Plutonium air transportable package Model PAT-1. Safety analysis report

International Nuclear Information System (INIS)

1978-02-01

The document is a Safety Analysis Report for the Plutonium Air Transportable Package, Model PAT-1, which was developed by Sandia Laboratories under contract to the Nuclear Regulatory Commission (NRC). The document describes the engineering tests and evaluations that the NRC staff used as a basis to determine that the package design meets the requirements specified in the NRC ''Qualification Criteria to Certify a Package for Air Transport of Plutonium'' (NUREG-0360). By virtue of its ability to meet the NRC Qualification Criteria, the package design is capable of safely withstanding severe aircraft accidents. The document also includes engineering drawings and specifications for the package. 92 figs, 29 tables

Safety Analysis Report for the KRI-ASM Transport Package

Energy Technology Data Exchange (ETDEWEB)

Bang, K. S.; Lee, J. C.; Kim, D. H.; Park, H. Y.; Kim, J. B.; Kim, H. J.; Seo, K. S

2005-11-15

Safety evaluation for the KRI-ASM transport package to transport safely I-131, which is produced at HANARO research reactor in KAERI, was carried out. In the safety analyses results for the KRI-ASM transport package, all the maximum stresses as well as the maximum temperature of the surface are lower than their allowable limits. The safety tests were performed by using the test model of the KRI-ASM transport package. Leak Test was performed after drop test and penetration test, the measured leakage rate was lower than allowable leakage rate. It is revealed that the containment integrity of the KRI-ASM transport package is maintained. Therefore, it shows that the integrity of the KRI-ASM transport package is well maintained.

Safety analysis report for packaging (onsite) sample pig transport system

International Nuclear Information System (INIS)

MCCOY, J.C.

1999-01-01

This Safety Analysis Report for Packaging (SARP) provides a technical evaluation of the Sample Pig Transport System as compared to the requirements of the U.S. Department of Energy, Richland Operations Office (RL) Order 5480.1, Change 1, Chapter III. The evaluation concludes that the package is acceptable for the onsite transport of Type B, fissile excepted radioactive materials when used in accordance with this document

Safety analysis report for packaging (onsite) sample pig transport system

Energy Technology Data Exchange (ETDEWEB)

MCCOY, J.C.

1999-03-16

This Safety Analysis Report for Packaging (SARP) provides a technical evaluation of the Sample Pig Transport System as compared to the requirements of the U.S. Department of Energy, Richland Operations Office (RL) Order 5480.1, Change 1, Chapter III. The evaluation concludes that the package is acceptable for the onsite transport of Type B, fissile excepted radioactive materials when used in accordance with this document.

Safety analysis report for packaging: neutron shipping cask, model 4T

International Nuclear Information System (INIS)

Peterson, R.T.

1977-01-01

This Safety Analysis Report for Packaging demonstrates that the neutron shipping cask can safely transport, in solid or powder form, all isotopes of uranium, plutonium, americium, curium, berkelium, californium, einsteinium, and fermium. The cask and its contents are described. It also evaluates transport conditions, structural parameters (e.g., load resistance, pressure and impact effects, lifting and tiedown devices), and shielding. Finally, it discusses compliance with Chapter 0529 of the Energy Research and Development Administration Manual, Safety Standards for the Packaging of Fissile and Other Radioactive Materials

Safety analysis report: packages cobalt-60 shipping cask (packaging of radioactive and fissile materials)

International Nuclear Information System (INIS)

Evans, J.E.; Langhaar, J.W.

1973-07-01

Safety Analysis Report DPSPU-73-124-1 replaces DPSPU-69-124-1 and Supplement 1 to permit shipment of 350,000 curies of 60 Co (maximum) in cobalt-60 shipping casks in compliance with 10 CFR Part 71, Packaging of Radioactive Materials for Transport

Spent fuel packaging and its safety analysis

International Nuclear Information System (INIS)

Takada, Kimitaka; Nakaoki, Kozo; Tamamura, Tadao; Matsuda, Fumio; Fukudome, Kazuyuki

1983-01-01

An all stainless steel B(U) type packaging is proposed to transport spent fuels discharged from research reactors and other radioactive materials. The package is used dry and provided with surface fins to absorb drop shock and to dissipate decay heat. Safety was analyzed for structural, thermal, containment shielding and criticality factors, and the integrity of the package was confirmed with the MARC-CDC, TRUMP, ORIGEN, QAD, ANISN, and KENO computer codes. (author)

Review of criticality safety and shielding analysis issues for transportation packages

International Nuclear Information System (INIS)

Parks, C.V.; Broadhead, B.L.

1995-01-01

The staff of the Nuclear Engineering Applications Section (NEAS) at Oak Ridge National Laboratory (ORNL) have been involved for over 25 years with the development and application of computational tools for use in analyzing the criticality safety and shielding features of transportation packages carrying radioactive material (RAM). The majority of the computational tools developed by ORNL/NEAS have been included within the SCALE modular code system (SCALE 1995). This code system has been used throughout the world for the evaluation of nuclear facility and package designs. With this development and application experience as a basis, this paper highlights a number of criticality safety and shielding analysis issues that confront the designer and reviewer of a new RAM package. Changes in the types and quantities of material that need to be shipped will keep these issues before the technical community and provide challenges to future package design and certification

Nuclear criticality safety analysis for the traveller PWR fuel shipping package

Energy Technology Data Exchange (ETDEWEB)

Vescovi, P.J.; Kent, N.A.; Casado, C.A. [Westinghouse Electric Co., LLC, Columbia, SC (United States)]|[ENUSA Industrias Avanzadas SA, Madrid (Spain)

2004-07-01

The Traveller PWR fresh fuel shipping package represents a radical departure from conventional PWR fuel package designs. Two immediately noticeable features of the Traveller are that it carries a single fuel assembly instead of two as do other package designs, and that it has built-in moderator, which forms part of the flux-trap system. The criticality safety case shows that the Traveller satisfies both U.S. and IAEA licensing requirements, and demonstrates that the package remains acceptably subcritical under normal conditions and hypothetical accident conditions of transport. This paper looks at the modeling techniques that were used to analyze the several accident scenarios that were considered, including: Lattice pitch expansion; Lattice pitch expansion along the fuel assembly length; Preferential flooding (selective flooding of different cavities); Differential flooding (varying water levels inside different cavities); Partial flooding (varying water density); Axial rod displacement; o Sensitivity studies of variable foam densities and boron content in packaging; Analysis for carrying loose rods in a rodbox; The criticality safety case for the Traveller proved to be a successful cooperative effort between ENUSA and Westinghouse.

Nuclear criticality safety analysis for the traveller PWR fuel shipping package

International Nuclear Information System (INIS)

Vescovi, P.J.; Kent, N.A.; Casado, C.A.

2004-01-01

The Traveller PWR fresh fuel shipping package represents a radical departure from conventional PWR fuel package designs. Two immediately noticeable features of the Traveller are that it carries a single fuel assembly instead of two as do other package designs, and that it has built-in moderator, which forms part of the flux-trap system. The criticality safety case shows that the Traveller satisfies both U.S. and IAEA licensing requirements, and demonstrates that the package remains acceptably subcritical under normal conditions and hypothetical accident conditions of transport. This paper looks at the modeling techniques that were used to analyze the several accident scenarios that were considered, including: Lattice pitch expansion; Lattice pitch expansion along the fuel assembly length; Preferential flooding (selective flooding of different cavities); Differential flooding (varying water levels inside different cavities); Partial flooding (varying water density); Axial rod displacement; o Sensitivity studies of variable foam densities and boron content in packaging; Analysis for carrying loose rods in a rodbox; The criticality safety case for the Traveller proved to be a successful cooperative effort between ENUSA and Westinghouse

Structural and Thermal Safety Analysis Report for the Type B Radioactive Waste Transport Package

Energy Technology Data Exchange (ETDEWEB)

Kim, D. H.; Seo, K. S.; Lee, J. C.; Bang, K. S

2007-09-15

We carried out structural safety evaluation for the type B radioactive waste transport package. Requirements for type B packages according to the related regulations such as IAEA Safety Standard Series No. TS-R-1, Korea Most Act. 2001-23 and US 10 CFR Part 71 were evaluated. General requirements for packages such as those for a lifting attachment, a tie-down attachment and pressure condition were considered. For the type B radioactive waste transport package, the structural, thermal and containment analyses were carried out under the normal transport conditions. Also the safety analysis were conducted under the accidental transport conditions. The 9 m drop test, 1 m puncture test, fire test and water immersion test under the accidental transport conditions were consecutively done. The type B radioactive waste transport packages were maintained the structural and thermal integrities.

Safety analysis report for packaging (onsite) doorstop samplecarrier system

Energy Technology Data Exchange (ETDEWEB)

Obrien, J.H.

1997-02-24

The Doorstop Sample Carrier System consists of a Type B certified N-55 overpack, U.S. Department of Transportation (DOT) specification or performance-oriented 208-L (55-gal) drum (DOT 208-L drum), and Doorstop containers. The purpose of the Doorstop Sample Carrier System is to transport samples onsite for characterization. This safety analysis report for packaging (SARP) provides the analyses and evaluation necessary to demonstrate that the Doorstop Sample Carrier System meets the requirements and acceptance criteria for both Hanford Site normal transport conditions and accident condition events for a Type B package. This SARP also establishes operational, acceptance, maintenance, and quality assurance (QA) guidelines to ensure that the method of transport for the Doorstop Sample Carrier System is performed safely in accordance with WHC-CM-2-14, Hazardous Material Packaging and Shipping.

Packaging review guide for reviewing safety analysis reports for packagings: Revision 0

International Nuclear Information System (INIS)

Fischer, L.E.; Chou, C.K.; Lloyd, W.R.; Mount, M.E.; Nelson, T.A.; Schwartz, M.W.; Witte, M.C.

1987-09-01

The Department of Energy (DOE) has established procedures for obtaining certification of packagings used by DOE and its contractors for the transport of radioactive materials. These certification review policies and procedures are established to ensure that DOE packaging designs and operations meet safety criteria at least equivalent to the standards prescribed by the Nuclear Regulatory Commission (NRC) certification process for packaging. The Packaging Review Guide (PRG) is not a DOE order, but has been prepared as guidance for the Packaging Certification Staff (PCS) under the Certifying Official, Office of Security Evaluations, or designated representatives. The principal purpose of the PRG is to assure the quality and uniformity of PCS reviews, and to present a well-defined base from which to evaluate proposed changes in the scope and requirements of reviews. The PRG also sets forth solutions and approaches determined to be acceptable in the past by the PCS in dealing with a specific safety issue or safety-related design area. These solutions and approaches are presented in this form so that reviewers can take consistent and well-understood positions as the same safety issues arise in future cases. An applicant submitting a SARP does not have to follow the solutions or approaches presented in the PRG. However, applicants should recognize that the PCS has spent substantial time and effort in reviewing and developing their positions for the issues. A corresponding amount of time and effort will probably be required to review and accept new or different solutions and approaches. Finally, it is also a purpose of the PRG to make information about DOE certification policy and procedures widely available to DOE field offices, DOE contractors, federal agencies, and interested members of the public. 7 refs., 15 figs., 14 tabs

Safety Evaluation for Packaging 101-SY Hydrogen Mitigation Mixer Pump package

International Nuclear Information System (INIS)

Carlstrom, R.F.

1994-01-01

This Safety Evaluation for Packaging (SEP) provides analysis and considered necessary to approve a one-time transfer of the 101-SY Hydrogen Mitigation Mixer Pump (HMMP). This SEP will demonstrate that the transfer of the HMMP in a new shipping container will provide an equivalent degree of safety as would be provided by packages meeting US Department of Transportation (DOT)/US Nuclear Regulatory Commission (NRC) requirements. This fulfills onsite, transportation requirements implemented by WHC-CM-2-14

Safety Evaluation for Packaging 101-SY Hydrogen Mitigation Mixer Pump package

Energy Technology Data Exchange (ETDEWEB)

Carlstrom, R.F.

1994-10-05

This Safety Evaluation for Packaging (SEP) provides analysis and considered necessary to approve a one-time transfer of the 101-SY Hydrogen Mitigation Mixer Pump (HMMP). This SEP will demonstrate that the transfer of the HMMP in a new shipping container will provide an equivalent degree of safety as would be provided by packages meeting US Department of Transportation (DOT)/US Nuclear Regulatory Commission (NRC) requirements. This fulfills onsite, transportation requirements implemented by WHC-CM-2-14.

An evaluation of a supplementary road safety package.

Science.gov (United States)

Guria, Jagadish; Leung, Joanne

2004-09-01

A Supplementary Road Safety Package (SRSP) was developed in New Zealand in 1995/1996 to supplement the compulsory breath test (CBT) and speed camera programmes introduced in 1993. A major feature of the package was the use of emotion and shock advertising campaigns not only to affect high risk driving attitudes and behaviours towards speeding and drink-driving but also to encourage the use of safety belts. Furthermore, the SRSP also emphasised targeting enforcement to these three areas. This package continued for 5 years. This paper estimates the effect of the package on road trauma. The analysis shows that the Package made substantial impact on road safety and saved over 285 lives over the 5-year period. Copyright 2003 Elsevier Ltd.

K Basins Spent Nuclear Fuel (SNF) Project Safety Analysis Report for Packaging (SARP) approval plan

International Nuclear Information System (INIS)

1995-01-01

This document delineates the plan for preparation, review, and approval of the K Basins Spent Nuclear Fuel (SNF) Packaging Design Criteria (PDC) document and the on-site Safety Analysis Report for Packaging (SARP). The packaging addressed in these documents is used to transport SNF in a Multi- canister Overpack (MCO) configuration

Safety analysis report for packaging (onsite) contaminated well cars

International Nuclear Information System (INIS)

Mercado, J.E.

1998-01-01

In support of past operations, railcars were used to ship irradiated fuel from the 100 Area fuel storage basins to the Plutonium Uranium Extraction (PUREX) Facility. There are two configurations for the packaging systems that transported the fuel: the Three-Well Cask Car, which is outfitted with three casks, and the taller, single well, New Production Reactor (NPR) Cask Car. In this document, these cask cars are referred to collectively as well cars. The purpose of this document is to evaluate and authorize the onsite transportation of well cars that contain significant levels of contamination. No irradiated fuel will be transported in the well cars. Neutron detection data confirmed that the well cars do not contain fuel. The intention is to move 14 retired well cars from their current locations in the 100 Area to a suitable storage location in the 200 Area. Each well car contains Type B quantities of radioactivity; so that the hazard of the transport operation is relatively low. This safety analysis report for packaging (SARP) provides the analyses and evaluations necessary to demonstrate that the contaminated well cars meet the transportation safety requirements of HNF-PRO-154, Responsibilities and Procedures for Hazardous Material Shipments for an onsite packaging. The scope of this document addresses the preparation and transportation of the contaminated well cars

Safety analysis report for packaging (onsite) decontaminated equipment self-container

International Nuclear Information System (INIS)

Boehnke, W.M.

1998-01-01

The purpose of this Safety Analysis Report for Packaging (SARP) is to demonstrate that specific decontaminated equipment can be safely used as its own self-container. As a Decontaminated Equipment Self-Container (also referred to as a self-container), no other packaging, such as a burial box, would be required to transport the equipment onsite. The self-container will consist of a piece of equipment or apparatus which has all readily removable interior contamination removed, all of its external openings sealed, and all external surfaces decontaminated to less than 2000 dpm/100 cm for gamma-emitting radionuclides and less than 220 dpm/100 CM2 for alpha-emitting radionuclides

Safety Analysis Report for Packaging: The unirradiated fuel shipping container USA/9853/AF

International Nuclear Information System (INIS)

1991-01-01

The HFBR Unirradiated Fuel Shipping Container was designed and fabricated at the Oak Ridge National Laboratory in 1978 for the transport of fuel for the High Flux Beam Reactor (HFBR) for Brookhaven National Laboratory. The package has been evaluated analytically, as well as the comparison to tests on similar packages, to demonstrate compliance with the applicable regulations governing packages in which radioactive and fissile materials are transported. The contents of this Safety Analysis Report for Packaging (SARP) are based on Regulatory Guide 7.9 (proposed Revision 2 - May 1986), 10 CFR Part 71, DOE Order 1540.2, DOE Order 5480.3, and 49 CFR Part 173

Safety analysis report for packaging (onsite) L3-181 N basin cask

International Nuclear Information System (INIS)

Adkins, H.E. Jr.

1996-01-01

Purpose of this Safety Analysis Report (SARP) is to authorize the onsite transfer of a Type B, Fissile Excepted, non-highway route controlled quantity in the L3-181 packaging from the N Basin to a storage/disposal facility within 200 West Area. This SARP provides the evaluation necessary to demonstrate that the L3-181 meets the requirements of the 'Hazardous Material Packaging and Shipping', WHC- CM-2-14, by meeting the applicable performance requirements for normal conditions of transport

Packaging review guide for reviewing safety analysis reports for packagings: Revision 1

International Nuclear Information System (INIS)

Fisher, L.E.; Chou, C.K.; Lloyd, W.R.; Mount, M.E.; Nelson, T.A.; Schwartz, M.W.; Witte, M.C.

1988-10-01

The Department of Energy (DOE) has established procedures for obtaining certification of packagings used by DOE and its contractors for the transport of radioactive materials. The principal purpose of this document is to assure the quality and uniformity of PCS reviews and to present a well-defined base from which to evaluate proposed changes in the scope and requirements of reviews. The Packaging Review Guide (PRG) also sets forth solutions and approaches determined to be acceptable in the past in dealing with a specific safety issue or safety-related design area. These solutions and approaches are presented in this form so that reviewers can take consistent and well-understood positions as the same safety issues arise in future cases. An applicant submitting a SARP does not have to follow the solutions or approaches presented. It is also a purpose of the PRG to make information about DOE certification policy and procedures widely available to DOE field offices, DOE contractors, federal agencies, and interested members of the public. 77 refs., 16 figs., 15 tabs

Safety analysis report for packaging, Oak Ridge Y-12 Plant, model DC-1 package with HEU oxide contents. Change pages for Rev.1

Energy Technology Data Exchange (ETDEWEB)

NONE

1995-01-18

This Safety Analysis Report for Packaging for the Oak Ridge Y-12 Plant for the Model DC-1 package with highly enriched uranium (HEU) oxide contents has been prepared in accordance with governing regulations form the Nuclear Regulatory Commission and the Department of Transportation and orders from the Department of energy. The fundamental safety requirements addressed by these regulations and orders pertain to the containment of radioactive material, radiation shielding, and nuclear subcriticality. This report demonstrates how these requirements are met.

Safety analysis report for packaging, Oak Ridge Y-12 Plant, model DC-1 package with HEU oxide contents. Change pages for Rev.1

International Nuclear Information System (INIS)

1995-01-01

This Safety Analysis Report for Packaging for the Oak Ridge Y-12 Plant for the Model DC-1 package with highly enriched uranium (HEU) oxide contents has been prepared in accordance with governing regulations form the Nuclear Regulatory Commission and the Department of Transportation and orders from the Department of energy. The fundamental safety requirements addressed by these regulations and orders pertain to the containment of radioactive material, radiation shielding, and nuclear subcriticality. This report demonstrates how these requirements are met

Safety during sea transport of radioactive materials. Probabilistic safety analysis of package fro sea surface fire accident

International Nuclear Information System (INIS)

Matsuoka, Takeshi; Obara, Isonori; Akutsu, Yukio; Aritomi, Masanori

2000-01-01

The ships carrying irradiated nuclear fuel, plutonium and high level radioactive wastes(INF materials) are designed to keep integrity of packaging based on the various safety and fireproof measures, even if the ship encounters a maritime fire accident. However, granted that the frequency is very low, realistic severe accidents should be evaluated. In this paper, probabilistic safety assessment method is applied to evaluate safety margin for severe sea fire accidents using event tree analysis. Based on our separate studies, the severest scenario was estimated as follows; an INF transport ship collides with oil tanker and induces a sea surface fire. Probability data such as ship's collision, oil leakage, ignition, escape from fire region, operations of cask cooling system and water flooding systems were also introduced from above mentioned studies. The results indicate that the probability of which packages cannot keep their integrity during the sea surface fire accident is very low and sea transport of INF materials is carried out very safely. (author)

Safety analysis report: packages. LP-50 tritium package (packaging of fissile and other radioactive materials). Final report

International Nuclear Information System (INIS)

Gates, A.A.; McCarthy, P.G.; Edl, J.W.

1975-04-01

Elemental tritium is shipped at low pressure in a stainless steel container (LP-50) sealed within an aluminum vessel and surrounded by a minimum of 4-in. thick Celotex insulation in a steel drum. The structural, thermal, containment, shielding, and criticality safety aspects of this package are evaluated. Procedures for loading and unloading, empty cask transport, acceptance testing and maintenance, and quality assurance requirements for the LP-50 package are described in detail. (U.S.)

Safety analysis report, packages. Drath and Schrader Double Lidded Drum (packaging of fissile and other radioactive materials). Final report

International Nuclear Information System (INIS)

Chalfant, G.G.

1985-07-01

The preceding Safety Analysis Report - Packages qualifies the Drath and Schrader Double Lidded Drum (see appendix E) as a Department of Transportation DOT 7A Type A packaging and/or ''Type A'' foreign made packaging. The allowable contents shall be: in solid form; non-fissile or exempt fissile material (as defined by 49 CFR 173.453); less than 700 pounds (318 kg) in weight; equal to or less than the A 1 or A 2 quantities of radioactive material as appropriate (see 49 CFR 173.435 for tables of A 1 /A 2 values); and hydrogen gas generation in radioactive waste shall be limited to a maximum of 2-1/2% and total gas pressure limited to 5 psig. Package marking shall be as specified in 49 CFR 178.350-3 or as specified by the foreign country of origin

Thermal analysis of transportation packaging for nuclear spent fuel

International Nuclear Information System (INIS)

Akamatsu, Hiroshi; Taniuchi, Hiroaki

1989-01-01

Safety analysis of transportation packaging for nuclear spent fuel comprises structural, thermal, containment, shielding and criticality factors, and the safety of a packaging is verified by these analyses. In thermal analysis, the temperature of each part of the packaging is calculated under normal and accident test conditions. As an example of thermal analysis, the temperature distribution of a packaging being subjected to a normal test was calculated by the TRUMP code and compared with measured data. (author)

Improving packaged food quality and safety. Part 1: synchrotron X-ray analysis.

Science.gov (United States)

López-Rubio, A; Hernandez-Muñoz, P; Catala, R; Gavara, R; Lagarón, J M

2005-10-01

The objective was to demonstrate, as an example of an application, the potential of synchrotron X-ray analysis to detect morphological alterations that can occur in barrier packaging materials and structures. These changes can affect the packaging barrier characteristics when conventional food preservation treatments are applied to packaged food. The paper presents the results of a number of experiments where time-resolved combined wide-angle X-ray scattering and small-angle X-ray scattering analysis as a function of temperature and humidity were applied to ethylene-vinyl alcohol co-polymers (EVOH), polypropylene (PP)/EVOH/PP structures, aliphatic polyketone terpolymer (PK) and amorphous polyamide (aPA) materials. A comparison between conventional retorting and high-pressure processing treatments in terms of morphologic alterations are also presented for EVOH. The impact of retorting on the EVOH structure contrasts with the good behaviour of the PK during this treatment and with that of aPA. However, no significant structural changes were observed by wide-angle X-ray scattering in the EVOH structures after high-pressure processing treatment. These structural observations have also been correlated with oxygen permeability measurements that are of importance when guaranteeing the intended levels of safety and quality of packaged food.

Safety Analysis Report for Packaging (SARP): USA/5790/BLF (ERDA-AL) and USA/5791/BLF (ERDA-AL)

International Nuclear Information System (INIS)

Griffin, J.F.; Bertram, R.E.; Blauvelt, R.K.; Edling, D.A.; Flanagan, T.M.; Peterson, J.B.; Prosser, D.L.

1976-01-01

The Safety Analysis Report for Packaging (SARP) satisfies the request of the U. S. Energy Research and Development Administration for a formal safety analysis of the two insulated drum shipping containers identified as USA/5790/BLF ERDA-AL and USA/5791/BLF ERDA-AL. Discussions of structural integrity, thermal resistance, radiation shielding and radiological safety, nuclear criticality safety, and quality control are included. Much of the information was previously submitted to ERDA/OSD/ALO and the Department of Transportation (DOT) and provided the basis for obtaining special permits DOT-SP-5790 and DOT-SP-5791 as well as the Interim Certificates of Compliance until the SARP could be prepared. Complete physical and technical descriptions of the packages are presented. Each package consists of a modified DOT Specification 2R cylindrical steel inner container centered within an insulated steel drum. The contents may be any radioactive materials which satisfy the requirements established in this SARP. A shipment of Plutonium-238 in the form of a solid oxide is evaluated in this SARP as an example. The results of the nuclear criticality safety analysis show how much of the fissile isotopes may be shipped as Fissile Class I, II, or III for each container. Design and development considerations, the test and evaluations required to prove the ability of the containers to withstand normal transportation conditions, and the sequence of four hypothetical accident conditions (free drop, puncture, thermal, and water immersion) are discussed

Safety analysis report for packaging onsite long-length contaminated equipment transport system

International Nuclear Information System (INIS)

McCormick, W.A.

1997-01-01

This safety analysis report for packaging describes the components of the long-length contaminated equipment (LLCE) transport system (TS) and provides the analyses, evaluations, and associated operational controls necessary for the safe use of the LLCE TS on the Hanford Site. The LLCE TS will provide a standardized, comprehensive approach for the disposal of approximately 98% of LLCE scheduled to be removed from the 200 Area waste tanks

Safety analysis report for packaging, onsite, long-length contaminated equipment transport system

Energy Technology Data Exchange (ETDEWEB)

McCormick, W.A.

1997-05-09

This safety analysis report for packaging describes the components of the long-length contaminated equipment (LLCE) transport system (TS) and provides the analyses, evaluations, and associated operational controls necessary for the safe use of the LLCE TS on the Hanford Site. The LLCE TS will provide a standardized, comprehensive approach for the disposal of approximately 98% of LLCE scheduled to be removed from the 200 Area waste tanks.

Safety analysis report for packages: packaging of fissile and other radioactive materials. Final report

International Nuclear Information System (INIS)

Chalfant, G.G.

1984-01-01

The 9965, 9966, 9967, and 9968 packages are designed for surface shipment of fissile and other radioactive materials where a high degree of containment (either single or double) is required. Provisions are made to add shielding material to the packaging as required. The package was physically tested to demonstrate that it meets the criteria specified in USDOE Order No. 5480.1, chapter III, dated 5/1/81, which invokes Title 10, Code of Federal Regulations, Part 71 (10 CFR 71), Packing and Transportation of Radioactive Material, and Title 49, Code of Federal Regulations, Part 100-179, Transportation. By restricting the maximum normal operating pressure of the packages to less than 7 kg/cm 2 (gauge) (99 to 54 psig), the packages will comply with Type B(U) regulations of the International Atomic Energy Agency (IAEA) in its Regulations for the Safe Transport of Radioactive Materials, Safety Series No. 6, 1973 Revised Edition, and may be used for export and import shipments. These packages have been assessed for transport of up to 14.5 kilograms of uranium, excluding uranium-233, or 4.4 kilograms of plutonium metal, oxides, or scrap having a maximum radioactive decay energy of 30 watts. Specific maximum package contents are given. This quantity and the configuration of uranium or plutonium metal cannot be made critical by any combination of hydrogeneous reflection and moderation regardless of the condition of the package. For a uranium-233 shipment, a separate criticality evaluation for the specific package is required

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

International Nuclear Information System (INIS)

Crow, S.R.

1996-01-01

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

Safety analysis report for packaging: neutron shipping cask, model 0.5T

International Nuclear Information System (INIS)

Peterson, R.T.

1976-01-01

The Safety Analysis Report for Packaging demonstrates that the neutron shipping cask can safely transport, in solid or powder form, all isotopes of uranium, plutonium, americium, curium, berkelium, californium, einsteinium, and fermium. The shipping cask and its contents are described. It also evaluates transport conditions, structural parameters (e.g., load resistance, pressure and impact effects, lifting and tiedown devices), and shielding. Finally, it discusses compliance with Chapter 0529 of the Energy Research and Development Administration Manual

Packaging Evaluation Approach to Improve Cosmetic Product Safety

Directory of Open Access Journals (Sweden)

Benedetta Briasco

2016-09-01

Full Text Available In the Regulation 1223/2009, evaluation of packaging has become mandatory to assure cosmetic product safety. In fact, the safety assessment of a cosmetic product can be successfully carried out only if the hazard deriving from the use of the designed packaging for the specific product is correctly evaluated. Despite the law requirement, there is too little information about the chemical-physical characteristics of finished packaging and the possible interactions between formulation and packaging; furthermore, different from food packaging, the cosmetic packaging is not regulated and, to date, appropriate guidelines are still missing. The aim of this work was to propose a practical approach to investigate commercial polymeric containers used in cosmetic field, especially through mechanical properties’ evaluation, from a safety point of view. First of all, it is essential to obtain complete information about raw materials. Subsequently, using an appropriate full factorial experimental design, it is possible to investigate the variables, like polymeric density, treatment, or type of formulation involved in changes to packaging properties or in formulation-packaging interaction. The variation of these properties can greatly affect cosmetic safety. In particular, mechanical properties can be used as an indicator of pack performances and safety. As an example, containers made of two types of polyethylene with different density, low-density polyethylene (LDPE and high-density polyethylene (HDPE, are investigated. Regarding the substances potentially extractable from the packaging, in this work the headspace solid-phase microextraction method (HSSPME was used because this technique was reported in the literature as suitable to detect extractables from the polymeric material here employed.

Safety analysis report for the TRUPACT-II shipping package (condensed version). Volume 1, Rev. 14

International Nuclear Information System (INIS)

1994-10-01

The condensed version of the TRUPACT-II Contact Handled Transuranic Waste Safety Analysis Report for Packaging (SARP) contains essential material required by TRUPACT-II users, plus additional contents (payload) information previously submitted to the U.S. Nuclear Regulatory Commission. All or part of the following sections, which are not required by users of the TRUPACT-II, are deleted from the condensed version: (i) structural analysis, (ii) thermal analysis, (iii) containment analysis, (iv) criticality analysis, (v) shielding analysis, and (vi) hypothetical accident test results

Safety analysis report for the TRUPACT-II shipping package (condensed version). Volume 1, Rev. 14

Energy Technology Data Exchange (ETDEWEB)

NONE

1994-10-01

The condensed version of the TRUPACT-II Contact Handled Transuranic Waste Safety Analysis Report for Packaging (SARP) contains essential material required by TRUPACT-II users, plus additional contents (payload) information previously submitted to the U.S. Nuclear Regulatory Commission. All or part of the following sections, which are not required by users of the TRUPACT-II, are deleted from the condensed version: (i) structural analysis, (ii) thermal analysis, (iii) containment analysis, (iv) criticality analysis, (v) shielding analysis, and (vi) hypothetical accident test results.

The procedures used to review safety analysis reports for packagings submitted to the US Department of Energy for certification

International Nuclear Information System (INIS)

Popper, G.F.; Raske, D.T.; Turula, P.

1988-01-01

This paper presents an overview of the procedures used at the Argonne National Laboratory (ANL) to review Safety Analysis Reports for Packagings (SARPs) submitted to the US Department of Energy (DOE) for issuance of a Certificate of Compliance. Prior to certification and shipment of a packaging for the transport of radioactive materials, a SARP must be prepared describing the design, contents, analyses, testing, and safety features of the packaging. The SARP must be reviewed to ensure that the specific packaging meets all DOE orders and federal regulations for safe transport. The ANL SARP review group provides an independent review and evaluation function for the DOE to ensure that the packaging meets all the prescribed requirements. This review involves many disciplines and includes evaluating the general information, drawings, construction details, operating procedures, maintenance and test programs, and the quality assurance plan for compliance with requirements. 14 refs., 6 figs

Packaging Evaluation Approach to Improve Cosmetic Product Safety

OpenAIRE

Benedetta Briasco; Priscilla Capra; Arianna Cecilia Cozzi; Barbara Mannucci; Paola Perugini

2016-01-01

In the Regulation 1223/2009, evaluation of packaging has become mandatory to assure cosmetic product safety. In fact, the safety assessment of a cosmetic product can be successfully carried out only if the hazard deriving from the use of the designed packaging for the specific product is correctly evaluated. Despite the law requirement, there is too little information about the chemical-physical characteristics of finished packaging and the possible interactions between formulation and packag...

Type B liquid package technical issues -- Experience with LR-56 safety analysis

International Nuclear Information System (INIS)

Smith, A.C.; Alstine, M.N. van; Gromada, R.J.; Hensel, S.J.; Gupta, N.K.

1997-01-01

In the course of the development of nuclear industry in France, shipment of Type B quantities (i.e., quantities having significant radiological consequences) of radioactive liquids between different, sites became necessary. Based on the experience acquired at the Commissariat a l'Energie Atomique (CEA) nuclear centers, a series of tanker trailers has been developed to meet this need. Similarly, as part of the ongoing program to process wastes to stable end forms, a need exists to move radioactive liquids at several DOE sites. The LR-56, developed by CEA to transport liquids of medium to high activity, was selected for these US applications, based on its design features and successful operating experience in France. No comparable Type B liquid packages are certified in the US Packages employed in transport of Type B quantities of liquids are either only suitable for small volumes, or are used within site boundaries with extensive administrative controls employed to insure that an adequate level of safety is maintained. The requirement is to provide safety equivalent to the level established by federal regulations in 10 CFR 71. Type B radioactive materials packages (RAM packages) are typically simple, rugged containers which are designed and fabricated in accordance with the ASME Boiler and Pressure Vessel Code to provide containment under the normal conditions of transport (NCT) and hypothetical accident conditions (HAC) established by the regulations. Packages designed for liquid contents must address a number of technical issues which are not common to packages for solid contents. This paper reviews the technical issues associated with Type B liquid packages from the perspective of the experience gained from the evaluation of the LR-56 for use at DOE sites

Food-packaging interactions influencing quality and safety.

Science.gov (United States)

Hotchkiss, J H

1997-01-01

Interactions between foods and packaging can be detrimental to quality and/or safety. Changes in product flavour due to aroma sorption and the transfer of undesirable flavours from packaging to foods are important mechanisms of deterioration when foods are packaged in polymer-based materials. Careful consideration must be given to those factors affecting such interactions when selecting packaging materials in order to maximize product quality, safety, and shelf-life while minimizing undesirable changes. Product considerations include sensitivity to flavour and related deteriorations, colour changes, vitamin loss, microbial activity, and amount of flavour available. Storage considerations include temperature, time, and processing method. Polymer considerations include type of polymer and processing method, volume or mass of polymer to product ratio, and whether the interaction is Fickian or non-Fickian. Methodology to determine the extent of such interactions must be developed. Direct interactions between food and packaging are not necessarily detrimental. The same principles governing undesirable interactions can be used to affect desirable outcomes. Examples include films which directly intercept or absorb oxygen, inhibit microorganisms, remove undesirable flavours by sorption, or indicate safety and product shelf-life.

Guide for reviewing safety analysis reports for packaging: Review of quality assurance requirements

International Nuclear Information System (INIS)

Moon, D.W.

1988-10-01

This review section describes quality assurance requirements applying to design, purchase, fabrication, handling, shipping, storing, cleaning, assembly, inspection, testing, operation, maintenance, repair, and modification of components of packaging which are important to safety. The design effort, operation's plans, and quality assurance requirements should be integrated to achieve a system in which the independent QA program is not overly stringent and the application of QA requirements is commensurate with safety significance. The reviewer must verify that the applicant's QA section in the SARP contains package-specific QA information required by DOE Orders and federal regulations that demonstrate compliance. 8 refs

Safety Analysis Report for packaging (onsite) steel waste package

International Nuclear Information System (INIS)

BOEHNKE, W.M.

2000-01-01

The steel waste package is used primarily for the shipment of remote-handled radioactive waste from the 324 Building to the 200 Area for interim storage. The steel waste package is authorized for shipment of transuranic isotopes. The maximum allowable radioactive material that is authorized is 500,000 Ci. This exceeds the highway route controlled quantity (3,000 A 2 s) and is a type B packaging

Safety Analysis Report for Packaging (SARP): USA/9507/BLF (ERDA--AL), Model AL-M1

International Nuclear Information System (INIS)

Watkins, R.A.; Bertram, R.E.; Blauvelt, R.K.; Edling, D.A.; Flanagan, T.M.; Griffin, J.F.; Rhinehammer, T.B.

1977-01-01

The SARP includes structural integrity, thermal resistance, radiation shielding and radiological safety, nuclear criticality safety, and quality control of three insulated drum shipping containers identified as USA/9507/BLF (ERDA-AL), also called AL-M1, configurations 1, 3, and 5. Complete physical and technical descriptions of the packages are presented. Each package consists of an inner container centered within an insulated steel drum. The contents are plutonium-239 and uranium-235 in configurations-1 and -3. The configuration-5 package is intended for shipment of up to 100,000 Ci of tritiated water immobilized on a sorbent such as molecular sieve

Safety Analysis Report for Packaging (SARP) for USA/5790/BLF (DOE-AL) and USA/5791/BLF (DOE-AL)

International Nuclear Information System (INIS)

Roome, L.G.; Watkins, R.A.; Bertram, R.E.; Kreider, H.B.

1980-01-01

This revised Safety Analysis Report for Packaging (SARP) includes discussions of structural integrity, thermal resistance, radiation shielding and radiological safety, nuclear criticality safety, and quality control of shipping containers. Much of the information was previously submitted to AEC/OSD/ALO and the Department of Transportation (DOT) and provided the basis for obtaining special permits DOT-SP-5790 and DOT-SP-5791 as well as the Interim Certificates of Compliance until the original SARP could be prepared and Certificates of Compliance issued by ERDA. This SARP revision incorporates information on certain design changes, the most significant of which relate to the inner container for the type 5790 package. Complete physical and technical descriptions of the packages are presented. Each package consists of a cylindrical steel inner container centered within an insulating steel drum assembly. The contents may be any radioactive materials which satisfy the requirements established in this SARP. A shipment of plutonium-238 in the form of a solid oxide is evaluated in this SARP as an example. The results of the nuclear criticality safety analysis show how much of the fissile isotopes may be shipped as Fissile Class I, II, or III for each container. Design and development considerations, the tests and evaluations required to prove the ability of the containers to withstand normal transportation conditions, and the sequence of four hypothetical accident conditions (free drop, puncture, thermal, and water immersion) are discussed. Tables, graphs, dimensional sketches, photographs, technical references, loading and shipping procedures, Mound Facility experience in using the containers, and copies of the DOE Certificates of Compliance are included. Internal reviews of the original and revised SARP's have been performed in compliance with the requirement of DOEM 5201-Part V

Safety Analysis Report for packaging (onsite) steel waste package

Energy Technology Data Exchange (ETDEWEB)

BOEHNKE, W.M.

2000-07-13

The steel waste package is used primarily for the shipment of remote-handled radioactive waste from the 324 Building to the 200 Area for interim storage. The steel waste package is authorized for shipment of transuranic isotopes. The maximum allowable radioactive material that is authorized is 500,000 Ci. This exceeds the highway route controlled quantity (3,000 A{sub 2}s) and is a type B packaging.

FFTF railroad tank car Safety Evaluation for Packaging

International Nuclear Information System (INIS)

Carlstrom, R.F.

1995-01-01

This Safety Evaluation for Packaging (SEP) provides evaluations considered necessary to approve transfer of the 8,000 gallon Liquid Waste Tank Car (LWTC) from Fast Flux Test Facility (FFTF) to the 200 Areas. This SEP will demonstrate that the transfer of the LWTC will provide an equivalent degree of safety as would be provided by packages meeting U.S. Department of Transportation (DOT) requirements. This fulfills onsite transportation requirements implemented in the Hazardous Material Packaging and Shipping, WHC-CM-2-14

Integrating environmental management into food safety and food packaging in Malaysia: review of the food regulation 1985

Science.gov (United States)

Nordin, N. H.; Hara, H.; Kaida, N.

2017-05-01

Food safety is an important issue that is related to public safety to prevent the toxicity threats of the food. Management through legal approach has been used in Malaysia as one of the predominant approaches to manage the environment. In this regard, the Food Regulation 1985 has been one of the mechanisms of environmental management through legal approach in controlling the safety of packaged food in food packaging industry in Malaysia. The present study aims to analyse and to explain the implementation of the Food Regulation 1985 in controlling the safety of packaged food in Malaysia and to integrate the concept of environmental management into the food safety issue. Qualitative analysis on the regulation document revealed that there are two main themes, general and specific, while their seven sub themes are included harmful packages, safety packages, reuse packages, polyvinyl chloride (PVC), alcoholic bottle, toys, money and others and iron powder. The implementation of the Food Regulation 1985 in controlling the safety of packaged food should not be regarded solely for regulation purposes but should be further developed for a broader sense of food safety from overcoming the food poisoning.

Regulatory and extra-regulatory testing to demonstrate radioactive material packaging safety

International Nuclear Information System (INIS)

Ammerman, D.J.

1997-01-01

Packages for the transportation of radioactive material must meet performance criteria to assure safety and environmental protection. The stringency of the performance criteria is based on the degree of hazard of the material being transported. Type B packages are used for transporting large quantities of radioisotopes (in terms of A 2 quantities). These packages have the most stringent performance criteria. Material with less than an A 2 quantity are transported in Type A packages. These packages have less stringent performance criteria. Transportation of LSA and SCO materials must be in open-quotes strong-tightclose quotes packages. The performance requirements for the latter packages are even less stringent. All of these package types provide a high level of safety for the material being transported. In this paper, regulatory tests that are used to demonstrate this safety will be described. The responses of various packages to these tests will be shown. In addition, the response of packages to extra-regulatory tests will be discussed. The results of these tests will be used to demonstrate the high level of safety provided to workers, the public, and the environment by packages used for the transportation of radioactive material

SCALE criticality safety verification and validation package

International Nuclear Information System (INIS)

Bowman, S.M.; Emmett, M.B.; Jordan, W.C.

