In-Package Chemistry Abstraction

DOE Office of Scientific and Technical Information (OSTI.GOV)

E. Thomas

2004-11-09

This report was developed in accordance with the requirements in ''Technical Work Plan for: Regulatory Integration Modeling and Analysis of the Waste Form and Waste Package'' (BSC 2004 [DIRS 171583]). The purpose of the in-package chemistry model is to predict the bulk chemistry inside of a breached waste package and to provide simplified expressions of that chemistry as function of time after breach to Total Systems Performance Assessment for the License Application (TSPA-LA). The scope of this report is to describe the development and validation of the in-package chemistry model. The in-package model is a combination of two models, amore » batch reactor model that uses the EQ3/6 geochemistry-modeling tool, and a surface complexation model that is applied to the results of the batch reactor model. The batch reactor model considers chemical interactions of water with the waste package materials and the waste form for commercial spent nuclear fuel (CSNF) waste packages and codisposed waste packages that contain both high-level waste glass (HLWG) and DOE spent fuel. The surface complexation model includes the impact of fluid-surface interactions (i.e., surface complexation) on the resulting fluid composition. The model examines two types of water influx: (1) the condensation of water vapor that diffuses into the waste package, and (2) seepage water that enters the waste package from the drift as a liquid. (1) Vapor Influx Case: The condensation of vapor onto the waste package internals is simulated as pure H2O and enters at a rate determined by the water vapor pressure for representative temperature and relative humidity conditions. (2) Water Influx Case: The water entering a waste package from the drift is simulated as typical groundwater and enters at a rate determined by the amount of seepage available to flow through openings in a breached waste package. TSPA-LA uses the vapor influx case for the nominal scenario for simulations where the waste package has been breached but the drip shield remains intact, so all of the seepage flow is diverted from the waste package. The chemistry from the vapor influx case is used to determine the stability of colloids and the solubility of radionuclides available for transport by diffusion, and to determine the degradation rates for the waste forms. TSPA-LA uses the water influx case for the seismic scenario, where the waste package has been breached and the drip shield has been damaged such that seepage flow is actually directed into the waste package. The chemistry from the water influx case that is a function of the flow rate is used to determine the stability of colloids and the solubility of radionuclides available for transport by diffusion and advection, and to determine the degradation rates for the CSNF and HLW glass. TSPA-LA does not use this model for the igneous scenario. Outputs from the in-package chemistry model implemented inside TSPA-LA include pH, ionic strength, and total carbonate concentration. These inputs to TSPA-LA will be linked to the following principle factors: dissolution rates of the CSNF and HLWG, dissolved concentrations of radionuclides, and colloid generation.« less

Decay heat of sodium fast reactor: Comparison of experimental measurements on the PHENIX reactor with calculations performed with the French DARWIN package

DOE Office of Scientific and Technical Information (OSTI.GOV)

Benoit, J. C.; Bourdot, P.; Eschbach, R.

2012-07-01

A Decay Heat (DH) experiment on the whole core of the French Sodium-Cooled Fast Reactor PHENIX has been conducted in May 2008. The measurements began an hour and a half after the shutdown of the reactor and lasted twelve days. It is one of the experiments used for the experimental validation of the depletion code DARWIN thereby confirming the excellent performance of the aforementioned code. Discrepancies between measured and calculated decay heat do not exceed 8%. (authors)

CORRELATIONS BETWEEN HOMOLOGUE CONCENTRATIONS OF PCDD/FS AND TOXIC EQUIVALENCY VALUES IN LABORATORY-, PACKAGE BOILER-, AND FIELD-SCALE INCINERATORS

EPA Science Inventory

The toxic equivalency (TEQ) values of polychlorinated dibenzo-p-dioxins and polychlorinated dibenzofurans (PCDD/Fs) are predicted with a model based on the homologue concentrations measured from a laboratory-scale reactor (124 data points), a package boiler (61 data points), and ...

Cleanup Verification Package for the 118-H-6:2, 105-H Reactor Ancillary Support Areas, Below-Grade Structures, and Underlying Soils; the 118-H-6:3, 105-H Reactor Fuel Storage Basin and Underlying Soils; The 118-H-6:3 Fuel Storage Basin Deep Zone Side Slope Soils; the 100-H-9, 100-H-10, and 100-H-13 French Drains; the 100-H-11 and 100-H-12 Expansion Box French Drains; and the 100-H-14 and 100-H-31 Surface Contamination Zones

DOE Office of Scientific and Technical Information (OSTI.GOV)

M. J. Appel

2006-06-29

This cleanup verification package documents completion of removal actions for the 105-H Reactor Ancillary Support Areas, Below-Grade Structures, and Underlying Soils (subsite 118-H-6:2); 105-H Reactor Fuel Storage Basin and Underlying Soils (118-H-6:3); and Fuel Storage Basin Deep Zone Side Slope Soils. This CVP also documents remedial actions for the following seven additional waste sties: French Drain C (100-H-9), French Drain D (100-H-10), Expansion Box French Drain E (100-H-11), Expansion Box French Drain F (100-H-12), French Drain G (100-H-13), Surface Contamination Zone H (100-H-14), and the Polychlorinated Biphenyl Surface Contamination Zone (100-H-31).

simBio: a Java package for the development of detailed cell models.

PubMed

Sarai, Nobuaki; Matsuoka, Satoshi; Noma, Akinori

2006-01-01

Quantitative dynamic computer models, which integrate a variety of molecular functions into a cell model, provide a powerful tool to create and test working hypotheses. We have developed a new modeling tool, the simBio package (freely available from ), which can be used for constructing cell models, such as cardiac cells (the Kyoto model from Matsuoka et al., 2003, 2004 a, b, the LRd model from Faber and Rudy, 2000, and the Noble 98 model from Noble et al., 1998), epithelial cells (Strieter et al., 1990) and pancreatic beta cells (Magnus and Keizer, 1998). The simBio package is written in Java, uses XML and can solve ordinary differential equations. In an attempt to mimic biological functional structures, a cell model is, in simBio, composed of independent functional modules called Reactors, such as ion channels and the sarcoplasmic reticulum, and dynamic variables called Nodes, such as ion concentrations. The interactions between Reactors and Nodes are described by the graph theory and the resulting graph represents a blueprint of an intricate cellular system. Reactors are prepared in a hierarchical order, in analogy to the biological classification. Each Reactor can be composed or improved independently, and can easily be reused for different models. This way of building models, through the combination of various modules, is enabled through the use of object-oriented programming concepts. Thus, simBio is a straightforward system for the creation of a variety of cell models on a common database of functional modules.

Black pepper powder microbiological quality improvement using DBD systems in atmospheric pressure

NASA Astrophysics Data System (ADS)

Grabowski, Maciej; Hołub, Marcin; Balcerak, Michał; Kalisiak, Stanisław; Dąbrowski, Waldemar

2015-07-01

Preliminary results are given regarding black pepper powder decontamination using dielectric barrier discharge (DBD) plasma in atmospheric pressure. Three different DBD reactor constructions were investigated, both packaged and unpackaged material was treated. Due to potential, industrial applications, in addition to microbiological results, water activity, loss of mass and the properties of packaging material, regarding barrier properties were investigated. Argon based treatment of packed pepper with DBD reactor configuration is proposed and satisfactory results are presented for treatment time of 5 min or less. Contribution to the topical issue "The 14th International Symposium on High Pressure Low Temperature Plasma Chemistry (HAKONE XIV)", edited by Nicolas Gherardi, Ronny Brandenburg and Lars Stollenwark

Dismantlement of the TSF-SNAP Reactor Assembly

DOE Office of Scientific and Technical Information (OSTI.GOV)

Peretz, Fred J

2009-01-01

This paper describes the dismantlement of the Tower Shielding Facility (TSF)?Systems for Nuclear Auxiliary Power (SNAP) reactor, a SNAP-10A reactor used to validate radiation source terms and shield performance models at Oak Ridge National Laboratory (ORNL) from 1967 through 1973. After shutdown, it was placed in storage at the Y-12 National Security Complex (Y-12), eventually falling under the auspices of the Highly Enriched Uranium (HEU) Disposition Program. To facilitate downblending of the HEU present in the fuel elements, the TSF-SNAP was moved to ORNL on June 24, 2006. The reactor assembly was removed from its packaging, inspected, and the sodium-potassiummore » (NaK) coolant was drained. A superheated steam process was used to chemically react the residual NaK inside the reactor assembly. The heat exchanger assembly was removed from the top of the reactor vessel, and the criticality safety sleeve was exchanged for a new safety sleeve that allowed for the removal of the vessel lid. A chain-mounted tubing cutter was used to separate the lid from the vessel, and the 36 fuel elements were removed and packaged in four U.S. Department of Transportation 2R/6M containers. The fuel elements were returned to Y-12 on July 13, 2006. The return of the fuel elements and disposal of all other reactor materials accomplished the formal objectives of the dismantlement project. In addition, a project model was established for the handling of a fully fueled liquid-metal?cooled reactor assembly. Current criticality safety codes have been benchmarked against experiments performed by Atomics International in the 1950s and 1960s. Execution of this project provides valuable experience applicable to future projects addressing space and liquid-metal-cooled reactors.« less

Reactor design and integration into a nuclear electric spacecraft

NASA Technical Reports Server (NTRS)

Phillips, W. M.; Koenig, D. R.

1978-01-01

One of the well-defined applications for nuclear power in space is nuclear electric propulsion (NEP). Mission studies have identified the optimum power level (400 kWe). A single Shuttle launch requirement and science-package integration have added additional constraints to the design. A reactor design which will meet these constraints has been studied. The reactor employs 90 fuel elements, each heat pipe cooled. Reactor control is obtained with BeO/B4C drums in a BeO reflector. The balance of the spacecraft is shielded from the reactor with LiH. Power conditioning and reactor control drum drives are located behind the LiH with the power conditioning. Launch safety, mechanical design and integration with the power conversion subsystem are discussed.

Low cost solar aray project: Experimental process system development unit for producing semiconductor-grade silicon using the silane-to-silicon process

NASA Technical Reports Server (NTRS)

1981-01-01

This phase consists of the engineering design, fabrication, assembly, operation, economic analysis, and process support R&D for an Experimental Process System Development Unit (EPSDU). The mechanical bid package was issued and the bid responses are under evaluation. Similarly, the electrical bid package was issued, however, responses are not yet due. The majority of all equipment is on order or has been received at the EPSDU site. The pyrolysis/consolidation process design package was issued. Preparation of process and instrumentation diagram for the free-space reactor was started. In the area of melting/consolidation, Kayex successfully melted chunk silicon and have produced silicon shot. The free-space reactor powder was successfully transported pneumatically from a storage bin to the auger feeder twenty-five feet up and was melted. The fluid-bed PDU has successfully operated at silane feed concentrations up to 21%. The writing of the operating manual has started. Overall, the design phase is nearing completion.

Work plan for improving the DARWIN2.3 depleted material balance calculation of nuclides of interest for the fuel cycle

NASA Astrophysics Data System (ADS)

Rizzo, Axel; Vaglio-Gaudard, Claire; Martin, Julie-Fiona; Noguère, Gilles; Eschbach, Romain

2017-09-01

DARWIN2.3 is the reference package used for fuel cycle applications in France. It solves the Boltzmann and Bateman equations in a coupling way, with the European JEFF-3.1.1 nuclear data library, to compute the fuel cycle values of interest. It includes both deterministic transport codes APOLLO2 (for light water reactors) and ERANOS2 (for fast reactors), and the DARWIN/PEPIN2 depletion code, each of them being developed by CEA/DEN with the support of its industrial partners. The DARWIN2.3 package has been experimentally validated for pressurized and boiling water reactors, as well as for sodium fast reactors; this experimental validation relies on the analysis of post-irradiation experiments (PIE). The DARWIN2.3 experimental validation work points out some isotopes for which the depleted concentration calculation can be improved. Some other nuclides have no available experimental validation, and their concentration calculation uncertainty is provided by the propagation of a priori nuclear data uncertainties. This paper describes the work plan of studies initiated this year to improve the accuracy of the DARWIN2.3 depleted material balance calculation concerning some nuclides of interest for the fuel cycle.

DOE Office of Scientific and Technical Information (OSTI.GOV)

None

Progress is reported on fundamental research in: crystal physics, reactions at metal surfaces, spectroscopy of ionic media, structure of metals, theory of alloying, physical properties, sintering, deformation of crystalline solids, x ray diffraction, metallurgy of superconducting materials, and electron microscope studies. Long-randge applied research studies were conducted for: zirconium metallurgy, materials compatibility, solid reactions, fuel element development, mechanical properties, non-destructive testing, and high-temperature materials. Reactor development support work was carried out for: gas-cooled reactor program, molten-salt reactor, high-flux isotope reactor, space-power program, thorium-utilization program, advanced-test reactor, Army Package Power Reactor, Enrico Fermi fast-breeder reactor, and water desalination program. Other programmore » activities, for which research was conducted, included: thermonuclear project, transuraniunn program, and post-irradiation examination laboratory. Separate abstracts were prepared for 30 sections of the report. (B.O.G.)« less

Chooz A, First Pressurized Water Reactor to be Dismantled in France - 13445

DOE Office of Scientific and Technical Information (OSTI.GOV)

Boucau, Joseph; Mirabella, C.; Nilsson, Lennart

2013-07-01

Nine commercial nuclear power plants have been permanently shut down in France to date, of which the Chooz A plant underwent an extensive decommissioning and dismantling program. Chooz Nuclear Power Station is located in the municipality of Chooz, Ardennes region, in the northeast part of France. Chooz B1 and B2 are 1,500 megawatt electric (MWe) pressurized water reactors (PWRs) currently in operation. Chooz A, a 305 MWe PWR implanted in two caves within a hill, began operations in 1967 and closed in 1991, and will now become the first PWR in France to be fully dismantled. EDF CIDEN (Engineering Centermore » for Dismantling and Environment) has awarded Westinghouse a contract for the dismantling of its Chooz A reactor vessel (RV). The project began in January 2010. Westinghouse is leading the project in a consortium with Nuvia France. The project scope includes overall project management, conditioning of the reactor vessel (RV) head, RV and RV internals segmentation, reactor nozzle cutting for lifting the RV out of the pit and seal it afterwards, dismantling of the RV thermal insulation, ALARA (As Low As Reasonably Achievable) forecast to ensure acceptable doses for the personnel, complementary vacuum cleaner to catch the chips during the segmentation work, needs and facilities, waste characterization and packaging, civil work modifications, licensing documentation. The RV and RV internals will be segmented based on the mechanical cutting technology that Westinghouse applied successfully for more than 13 years. The segmentation activities cover the cutting and packaging plan, tooling design and qualification, personnel training and site implementation. Since Chooz A is located inside two caves, the project will involve waste transportation from the reactor cave through long galleries to the waste buffer area. The project will end after the entire dismantling work is completed, and the waste storage is outside the caves and ready to be shipped either to the ANDRA (French National Radioactive Waste Management Agency) waste disposal facilities - (for low-level waste [LLW] and very low-level waste [VLLW], which are considered short lived) - or to the EDF Interim Storage Facility planned to be built on another site - (for low- and intermediate-level waste [LILW], which is considered long lived). The project has started with a detailed conceptual study that determines the step-by-step approach for dismantling the reactor and eventually supplying the packed containers ready for final disposal. All technical reports must be verified and approved by EDF and the French Nuclear Safety Authority before receiving the authorization to start the site work. The detailed conceptual study has been completed to date and equipment design and manufacturing is ongoing. This paper will present the conceptual design of the reactor internals segmentation and packaging process that will be implemented at Chooz A, including the planning, methodology, equipment, waste management, and packaging strategy. (authors)« less

324 Building B-Cell Pressurized Water Reactor Spent Fuel Packaging & Shipment RL Readiness Assessment Final Report [SEC 1 Thru 3

DOE Office of Scientific and Technical Information (OSTI.GOV)

HUMPHREYS, D C

A parallel readiness assessment (RA) was conducted by independent Fluor Hanford (FH) and U. S. Department of Energy, Richland Operations Office (RL) team to verify that an adequate state of readiness had been achieved for activities associated with the packaging and shipping of pressurized water reactor fuel assemblies from B-Cell in the 324 Building to the interim storage area at the Canister Storage Building in the 200 Area. The RL review was conducted in parallel with the FH review in accordance with the Joint RL/FH Implementation Plan (Appendix B). The RL RA Team members were assigned a FH RA Teammore » counterpart for the review. With this one-on-one approach, the RL RA Team was able to assess the FH Team's performance, competence, and adherence to the implementation plan and evaluate the level of facility readiness. The RL RA Team agrees with the FH determination that startup of the 324 Building B-Cell pressurized water reactor spent nuclear fuel packaging and shipping operations can safely proceed, pending completion of the identified pre-start items in the FH final report (see Appendix A), completion of the manageable list of open items included in the facility's declaration of readiness, and execution of the startup plan to operations.« less

Experimental validation of the DARWIN2.3 package for fuel cycle applications

DOE Office of Scientific and Technical Information (OSTI.GOV)

San-Felice, L.; Eschbach, R.; Bourdot, P.

2012-07-01

The DARWIN package, developed by the CEA and its French partners (AREVA and EDF) provides the required parameters for fuel cycle applications: fuel inventory, decay heat, activity, neutron, {gamma}, {alpha}, {beta} sources and spectrum, radiotoxicity. This paper presents the DARWIN2.3 experimental validation for fuel inventory and decay heat calculations on Pressurized Water Reactor (PWR). In order to validate this code system for spent fuel inventory a large program has been undertaken, based on spent fuel chemical assays. This paper deals with the experimental validation of DARWIN2.3 for the Pressurized Water Reactor (PWR) Uranium Oxide (UOX) and Mixed Oxide (MOX) fuelmore » inventory calculation, focused on the isotopes involved in Burn-Up Credit (BUC) applications and decay heat computations. The calculation - experiment (C/E-1) discrepancies are calculated with the latest European evaluation file JEFF-3.1.1 associated with the SHEM energy mesh. An overview of the tendencies is obtained on a complete range of burn-up from 10 to 85 GWd/t (10 to 60 GWcVt for MOX fuel). The experimental validation of the DARWIN2.3 package for decay heat calculation is performed using calorimetric measurements carried out at the Swedish Interim Spent Fuel Storage Facility for Pressurized Water Reactor (PWR) assemblies, covering a large burn-up (20 to 50 GWd/t) and cooling time range (10 to 30 years). (authors)« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

M. J. Appel

This cleanup verification package documents completion of remedial action for the 118-F-3, Minor Construction Burial Ground waste site. This site was an open field covered with cobbles, with no vegetation growing on the surface. The site received irradiated reactor parts that were removed during conversion of the 105-F Reactor from the Liquid 3X to the Ball 3X Project safety systems and received mostly vertical safety rod thimbles and step plugs.

Extension of the Bgl Broad Group Cross Section Library

NASA Astrophysics Data System (ADS)

Kirilova, Desislava; Belousov, Sergey; Ilieva, Krassimira

2009-08-01

The broad group cross-section libraries BUGLE and BGL are applied for reactor shielding calculation using the DOORS package based on discrete ordinates method and multigroup approximation of the neutron cross-sections. BUGLE and BGL libraries are problem oriented for PWR or VVER type of reactors respectively. They had been generated by collapsing the problem independent fine group library VITAMIN-B6 applying PWR and VVER one-dimensional radial model of the reactor middle plane using the SCALE software package. The surveillance assemblies (SA) of VVER-1000/320 are located on the baffle above the reactor core upper edge in a region where geometry and materials differ from those of the middle plane and the neutron field gradient is very high which would result in a different neutron spectrum. That is why the application of the fore-mentioned libraries for the neutron fluence calculation in the region of SA could lead to an additional inaccuracy. This was the main reason to study the necessity for an extension of the BGL library with cross-sections appropriate for the SA region. Comparative analysis of the neutron spectra of the SA region calculated by the VITAMIN-B6 and BGL libraries using the two-dimensional code DORT have been done with purpose to evaluate the BGL applicability for SA calculation.

Research reactor decommissioning experience - concrete removal and disposal -

DOE Office of Scientific and Technical Information (OSTI.GOV)

Manning, Mark R.; Gardner, Frederick W.

1990-07-01

Removal and disposal of neutron activated concrete from biological shields is the most significant operational task associated with research reactor decommissioning. During the period of 1985 thru 1989 Chem-Nuclear Systems, Inc. was the prime contractor for complete dismantlement and decommissioning of the Northrop TRIGA Mark F, the Virginia Tech Argonaut, and the Michigan State University TRIGA Mark I Reactor Facilities. This paper discusses operational requirements, methods employed, and results of the concrete removal, packaging, transport and disposal operations for these (3) research reactor decommissioning projects. Methods employed for each are compared. Disposal of concrete above and below regulatory release limitsmore » for unrestricted use are discussed. This study concludes that activated reactor biological shield concrete can be safely removed and buried under current regulations.« less

FFTF Passive Safety Test Data for Benchmarks for New LMR Designs

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wootan, David W.; Casella, Andrew M.

Liquid Metal Reactors (LMRs) continue to be considered as an attractive concept for advanced reactor design. Software packages such as SASSYS are being used to im-prove new LMR designs and operating characteristics. Significant cost and safety im-provements can be realized in advanced liquid metal reactor designs by emphasizing inherent or passive safety through crediting the beneficial reactivity feedbacks associ-ated with core and structural movement. This passive safety approach was adopted for the Fast Flux Test Facility (FFTF), and an experimental program was conducted to characterize the structural reactivity feedback. The FFTF passive safety testing pro-gram was developed to examine howmore » specific design elements influenced dynamic re-activity feedback in response to a reactivity input and to demonstrate the scalability of reactivity feedback results to reactors of current interest. The U.S. Department of En-ergy, Office of Nuclear Energy Advanced Reactor Technology program is in the pro-cess of preserving, protecting, securing, and placing in electronic format information and data from the FFTF, including the core configurations and data collected during the passive safety tests. Benchmarks based on empirical data gathered during operation of the Fast Flux Test Facility (FFTF) as well as design documents and post-irradiation examination will aid in the validation of these software packages and the models and calculations they produce. Evaluation of these actual test data could provide insight to improve analytical methods which may be used to support future licensing applications for LMRs« less

Manufacture and Testing of an Activation Foil Package for Use in AFIDS

DTIC Science & Technology

2005-03-01

Miller. Nuclides and Isotopes , 16th ed. Lockheed Martin, 2002. 4. Broadhead, Bryan. Sr. Development Staff, Reactor and Fuel Cycle Analysis ...alternative, the concept of using liquid nitrous oxide inside a reactor to simulate large volumes of air was investigated. Simulation using the...weapon. We analyzed whether N2O could replicate large volumes of air in neutron transport experiments since one cubic centimeter of liquid N2O

Irradiation Tests Supporting LEU Conversion of Very High Power Research Reactors in the US

DOE Office of Scientific and Technical Information (OSTI.GOV)

Woolstenhulme, N. E.; Cole, J. I.; Glagolenko, I.

The US fuel development team is developing a high density uranium-molybdenum alloy monolithic fuel to enable conversion of five high-power research reactors. Previous irradiation tests have demonstrated promising behavior for this fuel design. A series of future irradiation tests will enable selection of final fuel fabrication process and provide data to qualify the fuel at moderately-high power conditions for use in three of these five reactors. The remaining two reactors, namely the Advanced Test Reactor and High Flux Isotope Reactor, require additional irradiation tests to develop and demonstrate the fuel’s performance with even higher power conditions, complex design features, andmore » other unique conditions. This paper reviews the program’s current irradiation testing plans for these moderately-high irradiation conditions and presents conceptual testing strategies to illustrate how subsequent irradiation tests will build upon this initial data package to enable conversion of these two very-high power research reactors.« less

Testing of a Transport Cask for Research Reactor Spent Fuel - 13003

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mourao, Rogerio P.; Leite da Silva, Luiz; Miranda, Carlos A.

2013-07-01

Since the beginning of the last decade three Latin American countries that operate research reactors - Argentina, Brazil and Chile - have been joining efforts to improve the regional capability in the management of spent fuel elements from the TRIGA and MTR reactors operated in the region. A main drive in this initiative, sponsored by the International Atomic Energy Agency, is the fact that no definite solution regarding the back end of the research reactor fuel cycle has been taken by any of the participating country. However, any long-term solution - either disposition in a repository or storage away frommore » reactor - will involve at some stage the transportation of the spent fuel through public roads. Therefore, a licensed cask that provides adequate shielding, assurance of subcriticality, and conformance to internationally accepted safety, security and safeguards regimes is considered a strategic part of any future solution to be adopted at a regional level. As a step in this direction, a packaging for the transport of irradiated fuel for MTR and TRIGA research reactors was designed by the tri-national team and a half-scale model equipped with the MTR version of the internal basket was constructed in Argentina and Brazil and tested in Brazil. Three test campaigns have been carried out so far, covering both normal conditions of transportation and hypothetical accident conditions. After failing the tests in the first two test series, the specimen successfully underwent the last test sequence. A second specimen, incorporating the structural improvements in view of the previous tests results, will be tested in the near future. Numerical simulations of the free drop and thermal tests are being carried out in parallel, in order to validate the computational modeling that is going to be used as a support for the package certification. (authors)« less

Automation system for neutron activation analysis at the reactor IBR-2, Frank Laboratory of Neutron Physics, Joint Institute for Nuclear Research, Dubna, Russia.

PubMed

Pavlov, Sergey S; Dmitriev, Andrey Yu; Frontasyeva, Marina V

The present status of development of software packages and equipment designed for automation of NAA at the reactor IBR-2 of FLNP, JINR, Dubna, RF, is described. The NAA database, construction of sample changers and software for automation of spectra measurement and calculation of concentrations are presented. Automation of QC procedures is integrated in the software developed. Details of the design are shown.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Li, Meimei; Natesan, K.; Chen, Wei-Ying

This report provides an update on the understanding of the effect of sodium exposures on microstructure and tensile properties of Grade 91 (G91) steel in support of the design and operation of G91 components in sodium-cooled fast reactors (SFRs). The report is a Level 3 deliverable in FY17 (M3AT-17AN1602018), under the Work Package AT-17AN160201, “SFR Materials Testing” performed by the Argonne National Laboratory (ANL), as part of the Advanced Reactor Technologies Program.

AGC-2 Graphite Pre-irradiation Data Package

DOE Office of Scientific and Technical Information (OSTI.GOV)

David Swank; Joseph Lord; David Rohrbaugh

2010-08-01

The NGNP Graphite R&D program is currently establishing the safe operating envelope of graphite core components for a Very High Temperature Reactor (VHTR) design. The program is generating quantitative data necessary for predicting the behavior and operating performance of the new nuclear graphite grades. To determine the in-service behavior of the graphite for pebble bed and prismatic designs, the Advanced Graphite Creep (AGC) experiment is underway. This experiment is examining the properties and behavior of nuclear grade graphite over a large spectrum of temperatures, neutron fluences and compressive loads. Each experiment consists of over 400 graphite specimens that are characterizedmore » prior to irradiation and following irradiation. Six experiments are planned with the first, AGC-1, currently being irradiated in the Advanced Test Reactor (ATR) and pre-irradiation characterization of the second, AGC-2, completed. This data package establishes the readiness of 512 specimens for assembly into the AGC-2 capsule.« less

House OK's Russian aid

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rothstein, L.

1993-09-01

This article discusses the 2.5 Billion dollar aid package to Russia which House Appropriations Foreign Operations Subcommittee Chairman David Obey successfully defended on the House floor last June. Arizona Republican Jon Kyl offered an admendment that would cut 700 million from the package and was defeated with a 118 to 140 vote. The bill is currently in the hands of the Senate. The controversy over the bill and details concerning the aid package are discussed. The aid deal includes 250 million dollars for nuclear reactor safety and energy as well as environmental technical assistance, 655 million dollars to aid privatemore » sector development, and 704 million dollars for additional technical and economic assistance.« less

Methods and codes for neutronic calculations of the MARIA research reactor.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Andrzejewski, K.; Kulikowska, T.; Bretscher, M. M.

2002-02-18

The core of the MARIA high flux multipurpose research reactor is highly heterogeneous. It consists of beryllium blocks arranged in 6 x 8 matrix, tubular fuel assemblies, control rods and irradiation channels. The reflector is also heterogeneous and consists of graphite blocks clad with aluminum. Its structure is perturbed by the experimental beam tubes. This paper presents methods and codes used to calculate the MARIA reactor neutronics characteristics and experience gained thus far at IAE and ANL. At ANL the methods of MARIA calculations were developed in connection with the RERTR program. At IAE the package of programs was developedmore » to help its operator in optimization of fuel utilization.« less

Summary of the Advanced Reactor Design Criteria (ARDC) Phase 2 Activities

DOE Office of Scientific and Technical Information (OSTI.GOV)

Holbrook, Mark Raymond

This report provides an end-of-year summary reflecting the progress and status of proposed regulatory design criteria for advanced non-LWR designs in accordance with the Level 3 milestone in M3AT-15IN2001017 in work package AT-15IN200101. These criteria have been designated as ARDC, and they provide guidance to future applicants for addressing the GDC that are currently applied specifically to LWR designs. The report provides a summary of Phase 2 activities related to the various tasks associated with ARDC development and the subsequent development of example adaptations of ARDC for Sodium Fast Reactor (SFR) and modular High Temperature Gas-cooled Reactor (HTGR) designs.

Pyrolysis of aseptic packages (tetrapak) in a laboratory screw type reactor and secondary thermal/catalytic tar decomposition.

PubMed

Haydary, J; Susa, D; Dudáš, J

2013-05-01

Pyrolysis of aseptic packages (tetrapak cartons) in a laboratory apparatus using a flow screw type reactor and a secondary catalytic reactor for tar cracking was studied. The pyrolysis experiments were realized at temperatures ranging from 650 °C to 850 °C aimed at maximizing of the amount of the gas product and reducing its tar content. Distribution of tetrapak into the product yields at different conditions was obtained. The presence of H2, CO, CH4, CO2 and light hydrocarbons, HCx, in the gas product was observed. The Aluminum foil was easily separated from the solid product. The rest part of char was characterized by proximate and elemental analysis and calorimetric measurements. The total organic carbon in the tar product was estimated by elemental analysis of tars. Two types of catalysts (dolomite and red clay marked AFRC) were used for catalytic thermal tar decomposition. Three series of experiments (without catalyst in a secondary cracking reactor, with dolomite and with AFRC) at temperatures of 650, 700, 750, 800 and 850 °C were carried out. Both types of catalysts have significantly affected the content of tars and other components in pyrolytic gases. The effect of catalyst on the tetrapack distribution into the product yield on the composition of gas and on the total organic carbon in the tar product is presented in this work. Copyright © 2013 Elsevier Ltd. All rights reserved.

Reforming results of a novel radial reactor for a solid oxide fuel cell system with anode off-gas recirculation

NASA Astrophysics Data System (ADS)

Bosch, Timo; Carré, Maxime; Heinzel, Angelika; Steffen, Michael; Lapicque, François

2017-12-01

A novel reactor of a natural gas (NG) fueled, 1 kW net power solid oxide fuel cell (SOFC) system with anode off-gas recirculation (AOGR) is experimentally investigated. The reactor operates as pre-reformer, is of the type radial reactor with centrifugal z-flow, has the shape of a hollow cylinder with a volume of approximately 1 L and is equipped with two different precious metal wire-mesh catalyst packages as well as with an internal electric heater. Reforming investigations of the reactor are done stand-alone but as if the reactor would operate within the total SOFC system with AOGR. For the tests presented here it is assumed that the SOFC system runs on pure CH4 instead of NG. The manuscript focuses on the various phases of reactor operation during the startup process of the SOFC system. Startup process reforming experiments cover reactor operation points at which it runs on an oxygen to carbon ratio at the reactor inlet (ϕRI) of 1.2 with air supplied, up to a ϕRI of 2.4 without air supplied. As confirmed by a Monte Carlo simulation, most of the measured outlet gas concentrations are in or close to equilibrium.

Environmental impact assessment of a package type IFAS reactor during construction and operational phases: a life cycle approach.

PubMed

Singh, Nitin Kumar; Singh, Rana Pratap; Kazmi, Absar Ahmad

2017-05-01

In the present study, a life cycle assessment (LCA) approach was used to analyse the environmental impacts associated with the construction and operational phases of an integrated fixed-film activated sludge (IFAS) reactor treating municipal wastewater. This study was conducted within the boundaries of a research project that aimed to investigate the implementation related challenges of a package type IFAS reactor from an environmental perspective. Along with the LCA results of the construction phase, a comparison of the LCA results of seven operational phases is also presented in this study. The results showed that among all the inputs, the use of stainless steel in the construction phase caused the highest impact on environment, followed by electricity consumption in raw materials production. The impact of the construction phase on toxicity impact indicators was found to be significant compared to all operational phases. Among the seven operational phases of this study, the dissolved oxygen phase III, having a concentration of ∼4.5 mg/L, showed the highest impact on abiotic depletion, acidification, global warming, ozone layer depletion, human toxicity, fresh water eco-toxicity, marine aquatic eco-toxicity, terrestrial eco-toxicity, and photochemical oxidation. However, better effluent quality in this phase reduced the eutrophication load on environment.

Extreme temperature packaging: challenges and opportunities

NASA Astrophysics Data System (ADS)

Johnson, R. Wayne

2016-05-01

Consumer electronics account for the majority of electronics manufactured today. Given the temperature limits of humans, consumer electronics are typically rated for operation from -40°C to +85°C. Military applications extend the range to -65°C to +125°C while underhood automotive electronics may see +150°C. With the proliferation of the Internet of Things (IoT), the goal of instrumenting (sensing, computation, transmission) to improve safety and performance in high temperature environments such as geothermal wells, nuclear reactors, combustion chambers, industrial processes, etc. requires sensors, electronics and packaging compatible with these environments. Advances in wide bandgap semiconductors (SiC and GaN) allow the fabrication of high temperature compatible sensors and electronics. Integration and packaging of these devices is required for implementation into actual applications. The basic elements of packaging are die attach, electrical interconnection and the package or housing. Consumer electronics typically use conductive adhesives or low melting point solders for die attach, wire bonds or low melting solder for electrical interconnection and epoxy for the package. These materials melt or decompose in high temperature environments. This paper examines materials and processes for high temperature packaging including liquid transient phase and sintered nanoparticle die attach, high melting point wires for wire bonding and metal and ceramic packages. The limitations of currently available solutions will also be discussed.

Design of an ammonia closed-loop storage system in a CSP power plant with a power tower cavity receiver

NASA Astrophysics Data System (ADS)

Abdiwe, Ramadan; Haider, Markus

2017-06-01

In this study the thermochemical system using ammonia as energy storage carrier is investigated and a transient mathematical model using MATLAB software was developed to predict the behavior of the ammonia closed-loop storage system including but not limited to the ammonia solar reactor and the ammonia synthesis reactor. The MATLAB model contains transient mass and energy balances as well as chemical equilibrium model for each relevant system component. For the importance of the dissociation and formation processes in the system, a Computational Fluid Dynamics (CFD) simulation on the ammonia solar and synthesis reactors has been performed. The CFD commercial package FLUENT is used for the simulation study and all the important mechanisms for packed bed reactors are taken into account, such as momentum, heat and mass transfer, and chemical reactions. The FLUENT simulation reveals the profiles inside both reactors and compared them with the profiles from the MATLAB code.

Interim waste storage for the Integral Fast Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Benedict, R.W.; Phipps, R.D.; Condiff, D.W.

1991-01-01

The Integral Fast Reactor (IFR), which Argonne National Laboratory is developing, is an innovative liquid metal breeder reactor that uses metallic fuel and has a close coupled fuel recovery process. A pyrochemical process is used to separate the fission products from the actinide elements. These actinides are used to make new fuel for the reactor. As part of the overall IFR development program, Argonne has refurbished an existing Fuel Cycle Facility at ANL-West and is installing new equipment to demonstrate the remote reprocessing and fabrication of fuel for the Experimental Breeder Reactor II (EBR-II). During this demonstration the wastes thatmore » are produced will be treated and packaged to produce waste forms that would be typical of future commercial operations. These future waste forms would, assuming Argonne development goals are fulfilled, be essentially free of long half-life transuranic isotopes. Promising early results indicate that actinide extraction processes can be developed to strip these isotopes from waste stream and return them to the IFR type reactors for fissioning. 1 fig.« less

Thermal-Mechanical Stress Analysis of PWR Pressure Vessel and Nozzles under Grid Load-Following Mode: Interim Report on the Effect of Cyclic Hardening Material Properties and Pre-existing Cracks on Stress Analysis Results

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mohanty, Subhasish; Soppet, William; Majumdar, Saurin

This report provides an update on an assessment of environmentally assisted fatigue for light water reactor components under extended service conditions. This report is a deliverable under the work package for environmentally assisted fatigue as part of DOE’s Light Water Reactor Sustainability Program. In a previous report (September 2015), we presented tensile and fatigue test data and related hardening material properties for 508 low-alloys steel base metal and other reactor metals. In this report, we present thermal-mechanical stress analysis of the reactor pressure vessel and its hot-leg and cold-leg nozzles based on estimated material properties. We also present results frommore » thermal and thermal-mechanical stress analysis under reactor heat-up, cool-down, and grid load-following conditions. Analysis results are given with and without the presence of preexisting cracks in the reactor nozzles (axial or circumferential crack). In addition, results from validation stress analysis based on tensile and fatigue experiments are reported.« less

PR-EDB: Power Reactor Embrittlement Database - Version 3

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wang, Jy-An John; Subramani, Ranjit

2008-03-01

The aging and degradation of light-water reactor pressure vessels is of particular concern because of their relevance to plant integrity and the magnitude of the expected irradiation embrittlement. The radiation embrittlement of reactor pressure vessel materials depends on many factors, such as neutron fluence, flux, and energy spectrum, irradiation temperature, and preirradiation material history and chemical compositions. These factors must be considered to reliably predict pressure vessel embrittlement and to ensure the safe operation of the reactor. Large amounts of data from surveillance capsules are needed to develop a generally applicable damage prediction model that can be used for industrymore » standards and regulatory guides. Furthermore, the investigations of regulatory issues such as vessel integrity over plant life, vessel failure, and sufficiency of current codes, Standard Review Plans (SRPs), and Guides for license renewal can be greatly expedited by the use of a well-designed computerized database. The Power Reactor Embrittlement Database (PR-EDB) is such a comprehensive collection of data for U.S. designed commercial nuclear reactors. The current version of the PR-EDB lists the test results of 104 heat-affected-zone (HAZ) materials, 115 weld materials, and 141 base materials, including 103 plates, 35 forgings, and 3 correlation monitor materials that were irradiated in 321 capsules from 106 commercial power reactors. The data files are given in dBASE format and can be accessed with any personal computer using the Windows operating system. "User-friendly" utility programs have been written to investigate radiation embrittlement using this database. Utility programs allow the user to retrieve, select and manipulate specific data, display data to the screen or printer, and fit and plot Charpy impact data. The PR-EDB Version 3.0 upgrades Version 2.0. The package was developed based on the Microsoft .NET framework technology and uses Microsoft Access for backend data storage, and Microsoft Excel for plotting graphs. This software package is compatible with Windows (98 or higher) and has been built with a highly versatile user interface. PR-EDB Version 3.0 also contains an "Evaluated Residual File" utility for generating the evaluated processed files used for radiation embrittlement study.« less

Excore Modeling with VERAShift

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pandya, Tara M.; Evans, Thomas M.

It is important to be able to accurately predict the neutron flux outside the immediate reactor core for a variety of safety and material analyses. Monte Carlo radiation transport calculations are required to produce the high fidelity excore responses. Under this milestone VERA (specifically the VERAShift package) has been extended to perform excore calculations by running radiation transport calculations with Shift. This package couples VERA-CS with Shift to perform excore tallies for multiple state points concurrently, with each component capable of parallel execution on independent domains. Specifically, this package performs fluence calculations in the core barrel and vessel, or, performsmore » the requested tallies in any user-defined excore regions. VERAShift takes advantage of the general geometry package in Shift. This gives VERAShift the flexibility to explicitly model features outside the core barrel, including detailed vessel models, detectors, and power plant details. A very limited set of experimental and numerical benchmarks is available for excore simulation comparison. The Consortium for the Advanced Simulation of Light Water Reactors (CASL) has developed a set of excore benchmark problems to include as part of the VERA-CS verification and validation (V&V) problems. The excore capability in VERAShift has been tested on small representative assembly problems, multiassembly problems, and quarter-core problems. VERAView has also been extended to visualize these vessel fluence results from VERAShift. Preliminary vessel fluence results for quarter-core multistate calculations look very promising. Further development is needed to determine the details relevant to excore simulations. Validation of VERA for fluence and excore detectors still needs to be performed against experimental and numerical results.« less

Proceedings of the Conference on High-temperature Electronics

NASA Technical Reports Server (NTRS)

1981-01-01

The development of electronic devices for use in high temperature environments is addressed. The instrumentational needs of planetary exploration, fossil and nuclear power reactors, turbine engine monitoring, and well logging are defined. Emphasis is place on the fabrication and performance of materials and semiconductor devices, circuits and systems and packaging.

NDI (Nondestructive Inspection) Oriented Corrosion Control for Army Aircraft. Phase 1. Inspection Methods

DTIC Science & Technology

1989-07-01

Appendices A and B and are provided as cover sheets from each item rather than completc packages. The Pamplet Series materials were furnished as camera-ready...34Stational Neutron Radiography System for Aircraft Reliability and Maintainability." G. A. Technologies Brochure , Triga Reactor Division, San Diego

Development of Ultra-Fine Multigroup Cross Section Library of the AMPX/SCALE Code Packages

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jeon, Byoung Kyu; Sik Yang, Won; Kim, Kang Seog

The Consortium for Advanced Simulation of Light Water Reactors Virtual Environment for Reactor Applications (VERA) neutronic simulator MPACT is being developed by Oak Ridge National Laboratory and the University of Michigan for various reactor applications. The MPACT and simplified MPACT 51- and 252-group cross section libraries have been developed for the MPACT neutron transport calculations by using the AMPX and Standardized Computer Analyses for Licensing Evaluations (SCALE) code packages developed at Oak Ridge National Laboratory. It has been noted that the conventional AMPX/SCALE procedure has limited applications for fast-spectrum systems such as boiling water reactor (BWR) fuels with very highmore » void fractions and fast reactor fuels because of its poor accuracy in unresolved and fast energy regions. This lack of accuracy can introduce additional error sources to MPACT calculations, which is already limited by the Bondarenko approach for resolved resonance self-shielding calculation. To enhance the prediction accuracy of MPACT for fast-spectrum reactor analyses, the accuracy of the AMPX/SCALE code packages should be improved first. The purpose of this study is to identify the major problems of the AMPX/SCALE procedure in generating fast-spectrum cross sections and to devise ways to improve the accuracy. For this, various benchmark problems including a typical pressurized water reactor fuel, BWR fuels with various void fractions, and several fast reactor fuels were analyzed using the AMPX 252-group libraries. Isotopic reaction rates were determined by SCALE multigroup (MG) calculations and compared with continuous energy (CE) Monte Carlo calculation results. This reaction rate analysis revealed three main contributors to the observed differences in reactivity and reaction rates: (1) the limitation of the Bondarenko approach in coarse energy group structure, (2) the normalization issue of probability tables, and (3) neglect of the self-shielding effect of resonance-like cross sections at high energy range such as (n,p) cross section of Cl35. The first error source can be eliminated by an ultra-fine group (UFG) structure in which the broad scattering resonances of intermediate-weight nuclides can be represented accurately by a piecewise constant function. A UFG AMPX library was generated with modified probability tables and tested against various benchmark problems. The reactivity and reaction rates determined with the new UFG AMPX library agreed very well with respect to Monte Carlo Neutral Particle (MCNP) results. To enhance the lattice calculation accuracy without significantly increasing the computational time, performing the UFG lattice calculation in two steps was proposed. In the first step, a UFG slowing-down calculation is performed for the corresponding homogenized composition, and UFG cross sections are collapsed into an intermediate group structure. In the second step, the lattice calculation is performed for the intermediate group level using the condensed group cross sections. A preliminary test showed that the condensed library reproduces the results obtained with the UFG cross section library. This result suggests that the proposed two-step lattice calculation approach is a promising option to enhance the applicability of the AMPX/SCALE system to fast system analysis.« less

Mixed Oxide Fresh Fuel Package Auxiliary Equipment

DOE Office of Scientific and Technical Information (OSTI.GOV)

Yapuncich, F.; Ross, A.; Clark, R.H.

2008-07-01

The United States Department of Energy's National Nuclear Security Administration (NNSA) is overseeing the construction the Mixed Oxide (MOX) Fuel Fabrication Facility (MFFF) on the Savannah River Site. The new facility, being constructed by NNSA's contractor Shaw AREVA MOX Services, will fabricate fuel assemblies utilizing surplus plutonium as feedstock. The fuel will be used in designated commercial nuclear reactors. The MOX Fresh Fuel Package (MFFP), which has recently been licensed by the Nuclear Regulatory Commission (NRC) as a type B package (USA/9295/B(U)F-96), will be utilized to transport the fabricated fuel assemblies from the MFFF to the nuclear reactors. It wasmore » necessary to develop auxiliary equipment that would be able to efficiently handle the high precision fuel assemblies. Also, the physical constraints of the MFFF and the nuclear power plants require that the equipment be capable of loading and unloading the fuel assemblies both vertically and horizontally. The ability to reconfigure the load/unload evolution builds in a large degree of flexibility for the MFFP for the handling of many types of both fuel and non fuel payloads. The design and analysis met various technical specifications including dynamic and static seismic criteria. The fabrication was completed by three major fabrication facilities within the United States. The testing was conducted by Sandia National Laboratories. The unique design specifications and successful testing sequences will be discussed. (authors)« less

Qualification of Simulation Software for Safety Assessment of Sodium Cooled Fast Reactors. Requirements and Recommendations

DOE Office of Scientific and Technical Information (OSTI.GOV)

Brown, Nicholas R.; Pointer, William David; Sieger, Matt

2016-04-01

The goal of this review is to enable application of codes or software packages for safety assessment of advanced sodium-cooled fast reactor (SFR) designs. To address near-term programmatic needs, the authors have focused on two objectives. First, the authors have focused on identification of requirements for software QA that must be satisfied to enable the application of software to future safety analyses. Second, the authors have collected best practices applied by other code development teams to minimize cost and time of initial code qualification activities and to recommend a path to the stated goal.

Design and optimization of the heat rejection system for a liquid cooled thermionic space nuclear reactor power system

DOE Office of Scientific and Technical Information (OSTI.GOV)

Moriarty, M.P.

1993-01-15

The heat transport subsystem for a liquid metal cooled thermionic space nuclear power system was modelled using algorithms developed in support of previous nuclear power system study programs, which date back to the SNAP-10A flight system. The model was used to define the optimum dimensions of the various components in the heat transport subsystem subjected to the constraints of minimizing mass and achieving a launchable package that did not require radiator deployment. The resulting design provides for the safe and reliable cooling of the nuclear reactor in a proven lightweight design.

Design and optimization of the heat rejection system for a liquid cooled thermionic space nuclear reactor power system

NASA Astrophysics Data System (ADS)

Moriarty, Michael P.

1993-01-01

The heat transport subsystem for a liquid metal cooled thermionic space nuclear power system was modelled using algorithms developed in support of previous nuclear power system study programs, which date back to the SNAP-10A flight system. The model was used to define the optimum dimensions of the various components in the heat transport subsystem subjected to the constraints of minimizing mass and achieving a launchable package that did not require radiator deployment. The resulting design provides for the safe and reliable cooling of the nuclear reactor in a proven lightweight design.

Environmental Effect on Evolutionary Cyclic Plasticity Material Parameters of 316 Stainless Steel: An Experimental & Material Modeling Approach

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mohanty, Subhasish; Soppet, William K.; Majumdar, Saurin

2014-09-20

This report provides an update on an earlier assessment of environmentally assisted fatigue for light water reactor (LWR) materials under extended service conditions. This report is a deliverable under the work package for environmentally assisted fatigue in the Light Water Reactor Sustainability (LWRS) program. The overall objective of this LWRS project is to assess the degradation by environmentally assisted cracking/fatigue of LWR materials such as various alloy base metals and their welds used in reactor coolant system piping. This effort is to support the Department of Energy LWRS program for developing tools to understand the aging/failure mechanism and to predictmore » the remaining life of LWR components for anticipated 60-80 year operation.« less

Performance of compact fast pyrolysis reactor with Auger-type modules for the continuous liquid biofuel production

NASA Astrophysics Data System (ADS)

Nishimura, Shun; Ebitani, Kohki

2018-01-01

Development of a compact fast pyrolysis reactor constructed using Auger-type technology to afford liquid biofuel with high yield has been an interesting concept in support of local production for local consumption. To establish a widely useable module package, details of the performance of the developing compact module reactor were investigated. This study surveyed the properties of as-produced pyrolysis oil as a function of operation time, and clarified the recent performance of the developing compact fast pyrolysis reactor. Results show that after condensation in the scrubber collector, e.g. approx. 10 h for a 25 kg/h feedstock rate, static performance of pyrolysis oil with approximately 20 MJ/kg (4.8 kcal/g) calorific values were constantly obtained after an additional 14 h. The feeding speed of cedar chips strongly influenced the time for oil condensation process: i.e. 1.6 times higher feeding speed decreased the condensation period by half (approx. 5 h in the case of 40 kg/h). Increasing the reactor throughput capacity is an important goal for the next stage in the development of a compact fast pyrolysis reactor with Auger-type modules.

Thermal Hysteresis of MEMS Packaged Capacitive Pressure Sensor (CPS) Based 3C-SiC

NASA Astrophysics Data System (ADS)

Marsi, N.; Majlis, B. Y.; Mohd-Yasin, F.; Hamzah, A. A.; Mohd Rus, A. Z.

2016-11-01

Presented herein are the effects of thermal hysteresis analyses of the MEMS packaged capacitive pressure sensor (CPS). The MEMS CPS was employed on Si-on-3C-SiC wafer that was performed using the hot wall low-pressure chemical vapour deposition (LPCVD) reactors at the Queensland Micro and Nanotechnology Center (QMNC), Griffith University and fabricated using the bulk-micromachining process. The MEMS CPS was operated at an extreme temperature up to 500°C and high external pressure at 5.0 MPa. The thermal hysteresis phenomenon that causes the deflection, strain and stress on the 3C-SiC diaphragm spontaneously influence the MEMS CPS performances. The differences of temperature, hysteresis, and repeatability test were presented to demonstrate the functionality of the MEMS packaged CPS. As expected, the output hysteresis has a low hysteresis (less than 0.05%) which has the hardness greater than the traditional silicon. By utilizing this low hysteresis, it was revealed that the MEMS packaged CPS has high repeatability and stability of the sensor.

Design, Operation, and Modeling of a Vertical APCVD Reactor for Silicon Carbide Film Growth

NASA Technical Reports Server (NTRS)

DeAnna, Russell G.; Fleischman, Aaron J.; Zorman, Christian A.; Mehregany, Mehran

1998-01-01

An atmospheric pressure chemical vapor deposition (APCVD) reactor utilizing a unique vertical geometry which enables 3C-SiC films to be grown on two, 4-inch diameter Si wafers has been constructed. Contrary to expectations, 3C-SiC films grown in this reactor are thickest at the downstream end of the substrates. To better understand the reason for the thickness distribution on the wafers, an axisymmetric finite-element model of the gas flow in the reactor was constructed. The model uses the ANSYS53 Flowtran package and includes compressible and temperature-dependent fluid properties in laminar or turbulent flow. It does not include reaction chemistry or unsteady flow. The ANSYS53 results predict that the cool, inlet fluid falls through the inlet pipe and the warm, diffuser region like a jet. This jet impinges on top of the susceptor and gets diverted to the reactor side walls, where it flows to the bottom of the reactor, turns, and slowly rises along the face of the susceptor. This may explain why the SiC films are thickest at the downstream side of the wafers, as gas containing fresh reactants first passes over this region. Modeling results are presented for both one atmosphere and one half atmosphere reactor pressure.

76 FR 80410 - Advisory Committee on Reactor Safeguards; Meeting of the ACRS Subcommittee on Radiation...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-12-23

... Compliance with Packaging Requirements for Shipment and Receipt of Radioactive Material.'' The Subcommittee... Subcommittee on Radiation Protection and Nuclear Materials; Notice of Meeting The ACRS Subcommittee on Radiation Protection and Nuclear Materials will hold a meeting on January 18, 2012, Room T-2B3, 11545...

