered by accession number, and the other ordered by document number with-in each origin series (e.g. listing all INDC(SEC)-documents in one block). Reference to earlier INDSWG and ''interim'' INDC reports received between 1962 and 1967 are listed in report INDC/199 (dated November 1967)

Transport casks help solve spent fuel interim storage problems

International Nuclear Information System (INIS)

Dierkes, P.; Janberg, K.; Baatz, H.; Weinhold, G.

1980-01-01

Transport casks can be used as storage modules, combining the inherent safety of passive cooling with the absence of secondary radioactive waste and the flexibility to build up storage capacity according to actual requirements. In the Federal Republic of Germany, transport casks are being developed as a solution to its interim storage problems. Criteria for their design and licensing are outlined. Details are given of the casks and the storage facility. Tests are illustrated. (U.K.)

Issue on NPP-I and C important to safety-Data Communication

International Nuclear Information System (INIS)

Koo, I. S.; Hong, S. B.; Cho, J. W.; Choi, Y. S.; Lee, J. C.

2010-01-01

1. Issue on CDV and FDIS of IEC61500 - Nuclear Power Plants - Instrumentation and control important to safety -Data communication - Activities on IEC TC45, SC45A/WGA3. 2. As issue the requirements for safety data communication which is essential part of digital I and C systems, the fundamental technology for IT based nuclear I and C is established. 3. Approval and circulation of IED61500 CDV and FDIS - Issue of the international standard, IEC 61500. 4. Issue one IEC61500, three interim documents, three presentations and five technical support to industry, and participation in IEC TC45 and SC45A plenary meeting and intermediate meeting on SC45A/WGA3. 5. Based on IEC61500, an new project on wireless technologyes application to NPP will be proceeded

7 CFR 1735.75 - Interim financing.

Science.gov (United States)

2010-01-01

... 7 Agriculture 11 2010-01-01 2010-01-01 false Interim financing. 1735.75 Section 1735.75... Involving Loan Funds § 1735.75 Interim financing. (a) A borrower may submit a written request for RUS approval of interim financing if it is necessary to close an acquisition before the loan to finance the...

Interim staff position on environmental qualification of safety-related electrical equipment: including staff responses to public comments. Regulatory report

International Nuclear Information System (INIS)

Szukiewicz, A.J.

1981-07-01

This document provides the NRC staff positions regarding selected areas of environmental qualification of safety-related electrical equipment, in the resolution of Unresolved Safety Issue A-24, 'Qualification of Class IE Safety-Related Equipment.' The positions herein are applicable to plants that are or will be in the construction permit (CP) or operating license (OL) review process and that are required to satisfy the requirements set forth in either the 1971 or the 1974 version of IEEE-323 standard

105-H Reactor Interim Safe Storage Project Final Report

International Nuclear Information System (INIS)

Ison, E.G.

2008-01-01

The following information documents the decontamination and decommissioning of the 105-H Reactor facility, and placement of the reactor core into interim safe storage. The D and D of the facility included characterization, engineering, removal of hazardous and radiologically contaminated materials, equipment removal, decontamination, demolition of the structure, and restoration of the site. The ISS work also included construction of the safe storage enclosure, which required the installation of a new roofing system, power and lighting, a remote monitoring system, and ventilation components.

Angra-1 probabilistic safety study-phase B

International Nuclear Information System (INIS)

Fernandes Filho, T.L.; Gibelli, S.M.O.

1988-05-01

This study represents the Phase B of the Angra-1 Probabilistic Safety Study and is the the final report prepared for the IAEA under Research Contract No. 3423/R2/RB. The three main items covered in this report are the establishment of interim safety goals, analysis of Angra-1 operational experience and development of emergency procedures to address severe accidents. For establishment of interim safety goals a methodology for calculating consequences and risks associated to the Angra-1 operation was developed based on the available data and codes. The proposed safety goals refer to the individual risk of early fatality for people living in the vicinity of the plant, colective risk of cancer fatalities for people living near the plant, the propobability of core melt occurrence and the probability of dominant accident sequences. (author) [pt

Exemption of radiation sources and practices from regulatory control: Interim report

International Nuclear Information System (INIS)

1987-01-01

This document is an interim report on progress at the IAEA on exemption principles and their application to low-level radioactive waste disposal. In the first part of the document the general principles for the exemption of radiation sources and practices from regulatory control are described. The exempt quantities of low-level radioactive wastes for disposal to municipal landfill or by incineration including methods for their derivation and generic values are contained in the second part of the document. In the appendices the individual effective dose equivalents and committed effective dose equivalents by pathway for waste concentrations of 1 Bq.g -1 and some quoted limits from the literature on dust concentrations at the outlet of municipal waste incinerators are estimated

Experience with the licensing of the interim spent fuel storage facility modification

International Nuclear Information System (INIS)

Bezak, S.; Beres, J.

1999-01-01

After political and economical changes in the end of eighties, the utility operating the nuclear power plants in the Slovak Republic (SE, a.s.) decided to change the original scheme of the back-end of the nuclear fuel cycle; instead of reprocessing in the USSR/Russian Federation spent fuel will be stored in an interim spent fuel storage facility until the time of the final decision. As the best solution, a modification of the existing interim spent fuel storage facility has been proposed. Due to lack of legal documents for this area, the Regulatory Authority of the Slovak Republic (UJD SR) performed licensing procedures of the modification on the basis of recommendations by the IAEA, the US NRC and the relevant parts of the US CFR Title 10. (author)

13 CFR 120.890 - Source of interim financing.

Science.gov (United States)

2010-01-01

... 13 Business Credit and Assistance 1 2010-01-01 2010-01-01 false Source of interim financing. 120... Development Company Loan Program (504) Interim Financing § 120.890 Source of interim financing. A Project may use interim financing for all Project costs except the Borrower's contribution. Any source (including...

Global Spent Fuel Logistics Systems Study (GSFLS). Volume 2A. GSFLS visit findings (appendix). Interim report

International Nuclear Information System (INIS)

1978-01-01

This appendix is a part of the interim report documentation for the Global Spent Fuel Logistics System (GSFLS) study. This appendix provides the legal/regulatory reference material, supportive of Volume 2 - GSFLS Visit Finding and Evaluations; and certain background material on British Nuclear Fuel Limited

ITER final design report, cost review and safety analysis (FDR) and relevant documents

International Nuclear Information System (INIS)

1999-01-01

This volume contains the fourth major milestone report and documents associated with its acceptance, review and approval. This ITER Final Design Report, Cost Review and Safety Analysis was presented to the ITER Council at its 13th meeting in February 1998 and was approved at its extraordinary meeting on 25 June 1998. The contents include an outline of the ITER objectives, the ITER parameters and design overview as well as operating scenarios and plasma performance. Furthermore, design features, safety and environmental characteristics and schedule and cost estimates are given

Results from the Interim Salt Disposition Program Macrobatch 11 Tank 21H Acceptance Samples

Energy Technology Data Exchange (ETDEWEB)

Peters, T. B. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Bannochie, C. J. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2017-11-13

Savannah River National Laboratory (SRNL) analyzed samples from Tank 21H in support of verification of Macrobatch (Salt Batch) 11 for the Interim Salt Disposition Program (ISDP) for processing. This document reports characterization data on the samples of Tank 21H and fulfills the requirements of Deliverable 3 of the Technical Task Request (TTR).

Documents preparation and review

International Nuclear Information System (INIS)

1999-01-01

Ignalina Safety Analysis Group takes active role in assisting regulatory body VATESI to prepare various regulatory documents and reviewing safety reports and other documentation presented by Ignalina NPP in the process of licensing of unit 1. The list of main documents prepared and reviewed is presented

ETF interim design review

International Nuclear Information System (INIS)

Steiner, D.; Rutherford, P.H.

1980-01-01

A three-day ETF Interim Design Review was conducted on July 23-25, 1980, at the Sheraton Potomac Inn in Rockville, Maryland. The intent of the review was to provide a forum for an in-depth assessment and critique of all facets of the ETF design by members of the fusion community. The review began with an opening plenary session at which an overview of the ETF design was presented by D. Steiner, manager of the ETF Design Center, complemented by a physics overview by P.H. Rutherford, chairman of the ETF/INTOR Physics Committee. This was followed by six concurrent review sessions over the next day and a half. The review closed with a plenary session at which the Design Review Board presented its findings. This document consists of the viewgraphs for the opening plenary session and an edited version of the presentations made by Steiner and Rutherford

Savannah River interim waste management program plan: FY 1984. Revision 1

International Nuclear Information System (INIS)

1983-10-01

This document provides the program plan as requested by the Savannah River Operations Office of the Department of Energy. The plan was developed to provide a working knowledge of the nature and extent of the interim waste management programs being undertaken by Savannah River (SR) contractors for the Fiscal Year 1984. In addition, the document projects activities for several years beyond 1984 to adequately plan for safe handling and storage of radioactive wastes generated at Savannah River and for developing technology for improved management of low-level solid wastes. A revised plan will be issued prior to the beginning of the first quarter of each fiscal year. In this document, work descriptions and milestone schedules are current as of the date of publication. Budgets are based on available information as of June 1983

Global Spent Fuel Logistics Systems Study (GSFLS). Volume 2A. GSFLS visit findings (appendix). Interim report

Energy Technology Data Exchange (ETDEWEB)

None

1978-01-31

This appendix is a part of the interim report documentation for the Global Spent Fuel Logistics System (GSFLS) study. This appendix provides the legal/regulatory reference material, supportive of Volume 2 - GSFLS Visit Finding and Evaluations; and certain background material on British Nuclear Fuel Limited (BNFL).

The PDF4LHC Working Group Interim Report

CERN Document Server

Alekhin, Sergey; Ball, Richard D.; Bertone, Valerio; Blumlein, Johannes; Botje, Michiel; Butterworth, Jon; Cerutti, Francesco; Cooper-Sarkar, Amanda; de Roeck, Albert; Del Debbio, Luigi; Feltesse, Joel; Forte, Stefano; Glazov, Alexander; Guffanti, Alberto; Gwenlan, Claire; Huston, Joey; Jimenez-Delgado, Pedro; Lai, Hung-Liang; Latorre, Jose I.; McNulty, Ronan; Nadolsky, Pavel; Olaf Moch, Sven; Pumplin, Jon; Radescu, Voica; Rojo, Juan; Sjostrand, Torbjorn; Stirling, W.J.; Stump, Daniel; Thorne, Robert S.; Ubiali, Maria; Vicini, Alessandro; Watt, Graeme; Yuan, C.-P.

2011-01-01

This document is intended as a study of benchmark cross sections at the LHC (at 7 TeV) at NLO using modern parton distribution functions currently available from the 6 PDF fitting groups that have participated in this exercise. It also contains a succinct user guide to the computation of PDFs, uncertainties and correlations using available PDF sets. A companion note, also submitted to the archive, provides an interim summary of the current recommendations of the PDF4LHC working group for the use of parton distribution functions and of PDF uncertainties at the LHC, for cross section and cross section uncertainty calculations.

Hazard classification for the 200-ZP-1 Operable Unit Phase 2 and 3 interim remedial measure

International Nuclear Information System (INIS)

Oestreich, D.K.

1996-04-01

This safety assessment documents the Final Hazard Classification (FHC) for Phase 2 and 3 interim remedial measure (IRM) activities to be conducted in the 200 West Area of the Hanford Site. The 200-ZP-1 Phase 2 and 3 IRM activities will involve the air stripping of carbon tetrachloride (CCl 4 ) from extracted groundwater using a packed-bed stripper column followed by gas-phase adsorption of the CCl 4 from the stripper off-gas onto a granular activated carbon (GAC) bed. The stripper is designed to be operated at a feedwater flow rate of up to 1,893 L/min (500 gal/min) and to remove 13.6 kg/day (30.0 lb/day) of CCl 4 . For Phase 2, which includes the initial year of operation, it is planned to operate the stripper at 568 L/min (150 gal/min). The process flow diagram for the Phase 2 and 3 system is shown

Interim Control Strategy for the Test Area North/Technical Support Facility Sewage Treatment Facility Disposal Pond - Two-year Update

International Nuclear Information System (INIS)

L. V. Street

2007-01-01

The Idaho Cleanup Project has prepared this interim control strategy for the U.S. Department of Energy Idaho Operations Office pursuant to DOE Order 5400.5, Chapter 11.3e (1) to support continued discharges to the Test Area North/Technical Support Facility Sewage Treatment Facility Disposal Pond. In compliance with DOE Order 5400.5, a 2-year review of the Interim Control Strategy document has been completed. This submittal documents the required review of the April 2005 Interim Control Strategy. The Idaho Cleanup Project's recommendation is unchanged from the original recommendation. The Interim Control Strategy evaluates three alternatives: (1) re-route the discharge outlet to an uncontaminated area of the TSF-07; (2) construct a new discharge pond; or (3) no action based on justification for continued use. Evaluation of Alternatives 1 and 2 are based on the estimated cost and implementation timeframe weighed against either alternative's minimal increase in protection of workers, the public, and the environment. Evaluation of Alternative 3, continued use of the TSF-07 Disposal Pond under current effluent controls, is based on an analysis of four points: - Record of Decision controls will protect workers and the public - Risk of increased contamination is low - Discharge water will be eliminated in the foreseeable future - Risk of contamination spread is acceptable. The Idaho Cleanup Project recommends Alternative 3, no action other than continued implementation of existing controls and continued deactivation, decontamination, and dismantlement efforts at the Test Area North/Technical Support Facility

Burn site groundwater interim measures work plan.

Energy Technology Data Exchange (ETDEWEB)

Witt, Jonathan L. (North Wind, Inc., Idaho Falls, ID); Hall, Kevin A. (North Wind, Inc., Idaho Falls, ID)

2005-05-01

This Work Plan identifies and outlines interim measures to address nitrate contamination in groundwater at the Burn Site, Sandia National Laboratories/New Mexico. The New Mexico Environment Department has required implementation of interim measures for nitrate-contaminated groundwater at the Burn Site. The purpose of interim measures is to prevent human or environmental exposure to nitrate-contaminated groundwater originating from the Burn Site. This Work Plan details a summary of current information about the Burn Site, interim measures activities for stabilization, and project management responsibilities to accomplish this purpose.

24 CFR 35.1330 - Interim controls.

Science.gov (United States)

2010-04-01

... Lead-Paint Hazard Evaluation and Hazard Reduction Activities § 35.1330 Interim controls. Interim..., cleanable covering or coating, such as metal coil stock, plastic, polyurethane, or linoleum. (3) Surfaces...

The challenges facing the long term interim storage

Energy Technology Data Exchange (ETDEWEB)

Iracane, D. [CEA Sacaly, Dir. de la Simulation et des Outils Experimentaux-DSOE, 91 - Gif sur Yvette (France); Marvy, A. [CEA Saclay, Dir. du Developpement et de l' Innovation Nucleares-DDIN, 91 - Gif Sur Yvette (France)

2001-07-01

In France electricity generation by means of commercial nuclear power plants has come to a point where it contributes to the national demand at a level of 80%. The safety performance of the production system has also reached a high level of both maturity and reliability taking advantage of the cumulative effect of a 30 years long learning experience and ever more stringent safety requirements. The policy to reprocess spent fuel has been overriding but no final decision has yet been made regarding the ultimate disposition of the waste streams. Although studies on deep geological disposal are ongoing, France is also looking at whether and under which conditions a long-term interim storage may provide an effective flexibility to the fuel cycle back-end. We discuss thereafter the needs and the paramount objectives of this latter R and D program. Results are being framed as potential guiding criteria for decision makers and various stakeholders. In first part, we propose a general analysis which emphasises that a long term interim storage is more than a classical nuclear facility because it explicitly requires long-lasting control and creates a burden for Society during many generations. Then, in second part, we offer an overview of the technical results from the R and D program as they stand at the time of writing. As an answer to the Government request, a strong emphasis has been put on this research for three years. Conclusion is an attempt to outline the societal context in which future decisions will have to be made. (author)

The challenges facing the long term interim storage

International Nuclear Information System (INIS)

Iracane, D.; Marvy, A.

2001-01-01

In France electricity generation by means of commercial nuclear power plants has come to a point where it contributes to the national demand at a level of 80%. The safety performance of the production system has also reached a high level of both maturity and reliability taking advantage of the cumulative effect of a 30 years long learning experience and ever more stringent safety requirements. The policy to reprocess spent fuel has been overriding but no final decision has yet been made regarding the ultimate disposition of the waste streams. Although studies on deep geological disposal are ongoing, France is also looking at whether and under which conditions a long-term interim storage may provide an effective flexibility to the fuel cycle back-end. We discuss thereafter the needs and the paramount objectives of this latter R and D program. Results are being framed as potential guiding criteria for decision makers and various stakeholders. In first part, we propose a general analysis which emphasises that a long term interim storage is more than a classical nuclear facility because it explicitly requires long-lasting control and creates a burden for Society during many generations. Then, in second part, we offer an overview of the technical results from the R and D program as they stand at the time of writing. As an answer to the Government request, a strong emphasis has been put on this research for three years. Conclusion is an attempt to outline the societal context in which future decisions will have to be made. (author)

Progress and future direction for the interim safe storage and disposal of Hanford high level waste (HLW)

International Nuclear Information System (INIS)

Wodrich, D.D.

1996-01-01

This paper describes the progress made at the largest environmental cleanup program in the United States. Substantial advances in methods to start interim safe storage of Hanford Site high-level wastes, waste characterization to support both safety- and disposal-related information needs, and proceeding with cost-effective disposal by the US DOE and its Hanford Site contractors, have been realized. Challenges facing the Tank Waste Remediation System Program, which is charged with the dual and parallel missions of interim safe storage and disposal of the high-level tank waste stored at the Hanford Site, are described

Surveillance on The Safety and Efficacy of Ambrisentan (Volibris Tablet 2.5 mg) in Patients with Pulmonary Arterial Hypertension in Real Clinical Practice: Post-marketing Surveillance (Interim Analysis Report).

Science.gov (United States)

Takahashi, Tomohiko; Hayata, Satoru; Kobayashi, Akihiro; Onaka, Yuna; Ebihara, Takeshi; Hara, Terufumi

2018-03-01

Pulmonary arterial hypertension (PAH) is an intractable and rare disease and the accumulation of clinical evidence under real-world setting is needed. A post-marketing surveillance for the endothelin receptor antagonist ambrisentan (Volibris tablet) has been conducted by all-case investigation since September 2010. This paper is an interim report on the safety and efficacy of ambrisentan in 702 patients with PAH. PAH patients aged 15 years or older were subjected to the analysis. The safety analysis by overall cases or stratification of patient backgrounds and the efficacy analysis were investigated. Regarding patient characteristics, the 702 patients subjected to safety analysis included 543 (77.4%) women and 546 (77.8%) patients at WHO functional class II/III. The mean observational time was 392.7 days. A total of 324 adverse drug reaction (ADR) occurred in 204 (29.1%) patients. Common ADRs (≥ 2%) included anemia (4.6%), peripheral edema (4.1%), headache (3.6%), edema and face edema (2.6% each), abnormal hepatic function (2.3%), and epistaxis (2.1%). There were 82 serious ADRs occurring in 44 (6.3%) patients (385 serious adverse events in 184 (26.2%) patients). Although 11 (1.6%) interstitial lung disease (ILD) cases were reported, all were observed in patients with disease that may contribute to ILD and therefore it is difficult to assess if ambrisentan was associated with these events. There was no difference in safety in relation to the presence/absence of connective tissue disease-related PAH (CTD-PAH) or combination therapy. Among 677 patients subjected to efficacy analysis, those in whom hemodynamic status was determined before and after treatment showed improvement in the mean pulmonary arterial pressure and pulmonary vascular resistance after treatment. The interim results showed safety consistent with the known profile of ambrisentan in terms of the types and frequencies of ADRs in patients with PAH in real clinical practice, in comparison with

CMS Interim Memorandum of Understanding The Costs and How They are Calculated

CERN Document Server

Willers, Ian Malcolm

2001-01-01

This document explains how we arrived at the costs that are to be used in the Interim Memorandum of Understanding, iMoU. It integrates information taken from the Hoffmann review, PASTA and Lucas Taylor. I would like to thank Petr Moissenz for proof reading which helped me to improve the accuracy of the information in a short period of time. Also Claude Charlot for his valuable comments.

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

the integrated safety management system approach for having a uniform and consistent process: a method has been suggested by the U S . Department of Energy at Richland and the Project Hanford Procedures when fire hazard analyses and safety analyses are required. This process provides for a common basis approach in the development of the fire hazard analysis and the safety analysis. This process permits the preparers of both documents to jointly participate in the development of the hazard analysis process. This paper presents this method to implement the integrated safety management approach in the development of the fire hazard analysis and safety analysis that provides consistency of assumptions. consequences, design considerations, and other controls necessarily to protect workers, the public. and the environment

The Homestake Interim Laboratory and Homestake DUSEL

Science.gov (United States)

Lesko, Kevin T.

2011-12-01

The former Homestake gold mine in Lead South Dakota is proposed for the National Science Foundation's Deep Underground Science and Engineering Laboratory (DUSEL). The gold mine provides expedient access to depths in excess of 8000 feet below the surface (>7000 mwe). Homestake's long history of promoting scientific endeavours includes the Davis Solar Neutrino Experiment, a chlorine-based experiment that was hosted at the 4850 Level for more than 30 years. As DUSEL, Homestake would be uncompromised by competition with mining interests or other shared uses. The facility's 600-km of drifts would be available for conversion for scientific and educational uses. The State of South Dakota, under Governor Rounds' leadership, has demonstrated exceptionally strong support for Homestake and the creation of DUSEL. The State has provided funding totalling $46M for the preservation of the site for DUSEL and for the conversion and operation of the Homestake Interim Laboratory. Motivated by the strong educational and outreach potential of Homestake, the State contracted a Conversion Plan by world-recognized mine-engineering contractor to define the process of rehabilitating the facility, establishing the appropriate safety program, and regaining access to the facility. The State of South Dakota has established the South Dakota Science and Technology Authority to oversee the transfer of the Homestake property to the State and the rehabilitation and preservation of the facility. The Homestake Scientific Collaboration and the State of South Dakota's Science and Technology Authority has called for Letters of Interest from scientific, educational and engineering collaborations and institutions that are interested in hosting experiments and uses in the Homestake Interim Facility in advance of the NSF's DUSEL, to define experiments starting as early as 2007. The Homestake Program Advisory Committee has reviewed these Letters and their initial report has been released. Options for

The Homestake Interim Laboratory and Homestake DUSEL

International Nuclear Information System (INIS)

Lesko, Kevin T.

2011-01-01

The former Homestake gold mine in Lead South Dakota is proposed for the National Science Foundation's Deep Underground Science and Engineering Laboratory (DUSEL). The gold mine provides expedient access to depths in excess of 8000 feet below the surface (>7000 mwe). Homestake's long history of promoting scientific endeavours includes the Davis Solar Neutrino Experiment, a chlorine-based experiment that was hosted at the 4850 Level for more than 30 years. As DUSEL, Homestake would be uncompromised by competition with mining interests or other shared uses. The facility's 600-km of drifts would be available for conversion for scientific and educational uses. The State of South Dakota, under Governor Rounds' leadership, has demonstrated exceptionally strong support for Homestake and the creation of DUSEL. The State has provided funding totalling $46M for the preservation of the site for DUSEL and for the conversion and operation of the Homestake Interim Laboratory. Motivated by the strong educational and outreach potential of Homestake, the State contracted a Conversion Plan by world-recognized mine-engineering contractor to define the process of rehabilitating the facility, establishing the appropriate safety program, and regaining access to the facility. The State of South Dakota has established the South Dakota Science and Technology Authority to oversee the transfer of the Homestake property to the State and the rehabilitation and preservation of the facility. The Homestake Scientific Collaboration and the State of South Dakota's Science and Technology Authority has called for Letters of Interest from scientific, educational and engineering collaborations and institutions that are interested in hosting experiments and uses in the Homestake Interim Facility in advance of the NSF's DUSEL, to define experiments starting as early as 2007. The Homestake Program Advisory Committee has reviewed these Letters and their initial report has been released. Options for

BWR Spent Nuclear Fuel Integrity Research and Development Survey for UKABWR Spent Fuel Interim Storage

Energy Technology Data Exchange (ETDEWEB)

Bevard, Bruce Balkcom [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Mertyurek, Ugur [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Belles, Randy [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Scaglione, John M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2015-10-01

The objective of this report is to identify issues and support documentation and identify and detail existing research on spent fuel dry storage; provide information to support potential R&D for the UKABWR (United Kingdom Advanced Boiling Water Reactor) Spent Fuel Interim Storage (SFIS) Pre-Construction Safety Report; and support development of answers to questions developed by the regulator. Where there are gaps or insufficient data, Oak Ridge National Laboratory (ORNL) has summarized the research planned to provide the necessary data along with the schedule for the research, if known. Spent nuclear fuel (SNF) from nuclear power plants has historically been stored on site (wet) in spent fuel pools pending ultimate disposition. Nuclear power users (countries, utilities, vendors) are developing a suite of options and set of supporting analyses that will enable future informed choices about how best to manage these materials. As part of that effort, they are beginning to lay the groundwork for implementing longer-term interim storage of the SNF and the Greater Than Class C (CTCC) waste (dry). Deploying dry storage will require a number of technical issues to be addressed. For the past 4-5 years, ORNL has been supporting the U.S. Department of Energy (DOE) in identifying these key technical issues, managing the collection of data to be used in issue resolution, and identifying gaps in the needed data. During this effort, ORNL subject matter experts (SMEs) have become expert in understanding what information is publicly available and what gaps in data remain. To ensure the safety of the spent fuel under normal and frequent conditions of wet and subsequent dry storage, intact fuel must be shown to: 1.Maintain fuel cladding integrity; 2.Maintain its geometry for cooling, shielding, and subcriticality; 3.Maintain retrievability, and damaged fuel with pinhole or hairline cracks must be shown not to degrade further. Where PWR (pressurized water reactor) information is

End of mission report on seismic safety review mission for Belene NPP site

International Nuclear Information System (INIS)

Gurpinar, A.; Mohammadioun, B.; Schneider, H.; Serva, L.

1995-01-01

Upon the invitation of the Bulgarian government through the Committee for the Peaceful Uses of Atomic Energy and within the framework of the implementation of the Technical Cooperation project BUL/9/012 related to site and seismic of NPPs, a mission visited Sofia 3 - 7 July 1995. The mission constituted a follow-up of the interim review of subjects related to tectonic stability and seismic hazard characterization of the site which was performed in September 1993. The main objective of the mission was the final review of the subjects already reviewed in September 1993 as well as issues related to geotechnical engineering and foundation safety. The main terms of reference of the present mission was to verify the implementation of the recommendations of the Site Safety Review Mission of June 1990. This document gives findings on geology-tectonics, seismology and foundation safety. In the end conclusions and recommendations of the mission are presented

Safety evaluation of the Greifswald nuclear power plant, unit 1-4

International Nuclear Information System (INIS)

1990-06-01

The first interim report primarily deals with an evaluation of the pressurized components of the primary loops, especially with the embrittlement of the reactor pressure vessel material. In addition, first estimates concerning the safety design of the plants are made. The second interim report reflects the state of further studies relating to the safety design and the evaluation of operational experiences. The report includes a summarized assessment in which the recommendations cited in the technical chapters are evaluated and subdivided into three categories of backfitting measures. (orig.) [de

Interim Control Strategy for the Test Area North/Technical Support Facility Sewage Treatment Facility Disposal Pond - Two-year Update

Energy Technology Data Exchange (ETDEWEB)

L. V. Street

2007-04-01

The Idaho Cleanup Project has prepared this interim control strategy for the U.S. Department of Energy Idaho Operations Office pursuant to DOE Order 5400.5, Chapter 11.3e (1) to support continued discharges to the Test Area North/Technical Support Facility Sewage Treatment Facility Disposal Pond. In compliance with DOE Order 5400.5, a 2-year review of the Interim Control Strategy document has been completed. This submittal documents the required review of the April 2005 Interim Control Strategy. The Idaho Cleanup Project's recommendation is unchanged from the original recommendation. The Interim Control Strategy evaluates three alternatives: (1) re-route the discharge outlet to an uncontaminated area of the TSF-07; (2) construct a new discharge pond; or (3) no action based on justification for continued use. Evaluation of Alternatives 1 and 2 are based on the estimated cost and implementation timeframe weighed against either alternative's minimal increase in protection of workers, the public, and the environment. Evaluation of Alternative 3, continued use of the TSF-07 Disposal Pond under current effluent controls, is based on an analysis of four points: - Record of Decision controls will protect workers and the public - Risk of increased contamination is low - Discharge water will be eliminated in the foreseeable future - Risk of contamination spread is acceptable. The Idaho Cleanup Project recommends Alternative 3, no action other than continued implementation of existing controls and continued deactivation, decontamination, and dismantlement efforts at the Test Area North/Technical Support Facility.