1998-01-01

Verification and validation (V and V) are essential elements of software quality assurance (QA) for computer codes that are used for performing scientific calculations. V and V provides a means to ensure the reliability and accuracy of such software. As part of the SCALE QA and V and V plans, a general V and V package for the SCALE criticality safety codes has been assembled, tested and documented. The SCALE criticality safety V and V package is being made available to SCALE users through the Radiation Safety Information Computational Center (RSICC) to assist them in performing adequate V and V for their SCALE applications

Safety analysis report: packages 238Pu oxide shipping cask (packaging of fissile and other radioactive materials). Final report

International Nuclear Information System (INIS)

Evans, J.E.; Gates, A.A.

1975-06-01

Plutonium-238 (as PuO 2 powder) is shipped in triple-container stainless steel shipping casks in compliance with ERDA Manual Chapter 0529 (ERDAM 0529), Safety Standards for the Packaging of Fissile and Other Radioactive Materials. (U.S.)

Development on inelastic analysis acceptance criteria for radioactive material transportation packages

International Nuclear Information System (INIS)

Ammerman, D.J.; Ludwigsen, J.S.

1995-01-01

The response of radioactive material transportation packages to mechanical accident loadings can be more accurately characterized by non-linear dynamic analysis than by the ''Equivalent dynamic'' static elastic analysis typically used in the design of these packages. This more accurate characterization of the response can lead to improved package safety and design efficiency. For non-linear dynamic analysis to become the preferred method of package design analysis, an acceptance criterion must be established that achieves an equivalent level of safety as the currently used criterion defined in NRC Regulatory Guide 7.6 (NRC 1978). Sandia National Laboratories has been conducting a study of possible acceptance criteria to meet this requirement. In this paper non-linear dynamic analysis acceptance criteria based on stress, strain, and strain-energy-density will be discussed. An example package design will be compared for each of the design criteria, including the approach of NRC Regulatory Guide 7.6

Packaging design criteria for the Hanford Ecorok Packaging

International Nuclear Information System (INIS)

Mercado, M.S.

1996-01-01

The Hanford Ecorok Packaging (HEP) will be used to ship contaminated water purification filters from K Basins to the Central Waste Complex. This packaging design criteria documents the design of the HEP, its intended use, and the transportation safety criteria it is required to meet. This information will serve as a basis for the safety analysis report for packaging

Aspects of safety assessments for package with additional equipment components

International Nuclear Information System (INIS)

Reiche, I.; Boerst, F.M.; Krietsch, T.

2004-01-01

Many paragraphs in TS-R-1 contain the terms ''package'' or ''packaging''. These terms are defined in TS-R-1 paras 230 and 231 and explained in TS-G-1.1 paras 230.1 - 230.6. The importance of a consistent understanding of these definitions has been shown by recent discussions during the assessment of applications for package design approval. There was disagreement, if equipment components attached to the container body during transport, e.g. a transport frame, should be considered part of the package and taken into account in the safety assessment for the package. Discussions were also caused by the way inner design components are treated in the safety assessment of the package. This paper summarises the regulatory requirements to such additional equipment components and presents the way of their inclusion into the package design approval process in Germany

Use of nanotechnology in food processing, packaging and safety ...

African Journals Online (AJOL)

Use of nanotechnology in food processing, packaging and safety – review. ... application of nanotechnology in food packaging and food contact materials, ... developing active antimicrobial and antifungal surfaces, and sensing as well as ...

Analysis on the Industrial Design of Food Package and the Component of Hazardous Substance in the Packaging Material

OpenAIRE

Wei-Wen Huang

2015-01-01

Transferring the hazardous chemicals contained in food packaging materials into food would threaten the health of consumers, therefore, the related laws and regulations and the detection method of hazardous substance have been established at home and abroad to ensure the safety to use the food packaging material. According to the analysis on the hazardous component in the food packaging, a set of detection methods for hazardous substance in the food packaging was established in the paper and ...

Safety analysis report for packaging (onsite) for limited type Bmaterial in the CNS 14-215H cask

International Nuclear Information System (INIS)

Flanagan, B.D.

1997-01-01

The purpose of this Safety Analysis Report for Packaging is to provide the analyses and evaluations necessary to demonstrate that the CNS 14-215H cask provided by Chem-Nuclear Systems Inc. can safety transport greater than Type A quantities of radioactive material on the Hanford Site. The CNS 14-215H cask was chosen for its loading abilities, availability, and because it has a Certificate of Compliance (CoC) issued by the U.S. Nuclear Regulatory Commission (NRC) for transporting low specific activity in quantities greater than Type A material in commerce. Although the CDC does not cover greater than Type A material not meeting LSA requirements, it does allow for an established level of protection in determining the safety of transporting Type B material on the Hanford Site

Safety Evaluation of Radioactive Material Transport Package under Stacking Test Condition

International Nuclear Information System (INIS)

Lee, Ju Chan; Seo, Ki Seog; Yoo, Seong Yeon

2012-01-01

Radioactive waste transport package was developed to transport eight drums of low and intermediate level waste(LILW) in accordance with the IAEA and domestic related regulations. The package is classified with industrial package IP-2. IP-2 package is required to undergo a free drop test and a stacking test. After free drop and stacking tests, it should prevent the loss or dispersal of radioactive contents, and loss of shielding integrity which would result in more than 20 % increase in the radiation level at any external surface of the package. The objective of this study is to establish the safety test method and procedure for stacking test and to prove the structural integrities of the IP-2 package. Stacking test and analysis were performed with a compressive load equal to five times the weight of the package for a period of 24 hours using a full scale model. Strains and displacements were measured at the corner fitting of the package during the stacking test. The measured strains and displacements were compared with the analysis results, and there were good agreements. It is very difficult to measure the deflection at the container base, so the maximum deflection of the container base was calculated by the analysis method. The maximum displacement at the corner fitting and deflection at the container base were less than their allowable values. Dimensions of the test model, thickness of shielding material and bolt torque were measured before and after the stacking test. Throughout the stacking test, it was found that there were no loss or dispersal of radioactive contents and no loss of shielding integrity. Thus, the package was shown to comply with the requirements to maintain structural integrity under the stacking condition.

Intelligent Packaging Systems: Sensors and Nanosensors to Monitor Food Quality and Safety

Directory of Open Access Journals (Sweden)

Guillermo Fuertes

2016-01-01

Full Text Available The application of nanotechnology in different areas of food packaging is an emerging field that will grow rapidly in the coming years. Advances in food safety have yielded promising results leading to the development of intelligent packaging (IP. By these containers, it is possible to monitor and provide information of the condition of food, packaging, or the environment. This article describes the role of the different concepts of intelligent packaging. It is possible that this new technology could reach enhancing food safety, improving pathogen detection time, and controlling the quality of food and packaging throughout the supply chain.

Use of inelastic analysis to determine the response of packages to puncture accidents

International Nuclear Information System (INIS)

Ammerman, D.J.; Ludwigsen, J.S.

1996-01-01

The accurate analytical determination of the response of radioactive material transportation packages to the hypothetical puncture accident requires inelastic analysis techniques. Use of this improved analysis method recudes the reliance on empirical and approximate methods to determine the safety for puncture accidents. This paper will discuss how inelastic analysis techniques can be used to determine the stresses, strains and deformations resulting from puncture accidents for thin skin materials with different backing materials. A method will be discussed to assure safety for all of these types of packages

Metal food packaging design based on hazard analysis critical control point (HACCP system in canned food safety

Directory of Open Access Journals (Sweden)

Li Xingyi

2016-06-01

Full Text Available This study aims to design metal food packaging with hazard analysis critical control point (HACCP. First, theory of HACCP was introduced in detail. Taking empty cans provided by Wuxi Huapeng Food Packaging Company as an example, we studied migration of bisphenol compounds in coating of food can to food stimulant. Moreover, packaging design of luncheon meat can was taken as an example to confirm whether HACCP system could effectively control migration of phenolic substance. Results demonstrated that, coating of such empty were more likely to contain multiple bisphenol compounds such as bisphenol A (BPA, and bisphenol A diglycidyl ether (BADGE was considered as the leading bisphenol pollutant; food stimulant of different types, storage temperature and time could all impact migration of bisphenol compounds. HACCP system was proved to be effective in controlling hazards of phenolic substance in luncheon meat can and could reduce various phenolic substance indexes to an acceptable range. Therefore, HACCP can control migration of phenolic substance and recontamination of food and thus ensure food safety.

Beneficial uses shipping system (BUSS) cask, safety analysis report for packaging: Volumes 1 and 2

International Nuclear Information System (INIS)

Ferrell, P.C.

1997-01-01

The Beneficial Uses Shipping System (BUSS) cask Safety Analysis Report for Packaging (SARP) was originally prepared by Sandia National Laboratory (SNL). After the certification process was completed, the ownership of the BUSS cask and associated SARP was transferred from SNL to the DOE Hanford site in Richland, Washington. During timely renewal of the BUSS cask certificate of compliance, the SARP was revised to (1) respond to the timely renewal questions, (2) consolidate the previous revision made by SNL, and (3) bring the SARP into compliance with the 1996 version of 10 CFR 71. Since the BUSS cask is now the responsibility of RL, the SARP was reissued as a Hanford document

Overview of the DOE packaging certification process

Energy Technology Data Exchange (ETDEWEB)

Liu, Y.Y.; Carlson, R.D. [Argonne National Lab., IL (United States); Carlson, R.W. [Lawrence Livermore National Lab., CA (United States); Kapoor, A. [USDOE, Washington, DC (United States)

1995-12-31

This paper gives an overview of the DOE packaging certification process, which is implemented by the Office of Facility Safety Analysis, under the Assistance Secretary for Environment, Safety and Health, for packagings that are not used for weapons and weapons components, nor for naval nuclear propulsion. The overview will emphasize Type B packagings and the Safety Analysis Report for Packaging (SARP) review that parallels the NRC packaging review. Other important elements in the DOE packaging certification program, such as training, methods development, data bases, and technical assistance, are also emphasized, because they have contributed significantly to the improvement of the certification process since DOE consolidated its packaging certification function in 1985. The paper finishes with a discussion of the roles and functions of the DOE Packaging Safety Review Steering Committee, which is chartered to address issues and concerns of interest to the DOE packaging and transportation safety community. Two articles related to DOE packaging certification were published earlier on the SARP review procedures and the DOE Packaging Review Guide. These articles may be consulted for additional information.

Active and intelligent packaging: The indication of quality and safety.

Science.gov (United States)

Janjarasskul, Theeranun; Suppakul, Panuwat

2018-03-24

The food industry has been under growing pressure to feed an exponentially increasing world population and challenged to meet rigorous food safety law and regulation. The plethora of media consumption has provoked consumer demand for safe, sustainable, organic, and wholesome products with "clean" labels. The application of active and intelligent packaging has been commercially adopted by food and pharmaceutical industries as a solution for the future for extending shelf life and simplifying production processes; facilitating complex distribution logistics; reducing, if not eliminating the need for preservatives in food formulations; enabling restricted food packaging applications; providing convenience, improving quality, variety and marketing features; as well as providing essential information to ensure consumer safety. This chapter reviews innovations of active and intelligent packaging which advance packaging technology through both scavenging and releasing systems for shelf life extension, and through diagnostic and identification systems for communicating quality, tracking and brand protection.

A review of the safety features of 6M packagings for DOE programs

International Nuclear Information System (INIS)

1988-12-01

This report, prepared by a US Department of Energy (DOE) Task Force and organized for clarity into two-page modules, argues that the US Department of Transportation (DOT) Specification-6M packagings (hereafter referred to as 6M packaging, or simply 6M) merit continued DOE use and, if necessary, DOE certification. This report is designed to address the specific requirements of a Safety Analysis Report for Packaging (SARP). While not a SARP, this report constitutes a compilation of all available documentation on 6M packagings. The authors individually, and the Task Force collectively, believe their investigation provides justification for the continued use of 6M packagings because they meet criteria for quality assurance and for safety under normal and accident conditions as defined by the US Nuclear Regulatory Commission (NRC) regulations. This report may be used by DOE managers to assist in deliberations on future requirements for 6M packagings as they are required to support DOE programs. For the purpose of ready evaluation, this report includes categorical topics found in Nuclear Regulatory Guide 7.9, the topical guideline for SARPs. The format, however, will (it is hoped) pleasantly surprise customary reader expectations. For, while maintaining categorical headings and subheadings found in SARPs as a skeleton, the Task Force chose to adopt the document design principles developed by Hughes Aircraft in the 1960s, ''Sequential Thematic Organization of Publications'' (STOP). 37 figs

Safety analysis report for packaging (SARP) of the Oak Ridge National Laboratory Garden Carrier No. 2

International Nuclear Information System (INIS)

Klima, B.B.; Shappert, L.B.; Seagren, R.D.; Box, W.D.

1978-04-01

An analytical evaluation of the Oak Ridge National Laboratory Garden Carrier No. 2 was made to demonstrate its compliance with the regulations governing off-site radioactive material shipping packages. The evaluation encompassed five primary categories: structural integrity, thermal resistance, radiation shielding, nuclear criticality safety, and quality assurance. The results of the evaluation show that the cask complies with the applicable regulations. The package is designed to ship large quantities of fissile and radioactive materials as solids

Safety evaluation for packaging (onsite) SERF cask

International Nuclear Information System (INIS)

Edwards, W.S.

1997-01-01

This safety evaluation for packaging (SEP) documents the ability of the Special Environmental Radiometallurgy Facility (SERF) Cask to meet the requirements of WHC-CM-2-14, Hazardous Material Packaging and Shipping, for transfer of Type B quantities (up to highway route controlled quantities) of radioactive material within the 300 Area of the Hanford Site. This document shall be used to ensure that loading, tie down, transport, and unloading of the SERF Cask are performed in accordance with WHC-CM-2-14. This SEP is valid until October 1, 1999. After this date, an update or upgrade to this document is required

Safety analysis report for packaging (Oak Ridge Y-12 Plant Model DT-14A package for enriched uranium), Y/DD-326 Supplement No. 1 (Rev. 0)

International Nuclear Information System (INIS)

Cadelli, G.; Kennedy, W.R. Jr.

1986-01-01

This Supplement to the Safety Analysis Report Y/DD-326 [1.4.1] was prepared in accordance with US NRC Regulatory Guide 7.9 to address the Naval Fuel contents to be shipped in the DT-14A packaging. New chapters were written for Shielding, Critcality and Quality Assurance. Independent reviews were obtained for each of these chapters and for a product leakage analysis referenced in Chapter 7.0 of this document. Other chapters invoke the Y/DD-326 SARP with minor changes or additions as indicated in the Supplement. All references to form or composition of the fuel are classified Confidential NNPI in the interest of national security

Recent developments in intelligent packaging for enhancing food quality and safety.

Science.gov (United States)

Sohail, Muhammad; Sun, Da-Wen; Zhu, Zhiwei

2018-03-07

The role of packaging cannot be denied in the life cycle of any food product. Intelligent packaging is an emerging technology in the food packaging sector. Although it still needs its full emergence in the market, its importance has been proved for the maintenance of food quality and safety. The present review describes several aspects of intelligent packaging. It first highlights different tools used in intelligent packaging and elucidates the role of these packaging devices for maintaining the quality of different food items in terms of controlling microbial growth and gas concentration, and for providing convenience and easiness to its users in the form of time temperature indication. This review also discusses other intelligent packaging solutions in supply chain management of food products to control theft and counterfeiting conducts and broaden the image of the food companies in terms of branding and marketing. Overall, intelligent packaging can ensure food quality and safety in the food industry, however there are still some concerns over this emerging technology including high cost and legal aspects, and thus future work should be performed to overcome these problems for further promoting its applications in the food industry. Moreover, work should also be carried out to combine several single intelligent packaging devices into a single one, so that most of the benefits from this emerging technology can be achieved.

Safety evaluation for packaging (onsite) product removal can containers

International Nuclear Information System (INIS)

Burnside, M.E.

1998-01-01

Six Product Removal (PR) Cans and Containers are located within the Plutonium Finishing Plant. Each can is expected to contain a maximum of 3 g of residual radioactive material, consisting mainly of plutonium isotopes. The PR Can Containers were previously authorized by HNF-SD-TP-SEP-064, Rev. 0 (Boettger 1997), for the interarea transport of up to 3 g of plutonium. The purpose of this safety evaluation for packaging is to allow the transport of six PR Cans with their Containers from the Plutonium Finishing Plant to the 233 S Evaporator Facility. This safety evaluation for packaging is authorized for use until April 29, 1999, or until the shipment is made, whichever happens first

Safety analysis report for the TRUPACT-II shipping package (condensed version). Volume 2, Rev. 14

Energy Technology Data Exchange (ETDEWEB)

NONE

1994-10-01

This appendix determines the effective G values for payload shipping categories of contact handled transuranic (CH-TRU) waste materials, based on the radiolytic G values for waste materials that are discussed in detail in Appendix 3.6.8 of the Safety Analysis Report for the TRUPACT-II Shipping Package. The effective G values take into account self-absorption of alpha decay energy inside particulate contamination and the fraction of energy absorbed by nongas-generating materials. As described in Appendix 3.6.8, an effective G value, G{sub eff}, is defined by: G{sub eff} - {Sigma}{sub M} (F{sub M} x G{sub M}) F{sub M}-fraction of energy absorbed by material maximum G value for a material where the sum is over all materials present inside a waste container. The G value itself is determined primarily by the chemical properties of the material and its temperature. The value of F is determined primarily by the size of the particles containing the radionuclides, the distribution of radioactivity on the various materials present inside the waste container, and the stopping distance of alpha particles in air, in the waste materials, or in the waste packaging materials.

Safety analysis report for the TRUPACT-II shipping package (condensed version). Volume 2, Rev. 14

International Nuclear Information System (INIS)

1994-10-01

This appendix determines the effective G values for payload shipping categories of contact handled transuranic (CH-TRU) waste materials, based on the radiolytic G values for waste materials that are discussed in detail in Appendix 3.6.8 of the Safety Analysis Report for the TRUPACT-II Shipping Package. The effective G values take into account self-absorption of alpha decay energy inside particulate contamination and the fraction of energy absorbed by nongas-generating materials. As described in Appendix 3.6.8, an effective G value, G eff , is defined by: G eff - Σ M (F M x G M ) F M -fraction of energy absorbed by material maximum G value for a material where the sum is over all materials present inside a waste container. The G value itself is determined primarily by the chemical properties of the material and its temperature. The value of F is determined primarily by the size of the particles containing the radionuclides, the distribution of radioactivity on the various materials present inside the waste container, and the stopping distance of alpha particles in air, in the waste materials, or in the waste packaging materials

Safety evaluation for packaging CPC metal boxes

International Nuclear Information System (INIS)

Romano, T.

1995-01-01

This Safety Evaluation for Packaging (SEP) provides authorization for the use of Container Products Corporation (CPC) metal boxes, as described in this document, for the interarea shipment of radioactive contaminated equipment and debris for storage in the Central Waste Complex (CWC) or T Plant located in the 200 West Area. Authorization is granted until November 30, 1995. The CPC boxes included in this SEP were originally procured as US Department of Transportation (DOT) Specification 7A Type A boxes. A review of the documentation provided by the manufacturer revealed the documentation did not adequately demonstrate compliance to the 4 ft drop test requirement of 49 CFR 173.465(c). Preparation of a SEP is necessary to document the equivalent safety of the onsite shipment in lieu of meeting DOT packaging requirements until adequate documentation is received. The equivalent safety of the shipment is based on the fact that the radioactive contents consist of contaminated equipment and debris which are not dispersible. Each piece is wrapped in two layers of no less than 4 mil plastic prior to being placed in the box which has an additional 10 mil liner. Pointed objects and sharp edges are padded to prevent puncture of the plastic liner and wrapping

An updated status of Department of Energy safety reviews of packages for transporting radioactive material

International Nuclear Information System (INIS)

Kapoor, A.

1995-01-01

The Department of Energy conducts conformance reviews and issues Certificates of Compliance for Type B packaging for radioactive materials. Several offices within DOE perform these reviews which are required by the Department of Transportation to be to the regulations promulgated by the Nuclear Regulatory Commission or their safety equivalent. This paper focuses on one of these offices, the Office of Facility Safety Analysis, EH-32, which is responsible for reviewing and certifying packages other than those used for weapons and weapons component, for Naval Reactors, and for Civilian Radioactive Waste Management. This paper gives the background and organizational history of EH-32, discusses the version of regulations to which the packaging is reviewed, updates the status of these reviews, describes the effectiveness of the reviews, updates the training courses sponsored by EH-32, and mentions the new Quality Assurance Evaluations being started by EH-32

Evaluation of safety margin of packaging for radioactive materials transport during a severe fire

International Nuclear Information System (INIS)

Gilles, P.; Ringot, C.; Warniez, P.; Grall, L.; Perrot, J.

1986-06-01

A high safety is obtained by International regulations on radioactive materials transport. It is obtained by packaging design adapted to the potential risk. An important accident to consider is fire for two reasons: the probability of fire occuring for time and temperature higher than conditions applied to type B packaging (800 deg C, 1/2 hr) is not negligible, particularly for air or maritime transport. Safety margins are studied by computation and experimental tests. This report presents results obtained for different types of packagings. Results show a large safety margin [fr

Critical review of controlled release packaging to improve food safety and quality.

Science.gov (United States)

Chen, Xi; Chen, Mo; Xu, Chenyi; Yam, Kit L

2018-03-19

Controlled release packaging (CRP) is an innovative technology that uses the package to release active compounds in a controlled manner to improve safety and quality for a wide range of food products during storage. This paper provides a critical review of the uniqueness, design considerations, and research gaps of CRP, with a focus on the kinetics and mechanism of active compounds releasing from the package. Literature data and practical examples are presented to illustrate how CRP controls what active compounds to release, when and how to release, how much and how fast to release, in order to improve food safety and quality.

The IRSN experience feedback for the transport package design safety appraisals

International Nuclear Information System (INIS)

Sert, G.

2007-01-01

The activity of transportation of radioactive materials is in constant evolution; air transport of radio elements for medical use is growing rapidly as well as transport of instruments equipped with radioactive sources for inspections of buildings (controls of presence of lead in paintings) and in industry (non destructive examination of welding by gammagraphy, controls of density on building sites). Transports associated with the recycling of plutonium for the production of electricity by nuclear energy are now accomplished in routine. Globally, 900.000 packages are shipped each year in France; among them, approximately 100.000 packages belong to the category for which design approval is required. To maintain a high level of safety for this activity by limiting the probability of occurrence, the severity and consequences of the incidents and accidents, strict rules are implemented by users under the control of the Safety Authority According to the systematic approach of defence in depth, which is defined by the three principles of safety in design, of operational reliability and of effectiveness of emergency response, the robustness of the design of the package is of primary importance. It is based on regulatory requirements relating to the functions of safety (containment of radioactivity, protection against radiation and prevention of the risks of criticality) that must be ensured by the package in conditions of transport as well as in accident conditions. These rules and the way of applying them evolve with time. Indeed, on the one hand the regulation is reexamined periodically; on the other hand, the technical knowledge on the behaviour of the packages subject to the above mentioned conditions and the means of evaluation of this behaviour progress permanently

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

NARCIS (Netherlands)

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

2009-01-01

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

Recommendations for preparing the criticality safety evaluation of transportation packages

International Nuclear Information System (INIS)

Dyer, H.R.; Parks, C.V.

1997-04-01

This report provides recommendations on preparing the criticality safety section of an application for approval of a transportation package containing fissile material. The analytical approach to the evaluation is emphasized rather than the performance standards that the package must meet. Where performance standards are addressed, this report incorporates the requirements of 10 CFR Part 71. 12 refs., 6 figs., 8 tabs

The development of a packaging handbook

International Nuclear Information System (INIS)

Shappert, L.B.

1994-01-01

The Packaging Handbook, dealing with the development of packagings designed to carry radioactive material, is being written for DOE's Transportation and Packaging Safety Division. The primary goal of the Handbook is to provide sufficient technical information and guidance to improve the quality of Safety Analysis Reports on Type B Packagings (SARPs) that are submitted to DOE for certification. This paper provides an update on the status of the Handbook

Structural safety test and analysis of type IP-2 transport packages with bolted lid type and thick steel plate for radioactive waste drums in a NPP

International Nuclear Information System (INIS)

Kim, Dong Hak; Seo, Ki Seog; Lee, Sang Jin; Lee, Kyung Ho; Kim, Jeong Mook

2007-01-01

If a type IP-2 transport package were to be subjected to a free drop test and a penetration test under the normal conditions of transport, it should prevent a loss or dispersal of the radioactive contents and a more than 20% increase in the maximum radiation level at any external surface of the package. In this paper, we suggested the analytic method to evaluate the structural safety of a type IP-2 transport package using a thick steel plate for a structure part and a bolt for tying a bolt. Using an analysis a loss or disposal of the radioactive contents and a loss of shielding integrity were confirmed for two kinds of type IP-2 transport packages to transport radioactive waste drums from a waste facility to a temporary storage site in a nuclear power plant. Under the free drop condition the maximum average stress at the bolts and the maximum opening displacement of a lid were compared with the tensile stress of a bolt and the steps in a lid, which were made to avoid a streaming radiation in the shielding path, to evaluate a loss or dispersal of radioactive waste contents. Also a loss of shielding integrity was evaluated using the maximum decrease in a shielding thickness. To verify the impact dynamic analysis for free drop test condition and evaluate experimentally the safety of two kinds of type IP-2 transport packages, free drop tests were conducted with various drop directions

Safety Evaluation for Packaging (onsite) T Plant Canyon Items

International Nuclear Information System (INIS)

OBRIEN, J.H.

2000-01-01

This safety evaluation for packaging (SEP) evaluates and documents the ability to safely ship mostly unique inventories of miscellaneous T Plant canyon waste items (T-P Items) encountered during the canyon deck clean off campaign. In addition, this SEP addresses contaminated items and material that may be shipped in a strong tight package (STP). The shipments meet the criteria for onsite shipments as specified by Fluor Hanford in HNF-PRO-154, Responsibilities and Procedures for all Hazardous Material Shipments

Safety Evaluation for Packaging (onsite) T Plant Canyon Items

Energy Technology Data Exchange (ETDEWEB)

OBRIEN, J.H.

2000-07-14

This safety evaluation for packaging (SEP) evaluates and documents the ability to safely ship mostly unique inventories of miscellaneous T Plant canyon waste items (T-P Items) encountered during the canyon deck clean off campaign. In addition, this SEP addresses contaminated items and material that may be shipped in a strong tight package (STP). The shipments meet the criteria for onsite shipments as specified by Fluor Hanford in HNF-PRO-154, Responsibilities and Procedures for all Hazardous Material Shipments.

Safety analysis report for packaging (SARP) of the Oak Ridge National Laboratory Foamglas Shipping Container

International Nuclear Information System (INIS)

Klima, B.B.; Shappert, L.B.; Seagren, R.D.; Box, W.D.

1979-01-01

An analytical evaluation of the ORNL Foamglas Shipping Container was made to demonstrate its compliance with the regulations governing offsite radioactive material shipping packages. The evaluation encompassed five primary categories: structural integrity, thermal resistance, radiation shielding, nuclear criticality safety, and quality assurance. The results of the evaluation show that the container complies with the applicable regulations

Safety Analysis Report for Packaging (SARP) of the Oak Ridge National Laboratory Foamglas Shipping Container

International Nuclear Information System (INIS)

Klima, B.B.; Shappert, L.B.; Seagren, R.D.; Box, W.D.

1978-05-01

An analytical evaluation of the Oak Ridge National Laboratory (ORNL) Foamglas Shipping Container was made to demonstrate its compliance with the regulations governing offsite radioactive material shipping packages. The evaluation encompassed five primary categories: structural integrity, thermal resistance, radiation shielding, nuclear criticality safety, and quality assurance. The results of the evaluation show that the container complies with the applicable regulations

Specification of safety requirements for waste packages with respect to practicable quality control measures

International Nuclear Information System (INIS)

Gruendler, D.; Wurtinger, W.

1987-01-01

Waste packages for disposal in a repository in the Federal Republic of Germany have to meet safety requirements derived from site specific safety analyses. The examination of the waste packages with regard to compliance with these requirements is the main objective of quality control measures. With respect to quality control the requirements have to be specified in a way that practicable control measures can be applied. This is dealt with for the quality control of the activity inventory and the quality control of the waste form. The paper discusses the determination of the activity of hard-to-measure radionuclides and the specification of safety related requirements for the waste form and the packaging using typical examples

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

Energy Technology Data Exchange (ETDEWEB)

West, M H

2008-12-17

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

Demonstration tests for low level radioactive waste packaging safety

International Nuclear Information System (INIS)

Nagano, I.; Shimura, S.; Miki, T.; Tamamura, T.; Kunitomi, K.

1993-01-01

The transport packaging for low level radioactive waste (so-called the LLW packaging) has been developed to be utilized for transportation of LLW in 200 liter-drums from Japanese nuclear power stations to the LLW Disposal Center at Rokkashomura in Aomori Prefecture. Transportation is expected to start from December in 1992. We will explain the brief history of the development, technical features and specifications as well as two kinds of safety demonstration tests, namely one is '1.2 meter free drop test' and the other is 'ISO container standard test'. (J.P.N.)

Drug packaging in 2014: authorities should direct more efforts towards medication safety.

Science.gov (United States)

2015-05-01

In 2014, Prescrire examined the packaging quality of about 250 drugs. A few advances stand out, mainly involving recent drugs, but on the whole, the situation is worrisome in terms of medication safety. Although pharmaceutical companies and drug regulatory agencies seem to be taking more account of the risk of accidental poisoning in children, the level of protection remains low overall in the absence of stringent measures on the part of the authorities. New drugs too often have poor-quality or even dangerous packaging at the time of their market introduction. And the packaging quality of older drugs is disturbing. Pharmaceutical companies no longer invest in the packaging of these products, and agencies often fail to take advantage of the opportunities provided by their reassessment to improve the situation. The inappropriate labelling of certain injectable drugs remains a source of medication errors, sometimes resulting in very serious consequences. In 2014, signs of progress in the packaging of several drugs show that its role in medication safety is better appreciated. But the persistence of dangers in the pharmaceuticals market, created by "unfinished", overly complex or poor-quality packaging, raises the question of the responsibility of pharmaceutical companies and agencies for past and present accidents.

Safety analysis report for packaging: the ORNL DOT specification 6M - special form package

Energy Technology Data Exchange (ETDEWEB)

Schaich, R.W.

1982-07-01

The ORNL DOT Specification 6M - Special Form Package was fabricated at the Oak Ridge Nation al Laboratory (ORNL) for the transport of Type B solid non-fissile radioactive materials in special form. The package was evaluated on the basis of tests performed by the Dow Chemical Company, Rocky Flats Division, on the DOT-6M container and special form tests performed on a variety of stainless steel capsules at ORNL by Operations Division personnel. The results of these evaluations demonstrate that the package is in compliance with the applicable regulations for the transport of Type B quantities in special form of non-fissile radioactive materials.

Safety analysis report for packaging: the ORNL DOT specification 6M - special form package

International Nuclear Information System (INIS)

Schaich, R.W.

1982-07-01

The ORNL DOT Specification 6M - Special Form Package was fabricated at the Oak Ridge Nation al Laboratory (ORNL) for the transport of Type B solid non-fissile radioactive materials in special form. The package was evaluated on the basis of tests performed by the Dow Chemical Company, Rocky Flats Division, on the DOT-6M container and special form tests performed on a variety of stainless steel capsules at ORNL by Operations Division personnel. The results of these evaluations demonstrate that the package is in compliance with the applicable regulations for the transport of Type B quantities in special form of non-fissile radioactive materials

Safety analysis report for packaging: the ORNL DOT specification 6M - tritium trap package

International Nuclear Information System (INIS)

DeVore, J.R.

1984-04-01

The ORNL DOT Specification 6M--Tritium Trap Package was fabricated at the Oak Ridge National Laboratory (ORNL) for the transport of Type B quantities of tritium as solid uranium tritide. The package was evaluated on the basis of tests performed by the Dow Chemical Company, Rocky Flats Division, on the DOT-6M container, a drop test performed by the ORNL Operations Division, and International Atomic Energy Agency (IAEA) approvals on a similar tritium transport container. The results of these evaluations demonstrate that the package is in compliance with the applicable regulations for the transport of Type B quantities of tritium. 4 references, 8 figures

Classification of transportation packaging and dry spent fuel storage system components according to importance to safety

International Nuclear Information System (INIS)

Tyacke, M.J.; McConnell, J.W. Jr.; Ayers, A.L. Jr.; O'Connor, S.C.; Jankovich, J.P.

1996-01-01

The Idaho National Engineering Laboratory prepared a technical report for the Office of Nuclear Material Safety and Safeguards of the US Nuclear Regulatory Commission, entitled Classification of Transportation Packaging and Dry Spent Fuel Storage System Components According to Importance to Safety, NUREG/CR-6407. This paper provides the results of that report. It also presents the graded approach for classification of components used in transportation packagings and dry spent fuel storage systems. This approach provides a method for identifying the classification of components according to importance to safety within transportation packagings and dry spent fuel storage systems. Record retention requirements are discussed to identify the documentation necessary to validate that the individual components were fabricated in accordance with their assigned classification. A review of the existing regulations pertaining to transportation packagings and dry storage systems was performed to identify current requirements. The general types of transportation packagings and dry storage systems are identified. The methodology used in this paper is based on Regulatory Guide 7.10, Establishing Quality Assurance Programs for Packaging Used in the Transport of Radioactive Material. This paper also includes a list of generic components for each of the general types of transportation packagings and spent fuel storage systems, with a classification category assigned to each component. Several examples concerning the safety importance of components are presented

Safety analysis report: packages. LP-12 tritium package (packaging of fissile and other radioactive materials). Final report

International Nuclear Information System (INIS)

Gates, A.A.; McCarthy, P.G.; Edl, J.W.

1975-05-01

Elemental tritium is shipped at low pressure in a stainless steel container (LP-12) within an aluminum vessel and surrounded by 3.9 in.-thick Celotex insulation in a steel drum. Information is presented on the packaging design, evaluation of the structural, thermal, containment, shielding, and criticality characteristics of the package, procedures for loading, unloading, transporting, and testing the LP-12, and quality assurance requirements. (U.S.)

Review of the DOE Packaging and Transportation Safety Program

International Nuclear Information System (INIS)

Snyder, B.J.; Cece, J.M.

1992-12-01

This report documents the results of a year-long self-assessment of DOE-EH transportation and packaging safety activities. The self-assessment was initiated in September 1991 and concluded in August 1992. The self-assessment identified several significant issues, some of which have been resolved by EH. Also, improvements in the EH program were made during the course of the self-assessment. The report reflects the status of the EH transportation and packaging safety activities at the conclusion of the self-assessment. This report consists of several sections which discuss background, objectives and description of the review. Another section includes summary discussion and key conclusions. Appendix A, Issues, Observations and Recommendations, lists fifteen issues, including appropriate observations and recommendations. A Corrective Action Plan, which documents EH managements resolve to implement the agreed-upon recommendations, is included. The Corrective Action Plan reflects the status of completed and planned actions as of the date of the report

A probabilistic safety assessment of radioactive materials transport. A risk analysis of LLW package handling at harbor

International Nuclear Information System (INIS)

Watabe, Naohito; Suzuki, Hiroshi; Kouno, Yutaka

1997-01-01

The Probabilistic Safety Assessment (PSA) method for radioactive materials (RAM) transport has been developed by CRIEPI. A case study was executed for the purpose of studying the adaptability of the PSA method to LLW package handling, which is one of the processes of the actual transport. The main results of the case study were as follows; 1) Accident scenarios for falling of package were extracted from the 25 ton-crane system chart and package handling manual. 2) Protection methods for each drop accident scenario were confirmed. 3) Important points of the crane system were extracted. 4) Fault trees, which describe accident scenarios, were developed. 5) Probabilities for each basic event and the top event on fault trees were calculated. Consequently, 'falling of a package' on the package handling process by the 25 ton-crane was revealed to be extremely low. Among the four major stages of handling process, i.e. 'Rolling-up', 'Horizontal travelling' 'Rolling-down' and 'Contact with loading platform', the 'Rolling-down' process was found to be a major process with occupies more than 50% of the probability of the falling accidents. According to those results, it was concluded that PSA method is adaptable to package handling from the view points of extraction of weak points and review of the effect of vestment for facility. (author)

Safety analysis report for packaging (SARP) of the Oak Ridge National Laboratory. TRU curium shipping container

International Nuclear Information System (INIS)

Box, W.D.; Klima, B.B.; Seagren, R.D.; Shappert, L.B.; Aramayo, G.A.

1980-06-01

An analytical evaluation of the Oak Ridge National Laboratory Transuranium (TRU) Curium Shipping Container was made to demonstrate its compliance with the regulations governing offsite shipment of packages containing radioactive material. The evaluation encompassed five primary categories: structural integrity, thermal resistance, radiation shielding, nuclear criticality safety, and quality assurance. The results of the evaluation show that the container complies with the applicable regulations

Quality and safety aspects of reusable plastic food packaging materials : influence of reuse on intrinsic properties

NARCIS (Netherlands)

Jetten, J.; Kruijf, N. de

2002-01-01

The aim of the project was to develop a comprehensive package of quality assurance criteria for use by the industry and by regulatory authorities to ensure the quality and safety-in-use (sensory, microbiological and chemical safety) of reused plastics for food packaging. The paper describes the

The TE coupled RELAP5/PANTHER/COBRA code package and methodology for integrated PWR accident analysis

International Nuclear Information System (INIS)

Schneidesch, Christophe R.; Zhang, Jinzhao; Ammirabile, Luca; Dalleur, Jean-Paul

2006-01-01

At Tractebel Engineering (TE), a dynamic coupling has been developed between the best estimate thermal hydraulics system code RELAP5 and the 3-dimensional neutronics code PANTHER via the transient analysis code linkage program TALINK. An interface between PANTHER and the subchannel core thermal-hydraulic analysis code COBRA 3C has been established for on-line calculation of the Departure from Nucleate Boiling Ratio (DNBR). In addition to the standard RELAP5-PANTHER coupling, the fully dynamic coupling of the RELAP5/PANTHER/COBRA3C-TE code package can be activated for evaluation purposes in which the PANTHER close-channel thermal-hydraulics module is replaced by the COBRA3C-TE with cross flow modelling and extended T-H flow conditions capabilities. The qualification of the RELAP5-PANTHER coupling demonstrated the robustness achieved by the combined 3-D neutron kinetics/system T-H code package for transient simulations. The coupled TE code package has been approved by the Belgian Safety Authorities and is used at TE for analyzing asymmetric PWR accidents with strong core-system interactions. In particular, the TE coupled code package was first used to develop a main steam line break in hot shutdown conditions (SLBHZP) accident analysis methodology based on the TE deterministic bounding approach. This methodology has been reviewed and accepted by the Belgian Safety Authorities for specific applications. Those specific applications are related to the power up-rate and steam generator replacement project of the Doel 2 plant or to the Tihange-3 SLB accident re-analysis. A coupled feedwater line break (FLB) accident analysis methodology is currently being reviewed for application approval. The results of coupled thermal-hydraulic and neutronic analysis of SLB and FLB show that there exist important margins in the traditional final safety analysis report (FSAR) accident analysis. Those margins can be used to increase the operational flexibility of the plants. Moreover, the

The TE coupled RELAP5/PANTHER/COBRA code package and methodology for integrated PWR accident analysis

Energy Technology Data Exchange (ETDEWEB)

Schneidesch, Christophe R.; Zhang, Jinzhao; Ammirabile, Luca; Dalleur, Jean-Paul [Suez-Tractebel Engineering, Avenue Ariane 7, B-1200 Brussels (Belgium)

2006-07-01

At Tractebel Engineering (TE), a dynamic coupling has been developed between the best estimate thermal hydraulics system code RELAP5 and the 3-dimensional neutronics code PANTHER via the transient analysis code linkage program TALINK. An interface between PANTHER and the subchannel core thermal-hydraulic analysis code COBRA 3C has been established for on-line calculation of the Departure from Nucleate Boiling Ratio (DNBR). In addition to the standard RELAP5-PANTHER coupling, the fully dynamic coupling of the RELAP5/PANTHER/COBRA3C-TE code package can be activated for evaluation purposes in which the PANTHER close-channel thermal-hydraulics module is replaced by the COBRA3C-TE with cross flow modelling and extended T-H flow conditions capabilities. The qualification of the RELAP5-PANTHER coupling demonstrated the robustness achieved by the combined 3-D neutron kinetics/system T-H code package for transient simulations. The coupled TE code package has been approved by the Belgian Safety Authorities and is used at TE for analyzing asymmetric PWR accidents with strong core-system interactions. In particular, the TE coupled code package was first used to develop a main steam line break in hot shutdown conditions (SLBHZP) accident analysis methodology based on the TE deterministic bounding approach. This methodology has been reviewed and accepted by the Belgian Safety Authorities for specific applications. Those specific applications are related to the power up-rate and steam generator replacement project of the Doel 2 plant or to the Tihange-3 SLB accident re-analysis. A coupled feedwater line break (FLB) accident analysis methodology is currently being reviewed for application approval. The results of coupled thermal-hydraulic and neutronic analysis of SLB and FLB show that there exist important margins in the traditional final safety analysis report (FSAR) accident analysis. Those margins can be used to increase the operational flexibility of the plants. Moreover, the

Safety analysis report: packages. DOT specification 7A - Type A container Mark 15 sludge shipping package

International Nuclear Information System (INIS)

Zeh, C.W.

1985-03-01

Sludge or filter cake containing 1.1 wt % 235 U enriched uranium from Mark 15 fabrication will be packaged in 55-gallon containers and shipped from SRP to NLO, Fernald, Ohio for recovery of product. About 7 Metric Tons (MT) of filter cake will be produced from fabricating Mark 15 slugs each reactor charge. Packaged shipments of this material, consigned as exclusive use, will be shipped to NLO in Department of Transportation (DOT) Specification 7A - Type A packages under a DOE Certificate of Compliance for Fissile Class III shipments. ''Type A packaging'' is designed to retain containment and shielding integrity under normal conditions of transport. This report documents compliance of the package design, construction methods, material and test performance with DOT Specification 7A. This DOT 7A Type A package contains a carbon steel outer container which is a 0.060-in.-thick 55-gal, galvanized drum equipped with a gasketed closure. The outer container encloses a 0.090-in.-thick, open head, polyethylene liner with lid

Safety Analysis Report for Packaging (SARP) of the Oak Ridge National Laboratory TRU Californium Shipping Container

International Nuclear Information System (INIS)

Box, W.D.; Shappert, L.B.; Seagren, R.D.; Klima, B.B.; Jurgensen, M.C.; Hammond, C.R.; Watson, C.D.