Measurement and Analysis of Structural Integrity of Reactor Core Support Structure in Pressurized Water Reactor (PWR) Plant

NASA Astrophysics Data System (ADS)

Ansari, Saleem A.; Haroon, Muhammad; Rashid, Atif; Kazmi, Zafar

2017-02-01

Extensive calculation and measurements of flow-induced vibrations (FIV) of reactor internals were made in a PWR plant to assess the structural integrity of reactor core support structure against coolant flow. The work was done to meet the requirements of the Fukushima Response Action Plan (FRAP) for enhancement of reactor safety, and the regulatory guide RG-1.20. For the core surveillance measurements the Reactor Internals Vibration Monitoring System (IVMS) has been developed based on detailed neutron noise analysis of the flux signals from the four ex-core neutron detectors. The natural frequencies, displacement and mode shapes of the reactor core barrel (CB) motion were determined with the help of IVMS. The random pressure fluctuations in reactor coolant flow due to turbulence force have been identified as the predominant cause of beam-mode deflection of CB. The dynamic FIV calculations were also made to supplement the core surveillance measurements. The calculational package employed the computational fluid dynamics, mode shape analysis, calculation of power spectral densities of flow & pressure fields and the structural response to random flow excitation forces. The dynamic loads and stiffness of the Hold-Down Spring that keeps the core structure in position against upward coolant thrust were also determined by noise measurements. Also, the boron concentration in primary coolant at any time of the core cycle has been determined with the IVMS.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Haydary, J., E-mail: juma.haydary@stuba.sk; Susa, D.; Dudáš, J.

Highlights: ► Pyrolysis of aseptic packages was carried out in a laboratory flow reactor. ► Distribution of tetrapak into the product yields was obtained. ► Composition of the pyrolysis products was estimated. ► Secondary thermal and catalytic decomposition of tars was studied. ► Two types of catalysts (dolomite and red clay marked AFRC) were used. - Abstract: Pyrolysis of aseptic packages (tetrapak cartons) in a laboratory apparatus using a flow screw type reactor and a secondary catalytic reactor for tar cracking was studied. The pyrolysis experiments were realized at temperatures ranging from 650 °C to 850 °C aimed at maximizingmore » of the amount of the gas product and reducing its tar content. Distribution of tetrapak into the product yields at different conditions was obtained. The presence of H{sub 2}, CO, CH{sub 4}, CO{sub 2} and light hydrocarbons, HCx, in the gas product was observed. The Aluminum foil was easily separated from the solid product. The rest part of char was characterized by proximate and elemental analysis and calorimetric measurements. The total organic carbon in the tar product was estimated by elemental analysis of tars. Two types of catalysts (dolomite and red clay marked AFRC) were used for catalytic thermal tar decomposition. Three series of experiments (without catalyst in a secondary cracking reactor, with dolomite and with AFRC) at temperatures of 650, 700, 750, 800 and 850 °C were carried out. Both types of catalysts have significantly affected the content of tars and other components in pyrolytic gases. The effect of catalyst on the tetrapack distribution into the product yield on the composition of gas and on the total organic carbon in the tar product is presented in this work.« less

Neutron-Irradiated Samples as Test Materials for MPEX

DOE PAGES

Ellis, Ronald James; Rapp, Juergen

2015-10-09

Plasma Material Interaction (PMI) is a major concern in fusion reactor design and analysis. The Material-Plasma Exposure eXperiment (MPEX) will explore PMI under fusion reactor plasma conditions. Samples with accumulated displacements per atom (DPA) damage produced by fast neutron irradiations in the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL) will be studied in the MPEX facility. This paper presents assessments of the calculated induced radioactivity and resulting radiation dose rates of a variety of potential fusion reactor plasma-facing materials (such as tungsten). The scientific code packages MCNP and SCALE were used to simulate irradiation of themore » samples in HFIR including the generation and depletion of nuclides in the material and the subsequent composition, activity levels, gamma radiation fields, and resultant dose rates as a function of cooling time. A challenge of the MPEX project is to minimize the radioactive inventory in the preparation of the samples and the sample dose rates for inclusion in the MPEX facility.« less

Theoretical study of nanoparticle formation in thermal plasma processing: Nucleation, coagulation and aggregation

NASA Astrophysics Data System (ADS)

Mendoza Gonzalez, Norma Yadira

This work presents a mathematical modeling study of the synthesis of nanoparticles in radio frequency (RF) inductively coupled plasma (ICP) reactors. The purpose is to further investigate the influence of process parameters on the final size and morphology of produced particles. The proposed model involves the calculation of flow and temperature fields of the plasma gas. Evaporation of raw particles is also accounted with the particle trajectory and temperature history calculated with a Lagrangian approach. The nanoparticle formation is considered by homogeneous nucleation and the growth is caused by condensation and Brownian coagulation. The growth of fractal aggregates is considered by introducing a power law exponent Df. Transport of nanoparticles occurs by convection, thermophoresis and Brownian diffusion. The method of moments is used to solve the particle dynamics equation. The model is validated using experimental results from plasma reactors at laboratory scale. The results are presented in the following manner. First, use is made of the computational fluid dynamics software (CFD), Fluent 6.1 with a commercial companion package specifically developped for aerosols named: Fine Particle Model (FPM). This package is used to study the relationship between the operating parameters effect and the properties of the end products at the laboratory scale. Secondly, a coupled hybrid model for the synthesis of spherical particles and fractal aggregates is developped in place of the FPM package. Results obtained from this model will allow to identify the importance of each parameter in defining the morphology of spherical primary particles and fractal aggregates of nanoparticles. The solution of the model was made using the geometries and operating conditions of existing reactors at the Centre de Recherche en Energie, Plasma et Electrochimie (CREPE) of the Universite de Sherbrooke, for which experimental results were obtained experimentally. Additionally, this study demonstrates the importance of the flow and temperature fields on the growth of fractal particles; namely the aggregates.

Improved hydrocracker temperature control: Mobil quench zone technology

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sarli, M.S.; McGovern, S.J.; Lewis, D.W.

1993-01-01

Hydrocracking is a well established process in the oil refining industry. There are over 2.7 million barrels of installed capacity world-wide. The hydrocracking process comprises several families of highly exothermic reactions and the total adiabatic temperature rise can easily exceed 200 F. Reactor temperature control is therefore very important. Hydrocracking reactors are typically constructed with multiple catalyst beds in series. Cold recycle gas is usually injected between the catalyst beds to quench the reactions, thereby controlling overall temperature rise. The design of this quench zone is the key to good reactor temperature control, particularly when processing poorer quality, i.e., highermore » heat release, feeds. Mobil Research and Development Corporation (MRDC) has developed a robust and very effective quench zone technology (QZT) package, which is now being licensed to the industry for hydrocracking applications.« less

Thermal valorization of post-consumer film waste in a bubbling bed gasifier

DOE Office of Scientific and Technical Information (OSTI.GOV)

Martínez-Lera, S., E-mail: susanamartinezlera@gmail.com; Torrico, J.; Pallarés, J.

2013-07-15

Highlights: • Film waste from packaging is a common waste, a fraction of which is not recyclable. • Gasification can make use of the high energy value of the non-recyclable fraction. • This waste and two reference polymers were gasified in a bubbling bed reactor. • This experimental research proves technical feasibility of the process. • It also analyzes impact of composition and ER on the performance of the plant. - Abstract: The use of plastic bags and film packaging is very frequent in manifold sectors and film waste is usually present in different sources of municipal and industrial wastes.more » A significant part of it is not suitable for mechanical recycling but could be safely transformed into a valuable gas by means of thermal valorization. In this research, the gasification of film wastes has been experimentally investigated through experiments in a fluidized bed reactor of two reference polymers, polyethylene and polypropylene, and actual post-consumer film waste. After a complete experimental characterization of the three materials, several gasification experiments have been performed to analyze the influence of the fuel and of equivalence ratio on gas production and composition, on tar generation and on efficiency. The experiments prove that film waste and analogue polymer derived wastes can be successfully gasified in a fluidized bed reactor, yielding a gas with a higher heating value in a range from 3.6 to 5.6 MJ/m{sup 3} and cold gas efficiencies up to 60%.« less

Cooling Performance Analysis of ThePrimary Cooling System ReactorTRIGA-2000Bandung

NASA Astrophysics Data System (ADS)

Irianto, I. D.; Dibyo, S.; Bakhri, S.; Sunaryo, G. R.

2018-02-01

The conversion of reactor fuel type will affect the heat transfer process resulting from the reactor core to the cooling system. This conversion resulted in changes to the cooling system performance and parameters of operation and design of key components of the reactor coolant system, especially the primary cooling system. The calculation of the operating parameters of the primary cooling system of the reactor TRIGA 2000 Bandung is done using ChemCad Package 6.1.4. The calculation of the operating parameters of the cooling system is based on mass and energy balance in each coolant flow path and unit components. Output calculation is the temperature, pressure and flow rate of the coolant used in the cooling process. The results of a simulation of the performance of the primary cooling system indicate that if the primary cooling system operates with a single pump or coolant mass flow rate of 60 kg/s, it will obtain the reactor inlet and outlet temperature respectively 32.2 °C and 40.2 °C. But if it operates with two pumps with a capacity of 75% or coolant mass flow rate of 90 kg/s, the obtained reactor inlet, and outlet temperature respectively 32.9 °C and 38.2 °C. Both models are qualified as a primary coolant for the primary coolant temperature is still below the permitted limit is 49.0 °C.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Harmony, S.C.; Steiner, J.L.; Stumpf, H.J.

The PIUS advanced reactor is a 640-MWe pressurized water reactor developed by Asea Brown Boveri (ABB). A unique feature of the PIUS concept is the absence of mechanical control and shutdown rods. Reactivity is controlled by coolant boron concentration and the temperature of the moderator coolant. As part of the preapplication and eventual design certification process, advanced reactor applicants are required to submit neutronic and thermal-hydraulic safety analyses over a sufficient range of normal operation, transient conditions, and specified accident sequences. Los Alamos is supporting the US Nuclear Regulatory Commission`s preapplication review of the PIUS reactor. A fully one-dimensional modelmore » of the PIUS reactor has been developed for the Transient Reactor Analysis Code, TRACPF1/MOD2. Early in 1992, ABB submitted a Supplemental Information Package describing recent design modifications. An important feature of the PIUS Supplement design was the addition of an active scram system that will function for most transient and accident conditions. A one-dimensional Transient Reactor Analysis Code baseline calculation of the PIUS Supplement design were performed for a break in the main steam line at the outlet nozzle of the loop 3 steam generator. Sensitivity studies were performed to explore the robustness of the PIUS concept to severe off-normal conditions following a main steam line break. The sensitivity study results provide insights into the robustness of the design.« less

Reconstructing the direction of reactor antineutrinos via electron scattering in Gd-doped water Cherenkov detectors

DOE PAGES

Hellfeld, D.; Bernstein, A.; Dazeley, S.; ...

2017-01-01

The potential of elastic antineutrino-electron scattering (ν¯ e + e – → ν¯ e + e –) in a Gd-doped water Cherenkov detector to determine the direction of a nuclear reactor antineutrino flux was investigated using the recently proposed WATCHMAN antineutrino experiment as a baseline model. The expected scattering rate was determined assuming a 13 km standoff from a 3.758 GWt light water nuclear reactor. Background was estimated via independent simulations and by appropriately scaling published measurements from similar detectors. Many potential backgrounds were considered, including solar neutrinos, misidentified reactor-based inverse beta decay interactions, cosmogenic radionuclide and water-borne radon decays,more » and gamma rays from the photomultiplier tubes, detector walls, and surrounding rock. The detector response was modeled using a GEANT4-based simulation package. The results indicate that with the use of low radioactivity PMTs and sufficient fiducialization, water-borne radon and cosmogenic radionuclides pose the largest threats to sensitivity. The directional sensitivity was then analyzed as a function of radon contamination, detector depth, and detector size. Lastly, the results provide a list of theoretical conditions that, if satisfied in practice, would enable nuclear reactor antineutrino directionality in a Gd-doped water Cherenkov detector approximately 10 km from a large power reactor.« less

Risk Management for Sodium Fast Reactors.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Denman, Matthew R.; Groth, Katrina; Cardoni, Jeffrey N.

2015-01-01

Accident management is an important component to maintaining risk at acceptable levels for all complex systems, such as nuclear power plants. With the introduction of self - correcting, or inherently safe, reactor designs the focus has shifted from management by operators to allowing the syste m's design to manage the accident. While inherently and passively safe designs are laudable, extreme boundary conditions can interfere with the design attributes which facilitate inherent safety , thus resulting in unanticipated and undesirable end states. This report examines an inherently safe and small sodium fast reactor experiencing a beyond design basis seismic event withmore » the intend of exploring two issues : (1) can human intervention either improve or worsen the potential end states and (2) can a Bayes ian Network be constructed to infer the state of the reactor to inform (1). ACKNOWLEDGEMENTS The author s would like to acknowledge the U.S. Department of E nergy's Office of Nuclear Energy for funding this research through Work Package SR - 14SN100303 under the Advanced Reactor Concepts program. The authors also acknowledge the PRA teams at A rgonne N ational L aborator y , O ak R idge N ational L aborator y , and I daho N ational L aborator y for their continue d contributions to the advanced reactor PRA mission area.« less

Actinide removal from spent salts

DOEpatents

Hsu, Peter C.; von Holtz, Erica H.; Hipple, David L.; Summers, Leslie J.; Adamson, Martyn G.

2002-01-01

A method for removing actinide contaminants (uranium and thorium) from the spent salt of a molten salt oxidation (MSO) reactor is described. Spent salt is removed from the reactor and analyzed to determine the contaminants present and the carbonate concentration. The salt is dissolved in water, and one or more reagents are added to precipitate the thorium as thorium oxide and/or the uranium as either uranium oxide or as a diuranate salt. The precipitated materials are filtered, dried and packaged for disposal as radioactive waste. About 90% of the thorium and/or uranium present is removed by filtration. After filtration, salt solutions having a carbonate concentration >20% can be dried and returned to the reactor for re-use. Salt solutions containing a carbonate concentration <20% require further clean-up using an ion exchange column, which yields salt solutions that contain less than 0.1 ppm of thorium or uranium.

Metals removal from spent salts

DOEpatents

Hsu, Peter C.; Von Holtz, Erica H.; Hipple, David L.; Summers, Leslie J.; Brummond, William A.; Adamson, Martyn G.

2002-01-01

A method and apparatus for removing metal contaminants from the spent salt of a molten salt oxidation (MSO) reactor is described. Spent salt is removed from the reactor and analyzed to determine the contaminants present and the carbonate concentration. The salt is dissolved in water, and one or more reagents may be added to precipitate the metal oxide and/or the metal as either metal oxide, metal hydroxide, or as a salt. The precipitated materials are filtered, dried and packaged for disposal as waste or can be immobilized as ceramic pellets. More than about 90% of the metals and mineral residues (ashes) present are removed by filtration. After filtration, salt solutions having a carbonate concentration >20% can be spray-dried and returned to the reactor for re-use. Salt solutions containing a carbonate concentration <20% require further clean-up using an ion exchange column, which yields salt solutions that contain less than 1.0 ppm of contaminants.

Effective delayed neutron fraction and prompt neutron lifetime of Tehran research reactor mixed-core.

PubMed

Lashkari, A; Khalafi, H; Kazeminejad, H

2013-05-01

In this work, kinetic parameters of Tehran research reactor (TRR) mixed cores have been calculated. The mixed core configurations are made by replacement of the low enriched uranium control fuel elements with highly enriched uranium control fuel elements in the reference core. The MTR_PC package, a nuclear reactor analysis tool, is used to perform the analysis. Simulations were carried out to compute effective delayed neutron fraction and prompt neutron lifetime. Calculation of kinetic parameters is necessary for reactivity and power excursion transient analysis. The results of this research show that effective delayed neutron fraction decreases and prompt neutron lifetime increases with the fuels burn-up. Also, by increasing the number of highly enriched uranium control fuel elements in the reference core, the prompt neutron lifetime increases, but effective delayed neutron fraction does not show any considerable change.

Effective delayed neutron fraction and prompt neutron lifetime of Tehran research reactor mixed-core

PubMed Central

Lashkari, A.; Khalafi, H.; Kazeminejad, H.

2013-01-01

In this work, kinetic parameters of Tehran research reactor (TRR) mixed cores have been calculated. The mixed core configurations are made by replacement of the low enriched uranium control fuel elements with highly enriched uranium control fuel elements in the reference core. The MTR_PC package, a nuclear reactor analysis tool, is used to perform the analysis. Simulations were carried out to compute effective delayed neutron fraction and prompt neutron lifetime. Calculation of kinetic parameters is necessary for reactivity and power excursion transient analysis. The results of this research show that effective delayed neutron fraction decreases and prompt neutron lifetime increases with the fuels burn-up. Also, by increasing the number of highly enriched uranium control fuel elements in the reference core, the prompt neutron lifetime increases, but effective delayed neutron fraction does not show any considerable change. PMID:24976672

A Feasibility Study on Reactor Based Fission Neutron Radiography of 200-l Waste Packages

NASA Astrophysics Data System (ADS)

Bücherl, T.; Kalthoff, O.; von Gostomski, Ch. Lierse

This feasibility study investigates the applicability of fission neutrons for the non-destructive characterization of radioactive waste packages by means of neutron radiography. Based on a number of mock-up drums of different non-radioactive matrices, but being typical for radioactive waste generated in Europe, radiography measurements at the NECTAR and the ITS facility using fission neutrons and 60Co-gamma-rays, respectively, are performed. The resulting radiographs are compared and qualitatively assessed. In addition, a first approach for the stitching of the fission neutron radiographs to visualize the complete area of 200-l waste drums is performed. While the feasibility of fission neutrons is demonstrated successfully, fields for further improvements are identified.

Engineering Margin Factors Used in the Design of the VVER Fuel Cycles

NASA Astrophysics Data System (ADS)

Lizorkin, M. P.; Shishkov, L. K.

2017-12-01

The article describes methods for determination of the engineering margin factors currently used to estimate the uncertainties of the VVER reactor design parameters calculated via the KASKAD software package developed at the National Research Center Kurchatov Institute. These margin factors ensure the meeting of the operating (design) limits and a number of other restrictions under normal operating conditions.

Iridium-192 Production for Cancer Treatment

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rostelato, M.E.C.M.; Silva, C.P.G.; Rela, P.R.

2004-10-05

The purpose of this work is to settle a laboratory for Iridium -192 sources production, that is, to determine a wire activation method and to build a hot cell for the wires manipulation, quality control and packaging. The paper relates, mainly, the wire activation method and its quality control. The wire activation is carried out in our nuclear reactor, IEA- R1m.

Nuclear Computerized Library for Assessing Reactor Reliability (NUCLARR). Version 3.5, Quick Reference Guide

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gilbert, B.G.; Richards, R.E.; Reece, W.J.

1992-10-01

This Reference Guide contains instructions on how to install and use Version 3.5 of the NRC-sponsored Nuclear Computerized Library for Assessing Reactor Reliability (NUCLARR). The NUCLARR data management system is contained in compressed files on the floppy diskettes that accompany this Reference Guide. NUCLARR is comprised of hardware component failure data (HCFD) and human error probability (HEP) data, both of which are available via a user-friendly, menu driven retrieval system. The data may be saved to a file in a format compatible with IRRAS 3.0 and commercially available statistical packages, or used to formulate log-plots and reports of data retrievalmore » and aggregation findings.« less

Nuclear Computerized Library for Assessing Reactor Reliability (NUCLARR)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gilbert, B.G.; Richards, R.E.; Reece, W.J.

1992-10-01

This Reference Guide contains instructions on how to install and use Version 3.5 of the NRC-sponsored Nuclear Computerized Library for Assessing Reactor Reliability (NUCLARR). The NUCLARR data management system is contained in compressed files on the floppy diskettes that accompany this Reference Guide. NUCLARR is comprised of hardware component failure data (HCFD) and human error probability (HEP) data, both of which are available via a user-friendly, menu driven retrieval system. The data may be saved to a file in a format compatible with IRRAS 3.0 and commercially available statistical packages, or used to formulate log-plots and reports of data retrievalmore » and aggregation findings.« less

Report on Understanding and Predicting Effects of Thermal Aging on Microstructure and Tensile Properties of Grade 91 Steel for Structural Components

DOE Office of Scientific and Technical Information (OSTI.GOV)

Li, Meimei; Natesan, K.; Chen, Weiying

This report provides an update on understanding and predicting the effects of long-term thermal aging on microstructure and tensile properties of G91 to corroborate the ASME Code rules in strength reduction due to elevated temperature service. The research is to support the design and long-term operation of G91 structural components in sodium-cooled fast reactors (SFRs). The report is a Level 2 deliverable in FY17 (M2AT-17AN1602017), under the Work Package AT-17AN160201, “SFR Materials Testing” performed by the Argonne National Laboratory (ANL), as part of the Advanced Reactor Technologies Program.

Dynamic Modeling and Control of Nuclear Reactors Coupled to Closed-Loop Brayton Cycle Systems using SIMULINK{sup TM}

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wright, Steven A.; Sanchez, Travis

2005-02-06

The operation of space reactors for both in-space and planetary operations will require unprecedented levels of autonomy and control. Development of these autonomous control systems will require dynamic system models, effective control methodologies, and autonomous control logic. This paper briefly describes the results of reactor, power-conversion, and control models that are implemented in SIMULINK{sup TM} (Simulink, 2004). SIMULINK{sup TM} is a development environment packaged with MatLab{sup TM} (MatLab, 2004) that allows the creation of dynamic state flow models. Simulation modules for liquid metal, gas cooled reactors, and electrically heated systems have been developed, as have modules for dynamic power-conversion componentsmore » such as, ducting, heat exchangers, turbines, compressors, permanent magnet alternators, and load resistors. Various control modules for the reactor and the power-conversion shaft speed have also been developed and simulated. The modules are compiled into libraries and can be easily connected in different ways to explore the operational space of a number of potential reactor, power-conversion system configurations, and control approaches. The modularity and variability of these SIMULINK{sup TM} models provides a way to simulate a variety of complete power generation systems. To date, both Liquid Metal Reactors (LMR), Gas Cooled Reactors (GCR), and electric heaters that are coupled to gas-dynamics systems and thermoelectric systems have been simulated and are used to understand the behavior of these systems. Current efforts are focused on improving the fidelity of the existing SIMULINK{sup TM} modules, extending them to include isotopic heaters, heat pipes, Stirling engines, and on developing state flow logic to provide intelligent autonomy. The simulation code is called RPC-SIM (Reactor Power and Control-Simulator)« less

The NJOY Nuclear Data Processing System, Version 2016

DOE Office of Scientific and Technical Information (OSTI.GOV)

Macfarlane, Robert; Muir, Douglas W.; Boicourt, R. M.

The NJOY Nuclear Data Processing System, version 2016, is a comprehensive computer code package for producing pointwise and multigroup cross sections and related quantities from evaluated nuclear data in the ENDF-4 through ENDF-6 legacy card-image formats. NJOY works with evaluated files for incident neutrons, photons, and charged particles, producing libraries for a wide variety of particle transport and reactor analysis codes.

Thermoelectric-Driven Sustainable Sensing and Actuation Systems for Fault-Tolerant Nuclear Incidents

DOE Office of Scientific and Technical Information (OSTI.GOV)

Longtin, Jon

2016-02-08

The Fukushima Daiichi nuclear incident in March 2011 represented an unprecedented stress test on the safety and backup systems of a nuclear power plant. The lack of reliable information from key components due to station blackout was a serious setback, leaving sensing, actuation, and reporting systems unable to communicate, and safety was compromised. Although there were several independent backup power sources for required safety function on site, ultimately the batteries were drained and the systems stopped working. If, however, key system components were instrumented with self-powered sensing and actuation packages that could report indefinitely on the status of the system,more » then critical system information could be obtained while providing core actuation and control during off-normal status for as long as needed. This research project focused on the development of such a self-powered sensing and actuation system. The electrical power is derived from intrinsic heat in the reactor components, which is both reliable and plentiful. The key concept was based around using thermoelectric generators that can be integrated directly onto key nuclear components, including pipes, pump housings, heat exchangers, reactor vessels, and shielding structures, as well as secondary-side components. Thermoelectric generators are solid-state devices capable of converting heat directly into electricity. They are commercially available technology. They are compact, have no moving parts, are silent, and have excellent reliability. The key components to the sensor package include a thermoelectric generator (TEG), microcontroller, signal processing, and a wireless radio package, environmental hardening to survive radiation, flooding, vibration, mechanical shock (explosions), corrosion, and excessive temperature. The energy harvested from the intrinsic heat of reactor components can be then made available to power sensors, provide bi-directional communication, recharge batteries for other safety systems, etc. Such an approach is intrinsically fault tolerant: in the event that system temperatures increase, the amount of available energy will increase, which will make more power available for applications. The system can also be used during normal conditions to provide enhanced monitoring of key system components.« less

Validation of the new code package APOLLO2.8 for accurate PWR neutronics calculations

DOE Office of Scientific and Technical Information (OSTI.GOV)

Santamarina, A.; Bernard, D.; Blaise, P.

2013-07-01

This paper summarizes the Qualification work performed to demonstrate the accuracy of the new APOLLO2.S/SHEM-MOC package based on JEFF3.1.1 nuclear data file for the prediction of PWR neutronics parameters. This experimental validation is based on PWR mock-up critical experiments performed in the EOLE/MINERVE zero-power reactors and on P.I. Es on spent fuel assemblies from the French PWRs. The Calculation-Experiment comparison for the main design parameters is presented: reactivity of UOX and MOX lattices, depletion calculation and fuel inventory, reactivity loss with burnup, pin-by-pin power maps, Doppler coefficient, Moderator Temperature Coefficient, Void coefficient, UO{sub 2}-Gd{sub 2}O{sub 3} poisoning worth, Efficiency ofmore » Ag-In-Cd and B4C control rods, Reflector Saving for both standard 2-cm baffle and GEN3 advanced thick SS reflector. From this qualification process, calculation biases and associated uncertainties are derived. This code package APOLLO2.8 is already implemented in the ARCADIA new AREVA calculation chain for core physics and is currently under implementation in the future neutronics package of the French utility Electricite de France. (authors)« less

Magnet design with 100-kA HTS STARS conductors for the helical fusion reactor

NASA Astrophysics Data System (ADS)

Yanagi, N.; Terazaki, Y.; Ito, S.; Tamura, H.; Hamaguchi, S.; Mito, T.; Hashizume, H.; Sagara, A.

2016-12-01

The high-temperature superconducting (HTS) option is employed for the conceptual design of the LHD-type helical fusion reactor FFHR-d1. The 100-kA-class STARS (Stacked Tapes Assembled in Rigid Structure) conductor is used for the magnet system including the continuously wound helical coils. Protection of the magnet system in case of a quench is a crucial issue and the hot-spot temperature during an emergency discharge is estimated based on the zero-dimensional and one-dimensional analyses. The number of division of the coil winding package is examined to limit the voltage generation. For cooling the HTS magnet, helium gas flow is considered and its feasibility is examined by simple analysis as a first step.

TRACE Model for Simulation of Anticipated Transients Without Scram in a BWR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cheng L. Y.; Baek J.; Cuadra,A.

2013-11-10

A TRACE model has been developed for using theTRACE/PARCS computational package [1, 2] to simulate anticipated transients without scram (ATWS) events in a boiling water reactor (BWR). The model represents a BWR/5 housed in a Mark II containment. The reactor and the balance of plant systems are modeled in sufficient detail to enable the evaluation of plant responses and theeffectiveness of automatic and operator actions tomitigate this beyond design basis accident.The TRACE model implements features thatfacilitate the simulation of ATWS events initiated by turbine trip and closure of the main steam isolation valves (MSIV). It also incorporates control logic tomore » initiate actions to mitigate the ATWS events, such as water levelcontrol, emergency depressurization, and injection of boron via the standby liquid control system (SLCS). Two different approaches have been used to model boron mixing in the lower plenum of the reactor vessel: modulate coolant flow in the lower plenum by a flow valve, and use control logic to modular.« less

Performance of a natural gas fuel processor for residential PEFC system using a novel CO preferential oxidation catalyst

NASA Astrophysics Data System (ADS)

Echigo, Mitsuaki; Shinke, Norihisa; Takami, Susumu; Tabata, Takeshi

Natural gas fuel processors have been developed for 500 W and 1 kW class residential polymer electrolyte fuel cell (PEFC) systems. These fuel processors contain all the elements—desulfurizers, steam reformers, CO shift converters, CO preferential oxidation (PROX) reactors, steam generators, burners and heat exchangers—in one package. For the PROX reactor, a single-stage PROX process using a novel PROX catalyst was adopted. In the 1 kW class fuel processor, thermal efficiency of 83% at HHV was achieved at nominal output assuming a H 2 utilization rate in the cell stack of 76%. CO concentration below 1 ppm in the product gas was achieved even under the condition of [O 2]/[CO]=1.5 at the PROX reactor. The long-term durability of the fuel processor was demonstrated with almost no deterioration in thermal efficiency and CO concentration for 10,000 h, 1000 times start and stop cycles, 25,000 cycles of load change.

JPRS Report, Nuclear Developments

DTIC Science & Technology

1990-12-06

ban on that flail away at each other in maneuvers designed to nuclear tests (sic), even for peaceful purposes, across build up influence . Whatever...uranium and develop reactors for nuclear can help Mr. Collor nudge (away from the nuclear submarines. program) his fractious military along by suspending...two and a half years to behaviour to qualify for a certificate. WASHINGTON permit the second six-year aid package for 1988-93, was POST has meanwhile

Correlations between homologue concentrations of PCDD/Fs and toxic equivalency values in laboratory-, package boiler-, and field-scale incinerators.

PubMed

Iino, Fukuya; Takasuga, Takumi; Touati, Abderrahmane; Gullett, Brian K

2003-01-01

The toxic equivalency (TEQ) values of polychlorinated dibenzo-p-dioxins and polychlorinated dibenzofurans (PCDD/Fs) are predicted with a model based on the homologue concentrations measured from a laboratory-scale reactor (124 data points), a package boiler (61 data points), and operating municipal waste incinerators (114 data points). Regardless of the three scales and types of equipment, the different temperature profiles, sampling emissions and/or solids (fly ash), and the various chemical and physical properties of the fuels, all the PCDF plots showed highly linear correlations (R(2)>0.99). The fitting lines of the reactor and the boiler data were almost linear with slope of unity, whereas the slope of the municipal waste incinerator data was 0.86, which is caused by higher predicted values for samples with high measured TEQ. The strong correlation also implies that each of the 10 toxic PCDF congeners has a constant concentration relative to its respective total homologue concentration despite a wide range of facility types and combustion conditions. The PCDD plots showed significant scatter and poor linearity, which implies that the relative concentration of PCDD TEQ congeners is more sensitive to variations in reaction conditions than that of the PCDF congeners.

Properties of bio-oil generated by a pyrolysis of forest cedar residuals with the movable Auger-type reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Nishimura, Shun; Ebitani, Kohki, E-mail: ebitani@jaist.ac.jp; Miyazato, Akio

Our research project has developed the new movable reactor for bio-oil production in 2013 on the basis of Auger-type system. This package would be a great impact due to the concept of local production for local consumption in the hilly and mountainous area in not only Japan but also in the world. Herein, we would like to report the properties of the bio-oil generated by the developing Auger-type movable reactor. The synthesized bio-oil possessed C: 46.2 wt%, H: 6.5 wt%, N: wt%, S: <0.1 wt%, O: 46.8 wt% and H{sub 2}O: 18.4 wt%, and served a good calorific value ofmore » 18.1 MJ/kg. The spectroscopic and mass analyses such as FT-IR, GC-MS, {sup 13}C-NMR and FT-ICR MS supported that the bio-oil was composed by the fine mixtures of methoxy phenols and variety of alcohol or carboxylic acid functional groups. Thus, it is suggested that the bio-oil generated by the new movable Auger-type reactor has a significant potential as well as the existing bio-oil reported previously.« less

Disposition of Chicago Pile 5 (CP-5) Converter Tubes in the 10-160B Cask

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pancake, Daniel C.; Rock, Cynthia

This paper will focus on the unique characterization, packaging, and transportation issues associated with the disposition of the two CP-5 Converter Tube assemblies from Argonne National Laboratory. The converter tubes were constructed of combinations of HEU and alloys of zirconium, and were part of the original research facilities attached to the CP-5 reactor during operating evolutions. These assemblies were heavily irradiated during their operational lifetime, and were segregated from the balance of irradiated test specimens when the reactor was deactivated and slated for Decontamination and Demolition (D&D). In addition, the substantial contribution of fissile material to the assemblies’ inventory mademore » the potential disposition pathways extremely challenging. As a result, these items became part of Argonne’s legacy “nuclear footprint”, and were added to the Nuclear Footprint Reduction Project scope for disposition. The Project was responsible for the size reduction and characterization of these items, as well as the ultimate disposition. After negotiating a disposal pathway for these tubes, there were significant transportation issues that required a small team to overcome, in order to successfully ship these items to the Nevada National Security Site (NNSS). The Project team at Argonne, technical support from transportation specialists, licensing support from the 10-160B license owner, the Savanah River National Lab (SRNL) Packaging Certification Team (PCT, and the DOE EM-33 staff contributed to license and safety analysis report amendments that eventually authorized the shipment of the material. The paper will identify the organizations, and the specific actions, required to successfully make three “one of a kind” shipments of irradiated test specimen material. This will include the unique packaging configurations, contents modification for the cask license (via the Amendment process), criticality evaluations, and associated review and approval processes.« less

NRC ARDC Guidance Support Status Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Holbrook, Mark R.

This report provides a summary that reflects the progress and status of proposed regulatory design criteria for advanced non-light water reactor (LWR) designs in accordance with the Level 3 milestone M3AT-17IN2001013 in work package AT-17IN200101. These criteria have been designated as advanced reactor design criteria (ARDC) and they provide guidance to future applicants for addressing the general design criteria (GDC) that are currently applied specifically to LWR designs. This report provides a summary of Phase 2 activities related to the various tasks associated with ARDC development and the subsequent development of ARDC regulatory guidance for sodium fast reactor (SFR) andmore » modular high-temperature gas-cooled reactor (HTGR) designs. Status Report Organization: Section 2 discusses the origin of the GDC and their application to LWRs. Section 3 addresses the objective of this initiative and how it benefits the advanced non-LWR reactor vendors. Section 4 discusses the scope and structure of the initiative. Section 5 provides background on the U.S. Department of Energy (DOE) ARDC team’s original development of the proposed ARDC that were submitted to the NRC for consideration. Section 6 provides a summary of recent ARDC Phase 2 activities. Appendices A through E document the DOE ARDC team’s public comments on various sections of the NRC’s draft regulatory guide DG–1330, “Guidance for Developing Principal Design Criteria for Non-Light Water Reactors.”« less

Code Development in Coupled PARCS/RELAP5 for Supercritical Water Reactor

DOE PAGES

Hu, Po; Wilson, Paul

2014-01-01

The new capability is added to the existing coupled code package PARCS/RELAP5, in order to analyze SCWR design under supercritical pressure with the separated water coolant and moderator channels. This expansion is carried out on both codes. In PARCS, modification is focused on extending the water property tables to supercritical pressure, modifying the variable mapping input file and related code module for processing thermal-hydraulic information from separated coolant/moderator channels, and modifying neutronics feedback module to deal with the separated coolant/moderator channels. In RELAP5, modification is focused on incorporating more accurate water properties near SCWR operation/transient pressure and temperature in themore » code. Confirming tests of the modifications is presented and the major analyzing results from the extended codes package are summarized.« less

Dynamic analysis environment for nuclear forensic analyses

NASA Astrophysics Data System (ADS)

Stork, C. L.; Ummel, C. C.; Stuart, D. S.; Bodily, S.; Goldblum, B. L.

2017-01-01

A Dynamic Analysis Environment (DAE) software package is introduced to facilitate group inclusion/exclusion method testing, evaluation and comparison for pre-detonation nuclear forensics applications. Employing DAE, the multivariate signatures of a questioned material can be compared to the signatures for different, known groups, enabling the linking of the questioned material to its potential process, location, or fabrication facility. Advantages of using DAE for group inclusion/exclusion include built-in query tools for retrieving data of interest from a database, the recording and documentation of all analysis steps, a clear visualization of the analysis steps intelligible to a non-expert, and the ability to integrate analysis tools developed in different programming languages. Two group inclusion/exclusion methods are implemented in DAE: principal component analysis, a parametric feature extraction method, and k nearest neighbors, a nonparametric pattern recognition method. Spent Fuel Isotopic Composition (SFCOMPO), an open source international database of isotopic compositions for spent nuclear fuels (SNF) from 14 reactors, is used to construct PCA and KNN models for known reactor groups, and 20 simulated SNF samples are utilized in evaluating the performance of these group inclusion/exclusion models. For all 20 simulated samples, PCA in conjunction with the Q statistic correctly excludes a large percentage of reactor groups and correctly includes the true reactor of origination. Employing KNN, 14 of the 20 simulated samples are classified to their true reactor of origination.

System-Level Heat Transfer Analysis, Thermal- Mechanical Cyclic Stress Analysis, and Environmental Fatigue Modeling of a Two-Loop Pressurized Water Reactor. A Preliminary Study

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mohanty, Subhasish; Soppet, William; Majumdar, Saurin

This report provides an update on an assessment of environmentally assisted fatigue for light water reactor components under extended service conditions. This report is a deliverable in April 2015 under the work package for environmentally assisted fatigue under DOE's Light Water Reactor Sustainability program. In this report, updates are discussed related to a system level preliminary finite element model of a two-loop pressurized water reactor (PWR). Based on this model, system-level heat transfer analysis and subsequent thermal-mechanical stress analysis were performed for typical design-basis thermal-mechanical fatigue cycles. The in-air fatigue lives of components, such as the hot and cold legs,more » were estimated on the basis of stress analysis results, ASME in-air fatigue life estimation criteria, and fatigue design curves. Furthermore, environmental correction factors and associated PWR environment fatigue lives for the hot and cold legs were estimated by using estimated stress and strain histories and the approach described in NUREG-6909. The discussed models and results are very preliminary. Further advancement of the discussed model is required for more accurate life prediction of reactor components. This report only presents the work related to finite element modelling activities. However, in between multiple tensile and fatigue tests were conducted. The related experimental results will be presented in the year-end report.« less

Study the Cyclic Plasticity Behavior of 508 LAS under Constant, Variable and Grid-Load-Following Loading Cycles for Fatigue Evaluation of PWR Components

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mohanty, Subhasish; Barua, Bipul; Soppet, William K.

This report provides an update of an earlier assessment of environmentally assisted fatigue for components in light water reactors. This report is a deliverable in September 2016 under the work package for environmentally assisted fatigue under DOE’s Light Water Reactor Sustainability program. In an April 2016 report, we presented a detailed thermal-mechanical stress analysis model for simulating the stress-strain state of a reactor pressure vessel and its nozzles under grid-load-following conditions. In this report, we provide stress-controlled fatigue test data for 508 LAS base metal alloy under different loading amplitudes (constant, variable, and random grid-load-following) and environmental conditions (in airmore » or pressurized water reactor coolant water at 300°C). Also presented is a cyclic plasticity-based analytical model that can simultaneously capture the amplitude and time dependency of the component behavior under fatigue loading. Results related to both amplitude-dependent and amplitude-independent parameters are presented. The validation results for the analytical/mechanistic model are discussed. This report provides guidance for estimating time-dependent, amplitude-independent parameters related to material behavior under different service conditions. The developed mechanistic models and the reported material parameters can be used to conduct more accurate fatigue and ratcheting evaluation of reactor components.« less

FY 2017-Influence of Sodium Environment on the Tensile Properties of Advanced Alloys

DOE Office of Scientific and Technical Information (OSTI.GOV)

Natesan, K.; Li, Meimei; Chen, Wei-Ying

This report provides an update on the understanding of the effects of sodium exposures on tensile properties of advanced alloy 709 in support of the design and operation of structural components in sodium-cooled fast reactors (SFRs). The report is a Level 3 deliverable in FY17 (M3AT-17AN1602093), under the Work Package AT-17AN160209, “Sodium Compatibility” performed by Argonne National Laboratory (ANL), as part of Advanced Reactor Technologies Program. Three laboratory-size heats of Alloy 709 austenitic steel were investigated in liquid sodium environments at 550-650°C to understand its corrosion behaviour, microstructural evolution, and tensile properties. In addition, a commercial scale heat has beenmore » produced and hot-rolled into plates.« less

SAS2H Generated Isotopic Concentrations For B&W 15X15 PWR Assembly (SCPB:N/A)

DOE Office of Scientific and Technical Information (OSTI.GOV)

J.W. Davis

This analysis is prepared by the Mined Geologic Disposal System (MGDS) Waste Package Development Department (WPDD) to provide pressurized water reactor (PWR) isotopic composition data as a function of time for use in criticality analyses. The objectives of this evaluation are to generate burnup and decay dependant isotopic inventories and to provide these inventories in a form which can easily be utilized in subsequent criticality calculations.

Design Analysis of a Prepackaged Nuclear Power Plant for an Ice Cap Location

DTIC Science & Technology

1959-01-15

requirements and heating load 1.3 Site Conditions 1,U Air Transportability 1.5 Standby Power Availability 1.6 Building Structuree and Foundations 2,0...Skid with Reactor and Steam Generator Generator Weight Distribution Foundation Load Diagram (Secondary) Turbine Generator Package - Typical...Requirements and Heating Load The plant shall be capable of producing a minimum of 1500 Kw net ^ electrical energy at 4160/2400 volts, three phase

Intact and Degraded Criticality Calculations for the Codisposal of Shippingport LWBR Spent Nuclear Fuel in a Waste Package

DOE Office of Scientific and Technical Information (OSTI.GOV)

L.M. Montierth

2000-09-15

The objective of this calculation is to characterize the nuclear criticality safety concerns associated with the codisposal of the U.S. Department of Energy's (DOE) Shippingport Light Water Breeder Reactor (SP LWBR) Spent Nuclear Fuel (SNF) in a 5-Defense High-Level Waste (5-DHLW) Waste Package (WP), which is to be placed in a Monitored Geologic Repository (MGR). The scope of this calculation is limited to the determination of the effective neutron multiplication factor (K{sub eff}) for intact- and degraded-mode internal configurations of the codisposal WP containing Shippingport LWBR seed-type assemblies. The results of this calculation will be used to evaluate criticality issuesmore » and support the analysis that is planed to be performed to demonstrate the viability of the codisposal concept for the MGR. This calculation is associated with the waste package design and was performed in accordance with the DOE SNF Analysis Plan for FY 2000 (See Ref. 22). The document has been prepared in accordance with the Administrative Procedure AP-3.12Q, Calculations (Ref. 23).« less

BWR ASSEMBLY SOURCE TERMS FOR WASTE PACKAGE DESIGN

DOE Office of Scientific and Technical Information (OSTI.GOV)

T.L. Lotz

1997-02-15

This analysis is prepared by the Mined Geologic Disposal System (MGDS) Waste Package Development Department (WPDD) to provide boiling water reactor (BWR) assembly radiation source term data for use during Waste Package (WP) design. The BWR assembly radiation source terms are to be used for evaluation of radiolysis effects at the WP surface, and for personnel shielding requirements during assembly or WP handling operations. The objectives of this evaluation are to generate BWR assembly radiation source terms that bound selected groupings of BWR assemblies, with regard to assembly average burnup and cooling time, which comprise the anticipated MGDS BWR commercialmore » spent nuclear fuel (SNF) waste stream. The source term data is to be provided in a form which can easily be utilized in subsequent shielding/radiation dose calculations. Since these calculations may also be used for Total System Performance Assessment (TSPA), with appropriate justification provided by TSPA, or radionuclide release rate analysis, the grams of each element and additional cooling times out to 25 years will also be calculated and the data included in the output files.« less

Spent Nuclear Fuel Disposition

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wagner, John C.

One interdisciplinary field devoted to achieving the end-state of used nuclear fuel (UNF) through reuse and/or permanent disposal. The reuse option aims to make use of the remaining energy content in UNF and reduce the amount of long-lived radioactive materials that require permanent disposal. The planned approach in the U.S., as well as in many other countries worldwide, is direct permanent disposal in a deep geologic repository. Used nuclear fuel is fuel that has been irradiated in a nuclear reactor to the point where it is no longer capable of sustaining operational objectives. The vast majority (by mass) of UNFmore » is from electricity generation in commercial nuclear power reactors. Furthermore, the other main source of UNF in the U.S. is the Department of Energy’s (DOE) and other federal agencies’ operation of reactors in support of federal government missions, such as materials production, nuclear propulsion, research, testing, and training. Upon discharge from a reactor, UNF emits considerable heat from radioactive decay. Some period of active on-site cooling (e.g., 2 or more years) is typically required to facilitate efficient packaging and transportation to a disposition facility. Hence, the field of UNF disposition broadly includes storage, transportation and ultimate disposition. See also: Nuclear Fission (content/nuclear-fission/458400), Nuclear Fuels (/content/nuclear-fuels/458600), Nuclear Fuel Cycle (/content/nuclear-fuel-cycle/458500), Nuclear Fuels Reprocessing (/content/nuclear-fuels-reprocessing/458700), Nuclear Power (/content/nuclear-power/459600), Nuclear Reactor (/content/nuclear-reactor/460100), Radiation (/content/radiation/566300), and Radioactive Waste Management (/content/radioactive-waste-management/568900).« less

Spent Nuclear Fuel Disposition

DOE PAGES

Wagner, John C.

2016-05-22

One interdisciplinary field devoted to achieving the end-state of used nuclear fuel (UNF) through reuse and/or permanent disposal. The reuse option aims to make use of the remaining energy content in UNF and reduce the amount of long-lived radioactive materials that require permanent disposal. The planned approach in the U.S., as well as in many other countries worldwide, is direct permanent disposal in a deep geologic repository. Used nuclear fuel is fuel that has been irradiated in a nuclear reactor to the point where it is no longer capable of sustaining operational objectives. The vast majority (by mass) of UNFmore » is from electricity generation in commercial nuclear power reactors. Furthermore, the other main source of UNF in the U.S. is the Department of Energy’s (DOE) and other federal agencies’ operation of reactors in support of federal government missions, such as materials production, nuclear propulsion, research, testing, and training. Upon discharge from a reactor, UNF emits considerable heat from radioactive decay. Some period of active on-site cooling (e.g., 2 or more years) is typically required to facilitate efficient packaging and transportation to a disposition facility. Hence, the field of UNF disposition broadly includes storage, transportation and ultimate disposition. See also: Nuclear Fission (content/nuclear-fission/458400), Nuclear Fuels (/content/nuclear-fuels/458600), Nuclear Fuel Cycle (/content/nuclear-fuel-cycle/458500), Nuclear Fuels Reprocessing (/content/nuclear-fuels-reprocessing/458700), Nuclear Power (/content/nuclear-power/459600), Nuclear Reactor (/content/nuclear-reactor/460100), Radiation (/content/radiation/566300), and Radioactive Waste Management (/content/radioactive-waste-management/568900).« less

New developments in the McStas neutron instrument simulation package

NASA Astrophysics Data System (ADS)

Willendrup, P. K.; Knudsen, E. B.; Klinkby, E.; Nielsen, T.; Farhi, E.; Filges, U.; Lefmann, K.

2014-07-01

The McStas neutron ray-tracing software package is a versatile tool for building accurate simulators of neutron scattering instruments at reactors, short- and long-pulsed spallation sources such as the European Spallation Source. McStas is extensively used for design and optimization of instruments, virtual experiments, data analysis and user training. McStas was founded as a scientific, open-source collaborative code in 1997. This contribution presents the project at its current state and gives an overview of the main new developments in McStas 2.0 (December 2012) and McStas 2.1 (expected fall 2013), including many new components, component parameter uniformisation, partial loss of backward compatibility, updated source brilliance descriptions, developments toward new tools and user interfaces, web interfaces and a new method for estimating beam losses and background from neutron optics.

Dual Arm Work Package performance estimates and telerobot task network simulation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Draper, J.V.; Blair, L.M.

1997-02-01

This paper describes the methodology and results of a network simulation study of the Dual Arm Work Package (DAWP), to be employed for dismantling the Argonne National Laboratory CP-5 reactor. The development of the simulation model was based upon the results of a task analysis for the same system. This study was performed by the Oak Ridge National Laboratory (ORNL), in the Robotics and Process Systems Division. Funding was provided the US Department of Energy`s Office of Technology Development, Robotics Technology Development Program (RTDP). The RTDP is developing methods of computer simulation to estimate telerobotic system performance. Data were collectedmore » to provide point estimates to be used in a task network simulation model. Three skilled operators performed six repetitions of a pipe cutting task representative of typical teleoperation cutting operations.« less

Assessment of the prevailing physics codes: LEOPARD, LASER, and EPRI-CELL

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lan, J.S.

1981-01-01

In order to analyze core performance and fuel management, it is necessary to verify reactor physics codes in great detail. This kind of work not only serves the purpose of understanding and controlling the characteristics of each code, but also ensures the reliability as codes continually change due to constant modifications and machine transfers. This paper will present the results of a comprehensive verification of three code packages - LEOPARD, LASER, and EPRI-CELL.

Monte Carol-based validation of neutronic methodology for EBR-II analyses

DOE Office of Scientific and Technical Information (OSTI.GOV)

Liaw, J.R.; Finck, P.J.

1993-01-01

The continuous-energy Monte Carlo code VIM (Ref. 1) has been validated extensively over the years against fast critical experiments and other neutronic analysis codes. A high degree of confidence in VIM for predicting reactor physics parameters has been firmly established. This paper presents a numerical validation of two conventional multigroup neutronic analysis codes, DIF3D (Ref. 4) and VARIANT (Ref. 5), against VIM for two Experimental Breeder Reactor II (EBR-II) core loadings in detailed three-dimensional hexagonal-z geometry. The DIF3D code is based on nodal diffusion theory, and it is used in calculations for day-today reactor operations, whereas the VARIANT code ismore » based on nodal transport theory and is used with increasing frequency for specific applications. Both DIF3D and VARIANT rely on multigroup cross sections generated from ENDF/B-V by the ETOE-2/MC[sup 2]-II/SDX (Ref. 6) code package. Hence, this study also validates the multigroup cross-section processing methodology against the continuous-energy approach used in VIM.« less

Characterisation of imperial college reactor centre legacy waste using gamma-ray spectrometry

DOE Office of Scientific and Technical Information (OSTI.GOV)

Shuhaimi, Alif Imran Mohd

Waste characterisation is a principal component in waste management strategy. The characterisation includes identification of chemical, physical and radiochemical parameters of radioactive waste. Failure to determine specific waste properties may result in sentencing waste packages which are not compliant with the regulation of long term storage or disposal. This project involved measurement of intensity and energy of gamma photons which may be emitted by radioactive waste generated during decommissioning of Imperial College Reactor Centre (ICRC). The measurement will use High Purity Germanium (HPGe) as Gamma-ray detector and ISOTOPIC-32 V4.1 as analyser. In order to ensure the measurements provide reliable results,more » two quality control (QC) measurements using difference matrices have been conducted. The results from QC measurements were used to determine the accuracy of the ISOTOPIC software.« less

Dismantling of Loop-Type Channel Equipment of MR Reactor in NRC 'Kurchatov Institute' - 13040

DOE Office of Scientific and Technical Information (OSTI.GOV)

Volkov, Victor; Danilovich, Alexey; Zverkov, Yuri

2013-07-01

In 2009 the project of decommissioning of MR and RTF reactors was developed and approved by the Expert Authority of the Russian Federation (Gosexpertiza). The main objective of the decommissioning works identified in this project: - complete dismantling of reactor equipment and systems; - decontamination of reactor premises and site in accordance with the established sanitary and hygienic standards. At the preparatory stage (2008-2010) of the project the following works were executed: loop-type channels' dismantling in the storage pool; experimental fuel assemblies' removal from spent fuel repositories in the central hall; spent fuel assembly removal from the liquid-metal-cooled loop-type channelmore » of the reactor core and its placement into the SNF repository; and reconstruction of engineering support systems to the extent necessary for reactor decommissioning. The project assumes three main phases of dismantling and decontamination: - dismantling of equipment/pipelines of cooling circuits and loop-type channels, and auxiliary reactor equipment (2011-2012); - dismantling of equipment in underground reactor premises and of both MR and RTF in-vessel devices (2013-2014); - decontamination of reactor premises; rehabilitation of the reactor site; final radiation survey of reactor premises, loop-type channels and site; and issuance of the regulatory authorities' de-registration statement (2015). In 2011 the decommissioning license for the two reactors was received and direct MR decommissioning activities started. MR primary pipelines and loop-type facilities situated in the underground reactor hall were dismantled. Works were also launched to dismantle the loop-type channels' equipment in underground reactor premises; reactor buildings were reconstructed to allow removal of dismantled equipment; and the MR/RTF decommissioning sequence was identified. In autumn 2011 - spring 2012 results of dismantling activities performed are: - equipment from underground rooms (No. 66, 66A, 66B, 72, 64, 63) - as well as from water and gas loop corridors - was dismantled, with the total radwaste weight of 53 tons and the total removed activity of 5,0 x 10{sup 10} Bq; - loop-type channel equipment from underground reactor hall premises was dismantled; - 93 loop-type channels were characterized, chopped and removed, with radwaste of 2.6 x 10{sup 13} Bq ({sup 60}Co) and 1.5 x 10{sup 13} Bq ({sup 137}Cs) total activity removed from the reactor pool, fragmented and packaged. Some of this waste was placed into the high-level waste (HLW) repository of the Center. Dismantling works were executed with application of remotely operated mechanisms, which promoted decrease of radiation impact on the personnel. The average individual dose for the personnel was 1.9 mSv/year in 2011, and the collective dose is estimated as 0.0605 man x Sv/year. (authors)« less

Improvement of COBRA-TF for modeling of PWR cold- and hot-legs during reactor transients

NASA Astrophysics Data System (ADS)

Salko, Robert K.