Auditable Safety Analysis and Final Hazard Classification for the 105-N Reactor Zone and 109-N Steam Generator Zone Facility

International Nuclear Information System (INIS)

Kloster, G.L.

1998-07-01

This document is a graded auditable safety analysis (ASA) and final hazard classification (FHC) for the Reactor/Steam Generator Zone Segment. The Reactor/Steam Generator Zone Segment, part of the N Reactor Complex, that is also known as the Reactor Building and Steam Generator Cells. The installation of the modifications described within to support surveillance and maintenance activities are to be completed by July 1, 1999. The surveillance and maintenance activities addressed within are assumed to continue for the next 15- 20 years, until the initiation of facility D ampersand D (i.e., Interim Safe Storage). The graded ASA in this document is in accordance with EDPI-4.30-01, Rev. 1, Safety Analysis Documentation, (BHI-DE-1) and is consistent with guidance provided by the U.S. Department of Energy. This ASA describes the hazards within the facility and evaluates the adequacy of the measures taken to reduce, control, or mitigate the identified hazards. This document also serves as the FHC for the Reactor/Steam Generator Zone Segment. This FHC is developed through the use of bounding accident analyses that envelope the potential exposures to personnel

Safety assessment for spent fuel storage facilities

International Nuclear Information System (INIS)

1994-01-01

This Safety Practice has been prepared as part of the IAEA's programme on the safety assessment of interim spent fuel storage facilities which are not an integral part of an operating nuclear power plant. This report provides general guidance on the safety assessment process, discussing both deterministic and probabilistic assessment methods. It describes the safety assessment process for normal operation and anticipated operational occurrences and also related to accident conditions. 10 refs, 2 tabs

Basis for Interim Operation for Fuel Supply Shutdown Facility

International Nuclear Information System (INIS)

BENECKE, M.W.

2003-01-01

This document establishes the Basis for Interim Operation (BIO) for the Fuel Supply Shutdown Facility (FSS) as managed by the 300 Area Deactivation Project (300 ADP) organization in accordance with the requirements of the Project Hanford Management Contract procedure (PHMC) HNF-PRO-700, ''Safety Analysis and Technical Safety Requirements''. A hazard classification (Benecke 2003a) has been prepared for the facility in accordance with DOE-STD-1027-92 resulting in the assignment of Hazard Category 3 for FSS Facility buildings that store N Reactor fuel materials (303-B, 3712, and 3716). All others are designated Industrial buildings. It is concluded that the risks associated with the current and planned operational mode of the FSS Facility (uranium storage, uranium repackaging and shipment, cleanup, and transition activities, etc.) are acceptable. The potential radiological dose and toxicological consequences for a range of credible uranium storage building have been analyzed using Hanford accepted methods. Risk Class designations are summarized for representative events in Table 1.6-1. Mitigation was not considered for any event except the random fire event that exceeds predicted consequences based on existing source and combustible loading because of an inadvertent increase in combustible loading. For that event, a housekeeping program to manage transient combustibles is credited to reduce the probability. An additional administrative control is established to protect assumptions regarding source term by limiting inventories of fuel and combustible materials. Another is established to maintain the criticality safety program. Additional defense-in-depth controls are established to perform fire protection system testing, inspection, and maintenance to ensure predicted availability of those systems, and to maintain the radiological control program. It is also concluded that because an accidental nuclear criticality is not credible based on the low uranium enrichment

Interim action record of decision remedial alternative selection: TNX area groundwater operable unit

International Nuclear Information System (INIS)

Palmer, E.R.

1994-10-01

This document presents the selected interim remedial action for the TNX Area Groundwater Operable Unit at the Savannah River Site (SRS), which was developed in accordance with CERCLA of 1980, as amended by the Superfund Amendments and Reauthorization Act (SARA) of 1986, and to the extent practicable, the National Oil and Hazardous Substances Pollution contingency Plan (NCP). This decision is based on the Administrative Record File for this specific CERCLA unit

Waste Encapsulation and Storage Facility (WESF) Interim Status Closure Plan

International Nuclear Information System (INIS)

SIMMONS, F.M.

2000-01-01

This document describes the planned activities and performance standards for closing the Waste Encapsulation and Storage Facility (WESF). WESF is located within the 225B Facility in the 200 East Area on the Hanford Facility. Although this document is prepared based on Title 40 Code of Federal Regulations (CFR), Part 265, Subpart G requirements, closure of the storage unit will comply with Washington Administrative Code (WAC) 173-303-610 regulations pursuant to Section 5.3 of the Hanford Federal Facility Agreement and Consent Order (Tri-Party Agreement) Action Plan (Ecology et al. 1996). Because the intention is to clean close WESF, postclosure activities are not applicable to this interim status closure plan. To clean close the storage unit, it will be demonstrated that dangerous waste has not been left onsite at levels above the closure performance standard for removal and decontamination. If it is determined that clean closure is not possible or environmentally is impracticable, the interim status closure plan will be modified to address required postclosure activities. WESF stores cesium and strontium encapsulated salts. The encapsulated salts are stored in the pool cells or process cells located within 225B Facility. The dangerous waste is contained within a double containment system to preclude spills to the environment. In the unlikely event that a waste spill does occur outside the capsules, operating methods and administrative controls require that waste spills be cleaned up promptly and completely, and a notation made in the operating record. Because dangerous waste does not include source, special nuclear, and by-product material components of mixed waste, radionuclides are not within the scope of this documentation. The information on radionuclides is provided only for general knowledge

Documented Safety Analysis for the Waste Storage Facilities March 2010

Energy Technology Data Exchange (ETDEWEB)

Laycak, D T

2010-03-05

This Documented Safety Analysis (DSA) for the Waste Storage Facilities was developed in accordance with 10 CFR 830, Subpart B, 'Safety Basis Requirements,' and utilizes the methodology outlined in DOE-STD-3009-94, Change Notice 3. The Waste Storage Facilities consist of Area 625 (A625) and the Decontamination and Waste Treatment Facility (DWTF) Storage Area portion of the DWTF complex. These two areas are combined into a single DSA, as their functions as storage for radioactive and hazardous waste are essentially identical. The B695 Segment of DWTF is addressed under a separate DSA. This DSA provides a description of the Waste Storage Facilities and the operations conducted therein; identification of hazards; analyses of the hazards, including inventories, bounding releases, consequences, and conclusions; and programmatic elements that describe the current capacity for safe operations. The mission of the Waste Storage Facilities is to safely handle, store, and treat hazardous waste, transuranic (TRU) waste, low-level waste (LLW), mixed waste, combined waste, nonhazardous industrial waste, and conditionally accepted waste generated at LLNL (as well as small amounts from other DOE facilities).

Erlotinib 150 mg daily plus chemotherapy in advanced pancreatic cancer: an interim safety analysis of a multicenter, randomized, cross-over phase III trial of the 'Arbeitsgemeinschaft Internistische Onkologie'.

Science.gov (United States)

Boeck, Stefan; Vehling-Kaiser, Ursula; Waldschmidt, Dirk; Kettner, Erika; Märten, Angela; Winkelmann, Cornelia; Klein, Stefan; Kojouharoff, Georgi; Gauler, Thomas; Fischer von Weikersthal, Ludwig; Clemens, Michael R; Geissler, Michael; Greten, Tim F; Hegewisch-Becker, Susanna; Neugebauer, Sascha; Heinemann, Volker

2010-01-01

To date, only limited toxicity data are available for the combination of erlotinib with either capecitabine or gemcitabine as front-line therapy for advanced pancreatic cancer. Within a randomized phase III trial, 281 treatment-naive patients were randomly assigned between capecitabine (2000 mg/m/day, for 14 days, once every 3 weeks) plus erlotinib (150 mg/day, arm A) and gemcitabine (1000 mg/m as a 30-min infusion) plus erlotinib (150 mg/day, arm B). In case of treatment failure, patients were crossed over to a second-line treatment with the comparator cytostatic drug without erlotinib. The primary study endpoint was the time to treatment failure of second-line therapy (TTF2). This interim analysis of toxicity contains safety data from the first 127 randomized patients. During first-line therapy, patients received a median number of three treatment cycles (range 0-13) in both the arms. Regarding chemotherapy, a treatment delay was observed in 12% of the cycles in arm A and in 22% of the cycles in arm B. Dose reductions of the cytostatic drug were performed in 18 and 27% of treatment cycles, respectively. Erlotinib dose reductions were performed in 6 and 11% of all cycles. Grade 3/4 hematological toxicity was arms; major grade 3/4 toxicities in arms A and B were diarrhea (9 vs. 7%), skin rash (4 vs. 12%), and hand-foot syndrome (7 vs. 0%). No treatment-related death was observed. In conclusion, this interim safety analysis suggests that treatment with erlotinib 150 mg/day is feasible in combination with capecitabine or gemcitabine.

CCSS Literacy and Math Tools: An Interim Report on Implementation and Sustainability during the Pilot Year

Science.gov (United States)

Reumann-Moore, Rebecca; Lawrence, Nancy; Sanders, Felicia; Shaw, Kate; Christman, Jolley Bruce

2011-01-01

This document summarizes the findings from the initial round of research on the development and piloting of two types of instructional tools designed to support teachers' integration of the Common Core State Standards (CCSS) in literacy and math. In this interim report, Research for Action (RFA) presents key findings from the first half of the…

Interim Hanford Waste Management Plan

International Nuclear Information System (INIS)

1985-09-01

The September 1985 Interim Hanford Waste Management Plan (HWMP) is the third revision of this document. In the future, the HWMP will be updated on an annual basis or as major changes in disposal planning at Hanford Site require. The most significant changes in the program since the last release of this document in December 1984 include: (1) Based on studies done in support of the Hanford Defense Waste Environmental Impact Statement (HDW-EIS), the size of the protective barriers covering contaminated soil sites, solid waste burial sites, and single-shell tanks has been increased to provide a barrier that extends 30 m beyond the waste zone. (2) As a result of extensive laboratory development and plant testing, removal of transuranic (TRU) elements from PUREX cladding removal waste (CRW) has been initiated in PUREX. (3) The level of capital support in years beyond those for which specific budget projections have been prepared (i.e., fiscal year 1992 and later) has been increased to maintain Hanford Site capability to support potential future missions, such as the extension of N Reactor/PUREX operations. The costs for disposal of Hanford Site defense wastes are identified in four major areas in the HWMP: waste storage and surveillance, technology development, disposal operations, and capital expenditures

Engineering and safety features of modular vault dry storage

International Nuclear Information System (INIS)

Deacon, D.; Wheeler, D.J.

1984-01-01

This paper discusses the need for interim dry storage and reviews detailed features of the Modular Vault Dry storage concept. The concept meets three basic utility requirements. Firstly, the technology and safety features have been demonstrated on existing plant; secondly, it can be built and licensed in an acceptably short timescale; and thirdly, economic analysis shows that a modular vault dry store is often the cheapest option for interim storage

Dry interim storage of radioactive material in Germany

International Nuclear Information System (INIS)

Drobniewski, Christian; Palmes, Julia

2013-01-01

In accordance with the waste management concept in Germany, spent fuel is stored in interim storage facilities for a period of up to 40 years until deposition in a geological repository. In twelve on-site interim storages in the vicinity or directly on the sites of the nuclear power plants, spent fuel elements from reactor operation are stored after the necessary period of decay in wet storage basins inside the reactors. Additionally, three central interim storage facilities for storage of spent fuel of different origin are in operation. The German facilities realize the concept of dry interim storage in metallic transport and storage casks. The confinement of the radioactive material is ensured by the double lid system of the casks, of which the leak tightness is monitored constantly. The casks are constructed to provide adequate heat removal and shielding of gamma and neutron radiation. Usually the storage facilities are halls of thick concrete structures, which ensure the removal of the decay heat by natural convection. The main safety goal of the storage concept is to prevent unnecessary exposure of persons, material goods and environment to ionizing radiation. Moreover any exposure should be kept as low as reasonable achievable. To reach this goal the containment of the radioactive materials, the disposal of decay heat, the sub criticality and the shielding of ionizing radiation has to be demonstrated by the applicant and verified by the licensing authority. In particular accidents, incidents and disasters have to be considered in the facility and cask design. This includes mechanical impacts onto the cask, internal and external fire, and environmental effects like wind, rain, snowfall, flood, earthquakes and landslides. In addition civilizatoric influences like plane crashes and explosions have to be taken into account. In all mentioned cases the secure confinement of the radioactive materials has to be ensured. On-site storage facilities have to consider the

Interim Administrators in Higher Education: A National Study

Science.gov (United States)

Huff, Marie Thielke; Neubrander, Judy

2015-01-01

The focus of this paper is on the roles and experiences of interim administrators in higher education. A survey was given to current and recent interim administrators in four-year public universities and colleges across the United States. The goals were to identify the advantages and disadvantages of using and serving as interims, and to solicit…

Interim report

Energy Technology Data Exchange (ETDEWEB)

NONE

1985-06-01

This Interim Report summarizes the research and development activities of the Superconducting Super Collider project carried out from the completion of the Reference Designs Study (May 1984) to June 1985. It was prepared by the SSC Central Design Group in draft form on the occasion of the DOE Annual Review, June 19--21, 1985. Now largely organized by CDG Divisions, the bulk of each chapter documents the progress and accomplishments to date, while the final section(s) describe plans for future work. Chapter 1, Introduction, provides a basic brief description of the SSC, its physics justification, its origins, and the R&D organization set up to carry out the work. Chapter 2 gives a summary of the main results of the R&D program, the tasks assigned to the four magnet R&D centers, and an overview of the future plans. The reader wishing a quick look at the SSC Phase I effort can skim Chapter 1 and read Chapter 2. Subsequent chapters discuss in more detail the activities on accelerator physics, accelerator systems, magnets and cryostats, injector, detector R&D, conventional facilities, and project planning and management. The magnet chapter (5) documents in text and photographs the impressive progress in successful construction of many model magnets, the development of cryostats with low heat leaks, and the improvement in current-carrying capacity of superconducting strand. Chapter 9 contains the budgets and schedules of the COG Divisions, the overall R&D program, including the laboratories, and also preliminary projections for construction. Appendices provide information on the various panels, task forces and workshops held by the CDG in FY 1985, a bibliography of COG and Laboratory reports on SSC and SSC-related work, and on private industrial involvement in the project.

Plutonium storage criteria

Energy Technology Data Exchange (ETDEWEB)

Chung, D. [Scientech, Inc., Germantown, MD (United States); Ascanio, X. [Dept. of Energy, Germantown, MD (United States)

1996-05-01

The Department of Energy has issued a technical standard for long-term (>50 years) storage and will soon issue a criteria document for interim (<20 years) storage of plutonium materials. The long-term technical standard, {open_quotes}Criteria for Safe Storage of Plutonium Metals and Oxides,{close_quotes} addresses the requirements for storing metals and oxides with greater than 50 wt % plutonium. It calls for a standardized package that meets both off-site transportation requirements, as well as remote handling requirements from future storage facilities. The interim criteria document, {open_quotes}Criteria for Interim Safe Storage of Plutonium-Bearing Solid Materials{close_quotes}, addresses requirements for storing materials with less than 50 wt% plutonium. The interim criteria document assumes the materials will be stored on existing sites, and existing facilities and equipment will be used for repackaging to improve the margin of safety.

Release of radionuclides following severe accident in interim storage facility. Source term determination

International Nuclear Information System (INIS)

Morandi, S.; Mariani, M.; Giacobbo, F.; Covini, R.

2006-01-01

Among the severe accidents that can cause the release of radionuclides from an interim storage facility, with a consequent relevant radiological impact on the population, there is the impact of an aircraft on the facility. In this work, a safety assessment analysis for the case of an aircraft crash into an interim storage facility is tackled. To this aim a methodology, based upon DOE, IAEA and NUREG standard procedures and upon conservative yet realistic hypothesis, has been developed in order to evaluate the total radioactivity, source term, released to the biosphere in consequence of the impact, without recurring to the use of complicated numerical codes. The procedure consists in the identification of the accidental scenarios, in the evaluation of the consequent damage to the building structures and to the waste packages and in the determination of the total release of radionuclides through the building-atmosphere interface. The methodology here developed has been applied to the case of an aircraft crash into an interim storage facility currently under design. Results show that in case of perforation followed by a fire incident the total released activity would be greater of some orders of magnitude with respect to the case of mere perforation. (author)

Decision on performing interim analysis for comparative clinical trials.

Science.gov (United States)

Pak, Kyongsun; Jacobus, Susanna; Uno, Hajime

2017-09-01

In randomized-controlled trials, interim analyses are often planned for possible early trial termination to claim superiority or futility of a new therapy. While unblinding is necessary to conduct the formal interim analysis in blinded studies, blinded data also have information about the potential treatment difference between the groups. We developed a blinded data monitoring tool that enables investigators to predict whether they observe such an unblinded interim analysis results that supports early termination of the trial. Investigators may skip some of the planned interim analyses if an early termination is unlikely. We specifically focused on blinded, randomized-controlled studies to compare binary endpoints of a new treatment with a control. Assuming one interim analysis is planned for early termination for superiority or futility, we conducted extensive simulation studies to assess the impact of the implementation of our tool on the size, power, expected number of interim analyses, and bias in the treatment effect. The numerical study showed the proposed monitoring tool does not affect size or power, but dramatically reduces the expected number of interim analyses when the effect of the treatment difference is small. The tool serves as a useful reference when interpreting the summary of the blinded data throughout the course of the trial, without losing integrity of the study. This tool could potentially save the study resources and budget by avoiding unnecessary interim analyses.

Addendum to IFMIF-CDA interim report

Energy Technology Data Exchange (ETDEWEB)

Maekawa, Hiroshi; Ida, Mizuho [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; eds.

1996-08-01

During the second IFMIF-CDA Design Integration Workshop, the conceptual design and contents of `IFMIF-CDA Interim Report` were examined and discussed at both general and group meetings. Based on these discussion, the final IFMIF-CDA Report will be modified from the `Interim Report`. This report describes the outline of these modification. (author)

Addendum to IFMIF-CDA interim report

International Nuclear Information System (INIS)

Maekawa, Hiroshi; Ida, Mizuho

1996-08-01

During the second IFMIF-CDA Design Integration Workshop, the conceptual design and contents of 'IFMIF-CDA Interim Report' were examined and discussed at both general and group meetings. Based on these discussion, the final IFMIF-CDA Report will be modified from the 'Interim Report'. This report describes the outline of these modification. (author)

Interim initial state report for the safety assessment SR-Can

Energy Technology Data Exchange (ETDEWEB)

Pers, Karin (ed.) [Kemakta Konsult AB, Stockholm (Sweden)

2004-07-01

A thorough description of the initial state of the engineered parts of the repository system is one of the main bases for the SR-Can safety assessment. The initial state refers to the state at the time of deposition for the spent fuel and the engineered barriers and the natural, undisturbed state at the time of beginning of excavation for the repository for the geosphere and the biosphere. The repository system is based on the KBS-3 method, where copper canisters with a cast iron insert containing spent nuclear fuel are surrounded by bentonite clay and deposited at approximately 500 m depth in saturated, granitic rock. For the purpose of the safety assessment the engineered portion of the repository system has been divided into a number of consecutive barriers or sub-systems. The importance of a particular feature for safety has influenced the resolution into components. In principle, components close to the source term and those that play an important role for safety are treated in more detail than more peripheral components. For the option with 40 years of reactor operation, the quantity of BWR fuel is estimated at 7200 tonnes and the quantity of PWR fuel at 2300 tonnes. The fuel burn-up may vary from 15 MWd/kgU up to 60 MWd/kg. Geometric aspects of the fuel cladding tubes of importance in the safety assessment are, as a rule, handled sufficiently pessimistically in analyses of radionuclide transport that differences between different fuel types are irrelevant. The relative differences in radionuclide inventory with respect to burn-up are small. Deviations in inventory and deviating or damaged fuel are not considered in the SR-Can interim reporting but will be handled in the final reporting of SR-Can. The canister consists of an inner container, the insert of cast iron and an outer shell of copper. The cast iron insert provides mechanical stability and the copper shell protects against corrosion in the repository environment. The copper shell is 5 cm thick and

Interim initial state report for the safety assessment SR-Can

International Nuclear Information System (INIS)

Pers, Karin

2004-07-01

A thorough description of the initial state of the engineered parts of the repository system is one of the main bases for the SR-Can safety assessment. The initial state refers to the state at the time of deposition for the spent fuel and the engineered barriers and the natural, undisturbed state at the time of beginning of excavation for the repository for the geosphere and the biosphere. The repository system is based on the KBS-3 method, where copper canisters with a cast iron insert containing spent nuclear fuel are surrounded by bentonite clay and deposited at approximately 500 m depth in saturated, granitic rock. For the purpose of the safety assessment the engineered portion of the repository system has been divided into a number of consecutive barriers or sub-systems. The importance of a particular feature for safety has influenced the resolution into components. In principle, components close to the source term and those that play an important role for safety are treated in more detail than more peripheral components. For the option with 40 years of reactor operation, the quantity of BWR fuel is estimated at 7200 tonnes and the quantity of PWR fuel at 2300 tonnes. The fuel burn-up may vary from 15 MWd/kgU up to 60 MWd/kg. Geometric aspects of the fuel cladding tubes of importance in the safety assessment are, as a rule, handled sufficiently pessimistically in analyses of radionuclide transport that differences between different fuel types are irrelevant. The relative differences in radionuclide inventory with respect to burn-up are small. Deviations in inventory and deviating or damaged fuel are not considered in the SR-Can interim reporting but will be handled in the final reporting of SR-Can. The canister consists of an inner container, the insert of cast iron and an outer shell of copper. The cast iron insert provides mechanical stability and the copper shell protects against corrosion in the repository environment. The copper shell is 5 cm thick and

Retention of long-term interim restorations with sodium fluoride enriched interim cement

Science.gov (United States)

Strash, Carolyn

Purpose: Interim fixed dental prostheses, or "provisional restorations", are fabricated to restore teeth when definitive prostheses are made indirectly. Patients undergoing extensive prosthodontic treatment frequently require provisionalization for several months or years. The ideal interim cement would retain the restoration for as long as needed and still allow for ease of removal. It would also avoid recurrent caries by preventing demineralization of tooth structure. This study aims to determine if adding sodium fluoride varnish to interim cement may assist in the retention of interim restorations. Materials and methods: stainless steel dies representing a crown preparation were fabricated. Provisional crowns were milled for the dies using CAD/CAM technology. Crowns were provisionally cemented onto the dies using TempBond NE and NexTemp provisional cements as well as a mixture of TempBond NE and Duraphat fluoride varnish. Samples were stored for 24h then tested or thermocycled for 2500 or 5000 cycles before being tested. Retentive strength of each cement was recorded using a universal testing machine. Results: TempBond NE and NexTemp cements performed similarly when tested after 24h. The addition of Duraphat significantly decreased the retention when added to TempBond NE. NexTemp cement had high variability in retention over all tested time periods. Thermocycling for 2500 and 5000 cycles significantly decreased the retention of all cements. Conclusions: The addition of Duraphat fluoride varnish significantly decreased the retention of TempBond NE and is therefore not recommended for clinical use. Thermocycling significantly reduced the retention of TempBond NE and NexTemp. This may suggest that use of these cements for three months, as simulated in this study, is not recommended.

Hazardous Waste/Mixed Waste Treatment Building Safety Information Document (SID)

International Nuclear Information System (INIS)

Fatell, L.B.; Woolsey, G.B.

1993-01-01

This Safety Information Document (SID) provides a description and analysis of operations for the Hazardous Waste/Mixed Waste Disposal Facility Treatment Building (the Treatment Building). The Treatment Building has been classified as a moderate hazard facility, and the level of analysis performed and the methodology used are based on that classification. Preliminary design of the Treatment Building has identified the need for two separate buildings for waste treatment processes. The term Treatment Building applies to all these facilities. The evaluation of safety for the Treatment Building is accomplished in part by the identification of hazards associated with the facility and the analysis of the facility's response to postulated events involving those hazards. The events are analyzed in terms of the facility features that minimize the causes of such events, the quantitative determination of the consequences, and the ability of the facility to cope with each event should it occur. The SID presents the methodology, assumptions, and results of the systematic evaluation of hazards associated with operation of the Treatment Building. The SID also addresses the spectrum of postulated credible events, involving those hazards, that could occur. Facility features important to safety are identified and discussed in the SID. The SID identifies hazards and reports the analysis of the spectrum of credible postulated events that can result in the following consequences: Personnel exposure to radiation; Radioactive material release to the environment; Personnel exposure to hazardous chemicals; Hazardous chemical release to the environment; Events leading to an onsite/offsite fatality; and Significant damage to government property. The SID addresses the consequences to the onsite and offsite populations resulting from postulated credible events and the safety features in place to control and mitigate the consequences

Hazardous Waste/Mixed Waste Treatment Building Safety Information Document (SID)

Energy Technology Data Exchange (ETDEWEB)

Fatell, L.B.; Woolsey, G.B.

1993-04-15

This Safety Information Document (SID) provides a description and analysis of operations for the Hazardous Waste/Mixed Waste Disposal Facility Treatment Building (the Treatment Building). The Treatment Building has been classified as a moderate hazard facility, and the level of analysis performed and the methodology used are based on that classification. Preliminary design of the Treatment Building has identified the need for two separate buildings for waste treatment processes. The term Treatment Building applies to all these facilities. The evaluation of safety for the Treatment Building is accomplished in part by the identification of hazards associated with the facility and the analysis of the facility`s response to postulated events involving those hazards. The events are analyzed in terms of the facility features that minimize the causes of such events, the quantitative determination of the consequences, and the ability of the facility to cope with each event should it occur. The SID presents the methodology, assumptions, and results of the systematic evaluation of hazards associated with operation of the Treatment Building. The SID also addresses the spectrum of postulated credible events, involving those hazards, that could occur. Facility features important to safety are identified and discussed in the SID. The SID identifies hazards and reports the analysis of the spectrum of credible postulated events that can result in the following consequences: Personnel exposure to radiation; Radioactive material release to the environment; Personnel exposure to hazardous chemicals; Hazardous chemical release to the environment; Events leading to an onsite/offsite fatality; and Significant damage to government property. The SID addresses the consequences to the onsite and offsite populations resulting from postulated credible events and the safety features in place to control and mitigate the consequences.

The Nord interim store

International Nuclear Information System (INIS)

Leushacke, D.F.; Rittscher, D.

1996-01-01

In line with the decision taken in 1990 to shut down and decommission the Greifswald and Rheinsberg Nuclear Power Stations, the waste management concept of the Energiewerke Nord is based on direct and complete decommissioning of the six shut down reactor units within the next fifteen years. One key element of this concept is the construction and use of the Zwischenlager Nord (Nord Interim Store, ZLN) for holding the existing nuclear fuels and for interim and decay storage of the radioactive materials arising in decommissioning and demolition. The owner and operator of the store is Energiewerke Nord GmbH. The interim store has the functions of a processing and Energiewerke Nord GmbH. The interim store has the functions of a processing and treatment station and buffer store for the flows of residues arising. As a radioactive waste management station, it accommodates nuclear fuels, radioactive waste or residues which are not treated any further. It is used as a buffer store to allow the materials accumulating in disassembly to be stored temporarily before or after treatment in order to ensure continuous loading of the treatment plants. When operated as a processing station, the ZLN is able to handle nearly all types of radioactive waste and residues arising, except for nuclear fuels. These installations allow the treatment of radioactive residues to be separated from the demolition work both physically and in time. The possibilities of interium storage and buffer storage of untreated waste and waste packages make for high flexibility in logistics and waste management strategy. (orig.) [de

Area 5 Radioactive Waste Management Site Safety Assessment Document

International Nuclear Information System (INIS)

Horton, K.K.; Kendall, E.W.; Brown, J.J.