1980-01-01

An analytical evaluation of the Oak Ridge National Laboratory TRU Californium Shipping Container was made in order to demonstrate its compliance with the regulations governing off-site shipment of packages that contain radioactive material. The evaluation encompassed five primary categories: structural integrity, thermal resistance, radiation shielding, nuclear criticality safety, and quality assurance. The results of this evaluation demonstrate that the container complies with the applicable regulations

Safety evaluation for packaging 222-S laboratory cargo tank for onetime type B material shipment

International Nuclear Information System (INIS)

Nguyen, P.M.

1994-01-01

The purpose of this Safety Evaluation for Packaging (SEP) is to evaluate and document the safety of the onetime shipment of bulk radioactive liquids in the 222-S Laboratory cargo tank (222-S cargo tank). The 222-S cargo tank is a US Department of Transportation (DOT) MC-312 specification (DOT 1989) cargo tank, vehicle registration number HO-64-04275, approved for low specific activity (LSA) shipments in accordance with the DOT Title 49, Code of Federal Regulations (CFR). In accordance with the US Department of Energy, Richland Operations Office (RL) Order 5480.1A, Chapter III (RL 1988), an equivalent degree of safety shall be provided for onsite shipments as would be afforded by the DOT shipping regulations for a radioactive material package. This document demonstrates that this packaging system meets the onsite transportation safety criteria for a onetime shipment of Type B contents

Safety evaluation for packaging (onsite) plutonium recycle test reactor graphite cask

Energy Technology Data Exchange (ETDEWEB)

Romano, T.

1997-09-29

This safety evaluation for packaging (SEP) provides the evaluation necessary to demonstrate that the Plutonium Recycle Test Reactor (PRTR) Graphite Cask meets the requirements of WHC-CM-2-14, Hazardous Material Packaging and Shipping, for transfer of Type B, fissile, non-highway route controlled quantities of radioactive material within the 300 Area of the Hanford Site. The scope of this SEP includes risk, shieldling, criticality, and.tiedown analyses to demonstrate that onsite transportation safety requirements are satisfied. This SEP also establishes operational and maintenance guidelines to ensure that transport of the PRTR Graphite Cask is performed safely in accordance with WHC-CM-2-14. This SEP is valid until October 1, 1999. After this date, an update or upgrade to this document is required.

Safety evaluation for packaging (onsite) plutonium recycle test reactor graphite cask

International Nuclear Information System (INIS)

Romano, T.

1997-01-01

This safety evaluation for packaging (SEP) provides the evaluation necessary to demonstrate that the Plutonium Recycle Test Reactor (PRTR) Graphite Cask meets the requirements of WHC-CM-2-14, Hazardous Material Packaging and Shipping, for transfer of Type B, fissile, non-highway route controlled quantities of radioactive material within the 300 Area of the Hanford Site. The scope of this SEP includes risk, shieldling, criticality, and.tiedown analyses to demonstrate that onsite transportation safety requirements are satisfied. This SEP also establishes operational and maintenance guidelines to ensure that transport of the PRTR Graphite Cask is performed safely in accordance with WHC-CM-2-14. This SEP is valid until October 1, 1999. After this date, an update or upgrade to this document is required

Safety analysis report for packaging: the ORNL DOT Specification 20WC-5 - special form packaging

International Nuclear Information System (INIS)

Schaich, R.W.

1982-10-01

The ORNL DOT Specification 20WC-5 - Special Form Package was fabricated for the transport of large quantities of solid nonfissile radioactive materials in special form. The package was evaluated on the basis of tests performed at Sandia National Laboratories, Albuquerque, New Mexico on an identical fire and impact shield and special form tests performed on a variety of stainless steel capsules at ORNL by Operations Division personnel. The results of these evaluations demonstrate that the package is in compliance with the applicable regulations for the transport of large quantities of nonfissile radioactive materials in special form

Safety evaluation for packaging for 1720-DR sodium-filled tank

International Nuclear Information System (INIS)

Mercado, M.S.

1996-01-01

Preparations are under way to sell the sodium stored in the 1720-DR tank in the 1720-DR building. This will require that the tank, as well as the 1720-DR facility, be moved to the 300 Area, so that the sodium may be melted and transferred into a railroad tanker car. Because the sodium is a hazardous material and is being shipped in a nonspecification packaging, a safety evaluation for packaging (SEP) is required. This SEP approves the sodium-filled tank for a single shipment from the 105-DR area to the 300 Area

Materials of Criticality Safety Concern in Waste Packages

International Nuclear Information System (INIS)

Larson, S.L.; Day, B.A.

2006-01-01

10 CFR 71.55 requires in part that the fissile material package remain subcritical when considering 'the most reactive credible configuration consistent with the chemical and physical form of the material'. As waste drums and packages may contain unlimited types of materials, determination of the appropriately bounding moderator and reflector materials to ensure compliance with 71.55 requires a comprehensive analysis. Such an analysis was performed to determine the materials or elements that produce the most reactive configuration with regards to both moderation and reflection of a Pu-239 system. The study was originally performed for the TRUPACT-II shipping package and thus the historical fissile mass limit for the package, 325 g Pu-239, was used [1]. Reactivity calculations were performed with the SCALE package to numerically assess the moderation or reflection merits of the materials [2]. Additional details and results are given in SAIC-1322-001 [3]. The development of payload controls utilizing process knowledge to determine the classification of special moderator and/or reflector materials and the associated fissile mass limit is also addressed. (authors)

Commercial Spent Nuclear Fuel Waste Package Misload Analysis

International Nuclear Information System (INIS)

J.K. Knudson

2003-01-01

The purpose of this calculation is to estimate the probability of misloading a commercial spent nuclear fuel waste package with a fuel assembly(s) that has a reactivity (i.e., enrichment and/or burnup) outside the waste package design. The waste package designs are based on the expected commercial spent nuclear fuel assemblies and previous analyses (Macheret, P. 2001, Section 4.1 and Table 1). For this calculation, a misloaded waste package is defined as a waste package that has a fuel assembly(s) loaded into it with an enrichment and/or burnup outside the waste package design. An example of this type of misload is a fuel assembly designated for the 21-PWR Control Rod waste package being incorrectly loaded into a 21-PWR Absorber Plate waste package. This constitutes a misloaded 21-PWR Absorber Plate waste package, because the reactivity (i.e., enrichment and/or burnup) of a 21-PWR Control Rod waste package fuel assembly is outside the design of a 21-PWR Absorber Plate waste package. These types of misloads (i.e., fuel assembly with enrichment and/or burnup outside waste package design) are the only types that are evaluated in this calculation. This calculation utilizes information from ''Frequency of SNF Misload for Uncanistered Fuel Waste Package'' (CRWMS M and O 1998) as the starting point. The scope of this calculation is limited to the information available. The information is based on the whole population of fuel assemblies and the whole population of waste packages, because there is no information about the arrival of the waste stream at this time. The scope of this calculation deviates from that specified in ''Technical Work Plan for: Risk and Criticality Department'' (BSC 2002a, Section 2.1.30) in that only waste package misload is evaluated. The remaining issues identified (i.e., flooding and geometry reconfiguration) will be addressed elsewhere. The intended use of the calculation is to provide information and inputs to the Preclosure Safety Analysis

Structural and Shielding Safety of a Transport Package for Radioisotope Sealed Source Assembly

Energy Technology Data Exchange (ETDEWEB)

Seo, Kiseog; Cho, Ilje; Kim, Donghak [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

2006-07-01

As some kinds of radioisotope (RI) sealed source are produced by HANARO research reactor, a demand of RI transport package is increasing gradually. Foreign countries, which produce the various RIs, have the intrinsic model of the RI transport package. It is necessary to develop a RI and its transport package simultaneously. It is difficult to design a shielding part for this transport package because the passage for this source assembly should be provided from the center of shielding part to the outside of the package. In order to endure the accident conditions such as a 9 m drop and puncture, this transport package consists of the guide tubes, a gamma shield and a shock absorber. This paper describe that a shielding and structural safety of RI sealed source transport package are evaluated under the accident conditions.

Structural and Shielding Safety of a Transport Package for Radioisotope Sealed Source Assembly

International Nuclear Information System (INIS)

Seo, Kiseog; Cho, Ilje; Kim, Donghak

2006-01-01

As some kinds of radioisotope (RI) sealed source are produced by HANARO research reactor, a demand of RI transport package is increasing gradually. Foreign countries, which produce the various RIs, have the intrinsic model of the RI transport package. It is necessary to develop a RI and its transport package simultaneously. It is difficult to design a shielding part for this transport package because the passage for this source assembly should be provided from the center of shielding part to the outside of the package. In order to endure the accident conditions such as a 9 m drop and puncture, this transport package consists of the guide tubes, a gamma shield and a shock absorber. This paper describe that a shielding and structural safety of RI sealed source transport package are evaluated under the accident conditions

Classification of transportation packaging and dry spent fuel storage system components according to importance to safety

International Nuclear Information System (INIS)

McConnell, J.W., Jr; Ayers, A.L. Jr; Tyacke, M.J.

1996-02-01

This report provides a graded approach for classification of components used in transportation packaging and dry spent fuel storage systems. This approach provides a method for identifying, the classification of components according to importance to safety within transportation packagings and dry spent fuel storage systems. Record retention requirements are discussed to identify the documentation necessary to validate that the individual components were fabricated in accordance with their assigned classification. A review of the existing regulations pertaining to transportation packagings and dry storage systems was performed to identify current requirements The general types of transportation packagings and dry storage systems were identified. Discussions were held with suppliers and fabricators of packagings and storage systems to determine current practices. The methodology used in this report is based on Regulatory Guide 7.10, Establishing Quality Assurance Programs for Packaging Used in the Transport of Radioactive Material. This report also includes a list of generic components for each of the general types of transportation packagings and spent fuel storage systems. The safety importance of each component is discussed, and a classification category is assigned

Nano-food packaging: an overview of market, migration research, and safety regulations.

Science.gov (United States)

Bumbudsanpharoke, Nattinee; Ko, Seonghyuk

2015-05-01

Recently, food packages produced with nanoparticles, "nano-food packaging," have become more available in the current market. However, although the use of nanomaterials is increasing in food packaging applications, concern over toxicity affects consumer perceptions and acceptance. Quite a number of commercialized forms of nano-food packaging are coated or composited product with inorganic materials, for example, nanosilver and nanoclay as representative examples. Several studies have shown the possibility of nanomaterial migration from packaging or containers to foodstuff. The debate is still ongoing among researchers about the extent of migration and whether it is negligible and safe. Government agencies and stakeholders must hurry to determine use limitations and release conclusive legislation and regulations as soon as possible since nano-food packaging may have great impacts on human health. This paper aims to review the availability of nano-food packaging in the current market, report case studies on nanomaterial migration, and present the current status of safety regulations and management of nano-food packaging in leading countries across regions. This review should enable governments and researchers to develop further nanomaterial risk assessment studies. © 2015 Institute of Food Technologists®

Civilian use transport of radioactive substances on public road. Volume 1: Shipment accreditation and approval requests. Guide Nr 7, Revision 2 of 15 February 2016. Volume 2: safety file of package models, European guide 'Package Design Safety Report'. Civilian use transport of radioactive packages or substances on public road. Volume 3: Compliance of package models not subject to accreditation. Guide Nr 7, Revision 7 of the 2015/11/13

International Nuclear Information System (INIS)

2016-01-01

After having recalled the regulatory context and sanctions susceptible to be applied, the first volume presents the accreditation process for a package model: file content, tests programme, safety file, certification studies, documents to be produced, accreditation prorogation request, accreditation extension or package model modifications, instruction delays. Some peculiar cases are described. Models of accreditation certificate are provided, and obligations concerning packaging design, fabrication, use and maintenance are briefly discussed. The second volume is a European technical guide which is intended to assist in the preparation of the Package Design Safety Report (PDSR) to demonstrate compliance of a package design for the transport of radioactive material with the regulatory requirements. It covers package designs requiring competent authority approval, and also covers package designs not requiring competent authority approval. In its first two chapters, this document provides a generic structure and contents of a PDSR which applies to all package types. The contents are described in a comprehensive way to cover all important aspects. Some of these aspects may not be applicable to specific package type and details can be found in the annexes which provide further guidance for the scope of the contents of a PDSR, specifically for each package type. The third volume presents recommendations made by the ASN for all stakeholders to guarantee the compliance to regulation of package models which are 'not submitted to competent authority approval'. After an indication and a comment of the regulatory context, it presents requirements to be applied for the design of those package models, and then describes and comments the structure and content of a safety file for such package models (generalities, authorised contents, packaging description, safety demonstration, receipt, use and maintenance instruction, management system). The last part presents the

Normal conditions of transport thermal analysis and testing of a Type B drum package

International Nuclear Information System (INIS)

Jerrell, J.W.; Alstine, M.N. van; Gromada, R.J.

1995-01-01

Increasing the content limits of radioactive material packagings can save money and increase transportation safety by decreasing the total number of shipments required to transport large quantities of material. The contents of drum packages can be limited by unacceptable containment vessel pressures and temperatures due to the thermal properties of the insulation. The purpose of this work is to understand and predict the effects of insulation properties on containment system performance. The type B shipping container used in the study is a double containment fiberboard drum package. The package is primarily used to transport uranium and plutonium metals and oxides. A normal condition of transport (NCT) thermal test was performed to benchmark an NCT analysis of the package. A 21 W heater was placed in an instrumented package to simulate the maximum source decay heat. The package reached thermal equilibrium 120 hours after the heater was turned on. Testing took place indoors to minimize ambient temperature fluctuations. The thermal analysis of the package used fiberboard properties reported in the literature and resulted in temperature significantly greater than those measured during the test. Details of the NCT test will be described and transient temperatures at key thermocouple locations within the package will be presented. Analytical results using nominal fiberboard properties will be presented. Explanations of the results and the attempt to benchmark the analysis will be presented. The discovery that fiberboard has an anisotropic thermal conductivity and its effect on thermal performance will also be discussed

Safety evaluation for packaging (onsite) depleted uranium waste boxes

Energy Technology Data Exchange (ETDEWEB)

McCormick, W.A.

1997-08-27

This safety evaluation for packaging (SEP) allows the one-time shipment of ten metal boxes and one wooden box containing depleted uranium material from the Fast Flux Test Facility to the burial grounds in the 200 West Area for disposal. This SEP provides the analyses and operational controls necessary to demonstrate that the shipment will be safe for the onsite worker and the public.

Safety evaluation for packaging (onsite) depleted uranium waste boxes

International Nuclear Information System (INIS)

McCormick, W.A.

1997-01-01

This safety evaluation for packaging (SEP) allows the one-time shipment of ten metal boxes and one wooden box containing depleted uranium material from the Fast Flux Test Facility to the burial grounds in the 200 West Area for disposal. This SEP provides the analyses and operational controls necessary to demonstrate that the shipment will be safe for the onsite worker and the public

Safety analysis methodology for OPR 1000

International Nuclear Information System (INIS)

Hwang-Yong, Jun

2005-01-01

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

Safety Analysis Report for Packaging (SARP) of the Oak Ridge National Laboratory Garden Carrier No. 2. Revision 1

International Nuclear Information System (INIS)

Box, W.D.; Klima, B.B.; Seagren, R.D.; Shappert, L.B.; Watson, C.D.; Aramayo, G.A.

1979-08-01

An analytical evaluation of the Oak Ridge National Laboratory Garden Carrier No. 2 was made to demonstrate its compliance with the regulations governing off-site radioactive material shipping packages. The evaluation encompassed five primary categories: structural integrity, thermal resistance, radiation shielding, nuclear criticality safety, and quality assurance. The results of the evaluation show that the cask complies with the applicable regulations

Safety Analysis Report for Packaging (SARP) of the Oak Ridge National Laboratory Shipping Cask D-38. Revision 1

International Nuclear Information System (INIS)

Box, W.D.; Shappert, L.B.; Seagren, R.D.; Watson, C.D.; Hammond, C.R.; Klima, B.B.

1979-09-01

An analytical evaluation of the Oak Ridge National Laboratory Shipping Cask D-38 (solids shipments) was made to demonstrate its compliance with the regulations governing off-site radioactive material shipping packages. The evaluation encompassed five primary categories: structural integrity, thermal resistance, radiation shielding, nuclear criticality safety, and quality assurance. The results of the evaluation show that the cask complies with the applicable regulations

Safety Analysis Report for Packaging (SARP) of the Oak Ridge National Laboratory Garden Carrier No. 2. Revision 1

Energy Technology Data Exchange (ETDEWEB)

Box, W.D.; Klima, B.B.; Seagren, R.D.; Shappert, L.B.; Watson, C.D.; Aramayo, G.A.

1979-08-01

An analytical evaluation of the Oak Ridge National Laboratory Garden Carrier No. 2 was made to demonstrate its compliance with the regulations governing off-site radioactive material shipping packages. The evaluation encompassed five primary categories: structural integrity, thermal resistance, radiation shielding, nuclear criticality safety, and quality assurance. The results of the evaluation show that the cask complies with the applicable regulations.

Safety Analysis Report for Packaging (SARP) of the Oak Ridge National Laboratory Shipping Cask D-38. Revision 1

Energy Technology Data Exchange (ETDEWEB)

Box, W.D.; Shappert, L.B.; Seagren, R.D.; Watson, C.D.; Hammond, C.R.; Klima, B.B.

1979-09-01

An analytical evaluation of the Oak Ridge National Laboratory Shipping Cask D-38 (solids shipments) was made to demonstrate its compliance with the regulations governing off-site radioactive material shipping packages. The evaluation encompassed five primary categories: structural integrity, thermal resistance, radiation shielding, nuclear criticality safety, and quality assurance. The results of the evaluation show that the cask complies with the applicable regulations.

TopView - ATLAS top physics analysis package

CERN Document Server

Shibata, A

2007-01-01

TopView is a common analysis package which is widely used in the ATLAS top physics working group. The package is fully based on the official ATLAS software Athena and EventView and playing a central role in the collaborative analysis model. It is a functional package which accounts for a broad range issues in implementing physics analysis. As well as being a modular framework suitable as a common workplace for collaborators, TopView implements numerous analysis tools including a complete top-antitop reconstruction and single top reconstruction. The package is currently used to produce common ntuple from Monte Carlo production and future use cases are under rapid development. In this paper, the design and ideas behind TopView and the performance of the analyses implemented in the package are presented with detailed documentation of the contents and instruction for using the package.

ATMX-600 rail car safety analysis report for packaging (SARP)

International Nuclear Information System (INIS)

Adcock, F.E.; McCarthy, J.D.

1977-01-01

The ATMX-600 series rail car is used by Rockwell International, Rocky Flats Plant, for shipping low-level radioactive waste under the provisions of DOT Special Permit 5948. Fissile Class I shipments are authorized with the car loaded to capacity with drums containing up to 200 g of 239 Pu. Inner packaging may be polyethylene-lined steel drums or fiberglass-coated plywood crates. These massive double-walled steel cars provide the equivalent protection of a Type B package. Rapid loading and unloading of the 9- by 9- by 50-ft cargo space is accomplished by prepackaging the waste in standard 20-ft steel cargo containers. The ATMX-600 rail car will hold two cargo containers, each carrying seventy 55-gal drums and up to 44,800 lb gross weight. Improvements to inner packaging and current shipping practices are discussed

Development and experimental qualification of the new safety-criticality CRISTAL package

International Nuclear Information System (INIS)

Mattera, Ch.

1998-11-01

This thesis is concerned with Criticality-Safety studies related to the French Nuclear Fuel Cycle. We first describe the steps in the nuclear fuel cycle and the specific characteristics of these studies compared with those performed in Reactor Physics. In order to respond to the future requirements of the French Nuclear Program, we have developed a new package CRISTAL based on a recent cross sections library (CEA 93) and the newest accurate codes (APOLLO 2, MORET 4, TRIPOLI 4). The CRISTAL system includes two calculations routes: a design route which will be used by French Industry (COGEMA/SGN) and a reference route. To transfer this package to the French industry, we have elaborated calculation schemes for fissile solutions, dissolver media, transport casks and storage pools. Afterwards, these schemes have been used for the CRISTAL experimental validation. We have also contributed to the CRISTAL experimental database by reevaluating a French storage pool experiment: the CRISTO II experiment. This revaluation has been submitted to the OECD working group in order that this experiment can be used by international criticality safety engineers to validate calculations methods. This work represents a large contribution to the recommendation of accurate calculation schemes and to the experimental validation of the CRISTAL package. These studies came up to the French Industry expectations. (author)

Development and experimental testing of the new safety-criticality Cristal package

International Nuclear Information System (INIS)

Mattera, Ch.

1998-01-01

This thesis is concerned with Criticality-Safety studies related to the French Nuclear Fuel Cycle. We first describe the steps in the nuclear fuel cycle and the specific characteristics of these studies compared with those performed in Reactor Physics. In order to respond to the future requirements of the French Nuclear Program, we have developed a new package CRISTAL based on a recent cross sections library (CEA93) and the newest accurate codes (APOLLO2, MORET4, TRIPOLI4). The cristal system includes two calculations routes: a design route which will be used by French Industry (COGEMA/SGN) and a reference route.) To transfer this package to the French industry, we have elaborated calculation schemes for fissile solutions, dissolver media, transport casks and storage pools. Afterwards, these schemes have been used for the CRISTAL experimental validation. We have also contributed to the CRISTAL experimental database by reevaluating a French storage pool experiment: the CRISTO II experiment. This revaluation has been submitted to the OCDE working group in order that this experiment can be used by international criticality safety engineers to validate calculations methods. This work represents a large contribution to the recommendation of accurate calculation schemes and to the experimental validation of the CRISTAL package. These studies came up to the French Industry expectations. (author)

Safety demonstration analyses for severe accident of fresh nuclear fuel transport packages at JAERI

International Nuclear Information System (INIS)

Yamada, K.; Watanabe, K.; Nomura, Y.; Okuno, H.; Miyoshi, Y.

2004-01-01

It is expected in the near future that more and more fresh nuclear fuel will be transported in a variety of transport packages to cope with increasing demand from nuclear fuel cycle facilities. Accordingly, safety demonstration analyses of these methods are planned and conducted at JAERI under contract with the Ministry of Economy, Trade and Industry of Japan. These analyses are conducted part of a four year plan from 2001 to 2004 to verify integrity of packaging against leakage of radioactive material in the case of a severe accident envisioned to occur during transportation, for the purpose of gaining public acceptance of such nuclear fuel activities. In order to create the accident scenarios, actual transportation routes were surveyed, accident or incident records were tracked, international radioactive material transport regulations such as IAEA rules were investigated and, thus, accident conditions leading to mechanical damage and thermal failure were selected for inclusion in the scenario. As a result, the worst-case conditions of run-off-the-road accidents were incorporated, where there is impact against a concrete or asphalt surface. Fire accidents were assumed to occur after collision with a tank truck carrying lots of inflammable material or destruction by fire after collision inside a tunnel. The impact analyses were performed by using three-dimensional elements according to the general purpose impact analysis code LS-DYNA. Leak-tightness of the package was maintained even in the severe impact accident scenario. In addition, the thermal analyses were performed by using two-dimensional elements according to the general purpose finite element method computer code ABAQUS. As a result of these analyses, the integrity of the inside packaging component was found to be sufficient to maintain a leak-tight state, confirming its safety

SeDA: A software package for the statistical analysis of the instrument drift

International Nuclear Information System (INIS)

Lee, H. J.; Jang, S. C.; Lim, T. J.

2006-01-01

The setpoints for safety-related equipment are affected by many sources of an uncertainty. ANSI/ISA-S67.04.01-2000 [1] and ISA-RP6 7.04.02-2000 [2] suggested the statistical approaches for ensuring that the safety-related instrument setpoints were established and maintained within the technical specification limits [3]. However, Jang et al. [4] indicated that the preceding methodologies for a setpoint drift analysis might be insufficient to manage a setpoint drift on an instrumentation device and proposed new statistical analysis procedures for the management of a setpoint drift, based on the plant specific as-found/as-left data. Although IHPA (Instrument History Performance Analysis) is a widely known commercial software package to analyze an instrument setpoint drift, several steps in the new procedure cannot be performed by using it because it is based on the statistical approaches suggested in the ANSI/ISA-S67.04.01 -2000 [1] and ISA-RP67.04.02-2000 [2], In this paper we present a software package (SeDA: Setpoint Drift Analysis) that implements new methodologies, and which is easy to use, as it is accompanied by powerful graphical tools. (authors)

ATMX-600 rail car safety analysis report for packaging (SARP)

International Nuclear Information System (INIS)

Adcock, F.E.; McCarthy, J.D.

1977-01-01

The ATMX-600 series rail car is used by Rockwell International, Rocky Flats Plant, for shipping low-level radioactive waste under the provisions of DOT Special Permit 5948. Fissile Class I shipments are authorized with the car loaded to capacity with drums containing up to 200 grams of plutonium-239. Inner packaging may be polyethylene-lined steel drums or fiberglass-coated plywood crates. These massive double-walled steel cars provide the equivalent protection of a Type B package. Rapid loading and unloading of the 9- by 9- by 50-foot cargo space are accomplished by prepackaging the waste in standard 20-foot steel cargo containers. The ATMX-600 rail car will hold two cargo containers, each carrying seventy 55-gallon drums and up to 44,800 pounds gross weight. This report is a revision of an earlier document and describes improvements to inner packaging. It also reflects current shipping practices

Bacterial quality and safety of packaged fresh leafy vegetables at the retail level in Finland.

Science.gov (United States)

Nousiainen, L-L; Joutsen, S; Lunden, J; Hänninen, M-L; Fredriksson-Ahomaa, M

2016-09-02

Consumption of packaged fresh leafy vegetables, which are convenient ready-to-eat products, has increased during the last decade. The number of foodborne outbreaks associated with these products has concurrently increased. In our study, (1) label information, (2) O2/CO2 composition, (3) bacterial quality and (4) safety of 100 fresh leafy vegetables at the retail level were studied in Finland during 2013. Bacterial quality was studied using aerobic bacteria (AB) and coliform bacteria (CB) counts, and searching for the presence of Escherichia coli, Listeria and Yersinia. The safety was studied by the presence of Salmonella, ail-positive Yersinia, stx-positive E. coli (STEC) and Listeria monocytogenes using PCR and culturing. Important label information was unavailable on several packages originating from different companies. The packaging date was missing on all packages and the date of durability on 83% of the packages. Storage temperature was declared on 62% of the packages and 73% of the packages contained information about prewashing. The batch/lot number was missing on 29% of the packages. Very low oxygen (O2) (vegetable samples varying between 6.2 and 10.6 and 4.2-8.3logcfu/g, respectively. In most of the samples, the AB and CB counts exceeded 10(8) and 10(6)cfu/g, respectively. A positive correlation was observed between the AB and CB counts. E. coli was isolated from 15% of the samples and Yersinia from 33%. L. monocytogenes was isolated from two samples and ail-positive Y. enterocolitica in one. Using PCR, STEC was detected in seven samples, and Salmonella and ail-positive Y. enterocolitica in two samples each. The AB and CB mean values of products originating from different companies varied widely. High AB and CB counts and pathogenic bacteria were detected in ready-to-eat products not needing washing before use. Our study shows that the bacterial quality and safety of packaged fresh leafy vegetables is poor and label information on the packages is

Nuclear criticality safety assessment of ORR, NBS, and HFBR fuel element shipping package

International Nuclear Information System (INIS)

Thomas, J.T.

1979-01-01

A fuel element shipping package employing a borated-phenolic foam as a thermal insulating material is designed to transport as many as seven fuel elements for use in the Oak Ridge Research Reactor, the Brookhaven Fast Beam Reactor, or the National Bureau of Standards Reactor. This report presents the criticality safety evaluation and demonstrates that the requirements for a Fissile Class I package are satisfied by the design

Preliminary safety evaluation for the plutonium stabilization and packaging system

International Nuclear Information System (INIS)

Shapley, J.E.

1997-01-01

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

An evaluation of department of transportation specification packages

International Nuclear Information System (INIS)

Ratledge, J.E.; Rawl, R.R.

1993-01-01

Specification packages are broad families of package designs developed and authorized by the U.S. Department of Transportation (DOT) and the Nuclear Regulatory Commission (NRC) for transport of certain Type B and fissile radioactive materials, with each specification containing a number of designs of various sizes. The specification package designs have remained essentially unchanged in a changing regulatory environment. Changes to package designs or authorized contents under the DOT system can be accomplished by rule making action, but there has been little updating of the designs over the years. Many of the individual package designs are no longer supported by reasonably current safety analyses. Since the publication of these specifications, there have been changes in regulatory requirements and improvements in methods of testing and analysis. Additionally, contemplated revisions to the DOT and NRC regulations to bring design requirements into accord with IAEA Safety Series No. 6, 1985 Edition would eliminate fissile classes and require resistance to a crush test for small Type B packages meeting certain criteria. The NRC has requested that the Oak Ridge National Laboratory (ORNL) staff review the safety documentation of the specification packages to determine the possible need for further testing and analysis, modifications to the designs, and, perhaps, elimination of any designs for which there is insufficient demonstration of compliance with current and proposed requirements. This paper will present a summary of the technical data and information concerning the use of the packages that has been received to date. (author)

The LCDROOT Analysis Package

International Nuclear Information System (INIS)

Abe, Toshinori

2001-01-01

The North American Linear Collider Detector group has developed simulation and analysis program packages. LCDROOT is one of the packages, and is based on ROOT and the C++ programing language to maximally benefit from object oriented programming techniques. LCDROOT is constantly improved and now has a new topological vertex finder, ZVTOP3. In this proceeding, the features of the LCDROOT simulation are briefly described

New enhancements to SCALE for criticality safety analysis

International Nuclear Information System (INIS)

Hollenbach, D.F.; Bowman, S.M.; Petrie, L.M.; Parks, C.V.

1995-01-01

As the speed, available memory, and reliability of computer hardware increases and the cost decreases, the complexity and usability of computer software will increase, taking advantage of the new hardware capabilities. Computer programs today must be more flexible and user friendly than those of the past. Within available resources, the SCALE staff at Oak Ridge National Laboratory (ORNL) is committed to upgrading its computer codes to keep pace with the current level of technology. This paper examines recent additions and enhancements to the criticality safety analysis sections of the SCALE code package. These recent additions and enhancements made to SCALE can be divided into nine categories: (1) new analytical computer codes, (2) new cross-section libraries, (3) new criticality search sequences, (4) enhanced graphical capabilities, (5) additional KENO enhancements, (6) enhanced resonance processing capabilities, (7) enhanced material information processing capabilities, (8) portability of the SCALE code package, and (9) other minor enhancements, modifications, and corrections to SCALE. Each of these additions and enhancements to the criticality safety analysis capabilities of the SCALE code system are discussed below

Waste package reliability analysis

International Nuclear Information System (INIS)

Pescatore, C.; Sastre, C.

1983-01-01

Proof of future performance of a complex system such as a high-level nuclear waste package over a period of hundreds to thousands of years cannot be had in the ordinary sense of the word. The general method of probabilistic reliability analysis could provide an acceptable framework to identify, organize, and convey the information necessary to satisfy the criterion of reasonable assurance of waste package performance according to the regulatory requirements set forth in 10 CFR 60. General principles which may be used to evaluate the qualitative and quantitative reliability of a waste package design are indicated and illustrated with a sample calculation of a repository concept in basalt. 8 references, 1 table

Hanford Site radioactive hazardous materials packaging directory

International Nuclear Information System (INIS)

McCarthy, T.L.

1995-12-01

The Hanford Site Radioactive Hazardous Materials Packaging Directory (RHMPD) provides information concerning packagings owned or routinely leased by Westinghouse Hanford Company (WHC) for offsite shipments or onsite transfers of hazardous materials. Specific information is provided for selected packagings including the following: general description; approval documents/specifications (Certificates of Compliance and Safety Analysis Reports for Packaging); technical information (drawing numbers and dimensions); approved contents; areas of operation; and general information. Packaging Operations ampersand Development (PO ampersand D) maintains the RHMPD and may be contacted for additional information or assistance in obtaining referenced documentation or assistance concerning packaging selection, availability, and usage

Hanford Site radioactive hazardous materials packaging directory

Energy Technology Data Exchange (ETDEWEB)

McCarthy, T.L.

1995-12-01

The Hanford Site Radioactive Hazardous Materials Packaging Directory (RHMPD) provides information concerning packagings owned or routinely leased by Westinghouse Hanford Company (WHC) for offsite shipments or onsite transfers of hazardous materials. Specific information is provided for selected packagings including the following: general description; approval documents/specifications (Certificates of Compliance and Safety Analysis Reports for Packaging); technical information (drawing numbers and dimensions); approved contents; areas of operation; and general information. Packaging Operations & Development (PO&D) maintains the RHMPD and may be contacted for additional information or assistance in obtaining referenced documentation or assistance concerning packaging selection, availability, and usage.

Package leaflets of the most consumed medicines in Portugal: safety and regulatory compliance issues. A descriptive study

Directory of Open Access Journals (Sweden)

Carla Pires

Full Text Available CONTEXT AND OBJECTIVES: Package leaflets are necessary for safe use of medicines. The aims of the present study were: 1 to assess the compliance between the content of the package leaflets and the specifications of the pharmaceutical regulations; and 2 to identify potential safety issues for patients. DESIGN AND SETTING: Qualitative descriptive study, involving all the package leaflets of branded medicines from the three most consumed therapeutic groups in Portugal, analyzed in the Department of Pharmacoepidemiology, School of Pharmacy, University of Lisbon. METHODS: A checklist validated through an expert consensus process was used to gather the data. The content of each package leaflet in the sample was classified as compliant or non-compliant with compulsory regulatory issues (i.e. stated dosage and descriptions of adverse reactions and optional regulatory issues (i.e. adverse reaction frequency, symptoms and procedures in cases of overdose. RESULTS: A total of 651 package leaflets were identified. Overall, the package leaflets were found to be compliant with the compulsory regulatory issues. However, the optional regulatory issues were only addressed in around half of the sample of package leaflets, which made it possible to identify some situations of potentially compromised drug safety. CONCLUSION: Ideally, the methodologies for package leaflet approval should be reviewed and optimized as a way of ensuring the inclusion of the minimum essential information for safe use of medicines.

Y-12 defense programs. Nuclear Packaging Systems testing capabilities

International Nuclear Information System (INIS)

1995-06-01

The Nuclear Packaging Systems (NPS) Department can manage/accomplish any packaging task. The NPS organization is responsible for managing the design, testing, certification, procurement, operation, refurbishment, maintenance, and disposal of packaging used to transport radioactive materials, other hazardous materials, and general cargoes on public roads and within the Oak Ridge Y-12 Plant. Additionally, the NPS Department has developed a Quality Assurance plan for all packaging, design and procurement of nonweapon shipping containers for radioactive materials, and design and procurement of performance-oriented packaging for hazardous materials. Further, the NPS Department is responsible for preparation and submittal of Safety Analysis Reports for Packaging (SARP). The NPS Department coordinates shipping container procurement and safety certification activities that have lead-times of up to two years. A Packaging Testing Capabilities Table at the Oak Ridge complex is included as a table

Safety by design of printed multilayer materials intended for food packaging.

Science.gov (United States)

Domeño, Celia; Aznar, Margarita; Nerín, Cristina; Isella, Francesca; Fedeli, Mauro; Bosetti, Osvaldo

2017-07-01

Printing inks are commonly used in multilayer plastics materials used for food packaging, and compounds present in inks can migrate to the food either by diffusion through the multilayers or because of set-off phenomena. To avoid this problem, the right design of the packaging is crucial. This paper studies the safety by design of multilayer materials. First, the migration from four different multilayers manufactured using polyethylene terephthalate (PET), aluminium (Al) and polyethylene (PE) was determined. The structural differences among materials such as the presence of inks or lacquer coatings as well as the differences in layers position allowed the study of a safety-by-design approach. Sixty-nine different compounds were detected and identified; 49 of them were not included in the positive list of Regulation EU/10/2011 or in Swiss legislation and 15 belong to Cramer class III, which means that they have a theoretical high toxicity. Some of the compounds related to ink composition were pyrene, a compound commercially used to make dyes and dye precursors and the antioxidant Irganox 1300. The application of external lacquers decreased the concentration of some migrants but also brought the potential for new migrants coming from its composition. A final risk assessment of the material allowed evaluating food safety for different food simulants and confirm it.

Conceptual Assessment of a Fresh Fuel Transport Package for KJRR

Energy Technology Data Exchange (ETDEWEB)

Lee, Ju-Chan; Choi, W. S.; Bang, K. S.; Yu, S. H.; Park, J. S.; Yang, Y. Y. [KAERI, Daejeon (Korea, Republic of)

2016-05-15

The IAEA and domestic regulations stipulate that the fissile material transport package be subjected to the cumulative effects of a 9 m drop, 1 m puncture, 800 ℃ thermal and water leakage tests. A fissile material transport package should be maintained the subcriticality during the normal and accident conditions for contingency of leakage of water into or out of package, rearrangement of the contents, reduction of spaces and temperature changes. KAERI has been developing a fresh fuel transport package for Kijang research reactor (KJRR). This paper describes a conceptual design and preliminary safety analysis of the transport package for KJRR. The transport package was designed for shipment of a fresh fuel and a FM (Fission Molybdenum) target. Low-enriched uranium (LEU) of U-Mo fuel with U-235 enrichment of 19.75 w/o is used as a research reactor fuel. And LEU of UAlx-Al with U-235 enrichment of 19.75 w/o is used as a FM target material. The transport package was designed for shipment of a fresh fuel and a FM target. Safety analyses were conducted on all areas, including criticality, structural, and thermal fields. In the criticality analysis, effective neutron multiplication factors were below the criticality safety limit. In the structural analysis, the maximum stress satisfied the stress requirement stipulated in the ASME code. After 9 m free drop and 1 m puncture test, there was no significant deformation of fuel basket to cause a criticality. In the thermal analysis, the maximum temperatures at each part were lower than the allowable values.

Quality and safety aspects of reusable plastic food packaging materials : a European study to underpin future legislation

NARCIS (Netherlands)

Jetten, J.; Kruijf, N. de; Castle, L.

1999-01-01

The objective of this study was to develop a comprehensive package of quality assurance criteria for use by industry and regulatory authorities for ensuring the quality and safety-in-use (sensory, microbiological and chemical) of reused plastics for food packaging. The study included thermal

Radiological consequence analysis. Report on the work package 10. Preliminary safety analysis for the site Gorleben; Radiologische Konsequenzanalyse. Bericht zum Arbeitspaket 10. Vorlaeufige Sicherheitsanalyse fuer den Standort Gorleben

Energy Technology Data Exchange (ETDEWEB)

Larue, Juergen; Baltes, Bruno; Fischer, Heidemarie; Frieling, Gerd; Kock, Ingo; Navarro, Martin; Seher, Holger

2013-02-15

Work package 10 of the preliminary safety analysis for the site Gorleben concerns the analysis of release scenarios and the radiological consequences. The report includes the description of the applied methodology, the concept for the mathematical description of scenarios, the variety of scenarios for the radiological consequence analysis with the definition of the reference scenario. The database for the calculation covers geology and hydrology, the final repository concept and the process of the final disposal system. The used models were the one-phase transport model MARINE and the two-phase transport model TOUGH2. The report summarizes the results including an evaluation of the realization of the demonstration concept with respect to solution transport, fluid transport, robustness of the modeling results and analogy discussion to other disposal variants (i.e. borehole disposal and storage of transport or storage casks).

Packaging design criteria modified fuel spacer burial box. Revision 1

International Nuclear Information System (INIS)

Stevens, P.F.

1994-01-01

Various Hanford facilities must transfer large radioactively contaminated items to burial/storage. Presently, there are eighteen Fuel Spacer Burial Boxes (FSBBs) available on the Hanford Site for transport of such items. Previously, the FSBBS were transported from a rail car to the burial trench via a drag-off operation. To allow for the lifting of the boxes into the burial trench, it will be necessary to improve the packagings lifting attachments and provide structural reinforcement. Additional safety improvements to the packaging system will be provided by the addition of a positive closure system and package ventilation. FSBBs that are modified in such a manner are referred to as Modified Fuel Spacer Burial Boxes (MFSBs). The criteria provided by this PDC will be used to demonstrate that the transfer of the MFSB will provide an equivalent degree of safety as would be provided by a package meeting offsite transportation requirements. This fulfills the onsite transportation safety requirements implemented in WHC-CM-2-14, Hazardous Material Packaging and Shipping. A Safety Analysis Report for Packaging (SARP) will be prepared to evaluate the safety of the transfer operation. Approval of the SARP is required to authorize transfer. Criteria are also established to ensure burial requirements are met

Food safety concerns deriving from the use of silver based food packaging materials: a case study.

Directory of Open Access Journals (Sweden)

Alessandra ePezzuto

2015-10-01

Full Text Available The formulation of innovative packaging solutions, exerting a functional antimicrobial role in slowing down food spoilage, is expected to have a significant impact on the food industry, allowing both the maintenance of food safety criteria for longer periods and the reduction of food waste. Different materials are considered able to exert the required antimicrobial activity, among which are materials containing silver. However, challenges exist in the application of silver to food contact materials due to knowledge gaps in the production of ingredients, stability of delivery systems in food matrices and health risks caused by the same properties which also offer the benefits. Aims of the present study were to test the effectiveness and suitability of two packaging systems, one of which contained silver, for packaging and storing Stracchino cheese, a typical Italian fresh cheese, and to investigate if there was any potential for consumers to be exposed to silver, via migration from the packaging to the cheese. Results did not show any significant difference in the effectiveness of the packaging systems on packaged Stracchino cheese, excluding that the active packaging systems exerted an inhibitory effect on the growth of spoilage microorganisms. Moreover, silver migrated into the cheese matrix throughout the storage time (24 days. Silver levels in cheese finally exceeded the maximum established level for the migration of a non-authorised substance through a functional barrier (Commission Regulation (EC No. 450/2009. This result poses safety concerns and strongly suggests the need for more research aimed at better characterizing the new packaging materials in terms of their potential impacts on human health and the environment.