COBRA-TF is a two-phase, three-field (liquid, vapor, droplets) thermal-hydraulic modeling tool that has been developed by the Pacific Northwest Laboratory under sponsorship of the NRC. The code was developed for Light Water Reactor analysis starting in the 1980s; however, its development has continued to this current time. COBRA-TF still finds wide-spread use throughout the nuclear engineering field, including nuclear-power vendors, academia, and research institutions. It has been proposed that extension of the COBRA-TF code-modeling region from vessel-only components to Pressurized Water Reactor (PWR) coolant-line regions can lead to improved Loss-of-Coolant Accident (LOCA) analysis. Improved modeling is anticipated due to COBRA-TF's capability to independently model the entrained-droplet flow-field behavior, which has been observed to impact delivery to the core region[1]. Because COBRA-TF was originally developed for vertically-dominated, in-vessel, sub-channel flow, extension of the COBRA-TF modeling region to the horizontal-pipe geometries of the coolant-lines required several code modifications, including: • Inclusion of the stratified flow regime into the COBRA-TF flow regime map, along with associated interfacial drag, wall drag and interfacial heat transfer correlations, • Inclusion of a horizontal-stratification force between adjacent mesh cells having unequal levels of stratified flow, and • Generation of a new code-input interface for the modeling of coolant-lines. The sheer number of COBRA-TF modifications that were required to complete this work turned this project into a code-development project as much as it was a study of thermal-hydraulics in reactor coolant-lines. The means for achieving these tasks shifted along the way, ultimately leading the development of a separate, nearly completely independent one-dimensional, two-phase-flow modeling code geared toward reactor coolant-line analysis. This developed code has been named CLAP, for Coolant-Line-Analysis Package. Versions were created that were both coupled to COBRA-TF and standalone, with the most recent version being a standalone code. This code performs a separate, simplified, 1-D solution of the conservation equations while making special considerations for coolant-line geometry and flow phenomena. The end of this project saw a functional code package that demonstrates a stable numerical solution and that has gone through a series of Validation and Verification tests using the Two-Phase Testing Facility (TPTF) experimental data[2]. The results indicate that CLAP is under-performing RELAP5-MOD3 in predicting the experimental void of the TPTF facility in some cases. There is no apparent pattern, however, to point to a consistent type of case that the code fails to predict properly (e.g., low-flow, high-flow, discharging to full vessel, or discharging to empty vessel). Pressure-profile predictions are sometimes unrealistic, which indicates that there may be a problem with test-case boundary conditions or with the coupling of continuity and momentum equations in the solution algorithm. The code does predict the flow regime correctly for all cases with the stratification-force model off. Turning the stratification model on can cause the low-flow case void profiles to over-react to the force and the flow regime to transition out of stratified flow. The code would benefit from an increased amount of Validation & Verification testing. The development of CLAP was significant, as it is a cleanly written, logical representation of the reactor coolant-line geometry. It is stable and capable of modeling basic flow physics in the reactor coolant-line. Code development and debugging required the temporary removal of the energy equation and mass-transfer terms in governing equations. The reintroduction of these terms will allow future coupling to RELAP and re-coupling with COBRA-TF. Adding in more applicable entrainment and de-entrainment models would allow the capture of more advanced physics in the coolant-line that can be expected during Loss-of-Coolant Accident. One of the package's benefits is its ability to be used as a platform for future coolant-line model development and implementation, including capturing of the important de-entrainment behavior in reactor hot-legs (steam-binding effect) and flow convection in the upper-plenum region of the vessel.

A Freeware Path to Neutron Computed Tomography

NASA Astrophysics Data System (ADS)

Schillinger, Burkhard; Craft, Aaron E.

Neutron computed tomography has become a routine method at many neutron sources due to the availability of digital detection systems, powerful computers and advanced software. The commercial packages Octopus by Inside Matters and VGStudio by Volume Graphics have been established as a quasi-standard for high-end computed tomography. However, these packages require a stiff investment and are available to the users only on-site at the imaging facility to do their data processing. There is a demand from users to have image processing software at home to do further data processing; in addition, neutron computed tomography is now being introduced even at smaller and older reactors. Operators need to show a first working tomography setup before they can obtain a budget to build an advanced tomography system. Several packages are available on the web for free; however, these have been developed for X-rays or synchrotron radiation and are not immediately useable for neutron computed tomography. Three reconstruction packages and three 3D-viewers have been identified and used even for Gigabyte datasets. This paper is not a scientific publication in the classic sense, but is intended as a review to provide searchable help to make the described packages usable for the tomography community. It presents the necessary additional preprocessing in ImageJ, some workarounds for bugs in the software, and undocumented or badly documented parameters that need to be adapted for neutron computed tomography. The result is a slightly complicated, but surprisingly high-quality path to neutron computed tomography images in 3D, but not a replacement for the even more powerful commercial software mentioned above.

Development of Fuel Shuffling Module for PHISICS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Allan Mabe; Andrea Alfonsi; Cristian Rabiti

2013-06-01

PHISICS (Parallel and Highly Innovative Simulation for the INL Code System) [4] code toolkit has been in development at the Idaho National Laboratory. This package is intended to provide a modern analysis tool for reactor physics investigation. It is designed with the mindset to maximize accuracy for a given availability of computational resources and to give state of the art tools to the modern nuclear engineer. This is obtained by implementing several different algorithms and meshing approaches among which the user will be able to choose, in order to optimize his computational resources and accuracy needs. The software is completelymore » modular in order to simplify the independent development of modules by different teams and future maintenance. The package is coupled with the thermo-hydraulic code RELAP5-3D [3]. In the following the structure of the different PHISICS modules is briefly recalled, focusing on the new shuffling module (SHUFFLE), object of this paper.« less

Atmospheric-Pressure Cold Plasmas Used to Embed Bioactive Compounds in Matrix Material for Active Packaging of Fruits and Vegetables

NASA Astrophysics Data System (ADS)

Fernandez, Sulmer; Pedrow, Patrick; Powers, Joseph; Pitts, Marvin

2009-10-01

Active thin film packaging is a technology with the potential to provide consumers with new fruit and vegetable products-if the film can be applied without deactivating bioactive compounds.Atmospheric pressure cold plasma (APCP) processing can be used to activate monomer with concomitant deposition of an organic plasma polymerized matrix material and to immobilize a bioactive compound all at or below room temperature.Aims of this work include: 1) immobilize an antimicrobial in the matrix; 2) determine if the antimicrobial retains its functionality and 3) optimize the reactor design.The plasma zone will be obtained by increasing the voltage on an electrode structure until the electric field in the feed material (argon + monomer) yields electron avalanches. Results will be described using Red Delicious apples.Prospective matrix precursors are vanillin and cinnamic acid.A prospective bioactive compound is benzoic acid.

Silicon oxide permeation barrier coating of PET bottles and foils

NASA Astrophysics Data System (ADS)

Steves, Simon; Deilmann, Michael; Awakowicz, Peter

2009-10-01

Modern packaging materials such as polyethylene terephthalate (PET) have displaced established materials in many areas of food and beverage packaging. Plastic packing materials offer are various advantages concerning production and handling. PET bottles for instance are non-breakable and lightweight compared to glass and metal containers. However, PET offers poor barrier properties against gas permeation. Therefore, the shelf live of packaged food is reduced. Permeation of gases can be reduced by depositing transparent plasma polymerized silicon oxide (SiOx) barrier coatings. A microwave (2.45 GHz) driven low pressure plasma reactor is developed based on a modified Plasmaline antenna to treat PET foils or bottles. To increase the barrier properties of the coatings furthermore a RF substrate bias (13.56 MHz) is applied. The composition of the coatings is analyzed by means of Fourier transform infrared (FTIR) spectroscopy regarding carbon and hydrogen content. Influence of gas phase composition and substrate bias on chemical composition of the coatings is discussed. A strong relation between barrier properties and film composition is found: good oxygen barriers are observed as carbon content is reduced and films become quartz-like. Regarding oxygen permeation a barrier improvement factor (BIF) of 70 is achieved.

Thermal Destruction of TETS: Experiments and Modeling ...

EPA Pesticide Factsheets

Symposium Paper In the event of a contamination event involving chemical warfare agents (CWAs) or toxic industrial chemicals (TICs), large quantities of potentially contaminated materials, both indoor and outdoor, may be treated with thermal incineration during the site remediation process. Even if the CWAs or TICs of interest are not particularly thermally stable and might be expected to decompose readily in a high temperature combustion environment, the refractory nature of many materials found inside and outside buildings may present heat transfer challenges in an incineration system depending on how the materials are packaged and fed into the incinerator. This paper reports on a study to examine the thermal decomposition of a banned rodenticide, tetramethylene disulfotetramine (TETS) in a laboratory reactor, analysis of the results using classical reactor design theory, and subsequent scale-up of the results to a computer-simulation of a full-scale commercial hazardous waste incinerator processing ceiling tile contaminated with residual TETS.

SCORE-EVET: a computer code for the multidimensional transient thermal-hydraulic analysis of nuclear fuel rod arrays. [BWR; PWR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Benedetti, R. L.; Lords, L. V.; Kiser, D. M.

1978-02-01

The SCORE-EVET code was developed to study multidimensional transient fluid flow in nuclear reactor fuel rod arrays. The conservation equations used were derived by volume averaging the transient compressible three-dimensional local continuum equations in Cartesian coordinates. No assumptions associated with subchannel flow have been incorporated into the derivation of the conservation equations. In addition to the three-dimensional fluid flow equations, the SCORE-EVET code ocntains: (a) a one-dimensional steady state solution scheme to initialize the flow field, (b) steady state and transient fuel rod conduction models, and (c) comprehensive correlation packages to describe fluid-to-fuel rod interfacial energy and momentum exchange. Velocitymore » and pressure boundary conditions can be specified as a function of time and space to model reactor transient conditions such as a hypothesized loss-of-coolant accident (LOCA) or flow blockage.« less

Milestone M3FT-15OR0203112. Build redesigned HFIR rabbit capsules and make ready for insertion for irradiation in HFIR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Howard, Richard H; McDuffee, Joel Lee; Okuniewski, Maria A.

2015-09-01

This report details the fabrication and delivery of two Fuel Cycle Research and Development irradiation capsules (FCRP20 and FCRP03), with associated quality assurance documentation, to the High Flux Isotope Reactor. The capsules and documentation were delivered by September 30, 2015, thus meeting the deadline for milestone M3FT-15OR0203112. These irradiation experiments irradiate metal parallelepiped specimens that may consist of various compositions including uranium metal, steel, etc. This document contains a copy of the completed capsule fabrication request sheets, which detail all constituent components, pertinent drawings, etc., along with a detailed summary of the capsule assembly process performed by the Thermal Hydraulicsmore » and Irradiation Engineering Group (THIEG) in the Reactor and Nuclear Systems Division. A complete fabrication package record is maintained by THIEG and is available upon request.« less

Design consideration for a nuclear electric propulsion system

NASA Technical Reports Server (NTRS)

Phillips, W. M.; Pawlik, E. V.

1978-01-01

A study is currently underway to design a nuclear electric propulsion vehicle capable of performing detailed exploration of the outer-planets. Primary emphasis is on the power subsystem. Secondary emphasis includes integration into a spacecraft, and integration with the thrust subsystem and science package or payload. The results of several design iterations indicate an all-heat-pipe system offers greater reliability, elimination of many technology development areas and a specific weight of under 20 kg/kWe at the 400 kWe power level. The system is compatible with a single Shuttle launch and provides greater safety than could be obtained with designs using pumped liquid metal cooling. Two configurations, one with the reactor and power conversion forward on the spacecraft with the ion engines aft and the other with reactor, power conversion and ion engines aft were selected as dual baseline designs based on minimum weight, minimum required technology development and maximum growth potential and flexibility.

Tensile and Fatigue Testing and Material Hardening Model Development for 508 LAS Base Metal and 316 SS Similar Metal Weld under In-air and PWR Primary Loop Water Conditions

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mohanty, Subhasish; Soppet, William; Majumdar, Saurin

This report provides an update on an assessment of environmentally assisted fatigue for light water reactor components under extended service conditions. This report is a deliverable in September 2015 under the work package for environmentally assisted fatigue under DOE’s Light Water Reactor Sustainability program. In an April 2015 report we presented a baseline mechanistic finite element model of a two-loop pressurized water reactor (PWR) for systemlevel heat transfer analysis and subsequent thermal-mechanical stress analysis and fatigue life estimation under reactor thermal-mechanical cycles. In the present report, we provide tensile and fatigue test data for 508 low-alloy steel (LAS) base metal,more » 508 LAS heat-affected zone metal in 508 LAS–316 stainless steel (SS) dissimilar metal welds, and 316 SS-316 SS similar metal welds. The test was conducted under different conditions such as in air at room temperature, in air at 300 oC, and under PWR primary loop water conditions. Data are provided on materials properties related to time-independent tensile tests and time-dependent cyclic tests, such as elastic modulus, elastic and offset strain yield limit stress, and linear and nonlinear kinematic hardening model parameters. The overall objective of this report is to provide guidance to estimate tensile/fatigue hardening parameters from test data. Also, the material models and parameters reported here can directly be used in commercially available finite element codes for fatigue and ratcheting evaluation of reactor components under in-air and PWR water conditions.« less

System analyses on advanced nuclear fuel cycle and waste management

NASA Astrophysics Data System (ADS)

Cheon, Myeongguk

To evaluate the impacts of accelerator-driven transmutation of waste (ATW) fuel cycle on a geological repository, two mathematical models are developed: a reactor system analysis model and a high-level waste (HLW) conditioning model. With the former, fission products and residual trans-uranium (TRU) contained in HLW generated from a reference ATW plant operations are quantified and the reduction of TRU inventory included in commercial spent-nuclear fuel (CSNF) is evaluated. With the latter, an optimized waste loading and composition in solidification of HLW are determined and the volume reduction of waste packages associated with CSNF is evaluated. WACOM, a reactor system analysis code developed in this study for burnup calculation, is validated by ORIGEN2.1 and MCNP. WACOM is used to perform multicycle analysis for the reference lead-bismuth eutectic (LBE) cooled transmuter. By applying the results of this analysis to the reference ATW deployment scenario considered in the ATW roadmap, the HLW generated from the ATW fuel cycle is quantified and the reduction of TRU inventory contained in CSNF is evaluated. A linear programming (LP) model has been developed for determination of an optimized waste loading and composition in solidification of HLW. The model has been applied to a US-defense HLW. The optimum waste loading evaluated by the LP model was compared with that estimated by the Defense Waste Processing Facility (DWPF) in the US and a good agreement was observed. The LP model was then applied to the volume reduction of waste packages associated with CSNF. Based on the obtained reduction factors, the expansion of Yucca Mountain Repository (YMR) capacity is evaluated. It is found that with the reference ATW system, the TRU contained in CSNF could be reduced by a factor of ˜170 in terms of inventory and by a factor of ˜40 in terms of toxicity under the assumed scenario. The number of waste packages related to CSNF could be reduced by a factor of ˜8 in terms of volume and by factor of ˜10 on the basis of electricity generation when a sufficient cooling time for discharged spent fuel and zero process chemicals in HLW are assumed. The expansion factor of Yucca Mountain Repository capacity is estimated to be a factor of 2.4, much smaller than the reduction factor of CSNF waste packages, due to the existence of DOE-owned spent fuel and HLW. The YMR, however, could support 10 times greater electricity generation as long as the statutory capacity of DOE-owned SNF and HLW remains unchanged. This study also showed that the reduction of the number of waste packages could strongly be subject to the heat generation rate of HLW and the amount of process chemicals contained in HLW. For a greater reduction of the number of waste packages, a sufficient cooling time for discharged fuel and efforts to minimize the amount of process chemicals contained in HLW are crucial.

Conceptual design studies of the Electron Cyclotron launcher for DEMO reactor

NASA Astrophysics Data System (ADS)

Moro, Alessandro; Bruschi, Alex; Franke, Thomas; Garavaglia, Saul; Granucci, Gustavo; Grossetti, Giovanni; Hizanidis, Kyriakos; Tigelis, Ioannis; Tran, Minh-Quang; Tsironis, Christos

2017-10-01

A demonstration fusion power plant (DEMO) producing electricity for the grid at the level of a few hundred megawatts is included in the European Roadmap [1]. The engineering design and R&D for the electron cyclotron (EC), ion cyclotron and neutral beam systems for the DEMO reactor is being performed by Work Package Heating and Current Drive (WPHCD) in the framework of EUROfusion Consortium activities. The EC target power to the plasma is about 50 MW, in which the required power for NTM control and burn control is included. EC launcher conceptual design studies are here presented, showing how the main design drivers of the system have been taken into account (physics requirements, reactor relevant operations, issues related to its integration as in-vessel components). Different options for the antenna are studied in a parameters space including a selection of frequencies, injection angles and launch points to get the best performances for the antenna configuration, using beam tracing calculations to evaluate plasma accessibility and deposited power. This conceptual design studies comes up with the identification of possible limits, constraints and critical issues, essential in the selection process of launcher setup solution.

Containment Sodium Chemistry Models in MELCOR.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Louie, David; Humphries, Larry L.; Denman, Matthew R

To meet regulatory needs for sodium fast reactors’ future development, including licensing requirements, Sandia National Laboratories is modernizing MELCOR, a severe accident analysis computer code developed for the U.S. Nuclear Regulatory Commission (NRC). Specifically, Sandia is modernizing MELCOR to include the capability to model sodium reactors. However, Sandia’s modernization effort primarily focuses on the containment response aspects of the sodium reactor accidents. Sandia began modernizing MELCOR in 2013 to allow a sodium coolant, rather than water, for conventional light water reactors. In the past three years, Sandia has been implementing the sodium chemistry containment models in CONTAIN-LMR, a legacy NRCmore » code, into MELCOR. These chemistry models include spray fire, pool fire and atmosphere chemistry models. Only the first two chemistry models have been implemented though it is intended to implement all these models into MELCOR. A new package called “NAC” has been created to manage the sodium chemistry model more efficiently. In 2017 Sandia began validating the implemented models in MELCOR by simulating available experiments. The CONTAIN-LMR sodium models include sodium atmosphere chemistry and sodium-concrete interaction models. This paper presents sodium property models, the implemented models, implementation issues, and a path towards validation against existing experimental data.« less

Heuristic optimization of a continuous flow point-of-use UV-LED disinfection reactor using computational fluid dynamics.

PubMed

Jenny, Richard M; Jasper, Micah N; Simmons, Otto D; Shatalov, Max; Ducoste, Joel J

2015-10-15

Alternative disinfection sources such as ultraviolet light (UV) are being pursued to inactivate pathogenic microorganisms such as Cryptosporidium and Giardia, while simultaneously reducing the risk of exposure to carcinogenic disinfection by-products (DBPs) in drinking water. UV-LEDs offer a UV disinfecting source that do not contain mercury, have the potential for long lifetimes, are robust, and have a high degree of design flexibility. However, the increased flexibility in design options will add a substantial level of complexity when developing a UV-LED reactor, particularly with regards to reactor shape, size, spatial orientation of light, and germicidal emission wavelength. Anticipating that LEDs are the future of UV disinfection, new methods are needed for designing such reactors. In this research study, the evaluation of a new design paradigm using a point-of-use UV-LED disinfection reactor has been performed. ModeFrontier, a numerical optimization platform, was coupled with COMSOL Multi-physics, a computational fluid dynamics (CFD) software package, to generate an optimized UV-LED continuous flow reactor. Three optimality conditions were considered: 1) single objective analysis minimizing input supply power while achieving at least (2.0) log10 inactivation of Escherichia coli ATCC 11229; and 2) two multi-objective analyses (one of which maximized the log10 inactivation of E. coli ATCC 11229 and minimized the supply power). All tests were completed at a flow rate of 109 mL/min and 92% UVT (measured at 254 nm). The numerical solution for the first objective was validated experimentally using biodosimetry. The optimal design predictions displayed good agreement with the experimental data and contained several non-intuitive features, particularly with the UV-LED spatial arrangement, where the lights were unevenly populated throughout the reactor. The optimal designs may not have been developed from experienced designers due to the increased degrees of freedom offered by using UV-LEDs. The results of this study revealed that the coupled optimization routine with CFD was effective at significantly decreasing the engineer's design decision space and finding a potentially near-optimal UV-LED reactor solution. Published by Elsevier Ltd.

Rupture loop annex ion exchange RLAIX vault deactivation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ham, J.E.; Harris, D.L., Westinghouse Hanford

This engineering report documents the deactivation, stabilization and final conditions of the Rupture Loop Annex Ion Exchange (RLAIX) Vault located northwest of the 309 Building`s Plutonium Recycle Test Reactor (PRTR). Twelve ion exchange columns, piping debris, and column liquid were removed from the vault, packaged and shipped for disposal. The vault walls and floor were decontaminated, and portions of the vault were painted to fix loose contamination. Process piping and drains were plugged, and the cover blocks and rain cover were installed. Upon closure,the vault was empty, stabilized, isolated.

Transient plasma estimation: a noise cancelling/identification approach

DOE Office of Scientific and Technical Information (OSTI.GOV)

Candy, J.V.; Casper, T.; Kane, R.

1985-03-01

The application of a noise cancelling technique to extract energy storage information from sensors occurring during fusion reactor experiments on the Tandem Mirror Experiment-Upgrade (TMX-U) at the Lawrence Livermore National Laboratory (LLNL) is examined. We show how this technique can be used to decrease the uncertainty in the corresponding sensor measurements used for diagnostics in both real-time and post-experimental environments. We analyze the performance of algorithm on the sensor data and discuss the various tradeoffs. The algorithm suggested is designed using SIG, an interactive signal processing package developed at LLNL.

FY 2017 – Thermal Aging Effects on Advanced Structural Materials

DOE Office of Scientific and Technical Information (OSTI.GOV)

Li, Meimei; Natesan, K; Chen, Wei-Ying

This report provides an update on the evaluation of the effect of thermal aging on tensile properties of existing laboratory-sized heats of Alloy 709 austenitic stainless steel and the completion of effort on the thermal aging effect on the tensile properties of optimized G92 ferritic-martensitic steel. The report is a Level 3 deliverable in FY17 (M3AT-17AN1602081), under the Work Package AT-17AN160208, “Advanced Alloy Testing - ANL” performed by the Argonne National Laboratory (ANL), as part of the Advanced Reactor Technologies Program.

Structural properties of lead-lithium alloys

NASA Astrophysics Data System (ADS)

Khambholja, S. G.; Satikunvar, D. D.; Abhishek, Agraj; Thakore, B. Y.

2018-05-01

Lead-Lihtium alloys have found large number of applications as liquid metal coolants in nuclear reactors. Large number of experimental work is reported for this system. However, complete theoretical description is still rare. In this scenario, we in the present work report the study of ground state properties of Lead-Lithium system. The present study is performed using plane wave pseudopotential density functional theory as implemented in Quantum ESPRESSO package. The theoretical findings are in agreement with previously reported experimental data. Some conclusions are drawn based on present study, which will be helpful for a comprehensive study.

Heat flux estimates of power balance on Proto-MPEX with IR imaging

DOE Office of Scientific and Technical Information (OSTI.GOV)

Showers, M., E-mail: mshower1@vols.utk.edu; Oak Ridge National Laboratory, Oak Ridge, Tennessee 37831; Biewer, T. M.

The Prototype Material Plasma Exposure eXperiment (Proto-MPEX) at Oak Ridge National Laboratory (ORNL) is a precursor linear plasma device to the Material Plasma Exposure eXperiment (MPEX), which will study plasma material interactions (PMIs) for future fusion reactors. This paper will discuss the initial steps performed towards completing a power balance on Proto-MPEX to quantify where energy is lost from the plasma, including the relevant diagnostic package implemented. Machine operating parameters that will improve Proto-MPEX’s performance may be identified, increasing its PMI research capabilities.

Remaining Sites Verification Package for the 118-C-3:3, 105-C French Drains, Waste Site Reclassification Form 2006-016

DOE Office of Scientific and Technical Information (OSTI.GOV)

R. A. Carlson

2006-04-24

The 118-C-3:3 french drains received condensate from the steam heating system in the 105-C Reactor Building. The 118-C-3:3 french drain meets the remedial action objectives specified in the Remaining Sites ROD. The results demonstrate that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River.

Cold Atmospheric-Pressure Plasmas Applied to Active Packaging of Fruits and Vegetables

NASA Astrophysics Data System (ADS)

Pedrow, Patrick; Fernandez, Sulmer; Pitts, Marvin

2008-10-01

Active packaging of fruits and vegetables uses films that absorb molecules from or contribute molecules to the produce. Applying uniform film to specific parts of a plant will enhance safe and economic adoption of expensive biofilms and biochemicals which would damage the plant or surrounding environment if misapplied. The pilot application will be to apply wax film to apples, replacing hot wax which is expensive and lowers the textural quality of the apple. The plasma zone will be obtained by increasing the voltage on an electrode structure until the electric field in the feed material (Argon + monomer) is sufficiently high to yield electron avalanches. The ``corona onset criterion'' is used to design the cold plasma reactor. The apple will be placed in a treatment chamber downstream from the activation zone. Key physical properties of the film will be measured. The deposition rate will be optimized in terms of economics and fruit surface quality for the purpose of determining if the technique is competitive in food processing plants.

Impact investigation of reactor fuel operating parameters on reactivity for use in burnup credit applications

NASA Astrophysics Data System (ADS)

Sloma, Tanya Noel

When representing the behavior of commercial spent nuclear fuel (SNF), credit is sought for the reduced reactivity associated with the net depletion of fissile isotopes and the creation of neutron-absorbing isotopes, a process that begins when a commercial nuclear reactor is first operated at power. Burnup credit accounts for the reduced reactivity potential of a fuel assembly and varies with the fuel burnup, cooling time, and the initial enrichment of fissile material in the fuel. With regard to long-term SNF disposal and transportation, tremendous benefits, such as increased capacity, flexibility of design and system operations, and reduced overall costs, provide an incentive to seek burnup credit for criticality safety evaluations. The Nuclear Regulatory Commission issued Interim Staff Guidance 8, Revision 2 in 2002, endorsing burnup credit of actinide composition changes only; credit due to actinides encompasses approximately 30% of exiting pressurized water reactor SNF inventory and could potentially be increased to 90% if fission product credit were accepted. However, one significant issue for utilizing full burnup credit, compensating for actinide and fission product composition changes, is establishing a set of depletion parameters that produce an adequately conservative representation of the fuel's isotopic inventory. Depletion parameters can have a significant effect on the isotopic inventory of the fuel, and thus the residual reactivity. This research seeks to quantify the reactivity impact on a system from dominant depletion parameters (i.e., fuel temperature, moderator density, burnable poison rod, burnable poison rod history, and soluble boron concentration). Bounding depletion parameters were developed by statistical evaluation of a database containing reactor operating histories. The database was generated from summary reports of commercial reactor criticality data. Through depletion calculations, utilizing the SCALE 6 code package, several light water reactor assembly designs and in-core locations are analyzed in establishing a combination of depletion parameters that conservatively represent the fuel's isotopic inventory as an initiative to take credit for fuel burnup in criticality safety evaluations for transportation and storage of SNF.

DOE Office of Scientific and Technical Information (OSTI.GOV)

BRISC is a developmental prototype for a nextgeneration “systems-level” integrated performance and safety code (IPSC) for nuclear reactors. Its development served to demonstrate how a lightweight multi-physics coupling approach can be used to tightly couple the physics models in several different physics codes (written in a variety of languages) into one integrated package for simulating accident scenarios in a liquid sodium cooled “burner” nuclear reactor. For example, the RIO Fluid Flow and Heat transfer code developed at Sandia (SNL: Chris Moen, Dept. 08005) is used in BRISC to model fluid flow and heat transfer, as well as conduction heat transfermore » in solids. Because BRISC is a prototype, its most practical application is as a foundation or starting point for developing a true production code. The sub-codes and the associated models and correlations currently employed within BRISC were chosen to cover the required application space and demonstrate feasibility, but were not optimized or validated against experimental data within the context of their use in BRISC.« less

Conceptual design of the DEMO neutral beam injectors: main developments and R&D achievements

NASA Astrophysics Data System (ADS)

Sonato, P.; Agostinetti, P.; Bolzonella, T.; Cismondi, F.; Fantz, U.; Fassina, A.; Franke, T.; Furno, I.; Hopf, C.; Jenkins, I.; Sartori, E.; Tran, M. Q.; Varje, J.; Vincenzi, P.; Zanotto, L.

2017-05-01

The objectives of the nuclear fusion power plant DEMO, to be built after the ITER experimental reactor, are usually understood to lie somewhere between those of ITER and a ‘first of a kind’ commercial plant. Hence, in DEMO the issues related to efficiency and RAMI (reliability, availability, maintainability and inspectability) are among the most important drivers for the design, as the cost of the electricity produced by this power plant will strongly depend on these aspects. In the framework of the EUROfusion Work Package Heating and Current Drive within the Power Plant Physics and Development activities, a conceptual design of the neutral beam injector (NBI) for the DEMO fusion reactor has been developed by Consorzio RFX in collaboration with other European research institutes. In order to improve efficiency and RAMI aspects, several innovative solutions have been introduced in comparison to the ITER NBI, mainly regarding the beam source, neutralizer and vacuum pumping systems.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Stimpson, Shane G.

Activities to incorporate fuel performance capabilities into the Virtual Environment for Reactor Applications (VERA) are receiving increasing attention. The multiphysics emphasis is expanding as the neutronics (MPACT) and thermal-hydraulics (CTF) packages are becoming more mature. Capturing the finer details of fuel phenomena (swelling, densification, relocation, gap closure, etc.) is the natural next step in the VERA Core Simulator (VERA-CS) development process since these phenomena are currently not directly taken into account. While several codes could be used to accomplish this, the BISON fuel performance code being developed by the Idaho National Laboratory (INL) is the focus of ongoing work inmore » the Consortium for Advanced Simulation of Light Water Reactors (CASL). Built on INL’s MOOSE framework, BISON uses the finite element method for geometric representation and a Jacobian-free Newton-Krylov (JFNK) scheme to solve systems of partial differential equations for various fuel characteristic relationships. There are several modes of operation in BISON, but, for this work, it uses a 2D azimuthally symmetric (R-Z) smeared-pellet model.« less

On-site treatment of a motorway service area wastewater using a package sequencing batch reactor (SBR).

PubMed

Del Solar, J; Hudson, S; Stephenson, T

2005-01-01

A sequencing batch reactor (SBR) treating the effluent of a motorway service station in the south of England situated on a major tourist route was investigated. Wastewater from the kitchens, toilets and washrooms facilities was collected from the areas on each side of the motorway for treatment on-site. The SBR was designed for a population equivalent (p.e.) of 500, assuming an average flow of 100 m3/d, influent biochemical oxygen demand (BOD) of 300 mg/l, and influent suspended solids (SS) of 300 mg/l. Influent monitoring over 8 weeks revealed that the average flow was only 65 m3/d and the average influent BOD and SS were 480 mg/l and 473 mg/l respectively. This corresponded to a high sludge loading rate (F:M) of 0.42 d(-1) which accounted for poor performance. Therefore the cycle times were extended from 6 h to 7 h and effluent BOD improved from 79 to 27 mg/l.

Modeling solid thermal explosion containment on reactor HNIW and HMX.

PubMed

Lin, Chun-Ping; Chang, Chang-Ping; Chou, Yu-Chuan; Chu, Yung-Chuan; Shu, Chi-Min

2010-04-15

2,4,6,8,10,12-Hexanitro-2,4,6,8,10,12-hexaaza-isowurtzitane (HNIW), also known as CL-20 and octahydro-1,3,5,7-tetranitro-1,3,5,7-tetrazocine (HMX), are highly energetic materials which have been popular in national defense industries for years. This study established the models of thermal decomposition and thermal explosion hazard for HNIW and HMX. Differential scanning calorimetry (DSC) data were used for parameters determination of the thermokinetic models, and then these models were employed for simulation of thermal explosion in a 437L barrel reactor and a 24 kg cubic box package. Experimental results indicating the best storage conditions to avoid any violent runaway reaction of HNIW and HMX were also discovered. This study also developed an efficient procedure regarding creation of thermokinetics and assessment of thermal hazards of HNIW and HMX that could be applied to ensure safe storage conditions. 2009 Elsevier B.V. All rights reserved.

Study of Convection Heat Transfer in a Very High Temperature Reactor Flow Channel: Numerical and Experimental Results

DOE PAGES

Valentin, Francisco I.; Artoun, Narbeh; Anderson, Ryan; ...

2016-12-01

Very High Temperature Reactors (VHTRs) are one of the Generation IV gas-cooled reactor models proposed for implementation in next generation nuclear power plants. A high temperature/pressure test facility for forced and natural circulation experiments has been constructed. This test facility consists of a single flow channel in a 2.7 m (9’) long graphite column equipped with four 2.3kW heaters. Extensive 3D numerical modeling provides a detailed analysis of the thermal-hydraulic behavior under steady-state, transient, and accident scenarios. In addition, forced/mixed convection experiments with air, nitrogen and helium were conducted for inlet Reynolds numbers from 500 to 70,000. Our numerical resultsmore » were validated with forced convection data displaying maximum percentage errors under 15%, using commercial finite element package, COMSOL Multiphysics. Based on this agreement, important information can be extracted from the model, with regards to the modified radial velocity and property gas profiles. Our work also examines flow laminarization for a full range of Reynolds numbers including laminar, transition and turbulent flow under forced convection and its impact on heat transfer under various scenarios to examine the thermal-hydraulic phenomena that could occur during both normal operation and accident conditions.« less

A computationally simple model for determining the time dependent spectral neutron flux in a nuclear reactor core

NASA Astrophysics Data System (ADS)

Schneider, E. A.; Deinert, M. R.; Cady, K. B.

2006-10-01

The balance of isotopes in a nuclear reactor core is key to understanding the overall performance of a given fuel cycle. This balance is in turn most strongly affected by the time and energy-dependent neutron flux. While many large and involved computer packages exist for determining this spectrum, a simplified approach amenable to rapid computation is missing from the literature. We present such a model, which accepts as inputs the fuel element/moderator geometry and composition, reactor geometry, fuel residence time and target burnup and we compare it to OECD/NEA benchmarks for homogeneous MOX and UOX LWR cores. Collision probability approximations to the neutron transport equation are used to decouple the spatial and energy variables. The lethargy dependent neutron flux, governed by coupled integral equations for the fuel and moderator/coolant regions is treated by multigroup thermalization methods, and the transport of neutrons through space is modeled by fuel to moderator transport and escape probabilities. Reactivity control is achieved through use of a burnable poison or adjustable control medium. The model calculates the buildup of 24 actinides, as well as fission products, along with the lethargy dependent neutron flux and the results of several simulations are compared with benchmarked standards.

Furfural-based polymers for the sealing of reactor vessels dumped in the Arctic Kara Sea

DOE Office of Scientific and Technical Information (OSTI.GOV)

HEISER,J.H.; COWGILL,M.G.; SIVINTSEV,Y.V.

1996-10-07

Between 1965 and 1988, 16 naval reactor vessels were dumped in the Arctic Kara Sea. Six of the vessels contained spent nuclear fuel that had been damaged during accidents. In addition, a container holding {approximately} 60% of the damaged fuel from the No. 2 reactor of the atomic icebreaker Lenin was dumped in 1967. Before dumping, the vessels were filled with a solidification agent, Conservant F, in order to prevent direct contact between the seawater and the fuel and other activated components, thereby reducing the potential for release of radionuclides into the environment. The key ingredient in Conservant F ismore » furfural (furfuraldehyde). Other constituents vary, depending on specific property requirements, but include epoxy resin, mineral fillers, and hardening agents. In the liquid state (prior to polymerization) Conservant F is a low viscosity, homogeneous resin blend that provides long work times (6--9 hours). In the cured state, Conservant F provides resistance to water and radiation, has high adhesion properties, and results in minimal gas evolution. This paper discusses the properties of Conservant F in both its cured and uncured states and the potential performance of the waste packages containing spent nuclear fuel in the Arctic Kara Sea.« less

Electrons to Reactors Multiscale Modeling: Catalytic CO Oxidation over RuO 2

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sutton, Jonathan E.; Lorenzi, Juan M.; Krogel, Jaron T.

First-principles kinetic Monte Carlo (1p-kMC) simulations for CO oxidation on two RuO 2 facets, RuO 2(110) and RuO 2(111), were coupled to the computational fluid dynamics (CFD) simulations package MFIX, and reactor-scale simulations were then performed. 1p-kMC coupled with CFD has recently been shown as a feasible method for translating molecular scale mechanistic knowledge to the reactor scale, enabling comparisons to in situ and online experimental measurements. Only a few studies with such coupling have been published. This work incorporates multiple catalytic surface facets into the scale-coupled simulation, and three possibilities were investigated: the two possibilities of each facet individuallymore » being the dominant phase in the reactor, and also the possibility that both facets were present on the catalyst particles in the ratio predicted by an ab initio thermodynamics-based Wulff construction. When lateral interactions between adsorbates were included in the 1p-kMC simulations, the two surfaces, RuO 2(110) and RuO 2(111), were found to be of similar order-of-magnitude in activity for the pressure range of 1 × 10 –4 bar to 1 bar, with the RuO 2(110) surface-termination showing more simulated activity than the RuO 2(111) surface-termination. Coupling between the 1p-kMC and CFD was achieved with a lookup table generated by the error-based modified Shepard interpolation scheme. Isothermal reactor scale simulations were performed and compared to two separate experimental studies, conducted with reactant partial pressures of ≤0.1 bar. Simulations without an isothermality restriction were also conducted and showed that the simulated temperature gradient across the catalytic reactor bed is <0.5 K, which validated the use of the isothermality restriction for investigating the reactor-scale phenomenological temperature dependences. The approach with the Wulff construction based reactor simulations reproduced a trend similar to one experimental data set relatively well, with the (110) surface being more active at higher temperaures; in contrast, for the other experimental data set, our reactor simulations achieve surprisingly and perhaps fortuitously good agreement with the activity and phenomenological pressure dependence when it is assumed that the (111) facet is the only active facet present. Lastly, the active phase of catalytic CO oxidation over RuO 2 remains unsettled, but the present study presents proof of principle (and progress) toward more accurate multiscale modeling from electrons to reactors and new simulation results.« less

Electrons to Reactors Multiscale Modeling: Catalytic CO Oxidation over RuO 2

DOE PAGES

Sutton, Jonathan E.; Lorenzi, Juan M.; Krogel, Jaron T.; ...

2018-04-20

First-principles kinetic Monte Carlo (1p-kMC) simulations for CO oxidation on two RuO 2 facets, RuO 2(110) and RuO 2(111), were coupled to the computational fluid dynamics (CFD) simulations package MFIX, and reactor-scale simulations were then performed. 1p-kMC coupled with CFD has recently been shown as a feasible method for translating molecular scale mechanistic knowledge to the reactor scale, enabling comparisons to in situ and online experimental measurements. Only a few studies with such coupling have been published. This work incorporates multiple catalytic surface facets into the scale-coupled simulation, and three possibilities were investigated: the two possibilities of each facet individuallymore » being the dominant phase in the reactor, and also the possibility that both facets were present on the catalyst particles in the ratio predicted by an ab initio thermodynamics-based Wulff construction. When lateral interactions between adsorbates were included in the 1p-kMC simulations, the two surfaces, RuO 2(110) and RuO 2(111), were found to be of similar order-of-magnitude in activity for the pressure range of 1 × 10 –4 bar to 1 bar, with the RuO 2(110) surface-termination showing more simulated activity than the RuO 2(111) surface-termination. Coupling between the 1p-kMC and CFD was achieved with a lookup table generated by the error-based modified Shepard interpolation scheme. Isothermal reactor scale simulations were performed and compared to two separate experimental studies, conducted with reactant partial pressures of ≤0.1 bar. Simulations without an isothermality restriction were also conducted and showed that the simulated temperature gradient across the catalytic reactor bed is <0.5 K, which validated the use of the isothermality restriction for investigating the reactor-scale phenomenological temperature dependences. The approach with the Wulff construction based reactor simulations reproduced a trend similar to one experimental data set relatively well, with the (110) surface being more active at higher temperaures; in contrast, for the other experimental data set, our reactor simulations achieve surprisingly and perhaps fortuitously good agreement with the activity and phenomenological pressure dependence when it is assumed that the (111) facet is the only active facet present. Lastly, the active phase of catalytic CO oxidation over RuO 2 remains unsettled, but the present study presents proof of principle (and progress) toward more accurate multiscale modeling from electrons to reactors and new simulation results.« less

PWR upper/lower internals shield

DOE Office of Scientific and Technical Information (OSTI.GOV)

Homyk, W.A.

1995-03-01

During refueling of a nuclear power plant, the reactor upper internals must be removed from the reactor vessel to permit transfer of the fuel. The upper internals are stored in the flooded reactor cavity. Refueling personnel working in containment at a number of nuclear stations typically receive radiation exposure from a portion of the highly contaminated upper intervals package which extends above the normal water level of the refueling pool. This same issue exists with reactor lower internals withdrawn for inservice inspection activities. One solution to this problem is to provide adequate shielding of the unimmersed portion. The use ofmore » lead sheets or blankets for shielding of the protruding components would be time consuming and require more effort for installation since the shielding mass would need to be transported to a support structure over the refueling pool. A preferable approach is to use the existing shielding mass of the refueling pool water. A method of shielding was devised which would use a vacuum pump to draw refueling pool water into an inverted canister suspended over the upper internals to provide shielding from the normally exposed components. During the Spring 1993 refueling of Indian Point 2 (IP2), a prototype shield device was demonstrated. This shield consists of a cylindrical tank open at the bottom that is suspended over the refueling pool with I-beams. The lower lip of the tank is two feet below normal pool level. After installation, the air width of the natural shielding provided by the existing pool water. This paper describes the design, development, testing and demonstration of the prototype device.« less

Physical particularities of nuclear reactors using heavy moderators of neutrons

NASA Astrophysics Data System (ADS)

Kulikov, G. G.; Shmelev, A. N.

2016-12-01

In nuclear reactors, thermal neutron spectra are formed using moderators with small atomic weights. For fast reactors, inserting such moderators in the core may create problems since they efficiently decelerate the neutrons. In order to form an intermediate neutron spectrum, it is preferable to employ neutron moderators with sufficiently large atomic weights, using 233U as a fissile nuclide and 232Th and 231Pa as fertile ones. The aim of the work is to investigate the properties of heavy neutron moderators and to assess their advantages. The analysis employs the JENDL-4.0 nuclear data library and the SCALE program package for simulating the variation of fuel composition caused by irradiation in the reactor. The following main results are obtained. By using heavy moderators with small neutron moderation steps, one is able to (1) increase the rate of resonance capture, so that the amount of fertile material in the fuel may be reduced while maintaining the breeding factor of the core; (2) use the vacant space for improving the fuel-element properties by adding inert, strong, and thermally conductive materials and by implementing dispersive fuel elements in which the fissile material is self-replenished and neutron multiplication remains stable during the process of fuel burnup; and (3) employ mixtures of different fertile materials with resonance capture cross sections in order to increase the resonance-lattice density and the probability of resonance neutron capture leading to formation of fissile material. The general conclusion is that, by forming an intermediate neutron spectrum with heavy neutron moderators, one can use the fuel more efficiently and improve nuclear safety.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Harmon, K.M.; Lakey, L.T.; Leigh, I.W.

Worldwide activities related to nuclear fuel cycle and radioactive waste management programs are summarized. Several trends have developed in waste management strategy: All countries having to dispose of reprocessing wastes plan on conversion of the high-level waste (HLW) stream to a borosilicate glass and eventual emplacement of the glass logs, suitably packaged, in a deep geologic repository. Countries that must deal with plutonium-contaminated waste emphasize pluonium recovery, volume reduction and fixation in cement or bitumen in their treatment plans and expect to use deep geologic repositories for final disposal. Commercially available, classical engineering processing are being used worldwide to treatmore » and immobilize low- and intermediate-level wastes (LLW, ILW); disposal to surface structures, shallow-land burial and deep-underground repositories, such as played-out mines, is being done widely with no obvious technical problems. Many countries have established extensive programs to prepare for construction and operation of geologic repositories. Geologic media being studied fall into three main classes: argillites (clay or shale); crystalline rock (granite, basalt, gneiss or gabbro); and evaporates (salt formations). Most nations plan to allow 30 years or longer between discharge of fuel from the reactor and emplacement of HLW or spent fuel is a repository to permit thermal and radioactive decay. Most repository designs are based on the mined-gallery concept, placing waste or spent fuel packages into shallow holes in the floor of the gallery. Many countries have established extensive and costly programs of site evaluation, repository development and safety assessment. Two other waste management problems are the subject of major R and D programs in several countries: stabilization of uranium mill tailing piles; and immobilization or disposal of contaminated nuclear facilities, namely reactors, fuel cycle plants and R and D laboratories.« less

Production of an impermeable composite of irradiated graphite and glass by hot isostatic pressing as a long term leach resistant waste form

DOE Office of Scientific and Technical Information (OSTI.GOV)

Fachinger, Johannes; Muller, Walter; Marsat, Eric

2013-07-01

Around 250,000 tons of irradiated graphite (i-graphite) exists worldwide and can be considered as a current waste or future waste stream. The largest national i-graphite inventory is located in UK (∼ 100,000 tons) with significant quantities also in Russia and France [5]. Most of the i-graphite remains in the cores of shutdown nuclear reactors including the MAGNOX type in UK and the UNGG in France. Whilst there are still operational power reactors with graphite cores, such as the Russian RBMKs and the AGRs in UK, all of them will reach their end of life during the next two decades. Themore » most common reference waste management option of i-graphite is a wet or dry retrieval of the graphite blocks from the reactor core and the grouting of these blocks in a container without further conditioning. This produces large waste package volumes because the encapsulation capacity of the grout is limited and large cavities in the graphite blocks could reduce the packing densities. Packing densities from 0.5 to 1 tons per cubic meter have been assumed for grouting solutions. Furthermore the grout is permeable. This could over time allow the penetration of aqueous phases into the waste block and a potential dissolution and release of radionuclides. As a result particularly highly soluble radionuclides may not be retained by the grout. Vitrification could present an alternative, however a similar waste package volume increase may be expected since the encapsulation capacity of glass is potentially similar to or worse than that of grout. FNAG has developed a process for the production of a graphite-glass composite material called Impermeable Graphite Matrix (IGM) [3]. This process is also applicable to irradiated graphite which allows the manufacturing of an impermeable material without volume increase. Crushed i-graphite is mixed with 20 vol.% of glass and then pressed under vacuum at an elevated temperature in an axial hot vacuum press (HVP). The obtained product has zero or negligible porosity and a water impermeable structure. Structural analysis shows that the glass in the composite has replaced the pores in the graphite structure. The typical pore volume of a graphite material is in the range of 20 vol.%. Therefore no volume increase will occur in comparison with the former graphite material. This IGM material will allow the encapsulation of graphite with package densities larger than 1.5 ton per cubic meter. Therefore a huge volume saving can be achieved by such an alternative encapsulation method. Disposal performance is also enhanced since little or no leaching of radionuclides is observed due to the impermeability of the material NNL and FNAG have proved that IGM can be produced by hot isostatic pressing (HIP) which has several advantages for radioactive materials over the HVP process. - The sealed HIP container avoids the release of any radionuclides. - The outside of the waste package is not contaminated. - The HIP process time is shorter than the HVP process time. The isostatic press avoids anisotropic density distributions. - Simple filling of the HIP container has advantages over the filling of an axial die. (authors)« less

33 Shafts Category of Transuranic Waste Stored Below Ground within Area G

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hargis, Kenneth Marshall; Monk, Thomas H

This report compiles information to support the evaluation of alternatives and analysis of regulatory paths forward for the 33 shafts. The historical information includes a form completed by waste generators for each waste package (Reference 6) that included a waste description, estimates of Pu-239 and uranium-235 (U-235) based on an accounting technique, and calculations of mixed fission products (MFP) based on radiation measurements. A 1979 letter and questionnaire (Reference 7) provides information on waste packaging of hot cell waste and the configuration of disposal shafts as storage in the 33 Shafts was initiated. Tables of data by waste package weremore » developed during a review of historical documents that was performed in 2005 (Reference 8). Radiological data was coupled with material-type data to estimate the initial isotopic content of each waste package and an Oak Ridge National Laboratory computer code was used to calculate 2009 decay levels. Other sources of information include a waste disposal logbook for the 33 shafts (Reference 9), reports that summarize remote-handled waste generated at the CMR facility (Reference 10) and placement of waste in the 33 shafts (Reference 11), a report on decommissioning of the LAMPRE reactor (Reference 12), interviews with an employee and manager involved in placing waste in the 33 shafts (References 13 and 14), an interview with a long-time LANL employee involved in waste operations (Reference 15), a 2002 plan for disposition of remote-handled TRU waste (Reference 16), and photographs obtained during field surveys of several shafts in 2007. The WIPP Central Characterization Project (CCP) completed an Acceptable Knowledge (AK) summary report for 16 canisters of remote-handled waste from the CMR Facility that contains information relevant to the 33 Shafts on hot-cell operations and timeline (Reference 17).« less

Automated Work Package: Initial Wireless Communication Platform Design, Development, and Evaluation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Al Rashdan, Ahmad Yahya Mohammad; Agarwal, Vivek

The Department of Energy’s Light Water Reactor Sustainability Program is developing the scientific basis to ensure long-term reliability, productivity, safety, and security of the nuclear power industry in the United States. The Instrumentation, Information, and Control (II&C) pathway of the program aims to increase the role of advanced II&C technologies to achieve this objective. One of the pathway efforts at Idaho National Laboratory (INL) is to improve the work packages execution process by replacing the expensive, inefficient, bulky, complex, and error-prone paper-based work orders with automated work packages (AWPs). An AWP is an automated and dynamic presentation of the workmore » package designed to guide the user through the work process. It is loaded on a mobile device, such as a tablet, and is capable of communicating with plant equipment and systems to acquire plant and procedure states. The AWP replaces those functions where a computer is more efficient and reliable than a human. To enable the automatic acquisition of plant data, it is necessary to design and develop a prototype platform for data exchange between the field instruments and the AWP mobile devices. The development of the platform aims to reveal issues and solutions generalizable to large-scale implementation of a similar system. Topics such as bandwidth, robustness, response time, interference, and security are usually associated with wireless communication. These concerns, along with other requirements, are listed in an earlier INL report. Specifically, the targeted issues and performance aspects in this work are relevant to the communication infrastructure from the perspective of promptness, robustness, expandability, and interoperability with different technologies.« less

Anaerobic treatment for C and S removal in "zero-discharge" paper mills: effects of process design on S removal efficiencies.

PubMed

van Lier, J B; Lens, P N; Pol, L W

2001-01-01

Stringent environmental laws in Europe and Northern America lead to the development towards closure of the process water streams in pulp and paper mills. Application of a "zero-discharge" process is already a feasible option for the board and packaging paper industry, provided in-line treatment is applied. Concomitant energy conservation inside the mill results in process water temperatures of 50-60 degrees C. Thermophilic anaerobic treatment complemented with appropriate post-treatment is considered as the most cost-effective solution to meet re-use criteria of the process water and to keep its temperature. In the proposed closed-cycle, the anaerobic treatment step removes the largest fraction of the biodegradable COD and eliminates "S" as H2S from the process stream, without the use of additional chemicals. The anaerobic step is regarded as the only possible location to bleed "S" from the process water cycle. In laboratory experiments, the effect of upward liquid velocity (Vupw) and the specific gas loading rate (Vgas) on the S removal capacity of thermophilic anaerobic bio-reactors was investigated. Acidifying, sulphate reducing sludge bed reactors were fed with partly acidified synthetic paper mill wastewater and were operated at 55 degrees C and pH 6. The reactors were operated at organic loading rates up to 50 g COD.l-1.day-1 at COD/SO4(2-) ratios of 10. The effect of Vupw was researched by comparing the performance of a UASB reactor operated at 1.0 m.h-1 and an EGSB reactor, operated at 6.8 m.h-1. The Vupw had a strong effect on the fermentation patterns. In the UASB reactor, acidification yielded H2, acetate and propionate, leading to an accumulation of reducing equivalents. These were partly disposed of by the production of n-butyrate and n-valerate from propionate. In the EGSB reactor net acetate consumption was observed as well as high volumetric gas (CO2 and CH4) production rates. The higher gas production rates in the EGSB reactor resulted in higher S-stripping efficiencies. The effect of Vgas was further researched by comparing 2 UASB reactors which were sparged with N2 gas at a specific gas loading rate of 30 m3.m-2.day-1. In contrast to the regular UASB reactors, the gas-supplied UASB showed a more stable performance when the organic loading rates were increased. Also, the H2S stripping efficiency was 3-4 times higher in the gas-supplied UASB, reaching values of 67%. Higher values were not obtained owing to the relatively poor sulphate reduction efficiencies.