1980-02-01

The Area 5 Radioactive Waste Management Safety Assessment Document evaluates site characteristics, facilities and operating practices which contribute to the safe handling and storage/disposal of radioactive wastes at the Nevada Test Site. Physical geography, cultural factors, climate and meteorology, geology, hydrology (with emphasis on radionuclide migration), ecology, natural phenomena, and natural resources are discussed and determined to be suitable for effective containment of radionuclides. Also considered, as a separate section, are facilities and operating practices such as monitoring; storage/disposal criteria; site maintenance, equipment, and support; transportation and waste handling; and others which are adequate for the safe handling and storage/disposal of radioactive wastes. In conclusion, the Area 5 Radioactive Waste Management Site is suitable for radioactive waste handling and storage/disposal for a maximum of twenty more years at the present rate of utilization

Reactor operation safety information document

Energy Technology Data Exchange (ETDEWEB)

1990-01-01

The report contains a reactor facility description which includes K, P, and L reactor sites, structures, operating systems, engineered safety systems, support systems, and process and effluent monitoring systems; an accident analysis section which includes cooling system anomalies, radioactive materials releases, and anticipated transients without scram; a summary of onsite doses from design basis accidents; severe accident analysis (reactor core disruption); a description of operating contractor organization and emergency planning; and a summary of reactor safety evolution. (MB)

Advanced nuclear reactor public opinion project. Interim report

Energy Technology Data Exchange (ETDEWEB)

Benson, B.

1991-07-25

This Interim Report summarizes the findings of our first twenty in-depth interviews in the Advanced Nuclear Reactor Public Opinion Project. We interviewed 6 industry trade association officials, 3 industry attorneys, 6 environmentalists/nuclear critics, 3 state officials, and 3 independent analysts. In addition, we have had numerous shorter discussions with various individuals concerned about nuclear power. The report is organized into the four categories proposed at our April, 1991, Advisory Group meeting: safety, cost-benefit analysis, science education, and communications. Within each category, some change of focus from that of the Advisory Group has been required, to reflect the findings of our interviews. This report limits itself to describing our findings. An accompanying memo draws some tentative conclusions.

Interim report and accounts 1993/94

International Nuclear Information System (INIS)

1993-01-01

An interim set of accounts and reports is presented here for 1993/1994 for the health science company Amersham International. The company's research programs focus on developments in life science research, nuclear medicine and industrial quality and safety assurance, with particular expertise in the application of radioactivity to labelling and detection at the molecular level. This report which covers the half-year to 30 September 1993 shows promising financial results, with turnover, operating profits and earnings per share all having risen. All life science markets report growth although difficult trading conditions are being reported in Europe. Two products in the Healthcare business have achieved progress, a pain palliation agent for bone metastases has been launched in the United States, and European approval has been gained for a new technetium based heart imaging agent. Further growth is expected for the company. (UK)

Proceedings of the Topical Meeting on the safety of nuclear fuel cycle intermediate storage facilities

International Nuclear Information System (INIS)

1998-01-01

The CSNI Working Group on Fuel Cycle Safety held an International Topical Meeting on safety aspects of Intermediate Storage Facilities in Newby Bridge, England, from 28 to 30 October 1997. The main purpose of the meeting was to provide a forum for the exchange of information on the technical issues on the safety of nuclear fuel cycle facilities (intermediate storage). Titles of the papers are: An international view on the safety challenges to interim storage of spent fuel. Interim storage of intermediate and high-level waste in Belgium: a description and safety aspects. Encapsulated intermediate level waste product stores at Sellafield. Safety of interim storage facilities of spent fuel: the international dimension and the IAEA's activities. Reprocessing of irradiated fuel and radwaste conditioning at Belgoprocess site: an overview. Retrieval of wastes from interim storage silos at Sellafield. Outline of the fire and explosion of the bituminization facility and the activities of the investigation committee (STAIJAERI). The fire and explosion incident of the bituminization facility and the lessons learned from the incident. Study on the scenario of the fire incident and related analysis. Study on the scenario of the explosion incident and related analysis. Accident investigation board report on the May 14, 1997 chemical explosion at the plutonium reclamation facility, Hanford site, Richland, Washington. Dry interim storage of spent nuclear fuel elements in Germany. Safe and effective system for the bulk receipt and storage of light water reactor fuel prior to reprocessing. Receiving and storage of glass canisters at vitrified waste storage center of Japan Nuclear Fuel Ltd. Design and operational experience of dry cask storage systems. Sellafield MOX plant; Plant safety design (BNFL). The assessment of fault studies for intermediate term waste storage facilities within the UK nuclear regulatory regime. Non-active and active commissioning of the thermal oxide

Evaluation of Hose in Hose Transfer Line Service Life for Hanford's Interim Stabilization Program

International Nuclear Information System (INIS)

TORRES, T.D.

2000-01-01

RPP-6153, Engineering Task Plan for Hose-in-Hose Transfer System for the Interim Stabilization Program, defines the programmatic goals, functional requirements, and technical criteria for the development and subsequent installation of transfer line equipment to support Hanford's Interim Stabilization Program. RPP-6028, Specification for Hose in Hose Transfer Lines for Hanford's Interim Stabilization Program, has been issued to define the specific requirements for the design, manufacture, and verification of transfer line assemblies for specific waste transfer applications. Included in RPP-6028 are tables defining the chemical constituents of concern to which transfer lines will be exposed. Current Interim Stabilization Program planning forecasts that the at-grade transfer lines will be required to convey pumpable waste for as much as three years after commissioning. Prudent engineering dictates that the equipment placed in service have a working life in excess of this forecasted time period, with some margin to allow for future adjustments to the planned schedule. This document evaluates the effective service life of the Hose-in-Hose Transfer Lines, based on information submitted by the manufacturer and published literature. The effective service life of transfer line assemblies is a function of several factors. Foremost among these are process fluid characteristics, ambient environmental conditions, and the manufacturer's stated shelf life. This evaluation examines the manufacturer's certification of shelf life, the manufacturer's certifications of chemical compatibility with waste, and published literature on the effects of exposure to ionizing radiation on the mechanical properties of elastomeric materials to evaluate transfer line service life

Documenting Quality Improvement and Patient Safety Efforts: The Quality Portfolio. A Statement from the Academic Hospitalist Taskforce

OpenAIRE

Taylor, Benjamin B.; Parekh, Vikas; Estrada, Carlos A.; Schleyer, Anneliese; Sharpe, Bradley

2013-01-01

Physicians increasingly investigate, work, and teach to improve the quality of care and safety of care delivery. The Society of General Internal Medicine Academic Hospitalist Task Force sought to develop a practical tool, the quality portfolio, to systematically document quality and safety achievements. The quality portfolio was vetted with internal and external stakeholders including national leaders in academic medicine. The portfolio was refined for implementation to include an outlined fr...

Practice specific model regulations: Radiation safety of non-medical irradiation facilities. Interim report for comment

International Nuclear Information System (INIS)

2003-08-01

the infrastructure aimed at achieving its maximum efficiency, and extensively covers performance regulations. The BSS cover the application of ionizing radiation for all practices and interventions and are, therefore, basic and general in nature. Users must apply these basic requirements to their own particular practices. In this context, the preamble of the BSS states that: 'The Regulatory Authority may need to provide guidance on how certain regulatory requirements are to be fulfilled for various practices, for example in regulatory guideline documents.' There are certain requirements that, when applied to specific practices, can be fulfilled through virtually only one practical solution. In these cases, the regulatory authority would use a 'shall' statement for this solution. To meet other requirements, there may be more than one option. In these cases the regulatory authority would usually indicate the recommended option with a 'should' statement, which implies that licensees may choose another alternative provided that the level of safety is equivalent. This distinction has been maintained in this 'model regulations' for irradiation facilities in order to facilitate the decision of regulatory authorities on the degree of obligation

ISOE EG-SAM interim report - Report on behalf of the Sub expert Group

International Nuclear Information System (INIS)

Harris, Willie; Miller, David W.; Djeffal, Salah; Anderson, Ellen; Couasnon, Olivier; Hagemeyer, Derek; Sovijarvi, Jukka; Amaral, Marcos A.; Tarzia, J.P.; Schmidt, Claudia; Fritioff, Karin; Kaulard, Joerg; Lance, Benoit; Fritioff, Karin; Schieber, Caroline; Hayashida, Yoshihisa; Doty, Rick

2014-01-01

During its November 2012 meeting, the expert group decided to develop an interim (preliminary) report before the end of 2013 (with a general perspective and discussion of specific severe accident management worker dose issues), and to finalize the report by organizing the international workshop of 2014 to address national experiences, which will be incorporated to the report. The work of the EG-SAM focuses on radiation protection management and organization, radiation protection training and exercises related to severe accident management, facility configuration and readiness, worker protection, radioactive materials, contamination controls and logistics and key lessons learned especially from the TMI, Chernobyl and Fukushima Dai-ichi accidents. This interim report was completed through intensive work of all Group members nominated by the ISOE, and was accomplished during EG-SAM meetings through 2012-2013. This document gathers the different presentations given by the sub expert groups in charge of each chapter of the report

Operations and Maintenance Concept Plan for the Immobilized High Level Waste (IHLW) Interim Storage Facility

Energy Technology Data Exchange (ETDEWEB)

JANIN, L.F.

2000-08-30

This O&M Concept looks at the future operations and maintenance of the IHLW/CSB interim storage facility. It defines the overall strategy, objectives, and functional requirements for the portion of the building to be utilized by Project W-464. The concept supports the tasks of safety basis planning, risk mitigation, alternative analysis, decision making, etc. and will be updated as required to support the evolving design.

Operations and Maintenance Concept Plan for the Immobilized High-Level Waste (IHLW) Interim Storage Facility

International Nuclear Information System (INIS)

JANIN, L.F.

2000-01-01

This OandM Concept looks at the future operations and maintenance of the IHLW/CSB interim storage facility. It defines the overall strategy, objectives, and functional requirements for the portion of the building to be utilized by Project W-464. The concept supports the tasks of safety basis planning, risk mitigation, alternative analysis, decision making, etc. and will be updated as required to support the evolving design

Software Verification and Validation Report for the 244-AR Vault Interim Stabilization Ventilation System

International Nuclear Information System (INIS)

YEH, T.

2002-01-01

This document reports on the analysis, testing and conclusions of the software verification and validation for the 244-AR Vault Interim Stabilization ventilation system. Automation control system will use the Allen-Bradley software tools for programming and programmable logic controller (PLC) configuration. The 244-AR Interim Stabilization Ventilation System will be used to control the release of radioactive particles to the environment in the containment tent, located inside the canyon of the 244-AR facility, and to assist the waste stabilization efforts. The HVAC equipment, ducts, instruments, PLC hardware, the ladder logic executable software (documented code), and message display terminal are considered part of the temporary ventilation system. The system consists of a supply air skid, temporary ductwork (to distribute airflow), and two skid-mounted, 500-cfm exhausters connected to the east filter building and the vessel vent system. The Interim Stabilization Ventilation System is a temporary, portable ventilation system consisting of supply side and exhaust side. Air is supplied to the containment tent from an air supply skid. This skid contains a constant speed fan, a pre-filter, an electric heating coil, a cooling coil, and a constant flow device (CFD). The CFD uses a passive component that allows a constant flow of air to pass through the device. Air is drawn out of the containment tent, cells, and tanks by two 500-cfm exhauster skids running in parallel. These skids are equipped with fans, filters, stack, stack monitoring instrumentation, and a PLC for control. The 500CFM exhaust skids were fabricated and tested previously for saltwell pumping activities. The objective of the temporary ventilation system is to maintain a higher pressure to the containment tent, relative to the canyon and cell areas, to prevent contaminants from reaching the containment tent

Implementation plan for deployment of Federal Interim Storage facilities for commercial spent nuclear fuel

International Nuclear Information System (INIS)

1985-01-01

This document is the second annual report on plans for providing Federal Interim Storage (FIS) capacity. References are made to the first annual report as necessary (DOE/RW-0003, 1984). Background factors and aspects that were considered in the development of this deployment plan and activities and interactions considered to be required to implement an FIS program are discussed. The generic approach that the Department plans to follow in deploying FIS facilities is also described

Radiation shielding and dose rate evaluation at the interim storage facility for spent fuel from Cernavoda NPP

International Nuclear Information System (INIS)

Stanciu, Marcela; Mateescu, Silvia; Pantazi, Doina; Penescu, Maria

2000-01-01

At present studies necessary to license the Interim Storage Facility for the Spent Fuel (CANDU type) from Cernavoda NPP are developed in our country.The spent fuel from Cernavoda NPP is discharged into Spent Fuel Bay in Service Building of the plant, where it remains several years for cooling. After this period, the bundles of spent fuel are to be transferred to the Interim Storage Facility.The dry interim storage solution seems to be the most appropriate variant for Cernavoda NPP.The design of the Spent Fuel Interim Storage Facility must meet the applicable safety requirements in order to ensure radiological protection of the personnel, public and environment during all phases of the facility achievement. In this paper we intend to present the calculation of radiation shielding at the spent fuel interim storage facility for two technical solutions: - Concrete Monolithic Module and Concrete Storage Cask. In order to quantify the fuel composition after irradiation, the isotope generation and depletion code ORIGEN 2.1 has been used, taking into account a cooling time of 7 years and 9 years, respectively, for these two variants. The shielding calculations have been performed using the computer codes QAD-5K and MICROSHIELD-4. The evaluations refer only to gamma radiation because the resulting neutron source (from (α,n) reactions and spontaneous fission) is insignificant as compared to the gamma source. The final results consist in the minimum thickness of the shielding and the corresponding external dose rates, ensuring a design average dose rate based on national and international regulations. (authors)

FLAMMABLE GAS TECHNICAL BASIS DOCUMENT

Energy Technology Data Exchange (ETDEWEB)

KRIPPS, L.J.

2005-03-03

This document describes the qualitative evaluation of frequency and consequences for DST and SST representative flammable gas accidents and associated hazardous conditions without controls. The evaluation indicated that safety-significant structures, systems and components (SSCs) and/or technical safety requirements (TSRs) were required to prevent or mitigate flammable gas accidents. Discussion on the resulting control decisions is included. This technical basis document was developed to support WP-13033, Tank Farms Documented Safety Analysis (DSA), and describes the risk binning process for the flammable gas representative accidents and associated represented hazardous conditions. The purpose of the risk binning process is to determine the need for safety-significant structures, systems, and components (SSC) and technical safety requirement (TSR)-level controls for a given representative accident or represented hazardous condition based on an evaluation of the event frequency and consequence.

76 FR 58790 - Notice of Interim Approval

Science.gov (United States)

2011-09-22

... to the customers. Rate Scenario 3--Original Cumberland Marketing Policy The third rate alternative... an interim basis to the customers. Rate Scenario 3--Original Cumberland Marketing Policy The third... allocated on an interim basis to the customers. Rate Scenario 3--Original Cumberland Marketing Policy The...

The Interim Financial Statements: The Case of Greece

OpenAIRE

Rogdaki, E.I.; Kazantzis, Ch.

1999-01-01

The following paper refers to the accounting and auditing issues which emerge in the preparation of the interim financial statements of the companies: Firstly, the interim financial statements are defined as being the financial statements that provide useful information about the financial position and the financial results of a company which are realized and accrued during the fiscal year. The interim financial statements can be prepared on a monthly basis, on a quarterly basis or covering a...

78 FR 42012 - Safety Zone and Regulated Navigation Area; Chicago Sanitary and Ship Canal, Romeoville, IL

Science.gov (United States)

2013-07-15

... is issuing this Interim Rule to address two omissions from the regulatory text of the Safety zone and... boats, etc.). This revision is intended to make the regulatory text consistent with the discussion of... of Homeland Security IR Interim Rule NPRM Notice of Proposed Rulemaking RNA Regulated Navigation Area...

Safety Aspects for Vertical Wall Breakwaters

DEFF Research Database (Denmark)

Sørensen, John Dalsgaard; Burcharth, H. F.; Christiani, E.

1996-01-01

In this appendix some safety aspects in relation to vertical wall breakwaters are discussed. Breakwater structures such as vertical wall breakwaters are used under quite different conditions. The expected lifetime can be from 5 years (interim structure) to 100 years (permanent structure) and the ...

Effectiveness of interim remedial actions at a radioactive waste facility

International Nuclear Information System (INIS)

Devgun, J.S.; Beskid, N.J.; Peterson, J.M.; Seay, W.M.; McNamee, E.

1989-01-01

Over the past eight years, several interim remedial actions have been taken at the Niagara Falls Storage Site (NFSS), primarily to reduce radon and gamma radiation exposures and to consolidate radioactive waste into a waste containment facility. Interim remedial actions have included capping of vents, sealing of pipes, relocation of the perimeter fence (to limit radon risk), transfer and consolidation of waste, upgrading of storage buildings, construction of a clay cutoff wall (to limit the potential groundwater transport of contaminants), treatment and release of contaminated water, interim use of a synthetic liner, and emplacement of an interim clay cap. An interim waste containment facility was completed in 1986. 6 refs., 3 figs

Development of a probabilistic safety assessment framework for an interim dry storage facility subjected to an aircraft crash using best-estimate structural analysis

International Nuclear Information System (INIS)

Almomani, Belal; Jang, Dong Chan; Lee, Sang Hoon; Kang, Hyun Gook

2017-01-01

Using a probabilistic safety assessment, a risk evaluation framework for an aircraft crash into an interim spent fuel storage facility is presented. Damage evaluation of a detailed generic cask model in a simplified building structure under an aircraft impact is discussed through a numerical structural analysis and an analytical fragility assessment. Sequences of the impact scenario are shown in a developed event tree, with uncertainties considered in the impact analysis and failure probabilities calculated. To evaluate the influence of parameters relevant to design safety, risks are estimated for three specification levels of cask and storage facility structures. The proposed assessment procedure includes the determination of the loading parameters, reference impact scenario, structural response analyses of facility walls, cask containment, and fuel assemblies, and a radiological consequence analysis with dose–risk estimation. The risk results for the proposed scenario in this study are expected to be small relative to those of design basis accidents for best-estimated conservative values. The importance of this framework is seen in its flexibility to evaluate the capability of the facility to withstand an aircraft impact and in its ability to anticipate potential realistic risks; the framework also provides insight into epistemic uncertainty in the available data and into the sensitivity of the design parameters for future research

Development of a Probabilistic Safety Assessment Framework for an Interim Dry Storage Facility Subjected to an Aircraft Crash Using Best-Estimate Structural Analysis

Directory of Open Access Journals (Sweden)

Belal Almomani

2017-03-01

Full Text Available Using a probabilistic safety assessment, a risk evaluation framework for an aircraft crash into an interim spent fuel storage facility is presented. Damage evaluation of a detailed generic cask model in a simplified building structure under an aircraft impact is discussed through a numerical structural analysis and an analytical fragility assessment. Sequences of the impact scenario are shown in a developed event tree, with uncertainties considered in the impact analysis and failure probabilities calculated. To evaluate the influence of parameters relevant to design safety, risks are estimated for three specification levels of cask and storage facility structures. The proposed assessment procedure includes the determination of the loading parameters, reference impact scenario, structural response analyses of facility walls, cask containment, and fuel assemblies, and a radiological consequence analysis with dose–risk estimation. The risk results for the proposed scenario in this study are expected to be small relative to those of design basis accidents for best-estimated conservative values. The importance of this framework is seen in its flexibility to evaluate the capability of the facility to withstand an aircraft impact and in its ability to anticipate potential realistic risks; the framework also provides insight into epistemic uncertainty in the available data and into the sensitivity of the design parameters for future research.

Development of a probabilistic safety assessment framework for an interim dry storage facility subjected to an aircraft crash using best-estimate structural analysis

Energy Technology Data Exchange (ETDEWEB)

Almomani, Belal; Jang, Dong Chan [Dept. of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of); Lee, Sang Hoon [Dept. of Mechanical and Automotive Engineering, Keimyung University, Daegu (Korea, Republic of); Kang, Hyun Gook [Dept. of Mechanical, Aerospace and Nuclear Engineering, Rensselaer Polytechnic Institute, Troy (United States)

2017-03-15

Using a probabilistic safety assessment, a risk evaluation framework for an aircraft crash into an interim spent fuel storage facility is presented. Damage evaluation of a detailed generic cask model in a simplified building structure under an aircraft impact is discussed through a numerical structural analysis and an analytical fragility assessment. Sequences of the impact scenario are shown in a developed event tree, with uncertainties considered in the impact analysis and failure probabilities calculated. To evaluate the influence of parameters relevant to design safety, risks are estimated for three specification levels of cask and storage facility structures. The proposed assessment procedure includes the determination of the loading parameters, reference impact scenario, structural response analyses of facility walls, cask containment, and fuel assemblies, and a radiological consequence analysis with dose–risk estimation. The risk results for the proposed scenario in this study are expected to be small relative to those of design basis accidents for best-estimated conservative values. The importance of this framework is seen in its flexibility to evaluate the capability of the facility to withstand an aircraft impact and in its ability to anticipate potential realistic risks; the framework also provides insight into epistemic uncertainty in the available data and into the sensitivity of the design parameters for future research.

Implementation plan for deployment of Federal Interim Storage facilities for commercial spent nuclear fuel

International Nuclear Information System (INIS)

1986-12-01

This document is the third annual report on plans for providing Federal Interim Storage (FIS) capacity. References are made to the first and second annual reports, as necessary. Background factors and aspects that were considered in the development of this deployment plan and activities and interactions considered to be required to implement an FIS program are discussed. A generic description of the approach that the Department plans to follow in deploying FIS facilities is also described

Interim remedial action work plan for the cesium plots at Waste Area Grouping 13 at Oak Ridge National Laboratory, Oak Ridge, Tennessee

International Nuclear Information System (INIS)

1993-07-01

This remedial action work plan (RAWP) is issued under the Federal Facility Agreement to provide a basic approach for implementing the interim remedial action (IRA) described in Interim Record of Decision for the Oak Ridge National Laboratory Waste Area Grouping 13 Cesium Plots, Oak Ridge, Tennessee. This RAWP summarizes the interim record of decision (IROD) requirements and establishes the strategy for the implementation of the field activities. As documented in the IROD document, the primary goal of this action is to reduce the risk to human health and the environment resulting from current elevated levels of gamma radiation on the site and at areas accessible to the public adjacent to the site. The major steps of this IRA are to: Excavate cesium-contaminated soil; place the excavated soils in containers and transport to Waste Area Grouping (WAG) 6; and backfill excavated plots with clean fill materials. The actual remedial action will be performed by Department of Energy prime contractor, MK-Ferguson of Oak Ridge Company. Remediation of the cesium plots will require approximately 60 days to complete. During this time, all activities will be performed according to this RAWP and the applicable specifications, plans, and procedures referred to in this document. The IRA on WAG 13 will prevent a known source of cesium-contaminated soil from producing elevated levels of gamma radiation in areas accessible to the public, eliminate sources of contamination to the environment, and reduce the risks associated with surveillance and maintenance of the WAG 13 site

Interim licensing criteria for physical protection of certain storage of spent fuel

International Nuclear Information System (INIS)

Dwyer, P.A.

1994-11-01

This document presents interim criteria to be used in the physical protection licensing of certain spent fuel storage installations. Installations that will be reviewed under this criteria are those that store power reactor spent fuel at decommissioned power reactor sites; independent spent fuel storage installations located outside of the owner controlled area of operating nuclear power reactors; monitored retrievable storage installations owned by the Department of Energy, designed and constructed specifically for the storage, of spent fuel; the proposed geologic repository operations area; or permanently shutdown power reactors still holding a Part 50 license. This criteria applies to both dry cask and pool storage. However, the criteria in this document does not apply to the storage of spent fuel within the owner-controlled area of operating nuclear power reactors

Operable Unit 3: Proposed Plan/Environmental Assessment for interim remedial action

International Nuclear Information System (INIS)

1993-12-01

This document presents a Proposed Plan and an Environmental Assessment for an interim remedial action to be undertaken by the US Department of Energy (DOE) within Operable Unit 3 (OU3) at the Fernald Environmental Management Project (FEMP). This proposed plan provides site background information, describes the remedial alternatives being considered, presents a comparative evaluation of the alternatives and a rationnale for the identification of DOE's preferred alternative, evaluates the potential environmental and public health effects associated with the alternatives, and outlines the public's role in helping DOE and the EPA to make the final decision on a remedy

Second interim assessment of the Canadian concept for nuclear fuel waste disposal. Volume 1

International Nuclear Information System (INIS)

Wuschke, D.M.; Gillespie, P.A.; Main, D.E.

1985-07-01

The nuclear fuel waste disposal concept chosen for development and assessment in Canada involves the isolation of corrosion-resistant containers of waste in a vault located deep in plutonic rock. As the concept and the assessment tools are developed, periodic assessments are performed to permit evaluation of the methodology and provide feedback to those developing the concept. The ultimate goal of these assessments is to predict what impact the disposal system would have on man and the environment if the concept were implemented. The second assessment was performed in 1984 and is documented in the Second Interim assessment of the Canadian Concept for Nuclear Fuel Waste Disposal Volumes 1 to 4. This volume, entitled Summary, is a condensation of Volumes 2, 3 and 4. It briefly describes the Canadian nuclear fuel waste disposal concept, and the methods and results of the second interim pre-closure and post-closure assessments of that concept. 46 refs

Development of Onsite Transportation Safety Documents for Nevada Test Site

International Nuclear Information System (INIS)

Frank Hand; Willard Thomas; Frank Sciacca; Manny Negrete; Susan Kelley

2008-01-01

Department of Energy (DOE) Orders require each DOE site to develop onsite transportation safety documents (OTSDs). The Nevada Test Site approach divided all onsite transfers into two groups with each group covered by a standalone OTSD identified as Non-Nuclear and Nuclear. The Non-Nuclear transfers involve all radioactive hazardous material in less than Hazard Category (HC)-3 quantities and all chemically hazardous materials. The Nuclear transfers involve all radioactive material equal to or greater than HC-3 quantities and radioactive material mated with high explosives regardless of quantity. Both OTSDs comply with DOE O 460.1B requirements. The Nuclear OTSD also complies with DOE O 461.1A requirements and includes a DOE-STD-3009 approach to hazard analysis (HA) and accident analysis as needed. All Nuclear OTSD proposed transfers were determined to be non-equivalent and a methodology was developed to determine if 'equivalent safety' to a fully compliant Department of Transportation (DOT) transfer was achieved. For each HA scenario, three hypothetical transfers were evaluated: a DOT-compliant, uncontrolled, and controlled transfer. Equivalent safety is demonstrated when the risk level for each controlled transfer is equal to or less than the corresponding DOT-compliant transfer risk level. In this comparison the typical DOE-STD-3009 risk matrix was modified to reflect transportation requirements. Design basis conditions (DBCs) were developed for each non-equivalent transfer. Initial DBCs were based solely upon the amount of material present. Route-, transfer-, and site-specific conditions were evaluated and the initial DBCs revised as needed. Final DBCs were evaluated for each transfer's packaging and its contents

An Approach for Evaluating the Technical Quality of Interim Assessments

Science.gov (United States)

Li, Ying; Marion, Scott; Perie, Marianne; Gong, Brian

2010-01-01

Increasing numbers of schools and districts have expressed interest in interim assessment systems to prepare for summative assessments and to improve teaching and learning. However, with so many commercial interim assessments available, schools and districts are struggling to determine which interim assessment is most appropriate to their needs.…

CERN's new safety policy

CERN Multimedia

2014-01-01

The documents below, published on 29 September 2014 on the HSE website, together replace the document SAPOCO 42 as well as Safety Codes A1, A5, A9, A10, which are no longer in force. As from the publication date of these documents any reference made to the document SAPOCO 42 or to Safety Codes A1, A5, A9 and A10 in contractual documents or CERN rules and regulations shall be deemed to constitute a reference to the corresponding provisions of the documents listed below.   "The CERN Safety Policy" "Safety Regulation SR-SO - Responsibilities and organisational structure in matters of Safety at CERN" "General Safety Instruction GSI-SO-1 - Departmental Safety Officer (DSO)" "General Safety Instruction GSI-SO-2 - Territorial Safety Officer (TSO)" "General Safety Instruction GSI-SO-3 - Safety Linkperson (SLP)" "General Safety Instruction GSI-SO-4 - Large Experiment Group Leader In Matters of Safety (LEXGLI...

Disposal facility data for the interim performance

International Nuclear Information System (INIS)

Eiholzer, C.R.

1995-01-01

The purpose of this report is to identify and provide information on the waste package and disposal facility concepts to be used for the low-level waste tank interim performance assessment. Current concepts for the low-level waste form, canister, and the disposal facility will be used for the interim performance assessment. The concept for the waste form consists of vitrified glass cullet in a sulfur polymer cement matrix material. The waste form will be contained in a 2 x 2 x 8 meter carbon steel container. Two disposal facility concepts will be used for the interim performance assessment. These facility concepts are based on a preliminary disposal facility concept developed for estimating costs for a disposal options configuration study. These disposal concepts are based on vault type structures. None of the concepts given in this report have been approved by a Tank Waste Remediation Systems (TWRS) decision board. These concepts will only be used in th interim performance assessment. Future performance assessments will be based on approved designs

School Survey on Crime and Safety (SSOCS) 2000 Public-Use Data Files, User's Manual, and Detailed Data Documentation. [CD-ROM].