DOE Safety Metrics Indicator Program (SMIP) Fiscal Year 2000 Annual Report of Packaging- and Transportation-related Occurrences

International Nuclear Information System (INIS)

Dickerson, L.S.

2001-01-01

The U.S. Department of Energy (DOE) Occurrence Reporting and Processing System (ORPS) is an interactive computer system designed to support DOE-owned or -operated facilities in reporting and processing information concerning occurrences related to facility operations. The Oak Ridge National Laboratory has been charged by the DOE National Transportation Program Albuquerque (NTPA) with the responsibility of retrieving reports and information pertaining to packaging and transportation (P and T) incidents from the centralized ORPS database. These selected reports are analyzed for safety concerns, trends, potential impact on P and T operations, and ''lessons learned'' in P and T safety. To support this analysis and trending, the Safety Metrics Indicator Program (SMIP) was established by the NTPA in fiscal year (FY) 1998. Its chief goal is to augment historical reporting of occurrence-based information by providing (1) management notification of those incidents that require attention, (2) an accurate picture of contractors' P and T-related performance, and (3) meaningful statistics on occurrences at particular sites, including comparisons among different contractor sites and between DOE and the private sector. This annual report contains information on those P and T-related occurrences reported to the ORPS during the period from October 1, 1999, through September 30, 2000. Only those incidents that occur in preparation for transport, during transport, and during unloading of hazardous material are considered as packaging- or transportation-related occurrences

mediation: R Package for Causal Mediation Analysis

Directory of Open Access Journals (Sweden)

Dustin Tingley

2014-09-01

Full Text Available In this paper, we describe the R package mediation for conducting causal mediation analysis in applied empirical research. In many scientific disciplines, the goal of researchers is not only estimating causal effects of a treatment but also understanding the process in which the treatment causally affects the outcome. Causal mediation analysis is frequently used to assess potential causal mechanisms. The mediation package implements a comprehensive suite of statistical tools for conducting such an analysis. The package is organized into two distinct approaches. Using the model-based approach, researchers can estimate causal mediation effects and conduct sensitivity analysis under the standard research design. Furthermore, the design-based approach provides several analysis tools that are applicable under different experimental designs. This approach requires weaker assumptions than the model-based approach. We also implement a statistical method for dealing with multiple (causally dependent mediators, which are often encountered in practice. Finally, the package also offers a methodology for assessing causal mediation in the presence of treatment noncompliance, a common problem in randomized trials.

Safety analysis of the post-operational phase

International Nuclear Information System (INIS)

Berg, H.P.; Ehrlich, D.

1991-01-01

The safety analysis of normal operation covers an analytical study of the system parts ultimate repository - waste forms of the ultimate repository system under normal and accidental operation. On that basis a requirement concept has been developed which entails reactions on planning and design of the repository, and requirements of waste products, packagings and permissible activities. The procedure for the operational phase is explained giving the Konrad repository project as an example. (DG) [de

Legal analysis of information displayed on dental material packages: An exploratory research

Directory of Open Access Journals (Sweden)

Bhumika Rathore

2016-01-01

Full Text Available Introduction: Some of the dental materials possess occupational hazards, preprocedural errors, and patient allergies as suggested by evidence. With due consideration to safety of the patients and dental professionals, it is essential that the trade of these materials is in conformity with the law. Aim: To perform the legal analysis of the information displayed on the packaging of dental materials. Materials and Methods: The Bureau of Indian Standards sets guidelines for packaging and marketing of dental products in India. An exploratory cross-sectional study was performed using various search engines and websites to access the laws and regulations existing pertaining to dental materials packaging. Based on the data obtained, a unique packaging standardization checklist was developed. Dental laboratory and impression plasters, alginates, and endodontic instruments were surveyed for all the available brands. This study considered 16 brands of plasters and alginates and 42 brands of endodontic instruments for legal analysis. Legal analysis was performed using the direct observation checklist. Descriptive statistics were obtained using SPSS version 19. Results: The guidelines set by the Bureau of Indian Standards do exist but are not updated and stand as oblivious guards for marketing standards. Overall compliance to the guidelines was reported to be 18.5% by brands of alginates, 4.1% by plaster of Paris, and 11.11% by endodontic instruments. Wave One™ File reported maximum adherence with the guidelines as 66.7%. Conclusion: This study found lower rate of adherence to the guidelines, thus indicating insufficient information being disclosed to the consumers.

Interactions between sanitizers and packaging gas compositions and their effects on the safety and quality of fresh-cut onions (Allium cepa L.).

Science.gov (United States)

Page, Natalie; González-Buesa, Jaime; Ryser, Elliot T; Harte, Janice; Almenar, Eva

2016-02-02

Onions are one of the most widely utilized vegetables worldwide, with demand for fresh-cut onions steadily increasing. Due to heightened safety concerns and consumer demand, the implications of sanitizing and packaging on fresh-cut onion safety and quality need to be better understood. The objective of this study was to investigate the effect of produce sanitizers, in-package atmospheres, and their interactions on the growth of Salmonella Typhimurium, mesophilic aerobic bacteria, yeast and mold, and the physico-chemical quality of diced onions to determine the best sanitizer and in-package atmosphere combination for both safety and quality. Diced onions were inoculated or not with S. Typhimurium, sanitized in sodium hypochlorite, peroxyacetic acid, or liquid chlorine dioxide, and then packaged in either polylactic acid bags containing superatmospheric O2, elevated CO2/reduced O2, or air, or in polyethylene terephthalate snap-fit containers. Throughout 14 days of storage at 7 °C, packaged diced onions were assessed for their safety (S. Typhimurium), and quality (mesophilic aerobic bacteria, yeasts and molds, physico-chemical analyses, and descriptive and consumer acceptance sensory panels). While sanitizer affected (Ppackage atmosphere had a significant (Pinteractions between sanitizer and atmosphere that affected S. Typhimurium and pH were identified whereas 3-way interactions (sanitizer, atmosphere and time) were only observed for headspace CO2. Sodium hypochlorite and elevated CO2/reduced O2 was the best sanitizer and in-package atmosphere combination for enhancing the safety and quality of packaged diced onions. In addition, this combination led to diced onions acceptable for purchase after 2 weeks of storage by trained and consumer panels. Copyright © 2015 Elsevier B.V. All rights reserved.

Kinematics Simulation Analysis of Packaging Robot with Joint Clearance

Science.gov (United States)

Zhang, Y. W.; Meng, W. J.; Wang, L. Q.; Cui, G. H.

2018-03-01

Considering the influence of joint clearance on the motion error, repeated positioning accuracy and overall position of the machine, this paper presents simulation analysis of a packaging robot — 2 degrees of freedom(DOF) planar parallel robot based on the characteristics of high precision and fast speed of packaging equipment. The motion constraint equation of the mechanism is established, and the analysis and simulation of the motion error are carried out in the case of turning the revolute clearance. The simulation results show that the size of the joint clearance will affect the movement accuracy and packaging efficiency of the packaging robot. The analysis provides a reference point of view for the packaging equipment design and selection criteria and has a great significance on the packaging industry automation.

How to avoid errors in the design and fabrication of transportation packages

International Nuclear Information System (INIS)

Raske, D.T.

1996-01-01

The purpose of this paper is to discuss the errors and omissions most often identified when reviewing the design and fabrication of a packaging to transport high-level radioactive materials. The design and fabrication criteria recommended by the U.S. Department of Energy, Office of Facility Safety Analysis, for containment vessels of Type B commercial packagings containing high-level radioactive materials is based on the requirements of Section III, Division 1, Subsection NB of the ASME Boiler and Pressure Vessel Code. However, most packaging designers, engineers, and fabricators are intimidated by the sheer volume of requirements contained in the Code; as a result, the Code is not always followed and many requirements that do apply are often overlooked during preparation of the Safety Analysis Report for Packaging that constitutes the basis for evaluating the packaging for certification

Safety evaluation for packaging transportation of equipment for tank 241-C-106 waste sluicing system

International Nuclear Information System (INIS)

Calmus, D.B.

1994-01-01

A Waste Sluicing System (WSS) is scheduled for installation in nd waste storage tank 241-C-106 (106-C). The WSS will transfer high rating sludge from single shell tank 106-C to double shell waste tank 241-AY-102 (102-AY). Prior to installation of the WSS, a heel pump and a transfer pump will be removed from tank 106-C and an agitator pump will be removed from tank 102-AY. Special flexible receivers will be used to contain the pumps during removal from the tanks. After equipment removal, the flexible receivers will be placed in separate containers (packagings). The packaging and contents (packages) will be transferred from the Tank Farms to the Central Waste Complex (CWC) for interim storage and then to T Plant for evaluation and processing for final disposition. Two sizes of packagings will be provided for transferring the equipment from the Tank Farms to the interim storage facility. The packagings will be designated as the WSSP-1 and WSSP-2 packagings throughout the remainder of this Safety Evaluation for Packaging (SEP). The WSSP-1 packagings will transport the heel and transfer pumps from 106-C and the WSSP-2 packaging will transport the agitator pump from 102-AY. The WSSP-1 and WSSP-2 packagings are similar except for the length

Safety analysis report: packages. GPHS shipping package supplement 2 to the PISA shipping package (packaging of fissile and other radioactive materials). Final report

International Nuclear Information System (INIS)

Chalfant, G. G.

1981-06-01

Safety Analysis Report DPST-78-124-1 is amended to permit shipment of 6 General Purpose Heat Source (GPHS) capsules (max.). Each capsule contains an average of 2330 curies of 238 Pu, and each pair of capsules is contained in a welded stainless steel primary containment vessel, all of which are doubly contained in a flanged secondary containment vessel. This is in addition to the forms discussed in DPST-78-124-1 and Supplement 1

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

Energy Technology Data Exchange (ETDEWEB)

NONE

2013-08-15

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

Packaging- and transportation-related occurrence reports, October-December 1994

International Nuclear Information System (INIS)

Welch, M.J.; Dickerson, L.S.; Armstrong, C.J.

1995-02-01

The Oak Ridge National Laboratory (ORNL) Packaging and Transportation Safety Program (PATS), which is sponsored by the U.S. Department of Energy (DOE) Office of Environment, Safety and Health, Office of Facility Safety Analysis, EH-32, has been charged with the responsibility of retrieving reports and information pertaining to transportation or packaging incidents from the Occurrence Reporting and Processing System (ORPS). These selected reports are being analyzed for trends, impact on EH-32 policies and concerns, and lessons learned concerning transportation and packaging safety. This task is designed not only to keep EH-32 aware of current packaging and transportation incidents and potential transportation and packaging problems that may need attention on DOE sites but also to allow future dissemination of lessons learned to the Operations Offices and, subsequently, to management and operating contractors. This report, which covers the period from October 2 to December 31, 1994, covers the weekly tabular reports OR-94-40 through OR-94-52. These 12 reports, which contained a total of 75 occurrence reports (ORs) relating to packaging and transportation issues, were submitted to EH-32 for its information and use during this quarter. The 75 ORs that were selected from the hundreds reviewed are listed. The second column of Table I contains the PATS nature of occurrence (NOC) coding for the respective OR, and the third column lists the weekly report issue in which the OR was originally transmitted to DOE-Headquarters (HQ). The Lesson Learned bulletins produced this quarter are included. These two bulletins have been distributed to a large packaging and transportation safety audience and are included as a natural outgrowth of the quarterly reports

Test for radioactive material transport package safety

International Nuclear Information System (INIS)

Li Guoqiang; Zhao Bing; Zhang Jiangang; Wang Xuexin; Ma Anping

2012-01-01

Regulations on radioactive material transport in China were introduced. Test facilities and data acquiring instruments for radioactive material package in China Institute for Radiation Protection were also introduced in this paper, which were used in drop test and thermal test. Test facilities were constructed according to the requirements of IAEA's 'Regulations for the Safe Transport of Radioactive Material' (TS-R-l) and Chinese 'Regulations for the Safe Transport of Radioactive Material' (GB 11806-2004). Drop test facilities were used in free drop test, penetration test, mechanical test (free drop test Ⅰ, free drop test Ⅱ and free drop test Ⅲ) of type A and type B packages weighing less than thirteen tons. Thermal test of type B packages can be carried out in the thermal test facilities. Certification tests of type FCo70-YQ package, type 30A-HB-01 package, type SY-I package and type XAYT-I package according to regulations were done using these facilities. (authors)

Lessons learned during Type A Packaging testing

International Nuclear Information System (INIS)

O'Brien, J.H.; Kelly, D.L.

1995-11-01

For the past 6 years, the US Department of Energy (DOE) Office of Facility Safety Analysis (EH-32) has contracted Westinghouse Hanford Company (WHC) to conduct compliance testing on DOE Type A packagings. The packagings are tested for compliance with the U.S. Department of Transportation (DOT) Specification 7A, general packaging, Type A requirements. The DOE has shared the Type A packaging information throughout the nuclear materials transportation community. During testing, there have been recurring areas of packaging design that resulted in testing delays and/or initial failure. The lessons learned during the testing are considered a valuable resource. DOE requested that WHC share this resource. By sharing what is and can be encountered during packaging testing, individuals will hopefully avoid past mistakes

ITER interim design report package and relevant documents

International Nuclear Information System (INIS)

1996-01-01

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

Safety Evaluation for Packaging for onsite Transfer of plutonium recycle test reactor ion exchange columns

International Nuclear Information System (INIS)

Smith, R.J.

1995-01-01

The purpose of this Safety Evaluation for Packaging (SEP) is to authorize the use of three U.S. Department of Transportation (DOT) 7A, Type A metal boxes (Capital Industries Part No. S 0600-0600-1080- 0104) to package 12 Plutonium Recycle Test Reactor (PRTR) ion exchange columns as low-level waste (LLW). The packages will be transferred from the 309 Building in the 300 Area to low level waste burial in the 200 West Area. Revision 1 of WHC-SD-TP-SEP-035 (per ECN No. 621467) documents that the boxes containing ion exchange columns and grout will maintain the payload under normal conditions of transport if transferred without the box lids

Development of fresh fuel packaging for ATR demonstration reactor

International Nuclear Information System (INIS)

Kurakami, J.; Kurita, I.

1993-01-01

Related to development of the demonstration advanced thermal reactor, it is necessary and important to develop transport packaging which is used for transporting fresh fuel assemblies. Therefore, the packaging is now being developed in Power Reactor and Nuclear Fuel Development Corporation (PNC). Currently, PNC is fabricating two prototype packagings based on the final design, and land cruising and vibration tests, handling performance tests and prototype packaging tests will be executed with prototype packagings in order to experimentally confirm the soundness of packaging and its contents and the propriety of design technique. This paper describes the summary of general specifications and structures of this packaging and the summary of preliminary safety analysis of package. (J.P.N.)

Experimental study on transportation safety of package in side collision of heavy duty truck

International Nuclear Information System (INIS)

Suga, M.; Sasaki, T.

1989-01-01

The accidents in road transportation of package may be collision, fall and fire. It is necessary to examine all cases very carefully because collision might be caused by other vehicle. Collisions are classified into head-on collision, rear-end collision, side collision. A lot of experiments and analyses are reported on head-on collision, so the behavior of vehicle and package may be predicted without difficulty. Rear-end collisions bring about less impact and may be applied corresponding to the head-on collisions. About side collisions, few experiments or analyses are reported, and most of them are about passenger cars not about trucks. So it becomes important to study the transportation safety of package carried on a heavy duty truck when hit on the side by another truck similar in size

Safety Evaluation for Packaging for the N Reactor/single pass reactor fuel characterization shipments

International Nuclear Information System (INIS)

Stevens, P.F.

1994-01-01

The purpose of this Safety Evaluation for Packaging (SEP) is to authorize the ChemNuclear CNS 1-13G packaging to ship samples of irradiated fuel elements from the 100 K East and 100 K West basins to the Postirradiation Testing Laboratory (PTL) in support of the spent nuclear fuel characterization effort. It also authorizes the return of the fuel element samples to the 100 K East facility using the same packaging. The CNS 1-13G cask has been-chosen to transport the fuel because it has a Certificate of Compliance (CoC) issued by the US Nuclear Regulatory Commission (NRC) for transporting irradiated oxide and metal fuel in commerce. It is capable of being loaded and offloaded underwater and may be shipped with water in the payload compartment

Radioisotope thermoelectric generator transportation system safety analysis report for packaging. Volumes 1 and 2

International Nuclear Information System (INIS)

Ferrell, P.C.

1996-01-01

This SARP describes the RTG Transportation System Package, a Type B(U) packaging system that is used to transport an RTG or similar payload. The payload, which is included in this SARP, is a generic, enveloping payload that specifically encompasses the General Purpose Heat Source (GPHS) RTG payload. The package consists of two independent containment systems mounted on a shock isolation transport skid and transported within an exclusive-use trailer

Radioisotope thermoelectric generator transportation system safety analysis report for packaging. Volumes 1 and 2

Energy Technology Data Exchange (ETDEWEB)

Ferrell, P.C.

1996-04-18

This SARP describes the RTG Transportation System Package, a Type B(U) packaging system that is used to transport an RTG or similar payload. The payload, which is included in this SARP, is a generic, enveloping payload that specifically encompasses the General Purpose Heat Source (GPHS) RTG payload. The package consists of two independent containment systems mounted on a shock isolation transport skid and transported within an exclusive-use trailer.

DESIGN ANALYSIS FOR THE NAVAL SNF WASTE PACKAGE

International Nuclear Information System (INIS)

T.L. Mitchell

2000-01-01

The purpose of this analysis is to demonstrate the design of the naval spent nuclear fuel (SNF) waste package (WP) using the Waste Package Department's (WPD) design methodologies and processes described in the ''Waste Package Design Methodology Report'' (CRWMS MandO [Civilian Radioactive Waste Management System Management and Operating Contractor] 2000b). The calculations that support the design of the naval SNF WP will be discussed; however, only a sub-set of such analyses will be presented and shall be limited to those identified in the ''Waste Package Design Sensitivity Report'' (CRWMS MandO 2000c). The objective of this analysis is to describe the naval SNF WP design method and to show that the design of the naval SNF WP complies with the ''Naval Spent Nuclear Fuel Disposal Container System Description Document'' (CRWMS MandO 1999a) and Interface Control Document (ICD) criteria for Site Recommendation. Additional criteria for the design of the naval SNF WP have been outlined in Section 6.2 of the ''Waste Package Design Sensitivity Report'' (CRWMS MandO 2000c). The scope of this analysis is restricted to the design of the naval long WP containing one naval long SNF canister. This WP is representative of the WPs that will contain both naval short SNF and naval long SNF canisters. The following items are included in the scope of this analysis: (1) Providing a general description of the applicable design criteria; (2) Describing the design methodology to be used; (3) Presenting the design of the naval SNF waste package; and (4) Showing compliance with all applicable design criteria. The intended use of this analysis is to support Site Recommendation reports and assist in the development of WPD drawings. Activities described in this analysis were conducted in accordance with the technical product development plan (TPDP) ''Design Analysis for the Naval SNF Waste Package (CRWMS MandO 2000a)

Development of platform to compare different wall heat transfer packages for system analysis codes

International Nuclear Information System (INIS)

Kim, Min-Gil; Lee, Won Woong; Lee, Jeong Ik; Shin, Sung Gil

2016-01-01

System thermal hydraulic (STH) analysis code is used for analyzing and evaluating the safety of a designed nuclear system. The system thermal hydraulic analysis code typically solves mass, momentum and energy conservation equations for multiple phases with sets of selected empirical constitutive equations to close the problem. Several STH codes are utilized in academia, industry and regulators, such as MARS-KS, SPACE, RELAP5, COBRA-TF, TRACE, and so on. Each system thermal hydraulic code consists of different sets of governing equations and correlations. However, the packages and sets of correlations of each code are not compared quantitatively yet. Wall heat transfer mode transition maps of SPACE and MARS-KS have a little difference for the transition from wall nucleate heat transfer mode to wall film heat transfer mode. Both codes have the same heat transfer packages and correlations in most region except for wall film heat transfer mode. Most of heat transfer coefficients calculated for the range of selected variables of SPACE are the same with those of MARS-KS. For the intervals between 500K and 540K of wall temperature, MARS-KS selects the wall film heat transfer mode and Bromley correlation but SPACE select the wall nucleate heat transfer mode and Chen correlation. This is because the transition from nucleate boiling to film boiling of MARS-KS is earlier than SPACE. More detailed analysis of the heat transfer package and flow regime package will be followed in the near future

Safety evaluation for packaging transport of LSA-II liquids in MC-312 cargo tanks

Energy Technology Data Exchange (ETDEWEB)

Carlstrom, R.F.

1996-09-11

This safety evaluation for packaging authorizes the onsite transfer of bulk LSA-II radioactive liquids in the 222-S Laboratory Cargo Tank and Liquid Effluent Treatment Facility Cargo Tanks (which are U.S. Department of Transportation MC-312 specification cargo tanks) from their operating facilities to tank farm facilities.

Thermal analysis of NNWSI conceptual waste package designs

International Nuclear Information System (INIS)

Stein, W.; Hockman, J.N.; O'Neal, W.C.

1984-04-01

Lawrence Livermore National Laboratory is involved in the design and testing of high-level nuclear waste packages. Many of the aspects of waste package design and testing (e.g., corrosion and leaching) depend in part on the temperature history of the emplaced packages. This report discusses thermal modeling and analysis of various emplaced waste package conceptual designs including the models used, the assumptions and approximations made, and the results obtained. 16 references

Drop testing of the Westinghouse fresh nuclear fuel package

International Nuclear Information System (INIS)

Shappert, L.B.; Sanders, C.F.

1993-01-01

The Westinghouse Columbia Fuel Fabrication Facility has decided to develop and certify a new fresh fuel package design (type A, fissile) that has the capability to transport more highly enriched fuel than was previously possible. A prototype package was tested in support of the Safety Analysis Report of the Packaging (SARP). This paper provides detailed information on the tests and test results. A first prototype test was carried out at the STF, and the design did not give the safety margin that Westinghouse wanted for their containers. The data from the test were used to redesign the connection between the clamping frame and the pressure pad, and the tests were reinitiated. Three packages were then tested at the STF. All packages met the acceptance criteria and acceleration information was obtained that provided an indication of the behavior of the cradle and strongback which holds the fuel assemblies and nuclear poison in place. (J.P.N.)

The Innovative Approaches to Packaging – Comparison Analysis of Intelligent and Active Packaging Perceptions in Slovakia

Directory of Open Access Journals (Sweden)

Loucanova Erika

2017-06-01

Full Text Available Packaging has always served a practical function - to hold goods together and protect it when moving toward the customer through distribution channel. Today packaging is also a container for promoting the product and making it easier and safer to use. The sheer importance of packaging functions is still growing and consequently the interest of the company is to access to the packaging more innovative and creative. The paper deals with the innovative approaches to packaging resulting in the creation of packaging with interactive active features in the form of active and intelligent packaging. Using comparative analysis, we monitored the perception of the active packaging functions in comparison to intelligent packaging function among different age categories. We identified the age categories which are most interested in these functions.

Fragman: an R package for fragment analysis.

Science.gov (United States)

Covarrubias-Pazaran, Giovanny; Diaz-Garcia, Luis; Schlautman, Brandon; Salazar, Walter; Zalapa, Juan

2016-04-21

Determination of microsatellite lengths or other DNA fragment types is an important initial component of many genetic studies such as mutation detection, linkage and quantitative trait loci (QTL) mapping, genetic diversity, pedigree analysis, and detection of heterozygosity. A handful of commercial and freely available software programs exist for fragment analysis; however, most of them are platform dependent and lack high-throughput applicability. We present the R package Fragman to serve as a freely available and platform independent resource for automatic scoring of DNA fragment lengths diversity panels and biparental populations. The program analyzes DNA fragment lengths generated in Applied Biosystems® (ABI) either manually or automatically by providing panels or bins. The package contains additional tools for converting the allele calls to GenAlEx, JoinMap® and OneMap software formats mainly used for genetic diversity and generating linkage maps in plant and animal populations. Easy plotting functions and multiplexing friendly capabilities are some of the strengths of this R package. Fragment analysis using a unique set of cranberry (Vaccinium macrocarpon) genotypes based on microsatellite markers is used to highlight the capabilities of Fragman. Fragman is a valuable new tool for genetic analysis. The package produces equivalent results to other popular software for fragment analysis while possessing unique advantages and the possibility of automation for high-throughput experiments by exploiting the power of R.

mediation: R package for causal mediation analysis

OpenAIRE

Tingley, Dustin; Yamamoto, Teppei; Hirose, Kentaro; Keele, Luke; Imai, Kosuke

2012-01-01

In this paper, we describe the R package mediation for conducting causal mediation analysis in applied empirical research. In many scientific disciplines, the goal of researchers is not only estimating causal effects of a treatment but also understanding the process in which the treatment causally affects the outcome. Causal mediation analysis is frequently used to assess potential causal mechanisms. The mediation package implements a comprehensive suite of statistical tools for conducting su...

Former food products safety: microbiological quality and computer vision evaluation of packaging remnants contamination.

Science.gov (United States)

Tretola, M; Di Rosa, A R; Tirloni, E; Ottoboni, M; Giromini, C; Leone, F; Bernardi, C E M; Dell'Orto, V; Chiofalo, V; Pinotti, L

2017-08-01

The use of alternative feed ingredients in farm animal's diets can be an interesting choice from several standpoints, including safety. In this respect, this study investigated the safety features of selected former food products (FFPs) intended for animal nutrition produced in the framework of the IZS PLV 06/14 RC project by an FFP processing plant. Six FFP samples, both mash and pelleted, were analysed for the enumeration of total viable count (TVC) (ISO 4833), Enterobacteriaceae (ISO 21528-1), Escherichia coli (ISO 16649-1), coagulase-positive Staphylococci (CPS) (ISO 6888), presumptive Bacillus cereus and its spores (ISO 7932), sulphite-reducing Clostridia (ISO 7937), yeasts and moulds (ISO 21527-1), and the presence in 25 g of Salmonella spp. (ISO 6579). On the same samples, the presence of undesired ingredients, which can be identified as remnants of packaging materials, was evaluated by two different methods: stereomicroscopy according to published methods; and stereomicroscopy coupled with a computer vision system (IRIS Visual Analyzer VA400). All FFPs analysed were safe from a microbiological point of view. TVC was limited and Salmonella was always absent. When remnants of packaging materials were considered, the contamination level was below 0.08% (w/w). Of note, packaging remnants were found mainly from the 1-mm sieve mesh fractions. Finally, the innovative computer vision system demonstrated the possibility of rapid detection for the presence of packaging remnants in FFPs when combined with a stereomicroscope. In conclusion, the FFPs analysed in the present study can be considered safe, even though some improvements in FFP processing in the feeding plant can be useful in further reducing their microbial loads and impurity.

Safety evaluation for packaging for the transport of K Basin sludge samples in the PAS-1 cask

International Nuclear Information System (INIS)

SMITH, R.J.

1998-01-01

This safety evaluation for packaging authorizes the shipment of up to two 4-L sludge samples to and from the 325 Lab or 222-S Lab for characterization. The safety of this shipment is based on the current U.S. Department of Energy Certification of Compliance (CoC) for the PAS-1 cask, USA/9184/B(U) (DOE)

Safety evaluation for packaging for the transport of K Basin sludge samples in the PAS-1 cask

Energy Technology Data Exchange (ETDEWEB)

SMITH, R.J.

1998-11-17

This safety evaluation for packaging authorizes the shipment of up to two 4-L sludge samples to and from the 325 Lab or 222-S Lab for characterization. The safety of this shipment is based on the current U.S. Department of Energy Certification of Compliance (CoC) for the PAS-1 cask, USA/9184/B(U) (DOE).

Safety analysis report for packaging: the ORNL lithium hydroxide fire and impact shield

International Nuclear Information System (INIS)

Evans, J.H.; Eversole, R.E.; Just, R.A.; Schaich, R.W.

1984-07-01

The ORNL Lithium Hydroxide Fire and Impact Shield and its packaging were designed and fabricated at Oak Ridge National Laboratory to permit the transport of Type B quantities of radioactive material and limited quantities of fissionable material. The shield and its packaging were evaluated analytically and experimentally to determine its compliance with the applicable regulations governing containers in which radioactive and fissile materials are transported, and that evaluation is the subject of this report. Computational and test procedures were used to determine the structural integrity and thermal behavior of the shield relative to the general standards for normal conditions of transport and the standards for the hypothetical accident conditions. The results of the evaluation demonstrate that the shield and its packaging are in compliance with the applicable regulations. 16 references, 8 figures, 5 tables

Design and criticality considerations for 9977 and 9978 shipping packages

International Nuclear Information System (INIS)

Reed, R.; Biswas, D.; Abramczyk, G.

2009-01-01

Savannah River National Laboratory (SRNL) has developed two new, Type B, state-of-the-art, general purpose, fissile material Shipping Packages, designated as 9977 and 9978, as replacements for the U.S. Dept. of Transportation (DOT) specification 6M container. The packages accommodate plutonium, uranium, and other special nuclear materials in bulk quantities and in many forms with capabilities exceeding those of the 6M. A nuclear criticality safety evaluation demonstrates the safe configurations of the new shipping container for plutonium and uranium metal/oxide loading under various conditions for the Safety Analysis Report for Packaging (SARP). The evaluation is in compliance with the performance requirements of 10CFR 71, specifically 10CFR71.55 and 71.59. The criticality safety index (CSI) is 1.0 for most of the contents (Max. Pu Loading -4.4 kg). (authors)

Topical safety analysis report for the transportation of the NUHOMS reg-sign dry shielded canister

International Nuclear Information System (INIS)

1993-08-01

This Topical Safety Analysis Report (SAR) describes the design and the generic transportation licensing basis for utilizing the NUTECH HORIZONTAL MODULAR STORAGE (NUHOMS reg-sign) system dry shielded canister (DSC) containing twenty-four pressurized water reactor (PWR) spent fuel assemblies (SFA) in conjunction with a conceptually designed Transportation Cask. This SAR documents the design qualification of the NUHOMS reg-sign DSC as an integral part of a 10CFR71 Fissile Material Class III, Type B(M) Transportation Package. The package consists of the canister and a conceptual transportation cask (NUHOMS reg-sign Transportation Cask) with impact limiters. Engineering analysis is performed for the canister to confirm that the existing canister design complies with 10CFR71 transportation requirements. Evaluations and/or analyses is performed for criticality safety, shielding, structural, and thermal performance. Detailed engineering analysis for the transportation cask will be submitted in a future SAR requesting 10CFR71 certification of the complete waste package. Transportation operational considerations describe various operational aspects of the canister/transportation cask system. operational sequences are developed for canister transfer from storage to the transportation cask and interfaces with the cask auxiliary equipment for on- and off-site transport

Applications of nanotechnology in food packaging and food safety: barrier materials, antimicrobials and sensors.

Science.gov (United States)

Duncan, Timothy V

2011-11-01

In this article, several applications of nanomaterials in food packaging and food safety are reviewed, including: polymer/clay nanocomposites as high barrier packaging materials, silver nanoparticles as potent antimicrobial agents, and nanosensors and nanomaterial-based assays for the detection of food-relevant analytes (gasses, small organic molecules and food-borne pathogens). In addition to covering the technical aspects of these topics, the current commercial status and understanding of health implications of these technologies are also discussed. These applications were chosen because they do not involve direct addition of nanoparticles to consumed foods, and thus are more likely to be marketed to the public in the short term. Published by Elsevier Inc.

NUMERICAL ANALYSIS AND EXPERIMENTAL INVESTIGATION OF DISK SPRING CONFIGURATIONS WITH REGARD TO LOAD CAPACITY OF SAFETY PROGRESSIVE GEARS

Directory of Open Access Journals (Sweden)

Paweł LONKWIC

2016-09-01

Full Text Available The paper investigates the effect of various disk spring package configurations on brake load of safety progressive gears. The numerical analysis is performed using the Abaqus/CAE software and the designed 3D models. The numerical results are then verified in experimental tests. The tests also examine the effect of lubrication on brake load of spring packages. In addition, the paper investigates the work conditions of safety progressive gears at emergency braking. The experimental results show agreement with the numerical results.

Summary of the technical review of the safety analysis reports for packaging (SARP) for the transnuclear transport/storage casks: TN-BRP and TN-REG

International Nuclear Information System (INIS)

1986-07-01

The Safety Analysis Reports for Packaging for two spent fuel shipping casks were technically reviewed by the Oak Ridge National Laboratory. The casks were designed by Transnuclear, Inc., for shipment of 85 Big Rock Point boiling water reactor fuel elements and 40 R.E. Ginna pressurized water reactor fuel elements from West Valley, New York, to Idaho Falls, Idaho. The intent of the review was to ensure compliance of the casks with the requirements the applicable Federal Regulations contained in 10 CFR Pt. 71 and allow issuance of Department of Energy Certificates of Compliance for transport by the Department of Energy Idaho Operations Office. The review was performed by a team of Oak Ridge National Laboratory staff assembled for their expertise in criticality analysis, shielding, metallurgy, nondestructive testing, thermal analysis, structural analysis, and containment. This report describes the review processes, the findings in each technical area, and the overall conclusion that a Certificate of Compliance could be issued for the proposed single shipment under the specified conditions and constraints

Verification of the CONPAS (CONtainment Performance Analysis System) code package

International Nuclear Information System (INIS)

Kim, See Darl; Ahn, Kwang Il; Song, Yong Man; Choi, Young; Park, Soo Yong; Kim, Dong Ha; Jin, Young Ho.

1997-09-01

CONPAS is a computer code package to integrate the numerical, graphical, and results-oriented aspects of Level 2 probabilistic safety assessment (PSA) for nuclear power plants under a PC window environment automatically. For the integrated analysis of Level 2 PSA, the code utilizes four distinct, but closely related modules: (1) ET Editor, (2) Computer, (3) Text Editor, and (4) Mechanistic Code Plotter. Compared with other existing computer codes for Level 2 PSA, and CONPAS code provides several advanced features: computational aspects including systematic uncertainty analysis, importance analysis, sensitivity analysis and data interpretation, reporting aspects including tabling and graphic as well as user-friendly interface. The computational performance of CONPAS has been verified through a Level 2 PSA to a reference plant. The results of the CONPAS code was compared with an existing level 2 PSA code (NUCAP+) and the comparison proves that CONPAS is appropriate for Level 2 PSA. (author). 9 refs., 8 tabs., 14 figs

New nuclear package. At last a breakthrough for a European legal framework on nuclear safety?

International Nuclear Information System (INIS)

Schneider, Horst

2009-01-01

In 2003, the European Commission presented what it referred to as a nuclear package. Two draft directives were to cover nuclear safety and nuclear waste management in a legally binding sense on the level of the European Union. A separate directive on funds for decommissioning nuclear power plants and for waste management up to final storage, which had still been included in preliminary drafts in 2002, had been dropped and turned into recommendations in 2006. However, the nuclear package with the 2 draft directives found no sufficient majority in the Council in 2004. In November 2008, the Commission presented a new draft directive on nuclear safety, especially the safety of nuclear power plants. The Commission demands a European legal framework for the political acceptance of nuclear power. As far as procedures were concerned, the Commission had expressed its hope that the directive could be adopted by the summer of 2009. The draft directive has been thoroughly revised over the past four months. Shaping the European Union is a difficult matter. The improvement seems to be in the field of nuclear safety. It is to be hoped that a directive will be adopted in the end which will result in more acceptance, not just in arguments exchanged between the Commission and the member countries when it comes to transposition into national law and its execution. (orig.)

Road Safety Data, Collection, Transfer and Analysis DaCoTa. Deliverable 1.2: Road safety management investigation model and questionnaire.

NARCIS (Netherlands)

Dupont, H. Martensen, H. Papadimitriou, E. Yannis, G. Muhlrad, N. Jähi, H. Vallet, G. Giustiniani, G. Tripodi, A. Usami, D. Bax, C. Wijnen, W. Schöne, M.-L. Machata, K. Buttler, I. Zysinska, M. Talbot, R. Gitelman, V. & Hakkert, S. & Muhlrad, N. Gitelman, V. & Buttler, I. (Eds.)

2012-01-01

The aim of the DaCoTA Work Package 1 is to investigate road safety policy-making and management processes in Europe. In the Deliverables released previously, the Work Package 1 assessed the experts’ needs in terms of road safety knowledge, data and decision support tools (Deliverable 1.1/4.1), as

Flat-plate solar collector - installation package

Science.gov (United States)

1978-01-01

Package includes installation, operation and maintenance manual for collector, analysis of safety hazards, special handling instructions, materials list, installation drawings, and warranty and certification statement. Manual includes instructions for roof preparation and for preparing collector for installation. Several pages are devoted to major and minor repairs.

Safety-analysis report for packaging (SARP) general-purpose heat-source module 750-Watt shipping container

International Nuclear Information System (INIS)

Whitney, M.A.; Burgan, C.E.; Blauvelt, R.K.; Zocher, R.W.; Bronisz, S.E.

1981-01-01

The SARP includes discussions of structural integrity, thermal resistance, radiation shielding and radiological safety, nuclear criticality safety, and quality control. Extensive tests and evaluations were performed to show that the container will function effectively with respect to all required standards and when subjected to normal transportation conditions and the sequence of four hypothetical accident conditions (free drop, puncture, thermal, and water immersion). In addition, a steady state temperature profile and radiation profile were measured using two heat sources that very closely resemble the GPHS. This gave an excellent representation of the GPHS temperature and radiation profile. A nuclear criticality safety analysis determined that all safety requirements are met

The Evaluation of the Safety Benefits of Combined Passive and On-Board Active Safety Applications

Science.gov (United States)

Page, Yves; Cuny, Sophie; Zangmeister, Tobias; Kreiss, Jens-Peter; Hermitte, Thierry

2009-01-01

One of the objectives of the European TRACE project (TRaffic Accident Causation in Europe, 2006–2008) was to estimate the proportion of injury accidents that could be avoided and/or the proportion of injury accidents where the severity could be mitigated for on-the-market safety applications, if 100 % of the car fleet would be equipped with them. We have selected for evaluation the Electronic Stability Control (ESC) and the Emergency Brake Assist (EBA) applications. As for passive safety systems, recent cars are designed to offer overall safety protection. Car structure, load limiters, front airbags, side airbags, knee airbags, pretensioners, padding and non aggressive structures in the door panel, the dashboard, the windshield, the seats, and the head rest also contribute to applying more protection. The whole safety package is very difficult to evaluate separately, one element independently segmented from the others. We decided to consider evaluating the effectivenessof the whole passive safety package, This package,, for the sake of simplicity, was the number of stars awarded at the Euro NCAP testing. The challenges were to compare the effectiveness of some safety configuration SC I, with the effectiveness of a different safety configuration SC II. A safety configuration is understood as a package of safety functions. Ten comparisons have been carried out such as the evaluation of the safety benefit of a fifth star given that the car has four stars and an EBA. The main outcome of this analysis is that any addition of a passive or active safety function selected in this analysis is producing increased safety benefits. For example, if all cars were five stars fitted with EBA and ESC, instead of four stars without ESC and EBA, injury accidents would be reduced by 47.2% for severe injuries and 69.5% for fatal injuries. PMID:20184838

Incorporation of advanced accident analysis methodology into safety analysis reports

International Nuclear Information System (INIS)

2003-05-01

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

Packaging and transportation occurrence reporting

International Nuclear Information System (INIS)

Needels, T.S.

1996-01-01

The US Department of Energy (DOE) Order 231.1 calls for the maintenance of a database for all unclassified occurrence reports (ORs). ORS provide DOE with notice of incidents and accidents that endanger the public, workers, or DOE facility operations. To fulfill this policy, the DOE Occurrence Reporting and Processing System (ORPS) was established to require DOE facilities to report and process information concerning such events. The Oak Ridge National Laboratory (ORNL) provides DOE with data and analysis of occurrence related to packaging and transportation (P and T) safety. This program produces annual reports, lessons learned bulletins, and information for the packaging and transportation home page on the Internet. The analysis and reports provided can be used as a tool for oversight and a means for DOE sites to be proactive and anticipate problems through shared knowledge and lessons learned. To illustrate, some observable trends based on 3 years of the program are given. In summary, this program shows potential problem areas that need correcting, and possible breakdowns of safety

Safety evaluation for packaging (Onsite) transport of LSA-II liquids in MC-312 cargo tanks

International Nuclear Information System (INIS)

Carlstrom, R.F.

1996-01-01

This safety evaluation for packaging authorizes the onsite transfer of bulk LSA-II radioactive liquids in the 222-S Laboratory Cargo Tank and Liquid Effluent Treatment Facility Cargo Tanks (which are U.S. Department of Transportation MC-312 specification cargo tanks) from their operating facilities to tank farm facilities

Safety evaluation for packaging (onsite) for concrete-shielded RHTRU waste drum for the 327 postirradiation testing laboratory

Energy Technology Data Exchange (ETDEWEB)

Adkins, H.E.

1996-10-29

This safety evaluation for packaging authorizes onsite transport of Type B quantities of radioactive material in the Concrete- Shielded Remote-Handled Transuranic Waste (RH TRU) Drum per WHC-CM-2-14, Hazardous Material Packaging and Shipping. The drum will be used for transport of 327 Building legacy waste from the 300 Area to the Transuranic Waste Storage and Assay Facility in the 200 West Area and on to a Solid Waste Storage Facility, also in the 200 Area.

Safety balance: Analysis of safety systems

International Nuclear Information System (INIS)

Delage, M.; Giroux, C.