Comprehensive work plan for Building 3001 storage canal at the Oak Ridge National Laboratory, Oak Ridge, Tennessee

DOE Office of Scientific and Technical Information (OSTI.GOV)

NONE

1997-01-01

This Comprehensive Work Plan describes the method of accomplishment to replace the shielding protection of the water in the canal with a controlled low strength material (CLSM) 4. The canal was used during the operation of the Oak Ridge Graphite Reactor in the 1940s and 1950s to transport spent fuel slugs and irradiated test materials from the reactor, under water to the hot cell in Building 3019 for further processing, packaging, and handling. After the reactor was shut down, the canal was used until 1990 to store some irradiated materials until they could be transferred to a Solid Waste Storagemore » Area. This task has the following objectives and components: (1) minimize potential future risk to human health and the environment; (2) reduce surveillance and maintenance cost of the canal; (3) perform site preparation activities; (4) replace the water in the canal with a solid CLSM; (5) pump the water to the Process Waste Treatment System (PWTS) for further processing at the same rate that the CLSM is pumped under the water; (6) remove the water using a process that will protect the workers and the public in the visitors area from contamination while the CLSM is being pumped underneath the water; (7) painting a protective coating material over the CLSM after the CLSM has cured.« less

SFCOMPO-2.0: An OECD NEA database of spent nuclear fuel isotopic assays, reactor design specifications, and operating data

DOE PAGES

Michel-Sendis, F.; Gauld, I.; Martinez, J. S.; ...

2017-08-02

SFCOMPO-2.0 is the new release of the Organisation for Economic Co-operation and Development (OECD) Nuclear Energy Agency (NEA) database of experimental assay measurements. These measurements are isotopic concentrations from destructive radiochemical analyses of spent nuclear fuel (SNF) samples. We supplement the measurements with design information for the fuel assembly and fuel rod from which each sample was taken, as well as with relevant information on operating conditions and characteristics of the host reactors. These data are necessary for modeling and simulation of the isotopic evolution of the fuel during irradiation. SFCOMPO-2.0 has been developed and is maintained by the OECDmore » NEA under the guidance of the Expert Group on Assay Data of Spent Nuclear Fuel (EGADSNF), which is part of the NEA Working Party on Nuclear Criticality Safety (WPNCS). Significant efforts aimed at establishing a thorough, reliable, publicly available resource for code validation and safety applications have led to the capture and standardization of experimental data from 750 SNF samples from more than 40 reactors. These efforts have resulted in the creation of the SFCOMPO-2.0 database, which is publicly available from the NEA Data Bank. Our paper describes the new database, and applications of SFCOMPO-2.0 for computer code validation, integral nuclear data benchmarking, and uncertainty analysis in nuclear waste package analysis are briefly illustrated.« less

Reactor plasma facing component designs based on liquid metal concepts supported in porous systems

NASA Astrophysics Data System (ADS)

Tabarés, F. L.; Oyarzabal, E.; Martin-Rojo, A. B.; Tafalla, D.; de Castro, A.; Soleto, A.

2017-01-01

The use of liquid metals (LMs) as plasma facing components in fusion devices was proposed as early as 1970 for a field reversed concept and inertial fusion reactors. The idea was extensively developed during the APEX Project, at the turn of the century, and it is the subject at present of the biennial International Symposium on Lithium Applications (ISLA), whose fourth meeting took place in Granada, Spain at the end of September 2015. While liquid metal flowing concepts were specially addressed in USA research projects, the idea of embedding the metal in a capillary porous system (CPS) was put forwards by Russian teams in the 1990s, thus opening the possibility of static concepts. Since then, many ideas and accompanying experimental tests in fusion devices and laboratories have been produced, involving a large fraction of countries within the international fusion community. Within the EUROFusion Roadmap, these activities are encompassed into the working programs of the plasma facing components (PFC) and divertor tokamak test (DTT) packages. In this paper, a review of the state of the art in concepts based on the CPS set-up for a fusion reactor divertor target, aimed at preventing the ejection of the liquid metal by electro-magnetic (EM) forces generated under plasma operation, is described and required R+D activities on the topic, including ongoing work at CIEMAT specifically oriented to filling the remaining gaps, are stressed.

SFCOMPO-2.0: An OECD NEA database of spent nuclear fuel isotopic assays, reactor design specifications, and operating data

DOE Office of Scientific and Technical Information (OSTI.GOV)

Michel-Sendis, F.; Gauld, I.; Martinez, J. S.

SFCOMPO-2.0 is the new release of the Organisation for Economic Co-operation and Development (OECD) Nuclear Energy Agency (NEA) database of experimental assay measurements. These measurements are isotopic concentrations from destructive radiochemical analyses of spent nuclear fuel (SNF) samples. We supplement the measurements with design information for the fuel assembly and fuel rod from which each sample was taken, as well as with relevant information on operating conditions and characteristics of the host reactors. These data are necessary for modeling and simulation of the isotopic evolution of the fuel during irradiation. SFCOMPO-2.0 has been developed and is maintained by the OECDmore » NEA under the guidance of the Expert Group on Assay Data of Spent Nuclear Fuel (EGADSNF), which is part of the NEA Working Party on Nuclear Criticality Safety (WPNCS). Significant efforts aimed at establishing a thorough, reliable, publicly available resource for code validation and safety applications have led to the capture and standardization of experimental data from 750 SNF samples from more than 40 reactors. These efforts have resulted in the creation of the SFCOMPO-2.0 database, which is publicly available from the NEA Data Bank. Our paper describes the new database, and applications of SFCOMPO-2.0 for computer code validation, integral nuclear data benchmarking, and uncertainty analysis in nuclear waste package analysis are briefly illustrated.« less

Results of the engineering run of the Coherent Neutrino Nucleus Interaction Experiment (CONNIE)

NASA Astrophysics Data System (ADS)

Aguilar-Arevalo, A.; Bertou, X.; Bonifazi, C.; Butner, M.; Cancelo, G.; Castañeda Vázquez, A.; Cervantes Vergara, B.; Chavez, C. R.; Da Motta, H.; D'Olivo, J. C.; Dos Anjos, J.; Estrada, J.; Fernandez Moroni, G.; Ford, R.; Foguel, A.; Hernández Torres, K. P.; Izraelevitch, F.; Kavner, A.; Kilminster, B.; Kuk, K.; Lima, H. P., Jr.; Makler, M.; Molina, J.; Moreno-Granados, G.; Moro, J. M.; Paolini, E. E.; Sofo Haro, M.; Tiffenberg, J.; Trillaud, F.; Wagner, S.

2016-07-01

The CONNIE detector prototype is operating at a distance of 30 m from the core of a 3.8 GWth nuclear reactor with the goal of establishing Charge-Coupled Devices (CCD) as a new technology for the detection of coherent elastic neutrino-nucleus scattering. We report on the results of the engineering run with an active mass of 4 g of silicon. The CCD array is described, and the performance observed during the first year is discussed. A compact passive shield was deployed around the detector, producing an order of magnitude reduction in the background rate. The remaining background observed during the run was stable, and dominated by internal contamination in the detector packaging materials. The in-situ calibration of the detector using X-ray lines from fluorescence demonstrates good stability of the readout system. The event rates with the reactor ON and OFF are compared, and no excess is observed coming from nuclear fission at the power plant. The upper limit for the neutrino event rate is set two orders of magnitude above the expectations for the standard model. The results demonstrate the cryogenic CCD-based detector can be remotely operated at the reactor site with stable noise below 2 e- RMS and stable background rates. The success of the engineering test provides a clear path for the upgraded 100 g detector to be deployed during 2016.

High-Temperature Optical Sensor

NASA Technical Reports Server (NTRS)

Adamovsky, Grigory; Juergens, Jeffrey R.; Varga, Donald J.; Floyd, Bertram M.

2010-01-01

A high-temperature optical sensor (see Figure 1) has been developed that can operate at temperatures up to 1,000 C. The sensor development process consists of two parts: packaging of a fiber Bragg grating into a housing that allows a more sturdy thermally stable device, and a technological process to which the device is subjected to in order to meet environmental requirements of several hundred C. This technology uses a newly discovered phenomenon of the formation of thermally stable secondary Bragg gratings in communication-grade fibers at high temperatures to construct robust, optical, high-temperature sensors. Testing and performance evaluation (see Figure 2) of packaged sensors demonstrated operability of the devices at 1,000 C for several hundred hours, and during numerous thermal cycling from 400 to 800 C with different heating rates. The technology significantly extends applicability of optical sensors to high-temperature environments including ground testing of engines, flight propulsion control, thermal protection monitoring of launch vehicles, etc. It may also find applications in such non-aerospace arenas as monitoring of nuclear reactors, furnaces, chemical processes, and other hightemperature environments where other measurement techniques are either unreliable, dangerous, undesirable, or unavailable.

Pyrolysis of plastic packaging waste: A comparison of plastic residuals from material recovery facilities with simulated plastic waste.

PubMed

Adrados, A; de Marco, I; Caballero, B M; López, A; Laresgoiti, M F; Torres, A

2012-05-01

Pyrolysis may be an alternative for the reclamation of rejected streams of waste from sorting plants where packing and packaging plastic waste is separated and classified. These rejected streams consist of many different materials (e.g., polyethylene (PE), polypropylene (PP), polystyrene (PS), polyvinyl chloride (PVC), polyethylene terephthalate (PET), acrylonitrile butadiene styrene (ABS), aluminum, tetra-brik, and film) for which an attempt at complete separation is not technically possible or economically viable, and they are typically sent to landfills or incinerators. For this study, a simulated plastic mixture and a real waste sample from a sorting plant were pyrolyzed using a non-stirred semi-batch reactor. Red mud, a byproduct of the aluminum industry, was used as a catalyst. Despite the fact that the samples had a similar volume of material, there were noteworthy differences in the pyrolysis yields. The real waste sample resulted, after pyrolysis, in higher gas and solid yields and consequently produced less liquid. There were also significant differences noted in the compositions of the compared pyrolysis products. Copyright © 2011 Elsevier Ltd. All rights reserved.

Automated Work Packages Prototype: Initial Design, Development, and Evaluation. Light Water Reactor Sustainability Program

DOE Office of Scientific and Technical Information (OSTI.GOV)

Oxstrand, Johanna Helene; Ahmad Al Rashdan; Le Blanc, Katya Lee

The goal of the Automated Work Packages (AWP) project is to demonstrate how to enhance work quality, cost management, and nuclear safety through the use of advanced technology. The work described in this report is part of the digital architecture for a highly automated plant project of the technical program plan for advanced instrumentation, information, and control (II&C) systems technologies. This report addresses the DOE Milestone M2LW-15IN0603112: Describe the outcomes of field evaluations/demonstrations of the AWP prototype system and plant surveillance and communication framework requirements at host utilities. A brief background to the need for AWP research is provided, thenmore » two human factors field evaluation studies are described. These studies focus on the user experience of conducting a task (in this case a preventive maintenance and a surveillance test) while using an AWP system. The remaining part of the report describes an II&C effort to provide real time status updates to the technician by wireless transfer of equipment indications and a dynamic user interface.« less

Remaining Sites Verification Package for the 100-F-26:15 Miscellaneous Pipelines Associated with the 132-F-6, 1608-F Waste Water Pumping Station, Waste Site Reclassification Form 2007-031

DOE Office of Scientific and Technical Information (OSTI.GOV)

L. M. Dittmer

2008-03-18

The 100-F-26:15 waste site consisted of the remnant portions of underground process effluent and floor drain pipelines that originated at the 105-F Reactor. In accordance with this evaluation, the verification sampling results support a reclassification of this site to Interim Closed Out. The results of verification sampling show that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River.

Cold Trap Dismantling and Sodium Removal at a Fast Breeder Reactor - 12327

DOE Office of Scientific and Technical Information (OSTI.GOV)

Graf, A.; Petrick, H.; Stutz, U.

2012-07-01

The first German prototype Fast Breeder Nuclear Reactor (KNK) is currently being dismantled after being the only operating Fast Breeder-type reactor in Germany. As this reactor type used sodium as a coolant in its primary and secondary circuit, seven cold traps containing various amounts of partially activated sodium needed to be disposed of as part of the dismantling. The resulting combined difficulties of radioactive contamination and high chemical reactivity were handled by treating the cold traps differently depending on their size and the amount of sodium contained inside. Six small cold traps were processed onsite by cutting them up intomore » small parts using a band saw under a protective atmosphere. The sodium was then converted to sodium hydroxide by using water. The remaining large cold trap could not be handled in the same way due to its dimensions (2.9 m x 1.1 m) and the declared amount of sodium inside (1,700 kg). It was therefore manually dismantled inside a large box filled with a protective atmosphere, while the resulting pieces were packaged for later burning in a special facility. The experiences gained by KNK during this process may be advantageous for future dismantling projects in similar sodium-cooled reactors worldwide. The dismantling of a prototype fast breeder reactor provides the challenge not only to dismantle radioactive materials but also to handle sodium-contaminated or sodium-containing components. The treatment of sodium requires additional equipment and installations to ensure a safe handling. Since it is not permitted to bring sodium into a repository, all sodium has to be neutralized either through a controlled reaction with water or by incinerating. The resulting components can be disposed of as normal radioactive waste with no further conditions. The handling of sodium needs skilled and experienced workers to minimize the inherent risks. And the example of the disposal of the large KNK cold trap shows the interaction with others and also foreign decommissioning projects can provide solutions with were unknown before. (authors)« less

Thermal valorization of post-consumer film waste in a bubbling bed gasifier.

PubMed

Martínez-Lera, S; Torrico, J; Pallarés, J; Gil, A

2013-07-01

The use of plastic bags and film packaging is very frequent in manifold sectors and film waste is usually present in different sources of municipal and industrial wastes. A significant part of it is not suitable for mechanical recycling but could be safely transformed into a valuable gas by means of thermal valorization. In this research, the gasification of film wastes has been experimentally investigated through experiments in a fluidized bed reactor of two reference polymers, polyethylene and polypropylene, and actual post-consumer film waste. After a complete experimental characterization of the three materials, several gasification experiments have been performed to analyze the influence of the fuel and of equivalence ratio on gas production and composition, on tar generation and on efficiency. The experiments prove that film waste and analogue polymer derived wastes can be successfully gasified in a fluidized bed reactor, yielding a gas with a higher heating value in a range from 3.6 to 5.6 MJ/m3 and cold gas efficiencies up to 60%. Copyright © 2013 Elsevier Ltd. All rights reserved.

A novel miniaturized PCR multi-reactor array fabricated using flip-chip bonding techniques

NASA Astrophysics Data System (ADS)

Zou, Zhi-Qing; Chen, Xiang; Jin, Qing-Hui; Yang, Meng-Su; Zhao, Jian-Long

2005-08-01

This paper describes a novel miniaturized multi-chamber array capable of high throughput polymerase chain reaction (PCR). The structure of the proposed device is verified by using finite element analysis (FEA) to optimize the thermal performance, and then implemented on a glass-silicon substrate using a standard MEMS process and post-processing. Thermal analysis simulation and verification of each reactor cell is equipped with integrated Pt temperature sensors and heaters at the bottom of the reaction chamber for real-time accurate temperature sensing and control. The micro-chambers are thermally separated from each other, and can be controlled independently. The multi-chip array was packaged on a printed circuit board (PCB) substrate using a conductive polymer flip-chip bonding technique, which enables effective heat dissipation and suppresses thermal crosstalk between the chambers. The designed system has successfully demonstrated a temperature fluctuation of ±0.5 °C during thermal multiplexing of up to 2 × 2 chambers, a full speed of 30 min for 30 cycle PCR, as well as the capability of controlling each chamber digitally and independently.

Thermodynamic and kinetic modelling of fuel oxidation behaviour in operating defective fuel

NASA Astrophysics Data System (ADS)

Lewis, operating defective fuel B. J.; Thompson, W. T.; Akbari, F.; Thompson, D. M.; Thurgood, C.; Higgs, J.

2004-07-01

A theoretical treatment has been developed to predict the fuel oxidation behaviour in operating defective nuclear fuel elements. The equilibrium stoichiometry deviation in the hyper-stoichiometric fuel has been derived from thermodynamic considerations using a self-consistent set of thermodynamic properties for the U-O system, which emphasizes replication of solubilities and three-phase invariant conditions displayed in the U-O binary phase diagram. The kinetics model accounts for multi-phase transport including interstitial oxygen diffusion in the solid and gas-phase transport of hydrogen and steam in the fuel cracks. The fuel oxidation model is further coupled to a heat conduction model to account for the feedback effect of a reduced thermal conductivity in the hyper-stoichiometric fuel. A numerical solution has been developed using a finite-element technique with the FEMLAB software package. The model has been compared to available data from several in-reactor X-2 loop experiments with defective fuel conducted at the Chalk River Laboratories. The model has also been benchmarked against an O/U profile measurement for a spent defective fuel element discharged from a commercial reactor.

CURE: Clean use of reactor energy

DOE Office of Scientific and Technical Information (OSTI.GOV)

NONE

1990-05-01

This paper presents the results of a joint Westinghouse Hanford Company (Westinghouse Hanford)-Pacific Northwest Laboratory (PNL) study that considered the feasibility of treating radioactive waste before disposal to reduce the inventory of long-lived radionuclides, making the waste more suitable for geologic disposal. The treatment considered here is one in which waste would be chemically separated so that long-lived radionuclides can be treated using specific processes appropriate for the nuclide. The technical feasibility of enhancing repository performance by this type of treatment is considered in this report. A joint Westinghouse Hanford-PNL study group developed a concept called the Clean Use ofmore » Reactor Energy (CURE), and evaluated the potential of current technology to reduce the long-lived radionuclide content in waste from the nuclear power industry. The CURE process consists of three components: chemical separation of elements that have significant quantities of long-lived radioisotopes in the waste, exposure in a neutron flux to transmute the radioisotopes to stable nuclides, and packaging of radionuclides that cannot be transmuted easily for storage or geologic disposal. 76 refs., 32 figs., 24 tabs.« less

Design of an integrated fuel processor for residential PEMFCs applications

NASA Astrophysics Data System (ADS)

Seo, Yu Taek; Seo, Dong Joo; Jeong, Jin Hyeok; Yoon, Wang Lai

KIER has been developing a novel fuel processing system to provide hydrogen rich gas to residential PEMFCs system. For the effective design of a compact hydrogen production system, each unit process for steam reforming and water gas shift, has a steam generator and internal heat exchangers which are thermally and physically integrated into a single packaged hardware system. The newly designed fuel processor (prototype II) showed a thermal efficiency of 78% as a HHV basis with methane conversion of 89%. The preferential oxidation unit with two staged cascade reactors, reduces, the CO concentration to below 10 ppm without complicated temperature control hardware, which is the prerequisite CO limit for the PEMFC stack. After we achieve the initial performance of the fuel processor, partial load operation was carried out to test the performance and reliability of the fuel processor at various loads. The stability of the fuel processor was also demonstrated for three successive days with a stable composition of product gas and thermal efficiency. The CO concentration remained below 10 ppm during the test period and confirmed the stable performance of the two-stage PrOx reactors.

Experimental and Computational Study of Multiphase Flow Hydrodynamics in 2D Trickle Bed Reactors

NASA Astrophysics Data System (ADS)

Nadeem, H.; Ben Salem, I.; Kurnia, J. C.; Rabbani, S.; Shamim, T.; Sassi, M.

2014-12-01

Trickle bed reactors are largely used in the refining processes. Co-current heavy oil and hydrogen gas flow downward on catalytic particle bed. Fine particles in the heavy oil and/or soot formed by the exothermic catalytic reactions deposit on the bed and clog the flow channels. This work is funded by the refining company of Abu Dhabi and aims at mitigating pressure buildup due to fine deposition in the TBR. In this work, we focus on meso-scale experimental and computational investigations of the interplay between flow regimes and the various parameters that affect them. A 2D experimental apparatus has been built to investigate the flow regimes with an average pore diameter close to the values encountered in trickle beds. A parametric study is done for the development of flow regimes and the transition between them when the geometry and arrangement of the particles within the porous medium are varied. Liquid and gas flow velocities have also been varied to capture the different flow regimes. Real time images of the multiphase flow are captured using a high speed camera, which were then used to characterize the transition between the different flow regimes. A diffused light source was used behind the 2D Trickle Bed Reactor to enhance visualizations. Experimental data shows very good agreement with the published literature. The computational study focuses on the hydrodynamics of multiphase flow and to identify the flow regime developed inside TBRs using the ANSYS Fluent Software package. Multiphase flow inside TBRs is investigated using the "discrete particle" approach together with Volume of Fluid (VoF) multiphase flow modeling. The effect of the bed particle diameter, spacing, and arrangement are presented that may be used to provide guidelines for designing trickle bed reactors.

3D thermal modeling of TRISO fuel coupled with neutronic simulation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hu, Jianwei; Uddin, Rizwan

2010-01-01

The Very High Temperature Gas Reactor (VHTR) is widely considered as one of the top candidates identified in the Next Generation Nuclear Power-plant (NGNP) Technology Roadmap under the U.S . Depanment of Energy's Generation IV program. TRlSO particle is a common element among different VHTR designs and its performance is critical to the safety and reliability of the whole reactor. A TRISO particle experiences complex thermo-mechanical changes during reactor operation in high temperature and high burnup conditions. TRISO fuel performance analysis requires evaluation of these changes on micro scale. Since most of these changes are temperature dependent, 3D thermal modelingmore » of TRISO fuel is a crucial step of the whole analysis package. In this paper, a 3D numerical thermal model was developed to calculate temperature distribution inside TRISO and pebble under different scenarios. 3D simulation is required because pebbles or TRISOs are always subjected to asymmetric thermal conditions since they are randomly packed together. The numerical model was developed using finite difference method and it was benchmarked against ID analytical results and also results reported from literature. Monte-Carlo models were set up to calculate radial power density profile. Complex convective boundary condition was applied on the pebble outer surface. Three reactors were simulated using this model to calculate temperature distribution under different power levels. Two asymmetric boundary conditions were applied to the pebble to test the 3D capabilities. A gas bubble was hypothesized inside the TRISO kernel and 3D simulation was also carried out under this scenario. Intuition-coherent results were obtained and reported in this paper.« less

Irradiation of Wrought FeCrAl Tubes in the High Flux Isotope Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Linton, Kory D.; Field, Kevin G.; Petrie, Christian M.

The Advanced Fuels Campaign within the Nuclear Technology Research and Development program of the Department of Energy Office of Nuclear Energy is seeking to improve the accident tolerance of light water reactors. Alumina-forming ferritic alloys (e.g., FeCrAl) are one of the leading candidate materials for fuel cladding to replace traditional zirconium alloys because of the superior oxidation resistance of FeCrAl. However, there are still some unresolved questions regarding irradiation effects on the microstructure and mechanical properties of FeCrAl at end-of-life dose levels. In particular, there are concerns related to irradiation-induced embrittlement of FeCrAl alloys due to secondary phase formation. Tomore » address this issue, Oak Ridge National Laboratory has developed a new experimental design to irradiate shortened cladding tube specimens with representative 17×17 array pressurized water reactor diameter and thickness in the High Flux Isotope Reactor (HFIR) under relevant temperatures (300–350°C). Post-irradiation examination will include studies of dimensional change, microstructural changes, and mechanical performance. This report briefly summarizes the capsule design concept and the irradiation test matrix for six rabbit capsules. Each rabbit contains two FeCrAl alloy tube specimens. The specimens include Generation I and Generation II FeCrAl alloys with varying processing conditions, Cr concentrations, and minor alloying elements. The rabbits were successfully assembled, welded, evaluated, and delivered to the HFIR along with a complete quality assurance fabrication package. Pictures of the rabbit assembly process and detailed dimensional inspection of select specimens are included in this report. The rabbits were inserted into HFIR starting in cycle 472 (May 2017).« less

Health Physics Positions Data Base: Revision 1

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kerr, G.D.; Borges, T.; Stafford, R.S.

1994-02-01

The Health Physics Positions (HPPOS) Data Base of the Nuclear Regulatory Commission (NRC) is a collection of NRC staff positions on a wide range of topics involving radiation protection (health physics). It consists of 328 documents in the form of letters, memoranda, and excerpts from technical reports. The HPPOS Data Base was developed by the NRC Headquarters and Regional Offices to help ensure uniformity in inspections, enforcement, and licensing actions. Staff members of the Oak Ridge National Laboratory (ORNL) have assisted the NRC staff in summarizing the documents during the preparation of this NUREG report. These summaries are also beingmore » made available as a {open_quotes}stand alone{close_quotes} software package for IBM and IBM-compatible personal computers. The software package for this report is called HPPOS Version 2.0. A variety of indexing schemes were used to increase the usefulness of the NUREG report and its associated software. The software package and the summaries in the report are written in the context of the {open_quotes}new{close_quotes} 10 CFR Part 20 ({section}{section}20.1001--20.2401). The purpose of this NUREG report is to allow interested individuals to familiarize themselves with the contents of the HPPOS Data Base and with the basis of many NRC decisions and regulations. The HPPOS summaries and original documents are intended to serve as a source of information for radiation protection programs at nuclear research and power reactors, nuclear medicine, and other industries that either process or use nuclear materials.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rashdan, Ahmad Al; Oxstrand, Johanna; Agarwal, Vivek

As part of the ongoing efforts at the U.S. Department of Energy’s Light Water Reactor Sustainability Program, Idaho National Laboratory is conducting several pilot projects in collaboration with the nuclear industry to improve the reliability, safety, and economics of the nuclear power industry, especially as the nuclear power plants extend their operating licenses to 80 years. One of these pilot projects is the automated work package (AWP) pilot project. An AWP is an electronic intelligent and interactive work package. It uses plant condition, resources status, and user progress to adaptively drive the work process in a manner that increases efficiencymore » while reducing human error. To achieve this mission, the AWP acquires information from various systems of a nuclear power plant’s and incorporates several advanced instrumentation and control technologies along with modern human factors techniques. With the current rapid technological advancement, it is possible to envision several available or soon-to-be-available capabilities that can play a significant role in improving the work package process. As a pilot project, the AWP project develops a prototype of an expanding set of capabilities and evaluates them in an industrial environment. While some of the proposed capabilities are based on using technological advances in other applications, others are conceptual; thus, require significant research and development to be applicable in an AWP. The scope of this paper is to introduce a set of envisioned capabilities, their need for the industry, and the industry difficulties they resolve.« less

Numerical Bifurcation Analysis of Delayed Recycle Stream in a Continuously Stirred Tank Reactor

NASA Astrophysics Data System (ADS)

Gangadhar, Nalwala Rohitbabu; Balasubramanian, Periyasamy

2010-10-01

In this paper, we present the stability analysis of delay differential equations which arise as a result of transportation lag in the CSTR-mechanical separator recycle system. A first order irreversible elementary reaction is considered to model the system and is governed by the delay differential equations. The DDE-BIFTOOL software package is used to analyze the stability of the delay system. The present analysis reveals that the system exhibits delay independent stability for isothermal operation of the CSTR. In the absence of delay, the system is dynamically unstable for non-isothermal operation of the CSTR, and as a result of delay, the system exhibits delay dependent stability.

NARMER-1: a photon point-kernel code with build-up factors

NASA Astrophysics Data System (ADS)

Visonneau, Thierry; Pangault, Laurence; Malouch, Fadhel; Malvagi, Fausto; Dolci, Florence

2017-09-01

This paper presents an overview of NARMER-1, the new generation of photon point-kernel code developed by the Reactor Studies and Applied Mathematics Unit (SERMA) at CEA Saclay Center. After a short introduction giving some history points and the current context of development of the code, the paper exposes the principles implemented in the calculation, the physical quantities computed and surveys the generic features: programming language, computer platforms, geometry package, sources description, etc. Moreover, specific and recent features are also detailed: exclusion sphere, tetrahedral meshes, parallel operations. Then some points about verification and validation are presented. Finally we present some tools that can help the user for operations like visualization and pre-treatment.

Final Report on Developing Microstructure-Property Correlation in Reactor Materials using in situ High-Energy X-rays

DOE Office of Scientific and Technical Information (OSTI.GOV)

Li, Meimei; Almer, Jonathan D.; Yang, Yong

2016-01-01

This report provides a summary of research activities on understanding microstructure – property correlation in reactor materials using in situ high-energy X-rays. The report is a Level 2 deliverable in FY16 (M2CA-13-IL-AN_-0403-0111), under the Work Package CA-13-IL-AN_- 0403-01, “Microstructure-Property Correlation in Reactor Materials using in situ High Energy Xrays”, as part of the DOE-NE NEET Program. The objective of this project is to demonstrate the application of in situ high energy X-ray measurements of nuclear reactor materials under thermal-mechanical loading, to understand their microstructure-property relationships. The gained knowledge is expected to enable accurate predictions of mechanical performance of these materialsmore » subjected to extreme environments, and to further facilitate development of advanced reactor materials. The report provides detailed description of the in situ X-ray Radiated Materials (iRadMat) apparatus designed to interface with a servo-hydraulic load frame at beamline 1-ID at the Advanced Photon Source. This new capability allows in situ studies of radioactive specimens subject to thermal-mechanical loading using a suite of high-energy X-ray scattering and imaging techniques. We conducted several case studies using the iRadMat to obtain a better understanding of deformation and fracture mechanisms of irradiated materials. In situ X-ray measurements on neutron-irradiated pure metal and model alloy and several representative reactor materials, e.g. pure Fe, Fe-9Cr model alloy, 316 SS, HT-UPS, and duplex cast austenitic stainless steels (CASS) CF-8 were performed under tensile loading at temperatures of 20-400°C in vacuum. A combination of wide-angle X-ray scattering (WAXS), small-angle X-ray scattering (SAXS), and imaging techniques were utilized to interrogate microstructure at different length scales in real time while the specimen was subject to thermal-mechanical loading. In addition, in situ X-ray studies were complemented and benchmarked by ex situ characterization using advanced electron microscopy, atom probe tomography (APT) and micro/nano-indentation. The report presented in situ tensile test results on neutron-irradiated pure Fe, Fe-9Cr model alloy, 316 SS and CASS CF-8. These in situ experiments demonstrate the broad applications of the new capability in understanding several outstanding issues related to irradiated materials.« less

ORNL Pre-test Analyses of A Large-scale Experiment in STYLE

DOE Office of Scientific and Technical Information (OSTI.GOV)

Williams, Paul T; Yin, Shengjun; Klasky, Hilda B

Oak Ridge National Laboratory (ORNL) is conducting a series of numerical analyses to simulate a large scale mock-up experiment planned within the European Network for Structural Integrity for Lifetime Management non-RPV Components (STYLE). STYLE is a European cooperative effort to assess the structural integrity of (non-reactor pressure vessel) reactor coolant pressure boundary components relevant to ageing and life-time management and to integrate the knowledge created in the project into mainstream nuclear industry assessment codes. ORNL contributes work-in-kind support to STYLE Work Package 2 (Numerical Analysis/Advanced Tools) and Work Package 3 (Engineering Assessment Methods/LBB Analyses). This paper summarizes the current statusmore » of ORNL analyses of the STYLE Mock-Up3 large-scale experiment to simulate and evaluate crack growth in a cladded ferritic pipe. The analyses are being performed in two parts. In the first part, advanced fracture mechanics models are being developed and performed to evaluate several experiment designs taking into account the capabilities of the test facility while satisfying the test objectives. Then these advanced fracture mechanics models will be utilized to simulate the crack growth in the large scale mock-up test. For the second part, the recently developed ORNL SIAM-PFM open-source, cross-platform, probabilistic computational tool will be used to generate an alternative assessment for comparison with the advanced fracture mechanics model results. The SIAM-PFM probabilistic analysis of the Mock-Up3 experiment will utilize fracture modules that are installed into a general probabilistic framework. The probabilistic results of the Mock-Up3 experiment obtained from SIAM-PFM will be compared to those results generated using the deterministic 3D nonlinear finite-element modeling approach. The objective of the probabilistic analysis is to provide uncertainty bounds that will assist in assessing the more detailed 3D finite-element solutions and to also assess the level of confidence that can be placed in the best-estimate finiteelement solutions.« less

Low rank approach to computing first and higher order derivatives using automatic differentiation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Reed, J. A.; Abdel-Khalik, H. S.; Utke, J.

2012-07-01

This manuscript outlines a new approach for increasing the efficiency of applying automatic differentiation (AD) to large scale computational models. By using the principles of the Efficient Subspace Method (ESM), low rank approximations of the derivatives for first and higher orders can be calculated using minimized computational resources. The output obtained from nuclear reactor calculations typically has a much smaller numerical rank compared to the number of inputs and outputs. This rank deficiency can be exploited to reduce the number of derivatives that need to be calculated using AD. The effective rank can be determined according to ESM by computingmore » derivatives with AD at random inputs. Reduced or pseudo variables are then defined and new derivatives are calculated with respect to the pseudo variables. Two different AD packages are used: OpenAD and Rapsodia. OpenAD is used to determine the effective rank and the subspace that contains the derivatives. Rapsodia is then used to calculate derivatives with respect to the pseudo variables for the desired order. The overall approach is applied to two simple problems and to MATWS, a safety code for sodium cooled reactors. (authors)« less

Experiment Needs and Facilities Study Appendix A Transient Reactor Test Facility (TREAT) Upgrade

DOE Office of Scientific and Technical Information (OSTI.GOV)

None

The TREAT Upgrade effort is designed to provide significant new capabilities to satisfy experiment requirements associated with key LMFBR Safety Issues. The upgrade consists of reactor-core modifications to supply the physics performance needed for the new experiments, an Advanced TREAT loop with size and thermal-hydraulics capabilities needed for the experiments, associated interface equipment for loop operations and handling, and facility modifications necessary to accommodate operations with the Loop. The costs and schedules of the tasks to be accomplished under the TREAT Upgrade project are summarized. Cost, including contingency, is about 10 million dollars (1976 dollars). A schedule for execution ofmore » 36 months has been established to provide the new capabilities in order to provide timely support of the LMFBR national effort. A key requirement for the facility modifications is that the reactor availability will not be interrupted for more than 12 weeks during the upgrade. The Advanced TREAT loop is the prototype for the STF small-bundle package loop. Modified TREAT fuel elements contain segments of graphite-matrix fuel with graded uranium loadings similar to those of STF. In addition, the TREAT upgrade provides for use of STF-like stainless steel-UO{sub 2} TREAT fuel for tests of fully enriched fuel bundles. This report will introduce the Upgrade study by presenting a brief description of the scope, performance capability, safety considerations, cost schedule, and development requirements. This work is followed by a "Design Description". Because greatly upgraded loop performance is central to the upgrade, a description is given of Advanced TREAT loop requirements prior to description of the loop concept. Performance requirements of the upgraded reactor system are given. An extensive discussion of the reactor physics calculations performed for the Upgrade concept study is provided. Adequate physics performance is essential for performance of experiments with the Advanced TREAT loop, and the stress placed on these calculations reflects this. Additional material on performance and safety is provided. Backup calculations on calculations of plutonium-release limits are described. Cost and schedule information for the Upgrade are presented.« less

Efficiencies and Optimization of Weak Base Anion Ion-Exchange Resin for Groundwater Hexavalent Chromium Removal at Hanford

DOE Office of Scientific and Technical Information (OSTI.GOV)

Nesham, Dean O.; Ivarson, Kristine A.; Hanson, James P.

2014-02-03

The U.S. Department of Energy’s (DOE’s) contractor, CH2M HILL Plateau Remediation Company, has successfully converted a series of groundwater treatment facilities to use a new treatment resin that is delivering more than $3 million in annual cost savings and efficiency in treating groundwater contamination at the DOE Hanford Site in southeastern Washington State. During the production era, the nuclear reactors at the Hanford Site required a continuous supply of high-quality cooling water during operations. Cooling water consumption ranged from about 151,417 to 378,541 L/min (40,000 to 100,000 gal/min) per reactor, depending on specific operating conditions. Water from the Columbia Rivermore » was filtered and treated chemically prior to use as cooling water, including the addition of sodium dichromate as a corrosion inhibitor. Hexavalent chromium was the primary component of the sodium dichromate and was introduced into the groundwater at the Hanford Site as a result of planned and unplanned discharges from the reactors starting in 1944. Groundwater contamination by hexavalent chromium and other contaminants related to nuclear reactor operations resulted in the need for groundwater remedial actions within the Hanford Site reactor areas. Beginning in 1995, groundwater treatment methods were evaluated, leading to the use of pump-and-treat facilities with ion exchange using Dowex™ 21K, a regenerable, strong-base anion exchange resin. This required regeneration of the resin, which was performed offsite. In 2008, DOE recognized that regulatory agreements would require significant expansion for the groundwater chromium treatment capacity. As a result, CH2M HILL performed testing at the Hanford Site in 2009 and 2010 to demonstrate resin performance in the specific groundwater chemistry at different waste sites. The testing demonstrated that a weak-base anion, single-use resin, specifically ResinTech SIR-700 ®, was effective at removing chromium, had a significantly higher capacity, could be disposed of efficiently onsite, and would eliminate the complexities and programmatic risks from sampling, packaging, transportation, and return of resin for regeneration.« less

UP2 400 High Activity Oxide Legacy Waste Retrieval Project Scope and Progress-13048

DOE Office of Scientific and Technical Information (OSTI.GOV)

Chabeuf, Jean-Michel; Varet, Thierry

The High Activity Oxide facility (HAO) reprocessed sheared and dissolved 4500 metric tons of light water reactor fuel the fuel of the emerging light water reactor spent fuel between 1976 and 1998. Over the period, approximately 2200 tons of process waste, composed primarily of sheared hulls, was produced and stored in a vast silo in the first place, and in canisters stored in pools in subsequent years. Upon shutdown of the facility, AREVA D and D Division in La Hague launched a thorough investigation and characterization of the silos and pools content, which then served as input data for themore » definition of a legacy waste retrieval and reconditioning program. Basic design was conducted between 2005 and 2007, and was followed by an optimization phase which lead to the definition of a final scenario and budget, 12% under the initial estimates. The scenario planned for the construction of a retrieval and reconditioning cell to be built on top of the storage silo. The retrieved waste would then be rinsed and sorted, so that hulls could subsequently be sent to La Hague high activity compacting facility, while resins and sludge would be cemented within the retrieval cell. Detailed design was conducted successfully from 2008 until 2011, while a thorough research and development program was conducted in order to qualify each stage of the retrieval and reconditioning process, and assist in the elaboration of the final waste package specification. This R and D program was defined and conducted as a response and mitigation of the major project risks identified during the basic design process. Procurement and site preparatory works were then launched in 2011. By the end of 2012, R and D is nearly completed, the retrieval and reconditioning process have been secured, the final waste package specification is being completed, the first equipment for the retrieval cell is being delivered on site, while preparation works are allowing to free up space above and around the silo, to allow for construction which is scheduled to being during the first semester of 2013. The elaboration of the final waste package is still undergoing and expected to be completed by then end of 2013, following some final elements of R and D required to demonstrate the full compatibility of the package with deep geological repository. The HAO legacy waste retrieval project is so far the largest such project entering operational phase on the site of La Hague. It is on schedule, under budget, and in conformity with the delivery requirements set by the French Safety Authority, as well as other stakeholders. This project paves the way for the successful completion of AREVA La Hague other legacy waste retrieval projects, which are currently being drafted or already in active R and D phase. (authors)« less

Radiation hardness of Efratom M-100 rubidium frequency standard

NASA Technical Reports Server (NTRS)

English, T. C.; Vorwerk, H.; Rudie, N. J.

1983-01-01

The effects of nuclear radiation on rubidium gas cell frequency standards and components are presented, including the results of recent tests where a continuously operating rubidium frequency standard (Effratom, Model M-100) was subjected to simultaneous neutron/gamma radiation. At the highest neutron fluence 7.5 10 to the 12th power n/sq cm and total dose 11 krad(Si) tested, the unit operated satisfactorily; the total frequency change over the 2 1/2 hour test period due to all causes, including repeated retraction from and insertion into the reactor, was less than 1 x 10 to the -10th power. The effects of combined neutron/gamma radiation on rubidium frequency standard physics package components were also studied, and the results are presented.

Radiological controls integrated into design

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kindred, G.W.

1995-03-01

Radiological controls are required by law in the design of commercial nuclear power reactor facilities. These controls can be relatively minor or significant, relative to cost. To ensure that radiological controls are designed into a project, the health physicist (radiological engineer) must be involved from the beginning. This is especially true regarding keeping costs down. For every radiological engineer at a nuclear power plant there must be fifty engineers of other disciplines. The radiological engineer cannot be an expert on every discipline of engineering. However, he must be knowledgeable to the degree of how a design will impact the facilitymore » from a radiological perspective. This paper will address how to effectively perform radiological analyses with the goal of radiological controls integrated into the design package.« less

Safety Criticality Standards Using the French CRISTAL Code Package: Application to the AREVA NP UO{sub 2} Fuel Fabrication Plant

DOE Office of Scientific and Technical Information (OSTI.GOV)

Doucet, M.; Durant Terrasson, L.; Mouton, J.

2006-07-01

Criticality safety evaluations implement requirements to proof of sufficient sub critical margins outside of the reactor environment for example in fuel fabrication plants. Basic criticality data (i.e., criticality standards) are used in the determination of sub critical margins for all processes involving plutonium or enriched uranium. There are several criticality international standards, e.g., ARH-600, which is one the US nuclear industry relies on. The French Nuclear Safety Authority (DGSNR and its advising body IRSN) has requested AREVA NP to review the criticality standards used for the evaluation of its Low Enriched Uranium fuel fabrication plants with CRISTAL V0, the recentlymore » updated French criticality evaluation package. Criticality safety is a concern for every phase of the fabrication process including UF{sub 6} cylinder storage, UF{sub 6}-UO{sub 2} conversion, powder storage, pelletizing, rod loading, assembly fabrication, and assembly transportation. Until 2003, the accepted criticality standards were based on the French CEA work performed in the late seventies with the APOLLO1 cell/assembly computer code. APOLLO1 is a spectral code, used for evaluating the basic characteristics of fuel assemblies for reactor physics applications, which has been enhanced to perform criticality safety calculations. Throughout the years, CRISTAL, starting with APOLLO1 and MORET 3 (a 3D Monte Carlo code), has been improved to account for the growth of its qualification database and for increasing user requirements. Today, CRISTAL V0 is an up-to-date computational tool incorporating a modern basic microscopic cross section set based on JEF2.2 and the comprehensive APOLLO2 and MORET 4 codes. APOLLO2 is well suited for criticality standards calculations as it includes a sophisticated self shielding approach, a P{sub ij} flux determination, and a 1D transport (S{sub n}) process. CRISTAL V0 is the result of more than five years of development work focusing on theoretical approaches and the implementation of user-friendly graphical interfaces. Due to its comprehensive physical simulation and thanks to its broad qualification database with more than a thousand benchmark/calculation comparisons, CRISTAL V0 provides outstanding and reliable accuracy for criticality evaluations for configurations covering the entire fuel cycle (i.e. from enrichment, pellet/assembly fabrication, transportation, to fuel reprocessing). After a brief description of the calculation scheme and the physics algorithms used in this code package, results for the various fissile media encountered in a UO{sub 2} fuel fabrication plant will be detailed and discussed. (authors)« less

Benchmarking of Neutron Flux Parameters at the USGS TRIGA Reactor in Lakewood, Colorado

NASA Astrophysics Data System (ADS)

Alzaabi, Osama E.

The USGS TRIGA Reactor (GSTR) located at the Denver Federal Center in Lakewood Colorado provides opportunities to Colorado School of Mines students to do experimental research in the field of neutron activation analysis. The scope of this thesis is to obtain precise knowledge of neutron flux parameters at the GSTR. The Colorado School of Mines Nuclear Physics group intends to develop several research projects at the GSTR, which requires the precise knowledge of neutron fluxes and energy distributions in several irradiation locations. The fuel burn-up of the new GSTR fuel configuration and the thermal neutron flux of the core were recalculated since the GSTR core configuration had been changed with the addition of two new fuel elements. Therefore, a MCNP software package was used to incorporate the burn up of reactor fuel and to determine the neutron flux at different irradiation locations and at flux monitoring bores. These simulation results were compared with neutron activation analysis results using activated diluted gold wires. A well calibrated and stable germanium detector setup as well as fourteen samplers were designed and built to achieve accuracy in the measurement of the neutron flux. Furthermore, the flux monitoring bores of the GSTR core were used for the first time to measure neutron flux experimentally and to compare to MCNP simulation. In addition, International Atomic Energy Agency (IAEA) standard materials were used along with USGS national standard materials in a previously well calibrated irradiation location to benchmark simulation, germanium detector calibration and sample measurements to international standards.

Activation calculation for the dismantling and decommissioning of a light water reactor using MCNP™ with ADVANTG and ORIGEN-S

NASA Astrophysics Data System (ADS)

Schlömer, Luc; Phlippen, Peter-W.; Lukas, Bernard

2017-09-01

The decommissioning of a light water reactor (LWR), which is licensed under § 7 of the German Atomic Energy Act, following the post-operational phase requires a comprehensive licensing procedure including in particular radiation protection aspects and possible impacts to the environment. Decommissioning includes essential changes in requirements for the systems and components and will mainly lead to the direct dismantling. In this context, neutron induced activation calculations for the structural components have to be carried out to predict activities in structures and to estimate future costs for conditioning and packaging. To avoid an overestimation of the radioactive inventory and to calculate the expenses for decommissioning as accurate as possible, modern state-of-the-art Monte-Carlo-Techniques (MCNP™) are applied and coupled with present-day activation and decay codes (ORIGEN-S). In this context ADVANTG is used as weight window generator for MCNP™ i. e. as variance reduction tool to speed up the calculation in deep penetration problems. In this paper the calculation procedure is described and the obtained results are presented with a validation along with measured activities and photon dose rates measured in the post-operational phase. The validation shows that the applied calculation procedure is suitable for the determination of the radioactive inventory of a nuclear power plant. Even the measured gamma dose rates in the post-operational phase at different positions in the reactor building agree within a factor of 2 to 3 with the calculation results. The obtained results are accurate and suitable to support effectively the decommissioning planning process.

Trash to Gas: Converting Space Waste into Useful Supply Products

NASA Technical Reports Server (NTRS)

Tsoras, Alexandra

2013-01-01

The cost of sending mass into space with current propulsion technology is very expensive, making every item a crucial element of the space mission. It is essential that all materials be used to their fullest potential. Items like food, packaging, clothing, paper towels, gloves, etc., normally become trash and take up space after use. These waste materials are currently either burned up upon reentry in earth's atmosphere or sent on cargo return vehicles back to earth: a very wasteful method. The purpose of this project was to utilize these materials and create useful products like water and methane gas, which is used for rocket fuel, to further supply a deep space mission. The system used was a thermal degradation reactor with the configuration of a down-draft gasifier. The reactor was loaded with approximately 100g of trash simulant and heated with two external ceramic heaters with separate temperature control in order to create pyrolysis and gasification in one zone and incineration iri a second zone simultaneously. Trash was loaded into the top half of the reactor to undergo pyrolysis while the downdraft gas experienced gasification or incineration to treat tars and maximize the production of carbon dioxide. Minor products included carbon monoxide, methane, and other hydrocarbons. The carbon dioxide produced can be sent to a Sabatier reactor to convert the gas into methane, which can be used as rocket propellant. In order to maximize the carbon dioxide and useful gases produced, and minimize the unwanted tars and leftover ashen material, multiple experiments were performed with altered parameters such as differing temperatures, flow rates, and location of inlet air flow. According to the data received from these experiments, the process will be further scaled up and optimized to ultimately create a system that reduces trash buildup while at the same time providing enough useful gases to potentially fill a methane tank that could fuel a lunar ascent vehicle or other deep space mission.

Evaluation and Parameter Analysis of Burn up Calculations for the Assessment of Radioactive Waste - 13187

DOE Office of Scientific and Technical Information (OSTI.GOV)

Fast, Ivan; Aksyutina, Yuliya; Tietze-Jaensch, Holger

2013-07-01

Burn up calculations facilitate a determination of the composition and nuclear inventory of spent nuclear fuel, if operational history is known. In case this information is not available, the total nuclear inventory can be determined by means of destructive or, even on industrial scale, nondestructive measurement methods. For non-destructive measurements however only a few easy-to-measure, so-called key nuclides, are determined due to their characteristic gamma lines or neutron emission. From these measured activities the fuel burn up and cooling time are derived to facilitate the numerical inventory determination of spent fuel elements. Most regulatory bodies require an independent assessment ofmore » nuclear waste properties and their documentation. Prominent part of this assessment is a consistency check of inventory declaration. The waste packages often contain wastes from different types of spent fuels of different history and information about the secondary reactor parameters may not be available. In this case the so-called characteristic fuel burn up and cooling time are determined. These values are obtained from a correlations involving key-nuclides with a certain bandwidth, thus with upper and lower limits. The bandwidth is strongly dependent on secondary reactor parameter such as initial enrichment, temperature and density of the fuel and moderator, hence the reactor type, fuel element geometry and plant operation history. The purpose of our investigation is to look into the scaling and correlation limitations, to define and verify the range of validity and to scrutinize the dependencies and propagation of uncertainties that affect the waste inventory declarations and their independent verification. This is accomplished by numerical assessment and simulation of waste production using well accepted codes SCALE 6.0 and 6.1 to simulate the cooling time and burn up of a spent fuel element. The simulations are benchmarked against spent fuel from the real reactor Obrigheim in Germany for which sufficiently precise experimental reference data are available. (authors)« less

Geochemical Data Package for Performance Assessment Calculations Related to the Savannah River Site

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kaplan, Daniel I.

The Savannah River Site (SRS) disposes of low-level radioactive waste (LLW) and stabilizes high-level radioactive waste (HLW) tanks in the subsurface environment. Calculations used to establish the radiological limits of these facilities are referred to as Performance Assessments (PA), Special Analyses (SA), and Composite Analyses (CA). The objective of this document is to revise existing geochemical input values used for these calculations. This work builds on earlier compilations of geochemical data (2007, 2010), referred to a geochemical data packages. This work is being conducted as part of the on-going maintenance program of the SRS PA programs that periodically updates calculationsmore » and data packages when new information becomes available. Because application of values without full understanding of their original purpose may lead to misuse, this document also provides the geochemical conceptual model, the approach used for selecting the values, the justification for selecting data, and the assumptions made to assure that the conceptual and numerical geochemical models are reasonably conservative (i.e., bias the recommended input values to reflect conditions that will tend to predict the maximum risk to the hypothetical recipient). This document provides 1088 input parameters for geochemical parameters describing transport processes for 64 elements (>740 radioisotopes) potentially occurring within eight subsurface disposal or tank closure areas: Slit Trenches (ST), Engineered Trenches (ET), Low Activity Waste Vault (LAWV), Intermediate Level (ILV) Vaults, Naval Reactor Component Disposal Areas (NRCDA), Components-in-Grout (CIG) Trenches, Saltstone Facility, and Closed Liquid Waste Tanks. The geochemical parameters described here are the distribution coefficient, Kd value, apparent solubility concentration, k s value, and the cementitious leachate impact factor.« less

Uncertainty in the delayed neutron fraction in fuel assembly depletion calculations

NASA Astrophysics Data System (ADS)

Aures, Alexander; Bostelmann, Friederike; Kodeli, Ivan A.; Velkov, Kiril; Zwermann, Winfried

2017-09-01

This study presents uncertainty and sensitivity analyses of the delayed neutron fraction of light water reactor and sodium-cooled fast reactor fuel assemblies. For these analyses, the sampling-based XSUSA methodology is used to propagate cross section uncertainties in neutron transport and depletion calculations. Cross section data is varied according to the SCALE 6.1 covariance library. Since this library includes nu-bar uncertainties only for the total values, it has been supplemented by delayed nu-bar uncertainties from the covariance data of the JENDL-4.0 nuclear data library. The neutron transport and depletion calculations are performed with the TRITON/NEWT sequence of the SCALE 6.1 package. The evolution of the delayed neutron fraction uncertainty over burn-up is analysed without and with the consideration of delayed nu-bar uncertainties. Moreover, the main contributors to the result uncertainty are determined. In all cases, the delayed nu-bar uncertainties increase the delayed neutron fraction uncertainty. Depending on the fuel composition, the delayed nu-bar values of uranium and plutonium in fact give the main contributions to the delayed neutron fraction uncertainty for the LWR fuel assemblies. For the SFR case, the uncertainty of the scattering cross section of U-238 is the main contributor.

File-Based One-Way BISON Coupling Through VERA: User's Manual

DOE Office of Scientific and Technical Information (OSTI.GOV)

Stimpson, Shane G.