Science.gov (United States)

National Center for Education Statistics (ED), Washington, DC.

This CD-ROM contains the raw, public-use data from the 2000 School Survey on Crime and Safety (SSOCS) along with a User's Manual and Detailed Data Documentation. The data are provided in SAS, SPSS, STATA, and ASCII formats. The User's Manual and the Detailed Data Documentation are provided as .pdf files. (Author)

FLAMMABLE GAS TECHNICAL BASIS DOCUMENT

Energy Technology Data Exchange (ETDEWEB)

KRIPPS, L.J.

2005-02-18

This document describes the qualitative evaluation of frequency and consequences for double shell tank (DST) and single shell tank (SST) representative flammable gas accidents and associated hazardous conditions without controls. The evaluation indicated that safety-significant SSCs and/or TSRS were required to prevent or mitigate flammable gas accidents. Discussion on the resulting control decisions is included. This technical basis document was developed to support of the Tank Farms Documented Safety Analysis (DSA) and describes the risk binning process for the flammable gas representative accidents and associated represented hazardous conditions. The purpose of the risk binning process is to determine the need for safety-significant structures, systems, and components (SSC) and technical safety requirement (TSR)-level controls for a given representative accident or represented hazardous condition based on an evaluation of the event frequency and consequence.

Elder Abuse Demonstration Project. Third Interim Report to the Illinois General Assembly on Public Acts 83-1259 and 83-1432.

Science.gov (United States)

Illinois State Dept. on Aging, Springfield.

This document contains the third annual interim report of the Illinois Elder Abuse Demonstration Program. It discusses the overall intent of the demonstration program, trends and changes in the third year of the demonstration program compared with the results from the first two years of the program, and achievements and recommendations for a…

Dewey Decimal Classification Online Project: Interim Reports to the Council on Library Resources, April 1984, September 1984, and February 1985.

Science.gov (United States)

Markey, Karen; Demeyer, Anh N.

This research project focuses on the implementation and testing of the Dewey Decimal Classification (DDC) system as an online searcher's tool for subject access, browsing, and display in an online catalog. The research project comprises 12 activities. The three interim reports in this document cover the first seven of these activities: (1) obtain…

A document-driven method for certifying scientific computing software for use in nuclear safety analysis

International Nuclear Information System (INIS)

Smith, W. Spencer; Koothoor, Mimitha

2016-01-01

This paper presents a documentation and development method to facilitate the certification of scientific computing software used in the safety analysis of nuclear facilities. To study the problems faced during quality assurance and certification activities, a case study was performed on legacy software used for thermal analysis of a fuel pin in a nuclear reactor. Although no errors were uncovered in the code, 27 issues of incompleteness and inconsistency were found with the documentation. This work proposes that software documentation follow a rational process, which includes a software requirements specification following a template that is reusable, maintainable, and understandable. To develop the design and implementation, this paper suggests literate programming as an alternative to traditional structured programming. Literate programming allows for documenting of numerical algorithms and code together in what is termed the literate programmer's manual. This manual is developed with explicit traceability to the software requirements specification. The traceability between the theory, numerical algorithms, and implementation facilitates achieving completeness and consistency, as well as simplifies the process of verification and the associated certification

A document-driven method for certifying scientific computing software for use in nuclear safety analysis

Energy Technology Data Exchange (ETDEWEB)

Smith, W. Spencer; Koothoor, Mimitha [Computing and Software Department, McMaster University, Hamilton (Canada)

2016-04-15

This paper presents a documentation and development method to facilitate the certification of scientific computing software used in the safety analysis of nuclear facilities. To study the problems faced during quality assurance and certification activities, a case study was performed on legacy software used for thermal analysis of a fuel pin in a nuclear reactor. Although no errors were uncovered in the code, 27 issues of incompleteness and inconsistency were found with the documentation. This work proposes that software documentation follow a rational process, which includes a software requirements specification following a template that is reusable, maintainable, and understandable. To develop the design and implementation, this paper suggests literate programming as an alternative to traditional structured programming. Literate programming allows for documenting of numerical algorithms and code together in what is termed the literate programmer's manual. This manual is developed with explicit traceability to the software requirements specification. The traceability between the theory, numerical algorithms, and implementation facilitates achieving completeness and consistency, as well as simplifies the process of verification and the associated certification.

Safety analysis of spent fuel transport and storage casks under extreme impact conditions

International Nuclear Information System (INIS)

Wolff, D.; Wieser, G.; Ballheimer, V.; Voelzke, H.; Droste, B.

2005-01-01

Full text: Worldwide the security of transport and storage of spent fuel with respect to terrorism threats is a matter of concern. In Germany a spent nuclear fuel management program was developed by the government including a new concept of dry on-site interim storage instead of centralized interim storage. In order to minimize transports of spent fuel casks between nuclear power plants, reprocessing plants and central storage facilities, the operators of NPPs have to erect and to use interim storage facilities for spent nuclear fuel on the site or in the vicinity of nuclear power plants. Up to now, 11 on-site interim storage buildings, one storage tunnel and 4 on-site interim storage areas (preliminary cask storage till the on-site interim storage building is completed) have been licensed at 12 nuclear power plant sites. Inside the interim storage buildings the casks are kept in upright position, whereas at the preliminary interim storage areas horizontal storage of the casks on concrete slabs is used and each cask is covered by concrete elements. Storage buildings and concrete elements are designed only for gamma and neutron radiation shielding reasons and as weather protection. Therefore the security of spent fuel inside a dual purpose transport and storage cask depends on the inherent safety of the cask itself. For nearly three decades BAM has been investigating cask safety under severe accident conditions like drop tests from more than 9 m onto different targets and without impact limiters as well as artificially damaged prototype casks. Since the terror attacks of 11 September 2001 the determination of casks' inherent safety also under extreme impact conditions due to terrorist attacks has been of our increasing interest. With respect to spent fuel storage one of the most critical scenarios of a terrorist attack for a cask is the centric impact of a dynamic load onto the lid-seal-system caused e.g. by direct aircraft crash or its engine as well as by a

Interim Basis for PCB Sampling and Analyses

International Nuclear Information System (INIS)

BANNING, D.L.

2001-01-01

This document was developed as an interim basis for sampling and analysis of polychlorinated biphenyls (PCBs) and will be used until a formal data quality objective (DQO) document is prepared and approved. On August 31, 2000, the Framework Agreement for Management of Polychlorinated Biphenyls (PCBs) in Hanford Tank Waste was signed by the US. Department of Energy (DOE), the Environmental Protection Agency (EPA), and the Washington State Department of Ecology (Ecology) (Ecology et al. 2000). This agreement outlines the management of double shell tank (DST) waste as Toxic Substance Control Act (TSCA) PCB remediation waste based on a risk-based disposal approval option per Title 40 of the Code of Federal Regulations 761.61 (c). The agreement calls for ''Quantification of PCBs in DSTs, single shell tanks (SSTs), and incoming waste to ensure that the vitrification plant and other ancillary facilities PCB waste acceptance limits and the requirements of the anticipated risk-based disposal approval are met.'' Waste samples will be analyzed for PCBs to satisfy this requirement. This document describes the DQO process undertaken to assure appropriate data will be collected to support management of PCBs and is presented in a DQO format. The DQO process was implemented in accordance with the U.S. Environmental Protection Agency EPA QAlG4, Guidance for the Data Quality Objectives Process (EPA 1994) and the Data Quality Objectives for Sampling and Analyses, HNF-IP-0842/Rev.1 A, Vol. IV, Section 4.16 (Banning 1999)

Retained gas sampler interim safety assessment

Energy Technology Data Exchange (ETDEWEB)

Pasamehmetoglu, K.O.; Miller, W.O.; Unal, C.; Fujita, R.K.

1995-01-13

This safety assessment addresses the proposed action to install, operate, and remove a Retained Gas Sampler (RGS) in Tank 101-SY at Hanford. Purpose of the RGS is to help characterize the gas species retained in the tank waste; the information will be used to refine models that predict the gas-producing behavior of the waste tank. The RGS will take samples of the tank from top to bottom; these samples will be analyzed for gas constituents. The proposed action is required as part of an evaluation of mitigation concepts for eliminating episodic gas releases that result in high hydrogen concentrations in the tank dome space.

Retained gas sampler interim safety assessment

International Nuclear Information System (INIS)

Pasamehmetoglu, K.O.; Miller, W.O.; Unal, C.; Fujita, R.K.

1995-01-01

This safety assessment addresses the proposed action to install, operate, and remove a Retained Gas Sampler (RGS) in Tank 101-SY at Hanford. Purpose of the RGS is to help characterize the gas species retained in the tank waste; the information will be used to refine models that predict the gas-producing behavior of the waste tank. The RGS will take samples of the tank from top to bottom; these samples will be analyzed for gas constituents. The proposed action is required as part of an evaluation of mitigation concepts for eliminating episodic gas releases that result in high hydrogen concentrations in the tank dome space

The safety and effectiveness profile of eldecalcitol in a prospective, post-marketing observational study in Japanese patients with osteoporosis: interim report.

Science.gov (United States)

Saito, Hitoshi; Kakihata, Hiroyuki; Nishida, Yosuke; Yatomi, Sawako; Nihojima, Shigeru; Kobayashi, Yumiko; Tabata, Hidehiro; Nomura, Makoto

2017-07-01

This large-scale post-marketing surveillance study was conducted to assess the safety and effectiveness of eldecalcitol treatment in patients with osteoporosis in a Japanese clinical setting. A total of 3567 patients with osteoporosis were enrolled and received eldecalcitol 0.75 μg/day for 12 months. For this interim report, 3285 patients were eligible for analysis. Mean age was 74.9 ± 8.7 years; 86.8 % (2854/3285) were women. There were 142 reported adverse drug reactions (ADRs) in 129 patients (3.92 % of the total 3285 patients): the most common were hypercalcemia and increased blood calcium (0.88 %), renal impairment (0.27 %), abdominal discomfort (0.24 %), constipation (0.24 %), and pruritus (0.24 %). The incidence of ADRs was 5.10 % in men and 3.74 % in women. Although 10 serious ADRs were reported in 9 patients (0.27 %), no clinically significant safety issues were identified. Incidence of hypercalcemia or increased blood calcium was 8.47 % in patients with renal impairment and only 0.74 % in patients without renal impairment. At last observation, the incidence of new vertebral and nonvertebral fractures was 2.44 % and 1.70 %, respectively. There was a significant increase in bone mineral density at the lumbar spine and distal radius. The bone turnover markers BAP, serum NTX, urinary NTX, and TRACP-5b were suppressed by eldecalcitol treatment in both sexes. In conclusion, consistent with the findings of the phase III pivotal clinical trial, eldecalcitol was shown to have a favorable safety profile and effectiveness in Japanese patients with osteoporosis. However, periodic measurements of serum calcium were required to prevent occurrence of hypercalcemia during eldecalcitol treatment, especially in patients with renal impairment.

Lifecycle management for nuclear engineering project documents

International Nuclear Information System (INIS)

Zhang Li; Zhang Ming; Zhang Ling

2010-01-01

The nuclear engineering project documents with great quantity and various types of data, in which the relationships of each document are complex, the edition of document update frequently, are managed difficultly. While the safety of project even the nuclear safety is threatened seriously by the false documents and mistakes. In order to ensure the integrality, veracity and validity of project documents, the lifecycle theory of document is applied to build documents center, record center, structure and database of document lifecycle management system. And the lifecycle management is used to the documents of nuclear engineering projects from the production to pigeonhole, to satisfy the quality requirement of nuclear engineering projects. (authors)

76 FR 67361 - Visas: Documentation of Immigrants Under the Immigration and Nationality Act, as Amended

Science.gov (United States)

2011-11-01

... DEPARTMENT OF STATE 22 CFR Part 42 [Public Notice 7391] RIN 1400-AC86 Visas: Documentation of Immigrants Under the Immigration and Nationality Act, as Amended AGENCY: State Department. ACTION: Interim final rule. SUMMARY: This rule amends the Department of State's regulations relating to adoptions in...

The probabilistic risk analysis of external hazards of an interim storage for spent nuclear fuel in Olkiluoto

International Nuclear Information System (INIS)

Puukka, Tiia

2014-01-01

Due to natural disasters occurred in the world and the experiences perceived of the Fukushima nuclear accident, the particular knowledge of the role and influence of external hazards in the safety of interim storage of spent nuclear fuel has been emphasized. For that reason it is substantial that they are included in the probabilistic risk assessment (PRA) of the interim storage facility. This is also required by the Regulatory Guides issued by The Finnish Radiation and Nuclear Safety Authority STUK. To enhance safety culture and nuclear safety in Olkiluoto, The Finnish utility Teollisuuden Voima Oyj has recently completed an analysis of external natural (seismic events are studied as a separate analysis) and unintentional human-induced risks associated with the spent fuel pool cooling and decay heat removal systems as part of the full-scope PRA study for the interim storage of spent fuel (KPA store). The analysis had four goals to achieve: (1) to determine the definition of an initiating event in the context of the KPA store, (2) to identify all potential external hazards and hazard combinations, (3) to perform a qualitative screening analysis based on frequency-strength analysis and detailed plant responses analysis and (4) to model the hazards passed the screening analysis so that model can be used as a risk analysis tool in the risk informed decision making and operating procedures. The assessment carried out included the analysis of operation procedures of decay heat removal, the study of external hazards related initiating events included in the PRA of the OL1 and OL2 nuclear power plants and their dependencies on the initiating events of the KPA store. All external hazards related initiating events were modeled using fault tree linking method. The main result and conclusion of this study was that using the screening analysis, initiating events caused by external hazards that could lead to leakage of the spent fuel pools or that could pose a threat to the

T-TY Tank Farm Interim Surface Barrier Demonstration - Vadose Zone Monitoring Plan

International Nuclear Information System (INIS)

Zhang, Z.F.; Strickland, Christopher E.; Field, Jim G.; Parker, Danny L.

2010-01-01

The Hanford Site has 149 underground single-shell tanks that store hazardous radioactive waste. Many of these tanks and their associated infrastructure (e.g., pipelines, diversion boxes) have leaked. Some of the leaked waste has entered the groundwater. The largest known leak occurred from the T-106 Tank of the 241-T Tank Farm in 1973. Five tanks are assumed to have leaked in the TY Farm. Many of the contaminants from those leaks still reside within the vadose zone within the T and TY Tank Farms. The Department of Energy's Office of River Protection seeks to minimize the movement of these contaminant plumes by placing interim barriers on the ground surface. Such barriers are expected to prevent infiltrating water from reaching the plumes and moving them further. The soil water regime is monitored to determine the effectiveness of the interim surface barriers. Soil-water content and water pressure are monitored using off-the-shelf equipment that can be installed by the hydraulic hammer technique. Four instrument nests were installed in the T Farm in fiscal year (FY) 2006 and FY2007; two nests were installed in the TY Farm in FY2010. Each instrument nest contains a neutron probe access tube, a capacitance probe, and four heat-dissipation units. A meteorological station has been installed at the north side of the fence of the T Farm. This document summarizes the monitoring methods, the instrument calibration and installation, and the vadose zone monitoring plan for interim barriers in T farm and TY Farm.

Final Safety Analysis Document for Building 693 Chemical Waste Storage Building at Lawrence Livermore National Laboratory

International Nuclear Information System (INIS)

Salazar, R.J.; Lane, S.

1992-02-01

This Safety Analysis Document (SAD) for the Lawrence Livermore National Laboratory (LLNL) Building 693, Chemical Waste Storage Building (desipated as Building 693 Container Storage Unit in the Laboratory's RCRA Part B permit application), provides the necessary information and analyses to conclude that Building 693 can be operated at low risk without unduly endangering the safety of the building operating personnel or adversely affecting the public or the environment. This Building 693 SAD consists of eight sections and supporting appendices. Section 1 presents a summary of the facility designs and operations and Section 2 summarizes the safety analysis method and results. Section 3 describes the site, the facility desip, operations and management structure. Sections 4 and 5 present the safety analysis and operational safety requirements (OSRs). Section 6 reviews Hazardous Waste Management's (HWM) Quality Assurance (QA) program. Section 7 lists the references and background material used in the preparation of this report Section 8 lists acronyms, abbreviations and symbols. Appendices contain supporting analyses, definitions, and descriptions that are referenced in the body of this report

Evaluation of Hose in Hose Transfer Line Service Life for Hanford's Interim Stabilization Program

Energy Technology Data Exchange (ETDEWEB)

TORRES, T.D.

2000-08-24

RPP-6153, Engineering Task Plan for Hose-in-Hose Transfer System for the Interim Stabilization Program, defines the programmatic goals, functional requirements, and technical criteria for the development and subsequent installation of transfer line equipment to support Hanford's Interim Stabilization Program. RPP-6028, Specification for Hose in Hose Transfer Lines for Hanford's Interim Stabilization Program, has been issued to define the specific requirements for the design, manufacture, and verification of transfer line assemblies for specific waste transfer applications. Included in RPP-6028 are tables defining the chemical constituents of concern to which transfer lines will be exposed. Current Interim Stabilization Program planning forecasts that the at-grade transfer lines will be required to convey pumpable waste for as much as three years after commissioning. Prudent engineering dictates that the equipment placed in service have a working life in excess of this forecasted time period, with some margin to allow for future adjustments to the planned schedule. This document evaluates the effective service life of the Hose-in-Hose Transfer Lines, based on information submitted by the manufacturer and published literature. The effective service life of transfer line assemblies is a function of several factors. Foremost among these are process fluid characteristics, ambient environmental conditions, and the manufacturer's stated shelf life. This evaluation examines the manufacturer's certification of shelf life, the manufacturer's certifications of chemical compatibility with waste, and published literature on the effects of exposure to ionizing radiation on the mechanical properties of elastomeric materials to evaluate transfer line service life.

CMM Interim Check Design of Experiments (U)

Energy Technology Data Exchange (ETDEWEB)

Montano, Joshua Daniel [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2015-07-29

Coordinate Measuring Machines (CMM) are widely used in industry, throughout the Nuclear Weapons Complex and at Los Alamos National Laboratory (LANL) to verify part conformance to design definition. Calibration cycles for CMMs at LANL are predominantly one year in length and include a weekly interim check to reduce risk. The CMM interim check makes use of Renishaw’s Machine Checking Gauge which is an off-the-shelf product simulates a large sphere within a CMM’s measurement volume and allows for error estimation. As verification on the interim check process a design of experiments investigation was proposed to test a couple of key factors (location and inspector). The results from the two-factor factorial experiment proved that location influenced results more than the inspector or interaction.

Interim Hanford Waste Management Technology Plan

International Nuclear Information System (INIS)

1985-09-01

The Interim Hanford Waste Management Technology Plan (HWMTP) is a companion document to the Interim Hanford Waste Management Plan (HWMP). A reference plan for management and disposal of all existing and certain projected future radioactive Hanford Site Defense Wastes (HSDW) is described and discussed in the HWMP. Implementation of the reference plan requires that various open technical issues be satisfactorily resolved. The principal purpose of the HWMTP is to present detailed descriptions of the technology which must be developed to close each of the technical issues associated with the reference plan identified in the HWMP. If alternative plans are followed, however, technology development efforts including costs and schedules must be changed accordingly. Technical issues addressed in the HWMTP and HWMP are those which relate to disposal of single-shell tank wastes, contaminated soil sites, solid waste burial sites, double-shell tank wastes, encapsulated 137 CsCl and 90 SrF 2 , stored and new solid transuranic (TRU) wastes, and miscellaneous wastes such as contaminated sodium metal. Among the high priority issues to be resolved are characterization of various wastes including early determination of the TRU content of future cladding removal wastes; completion of development of vitrification (Hanford Waste Vitrification Plant) and grout technology; control of subsidence in buried waste sites; and development of criteria and standards including performance assessments of systems proposed for disposal of HSDW. Estimates of the technology costs shown in this report are made on the basis that all identified tasks for all issues associated with the reference disposal plan must be performed. Elimination of, consolidation of, or reduction in the scope of individual tasks will, of course, be reflected in corresponding reduction of overall technology costs

Current status of the first interim spent fuel storage facility in Japan

International Nuclear Information System (INIS)

Shinbo, Hitoshi; Kondo, Mitsuru

2008-01-01

In Japan, storage of spent fuels outside nuclear power plants was enabled as a result of partial amendments to the Nuclear Reactor Regulation Law in June 2000. Five months later, Mutsu City in Aomori Prefecture asked the Tokyo Electric Power Company (TEPCO) to conduct technical surveys on siting of the interim spent fuel storage facility (we call it 'Recyclable-Fuel Storage Center'). In April 2003, TEPCO submitted the report on siting feasibility examination, concluded that no improper engineering data for siting, construction of the facility will be possible from engineering viewpoint. Siting Activities for publicity and public acceptance have been continued since then. After these activities, Aomori Prefecture and Mutsu City approved siting of the Recyclable Fuel Storage Center in October 2005. Aomori Prefecture, Mutsu City, TEPCO and Japan Atomic Power Company (JAPC) signed an agreement on the interim spent fuel storage Facility. A month later, TEPCO and JAPC established Recyclable-Fuel Storage Company (RFS) in Mutsu City through joint capital investment, specialized in the first interim spent fuel storage Facility in Japan. In May 2007, we made an application for establishment permit, following safety review by regulatory authorities. In March 2008, we started the preparatory construction. RFS will safely store of spent fuels of TEPCO and JAPC until they will be reprocessed. Final storage capacity will be 5,000 ton-U. First we will construct the storage building of 3,000 ton-U to be followed by second building. We aim to start operation by 2010. (author)

Applications for electronic documents

International Nuclear Information System (INIS)

Beitel, G.A.

1995-01-01

This paper discusses the application of electronic media to documents, specifically Safety Analysis Reports (SARs), prepared for Environmental Restoration and Waste Management (ER ampersand WM) programs being conducted for the Department of Energy (DOE) at the Idaho National Engineering Laboratory (INEL). Efforts are underway to upgrade our document system using electronic format. To satisfy external requirements (DOE, State, and Federal), ER ampersand WM programs generate a complement of internal requirements documents including a SAR and Technical Safety Requirements along with procedures and training materials. Of interest, is the volume of information and the difficulty in handling it. A recently prepared ER ampersand WM SAR consists of 1,000 pages of text and graphics; supporting references add 10,000 pages. Other programmatic requirements documents consist of an estimated 5,000 pages plus references

Introducing Systematic Aging Management for Interim Storage Facilities in Germany

International Nuclear Information System (INIS)

Spieth-Achtnich, Angelika; Schmidt, Gerhard

2014-01-01

In Germany twelve at-reactor and three central (away from reactor) dry storage facilities are in operation, where the fuel is stored in combined transport-and-storage casks. The safety of the storage casks and facilities has been approved and is licensed for up to 40 years operating time. If the availability of a final disposal facility for the stored wastes (spent fuel and high-level wastes from reprocessing) will be further delayed the renewal of the licenses can become necessary in future. Since 2001 Germany had a regulatory guideline for at-reactor dry interim storage of spent fuel. In this guideline some elements of ageing were implemented, but no systematic approach was made for a state-of-the-art ageing management. Currently the guideline is updated to include all kind of storage facilities (central storages as well) and all kinds of high level waste (also waste from reprocessing). Draft versions of the update are under discussion. In these drafts a systematic ageing management is seen as an instrument to upgrade the available technical knowledge base for possible later regulatory decisions, should it be necessary to prolong storage periods to beyond the currently approved limits. It is further recognized as an instrument to prevent from possible and currently unrecognized ageing mechanisms. The generation of information on ageing can be an important basis for the necessary safety-relevant verifications for long term storage. For the first time, the demands for a systematic monitoring of ageing processes for all safety-related components of the storage system are described. In addition, for inaccessible container components such as the seal system, the neutron shielding, the baskets and the waste inventory, the development of a monitoring program is recommended. The working draft to the revised guideline also contains recommendations on non-technical ageing issues such as the long-term preservation of knowledge, long term personnel planning and long term

Establishment of a rationalized safety assurance logic aiming at FBRs with enhanced social acceptance (1). Interim report of CEA/JNC collaboration NWP-5(a) from 1999 to 2001: common view and JNC's contribution

International Nuclear Information System (INIS)

Niwa, Hajime; Tobita, Yoshiharu; Kurisaka, Kenichi; Kubo, Shigenobu; Kamiyama, Kenji

2001-12-01

This is an interim report describing the progress and the results of the collaborative research works between JNC and CEA on the safety logic in future fast reactors under the title of 'Establishment of a Rationalized Safety Assurance Logic Aiming at FBRs with Enhanced Social Acceptance' from 1999 to 2001. This contains JNC's contribution and common view of both partners. (1) Safety goals are proposed from JNC and CEA. Significant coherency is found such as to keep defense-in depth concept, mitigation measures against core melt are taken into account for containment design, evacuation free' concept is pursued, quantitative safety target is also considered as well as deterministic approach, and improvement of social acceptance is considered from the development stage of the fuel cycle including nuclear power plants. (2) Safety characteristics of each candidate coolant were compared and discussed. Gas-cooled fast reactor is a common interest area. Discussions are focused on: safety design requirements, safety evaluation events list, transient behavior analysis, core catcher designs, and so on. (3) JNC's results include criticality map for predicting CDA behavior and consequences, and CDA analysis results of lead-cooled and gas-cooled fast reactors with SIMMER-III. The collaboration on the action NWP-5a is recognized as being of great importance for the orientation of the innovative design studies. (author)

Evaluation of ERA-Interim precipitation data in complex terrain

Science.gov (United States)

Gao, Lu; Bernhardt, Matthias; Schulz, Karsten

2013-04-01

Precipitation controls a large variety of environmental processes, which is an essential input parameter for land surface models e.g. in hydrology, ecology and climatology. However, rain gauge networks provides the necessary information, are commonly sparse in complex terrains, especially in high mountainous regions. Reanalysis products (e.g. ERA-40 and NCEP-NCAR) as surrogate data are increasing applied in the past years. Although they are improving forward, previous studies showed that these products should be objectively evaluated due to their various uncertainties. In this study, we evaluated the precipitation data from ERA-Interim, which is a latest reanalysis product developed by ECMWF. ERA-Interim daily total precipitation are compared with high resolution gridded observation dataset (E-OBS) at 0.25°×0.25° grids for the period 1979-2010 over central Alps (45.5-48°N, 6.25-11.5°E). Wet or dry day is defined using different threshold values (0.5mm, 1mm, 5mm, 10mm and 20mm). The correspondence ratio (CR) is applied for frequency comparison, which is the ratio of days when precipitation occurs in both ERA-Interim and E-OBS dataset. The result shows that ERA-Interim captures precipitation occurrence very well with a range of CR from 0.80 to 0.97 for 0.5mm to 20mm thresholds. However, the bias of intensity increases with rising thresholds. Mean absolute error (MAE) varies between 4.5 mm day-1 and 9.5 mm day-1 in wet days for whole area. In term of mean annual cycle, ERA-Interim almost has the same standard deviation of the interannual variability of daily precipitation with E-OBS, 1.0 mm day-1. Significant wet biases happened in ERA-Interim throughout warm season (May to August) and dry biases in cold season (November to February). The spatial distribution of mean annual daily precipitation shows that ERA-Interim significant underestimates precipitation intensity in high mountains and northern flank of Alpine chain from November to March while pronounced

Implementation in Russia and the European Union of International Safety Standards of Identity Documents with Biometric Data: Legal Regulation and Perspectives

Directory of Open Access Journals (Sweden)

Alexander Grigoryevich Volevodz

2015-01-01

Full Text Available The article contains the findings of a research into particular aspects of use of identity documents with personal biometric data. It considers the international safety standards of documents with biometric data worked out by the International Civil Aviation Organization (ICAO, pursuant to which those data should be included into machine-readable documents used by their holders for travel to various states. It contains the information on the implementation of these international standards in Russian and European Union law. The author has substantiated a conclusion to the effect that the procedure established in Russia for production and issuance, as well as for use of international, diplomatic and service passports identifying the Russian Federation citizen outside the Russian Federation territory, containing electronic information carriers with personal and biometric personal data, currently conforms to the international safety standards of documents with biometric data. The article surveys the experience of introducing domestic biometric identity documents - electronic passports in various countries of the world, and the problems arising therefrom. It substantiates the advantages and disadvantages of determining a passport of the Russian Federation citizen issued in the form of an identity card with an electronic information carrier, as the main document of the Russian Federation citizen identifying him domestically within the country's territory.