1990-12-01

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

WAPDEG Analysis of Waste Package and Drip shield Degradation

International Nuclear Information System (INIS)

K. Mon

2004-01-01

As directed by ''Technical Work Plan for: Regulatory Integration Modeling and Analysis of the Waste Form and Waste Package'' (BSC 2004 [DIRS 171583]), an analysis of the degradation of the engineered barrier system (EBS) drip shields and waste packages at the Yucca Mountain repository is developed. The purpose of this activity is to provide the TSPA with inputs and methodologies used to evaluate waste package and drip shield degradation as a function of exposure time under exposure conditions anticipated in the repository. This analysis provides information useful to satisfy ''Yucca Mountain Review Plan, Final Report'' (NRC 2003 [DIRS 163274]) requirements. Several features, events, and processes (FEPs) are also discussed (Section 6.2, Table 15). The previous revision of this report was prepared as a model report in accordance with AP-SIII.10Q, Models. Due to changes in the role of this report since the site recommendation, it no longer contains model development. This revision is prepared as a scientific analysis in accordance with AP-SIII.9Q, ''Scientific Analyses'' and uses models previously validated in (1) ''Stress Corrosion Cracking of the Drip Shield, the Waste Package Outer Barrier, and the Stainless Steel Structural Material'' (BSC 2004 [DIRS 169985]); (2) ''General Corrosion and Localized Corrosion of Waste Package Outer Barrier'' (BSC 2004 [DIRS 169984]); and (3) ''General Corrosion and Localized Corrosion of Drip Shield'' (BSC 2004 [DIRS 169845]). The integrated waste package degradation (IWPD) analysis presented in this report treats several implementation-related issues, such as defining the number and size of patches per waste package that undergo stress corrosion cracking; recasting the weld flaw analysis in a form as implemented in the Closure Weld Defects (CWD) software; and, general corrosion rate manipulations (e.g., change of scale in Section 6.3.4). The weld flaw portion of this report takes input from an engineering calculation (BSC 2004

RepoTREND. The program package for the integrated long term safety analysis of final repository systems. Version 4.5 (State March 2016); RepoTREND. Das Programmpaket zur integrierten Langzeitsicherheitsanalyse von Endlagersystemen. Version 4.5 (Stand Maerz 2016)

Energy Technology Data Exchange (ETDEWEB)

Reiche, Tatiana

2016-04-15

The long-term safety analysis is the analysis of final repository behavior after closure includes the spreading of pollutants into the biosphere (mobilization and release of pollutants into the near field, radionuclide migration through the geosphere, radiation exposure in the biosphere) and the radiological consequences. The report describes the program package RepoTREND, the respective modules (near field, GeoTREND, BioTREND and probabilistic analyses), sequencing and postprocessing and the quality management.

Program packages for dynamics systems analysis and design

International Nuclear Information System (INIS)

Athani, V.V.

1976-01-01

The development of computer program packages for dynamic system analysis and design are reported. The purpose of developing these program packages is to take the burden of writing computer programs off the mind of the system engineer and to enable him to concentrate on his main system analysis and design work. Towards this end, four standard computer program packages have been prepared : (1) TFANA - starting from system transfer function this program computes transient response, frequency response, root locus and stability by Routh Hurwitz criterion, (2) TFSYN - classical synthesis using algebraic method of Shipley, (3) MODANA - starting from state equations of the system this program computes solution of state equations, controllability, observability and stability, (4) OPTCON - This program obtains solutions of (i) linear regulator problem, (ii) servomechanism problems and (iii) problem of pole placement. The paper describes these program packages with the help of flowcharts and illustrates their use with the help of examples. (author)

Safety evaluation for packaging (onsite) concrete-lined waste packaging

Energy Technology Data Exchange (ETDEWEB)

Romano, T.

1997-09-25

The Pacific Northwest National Laboratory developed a package to ship Type A, non-transuranic, fissile excepted quantities of liquid or solid radioactive material and radioactive mixed waste to the Central Waste Complex for storage on the Hanford Site.

Benchmarking of the computer code and the thirty foot side drop analysis for the Shippingport (RPV/NST package)

International Nuclear Information System (INIS)

Bumpus, S.E.; Gerhard, M.A.; Hovingh, J.; Trummer, D.J.; Witte, M.C.

1989-01-01

This paper presents the benchmarking of a finite element computer code and the subsequent results from the code simulating the 30 foot side drop impact of the RPV/NST transport package from the decommissioned Shippingport Nuclear Power Station. The activated reactor pressure vessel (RPV), thermal shield, and other reactor external components were encased in concrete contained by the neutron shield tank (NST) and a lifting skirt. The Shippingport RPV/NST package, a Type B Category II package, weighs approximately 900 tons and has 17.5 ft diameter and 40.7 ft. length. For transport of the activated components from Shippingport to the burial site, the Safety Analysis Report for Packaging (SARP) demonstrated that the package can withstand the hypothetical accidents of DOE Order 5480.3 including 10 CFR 71. Mathematical simulations of these accidents can substitute for actual tests if the simulated results satisfy the acceptance criteria. Any such mathematical simulation, including the modeling of the materials, must be benchmarked to experiments that duplicate the loading conditions of the tests. Additional confidence in the simulations is justified if the test specimens are configured similar to the package

Safety margins in deterministic safety analysis

International Nuclear Information System (INIS)

Viktorov, A.

2011-01-01

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

Modelling approach to evaluate safety of LILW-SL disposal in slovenia considering different waste packaging options

International Nuclear Information System (INIS)

Perko, J.; Mallants, D.

2007-01-01

The long-term safety of radioactive waste repositories is usually demonstrated by means of a safety assessment which normally includes modelling of radionuclide release from a multi-barrier surface or deep repository to the geosphere and biosphere. The present quantitative evaluation performed emphasizes on contrasting disposal options under consideration in Slovenia and concerns siting, disposal concept (deep versus surface), and waste packaging. The assessment has identified a number of conditions that would lead to acceptable waste disposal solutions, while at the same time results also revealed options that would result in exceeding the radiological criteria. Results presented are the output of a collective effort of a Quintessa-led Consortium with SCK-CEN and Belgatom, in the framework of a recent PHARE project. The key objective of this work was to identify the preferred disposal concept and packaging option from a number of alternatives being considered by the Slovenian radioactive waste management agency (ARAO) for low and intermediate level short-lived waste (LILW-SL). The emphasis of the assessment was the consideration of several waste treatment and packaging options in an attempt to identify the minimum required containment characteristics which would result in safe disposal and the cost-benefit of additional safety measures. Waste streams for which alternative treatment and packaging solutions were developed and evaluated include decommissioning waste and NPP operational wastes containing drums with unconditioned ion exchange resins in overpacked tube type containers (TTCs). For the former the disposal options under consideration were either direct disposal of loose pieces grouted into a vault or use of high integrity containers. For the latter three options were foreseen. The first is overpacking of resin containing TTCs grouted into high integrity containers, the second option is complete treatment with hydration, neutralisation, and cementation of

Study on transport safety of refresh MOX fuel. Radiation dose from package hypothetically submerged into sea

International Nuclear Information System (INIS)

Tsumune, Daisuke; Suzuki; Hiroshi; Saegusa, Toshiari; Maruyama, Koki; Ito, Chihiro; Watabe, Naoto

1999-01-01

The sea transport of fresh MOX fuel from Europe to Japan is under planning. For the structure and equipment of transport ships for fresh MOX fuels, there is a special safety standard called the INF Code of IMO (International Maritime Organization). For transport of radioactive materials, there is a safety standard stipulated in Regulations for the Safe Transport of Radioactive Material issued by IAEA (International Atomic Energy Agency). Under those code and standard, fresh MOX fuel will be transported safely on the sea. However, a dose assessment has been made by assuming that a fresh MOX fuel package might be sunk into the sea by unexpected reasons. In the both cases for a package sunk at the coastal region and for that sunk at the ocean, the evaluated result of the dose equivalent by radiation exposure to the public are far below the dose equivalent limit of the ICRP recommendation (1 mSv/year). (author)

DOE-EM-45 Packaging Operations And Maintenance Course

International Nuclear Information System (INIS)

Watkins, R.; England, J.

2010-01-01

Savannah River National Laboratory - Savannah River Packaging Technology (SRNL-SRPT) delivered the inaugural offering of the Packaging Operations and Maintenance Course for DOE-EM-45's Packaging Certification Program (PCP) at the University of South Carolina Aiken on September 1 and 2, 2009. Twenty-nine students registered, attended, and completed this training. The DOE-EM-45 Packaging Certification Program (PCP) sponsored the presentation of a new training course, Packaging Maintenance and Operations, on September 1-2, 2009 at the University of South Carolina Aiken (USC-Aiken) campus in Aiken, SC. The premier offering of the course was developed and presented by the Savannah River National Laboratory, and attended by twenty-nine students across the DOE, NNSA and private industry. This training informed package users of the requirements associated with handling shipping containers at a facility (user) level and provided a basic overview of the requirements typically outlined in Safety Analysis Report for Packaging (SARP) Chapters 1, 7, and 8. The course taught packaging personnel about the regulatory nature of SARPs to help reduce associated and often costly packaging errors. Some of the topics covered were package contents, loading, unloading, storage, torque requirements, maintaining records, how to handle abnormal conditions, lessons learned, leakage testing (including demonstration), and replacement parts. The target audience for this course was facility operations personnel, facility maintenance personnel, and field quality assurance personnel who are directly involved in the handling of shipping containers. The training also aimed at writers of SARP Chapters 1, 7, and 8, package designers, and anyone else involved in radioactive material packaging and transportation safety. Student feedback and critiques of the training were very positive. SRNL will offer the course again at USC Aiken in September 2010.

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

International Nuclear Information System (INIS)

1996-01-01

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

Topical safety analysis report for the transportation of the NUHOMS{reg_sign} dry shielded canister. Volume 1

Energy Technology Data Exchange (ETDEWEB)

None

1993-08-01

This Topical Safety Analysis Report (SAR) describes the design and the generic transportation licensing basis for utilizing the NUTECH HORIZONTAL MODULAR STORAGE (NUHOMS{reg_sign}) system dry shielded canister (DSC) containing twenty-four pressurized water reactor (PWR) spent fuel assemblies (SFA) in conjunction with a conceptually designed Transportation Cask. This SAR documents the design qualification of the NUHOMS{reg_sign} DSC as an integral part of a 10CFR71 Fissile Material Class III, Type B(M) Transportation Package. The package consists of the canister and a conceptual transportation cask (NUHOMS{reg_sign} Transportation Cask) with impact limiters. Engineering analysis is performed for the canister to confirm that the existing canister design complies with 10CFR71 transportation requirements. Evaluations and/or analyses is performed for criticality safety, shielding, structural, and thermal performance. Detailed engineering analysis for the transportation cask will be submitted in a future SAR requesting 10CFR71 certification of the complete waste package. Transportation operational considerations describe various operational aspects of the canister/transportation cask system. operational sequences are developed for canister transfer from storage to the transportation cask and interfaces with the cask auxiliary equipment for on- and off-site transport.

Safety evaluation for packaging (onsite) for the Pacific Northwest National Laboratory HEPA filter box

International Nuclear Information System (INIS)

McCoy, J.C.

1998-01-01

This safety evaluation for packaging (SEP) evaluates and documents the safe onsite transport of eight high-efficiency particulate air (HEPA) filters in the Pacific Northwest National Laboratory HEPA Filter Box from the 300 Area of the Hanford Site to the Central Waste Complex and on to burial in the 200 West Area. Use of this SEP is authorized for 1 year from the date of release

Safety analysis fundamentals

International Nuclear Information System (INIS)

Wright, A.C.D.

2002-01-01

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

Novel food packaging systems with natural antimicrobial agents.

Science.gov (United States)

Irkin, Reyhan; Esmer, Ozlem Kizilirmak

2015-10-01

A new type of packaging that combines food packaging materials with antimicrobial substances to control microbial surface contamination of foods to enhance product microbial safety and to extend shelf-life is attracting interest in the packaging industry. Several antimicrobial compounds can be combined with different types of packaging materials. But in recent years, since consumer demand for natural food ingredients has increased because of safety and availability, these natural compounds are beginning to replace the chemical additives in foods and are perceived to be safer and claimed to alleviate safety concerns. Recent research studies are mainly focused on the application of natural antimicrobials in food packaging system. Biologically derived compounds like bacteriocins, phytochemicals, enzymes can be used in antimicrobial food packaging. The aim of this review is to give an overview of most important knowledge about application of natural antimicrobial packagings with model food systems and their antimicrobial effects on food products.

Nuclear criticality safety analysis of a spent fuel waste package in a tuff repository

International Nuclear Information System (INIS)

Weren, B.H.; Capo, M.A.; O'Neal, W.C.

1983-12-01

An assessment has been performed of the criticality potential associated with the disposal of spent fuel in a tuff geology above the water table. Eleven potential configurations were defined which cover a vast range of geometries and conditions from the nominal configuration at emplacement to a hypothetical configuration thousands of years after emplacement in which the structure is gone, the fuel pellets disintegrated and the borehole flooded. Of these eleven configurations, four have been evaluated at this time. The results of this evaluation indicate that even with very conservative assumptions (4.5 w/o fresh fuel), criticality is not a problem for the nominal configuration either dry or fully flooded. In the cases where the condition of the waste package is assumed to have severely deteriorated, over long times, calculations were performed with less conservative assumptions (depleted fuel). An assessment of these calculations indicates that criticality safety could be demonstrated if the depletion of the fissile inventory during fuel irradiation is taken into account. A detailed discussion of the calculations performed is presented in this report. Also included are a description of the configurations which were considered, the analytical methods and models used, and a discussion of additional related work which should be performed. 15 references, 11 figures, 8 tables

The integrated code system CASCADE-3D for advanced core design and safety analysis

International Nuclear Information System (INIS)

Neufert, A.; Van de Velde, A.

1999-01-01

The new program system CASCADE-3D (Core Analysis and Safety Codes for Advanced Design Evaluation) links some of Siemens advanced code packages for in-core fuel management and accident analysis: SAV95, PANBOX/COBRA and RELAP5. Consequently by using CASCADE-3D the potential of modern fuel assemblies and in-core fuel management strategies can be much better utilized because safety margins which had been reduced due to conservative methods are now predicted more accurately. By this innovative code system the customers can now take full advantage of the recent progress in fuel assembly design and in-core fuel management.(author)

Safety and Feasibility of a Ketamine Package to Support Emergency and Essential Surgery in Kenya when No Anesthetist is Available: An Analysis of 1216 Consecutive Operative Procedures.

Science.gov (United States)

Burke, Thomas F; Suarez, Sebastian; Sessler, Daniel I; Senay, Ayla; Yusufali, Taha; Masaki, Charles; Guha, Moytrayee; Rogo, Debora; Jani, Pankaj; Nelson, Brett D; Rogo, Khama

2017-12-01

Lack of access to emergency and essential surgery is widespread in low- and middle-income countries. Scarce anesthesia services contribute to this unmet need. The aim of this study was to evaluate the safety and feasibility of the Every Second Matters for Emergency and Essential Surgery-Ketamine (ESM-Ketamine) package for emergency and essential procedures when no anesthetist was available. From November 2013 to September 2017, the ESM-Ketamine package was used for patients requiring emergency or life-improving surgeries in fifteen selected facilities across Kenya when no anesthetist was available. A mixed-methods approach was used to assess safety and feasibility of the ESM-Ketamine package, including demand, acceptability, and practicality. The primary outcome was ketamine-related adverse events. Key-informant interviews captured perceptions of providers, hospital administrators, and surgeons/proceduralists. Non-anesthetist mid-level providers used ESM-Ketamine for 1216 surgical procedures across the fifteen study facilities. The median ketamine dose was 2.1 mg/kg. Brief (30 s) oxygen desaturations occurred in seven patients (0.6%). There were 157 (13%) reported cases of hallucinations and agitation which were treated with diazepam. All patients recovered uneventfully, and no ketamine-related deaths were reported. Twenty-seven key-informant interviews showed strong support for the program with four main themes: financial considerations, provision of services, staff impact, and scaling considerations. The ESM-Ketamine package appears safe and feasible and is capable of expanding access to emergency and essential surgeries in rural Kenya when no anesthetist is available.

Waste package performance analysis

International Nuclear Information System (INIS)

Lester, D.H.; Stula, R.T.; Kirstein, B.E.

1982-01-01

A performance assessment model for multiple barrier packages containing unreprocessed spent fuel has been applied to several package designs. The resulting preliminary assessments were intended for use in making decisions about package development programs. A computer model called BARIER estimates the package life and subsequent rate of release of selected nuclides. The model accounts for temperature, pressure (and resulting stresses), bulk and localized corrosion, and nuclide retardation by the backfill after water intrusion into the waste form. The assessment model assumes a post-closure, flooded, geologic repository. Calculations indicated that, within the bounds of model assumptions, packages could last for several hundred years. Intact backfills of appropriate design may be capable of nuclide release delay times on the order of 10 7 yr for uranium, plutonium, and americium. 8 references, 6 figures, 9 tables

Reactor physics computer code development for neutronic design, fuel-management, reactor operation and safety analysis of PHWRs

International Nuclear Information System (INIS)

Rastogi, B.P.

1989-01-01

This report discusses various reactor physics codes developed for neutronic design, fuel-management, reactor operation and safety analysis of PHWRs. These code packages have been utilized for nuclear design of 500 MWe and new 235 MWe PHWRs. (author)

Active food packaging technologies.

Science.gov (United States)

Ozdemir, Murat; Floros, John D

2004-01-01

Active packaging technologies offer new opportunities for the food industry, in the preservation of foods. Important active packaging systems currently known to date, including oxygen scavengers, carbon dioxide emitters/absorbers, moisture absorbers, ethylene absorbers, ethanol emitters, flavor releasing/absorbing systems, time-temperature indicators, and antimicrobial containing films, are reviewed. The principle of operation of each active system is briefly explained. Recent technological advances in active packaging are discussed, and food related applications are presented. The effects of active packaging systems on food quality and safety are cited.

DHLW Glass Waste Package Criticality Analysis (SCPB:N/A)

International Nuclear Information System (INIS)

Davis, J.W.

1996-01-01

This analysis is prepared by the Mined Geologic Disposal System (MGDS) Waste Package Development Department (WPDD) to determine the viability of the Defense High-Level Waste (DHLW) Glass waste package concept with respect to criticality regulatory requirements in compliance with the goals of the Waste Package Implementation Plan (Ref. 5.1) for conceptual design. These design calculations are performed in sufficient detail to provide a comprehensive comparison base with other design alternatives. The objective of this evaluation is to show to what extent the concept meets the regulatory requirements or indicate additional measures that are required for the intact waste package

COMPACTION OF FIBERBOARD IN A 9975 SHIPPING PACKAGE

Energy Technology Data Exchange (ETDEWEB)

Stefek, T.; Daugherty, W.; Estochen, E.; Leduc, D.

2011-05-11

Compaction of lower layers in the fiberboard overpack has been observed in 9975 packages that contain elevated moisture. Lab testing has resulted in a better understanding of (1) the relationship between the fiberboard moisture level and compaction of the lower fiberboard assembly, and (2) the behavior of the fiberboard during transport. In laboratory tests, higher moisture content has been shown to correspond to higher total compaction of fiberboard material, greater rate of compaction, and continued compaction over a longer period of time. In addition, laboratory tests have shown that the application of a dynamic load results in higher fiberboard compaction. The test conditions and sample geometric/loading configurations were chosen to simulate the regulatory requirements for 9975 package input dynamic loading. Dynamic testing was conducted over a period of several months to acquire immediate and cumulative changes in geometric data for various moisture levels. Currently, one sample set has undergone a complete dynamic test regimen, while testing of another set is still in-progress. The dynamic input, data acquisition, test effects on sample dynamic parameters, and interim results from this test program will be summarized and compared to regulatory specifications for dynamic loading. This will provide a basis from which to evaluate the impact of moisture and fiberboard compaction on the safety basis for transportation (Safety Analysis Report for Packaging) and storage (facility Documented Safety Analysis) at the Savannah River Site (SRS).

Sensitivity analysis of reactor safety parameters with automated adjoint function generation

International Nuclear Information System (INIS)

Kallfelz, J.M.; Horwedel, J.E.; Worley, B.A.

1992-01-01

A project at the Paul Scherrer Institute (PSI) involves the development of simulation models for the transient analysis of the reactors in Switzerland (STARS). This project, funded in part by the Swiss Federal Nuclear Safety Inspectorate, also involves the calculation and evaluation of certain transients for Swiss light water reactors (LWRs). For best-estimate analyses, a key element in quantifying reactor safety margins is uncertainty evaluation to determine the uncertainty in calculated integral values (responses) caused by modeling, calculational methodology, and input data (parameters). The work reported in this paper is a joint PSI/Oak Ridge National Laboratory (ORNL) application to a core transient analysis code of an ORNL software system for automated sensitivity analysis. The Gradient-Enhanced Software System (GRESS) is a software package that can in principle enhance any code so that it can calculate the sensitivity (derivative) to input parameters of any integral value (response) calculated in the original code. The studies reported are the first application of the GRESS capability to core neutronics and safety codes

Advancements in meat packaging.

Science.gov (United States)

McMillin, Kenneth W

2017-10-01

Packaging of meat provides the same or similar benefits for raw chilled and processed meats as other types of food packaging. Although air-permeable packaging is most prevalent for raw chilled red meat, vacuum and modified atmosphere packaging offer longer shelf life. The major advancements in meat packaging have been in the widely used plastic polymers while biobased materials and their integration into composite packaging are receiving much attention for functionality and sustainability. At this time, active and intelligent packaging are not widely used for antioxidant, antimicrobial, and other functions to stabilize and enhance meat properties although many options are being developed and investigated. The advances being made in nanotechnology will be incorporated into food packaging and presumably into meat packaging when appropriate and useful. Intelligent packaging using sensors for transmission of desired information and prompting of subsequent changes in packaging materials, environments or the products to maintain safety and quality are still in developmental stages. Copyright © 2017 Elsevier Ltd. All rights reserved.

Extension of ship accident analysis to multiple-package shipments

International Nuclear Information System (INIS)

Mills, G.S.; Neuhauser, K.S.

1997-11-01

Severe ship accidents and the probability of radioactive material release from spent reactor fuel casks were investigated previously. Other forms of RAM, e.g., plutonium oxide powder, may be shipped in large numbers of packagings rather than in one to a few casks. These smaller, more numerous packagings are typically placed in ISO containers for ease of handling, and several ISO containers may be placed in one of several holds of a cargo ship. In such cases, the size of a radioactive release resulting from a severe collision with another ship is determined not by the likelihood of compromising a single, robust package but by the probability that a certain fraction of 10's or 100's of individual packagings is compromised. The previous analysis involved a statistical estimation of the frequency of accidents which would result in damage to a cask located in one of seven cargo holds in a collision with another ship. The results were obtained in the form of probabilities (frequencies) of accidents of increasing severity and of release fractions for each level of severity. This paper describes an extension of the same general method in which the multiple packages are assumed to be compacted by an intruding ship's bow until there is no free space in the hold. At such a point, the remaining energy of the colliding ship is assumed to be dissipated by progressively crushing the RAM packagings and the probability of a particular fraction of package failures is estimated by adaptation of the statistical method used previously. The parameters of a common, well characterized packaging, the 6M with 2R inner containment vessel, were employed as an illustrative example of this analysis method. However, the method is readily applicable to other packagings for which crush strengths have been measured or can be estimated with satisfactory confidence

Extension of ship accident analysis to multiple-package shipments

International Nuclear Information System (INIS)

Mills, G.S.; Neuhauser, K.S.

1998-01-01

Severe ship accidents and the probability of radioactive material release from spent reactor fuel casks were investigated previously (Spring, 1995). Other forms of RAM, e.g., plutonium oxide powder, may be shipped in large numbers of packagings rather than in one to a few casks. These smaller, more numerous packagings are typically placed in ISO containers for ease of handling, and several ISO containers may be placed in one of several holds of a cargo ship. In such cases, the size of a radioactive release resulting from a severe collision with another ship is determined not by the likelihood of compromising a single, robust package but by the probability that a certain fraction of 10's or 100's of individual packagings is compromised. The previous analysis (Spring, 1995) involved a statistical estimation of the frequency of accidents which would result in damage to a cask located in one of seven cargo holds in a collision with another ship. The results were obtained in the form of probabilities (frequencies) of accidents of increasing severity and of release fractions for each level of severity. This paper describes an extension of the same general method in which the multiple packages are assumed to be compacted by an intruding ship's bow until there is no free space in the hold. At such a point, the remaining energy of the colliding ship is assumed to be dissipated by progressively crushing the RAM packagings and the probability of a particular fraction of package failures is estimated by adaptation of the statistical method used previously. The parameters of a common, well-characterized packaging, the 6M with 2R inner containment vessel, were employed as an illustrative example of this analysis method. However, the method is readily applicable to other packagings for which crush strengths have been measured or can be estimated with satisfactory confidence. (authors)

WGCNA: an R package for weighted correlation network analysis.

Science.gov (United States)

Langfelder, Peter; Horvath, Steve

2008-12-29

Correlation networks are increasingly being used in bioinformatics applications. For example, weighted gene co-expression network analysis is a systems biology method for describing the correlation patterns among genes across microarray samples. Weighted correlation network analysis (WGCNA) can be used for finding clusters (modules) of highly correlated genes, for summarizing such clusters using the module eigengene or an intramodular hub gene, for relating modules to one another and to external sample traits (using eigengene network methodology), and for calculating module membership measures. Correlation networks facilitate network based gene screening methods that can be used to identify candidate biomarkers or therapeutic targets. These methods have been successfully applied in various biological contexts, e.g. cancer, mouse genetics, yeast genetics, and analysis of brain imaging data. While parts of the correlation network methodology have been described in separate publications, there is a need to provide a user-friendly, comprehensive, and consistent software implementation and an accompanying tutorial. The WGCNA R software package is a comprehensive collection of R functions for performing various aspects of weighted correlation network analysis. The package includes functions for network construction, module detection, gene selection, calculations of topological properties, data simulation, visualization, and interfacing with external software. Along with the R package we also present R software tutorials. While the methods development was motivated by gene expression data, the underlying data mining approach can be applied to a variety of different settings. The WGCNA package provides R functions for weighted correlation network analysis, e.g. co-expression network analysis of gene expression data. The R package along with its source code and additional material are freely available at http://www.genetics.ucla.edu/labs/horvath/CoexpressionNetwork/Rpackages/WGCNA.

Safety analysis report for packaging (onsite) for the Waste Encapsulation and Storage Facility ion exchange module

International Nuclear Information System (INIS)

Romano, T.

1997-01-01

The Waste Encapsulation and Storage Facility (WESF) is in need of providing an emergency ion exchange system to remove cesium or strontium from the pool cell in the event of a capsule failure. The emergency system is call the WESF Emergency Ion Exchange System and the packaging is called the WESF ion exchange module (WIXM). The packaging system will meet the onsite transportation requirements for a Type B, highway route controlled quantity package as well as disposal requirements for Category 3 waste

Eliminating HIV-1 Packaging Sequences from Lentiviral Vector Proviruses Enhances Safety and Expedites Gene Transfer for Gene Therapy.

Science.gov (United States)

Vink, Conrad A; Counsell, John R; Perocheau, Dany P; Karda, Rajvinder; Buckley, Suzanne M K; Brugman, Martijn H; Galla, Melanie; Schambach, Axel; McKay, Tristan R; Waddington, Simon N; Howe, Steven J

2017-08-02

Lentiviral vector genomic RNA requires sequences that partially overlap wild-type HIV-1 gag and env genes for packaging into vector particles. These HIV-1 packaging sequences constitute 19.6% of the wild-type HIV-1 genome and contain functional cis elements that potentially compromise clinical safety. Here, we describe the development of a novel lentiviral vector (LTR1) with a unique genomic structure designed to prevent transfer of HIV-1 packaging sequences to patient cells, thus reducing the total HIV-1 content to just 4.8% of the wild-type genome. This has been achieved by reconfiguring the vector to mediate reverse-transcription with a single strand transfer, instead of the usual two, and in which HIV-1 packaging sequences are not copied. We show that LTR1 vectors offer improved safety in their resistance to remobilization in HIV-1 particles and reduced frequency of splicing into human genes. Following intravenous luciferase vector administration to neonatal mice, LTR1 sustained a higher level of liver transgene expression than an equivalent dose of a standard lentivirus. LTR1 vectors produce reverse-transcription products earlier and start to express transgenes significantly quicker than standard lentiviruses after transduction. Finally, we show that LTR1 is an effective lentiviral gene therapy vector as demonstrated by correction of a mouse hemophilia B model. Copyright © 2017 The Author(s). Published by Elsevier Inc. All rights reserved.

Normal Condition on Transport Thermal Analysis and Testing of a Type B Drum Package

International Nuclear Information System (INIS)

Jerrell, J.W.; van Alstine, M.N.; Gromada, R.J.

1995-01-01

Increasing the content limits of radioactive material packagings can save money and increase transportation safety by decreasing the total number of shipments required to transport large quantities of material. The contents of drum packages can be limited by unacceptable containment vessel pressures and temperatures due to the thermal properties of the insulation. The purpose of this work is to understand and predict the effects of insulation properties on containment system performance

CH Packaging Program Guidance

International Nuclear Information System (INIS)

2005-01-01

The purpose of this document is to provide the technical requirements for preparation for use, operation, inspection, and maintenance of a Transuranic Package Transporter Model II (TRUPACT-II), a HalfPACT shipping package, and directly related components. This document complies with the minimum requirements as specified in the TRUPACT-II Safety Analysis Report for Packaging (SARP), HalfPACT SARP, and U.S. Nuclear Regulatory Commission (NRC) Certificates of Compliance (C of C) 9218 and 9279, respectively. In the event of a conflict between this document and the SARP or C of C, the C of C shall govern. The C of Cs state: ''each package must be prepared for shipment and operated in accordance with the procedures described in Chapter 7.0, Operating Procedures, of the application.'' They further state: ''each package must be tested and maintained in accordance with the procedures described in Chapter 8.0, Acceptance Tests and Maintenance Program of the Application.'' Chapter 9.0 of the SARP charges the Waste Isolation Pilot Plant (WIPP) management and operating (M and O) contractor with assuring packaging is used in accordance with the requirements of the C of C. Because the packaging is NRC-approved, users need to be familiar with Title 10 Code of Federal Regulations (CFR) 71.8. Any time a user suspects or has indications that the conditions of approval in the C of C were not met, the Carlsbad Field Office (CBFO) shall be notified immediately. The CBFO will evaluate the issue and notify the NRC if required.

Edible packaging materials.

Science.gov (United States)

Janjarasskul, Theeranun; Krochta, John M

2010-01-01

Research groups and the food and pharmaceutical industries recognize edible packaging as a useful alternative or addition to conventional packaging to reduce waste and to create novel applications for improving product stability, quality, safety, variety, and convenience for consumers. Recent studies have explored the ability of biopolymer-based food packaging materials to carry and control-release active compounds. As diverse edible packaging materials derived from various by-products or waste from food industry are being developed, the dry thermoplastic process is advancing rapidly as a feasible commercial edible packaging manufacturing process. The employment of nanocomposite concepts to edible packaging materials promises to improve barrier and mechanical properties and facilitate effective incorporation of bioactive ingredients and other designed functions. In addition to the need for a more fundamental understanding to enable design to desired specifications, edible packaging has to overcome challenges such as regulatory requirements, consumer acceptance, and scaling-up research concepts to commercial applications.

The development of a safety analysis methodology for the optimized power reactor 1000

International Nuclear Information System (INIS)

Hwang-Yong, Jun; Yo-Han, Kim

2005-01-01

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

Safety evaluation report for packaging (onsite) concrete-lined waste packaging

International Nuclear Information System (INIS)

Romano, T.

1997-01-01

The Pacific Northwest National Laboratory developed a package to ship Type A, non-transuranic, fissile excepted quantities of liquid or solid radioactive material and radioactive mixed waste to the Central Waste Complex for storage on the Hanford Site

MEASURE/ANOMTEST. Anomaly detection software package for the Dodewaard power plant facility. Supplement 1. Extension of measurement analysis part, addition of plot package

International Nuclear Information System (INIS)

Schoonewelle, H.

1995-01-01

The anomaly detection software package installed at the Dodewaard nuclear power plant has been revised with respect to the part of the measurement analysis. A plot package has been added to the package. Signals in which an anomaly has been detected are automatically plotted including the uncertainty margins of the signals. This report gives a description of the revised measurement analysis part and the plot package. Each new routine of the plot package is described briefly and the new input and output files are given. (orig.)

Thirty years of transport package development for spent fuels

International Nuclear Information System (INIS)

Cory, A.R.

2005-01-01

By June 2005, when shipments of spent fuel for reprocessing from Germany are concluded, BNFL flask types will have been responsible for transporting more than 2000 tonnes of heavy metal in Europe in the form of spent fuel. Several thousand more tonnes of spent fuel have been transported by sea from Japan over the last thirty years. The design of spent fuel packages has not stood still for that time. In order to anticipate the changing needs of the nuclear power generation industry, advances have been made both in package design and analysis. Thirty years ago spent fuel burnup and initial enrichment were considerably lower, which was reflected in the different demands placed on the shielding design of packages, and in the design of the internal basket to separate the fuel assemblies. Technical development of both 'wet' (water-filled cavity) and 'dry' packages has progressed in parallel, and the relative merits and peculiarities of each type is explored. BNFL has considerable experience in the operation of both types, and is well placed to comment on practical and functional issues associated with both types. While there have been certain evolutionary changes affecting package design, there have also been more significant changes in the Design Safety Case. These have sometimes been necessary to meet changes in IAEA Regulations, or the challenges posed by the regulators themselves. In other cases advantage has been taken of improvements in analytical techniques to demonstrate increased margins of operational safety. Where possible these margins have also been increased by other means, such as taking advantage of commercial trends to reduce package thermal loads. A key factor over the last thirty years has been the increasing influence of the Regulating Authorities and the development of the IAEA Regulations. The various Competent Authorities now tend to have a higher proportion of technical experts, often recruited from the nuclear industry, and are thus more able to

Design, development and production of the TNF-XI new powder package

International Nuclear Information System (INIS)

Naigeon, P.; Brut, S.; Fujiwara, T.; Cohen, M.E.

2004-01-01

The TNF-XI was jointly developed by COGEMA LOGISTICS of France and Nuclear Fuel Industries, Ltd. (NFI) of Japan. The design and development of this package were started in 2000 and it was first used in 2003. The package design was based on the COGEMA LOGISTICS TN trademark UO2 package, which is cylindrical. To optimize the use of space and to facilitate operations, the TNF-XI incorporates four cavities. Each cavity is similar to one TN trademark U02. The overall shape of the package is approximately a one-meter cube, which allows it to be easily arranged and stacked in a transport container. It has a total weight of nearly 1 metric ton when loaded with 300 kg of uraniferous oxide material. The powder (or pellets) is placed inside NFI pails, and three pails are placed inside each cavity. The test results and analyses were documented in the French, Japanese, and United States 'Safety Analysis Reports' in order to get approvals in these countries. Additionally, the French license was validated in other countries. The 'Safety Analysis Reports' were prepared by COGEMA LOGISTICS in France, NFI in Japan and PACKAGING TECHNOLOGY, Inc. in the USA. In order to transform the TNF-XI design into a fleet of 800 packagings, a successful mass production process had to be developed and put into place by COGEMA LOGISTICS and the fabricator MECAGEST. Adopting mass production methods was challenging - not only because of the large quantity of packages, but also because of the rate of production (50 units per month minimum), quality requirements and the need to keep low packaging costs. This resulted in a significant challenge for mass producing the packaging: what was needed was a robust, efficient, and well-organized process. During the development of the manufacturing process, we also worked on shortening the learning curve

Criticality safety of spent fuel casks considering water inleakage

International Nuclear Information System (INIS)

Osgood, N.L.; Withee, C.J.; Easton, E.P.

2004-01-01

A fundamental safety design parameter for all fissile material packages is that a single package must be critically safe even if water leaks into the containment system. In addition, criticality safety must be assured for arrays of packages under normal conditions of transport (undamaged packages) and under hypothetical accident conditions (damaged packages). The U.S. Nuclear Regulatory Commission staff has revised the review protocol for demonstrating criticality safety for spent fuel casks. Previous review guidance specified that water inleakage be considered under accident conditions. This practice was based on the fact that the leak tightness of spent fuel casks is typically demonstrated by use of structural analysis and not by physical testing. In addition, since a single package was shown to be safe with water inleakage, it was concluded that this analysis was also applicable to an array of damaged packages, since the heavy shield walls in spent fuel casks neutronically isolate each cask in the array. Inherent in this conclusion is that the fuel assembly geometry does not change significantly, even under drop test conditions. Requests for shipping fuel with burnup exceeding 40 GWd/MTU, including very high burnups exceeding 60 GWD/MTU, caused a reassessment of this assumption. Fuel cladding structural strength and ductility were not clearly predictable for these higher burnups. Therefore the single package analysis for an undamaged package may not be applicable for the damaged package. NRC staff developed a new practice for review of spent fuel casks under accident conditions. The practice presents two methods for approval that would allow an assessment of potential reconfiguration of the fuel assembly under accident conditions, or, alternatively, a demonstration of the water-exclusion boundary through physical testing

Nanotechnology: An Untapped Resource for Food Packaging.

Science.gov (United States)

Sharma, Chetan; Dhiman, Romika; Rokana, Namita; Panwar, Harsh

2017-01-01

Food commodities are packaged and hygienically transported to protect and preserve them from any un-acceptable alteration in quality, before reaching the end-consumer. Food packaging continues to evolve along-with the innovations in material science and technology, as well as in light of consumer's demand. Presently, the modern consumers of competitive economies demands for food with natural quality, assured safety, minimal processing, extended shelf-life and ready-to-eat concept. Innovative packaging systems, not only ascertains transit preservation and effective distribution, but also facilitates communication at the consumer levels. The technological advances in the domain of food packaging in twenty-first century are mainly chaired by nanotechnology, the science of nano-materials. Nanotechnology manipulates and creates nanometer scale materials, of commercial and scientific relevance. Introduction of nanotechnology in food packaging sector has significantly addressed the food quality, safety and stability concerns. Besides, nanotechnology based packaging intimate's consumers about the real time quality of food product. Additionally, nanotechnology has been explored for controlled release of preservatives/antimicrobials, extending the product shelf life within the package. The promising reports for nanotechnology interventions in food packaging have established this as an independent priority research area. Nanoparticles based food packages offer improved barrier and mechanical properties, along with food preservation and have gained welcoming response from market and end users. In contrary, recent advances and up-liftment in this area have raised various ethical, environmental and safety concerns. Policies and regulation regarding nanoparticles incorporation in food packaging are being reviewed. This review presents the existing knowledge, recent advances, concerns and future applications of nanotechnology in food packaging sector.

Safety analysis report for packaging Lawrence Livermore Laboratories shipping containers

International Nuclear Information System (INIS)

Evans, J.H.

1975-12-01

The Lawrence Livermore Laboratories shipping containers were designed at Oak Ridge National Laboratory for use in transporting weapons and nuclear components. The design for the containers was evaluated to show compliance with applicable regulations governing packages in which radioactive and fissile materials are transported. Computational procedures were used to determine the structural integrity and thermal behavior of the containers relative to the standards for the normal conditions of transport. A full-scale container test model was destructively tested to verify compliance with the standards for the accident conditions. The results of the analytical evaluations and the tests demonstrate that the design for the Lawrence Livermore Laboratories shipping containers is in compliance with the applicable regulations

16 CFR 1700.15 - Poison prevention packaging standards.

Science.gov (United States)

2010-01-01

... 16 Commercial Practices 2 2010-01-01 2010-01-01 false Poison prevention packaging standards. 1700.15 Section 1700.15 Commercial Practices CONSUMER PRODUCT SAFETY COMMISSION POISON PREVENTION PACKAGING ACT OF 1970 REGULATIONS POISON PREVENTION PACKAGING § 1700.15 Poison prevention packaging...

River Data Package for the 2004 Composite Analysis

Energy Technology Data Exchange (ETDEWEB)

Rakowski, Cynthia L.; Guensch, Gregory R.; Patton, Gregory W.

2004-08-01

Beginning in fiscal year 2003, the DOE Richland Operations Office initiated activities, including the development of data packages, to support the 2004 Composite Analysis. The river data package provides calculations of flow and transport in the Columbia River system. This document presents the data assembled to run the river module components for the section of the Columbia River from Vernita Bridge to the confluence with the Yakima River.

Documentation and analysis for packaging limited quantity ice chests

International Nuclear Information System (INIS)

Nguyen, P.M.

1995-01-01

The purpose of this Documentation and Analysis for Packaging (DAP) is to document that ice chests meet the intent of the International Air Transport Association (IATA) and the U.S. Department of Transportation (DOT) Code of Federal Regulations as strong, tight containers for the packaging of limited quantities for transport. This DAP also outlines the packaging method used to protect the sample bottles from breakage. Because the ice chests meet the DOT requirements, they can be used to ship LTD QTY on the Hanford Site

Certification test for safety of new fuel transportation package

International Nuclear Information System (INIS)

Aritomi, Masanori; Sugawa, Osami; Suga, Masao.

1993-01-01

The objective of this certification test is to prove the safety of new fuel transportation package against a fire of actual size caused by traffic accidents. After the fire test, the fuel assemblies were covered with coal-tar like material vaporized from anti-shock material used in the container. Surface color of BWR-type fuel assembly was dark grey that is supposed to be the color of oxide of Zircaloy. As for PWR-type fuel assembly, the condition encountered during fire test caused no change to the outlook of the rod element. Both the BWR and PWR type fuel rod elements showed no deformation and were completely sound. Therefore it may be concluded that the container protected the mimic fuel assemblies against fire of 30 minutes duration and caused no damage. This report is the result of the above experiments and examinations, and we appreciate the cooperation of those who are concerned. (J.P.N.)

IMAP: A complete Ion Micro-Analysis Package for the nuclear microprobe

International Nuclear Information System (INIS)

Antolak, A.J.; Hildner, M.L.; Morse, D.H.; Bench, G.S.

1993-01-01

Microprobe techniques using scanned, focused MeV ions are routinely used in Livermore for materials characterization. Comprehensive data analysis with these techniques is accomplished with the computer software package IMAP, for Ion Micro-Analysis Package. IMAP consists of a set of command language procedures for data processing and quantitative spectral analysis. Deconvolution of the data is achieved by spawning sub-processes within IMAP which execute analysis codes for each specific microprobe technique. IMAP is structured to rapidly analyze individual spectra or multi-dimensional data blocks which classify individual events by the two scanning dimensions, the energy of the detected radiation and, when necessary, one sample rotation dimension. Several examples are presented to demonstrate the utility of the package

Dualities in the analysis of phage DNA packaging motors

Science.gov (United States)

Serwer, Philip; Jiang, Wen

2012-01-01

The DNA packaging motors of double-stranded DNA phages are models for analysis of all multi-molecular motors and for analysis of several fundamental aspects of biology, including early evolution, relationship of in vivo to in vitro biochemistry and targets for anti-virals. Work on phage DNA packaging motors both has produced and is producing dualities in the interpretation of data obtained by use of both traditional techniques and the more recently developed procedures of single-molecule analysis. The dualities include (1) reductive vs. accretive evolution, (2) rotation vs. stasis of sub-assemblies of the motor, (3) thermal ratcheting vs. power stroking in generating force, (4) complete motor vs. spark plug role for the packaging ATPase, (5) use of previously isolated vs. new intermediates for analysis of the intermediate states of the motor and (6) a motor with one cycle vs. a motor with two cycles. We provide background for these dualities, some of which are under-emphasized in the literature. We suggest directions for future research. PMID:23532204

Multiple-Group Analysis Using the sem Package in the R System

Science.gov (United States)

Evermann, Joerg

2010-01-01

Multiple-group analysis in covariance-based structural equation modeling (SEM) is an important technique to ensure the invariance of latent construct measurements and the validity of theoretical models across different subpopulations. However, not all SEM software packages provide multiple-group analysis capabilities. The sem package for the R…

Grandfathering of competent authority approved packages

International Nuclear Information System (INIS)

Osgood, N.L.