Activities to incorporate fuel performance capabilities into the Virtual Environment for Reactor Applications (VERA) are receiving increasing attention [1–6]. The multiphysics emphasis is expanding as the neutronics (MPACT) and thermal-hydraulics (CTF) packages are becoming more mature. Capturing the finer details of fuel phenomena (swelling, densification, relocation, gap closure, etc.) is the natural next step in the VERA development process since these phenomena are currently not directly taken into account. While several codes could be used to accomplish this, the BISON fuel performance code [8,9] being developed by the Idaho National Laboratory (INL) is the focus of ongoing work in themore » Consortium for Advanced Simulation of Light Water Reactors (CASL). Built on INL’s MOOSE framework [10], BISON uses the finite element method for geometric representation and a Jacobian-free Newton-Krylov (JFNK) scheme to solve systems of partial differential equations for various fuel characteristic relationships. There are several modes of operation in BISON, but, this work uses a 2D azimuthally symmetric (R-Z) smeared-pellet model. This manual is intended to cover (1) the procedure pertaining to the standalone BISON one-way coupling from VERA and (2) the procedure to generate BISON fuel temperature tables that VERA can use.« less

VALVES FOR THE HIGH PRESSURE-HIGH TEMPERATURE (HP-HT) FLUORINATION SYSTEM. (Engineering Materials)

DOE Office of Scientific and Technical Information (OSTI.GOV)

None

1963-10-31

This package contains two drawings of valves which eliminate errors in the gravimetric oxide dilution procedure of U/sup 235/ measurement. Isotopic contaminatioNonen in the high pressure fluorination reactor was corrected by changing the manner in which the Cu tubing joins the valve and by modification of the bellows. The compact inlet system was modified to improve the precision of the spectrometer analyses. Changes were raade in the basic leak and the air operator, which is a diaphragm-type valve, so that the setting of the flow level is controlled by the closure spring adjustment screw. This capillary-type leak has increased controlmore » range and sraooth control characteristics. It is simple to construct, is remotely operated and is free from corrosion failure. (F.S.)« less

Remaining Sites Verification Package for the 100-F-26:12, 1.8-m (72-in.) Main Process Sewer Pipeline, Waste Site Reclassification Form 2007-034

DOE Office of Scientific and Technical Information (OSTI.GOV)

J. M. Capron

2008-04-29

The 100-F-26:12 waste site was an approximately 308-m-long, 1.8-m-diameter east-west-trending reinforced concrete pipe that joined the North Process Sewer Pipelines (100-F-26:1) and the South Process Pipelines (100-F-26:4) with the 1.8-m reactor cooling water effluent pipeline (100-F-19). In accordance with this evaluation, the verification sampling results support a reclassification of this site to Interim Closed Out. The results of verification sampling show that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River.

A chemical and fluid dynamic study of the chemical vapor deposition of aluminum nitride in a vertical reactor

NASA Astrophysics Data System (ADS)

Bather, Wayne Anthony

The metalorganic chemical vapor deposition (MOCVD) growth of compound semiconductors has become important in producing many high performance electronic and optoelectronic devices from the wide bandgaps III-V nitrides, for example, aluminum nitride (AlN). A systematic theoretical and experimental investigation of the chemistry and mass transport process in a MOCVD system can yield predictive models of the deposition process. The chemistries and fluid dynamics of the MOCVD growth of AlN in a vertical reactor is analyzed and characterized in order to parameterize and model the deposition process. A Fourier Transform Infrared (FTIR) spectroscopic study of the predeposition reactions between trimethylaluminum (TMAl) and ammonia (NHsb3) is carried out in a static gas cell to examine the primary homogeneous gas phase reactions, pyrolysis of the reactants, and adduct formation, possibly accompanied by elimination reactions. A series of reactions, based on laboratory studies and literature review, is then proposed to model the deposition process. All pertinent kinetic, thermochemical, and transport properties were obtained. Utilizing a mass transport model, we performed computational fluid dynamics calculations using the FLUENT software package. We determined temperature, velocity, and concentration profiles, along with deposition rates inside the experimental vertical CVD reactor in the Howard University Material Science Research Center of Excellence. Experimental deposition rate data were found to be in good agreement with those predicted from the simulations, thus validating the proposed model. The control of the homogeneous gas phase reaction leading to the formation and subsequent decomposition of the adduct is critical to the formation of device-grade AlN films. Many basic processes occurring during MOCVD of AlN are still not completely understood, and none of the detailed surface reaction mechanisms are known.

Digital Signal Processing Methods for Safety Systems Employed in Nuclear Power Industry

NASA Astrophysics Data System (ADS)

Popescu, George

Some of the major safety concerns in the nuclear power industry focus on the readiness of nuclear power plant safety systems to respond to an abnormal event, the security of special nuclear materials in used nuclear fuels, and the need for physical security to protect personnel and reactor safety systems from an act of terror. Routine maintenance and tests of all nuclear reactor safety systems are performed on a regular basis to confirm the ability of these systems to operate as expected. However, these tests do not determine the reliability of these safety systems and whether the systems will perform for the duration of an accident and whether they will perform their tasks without failure after being engaged. This research has investigated the progression of spindle asynchronous error motion determined from spindle accelerations to predict bearings failure onset. This method could be applied to coolant pumps that are essential components of emergency core cooling systems at all nuclear power plants. Recent security upgrades mandated by the Nuclear Regulatory Commission and the Department of Homeland Security have resulted in implementation of multiple physical security barriers around all of the commercial and research nuclear reactors in the United States. A second part of this research attempts to address an increased concern about illegal trafficking of Special Nuclear Materials (SNM). This research describes a multi element scintillation detector system designed for non - invasive (passive) gamma ray surveillance for concealed SNM that may be within an area or sealed in a package, vehicle or shipping container. Detection capabilities of the system were greatly enhanced through digital signal processing, which allows the combination of two very powerful techniques: 1) Compton Suppression (CS) and 2) Pulse Shape Discrimination (PSD) with less reliance on complicated analog instrumentation.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kreitman, Paul J.; Sirianni, Steve R.; Pillard, Mark M.

Entergy recently performed an Extended Power Up-rate (EPU) on their Grand Gulf Nuclear Station, near Port Gibson, Mississippi. To support the EPU, a new Steam Dryer Assembly was installed during the last refueling outage. Due to limited access into the containment, the large Replacement Steam Dryer (RSD) had to be brought into the containment in pieces and then final assembly was completed on the refueling floor before installation into the reactor. Likewise, the highly contaminated Original Steam Dryer (OSD) had to be segmented into manageable sections, loaded into specially designed shielded containers, and rigged out of containment where they willmore » be safely stored until final disposal is accomplished at an acceptable waste repository. Westinghouse Nuclear Services was contracted by Entergy to segment, package and remove the OSD from containment. This work was performed on critical path during the most recent refueling outage. The segmentation was performed underwater to minimize radiation exposure to the workers. Special hydraulic saws were developed for the cutting operations based on Westinghouse designs previously used in Sweden to segment ABB Reactor Internals. The mechanical cutting method was selected because of its proven reliability and the minimal cutting debris that is generated by the process. Maintaining stability of the large OSD sections during cutting was accomplished using a custom built support stand that was installed into the Moisture Separator Pool after the Moisture Separator was installed back in the reactor vessel. The OSD was then moved from the Steam Dryer Pool to the Moisture Separator Pool for segmentation. This scenario resolved the logistical challenge of having two steam dryers and a moisture separator in containment simultaneously. A water filtration/vacuum unit was supplied to maintain water clarity during the cutting and handling operations and to collect the cutting chips. (authors)« less

Code qualification of structural materials for AFCI advanced recycling reactors.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Natesan, K.; Li, M.; Majumdar, S.

2012-05-31

This report summarizes the further findings from the assessments of current status and future needs in code qualification and licensing of reference structural materials and new advanced alloys for advanced recycling reactors (ARRs) in support of Advanced Fuel Cycle Initiative (AFCI). The work is a combined effort between Argonne National Laboratory (ANL) and Oak Ridge National Laboratory (ORNL) with ANL as the technical lead, as part of Advanced Structural Materials Program for AFCI Reactor Campaign. The report is the second deliverable in FY08 (M505011401) under the work package 'Advanced Materials Code Qualification'. The overall objective of the Advanced Materials Codemore » Qualification project is to evaluate key requirements for the ASME Code qualification and the Nuclear Regulatory Commission (NRC) approval of structural materials in support of the design and licensing of the ARR. Advanced materials are a critical element in the development of sodium reactor technologies. Enhanced materials performance not only improves safety margins and provides design flexibility, but also is essential for the economics of future advanced sodium reactors. Code qualification and licensing of advanced materials are prominent needs for developing and implementing advanced sodium reactor technologies. Nuclear structural component design in the U.S. must comply with the ASME Boiler and Pressure Vessel Code Section III (Rules for Construction of Nuclear Facility Components) and the NRC grants the operational license. As the ARR will operate at higher temperatures than the current light water reactors (LWRs), the design of elevated-temperature components must comply with ASME Subsection NH (Class 1 Components in Elevated Temperature Service). However, the NRC has not approved the use of Subsection NH for reactor components, and this puts additional burdens on materials qualification of the ARR. In the past licensing review for the Clinch River Breeder Reactor Project (CRBRP) and the Power Reactor Innovative Small Module (PRISM), the NRC/Advisory Committee on Reactor Safeguards (ACRS) raised numerous safety-related issues regarding elevated-temperature structural integrity criteria. Most of these issues remained unresolved today. These critical licensing reviews provide a basis for the evaluation of underlying technical issues for future advanced sodium-cooled reactors. Major materials performance issues and high temperature design methodology issues pertinent to the ARR are addressed in the report. The report is organized as follows: the ARR reference design concepts proposed by the Argonne National Laboratory and four industrial consortia were reviewed first, followed by a summary of the major code qualification and licensing issues for the ARR structural materials. The available database is presented for the ASME Code-qualified structural alloys (e.g. 304, 316 stainless steels, 2.25Cr-1Mo, and mod.9Cr-1Mo), including physical properties, tensile properties, impact properties and fracture toughness, creep, fatigue, creep-fatigue interaction, microstructural stability during long-term thermal aging, material degradation in sodium environments and effects of neutron irradiation for both base metals and weld metals. An assessment of modified versions of Type 316 SS, i.e. Type 316LN and its Japanese version, 316FR, was conducted to provide a perspective for codification of 316LN or 316FR in Subsection NH. Current status and data availability of four new advanced alloys, i.e. NF616, NF616+TMT, NF709, and HT-UPS, are also addressed to identify the R&D needs for their code qualification for ARR applications. For both conventional and new alloys, issues related to high temperature design methodology are described to address the needs for improvements for the ARR design and licensing. Assessments have shown that there are significant data gaps for the full qualification and licensing of the ARR structural materials. Development and evaluation of structural materials require a variety of experimental facilities that have been seriously degraded in the past. The availability and additional needs for the key experimental facilities are summarized at the end of the report. Detailed information covered in each Chapter is given.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Marques, J.G.; Ramos, A.R.; Fernandes, A.C.

The behavior of electronic components and circuits under radiation is a concern shared by the nuclear industry, the space community and the high-energy physics community. Standard commercial components are used as much as possible instead of radiation hard components, since they are easier to obtain and allow a significant reduction of costs. However, these standard components need to be tested in order to determine their radiation tolerance. The Portuguese Research Reactor (RPI) is a 1 MW pool-type reactor, operating since 1961. The irradiation of electronic components and circuits is one area where a 1 MW reactor can be competitive, sincemore » the fast neutron fluences required for testing are in most cases well below 10{sup 16} n/cm{sup 2}. A program was started in 1999 to test electronics components and circuits for the LHC facility at CERN, initially using a dedicated in-pool irradiation device and later a beam line with tailored neutron and gamma filters. Neutron filters are essential to reduce the intensity of the thermal neutron flux, which does not produce significant defects in electronic components but produces unwanted radiation from activation of contacts and packages of integrated circuits and also of the printed circuit boards. In irradiations performed within the line-of-sight of the core of a fission reactor there is simultaneous gamma radiation which complicates testing in some cases. Filters can be used to reduce its importance and separate testing with a pure gamma radiation source can contribute to clarify some irradiation results. Practice has shown the need to introduce several improvements to the procedures and facilities over the years. We will review improvements done in the following areas: - Optimization of neutron and gamma filters; - Dosimetry procedures in mixed neutron / gamma fields; - Determination of hardness parameter and 1 MeV-equivalent neutron fluence; - Temperature measurement and control during irradiation; - Follow-up of reactor power operational fluctuations; - Study of gamma radiation effects only. The fission neutron spectrum can be limitative for some of the tests, as most neutrons are in the 1-2 MeV energy range. Significant progress has been made lately in compact neutron generators using D-D and D-T fusion reactions, achieving higher neutron fluxes and longer lifetime than previously available. The advantages of using compact neutron generators for testing of electronic components and circuits will be also discussed. (authors)« less

Advanced Test Reactor Core Modeling Update Project Annual Report for Fiscal Year 2012

DOE Office of Scientific and Technical Information (OSTI.GOV)

David W. Nigg, Principal Investigator; Kevin A. Steuhm, Project Manager

Legacy computational reactor physics software tools and protocols currently used for support of Advanced Test Reactor (ATR) core fuel management and safety assurance, and to some extent, experiment management, are inconsistent with the state of modern nuclear engineering practice, and are difficult, if not impossible, to properly verify and validate (V&V) according to modern standards. Furthermore, the legacy staff knowledge required for application of these tools and protocols from the 1960s and 1970s is rapidly being lost due to staff turnover and retirements. In late 2009, the Idaho National Laboratory (INL) initiated a focused effort, the ATR Core Modeling Updatemore » Project, to address this situation through the introduction of modern high-fidelity computational software and protocols. This aggressive computational and experimental campaign will have a broad strategic impact on the operation of the ATR, both in terms of improved computational efficiency and accuracy for support of ongoing DOE programs as well as in terms of national and international recognition of the ATR National Scientific User Facility (NSUF). The ATR Core Modeling Update Project, targeted for full implementation in phase with the next anticipated ATR Core Internals Changeout (CIC) in the 2014-2015 time frame, began during the last quarter of Fiscal Year 2009, and has just completed its third full year. Key accomplishments so far have encompassed both computational as well as experimental work. A new suite of stochastic and deterministic transport theory based reactor physics codes and their supporting nuclear data libraries (HELIOS, KENO6/SCALE, NEWT/SCALE, ATTILA, and an extended implementation of MCNP5) has been installed at the INL under various licensing arrangements. Corresponding models of the ATR and ATRC are now operational with all five codes, demonstrating the basic feasibility of the new code packages for their intended purpose. Of particular importance, a set of as-run core depletion HELIOS calculations for all ATR cycles since August 2009, Cycle 145A through Cycle 151B, was successfully completed during 2012. This major effort supported a decision late in the year to proceed with the phased incorporation of the HELIOS methodology into the ATR Core Safety Analysis Package (CSAP) preparation process, in parallel with the established PDQ-based methodology, beginning late in Fiscal Year 2012. Acquisition of the advanced SERPENT (VTT-Finland) and MC21 (DOE-NR) Monte Carlo stochastic neutronics simulation codes was also initiated during the year and some initial applications of SERPENT to ATRC experiment analysis were demonstrated. These two new codes will offer significant additional capability, including the possibility of full-3D Monte Carlo fuel management support capabilities for the ATR at some point in the future. Finally, a capability for rigorous sensitivity analysis and uncertainty quantification based on the TSUNAMI system has been implemented and initial computational results have been obtained. This capability will have many applications as a tool for understanding the margins of uncertainty in the new models as well as for validation experiment design and interpretation.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Vogel, Sven C.; Losko, Adrian Simon; Pokharel, Reeju

The goal of the Advanced Non-destructive Fuel Examination (ANDE) work package is the development and application of non-destructive neutron imaging and scattering techniques to ceramic and metallic nuclear fuels, ultimately also to irradiated fuels. The results of these characterizations provide complete pre- and post-irradiation on length scales ranging from mm to nm, guide destructive examination, and inform modelling efforts. Besides technique development and application to samples to be irradiated, the ANDE work package also examines possible technologies to provide these characterization techniques pool-side, e.g. at the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) using laser-driven intense pulsed neutronmore » and gamma sources. Neutron tomography and neutron diffraction characterizations were performed on nine pellets; four UN/ U-Si composite formulations (two enrichment levels), three pure U 3Si 5 reference formulations (two enrichment levels), and two reject pellets with visible flaws (to qualify the technique). The 235U enrichments ranged from 0.2 to 8.8 wt. %. The nitride/silicide composites are candidate compositions for use as Accident Tolerant Fuel (ATF). The monophase U 3Si 5 material was included as a reference. Pellets from the same fabrication batches will be inserted in the Advanced Test Reactor at Idaho during 2016. We have also proposed a data format to build a database for characterization results of individual pellets. Neutron data reported in this report were collected in the LANSCE run cycle that started in September 2015 and ended in March 2016. This report provides the results for the characterized samples and discussion in the context of ANDE and APIE. We quantified the gamma spectra of several samples in their received state as well as after neutron irradiation to ensure that the neutron irradiation does not add significant activation that would complicate shipment and handling. We demonstrated synchrotron-based 3D X-ray microscopy on the composite fuel materials, providing unparalleled level of detail on the 3D microstructure. Furthermore, we initiated development of shielding containers allowing the characterizations presented herein while allowing handling of irradiated samples.« less

NEET Micro-Pocket Fission Detector. Final Project report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Unruh, T.; Rempe, Joy; McGregor, Douglas

2014-09-01

A collaboration between the Idaho National Laboratory (INL), the Kansas State University (KSU), and the French Alternative Energies and Atomic Energy Commission, Commissariat à l'Énergie Atomique et aux Energies Alternatives, (CEA), is funded by the Nuclear Energy Enabling Technologies (NEET) program to develop and test Micro-Pocket Fission Detectors (MPFDs), which are compact fission chambers capable of simultaneously measuring thermal neutron flux, fast neutron flux and temperature within a single package. When deployed, these sensors will significantly advance flux detection capabilities for irradiation tests in US Material Test Reactors (MTRs). Ultimately, evaluations may lead to a more compact, more accurate, andmore » longer lifetime flux sensor for critical mock-ups, and high performance reactors, allowing several Department of Energy Office of Nuclear Energy (DOE-NE) programs to obtain higher accuracy/higher resolution data from irradiation tests of candidate new fuels and materials. Specifically, deployment of MPFDs will address several challenges faced in irradiations performed at MTRs: Current fission chamber technologies do not offer the ability to measure fast flux, thermal flux and temperature within a single compact probe; MPFDs offer this option. MPFD construction is very different than current fission chamber construction; the use of high temperature materials allow MPFDs to be specifically tailored to survive harsh conditions encountered in-core of high performance MTRs. The higher accuracy, high fidelity data available from the compact MPFD will significantly enhance efforts to validate new high-fidelity reactor physics codes and new multi-scale, multi-physics codes. MPFDs can be built with variable sensitivities to survive the lifetime of an experiment or fuel assembly in some MTRs, allowing for more efficient and cost effective power monitoring. The small size of the MPFDs allows multiple sensors to be deployed, offering the potential to accurately measure the flux and temperature profiles in the reactor. This report summarizes the status at the end of year two of this three year project. As documented in this report, all planned accomplishments for developing this unique new, compact, multipurpose sensor have been completed.« less

Fundamentals, current state of the development of, and prospects for further improvement of the new-generation thermal-hydraulic computational HYDRA-IBRAE/LM code for simulation of fast reactor systems

NASA Astrophysics Data System (ADS)

Alipchenkov, V. M.; Anfimov, A. M.; Afremov, D. A.; Gorbunov, V. S.; Zeigarnik, Yu. A.; Kudryavtsev, A. V.; Osipov, S. L.; Mosunova, N. A.; Strizhov, V. F.; Usov, E. V.

2016-02-01

The conceptual fundamentals of the development of the new-generation system thermal-hydraulic computational HYDRA-IBRAE/LM code are presented. The code is intended to simulate the thermalhydraulic processes that take place in the loops and the heat-exchange equipment of liquid-metal cooled fast reactor systems under normal operation and anticipated operational occurrences and during accidents. The paper provides a brief overview of Russian and foreign system thermal-hydraulic codes for modeling liquid-metal coolants and gives grounds for the necessity of development of a new-generation HYDRA-IBRAE/LM code. Considering the specific engineering features of the nuclear power plants (NPPs) equipped with the BN-1200 and the BREST-OD-300 reactors, the processes and the phenomena are singled out that require a detailed analysis and development of the models to be correctly described by the system thermal-hydraulic code in question. Information on the functionality of the computational code is provided, viz., the thermalhydraulic two-phase model, the properties of the sodium and the lead coolants, the closing equations for simulation of the heat-mass exchange processes, the models to describe the processes that take place during the steam-generator tube rupture, etc. The article gives a brief overview of the usability of the computational code, including a description of the support documentation and the supply package, as well as possibilities of taking advantages of the modern computer technologies, such as parallel computations. The paper shows the current state of verification and validation of the computational code; it also presents information on the principles of constructing of and populating the verification matrices for the BREST-OD-300 and the BN-1200 reactor systems. The prospects are outlined for further development of the HYDRA-IBRAE/LM code, introduction of new models into it, and enhancement of its usability. It is shown that the program of development and practical application of the code will allow carrying out in the nearest future the computations to analyze the safety of potential NPP projects at a qualitatively higher level.

Heuristic rules embedded genetic algorithm for in-core fuel management optimization

NASA Astrophysics Data System (ADS)

Alim, Fatih

The objective of this study was to develop a unique methodology and a practical tool for designing loading pattern (LP) and burnable poison (BP) pattern for a given Pressurized Water Reactor (PWR) core. Because of the large number of possible combinations for the fuel assembly (FA) loading in the core, the design of the core configuration is a complex optimization problem. It requires finding an optimal FA arrangement and BP placement in order to achieve maximum cycle length while satisfying the safety constraints. Genetic Algorithms (GA) have been already used to solve this problem for LP optimization for both PWR and Boiling Water Reactor (BWR). The GA, which is a stochastic method works with a group of solutions and uses random variables to make decisions. Based on the theories of evaluation, the GA involves natural selection and reproduction of the individuals in the population for the next generation. The GA works by creating an initial population, evaluating it, and then improving the population by using the evaluation operators. To solve this optimization problem, a LP optimization package, GARCO (Genetic Algorithm Reactor Code Optimization) code is developed in the framework of this thesis. This code is applicable for all types of PWR cores having different geometries and structures with an unlimited number of FA types in the inventory. To reach this goal, an innovative GA is developed by modifying the classical representation of the genotype. To obtain the best result in a shorter time, not only the representation is changed but also the algorithm is changed to use in-core fuel management heuristics rules. The improved GA code was tested to demonstrate and verify the advantages of the new enhancements. The developed methodology is explained in this thesis and preliminary results are shown for the VVER-1000 reactor hexagonal geometry core and the TMI-1 PWR. The improved GA code was tested to verify the advantages of new enhancements. The core physics code used for VVER in this research is Moby-Dick, which was developed to analyze the VVER by SKODA Inc. The SIMULATE-3 code, which is an advanced two-group nodal code, is used to analyze the TMI-1.

Preliminary Concept of Operations for the Spent Fuel Management System--WM2017

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cumberland, Riley M; Adeniyi, Abiodun Idowu; Howard, Rob L

The Nuclear Fuels Storage and Transportation Planning Project (NFST) within the U.S. Department of Energy s Office of Nuclear Energy is tasked with identifying, planning, and conducting activities to lay the groundwork for developing interim storage and transportation capabilities in support of an integrated waste management system. The system will provide interim storage for commercial spent nuclear fuel (SNF) from reactor sites and deliver it to a repository. The system will also include multiple subsystems, potentially including; one or more interim storage facilities (ISF); one or more repositories; facilities to package and/or repackage SNF; and transportation systems. The project teammore » is analyzing options for an integrated waste management system. To support analysis, the project team has developed a Concept of Operations document that describes both the potential integrated system and inter-dependencies between system components. The goal of this work is to aid systems analysts in the development of consistent models across the project, which involves multiple investigators. The Concept of Operations document will be updated periodically as new developments emerge. At a high level, SNF is expected to travel from reactors to a repository. SNF is first unloaded from reactors and placed in spent fuel pools for wet storage at utility sites. After the SNF has cooled enough to satisfy loading limits, it is placed in a container at reactor sites for storage and/or transportation. After transportation requirements are met, the SNF is transported to an ISF to store the SNF until a repository is developed or directly to a repository if available. While the high level operation of the system is straightforward, analysts must evaluate numerous alternative options. Alternative options include the number of ISFs (if any), ISF design, the stage at which SNF repackaging occurs (if any), repackaging technology, the types of containers used, repository design, component sizing, and timing of events. These alternative options arise due to technological, economic, or policy considerations. As new developments regularly emerge, the operational concepts will be periodically updated. This paper gives an overview of the different potential alternatives identified in the Concept of Operations document at a conceptual level.« less

Power Balance Analysis of the Prototype-Material Plasma Exposure eXperiment

NASA Astrophysics Data System (ADS)

Showers, M. A.; Biewer, T. M.; Caneses, J. F.; Caughman, J. B. O.; Lumsdaine, A.; Owen, L.; Rapp, J.; Youchison, D.; Beers, C. J.; Donovan, D. C.; Kafle, N.; Ray, H. B.

2017-10-01

The Prototype-Material Plasma Exposure eXperiment (Proto-MPEX) is a test bed for the plasma source concept for the planned Material Plasma Exposure eXperiment (MPEX), a steady-state linear device studying plasma material interactions for fusion reactors. A power balance of Proto-MPEX attempts to identify machine operating parameters that will improve Proto-MPEX's performance, potentially impacting the MPEX design concept. A power balance has been performed utilizing an extensive diagnostic suite to identify mechanisms and locations of power loss from the main plasma. The diagnostic package includes infrared cameras, double Langmuir probes, fluoroptic probes, Mach probes, a Thomson scattering diagnostic, a McPherson spectrometer and in-vessel thermocouples. Radiation losses are estimated with absolute calibrated spectroscopic signals. This work was supported by the U.S. D.O.E. contract DE-AC05-00OR22725.

Multiphysics Modeling of an Annular Linear Induction Pump With Applications to Space Nuclear Power Systems

NASA Technical Reports Server (NTRS)

Kilbane, J.; Polzin, K. A.

2014-01-01

An annular linear induction pump (ALIP) that could be used for circulating liquid-metal coolant in a fission surface power reactor system is modeled in the present work using the computational COMSOL Multiphysics package. The pump is modeled using a two-dimensional, axisymmetric geometry and solved under conditions similar to those used during experimental pump testing. Real, nonlinear, temperature-dependent material properties can be incorporated into the model for both the electrically-conducting working fluid in the pump (NaK-78) and structural components of the pump. The intricate three-phase coil configuration of the pump is implemented in the model to produce an axially-traveling magnetic wave that is qualitatively similar to the measured magnetic wave. The model qualitatively captures the expected feature of a peak in efficiency as a function of flow rate.

Remaining Sites Verification Package for the 116-C-3, 105-C Chemical Waste Tanks, Waste Site Reclassification Form 2008-002

DOE Office of Scientific and Technical Information (OSTI.GOV)

L. M. Dittmer

2008-01-31

The 116-C-3 waste site consisted of two underground storage tanks designed to receive mixed waste from the 105-C Reactor Metals Examination Facility chemical dejacketing process. Confirmatory evaluation and subsequent characterization of the site determined that the southern tank contained approximately 34,000 L (9,000 gal) of dejacketing wastes, and that the northern tank was unused. In accordance with this evaluation, the verification sampling and modeling results support a reclassification of this site to Interim Closed Out. The results of verification sampling demonstrate that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils.more » The results also show that residual contaminant concentrations are protective of groundwater and the Columbia River.« less

Pyrolysis behavior of different type of materials contained in the rejects of packaging waste sorting plants.

PubMed

Adrados, A; De Marco, I; Lopez-Urionabarrenechea, A; Caballero, B M; Laresgoiti, M F

2013-01-01

In this paper rejected streams coming from a waste packaging material recovery facility have been characterized and separated into families of products of similar nature in order to determine the influence of different types of ingredients in the products obtained in the pyrolysis process. The pyrolysis experiments have been carried out in a non-stirred batch 3.5 dm(3) reactor, swept with 1 L min(-1) N(2), at 500°C for 30 min. Pyrolysis liquids are composed of an organic phase and an aqueous phase. The aqueous phase is greater as higher is the cellulosic material content in the sample. The organic phase contains valuable chemicals as styrene, ethylbenzene and toluene, and has high heating value (HHV) (33-40 MJ kg(-1)). Therefore they could be used as alternative fuels for heat and power generation and as a source of valuable chemicals. Pyrolysis gases are mainly composed of hydrocarbons but contain high amounts of CO and CO(2); their HHV is in the range of 18-46 MJ kg(-1). The amount of COCO(2) increases, and consequently HHV decreases as higher is the cellulosic content of the waste. Pyrolysis solids are mainly composed of inorganics and char formed in the process. The cellulosic materials lower the quality of the pyrolysis liquids and gases, and increase the production of char. Copyright © 2012 Elsevier Ltd. All rights reserved.

Savannah River Site Spent Nuclear Fuel Management Final Environmental Impact Statement

DOE Office of Scientific and Technical Information (OSTI.GOV)

N /A

The proposed DOE action considered in this environmental impact statement (EIS) is to implement appropriate processes for the safe and efficient management of spent nuclear fuel and targets at the Savannah River Site (SRS) in Aiken County, South Carolina, including placing these materials in forms suitable for ultimate disposition. Options to treat, package, and store this material are discussed. The material included in this EIS consists of approximately 68 metric tons heavy metal (MTHM) of spent nuclear fuel 20 MTHM of aluminum-based spent nuclear fuel at SRS, as much as 28 MTHM of aluminum-clad spent nuclear fuel from foreign andmore » domestic research reactors to be shipped to SRS through 2035, and 20 MTHM of stainless-steel or zirconium-clad spent nuclear fuel and some Americium/Curium Targets stored at SRS. Alternatives considered in this EIS encompass a range of new packaging, new processing, and conventional processing technologies, as well as the No Action Alternative. A preferred alternative is identified in which DOE would prepare about 97% by volume (about 60% by mass) of the aluminum-based fuel for disposition using a melt and dilute treatment process. The remaining 3% by volume (about 40% by mass) would be managed using chemical separation. Impacts are assessed primarily in the areas of water resources, air resources, public and worker health, waste management, socioeconomic, and cumulative impacts.« less

Disposal Of Irradiated Cadmium Control Rods From The Plumbrook Reactor Facility

DOE Office of Scientific and Technical Information (OSTI.GOV)

Posivak, E.J.; Berger, S.R.; Freitag, A.A.

2008-07-01

Innovative mixed waste disposition from NASA's Plum Brook Reactor Facility was accomplished without costly repackaging. Irradiated characteristic hardware with contact dose rates as high as 8 Sv/hr was packaged in a HDPE overpack and stored in a Secure Environmental Container during earlier decommissioning efforts, awaiting identification of a suitable pathway. WMG obtained regulatory concurrence that the existing overpack would serve as the macro-encapsulant per 40CFR268.45 Table 1.C. The overpack vent was disabled and the overpack was placed in a stainless steel liner to satisfy overburden slumping requirements. The liner was sealed and placed in shielded shoring for transport to themore » disposal site in a US DOT Type A cask. Disposition via this innovative method avoided cost, risk, and dose associated with repackaging the high dose irradiated characteristic hardware. In conclusion: WMG accomplished what others said could not be done. Large D and D contractors advised NASA that the cadmium control rods could only be shipped to the proposed Yucca mountain repository. NASA management challenged MOTA to find a more realistic alternative. NASA and MOTA turned to WMG to develop a methodology to disposition the 'hot and nasty' waste that presumably had no path forward. Although WMG lead a team that accomplished the 'impossible', the project could not have been completed with out the patient, supportive management by DOE-EM, NASA, and MOTA. (authors)« less

Refuse Derived Fuel (RDF) production and gasification in a pilot plant integrated with an Otto cycle ICE through Aspen plus™ modelling: Thermodynamic and economic viability.

PubMed

Násner, Albany Milena Lozano; Lora, Electo Eduardo Silva; Palacio, José Carlos Escobar; Rocha, Mateus Henrique; Restrepo, Julian Camilo; Venturini, Osvaldo José; Ratner, Albert

2017-11-01

This work deals with the development of a Refuse Derived Fuel (RDF) gasification pilot plant using air as a gasification agent. A downdraft fixed bed reactor is integrated with an Otto cycle Internal Combustion Engine (ICE). Modelling was carried out using the Aspen Plus™ software to predict the ideal operational conditions for maximum efficiency. Thermodynamics package used in the simulation comprised the Non-Random Two-Liquid (NRTL) model and the Hayden-O'Connell (HOC) equation of state. As expected, the results indicated that the Equivalence Ratio (ER) has a direct influence over the gasification temperature and the composition of the Raw Produced Gas (RPG), and effects of ER over the Lower Heating Value (LHV) and Cold Gasification Efficiency (CGE) of the RPG are also discussed. A maximum CGE efficiency of 57-60% was reached for ER values between 0.25 and 0.3, also an average reactor temperature values in the range of 680-700°C, with a peak LHV of 5.8MJ/Nm 3 . RPG was burned in an ICE, reaching an electrical power of 50kW el . The economic assessment of the pilot plant implementation was also performed, showing the project is feasible, with power above 120kW el with an initial investment of approximately US$ 300,000. Copyright © 2017 Elsevier Ltd. All rights reserved.

Characterization of Novel Calorimeters in the Annular Core Research Reactor

NASA Astrophysics Data System (ADS)

Hehr, Brian D.; Parma, Edward J.; Peters, Curtis D.; Naranjo, Gerald E.; Luker, S. Michael

2016-02-01

A series of pulsed irradiation experiments have been performed in the central cavity of Sandia National Laboratories' Annular Core Research Reactor (ACRR) to characterize the responses of a set of elemental calorimeter materials including Si, Zr, Sn, Ta, W, and Bi. Of particular interest was the perturbing effect of the calorimeter itself on the ambient radiation field - a potential concern in dosimetry applications. By placing the calorimeter package into a neutron-thermalizing lead/polyethylene (LP) bucket and irradiating both with and without a cadmium wrapper, it was demonstrated that prompt capture gammas generated inside the calorimeters can be a significant contributor to the measured dose in the active disc region. An MCNP model of the experimental setup was shown to replicate measured dose responses to within 10%. The internal (n,γ) contribution was found to constitute as much as 50% of the response inside the LP bucket and up to 20% inside the nominal (unmodified) cavity environment, with Ta and W exhibiting the largest enhancement due to their sizable (n,γ) cross sections. Capture reactions in non-disc components of the calorimeter were estimated to be responsible for up to a few percent of the measured response. This work was supported by the United States Department of Energy under Contract DE-AC04-94AL85000. Sandia is a multiprogram laboratory operated by Sandia Corporation, a Lockheed Martin Company, for the United States Department of Energy.

IJS procedure for RELAP5 to TRACE input model conversion using SNAP

DOE Office of Scientific and Technical Information (OSTI.GOV)

Prosek, A.; Berar, O. A.

2012-07-01

The TRAC/RELAP Advanced Computational Engine (TRACE) advanced, best-estimate reactor systems code developed by the U.S. Nuclear Regulatory Commission comes with a graphical user interface called Symbolic Nuclear Analysis Package (SNAP). Much of efforts have been done in the past to develop the RELAP5 input decks. The purpose of this study is to demonstrate the Institut 'Josef Stefan' (IJS) conversion procedure from RELAP5 to TRACE input model of BETHSY facility. The IJS conversion procedure consists of eleven steps and is based on the use of SNAP. For calculations of the selected BETHSY 6.2TC test the RELAP5/MOD3.3 Patch 4 and TRACE V5.0more » Patch 1 were used. The selected BETHSY 6.2TC test was 15.24 cm equivalent diameter horizontal cold leg break in the reference pressurized water reactor without high pressure and low pressure safety injection. The application of the IJS procedure for conversion of BETHSY input model showed that it is important to perform the steps in proper sequence. The overall calculated results obtained with TRACE using the converted RELAP5 model were close to experimental data and comparable to RELAP5/MOD3.3 calculations. Therefore it can be concluded, that proposed IJS conversion procedure was successfully demonstrated on the BETHSY integral test facility input model. (authors)« less

RELAP-7 Level 2 Milestone Report: Demonstration of a Steady State Single Phase PWR Simulation with RELAP-7

DOE Office of Scientific and Technical Information (OSTI.GOV)

David Andrs; Ray Berry; Derek Gaston

The document contains the simulation results of a steady state model PWR problem with the RELAP-7 code. The RELAP-7 code is the next generation nuclear reactor system safety analysis code being developed at Idaho National Laboratory (INL). The code is based on INL's modern scientific software development framework - MOOSE (Multi-Physics Object-Oriented Simulation Environment). This report summarizes the initial results of simulating a model steady-state single phase PWR problem using the current version of the RELAP-7 code. The major purpose of this demonstration simulation is to show that RELAP-7 code can be rapidly developed to simulate single-phase reactor problems. RELAP-7more » is a new project started on October 1st, 2011. It will become the main reactor systems simulation toolkit for RISMC (Risk Informed Safety Margin Characterization) and the next generation tool in the RELAP reactor safety/systems analysis application series (the replacement for RELAP5). The key to the success of RELAP-7 is the simultaneous advancement of physical models, numerical methods, and software design while maintaining a solid user perspective. Physical models include both PDEs (Partial Differential Equations) and ODEs (Ordinary Differential Equations) and experimental based closure models. RELAP-7 will eventually utilize well posed governing equations for multiphase flow, which can be strictly verified. Closure models used in RELAP5 and newly developed models will be reviewed and selected to reflect the progress made during the past three decades. RELAP-7 uses modern numerical methods, which allow implicit time integration, higher order schemes in both time and space, and strongly coupled multi-physics simulations. RELAP-7 is written with object oriented programming language C++. Its development follows modern software design paradigms. The code is easy to read, develop, maintain, and couple with other codes. Most importantly, the modern software design allows the RELAP-7 code to evolve with time. RELAP-7 is a MOOSE-based application. MOOSE (Multiphysics Object-Oriented Simulation Environment) is a framework for solving computational engineering problems in a well-planned, managed, and coordinated way. By leveraging millions of lines of open source software packages, such as PETSC (a nonlinear solver developed at Argonne National Laboratory) and LibMesh (a Finite Element Analysis package developed at University of Texas), MOOSE significantly reduces the expense and time required to develop new applications. Numerical integration methods and mesh management for parallel computation are provided by MOOSE. Therefore RELAP-7 code developers only need to focus on physics and user experiences. By using the MOOSE development environment, RELAP-7 code is developed by following the same modern software design paradigms used for other MOOSE development efforts. There are currently over 20 different MOOSE based applications ranging from 3-D transient neutron transport, detailed 3-D transient fuel performance analysis, to long-term material aging. Multi-physics and multiple dimensional analyses capabilities can be obtained by coupling RELAP-7 and other MOOSE based applications and by leveraging with capabilities developed by other DOE programs. This allows restricting the focus of RELAP-7 to systems analysis-type simulations and gives priority to retain and significantly extend RELAP5's capabilities.« less

Microchannel Reactor System Design & Demonstration For On-Site H2O2 Production by Controlled H2/O2 Reaction

DOE Office of Scientific and Technical Information (OSTI.GOV)

Adeniyi Lawal

We successfully demonstrated an innovative hydrogen peroxide (H2O2) production concept which involved the development of flame- and explosion-resistant microchannel reactor system for energy efficient, cost-saving, on-site H2O2 production. We designed, fabricated, evaluated, and optimized a laboratory-scale microchannel reactor system for controlled direct combination of H2 and O2 in all proportions including explosive regime, at a low pressure and a low temperature to produce about 1.5 wt% H2O2 as proposed. In the second phase of the program, as a prelude to full-scale commercialization, we demonstrated our H2O2 production approach by ‘numbering up’ the channels in a multi-channel microreactor-based pilot plant tomore » produce 1 kg/h of H2O2 at 1.5 wt% as demanded by end-users of the developed technology. To our knowledge, we are the first group to accomplish this significant milestone. We identified the reaction pathways that comprise the process, and implemented rigorous mechanistic kinetic studies to obtain the kinetics of the three main dominant reactions. We are not aware of any such comprehensive kinetic studies for the direct combination process, either in a microreactor or any other reactor system. We showed that the mass transfer parameter in our microreactor system is several orders of magnitude higher than what obtains in the macroreactor, attesting to the superior performance of microreactor. A one-dimensional reactor model incorporating the kinetics information enabled us to clarify certain important aspects of the chemistry of the direct combination process as detailed in section 5 of this report. Also, through mathematical modeling and simulation using sophisticated and robust commercial software packages, we were able to elucidate the hydrodynamics of the complex multiphase flows that take place in the microchannel. In conjunction with the kinetics information, we were able to validate the experimental data. If fully implemented across the whole industry as a result of our technology demonstration, our production concept is expected to save >5 trillion Btu/year of steam usage and >3 trillion Btu/year in electric power consumption. Our analysis also indicates >50 % reduction in waste disposal cost and ~10% reduction in feedstock energy. These savings translate to ~30% reduction in overall production and transportation costs for the $1B annual H2O2 market.« less

Automated Work Package: Conceptual Design and Data Architecture

DOE Office of Scientific and Technical Information (OSTI.GOV)

Al Rashdan, Ahmad; Oxstrand, Johanna; Agarwal, Vivek

The automated work package (AWP) is one of the U.S. Department of Energy’s (DOE) Light Water Reactor Sustainability Program efforts to enhance the safety and economics of the nuclear power industry. An AWP is an adaptive and interactive work package that intelligently drives the work process according to the plant condition, resources status, and users progress. The AWP aims to automate several manual tasks of the work process to enhance human performance and reduce human errors. Electronic work packages (eWPs), studied by the Electric Power Research Institute (EPRI), are work packages that rely to various extent on electronic data processingmore » and presentation. AWPs are the future of eWPs. They are envisioned to incorporate the advanced technologies of the future, and thus address the unresolved deficiencies associated with the eWPs in a nuclear power plant. In order to define the AWP, it is necessary to develop an ideal envisioned scenario of the future work process without any current technology restriction. The approach followed to develop this scenario is specific to every stage of the work process execution. The scenario development resulted in fifty advanced functionalities that can be part of the AWP. To rank the importance of these functionalities, a survey was conducted involving several U.S. nuclear utilities. The survey aimed at determining the current need of the nuclear industry with respect to the current work process, i.e. what the industry is satisfied with, and where the industry envisions potential for improvement. The survey evaluated the most promising functionalities resulting from the scenario development. The results demonstrated a significant desire to adopt the majority of these functionalities. The results of the survey are expected to drive the Idaho National Laboratory (INL) AWP research and development (R&D). In order to facilitate this mission, a prototype AWP is needed. Since the vast majority of earlier efforts focused on the frontend aspects of the AWP, the backend data architecture was researched and developed in this effort. The backend design involved data architecture aspects. It was realized through this effort that the key aspects of this design are hierarchy, data configuration and live information, data templates and instances, the flow of work package execution, the introduction of properties, and the means to interface the backend to the frontend. After the backend design was developed, a data structure was built to reflect the developed data architecture. The data structure was developed to accommodate the fifty functionalities identified by the envisioned scenario development. The data structure was evaluated by incorporating an example work order from the nuclear power industry. The implementation resulted in several optimization iterations of the data structure. In addition, the rearrangement of the work order information to fit the data structure highlighted several possibilities for improvement in the current work order design, and significantly reduced the size of the work order.« less

Processing of thermionic power on an electrically propelled spacecraft

NASA Technical Reports Server (NTRS)

Macie, T. W.

1973-01-01

A study to define the power processing equipment required between a thermionic reactor and an array of mercury-ion thrusters for a nuclear electric propulsion system is reported. Observations and recommendations that resulted from this study were: (1) the preferred thermionic-fuel-element source voltages are 23 V or higher; (2) transistor characteristics exert a strong effect on power processor mass; (3) the power processor mass could be considerably reduced should the magnetic materials that exhibit low losses at high frequencies, that have a high Curie point, and that can operate at 15 to 20 kG become avaliable; (4) electrical component packaging on the radiator could reduce the area that is sensitive to meteoroid penetration, thereby reducing the meteoroid shielding mass requirement; (5) an experimental model of the power processor design should be built and tested to verify the efficiencies, masses, and all the automatic operational aspects of the design.

Aging of electronics with application to nuclear power plant instrumentation. [PWR; BWR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Johnson, Jr, R T; Thome, F V; Craft, C M

1983-01-01

A survey to identify areas of needed research to understand aging mechanisms for electronics in nuclear power plant instrumentation has been completed. The emphasis was on electronic components such as semiconductors, capacitors, and resistors used in safety-related instrumentation in the reactor containment area. The environmental and operational stress factors which may produce degradation during long-term operation were identified. Some attention was also given to humidity effects as related to seals and encapsulants, and failures in printed circuit boards and bonds and solder joints. Results suggest that neutron as well as gamma irradiations should be considered in simulating the aging environmentmore » for electronic components. Radiation dose-rate effects in semiconductor devices and organic capacitors need to be further investigated, as well as radiation-voltage bias synergistic effects in semiconductor devices and leakage and permeation of moisture through seals in electronics packages.« less

Technology of Producing the Contact Connections of Superconductor Metal-Sheathed Cable

NASA Astrophysics Data System (ADS)

Jakubowski, Andrzej

2017-06-01

The technology of producing the current contact connections on the superconductor cable edges is presented. This lead cable is used as one of the major elements of the magnetic system in thermonuclear reactor construction, actuality for modern world energy. The technology is realized by the radial draft of metal thin-walled tube on the conductor's package. The filling of various profiles by round section wire is optimized. Geometrical characteristics of the dangerous crosssection (as a broken ring) of thin-walled tube injured by the sector cut-out are accounted. The comparative strength calculation of the solid and injured tubes at a longitudinal compression and lateral bending is acted. The radial draft mechanism of cylindrical thin-walled sheath with the wire packing is designed. The necessity to use the nonlinear theory for the sheaths calculate is set. The resilient co-operation of wires as the parallel located cylinders with the contact stripes of rectangular form is considered.

Prediction of normalized biodiesel properties by simulation of multiple feedstock blends.

PubMed

García, Manuel; Gonzalo, Alberto; Sánchez, José Luis; Arauzo, Jesús; Peña, José Angel

2010-06-01

A continuous process for biodiesel production has been simulated using Aspen HYSYS V7.0 software. As fresh feed, feedstocks with a mild acid content have been used. The process flowsheet follows a traditional alkaline transesterification scheme constituted by esterification, transesterification and purification stages. Kinetic models taking into account the concentration of the different species have been employed in order to simulate the behavior of the CSTR reactors and the product distribution within the process. The comparison between experimental data found in literature and the predicted normalized properties, has been discussed. Additionally, a comparison between different thermodynamic packages has been performed. NRTL activity model has been selected as the most reliable of them. The combination of these models allows the prediction of 13 out of 25 parameters included in standard EN-14214:2003, and confers simulators a great value as predictive as well as optimization tool. (c) 2010 Elsevier Ltd. All rights reserved.

Development of a New 47-Group Library for the CASL Neutronics Simulators

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kim, Kang Seog; Williams, Mark L; Wiarda, Dorothea

The CASL core simulator MPACT is under development for the neutronics and thermal-hydraulics coupled simulation for the pressurized light water reactors. The key characteristics of the MPACT code include a subgroup method for resonance self-shielding, and a whole core solver with a 1D/2D synthesis method. The ORNL AMPX/SCALE code packages have been significantly improved to support various intermediate resonance self-shielding approximations such as the subgroup and embedded self-shielding methods. New 47-group AMPX and MPACT libraries based on ENDF/B-VII.0 have been generated for the CASL core simulator MPACT of which group structure comes from the HELIOS library. The new 47-group MPACTmore » library includes all nuclear data required for static and transient core simulations. This study discusses a detailed procedure to generate the 47-group AMPX and MPACT libraries and benchmark results for the VERA progression problems.« less

Implementation, capabilities, and benchmarking of Shift, a massively parallel Monte Carlo radiation transport code

DOE PAGES

Pandya, Tara M.; Johnson, Seth R.; Evans, Thomas M.; ...

2015-12-21

This paper discusses the implementation, capabilities, and validation of Shift, a massively parallel Monte Carlo radiation transport package developed and maintained at Oak Ridge National Laboratory. It has been developed to scale well from laptop to small computing clusters to advanced supercomputers. Special features of Shift include hybrid capabilities for variance reduction such as CADIS and FW-CADIS, and advanced parallel decomposition and tally methods optimized for scalability on supercomputing architectures. Shift has been validated and verified against various reactor physics benchmarks and compares well to other state-of-the-art Monte Carlo radiation transport codes such as MCNP5, CE KENO-VI, and OpenMC. Somemore » specific benchmarks used for verification and validation include the CASL VERA criticality test suite and several Westinghouse AP1000 ® problems. These benchmark and scaling studies show promising results.« less

ANNETTE Project: Contributing to The Nuclearization of Fusion

NASA Astrophysics Data System (ADS)

Ambrosini, W.; Cizelj, L.; Dieguez Porras, P.; Jaspers, R.; Noterdaeme, J.; Scheffer, M.; Schoenfelder, C.

2018-01-01

The ANNETTE Project (Advanced Networking for Nuclear Education and Training and Transfer of Expertise) is well underway, and one of its work packages addresses the design, development and implementation of nuclear fusion training. A systematic approach is used that leads to the development of new training courses, based on identified nuclear competences needs of the work force of (future) fusion reactors and on the current availability of suitable training courses. From interaction with stakeholders involved in the ITER design and construction or the JET D-T campaign, it became clear that the lack of nuclear safety culture awareness already has an impact on current projects. Through the collaboration between the European education networks in fission (ENEN) and fusion (FuseNet) in the ANNETTE project, this project is well positioned to support the development of nuclear competences for ongoing and future fusion projects. Thereby it will make a clear contribution to the realization of fusion energy.

Columbia, OV-102, forward middeck locker experiments and meal tray assemblies

NASA Technical Reports Server (NTRS)

1982-01-01

Overall view of forward middeck locker shows Continuous Flow Electrophoresis System (CFES) experiment control and monitoring module and sample storage module (on port side) and Monodisperse Latex Reactor (MLR) (on starboard side). Water Dispenser Kit water gun (above CFES module) and meal tray assemblies covered with snack food packages and beverage containers appear around the two experiments. Thanks to a variety of juices and other food items, this array in the middeck probably represents the most colorful area onboard the Earth-orbiting Columbia, Orbiter Vehicle (OV) 102. Most of the meal items have been carefully fastened to meal tray assemblies (foodtrays) and locker doors (or both). What has not been attached by conventional methods has been safely 'tucked' under something heavy (note jacket shoved into space occupied MLR). MLR is making its second flight and is designed to test the flexibility of making large-size, monodisperse (same size), polystyrene latex micro-spheres using

Compact gasoline fuel processor for passenger vehicle APU

NASA Astrophysics Data System (ADS)

Severin, Christopher; Pischinger, Stefan; Ogrzewalla, Jürgen

Due to the increasing demand for electrical power in today's passenger vehicles, and with the requirements regarding fuel consumption and environmental sustainability tightening, a fuel cell-based auxiliary power unit (APU) becomes a promising alternative to the conventional generation of electrical energy via internal combustion engine, generator and battery. It is obvious that the on-board stored fuel has to be used for the fuel cell system, thus, gasoline or diesel has to be reformed on board. This makes the auxiliary power unit a complex integrated system of stack, air supply, fuel processor, electrics as well as heat and water management. Aside from proving the technical feasibility of such a system, the development has to address three major barriers:start-up time, costs, and size/weight of the systems. In this paper a packaging concept for an auxiliary power unit is presented. The main emphasis is placed on the fuel processor, as good packaging of this large subsystem has the strongest impact on overall size. The fuel processor system consists of an autothermal reformer in combination with water-gas shift and selective oxidation stages, based on adiabatic reactors with inter-cooling. The configuration was realized in a laboratory set-up and experimentally investigated. The results gained from this confirm a general suitability for mobile applications. A start-up time of 30 min was measured, while a potential reduction to 10 min seems feasible. An overall fuel processor efficiency of about 77% was measured. On the basis of the know-how gained by the experimental investigation of the laboratory set-up a packaging concept was developed. Using state-of-the-art catalyst and heat exchanger technology, the volumes of these components are fixed. However, the overall volume is higher mainly due to mixing zones and flow ducts, which do not contribute to the chemical or thermal function of the system. Thus, the concept developed mainly focuses on minimization of those component volumes. Therefore, the packaging utilizes rectangular catalyst bricks and integrates flow ducts into the heat exchangers. A concept is presented with a 25 l fuel processor volume including thermal isolation for a 3 kW el auxiliary power unit. The overall size of the system, i.e. including stack, air supply and auxiliaries can be estimated to 44 l.

Gap Analysis of Material Properties Data for Ferritic/Martensitic HT-9 Steel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Brown, Neil R.; Serrano De Caro, Magdalena; Rodriguez, Edward A.