12 CFR 541.18 - Interim Federal savings association.

Science.gov (United States)

2010-01-01

... an existing savings and loan holding company or to facilitate any other transaction the Office may... 12 Banks and Banking 5 2010-01-01 2010-01-01 false Interim Federal savings association. 541.18... REGULATIONS AFFECTING FEDERAL SAVINGS ASSOCIATIONS § 541.18 Interim Federal savings association. The term...

TWRS safety SSCs: Requirements and characteristics

International Nuclear Information System (INIS)

Smith-Fewell, M.A.

1997-01-01

Safety Systems, Structures, and Components (SSCs) have been identified from hazard and accident analyses. These analyses were performed to support the Tank Waste Remediation System (TWRS) Final Safety Analysis Report (FSAR) and Basis for Interim Operation (BID). The text identifies and evaluates the SSCs and their supporting SSCs to show that they either prevent the occurrence of the accident or mitigate the consequences of the accident to below the acceptance guidelines. The requirements for the SSCs to fulfill these tasks are described

Safety and efficacy of ipragliflozin in Japanese patients with type 2 diabetes in real-world clinical practice: interim results of the STELLA-LONG TERM post-marketing surveillance study.

Science.gov (United States)

Nakamura, Ichiro; Maegawa, Hiroshi; Tobe, Kazuyuki; Tabuchi, Hiromi; Uno, Satoshi

2018-02-01

Data regarding the efficacy and safety of sodium-glucose cotransporter 2 inhibitors in the real-world setting in Japan are limited. The STELLA-LONG TERM study is an ongoing 3-year post-marketing surveillance study of ipragliflozin in type 2 diabetes (T2D) patients. Here, we report the interim results (including 3-, 12-, and 24-month data). All Japanese patients with T2D who were first prescribed ipragliflozin between 17 July 2014 and 16 October 2015 at participating centers in Japan were registered in STELLA-LONG TERM. At 3, 12, and 24 months, the safety analysis set comprised 11,053, 5475, and 138 patients, respectively; the efficacy analysis set comprised 8757 patients. Ipragliflozin treatment resulted in statistically significant improvements versus baseline in hemoglobin A1c, fasting plasma glucose concentration, body weight, blood pressure, heart rate, and serum concentrations of low-density lipoprotein cholesterol and triglycerides. The adverse drug reaction incidence rate was 10.71%, the most common reactions being renal and urinary disorders (5.06%), infections and infestations (1.24%), and skin and subcutaneous tissue disorders (1.14%). Ipragliflozin was well tolerated and effective in Japanese patients with T2D; no new safety issues were identified.

Radiation Protection and Safety of Radiation Sources: International Basic Safety Standards. General Safety Requirements. Pt. 3 (Chinese Edition)

International Nuclear Information System (INIS)

2014-01-01

This publication is the new edition of the International Basic Safety Standards. The edition is co-sponsored by seven other international organizations — European Commission (EC/Euratom), FAO, ILO, OECD/NEA, PAHO, UNEP and WHO. It replaces the interim edition that was published in November 2011 and the previous edition of the International Basic Safety Standards which was published in 1996. It has been extensively revised and updated to take account of the latest finding of the United Nations Scientific Committee on the Effects of Atomic Radiation, and the latest recommendations of the International Commission on Radiological Protection. The publication details the requirements for the protection of people and the environment from harmful effects of ionizing radiation and for the safety of radiation sources. All circumstances of radiation exposure are considered

Radiation protection and safety of radiation sources: International basic safety standards. General safety requirements. Pt. 3 (French Edition)

International Nuclear Information System (INIS)

2016-01-01

This publication is the new edition of the International Basic Safety Standards. The edition is co-sponsored by seven other international organizations — European Commission (EC/Euratom), FAO, ILO, OECD/NEA, PAHO, UNEP and WHO. It replaces the interim edition that was published in November 2011 and the previous edition of the International Basic Safety Standards which was published in 1996. It has been extensively revised and updated to take account of the latest finding of the United Nations Scientific Committee on the Effects of Atomic Radiation, and the latest recommendations of the International Commission on Radiological Protection. The publication details the requirements for the protection of people and the environment from harmful effects of ionizing radiation and for the safety of radiation sources. All circumstances of radiation exposure are considered

Radiation Protection and Safety of Radiation Sources: International Basic Safety Standards. General Safety Requirements. Pt. 3 (Arabic Edition)

International Nuclear Information System (INIS)

2015-01-01

This publication is the new edition of the International Basic Safety Standards. The edition is co-sponsored by seven other international organizations — European Commission (EC/Euratom), FAO, ILO, OECD/NEA, PAHO, UNEP and WHO. It replaces the interim edition that was published in November 2011 and the previous edition of the International Basic Safety Standards which was published in 1996. It has been extensively revised and updated to take account of the latest finding of the United Nations Scientific Committee on the Effects of Atomic Radiation, and the latest recommendations of the International Commission on Radiological Protection. The publication details the requirements for the protection of people and the environment from harmful effects of ionizing radiation and for the safety of radiation sources. All circumstances of radiation exposure are considered

Plutonium uranium extraction (PUREX) end state basis for interim operation (BIO) for surveillance and maintenance

Energy Technology Data Exchange (ETDEWEB)

DODD, E.N.

1999-05-12

This Basis for Interim Operation (BIO) was developed for the PUREX end state condition following completion of the deactivation project. The deactivation project has removed or stabilized the hazardous materials within the facility structure and equipment to reduce the hazards posed by the facility during the surveillance and maintenance (S and M) period, and to reduce the costs associated with the S and M. This document serves as the authorization basis for the PUREX facility, excluding the storage tunnels, railroad cut, and associated tracks, for the deactivated end state condition during the S and M period. The storage tunnels, and associated systems and areas, are addressed in WHC-SD-HS-SAR-001, Rev. 1, PUREX Final Safety Analysis Report. During S and M, the mission of the facility is to maintain the conditions and equipment in a manner that ensures the safety of the workers, environment, and the public. The S and M phase will continue until the final decontamination and decommissioning (D and D) project and activities are begun. Based on the methodology of DOE-STD-1027-92, Hazards Categorization and Accident Analysis Techniques for Compliance with DOE Order 5480.23, Nuclear Safety Analysis Reports, the final facility hazards category is identified as hazards category This considers the remaining material inventories, form and distribution of the material, and the energies present to initiate events of concern. Given the current facility configuration, conditions, and authorized S and M activities, there are no operational events identified resulting in significant hazard to any of the target receptor groups (e.g., workers, public, environment). The only accident scenarios identified with consequences to the onsite co-located workers were based on external natural phenomena, specifically an earthquake. The dose consequences of these events are within the current risk evaluation guidelines and are consistent with the expectations for a hazards category 2

Plutonium uranium extraction (PUREX) end state basis for interim operation (BIO) for surveillance and maintenance

International Nuclear Information System (INIS)

DODD, E.N.

1999-01-01

This Basis for Interim Operation (BIO) was developed for the PUREX end state condition following completion of the deactivation project. The deactivation project has removed or stabilized the hazardous materials within the facility structure and equipment to reduce the hazards posed by the facility during the surveillance and maintenance (S and M) period, and to reduce the costs associated with the S and M. This document serves as the authorization basis for the PUREX facility, excluding the storage tunnels, railroad cut, and associated tracks, for the deactivated end state condition during the S and M period. The storage tunnels, and associated systems and areas, are addressed in WHC-SD-HS-SAR-001, Rev. 1, PUREX Final Safety Analysis Report. During S and M, the mission of the facility is to maintain the conditions and equipment in a manner that ensures the safety of the workers, environment, and the public. The S and M phase will continue until the final decontamination and decommissioning (D and D) project and activities are begun. Based on the methodology of DOE-STD-1027-92, Hazards Categorization and Accident Analysis Techniques for Compliance with DOE Order 5480.23, Nuclear Safety Analysis Reports, the final facility hazards category is identified as hazards category This considers the remaining material inventories, form and distribution of the material, and the energies present to initiate events of concern. Given the current facility configuration, conditions, and authorized S and M activities, there are no operational events identified resulting in significant hazard to any of the target receptor groups (e.g., workers, public, environment). The only accident scenarios identified with consequences to the onsite co-located workers were based on external natural phenomena, specifically an earthquake. The dose consequences of these events are within the current risk evaluation guidelines and are consistent with the expectations for a hazards category 2

Single-shell tank interim stabilization project plan

Energy Technology Data Exchange (ETDEWEB)

Ross, W.E.

1998-03-27

Solid and liquid radioactive waste continues to be stored in 149 single-shell tanks at the Hanford Site. To date, 119 tanks have had most of the pumpable liquid removed by interim stabilization. Thirty tanks remain to be stabilized. One of these tanks (C-106) will be stabilized by retrieval of the tank contents. The remaining 29 tanks will be interim stabilized by saltwell pumping. In the summer of 1997, the US Department of Energy (DOE) placed a moratorium on the startup of additional saltwell pumping systems because of funding constraints and proposed modifications to the Hanford Federal Facility Agreement and Consent Order (Tri-Party Agreement) milestones to the Washington State Department of Ecology (Ecology). In a letter dated February 10, 1998, Final Determination Pursuant to Hanford Federal Facility Agreement and Consent Order (Tri-Party Agreement) in the Matter of the Disapproval of the DOE`s Change Control Form M-41-97-01 (Fitzsimmons 1998), Ecology disapproved the DOE Change Control Form M-41-97-01. In response, Fluor Daniel Hanford, Inc. (FDH) directed Lockheed Martin Hanford Corporation (LNMC) to initiate development of a project plan in a letter dated February 25, 1998, Direction for Development of an Aggressive Single-Shell Tank (SST) Interim Stabilization Completion Project Plan in Support of Hanford Federal Facility Agreement and Consent Order (Tri-Party Agreement). In a letter dated March 2, 1998, Request for an Aggressive Single-Shell Tank (SST) Interim Stabilization Completion Project Plan, the DOE reaffirmed the need for an aggressive SST interim stabilization completion project plan to support a finalized Tri-Party Agreement Milestone M-41 recovery plan. This project plan establishes the management framework for conduct of the TWRS Single-Shell Tank Interim Stabilization completion program. Specifically, this plan defines the mission needs and requirements; technical objectives and approach; organizational structure, roles, responsibilities

Single-shell tank interim stabilization project plan

International Nuclear Information System (INIS)

Ross, W.E.

1998-01-01

Solid and liquid radioactive waste continues to be stored in 149 single-shell tanks at the Hanford Site. To date, 119 tanks have had most of the pumpable liquid removed by interim stabilization. Thirty tanks remain to be stabilized. One of these tanks (C-106) will be stabilized by retrieval of the tank contents. The remaining 29 tanks will be interim stabilized by saltwell pumping. In the summer of 1997, the US Department of Energy (DOE) placed a moratorium on the startup of additional saltwell pumping systems because of funding constraints and proposed modifications to the Hanford Federal Facility Agreement and Consent Order (Tri-Party Agreement) milestones to the Washington State Department of Ecology (Ecology). In a letter dated February 10, 1998, Final Determination Pursuant to Hanford Federal Facility Agreement and Consent Order (Tri-Party Agreement) in the Matter of the Disapproval of the DOE's Change Control Form M-41-97-01 (Fitzsimmons 1998), Ecology disapproved the DOE Change Control Form M-41-97-01. In response, Fluor Daniel Hanford, Inc. (FDH) directed Lockheed Martin Hanford Corporation (LNMC) to initiate development of a project plan in a letter dated February 25, 1998, Direction for Development of an Aggressive Single-Shell Tank (SST) Interim Stabilization Completion Project Plan in Support of Hanford Federal Facility Agreement and Consent Order (Tri-Party Agreement). In a letter dated March 2, 1998, Request for an Aggressive Single-Shell Tank (SST) Interim Stabilization Completion Project Plan, the DOE reaffirmed the need for an aggressive SST interim stabilization completion project plan to support a finalized Tri-Party Agreement Milestone M-41 recovery plan. This project plan establishes the management framework for conduct of the TWRS Single-Shell Tank Interim Stabilization completion program. Specifically, this plan defines the mission needs and requirements; technical objectives and approach; organizational structure, roles, responsibilities

Environmental qualification - walkdowns: The documentation of configuration information for safety related components, equipment and systems

International Nuclear Information System (INIS)

Melmer, J.; Waters, M.

1995-01-01

Environmental Qualification walkdowns are conducted to collect field data to verify/validate/document configurations of safety related equipment and systems. This paper describes the process for conducting walkdowns and the justification for using an electronic format. The following are described: a) Background; b) Preparing, executing and processing walkdowns; c) Hardware/software; d) Impact of a paperless system on walkdown execution, maintenance and work planning; e) Other applications for the technology

40 CFR 80.141 - Interim detergent gasoline program.

Science.gov (United States)

2010-07-01

... 40 Protection of Environment 16 2010-07-01 2010-07-01 false Interim detergent gasoline program. 80... (CONTINUED) REGULATION OF FUELS AND FUEL ADDITIVES Detergent Gasoline § 80.141 Interim detergent gasoline... apply to: (i) All gasoline sold or transferred to a party who sells or transfers gasoline to the...

KSC Press Site Transformer Bldg. (K7-1205c) SWMU 074 Interim Measure Work Plan

Science.gov (United States)

Starr, A. Scott; Applegate, Joe

2014-01-01

This document presents and discusses the Interim Measure (IM) Work Plan for the Press Site Transformer Building (K7-1205C). The purpose of the proposed IM activities is to remove soil affected with polychlorinated biphenyls (PCBs) greater than the Florida Department of Environmental Protection (FDEP) residential direct-exposure Soil Cleanup Target Level (R-SCTL) of 0.5 milligrams per kilogram and encapsulate concrete exhibiting PCB concentration greater than the Toxic Substance Control Act (TSCA) threshold of 50 milligrams per kilogram.

Environmental information document defense waste processing facility

International Nuclear Information System (INIS)

1981-07-01

This report documents the impact analysis of a proposed Defense Waste Processing Facility (DWPF) for immobilizing high-level waste currently being stored on an interim basis at the Savannah River Plant (SRP). The DWPF will process the waste into a form suitable for shipment to and disposal in a federal repository. The DWPF will convert the high-level waste into: a leach-resistant form containing above 99.9% of all the radioactivity, and a residue of slightly contaminated salt. The document describes the SRP site and environs, including population, land and water uses; surface and subsurface soils and waters; meteorology; and ecology. A conceptual integrated facility for concurrently producing glass waste and saltcrete is described, and the environmental effects of constructing and operating the facility are presented. Alternative sites and waste disposal options are addressed. Also environmental consultations and permits are discussed

Technical basis document for natural event hazards

International Nuclear Information System (INIS)

CARSON, D.M.

2003-01-01

This technical basis document was developed to support the Tank Farms Documented Safety Analysis (DSA), and describes the risk binning process and the technical basis for assigning risk bins for natural event hazards (NEH)-initiated representative accident and associated represented hazardous conditions. The purpose of the risk binning process is to determine the need for safety-significant structures, systems, and components (SSC) and technical safety requirement (TSR)-level controls for a given representative accident or represented hazardous conditions based on an evaluation of the frequency and consequence. Note that the risk binning process is not applied to facility workers, because all facility worker hazardous conditions are considered for safety-significant SSCs and/or TSR-level controls. Determination of the need for safety-class SSCs was performed in accordance with DOE-STD-3009-94, ''Preparation Guide for US Department of Energy Nonreactor Nuclear Facility Documented Safety Analyses'', as described in this report

Specification ''I'' of the CEFRI concerning the interim job enterprises proposing personnel of A or B category to work in nuclear facilities

CERN Document Server

Int. At. Energy Agency Wien

2002-01-01

This document aims to specify the organization dispositions which have to bee taken by the interim job enterprises proposing personnel of A or B category to work in nuclear facilities. These dispositions should allow to respect the demands of the CEFRI in matter of formation, medical control and personnel dosimetry. (A.L.B.)

Safety significance of ATR passive safety response attributes

International Nuclear Information System (INIS)

Atkinson, S.A.

1990-01-01

The Advanced Test Reactor (ATR) at the Idaho National Engineering Laboratory was designed with some passive safety response attributes which contribute to the safety of the facility. The three passive safety attributes being evaluated in the paper are: 1) In-core and in-vessel natural convection cooling, 2) a passive heat sink capability of the ATR primary coolant system (PCS) for the transfer of decay power from the uninsulated piping to the confinement, and 3) gravity feed of emergency coolant makeup. The safety significance of the ATR passive safety response attributes is that the reactor can passively respond to most transients, given a reactor scram, to provide adequate decay power removal and a significant time for operator action should the normal active heat removal systems and their backup systems both fail. The ATR Interim Level 1 Probabilistic Risk Assessment (PRA) models and results were used to evaluate the significance to ATR fuel damage frequency (or probability) of the above three passive response attributes. The results of the evaluation indicate that the first attribute is a major safety characteristic of the ATR. The second attribute has a noticeable but only minor safety significance. The third attribute has no significant influence on the ATR firewater injection system (emergency coolant system)

Design review report FFTF interim storage cask

International Nuclear Information System (INIS)

Scott, P.L.

1995-01-01

Final Design Review Report for the FFTF Interim Storage Cask. The Interim Storage Cask (ISC) will be used for long term above ground dry storage of FFTF irradiated fuel in Core Component Containers (CCC)s. The CCC has been designed and will house assemblies that have been sodium washed in the IEM Cell. The Solid Waste Cask (SWC) will transfer a full CCC from the IEM Cell to the RSB Cask Loading Station where the ISC will be located to receive it. Once the loaded ISC has been sealed at the RSB Cask Loading Station, it will be transferred by facility crane to the DSWC Transporter. After the ISC has been transferred to the Interim Storage Area (ISA), which is yet to be designed, a mobile crane will be used to place the ISC in its final storage location

Choosing a spent fuel interim storage system

International Nuclear Information System (INIS)

Roland, V.; Hunter, I.

2001-01-01

The Transnucleaire Group has developed different modular solutions to address spent fuel interim storage needs of NPP. These solutions, that are present in Europe, USA and Asia are metal casks (dual purpose or storage only) of the TN 24 family and the NUHOMS canister based system. It is not always simple for an operator to sort out relevant choice criteria. After explaining the basic designs involved on the examples of the TN 120 WWER dual purpose cask and the NUHOMS 56 WWER for WWER 440 spent fuel, we shall discuss the criteria that govern the choice of a given spent fuel interim storage system from the stand point of the operator. In conclusion, choosing and implementing an interim storage system is a complex process, whose implications can be far reaching for the long-term success of a spent fuel management policy. (author)

Hanford surplus facilities hazards identification document

International Nuclear Information System (INIS)

Egge, R.G.

1997-01-01

This document provides general safety information needed by personnel who enter and work in surplus facilities managed by Bechtel Hanford, Inc. The purpose of the document is to enhance access control of surplus facilities, educate personnel on the potential hazards associated with these facilities prior to entry, and ensure that safety precautions are taken while in the facility

78 FR 67442 - Congestion Mitigation and Air Quality Improvement Program Interim Guidance

Science.gov (United States)

2013-11-12

...] Congestion Mitigation and Air Quality Improvement Program Interim Guidance AGENCY: Federal Highway... Comment. SUMMARY: The FHWA is issuing Interim Guidance on the Congestion Mitigation and Air Quality.../environment/air_quality/cmaq/policy_and_guidance/2008_guidance/ guidance/. DATES: This Interim Guidance is...

47 CFR 51.715 - Interim transport and termination pricing.

Science.gov (United States)

2010-10-01

... 47 Telecommunication 3 2010-10-01 2010-10-01 false Interim transport and termination pricing. 51.715 Section 51.715 Telecommunication FEDERAL COMMUNICATIONS COMMISSION (CONTINUED) COMMON CARRIER... Telecommunications Traffic § 51.715 Interim transport and termination pricing. (a) Upon request from a...

ITER Conceptual design: Interim report

International Nuclear Information System (INIS)

1990-01-01

This interim report describes the results of the International Thermonuclear Experimental Reactor (ITER) Conceptual Design Activities after the first year of design following the selection of the ITER concept in the autumn of 1988. Using the concept definition as the basis for conceptual design, the Design Phase has been underway since October 1988, and will be completed at the end of 1990, at which time a final report will be issued. This interim report includes an executive summary of ITER activities, a description of the ITER device and facility, an operation and research program summary, and a description of the physics and engineering design bases. Included are preliminary cost estimates and schedule for completion of the project

Interim radiological safety standards and evaluation procedures for subseabed high-level waste disposal

International Nuclear Information System (INIS)

Klett, R.D.

1997-06-01

The Seabed Disposal Project (SDP) was evaluating the technical feasibility of high-level nuclear waste disposal in deep ocean sediments. Working standards were needed for risk assessments, evaluation of alternative designs, sensitivity studies, and conceptual design guidelines. This report completes a three part program to develop radiological standards for the feasibility phase of the SDP. The characteristics of subseabed disposal and how they affect the selection of standards are discussed. General radiological protection standards are reviewed, along with some new methods, and a systematic approach to developing standards is presented. The selected interim radiological standards for the SDP and the reasons for their selection are given. These standards have no legal or regulatory status and will be replaced or modified by regulatory agencies if subseabed disposal is implemented. 56 refs., 29 figs., 15 tabs

A comparison of the value relevance of interim and annual financial statements

Directory of Open Access Journals (Sweden)

Mbalenhle Zulu

2017-03-01

Aim: It explores whether the value relevance of interim financial statements is higher than the value relevance of annual financial statements. Finally, it investigates whether accounting information published in interim and annual financial statements has incremental value relevance. Setting: Data for the period from 1999 to 2012 were collected from a sample of non-financial companies listed on the Johannesburg Stock Exchange. Method: The Ohlson model to investigate the value relevance of accounting information was used for the study. Results: The results show that interim book value of equity is value relevant while interim earnings are not. Interim financial statements appear to have higher value relevance than annual financial statements. The value relevance of interim and annual accounting information has remained fairly constant over the sample period. Incremental comparisons provide evidence that additional book value of equity and earnings that accrue to a company between interim and annual reporting dates are value relevant. Conclusion: The study was conducted over a long sample period (1999–2012, in an era when a technology-driven economy and more timely reporting media could have had an effect on the value relevance of published accounting information. To the best of our knowledge, this is the first study to evaluate and compare the value relevance of published interim and annual financial statements.

CERCLA document flow: Compressing the schedule, saving costs, and expediting review at the Savannah River Site

International Nuclear Information System (INIS)

Hoffman, W.D.

1991-01-01

The purpose of this paper is to convey the logic of the CERCLA document flow including Work Plans, Characterization Studies, Risk Assessments, Remedial Investigations, Feasibility Studies, proposed plans, and Records of Decision. The intent is to show how schedules at the Savannah River Site are being formulated to accomplish work using an observational approach where carefully planned tasks can be initiated early and carried out in parallel. This paper will share specific proactive experience in working with the EPA to expedite projects, begin removal actions, take interim actions, speed document flow, and eliminate unnecessary documents from the review cycle

Canister storage building design basis accident analysis documentation

International Nuclear Information System (INIS)

KOPELIC, S.D.

1999-01-01

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

Decommissioning costs of WWER-440 nuclear power plants. Interim report: Data collection and preliminary evaluations

International Nuclear Information System (INIS)

2002-11-01

Based on the interest in decommissioning costs within Member States, especially in WWER- 440 operating countries that face the complex decision about continued operation vs. decommissioning in the near future, the IAEA launched the task to prepare a technical document on decommissioning costs of WWER-440 nuclear power plants. The main objectives of this publication were to present the decommissioning costs of WWER-440 NPPs in a uniform manner, i.e. using the cost item and cost group system of the Interim Technical Document on Nuclear Decommissioning 'A Proposed Standardised List of Items for Costing Purposes' developed jointly by the EC, the IAEA and the OECD Nuclear Energy Agency (NEA), and providing, as such, a basis for understanding decommissioning costs differences. Member States operating WWER-440 NPPs or having such units under shutdown or even under decommissioning conditions have been requested to provide cost estimates and other input data in order to facilitate understanding of their cost figures. Both decommissioning options, i.e. immediate decommissioning and safe enclosure, have been considered. In the aforementioned joint Interim Technical Document, cost items related to activities that are carried out with a similar emphasis, whether or not tied to a similar time schedule for decommissioning, or that are based on overall activities that cannot be categorised in a specific time period, are grouped as follows: pre-decommissioning actions; facility shutdown activities; procurement of general equipment and material; dismantling activities; waste processing, storage and disposal; site security, surveillance and maintenance; site restoration, cleanup and landscaping; project management, engineering and site support; research and development; fuel and nuclear material; other costs. Before starting implementation of the study, agreement was obtained on general financial, technical and social boundary conditions that should be used in order to facilitate

FLUOR HANFORD SAFETY MANAGEMENT PROGRAMS

Energy Technology Data Exchange (ETDEWEB)

GARVIN, L. J.; JENSEN, M. A.

2004-04-13

This document summarizes safety management programs used within the scope of the ''Project Hanford Management Contract''. The document has been developed to meet the format and content requirements of DOE-STD-3009-94, ''Preparation Guide for US. Department of Energy Nonreactor Nuclear Facility Documented Safety Analyses''. This document provides summary descriptions of Fluor Hanford safety management programs, which Fluor Hanford nuclear facilities may reference and incorporate into their safety basis when producing facility- or activity-specific documented safety analyses (DSA). Facility- or activity-specific DSAs will identify any variances to the safety management programs described in this document and any specific attributes of these safety management programs that are important for controlling potentially hazardous conditions. In addition, facility- or activity-specific DSAs may identify unique additions to the safety management programs that are needed to control potentially hazardous conditions.

Single-shell tank interim stabilization risk analysis

International Nuclear Information System (INIS)

Basche, A.D.

1998-01-01

The purpose of the Single-Shell Tank (SST) Interim Stabilization Risk Analysis is to provide a cost and schedule risk analysis of HNF-2358, Rev. 1, Single-Shell Tank Interim Stabilization Project Plan (Project Plan) (Ross et al. 1998). The analysis compares the required cost profile by fiscal year (Section 4.2) and revised schedule completion date (Section 4.5) to the Project Plan. The analysis also evaluates the executability of the Project Plan and recommends a path forward for risk mitigation

TWRS HLW interim storage facility search and evaluation

Energy Technology Data Exchange (ETDEWEB)

Calmus, R.B., Westinghouse Hanford

1996-05-16

The purpose of this study was to identify and provide an evaluation of interim storage facilities and potential facility locations for the vitrified high-level waste (HLW) from the Phase I demonstration plant and Phase II production plant. In addition, interim storage facilities for solidified separated radionuclides (Cesium and Technetium) generated during pretreatment of Phase I Low-Level Waste Vitrification Plant feed was evaluated.

Traceability of Software Safety Requirements in Legacy Safety Critical Systems

Science.gov (United States)

Hill, Janice L.

2007-01-01

How can traceability of software safety requirements be created for legacy safety critical systems? Requirements in safety standards are imposed most times during contract negotiations. On the other hand, there are instances where safety standards are levied on legacy safety critical systems, some of which may be considered for reuse for new applications. Safety standards often specify that software development documentation include process-oriented and technical safety requirements, and also require that system and software safety analyses are performed supporting technical safety requirements implementation. So what can be done if the requisite documents for establishing and maintaining safety requirements traceability are not available?

Canister storage building design basis accident analysis documentation

Energy Technology Data Exchange (ETDEWEB)

KOPELIC, S.D.

1999-02-25

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

Recommendations: Procedure to develop a preliminary safety report as part of the radioactive waste repository construction licensing process

International Nuclear Information System (INIS)

2003-01-01

The structure of a preliminary safety report for the title purpose should be as follows: A. Textual part: 1. General (Introduction, Basic information about the construction, Timetable); 2. Site information (Siting, Geography and demography, Meteorology and climatic situation, Hydrology, Geology and hydrogeology); 3. Repository design description (Basic function and performance requirements, Design, Auxiliary systems, Fire prevention/protection, Emergency plans); 4. Operation of the repository (Waste acceptance and inspection, Waste handling and interim storage, Waste disposal, Operating monitoring), 5. Health and environmental impact assessment (Radionuclide inventory, Radionuclide transport paths and mechanisms of release into the environment, Radionuclide release in normal and emergency situations, Radiation protection - health impact assessment and regulatory compliance, Draft operating limits and conditions, Proposed ways of assuring physical protection, Uncertainty assessment), 6. Safe repository shutdown/decommissioning concept, 7 Quality assurance assessment, 8. List of selected equipment. B. Annexes: Maps, Drawings, Diagrams, Miscellaneous; C. Documentation: Previous safety report amendments, Protocols, Miscellaneous. (P.A.)