2004-01-01

International Atomic Energy Agency transportation regulations are reviewed and revised on a periodic basis as new technical and scientific information becomes available. The 1996 Edition of the Regulations for the Safe Transport of Radioactive Materials in TS-R-1 includes provisions for the use of package designs approved to previous editions of the regulations. This assures that there is no disruption of transport when the regulations are updated and revised. The revision of package design standards may make certain designs obsolete, though not necessarily unsafe. The U.S. Nuclear Regulatory Commission is the agency in the United States that certifies transportation packages for Type B and fissile materials. NRC regulations include grandfathering provisions that are comparable to and compatible with the IAEA standards. NRC staff is promoting a new system that would eliminate the need to grandfather package designs. Under the new method, any new or revised provision of the regulations that affects package standards would include its own transitional arrangements. In this way, each change would be evaluated for its safety importance. Changes in the package standards that are important to safety would be implemented immediately upon the regulations coming into force. Other changes, that do not significantly affect safety, would have longer implementation periods. In this way, all packages in use would be compatible with the regulations in force, and no specific grandfathering provisions for older designs would be needed. NRC staff has concluded that the package design standards are mature and have been shown to be protective over the past 40 years of shipping experience. We predict that future changes in package design standards will not be substantive in terms of resulting in significant changes in physical performance of a package in transport, including actual transportation accidents. The benefits of the new system would be a more predictable regulatory structure

49 CFR 173.465 - Type A packaging tests.

Science.gov (United States)

2010-10-01

... 49 Transportation 2 2010-10-01 2010-10-01 false Type A packaging tests. 173.465 Section 173.465 Transportation Other Regulations Relating to Transportation PIPELINE AND HAZARDOUS MATERIALS SAFETY... SHIPMENTS AND PACKAGINGS Class 7 (Radioactive) Materials § 173.465 Type A packaging tests. (a) The packaging...

PyPathway: Python Package for Biological Network Analysis and Visualization.

Science.gov (United States)

Xu, Yang; Luo, Xiao-Chun

2018-05-01

Life science studies represent one of the biggest generators of large data sets, mainly because of rapid sequencing technological advances. Biological networks including interactive networks and human curated pathways are essential to understand these high-throughput data sets. Biological network analysis offers a method to explore systematically not only the molecular complexity of a particular disease but also the molecular relationships among apparently distinct phenotypes. Currently, several packages for Python community have been developed, such as BioPython and Goatools. However, tools to perform comprehensive network analysis and visualization are still needed. Here, we have developed PyPathway, an extensible free and open source Python package for functional enrichment analysis, network modeling, and network visualization. The network process module supports various interaction network and pathway databases such as Reactome, WikiPathway, STRING, and BioGRID. The network analysis module implements overrepresentation analysis, gene set enrichment analysis, network-based enrichment, and de novo network modeling. Finally, the visualization and data publishing modules enable users to share their analysis by using an easy web application. For package availability, see the first Reference.

CH Packaging Program Guidance

International Nuclear Information System (INIS)

Washington TRU Solutions LLC

2003-01-01

The purpose of this document is to provide the technical requirements for preparation for use, operation, inspection, and maintenance of a Transuranic Package Transporter Model II (TRUPACT-II), a HalfPACT shipping package, and directly related components. This document complies with the minimum requirements as specified in the TRUPACT-II Safety Analysis Report for Packaging (SARP), HalfPACT SARP, and Nuclear Regulatory Commission (NRC) Certificates of Compliance (C of C) 9218 and 9279, respectively. In the event of a conflict between this document and the SARP or C of C, the C of C shall govern. The C of Cs state: ''each package must be prepared for shipment and operated in accordance with the procedures described in Chapter 7.0, Operating Procedures, of the application.'' They further state: ''each package must be tested and maintained in accordance with the procedures described in Chapter 8.0, Acceptance Tests and Maintenance Program of the Application.'' Chapter 9.0 of the SARP charges the WIPP management and operating (M and O) contractor with assuring packaging is used in accordance with the requirements of the C of C. Because the packaging is NRC-approved, users need to be familiar with 10 CFR 71.11. Any time a user suspects or has indications that the conditions of approval in the C of C were not met, the Carlsbad Field Office (CBFO) shall be notified immediately. CBFO will evaluate the issue and notify the NRC if required. This document provides the instructions to be followed to operate, maintain, and test the TRUPACT-II and HalfPACT packaging. The intent of these instructions is to standardize operations. All users will follow these instructions or equivalent instructions that assure operations are safe and meet the requirements of the SARPs

CH Packaging Program Guidance

International Nuclear Information System (INIS)

Washington TRU Solutions LLC

2002-01-01

The purpose of this document is to provide the technical requirements for preparation for use, operation, inspection, and maintenance of a Transuranic Package Transporter Model II (TRUPACT-II), a HalfPACT Shipping Package, and directly related components. This document complies with the minimum requirements as specified in TRUPACT-II Safety Analysis Report for Packaging (SARP), HalfPACT SARP, and Nuclear Regulatory Commission (NRC) Certificates of Compliance (C of C) 9218 and 9279, respectively. In the event there is a conflict between this document and the SARP or C of C, the SARP and/or C of C shall govern. C of Cs state: ''each package must be prepared for shipment and operated in accordance with the procedures described in Chapter 7.0, Operating Procedures, of the application.'' They further state: ''each package must be tested and maintained in accordance with the procedures described in Chapter 8.0, Acceptance Tests and Maintenance Program of the Application.'' Chapter 9.0 of the SAR P charges the WIPP Management and Operation (M and O) contractor with assuring packaging is used in accordance with the requirements of the C of C. Because the packaging is NRC-approved, users need to be familiar with 10 CFR 71.11. Any time a user suspects or has indications that the conditions of approval in the C of C were not met, the Carlsbad Field Office (CBFO) shall be notified immediately. CBFO will evaluate the issue and notify the NRC if required. This document details the instructions to be followed to operate, maintain, and test the TRUPACT-II and HalfPACT packaging. The intent of these instructions is to standardize these operations. All users will follow these instructions or equivalent instructions that assure operations are safe and meet the requirements of the SARPs

Shortened outage duration and increased safety with head assembly upgrade packages

International Nuclear Information System (INIS)

Leanne, M.; Lisien, P.E.; Plute, K.; Duran, J.

2007-01-01

To significantly reduce outage critical path duration and personnel radiation exposure, and to increase personnel safety, Westinghouse Electric Co., LLC has designed and installed upgrades to the existing head assemblies of operating pressurized water reactors. These upgrades are known as Head Assembly Upgrade Packages (HAUPs) or Simplified Head Assemblies (SHAs). Custom configurations are created from a set of standard elements to optimize the design for each unique containment, head assembly configuration, and licensing basis. Two primary options are available for implementation: a full HAUP or targeted component and system upgrades. Plants may achieve much of the outage savings, dose reduction, and safety improvements even with a more limited hardware scope. A range of improvements can be offered from integral missile shields, to redesigned duct work, radiation shields, and cable layout and connection optimization. The hardware changes are customized to target the scope that adds the most value for a given plant. While combining upgrades with a reactor vessel head (RVH) replacement adds some flexibility, it is not necessary. Some plants have chosen to implement targeted upgrades prior to a replacement RVH outage and then complete the remainder of the full HAUP during the replacement RVH outage. Three-dimensional computer aided design tools are used in the conceptual and detailed design phases to identify and avoid interferences between existing and replacement plant components. State-of-the-art computational fluid dynamics models for control drive mechanism (CDM) cooling systems are used to demonstrate the ability to maintain or improve the original design performance while greatly simplifying the disassembly/re-assembly activities. Likewise, state-of-the-art finite element analysis methods allow optimization of structural components while meeting code limits for design basis accident conditions. (authors)

Intercomparison of gamma ray analysis software packages

International Nuclear Information System (INIS)

1998-04-01

The IAEA undertook an intercomparison exercise to review available software for gamma ray spectra analysis. This document describes the methods used in the intercomparison exercise, characterizes the software packages reviewed and presents the results obtained. Only direct results are given without any recommendation for a particular software or method for gamma ray spectra analysis

Network meta-analysis using R: a review of currently available automated packages.

Directory of Open Access Journals (Sweden)

Binod Neupane

Full Text Available Network meta-analysis (NMA--a statistical technique that allows comparison of multiple treatments in the same meta-analysis simultaneously--has become increasingly popular in the medical literature in recent years. The statistical methodology underpinning this technique and software tools for implementing the methods are evolving. Both commercial and freely available statistical software packages have been developed to facilitate the statistical computations using NMA with varying degrees of functionality and ease of use. This paper aims to introduce the reader to three R packages, namely, gemtc, pcnetmeta, and netmeta, which are freely available software tools implemented in R. Each automates the process of performing NMA so that users can perform the analysis with minimal computational effort. We present, compare and contrast the availability and functionality of different important features of NMA in these three packages so that clinical investigators and researchers can determine which R packages to implement depending on their analysis needs. Four summary tables detailing (i data input and network plotting, (ii modeling options, (iii assumption checking and diagnostic testing, and (iv inference and reporting tools, are provided, along with an analysis of a previously published dataset to illustrate the outputs available from each package. We demonstrate that each of the three packages provides a useful set of tools, and combined provide users with nearly all functionality that might be desired when conducting a NMA.

Drop testing of the Westinghouse fresh nuclear fuel package

International Nuclear Information System (INIS)

Shappert, L.B.; Sanders, C.F.

1992-01-01

In recent years, the Westinghouse Columbia Fuel Fabrication Facility has been faced with increasing pressure from utilities that wished to take the fuel in their nuclear power plants to higher burnups. To help accommodate this trend, Westinghouse has determined that it needs the ability to increase the enrichment of the fresh fuel it delivers to its customers. One critical step in this process is to certify a new (Type A, fissile) fresh fuel package design that has the capability to transport fuel with a higher enrichment than was previously available. A prototype package was tested in support of the Safety Analysis Report of the Packaging. This paper provides detailed information on those tests and their results

Time temperature indicators as devices intelligent packaging

Directory of Open Access Journals (Sweden)

Adriana Pavelková

2013-01-01

Full Text Available Food packaging is an important part of food production. Temperature is a one of crucial factor which affecting the quality and safety of food products during distribution, transport and storage. The one way of control of food quality and safety is the application of new packaging systems, which also include the intelligent or smart packaging. Intelligent packaging is a packaging system using different indicators for monitoring the conditions of production, but in particular the conditions during transport and storage. Among these indicators include the time-temperature indicators to monitor changes in temperature, which is exposed the product and to inform consumers about the potential risks associated with consumption of these products. Time temperature indicators are devices that show an irreversible change in a physical characteristic, usually color or shape, in response to temperature history. Some are designed to monitor the evolution of temperature with time along the distribution chain and others are designed to be used in the consumer packages.

Thermal characteristic of insulation for optimum design of RI transport package

International Nuclear Information System (INIS)

Lee, J. C.; Bang, K. S.; Seo, K. S.

2002-01-01

A package to transport the high level radioactive materials in required to withstand the hypothetical accident conditions as well as normal transport conditions according to IAEA and domestic regulations. The regulations require that the package should maintain the shielding, thermal and structural integrities to release no radioactive material. Thermal characteristics of insulations were evaluated and optimum insulation thickness was deduced for RI transport package. The package has a maximum capacity of 600 Curies for Ir-192 sealed source. The insulation thickness was decided with 10 mm of polyurethane form to maintain the thermal safety under fire accident condition. Thermal analysis was carried out for RI transport package, and it was shown that the thermal integrity of the package was maintained. The results obtained this study will be applied to a basic data for design of RI transport cask

Analysis and evaluation of a radioactive waste package retrieved from the Farallon Islands 900-meter disposal site

International Nuclear Information System (INIS)

Colombo, P.; Kendig, M.W.

1990-09-01

The Environmental Protection Agency (EPA) was given a Congressional mandate to develop criteria and regulations governing the ocean disposal of all forms of waste. The EPA taken an active role both nationally and within the international nuclear regulatory community to develop the effective controls necessary to protect the health and safety of man and the marine environment. The EPA Office of Radiation Programs (ORP) first initiated feasibility studies to determine whether current technologies could be applied toward determining the fate of radioactive waste disposed of in the past. After successfully locating actual radioactive waste packages in formerly used disposal sites, in the United States, the Office of Radiation Programs developed an intensive program of site characterization studies to examine biological, chemical and physical characteristics including evaluations of the concentration and distribution of radionuclides within these sites, and has conducted a performance evaluation of past packaging techniques and materials. Brookhaven National Laboratory (BNL) has performed container corrosion and matrix analysis studies on the recovered radioactive waste packages. This report presents the final results of laboratory analyses performed. 17 refs., 40 figs., 7 tabs

Analysis and evaluation of a radioactive waste package retrieved from the Farallon Islands 900-meter disposal site

Energy Technology Data Exchange (ETDEWEB)

Colombo, P.; Kendig, M.W.

1990-09-01

The Environmental Protection Agency (EPA) was given a Congressional mandate to develop criteria and regulations governing the ocean disposal of all forms of waste. The EPA taken an active role both nationally and within the international nuclear regulatory community to develop the effective controls necessary to protect the health and safety of man and the marine environment. The EPA Office of Radiation Programs (ORP) first initiated feasibility studies to determine whether current technologies could be applied toward determining the fate of radioactive waste disposed of in the past. After successfully locating actual radioactive waste packages in formerly used disposal sites, in the United States, the Office of Radiation Programs developed an intensive program of site characterization studies to examine biological, chemical and physical characteristics including evaluations of the concentration and distribution of radionuclides within these sites, and has conducted a performance evaluation of past packaging techniques and materials. Brookhaven National Laboratory (BNL) has performed container corrosion and matrix analysis studies on the recovered radioactive waste packages. This report presents the final results of laboratory analyses performed. 17 refs., 40 figs., 7 tabs.

PIXAN: the Lucas Heights PIXE analysis computer package

International Nuclear Information System (INIS)

Clayton, E.

1986-11-01

To fully utilise the multielement capability and short measurement time of PIXE it is desirable to have an automated computer evaluation of the measured spectra. Because of the complex nature of PIXE spectra, a critical step in the analysis is the data reduction, in which the areas of characteristic peaks in the spectrum are evaluated. In this package the computer program BATTY is presented for such an analysis. The second step is to determine element concentrations, knowing the characteristic peak areas in the spectrum. This requires a knowledge of the expected X-ray yield for that element in the sample. The computer program THICK provides that information for both thick and thin PIXE samples. Together, these programs form the package PIXAN used at Lucas Heights for PIXE analysis

Antimicrobial food packaging: potential and pitfalls

Science.gov (United States)

Malhotra, Bhanu; Keshwani, Anu; Kharkwal, Harsha

2015-01-01

Nowadays food preservation, quality maintenance, and safety are major growing concerns of the food industry. It is evident that over time consumers’ demand for natural and safe food products with stringent regulations to prevent food-borne infectious diseases. Antimicrobial packaging which is thought to be a subset of active packaging and controlled release packaging is one such promising technology which effectively impregnates the antimicrobial into the food packaging film material and subsequently delivers it over the stipulated period of time to kill the pathogenic microorganisms affecting food products thereby increasing the shelf life to severe folds. This paper presents a picture of the recent research on antimicrobial agents that are aimed at enhancing and improving food quality and safety by reduction of pathogen growth and extension of shelf life, in a form of a comprehensive review. Examination of the available antimicrobial packaging technologies is also presented along with their significant impact on food safety. This article entails various antimicrobial agents for commercial applications, as well as the difference between the use of antimicrobials under laboratory scale and real time applications. Development of resistance amongst microorganisms is considered as a future implication of antimicrobials with an aim to come up with actual efficacies in extension of shelf life as well as reduction in bacterial growth through the upcoming and promising use of antimicrobials in food packaging for the forthcoming research down the line. PMID:26136740

Analysis on descriptions of precautionary statements in package inserts of medicines

Directory of Open Access Journals (Sweden)

Tsuchiya F

2012-02-01

Full Text Available Keita Nabeta1, Masaomi Kimura2, Michiko Ohkura2, Fumito Tsuchiya31Graduate School of Engineering and Science, Shibaura Institute of Technology, Toyosu 3-7-5, Koto-ku, Tokyo, 135-8548 Japan; 2Faculty of Engineering, Shibaura Institute of Technology, Toyosu 3-7-5, Koto-ku, Tokyo, 135-8548 Japan; 3School of Pharmacy, International University of Health and Welfare, Minami-Aoyama 1-24-1, Minato-ku, Tokyo, 107-0062 JapanBackground: To prevent medical accidents, users must be informed of the cautions written in medical package inserts. To realize countermeasures by utilizing information systems, we must also implement a drug information database. However, this is not easy to develop, since the descriptions in package inserts are too complex and their information poorly structured. It is necessary to analyze package insert information and propose a data structure.Methods: We analyzed the descriptions of 'precautions for application' in package inserts via text mining methods. In order to summarize statements, we applied dependency analysis to statements and visualized their relations between predicate words and other words. Furthermore, we extracted words representing timing to execute the order.Results: We found that there are four types of statements: direct orders such as "使用する" (use, causative orders such as "使用させる" (make someone use, direct interdictions such as "使用しない" (do not use, and causative interdictions such as "使用させない" (do not make user use. As for words representing timing, we extracted six groups: "at the time of delivery," "at the time of preparation," "in use," "after use," and "at the time of storage." From these results, we obtained points of consideration concerning the subjects of orders in the statements and timing of their execution.Conclusion: From the obtained knowledge, we can define the information structure used to describe the precautionary statement. It should contain information such

Role of waste packages in the safety of a high level waste repository in a deep geological formation

International Nuclear Information System (INIS)

Bretheau, F.; Lewi, J.

1990-06-01

The safety of a radioactive waste disposal facility lays on the three following barriers placed between the radioactive materials and the biosphere: the waste package; the engineered barriers; the geological barrier. The function assigned to each of these barriers in the performance assessment is an option taken by the organization responsible for waste disposal management (ANDRA in France), which must show that: expected performances of each barrier (confinement ability, life-time, etc.) are at least equal to those required to fulfill the assigned function; radiation protection requirements are met in all situations considered as credible, whether they be the normal situation or random event situations. The French waste management strategy is based upon two types of disposal depending on the nature and activity of waste packages: - surface disposal intended for low and medium level wastes having half-lives of about 30 years or less and alpha activity less than 3.7 MBq/kg (0.1 Ci/t), for individual packages and less than 0.37 MBq/kg (0.01 Ci/t) in the average. Deep geological disposal intended for TRU and high level wastes. The conditions of acceptance of packages in a surface disposal site are subject to the two fundamental safety rules no. I.2 and III.2.e. The present paper is only dealing with deep geological disposal. For deep geological repositories, three stages are involved: stage preceding definitive disposal (intermediate storage, transportation, handling, setting up in the disposal cavities); stage subsequent to definitive sealing of the disposal cavities but prior to the end of operation of the repository; stage subsequent to closure of the repository. The role of the geological barrier has been determined as the essential part of long term radioactivity confinement, by a working group, set up by the French safety authorities. Essential technical criteria relating to the choice of a site so defined by this group, are the following: very low permeability

Evaluation of static analysis tools used to assess software important to nuclear power plant safety

Energy Technology Data Exchange (ETDEWEB)

Ourghanlian, Alain [EDF Lab CHATOU, Simulation and Information Technologies for Power Generation Systems Department, EDF R and D, Cedex (France)

2015-03-15

We describe a comparative analysis of different tools used to assess safety-critical software used in nuclear power plants. To enhance the credibility of safety assessments and to optimize safety justification costs, Electricit e de France (EDF) investigates the use of methods and tools for source code semantic analysis, to obtain indisputable evidence and help assessors focus on the most critical issues. EDF has been using the PolySpace tool for more than 10 years. Currently, new industrial tools based on the same formal approach, Abstract Interpretation, are available. Practical experimentation with these new tools shows that the precision obtained on one of our shutdown systems software packages is substantially improved. In the first part of this article, we present the analysis principles of the tools used in our experimentation. In the second part, we present the main characteristics of protection-system software, and why these characteristics are well adapted for the new analysis tools.

Authorised radioactive contents of packages: how to maximise simultaneously safety and flexibility?

Energy Technology Data Exchange (ETDEWEB)

Malesys, P. [COGEMA Logistics (AREVA Group), Saint-Quentin-en-Yvelines (France); Field, G. [Packaging Technology, Inc. (AREVA Group), Tacoma (United States)

2004-07-01

Packages for the transport of radioactive material are required to comply with national and / or international regulations. These regulations are widely based on the requirements set forth by the International Atomic Energy Agency (IAEA) in the ''Regulations for the Safe Transport of Radioactive Material''. ''The objective of these Regulations is to protect persons, property and the environment from the effects of radiation during the transport of radioactive material. This protection is achieved by requiring: (a) containment of the radioactive contents; (b) control of external radiation levels; (c) prevention of criticality; and (d) prevention of damage caused by heat.'' Some of the package designs require an approval by the applicable Competent Authority(ies). The application for package approval includes ''a specification of the authorised radioactive content, including any restrictions on the radioactive contents which might not be obvious from the nature of the packaging''. The required ''specification of the authorised radioactive content'' must include assurance that the material that is actually transported meets the four above-mentioned requirements. For that purpose, the choice of the parameters of the specification is crucial: they should be of such a nature that they can be checked before shipment to allow demonstration of compliance with the above-mentioned requirements. From our experience, it is appropriate to consider parameters that have the most direct influence on the compliance with the requirements. It is worthwhile to consider, for instance and when possible, direct measurement for the assessment of compliance with a requirement. A typical example is measurement of radiation exposure: it can provide a confident level of verification of the external radiation level, regardless of other parameters such as the activity. The paper presents our experience, particularly

Authorised radioactive contents of packages: how to maximise simultaneously safety and flexibility?

International Nuclear Information System (INIS)

Malesys, P.; Field, G.

2004-01-01

Packages for the transport of radioactive material are required to comply with national and / or international regulations. These regulations are widely based on the requirements set forth by the International Atomic Energy Agency (IAEA) in the ''Regulations for the Safe Transport of Radioactive Material''. ''The objective of these Regulations is to protect persons, property and the environment from the effects of radiation during the transport of radioactive material. This protection is achieved by requiring: (a) containment of the radioactive contents; (b) control of external radiation levels; (c) prevention of criticality; and (d) prevention of damage caused by heat.'' Some of the package designs require an approval by the applicable Competent Authority(ies). The application for package approval includes ''a specification of the authorised radioactive content, including any restrictions on the radioactive contents which might not be obvious from the nature of the packaging''. The required ''specification of the authorised radioactive content'' must include assurance that the material that is actually transported meets the four above-mentioned requirements. For that purpose, the choice of the parameters of the specification is crucial: they should be of such a nature that they can be checked before shipment to allow demonstration of compliance with the above-mentioned requirements. From our experience, it is appropriate to consider parameters that have the most direct influence on the compliance with the requirements. It is worthwhile to consider, for instance and when possible, direct measurement for the assessment of compliance with a requirement. A typical example is measurement of radiation exposure: it can provide a confident level of verification of the external radiation level, regardless of other parameters such as the activity. The paper presents our experience, particularly in France and in the USA, to demonstrate that it is possible to adequately

Reliability test and failure analysis of high power LED packages

International Nuclear Information System (INIS)

Chen Zhaohui; Zhang Qin; Wang Kai; Luo Xiaobing; Liu Sheng

2011-01-01

A new type application specific light emitting diode (LED) package (ASLP) with freeform polycarbonate lens for street lighting is developed, whose manufacturing processes are compatible with a typical LED packaging process. The reliability test methods and failure criterions from different vendors are reviewed and compared. It is found that test methods and failure criterions are quite different. The rapid reliability assessment standards are urgently needed for the LED industry. 85 0 C/85 RH with 700 mA is used to test our LED modules with three other vendors for 1000 h, showing no visible degradation in optical performance for our modules, with two other vendors showing significant degradation. Some failure analysis methods such as C-SAM, Nano X-ray CT and optical microscope are used for LED packages. Some failure mechanisms such as delaminations and cracks are detected in the LED packages after the accelerated reliability testing. The finite element simulation method is helpful for the failure analysis and design of the reliability of the LED packaging. One example is used to show one currently used module in industry is vulnerable and may not easily pass the harsh thermal cycle testing. (semiconductor devices)

Larch: An Analysis Package for XAFS and Related Spectroscopies

International Nuclear Information System (INIS)

Newville, Matthew

2013-01-01

LARCH, a package of analysis tools for XAFS and related spectroscopies is presented. A complete rewrite of the ifeffit package, the initial release of larch preserves the core XAFS analysis procedures such as normalization, background subtraction, Fourier transforms, fitting of XANES spectra, and fitting of experimental spectra to a sum of feff Paths, with few algorithmic changes made in comparison to IFEFFIT. LARCH is written using Python and its packages for scientific programming, which gives significant improvements over IFEFFIT in the ability to handle multi-dimensional and large data sets, write complex analysis scripts, visualize data, add new functionality, and customize existing capabilities. Like the earlier version, larch can run from an interactive command line or in batch-mode, but larch can also be run as a server and accessed from clients using standard inter-process communication techniques available in a variety of computer languages. larch is freely available under an open source license. Examples of using larch are shown, future directions for development are discussed, and collaborations for adding new capabilities are actively sought.

Evaluation of low and intermediate level radioactive solidified waste forms and packages

International Nuclear Information System (INIS)

1990-10-01

Evaluation of low and intermediate level radioactive waste forms and packages with respect to compliance with quality and safety requirements for transport, interim storage and disposal has become a very important part of the radioactive waste management strategy in many countries. The evaluation of waste forms and packages provides precise basic data for regulatory bodies to establish safety requirements, and implement quality control and quality assurance procedures for radioactive waste management programmes. The requirements depend very much upon the disposal option selected, treatment technology used, waste form characteristics, package quality and other factors. The regulatory requirements can also influence the methodology of waste form/package evaluation together with selection and analysis of data for quality control and safety assurance. A coordinated research programme started at the end of 1985 and brought together 12 participants from 11 countries. The results of the programme and each particular project were discussed at three Research Coordination Meetings held in Cairo, Egypt, in May, 1986; in Beijing, China, in April, 1998; and at Harwell Laboratory, United Kingdom, in November, 1989. This document summarises the salient features and results achieved during the four year investigation and a recommendation for future work in this area. Refs, figs and tabs

Performance analysis of conceptual waste package designs in salt repositories

International Nuclear Information System (INIS)

Jansen, G. Jr.; Raines, G.E.; Kircher, J.F.

1984-01-01

A performance analysis of commercial high-level waste and spent fuel conceptual package designs in reference repositories in three salt formations was conducted with the WAPPA waste package code. Expected conditions for temperature, stress, brine composition, radiation level, and brine flow rate were used as boundary conditions to compute expected corrosion of a thick-walled overpack of 1025 wrought steel. In all salt formations corrosion by low Mg salt-dissolution brines typical of intrusion scenarios was too slow to cause the package to fail for thousands of years after burial. In high Mg brines judged typical of thermally migrating brines in bedded salt formations, corrosion rates which would otherwise have caused the packages to fail within a few hundred years were limited by brine availability. All of the brine reaching the package was consumed by reaction with the iron in the overpack, thus preventing further corrosion. Uniform brine distribution over the package surface was an important factor in predicting long package lifetimes for the high Mg brines. 14 references, 15 figures

CH Packaging Program Guidance

International Nuclear Information System (INIS)

2006-01-01

The purpose of this document is to provide the technical requirements for preparation for use, operation, inspection, and maintenance of a Transuranic Package Transporter Model II (TRUPACT-II), a HalfPACT shipping package, and directly related components. This document complies with the minimum requirements as specified in the TRUPACT-II Safety Analysis Report for Packaging (SARP), HalfPACT SARP, and U.S. Nuclear Regulatory Commission (NRC) Certificates of Compliance (C of C) 9218 and 9279, respectively. In the event of a conflict between this document and the SARP or C of C, the C of C shall govern. The C of Cs state: 'each package must be prepared for shipment and operated in accordance with the procedures described in Chapter 7.0, Operating Procedures, of the application.' They further state: 'each package must be tested and maintained in accordance with the procedures described in Chapter 8.0, Acceptance Tests and Maintenance Program of the Application.' Chapter 9.0 of the SARP charges the U.S. Department of Energy (DOE) or the Waste Isolation Pilot Plant| (WIPP) management and operating (M and O) contractor with assuring packaging is used in accordance with the requirements of the C of C. Because the packaging is NRC-approved, users need to be familiar with Title 10 Code of Federal Regulations(CFR) 71.8. Any time a user suspects or has indications that the conditions of approval in the C of C were not met, the Carlsbad Field Office (CBFO) shall be notified immediately. The CBFO will evaluate the issue and notify the NRC if required.In accordance with 10 CFR Part 71, certificate holders, packaging users, and contractors or subcontractors who use, design, fabricate, test, maintain, or modify the packaging shall post copies of (1) 10 CFR Part 21 regulations, (2) Section 206 of the Energy Reorganization Act of 1974, and (3) NRC Form 3, Notice to Employees. These documents must be posted in a conspicuous location where the activities subject to these regulations

CH Packaging Program Guidance

International Nuclear Information System (INIS)

2007-01-01

The purpose of this document is to provide the technical requirements for preparation for use, operation, inspection, and maintenance of a Transuranic Package Transporter Model II (TRUPACT-II), a HalfPACT shipping package, and directly related components. This document complies with the minimum requirements as specified in the TRUPACT-II Safety Analysis Report for Packaging (SARP), HalfPACT SARP, and U.S. Nuclear Regulatory Commission (NRC) Certificates of Compliance (C of C) 9218 and 9279, respectively. In the event of a conflict between this document and the SARP or C of C, the C of C shall govern. The C of Cs state: 'each package must be prepared for shipment and operated in accordance with the procedures described in Chapter 7.0, Operating Procedures, of the application.' They further state: 'each package must be tested and maintained in accordance with the procedures described in Chapter 8.0, Acceptance Tests and Maintenance Program of the Application.' Chapter 9.0 of the SARP charges the U.S. Department of Energy (DOE) or the Waste Isolation Pilot Plant (WIPP) management and operating (M and O) contractor with assuring packaging is used in accordance with the requirements of the C of C. Because the packaging is NRC-approved, users need to be familiar with Title 10 Code of Federal Regulations (CFR) 71.8. Any time a user suspects or has indications that the conditions of approval in the C of C were not met, the Carlsbad Field Office (CBFO) shall be notified immediately. The CBFO will evaluate the issue and notify the NRC if required.In accordance with 10 CFR Part 71, certificate holders, packaging users, and contractors or subcontractors who use, design, fabricate, test, maintain, or modify the packaging shall post copies of (1) 10 CFR Part 21 regulations, (2) Section 206 of the Energy Reorganization Act of 1974, and (3) NRC Form 3, Notice to Employees. These documents must be posted in a conspicuous location where the activities subject to these regulations

Procurement strategic analysis of nuclear safety equipment

International Nuclear Information System (INIS)

Wu Caixia; Yang Haifeng; Li Xiaoyang; Li Shixin

2013-01-01

The nuclear power development plan in China puts forward a challenge on procurement of nuclear safety equipment. Based on the characteristics of the procurement of nuclear safety equipment, requirements are raised for procurement process, including further clarification of equipment technical specification, establishment and improvement of the expert database of the nuclear power industry, adoption of more reasonable evaluation method and establishment of a unified platform for nuclear power plants to procure nuclear safety equipment. This paper makes recommendation of procurement strategy for nuclear power production enterprises from following aspects, making a plan of procurement progress, dividing procurement packages rationally, establishing supplier database through qualification review and implementing classified management, promoting localization process of key equipment continually and further improving the system and mechanism of procurement of nuclear safety equipment. (authors)

Flammability Analysis For Actinide Oxides Packaged In 9975 Shipping Containers

Energy Technology Data Exchange (ETDEWEB)

Laurinat, James E.; Askew, Neal M.; Hensel, Steve J.

2013-03-21

Packaging options are evaluated for compliance with safety requirements for shipment of mixed actinide oxides packaged in a 9975 Primary Containment Vessel (PCV). Radiolytic gas generation rates, PCV internal gas pressures, and shipping windows (times to reach unacceptable gas compositions or pressures after closure of the PCV) are calculated for shipment of a 9975 PCV containing a plastic bottle filled with plutonium and uranium oxides with a selected isotopic composition. G-values for radiolytic hydrogen generation from adsorbed moisture are estimated from the results of gas generation tests for plutonium oxide and uranium oxide doped with curium-244. The radiolytic generation of hydrogen from the plastic bottle is calculated using a geometric model for alpha particle deposition in the bottle wall. The temperature of the PCV during shipment is estimated from the results of finite element heat transfer analyses.

16 CFR 1702.13 - Labeling and packaging samples.

Science.gov (United States)

2010-01-01

... 16 Commercial Practices 2 2010-01-01 2010-01-01 false Labeling and packaging samples. 1702.13 Section 1702.13 Commercial Practices CONSUMER PRODUCT SAFETY COMMISSION POISON PREVENTION PACKAGING ACT OF 1970 REGULATIONS PETITIONS FOR EXEMPTIONS FROM POISON PREVENTION PACKAGING ACT REQUIREMENTS; PETITION...

Linking Safety Analysis to Safety Requirements

DEFF Research Database (Denmark)

Hansen, Kirsten Mark

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

Safety analysis for research reactors

International Nuclear Information System (INIS)

2008-01-01

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

DMAIC Application and Fault Analysis of Metal Packaging in the Canning Industry

Directory of Open Access Journals (Sweden)

Rafael Santos de Souza

2017-11-01

Full Text Available In the food industry, quality assurance is strongly associated with consumer safety and the risk of compromising the purity of products. Specifically, in the canning industries, rusty, dented or bulging cans may contain bacteria and cause foodborne illness. This article discusses the application of DMAIC and failure analysis to reduce the number of crushed cans in a canning industry. The methodology was of an applied nature, with an exploratory objective, a qualitative and quantitative approach and a case study method used. Through statistical analysis, it was found that the amount of dented packaging in the production process, which comprises the steps of filling, seaming and basketing, was high. The step with the highest number of crushed cans between three stages was seaming. This was followed by the application of FMEA failure analysis related to the use of an automated seaming machine. Finally, an action plan was roposed to solve the problem of cans crushed in the seaming step and the results demonstrate the improvements obtained before the target after 4 months (August-November 2015.

Analysis of counting data: Development of the SATLAS Python package

Science.gov (United States)

Gins, W.; de Groote, R. P.; Bissell, M. L.; Granados Buitrago, C.; Ferrer, R.; Lynch, K. M.; Neyens, G.; Sels, S.

2018-01-01

For the analysis of low-statistics counting experiments, a traditional nonlinear least squares minimization routine may not always provide correct parameter and uncertainty estimates due to the assumptions inherent in the algorithm(s). In response to this, a user-friendly Python package (SATLAS) was written to provide an easy interface between the data and a variety of minimization algorithms which are suited for analyzinglow, as well as high, statistics data. The advantage of this package is that it allows the user to define their own model function and then compare different minimization routines to determine the optimal parameter values and their respective (correlated) errors. Experimental validation of the different approaches in the package is done through analysis of hyperfine structure data of 203Fr gathered by the CRIS experiment at ISOLDE, CERN.

Antimicrobial packaging for fresh-cut fruits

Science.gov (United States)

Fresh-cut fruits are minimally processed produce which are consumed directly at their fresh stage without any further kill step. Microbiological quality and safety are major challenges to fresh-cut fruits. Antimicrobial packaging is one of the innovative food packaging systems that is able to kill o...

Isotopic analysis of radioactive waste packages (an inexpensive approach)

International Nuclear Information System (INIS)

Padula, D.A.; Richmond, J.S.

1983-01-01

A computer printout of the isotopic analysis for all radioactive waste packages containing resins, or other aqueous filter media is now required at the disposal sites at Barnwell, South Carolina, and Beatty, Nevada. Richland, Washington requires an isotopic analysis for all radioactive waste packages. The NRC (Nuclear Regulatory Commission), through 10 CFR 61, will require shippers of radioactive waste to classify and label for disposal all radioactive waste forms. These forms include resins, filters, sludges, and dry active waste (trash). The waste classification is to be based upon 10 CFR 61 (Section 1-7). The isotopes upon which waste classification is to be based are tabulated. 7 references, 8 tables

RECENT TRENDS IN PACKAGING SYSTEMS FOR PHARMACEUTICAL PRODUCTS

Directory of Open Access Journals (Sweden)

Renata Dobrucka

2014-12-01

Full Text Available Background:  In recent years, pharmaceutical packaging market was one of the fastest growing areas of the packaging industry. At the same time the packaging manufacturers put high demands on quality and safety. Methods: Review of innovations in packaging systems for pharmaceutical products was made including newest information of researches and achievements of recent years. Results and conclusion: Observed in recent years the development of pharmaceutical packaging market expanded due to with the huge technological advances that allow introduction of new packaging. Also, in this study presented intelligent packaging in pharmacy and innovation in child-resistance packaging.

Advanced Manufacturing Systems in Food Processing and Packaging Industry

International Nuclear Information System (INIS)

Sani, Mohd Shafie; Aziz, Faieza Abdul

2013-01-01

In this paper, several advanced manufacturing systems in food processing and packaging industry are reviewed, including: biodegradable smart packaging and Nano composites, advanced automation control system consists of fieldbus technology, distributed control system and food safety inspection features. The main purpose of current technology in food processing and packaging industry is discussed due to major concern on efficiency of the plant process, productivity, quality, as well as safety. These application were chosen because they are robust, flexible, reconfigurable, preserve the quality of the food, and efficient.

Advanced Manufacturing Systems in Food Processing and Packaging Industry

Science.gov (United States)

Shafie Sani, Mohd; Aziz, Faieza Abdul

2013-06-01

In this paper, several advanced manufacturing systems in food processing and packaging industry are reviewed, including: biodegradable smart packaging and Nano composites, advanced automation control system consists of fieldbus technology, distributed control system and food safety inspection features. The main purpose of current technology in food processing and packaging industry is discussed due to major concern on efficiency of the plant process, productivity, quality, as well as safety. These application were chosen because they are robust, flexible, reconfigurable, preserve the quality of the food, and efficient.

Generic Degraded Configuration Probability Analysis for the Codisposal Waste Package

International Nuclear Information System (INIS)

S.F.A. Deng; M. Saglam; L.J. Gratton

2001-01-01

In accordance with the technical work plan, ''Technical Work Plan For: Department of Energy Spent Nuclear Fuel Work Packages'' (CRWMS M and O 2000c), this Analysis/Model Report (AMR) is developed for the purpose of screening out degraded configurations for U.S. Department of Energy (DOE) spent nuclear fuel (SNF) types. It performs the degraded configuration parameter and probability evaluations of the overall methodology specified in the ''Disposal Criticality Analysis Methodology Topical Report'' (YMP 2000, Section 3) to qualifying configurations. Degradation analyses are performed to assess realizable parameter ranges and physical regimes for configurations. Probability calculations are then performed for configurations characterized by k eff in excess of the Critical Limit (CL). The scope of this document is to develop a generic set of screening criteria or models to screen out degraded configurations having potential for exceeding a criticality limit. The developed screening criteria include arguments based on physical/chemical processes and probability calculations and apply to DOE SNF types when codisposed with the high-level waste (HLW) glass inside a waste package. The degradation takes place inside the waste package and is long after repository licensing has expired. The emphasis of this AMR is on degraded configuration screening and the probability analysis is one of the approaches used for screening. The intended use of the model is to apply the developed screening criteria to each DOE SNF type following the completion of the degraded mode criticality analysis internal to the waste package

Atmospheric Data Package for the Composite Analysis

Energy Technology Data Exchange (ETDEWEB)

Napier, Bruce A.; Ramsdell, James V.

2005-09-01

The purpose of this data package is to summarize our conceptual understanding of atmospheric transport and deposition, describe how this understanding will be simplified for numerical simulation as part of the Composite Analysis (i.e., implementation model), and finally to provide the input parameters needed for the simulations.

A new package: MySAS for small angle scattering data analysis

International Nuclear Information System (INIS)

Huang Chaoqiang; Xia Qingzhong; Yan Guanyun; Sun Guang'ai; Chen Bo

2010-01-01

In this paper, A MySAS package, which is verified on Windows XP, can easily convert two-dimensional data in small angle neutron and X-ray scattering analysis, operate individually and execute one particular operation as numerical data reduction or analysis, and graphical visualization. This MySAS package can implement the input and output routines via scanning certain properties, thus recalling completely sets of repetition input and selecting the input files. On starting from the two-dimensional files, the MySAS package can correct the anisotropic or isotropic data for physical interpretation and select the relevant pixels. Over 50 model functions are fitted by the POWELL code using χ 2 as the figure of merit function. (authors)

Packaging- and transportation-related occurrence reports. Fiscal year 1996 annual report

International Nuclear Information System (INIS)

Dickerson, L.S.; Welch, M.J.