2012-08-28

The US Department of Energy (DOE), Office of Nuclear Energy (NE), is supporting the development of an ASME Code Case for adoption of 12Cr-1Mo-VW ferritic/martensitic (F/M) steel, commonly known as HT-9, primarily for use in elevated temperature design of liquid-metal fast reactors (LMFR) and components. In 2011, Los Alamos National Laboratory (LANL) nuclear engineering staff began assisting in the development of a small modular reactor (SMR) design concept, previously known as the Hyperion Module, now called the Gen4 Module. LANL staff immediately proposed HT-9 for the reactor vessel and components, as well as fuel clad and ducting, due to itsmore » superior thermal qualities. Although the ASME material Code Case, for adoption of HT-9 as an approved elevated temperature material for LMFR service, is the ultimate goal of this project, there are several key deliverables that must first be successfully accomplished. The most important key deliverable is the research, accumulation, and documentation of specific material parameters; physical, mechanical, and environmental, which becomes the basis for an ASME Code Case. Time-independent tensile and ductility data and time-dependent creep and creep-rupture behavior are some of the material properties required for a successful ASME Code case. Although this report provides a cursory review of the available data, a much more comprehensive study of open-source data would be necessary. This report serves three purposes: (a) provides a list of already existing material data information that could ultimately be made available to the ASME Code, (b) determines the HT-9 material properties data missing from available sources that would be required and (c) estimates the necessary material testing required to close the gap. Ultimately, the gap analysis demonstrates that certain material properties testing will be required to fulfill the necessary information package for an ASME Code Case.« less

Pellet-clad mechanical interaction screening using VERA applied to Watts Bar Unit 1, Cycles 1–3

DOE PAGES

Stimpson, Shane; Powers, Jeffrey; Clarno, Kevin; ...

2017-12-22

The Consortium for Advanced Simulation of Light Water Reactors (CASL) aims to provide high-fidelity multiphysics simulations of light water nuclear reactors. To accomplish this, CASL is developing the Virtual Environment for Reactor Applications (VERA), which is a suite of code packages for thermal hydraulics, neutron transport, fuel performance, and coolant chemistry. As VERA continues to grow and expand, there has been an increased focus on incorporating fuel performance analysis methods. One of the primary goals of CASL is to estimate local cladding failure probability through pellet-clad interaction, which consists of both pellet-clad mechanical interaction (PCMI) and stress corrosion cracking. Estimatingmore » clad failure is important to preventing release of fission products to the primary system and accurate estimates could prove useful in establishing less conservative power ramp rates or when considering load-follow operations.While this capability is being pursued through several different approaches, the procedure presented in this article focuses on running independent fuel performance calculations with BISON using a file-based one-way coupling based on multicycle output data from high fidelity, pin-resolved coupled neutron transport–thermal hydraulics simulations. This type of approach is consistent with traditional fuel performance analysis methods, which are typically separate from core simulation analyses. A more tightly coupled approach is currently being developed, which is the ultimate target application in CASL.Recent work simulating 12 cycles of Watts Bar Unit 1 with VERA core simulator are capitalized upon, and quarter-core BISON results for parameters of interest to PCMI (maximum centerline fuel temperature, maximum clad hoop stress, and minimum gap size) are presented for Cycles 1–3. In conclusion, based on these results, this capability demonstrates its value and how it could be used as a screening tool for gathering insight into PCMI, singling out limiting rods for further, more detailed analysis.« less

Pellet-clad mechanical interaction screening using VERA applied to Watts Bar Unit 1, Cycles 1–3

DOE Office of Scientific and Technical Information (OSTI.GOV)

Stimpson, Shane; Powers, Jeffrey; Clarno, Kevin

The Consortium for Advanced Simulation of Light Water Reactors (CASL) aims to provide high-fidelity multiphysics simulations of light water nuclear reactors. To accomplish this, CASL is developing the Virtual Environment for Reactor Applications (VERA), which is a suite of code packages for thermal hydraulics, neutron transport, fuel performance, and coolant chemistry. As VERA continues to grow and expand, there has been an increased focus on incorporating fuel performance analysis methods. One of the primary goals of CASL is to estimate local cladding failure probability through pellet-clad interaction, which consists of both pellet-clad mechanical interaction (PCMI) and stress corrosion cracking. Estimatingmore » clad failure is important to preventing release of fission products to the primary system and accurate estimates could prove useful in establishing less conservative power ramp rates or when considering load-follow operations.While this capability is being pursued through several different approaches, the procedure presented in this article focuses on running independent fuel performance calculations with BISON using a file-based one-way coupling based on multicycle output data from high fidelity, pin-resolved coupled neutron transport–thermal hydraulics simulations. This type of approach is consistent with traditional fuel performance analysis methods, which are typically separate from core simulation analyses. A more tightly coupled approach is currently being developed, which is the ultimate target application in CASL.Recent work simulating 12 cycles of Watts Bar Unit 1 with VERA core simulator are capitalized upon, and quarter-core BISON results for parameters of interest to PCMI (maximum centerline fuel temperature, maximum clad hoop stress, and minimum gap size) are presented for Cycles 1–3. In conclusion, based on these results, this capability demonstrates its value and how it could be used as a screening tool for gathering insight into PCMI, singling out limiting rods for further, more detailed analysis.« less

Production of Hydrogen by Superadiabatic Decomposition of Hydrogen Sulfide - Final Technical Report for the Period June 1, 1999 - September 30, 2000

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rachid B. Slimane; Francis S. Lau; Javad Abbasian

2000-10-01

The objective of this program is to develop an economical process for hydrogen production, with no additional carbon dioxide emission, through the thermal decomposition of hydrogen sulfide (H{sub 2}S) in H{sub 2}S-rich waste streams to high-purity hydrogen and elemental sulfur. The novel feature of the process being developed is the superadiabatic combustion (SAC) of part of the H{sub 2}S in the waste stream to provide the thermal energy required for the decomposition reaction such that no additional energy is required. The program is divided into two phases. In Phase 1, detailed thermochemical and kinetic modeling of the SAC reactor withmore » H{sub 2}S-rich fuel gas and air/enriched air feeds is undertaken to evaluate the effects of operating conditions on exit gas products and conversion efficiency, and to identify key process parameters. Preliminary modeling results are used as a basis to conduct a thorough evaluation of SAC process design options, including reactor configuration, operating conditions, and productivity-product separation schemes, with respect to potential product yields, thermal efficiency, capital and operating costs, and reliability, ultimately leading to the preparation of a design package and cost estimate for a bench-scale reactor testing system to be assembled and tested in Phase 2 of the program. A detailed parametric testing plan was also developed for process design optimization and model verification in Phase 2. During Phase 2 of this program, IGT, UIC, and industry advisors UOP and BP Amoco will validate the SAC concept through construction of the bench-scale unit and parametric testing. The computer model developed in Phase 1 will be updated with the experimental data and used in future scale-up efforts. The process design will be refined and the cost estimate updated. Market survey and assessment will continue so that a commercial demonstration project can be identified.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dana L. Kelly

Typical engineering systems in applications with high failure consequences such as nuclear reactor plants often employ redundancy and diversity of equipment in an effort to lower the probability of failure and therefore risk. However, it has long been recognized that dependencies exist in these redundant and diverse systems. Some dependencies, such as common sources of electrical power, are typically captured in the logic structure of the risk model. Others, usually referred to as intercomponent dependencies, are treated implicitly by introducing one or more statistical parameters into the model. Such common-cause failure models have limitations in a simulation environment. In addition,more » substantial subjectivity is associated with parameter estimation for these models. This paper describes an approach in which system performance is simulated by drawing samples from the joint distributions of dependent variables. The approach relies on the notion of a copula distribution, a notion which has been employed by the actuarial community for ten years or more, but which has seen only limited application in technological risk assessment. The paper also illustrates how equipment failure data can be used in a Bayesian framework to estimate the parameter values in the copula model. This approach avoids much of the subjectivity required to estimate parameters in traditional common-cause failure models. Simulation examples are presented for failures in time. The open-source software package R is used to perform the simulations. The open-source software package WinBUGS is used to perform the Bayesian inference via Markov chain Monte Carlo sampling.« less

EARLY ENTRANCE COPRODUCTION PLANT

DOE Office of Scientific and Technical Information (OSTI.GOV)

David Storm; Govanon Nongbri; Steve Decanio

2004-01-12

The overall objective of this project is the three phase development of an Early Entrance Coproduction Plant (EECP) which uses petroleum coke to produce at least one product from at least two of the following three categories: (1) electric power (or heat), (2) fuels, and (3) chemicals using ChevronTexaco's proprietary gasification technology. The objective of Phase I is to determine the feasibility and define the concept for the EECP located at a specific site; develop a Research, Development, and Testing (RD&T) Plan to mitigate technical risks and barriers; and prepare a Preliminary Project Financing Plan. The objective of Phase IImore » is to implement the work as outlined in the Phase I RD&T Plan to enhance the development and commercial acceptance of coproduction technology. The objective of Phase III is to develop an engineering design package and a financing and testing plan for an EECP located at a specific site. The project's intended result is to provide the necessary technical, economic, and environmental information needed by industry to move the EECP forward to detailed design, construction, and operation. The partners in this project are Texaco Energy Systems LLC or TES (a subsidiary of ChevronTexaco), General Electric (GE), Praxair, and Kellogg Brown & Root (KBR) in addition to the U.S. Department of Energy (DOE). TES is providing gasification technology and Fischer-Tropsch (F-T) technology developed by Rentech, Inc., GE is providing combustion turbine technology, Praxair is providing air separation technology, and KBR is providing engineering. During Phase I, a design basis for the Fischer-Tropsch Synthesis section was developed based on limited experience with the specified feed gas and operating conditions. The objective of this Task in Phase II RD&T work was to confirm the performance of the F-T reactor at the set design conditions. Although much of the research, development, and testing work were done by TES outside of this project, several important issues were addressed in this phase of the project. They included Rejuvenation/Regeneration of the Fischer-Tropsch Catalyst, online Catalyst Withdrawal and Addition from the synthesis reactor, and the Fischer-Tropsch Design Basis Confirmation. In Phase III the results from these RD&T work will be incorporated in developing the engineering design package. This Topical Report documents the Phase II RD&T work that was completed for this task.« less

An Approach for Validating Actinide and Fission Product Burnup Credit Criticality Safety Analyses--Criticality (keff) Predictions

DOE Office of Scientific and Technical Information (OSTI.GOV)

Scaglione, John M; Mueller, Don; Wagner, John C

2011-01-01

One of the most significant remaining challenges associated with expanded implementation of burnup credit in the United States is the validation of depletion and criticality calculations used in the safety evaluation - in particular, the availability and use of applicable measured data to support validation, especially for fission products. Applicants and regulatory reviewers have been constrained by both a scarcity of data and a lack of clear technical basis or approach for use of the data. U.S. Nuclear Regulatory Commission (NRC) staff have noted that the rationale for restricting their Interim Staff Guidance on burnup credit (ISG-8) to actinide-only ismore » based largely on the lack of clear, definitive experiments that can be used to estimate the bias and uncertainty for computational analyses associated with using burnup credit. To address the issue of validation, the NRC initiated a project with the Oak Ridge National Laboratory to (1) develop and establish a technically sound validation approach (both depletion and criticality) for commercial spent nuclear fuel (SNF) criticality safety evaluations based on best-available data and methods and (2) apply the approach for representative SNF storage and transport configurations/conditions to demonstrate its usage and applicability, as well as to provide reference bias results. The purpose of this paper is to describe the criticality (k{sub eff}) validation approach, and resulting observations and recommendations. Validation of the isotopic composition (depletion) calculations is addressed in a companion paper at this conference. For criticality validation, the approach is to utilize (1) available laboratory critical experiment (LCE) data from the International Handbook of Evaluated Criticality Safety Benchmark Experiments and the French Haut Taux de Combustion (HTC) program to support validation of the principal actinides and (2) calculated sensitivities, nuclear data uncertainties, and the limited available fission product LCE data to predict and verify individual biases for relevant minor actinides and fission products. This paper (1) provides a detailed description of the approach and its technical bases, (2) describes the application of the approach for representative pressurized water reactor and boiling water reactor safety analysis models to demonstrate its usage and applicability, (3) provides reference bias results based on the prerelease SCALE 6.1 code package and ENDF/B-VII nuclear cross-section data, and (4) provides recommendations for application of the results and methods to other code and data packages.« less

Integrated System for Retrieval, Transportation and Consolidated Storage of Used Nuclear Fuel in the US - 13312

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bracey, William; Bondre, Jayant; Shelton, Catherine

2013-07-01

The current inventory of used nuclear fuel assemblies (UNFAs) from commercial reactor operations in the United States totals approximately 65,000 metric tons or approximately 232,000 UNFAs primarily stored at the 104 operational reactors in the US and a small number of decommissioned reactors. This inventory is growing at a rate of roughly 2,000 to 2,400 metric tons each year, (Approx. 7,000 UNFAs) as a result of ongoing commercial reactor operations. Assuming an average of 10 metric tons per storage/transportation casks, this inventory of commercial UNFAs represents about 6,500 casks with an additional of about 220 casks every year. In Januarymore » 2010, the Blue Ribbon Commission (BRC) [1] was directed to conduct a comprehensive review of policies for managing the back end of the nuclear fuel cycle and recommend a new plan. The BRC issued their final recommendations in January 2012. One of the main recommendations is for the United States to proceed promptly to develop one or more consolidated storage facilities (CSF) as part of an integrated, comprehensive plan for safely managing the back end of the nuclear fuel cycle. Based on its extensive experience in storage and transportation cask design, analysis, licensing, fabrication, and operations including transportation logistics, Transnuclear, Inc. (TN), an AREVA Subsidiary within the Logistics Business Unit, is engineering an integrated system that will address the complete process of commercial UNFA management. The system will deal with UNFAs in their current storage mode in various configurations, the preparation including handling and additional packaging where required and transportation of UNFAs to a CSF site, and subsequent storage, operation and maintenance at the CSF with eventual transportation to a future repository or recycling site. It is essential to proceed by steps to ensure that the system will be the most efficient and serve at best its purpose by defining: the problem to be resolved, the criteria to evaluate the solutions, and the alternative solutions. The complexity of the project is increasing with time (more fuel assemblies, new storage systems, deteriorating logistics infrastructure at some sites, etc.) but with the uncertainty on the final disposal path, flexibility and simplicity will be critical. (authors)« less

High burn-up spent nuclear fuel transport reliability investigation

DOE PAGES

Wang, Jy-An; Wang, Hong; Jiang, Hao; ...

2018-04-15

Transportation packages for spent nuclear fuel (SNF) must meet safety requirements under normal and accident conditions as specified by federal regulations. During road or rail transportation, SNF will experience unique conditions that could affect the structural integrity of the cladding due to vibrational and impact loading. Lack of SNF inertia-induced dynamic fatigue data, especially for the high burn-up (HBU) SNF systems, has brought significant challenges to quantify the reliability of SNF during transportation with a high degree of confidence. To address this shortcoming, Oak Ridge National Laboratory (ORNL) developed a SNF vibration testing protocol without fuel pellets removal, which hasmore » provided significant insight regarding the dynamics of mechanical interactions between pellet and cladding. This research has provided a detailed understanding about the effect of loading rate and loading mode on the fatigue damage evolution of HBU SNF under normal conditions of transport (NCT). Static and dynamic loading experimental data were generated for SNF under simulated transportation environments using a cyclic integrated reversible-bending fatigue tester (CIRFT), an enabling hot-cell testing technology developed at ORNL. SNF flexural tensile strength and fatigue S-N data from pressurized water reactors (PWRs) and boiling water reactor (BWR) HBU SNF are presented in this paper, including the potential effects of pellet-cladding interface bonding, hydride reorientation, and thermal annealing to SNF vibration reliability. The data presented here can be used to meet the nuclear industry and U.S. Nuclear Regulatory Commission needs in safety of SNF transportation operations.« less

High burn-up spent nuclear fuel transport reliability investigation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wang, Jy-An; Wang, Hong; Jiang, Hao

Transportation packages for spent nuclear fuel (SNF) must meet safety requirements under normal and accident conditions as specified by federal regulations. During road or rail transportation, SNF will experience unique conditions that could affect the structural integrity of the cladding due to vibrational and impact loading. Lack of SNF inertia-induced dynamic fatigue data, especially for the high burn-up (HBU) SNF systems, has brought significant challenges to quantify the reliability of SNF during transportation with a high degree of confidence. To address this shortcoming, Oak Ridge National Laboratory (ORNL) developed a SNF vibration testing protocol without fuel pellets removal, which hasmore » provided significant insight regarding the dynamics of mechanical interactions between pellet and cladding. This research has provided a detailed understanding about the effect of loading rate and loading mode on the fatigue damage evolution of HBU SNF under normal conditions of transport (NCT). Static and dynamic loading experimental data were generated for SNF under simulated transportation environments using a cyclic integrated reversible-bending fatigue tester (CIRFT), an enabling hot-cell testing technology developed at ORNL. SNF flexural tensile strength and fatigue S-N data from pressurized water reactors (PWRs) and boiling water reactor (BWR) HBU SNF are presented in this paper, including the potential effects of pellet-cladding interface bonding, hydride reorientation, and thermal annealing to SNF vibration reliability. The data presented here can be used to meet the nuclear industry and U.S. Nuclear Regulatory Commission needs in safety of SNF transportation operations.« less

Deterministic Local Sensitivity Analysis of Augmented Systems - II: Applications to the QUENCH-04 Experiment Using the RELAP5/MOD3.2 Code System

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ionescu-Bujor, Mihaela; Jin Xuezhou; Cacuci, Dan G.

2005-09-15

The adjoint sensitivity analysis procedure for augmented systems for application to the RELAP5/MOD3.2 code system is illustrated. Specifically, the adjoint sensitivity model corresponding to the heat structure models in RELAP5/MOD3.2 is derived and subsequently augmented to the two-fluid adjoint sensitivity model (ASM-REL/TF). The end product, called ASM-REL/TFH, comprises the complete adjoint sensitivity model for the coupled fluid dynamics/heat structure packages of the large-scale simulation code RELAP5/MOD3.2. The ASM-REL/TFH model is validated by computing sensitivities to the initial conditions for various time-dependent temperatures in the test bundle of the Quench-04 reactor safety experiment. This experiment simulates the reflooding with water ofmore » uncovered, degraded fuel rods, clad with material (Zircaloy-4) that has the same composition and size as that used in typical pressurized water reactors. The most important response for the Quench-04 experiment is the time evolution of the cladding temperature of heated fuel rods. The ASM-REL/TFH model is subsequently used to perform an illustrative sensitivity analysis of this and other time-dependent temperatures within the bundle. The results computed by using the augmented adjoint sensitivity system, ASM-REL/TFH, highlight the reliability, efficiency, and usefulness of the adjoint sensitivity analysis procedure for computing time-dependent sensitivities.« less

Analyses of the reflector tank, cold source, and beam tube cooling for ANS reactor. [Advanced Neutron Source (ANS)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Marland, S.

1992-07-01

This report describes my work as an intern with Martin Marietta Energy Systems, Inc., in the summer of 1991. I was assigned to the Reactor Technology Engineering Department, working on the Advanced Neutron Source (ANS). My first project was to select and analyze sealing systems for the top of the diverter/reflector tank. This involved investigating various metal seals and calculating the forces necessary to maintain an adequate seal. The force calculations led to an analysis of several bolt patterns and lockring concepts that could be used to maintain a seal on the vessel. Another project involved some pressure vessel stressmore » calculations and the calculation of the center of gravity for the cold source assembly. I also completed some sketches of possible cooling channel patterns for the inner vessel of the cold source. In addition, I worked on some thermal design analyses for the reflector tank and beam tubes, including heat transfer calculations and assisting in Patran and Pthermal analyses. To supplement the ANS work, I worked on other projects. I completed some stress/deflection analyses on several different beams. These analyses were done with the aid of CAASE, a beam-analysis software package. An additional project involved bending analysis on a carbon removal system. This study was done to find the deflection of a complex-shaped beam when loaded with a full waste can.« less

Impact of minor actinide recycling on sustainable fuel cycle options

DOE Office of Scientific and Technical Information (OSTI.GOV)

Heidet, F.; Kim, T. K.; Taiwo, T. A.

The recent Evaluation and Screening study chartered by the U.S. Department of Energy, Office of Nuclear Energy, has identified four fuel cycle options as being the most promising. Among these four options, the two single-stage fuel cycles rely on a fast reactor and are differing in the fact that in one case only uranium and plutonium are recycled while in the other case minor actinides are also recycled. The two other fuel cycles are two-stage and rely on both fast and thermal reactors. They also differ in the fact that in one case only uranium and plutonium are recycled whilemore » in the other case minor actinides are also recycled. The current study assesses the impact of recycling minor actinides on the reactor core design, its performance characteristics, and the characteristics of the recycled material and waste material. The recycling of minor actinides is found not to affect the reactor core performance, as long as the same cycle length, core layout and specific power are being used. One notable difference is that the required transuranics (TRU) content is slightly increased when minor actinides are recycled. The mass flows are mostly unchanged given a same specific power and cycle length. Although the material mass flows and reactor performance characteristics are hardly affected by recycling minor actinides, some differences are observed in the waste characteristics between the two fuel cycles considered. The absence of minor actinides in the waste results in a different buildup of decay products, and in somewhat different behaviors depending on the characteristic and time frame considered. Recycling of minor actinides is found to result in a reduction of the waste characteristics ranging from 10% to 90%. These results are consistent with previous studies in this domain and depending on the time frame considered, packaging conditions, repository site, repository strategy, the differences observed in the waste characteristics could be beneficial and help improve the repository performance. On the other hand, recycling minor actinides also results in an increase of the recycled fuel characteristics and therefore of the charged fuel. The radioactivity is slightly increased while the decay heat and radiotoxicities are very significantly increased. Despite these differences, the characteristics of the fuel at time of discharge remain similar whether minor actinides are recycled or not, with the exception of the inhalation radiotoxicity which is significantly larger with minor actinide recycling. After some cooling the characteristics of the discharged fuel become larger when minor actinides are recycled, potentially affecting the reprocessing plant requirements. Recycling minor actinides has a negative impact on the characteristics of the fresh fuel and will make it more challenging to fabricate fuel containing minor actinides.« less

Secure Retrieval of FFTF Testing, Design, and Operating Information

DOE Office of Scientific and Technical Information (OSTI.GOV)

Butner, R. Scott; Wootan, David W.; Omberg, Ronald P.

One of the goals of the Advanced Fuel Cycle Initiative (AFCI) is to preserve the knowledge that has been gained in the United States on Liquid Metal Reactors (LMR). In addition, preserving LMR information and knowledge is part of a larger international collaborative activity conducted under the auspices of the International Atomic Energy Agency (IAEA). A similar program is being conducted for EBR-II at the Idaho Nuclear Laboratory (INL) and international programs are also in progress. Knowledge preservation at the FFTF is focused on the areas of design, construction, startup, and operation of the reactor. As the primary function ofmore » the FFTF was testing, the focus is also on preserving information obtained from irradiation testing of fuels and materials. This information will be invaluable when, at a later date, international decisions are made to pursue new LMRs. In the interim, this information may be of potential use for international exchanges with other LMR programs around the world. At least as important in the United States, which is emphasizing large-scale computer simulation and modeling, this information provides the basis for creating benchmarks for validating and testing these large scale computer programs. Although the preservation activity with respect to FFTF information as discussed below is still underway, the team of authors above is currently retrieving and providing experimental and design information to the LMR modeling and simulation efforts for use in validating their computer models. On the Hanford Site, the FFTF reactor plant is one of the facilities intended for decontamination and decommissioning consistent with the cleanup mission on this site. The reactor facility has been deactivated and is being maintained in a cold and dark minimal surveillance and maintenance mode until final decommissioning is pursued. In order to ensure protection of information at risk, the program to date has focused on sequestering and secure retrieval. Accomplishments include secure retrieval of: more than 400 boxes of FFTF information, several hundred microfilm reels including Clinch River Breeder Reactor (CRBR) information, and 40 boxes of information on the Fuels and Materials Examination Facility (FMEF). All information preserved to date is now being stored and categorized consistent with the IAEA international standardized taxonomy. Earlier information largely related to irradiation testing is likewise being categorized. The fuel test results information exists in several different formats depending upon the final stage of the test evaluation. In some cases there is information from both non-destructive and destructive examination while in other cases only non-destructive results are available. Non-destructive information would include disassembly records, dimensional profilometry, gamma spectrometry, and neutron radiography. Information from destructive examinations would include fission gas analysis, metallography, and photomicrographs. Archiving of FFTF data, including both the reactor plant and the fuel test information, is being performed in coordination with other data archiving efforts underway under the aegis of the AFCI program. In addition to the FFTF efforts, archiving of data from the EBR-II reactor is being carried out by INL. All material at risk associated with FFTF documentation has been secured in a timely manner consistent with the stated plan. This documentation is now being categorized consistent with internationally agreed upon IAEA standards. Documents are being converted to electronic format for transfer to a large searchable electronic database being developed by INL. In addition, selected FFTF information is being used to generate test cases for large-scale simulation modeling efforts and for providing Design Data Need (DDN) packages as requested by the AFCI program.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Natesan, K.; Momozaki, Y.; Li, M.

This report gives a description of the activities in design, fabrication, construction, and assembling of a pumped sodium loop for the sodium compatibility studies on advanced structural materials. The work is the Argonne National Laboratory (ANL) portion of the effort on the work project entitled, 'Sodium Compatibility of Advanced Fast Reactor Materials,' and is a part of Advanced Materials Development within the Reactor Campaign. The objective of this project is to develop information on sodium corrosion compatibility of advanced materials being considered for sodium reactor applications. This report gives the status of the sodium pumped loop at Argonne National Laboratory,more » the specimen details, and the technical approach to evaluate the sodium compatibility of advanced structural alloys. This report is a deliverable from ANL in FY2010 (M2GAN10SF050302) under the work package G-AN10SF0503 'Sodium Compatibility of Advanced Fast Reactor Materials.' Two reports were issued in 2009 (Natesan and Meimei Li 2009, Natesan et al. 2009) which examined the thermodynamic and kinetic factors involved in the purity of liquid sodium coolant for sodium reactor applications as well as the design specifications for the ANL pumped loop for testing advanced structural materials. Available information was presented on solubility of several metallic and nonmetallic elements along with a discussion of the possible mechanisms for the accumulation of impurities in sodium. That report concluded that the solubility of many metals in sodium is low (<1 part per million) in the temperature range of interest in sodium reactors and such trace amounts would not impact the mechanical integrity of structural materials and components. The earlier report also analyzed the solubility and transport mechanisms of nonmetallic elements such as oxygen, nitrogen, carbon, and hydrogen in laboratory sodium loops and in reactor systems such as Experimental Breeder Reactor-II, Fast Flux Test Facility, and Clinch River Breeder Reactor. Among the nonmetallic elements discussed, oxygen is deemed controllable and its concentration in sodium can be maintained in sodium for long reactor life by using cold-trap method. It was concluded that among the cold-trap and getter-trap methods, the use of cold trap is sufficient to achieve oxygen concentration of the order of 1 part per million. Under these oxygen conditions in sodium, the corrosion performance of structural materials such as austenitic stainless steels and ferritic steels will be acceptable at a maximum core outlet sodium temperature of {approx}550 C. In the current sodium compatibility studies, the oxygen concentration in sodium will be controlled and maintained at {approx}1 ppm by controlling the cold trap temperature. The oxygen concentration in sodium in the forced convection sodium loop will be controlled and monitored by maintaining the cold trap temperature in the range of 120-150 C, which would result in oxygen concentration in the range of 1-2 ppm. Uniaxial tensile specimens are being exposed to flowing sodium and will be retrieved and analyzed for corrosion and post-exposure tensile properties. Advanced materials for sodium exposure include austenitic alloy HT-UPS and ferritic-martensitic steels modified 9Cr-1Mo and NF616. Among the nonmetallic elements in sodium, carbon was assessed to have the most influence on structural materials since carbon, as an impurity, is not amenable to control and maintenance by any of the simple purification methods. The dynamic equilibrium value for carbon in sodium systems is dependent on several factors, details of which were discussed in the earlier report. The current sodium compatibility studies will examine the role of carbon concentration in sodium on the carburization-decarburization of advanced structural materials at temperatures up to 650 C. Carbon will be added to the sodium by exposure of carbon-filled iron tubes, which over time will enable carbon to diffuse through iron and dissolve into sodium. The method enables addition of dissolved carbon (without carbon particulates) in sodium that is of interest for materials compatibility evaluation. The removal of carbon from the sodium will be accomplished by exposing carbon-gettering alloys such as refractory metals that have a high partitioning coefficient for carbon and also precipitate carbides, thereby decreasing the carbon concentration in sodium.« less

On the Numerical Analysis of Decay Rate Enhancement in Metallic Environment

NASA Astrophysics Data System (ADS)

Mehedinteanu, S.

2007-10-01

Motivated on the very recent experiments to determine the acceleration of the alpha decay of meta-stable radionuclides in metallic environment some work has been done to strengthten the importance in the process of electrons screening in metals. Thus, by combining the Gamow decay theory with electrostatic screening in Debye-Hückel approximation (jellium model) a formula for ``the shift'' in screening energy which enters in the decay enhancement factor expression that copes well with these experiments has been derived. It was established that to simulate the poly-atoms system containing decaying isotopes in QM&MD codes calculations, and to include ``the screening energy shift'' of protons, decay alpha, beta+ particles due to all surrounding interacting effects, it is sufficiently only to substitute the code ruly pseudo-potential input for hydrogen-like atoms (including alpha) by a screened Coulomb potential as from the well-known Gamow alpha decay theory. For demonstration is used the QM&MD code package which usually performs density-functional theory (DFT) total-energy calculations for materials ranging from insulators to transition metals. This package employs first-principles pseudo-potentials and a plane-wave basis-set, and it was used to do a special calculus for some metal environments (Pd) where protons-deuterons are implanted or when it is alloyed with a radionuclide-like isotopes (174Hf72), the results compare well with the existing experiments on the decay enhancement. These works give further arguments for a cheap solution to remove the transuranic waste (involving all alpha-decay) of used-up rods of fission reactors in a time period of a few years.

Unconventional Staging Package Selection Leads to Cost Savings

DOE Office of Scientific and Technical Information (OSTI.GOV)

,

2012-06-07

In late 2010, U.S. Department of Energy (DOE) Deputy Secretary of Energy, Daniel Poneman, directed that an analysis be conducted on the U-233 steel-clad, Zero Power Reactor (ZPR) fuel plates that were stored at Oak Ridge National Laboratory (ORNL), focusing on cost savings and any potential DOE programmatic needs for the special nuclear material (SNM). The NA-162 Nuclear Criticality Safety Program requested retention of these fuel plates for use in experiments at the Nevada National Security Site (NNSS). A Secretarial Initiative challenged ORNL to make the first shipment to the NNSS by the end of the 2011 calendar year, andmore » this effort became known as the U-233 Project Accelerated Shipping Campaign. To meet the Secretarial Initiative, National Security Technologies, LLC (NSTec), the NNSS Management and Operations contractor, was asked to facilitate the receipt and staging of the U-233 fuel plates in the Device Assembly Facility (DAF). Because there were insufficient staging containers available for the fuel plates, NSTec conducted an analysis of alternatives. The project required a staging method that would reduce the staging footprint while addressing nuclear criticality safety and radiation exposure concerns. To accommodate an intermediate staging method of approximately five years, the NSTec project team determined that a unique and unconventional staging package, the AT-400R, was available to meet the project requirements. By using the AT-400R containers, NSTec was able to realize a cost savings of approximately $10K per container, a total cost savings of nearly $450K.« less

NEET Enhanced Micro-Pocket Fission Detector for High Temperature Reactors - FY16 Status Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Unruh, Troy; Reichenberger, Michael; Stevenson, Sarah

2016-09-01

A collaboration between the Idaho National Laboratory (INL), the Kansas State University (KSU), and the French Atomic Energy Agency, Commissariat à l'Énergie Atomique et aux Energies Alternatives, (CEA), has been initiated by the Nuclear Energy Enabling Technologies (NEET) Advanced Sensors and Instrumentation (ASI) program for developing and testing High Temperature Micro-Pocket Fission Detectors (HT MPFD), which are compact fission chambers capable of simultaneously measuring thermal neutron flux, fast neutron flux and temperature within a single package for temperatures up to 800 °C. The MPFD technology utilizes a small, multi-purpose, robust, in-core fission chambers and thermocouple. As discussed within this report,more » the small size, variable sensitivity, and increased accuracy of the MPFD technology represent a revolutionary improvement over current methods used to support irradiations in US Material Test Reactors (MTRs). Previous research conducted through NEET ASI1-3 has shown that the MPFD technology could be made robust and was successfully tested in a reactor core. This new project will further the MPFD technology for higher temperature regimes and other reactor applications by developing a HT MPFD suitable for temperatures up to 800 °C. This report summarizes the research progress for year two of this three year project. Highlights from research accomplishments include: • Continuation of a joint collaboration between INL, KSU, and CEA. Note that CEA is participating at their own expense because of interest in this unique new sensor. • An updated parallel wire HT MPFD design was developed. • Program support for HT MPFD deployments was given to Accident Tolerant Fuels (ATF) and Advanced Gas-cooled Reactor (AGR) irradiation test programs. • Quality approved materials for HT MPFD construction were procured by irradiation test programs for upcoming deployments. • KSU improved and performed electrical contact and fissile material plating. • KSU delivered fissile HT MPFD parts to INL for final construction of HT MPFD prototype. • A prototype HT MPFD was constructed and analyzed at INL. • The HT MPFD has been modeled in MCNP to optimize the amount of fissile material deposition. • The HT MPFD has been modeled in MCNP to optimize the sensor location in the irradiation test. • The fissile material deposition is undergoing independent verifications. • Detector amplifier electronics have been revised and tested by KSU. • Several project meetings were held at INL and KSU to discuss the roles and responsibilities between INL, KSU, and CEA for development and deployment of the HT MPFDs. As documented in this report, FY16 funding has allowed the project to meet year two planned accomplishments to develop a HT MPFD. In addition, the accomplishments of this project have attracted independent funding from other Department of Energy Office of Nuclear Energy (DOE-NE) programs for MTR irradiations of the MPFD technology. These are significant opportunities for this NEET Enhanced Micro-Pocket Fission Detector for High Temperature Reactors project because the irradiation expense of these experiments could not be included in the original project scope.« less

How we shipped our flip and standard too

DOE Office of Scientific and Technical Information (OSTI.GOV)

Deigl, H.J.; Feltz, D.E.

1984-07-01

This paper highlights the planning and handling activities for the shipment of irradiated TRIGA fuel from Texas A and M University to the Argonne National Lab/West (ANL/West) reactor facility at Idaho Falls, Idaho. Attention is focused on the enormous time spent on the planning and preparations prior to the shipment. The actual handling time at the NSCR for three shipping packages containing a total 51 elements was only 4 days, but, the time spent in planning and preparation exceeded 16 months. The fuel was transferred for shipment without incident - and from a health physics standpoint the exercise went verymore » well. Whole body exposures and hand doses were minimal for such a large undertaking. ANL/West health physicists reported contamination of the lifting devices for the HFIR when they received the cask. These pieces were wipe tested and contamination was found to be less than 200 dpm. If they were contaminated we were extremely fortunate during handling not to contaminate our facility or personnel.« less

Chemical recycling of plastic wastes made from polyethylene (LDPE and HDPE) and polypropylene (PP).

PubMed

Achilias, D S; Roupakias, C; Megalokonomos, P; Lappas, A A; Antonakou, Epsilon V

2007-11-19

The recycling of either model polymers or waste products based on low-density polyethylene (LDPE), high-density polyethylene (HDPE) or polypropylene (PP) is examined using the dissolution/reprecipitation method, as well as pyrolysis. In the first technique, different solvents/non-solvents were examined at different weight percent amounts and temperatures using as raw material both model polymers and commercial products (packaging film, bags, pipes, food-retail outlets). The recovery of polymer in every case was greater than 90%. FT-IR spectra and tensile mechanical properties of the samples before and after recycling were measured. Furthermore, catalytic pyrolysis was carried out in a laboratory fixed bed reactor with an FCC catalyst using again model polymers and waste products as raw materials. Analysis of the derived gases and oils showed that pyrolysis gave a mainly aliphatic composition consisting of a series of hydrocarbons (alkanes and alkenes), with a great potential to be recycled back into the petrochemical industry as a feedstock for the production of new plastics or refined fuels.

Current drive with combined electron cyclotron wave and high harmonic fast wave in tokamak plasmas

NASA Astrophysics Data System (ADS)

Li, J. C.; Gong, X. Y.; Dong, J. Q.; Wang, J.; Zhang, N.; Zheng, P. W.; Yin, C. Y.

2016-12-01

The current driven by combined electron cyclotron wave (ECW) and high harmonic fast wave is investigated using the GENRAY/CQL3D package. It is shown that no significant synergetic current is found in a range of cases with a combined ECW and fast wave (FW). This result is consistent with a previous study [Harvey et al., in Proceedings of IAEA TCM on Fast Wave Current Drive in Reactor Scale Tokamaks (Synergy and Complimentarily with LHCD and ECRH), Arles, France, IAEA, Vienna, 1991]. However, a positive synergy effect does appear with the FW in the lower hybrid range of frequencies. This positive synergy effect can be explained using a picture of the electron distribution function induced by the ECW and a very high harmonic fast wave (helicon). The dependence of the synergy effect on the radial position of the power deposition, the wave power, the wave frequency, and the parallel refractive index is also analyzed, both numerically and physically.

Approach for validating actinide and fission product compositions for burnup credit criticality safety analyses

DOE PAGES

Radulescu, Georgeta; Gauld, Ian C.; Ilas, Germina; ...

2014-11-01

This paper describes a depletion code validation approach for criticality safety analysis using burnup credit for actinide and fission product nuclides in spent nuclear fuel (SNF) compositions. The technical basis for determining the uncertainties in the calculated nuclide concentrations is comparison of calculations to available measurements obtained from destructive radiochemical assay of SNF samples. Probability distributions developed for the uncertainties in the calculated nuclide concentrations were applied to the SNF compositions of a criticality safety analysis model by the use of a Monte Carlo uncertainty sampling method to determine bias and bias uncertainty in effective neutron multiplication factor. Application ofmore » the Monte Carlo uncertainty sampling approach is demonstrated for representative criticality safety analysis models of pressurized water reactor spent fuel pool storage racks and transportation packages using burnup-dependent nuclide concentrations calculated with SCALE 6.1 and the ENDF/B-VII nuclear data. Furthermore, the validation approach and results support a recent revision of the U.S. Nuclear Regulatory Commission Interim Staff Guidance 8.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bourke, Mark Andrew; Vogel, Sven C.; Voit, Stewart Lancaster

Tomographic imaging and diffraction measurements were performed on nine pellets; four UN/ U Si composite formulations (two enrichment levels), three pure U 3Si 5 reference formulations (two enrichment levels) and two reject pellets with visible flaws (to qualify the technique). The U-235 enrichments ranged from 0.2 to 8.8 wt.%. The nitride/silicide composites are candidate compositions for use as Accident Tolerant Fuel (ATF). The monophase U 3Si 5 material was included as a reference. Pellets from the same fabrication batches will be inserted in the Advanced Test Reactor at Idaho during 2016. The goal of the Advanced Non-destructive Fuel Examination workmore » package is the development and application of non-destructive neutron imaging and scattering techniques to ceramic and metallic nuclear fuels. Data reported in this report were collected in the LANSCE run cycle that started in September 2015 and ended in March 2016. Data analysis is ongoing; thus, this report provides a preliminary review of the measurements and provides an overview of the characterized samples.« less

Thermal Neutron Capture onto the Stable Tungsten Isotopes

NASA Astrophysics Data System (ADS)

Hurst, A. M.; Firestone, R. B.; Sleaford, B. W.; Summers, N. C.; Revay, Zs.; Szentmiklósi, L.; Belgya, T.; Basunia, M. S.; Capote, R.; Choi, H.; Dashdorj, D.; Escher, J.; Krticka, M.; Nichols, A.

2012-02-01

Thermal neutron-capture measurements of the stable tungsten isotopes have been carried out using the guided thermal-neutron beam at the Budapest Reactor. Prompt singles spectra were collected and analyzed using the HYPERMET γ-ray analysis software package for the compound tungsten systems 183W, 184W, and 187W, prepared from isotopically-enriched samples of 182W, 183W, and 186W, respectively. These new data provide both confirmation and new insights into the decay schemes and structure of the tungsten isotopes reported in the Evaluated Gamma-ray Activation File based upon previous elemental analysis. The experimental data have also been compared to Monte Carlo simulations of γ-ray emission following the thermal neutron-capture process using the statistical-decay code DICEBOX. Together, the experimental cross sections and modeledfeeding contribution from the quasi continuum, have been used to determine the total radiative thermal neutron-capture cross sections for the tungsten isotopes and provide improved decay-scheme information for the structural- and neutron-data libraries.

ADVANCED NUCLEAR FUEL CYCLE EFFECTS ON THE TREATMENT OF UNCERTAINTY IN THE LONG-TERM ASSESSMENT OF GEOLOGIC DISPOSAL SYSTEMS - EBS INPUT

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sutton, M; Blink, J A; Greenberg, H R

2012-04-25

The Used Fuel Disposition (UFD) Campaign within the Department of Energy's Office of Nuclear Energy (DOE-NE) Fuel Cycle Technology (FCT) program has been tasked with investigating the disposal of the nation's spent nuclear fuel (SNF) and high-level nuclear waste (HLW) for a range of potential waste forms and geologic environments. The planning, construction, and operation of a nuclear disposal facility is a long-term process that involves engineered barriers that are tailored to both the geologic environment and the waste forms being emplaced. The UFD Campaign is considering a range of fuel cycles that in turn produce a range of wastemore » forms. The UFD Campaign is also considering a range of geologic media. These ranges could be thought of as adding uncertainty to what the disposal facility design will ultimately be; however, it may be preferable to thinking about the ranges as adding flexibility to design of a disposal facility. For example, as the overall DOE-NE program and industrial actions result in the fuel cycles that will produce waste to be disposed, and the characteristics of those wastes become clear, the disposal program retains flexibility in both the choice of geologic environment and the specific repository design. Of course, other factors also play a major role, including local and State-level acceptance of the specific site that provides the geologic environment. In contrast, the Yucca Mountain Project (YMP) repository license application (LA) is based on waste forms from an open fuel cycle (PWR and BWR assemblies from an open fuel cycle). These waste forms were about 90% of the total waste, and they were the determining waste form in developing the engineered barrier system (EBS) design for the Yucca Mountain Repository design. About 10% of the repository capacity was reserved for waste from a full recycle fuel cycle in which some actinides were extracted for weapons use, and the remaining fission products and some minor actinides were encapsulated in borosilicate glass. Because the heat load of the glass was much less than the PWR and BWR assemblies, the glass waste form was able to be co-disposed with the open cycle waste, by interspersing glass waste packages among the spent fuel assembly waste packages. In addition, the Yucca Mountain repository was designed to include some research reactor spent fuel and naval reactor spent fuel, within the envelope that was set using the commercial reactor assemblies as the design basis waste form. This milestone report supports Sandia National Laboratory milestone M2FT-12SN0814052, and is intended to be a chapter in that milestone report. The independent technical review of this LLNL milestone was performed at LLNL and is documented in the electronic Information Management (IM) system at LLNL. The objective of this work is to investigate what aspects of quantifying, characterizing, and representing the uncertainty associated with the engineered barrier are affected by implementing different advanced nuclear fuel cycles (e.g., partitioning and transmutation scenarios) together with corresponding designs and thermal constraints.« less

Tritium pellet injector for the tokamak fusion test reactor

NASA Astrophysics Data System (ADS)

Gouge, M. J.; Baylor, L. R.; Combs, S. K.; Fisher, P. W.; Foust, C. R.; Milora, S. L.

The tritium pellet injector (TPI) for the Tokamak Fusion Test Reactor (TFTR) will provide a tritium pellet fueling capability with pellet speeds in the 1- to 3-km/s range for the TFTR deuterium-tritium (D-T) plasma phase. An existing deuterium pellet injector (DPI) was modified at Oak Ridge National Laboratory (ORNL) to provide a four-shot, tritium-compatible, pipe-gun configuration with three upgraded single-stage pneumatic guns and a two-stage light gas gun driver. The TPI was designed for frozen pellets ranging in size from 3 to 4 mm in diameter in arbitrarily programmable firing sequences at tritium pellet speeds up to approximately 1.5 km/s for the three single-stage drivers and 2.5 to 3 km/s for the two-stage driver. Injector operation is controlled by a programmable logic controller (PLC). The new pipe-gun injector assembly was installed in the modified DPI guard vacuum box, and modifications were also made to the internals of the DPI vacuum injection line, including a new pellet diagnostics package. Assembly of these modified parts with existing DPI components was then completed and the TPI was tested at ORNL with deuterium pellets. Results of the testing program at ORNL are described. The TPI has been installed and operated on TFTR in support of the FY-92 deuterium plasma run period. In 1993, the tritium pellet injector will be retrofitted with a D-T fuel manifold and tritium gloveboxes and integrated into TFTR tritium processing systems to provide full tritium pellet capability.

Inter-Disciplinary Collaboration in Support of the Post-Standby TREAT Mission

DOE Office of Scientific and Technical Information (OSTI.GOV)

DeHart, Mark; Baker, Benjamin; Ortensi, Javier

Although analysis methods have advanced significantly in the last two decades, high fidelity multi- physics methods for reactors systems have been under development for only a few years and are not presently mature nor deployed. Furthermore, very few methods provide the ability to simulate rapid transients in three dimensions. Data for validation of advanced time-dependent multi- physics is sparse; at TREAT, historical data were not collected for the purpose of validating three-dimensional methods, let alone multi-physics simulations. Existing data continues to be collected to attempt to simulate the behavior of experiments and calibration transients, but it will be insufficient formore » the complete validation of analysis methods used for TREAT transient simulations. Hence, a 2018 restart will most likely occur without the direct application of advanced modeling and simulation methods. At present, the current INL modeling and simulation team plans to work with TREAT operations staff in performing reactor simulations with MAMMOTH, in parallel with the software packages currently being used in preparation for core restart (e.g., MCNP5, RELAP5, ABAQUS). The TREAT team has also requested specific measurements to be performed during startup testing, currently scheduled to run from February to August of 2018. These startup measurements will be crucial in validating the new analysis methods in preparation for ultimate application for TREAT operations and experiment design. This document describes the collaboration between modeling and simulation staff and restart, operations, instrumentation and experiment development teams to be able to effectively interact and achieve successful validation work during restart testing.« less

Scalable Methods for Uncertainty Quantification, Data Assimilation and Target Accuracy Assessment for Multi-Physics Advanced Simulation of Light Water Reactors

NASA Astrophysics Data System (ADS)

Khuwaileh, Bassam

High fidelity simulation of nuclear reactors entails large scale applications characterized with high dimensionality and tremendous complexity where various physics models are integrated in the form of coupled models (e.g. neutronic with thermal-hydraulic feedback). Each of the coupled modules represents a high fidelity formulation of the first principles governing the physics of interest. Therefore, new developments in high fidelity multi-physics simulation and the corresponding sensitivity/uncertainty quantification analysis are paramount to the development and competitiveness of reactors achieved through enhanced understanding of the design and safety margins. Accordingly, this dissertation introduces efficient and scalable algorithms for performing efficient Uncertainty Quantification (UQ), Data Assimilation (DA) and Target Accuracy Assessment (TAA) for large scale, multi-physics reactor design and safety problems. This dissertation builds upon previous efforts for adaptive core simulation and reduced order modeling algorithms and extends these efforts towards coupled multi-physics models with feedback. The core idea is to recast the reactor physics analysis in terms of reduced order models. This can be achieved via identifying the important/influential degrees of freedom (DoF) via the subspace analysis, such that the required analysis can be recast by considering the important DoF only. In this dissertation, efficient algorithms for lower dimensional subspace construction have been developed for single physics and multi-physics applications with feedback. Then the reduced subspace is used to solve realistic, large scale forward (UQ) and inverse problems (DA and TAA). Once the elite set of DoF is determined, the uncertainty/sensitivity/target accuracy assessment and data assimilation analysis can be performed accurately and efficiently for large scale, high dimensional multi-physics nuclear engineering applications. Hence, in this work a Karhunen-Loeve (KL) based algorithm previously developed to quantify the uncertainty for single physics models is extended for large scale multi-physics coupled problems with feedback effect. Moreover, a non-linear surrogate based UQ approach is developed, used and compared to performance of the KL approach and brute force Monte Carlo (MC) approach. On the other hand, an efficient Data Assimilation (DA) algorithm is developed to assess information about model's parameters: nuclear data cross-sections and thermal-hydraulics parameters. Two improvements are introduced in order to perform DA on the high dimensional problems. First, a goal-oriented surrogate model can be used to replace the original models in the depletion sequence (MPACT -- COBRA-TF - ORIGEN). Second, approximating the complex and high dimensional solution space with a lower dimensional subspace makes the sampling process necessary for DA possible for high dimensional problems. Moreover, safety analysis and design optimization depend on the accurate prediction of various reactor attributes. Predictions can be enhanced by reducing the uncertainty associated with the attributes of interest. Accordingly, an inverse problem can be defined and solved to assess the contributions from sources of uncertainty; and experimental effort can be subsequently directed to further improve the uncertainty associated with these sources. In this dissertation a subspace-based gradient-free and nonlinear algorithm for inverse uncertainty quantification namely the Target Accuracy Assessment (TAA) has been developed and tested. The ideas proposed in this dissertation were first validated using lattice physics applications simulated using SCALE6.1 package (Pressurized Water Reactor (PWR) and Boiling Water Reactor (BWR) lattice models). Ultimately, the algorithms proposed her were applied to perform UQ and DA for assembly level (CASL progression problem number 6) and core wide problems representing Watts Bar Nuclear 1 (WBN1) for cycle 1 of depletion (CASL Progression Problem Number 9) modeled via simulated using VERA-CS which consists of several multi-physics coupled models. The analysis and algorithms developed in this dissertation were encoded and implemented in a newly developed tool kit algorithms for Reduced Order Modeling based Uncertainty/Sensitivity Estimator (ROMUSE).

Foundations of Nuclear Geophysics

NASA Astrophysics Data System (ADS)

Herndon, J. M.; Hollenbach, D. F.

2002-05-01

Herndon suggested that the inner core of the Earth consists, not of partially crystallized iron metal, but of nickel silicide. He has shown by fundamental mass ratios that i) the Earth as a whole, especially the inner 82%, has a state of oxidation like primitive enstatite chondrites, and ii) the lower mantle and core are similar in composition to the Abee enstatite chondrite. By analogy with Abee data, CaS and MgS precipitates from the core are expected to collect at the core-mantle boundary and, significantly, a major fraction of the actinides are expected to precipitate from the core and to collect at the center of the Earth. Herndon demonstrated the feasibility of a nuclear fission reactor at the center of the Earth as the energy source for the geomagnetic field and described a natural mechanism that would lead to variations in energy production and thus variations in the geomagnetic field. Hollenbach and Herndon produced numerical simulations of the operation of the geo-reactor over the lifetime of the Earth using the state-of-the-art, validated, industry standard SCALE code package developed at Oak Ridge National Laboratory. The results clearly demonstrate that such a geo-reactor would i) function as a fast-neutron breeder reactor; ii) under appropriate conditions, operate over the entire period of geologic time; iii) function in such a manner as to yield variable and/or intermittent output; iv) generate energy at levels in the range generally accepted by the geophysics community; and, v) produce He-3 and He-4 in ratios that are in the range observed from deep-mantle sources. Deep-source He-3, the authors submit, is evidence of in-core sustained nuclear fission, rather than the out-gassing of primordial He-3; which in turn is evidence of large amounts of uranium residing in the Earth's core; which in turn is evidence that the core has a state of oxidation like the corresponding matter in primitive enstatite chondrites. The factors affecting He-3/He-4 ratios, their causes and implications, will be discussed in the presentation. Also, the current state of investigations into additional deep-Earth nuclear fission signatures will be presented. References: J. M. Herndon, Proc. R. Roc. London, Ser. A, 368 (1979) 495; J. Geomagn. Geoelectr. 45 (1993) 423; Proc. R. Soc. London, Ser. A, 445 (1994) 453; Proc. Nat. Acad. Sci. (USA) 93 (1996) 646. Hollenbach, D. F. and J. M. Herndon, Proc. Nat. Acad. Sci. (USA) 98 (2001) 11085.

PM-1 NUCLEAR POWER PLANT PROGRAM. Quarterly Progress Report No. 2 for June 1 to August 31, 1959

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sieg, J.S.; Smith, E.H.

1959-10-01

The objective of the contract is the design, development, fabrication, installation, and initial testing and operation of a prepackaged air- transportable pressurized water reactor nuclear power plant, the PM-1. The specified output is 1 Mwe and 7 million Btu/hr of heat. The plant is to be operational by March 1962. The principal efforts were completion of the plant parametric study and preparation of the preliminary design. A summary of design parameters is given. Systems development work included study and selection of packages for full-scale testing, a survey of in-core instrumentation techniques, control and instrumentation development, and development of components formore » the steam generator, condenser, and turbine generator, which are not commercially available. Reactor development work included completion of the parametric zeropower experiments and preparrtions for a flexible zeropower test program, a revision of plans for irradiation testing PM-1 fuel elements, initiation of a reactor flow test program, outliring of a heat tnansfer test program, completion of the seven-tube test section (SETCH-1) tests, and evaluation of control rod actuators leading to specification of a magnetic jack-type control rod drive similar to that reported in ANL-5768. Completion of the prelimirary design led to initiation of the final design effort, which will be the principal activity during the next two project quarters. Preparations for core fabrication included procurement of core cladding material for the zero-power teat core, arrangement with a subcontractor to convent UF/sub 6/ to UO/sub 2/ and to commence delivery of the oxide during the next quarter, development of fuel element fabrication and ultrasonic testing techniques, study of control rod materials, UO/sub 2/ recovery techniques, and boron analysis methods. Preliminary work on site preparation was pursued with receipt of USAEC approval for a location on the eastern slope of Warren Peak at Sundance, Wyoming. A survey of this site is underway. A preliminary Hazards Summary Report is in preparation. (For preceding period see MND-M-1812.) (auth)« less

49 CFR 173.24a - Additional general requirements for non-bulk packagings and packages.

Code of Federal Regulations, 2010 CFR

2010-10-01

... subchapter. (b) Non-bulk packaging filling limits. (1) A single or composite non-bulk packaging may be filled... gross mass marked on the packaging. (3) A single or composite non-bulk packaging which is tested and... marked on the packaging, or 1.2 if not marked. In addition: (i) A single or composite non-bulk packaging...