Gaz de France interim financial report 2007

International Nuclear Information System (INIS)

2007-01-01

This financial report contains the unaudited condensed financial statements of Gaz de France Group for the first half ended June 30, 2007, which were reviewed by the audit committee on August 27, 2007 and by the board of directors at its meeting on August 28, 2007. It includes forward-looking statements concerning the objectives, strategies, financial position, future operating results and the operations of Gaz de France Group. These statements reflect the Group's current perception of its activities and the markets in which it operates, as well as various estimates and assumptions considered to be reasonable. Content: interim management report (highlights of the first half of 2007, revenues and results for the period, financial structure, data on outstanding stock, outlook); interim consolidated financial statements (consolidated statements of income, consolidated balance sheets, consolidated statements of cash flows, recognized income and expenses, statements of changes in shareholders' equity, note to the consolidated financial statements); statement by the person responsible for the interim financial report; statutory auditors' report. (J.S.)

Industrial complementarities between interim storage and reversible geological repository - 59237

International Nuclear Information System (INIS)

Hoorelbeke, Jean-Michel

2012-01-01

The French Act voted in 2006 made the choice of deep geological disposal as the reference option for the long term management of high level (HLW) and intermediate level long-lived waste. The CIGEO repository project aims at avoiding or limiting burden to future generations, which could not be achieved by the extension in time of interim storage. The reversibility as provided by the Act will maintain a liberty of choice for waste management on a duration which is comparable to new storage facility. Interim storage is required to accommodate waste as long as the repository is not available. The commissioning of the repository in 2025 will not suppress needs for interim storage. The paper describes the complementarities between existing and future interim storage facilities and the repository project: repository operational issues and planning, HLW thermal decay, support for the reversibility, etc. It shows opportunities to prepare a global optimization of waste management including the utilization at best of storage capacities and the planning of waste emplacement in the repository in such a way to facilitate operational conditions and to limit cost. Preliminary simulations of storage-disposal scenarios are presented. Thanks to an optimal use of the waste management system, provision can be made for a progressive increase of waste emplacement flow during the first operation phase of the repository. It is then possible to stabilize the industrial activity level of the repository site. An optimal utilization of interim storage can also limit the diversity of waste packages emplaced simultaneously, which facilitates the operation of the repository. 60 years minimum interim storage duration is generally required with respect to HLW thermal output. Extending this interim storage period may reduce the underground footprint of the repository. Regarding reversibility, the capability to manage waste packages potentially retrieved from the repository should be analyzed. The

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

International Nuclear Information System (INIS)

Schlosser, R.L.

1998-01-01

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

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

International Nuclear Information System (INIS)

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

2016-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-10-15

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

Use of alternative dispute resolution--HHS. Notice of interim policy.

Science.gov (United States)

1992-10-27

The Department has developed an interim policy to address the use of alternative dispute resolution (ADR) as required by the Administrative Dispute Resolution Act (ADR Act), Public Law No. 101-552. This interim policy also responds to the Negotiated Rulemaking Act, Public Law No. 101-648, and relevant elements of the Executive Order on Civil Justice Reform (E.O. 12778). The Department is adopting an interim policy because we need a baseline of experience and knowledge from our own pilot activities and those of other agencies before finalizing a policy.

Modernization and refurbishment of the Central Interim Storage

International Nuclear Information System (INIS)

Mele, I.; Zeleznik, N.

2002-01-01

The Central Interim Storage for radioactive waste in Brinje, being put into operation in 1986, needs refurbishment and modernization in order to meet the up-to-date operational and safety requirements and to ensure the normal and undisturbed acceptance of radioactive waste from small producers in the future. Because of the waste, being already stored in the storage, the lack of reprocessing capacities and the lack of auxiliary room, the refurbishment and modernization is a complex problem, which needs to be addressed with care. The plan of refurbishment and modernization requires an integral approach, covering all different aspects of renewal and reconstruction. The implementation plan, however, must be based on the actual state of the storage and real conditions for the implementations: from technical to financial. In this paper the project for refurbishment and modernization of the storage, and some activities that have already been implemented, are presented.(author)

50 CFR 660.720 - Interim protection for sea turtles.

Science.gov (United States)

2010-10-01

... 50 Wildlife and Fisheries 9 2010-10-01 2010-10-01 false Interim protection for sea turtles. 660.720 Section 660.720 Wildlife and Fisheries FISHERY CONSERVATION AND MANAGEMENT, NATIONAL OCEANIC AND... Migratory Fisheries § 660.720 Interim protection for sea turtles. (a) Until the effective date of §§ 660.707...

Explanation of Significant Differences for the Record of Decision for Interim Actions in Zone 1, East Tennessee Technology Park, Oak Ridge, Tennessee

Energy Technology Data Exchange (ETDEWEB)

Bechtel Jacobs

2011-02-01

Zone 1 is a 1400-acre area outside the fence of the main plant at The East Tennessee Technology Park (ETTP) in Oak Ridge, Tennessee. The Record of Decision for Interim Actions in Zone, ETTP (Zone 1 Interim ROD) (DOE 2002) identifies the remedial actions for contaminated soil, buried waste, and subsurface infrastructure necessary to protect human health and to limit further contamination of groundwater. Since the Zone 1 Interim Record of Decision (ROD) was signed, new information has been obtained that requires the remedy to be modified as follows: (1) Change the end use in Contractor's Spoil Area (CSA) from unrestricted industrial to recreational; (2) Remove Exposure Units (EU5) ZI-50, 51, and 52 from the scope of the Zone I Interim ROD; (3) Change the end use of the duct bank corridor from unrestricted industrial to restricted industrial; and (4) Remove restriction for the disturbance of soils below 10 feet in Exposure Unit (EU) Z1-04. In accordance with 40 Code of Federal Regulations (CFR) 300.435, these scope modifications are a 'significant' change to the Zone 1 Interim ROD. In accordance with CERCLA Sect. 117 (c) and 40 CFR 300.435 (c)(2)(i), such a significant change is documented with an Explanation of Significant Differences (ESD). The purpose of this ESD is to make the changes listed above. This ESD is part of the Administrative Record file, and it, and other information supporting the selected remedy, can be found at the DOE Information Center, 475 Oak Ridge Turnpike, Oak Ridge, Tennessee 37830, from 8:00 a.m. to 5:00 p.m., Monday through Friday. The ORR is located in Roane and Anderson counties, within and adjacent to the corporate city limits of Oak Ridge, Tennessee. ETTP is located in Roane County near the northwest corner of the ORR. ETTP began operation during World War II as part of the Manhattan Project. The original mission of ETTP was to produce enriched uranium for use in atomic weapons. The plant produced enriched uranium from

Continuation of the summarizing interim report on previous results of the Gorleben site survey as of May 1983

International Nuclear Information System (INIS)

1990-04-01

In addition to results from the 1983 interim report, this report contains, in order to supplement the surface explorations, seismic reflection measurements, hydrogeologic and seismologic investigations, sorption experiments, and studies of glacial development in the site region and of long-term safety of final waste repositories in salt domes. The site's high grade of suitability for becoming a final radioactive waste repository, the legal basis as well as quality assurance are evaluated. (orig.) [de

Documents and legal texts

International Nuclear Information System (INIS)

2014-01-01

This section of the Bulletin presents the recently published documents and legal texts sorted by country: - Brazil: Resolution No. 169 of 30 April 2014. - Japan: Act Concerning Exceptions to Interruption of Prescription Pertaining to Use of Settlement Mediation Procedures by the Dispute Reconciliation Committee for Nuclear Damage Compensation in relation to Nuclear Damage Compensation Disputes Pertaining to the Great East Japan Earthquake (Act No. 32 of 5 June 2013); Act Concerning Measures to Achieve Prompt and Assured Compensation for Nuclear Damage Arising from the Nuclear Plant Accident following the Great East Japan Earthquake and Exceptions to the Extinctive Prescription, etc. of the Right to Claim Compensation for Nuclear Damage (Act No. 97 of 11 December 2013); Fourth Supplement to Interim Guidelines on Determination of the Scope of Nuclear Damage Resulting from the Accident at the Tokyo Electric Power Company Fukushima Daiichi and Daini Nuclear Power Plants (Concerning Damages Associated with the Prolongation of Evacuation Orders, etc.); Outline of 'Fourth Supplement to Interim Guidelines (Concerning Damages Associated with the Prolongation of Evacuation Orders, etc.)'. - OECD Nuclear Energy Agency: Decision and Recommendation of the Steering Committee Concerning the Application of the Paris Convention to Nuclear Installations in the Process of Being Decommissioned; Joint Declaration on the Security of Supply of Medical Radioisotopes. - United Arab Emirates: Federal Decree No. (51) of 2014 Ratifying the Convention on Supplementary Compensation for Nuclear Damage; Ratification of the Federal Supreme Council of Federal Decree No. (51) of 2014 Ratifying the Convention on Supplementary Compensation for Nuclear Damage

IAEA safety fundamentals: the safety of nuclear installations and the defence in depth concept

International Nuclear Information System (INIS)

Aro, I.

2005-01-01

This presentation is a replica of the similar presentation provided by the IAEA Basic Professional Training Course on Nuclear Safety. The presentation utilizes the IAEA Safety Series document No. 110, Safety Fundamentals: the Safety of Nuclear Installations. The objective of the presentation is to provide the basic rationale for actions in provision of nuclear safety. The presentation also provides basis to understand national nuclear safety requirements. There are three Safety Fundamentals documents in the IAEA Safety Series: one for nuclear safety, one for radiation safety and one for waste safety. The IAEA is currently revising its Safety Fundamentals by combining them into one general Safety Fundamentals document. The IAEA Safety Fundamentals are not binding requirements to the Member States. But, a very similar text has been provided in the Convention on Nuclear Safety which is legally binding for the Member State after ratification by the Parliament. This presentation concentrates on nuclear safety. The Safety Fundamentals documents are the 'policy documents' of the IAEA Safety Standards Series. They state the basic objectives, concepts and principles involved in ensuring protection and safety in the development and application of atomic energy for peaceful purposes. They will state - without providing technical details and without going into the application of principles - the rationale for actions necessary in meeting Safety Requirements. Chapter 7 of this presentation describes the basic features of defence in depth concept which is referred to in the Safety Fundamentals document. The defence in depth concept is a key issue in reaching high level of safety specifically at the design stage but as the reader can see the extended concept also refers to the operational stage. The appendix has been taken directly from the IAEA Basic Professional Training Course on Nuclear Safety and applied to the Finnish conditions. The text originates from the references

Safe interim storage of Hanford tank wastes, draft environmental impact statement, Hanford Site, Richland, Washington

Energy Technology Data Exchange (ETDEWEB)

1994-07-01

This Draft EIS is prepared pursuant to the National Environmental Policy Act (NEPA) and the Washington State Environmental Policy Act (SEPA). DOE and Ecology have identified the need to resolve near-term tank safety issues associated with Watchlist tanks as identified pursuant to Public Law (P.L.) 101-510, Section 3137, ``Safety Measures for Waste Tanks at Hanford Nuclear Reservation,`` of the National Defense Authorization Act for Fiscal Year 1991, while continuing to provide safe storage for other Hanford wastes. This would be an interim action pending other actions that could be taken to convert waste to a more stable form based on decisions resulting from the Tank Waste Remediation System (TWRS) EIS. The purpose for this action is to resolve safety issues concerning the generation of unacceptable levels of hydrogen in two Watchlist tanks, 101-SY and 103-SY. Retrieving waste in dilute form from Tanks 101-SY and 103-SY, hydrogen-generating Watchlist double shell tanks (DSTs) in the 200 West Area, and storage in new tanks is the preferred alternative for resolution of the hydrogen safety issues.

Safe interim storage of Hanford tank wastes, draft environmental impact statement, Hanford Site, Richland, Washington

International Nuclear Information System (INIS)

1994-07-01

This Draft EIS is prepared pursuant to the National Environmental Policy Act (NEPA) and the Washington State Environmental Policy Act (SEPA). DOE and Ecology have identified the need to resolve near-term tank safety issues associated with Watchlist tanks as identified pursuant to Public Law (P.L.) 101-510, Section 3137, ''Safety Measures for Waste Tanks at Hanford Nuclear Reservation,'' of the National Defense Authorization Act for Fiscal Year 1991, while continuing to provide safe storage for other Hanford wastes. This would be an interim action pending other actions that could be taken to convert waste to a more stable form based on decisions resulting from the Tank Waste Remediation System (TWRS) EIS. The purpose for this action is to resolve safety issues concerning the generation of unacceptable levels of hydrogen in two Watchlist tanks, 101-SY and 103-SY. Retrieving waste in dilute form from Tanks 101-SY and 103-SY, hydrogen-generating Watchlist double shell tanks (DSTs) in the 200 West Area, and storage in new tanks is the preferred alternative for resolution of the hydrogen safety issues

SAFETY INSTRUCTION AND SAFETY NOTE

CERN Multimedia

TIS Secretariat

2002-01-01

Please note that the SAFETY INSTRUCTION N0 49 (IS 49) and the SAFETY NOTE N0 28 (NS 28) entitled respectively 'AVOIDING CHEMICAL POLLUTION OF WATER' and 'CERN EXHIBITIONS - FIRE PRECAUTIONS' are available on the web at the following urls: http://edms.cern.ch/document/335814 and http://edms.cern.ch/document/335861 Paper copies can also be obtained from the TIS Divisional Secretariat, email: TIS.Secretariat@cern.ch

Wayne Interim Storage Site annual environmental report for calendar year 1991, Wayne, New Jersey. [Wayne Interim Storage Site

Energy Technology Data Exchange (ETDEWEB)

None

1992-09-01

This document describes the envirormental monitoring program at the Wayne Interim Storage Site (WISS) and surrounding area, implementation of the program, and monitoring results for 1991. Environmental monitoring of WISS and surrounding area began in 1984 when Congress added the site to the US Department of Energy's (DOE) Formerly Utilized Sites Remedial Action Program (FUSRAP). FUSRAP is a DOE program to decontaminate or otherwise control sites where residual radioactive materials remain from the early years of the nation's atomic energy program or from commercial operations causing conditions that Congress has authorized DOE to remedy. WISS is a National Priorities List site. The environmental monitoring program at WISS includes sampling networks for radon and thoron concentrations in air; external gamma radiation exposure; and radium-226, radium-228, thorium-232, and total uranium concentrations in surface water, sediment, and groundwater. Several nonradiological parameters are also measured in groundwater. Monitoring results are compared with applicable Environmental Protection Agency standards, DOE derived concentration guides, dose limits, and other requirements in DOE orders. Environmental standards are established to protect public health and the environment.

Maywood Interim Storage Site annual environmental report for calendar year 1991, Maywood, New Jersey. [Maywood Interim Storage Site

Energy Technology Data Exchange (ETDEWEB)

None

1992-09-01

This document describes the environmental monitoring program at the Maywood Interim Storage Site (MISS) and surrounding area, implementation of the program, and monitoring results for 1991. Environmental monitoring of MISS began in 1984 when congress added the site to the US Department of Energy's (DOE) Formerly Utilized Sites Remedial Action Program (FUSRAP). FUSRAP is a DOE program to identify and decontaminate or otherwise control sites where residual radioactive materials remain from the early years of the nation's atomic energy program or from commercial operations causing conditions that Congress has authorized DOE to remedy. The environmental monitoring program at MISS includes sampling networks for radon and thoron concentrations in air; external gamma radiation-exposure; and total uranium, radium-226, radium-228, thorium-232, and thorium-230 concentrations in surface water, sediment, and groundwater. Additionally, several nonradiological parameters are measured in surface water, sediment, and groundwater. Monitoring results are compared with applicable Environmental Protection Agency standards, DOE derived concentration guides (DCGs), dose limits, and other requirements in DOE orders. Environmental standards are established to protect public health and the environment.

Safety analysis of exothermic reaction hazards associated with the organic liquid layer in tank 241-C-103

International Nuclear Information System (INIS)

Postma, A.K.; Bechtold, D.B.; Borsheim, G.L.; Grisby, J.M.; Guthrie, R.L.; Kummerer, M.; Turner, D.A.; Plys, M.G.

1994-03-01

Safety hazards associated with the interim storage of a potentially flammable organic liquid in waste Tank C-103 are identified and evaluated. The technical basis for closing the unreviewed safety question (USQ) associated with the floating liquid organic layer in this tank is presented

Safety analysis of exothermic reaction hazards associated with the organic liquid layer in tank 241-C-103

Energy Technology Data Exchange (ETDEWEB)

Postma, A.K.; Bechtold, D.B.; Borsheim, G.L.; Grisby, J.M.; Guthrie, R.L.; Kummerer, M.; Turner, D.A. [Westinghouse Hanford Co., Richland, WA (United States); Plys, M.G. [Fauske and Associates, Inc., Burr Ridge, IL (United States)

1994-03-01

Safety hazards associated with the interim storage of a potentially flammable organic liquid in waste Tank C-103 are identified and evaluated. The technical basis for closing the unreviewed safety question (USQ) associated with the floating liquid organic layer in this tank is presented.

General safety considerations

Energy Technology Data Exchange (ETDEWEB)

NONE

1998-09-01

This document presents the full filling of the Brazilian obligations under the Convention on Nuclear Safety. The Chapter 4 of the document contains some details about the priority to safety, financial and human resources, human factors, quality assurance, safety assessment and verification, radiation protection and emergency preparedness.

General safety considerations

International Nuclear Information System (INIS)

2001-01-01

This document presents the full filling of the Brazilian obligations under the Convention on Nuclear Safety. The Chapter 4 of the document contains some details about the priority to safety, financial and human resources, human factors, quality assurance, safety assessment and verification, radiation protection and emergency preparedness

General safety considerations

Energy Technology Data Exchange (ETDEWEB)

NONE

2001-09-01

This document presents the full filling of the Brazilian obligations under the Convention on Nuclear Safety. The Chapter 4 of the document contains some details about the priority to safety, financial and human resources, human factors, quality assurance, safety assessment and verification, radiation protection and emergency preparedness.

General safety considerations

International Nuclear Information System (INIS)

1998-01-01

This document presents the full filling of the Brazilian obligations under the Convention on Nuclear Safety. The Chapter 4 of the document contains some details about the priority to safety, financial and human resources, human factors, quality assurance, safety assessment and verification, radiation protection and emergency preparedness

Classification of Aeronautics System Health and Safety Documents

Data.gov (United States)

National Aeronautics and Space Administration — Most complex aerospace systems have many text reports on safety, maintenance, and associated issues. The Aviation Safety Reporting System (ASRS) spans several...

Materials behavior in interim storage of spent fuel

International Nuclear Information System (INIS)

Johnson, A.B. Jr.; Bailey, W.J.; Gilbert, E.R.; Inman, S.C.

1982-01-01

Interim storage has emerged as the only current spent-fuel management method in the US and is essential in all countries with nuclear reactors. Materials behavior is a key aspect in licensing interim-storage facilities for several decades of spent-fuel storage. This paper reviews materials behavior in wet storage, which is licensed for light-water reactor (LWR) fuel, and dry storage, for which a licensing position for LWR fuel is developing

Criteria for preparation and evaluation of radiological emergency response plans and preparedness in support of nuclear power plants. Interim report

International Nuclear Information System (INIS)

1980-01-01

The purpose of this document is to provide a common reference and interim guidance source for: state and local governments and nuclear facility operators in the development of radiological emergency response plans and preparedness in support of nuclear power plants; and Nuclear Regulatory Commission (NRC), Federal Emergency Management Agency (FEMA) and other Federal agency personnel engaged in the review of state, local government, and licensee plans and preparedness

78 FR 49782 - Interim Staff Guidance on Changes During Construction

Science.gov (United States)

2013-08-15

... Construction AGENCY: Nuclear Regulatory Commission. ACTION: Draft interim staff guidance; request for comment... During Construction.'' This ISG provides guidance to the NRC staff on the Preliminary Amendment Request...-ISG-025 ``Interim Staff Guidance on Changes during Construction under 10 CFR Part 52'' is available...

Canister Storage Building (CSB) Design Basis Accident Analysis Documentation

International Nuclear Information System (INIS)

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

2000-01-01

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

Canister Storage Building (CSB) Design Basis Accident Analysis Documentation

International Nuclear Information System (INIS)

CROWE, R.D.

1999-01-01

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

Interim long-term surveillance plan for the Cheney disposal site near, Grand Junction, Colorado

International Nuclear Information System (INIS)

1997-08-01

This interim long-term surveillance plan (LTSP) describes the U.S. Department of Energy's (DOE) long-term care program for the Uranium Mill Tailings Remedial Action (UMTRA) Project Cheney Disposal Site in Mesa County near Grand Junction, Colorado. This LSTP describes the long-term surveillance program the DOE will implement to ensure the Cheney disposal site performs as designed and is cared for in a manner that protects the public health and safety and the environment. Before each disposal site is licensed for custody and long-term care, the Nuclear Regulatory Commission (NRC) requires the DOE to submit such a site-specific LTSP

Model environmental assessment for a property-cleanup/interim-storage remedial action at a formerly utilized site

International Nuclear Information System (INIS)

Merry-Libby, P.

1982-07-01

This document has been prepared as a model for the preparation of an Environmental Assessment (EA) for a property-cleanup/interim-storage type of remedial action under the Formerly Utilized Sites Remedial Action Program (FUSRAP) of the US Department of Energy (DOE). For major federal actions significantly affecting the quality of the human environment, an Environmental Impact Statement (EIS) must be prepared to aid DOE in making its decision. However, when it is not clear that an action is major and the impacts are significant, an EA may be prepared to determine whether to prepare an EIS or a finding of no significant impact (FONSI). If it is likely that an action may be major and the impacts significant, it is usually more cost-effective and timely to directly prepare an EIS. If it is likely that a FONSI can be reached after some environmental assessment, as DOE believes may be the case for most property-cleanup/interim-storage remedial actions, preparation of site-specific EAs is an effective means of compliance with NEPA

Safety and effectiveness of 24-week treatment with iguratimod, a new oral disease-modifying antirheumatic drug, for patients with rheumatoid arthritis: interim analysis of a post-marketing surveillance study of 2679 patients in Japan.

Science.gov (United States)

Mimori, Tsuneyo; Harigai, Masayoshi; Atsumi, Tatsuya; Fujii, Takao; Kuwana, Masataka; Matsuno, Hiroaki; Momohara, Shigeki; Takei, Syuji; Tamura, Naoto; Takasaki, Yoshinari; Ikeuchi, Satoshi; Kushimoto, Satoru; Koike, Takao

2017-09-01

To determine the real-world safety and effectiveness of iguratimod (IGU) for rheumatoid arthritis (RA), a 52-week, Japanese, post-marketing surveillance study was conducted. An interim analysis at week 24 was performed. This study included all RA patients who received IGU following its introduction to the market. All adverse events (AEs) and adverse drug reactions (ADRs) were collected. Effectiveness was evaluated by the change in Disease Activity Score 28-C-reactive protein (DAS28-CRP) from baseline to week 24. Safety was analyzed in 2679 patients. The overall incidences of AEs, ADRs, and serious ADRs were 38.41, 31.65, and 3.21%, respectively; the most commonly reported serious ADRs were pneumonia/bacterial pneumonia, interstitial lung disease, and Pneumocystis jiroveci pneumonia. Concomitant glucocorticoid use and comorbid conditions associated with respiratory disease were identified as risk factors for serious infections. Pulmonary alveolar hemorrhage and increased international normalized ratio of prothrombin time were observed with concomitant use of IGU and warfarin. The DAS28-CRP decreased from baseline to week 24. Although a safety concern was identified with concomitant use of IGU and warfarin, this real-world study showed no other new safety concerns and similar effectiveness to clinical trials. IGU is a new therapeutic option for RA patients.

Assessment by peer review of the effectiveness of a regulatory programme for radiation safety. Interim report for comment; Evaluacion mediante examen por pares de la efectividad de un programa regulador para la seguridad radiologica. Informe provisional para formular comentarios

Energy Technology Data Exchange (ETDEWEB)

NONE

2002-06-01

This document covers assessment of those aspects of a radiation protection and safety infrastructure that are implemented by the Regulatory Authority for radiation sources and practices using such sources and necessarily includes those ancillary technical services, such as dosimetry services, which directly affect the ability of the Regulatory Authority to discharge its responsibilities. The focus of the guidance in this TECDOC is on assessment of a regulatory programme intended to implement the BSS. The BSS address transportation and waste safety mainly by reference to other IAEA documents. When conducting an assessment, the Review Team members should be aware of the latest IAEA documents (or similar national documents) concerning transportation and waste safety and, if appropriate, nuclear safety, and take them into account to the extent applicable when assessing the effectiveness of the regulatory programme governing radiation protection and safety of radiation source practices in a particular State.

Spent fuel interim management: 1995 update

International Nuclear Information System (INIS)

Anderson, C.K.

1995-01-01

The problems of interim away-from-reactor spent fuel storage and storage in spent fuel pools at the reactor site are discussed. An overview of the state-of-the-art in the USA, Europe, and Japan is presented. The technical facilities for away-from-reactor storage are briefly described, including wet storage pools, interactive concrete systems, metallic containers, and passive concrete systems. Reprocessing technologies are mostly at the design stage only. It is predicted that during the 20 years to come, about 50 000 tonnes of spent fuel will be stored at reactor sites regardless of the advance of spent fuel reprocessing or interim storage projects. (J.B.). 4 tabs., 2 figs

General certification procedure of enterprises and interim job enterprises

CERN Document Server

Int. At. Energy Agency, Wien

2002-01-01

This procedure defines the certification global process of enterprises employing workers of A or B category for nuclear facilities and interim job enterprises proposing workers of A or B category for nuclear facilities. This certification proves the enterprises ability to satisfy the specification ''E'' of the CEFRI and the interim job enterprises to satisfy the specification ''I'' of the CEFRI. (A.L.B.)

Results from the interim salt disposition program macrobatch 10 tank ²¹H qualification samples

Energy Technology Data Exchange (ETDEWEB)

Peters, T. B. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Bannochie, C. J. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2017-02-23

Savannah River National Laboratory (SRNL) analyzed samples from Tank 21H in support of qualification of Macrobatch (Salt Batch) 10 for the Interim Salt Disposition Program (ISDP). This document reports characterization data on the samples of Tank 21H and fulfills the requirements of Deliverable 3 of the Technical Task Request (TTR). Further work will report the results of the Extraction-Scrub-Strip (ESS) testing (Task 5 of the TTR) using the Tank 21H material. Task 4 of the TTR (MST Strike) will not be completed for Salt Batch 10.

TECHNICAL BASIS DOCUMENT FOR NATURAL EVENT HAZARDS

International Nuclear Information System (INIS)

KRIPPS, L.J.

2006-01-01

This technical basis document was developed to support the documented safety analysis (DSA) and describes the risk binning process and the technical basis for assigning risk bins for natural event hazard (NEH)-initiated accidents. The purpose of the risk binning process is to determine the need for safety-significant structures, systems, and components (SSC) and technical safety requirement (TSR)-level controls for a given representative accident or represented hazardous conditions based on an evaluation of the frequency and consequence. Note that the risk binning process is not applied to facility workers, because all facility worker hazardous conditions are considered for safety-significant SSCs and/or TSR-level controls

Tribal child welfare. Interim final rule.

Science.gov (United States)

2012-01-06

The Administration for Children and Families (ACF) is issuing this interim final rule to implement statutory provisions related to the Tribal title IV-E program. Effective October 1, 2009, section 479B(b) of the Social Security Act (the Act) authorizes direct Federal funding of Indian Tribes, Tribal organizations, and Tribal consortia that choose to operate a foster care, adoption assistance and, at Tribal option, a kinship guardianship assistance program under title IV-E of the Act. The Fostering Connections to Success and Increasing Adoptions Act of 2008 requires that ACF issue interim final regulations which address procedures to ensure that a transfer of responsibility for the placement and care of a child under a State title IV-E plan to a Tribal title IV-E plan occurs in a manner that does not affect the child's eligibility for title IV-E benefits or medical assistance under title XIX of the Act (Medicaid) and such services or payments; in-kind expenditures from third-party sources for the Tribal share of administration and training expenditures under title IV-E; and other provisions to carry out the Tribal-related amendments to title IV-E. This interim final rule includes these provisions and technical amendments necessary to implement a Tribal title IV-E program.

Lamar Low-Level Jet Program Interim Report

Energy Technology Data Exchange (ETDEWEB)

Kelley, N.; Shirazi, M.; Jager, D.; Wilde, S.; Adams, J.; Buhl, M.; Sullivan, P.; Patton, E.