1997-02-01

The Oak Ridge National Laboratory (ORNL), through its support to the US Department of Energy's (DOE's) Office of Transportation, Emergency Management, and Analytical Services (EM-76), retrieves reports and information pertaining to transportation and packaging occurrences from the centralized Occurrence Reporting and Processing System (ORPS) database. These selected reports are analyzed for trends, impact on packaging and transportation operations and safety concerns, and lessons learned (LL) in transportation and packaging safety. Some selected reports are reviewed to evaluate the corrective actions being conducted. This report contains an analysis of 246 occurrences identified as packaging- or transportation-related during fiscal year (FY) 1996, with supporting data from calendar year (CY) 1991 through 1995 which provide the basis for trending. The overall number of packaging- and transportation-related occurrences remains a small percentage of the total occurrences in the DOE system, through it is relatively higher this year (∼6%) than previous years when transportation occurrences were approximately 3% of the total. The decrease in the total number of occurrences may be the result of the rollup provisions of the new DOE Order 232.1, and the comparative increase in packaging- and transportation-related occurrence reports (ORs) is only a reflection of the decrease in the overall total. There does not appear to be a correlation between the total number of offsite hazardous materials shipments and the number of reported occurrences. The offsite occurrences, while few in number, are consistent for the major shippers and contractors

INTRA graphical package - INTRA-Graph 1.0

International Nuclear Information System (INIS)

Hofman, D.; Edlund, O.

2001-04-01

INTRA-Graph 1.0 has been developed at Studsvik Eco and Safety AB in the frame of the European Fusion Technology Programme for application in the safety analysis using the INTRA code. INTRA-Graph 1.0 is a graphical package producing 2-dimensional plots of results generated by the INTRA code. INTRA-Graph 1.0 has been developed by extending the Grace package source code, distributed under the terms of GNU General Public License. The changes in the Grace source files are limited to provide easy updates of the INTRA-Graph when a new version of Grace will be released. The INTRA-related functionality has been implemented in new source files. The present report describes and gives complete listing of these files. The changes in the Grace source files are also described and the listing of the changed parts of the files is presented. The report gives detailed explanations and examples of files required for installation and configuration of INTRA-Graph on the different types of Unix workstations

WASTE PACKAGE TRANSPORTER DESIGN

International Nuclear Information System (INIS)

Weddle, D.C.; Novotny, R.; Cron, J.

1998-01-01

The purpose of this Design Analysis is to develop preliminary design of the waste package transporter used for waste package (WP) transport and related functions in the subsurface repository. This analysis refines the conceptual design that was started in Phase I of the Viability Assessment. This analysis supports the development of a reliable emplacement concept and a retrieval concept for license application design. The scope of this analysis includes the following activities: (1) Assess features of the transporter design and evaluate alternative design solutions for mechanical components. (2) Develop mechanical equipment details for the transporter. (3) Prepare a preliminary structural evaluation for the transporter. (4) Identify and recommend the equipment design for waste package transport and related functions. (5) Investigate transport equipment interface tolerances. This analysis supports the development of the waste package transporter for the transport, emplacement, and retrieval of packaged radioactive waste forms in the subsurface repository. Once the waste containers are closed and accepted, the packaged radioactive waste forms are termed waste packages (WP). This terminology was finalized as this analysis neared completion; therefore, the term disposal container is used in several references (i.e., the System Description Document (SDD)) (Ref. 5.6). In this analysis and the applicable reference documents, the term ''disposal container'' is synonymous with ''waste package''

Prediction of Heat Transfer Performance on Horizontal U-Shaped Heat Exchanger in Passive Safety System Using MARS

Energy Technology Data Exchange (ETDEWEB)

Jeon, Seong-Su; Hong, Soon-Joon [FNC Tech, Yongin (Korea, Republic of); Cho, Hyoung-Kyu; Park, Goon-Cherl [Seoul National University, Seoul (Korea, Republic of)

2015-10-15

The design and the safety analysis of the passive safety systems are performed mainly using the best-estimate thermal-hydraulic analysis codes such as RELAP5 and MARS. This study developed the heat transfer model package for the horizontal U-shaped HX submerged in a pool by improving the horizontal in-tube condensation model and developing the outside-tube natural convective nucleate boiling model. This paper presents the HX model package and the validation results against the passive safety system-related experimental data of PASCAL and ATLAS-PAFS. This study developed the heat transfer model package of the horizontal U-shaped HX submerged in a pool in order to obtain a reliable prediction of the HX heat removal performance of the passive safety system, especially PAFS, using MARS. From the validation results, the proposed model package provided the improved prediction of HX performance (condensation, natural convective nucleate boiling, and heat removal rate of the HX) compared to the default model in MARS.

RepoSTAR. A Code package for control and evaluation of statistical calculations with the program package RepoTREND; RepoSTAR. Ein Codepaket zur Steuerung und Auswertung statistischer Rechenlaeufe mit dem Programmpaket RepoTREND

Energy Technology Data Exchange (ETDEWEB)

Becker, Dirk-Alexander

2016-05-15

The program package RepoTREND for the integrated long terms safety analysis of final repositories allows besides deterministic studies of defined problems also statistical or probabilistic analyses. Probabilistic uncertainty and sensitivity analyses are realized in the program package repoTREND by a specific statistic frame called RepoSTAR. The report covers the following issues: concept, sampling and data supply of single simulations, evaluation of statistical calculations with the program RepoSUN.

Implicit Communication of Food Product Healthfulness through Package Design: A Content Analysis

DEFF Research Database (Denmark)

Festila, Alexandra; Chrysochou, Polymeros

2018-01-01

of twelve food product categories across two countries (Denmark and USA), our findings indicate that: (a) implicit package design elements (colors, imagery, material, shape) differ between health-positioned and regular products, and (b) these differences are product category-specific rather than universal......How do food companies use package design to communicate healthfulness? The present study addresses this question by investigating the most typical implicit package design elements used by food companies for their health-positioned food products. Using a content analysis on the packaging design...

Quality assurance guidance for TRUPACT-II [Transuranic Package Transporter-II] payload control

International Nuclear Information System (INIS)

1989-10-01

The Transuranic Package Transporter-II (TRUPACT-II) Safety Analysis Report for Packaging (SARP) approved by the Nuclear Regulatory Commission (NRC), discusses authorized methods for payload control in Appendix 1.3.7 and the Quality Assurance (QA) requirements in Section 9.3. Subsection 9.3.2.1 covers maintenance and use of the TRUPACT-II and the specific QA requirements are given in DOE/WIPP 89-012. Subsection 9.3.2.2 covers payload compliance, for which this document was written. 6 refs

Teaching preschool children to report suspicious packages to adults.

Science.gov (United States)

May, Michael E; Shayter, Ashley M; Schmick, Ayla; Barron, Becky; Doherty, Meghan; Johnson, Matthew

2018-05-16

Law enforcement agencies stress that public reporting of terror-related crime is the predominant means for disrupting these actions. However, schools may be unprepared because the majority of the populace may not understand the threat of suspicious materials or what to do when they are found on school grounds. The purpose of this study was to systematically teach preschool children to identify and report suspicious packages across three experiments. In the first experiment, we used multiple exemplar training to teach children to identify the characteristics of safe and unsafe packages. In the second experiment, we taught participants to identify the locations where packages should be considered unsafe. Finally, in the third experiment, we used behavioral skills training to teach participants to avoid touching unsafe packages, leave the area where they were located, and report their discovery to an adult. Results suggest the participants quickly developed these skills. Implications for safety skills in young school children are discussed. © 2018 Society for the Experimental Analysis of Behavior.

Advances in Packaging Methods, Processes and Systems

Directory of Open Access Journals (Sweden)

Nitaigour Premchand Mahalik

2014-10-01

Full Text Available The food processing and packaging industry is becoming a multi-trillion dollar global business. The reason is that the recent increase in incomes in traditionally less economically developed countries has led to a rise in standards of living that includes a significantly higher consumption of packaged foods. As a result, food safety guidelines have been more stringent than ever. At the same time, the number of research and educational institutions—that is, the number of potential researchers and stakeholders—has increased in the recent past. This paper reviews recent developments in food processing and packaging (FPP, keeping in view the aforementioned advancements and bearing in mind that FPP is an interdisciplinary area in that materials, safety, systems, regulation, and supply chains play vital roles. In particular, the review covers processing and packaging principles, standards, interfaces, techniques, methods, and state-of-the-art technologies that are currently in use or in development. Recent advances such as smart packaging, non-destructive inspection methods, printing techniques, application of robotics and machineries, automation architecture, software systems and interfaces are reviewed.

Development of Specifications for Radioactive Waste Packages

International Nuclear Information System (INIS)

2006-10-01

The main objective of this publication is to provide guidelines for the development of waste package specifications that comply with waste acceptance requirements for storage and disposal of radioactive waste. It will assist waste generators and waste package producers in selecting the most significant parameters and in developing and implementing specifications for each individual type of waste and waste package. This publication also identifies and reviews the activities and technical provisions that are necessary to meet safety requirements; in particular, selection of the significant safety parameters and preparation of specifications for waste forms, waste containers and waste packages using proven approaches, methods and technologies. This report provides guidance using a systematic, stepwise approach, integrating the technical, organizational and administrative factors that need to be considered at each step of planning and implementing waste package design, fabrication, approval, quality assurance and control. The report reflects the considerable experience and knowledge that has been accumulated in the IAEA Member States and is consistent with the current international requirements, principles, standards and guidance for the safe management of radioactive waste

Development of Specifications for Radioactive Waste Packages

Energy Technology Data Exchange (ETDEWEB)

NONE

2006-10-15

The main objective of this publication is to provide guidelines for the development of waste package specifications that comply with waste acceptance requirements for storage and disposal of radioactive waste. It will assist waste generators and waste package producers in selecting the most significant parameters and in developing and implementing specifications for each individual type of waste and waste package. This publication also identifies and reviews the activities and technical provisions that are necessary to meet safety requirements; in particular, selection of the significant safety parameters and preparation of specifications for waste forms, waste containers and waste packages using proven approaches, methods and technologies. This report provides guidance using a systematic, stepwise approach, integrating the technical, organizational and administrative factors that need to be considered at each step of planning and implementing waste package design, fabrication, approval, quality assurance and control. The report reflects the considerable experience and knowledge that has been accumulated in the IAEA Member States and is consistent with the current international requirements, principles, standards and guidance for the safe management of radioactive waste.

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-05-15

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

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

International Nuclear Information System (INIS)

Kim, Yun Goo; Oh, Eung Se

2016-01-01

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

Generic Degraded Congiguration Probability Analysis for DOE Codisposal Waste Package

Energy Technology Data Exchange (ETDEWEB)

S.F.A. Deng; M. Saglam; L.J. Gratton

2001-05-23

In accordance with the technical work plan, ''Technical Work Plan For: Department of Energy Spent Nuclear Fuel Work Packages'' (CRWMS M&O 2000c), this Analysis/Model Report (AMR) is developed for the purpose of screening out degraded configurations for U.S. Department of Energy (DOE) spent nuclear fuel (SNF) types. It performs the degraded configuration parameter and probability evaluations of the overall methodology specified in the ''Disposal Criticality Analysis Methodology Topical Report'' (YMP 2000, Section 3) to qualifying configurations. Degradation analyses are performed to assess realizable parameter ranges and physical regimes for configurations. Probability calculations are then performed for configurations characterized by k{sub eff} in excess of the Critical Limit (CL). The scope of this document is to develop a generic set of screening criteria or models to screen out degraded configurations having potential for exceeding a criticality limit. The developed screening criteria include arguments based on physical/chemical processes and probability calculations and apply to DOE SNF types when codisposed with the high-level waste (HLW) glass inside a waste package. The degradation takes place inside the waste package and is long after repository licensing has expired. The emphasis of this AMR is on degraded configuration screening and the probability analysis is one of the approaches used for screening. The intended use of the model is to apply the developed screening criteria to each DOE SNF type following the completion of the degraded mode criticality analysis internal to the waste package.

Overview of the DOE-EM Packaging Certification Program

International Nuclear Information System (INIS)

Feldman, M.R.; Bennett, M.E.; Shuler, J.M.

2009-01-01

The U.S. Department of Transportation, in 49 CFR 173.7(d) grants the U.S. Department of Energy (DOE) the power to use 'packagings made by or under the direction of the U.S. Department of Energy... for the transportation of Class 7 materials when evaluated, approved and certified by the Department of Energy against packaging standards equivalent to those specified in 10 CFR part 71'. Via DOE Order 460.1B, DOE has established the DOE Packaging Certification Program (PCP) within the Department of Environmental Management for purposes including the certification of radioactive materials packages for DOE use. This paper will provide an overview of the programs and activities currently undertaken by the PCP in support of the safe transport of radioactive materials, including technical review of Safety Analysis Reports for Packaging, development of guidance documents and training courses, a quality assurance audit and field assessment program, database and docket management, and testing and test methodology development. The paper will also highlight the various organizations currently utilized by the PCP to meet the requirements of DOE O 460.1B, as well as some creative and effective methods that are being used to meet program objectives. The DOE Package Certification Program's primary function is to perform technical reviews of SARPs in support of the packaging certification process to ensure that the maximum protection is afforded to the public, all federal regulations are met, and the process is as time-effective and cost-effective as possible. Five additional specific functions are also supported by the PCP: development of guidance documents, training courses, a QA audit and field assessment program, database and docket management, and testing methods development. Each of these functions individually contributes to the overall mission of the PCP as defined in DOE O 460.1B. Taken as a whole, these functions represent a robust program to ensure the safety of workers

Transportation package design using numerical optimization

International Nuclear Information System (INIS)

Harding, D.C.; Witkowski, W.R.

1991-01-01

The purpose of this overview is twofold: first, to outline the theory and basic elements of numerical optimization; and second, to show how numerical optimization can be applied to the transportation packaging industry and used to increase efficiency and safety of radioactive and hazardous material transportation packages. A more extensive review of numerical optimization and its applications to radioactive material transportation package design was performed previously by the authors (Witkowski and Harding 1992). A proof-of-concept Type B package design is also presented as a simplified example of potential improvements achievable using numerical optimization in the design process

Towards an Industrial Application of Statistical Uncertainty Analysis Methods to Multi-physical Modelling and Safety Analyses

International Nuclear Information System (INIS)

Zhang, Jinzhao; Segurado, Jacobo; Schneidesch, Christophe

2013-01-01

Since 1980's, Tractebel Engineering (TE) has being developed and applied a multi-physical modelling and safety analyses capability, based on a code package consisting of the best estimate 3D neutronic (PANTHER), system thermal hydraulic (RELAP5), core sub-channel thermal hydraulic (COBRA-3C), and fuel thermal mechanic (FRAPCON/FRAPTRAN) codes. A series of methodologies have been developed to perform and to license the reactor safety analysis and core reload design, based on the deterministic bounding approach. Following the recent trends in research and development as well as in industrial applications, TE has been working since 2010 towards the application of the statistical sensitivity and uncertainty analysis methods to the multi-physical modelling and licensing safety analyses. In this paper, the TE multi-physical modelling and safety analyses capability is first described, followed by the proposed TE best estimate plus statistical uncertainty analysis method (BESUAM). The chosen statistical sensitivity and uncertainty analysis methods (non-parametric order statistic method or bootstrap) and tool (DAKOTA) are then presented, followed by some preliminary results of their applications to FRAPCON/FRAPTRAN simulation of OECD RIA fuel rod codes benchmark and RELAP5/MOD3.3 simulation of THTF tests. (authors)

WASTE PACKAGE TRANSPORTER DESIGN

Energy Technology Data Exchange (ETDEWEB)

D.C. Weddle; R. Novotny; J. Cron

1998-09-23

The purpose of this Design Analysis is to develop preliminary design of the waste package transporter used for waste package (WP) transport and related functions in the subsurface repository. This analysis refines the conceptual design that was started in Phase I of the Viability Assessment. This analysis supports the development of a reliable emplacement concept and a retrieval concept for license application design. The scope of this analysis includes the following activities: (1) Assess features of the transporter design and evaluate alternative design solutions for mechanical components. (2) Develop mechanical equipment details for the transporter. (3) Prepare a preliminary structural evaluation for the transporter. (4) Identify and recommend the equipment design for waste package transport and related functions. (5) Investigate transport equipment interface tolerances. This analysis supports the development of the waste package transporter for the transport, emplacement, and retrieval of packaged radioactive waste forms in the subsurface repository. Once the waste containers are closed and accepted, the packaged radioactive waste forms are termed waste packages (WP). This terminology was finalized as this analysis neared completion; therefore, the term disposal container is used in several references (i.e., the System Description Document (SDD)) (Ref. 5.6). In this analysis and the applicable reference documents, the term ''disposal container'' is synonymous with ''waste package''.

Quality assurance requirements for packaging and transportation of radioactive materials

International Nuclear Information System (INIS)

Barker, R.F.; MacDonald, C.E.; Doda, R.J.

1978-01-01

This paper discusses the new quality assurance regulations of the Nuclear Regulatory Commission (NRC) for packaging and transportation of radioactive materials. These regulations became effective on October 18, 1977. Background information concerning these regulations and packaging and transportation history is included. The quality assurance program is described with indications of how it is composed of general (administrative) provisions which must meet the 18 quality assurance criteria and be approved by the NRC; specific provisions which appear in the DOT and NRC regulations and in the individual package design approval; and other specific procedures which are not required by regulations but which are necessary for the proper control of quality. The quality assurance program is to be developed using a graded approach for the application of pertinent criteria and optimizing the required degree of safety and control efforts involved in achieving this level of safety. The licensee-user is responsible for all phases of quality assurance for packaging activities including: design, manufacture, test, use, maintenance and repair. The package design phase is considered to be particularly important in producing adequate safety in operational activities concerning packaging and transportation of radioactive materials

Risk Analysis and Decision-Making Software Package (1997 Version) User Manual

Energy Technology Data Exchange (ETDEWEB)

Chung, F.T.H.

1999-02-11

This manual provides instructions for using the U.S. Department of Energy's (DOE) risk analysis and decision making software (1997 version) developed at BDM Petroleum Technologies by BDM-Oklahoma, Inc. for DOE, under contract No. DE-AC22-94PC91OO8. This software provides petroleum producers with a simple, handy tool for exploration and production risk analysis and decision-making. It collects useful risk analysis tools in one package so that users do not have to use several programs separately. The software is simple to use, but still provides many functions. The 1997 version of the software package includes the following tools: (1) Investment risk (Gambler's ruin) analysis; (2) Monte Carlo simulation; (3) Best fit for distribution functions; (4) Sample and rank correlation; (5) Enhanced oil recovery method screening; and (6) artificial neural network. This software package is subject to change. Suggestions and comments from users are welcome and will be considered for future modifications and enhancements of the software. Please check the opening screen of the software for the current contact information. In the future, more tools will be added to this software package. This manual includes instructions on how to use the software but does not attempt to fully explain the theory and algorithms used to create it.

A software package for biomedical image processing and analysis

International Nuclear Information System (INIS)

Goncalves, J.G.M.; Mealha, O.

1988-01-01

The decreasing cost of computing power and the introduction of low cost imaging boards justifies the increasing number of applications of digital image processing techniques in the area of biomedicine. There is however a large software gap to be fulfilled, between the application and the equipment. The requirements to bridge this gap are twofold: good knowledge of the hardware provided and its interface to the host computer, and expertise in digital image processing and analysis techniques. A software package incorporating these two requirements was developed using the C programming language, in order to create a user friendly image processing programming environment. The software package can be considered in two different ways: as a data structure adapted to image processing and analysis, which acts as the backbone and the standard of communication for all the software; and as a set of routines implementing the basic algorithms used in image processing and analysis. Hardware dependency is restricted to a single module upon which all hardware calls are based. The data structure that was built has four main features: hierchical, open, object oriented, and object dependent dimensions. Considering the vast amount of memory needed by imaging applications and the memory available in small imaging systems, an effective image memory management scheme was implemented. This software package is being used for more than one and a half years by users with different applications. It proved to be an excellent tool for helping people to get adapted into the system, and for standardizing and exchanging software, yet preserving flexibility allowing for users' specific implementations. The philosophy of the software package is discussed and the data structure that was built is described in detail

Effect of Various Packaging Methods on Small-Scale Hanwoo (Korean Native Cattle) during Refrigerated Storage

Science.gov (United States)

Yu, Hwan Hee; Song, Myung Wook; Kim, Tae-Kyung; Choi, Yun-Sang; Cho, Gyu Yong; Lee, Na-Kyoung; Paik, Hyun-Dong

2018-01-01

Abstract The objective of this study was to investigate comparison of physicochemical, microbiological, and sensory characteristics of Hanwoo eye of round by various packaging methods [wrapped packaging (WP), modified atmosphere packaging (MAP), vacuum packaging (VP) with three different vacuum films, and vacuum skin packaging (VSP)] at a small scale. Packaged Hanwoo beef samples were stored in refrigerated conditions (4±1°C) for 28 days. Packaged beef was sampled on days 0, 7, 14, 21, and 28. Physicochemical [pH, surface color, thiobarbituric acid reactive substances (TBARS), and volatile basic nitrogen (VBN) values], microbiological, and sensory analysis of packaged beef samples were performed. VP and VSP samples showed low TBARS and VBN values, and pH and surface color did not change substantially during the 28-day period. For VSP, total viable bacteria, psychrotrophic bacteria, lactic acid bacteria, and coliform counts were lower than those for other packaging systems. Salmonella spp. and Escherichia coli O157:H7 were not detected in any packaged beef samples. A sensory analysis showed that the scores for appearance, flavor, color, and overall acceptability did not change significantly until day 7. In total, VSP was effective with respect to significantly higher a* values, physicochemical stability, and microbial safety in Hanwoo packaging (p<0.05). PMID:29805283

AMDA: an R package for the automated microarray data analysis

Directory of Open Access Journals (Sweden)

Foti Maria

2006-07-01

Full Text Available Abstract Background Microarrays are routinely used to assess mRNA transcript levels on a genome-wide scale. Large amount of microarray datasets are now available in several databases, and new experiments are constantly being performed. In spite of this fact, few and limited tools exist for quickly and easily analyzing the results. Microarray analysis can be challenging for researchers without the necessary training and it can be time-consuming for service providers with many users. Results To address these problems we have developed an automated microarray data analysis (AMDA software, which provides scientists with an easy and integrated system for the analysis of Affymetrix microarray experiments. AMDA is free and it is available as an R package. It is based on the Bioconductor project that provides a number of powerful bioinformatics and microarray analysis tools. This automated pipeline integrates different functions available in the R and Bioconductor projects with newly developed functions. AMDA covers all of the steps, performing a full data analysis, including image analysis, quality controls, normalization, selection of differentially expressed genes, clustering, correspondence analysis and functional evaluation. Finally a LaTEX document is dynamically generated depending on the performed analysis steps. The generated report contains comments and analysis results as well as the references to several files for a deeper investigation. Conclusion AMDA is freely available as an R package under the GPL license. The package as well as an example analysis report can be downloaded in the Services/Bioinformatics section of the Genopolis http://www.genopolis.it/

LFSTAT - An R-Package for Low-Flow Analysis

Science.gov (United States)

Koffler, D.; Laaha, G.

2012-04-01

When analysing daily streamflow data focusing on low flow and drought, the state of the art is well documented in the Manual on Low-Flow Estimation and Prediction [1] published by the WMO. While it is clear what has to be done, it is not so clear how to preform the analysis and make the calculation as reproducible as possible. Our software solution expands the high preforming statistical open source software package R to analyse daily stream flow data focusing on low-flows. As command-line based programs are not everyone's preference, we also offer a plug-in for the R-Commander, an easy to use graphical user interface (GUI) to analyse data in R. Functionality includes estimation of the most important low-flow indices. Beside standardly used flow indices also BFI and Recession constants can be computed. The main applications of L-moment based Extreme value analysis and regional frequency analysis (RFA) are available. Calculation of streamflow deficits is another important feature. The most common graphics are prepared and can easily be modified according to the users preferences. Graphics include hydrographs for different periods, flexible streamflow deficit plots, baseflow visualisation, flow duration curves as well as double mass curves just to name a few. The package uses a S3-class called lfobj (low-flow objects). Once this objects are created, analysis can be preformed by mouse-click, and a script can be saved to make the analysis easy reproducible. At the moment we are offering implementation of all major methods proposed in the WMO manual on Low-flow Estimation and Predictions. Future plans include e.g. report export in odt-file using odf-weave. We hope to offer a tool to ease and structure the analysis of stream flow data focusing on low-flows and to make analysis transparent and communicable. The package is designed for hydrological research and water management practice, but can also be used in teaching students the first steps in low-flow hydrology.

Semianalytic Design Sensitivity Analysis of Nonlinear Structures With a Commercial Finite Element Package

International Nuclear Information System (INIS)

Lee, Tae Hee; Yoo, Jung Hun; Choi, Hyeong Cheol

2002-01-01

A finite element package is often used as a daily design tool for engineering designers in order to analyze and improve the design. The finite element analysis can provide the responses of a system for given design variables. Although finite element analysis can quite well provide the structural behaviors for given design variables, it cannot provide enough information to improve the design such as design sensitivity coefficients. Design sensitivity analysis is an essential step to predict the change in responses due to a change in design variables and to optimize a system with the aid of the gradient-based optimization techniques. To develop a numerical method of design sensitivity analysis, analytical derivatives that are based on analytical differentiation of the continuous or discrete finite element equations are effective but analytical derivatives are difficult because of the lack of internal information of the commercial finite element package such as shape functions. Therefore, design sensitivity analysis outside of the finite element package is necessary for practical application in an industrial setting. In this paper, the semi-analytic method for design sensitivity analysis is used for the development of the design sensitivity module outside of a commercial finite element package of ANSYS. The direct differentiation method is employed to compute the design derivatives of the response and the pseudo-load for design sensitivity analysis is effectively evaluated by using the design variation of the related internal nodal forces. Especially, we suggest an effective method for stress and nonlinear design sensitivity analyses that is independent of the commercial finite element package is also discussed. Numerical examples are illustrated to show the accuracy and efficiency of the developed method and to provide insights for implementation of the suggested method into other commercial finite element packages

Safety evaluation for packaging (onsite) for the concrete-shielded RH TRU drum for the 327 Postirradiation Testing Laboratory

International Nuclear Information System (INIS)

Smith, R.J.

1998-01-01

This safety evaluation for packaging authorizes onsite transport of Type B quantities of radioactive material in the Concrete Shielded Remote-Handled Transuranic Waste (RH TRU) Drum per HNF-PRO-154, Responsibilities and Procedures for all Hazardous Material Shipments. The drum will be used for transport of 327 Building legacy waste from the 300 Area to a solid waste storage facility on the Hanford Site

Naval Waste Package Design Report

International Nuclear Information System (INIS)

M.M. Lewis

2004-01-01

A design methodology for the waste packages and ancillary components, viz., the emplacement pallets and drip shields, has been developed to provide designs that satisfy the safety and operational requirements of the Yucca Mountain Project. This methodology is described in the ''Waste Package Design Methodology Report'' Mecham 2004 [DIRS 166168]. To demonstrate the practicability of this design methodology, four waste package design configurations have been selected to illustrate the application of the methodology. These four design configurations are the 21-pressurized water reactor (PWR) Absorber Plate waste package, the 44-boiling water reactor (BWR) waste package, the 5-defense high-level waste (DHLW)/United States (U.S.) Department of Energy (DOE) spent nuclear fuel (SNF) Co-disposal Short waste package, and the Naval Canistered SNF Long waste package. Also included in this demonstration is the emplacement pallet and continuous drip shield. The purpose of this report is to document how that design methodology has been applied to the waste package design configurations intended to accommodate naval canistered SNF. This demonstrates that the design methodology can be applied successfully to this waste package design configuration and support the License Application for construction of the repository

TropFishR: an R package for fisheries analysis with length-frequency data

DEFF Research Database (Denmark)

Mildenberger, Tobias; Taylor, M. H.; Wolff, A.M.

2017-01-01

1. The R package TropFishR is a new analysis toolbox compiling single-species stock assessment methods specifically designed for data-limited fisheries analysis using length-frequency data. 2. It includes methods for (i) estimating biological stock characteristics such as growth and mortality par...... introduces the package and demonstrates the functionality of a selection of its core methods. 4. TropFishR modernises traditional stock assessment methods by easing application and development and by combining it with advanced statistical approaches......1. The R package TropFishR is a new analysis toolbox compiling single-species stock assessment methods specifically designed for data-limited fisheries analysis using length-frequency data. 2. It includes methods for (i) estimating biological stock characteristics such as growth and mortality...

General specifications for the development of a USL NASA PC R and D statistical analysis support package

Science.gov (United States)

Dominick, Wayne D. (Editor); Bassari, Jinous; Triantafyllopoulos, Spiros

1984-01-01

The University of Southwestern Louisiana (USL) NASA PC R and D statistical analysis support package is designed to be a three-level package to allow statistical analysis for a variety of applications within the USL Data Base Management System (DBMS) contract work. The design addresses usage of the statistical facilities as a library package, as an interactive statistical analysis system, and as a batch processing package.

CH Packaging Program Guidance

International Nuclear Information System (INIS)

2008-01-01

The purpose of this document is to provide the technical requirements for preparation for use, operation, inspection, and maintenance of a Transuranic Package Transporter Model II (TRUPACT-II), a HalfPACT shipping package, and directly related components. This document complies with the minimum requirements as specified in the TRUPACT-II Safety Analysis Report for Packaging (SARP), HalfPACT SARP, and U.S. Nuclear Regulatory Commission (NRC) Certificates of Compliance (C of C) 9218 and 9279, respectively. In the event of a conflict between this document and the SARP or C of C, the C of C shall govern. The C of Cs state: 'each package must be prepared for shipment and operated in accordance with the procedures described in Chapter 7.0, Operating Procedures, of the pplication.' They further state: 'each package must be tested and maintained in accordance with the procedures described in Chapter 8.0, Acceptance Tests and Maintenance Program of the Application.' Chapter 9.0 of the SARP charges the U.S. Department of Energy (DOE) or the Waste Isolation Pilot Plant (WIPP) management and operating (M and O) contractor with assuring packaging is used in accordance with the requirements of the C of C. Because the packaging is NRC-approved, users need to be familiar with Title 10 Code of Federal Regulations (CFR) 71.8. Any time a user suspects or has indications that the conditions of approval in the C of C were not met, the Carlsbad Field Office (CBFO) shall be notified immediately. The CBFO will evaluate the issue and notify the NRC if required. In accordance with 10 CFR Part 71, certificate holders, packaging users, and contractors or subcontractors who use, design, fabricate, test, maintain, or modify the packaging shall post copies of (1) 10 CFR Part 21 regulations, (2) Section 206 of the Energy Reorganization Act of 1974, and (3) NRC Form 3, Notice to Employees. These documents must be posted in a conspicuous location where the activities subject to these regulations

CH Packaging Program Guidance

International Nuclear Information System (INIS)

2009-01-01

The purpose of this document is to provide the technical requirements for preparation for use, operation, inspection, and maintenance of a Transuranic Package Transporter Model II (TRUPACT-II), a HalfPACT shipping package, and directly related components. This document complies with the minimum requirements as specified in the TRUPACT-II Safety Analysis Report for Packaging (SARP), HalfPACT SARP, and U.S. Nuclear Regulatory Commission (NRC) Certificates of Compliance (C of C) 9218 and 9279, respectively. In the event of a conflict between this document and the SARP or C of C, the C of C shall govern. The C of Cs state: 'each package must be prepared for shipment and operated in accordance with the procedures described in Chapter 7.0, Operating Procedures, of the application.' They further state: 'each package must be tested and maintained in accordance with the procedures described in Chapter 8.0, Acceptance Tests and Maintenance Program of the Application.' Chapter 9.0 of the SARP charges the U.S. Department of Energy (DOE) or the Waste Isolation Pilot Plant (WIPP) management and operating (M and O) contractor with assuring packaging is used in accordance with the requirements of the C of C. Because the packaging is NRC-approved, users need to be familiar with Title 10 Code of Federal Regulations (CFR) 71.8. Any time a user suspects or has indications that the conditions of approval in the C of C were not met, the Carlsbad Field Office (CBFO) shall be notified immediately. The CBFO will evaluate the issue and notify the NRC if required. In accordance with 10 CFR Part 71, certificate holders, packaging users, and contractors or subcontractors who use, design, fabricate, test, maintain, or modify the packaging shall post copies of (1) 10 CFR Part 21 regulations, (2) Section 206 of the Energy Reorganization Act of 1974, and (3) NRC Form 3, Notice to Employees. These documents must be posted in a conspicuous location where the activities subject to these regulations

RH Packaging Program Guidance

International Nuclear Information System (INIS)

2008-01-01

The purpose of this program guidance document is to provide the technical requirements for use, operation, inspection, and maintenance of the RH-TRU 72-B Waste Shipping Package (also known as the 'RH-TRU 72-B cask') and directly related components. This document complies with the requirements as specified in the RH-TRU 72-B Safety Analysis Report for Packaging (SARP), and Nuclear Regulatory Commission (NRC) Certificate of Compliance (C of C) 9212. If there is a conflict between this document and the SARP and/or C of C, the C of C shall govern. The C of C states: 'each package must be prepared for shipment and operated in accordance with the procedures described in Chapter 7.0, Operating Procedures, of the application.' It further states: 'each package must be tested and maintained in accordance with the procedures described in Chapter 8.0, Acceptance Tests and Maintenance Program of the Application.' Chapter 9.0 of the SARP tasks the Waste Isolation Pilot Plant (WIPP) Management and Operating (M and O) Contractor with assuring the packaging is used in accordance with the requirements of the C of C. Because the packaging is NRC-approved, users need to be familiar with Title 10 Code of Federal Regulations (CFR) 71.8, 'Deliberate Misconduct.' Any time a user suspects or has indications that the conditions of approval in the C of C were not met, the U.S. Department of Energy (DOE) Carlsbad Field Office (CBFO) shall be notified immediately. The CBFO will evaluate the issue and notify the NRC if required.In accordance with 10 CFR Part 71, 'Packaging and Transportation of Radioactive Material,' certificate holders, packaging users, and contractors or subcontractors who use, design, fabricate, test, maintain, or modify the packaging shall post copies of (1) 10 CFR Part 21, 'Reporting of Defects and Noncompliance,' regulations, (2) Section 206 of the Energy Reorganization Act of 1974, and (3) NRC Form 3, Notice to Employees. These documents must be posted in a conspicuous

RH Packaging Program Guidance

International Nuclear Information System (INIS)

2006-01-01

The purpose of this program guidance document is to provide the technical requirements for use, operation, inspection, and maintenance of the RH-TRU 72-B Waste Shipping Package and directly related components. This document complies with the requirements as specified in the RH-TRU 72-B Safety Analysis Report for Packaging (SARP), and Nuclear Regulatory Commission (NRC) Certificate of Compliance (C of C) 9212. If there is a conflict between this document and the SARP and/or C of C, the C of C shall govern. The C of C states: 'each package must be prepared for shipment and operated in accordance with the procedures described in Chapter 7.0, Operating Procedures, of the application.' It further states: 'each package must be tested and maintained in accordance with the procedures described in Chapter 8.0, Acceptance Tests and Maintenance Program of the Application.' Chapter 9.0 of the SARP tasks the Waste Isolation Pilot Plant (WIPP) Management and Operating (M and O) Contractor with assuring the packaging is used in accordance with the requirements of the C of C. Because the packaging is NRC-approved, users need to be familiar with 10 Code of Federal Regulations (CFR) 1.8, 'Deliberate Misconduct.' Any time a user suspects or has indications that the conditions of approval in the C of C were not met, the U.S. Department of Energy (DOE) Carlsbad Field Office (CBFO) shall be notified immediately. CBFO will evaluate the issue and notify the NRC if required. In accordance with 10 CFR Part 71, 'Packaging and Transportation of Radioactive Material,' certificate holders, packaging users, and contractors or subcontractors who use, design, fabricate, test, maintain, or modify the packaging shall post copies of (1) 10 CFR Part 21, 'Reporting of Defects and Noncompliance,' regulations, (2) Section 206 of the Energy Reorganization Act of 1974, and (3) NRC Form 3, Notice to Employees. These documents must be posted in a conspicuous location where the activities subject to these

The Importance of Take-Out Food Packaging Attributes: Conjoint Analysis and Quality Function Deployment Approach

Science.gov (United States)

Lestari Widaningrum, Dyah

2014-03-01

This research aims to investigate the importance of take-out food packaging attributes, using conjoint analysis and QFD approach among consumers of take-out food products in Jakarta, Indonesia. The conjoint results indicate that perception about packaging material (such as paper, plastic, and polystyrene foam) plays the most important role overall in consumer perception. The clustering results that there is strong segmentation in which take-out food packaging material consumer consider most important. Some consumers are mostly oriented toward the colour of packaging, while another segment of customers concerns on packaging shape and packaging information. Segmentation variables based on packaging response can provide very useful information to maximize image of products through the package's impact. The results of House of Quality development described that Conjoint Analysis - QFD is a useful combination of the two methodologies in product development, market segmentation, and the trade off between customers' requirements in the early stages of HOQ process

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

International Nuclear Information System (INIS)

2014-01-01

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

Experimental And Numerical Investigation Of The Flow Analysis Of The Water-Saving Safety Valve

Directory of Open Access Journals (Sweden)

Muhammed Safa Kamer

2015-08-01

Full Text Available Abstract In this study an auto-mechanical safety valve was designed and manufactured in order to prevent possible wastage of water and water raid after instantaneous water cuts during water usage in places where water use is widespread. Safety valve is activated and it switches off the line when water is cut off when mains pressure is equal to atmospheric pressure and as it does not allow water to pass when it comes back it saves water and prevents the formation of raids. An experiment set was conducted in order to measure the pressure drop between the inlet and outlet of the safety valve and it was found that with the increased flow rate the pressure drop increases. The three-dimensional flow analysis of the safety valve was carried out with Ansys-Fluent software package and the results obtained were compared with experimental data and a good harmony was achieved.

STATUS REPORT FOR MOISTURE EFFECTS ON COMPACTION OF FIBERBOARD IN A 9975 SHIPPING PACKAGE

Energy Technology Data Exchange (ETDEWEB)

Stefek, T.; Daugherty, W.; Estochen, E.

2011-06-23

Compaction of lower layers in the fiberboard overpack has been observed in 9975 packages that contain elevated moisture. Lab testing has resulted in a better understanding of the relationship between the fiberboard moisture level and compaction of the lower fiberboard assembly, and the behavior of the fiberboard during transport. In laboratory tests, higher moisture content has been shown to correspond to higher total compaction of fiberboard material, greater rate of compaction, and continued compaction over a longer period of time. In addition, laboratory tests have shown that the application of a dynamic load results in higher fiberboard compaction. The test conditions and sample geometric/loading configurations were chosen to simulate the regulatory requirements for 9975 package input dynamic loading. Dynamic testing was conducted over a period of six months to acquire immediate and cumulative changes in geometric data for various moisture levels. Currently, one sample set has undergone a complete dynamic test regimen, while testing of another set is still in-progress. The dynamic input, data acquisition, test effects on sample dynamic parameters, and interim results from this test program are summarized and compared to regulatory specifications for dynamic loading. This will provide a basis from which to evaluate the impact of moisture and fiberboard compaction on the safety basis for transportation (Safety Analysis Report for Packaging) and storage (facility Documented Safety Analysis) at the Savannah River Site (SRS).

K Basin safety analysis

International Nuclear Information System (INIS)

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

1994-01-01

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

Safety analysis for 'Fugen'

International Nuclear Information System (INIS)

1997-10-01

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

PAT-1 safety analysis report addendum.

Energy Technology Data Exchange (ETDEWEB)

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

2010-09-01

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

Biocidal packaging for pharmaceuticals, foods, and other perishables.

Science.gov (United States)

Larson, Alyssa M; Klibanov, Alexander M

2013-01-01

Many consumer goods must be protected from bacterial and fungal colonization to ensure their integrity and safety. By making these items' packaging biocidal, the interior environment can be preserved from microbial spoilage without altering the products themselves. Herein we briefly review this concept, referred to as active packaging, and discuss existing methods for constructing active packaging systems. They are based on either packaging materials that release biocides or those that are themselves intrinsically biocidal (or biostatic), with numerous variations within each category.

The IronChip evaluation package: a package of perl modules for robust analysis of custom microarrays

Directory of Open Access Journals (Sweden)

Brazma Alvis

2010-03-01

Full Text Available Abstract Background Gene expression studies greatly contribute to our understanding of complex relationships in gene regulatory networks. However, the complexity of array design, production and manipulations are limiting factors, affecting data quality. The use of customized DNA microarrays improves overall data quality in many situations, however, only if for these specifically designed microarrays analysis tools are available. Results The IronChip Evaluation Package (ICEP is a collection of Perl utilities and an easy to use data evaluation pipeline for the analysis of microarray data with a focus on data quality of custom-designed microarrays. The package has been developed for the statistical and bioinformatical analysis of the custom cDNA microarray IronChip but can be easily adapted for other cDNA or oligonucleotide-based designed microarray platforms. ICEP uses decision tree-based algorithms to assign quality flags and performs robust analysis based on chip design properties regarding multiple repetitions, ratio cut-off, background and negative controls. Conclusions ICEP is a stand-alone Windows application to obtain optimal data quality from custom-designed microarrays and is freely available here (see "Additional Files" section and at: http://www.alice-dsl.net/evgeniy.vainshtein/ICEP/

Development and experimental qualification of the new safety-criticality CRISTAL package; Developpement et qualification experimentale du nouveau formulaire de surete-criticite Cristal

Energy Technology Data Exchange (ETDEWEB)

Mattera, Ch

1998-11-01

This thesis is concerned with Criticality-Safety studies related to the French Nuclear Fuel Cycle. We first describe the steps in the nuclear fuel cycle and the specific characteristics of these studies compared with those performed in Reactor Physics. In order to respond to the future requirements of the French Nuclear Program, we have developed a new package CRISTAL based on a recent cross sections library (CEA 93) and the newest accurate codes (APOLLO 2, MORET 4, TRIPOLI 4). The CRISTAL system includes two calculations routes: a design route which will be used by French Industry (COGEMA/SGN) and a reference route. To transfer this package to the French industry, we have elaborated calculation schemes for fissile solutions, dissolver media, transport casks and storage pools. Afterwards, these schemes have been used for the CRISTAL experimental validation. We have also contributed to the CRISTAL experimental database by reevaluating a French storage pool experiment: the CRISTO II experiment. This revaluation has been submitted to the OECD working group in order that this experiment can be used by international criticality safety engineers to validate calculations methods. This work represents a large contribution to the recommendation of accurate calculation schemes and to the experimental validation of the CRISTAL package. These studies came up to the French Industry expectations. (author)

Reactor Safety Analysis

International Nuclear Information System (INIS)

Arien, B.