49 CFR 173.24a - Additional general requirements for non-bulk packagings and packages.

Code of Federal Regulations, 2012 CFR

2012-10-01

... subchapter. (b) Non-bulk packaging filling limits. (1) A single or composite non-bulk packaging may be filled... gross mass marked on the packaging. (3) A single or composite non-bulk packaging which is tested and... marked on the packaging, or 1.2 if not marked. In addition: (i) A single or composite non-bulk packaging...

49 CFR 173.24a - Additional general requirements for non-bulk packagings and packages.

Code of Federal Regulations, 2013 CFR

2013-10-01

... subchapter. (b) Non-bulk packaging filling limits. (1) A single or composite non-bulk packaging may be filled... gross mass marked on the packaging. (3) A single or composite non-bulk packaging which is tested and... marked on the packaging, or 1.2 if not marked. In addition: (i) A single or composite non-bulk packaging...

Generic repository design concepts and thermal analysis (FY11).

DOE Office of Scientific and Technical Information (OSTI.GOV)

Howard, Robert; Dupont, Mark; Blink, James A.

2011-08-01

Reference concepts for geologic disposal of used nuclear fuel and high-level radioactive waste in the U.S. are developed, including geologic settings and engineered barriers. Repository thermal analysis is demonstrated for a range of waste types from projected future, advanced nuclear fuel cycles. The results show significant differences among geologic media considered (clay/shale, crystalline rock, salt), and also that waste package size and waste loading must be limited to meet targeted maximum temperature values. In this study, the UFD R&D Campaign has developed a set of reference geologic disposal concepts for a range of waste types that could potentially be generatedmore » in advanced nuclear FCs. A disposal concept consists of three components: waste inventory, geologic setting, and concept of operations. Mature repository concepts have been developed in other countries for disposal of spent LWR fuel and HLW from reprocessing UNF, and these serve as starting points for developing this set. Additional design details and EBS concepts will be considered as the reference disposal concepts evolve. The waste inventory considered in this study includes: (1) direct disposal of SNF from the LWR fleet, including Gen III+ advanced LWRs being developed through the Nuclear Power 2010 Program, operating in a once-through cycle; (2) waste generated from reprocessing of LWR UOX UNF to recover U and Pu, and subsequent direct disposal of used Pu-MOX fuel (also used in LWRs) in a modified-open cycle; and (3) waste generated by continuous recycling of metal fuel from fast reactors operating in a TRU burner configuration, with additional TRU material input supplied from reprocessing of LWR UOX fuel. The geologic setting provides the natural barriers, and establishes the boundary conditions for performance of engineered barriers. The composition and physical properties of the host medium dictate design and construction approaches, and determine hydrologic and thermal responses of the disposal system. Clay/shale, salt, and crystalline rock media are selected as the basis for reference mined geologic disposal concepts in this study, consistent with advanced international repository programs, and previous investigations in the U.S. The U.S. pursued deep geologic disposal programs in crystalline rock, shale, salt, and volcanic rock in the years leading up to the Nuclear Waste Policy Act, or NWPA (Rechard et al. 2011). The 1987 NWPA amendment act focused the U.S. program on unsaturated, volcanic rock at the Yucca Mountain site, culminating in the 2008 license application. Additional work on unsaturated, crystalline rock settings (e.g., volcanic tuff) is not required to support this generic study. Reference disposal concepts are selected for the media listed above and for deep borehole disposal, drawing from recent work in the U.S. and internationally. The main features of the repository concepts are discussed in Section 4.5 and summarized in Table ES-1. Temperature histories at the waste package surface and a specified distance into the host rock are calculated for combinations of waste types and reference disposal concepts, specifying waste package emplacement modes. Target maximum waste package surface temperatures are identified, enabling a sensitivity study to inform the tradeoff between the quantity of waste per disposal package, and decay storage duration, with respect to peak temperature at the waste package surface. For surface storage duration on the order of 100 years or less, waste package sizes for direct disposal of SNF are effectively limited to 4-PWR configurations (or equivalent size and output). Thermal results are summarized, along with recommendations for follow-on work including adding additional reference concepts, verification and uncertainty analysis for thermal calculations, developing descriptions of surface facilities and other system details, and cost estimation to support system-level evaluations.« less

Detecting small holes in packages

DOEpatents

Kronberg, James W.; Cadieux, James R.

1996-01-01

A package containing a tracer gas, and a method for determining the presence of a hole in the package by sensing the presence of the gas outside the package. The preferred tracer gas, especially for food packaging, is sulfur hexafluoride. A quantity of the gas is added to the package and the package is closed. The concentration of the gas in the atmosphere outside the package is measured and compared to a predetermined value of the concentration of the gas in the absence of the package. A measured concentration greater than the predetermined value indicates the presence of a hole in the package. Measuring may be done in a chamber having a lower pressure than that in the package.

Green Packaging Management of Logistics Enterprises

NASA Astrophysics Data System (ADS)

Zhang, Guirong; Zhao, Zongjian

From the connotation of green logistics management, we discuss the principles of green packaging, and from the two levels of government and enterprises, we put forward a specific management strategy. The management of green packaging can be directly and indirectly promoted by laws, regulations, taxation, institutional and other measures. The government can also promote new investment to the development of green packaging materials, and establish specialized institutions to identify new packaging materials, standardization of packaging must also be accomplished through the power of the government. Business units of large scale through the packaging and container-based to reduce the use of packaging materials, develop and use green packaging materials and easy recycling packaging materials for proper packaging.

Assessment of Quality Assurance Measures for Radioactive Material Transport Packages not Requiring Competent Authority Design Approval - 13282

DOE Office of Scientific and Technical Information (OSTI.GOV)

Komann, Steffen; Groeke, Carsten; Droste, Bernhard

The majority of transports of radioactive materials are carried out in packages which don't need a package design approval by a competent authority. Low-active radioactive materials are transported in such packages e.g. in the medical and pharmaceutical industry and in the nuclear industry as well. Decommissioning of NPP's leads to a strong demand for packages to transport low and middle active radioactive waste. According to IAEA regulations the 'non-competent authority approved package types' are the Excepted Packages and the Industrial Packages of Type IP-1, IP-2 and IP-3 and packages of Type A. For these types of packages an assessment bymore » the competent authority is required for the quality assurance measures for the design, manufacture, testing, documentation, use, maintenance and inspection (IAEA SSR 6, Chap. 306). In general a compliance audit of the manufacturer of the packaging is required during this assessment procedure. Their regulatory level in the IAEA regulations is not comparable with the 'regulatory density' for packages requiring competent authority package design approval. Practices in different countries lead to different approaches within the assessment of the quality assurance measures in the management system as well as in the quality assurance program of a special package design. To use the package or packaging in a safe manner and in compliance with the regulations a management system for each phase of the life of the package or packaging is necessary. The relevant IAEA-SSR6 chap. 801 requires documentary verification by the consignor concerning package compliance with the requirements. (authors)« less

Smart packaging systems for food applications: a review.

PubMed

Biji, K B; Ravishankar, C N; Mohan, C O; Srinivasa Gopal, T K

2015-10-01

Changes in consumer preference for safe food have led to innovations in packaging technologies. This article reviews about different smart packaging systems and their applications in food packaging, packaging research with latest innovations. Active and intelligent packing are such packaging technologies which offer to deliver safer and quality products. Active packaging refers to the incorporation of additives into the package with the aim of maintaining or extending the product quality and shelf life. The intelligent systems are those that monitor the condition of packaged food to give information regarding the quality of the packaged food during transportation and storage. These technologies are designed to the increasing demand for safer foods with better shelf life. The market for active and intelligent packaging systems is expected to have a promising future by their integration into packaging materials or systems.

Detecting small holes in packages

DOEpatents

Kronberg, J.W.; Cadieux, J.R.

1996-03-19

A package containing a tracer gas, and a method for determining the presence of a hole in the package by sensing the presence of the gas outside the package are disclosed. The preferred tracer gas, especially for food packaging, is sulfur hexafluoride. A quantity of the gas is added to the package and the package is closed. The concentration of the gas in the atmosphere outside the package is measured and compared to a predetermined value of the concentration of the gas in the absence of the package. A measured concentration greater than the predetermined value indicates the presence of a hole in the package. Measuring may be done in a chamber having a lower pressure than that in the package. 3 figs.

A Review of Patents for the Smart Packaging of Meat and Muscle-based Food Products.

PubMed

Holman, Benjamin; Kerry, Joseph P; Hopkins, David L

2017-10-31

Meat packaging once acted primarily as an inert barrier to protect its contents against contamination and this function has shifted. Packaging now includes complementary functions that improve product quality, longevity and customer/retail appeal. The devices and methods applied to achieve these functions may be categorised as smart packaging, which includes intelligent packaging, devised to monitor and communicate packaged content status, and active packaging, to provide passive adjustment of in-pack conditions from its interactions with the packaged meat. Smart packaging examples already available from recent patents include antimicrobial and antioxidant packaging coatings and inserts; sensors or indicators that identify spoilage and freshness; functional engineering customisations; improvements to packaging integrity; leak or tamper detectors; and, environmentally sustainable options. Together, these inventions respond to industry and customer demands for meat packaging and are therefore the focus of this review, in which we discuss their applications and limitations in meat packaging. Copyright© Bentham Science Publishers; For any queries, please email at epub@benthamscience.org.

Active and intelligent packaging systems for a modern society.

PubMed

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.

Packaging and Embedded Electronics for the Next Generation

NASA Technical Reports Server (NTRS)

Sampson, Michael J.

2010-01-01

This viewgraph presentation describes examples of electronic packaging that protects an electronic element from handling, contamination, shock, vibration and light penetration. The use of Hermetic and non-hermetic packaging is also discussed. The topics include: 1) What is Electronic Packaging? 2) Why Package Electronic Parts? 3) Evolution of Packaging; 4) General Packaging Discussion; 5) Advanced non-hermetic packages; 6) Discussion of Hermeticity; 7) The Class Y Concept and Possible Extensions; 8) Embedded Technologies; and 9) NEPP Activities.

Staged Catalytic Partial Oxidation (SCPO) System - The State of Art Integrated Short Contact Time Hydrogen Generator

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ke Liu; Jin Ki Hong; Wei Wei

Research and development on hydrogen and syngas production have great potential in addressing the following challenges in energy arena: (1) produce more clean fuels to meet the increasing demands for clean liquid and gaseous fuels for transportation and electricity generation, (2) increase the efficiency of energy utilization for fuels and electricity production, and (3) eliminate the pollutants and decouple the link between energy utilization and greenhouse gas emissions in end-use systems [Song, 2006, Liu, Song & Subramani 2009]. In this project, GE Global Research (GEGR) collaborated with Argonne National Laboratory (ANL) and the University of Minnesota (UoMn), developed and demonstratedmore » a low cost, compact staged catalytic partial oxidation (SCPO) technology for distributed hydrogen generation. GEGR analyzed different reforming system designs, and developed the SCPO reforming system which is a unique technology staging and integrating 3 different short contact time catalysts in a single, compact reactor: catalytic partial oxidation (CPO), steam methane reforming (SMR) and water-gas shift (WGS). This integration is demonstrated via the fabrication of a prototype scale unit of each key technology. Approaches for key technical challenges of the program includes: · Analyzed different system designs · Designed the SCPO hydrogen production system · Developed highly active and sulfur tolerant CPO catalysts · Designed and built different pilot-scale reactors to demonstrate each key technology · Evaluated different operating conditions · Quantified the efficiency and cost of the system · Developed process design package (PDP) for 1500 kg H2/day distributed H2 production unit. SCPO met the Department of Energy (DOE) and GE’s cost and efficiency targets for distributed hydrogen production.« less

New Multi-group Transport Neutronics (PHISICS) Capabilities for RELAP5-3D and its Application to Phase I of the OECD/NEA MHTGR-350 MW Benchmark

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gerhard Strydom; Cristian Rabiti; Andrea Alfonsi

2012-10-01

PHISICS is a neutronics code system currently under development at the Idaho National Laboratory (INL). Its goal is to provide state of the art simulation capability to reactor designers. The different modules for PHISICS currently under development are a nodal and semi-structured transport core solver (INSTANT), a depletion module (MRTAU) and a cross section interpolation (MIXER) module. The INSTANT module is the most developed of the mentioned above. Basic functionalities are ready to use, but the code is still in continuous development to extend its capabilities. This paper reports on the effort of coupling the nodal kinetics code package PHISICSmore » (INSTANT/MRTAU/MIXER) to the thermal hydraulics system code RELAP5-3D, to enable full core and system modeling. This will enable the possibility to model coupled (thermal-hydraulics and neutronics) problems with more options for 3D neutron kinetics, compared to the existing diffusion theory neutron kinetics module in RELAP5-3D (NESTLE). In the second part of the paper, an overview of the OECD/NEA MHTGR-350 MW benchmark is given. This benchmark has been approved by the OECD, and is based on the General Atomics 350 MW Modular High Temperature Gas Reactor (MHTGR) design. The benchmark includes coupled neutronics thermal hydraulics exercises that require more capabilities than RELAP5-3D with NESTLE offers. Therefore, the MHTGR benchmark makes extensive use of the new PHISICS/RELAP5-3D coupling capabilities. The paper presents the preliminary results of the three steady state exercises specified in Phase I of the benchmark using PHISICS/RELAP5-3D.« less

Investigation of Bio-Regenerative Life Support and Trash-To-Gas Experiment on a 4 Month Mars Simulation Mission

NASA Technical Reports Server (NTRS)

Caraccio, Anne; Poulet, Lucie; Hintze, Paul E.; Miles, John D.

2014-01-01

Future crewed missions to other planets or deep space locations will require regenerative Life Support Systems (LSS) as well as recycling processes for mission waste. Constant resupply of many commodity materials will not be a sustainable option for deep space missions, nor will storing trash on board a vehicle or at a lunar or Martian outpost. The habitable volume will decline as the volume of waste increases. A complete regenerative environmentally controlled life support system (ECLSS) on an extra-terrestrial outpost will likely include physico-chemical and biological technologies, such as bioreactors and greenhouse modules. Physico-chemical LSS do not enable food production and bio-regenerative LSS are not stable enough to be used alone in space. Mission waste that cannot be recycled into the bio-regenerative ECLSS can include excess food, food packaging, clothing, tape, urine and fecal waste. This waste will be sent to a system for converting the trash into the high value products. Two crew members on a 120 day Mars analog simulation, in collaboration with Kennedy Space Centers (KSC) Trash to Gas (TtG) project investigated a semi-closed loop system that treated non-edible biomass and other logistical waste for volume reduction and conversion into useful commodities. The purposes of this study are to show the how plant growth affects the amount of resources required by the habitat and how spent plant material can be recycled. Real-time data was sent to the reactor at KSC in Florida for replicating the analog mission waste for laboratory operation. This paper discusses the 120 day mission plant growth activity, logistical and plant waste management, power and water consumption effects of the plant and logistical waste, and potential energy conversion techniques using KSCs TtG reactor technology.

Drying results of K-Basin fuel element 1990 (Run 1)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Marschman, S.C.; Abrefah, J.; Klinger, G.S.

1998-06-01

The water-filled K-Basins in the Hanford 100-Area have been used to store N-Reactor spent nuclear fuel (SNF) since the 1970s. Because some leaks in the basins have been detected and some of the fuel is breached due to handling damage and corrosion, efforts are underway to remove the fuel elements from wet storage. An Integrated Process Strategy (IPS) has been developed to package, dry, transport, and store these metallic uranium fuels in an interim storage facility on the Hanford Site (WHC 1995). Information required to support the development of the drying processes, and the required safety analyses, is being obtainedmore » from characterization tests conducted on fuel elements removed from the K-Basins. A series of whole element drying tests (reported in separate documents, see Section 8.0) have been conducted by Pacific Northwest National Laboratory (PNNL) on several intact and damaged fuel elements recovered from both the K-East and K-West Basins. This report documents the results of the first of those tests (Run 1), which was conducted on an N-Reactor inner fuel element (1990) that had been stored underwater in the K-West Basin (see Section 2.0). This fuel element was subjected to a combination of low- and high-temperature vacuum drying treatments that were intended to mimic, wherever possible, the fuel treatment strategies of the IPS. The testing was conducted in the Whole Element Furnace Testing System, described in Section 3.0, located in the Postirradiation Testing Laboratory (PTL, 327 Building). The test conditions and methodology are given in Section 4.0, and the experimental results provided in Section 5.0. These results are further discussed in Section 6.0.« less

Energy efficiency in industrial mixing and cooling of non-Newtonian fluid in a stirred tank reactor

NASA Astrophysics Data System (ADS)

Baghli, Houda; Benyettou, Mohamed; Tchouar, Noureddine; Merah, Abdelkrim; Djafri, Mohammed

2018-05-01

This paper study the energy efficiency of the mixing and cooling of a non-Newtonian fluid manufactured on an industrial scale in a stirred tank reactor equipped with jacketed cooling side. The purpose of this study is to optimize the heat transfer to degrease the cooling time and recommend a technologic innovation to realize this purpose without altering the quality of this product. First the different production processes are analyzed. The decrease of the shear stress with time indicates that this fluid is non-Newtonian and has to be characterized. The rheological behavior of this fluid is determined by a series of viscosimetric measurements, at different shear rates (30 to 400 s-1), and at different temperatures in the range (20° C to 80 °C), representing the stress and temperature conditions recorded during production, storage and packaging cycles of this product. Experimental results show that the nature of the fluid is pseudo-plastic with flow behavior index n<1 and follow the power law model, with the influence of temperature on flow consistency index K. A thermo-dependent model is given to express this rheological parameters and viscosity of this fluid as a function of temperature, valid for the fluid temperature between 20 to 80 °C. This rheological model is used to achieve the heat transfer simulation in the industrial stirred tank with an anchor impeller mixing. Simulation results shows that the cooling time by mixing can be the quarter by reducing the stirring speed to 125 rpm, and decreasing the coolant temperature to 20°C and therefore reduce energy consumption. A technologic integration of a natural cooling thermo-siphon devise outside the process is proposed to afford a cooling fluid below 20°C.

Results of a nuclear power plant application of A New Technique for Human Error Analysis (ATHEANA)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Whitehead, D.W.; Forester, J.A.; Bley, D.C.

1998-03-01

A new method to analyze human errors has been demonstrated at a pressurized water reactor (PWR) nuclear power plant. This was the first application of the new method referred to as A Technique for Human Error Analysis (ATHEANA). The main goals of the demonstration were to test the ATHEANA process as described in the frame-of-reference manual and the implementation guideline, test a training package developed for the method, test the hypothesis that plant operators and trainers have significant insight into the error-forcing-contexts (EFCs) that can make unsafe actions (UAs) more likely, and to identify ways to improve the method andmore » its documentation. A set of criteria to evaluate the success of the ATHEANA method as used in the demonstration was identified. A human reliability analysis (HRA) team was formed that consisted of an expert in probabilistic risk assessment (PRA) with some background in HRA (not ATHEANA) and four personnel from the nuclear power plant. Personnel from the plant included two individuals from their PRA staff and two individuals from their training staff. Both individuals from training are currently licensed operators and one of them was a senior reactor operator on shift until a few months before the demonstration. The demonstration was conducted over a 5-month period and was observed by members of the Nuclear Regulatory Commission`s ATHEANA development team, who also served as consultants to the HRA team when necessary. Example results of the demonstration to date, including identified human failure events (HFEs), UAs, and EFCs are discussed. Also addressed is how simulator exercises are used in the ATHEANA demonstration project.« less

Hydrothermal Testing of K Basin Sludge and N Reactor Fuel at Sludge Treatment Project Operating Conditions

DOE Office of Scientific and Technical Information (OSTI.GOV)

Delegard, Calvin H.; Schmidt, Andrew J.; Thornton, Brenda M.

The Sludge Treatment Project (STP), managed for the U. S. DOE by Fluor Hanford (FH), was created to design and operate a process to eliminate uranium metal from K Basin sludge prior to packaging for Waste Isolation Pilot Plant (WIPP). The STP process uses high temperature liquid water to accelerate the reaction, produce uranium dioxide from the uranium metal, and safely discharge the hydrogen. Under nominal process conditions, the sludge will be heated in pressurized water at 185°C for as long as 72 hours to assure the complete reaction (corrosion) of up to 0.25-inch diameter uranium metal pieces. Under contractmore » to FH, the Pacific Northwest National Laboratory (PNNL) conducted bench-scale testing of the STP hydrothermal process in November and December 2006. Five tests (~50 ml each) were conducted in sealed, un-agitated reaction vessels under the hydrothermal conditions (e.g., 7 to 72 h at 185°C) of the STP corrosion process using radioactive sludge samples collected from the K East Basin and particles/coupons of N Reactor fuel also taken from the K Basins. The tests were designed to evaluate and understand the chemical changes that may be occurring and the effects that any changes would have on sludge rheological properties. The tests were not designed to evaluate engineering aspects of the process. The hydrothermal treatment affected the chemical and physical properties of the sludge. In each test, significant uranium compound phase changes were identified, resulting from dehydration and chemical reduction reactions. Physical properties of the sludge were significantly altered from their initial, as-settled sludge values, including, shear strength, settled density, weight percent water, and gas retention.« less

Packaging Your Training Materials

ERIC Educational Resources Information Center

Espeland, Pamela

1977-01-01

The types of packaging and packaging materials to use for training materials should be determined during the planning of the training programs, according to the packaging market. Five steps to follow in shopping for packaging are presented, along with a list of packaging manufacturers. (MF)

Current topics in active and intelligent food packaging for preservation of fresh foods.

PubMed

Lee, Seung Yuan; Lee, Seung Jae; Choi, Dong Soo; Hur, Sun Jin

2015-11-01

The purpose of this review is to provide an overview of current packaging systems, e.g. active packaging and intelligent packaging, for various foods. Active packaging, such as modified atmosphere packaging (MAP), extends the shelf life of fresh produce, provides a high-quality product, reduces economic losses, including those caused by delay of ripening, and improves appearance. However, in active packaging, several variables must be considered, such as temperature control and different gas formulations with different product types and microorganisms. Active packaging refers to the incorporation of additive agents into packaging materials with the purpose of maintaining or extending food product quality and shelf life. Intelligent packaging is emerging as a potential advantage in food processing and is an especially useful tool for tracking product information and monitoring product conditions. Moreover, intelligent packaging facilitates data access and information exchange by altering conditions inside or outside the packaging and product. In spite of these advantages, few of these packaging systems are commercialized because of high cost, strict safety and hygiene regulations or limited consumer acceptance. Therefore more research is needed to develop cheaper, more easily applicable and effective packaging systems for various foods. © 2015 Society of Chemical Industry.

Color stability of ground beef packaged in a low carbon monoxide atmosphere or vacuum.

PubMed

Jeong, Jong Youn; Claus, James R

2011-01-01

Ground beef was either packaged in an atmosphere of 0.4% CO, 30% CO₂, and 69.6% N₂ (CO-MAP) or vacuum. After storage (48 h, 2-3°C), packages of CO-MAP and vacuum were opened and overwrapped with polyvinyl chloride. Other CO-MAP and vacuum packages were left intact. Packages were initially displayed for 7 days (2-3°C). Intact packages were further displayed up to 35 days before being opened and displayed (1 or 3 days). Intact CO-MAP packaged ground beef was always more red than intact vacuum-packaged ground beef. Color was relatively stable for both types of intact packages over 35 days of display. Upon opening CO-MAP packaged ground beef, the red color decreased slower than in ground beef from vacuum packages. Published by Elsevier Ltd.

Packaging Concerns and Techniques for Large Devices: Challenges for Complex Electronics

NASA Technical Reports Server (NTRS)

LaBel, Kenneth A.; Sampson, Michael J.

2010-01-01

NASA is going to have to accept the use of non-hermetic packages for complex devices. There are a large number of packaging options available. Space application subjects the packages to stresses that they were probably not designed for (vacuum for instance). NASA has to find a way of having assurance in the integrity of the packages. There are manufacturers interested in qualifying non-hermetic packages to MIL-PRF-38535 Class V. Government space users are agreed that Class V should be for hermetic packages only. NASA is working on a new Class for non-hermetic packages for M38535 Appendix B, "Class Y". Testing for package integrity will be required but can be package specific as described by a Package Integrity Test Plan. The plan is developed by the manufacturer and approved by DSCC and government space.

19 CFR 191.13 - Packaging materials.

Code of Federal Regulations, 2010 CFR

2010-04-01

... 19 Customs Duties 2 2010-04-01 2010-04-01 false Packaging materials. 191.13 Section 191.13 Customs... (CONTINUED) DRAWBACK General Provisions § 191.13 Packaging materials. (a) Imported packaging material... packaging material when used to package or repackage merchandise or articles exported or destroyed pursuant...

49 CFR 178.905 - Large Packaging identification codes.

Code of Federal Regulations, 2010 CFR

2010-10-01

... 49 Transportation 2 2010-10-01 2010-10-01 false Large Packaging identification codes. 178.905... FOR PACKAGINGS Large Packagings Standards § 178.905 Large Packaging identification codes. Large packaging code designations consist of: two numerals specified in paragraph (a) of this section; followed by...

Advancements in meat packaging.

PubMed

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.

Packaging Concerns/Techniques for Large Devices

NASA Technical Reports Server (NTRS)

Sampson, Michael J.

2009-01-01

This slide presentation reviews packaging challenges and options for electronic parts. The presentation includes information about non-hermetic packages, space challenges for packaging and complex package variations.

Transformation of food packaging from passive to innovative via nanotechnology: concepts and critiques.

PubMed

Mlalila, Nichrous; Kadam, Dattatreya M; Swai, Hulda; Hilonga, Askwar

2016-09-01

In recent decades, there is a global advancement in manufacturing industry due to increased applications of nanotechnology. Food industry also has been tremendously changing from passive packaging to innovative packaging, to cope with global trends, technological advancements, and consumer preferences. Active research is taking place in food industry and other scientific fields to develop innovative packages including smart, intelligent and active food packaging for more effective and efficient packaging materials with balanced environmental issues. However, in food industry the features behind smart packaging are narrowly defined to be distinguished from intelligent packaging as in other scientific fields, where smart materials are under critical investigations. This review presents some scientific concepts and features pertaining innovative food packaging. The review opens new research window in innovative food packaging to cover the existing disparities for further precise research and development of food packaging industry.

IFT Scientific Status Summary 2008: Innovative Food Packaging Solutions

USDA-ARS?s Scientific Manuscript database

Food and beverage packaging comprises 55-65% of the $110 billion value of packaging in the United States. This review provides a summary of innovative technology developments in food packaging. The expanded role of food and beverage packaging is reviewed. Active and intelligent food packaging, ba...

49 CFR 178.920 - Standards for metal Large Packagings.

Code of Federal Regulations, 2011 CFR

2011-10-01

... 49 Transportation 3 2011-10-01 2011-10-01 false Standards for metal Large Packagings. 178.920... PACKAGINGS Large Packagings Standards § 178.920 Standards for metal Large Packagings. (a) The provisions in this section apply to metal Large Packagings intended to contain liquids and solids. Metal Large...

49 CFR 178.920 - Standards for metal Large Packagings.

Code of Federal Regulations, 2010 CFR

2010-10-01

... 49 Transportation 2 2010-10-01 2010-10-01 false Standards for metal Large Packagings. 178.920... FOR PACKAGINGS Large Packagings Standards § 178.920 Standards for metal Large Packagings. (a) The provisions in this section apply to metal Large Packagings intended to contain liquids and solids. Metal...

78 FR 17890 - Energy Efficiency Program for Commercial and Industrial Equipment: Public Meeting and...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-03-25

... Document for Packaged Terminal Air Conditioners and Packaged Terminal Heat Pumps AGENCY: Office of Energy... must identify the framework document for packaged terminal air conditioners and packaged terminal heat... packaged terminal air conditioners and packaged terminal heat pumps. 78 FR 12252. The document provided for...

49 CFR 178.925 - Standards for rigid plastic Large Packagings.

Code of Federal Regulations, 2011 CFR

2011-10-01

... 49 Transportation 3 2011-10-01 2011-10-01 false Standards for rigid plastic Large Packagings. 178... FOR PACKAGINGS Large Packagings Standards § 178.925 Standards for rigid plastic Large Packagings. (a) The provisions in this section apply to rigid plastic Large Packagings intended to contain liquids and...

49 CFR 178.925 - Standards for rigid plastic Large Packagings.

Code of Federal Regulations, 2012 CFR

2012-10-01

... 49 Transportation 3 2012-10-01 2012-10-01 false Standards for rigid plastic Large Packagings. 178... FOR PACKAGINGS Large Packagings Standards § 178.925 Standards for rigid plastic Large Packagings. (a) The provisions in this section apply to rigid plastic Large Packagings intended to contain liquids and...

49 CFR 178.925 - Standards for rigid plastic Large Packagings.

Code of Federal Regulations, 2014 CFR

2014-10-01

... 49 Transportation 3 2014-10-01 2014-10-01 false Standards for rigid plastic Large Packagings. 178... FOR PACKAGINGS Large Packagings Standards § 178.925 Standards for rigid plastic Large Packagings. (a) The provisions in this section apply to rigid plastic Large Packagings intended to contain liquids and...

49 CFR 178.925 - Standards for rigid plastic Large Packagings.

Code of Federal Regulations, 2013 CFR

2013-10-01

... 49 Transportation 3 2013-10-01 2013-10-01 false Standards for rigid plastic Large Packagings. 178... FOR PACKAGINGS Large Packagings Standards § 178.925 Standards for rigid plastic Large Packagings. (a) The provisions in this section apply to rigid plastic Large Packagings intended to contain liquids and...

49 CFR 178.350 - Specification 7A; general packaging, Type A.

Code of Federal Regulations, 2010 CFR

2010-10-01

... 49 Transportation 2 2010-10-01 2010-10-01 false Specification 7A; general packaging, Type A. 178... FOR PACKAGINGS Specifications for Packagings for Class 7 (Radioactive) Materials § 178.350 Specification 7A; general packaging, Type A. (a) Each packaging must meet all applicable requirements of subpart...

Package for integrated optic circuit and method

DOEpatents

Kravitz, Stanley H.; Hadley, G. Ronald; Warren, Mial E.; Carson, Richard F.; Armendariz, Marcelino G.

1998-01-01

A structure and method for packaging an integrated optic circuit. The package comprises a first wall having a plurality of microlenses formed therein to establish channels of optical communication with an integrated optic circuit within the package. A first registration pattern is provided on an inside surface of one of the walls of the package for alignment and attachment of the integrated optic circuit. The package in one embodiment may further comprise a fiber holder for aligning and attaching a plurality of optical fibers to the package and extending the channels of optical communication to the fibers outside the package. In another embodiment, a fiber holder may be used to hold the fibers and align the fibers to the package. The fiber holder may be detachably connected to the package.

Package for integrated optic circuit and method

DOEpatents

Kravitz, S.H.; Hadley, G.R.; Warren, M.E.; Carson, R.F.; Armendariz, M.G.

1998-08-04

A structure and method are disclosed for packaging an integrated optic circuit. The package comprises a first wall having a plurality of microlenses formed therein to establish channels of optical communication with an integrated optic circuit within the package. A first registration pattern is provided on an inside surface of one of the walls of the package for alignment and attachment of the integrated optic circuit. The package in one embodiment may further comprise a fiber holder for aligning and attaching a plurality of optical fibers to the package and extending the channels of optical communication to the fibers outside the package. In another embodiment, a fiber holder may be used to hold the fibers and align the fibers to the package. The fiber holder may be detachably connected to the package. 6 figs.

Technical Review Report for the Model 9978-96 Package Safety Analysis Report for Packaging (S-SARP-G-00002, Revision 1, March 2009)

DOE Office of Scientific and Technical Information (OSTI.GOV)

West, M

2009-03-06

This Technical Review Report (TRR) documents the review, performed by Lawrence Livermore National Laboratory (LLNL) Staff, at the request of the Department of Energy (DOE), on the 'Safety Analysis Report for Packaging (SARP), Model 9978 B(M)F-96', Revision 1, March 2009 (S-SARP-G-00002). The Model 9978 Package complies with 10 CFR 71, and with 'Regulations for the Safe Transport of Radioactive Material-1996 Edition (As Amended, 2000)-Safety Requirements', International Atomic Energy Agency (IAEA) Safety Standards Series No. TS-R-1. The Model 9978 Packaging is designed, analyzed, fabricated, and tested in accordance with Section III of the American Society of Mechanical Engineers Boiler and Pressuremore » Vessel Code (ASME B&PVC). The review presented in this TRR was performed using the methods outlined in Revision 3 of the DOE's 'Packaging Review Guide (PRG) for Reviewing Safety Analysis Reports for Packages'. The format of the SARP follows that specified in Revision 2 of the Nuclear Regulatory Commission's Regulatory Guide 7.9, i.e., 'Standard Format and Content of Part 71 Applications for Approval of Packages for Radioactive Material'. Although the two documents are similar in their content, they are not identical. Formatting differences have been noted in this TRR, where appropriate. The Model 9978 Packaging is a single containment package, using a 5-inch containment vessel (5CV). It uses a nominal 35-gallon drum package design. In comparison, the Model 9977 Packaging uses a 6-inch containment vessel (6CV). The Model 9977 and Model 9978 Packagings were developed concurrently, and they were referred to as the General Purpose Fissile Material Package, Version 1 (GPFP). Both packagings use General Plastics FR-3716 polyurethane foam as insulation and as impact limiters. The 5CV is used as the Primary Containment Vessel (PCV) in the Model 9975-96 Packaging. The Model 9975-96 Packaging also has the 6CV as its Secondary Containment Vessel (SCV). In comparison, the Model 9975 Packagings use Celotex{trademark} for insulation and as impact limiters. To provide a historical perspective, it is noted that the Model 9975-96 Packaging is a 35-gallon drum package design that has evolved from a family of packages designed by DOE contractors at the Savannah River Site. Earlier package designs, i.e., the Model 9965, the Model 9966, the Model 9967, and the Model 9968 Packagings, were originally designed and certified in the early 1980s. In the 1990s, updated package designs that incorporated design features consistent with the then-newer safety requirements were proposed. The updated package designs at the time were the Model 9972, the Model 9973, the Model 9974, and the Model 9975 Packagings, respectively. The Model 9975 Package was certified by the Packaging Certification Program, under the Office of Safety Management and Operations. The Model 9978 Package has six Content Envelopes: C.1 ({sup 238}Pu Heat Sources), C.2 ( Pu/U Metals), C.3 (Pu/U Oxides, Reserved), C.4 (U Metal or Alloy), C.5 (U Compounds), and C.6 (Samples and Sources). Per 10 CFR 71.59 (Code of Federal Regulations), the value of N is 50 for the Model 9978 Package leading to a Criticality Safety Index (CSI) of 1.0. The Transport Index (TI), based on dose rate, is calculated to be a maximum of 4.1.« less

49 CFR 173.24 - General requirements for packagings and packages.

Code of Federal Regulations, 2011 CFR

2011-10-01

...) New packagings and packagings which are reused; and (3) Specification and non-specification packagings..., sufficient ullage (outage) must be left to ensure that neither leakage nor permanent distortion of the...

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ramuhalli, Pradeep; Hirt, Evelyn H.; Dib, Gerges

This project involved the development of enhanced risk monitors (ERMs) for active components in Advanced Reactor (AdvRx) designs by integrating real-time information about equipment condition with risk monitors. Health monitoring techniques in combination with predictive estimates of component failure based on condition and risk monitors can serve to indicate the risk posed by continued operation in the presence of detected degradation. This combination of predictive health monitoring based on equipment condition assessment and risk monitors can also enable optimization of maintenance scheduling with respect to the economics of plant operation. This report summarizes PNNL’s multi-year project on the development andmore » evaluation of an ERM concept for active components while highlighting FY2016 accomplishments. Specifically, this report provides a status summary of the integration and demonstration of the prototypic ERM framework with the plant supervisory control algorithms being developed at Oak Ridge National Laboratory (ORNL), and describes additional case studies conducted to assess sensitivity of the technology to different quantities. Supporting documentation on the software package to be provided to ONRL is incorporated in this report.« less

Summary of BISON Development and Validation Activities - NEAMS FY16 Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Williamson, R. L.; Pastore, G.; Gamble, K. A.

This summary report contains an overview of work performed under the work package en- titled “FY2016 NEAMS INL-Engineering Scale Fuel Performance (BISON)” A first chapter identifies the specific FY-16 milestones, providing a basic description of the associated work and references to related detailed documentation. Where applicable, a representative technical result is provided. A second chapter summarizes major additional accomplishments, which in- clude: 1) publication of a journal article on solution verification and validation of BISON for LWR fuel, 2) publication of a journal article on 3D Missing Pellet Surface (MPS) analysis of BWR fuel, 3) use of BISON to designmore » a unique 3D MPS validation experiment for future in- stallation in the Halden research reactor, 4) participation in an OECD benchmark on Pellet Clad Mechanical Interaction (PCMI), 5) participation in an OECD benchmark on Reactivity Insertion Accident (RIA) analysis, 6) participation in an OECD activity on uncertainity quantification and sensitivity analysis in nuclear fuel modeling and 7) major improvements to BISON’s fission gas behavior models. A final chapter outlines FY-17 future work.« less

A parallel multi-domain solution methodology applied to nonlinear thermal transport problems in nuclear fuel pins

DOE PAGES

Philip, Bobby; Berrill, Mark A.; Allu, Srikanth; ...

2015-01-26

We describe an efficient and nonlinearly consistent parallel solution methodology for solving coupled nonlinear thermal transport problems that occur in nuclear reactor applications over hundreds of individual 3D physical subdomains. Efficiency is obtained by leveraging knowledge of the physical domains, the physics on individual domains, and the couplings between them for preconditioning within a Jacobian Free Newton Krylov method. Details of the computational infrastructure that enabled this work, namely the open source Advanced Multi-Physics (AMP) package developed by the authors are described. The details of verification and validation experiments, and parallel performance analysis in weak and strong scaling studies demonstratingmore » the achieved efficiency of the algorithm are presented. Moreover, numerical experiments demonstrate that the preconditioner developed is independent of the number of fuel subdomains in a fuel rod, which is particularly important when simulating different types of fuel rods. Finally, we demonstrate the power of the coupling methodology by considering problems with couplings between surface and volume physics and coupling of nonlinear thermal transport in fuel rods to an external radiation transport code.« less

Characterization of the CALIBAN Critical Assembly Neutron Spectra using Several Adjustment Methods Based on Activation Foils Measurement

NASA Astrophysics Data System (ADS)

Casoli, Pierre; Grégoire, Gilles; Rousseau, Guillaume; Jacquet, Xavier; Authier, Nicolas

2016-02-01

CALIBAN is a metallic critical assembly managed by the Criticality, Neutron Science and Measurement Department located on the French CEA Center of Valduc. The reactor is extensively used for benchmark experiments dedicated to the evaluation of nuclear data, for electronic hardening or to study the effect of the neutrons on various materials. Therefore CALIBAN irradiation characteristics and especially its central cavity neutron spectrum have to be very accurately evaluated. In order to strengthen our knowledge of this spectrum, several adjustment methods based on activation foils measurements are being studied for a few years in the laboratory. Firstly two codes included in the UMG package have been tested and compared: MAXED and GRAVEL. More recently, the CALIBAN cavity spectrum has been studied using CALMAR, a new adjustment tool currently under development at the CEA Center of Cadarache. The article will discuss and compare the results and the quality of spectrum rebuilding obtained with the UMG codes and with the CALMAR software, from a set of activation measurements carried out in the CALIBAN irradiation cavity.

A Power Conversion Concept for the Jupiter Icy Moons Orbiter

NASA Technical Reports Server (NTRS)

Mason, Lee S.

2003-01-01

The Jupiter Icy Moons Orbiter (JIMO) mission is currently under study by the Office of Space Science under the Project Prometheus Program. JIMO is examining the use of Nuclear Electric Propulsion (NEP) to carry scientific payloads to three Jovian moons. A potential power system concept includes dual 100 kWe Brayton converters, a deployable pumped loop heat rejection subsystem, and a 400 Vac Power Management and Distribution (PMAD) bus. Many trades were performed in aniving at this candidate power system concept. System-level studies examined design and off-design operating modes, determined startup requirements, evaluated subsystem redundancy options, and quantified the mass and radiator area of reactor power systems from 20 to 200 kWe. In the Brayton converter subsystem, studies were performed to investigate converter packaging options, and assess the induced torque effects on spacecraft dynamics due to rotating machinery. In the heat rejection subsystem, design trades were conducted on heat transport approaches, material and fluid options, and deployed radiator geometries. In the PMAD subsystem, the overall electrical architecture was defined and trade studies examined distribution approaches, voltage levels, and cabling options.

ELECTRONUCLEAR RESEARCH DIVISION SEMIANNUAL PROGRESS REPORT FOR PERIOD ENDING MARCH 20, 1955

DOE Office of Scientific and Technical Information (OSTI.GOV)

Howard, F.T. ed.

1955-06-24

The 86-in. cyclotron is being modified to provide for deflection of the proton beam. Radioisotope production and cyclotron operation before shut-down are summarized. With the use of the 63-in. cyclotron, the absolute values of the electron capture and loss cross sections for elastic scattering of N by N was measured at energies from 13 to 22 Mev. A double-focusing 90 deg magnet is being designed for use in identifying the reaction products from N-induced nuclear reactions. The 44-in. cyclotron is being revised to provide for the acceleration of protons to 1.5 and 5 Mev. The feasibility of converting the 44-more » in. cyclotron to a 48-in. heavy-particle cyclotron is being studied, and design specifications are given. The production of Pu isotopes by electromagnetic separation, Pu recycle chemistry, and product processing are discussed. The Army Package Power Reactor program is summarized. APPR-type fuel assemblies have been fabricated for irradiation experiments and are being corrosion tested. Feasibility studies of a fixed-frequency 1-bev accelerator are reported. (W.L.H.)« less

On the Rocks: Microbiological Quality and Microbial Diversity of Packaged Ice in Southern California.

PubMed

Lee, Kun Ho; Ab Samad, Liana S; Lwin, Phillip M; Riedel, Stefan F; Magin, Ashley; Bashir, Mina; Vaishampayan, Parag A; Lin, Wei-Jen

2017-06-01

Ice is defined as a food and is frequently used in direct contact with food and beverages. Packaged ice is commercially produced and can be easily found in grocery and convenience stores. However, the quality and safety of packaged ice products is not consistent. The Packaged Ice Quality Control Standards manual (PIQCS) published by the International Packaged Ice Association provides the quality and processing standards for packaged ice produced by its members. Packaged ice produced on the premise of stores (on-site packaged ice) is not required to be in compliance with these standards. In this study, packaged ice produced by manufacturing plants or by in-store bagger (ISB) machines and on-site packaged ice were compared for their microbiological quality and microbial diversity. Our results revealed that 19% of the 120 on-site packaged ice samples did not meet the PIQCS microbial limit of 500 CFU/mL (or g) and also the absence of coliforms and Escherichia coli . Staphylococci were found in 34% of the on-site packaged ice samples, most likely through contamination from the packaging workers. None of the ISB and manufactured packaged ice samples had unacceptable microbial levels, and all were devoid of staphylococci. Salmonella was absent in all samples analyzed in this study. Microbial community analysis of ice based on 16S/18S rRNA targeted sequencing revealed a much higher microbial diversity and abundance in the on-site packaged ice than in the ISB ice. Proteobacteria, especially Alphaproteobacteria and Betaproteobacteria, were the dominant bacterial groups in all samples tested. Most of these bacteria were oligotrophic; however, a few opportunistic or potential pathogens were found at low levels in the on-site packaged ice but not in the ISB packaged ice. The types of microbes identified may provide information needed to investigate potential sources of contamination. Our data also suggest a need for enforcement of processing standards during the on-site packaging of ice.

Food packages for Space Shuttle

NASA Technical Reports Server (NTRS)

Fohey, M. F.; Sauer, R. L.; Westover, J. B.; Rockafeller, E. F.

1978-01-01

The paper reviews food packaging techniques used in space flight missions and describes the system developed for the Space Shuttle. Attention is directed to bite-size food cubes used in Gemini, Gemini rehydratable food packages, Apollo spoon-bowl rehydratable packages, thermostabilized flex pouch for Apollo, tear-top commercial food cans used in Skylab, polyethylene beverage containers, Skylab rehydratable food package, Space Shuttle food package configuration, duck-bill septum rehydration device, and a drinking/dispensing nozzle for Space Shuttle liquids. Constraints and testing of packaging is considered, a comparison of food package materials is presented, and typical Shuttle foods and beverages are listed.

Re-design of apple pia packaging using quality function deployment method

NASA Astrophysics Data System (ADS)

Pulungan, M. H.; Nadira, N.; Dewi, I. A.

2018-03-01

This study was aimed to identify the attributes for premium apple pia packaging, to determine the technical response to be carried out by Permata Agro Mandiri Small and Medium Enterprise (SME) and to design a new apple pie packaging acceptable by the SME. The Quality Function Deployment (QFD) method was employed to improve the apple pia packaging design, which consisted of seven stages in data analysis. The results indicated that whats attribute required by the costumers include graphic design, dimensions, capacity, shape, strength, and resistance of packaging. While, the technical responses to be conducted by the SMEs were as follows: attractive visual packaging designs, attractive colors, clear images and information, packaging size dimensions, a larger capacity packaging (more product content), ergonomic premium packaging, not easily torn, and impact resistant packaging materials. The findings further confirmed that the design of premium apple pia packaging accepted by the SMES was the one with the capacity of ten apple pia or 200 g weight, and with rectangular or beam shape form. The packaging material used was a duplex carton with 400 grammage (g/m2), the outer part of the packaging was coated with plastic and the inside was added with duplex carton. The acceptable packaging dimension was 30 cm x 5 cm x 3 cm (L x W x H) with a mix of black and yellow color in the graphical design.

Child-resistant and tamper-resistant packaging: A systematic review to inform tobacco packaging regulation.

PubMed

Jo, Catherine L; Ambs, Anita; Dresler, Carolyn M; Backinger, Cathy L

2017-02-01

We aimed to investigate the effects of special packaging (child-resistant, adult-friendly) and tamper-resistant packaging on health and behavioral outcomes in order to identify research gaps and implications for packaging standards for tobacco products. We searched seven databases for keywords related to special and tamper-resistant packaging, consulted experts, and reviewed citations of potentially relevant studies. 733 unique papers were identified. Two coders independently screened each title and abstract for eligibility. They then reviewed the full text of the remaining papers for a second round of eligibility screening. Included studies investigated a causal relationship between type of packaging or packaging regulation and behavioral or health outcomes and had a study population composed of consumers. Studies were excluded on the basis of publication type, if they were not peer-reviewed, and if they had low external validity. Two reviewers independently coded each paper for study and methodological characteristics and limitations. Discrepancies were discussed and resolved. The review included eight studies: four assessing people's ability to access the contents of different packaging types and four evaluating the impact of packaging requirements on health-related outcomes. Child-resistant packaging was generally more difficult to open than non-child-resistant packaging. Child-resistant packaging requirements have been associated with reductions in child mortality. Child-resistant packaging holds the expectation to reduce tobacco product poisonings among children under six. Published by Elsevier Inc.

Child-resistant and tamper-resistant packaging: A systematic review to inform tobacco packaging regulation

PubMed Central

Jo, Catherine L.; Ambs, Anita; Dresler, Carolyn M.; Backinger, Cathy L.

2017-01-01

Objective We aimed to investigate the effects of special packaging (child-resistant, adult-friendly) and tamper-resistant packaging on health and behavioral outcomes in order to identify research gaps and implications for packaging standards for tobacco products. Methods We searched seven databases for keywords related to special and tamper-resistant packaging, consulted experts, and reviewed citations of potentially relevant studies. 733 unique papers were identified. Two coders independently screened each title and abstract for eligibility. They then reviewed the full text of the remaining papers for a second round of eligibility screening. Included studies investigated a causal relationship between type of packaging or packaging regulation and behavioral or health outcomes and had a study population composed of consumers. Studies were excluded on the basis of publication type, if they were not peer-reviewed, and if they had low external validity. Two reviewers independently coded each paper for study and methodological characteristics and limitations. Discrepancies were discussed and resolved. Results The review included eight studies: four assessing people’s ability to access the contents of different packaging types and four evaluating the impact of packaging requirements on health-related outcomes. Child-resistant packaging was generally more difficult to open than non-child-resistant packaging. Child-resistant packaging requirements have been associated with reductions in child mortality. Conclusions Child-resistant packaging holds the expectation to reduce tobacco product poisonings among children under six. PMID:27939602

49 CFR 173.459 - Mixing of fissile material packages with non-fissile or fissile-excepted material packages.

Code of Federal Regulations, 2012 CFR

2012-10-01

... 49 Transportation 2 2012-10-01 2012-10-01 false Mixing of fissile material packages with non... (Radioactive) Materials § 173.459 Mixing of fissile material packages with non-fissile or fissile-excepted material packages. Mixing of fissile material packages with other types of Class 7 (radioactive) materials...

49 CFR 173.459 - Mixing of fissile material packages with non-fissile or fissile-excepted material packages.

Code of Federal Regulations, 2013 CFR

2013-10-01

... 49 Transportation 2 2013-10-01 2013-10-01 false Mixing of fissile material packages with non... (Radioactive) Materials § 173.459 Mixing of fissile material packages with non-fissile or fissile-excepted material packages. Mixing of fissile material packages with other types of Class 7 (radioactive) materials...

49 CFR 173.459 - Mixing of fissile material packages with non-fissile or fissile-excepted material packages.

Code of Federal Regulations, 2011 CFR

2011-10-01

... 49 Transportation 2 2011-10-01 2011-10-01 false Mixing of fissile material packages with non... (Radioactive) Materials § 173.459 Mixing of fissile material packages with non-fissile or fissile-excepted material packages. Mixing of fissile material packages with other types of Class 7 (radioactive) materials...

49 CFR 173.459 - Mixing of fissile material packages with non-fissile or fissile-excepted material packages.

Code of Federal Regulations, 2010 CFR

2010-10-01

... 49 Transportation 2 2010-10-01 2010-10-01 false Mixing of fissile material packages with non... (Radioactive) Materials § 173.459 Mixing of fissile material packages with non-fissile or fissile-excepted material packages. Mixing of fissile material packages with other types of Class 7 (radioactive) materials...

49 CFR 173.459 - Mixing of fissile material packages with non-fissile or fissile-excepted material packages.

Code of Federal Regulations, 2014 CFR

2014-10-01

... 49 Transportation 2 2014-10-01 2014-10-01 false Mixing of fissile material packages with non... (Radioactive) Materials § 173.459 Mixing of fissile material packages with non-fissile or fissile-excepted material packages. Mixing of fissile material packages with other types of Class 7 (radioactive) materials...

77 FR 70198 - Self-Regulatory Organizations; The NASDAQ Stock Market LLC; Notice of Filing and Immediate...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-11-23

....com Trading and Compliance Data Package service (``Data Package'').\\5\\ The Data Package service... reports provided by the Data Package. The equity trade journal report of the Data Package provides trade... the Data Package report, but requires further segregation and arrangement of the data so that it is...

Integrated Approach to Industrial Packaging Design

NASA Astrophysics Data System (ADS)

Vorobeva, O.

2017-11-01

The article reviews studies in the field of industrial packaging design. The major factors which influence technological, ergonomic, economic and ecological features of packaging are established. The main modern trends in packaging design are defined, the principles of marketing communications and their influence on consumers’ consciousness are indicated, and the function of packaging as a transmitter of brand values is specified. Peculiarities of packaging technology and printing techniques in modern printing industry are considered. The role of designers in the stage-by-stage development of the construction, form and graphic design concept of packaging is defined. The examples of authentic packaging are given and the mention of the tetrahedron packaging history is made. At the end of the article, conclusions on the key research aspects are made.