2004-01-01

This interim report presents the results to date from the Lamar Low-Level Jet Program (LLLJP) that has been established as joint effort among the U.S. Department of Energy (DOE), the National Wind Technology Center (NWTC) of the National Renewable Energy Laboratory (NREL), and General Electric Wind Energy (GE Wind). The purpose of this project is to develop an understanding of the influence of nocturnal low-level jet streams on the inflow turbulence environment and the documenting of any potential operating impacts on current large wind turbines and the Low Wind Speed Turbine (LWST) designs of the future. A year's record of detailed nocturnal turbulence measurements has been collected from NREL instrumentation installed on the GE Wind 120-m tower in southeastern Colorado and supplemented with mean wind profile data collected using an acoustic wind profiler or SODAR (Sound Detection and Ranging). The analyses of measurements taken as part of a previous program conducted at the NWTC have been used to aid in the interpretation of the results of representative case studies of data collected from the GE Wind tower.

ITER safety

International Nuclear Information System (INIS)

Raeder, J.; Piet, S.; Buende, R.

1991-01-01

As part of the series of publications by the IAEA that summarize the results of the Conceptual Design Activities for the ITER project, this document describes the ITER safety analyses. It contains an assessment of normal operation effluents, accident scenarios, plasma chamber safety, tritium system safety, magnet system safety, external loss of coolant and coolant flow problems, and a waste management assessment, while it describes the implementation of the safety approach for ITER. The document ends with a list of major conclusions, a set of topical remarks on technical safety issues, and recommendations for the Engineering Design Activities, safety considerations for siting ITER, and recommendations with regard to the safety issues for the R and D for ITER. Refs, figs and tabs

Presidential Transition: The Experience of Two Community College Interim Presidents

Science.gov (United States)

Thompson, Matthew D.

2010-01-01

The purpose of this qualitative case study was to understand the experiences of two community college interim presidents, their characteristics, and how they led institutions following an abrupt presidential departure. There were two fundamental questions framing this research study, 1. How do two interim community college presidents lead…

30 CFR 827.13 - Coal preparation plants: Interim performance standards.

Science.gov (United States)

2010-07-01

... 30 Mineral Resources 3 2010-07-01 2010-07-01 false Coal preparation plants: Interim performance...-COAL PREPARATION PLANTS NOT LOCATED WITHIN THE PERMIT AREA OF A MINE § 827.13 Coal preparation plants: Interim performance standards. (a) Persons operating or who have operated coal preparation plants after...

Fast Flux Test Facility interim examination and maintenance cell: Past, present, and future

International Nuclear Information System (INIS)

Vincent, J.R.

1990-09-01

The Fast Flux Test Facility Interim Examination and Maintenance Cell was designed to perform interim examination and/or disassembly of experimental core components for final analysis elsewhere, as well as maintenance of sodium-wetted or neutron-activated internal reactor parts and plant support hardware. The Interim Examination and Maintenance Cell equipment developed and used for the first ten years of operation has been primarily devoted to the disassembly and examination of core component test assemblies. While no major reactor equipment has required remote repair or maintenance, the Interim Examina Examination and Maintenance Cell has served as the remote repair facility for its own in-cell equipment, and several innovative remote repairs have been accomplished. The Interim Examination and Maintenance Cell's demonstrated versatility has shown its capability to support a challenging future. 12 refs., 9 figs

New York State interim waste management cost evaluation

International Nuclear Information System (INIS)

Ma, M.S.; Watts, R.J.; Jorgensen, J.R.; Rochester Gas and Electric Corp., NY)

1985-01-01

The purpose of this study is to investigate and quantify the comparative costs associated with including or excluding Class A utility wastes at a centralized interim waste management facility in New York State. The objective of the study is to assess the unit costs and total statewide costs associated with two distinct scenarios: (1) the case where non-utility Class A LLRW is received, incinerated and stored at the centralized interim facility, and utility Class A wastes are held without incineration at respective nuclear power plant interim onsite facilities without incineration; and (2) the alternative case where both utility and non-utility Class A wastes are accepted, incinerated and stored at the centralized facility. Unit costs to waste generators are estimated for each of the two cases described. This is followed by an estimation of the statewide cost impact to the public. The cost impact represents the cost differential resulting from the exclusion of utility Class A waste from the centralized NYS interim waste management facility. The principal factors comprising the cost differential include (1) higher unit disposal fees charged to non-utility waste generators, which are passed along in the costs of products and services; and (2) costs to utilities due to construction of additional onsite storage capacity, which in turn are charged to electric rate payers

216-T-4 interim stabilization final report

International Nuclear Information System (INIS)

Smith, D.L.

1996-01-01

This report provides a general description of the activities performed for the interim stabilization of the 216-T-4-1 ditch, 216-T-4-2 ditch, and 216-T-4-2 pond. Interim stabilization was required to reduce the amount of surface-contaminated acres and to minimize the migration of radioactive contamination. Work associated with the 216-T4-1 ditch and 216-T-4-2 pond was performed by the Radiation Area Remedial Action (RARA) Project. Work associated with the 216-T-4-2 ditch was done concurrently but was funded by Westinghouse Hanford Company (WHC) Tank Waste Remediation Systems (TWRS)

Lessons for outsourcing and interim management relationships.

Science.gov (United States)

Macko, W; Kostyack, P T

1999-01-01

Few decisions can affect an organization more than the selection of an outsourcing or interim management partner. More and more health care organizations face such decisions in today's competitive market in order to face new business needs. Making these relationships successful can be important for health care organizations seeking competitive advantages or seeking immediately accessible management support. These relationships, however, require careful partner selection and development. Success in outsourcing and interim management relationships is contingent upon a thorough selection process, a strong contract that has clearly and explicitly detailed responsibilities and a culture-sensitive business rapport between the client and selected partner.

PROJECT W-551 INTERIM PRETREATMENT SYSTEM PRECONCEPTUAL CANDIDATE TECHNOLOGY DESCRIPTIONS

Energy Technology Data Exchange (ETDEWEB)

MAY TH

2008-08-12

The Office of River Protection (ORP) has authorized a study to recommend and select options for interim pretreatment of tank waste and support Waste Treatment Plant (WTP) low activity waste (LAW) operations prior to startup of all the WTP facilities. The Interim Pretreatment System (IPS) is to be a moderately sized system which separates entrained solids and 137Cs from tank waste for an interim time period while WTP high level waste vitrification and pretreatment facilities are completed. This study's objective is to prepare pre-conceptual technology descriptions that expand the technical detail for selected solid and cesium separation technologies. This revision includes information on additional feed tanks.

Safety technical considerations on the 2012 periodic safety verification of the Beznau nuclear power plant

International Nuclear Information System (INIS)

2016-12-01

According to nuclear legislation, the owner of an operational license for a nuclear power plant has to provide a periodic safety verification (PSU) every 10 years. The 'North Eastern Power Plants' company (NOK), today AXPO Power AG already performed such a PSU for the Beznau-2 nuclear reactor block (KKB2) in 2002. The Beznau-1 nuclear reactor block (KKB1) received its definitive operational license in October 1970, after test operation during 7 months. After the license for test operation received on July 16 th , 1971, the operational license of KKB2 was renewed several times, each time for a certain period of validity. In 1991, NOK requested a definitive operational license for KKB2, but in 1994 the Swiss Federal Council lengthened the license for only 10 years. Moreover, it laid down that NOK has to periodically report on the safety of the facility. With its letter of August 23 rd , 1998, the Federal Office of Energy defined the documents to be produced for the PSU. The extent of the PSU was defined in such a way that many documents concern the whole power plant, i.e. both nuclear reactor blocks. On December 3 rd , 2004, the Swiss Federal Council granted KKB2 an operational license of limited validity. The present report reviews the 2012 PSU, which covers the time interval from January 1 st , 2002, to December 31 st , 2011, from the point of view of safety. It contains documents for the evaluation of both reactor blocks at KKB. The Beznau interim storage pool was also taken into consideration; it is situated on the KKB site, but, according to a decision of the Swiss Federal Council of May 23 rd , 1991, it has an independent operational license. The evaluation of ageing surveillance takes the whole operational period of the facility into account, i.e. the ageing mechanisms acting as from the beginning of the operation. Moreover, important developments that occurred after the surveillance time interval have been taken into account, especially the status

Do Interim Assessments Reduce the Race and SES Achievement Gaps?

Science.gov (United States)

Konstantopoulos, Spyros; Li, Wei; Miller, Shazia R.; van der Ploeg, Arie

2017-01-01

The authors examined differential effects of interim assessments on minority and low socioeconomic status students' achievement in Grades K-6. They conducted a large-scale cluster randomized experiment in 2009-2010 to evaluate the impact of Indiana's policy initiative introducing interim assessments statewide. The authors used 2-level models to…

Analysis of consequences of postulated solvent fires in Hanford site waste tanks

Energy Technology Data Exchange (ETDEWEB)

Cowley, W.L., Westinghouse Hanford

1996-08-12

This document contains the calculations that support the accident analyses for accidents involving organic solvents. This work was performed to support the Basis for Interim Operation (BIO) and the Final Safety Analysis Report (FSAR) for Tank Waste Remediation Systems (TWRS).

Interim Report by Asia International Grid Connection Study Group

Science.gov (United States)

Omatsu, Ryo

2018-01-01

The Asia International Grid Connection Study Group Interim Report examines the feasibility of developing an international grid connection in Japan. The Group has investigated different cases of grid connections in Europe and conducted research on electricity markets in Northeast Asia, and identifies the barriers and challenges for developing an international grid network including Japan. This presentation introduces basic contents of the interim report by the Study Group.

Preparation for tritiated waste management of fusion facilities: Interim storage WAC

Energy Technology Data Exchange (ETDEWEB)

Decanis, C., E-mail: christelle.decanis@cea.fr [CEA, DEN, Centre de Cadarache, F-13108 Saint-Paul-lez-Durance (France); Canas, D. [CEA, DEN/DADN, Centre de Saclay, F-91191 Gif-sur-Yvette cedex (France); Derasse, F. [CEA, DEN, Centre de Cadarache, F-13108 Saint-Paul-lez-Durance (France); Pamela, J. [CEA, Agence ITER-France, F-13108 Saint-Paul-lez-Durance (France)

2016-11-01

Highlights: • Fusion devices including ITER will generate tritiated waste. • Interim storage is the reference solution offering an answer for all types of tritiated radwaste. • Interim storage is a buffer function in the process management and definition of the waste acceptance criteria (WAC) is a key milestone in the facility development cycle. • Defining WAC is a relevant way to identify ahead of time the studies to be launched and the required actions to converge on a detailed design for example material specific studies, required treatment, interfaces management, modelling and monitoring studies. - Abstract: Considering the high mobility of tritium through the package in which it is contained, the new 50-year storage concepts proposed by the French Alternative Energies and Atomic Energy Commission (CEA) currently provide a solution adapted to the management of waste with tritium concentrations higher than the accepted limits in the disposals. The 50-year intermediate storage corresponds to 4 tritium radioactive periods i.e., a tritium reduction by a factor 16. This paper details the approach implemented to define the waste acceptance criteria (WAC) for an interim storage facility that not only takes into account the specificity of tritium provided by the reference scheme for the management of tritiated waste in France, but also the producers’ needs, the safety analysis of the facility and Andra’s disposal requirements. This will lead to define a set of waste specifications that describe the generic criteria such as acceptable waste forms, general principles and specific issues, e.g. conditioning, radioactive content, tritium content, waste tracking system, and quality control. This approach is also a way to check in advance, during the design phase of the waste treatment chain, how the future waste could be integrated into the overall waste management routes and identify possible key points that need further investigations (design changes, selection

Canister Storage Building (CSB) Design Basis Accident Analysis Documentation

Energy Technology Data Exchange (ETDEWEB)

CROWE, R.D.

1999-09-09

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

Fiscal 1976 Sunshine Project research report. Interim report (hydrogen energy); 1976 nendo chukan hokokushoshu. Suiso energy

Energy Technology Data Exchange (ETDEWEB)

NONE

1976-11-01

This report summarizes the Sunshine Project research interim reports on hydrogen energy of every organizations. The report includes research items, laboratories, institutes and enterprises concerned, research targets, research plans, and progress conditions. The research items are as follows. (1) Hydrogen production technology (electrolysis, high- temperature high-pressure water electrolysis, 4 kinds of thermochemical techniques, direct thermolysis). (2) Hydrogen transport and storage technology (2 kinds of solidification techniques). (3) Hydrogen use technology (combustion technology, fuel cell, solid electrolyte fuel cell, fuel cell power system, hydrogen fuel engine). (4) Hydrogen safety measures technology (disaster preventive technology for gaseous and liquid hydrogen, preventing materials from embrittlement due to hydrogen, hydrogen refining, transport and storage systems, their safety technology). (5) Hydrogen energy system (hydrogen energy system, hydrogen use subsystems, peripheral technologies). (NEDO)

75 FR 36288 - Amended Safety Zone and Regulated Navigation Area, Chicago Sanitary and Ship Canal, Romeoville, IL

Science.gov (United States)

2010-06-25

...The Coast Guard is revising its safety zone and Regulated Navigation Area (RNA) on the Chicago Sanitary and Ship Canal (CSSC) near Romeoville, IL. This revised temporary interim rule reduces the areas covered by the safety zone and RNA, and places additional restrictions on vessels that may transit the RNA.

Preliminary safety evaluation, based on initial site investigation data. Planning document

International Nuclear Information System (INIS)

Hedin, Allan

2002-12-01

This report is a planning document for the preliminary safety evaluations (PSE) to be carried out at the end of the initial stage of SKBs ongoing site investigations for a deep repository for spent nuclear fuel. The main purposes of the evaluations are to determine whether earlier judgements of the suitability of the candidate area for a deep repository with respect to long-term safety holds up in the light of borehole data and to provide feed-back to continued site investigations and site specific repository design. The preliminary safety evaluations will be carried out by a safety assessment group, based on a site model, being part of a site description, provided by a site modelling group and a repository layout within that model suggested by a repository engineering group. The site model contains the geometric features of the site as well as properties of the host rock. Several alternative interpretations of the site data will likely be suggested. Also the biosphere is included in the site model. A first task for the PSE will be to compare the rock properties described in the site model to previously established criteria for a suitable host rock. This report gives an example of such a comparison. In order to provide more detailed feedback, a number of thermal, hydrological, mechanical and chemical analyses of the site will also be included in the evaluation. The selection of analyses is derived from the set of geosphere and biosphere analyses preliminarily planned for the comprehensive safety assessment named SR-SITE, which will be based on a complete site investigation. The selection is dictated primarily by the expected feedback to continued site investigations and by the availability of data after the PSE. The repository engineering group will consider several safety related factors in suggesting a repository layout: Thermal calculations will be made to determine a minimum distance between canisters avoiding canister surface temperatures above 100 deg C

The Interim Financial Reporting in the Spirit of the IAS 34 Norm

OpenAIRE

Ovidia Doinea

2008-01-01

The role of an interim financial reporting is to allow the information users to acknowledge the activity of an entity on period shorter than financial exercise from the perspective of the available profits and cash flows generated as well as from the point of view of its financial position and liquidity. The interim financial reporting includes a complete or condensed set of financial statements which target to update the last financial reporting, usually the annual report. The interim financ...

Safety Basis Report

International Nuclear Information System (INIS)

R.J. Garrett

2002-01-01

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

Safety Basis Report

Energy Technology Data Exchange (ETDEWEB)

R.J. Garrett

2002-01-14

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

Management of Documents and Information in BSC Secretariat

International Nuclear Information System (INIS)

Sumathi, E.; Jayarajan, K.

2017-01-01

The regulatory and safety function of BARC facilities is being carried out by BARC Safety Framework with BARC Safety Council as an apex body. Presently, about one hundred safety committees and task forces are functional in the framework. BSC Secretariat (BSCS) provides technical and administrative support to the BARC Safety Framework for the regulatory activities in BARC. Important documents/records related to committee decisions and facilities are maintained in BSCS, through an established documentation and record keeping system. The compliance of regulatory recommendations is verified during regulatory inspections and subsequent submissions made by the facility authority. This supports the effective regulatory decision making of various committees. This article elaborates the maintenance of records and information at BSC

A Generic Software Safety Document Generator

Science.gov (United States)

Denney, Ewen; Venkatesan, Ram Prasad

2004-01-01

Formal certification is based on the idea that a mathematical proof of some property of a piece of software can be regarded as a certificate of correctness which, in principle, can be subjected to external scrutiny. In practice, however, proofs themselves are unlikely to be of much interest to engineers. Nevertheless, it is possible to use the information obtained from a mathematical analysis of software to produce a detailed textual justification of correctness. In this paper, we describe an approach to generating textual explanations from automatically generated proofs of program safety, where the proofs are of compliance with an explicit safety policy that can be varied. Key to this is tracing proof obligations back to the program, and we describe a tool which implements this to certify code auto-generated by AutoBayes and AutoFilter, program synthesis systems under development at the NASA Ames Research Center. Our approach is a step towards combining formal certification with traditional certification methods.

A web-based endpoint adjudication system for interim analyses in clinical trials.

Science.gov (United States)

Nolen, Tracy L; Dimmick, Bill F; Ostrosky-Zeichner, Luis; Kendrick, Amy S; Sable, Carole; Ngai, Angela; Wallace, Dennis

2009-02-01

A data monitoring committee (DMC) is often employed to assess trial progress and review safety data and efficacy endpoints throughout a trail. Interim analyses performed for the DMC should use data that are as complete and verified as possible. Such analyses are complicated when data verification involves subjective study endpoints or requires clinical expertise to determine each subject's status with respect to the study endpoint. Therefore, procedures are needed to obtain adjudicated data for interim analyses in an efficient manner. In the past, methods for handling such data included using locally reported results as surrogate endpoints, adjusting analysis methods for unadjudicated data, or simply performing the adjudication as rapidly as possible. These methods all have inadequacies that make their sole usage suboptimal. For a study of prophylaxis for invasive candidiasis, adjudication of both study eligibility criteria and clinical endpoints prior to two interim analyses was required. Because the study was expected to enroll at a moderate rate and the sponsor required adjudicated endpoints to be used for interim analyses, an efficient process for adjudication was required. We created a web-based endpoint adjudication system (WebEAS) that allows for expedited review by the endpoint adjudication committee (EAC). This system automatically identifies when a subject's data are complete, creates a subject profile from the study data, and assigns EAC reviewers. The reviewers use the WebEAS to review the subject profile and submit their completed review form. The WebEAS then compares the reviews, assigns an additional review as a tiebreaker if needed, and stores the adjudicated data. The study for which this system was originally built was administratively closed after 10 months with only 38 subjects enrolled. The adjudication process was finalized and the WebEAS system activated prior to study closure. Some website accessibility issues presented initially. However

Nuclear criticality safety guide

International Nuclear Information System (INIS)

Pruvost, N.L.; Paxton, H.C.

1996-09-01

This technical reference document cites information related to nuclear criticality safety principles, experience, and practice. The document also provides general guidance for criticality safety personnel and regulators

Nuclear criticality safety guide

Energy Technology Data Exchange (ETDEWEB)

Pruvost, N.L.; Paxton, H.C. [eds.

1996-09-01

This technical reference document cites information related to nuclear criticality safety principles, experience, and practice. The document also provides general guidance for criticality safety personnel and regulators.

Technical basis document for external events

International Nuclear Information System (INIS)

OBERG, B.D.

2003-01-01

This document supports the Tank Farms Documented Safety Analysis and presents the technical basis for the FR-equencies of externally initiated accidents. The consequences of externally initiated events are discussed in other documents that correspond to the accident that was caused by the external event. The external events include aircraft crash, vehicle accident, range fire, and rail accident

Phase V storage (Project W-112) Central Waste Complex operational readiness review, final report

International Nuclear Information System (INIS)

Wight, R.H.

1997-01-01

This document is the final report for the RFSH conducted, Contractor Operational Readiness Review (ORR) for the Central Waste Complex (CWC) Project W-112 and Interim Safety Basis implementation. As appendices, all findings, observations, lines of inquiry and the implementation plan are included

Phase 5 storage (Project W-112) Central Waste Complex operational readiness review, final report

Energy Technology Data Exchange (ETDEWEB)

Wight, R.H.

1997-05-30

This document is the final report for the RFSH conducted, Contractor Operational Readiness Review (ORR) for the Central Waste Complex (CWC) Project W-112 and Interim Safety Basis implementation. As appendices, all findings, observations, lines of inquiry and the implementation plan are included.

Time-dependent crack growth in Alloy 718: An interim assessment

International Nuclear Information System (INIS)

James, L.A.

1982-08-01

Previous results on the time-dependent nature of fatigue-crack propagation (FCP) in Alloy 718 at elevated temperatures were reviewed. Additional experiments were conducted to further define certain aspects of the time-dependent crack growth behavior. it was found that loading waveform influenced FCP behavior, with tensile hold-times producing higher growth rates than continuous cycling at the same frequency. Crack growth rates under hold-time conditions tended to increase with decreasing grain size. Finally, experiments were conducted which tended to cast some doubt upon the ability of linear-elastic fracture mechanics (LEFM) techniques to characterize cracking behavior in this alloy under hold-time conditions. However, since a superior correlating parameter has not yet been proven, it is suggested that LEFM methods be used in the interim with appropriate safety factors to account for the potential errors. 34 refs., 10 figs., 4 tabs

Commentary: Interim leadership of academic departments at U.S. medical schools.

Science.gov (United States)

Grigsby, R Kevin; Aber, Robert C; Quillen, David A

2009-10-01

Medical schools and teaching hospitals are experiencing more frequent turnover of department chairs. Loss of a department chair creates instability in the department and may have a negative effect on the organization at large. Interim leadership of academic departments is common, and interim chairs are expected to immediately demonstrate skills and leadership abilities. However, little is known about how persons are prepared to assume the interim chair role. Newer competencies for effective leadership include an understanding of the business of medicine, interpersonal and communication skills, the ability to deal with conflict and solve adaptive challenges, and the ability to build and work on teams. Medical schools and teaching hospitals need assistance to meet the unique training and support needs of persons serving as interim leaders. For example, the Association of American Medical Colleges and individual chair societies can develop programs to allow current chairs to reflect on their present positions and plan for the future. Formal leadership training, mentorship opportunities, and conscientious succession planning are good first steps in preparing to meet the needs of academic departments during transitions in leadership. Also, interim leadership experience may be useful as a means for "opening the door" to underrepresented persons, including women, and increasing the diversity of the leadership team.

Annual Report 1998 concerning the nuclear safety and radiological protection in the Swiss nuclear installations

Energy Technology Data Exchange (ETDEWEB)

NONE

1999-05-01

The report presents detailed information about the nuclear safety and radiological protection in the Swiss nuclear power plants, the central interim storage at Wuerenlingen, the Paul Scherrer Institute (PSI) and other nuclear installations in Switzerland.

Annual Report 1998 concerning the nuclear safety and radiological protection in the Swiss nuclear installations

International Nuclear Information System (INIS)

1999-05-01

The report presents detailed information about the nuclear safety and radiological protection in the Swiss nuclear power plants, the central interim storage at Wuerenlingen, the Paul Scherrer Institute (PSI) and other nuclear installations in Switzerland

Annual Report 1999 concerning the nuclear safety and radiological protection in the Swiss nuclear installations

International Nuclear Information System (INIS)

2000-08-01

The report presents detailed information about the nuclear safety and radiological protection in the Swiss nuclear power plants, the central interim storage at Wuerenlingen, the Paul Scherrer Institute (PSI) and other nuclear installations in Switzerland

Efficacy and Safety of Rituximab in Children with Refractory Nephrotic Syndrome; A Multicenter Clinical Trial

Directory of Open Access Journals (Sweden)

Yo Han Ahn

2014-06-01

Conclusions: In this interim analysis of clinical trial to evaluate the efficacy and safety of RTX in children with refractory NS, RTX treatment for refractory NS was safe and effective, especially in patients with DNS.

Annual Report 1999 concerning the nuclear safety and radiological protection in the Swiss nuclear installations

Energy Technology Data Exchange (ETDEWEB)

NONE

2000-08-15

The report presents detailed information about the nuclear safety and radiological protection in the Swiss nuclear power plants, the central interim storage at Wuerenlingen, the Paul Scherrer Institute (PSI) and other nuclear installations in Switzerland.

29 CFR 1905.7 - Form of documents; subscription; copies.

Science.gov (United States)

2010-07-01

... UNDER THE WILLIAMS-STEIGER OCCUPATIONAL SAFETY AND HEALTH ACT OF 1970 General § 1905.7 Form of documents... 29 Labor 5 2010-07-01 2010-07-01 false Form of documents; subscription; copies. 1905.7 Section 1905.7 Labor Regulations Relating to Labor (Continued) OCCUPATIONAL SAFETY AND HEALTH ADMINISTRATION...

Radiation analysis for a generic centralized interim storage facility

International Nuclear Information System (INIS)

Gillespie, S.G.; Lopez, P.; Eble, R.G.

1997-01-01

This paper documents the radiation analysis performed for the storage area of a generic Centralized Interim Storage Facility (CISF) for commercial spent nuclear fuel (SNF). The purpose of the analysis is to establish the CISF Protected Area and Restricted Area boundaries by modeling a representative SNF storage array, calculating the radiation dose at selected locations outside the storage area, and comparing the results with regulatory radiation dose limits. The particular challenge for this analysis is to adequately model a large (6000 cask) storage array with a reasonable amount of analysis time and effort. Previous analyses of SNF storage systems for Independent Spent Fuel Storage Installations at nuclear plant sites (for example in References 5.1 and 5.2) had only considered small arrays of storage casks. For such analyses, the dose contribution from each storage cask can be modeled individually. Since the large number of casks in the CISF storage array make such an approach unrealistic, a simplified model is required

Documentation: Records and Reports.

Science.gov (United States)

Akers, Michael J

2017-01-01

This article deals with documentation to include the beginning of documentation, the requirements of Good Manufacturing Practice reports and records, and the steps that can be taken to minimize Good Manufacturing Practice documentation problems. It is important to remember that documentation for 503a compounding involves the Formulation Record, Compounding Record, Standard Operating Procedures, Safety Data Sheets, etc. For 503b outsourcing facilities, compliance with Current Good Manufacturing Practices is required, so this article is applicable to them. For 503a pharmacies, one can see the development and modification of Good Manufacturing Practice and even observe changes as they are occurring in 503a documentation requirements and anticipate that changes will probably continue to occur. Copyright© by International Journal of Pharmaceutical Compounding, Inc.

Nuclear power plants documentation system

International Nuclear Information System (INIS)

Schwartz, E.L.

1991-01-01

Since the amount of documents (type and quantity) necessary for the entire design of a NPP is very large, this implies that an overall and detailed identification, filling and retrieval system shall be implemented. This is even more applicable to the FINAL QUALITY DOCUMENTATION of the plant, as stipulated by IAEA Safety Codes and related guides. For such a purpose it was developed a DOCUMENTATION MANUAL, which describes in detail the before mentioned documentation system. Here we present the expected goals and results which we have to reach for Angra 2 and 3 Project. (author)

Preliminary Safety Information Document for the Standard MHTGR. Volume 1, (includes latest Amendments)

Energy Technology Data Exchange (ETDEWEB)

None

1986-01-01

With NRC concurrence, the Licensing Plan for the Standard HTGR describes an application program consistent with 10CFR50, Appendix O to support a US Nuclear Regulatory Commission (NRC) review and design certification of an advanced Standard modular High Temperature Gas-Cooled Reactor (MHTGR) design. Consistent with the NRC's Advanced Reactor Policy, the Plan also outlines a series of preapplication activities which have as an objective the early issuance of an NRC Licensability Statement on the Standard MHTGR conceptual design. This Preliminary Safety Information Document (PSID) has been prepared as one of the submittals to the NRC by the US Department of Energy in support of preapplication activities on the Standard MHTGR. Other submittals to be provided include a Probabilistic Risk Assessment, a Regulatory Technology Development Plan, and an Emergency Planning Bases Report.

Safety of operations in the manufacture of driver fuel for the first charge of the Dragon Reactor and modifications to the safety document for the Dragon Fuel Element Production Building

International Nuclear Information System (INIS)

Beutler, H.; Cross, J.; Flamm, J.

1965-01-01

The manufacture of the zirconium containing 'driver' fuel and fuel elements for the First Charge of the Dragon Reactor Experiment has been completed without incident. This is a report on the safety of operations in the Dragon Fuel Element Production Building during an approximately six month period when the 'driver' fuel was manufactured and 25 elements containing this fuel were assembled and exported to the Reactor Building. The opportunity is taken to bring the Safety Document up-to-date and to report on any significant operational failures of equipment. (author)

RELEASE OF DRIED RADIOACTIVE WASTE MATERIALS TECHNICAL BASIS DOCUMENT

International Nuclear Information System (INIS)

KOZLOWSKI, S.D.