2000-01-01

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

Designing packaging to support the safe use of medicines at home.

Science.gov (United States)

Ward, James; Buckle, Peter; John Clarkson, P

2010-09-01

In the light of a number of serious incidents in the UK resulting from accidental overdoses of methotrexate, this study investigated how the design of methotrexate packaging can influence patient safety, and aimed to collect evidence to provide a basis for the development of new concepts for revised designs by the pharmaceutical industry. The research found that patients using methotrexate experience a number of difficulties in using their medicines packaging, and as a result, resourcefully adopt a variety of "coping strategies" which may increase the risk of dosing errors. By investigating both the practice of methotrexate users, and the design of the system that supports methotrexate use, additional problems were observed across the healthcare system, meaning that the function of medicines packaging in ensuring safety may be even more critical than first suspected. As a result of this research the National Patient Safety Agency responded with a UK-wide programme of work to improve safety for patients, and continues to work with the pharmaceutical industry to develop more user-friendly packaging and labelling. 2010 Elsevier Ltd. All rights reserved.

Antimicrobial packaging with natural compunds - a review

Directory of Open Access Journals (Sweden)

Renata Dobrucka

2016-12-01

Full Text Available Background:  Packaging problems are an integral part of logistics and the implementation of packaging significantly affects the effectiveness of logistics processes, as a factor which increases the safety and the quality of products being transported. Active packaging is an area of technology needed to meet the requirements of the contemporary consumer. Active packaging creates additional opportunities in systems for packing goods, as well as offering a solution in which the packaging, the product and surroundings interact. Furthermore, active packaging allows packaging to interact with food and the environment and play a dynamic role in food preservation. The main role of antimicrobial packaging is to inhibit the growth of microorganisms that reduce the quality of the packaged product. Methods: The application of natural antimicrobial agents appears to be safe for food products. Also, these compounds have potential applications as a natural preservative in the food packaging industry. This study presents some antibacterial agents, namely chitosan, nisin and pectins. Results and conclusion: Natural substances used in active packaging can eliminate the danger of chemical substances migrating to food.

Active and intelligent food packaging: legal aspects and safety concerns

NARCIS (Netherlands)

Dainelli, D.; Gontard, N.; Spyropoulos, D.; Zondervan-van den Beuken, E.; Tobback, P.

2008-01-01

'Active and intelligent' (A&I) food packaging is based on a deliberate interaction of the packaging with the food and/or its direct environment. This article presents: (i) the main types of materials developed for food contact; (ii) the global market and the future trends of active and intelligent

SPHINX experimenters information package

International Nuclear Information System (INIS)

Zarick, T.A.

1996-08-01

This information package was prepared for both new and experienced users of the SPHINX (Short Pulse High Intensity Nanosecond X-radiator) flash X-Ray facility. It was compiled to help facilitate experiment design and preparation for both the experimenter(s) and the SPHINX operational staff. The major areas covered include: Recording Systems Capabilities,Recording System Cable Plant, Physical Dimensions of SPHINX and the SPHINX Test cell, SPHINX Operating Parameters and Modes, Dose Rate Map, Experiment Safety Approval Form, and a Feedback Questionnaire. This package will be updated as the SPHINX facilities and capabilities are enhanced

Reliability analysis of PLC safety equipment

Energy Technology Data Exchange (ETDEWEB)

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

2006-06-15

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

Reliability analysis of PLC safety equipment

International Nuclear Information System (INIS)

Yu, J.; Kim, J. Y.

2006-06-01

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

Radioisotope Thermoelectric Generator Transporation System licensed hardware second certification test series and package shock mount system test

International Nuclear Information System (INIS)

Ferrell, P.C.; Moody, D.A.

1995-10-01

This paper presents a summary of two separate drop test a e performed in support of the Radioisotope Thermoelectric Generator (RTG) Transportation System (RTGTS). The first portion of this paper presents the second series of drop testing required to demonstrate that the RTG package design meets the requirements of Title 10, Code of Federal Regulations, ''Part 71'' (10 CFR 71). Results of the first test series, performed in July 1994, demonstrated that some design changes were necessary. The package design was modified to improve test performance and the design changes were incorporated into the Safety Analysis Report for Packaging (SARP). The second full-size certification test article (CTA-2) incorporated the modified design and was tested at the US Department of Energy's (DOE) Hanford Site near Richland, Washington. With the successful completion of the test series, and pending DOE Office of Facility Safety Analysis approval of the SARP, a certificate of compliance will be issued for the RTG package allowing its use. The second portion of this paper presents the design and testing of the RTG Package Mount System. The RTG package mount was designed to protect the RTG from excessive vibration during transport, provide shock protection during on/off loading, and provide a mechanism for moving the RTG package with a forklift. Military Standard (MIL-STD) 810E, Transit Drop Procedure (DOE 1989), was used to verify that the shock limiting system limited accelerations in excess of 15 G's at frequencies below 150 Hz. Results of the package mount drop tests indicate that an impact force of 15 G's was not exceeded in any test from a free drop height of 457 mm (18 in.)

10 CFR 76.87 - Technical safety requirements.

Science.gov (United States)

2010-01-01

...: (1) Effects of natural phenomena; (2) Building and process ventilation and offgas; (3) Criticality...; (8) Environmental protection; (9) Packaging and transporting nuclear materials; (10) Accident analysis; (11) Chemical safety; (12) Sharing of facilities, structures, systems and components; (13...

Transportation package design using numerical optimization

International Nuclear Information System (INIS)

Harding, D.C.; Witkowski, W.R.

1992-01-01

The design of structures and engineering systems has always been an iterative process whose complexity was dependent upon the boundary conditions, constraints and available analytical tools. Transportation packaging design is no exception with structural, thermal and radiation shielding constraints based on regulatory hypothetical accident conditions. Transportation packaging design is often accomplished by a group of specialists, each designing a single component based on one or more simple criteria, pooling results with the group, evaluating the open-quotes pooledclose quotes design, and then reiterating the entire process until a satisfactory design is reached. The manual iterative methods used by the designer/analyst can be summarized in the following steps: design the part, analyze the part, interpret the analysis results, modify the part, and re-analyze the part. The inefficiency of this design practice and the frequently conservative result suggests the need for a more structured design methodology, which can simultaneously consider all of the design constraints. Numerical optimization is a structured design methodology whose maturity in development has allowed it to become a primary design tool in many industries. The purpose of this overview is twofold: first, to outline the theory and basic elements of numerical optimization; and second, to show how numerical optimization can be applied to the transportation packaging industry and used to increase efficiency and safety of radioactive and hazardous material transportation packages. A more extensive review of numerical optimization and its applications to radioactive material transportation package design was performed previously by the authors (Witkowski and Harding 1992). A proof-of-concept Type B package design is also presented as a simplified example of potential improvements achievable using numerical optimization in the design process

The UK core performance code package

International Nuclear Information System (INIS)

Hutt, P.K.; Gaines, N.; McEllin, M.; White, R.J.; Halsall, M.J.

1991-01-01

Over the last few years work has been co-ordinated by Nuclear Electric, originally part of the Central Electricity Generating Board, with contributions from the United Kingdom Atomic Energy Authority and British Nuclear Fuels Limited, to produce a generic, easy-to-use and integrated package of core performance codes able to perform a comprehensive range of calculations for fuel cycle design, safety analysis and on-line operational support for Light Water Reactor and Advanced Gas Cooled Reactor plant. The package consists of modern rationalized generic codes for lattice physics (WIMS), whole reactor calculations (PANTHER), thermal hydraulics (VIPRE) and fuel performance (ENIGMA). These codes, written in FORTRAN77, are highly portable and new developments have followed modern quality assurance standards. These codes can all be run ''stand-alone'' but they are also being integrated within a new UNIX-based interactive system called the Reactor Physics Workbench (RPW). The RPW provides an interactive user interface and a sophisticated data management system. It offers quality assurance features to the user and has facilities for defining complex calculational sequences. The Paper reviews the current capabilities of these components, their integration within the package and outlines future developments underway. Finally, the Paper describes the development of an on-line version of this package which is now being commissioned on UK AGR stations. (author)

From waste packages acceptance criteria to waste packages acceptance process at the Centre de l'Aube disposal facility

International Nuclear Information System (INIS)

Dutzer, M.

2003-01-01

The Centre de l'Aube disposal facility has now been operated for 10 years. At the end of 2001, about 124,000 m3 of low and intermediate level short lived waste packages, representing 180,000 packages, have been disposed, for a total capacity of 1,000,000 m3. The flow of waste packages is now between 12 and 15,000 m3 per year, that is one third of the flow that was taken into account for the design of the repository. It confirms the efforts by waste generators to minimise waste production. This flow represents 25 to 30,000 packages, 50% are conditioned into the compaction facility of the repository, so that 17,000 packages are disposed per year. 54 disposal vaults have been closed. In 1996-1999, the safety assessment of the repository have been reviewed, taking into account the experience of operation. This assessment was investigated by the regulatory body and, subsequently, a so-called 'definitive license' to operate was granted to ANDRA on September 2, 1999 with updated licensing requirements. Another review will be performed in 2004. To ensure a better consistency with the safety assessment of the facility, Andra issued new technical requirements for waste packages at the end of 2000. Discussions with waste generators also showed that the waste package acceptance process should be improved to provide a more precise definition of operational criteria to comply with in waste conditioning facilities. Consequently, a new approach has been implemented since 2000. (orig.)

Reactor safety analysis

International Nuclear Information System (INIS)

Arien, B.

1998-01-01

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

Packaging- and transportation-related occurrence reports: 1993 annual report

International Nuclear Information System (INIS)

Welch, M.J.; Dickerson, L.S.; Jennings, S.D.

1994-06-01

The US Department of Energy (DOE) Occurrence Reporting and Processing System (ORPS) is an interactive computer system designed to support DOE-owned or -operated facilities in reporting and processing of information concerning occurrences related to facility operations. The requirements for reporting and the extent of the occurrences to be reported are defined in DOE Order 5000.3B, Occurrence Reporting and Processing of Operations Information (hereafter referred to as DOE 5000.3B). The centralized data base, which is managed by the Idaho National Engineering Laboratory (INEL), provides computerized support for the collection, distribution, updating, analysis, and sign-off of information in the occurrence reports (ORs). The Oak Ridge National Laboratory (ORNL) Packaging and Transportation Safety (PATS) Program has been made responsible for retrieving reports and information pertaining to transportation and packaging incidents/accidents from the centralized ORPS data base. This annual report details the methodology that PATS uses to conduct searches of the ORPS for pertinent information, the form of the reporting to EH-332, review and examination of trends observed in ORs related to transportation and packaging safety, a presentation and discussion of the root-cause codes of ORPS and the nature of occurrence codes of PATS, timely processing of notification reports to final stage, and analysis of 10% of the reported ORs that were finalized to determine whether the actions taken to close out the occurrences were sufficient to ensure remediation of the incident and to prevent a recurrence. Data in the report are presented by calendar years

Characterizing and packaging BN-350 spent fuel for long-term dry storage

International Nuclear Information System (INIS)

Lambert, J. D. B.; Bolshinsky, I.; Haues, S.L.; Allen, K.J.; Howden, E.A.; Hill, R.N.; Planchon, H.P.; Staples, P.; Karaulov, V.N.; Blynskij, A.P.; Yakovlev, I.K.; Maev, V.; Dumchev, I. A.

2000-01-01

The Republic of Kazakhstan is being assisted by the U.S. Department of Energy in preparing spent fuel from the BN-350 fast reactor for long term dry storage. Argonne National Laboratory was assigned responsibility for the physical and nuclear characterization of the spent fuel, for the design and safety analysis of 6-pac and 4-pac canisters used to contain spent fuel assemblies for storage, and for the design, testing and installation of a closure station at the reactor in which the canisters of fuel are dried, filled with inert gas and welded shut. This paper briefly describes the specialized components and equipment used, the process followed, and experience gained in packaging the spent fuel. Olsen et al and Schaefer separately discuss overall safety and criticality considerations of the packaging process in parallel papers to this conference

Consignor's part and multimodality of the packages of RAM

International Nuclear Information System (INIS)

Grenier, M.

1993-01-01

One of the essential principles of Safety in transport of RAM as it was carried on, until the present days, is that the safety comes firstly from the packaging, and not from a special consideration of the mode of transport or anything else. A natural consequence was that the packaging remains the same, whatever is the used mode of transport, another that the transport mode does not take part, otherly than by its normal safety in the carriage itself in the protection against the radioactive hazard. Some recent actions, nevertheless, could be considered as putting some doubt on the implementation of this principles in the regulations and sometimes on these principles themselves. However, these latter come from the fact that the safety of the transport lays essentially in the hands of the consignor. He is indeed the only one who knows well enough the material and who is able to assess the adequation of the packaging and to warrant it to the Competent Authority when necessary. Besides, the packaging is the only direct grasp he has on the shipment, outside the special case of exclusive use. This situation is not completely incompatible with a new possible graduation of the packaging according to the transport mode, under the condition that this graduation be unidimensional which means the same in every field of testing. The consignor must, then, select the packaging corresponding to the most demanding of the possible modes. It is certainly much more difficult to adapt the mode of transport to the shipped materials themselves. All the transport has then to be in the hands of the consignor. That perhaps, can be relevant to industrial nuclear transport (eg: irradiated fuels, wastes), but not to the carriage of radioactive materials in general, which is the scope of IAEA regulations. (author)

Release Mathematical Model of Active Agent from Packaging Material into Food

Directory of Open Access Journals (Sweden)

Xiuling Huang

2013-01-01

Full Text Available Active packaging is an innovative packaging technology by which active compounds are released from the package to enhance the quality and microbial safety for a wide range of foods. The problem of active ingredient release through the bilayer packaging food system is studied from a theoretical viewpoint. A release model is built to provide predictions of concentration and amount of active ingredient. The equations are built based on Fickian diffusion, and numerical solutions are obtained through finite difference. Different diffusion coefficients DP and DC of active ingredient in different packaging layers, partition coefficient kCP at the interface of outer layer and inner layer, partition coefficient kFC at the interface of inner layer and food, and mass transfer coefficient hm at the interface of inner layer and food are considered in the model. The effects of kCP, thicknesses of outer layer and inner layer, CP0, DP, DC, kFC, and hm on the release are discussed. Corresponding conclusions and analysis are given.

Hazardous materials package performance regulations

International Nuclear Information System (INIS)

Russell, N.A.; Glass, R.E.; McClure, J.D.; Finley, N.C.

1992-01-01

The hazardous materials (hazmat) packaging development and certification process is currently defined by two different regulatory philosophies, one based on specification packagings and the other based on performance standards. With specification packagings, a packaging is constructed according to an agreed set of design specifications. In contrast, performance standards do not specify the packaging design; they specify performance standards that a packaging design must be able to pass before it can be certified for transport. The packaging can be designed according to individual needs as long as it meets these performance standards. Performance standards have been used nationally and internationally for about 40 years to certify radioactive materials (RAM) packagings. It is reasonable to state that for RAM transport, performance specifications have maintained transport safety. A committee of United Nation's experts recommended the performance standard philosophy as the preferred regulation method for hazmat packaging. Performance standards for hazmat packagings smaller than 118 gallons have been adopted in 49CFR178. Packagings for materials that are classified as toxic-by-inhalation must comply with the performance standards by October 1, 1993, and packagings for all other classes of hazardous materials covered must comply by October 1, 1996. For packages containing bulk (in excess of 188 gallons) quantities of materials that are extremely toxic by inhalation, there currently are no performance requirements. This paper discusses a Hazmat Packaging Performance Evaluation (HPPE) project to look at the subset of bulk packagings that are larger than 2000 gallons. The objectives of this project are the evaluate current hazmat specification packagings and develop supporting documentation for determining performance requirements for packagings in excess of 2000 gallons that transport hazardous materials that have been classified as extremely toxic by inhalation (METBI)

Improvement of Level-1 PSA computer code package -A study for nuclear safety improvement-

International Nuclear Information System (INIS)

Park, Chang Kyu; Kim, Tae Woon; Ha, Jae Joo; Han, Sang Hoon; Cho, Yeong Kyun; Jeong, Won Dae; Jang, Seung Cheol; Choi, Young; Seong, Tae Yong; Kang, Dae Il; Hwang, Mi Jeong; Choi, Seon Yeong; An, Kwang Il

1994-07-01

This year is the second year of the Government-sponsored Mid- and Long-Term Nuclear Power Technology Development Project. The scope of this subproject titled on 'The Improvement of Level-1 PSA Computer Codes' is divided into three main activities : (1) Methodology development on the under-developed fields such as risk assessment technology for plant shutdown and external events, (2) Computer code package development for Level-1 PSA, (3) Applications of new technologies to reactor safety assessment. At first, in the area of PSA methodology development, foreign PSA reports on shutdown and external events have been reviewed and various PSA methodologies have been compared. Level-1 PSA code KIRAP and CCF analysis code COCOA are converted from KOS to Windows. Human reliability database has been also established in this year. In the area of new technology applications, fuzzy set theory and entropy theory are used to estimate component life and to develop a new measure of uncertainty importance. Finally, in the field of application study of PSA technique to reactor regulation, a strategic study to develop a dynamic risk management tool PEPSI and the determination of inspection and test priority of motor operated valves based on risk importance worths have been studied. (Author)

Natural additives and agricultural wastes in biopolymer formulations for food packaging

Science.gov (United States)

Valdés, Arantzazu; Mellinas, Ana Cristina; Ramos, Marina; Garrigós, María Carmen; Jiménez, Alfonso

2014-02-01

The main directions in food packaging research are targeted towards improvements in food quality and food safety. For this purpose, food packaging providing longer product shelf-life, as well as the monitoring of safety and quality based upon international standards, is desirable. New active packaging strategies represent a key area of development in new multifunctional materials where the use of natural additives and/or agricultural wastes is getting increasing interest. The development of new materials, and particularly innovative biopolymer formulations, can help to address these requirements and also with other packaging functions such as: food protection and preservation, marketing and smart communication to consumers. The use of biocomposites for active food packaging is one of the most studied approaches in the last years on materials in contact with food. Applications of these innovative biocomposites could help to provide new food packaging materials with improved mechanical, barrier, antioxidant and antimicrobial properties. From the food industry standpoint, concerns such as the safety and risk associated with these new additives, migration properties and possible human ingestion and regulations need to be considered. The latest innovations in the use of these innovative formulations to obtain biocomposites are reported in this review. Legislative issues related to the use of natural additives and agricultural wastes in food packaging systems are also discussed.

Natural additives and agricultural wastes in biopolymer formulations for food packaging.

Science.gov (United States)

Valdés, Arantzazu; Mellinas, Ana Cristina; Ramos, Marina; Garrigós, María Carmen; Jiménez, Alfonso

2014-01-01

The main directions in food packaging research are targeted toward improvements in food quality and food safety. For this purpose, food packaging providing longer product shelf-life, as well as the monitoring of safety and quality based upon international standards, is desirable. New active packaging strategies represent a key area of development in new multifunctional materials where the use of natural additives and/or agricultural wastes is getting increasing interest. The development of new materials, and particularly innovative biopolymer formulations, can help to address these requirements and also with other packaging functions such as: food protection and preservation, marketing and smart communication to consumers. The use of biocomposites for active food packaging is one of the most studied approaches in the last years on materials in contact with food. Applications of these innovative biocomposites could help to provide new food packaging materials with improved mechanical, barrier, antioxidant, and antimicrobial properties. From the food industry standpoint, concerns such as the safety and risk associated with these new additives, migration properties and possible human ingestion and regulations need to be considered. The latest innovations in the use of these innovative formulations to obtain biocomposites are reported in this review. Legislative issues related to the use of natural additives and agricultural wastes in food packaging systems are also discussed.

Natural additives and agricultural wastes in biopolymer formulations for food packaging

Directory of Open Access Journals (Sweden)

Arantzazu eValdés

2014-02-01

Full Text Available The main directions in food packaging research are targeted towards improvements in food quality and food safety. For this purpose, food packaging providing longer product shelf-life, as well as the monitoring of safety and quality based upon international standards, is desirable. New active packaging strategies represent a key area of development in new multifunctional materials where the use of natural additives and/or agricultural wastes is getting increasing interest. The development of new materials, and particularly innovative biopolymer formulations, can help to address these requirements and also with other packaging functions such as: food protection and preservation, marketing and smart communication to consumers. The use of biocomposites for active food packaging is one of the most studied approaches in the last years on materials in contact with food. Applications of these innovative biocomposites could help to provide new food packaging materials with improved mechanical, barrier, antioxidant and antimicrobial properties. From the food industry standpoint, concerns such as the safety and risk associated with these new additives, migration properties and possible human ingestion and regulations need to be considered. The latest innovations in the use of these innovative formulations to obtain biocomposites are reported in this review. Legislative issues related to the use of natural additives and agricultural wastes in food packaging systems are also discussed.

Completion of the radioactive materials packaging handbook

International Nuclear Information System (INIS)

Shappert, L.B.

1998-01-01

'The Radioactive Materials Packaging Handbook: Design, Operation and Maintenance', which will serve as a replacement for the 'Cask Designers Guide'(1970), has now been completed and submitted to the Oak Ridge National Laboratory (ORNL) electronics publishing group for layout and printing; it is scheduled to be printed in late spring 1998. The Handbook, written by experts in their particular fields, is a compilation of technical chapters that address the design aspects of a package intended for transporting radioactive material in normal commerce; it was prepared under the direction of M. E. Wangler of the US DOE and is intended to provide a wealth of technical guidance that will give designers a better understanding of the regulatory approval process, preferences of regulators on specific aspects of package design, and the types of analyses that should be considered when designing a package to carry radioactive materials. Even though the Handbook is concerned with both small and large packagings, most of the emphasis is placed on large packagings that are capable of transporting fissile, radioactive sources (e.g. spent fuels). The safety analysis reports for packagings (SARPs) must address the widest range of technical topics in order to meet United States and/or international regulations, all of which are covered in the Handbook. One of the primary goals of the Handbook is to provide information which would guide designers of radioactive materials packages to make decisions that would most likely be acceptable to regulatory agencies during the approval process of the packaging. It was therefore important to find those authors who not only were experts in one or more of the areas that are addressed in a SARP, but who also had been exposed to the regulatory process or had operational experience dealing with a wide variety of package types. Twenty-five such people have contributed their time and talents to the development of this document, mostly on a volunteer basis

Ceramic ball grid array package stress analysis

Science.gov (United States)

Badri, S. H. B. S.; Aziz, M. H. A.; Ong, N. R.; Sauli, Z.; Alcain, J. B.; Retnasamy, V.

2017-09-01

The ball grid array (BGA), a form of chip scale package (CSP), was developed as one of the most advanced surface mount devices, which may be assembled by an ordinary surface ball bumps are used instead of plated nickel and gold (Ni/Au) bumps. Assembly and reliability of the BGA's printed circuit board (PCB), which is soldered by conventional surface mount technology is considered in this study. The Ceramic Ball Grid Array (CBGA) is a rectangular ceramic package or square-shaped that will use the solder ball for external electrical connections instead of leads or wire for connections. The solder balls will be arranged in an array or grid at the bottom of the ceramic package body. In this study, ANSYS software is used to investigate the stress on the package for 2 balls and 4 balls of the CBGA package with the various force range of 1-3 Newton applied to the top of the die, top of the substrate and side of the substrate. The highest maximum stress was analyzed and the maximum equivalent stress was observed on the solder ball and the die. From the simulation result, the CBGA package with less solder balls experience higher stress compared to the package with many solder balls. Therefore, less number of solder ball on the CBGA package results higher stress and critically affect the reliability of the solder balls itself, substrate and die which can lead to the solder crack and also die crack.

A probabilistic risk-analysis of the transport of small radioactive material type B packages in France

International Nuclear Information System (INIS)

Hubert, P.; Pages, P.

1982-01-01

The assessment of the accidental risk due to the road transportation of a small package containing γ-ray emitters is performed in France. Analyzing records of road transportation accidents, modeling the package behaviour and estimating the importance of the involved population are the three main steps of the study. The interest of such an anlysis relies on the relative simplicity of the model and the availability of statistical data. This allows modelling of the whole process and study of the various sensitivities. It is also of pratical interest when assessing the cost-effectiveness of some safety/protection measures

The arison data acquisition and elaboration software package running on Hewlett-Packard minicomputer

International Nuclear Information System (INIS)

Diamantidis, Z.

1987-01-01

In this article, a data acquisition and elaboration system is described consisting in a PCM data acquisition and a spectrum analyser system and their data elaboration package for reactor safety or other general purposes. Measurements on temperature, noise fluctuations of temperature or other noise analysis dynamics, pressure, etc. Time series and their conversion in engineering units and their statistical and frequency analysis is provided

Bayesian cost-effectiveness analysis with the R package BCEA

CERN Document Server

Baio, Gianluca; Heath, Anna

2017-01-01

The book provides a description of the process of health economic evaluation and modelling for cost-effectiveness analysis, particularly from the perspective of a Bayesian statistical approach. Some relevant theory and introductory concepts are presented using practical examples and two running case studies. The book also describes in detail how to perform health economic evaluations using the R package BCEA (Bayesian Cost-Effectiveness Analysis). BCEA can be used to post-process the results of a Bayesian cost-effectiveness model and perform advanced analyses producing standardised and highly customisable outputs. It presents all the features of the package, including its many functions and their practical application, as well as its user-friendly web interface. The book is a valuable resource for statisticians and practitioners working in the field of health economics wanting to simplify and standardise their workflow, for example in the preparation of dossiers in support of marketing authorisation, or acade...

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

Unit-of-Use Versus Traditional Bulk Packaging

Directory of Open Access Journals (Sweden)

Tiffany So

2012-01-01

Full Text Available Background: The choice between unit-of-use versus traditional bulk packaging in the US has long been a continuous debate for drug manufacturers and pharmacies in order to have the most efficient and safest practices. Understanding the benefits of using unit-of-use packaging over bulk packaging by US drug manufacturers in terms of workflow efficiency, economical costs and medication safety in the pharmacy is sometimes challenging.Methods: A time-saving study comparing the time saved using unit-of-use packaging versus bulk packaging, was examined. Prices between unit-of-use versus bulk packages were compared by using the Red Book: Pharmacy’s Fundamental Reference. Other articles were reviewed on the topics of counterfeiting, safe labeling, and implementation of unit-of-use packaging. Lastly, a cost-saving study was reviewed showing how medication adherence, due to improved packaging, could be cost-effective for patients.Results: When examining time, costs, medication adherence, and counterfeiting arguments, unit-of-use packaging proved to be beneficial for patients in all these terms.

Unit-of-Use Versus Traditional Bulk Packaging

Directory of Open Access Journals (Sweden)

Tiffany So

2012-01-01

Full Text Available Background: The choice between unit-of-use versus traditional bulk packaging in the US has long been a continuous debate for drug manufacturers and pharmacies in order to have the most efficient and safest practices. Understanding the benefits of using unit-of-use packaging over bulk packaging by US drug manufacturers in terms of workflow efficiency, economical costs and medication safety in the pharmacy is sometimes challenging. Methods: A time-saving study comparing the time saved using unit-of-use packaging versus bulk packaging, was examined. Prices between unit-of-use versus bulk packages were compared by using the Red Book: Pharmacy's Fundamental Reference. Other articles were reviewed on the topics of counterfeiting, safe labeling, and implementation of unit-of-use packaging. Lastly, a cost-saving study was reviewed showing how medication adherence, due to improved packaging, could be cost-effective for patients. Results: When examining time, costs, medication adherence, and counterfeiting arguments, unit-of-use packaging proved to be beneficial for patients in all these terms.   Type: Student Project

Periodic safety review of the HTR-10 safety analysis

International Nuclear Information System (INIS)

Chen Fubing; Zheng Yanhua; Shi Lei; Li Fu

2015-01-01

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

Development and experimental testing of the new safety-criticality Cristal package; Developpement et qualification experimentale du nouveau formulaire de surete-criticite Cristal

Energy Technology Data Exchange (ETDEWEB)

Mattera, Ch

1998-11-10

This thesis is concerned with Criticality-Safety studies related to the French Nuclear Fuel Cycle. We first describe the steps in the nuclear fuel cycle and the specific characteristics of these studies compared with those performed in Reactor Physics. In order to respond to the future requirements of the French Nuclear Program, we have developed a new package CRISTAL based on a recent cross sections library (CEA93) and the newest accurate codes (APOLLO2, MORET4, TRIPOLI4). The cristal system includes two calculations routes: a design route which will be used by French Industry (COGEMA/SGN) and a reference route.) To transfer this package to the French industry, we have elaborated calculation schemes for fissile solutions, dissolver media, transport casks and storage pools. Afterwards, these schemes have been used for the CRISTAL experimental validation. We have also contributed to the CRISTAL experimental database by reevaluating a French storage pool experiment: the CRISTO II experiment. This revaluation has been submitted to the OCDE working group in order that this experiment can be used by international criticality safety engineers to validate calculations methods. This work represents a large contribution to the recommendation of accurate calculation schemes and to the experimental validation of the CRISTAL package. These studies came up to the French Industry expectations. (author) 70 refs.

Safety analysis of nuclear power plants

International Nuclear Information System (INIS)

Selvatici, E.

1981-01-01

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

10 CFR 70.62 - Safety program and integrated safety analysis.

Science.gov (United States)

2010-01-01

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

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

International Nuclear Information System (INIS)

COLLOPY, M.T.

1999-01-01

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

iGC-an integrated analysis package of gene expression and copy number alteration.

Science.gov (United States)

Lai, Yi-Pin; Wang, Liang-Bo; Wang, Wei-An; Lai, Liang-Chuan; Tsai, Mong-Hsun; Lu, Tzu-Pin; Chuang, Eric Y

2017-01-14

With the advancement in high-throughput technologies, researchers can simultaneously investigate gene expression and copy number alteration (CNA) data from individual patients at a lower cost. Traditional analysis methods analyze each type of data individually and integrate their results using Venn diagrams. Challenges arise, however, when the results are irreproducible and inconsistent across multiple platforms. To address these issues, one possible approach is to concurrently analyze both gene expression profiling and CNAs in the same individual. We have developed an open-source R/Bioconductor package (iGC). Multiple input formats are supported and users can define their own criteria for identifying differentially expressed genes driven by CNAs. The analysis of two real microarray datasets demonstrated that the CNA-driven genes identified by the iGC package showed significantly higher Pearson correlation coefficients with their gene expression levels and copy numbers than those genes located in a genomic region with CNA. Compared with the Venn diagram approach, the iGC package showed better performance. The iGC package is effective and useful for identifying CNA-driven genes. By simultaneously considering both comparative genomic and transcriptomic data, it can provide better understanding of biological and medical questions. The iGC package's source code and manual are freely available at https://www.bioconductor.org/packages/release/bioc/html/iGC.html .

German competent authority guidance in FE methods applications for package design assessment

Energy Technology Data Exchange (ETDEWEB)

Voelzke, H.; Wieser, G.; Zencker, U.; Qiao Linan; Ballheimer, V. [Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin (Germany)

2004-07-01

The development of new methods in analysing package designs by using the finite element method (FEM) is of increasing importance. Package designers are more and more applying the growing opportunities of numerical methods to perform safety assessments for their products which requires suited methods also for competent authorities like BAM to verify the applicants' results. This presentation gives an topical overview of the experiences and tendencies within the complex field of finite element design testing. There are at first some general and more formal aspects concerning the correct finite element program selection and documentation of modelling, material properties, boundaries and calculation results including their interpretation. To give a reliable basis to the applicants in Germany BAM has drawn up and published a Finite Element Guideline recently. Secondly, the paper discusses actual technical questions which are of a wide interest and range from mechanical reflections on cask drop and extreme impact scenarios to thermal reflections on decay heat removal and fire scenarios. Examples from BAM work on FE-development activities are shown to demonstrate the great opportunities as well as the difficulties of using finite element methods for package safety analysis and design testing.

German competent authority guidance in FE methods applications for package design assessment

International Nuclear Information System (INIS)

Voelzke, H.; Wieser, G.; Zencker, U.; Qiao Linan; Ballheimer, V.

2004-01-01

The development of new methods in analysing package designs by using the finite element method (FEM) is of increasing importance. Package designers are more and more applying the growing opportunities of numerical methods to perform safety assessments for their products which requires suited methods also for competent authorities like BAM to verify the applicants' results. This presentation gives an topical overview of the experiences and tendencies within the complex field of finite element design testing. There are at first some general and more formal aspects concerning the correct finite element program selection and documentation of modelling, material properties, boundaries and calculation results including their interpretation. To give a reliable basis to the applicants in Germany BAM has drawn up and published a Finite Element Guideline recently. Secondly, the paper discusses actual technical questions which are of a wide interest and range from mechanical reflections on cask drop and extreme impact scenarios to thermal reflections on decay heat removal and fire scenarios. Examples from BAM work on FE-development activities are shown to demonstrate the great opportunities as well as the difficulties of using finite element methods for package safety analysis and design testing

An R package for the integrated analysis of metabolomics and spectral data.

Science.gov (United States)

Costa, Christopher; Maraschin, Marcelo; Rocha, Miguel

2016-06-01

Recently, there has been a growing interest in the field of metabolomics, materialized by a remarkable growth in experimental techniques, available data and related biological applications. Indeed, techniques as nuclear magnetic resonance, gas or liquid chromatography, mass spectrometry, infrared and UV-visible spectroscopies have provided extensive datasets that can help in tasks as biological and biomedical discovery, biotechnology and drug development. However, as it happens with other omics data, the analysis of metabolomics datasets provides multiple challenges, both in terms of methodologies and in the development of appropriate computational tools. Indeed, from the available software tools, none addresses the multiplicity of existing techniques and data analysis tasks. In this work, we make available a novel R package, named specmine, which provides a set of methods for metabolomics data analysis, including data loading in different formats, pre-processing, metabolite identification, univariate and multivariate data analysis, machine learning, and feature selection. Importantly, the implemented methods provide adequate support for the analysis of data from diverse experimental techniques, integrating a large set of functions from several R packages in a powerful, yet simple to use environment. The package, already available in CRAN, is accompanied by a web site where users can deposit datasets, scripts and analysis reports to be shared with the community, promoting the efficient sharing of metabolomics data analysis pipelines. Copyright © 2016 Elsevier Ireland Ltd. All rights reserved.

Development of waste packages for TRU-disposal. 5. Development of cylindrical metal package for TRU wastes

International Nuclear Information System (INIS)

Mine, Tatsuya; Mizubayashi, Hiroshi; Asano, Hidekazu; Owada, Hitoshi; Otsuki, Akiyoshi

2005-01-01

Development of the TRU waste package for hulls and endpieces compression canisters, which include long-lived and low sorption nuclides like C-14 is essential and will contribute a lot to a reasonable enhancement of safety and economy of the TRU-disposal system. The cylindrical metal package made of carbon steel for canisters to enhance the efficiency of the TRU-disposal system and to economically improve their stacking conditions was developed. The package is a welded cylindrical construction with a deep drawn upper cover and a disc plate for a bottom cover. Since the welding is mainly made only for an upper cover and a bottom disc plate, this package has a better containment performance for radioactive nuclide and can reduce the cost for construction and manufacturing including its welding control. Furthermore, this package can be laid down in pile for stacking in the circular cross section disposal tunnel for the sedimentary rock, which can drastically minimize the space for disposal tunnel as mentioned previously in TRU report. This paper reports the results of the study for application of newly developed metal package into the previous TRU-disposal system and for the stacking equipment for the package. (author)

Active and intelligent packaging systems for a modern society.

Science.gov (United States)

Realini, Carolina E; Marcos, Begonya

2014-11-01

Active and intelligent packaging systems are continuously evolving in response to growing challenges from a modern society. This article reviews: (1) the different categories of active and intelligent packaging concepts and currently available commercial applications, (2) latest packaging research trends and innovations, and (3) the growth perspectives of the active and intelligent packaging market. Active packaging aiming at extending shelf life or improving safety while maintaining quality is progressing towards the incorporation of natural active agents into more sustainable packaging materials. Intelligent packaging systems which monitor the condition of the packed food or its environment are progressing towards more cost-effective, convenient and integrated systems to provide innovative packaging solutions. Market growth is expected for active packaging with leading shares for moisture absorbers, oxygen scavengers, microwave susceptors and antimicrobial packaging. The market for intelligent packaging is also promising with strong gains for time-temperature indicator labels and advancements in the integration of intelligent concepts into packaging materials. Copyright © 2014 Elsevier Ltd. All rights reserved.

Material flow analysis for an industry - A case study in packaging

Science.gov (United States)

Amey, E.B.; Sandgren, K.

1996-01-01

The basic materials used in packaging are glass, metals (primarily aluminum and steel), an ever-growing range of plastics, paper and paperboard, wood, textiles for bags, and miscellaneous other materials (such as glues, inks, and other supplies). They are fabricated into rigid, semi-rigid, or flexible containers. The most common forms of these containers include cans, drums, bottles, cartons, boxes, bags, pouches, and wraps. Packaging products are, for the most part, low cost, bulky products that are manufactured close to their customers. There is virtually no import or export of packaging products. A material flow analysis can be developed that looks at all inputs to an industrial sector, inventories the losses in processing, and tracks the fate of the material after its useful life. An example is presented that identifies the material inputs to the packaging industry, and addresses the ultimate fate of the materials used. ?? 1996 International Association for Mathematical Geology.

IAEA Review for Gap Analysis of Safety Analysis Capability

International Nuclear Information System (INIS)

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

2014-01-01

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

Application of active packaging systems in probiotic foods

Directory of Open Access Journals (Sweden)

Renata Dobrucka

2013-09-01

Full Text Available Background: The packaging of the product has an important role in the protection of the stability of the final product. The use of active packaging system is due to play an increasingly important role by offering numerous and innovative solutions for extending the shelf-life or improve food quality and safety. Methods: On the basis of broad review of the current state of the art in world literature, application of packaging systems in probiotics foods was discussed. Results: In this study presented research and development in packaging systems for probiotics foods, using suitable materials with combine passive with active packaging solutions. Conclusion: Active packages with incorporated oxygen barrier materials or films with selective permeability properties also have potential applications in the packaging of probiotic food products. This is a broad field of research for scientists and industry.

SECOND WASTE PACKAGE PROBABILISTIC CRITICALITY ANALYSIS: GENERATION AND EVALUATION OF INTERNAL CRITICIALITY CONFIGURATIONS

Energy Technology Data Exchange (ETDEWEB)

P. Gottlieb, J.R. Massari, J.K. McCoy

1996-03-27

This analysis is prepared by the Mined Geologic Disposal System (MGDS) Waste Package Development (WPD) department to provide an evaluation of the criticality potential within a waste package having sonic or all of its contents degraded by corrosion and removal of neutron absorbers. This analysis is also intended to provide an estimate of the consequences of any internal criticality, particularly in terms of any increase in radionuclide inventory. These consequence estimates will be used as part of the WPD input to the Total System Performance Assessment. The ultimate objective of this analysis is to augment the information gained from the Initial Waste Package Probabilistic Criticality Analyses (Ref. 5.8 and 5.9, hereafter referred to as IPA) to a degree which will support preliminary waste package design recommendations intended to reduce the risk of waste package criticality and the risk to total repository system performance posed by the consequences of any criticality. The IPA evaluated the criticality potential under the assumption that the waste package basket retained its structural integrity, so that the assemblies retained their initial separation, even when the neutron absorbers had been leached from the basket. This analysis is based on the more realistic condition that removal of the neutron absorbers is a consequence of the corrosion of the steel in which they are contained, which has the additional consequence of reducing the structural support between assemblies. The result is a set of more reactive configurations having a smaller spacing between assemblies, or no inter-assembly spacing at all. Another difference from the IPA is the minimal attention to probabilistic evaluation given in this study. Although the IPA covered a time horizon to 100,000 years, the lack of consideration of basket degradation modes made it primarily applicable to the first 10,000 years. In contrast, this study, by focusing on the degraded modes of the basket, is primarily

Test and evaluation document for DOT Specification 7A Type A Packaging. Revision 3

Energy Technology Data Exchange (ETDEWEB)

NONE

1996-01-30

The US Department of Energy (DOE) has been conducting, through several of its operating contractors, an evaluation and testing program to qualify Type A radioactive material packagings per US Department of Transportation (DOT) Specification 7A (DOT-7A) of the Code of Federal Regulations (CFR), Title 49, Part 178 (49 CFR 178). The program is currently administered by the DOE, Office of Facility Safety Analysis, DOE/EH-32, at DOE-Headquarters (DOE-HQ) in Germantown, Maryland. This document summarizes the evaluation and testing performed for all of the packagings successfully qualified in this program.

Test and evaluation document for DOT Specification 7A Type A Packaging. Revision 3

International Nuclear Information System (INIS)

1996-01-01

The US Department of Energy (DOE) has been conducting, through several of its operating contractors, an evaluation and testing program to qualify Type A radioactive material packagings per US Department of Transportation (DOT) Specification 7A (DOT-7A) of the Code of Federal Regulations (CFR), Title 49, Part 178 (49 CFR 178). The program is currently administered by the DOE, Office of Facility Safety Analysis, DOE/EH-32, at DOE-Headquarters (DOE-HQ) in Germantown, Maryland. This document summarizes the evaluation and testing performed for all of the packagings successfully qualified in this program

Recommended nuclear criticality safety experiments in support of the safe transportation of fissile material

International Nuclear Information System (INIS)

Tollefson, D.A.; Elliott, E.P.; Dyer, H.R.; Thompson, S.A.

1993-01-01

Validation of computer codes and nuclear data (cross-section) libraries using benchmark quality critical (or certain subcritical) experiments is an essential part of a nuclear criticality safety evaluation. The validation results establish the credibility of the calculational tools for use in evaluating a particular application. Validation of the calculational tools is addressed in several American National Standards Institute/American Nuclear Society (ANSI/ANS) standards, with ANSI/ANS-8.1 being the most relevant. Documentation of the validation is a required part of all safety analyses involving significant quantities of fissile materials. In the case of transportation of fissile materials, the safety analysis report for packaging (SARP) must contain a thorough discussion of benchmark experiments, detailing how the experiments relate to the significant packaging and contents materials (fissile, moderating, neutron absorbing) within the package. The experiments recommended in this paper are needed to address certain areas related to transportation of unirradiated fissile materials in drum-type containers (packagings) for which current data are inadequate or are lacking