2007-01-01

This technical basis document was developed to support RPP-23429, Preliminary Documented Safety Analysis for the Demonstration Bulk Vitrification System (PDSA) and RPP-23479, Preliminary Documented Safety Analysis for the Contact-Handled Transuranic Mixed (CH-TRUM) Waste Facility. The main document describes the risk binning process and the technical basis for assigning risk bins to the representative accidents involving the release of dried radioactive waste materials from the Demonstration Bulk Vitrification System (DBVS) and to the associated represented hazardous conditions. Appendices D through F provide the technical basis for assigning risk bins to the representative dried waste release accident and associated represented hazardous conditions for the Contact-Handled Transuranic Mixed (CH-TRUM) Waste Packaging Unit (WPU). The risk binning process uses an evaluation of the frequency and consequence of a given representative accident or represented hazardous condition to determine the need for safety structures, systems, and components (SSC) and technical safety requirement (TSR)-level controls. A representative accident or a represented hazardous condition is assigned to a risk bin based on the potential radiological and toxicological consequences to the public and the collocated worker. Note that the risk binning process is not applied to facility workers because credible hazardous conditions with the potential for significant facility worker consequences are considered for safety-significant SSCs and/or TSR-level controls regardless of their estimated frequency. The controls for protection of the facility workers are described in RPP-23429 and RPP-23479. Determination of the need for safety-class SSCs was performed in accordance with DOE-STD-3009-94, Preparation Guide for US. Department of Energy Nonreactor Nuclear Facility Documented Safety Analyses, as described below

Safety significance of ATR [Advanced Test Reactor] passive safety response attributes

International Nuclear Information System (INIS)

Atkinson, S.A.

1989-01-01

The Advanced Test Reactor (ATR) at the Idaho National Engineering Laboratory was designed with some passive safety response attributes which contribute to the safety posture of the facility. The three passive safety attributes being evaluated in the paper are: (1) In-core and in-vessel natural convection cooling, (2) a passive heat sink capability of the ATR primary coolant system (PCS) for the transfer of decay power from the uninsulated piping to the confinement, and (3) gravity feed of emergency coolant makeup. The safety significance of the ATR passive safety response attributes is that the reactor can passively respond for most transients, given a reactor scram, to provide adequate decay power removal and a significant time for operator action should the normal active heat removal systems and their backup systems both fail. The ATR Interim Level 1 Probabilistic Risk Assessment (PRA) model ands results were used to evaluate the significance to ATR fuel damage frequency (or probability) of the above three passive response attributes. The results of the evaluation indicate that the first attribute is a major safety characteristic of the ATR. The second attribute has a noticeable but only minor safety significance. The third attribute has no significant influence on the ATR Level 1 PRA because of the diversity and redundancy of the ATR firewater injection system (emergency coolant system). 8 refs., 4 figs., 1 tab

New set of Chemical Safety rules

CERN Multimedia

HSE Unit

2011-01-01

A new set of four Safety Rules was issued on 28 March 2011: Safety Regulation SR-C ver. 2, Chemical Agents (en); General Safety Instruction GSI-C1, Prevention and Protection Measures (en); General Safety Instruction GSI-C2, Explosive Atmospheres (en); General Safety Instruction GSI-C3, Monitoring of Exposure to Hazardous Chemical Agents in Workplace Atmospheres (en). These documents form part of the CERN Safety Rules and are issued in application of the “Staff Rules and Regulations” and of document SAPOCO 42. These documents set out the minimum requirements for the protection of persons from risks to their occupational safety and health arising, or likely to arise, from the effects of hazardous chemical agents that are present in the workplace or used in any CERN activity. Simultaneously, the HSE Unit has published seven Safety Guidelines and six Safety Forms. These documents are available from the dedicated Web page “Chemical, Cryogenic and Biological Safety&...

In vitro evaluation of the marginal integrity of CAD/CAM interim crowns.

Science.gov (United States)

Kelvin Khng, Kwang Yong; Ettinger, Ronald L; Armstrong, Steven R; Lindquist, Terry; Gratton, David G; Qian, Fang

2016-05-01

The accuracy of interim crowns made with computer-aided design and computer-aided manufacturing (CAD/CAM) systems has not been well investigated. The purpose of this in vitro study was to evaluate the marginal integrity of interim crowns made by CAD/CAM compared with that of conventional polymethylmethacrylate (PMMA) crowns. A dentoform mandibular left second premolar was prepared for a ceramic crown and scanned for the fabrication of 60 stereolithical resin dies, half of which were scanned to fabricate 15 Telio CAD-CEREC and 15 Paradigm MZ100-E4D-E4D crowns. Fifteen Caulk and 15 Jet interim crowns were made on the remaining resin dies. All crowns were cemented with Tempgrip under a 17.8-N load, thermocycled for 1000 cycles, placed in 0.5% acid fuschin for 24 hours, and embedded in epoxy resin before sectioning from the mid-buccal to mid-lingual surface. The marginal discrepancy was measured using a traveling microscope, and dye penetration was measured as a percentage of the overall length under the crown. The mean vertical marginal discrepancy of the conventionally made interim crowns was greater than for the CAD/CAM crowns (P=.006), while no difference was found for the horizontal component (P=.276). The mean vertical marginal discrepancy at the facial surface of the Caulk crowns was significantly greater than that of the other 3 types of interim crowns (Pmargin, the mean horizontal component of the Telio crowns was significantly larger than that of the other 3 types, with no difference at the lingual margins (P=.150). The mean percentage dye penetration for the Paradigm MZ100-E4D crowns was significantly greater and for Jet crowns significantly smaller than for the other 3 crowns (Pmarginal discrepancies of the Jet interim crowns at the facial surface and with the horizontal marginal discrepancies of the Caulk interim crowns at the lingual surface (Pmarginal discrepancy was found with the interim crowns fabricated by CAD/CAM as compared with PMMA crowns

Interim Action Proposed Plan for the old radioactive waste burial ground (643-E)

International Nuclear Information System (INIS)

McFalls, S.

1995-12-01

This Interim Action Proposed (IAPP) is issued by the U.S. Department of Energy (DOE), which functions as the lead agency for SRS remedial activities, and with concurrence by the U.S. Environmental Protection Agency (EPA) and the South Carolina Department of Health and Environmental Control (SCDHEC). The purpose of this IAPP is to describe the preferred interim remedial action for addressing the Old Radioactive Waste Burial Ground (ORWBG) unit located in the Burial Ground Complex (BGC) at the Savannah River Site (SRS) in Aiken, South Carolina. On December 21, 1989, SRS was included on the National Priorities List (NPL). In accordance with Section 120 of the Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA), DOE has negotiated a Federal Facility Agreement (FFA, 1993) with EPA and SCDHEC to coordinate remedial activities at SRS. Public participation requirements are listed in Sections 113 and 117 of CERCLA. These requirements include establishment of an Administrative Record File that documents the selection of remedial alternatives and allows for review and comment by the public regarding those alternatives. The SRS Public Involvement Plan (PIP) (DOE, 1994) is designed to facilitate public involvement in the decision-making process for permitting closure, and the selection of remedial alternatives. Section 117(a) of CERCLA, 1980, as amended, requires publication of a notice of any proposed remedial action

Project management plan for Reactor 105-C Interim Safe Storage project

International Nuclear Information System (INIS)

Plagge, H.A.

1996-09-01

Reactor 105-C (located on the Hanford Site in Richland, Washington) will be placed into an interim safe storage condition such that (1) interim inspection can be limited to a 5-year frequency; (2) containment ensures that releases to the environmental are not credible under design basis conditions; and (3) final safe storage configuration shall not preclude or significantly increase the cost for any decommissioning alternatives for the reactor assembly.This project management plan establishes plans, organizational responsibilities, control systems, and procedures for managing the execution of Reactor 105-C interim safe storage activities to meet programmatic requirements within authorized funding and approved schedules

Urgent recommendation. Interim report

Energy Technology Data Exchange (ETDEWEB)

Nakano, Masayuki [International Affairs and Safeguards Division, Atomic Energy Bureau, Science and Technology Agency, Tokyo (Japan)

2000-12-01

The Investigation Committee for Critical Accident at Uranium Processing Plant was founded immediately after the accident to investigate the cause of the accident and to establish measures to prevent the similar accident. On September 30, 1999 around 10:35, the Japan's first criticality accident occurred at JCO Co. Ltd. Uranium processing plant (auxiliary conversion plant) located at Tokai-mura Ibaraki-ken. The criticality continued on and off for approximately 20 hours after the first instantaneous criticality. The accident led the recommendation of tentative evacuation and sheltering indoors for residents living in the neighborhood. The serious exposure to neutrons happened to three workers. The dominant effect is dose due to neutrons and gamma rays from the precipitation tank. When the accident took place, three workers dissolved sequentially about 2.4 kg uranium powder with 18.8 % enrichment in the 10-litter bucket with nitric acid. The procedure of homogenization of uranium nitrate was supposed to be controlled using the shape-limited narrow storage column. Actually, however, the thick and large precipitation tank was used. As a result, about 16.6 kg of uranium was fed into the tank, which presumably caused criticality. The first notification by JCO was delayed and the following communication was not smooth. This led to the delay of correct understanding of the situation and made the initial proper response difficult, then followed by insufficient communication between the nation, prefecture, and local authority. Urgent recommendations were made on the following items; (1) Safety measures to be taken at the accident site, (2) health cares for residents and others, (3) Comprehensive safety securing at nuclear operators such as Establishment of the effective audit system, Safety education for employees and Qualification and licensing system, Safety related documents, etc. (4) Reconstruction of the government's safety regulations such as How safety

Urgent recommendation. Interim report

International Nuclear Information System (INIS)

Nakano, Masayuki

2000-01-01

The Investigation Committee for Critical Accident at Uranium Processing Plant was founded immediately after the accident to investigate the cause of the accident and to establish measures to prevent the similar accident. On September 30, 1999 around 10:35, the Japan's first criticality accident occurred at JCO Co. Ltd. Uranium processing plant (auxiliary conversion plant) located at Tokai-mura Ibaraki-ken. The criticality continued on and off for approximately 20 hours after the first instantaneous criticality. The accident led the recommendation of tentative evacuation and sheltering indoors for residents living in the neighborhood. The serious exposure to neutrons happened to three workers. The dominant effect is dose due to neutrons and gamma rays from the precipitation tank. When the accident took place, three workers dissolved sequentially about 2.4 kg uranium powder with 18.8 % enrichment in the 10-litter bucket with nitric acid. The procedure of homogenization of uranium nitrate was supposed to be controlled using the shape-limited narrow storage column. Actually, however, the thick and large precipitation tank was used. As a result, about 16.6 kg of uranium was fed into the tank, which presumably caused criticality. The first notification by JCO was delayed and the following communication was not smooth. This led to the delay of correct understanding of the situation and made the initial proper response difficult, then followed by insufficient communication between the nation, prefecture, and local authority. Urgent recommendations were made on the following items; (1) Safety measures to be taken at the accident site, (2) health cares for residents and others, (3) Comprehensive safety securing at nuclear operators such as Establishment of the effective audit system, Safety education for employees and Qualification and licensing system, Safety related documents, etc. (4) Reconstruction of the government's safety regulations such as How safety regulation

Interim FDG-PET Scan in Hodgkin's Lymphoma: Hopes and Caveats

Directory of Open Access Journals (Sweden)

M. André

2011-01-01

Full Text Available FDG-PET has recently emerged as an important tool for the management of Hodgkins lymphoma. Although its use for initial staging and response evaluation at the end of treatment is well established, the place of interim PET for response assessment and subsequent treatment tailoring is still quite controversial. The use of interim PET after a few cycles of chemotherapy may allow treatment reduction for good responders, leading to lesser treatment toxicities as well as early treatment adaptation for bad responders with a potential higher chance for cure. Interpretation of interim PET is a rapidly moving field. Actually, visual interpretation is preferred over quantitative interpretation in this situation. The notion of minimal residual uptake emerged for faint persisting FDG uptake, but has evolved during the recent years. Guidelines using mediastinum and liver as references have been proposed at the expert meeting in Deauville 2009. Actually, several trials are ongoing both for localised and advanced disease to evaluate the FDG-PET potential for early treatment monitoring and tailoring. Until the results of these prospective randomized trials become available, treatment changes according to the interim PET results should remain inappropriate and limited to well-conducted clinical trials.

42 CFR 93.401 - Interaction with other offices and interim actions.

Science.gov (United States)

2010-10-01

... 42 Public Health 1 2010-10-01 2010-10-01 false Interaction with other offices and interim actions. 93.401 Section 93.401 Public Health PUBLIC HEALTH SERVICE, DEPARTMENT OF HEALTH AND HUMAN SERVICES... Human Services General Information § 93.401 Interaction with other offices and interim actions. (a) ORI...

42 CFR 417.572 - Budget and enrollment forecast and interim reports.

Science.gov (United States)

2010-10-01

... 42 Public Health 3 2010-10-01 2010-10-01 false Budget and enrollment forecast and interim reports... PLANS, AND HEALTH CARE PREPAYMENT PLANS Medicare Payment: Cost Basis § 417.572 Budget and enrollment forecast and interim reports. (a) Annual submittal. The HMO or CMP must submit an annual operating budget...

Developing new transportable storage casks for interim dry storage

International Nuclear Information System (INIS)

Hayashi, K.; Iwasa, K.; Araki, K.; Asano, R.

2004-01-01

Transportable storage metal casks are to be consistently used during transport and storage for AFR interim dry storage facilities planning in Japan. The casks are required to comply with the technical standards of regulations for both transport (hereinafter called ''transport regulation'') and storage (hereafter called ''storage regulation'') to maintain safety functions (heat transfer, containment, shielding and sub-critical control). In addition to these requirements, it is not planned in normal state to change the seal materials during storage at the storage facility, therefore it is requested to use same seal materials when the casks are transported after storage period. The dry transportable storage metal casks that satisfy the requirements have been developed to meet the needs of the dry storage facilities. The basic policy of this development is to utilize proven technology achieved from our design and fabrication experience, to carry out necessary verification for new designs and to realize a safe and rational design with higher capacity and efficient fabrication

Early interim 18F-FDG PET in Hodgkin's lymphoma: evaluation on 304 patients

International Nuclear Information System (INIS)

Zinzani, Pier Luigi; Stefoni, Vittorio; Broccoli, Alessandro; Argnani, Lisa; Baccarani, Michele; Rigacci, Luigi; Puccini, Benedetta; Castagnoli, Antonio; Vaggelli, Luca; Zanoni, Lucia; Fanti, Stefano

2012-01-01

The use of early (interim) PET restaging during first-line therapy of Hodgkin's lymphoma (HL) in clinical practice has considerably increased because of its ability to provide early recognition of treatment failure allowing patients to be transferred to more intensive treatment regimens. Between June 1997 and June 2009, 304 patients with newly diagnosed HL (147 early stage and 157 advanced stage) were treated with the ABVD regimen at two Italian institutions. Patients underwent PET staging and restaging at baseline, after two cycles of therapy and at the end of the treatment. Of the 304 patients, 53 showed a positive interim PET scan and of these only 13 (24.5%) achieved continuous complete remission (CCR), whereas 251 patients showed a negative PET scan and of these 231 (92%) achieved CCR. Comparison between interim PET-positive and interim PET-negative patients indicated a significant association between PET findings and 9-year progression-free survival and 9-year overall survival, with a median follow-up of 31 months. Among the early-stage patients, 19 had a positive interim PET scan and only 4 (21%) achieved CCR; among the 128 patients with a negative interim PET scan, 122 (97.6%) achieved CCR. Among the advanced-stage patients, 34 showed a persistently positive PET scan with only 9 (26.4%) achieving CCR, whereas 123 showed a negative interim PET scan with 109 (88.6%) achieving CCR. Our results demonstrate the role of an early PET scan as a significant step forward in the management of patients with early-stage or advanced-stage HL. (orig.)

10 CFR 431.401 - Petitions for waiver, and applications for interim waiver, of test procedure.

Science.gov (United States)

2010-01-01

... Renewable Energy, U.S. Department of Energy. Each Application for Interim Waiver must reference the Petition... Renewable Energy. (e) Provisions specific to interim waivers—(1) Disposition of application. If... 10 Energy 3 2010-01-01 2010-01-01 false Petitions for waiver, and applications for interim waiver...

Fit of interim crowns fabricated using photopolymer-jetting 3D printing.

Science.gov (United States)

Mai, Hang-Nga; Lee, Kyu-Bok; Lee, Du-Hyeong

2017-08-01

The fit of interim crowns fabricated using 3-dimensional (3D) printing is unknown. The purpose of this in vitro study was to evaluate the fit of interim crowns fabricated using photopolymer-jetting 3D printing and to compare it with that of milling and compression molding methods. Twelve study models were fabricated by making an impression of a metal master model of the mandibular first molar. On each study model, interim crowns (N=36) were fabricated using compression molding (molding group, n=12), milling (milling group, n=12), and 3D polymer-jetting methods. The crowns were prepared as follows: molding group, overimpression technique; milling group, a 5-axis dental milling machine; and polymer-jetting group using a 3D printer. The fit of interim crowns was evaluated in the proximal, marginal, internal axial, and internal occlusal regions by using the image-superimposition and silicone-replica techniques. The Mann-Whitney U test and Kruskal-Wallis tests were used to compare the results among groups (α=.05). Compared with the molding group, the milling and polymer-jetting groups showed more accurate results in the proximal and marginal regions (P3D printing significantly enhanced the fit of interim crowns, particularly in the occlusal region. Copyright © 2016 Editorial Council for the Journal of Prosthetic Dentistry. Published by Elsevier Inc. All rights reserved.

Cold Vacuum Drying Facility Design Basis Accident Analysis Documentation

International Nuclear Information System (INIS)

PIEPHO, M.G.

1999-01-01

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

Annual report 1996 concerning the nuclear safety and radiological protection in the Swiss nuclear installations

Energy Technology Data Exchange (ETDEWEB)

NONE

1997-05-01

The report presents detailed information about the nuclear safety and radiological protection in the Swiss nuclear power plants, the central interim storage at Wuerenlingen, the Paul Scherrer Institute (PSI) and other nuclear installations in Switzerland. figs., tabs., refs.

Annual report 1996 concerning the nuclear safety and radiological protection in the Swiss nuclear installations

International Nuclear Information System (INIS)

1997-05-01

The report presents detailed information about the nuclear safety and radiological protection in the Swiss nuclear power plants, the central interim storage at Wuerenlingen, the Paul Scherrer Institute (PSI) and other nuclear installations in Switzerland. figs., tabs., refs

Mixing of incompatible materials in waste tanks technical basis document

International Nuclear Information System (INIS)

SANDGREN, K.R.

2003-01-01

This technical basis document was developed to support the Tank Farms Documented Safety Analysis (DSA) and describes the risk binning process, the technical basis for assigning risk bins, and the controls selected for the mixing of incompatible materials representative accident and associated represented hazardous conditions. The purpose of the risk binning process is to determine the need for safety-significant structures, systems, and components (SSCs) and/or technical safety requirement (TSR)-level controls for a given representative accident or represented hazardous conditions based on an evaluation of the FR-equency and consequence. Note that the risk binning process is not applied to facility workers, because all facility worker hazardous conditions are considered for safety-significant SSCs and/or TSR level controls. Determination of the need for safety-class SSCs was performed in accordance with DOE-STD-3009-94, ''Preparation Guide for US Department of Energy Nonreactor Nuclear Facility Documented Safety Analyses'', as described in this report

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

Plan for safety case of spent fuel repository at Olkiluoto

International Nuclear Information System (INIS)

Vieno, T.; Ikonen, A.T.K.

2005-02-01

Posiva aims to present the Safety Case supporting the construction license application of the spent fuel repository at Olkiluoto by 2012. An outline and preliminary assessments will be presented in 2009. Interim reporting and an update of the Safety Case plan will be presented in 2006, as required by the authorities. The KBS-3 disposal concept aims at long-term isolation and containment of spent fuel assemblies in durable copper-iron canisters emplaced in a repository to be constructed at a depth between 400 and 600 metres in crystalline bedrock. By 2012, studies on the KBS-3 disposal concept and site investigations at Olkiluoto will have been continued over about thirty years. The construction of an underground rock characterisation facility (called ONKALO) was started in June 2004. The investigations are carried out in close cooperation with the Swedish SKB developing and assessing the same disposal concept at candidate sites, resembling Olkiluoto, at the other side of the Baltic Sea. A safety case is the synthesis of evidence, analyses and arguments that quantify and substantiate the safety, and the level of expert confidence in the safety, of a planned repository. Posiva's Safety Case will be organised in a portfolio including ten main reports, which will be periodically updated according the overall schedule presented in the plan. The Site report describing the present state and past evolution of the Olkiluoto site, as well as the disturbances caused by the construction of ONKALO and the first stage of the repository, forms the geoscientific basis of the Safety Case. The engineering basis is provided by the reports on the Characteristics of spent fuel, Canister design, and Repository design. The Process report containing descriptions and analyses of features, events and processes potentially affecting the disposal system, and the report on the Evolution of site and repository form the scientific basis of the Safety Case. The latter report will describe and

A Non-Traditional Interim Project.

Science.gov (United States)

Brown, Diane; Ward, Dorothy

1980-01-01

Describes a project initiated by the Foreign Language Department of Birmingham-Southern College for their Interim term and discusses an interdisciplinary course focusing on Medieval Europe. The course included presentations on German and French language and literature, as well as lectures on the arts, philosophy, and family life of the period.…

Interim-status groundwater monitoring plan for the 216-B-63 trench. Revision 1

Energy Technology Data Exchange (ETDEWEB)

Sweeney, M.D.

1995-06-13

This document outlines the groundwater monitoring plan for interim-status detection-level monitoring of the 216-B-63 Trench. This is a revision of the initial groundwater monitoring plan prepared for Westinghouse Hanford Company (WHC) by Bjornstad and Dudziak (1989). The 216-B-63 Trench, located at the Hanford Site in south-central Washington State, is an open, unlined, earthern trench approximately 1.2 m (4 ft) wide at the bottom, 427 m (1400 ft) long, and 3 m (10 ft) deep that received wastewater containing hazardous waste and radioactive materials from B Plant, located in the 200 East Area. Liquid effluent discharge to the 216-B-63 Trench began in March 1970 and ceased in February 1992. The trench is now managed by Waste Tank Operations.

Interim-status groundwater monitoring plan for the 216-B-63 trench. Revision 1

International Nuclear Information System (INIS)

Sweeney, M.D.

1995-01-01

This document outlines the groundwater monitoring plan for interim-status detection-level monitoring of the 216-B-63 Trench. This is a revision of the initial groundwater monitoring plan prepared for Westinghouse Hanford Company (WHC) by Bjornstad and Dudziak (1989). The 216-B-63 Trench, located at the Hanford Site in south-central Washington State, is an open, unlined, earthern trench approximately 1.2 m (4 ft) wide at the bottom, 427 m (1400 ft) long, and 3 m (10 ft) deep that received wastewater containing hazardous waste and radioactive materials from B Plant, located in the 200 East Area. Liquid effluent discharge to the 216-B-63 Trench began in March 1970 and ceased in February 1992. The trench is now managed by Waste Tank Operations

Safety aspects of long-term dry interim storage of type-B spent fuel and HLW transport casks

International Nuclear Information System (INIS)

Wolff, D.; Probst, U.; Voelzke, H.; Droste, B.; Roedel, R.

2004-01-01

Based on the German decision to minimise transports of spent fuel casks between nuclear power plants, reprocessing plants and central storage facilities several on-site storage facilities have been licensed till the end of 2003. Because of the large amount of type-B transport casks which are going to be used for long-term interim storage the question of time limited type-B license maintenance during the storage period of up to 40 years has been discussed under different aspects. This paper describes present technical aspects of the discussion. A main aspect of transport cask qualification for interim storage is the long-term behaviour of the metallic seal lid system. Concerning this results from current experimental long-term tests with metallic ''Helicoflex''-seals in which pool water is enclosed are presented. The test series has been performed by the Federal Institute for Materials Research and Testing (BAM) on behalf of the Federal Office for Radiation Protection (BfS) since 2001. Finally, the paper presents a German concept for an authorities' and technical experts' exchange of experience, know-how and state of the art referring to cask dispatch in nuclear facilities. BAM has taken over a central role in this so-called ''co-ordinating institution for cask dispatching information'' (''KOBAF'') which contains an online data base and a technical working group meeting twice a year. The goal is to keep comparable technical standards for all nuclear sites and storage facilities which are going to load and dispatch casks of the same or similar types under the responsibility of different German state governments for the next decades

Safety relevant aspects of the long-term intermediate storage of spent fuel elements and vitrified high-level radioactive wastes; Sicherheitstechnische Aspekte der langfristigen Zwischenlagerung von bestrahlten Brennelementen und verglastem HAW

Energy Technology Data Exchange (ETDEWEB)

Ellinger, A.; Geupel, S.; Gewehr, K.; Gmal, B.; Hannstein, V.; Hummelsheim, K.; Kilger, R.; Wagner, M. [Gesellschaft fuer Anlagen- und Reaktorsicherheit mbH (GRS), Koeln (Germany); Schmidt, G.; Spieth-Achtnich, A. [Oeko-Institut e.V. - Institut fuer angewandte Oekolgie (Germany)

2010-04-15

The currently in Germany pursued concept for management of spent fuel from nuclear power plants provides intermediate dry cask storage at the NPP sites until direct disposal in a deep geologic repository. In addition the earlier commissioned centralized dry storage facilities are being used for storage of high level radioactive waste returned from foreign reprocessing of German spent fuel performed so far. The dry interim storage facilities are licensed for 40 years of operation time. According to the German regulations a full scope periodic safety review is not required so far, neither practical experience on dry storage for this period of time is available. With regard to this background the report at hand is dealing with long term effects, which may affect safety of the interim storage during the 40 years period or beyond if appropriate, and with the question, whether additional analyses or monitoring measures may be required. Therefore relevant publications have been evaluated, calculations have been performed as well as a systematic screening with regard to loads and possible ageing effects has been applied to structures and components important for safety of the storage, in order to identify relevant long term effects, which may not have been considered sufficiently so far and to provide proposals for an improved ageing management. The report firstly provides an overview on the current state of technology describing shortly the national and international practice and experience. In the following chapters safety aspects of interim storage with regard to time dependent effects and variations are being analyzed and discussed. Among this not only technical aspects like the long term behavior of fuel elements, canisters and storage systems are addressed, but also operational long term aspects regarding personnel planning, know how conservation, documentation and quality management are taken into account. A separate chapter is dedicated to developing and describing

Interim and final storage casks

International Nuclear Information System (INIS)

Stumpfrock, L.; Kockelmann, H.

2012-01-01

The disposal of radioactive waste is a huge social challenge in Germany and all over the world. As is well known the search for a site for a final repository for high-level waste in Germany is not complete. Therefore, interim storage facilities for radioactive waste were built at plant sites in Germany. The waste is stored in these storage facilities in appropriate storage and transport casks until the transport in a final repository can be carried out. Licensing of the storage and transport casks aimed for use in the public space is done according to the traffic laws and for handling in the storage facility according to nuclear law. Taking into account the activity of the waste to be stored, different containers are in use, so that experience is available from the licensing and operation in interim storage facilities. The large volume of radioactive waste to be disposed of after the shut-down of power generation in nuclear power stations makes it necessary for large quantities of licensed storage and transport casks to be provided soon.

Reducing Health Services for Refugees Through Reforms to the Interim Federal Health Program

Directory of Open Access Journals (Sweden)

Andrew C. Stevenson

2018-04-01

Full Text Available Since 1957 the Interim Federal Health Program (IFHP has provided temporary health care coverage to refugees and refugee claimants, but in 2012 the Conservative government reformed the IFHP, reducing, or eliminating access to health services for these groups. The government framed the changes around fairness and safety, stating that it would save tax payers $100 million over five years, reduce incentive for migrants with unfounded refugee claims from coming to Canada, protect public health and safety, and defend the integrity of the immigration system. With a Conservative majority, the reform was easily implemented despite a lack of evidence supporting these claims. In 2014, the Federal Court rejected the government's notion of fairness and safety, ruling that the cuts were cruel and unusual treatment of an already vulnerable population. The government appealed this ruling but, in 2016, the Liberals took power and restored funding to the IFHP to pre-2012 levels. Ad hoc evaluations predicted inequitable and adverse impacts on refugees, negative impacts on health, and increased costs to refugees, provincial governments, and health providers. Overall the threats and weaknesses of this reform clearly outweighed the few and unconvincing opportunities and strengths of the program, leading to its demise.