Transmissible Gastroenteritis Coronavirus Genome Packaging Signal Is Located at the 5′ End of the Genome and Promotes Viral RNA Incorporation into Virions in a Replication-Independent Process

PubMed Central

Morales, Lucia; Mateos-Gomez, Pedro A.; Capiscol, Carmen; del Palacio, Lorena; Sola, Isabel

2013-01-01

Preferential RNA packaging in coronaviruses involves the recognition of viral genomic RNA, a crucial process for viral particle morphogenesis mediated by RNA-specific sequences, known as packaging signals. An essential packaging signal component of transmissible gastroenteritis coronavirus (TGEV) has been further delimited to the first 598 nucleotides (nt) from the 5′ end of its RNA genome, by using recombinant viruses transcribing subgenomic mRNA that included potential packaging signals. The integrity of the entire sequence domain was necessary because deletion of any of the five structural motifs defined within this region abrogated specific packaging of this viral RNA. One of these RNA motifs was the stem-loop SL5, a highly conserved motif in coronaviruses located at nucleotide positions 106 to 136. Partial deletion or point mutations within this motif also abrogated packaging. Using TGEV-derived defective minigenomes replicated in trans by a helper virus, we have shown that TGEV RNA packaging is a replication-independent process. Furthermore, the last 494 nt of the genomic 3′ end were not essential for packaging, although this region increased packaging efficiency. TGEV RNA sequences identified as necessary for viral genome packaging were not sufficient to direct packaging of a heterologous sequence derived from the green fluorescent protein gene. These results indicated that TGEV genome packaging is a complex process involving many factors in addition to the identified RNA packaging signal. The identification of well-defined RNA motifs within the TGEV RNA genome that are essential for packaging will be useful for designing packaging-deficient biosafe coronavirus-derived vectors and providing new targets for antiviral therapies. PMID:23966403

Influence of gag and RRE Sequences on HIV-1 RNA Packaging Signal Structure and Function.

PubMed

Kharytonchyk, Siarhei; Brown, Joshua D; Stilger, Krista; Yasin, Saif; Iyer, Aishwarya S; Collins, John; Summers, Michael F; Telesnitsky, Alice

2018-07-06

The packaging signal (Ψ) and Rev-responsive element (RRE) enable unspliced HIV-1 RNAs' export from the nucleus and packaging into virions. For some retroviruses, engrafting Ψ onto a heterologous RNA is sufficient to direct encapsidation. In contrast, HIV-1 RNA packaging requires 5' leader Ψ elements plus poorly defined additional features. We previously defined minimal 5' leader sequences competitive with intact Ψ for HIV-1 packaging, and here examined the potential roles of additional downstream elements. The findings confirmed that together, HIV-1 5' leader Ψ sequences plus a nuclear export element are sufficient to specify packaging. However, RNAs trafficked using a heterologous export element did not compete well with RNAs using HIV-1's RRE. Furthermore, some RNA additions to well-packaged minimal vectors rendered them packaging-defective. These defects were rescued by extending gag sequences in their native context. To understand these packaging defects' causes, in vitro dimerization properties of RNAs containing minimal packaging elements were compared to RNAs with sequence extensions that were or were not compatible with packaging. In vitro dimerization was found to correlate with packaging phenotypes, suggesting that HIV-1 evolved to prevent 5' leader residues' base pairing with downstream residues and misfolding of the packaging signal. Our findings explain why gag sequences have been implicated in packaging and show that RRE's packaging contributions appear more specific than nuclear export alone. Paired with recent work showing that sequences upstream of Ψ can dictate RNA folds, the current work explains how genetic context of minimal packaging elements contributes to HIV-1 RNA fate determination. Copyright © 2018 Elsevier Ltd. All rights reserved.

NEET Enhanced Micro Pocket Fission Detector for High Temperature Reactors - FY15 Status Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Unruh, Troy; McGregor, Douglas; Ugorowski, Phil

2015-09-01

A new project, that is a collaboration between the Idaho National Laboratory (INL), the Kansas State University (KSU), and the French Atomic Energy Agency, Commissariat à l'Énergie Atomique et aux Energies Alternatives, (CEA), has been initiated by the Nuclear Energy Enabling Technologies (NEET) Advanced Sensors and Instrumentation (ASI) program for developing and testing High Temperature Micro-Pocket Fission Detectors (HT MPFD), which are compact fission chambers capable of simultaneously measuring thermal neutron flux, fast neutron flux and temperature within a single package for temperatures up to 800 °C. The MPFD technology utilizes a small, multi-purpose, robust, in-core parallel plate fission chambermore » and thermocouple. As discussed within this report, the small size, variable sensitivity, and increased accuracy of the MPFD technology represent a revolutionary improvement over current methods used to support irradiations in US Material Test Reactors (MTRs). Previous research conducted through NEET ASI1-3 has shown that the MPFD technology could be made robust and was successfully tested in a reactor core. This new project will further the MPFD technology for higher temperature regimes and other reactor applications by developing a HT MPFD suitable for temperatures up to 800 °C. This report summarizes the research progress for year one of this three year project. Highlights from research accomplishments include: A joint collaboration was initiated between INL, KSU, and CEA. Note that CEA is participating at their own expense because of interest in this unique new sensor. An updated HT MPFD design was developed. New high temperature-compatible materials for HT MPFD construction were procured. Construction methods to support the new design were evaluated at INL. Laboratory evaluations of HT MPFD were initiated. Electrical contact and fissile material plating has been performed at KSU. Updated detector electronics are undergoing evaluations at KSU. A project meeting was held at KSU to discuss the roles and responsibilities between INL and KSU for development of the HT MPFDs. Provide input to various irradiation programs for installation of the MPFD technology in irradiation tests. As documented in this report, FY15 funding has allowed the project to meet year one planned accomplishments to develop a HT MPFD that offers US MTR users enhanced capabilities for real-time measurement of flux and temperature with a single detector. In addition, the accomplishments of this project have attracted funding from other Department of Energy Office of Nuclear Energy (DOE-NE) programs for additional applications. The work in those programs will build on current activities completed in this NEETASI HT MPFD project, but the MPFD will be specifically tailored to meet their program needs.« less

JPRS Report, Science & Technology, China: Energy.

DTIC Science & Technology

1992-03-30

breeder reactors should become...the primary type of reactors . In developing breeder reactors , we should follow the path of using metal fuel. Breeder reactors give us more time to...first reactor used for power generation was a fast reactor : the " Breeder 1" reactor at the Idaho National Reactor Test Center which was used to

Components of Adenovirus Genome Packaging

PubMed Central

Ahi, Yadvinder S.; Mittal, Suresh K.

2016-01-01

Adenoviruses (AdVs) are icosahedral viruses with double-stranded DNA (dsDNA) genomes. Genome packaging in AdV is thought to be similar to that seen in dsDNA containing icosahedral bacteriophages and herpesviruses. Specific recognition of the AdV genome is mediated by a packaging domain located close to the left end of the viral genome and is mediated by the viral packaging machinery. Our understanding of the role of various components of the viral packaging machinery in AdV genome packaging has greatly advanced in recent years. Characterization of empty capsids assembled in the absence of one or more components involved in packaging, identification of the unique vertex, and demonstration of the role of IVa2, the putative packaging ATPase, in genome packaging have provided compelling evidence that AdVs follow a sequential assembly pathway. This review provides a detailed discussion on the functions of the various viral and cellular factors involved in AdV genome packaging. We conclude by briefly discussing the roles of the empty capsids, assembly intermediates, scaffolding proteins, portal vertex and DNA encapsidating enzymes in AdV assembly and packaging. PMID:27721809

16 CFR 1700.15 - Poison prevention packaging standards.

Code of Federal Regulations, 2012 CFR

2012-01-01

... 16 Commercial Practices 2 2012-01-01 2012-01-01 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...

16 CFR 1700.15 - Poison prevention packaging standards.

Code of Federal Regulations, 2014 CFR

2014-01-01

... 16 Commercial Practices 2 2014-01-01 2014-01-01 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...

16 CFR 1700.15 - Poison prevention packaging standards.

Code of Federal Regulations, 2010 CFR

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...

16 CFR 1700.15 - Poison prevention packaging standards.

Code of Federal Regulations, 2011 CFR

2011-01-01

... 16 Commercial Practices 2 2011-01-01 2011-01-01 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...

16 CFR 1700.3 - Establishment of standards for special packaging.

Code of Federal Regulations, 2011 CFR

2011-01-01

... standard; (2) Available scientific, medical, and engineering data concerning special packaging and... such standards, the Commission shall not prescribe specific packaging designs, product content, package... such substance in a package which the Commission determines is unnecessarily attractive to children. (e...

16 CFR 1700.3 - Establishment of standards for special packaging.

Code of Federal Regulations, 2010 CFR

2010-01-01

... standard; (2) Available scientific, medical, and engineering data concerning special packaging and... such standards, the Commission shall not prescribe specific packaging designs, product content, package... such substance in a package which the Commission determines is unnecessarily attractive to children. (e...

16 CFR 1700.3 - Establishment of standards for special packaging.

Code of Federal Regulations, 2012 CFR

2012-01-01

... standard; (2) Available scientific, medical, and engineering data concerning special packaging and... such standards, the Commission shall not prescribe specific packaging designs, product content, package... such substance in a package which the Commission determines is unnecessarily attractive to children. (e...

16 CFR 1700.3 - Establishment of standards for special packaging.

Code of Federal Regulations, 2014 CFR

2014-01-01

... standard; (2) Available scientific, medical, and engineering data concerning special packaging and... such standards, the Commission shall not prescribe specific packaging designs, product content, package... such substance in a package which the Commission determines is unnecessarily attractive to children. (e...

49 CFR 173.25 - Authorized packagings and overpacks.

Code of Federal Regulations, 2010 CFR

2010-10-01

... 49 Transportation 2 2010-10-01 2010-10-01 false Authorized packagings and overpacks. 173.25...-GENERAL REQUIREMENTS FOR SHIPMENTS AND PACKAGINGS Preparation of Hazardous Materials for Transportation § 173.25 Authorized packagings and overpacks. (a) Authorized packages containing hazardous materials may...

40 CFR 157.27 - Unit packaging.

Code of Federal Regulations, 2010 CFR

2010-07-01

... 40 Protection of Environment 23 2010-07-01 2010-07-01 false Unit packaging. 157.27 Section 157.27 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) PESTICIDE PROGRAMS PACKAGING REQUIREMENTS FOR PESTICIDES AND DEVICES Child-Resistant Packaging § 157.27 Unit packaging. Pesticide products...

49 CFR 173.23 - Previously authorized packaging.

Code of Federal Regulations, 2010 CFR

2010-10-01

... 49 Transportation 2 2010-10-01 2010-10-01 false Previously authorized packaging. 173.23 Section... REQUIREMENTS FOR SHIPMENTS AND PACKAGINGS Preparation of Hazardous Materials for Transportation § 173.23 Previously authorized packaging. (a) When the regulations specify a packaging with a specification marking...

High-performance packaging for monolithic microwave and millimeter-wave integrated circuits

NASA Technical Reports Server (NTRS)

Shalkhauser, K. A.; Li, K.; Shih, Y. C.

1992-01-01

Packaging schemes are developed that provide low-loss, hermetic enclosure for enhanced monolithic microwave and millimeter-wave integrated circuits. These package schemes are based on a fused quartz substrate material offering improved RF performance through 44 GHz. The small size and weight of the packages make them useful for a number of applications, including phased array antenna systems. As part of the packaging effort, a test fixture was developed to interface the single chip packages to conventional laboratory instrumentation for characterization of the packaged devices.

Reliability of high I/O high density CCGA interconnect electronic packages under extreme thermal environments

NASA Astrophysics Data System (ADS)

Ramesham, Rajeshuni

2012-03-01

Ceramic column grid array (CCGA) packages have been increasing in use based on their advantages such as high interconnect density, very good thermal and electrical performances, compatibility with standard surfacemount packaging assembly processes, and so on. CCGA packages are used in space applications such as in logic and microprocessor functions, telecommunications, payload electronics, and flight avionics. As these packages tend to have less solder joint strain relief than leaded packages or more strain relief over lead-less chip carrier packages, the reliability of CCGA packages is very important for short-term and long-term deep space missions. We have employed high density CCGA 1152 and 1272 daisy chained electronic packages in this preliminary reliability study. Each package is divided into several daisy-chained sections. The physical dimensions of CCGA1152 package is 35 mm x 35 mm with a 34 x 34 array of columns with a 1 mm pitch. The dimension of the CCGA1272 package is 37.5 mm x 37.5 mm with a 36 x 36 array with a 1 mm pitch. The columns are made up of 80% Pb/20%Sn material. CCGA interconnect electronic package printed wiring polyimide boards have been assembled and inspected using non-destructive x-ray imaging techniques. The assembled CCGA boards were subjected to extreme temperature thermal atmospheric cycling to assess their reliability for future deep space missions. The resistance of daisy-chained interconnect sections were monitored continuously during thermal cycling. This paper provides the experimental test results of advanced CCGA packages tested in extreme temperature thermal environments. Standard optical inspection and x-ray non-destructive inspection tools were used to assess the reliability of high density CCGA packages for deep space extreme temperature missions.

Polymer dispensing and embossing technology for the lens type LED packaging

NASA Astrophysics Data System (ADS)

Chien, Chien-Lin Chang; Huang, Yu-Che; Hu, Syue-Fong; Chang, Chung-Min; Yip, Ming-Chuen; Fang, Weileun

2013-06-01

This study presents a ring-type micro-structure design on the substrate and its corresponding micro fabrication processes for a lens-type light-emitting diode (LED) package. The dome-type or crater-type silicone lenses are achieved by a dispensing and embossing process rather than a molding process. Silicone with a high viscosity and thixotropy index is used as the encapsulant material. The ring-type micro structure is adopted to confine the dispensed silicone encapsulant so as to form the packaged lens. With the architecture and process described, this LED package technology herein has three merits: (1) the flexibility of lens-type LED package designs is enhanced; (2) a dome-type package design is used to enhance the intensity; (3) a crater-type package design is used to enhance the view angle. Measurement results show the ratio between the lens height and lens radius can vary from 0.4 to 1 by changing the volume of dispensed silicone. The view angles of dome-type and crater-type packages can reach 155° ± 5° and 175° ± 5°, respectively. As compared with the commercial plastic leaded chip carrier-type package, the luminous flux of a monochromatic blue light LED is improved by 15% by the dome-type package (improved by 7% by the crater-type package) and the luminous flux of a white light LED is improved by 25% by the dome-type package (improved by 13% by the crater-type package). The luminous flux of monochromatic blue light LED and white light LED are respectively improved by 8% and 12% by the dome-type package as compare with the crater-type package.

Method of fabricating a microelectronic device package with an integral window

DOEpatents

Peterson, Kenneth A.; Watson, Robert D.

2003-01-01

A method of fabricating a microelectronic device package with an integral window for providing optical access through an aperture in the package. The package is made of a multilayered insulating material, e.g., a low-temperature cofired ceramic (LTCC) or high-temperature cofired ceramic (HTCC). The window is inserted in-between personalized layers of ceramic green tape during stackup and registration. Then, during baking and firing, the integral window is simultaneously bonded to the sintered ceramic layers of the densified package. Next, the microelectronic device is flip-chip bonded to cofired thick-film metallized traces on the package, where the light-sensitive side is optically accessible through the window. Finally, a cover lid is attached to the opposite side of the package. The result is a compact, low-profile package, flip-chip bonded, hermetically-sealed package having an integral window.

Sensory impacts of food-packaging interactions.

PubMed

Duncan, Susan E; Webster, Janet B

2009-01-01

Sensory changes in food products result from intentional or unintentional interactions with packaging materials and from failure of materials to protect product integrity or quality. Resolving sensory issues related to plastic food packaging involves knowledge provided by sensory scientists, materials scientists, packaging manufacturers, food processors, and consumers. Effective communication among scientists and engineers from different disciplines and industries can help scientists understand package-product interactions. Very limited published literature describes sensory perceptions associated with food-package interactions. This article discusses sensory impacts, with emphasis on oxidation reactions, associated with the interaction of food and materials, including taints, scalping, changes in food quality as a function of packaging, and examples of material innovations for smart packaging that can improve sensory quality of foods and beverages. Sensory evaluation is an important tool for improved package selection and development of new materials.

Efficacy of Antimicrobial Agents for Food Contact Applications: Biological Activity, Incorporation into Packaging, and Assessment Methods: A Review.

PubMed

Mousavi Khaneghah, Amin; Hashemi, Seyed Mohammad Bagher; Eş, Ismail; Fracassetti, Daniela; Limbo, Sara

2018-07-01

Interest in the utilization of antimicrobial active packaging for food products has increased in recent years. Antimicrobial active packaging involves the incorporation of antimicrobial compounds into packaging materials, with the aim of maintaining or extending food quality and shelf life. Plant extracts, essential oils, organic acids, bacteriocins, inorganic substances, enzymes, and proteins are used as antimicrobial agents in active packaging. Evaluation of the antimicrobial activity of packaging materials using different methods has become a critical issue for both food safety and the commercial utilization of such packaging technology. This article reviews the different types of antimicrobial agents used for active food packaging materials, the main incorporation techniques, and the assessment methods used to examine the antimicrobial activity of packaging materials, taking into account their safety as food contact materials.

78 FR 1101 - Hazardous Materials: Harmonization With the United Nations Recommendations on the Transport of...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-01-07

... packaging maintains an equivalent level of performance to the originally tested packaging design must be... material, packing group assignments, special provisions, packaging authorizations, packaging sections, air... responsibilities related to packaging design variation, manufacturer notification, and recordkeeping requirements...

16 CFR § 1700.3 - Establishment of standards for special packaging.

Code of Federal Regulations, 2013 CFR

2013-01-01

... such standard; (2) Available scientific, medical, and engineering data concerning special packaging and... such standards, the Commission shall not prescribe specific packaging designs, product content, package... such substance in a package which the Commission determines is unnecessarily attractive to children. (e...

Packaging Technologies for 500C SiC Electronics and Sensors

NASA Technical Reports Server (NTRS)

Chen, Liang-Yu

2013-01-01

Various SiC electronics and sensors are currently under development for applications in 500C high temperature environments such as hot sections of aerospace engines and the surface of Venus. In order to conduct long-term test and eventually commercialize these SiC devices, compatible packaging technologies for the SiC electronics and sensors are required. This presentation reviews packaging technologies developed for 500C SiC electronics and sensors to address both component and subsystem level packaging needs for high temperature environments. The packaging system for high temperature SiC electronics includes ceramic chip-level packages, ceramic printed circuit boards (PCBs), and edge-connectors. High temperature durable die-attach and precious metal wire-bonding are used in the chip-level packaging process. A high temperature sensor package is specifically designed to address high temperature micro-fabricated capacitive pressure sensors for high differential pressure environments. This presentation describes development of these electronics and sensor packaging technologies, including some testing results of SiC electronics and capacitive pressure sensors using these packaging technologies.

Biofilm reactors for industrial bioconversion processes: employing potential of enhanced reaction rates

PubMed Central

Qureshi, Nasib; Annous, Bassam A; Ezeji, Thaddeus C; Karcher, Patrick; Maddox, Ian S

2005-01-01

This article describes the use of biofilm reactors for the production of various chemicals by fermentation and wastewater treatment. Biofilm formation is a natural process where microbial cells attach to the support (adsorbent) or form flocs/aggregates (also called granules) without use of chemicals and form thick layers of cells known as "biofilms." As a result of biofilm formation, cell densities in the reactor increase and cell concentrations as high as 74 gL-1 can be achieved. The reactor configurations can be as simple as a batch reactor, continuous stirred tank reactor (CSTR), packed bed reactor (PBR), fluidized bed reactor (FBR), airlift reactor (ALR), upflow anaerobic sludge blanket (UASB) reactor, or any other suitable configuration. In UASB granular biofilm particles are used. This article demonstrates that reactor productivities in these reactors have been superior to any other reactor types. This article describes production of ethanol, butanol, lactic acid, acetic acid/vinegar, succinic acid, and fumaric acid in addition to wastewater treatment in the biofilm reactors. As the title suggests, biofilm reactors have high potential to be employed in biotechnology/bioconversion industry for viable economic reasons. In this article, various reactor types have been compared for the above bioconversion processes. PMID:16122390

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

PubMed Central

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

Botulism challenge studies of a modified atmosphere package for fresh mussels: inoculated pack studies.

PubMed

Newell, C R; Ma, Li; Doyle, Michael

2012-06-01

A series of botulism challenge studies were performed to determine the possibility of production of botulinum toxin in mussels (Mytilus edulis) held under a commercial high-oxygen (60 to 65% O(2)), modified atmosphere packaging (MAP) condition. Spore mixtures of six strains of nonproteolytic Clostridium botulinum were introduced into mussel MAP packages receiving different packaging buffers with or without the addition of lactic acid bacteria. Dye studies and package flipping trials were conducted to ensure internalization of spores by packed mussels. Inoculated mussel packages were stored at normal (4°C) and abusive (12°C) temperatures for 21 and 13 days, respectively, which were beyond the packaged mussels' intended shelf life. Microbiological and chemical analyses were conducted at predetermined intervals (a total of five sampling times at each temperature), including total aerobic plate counts, C. botulinum counts, lactic acid bacterial counts, package headspace gas composition, pH of packaging buffer and mussel meat, and botulinum toxin assays of packaging buffer and mussel meat. Results revealed that C. botulinum inoculated in fresh mussels packed under MAP packaging did not produce toxin, even at an abusive storage temperature and when held beyond their shelf life. No evidence was found that packaging buffers or gas composition influenced the lack of botulinum toxin production in packed mussels.

Physical and computational studies of slag behavior in an entrained flow gasifier

NASA Astrophysics Data System (ADS)

Pummill, Randy

This work details an investigation of how to modify slag flow so as to maintain a clear line of sight across the reaction section of an entrained-flow coal gasifier. Physical and computational models were developed to study methods of diverting the molten slag that flows vertically down the walls of the reactor. The physical models employed silicone oil of varying viscosity. The computational models were developed using the Fluent software package. Based on the insight gained from the results of the models, two devices were created and tested in a pilot scale gasifier located at the University of Utah. The first method of slag diversion studied employed a gas jet to impact the slag film and cause it to flow around a sight port in the gasifier wall. By studying the film and jet interactions, it was discovered that the resulting behavior of such a system can be described by a dimensionless ratio of the kinetic energy of the jet and the surface energy of the film. The development of the dimensionless number, called a Lotte number in this work, is presented in detail. Generally, viscous films will be broken by a jet when the Lotte number is greater than 5 and will reclose when the Lotte number falls below a value of 1.5. The second slag diversion method studied used a round alumina tube protruding horizontally into the reaction section to break up the film. As the film impacts the tube, it progresses horizontally along the length of the tube before resuming the downward flow. The models helped to establish how far the tube should protrude into the reactor in order to successfully break up the slag flow. Slag diversion devices were constructed and installed on a pilot scale gasifier. The jet diversion method was found to require an unreasonably large amount of purge gas to be successful and the metal jet suffered from the high temperature of the reactor despite the cooling effect of the gas. The tube diversion method worked very well for a series of experiments. However, erosion of the alumina tube in the reaction section remains an impediment to using such a device in an industrial setting. A design using a water-cooled tube is suggested.

Natural biopolimers in organic food packaging

NASA Astrophysics Data System (ADS)

Wieczynska, Justyna; Cavoski, Ivana; Chami, Ziad Al; Mondelli, Donato; Di Donato, Paola; Di Terlizzi, Biagio

2014-05-01

Concerns on environmental and waste problems caused by use of non-biodegradable and non-renewable based plastic packaging have caused an increase interest in developing biodegradable packaging using renewable natural biopolymers. Recently, different types of biopolymers like starch, cellulose, chitosan, casein, whey protein, collagen, egg white, soybean protein, corn zein, gelatin and wheat gluten have attracted considerable attention as potential food packaging materials. Recyclable or biodegradable packaging material in organic processing standards is preferable where possible but specific principles of packaging are not precisely defined and standards have to be assessed. There is evidence that consumers of organic products have specific expectations not only with respect to quality characteristics of processed food but also in social and environmental aspects of food production. Growing consumer sophistication is leading to a proliferation in food eco-label like carbon footprint. Biopolymers based packaging for organic products can help to create a green industry. Moreover, biopolymers can be appropriate materials for the development of an active surfaces designed to deliver incorporated natural antimicrobials into environment surrounding packaged food. Active packaging is an innovative mode of packaging in which the product and the environment interact to prolong shelf life or enhance safety or sensory properties, while maintaining the quality of the product. The work will discuss the various techniques that have been used for development of an active antimicrobial biodegradable packaging materials focusing on a recent findings in research studies. With the current focus on exploring a new generation of biopolymer-based food packaging materials with possible applications in organic food packaging. Keywords: organic food, active packaging, biopolymers , green technology

Evaluating Penetration Ability of Plodia interpunctella (Hübner) (Lepidoptera: Pyralidae) Larvae into Multilayer Polypropylene Packages.

PubMed

Scheff, Deanna S; Sehgal, Blossom; Subramanyam, Bhadriraju

2018-04-18

The larvae of the Indian meal moth, Plodia interpunctella (Hübner), can invade or penetrate packaging materials and infest food products. Energy bars with three polypropylene packaging types were challenged with eggs (first instars), third instars, and fifth instars of P. interpunctella to determine package resistance at 28 °C and 65% r.h. The packing types were also challenged with two male and two female pupae of P. interpunctella under similar conditions in order to determine which package provided the greatest protection against larval penetration. Samples infested with eggs, third instars, and pupae were evaluated after 21 days and 42 days to count the number of larvae, pupae, and adults found inside the packages. Packages challenged with fifth instars were observed after 21 days to count the number of larvae, pupae, and adults inside each package. The number and diameter of the holes were determined in each package, followed by the amount of damage sustained to the energy bar. Third and fifth instars showed a higher tendency to penetrate all of the packaging types. First instars showed a reduction in package penetration ability compared with third and fifth instars. The increase in exposure time resulted in an increase in the damage sustained to the energy bars. Among packaging types, the thickest package (Test A) was most resilient to penetration by all of the larval stages. In conclusion, energy bar manufacturers need to invest more effort into improving packaging designs, creating thicker gauge films, or advancing odor barrier technology, in order to prevent penetration and infestation by P. interpunctella larvae.

77 FR 22504 - Hazardous Materials; Packages Intended for Transport by Aircraft

Federal Register 2010, 2011, 2012, 2013, 2014

2012-04-16

... material to absorb the entire contents of the inner packaging, before being placed in its outer package... combination packaging intended for the air transportation of liquid hazardous materials is capable of..., leakproof receptacle or intermediate packaging containing sufficient absorbent material to absorb the entire...

40 CFR 157.30 - Voluntary use of child-resistant packaging.

Code of Federal Regulations, 2010 CFR

2010-07-01

... packaging. 157.30 Section 157.30 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) PESTICIDE PROGRAMS PACKAGING REQUIREMENTS FOR PESTICIDES AND DEVICES Child-Resistant Packaging § 157.30 Voluntary use of child-resistant packaging. A registrant whose product is not required to be in child...

49 CFR 173.428 - Empty Class 7 (radioactive) materials packaging.

Code of Federal Regulations, 2010 CFR

2010-10-01

... 49 Transportation 2 2010-10-01 2010-10-01 false Empty Class 7 (radioactive) materials packaging... SHIPPERS-GENERAL REQUIREMENTS FOR SHIPMENTS AND PACKAGINGS Class 7 (Radioactive) Materials § 173.428 Empty Class 7 (radioactive) materials packaging. A packaging which previously contained Class 7 (radioactive...

49 CFR 173.29 - Empty packagings.

Code of Federal Regulations, 2010 CFR

2010-10-01

... 49 Transportation 2 2010-10-01 2010-10-01 false Empty packagings. 173.29 Section 173.29... SHIPMENTS AND PACKAGINGS Preparation of Hazardous Materials for Transportation § 173.29 Empty packagings. (a) General. Except as otherwise provided in this section, an empty packaging containing only the residue of a...

Control of reactor coolant flow path during reactor decay heat removal

DOEpatents

Hunsbedt, Anstein N.

1988-01-01

An improved reactor vessel auxiliary cooling system for a sodium cooled nuclear reactor is disclosed. The sodium cooled nuclear reactor is of the type having a reactor vessel liner separating the reactor hot pool on the upstream side of an intermediate heat exchanger and the reactor cold pool on the downstream side of the intermediate heat exchanger. The improvement includes a flow path across the reactor vessel liner flow gap which dissipates core heat across the reactor vessel and containment vessel responsive to a casualty including the loss of normal heat removal paths and associated shutdown of the main coolant liquid sodium pumps. In normal operation, the reactor vessel cold pool is inlet to the suction side of coolant liquid sodium pumps, these pumps being of the electromagnetic variety. The pumps discharge through the core into the reactor hot pool and then through an intermediate heat exchanger where the heat generated in the reactor core is discharged. Upon outlet from the heat exchanger, the sodium is returned to the reactor cold pool. The improvement includes placing a jet pump across the reactor vessel liner flow gap, pumping a small flow of liquid sodium from the lower pressure cold pool into the hot pool. The jet pump has a small high pressure driving stream diverted from the high pressure side of the reactor pumps. During normal operation, the jet pumps supplement the normal reactor pressure differential from the lower pressure cold pool to the hot pool. Upon the occurrence of a casualty involving loss of coolant pump pressure, and immediate cooling circuit is established by the back flow of sodium through the jet pumps from the reactor vessel hot pool to the reactor vessel cold pool. The cooling circuit includes flow into the reactor vessel liner flow gap immediate the reactor vessel wall and containment vessel where optimum and immediate discharge of residual reactor heat occurs.

49 CFR 173.162 - Gallium.

Code of Federal Regulations, 2014 CFR

2014-10-01

... 49 Transportation 2 2014-10-01 2014-10-01 false Gallium. 173.162 Section 173.162 Transportation... PACKAGINGS Non-bulk Packaging for Hazardous Materials Other Than Class 1 and Class 7 § 173.162 Gallium. (a) Except when packaged in cylinders or steel flasks, gallium must be packaged in packagings which meet the...

49 CFR 173.162 - Gallium.

Code of Federal Regulations, 2012 CFR

2012-10-01

... 49 Transportation 2 2012-10-01 2012-10-01 false Gallium. 173.162 Section 173.162 Transportation... PACKAGINGS Non-bulk Packaging for Hazardous Materials Other Than Class 1 and Class 7 § 173.162 Gallium. (a) Except when packaged in cylinders or steel flasks, gallium must be packaged in packagings which meet the...

49 CFR 173.162 - Gallium.

Code of Federal Regulations, 2013 CFR

2013-10-01

... 49 Transportation 2 2013-10-01 2013-10-01 false Gallium. 173.162 Section 173.162 Transportation... PACKAGINGS Non-bulk Packaging for Hazardous Materials Other Than Class 1 and Class 7 § 173.162 Gallium. (a) Except when packaged in cylinders or steel flasks, gallium must be packaged in packagings which meet the...

10 CFR 60.135 - Criteria for the waste package and its components.

Code of Federal Regulations, 2010 CFR

2010-01-01

... Section 60.135 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) DISPOSAL OF HIGH-LEVEL RADIOACTIVE WASTES... for the waste package and its components. (a) High-level-waste package design in general. (1) Packages... package's permanent written records. (c) Waste form criteria for HLW. High-level radioactive waste that is...

16 CFR § 1700.15 - Poison prevention packaging standards.

Code of Federal Regulations, 2013 CFR

2013-01-01

... 16 Commercial Practices 2 2013-01-01 2013-01-01 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...

19 CFR 10.2022 - Retail packaging materials and containers.

Code of Federal Regulations, 2014 CFR

2014-04-01

... 19 Customs Duties 1 2014-04-01 2014-04-01 false Retail packaging materials and containers. 10.2022... Trade Promotion Agreement Rules of Origin § 10.2022 Retail packaging materials and containers. (a) Effect on tariff shift rule. Packaging materials and containers in which a good is packaged for retail...

49 CFR 173.36 - Hazardous materials in Large Packagings.

Code of Federal Regulations, 2012 CFR

2012-10-01

... Packagings (e.g., 51H) are only authorized for use with flexible inner packagings. (3) Friction. The nature and thickness of the outer packaging must be such that friction during transportation is not likely to... transportation in inner packagings appropriately resistant to an increase of internal pressure likely to develop...

16 CFR 305.15 - Labeling for lighting products.

Code of Federal Regulations, 2011 CFR

2011-01-01

...) Package labeling. For purposes of labeling under this section, packaging for such fluorescent lamp... individually or in small numbers. The encircled capital letter “E” on packages containing fluorescent lamp... ink, on the surface of the package on which printing or a label normally appears. If the package...

21 CFR 801.437 - User labeling for devices that contain natural rubber.

Code of Federal Regulations, 2014 CFR

2014-04-01

... User labeling for devices that contain natural rubber. (a) Data in the Medical Device Reporting System... of the device packaging, the outside package, container or wrapper, and the immediate device package... panel of the device packaging, the outside package, container or wrapper, and the immediate device...

21 CFR 801.437 - User labeling for devices that contain natural rubber.

Code of Federal Regulations, 2011 CFR

2011-04-01

... User labeling for devices that contain natural rubber. (a) Data in the Medical Device Reporting System... of the device packaging, the outside package, container or wrapper, and the immediate device package... panel of the device packaging, the outside package, container or wrapper, and the immediate device...

49 CFR 173.415 - Authorized Type A packages.

Code of Federal Regulations, 2011 CFR

2011-10-01

... 49 Transportation 2 2011-10-01 2011-10-01 false Authorized Type A packages. 173.415 Section 173... REQUIREMENTS FOR SHIPMENTS AND PACKAGINGS Class 7 (Radioactive) Materials § 173.415 Authorized Type A packages. The following packages are authorized for shipment if they do not contain quantities exceeding A1 or...

49 CFR 173.415 - Authorized Type A packages.

Code of Federal Regulations, 2013 CFR

2013-10-01

... 49 Transportation 2 2013-10-01 2013-10-01 false Authorized Type A packages. 173.415 Section 173... REQUIREMENTS FOR SHIPMENTS AND PACKAGINGS Class 7 (Radioactive) Materials § 173.415 Authorized Type A packages. The following packages are authorized for shipment if they do not contain quantities exceeding A1 or...

21 CFR 801.437 - User labeling for devices that contain natural rubber.

Code of Federal Regulations, 2013 CFR

2013-04-01

... User labeling for devices that contain natural rubber. (a) Data in the Medical Device Reporting System... of the device packaging, the outside package, container or wrapper, and the immediate device package... panel of the device packaging, the outside package, container or wrapper, and the immediate device...

49 CFR 173.415 - Authorized Type A packages.

Code of Federal Regulations, 2010 CFR

2010-10-01

... 49 Transportation 2 2010-10-01 2010-10-01 false Authorized Type A packages. 173.415 Section 173... REQUIREMENTS FOR SHIPMENTS AND PACKAGINGS Class 7 (Radioactive) Materials § 173.415 Authorized Type A packages. The following packages are authorized for shipment if they do not contain quantities exceeding A1 or...

21 CFR 801.437 - User labeling for devices that contain natural rubber.

Code of Federal Regulations, 2012 CFR

2012-04-01

... User labeling for devices that contain natural rubber. (a) Data in the Medical Device Reporting System... of the device packaging, the outside package, container or wrapper, and the immediate device package... panel of the device packaging, the outside package, container or wrapper, and the immediate device...

21 CFR 801.437 - User labeling for devices that contain natural rubber.

Code of Federal Regulations, 2010 CFR

2010-04-01

... User labeling for devices that contain natural rubber. (a) Data in the Medical Device Reporting System... of the device packaging, the outside package, container or wrapper, and the immediate device package... panel of the device packaging, the outside package, container or wrapper, and the immediate device...

49 CFR 173.415 - Authorized Type A packages.

Code of Federal Regulations, 2012 CFR

2012-10-01

... 49 Transportation 2 2012-10-01 2012-10-01 false Authorized Type A packages. 173.415 Section 173... REQUIREMENTS FOR SHIPMENTS AND PACKAGINGS Class 7 (Radioactive) Materials § 173.415 Authorized Type A packages. The following packages are authorized for shipment if they do not contain quantities exceeding A1 or...

7 CFR 58.340 - Printing and packaging.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 7 Agriculture 3 2010-01-01 2010-01-01 false Printing and packaging. 58.340 Section 58.340... Procedures § 58.340 Printing and packaging. Printing and packaging of consumer size containers of butter... packaging equipment should be provided. The outside cartons should be removed from bulk butter in a room...

9 CFR 112.10 - Special packaging and labeling.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 9 Animals and Animal Products 1 2010-01-01 2010-01-01 false Special packaging and labeling. 112.10... AGRICULTURE VIRUSES, SERUMS, TOXINS, AND ANALOGOUS PRODUCTS; ORGANISMS AND VECTORS PACKAGING AND LABELING § 112.10 Special packaging and labeling. A biological product, which requires special packaging and/or...

21 CFR 820.130 - Device packaging.

Code of Federal Regulations, 2010 CFR

2010-04-01

... 21 Food and Drugs 8 2010-04-01 2010-04-01 false Device packaging. 820.130 Section 820.130 Food and... QUALITY SYSTEM REGULATION Labeling and Packaging Control § 820.130 Device packaging. Each manufacturer shall ensure that device packaging and shipping containers are designed and constructed to protect the...

7 CFR 58.151 - Packaging and repackaging.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 7 Agriculture 3 2010-01-01 2010-01-01 false Packaging and repackaging. 58.151 Section 58.151... Specifications for Dairy Plants Approved for USDA Inspection and Grading Service 1 Packaging and General Identification § 58.151 Packaging and repackaging. (a) Packaging dairy products or cutting and repackaging all...

49 CFR 173.206 - Packaging requirements for chlorosilanes.

Code of Federal Regulations, 2010 CFR

2010-10-01

... 49 Transportation 2 2010-10-01 2010-10-01 false Packaging requirements for chlorosilanes. 173.206...-GENERAL REQUIREMENTS FOR SHIPMENTS AND PACKAGINGS Non-bulk Packaging for Hazardous Materials Other Than Class 1 and Class 7 § 173.206 Packaging requirements for chlorosilanes. (a) When § 172.101 of this...

7 CFR 58.444 - Packaging and repackaging.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 7 Agriculture 3 2010-01-01 2010-01-01 false Packaging and repackaging. 58.444 Section 58.444... Procedures § 58.444 Packaging and repackaging. (a) Packaging rindless cheese or cutting and repackaging all styles of bulk cheese shall be conducted under rigid sanitary conditions. The atmosphere of the packaging...

19 CFR 10.461 - Retail packaging materials and containers.

Code of Federal Regulations, 2010 CFR

2010-04-01

... 19 Customs Duties 1 2010-04-01 2010-04-01 false Retail packaging materials and containers. 10.461... Free Trade Agreement Rules of Origin § 10.461 Retail packaging materials and containers. Packaging... requirement, the value of such packaging materials and containers will be taken into account as originating or...

7 CFR 58.53 - Supervisor of packaging required.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 7 Agriculture 3 2010-01-01 2010-01-01 false Supervisor of packaging required. 58.53 Section 58.53... Packaging Products with Official Identification § 58.53 Supervisor of packaging required. The official....54 through 58.57, shall be done only under the supervision of a supervisor of packaging. The...

21 CFR 310.515 - Patient package inserts for estrogens.

Code of Federal Regulations, 2011 CFR

2011-04-01

... 21 Food and Drugs 5 2011-04-01 2011-04-01 false Patient package inserts for estrogens. 310.515... package inserts for estrogens. (a) Requirement for a patient package insert. FDA concludes that the safe... patient package insert containing information concerning the drug's benefits and risks. An estrogen drug...

Packaging for Food Service

NASA Technical Reports Server (NTRS)

Stilwell, E. J.

1985-01-01

Most of the key areas of concern in packaging the three principle food forms for the space station were covered. It can be generally concluded that there are no significant voids in packaging materials availability or in current packaging technology. However, it must also be concluded that the process by which packaging decisions are made for the space station feeding program will be very synergistic. Packaging selection will depend heavily on the preparation mechanics, the preferred presentation and the achievable disposal systems. It will be important that packaging be considered as an integral part of each decision as these systems are developed.

Environmental Assessment of Packaging: The Consumer Point of View

PubMed

Van Dam YK

1996-09-01

When marketing environmentally responsible packaged products, the producer is confronted with consumer beliefs concerning the environmental friendliness of packaging materials. When making environmentally conscious packaging decisions, these consumer beliefs should be taken into account alongside the technical guidelines. Dutch consumer perceptions of the environmental friendliness of packaged products are reported and compared with the results of a life-cycle analysis assessment. It is shown that consumers judge environmental friendliness mainly from material and returnability. Furthermore, the consumer perception of the environmental friendliness of packaging material is based on the postconsumption waste, whereas the environmental effects of production are ignored. From the consumer beliefs concerning environmental friendliness implications are deduced for packaging policy and for environmental policy.KEY WORDS: Consumer behavior; Environment; Food; Packaging; Perception; Waste

Bi-level microelectronic device package with an integral window

DOEpatents

Peterson, Kenneth A.; Watson, Robert D.

2004-01-06

A package with an integral window for housing a microelectronic device. The integral window is bonded directly to the package without having a separate layer of adhesive material disposed in-between the window and the package. The device can be a semiconductor chip, CCD chip, CMOS chip, VCSEL chip, laser diode, MEMS device, or IMEMS device. The multilayered package can be formed of a LTCC or HTCC cofired ceramic material, with the integral window being simultaneously joined to the package during LTCC or HTCC processing. The microelectronic device can be flip-chip bonded so that the light-sensitive side is optically accessible through the window. The package has at least two levels of circuits for making electrical interconnections to a pair of microelectronic devices. The result is a compact, low-profile package having an integral window that is hermetically sealed to the package prior to mounting and interconnecting the microelectronic device(s).

Why Packaging Is Commercially Vital for Tobacco Corporations.

PubMed

Barraclough, Simon; Gleeson, Deborah

2017-03-01

This study analyses what British American Tobacco (BAT) and its 4 publicly listed Asian subsidiary companies have told their shareholders about the commercial value of tobacco packaging. The discourse on packaging in BAT annual reports was analyzed, revealing themes of modernization, rejuvenation, internationalism, heritage, innovation, value for money, and competitive edge. Packaging was credited with providing existing brands with a competitive edge and enabling the successful "launch" of new ones. Since advertising, sponsorship, and free samples were prohibited in many countries, packaging has become more important for advertising. New brands and brand variants have proliferated. BAT companies have allocated considerable resources to regularly altering packaging for marketing purposes. Clearly, restrictions on packaging will substantially detract from the promotion of the company's brands. The findings provide further evidence from industry sources of the vital function of packaging and further justify plain packaging as an essential part of any comprehensive tobacco control policy.

Food packaging history and innovations.

PubMed

Risch, Sara J

2009-09-23

Food packaging has evolved from simply a container to hold food to something today that can play an active role in food quality. Many packages are still simply containers, but they have properties that have been developed to protect the food. These include barriers to oxygen, moisture, and flavors. Active packaging, or that which plays an active role in food quality, includes some microwave packaging as well as packaging that has absorbers built in to remove oxygen from the atmosphere surrounding the product or to provide antimicrobials to the surface of the food. Packaging has allowed access to many foods year-round that otherwise could not be preserved. It is interesting to note that some packages have actually allowed the creation of new categories in the supermarket. Examples include microwave popcorn and fresh-cut produce, which owe their existence to the unique packaging that has been developed.

The comparison of antimicrobial packaging properties with different applications incorporation method of active material

NASA Astrophysics Data System (ADS)

Anwar, R. W.; Sugiarto; Warsiki, E.

2018-03-01

Contamination after the processing of products during storage, distribution and marketing is one of the main causes of food safety issues. Handling of food products after processing can be done during the packaging process. Antimicrobial (AM) active packaging is one of the concept of packaging product development by utilize the interaction between the product and the packaging environment that can delay the bacterial damage by killing or reducing bacterial growth. The active system is formed by incorporating an antimicrobial agent against a packaging matrix that will function as a carrier. Many incorporation methods have been developed in this packaging-making concept which were direct mixing, polishing, and encapsulation. The aims of this research were to examine the different of the AM packaging performances including its stability and effectiveness of its function that would be produced by three different methods. The stability of the packaging function was analyzed by looking at the diffusivity of the active ingredient to the matrix using SEM. The effectiveness was analyzed by the ability of the packaging to prevent the growing of the microbial. The results showed that different incorporation methods resulted on different characteristics of the AM packaging.

Application of Au-Sn eutectic bonding in hermetic radio-frequency microelectromechanical system wafer level packaging

NASA Astrophysics Data System (ADS)

Wang, Qian; Choa, Sung-Hoon; Kim, Woonbae; Hwang, Junsik; Ham, Sukjin; Moon, Changyoul

2006-03-01

Development of packaging is one of the critical issues toward realizing commercialization of radio-frequency-microelectromechanical system (RF-MEMS) devices. The RF-MEMS package should be designed to have small size, hermetic protection, good RF performance, and high reliability. In addition, packaging should be conducted at sufficiently low temperature. In this paper, a low-temperature hermetic wafer level packaging scheme for the RF-MEMS devices is presented. For hermetic sealing, Au-Sn eutectic bonding technology at temperatures below 300°C is used. Au-Sn multilayer metallization with a square loop of 70 µm in width is performed. The electrical feed-through is achieved by the vertical through-hole via filling with electroplated Cu. The size of the MEMS package is 1 mm × 1 mm × 700 µm. The shear strength and hermeticity of the package satisfies the requirements of MIL-STD-883F. Any organic gases or contamination are not observed inside the package. The total insertion loss for the packaging is 0.075 dB at 2 GHz. Furthermore, the robustness of the package is demonstrated by observing no performance degradation and physical damage of the package after several reliability tests.

Development of low cost, high reliability sealing techniques for hybrid microcircuit packages. Phase 2, supplement 1: Moisture permeation of adhesive-sealed hybrid microcircuit packages

NASA Technical Reports Server (NTRS)

Perkins, K. L.; Licari, J. J.

1978-01-01

The susceptibility of adhesive-sealed ceramic packages to moisture permeation was investigated. The two adhesives, Ablebond 789-1 and Epo-Tek H77, were evaluated as package sealants. These adhesives were previously selected as the most promising candidates for this application from a group of ten adhesives. Ceramic packages sealed with these adhesives were exposed to temperature-humidity conditions of 25 C/98 percent RH, 50 C/60 percent RH, 50 C/98 percent RH, and 85 C/85 percent RH and their moisture contents using were monitored solid state moisture sensors sealed inside them. Five packages were tested at each of these exposures - two ceramic packages sealed with each of the two adhesives and one seam-sealed gold-plated Kovar package. This latter package was included to serve as a control. The results showed that the adhesive-sealed packages were not hermetic to moisture. The rates at which moisture entered the packages increased with the severity of the exposure environments (i.e., higher temperatures and higher moisture vapor pressures) with greater dependence on temperature than on moisture vapor pressure.

High-performance packaging for monolithic microwave and millimeter-wave integrated circuits

NASA Technical Reports Server (NTRS)

Shalkhauser, K. A.; Li, K.; Shih, Y. C.

1992-01-01

Packaging schemes were developed that provide low-loss, hermetic enclosure for advanced monolithic microwave and millimeter-wave integrated circuits (MMICs). The package designs are based on a fused quartz substrate material that offers improved radio frequency (RF) performance through 44 gigahertz (GHz). The small size and weight of the packages make them appropriate for a variety of applications, including phased array antenna systems. Packages were designed in two forms; one for housing a single MMIC chip, the second in the form of a multi-chip phased array module. The single chip array module was developed in three separate sizes, for chips of different geometry and frequency requirements. The phased array module was developed to address packaging directly for antenna applications, and includes transmission line and interconnect structures to support multi-element operation. All packages are fabricated using fused quartz substrate materials. As part of the packaging effort, a test fixture was developed to interface the single chip packages to conventional laboratory instrumentation for characterization of the packaged devices. The package and test fixture designs were both developed in a generic sense, optimizing performance for a wide range of possible applications and devices.

Essential interventions: implementation strategies and proposed packages of care

PubMed Central

2014-01-01

In an effort to accelerate progress towards achieving Millennium Development Goal (MDG) 4 and 5, provision of essential reproductive, maternal, newborn and child health (RMNCH) interventions is being considered. Not only should a state-of-the-art approach be taken for services delivered to the mother, neonate and to the child, but services must also be deployed across the household to hospital continuum of care approach and in the form of packages. The paper proposed several packages for improved maternal, newborn and child health that can be delivered across RMNCH continuum of care. These packages include: supportive care package for women to promote awareness related to healthy pre-pregnancy and pregnancy interventions; nutritional support package for mother to improve supplementation of essential nutrients and micronutrients; antenatal care package to detect, treat and manage infectious and noninfectious diseases and promote immunization; high risk care package to manage preeclampsia and eclampsia in pregnancy; childbirth package to promote support during labor and importance of skilled birth attendance during labor; essential newborn care package to support healthy newborn care practices; and child health care package to prevent and manage infections. This paper further discussed the implementation strategies for employing these interventions at scale. PMID:25178110

Healthy choice?: Exploring how children evaluate the healthfulness of packaged foods.

PubMed

Elliott, Charlene; Brierley, Meaghan

2012-11-06

Today's supermarket contains hundreds of packaged foods specifically targeted at children. Yet research has shown that children are confused by the various visual messages found on packaged food products. This study explores children's nutrition knowledge with regard to packaged food products, to uncover strengths and difficulties they have in evaluating the healthfulness of these foods. Focus groups were conducted with children (grades 1-6). Particular attention was paid to the ways children made use of what they know about nutrition when faced with the visual elements and appeals presented on food packaging. Children relied heavily on packages' written and visual aspects--including colour, images, spokes-characters, front-of-package claims--to assess the healthfulness of a food product. These elements interfere with children's ability to make healthy choices when it comes to packaged foods. Choosing healthy packaged foods is challenging for children due to competing sets of knowledge: one pertains to their understanding of visual, associational cues; the other, to translating their understanding of nutrition to packaged foods. Canada's Food Guide, along with the curriculum taught to Canadian children at schools, does not appear to provide children with the tools necessary to navigate a food environment dominated by packaged foods.

Fabrication and Performance of MEMS-Based Pressure Sensor Packages Using Patterned Ultra-Thick Photoresists

PubMed Central

Chen, Lung-Tai; Chang, Jin-Sheng; Hsu, Chung-Yi; Cheng, Wood-Hi

2009-01-01

A novel plastic packaging of a piezoresistive pressure sensor using a patterned ultra-thick photoresist is experimentally and theoretically investigated. Two pressure sensor packages of the sacrifice-replacement and dam-ring type were used in this study. The characteristics of the packaged pressure sensors were investigated by using a finite-element (FE) model and experimental measurements. The results show that the thermal signal drift of the packaged pressure sensor with a small sensing-channel opening or with a thin silicon membrane for the dam-ring approach had a high packaging induced thermal stress, leading to a high temperature coefficient of span (TCO) response of −0.19% span/°C. The results also show that the thermal signal drift of the packaged pressure sensors with a large sensing-channel opening for sacrifice-replacement approach significantly reduced packaging induced thermal stress, and hence a low TCO response of −0.065% span/°C. However, the packaged pressure sensors of both the sacrifice-replacement and dam-ring type still met the specification −0.2% span/°C of the unpackaged pressure sensor. In addition, the size of proposed packages was 4 × 4 × 1.5 mm3 which was about seven times less than the commercialized packages. With the same packaging requirement, the proposed packaging approaches may provide an adequate solution for use in other open-cavity sensors, such as gas sensors, image sensors, and humidity sensors. PMID:22454580

Which community care for patients with schizophrenic disorders? Packages of care provided by Departments of Mental Health in Lombardy (Italy).

PubMed

Lora, Antonio; Cosentino, Ugo; Gandini, Anna; Zocchetti, Carlo

2007-01-01

The treatment of schizophrenic disorders is the most important challenge for community care. The analysis focuses on packages of care provided to 23.602 patients with a ICD-10 diagnosis of schizophrenic disorder and treated in 2001 by the Departments of Mental Health in Lombardy, Italy. Packages of care refer to a mix of treatments provided to each patient during the year by different settings. Direct costs of the packages were calculated. Linear Discriminant Analysis has been used to link socio-demographic and diagnostic sub-groups of the patients to packages of care. People with schizophrenic disorders received relatively few care packages: only four packages involved more than 5%. Two thirds of the patients received only care provided by Community Mental Health Centres. In the other two packages with a percentage over 5%, the activity was provided by CMHCs, jointly with General Hospitals or Day Care Facilities. Complex care packages were rare (only 6%). As well as the intensity, also the variety of care provided by CMHCs increased with the complexity of care packages. In Lombardy more than half of the resources were spent for schizophrenia. The range of the costs per package was very wide. LDA failed to link characteristics of the patients to packages of care. Care packages are useful tools to understand better how mental health system works, how resources have been spent and to point out problems in the quality of care.

GENERAL PURPOSE ADA PACKAGES

NASA Technical Reports Server (NTRS)

Klumpp, A. R.

1994-01-01

Ten families of subprograms are bundled together for the General-Purpose Ada Packages. The families bring to Ada many features from HAL/S, PL/I, FORTRAN, and other languages. These families are: string subprograms (INDEX, TRIM, LOAD, etc.); scalar subprograms (MAX, MIN, REM, etc.); array subprograms (MAX, MIN, PROD, SUM, GET, and PUT); numerical subprograms (EXP, CUBIC, etc.); service subprograms (DATE_TIME function, etc.); Linear Algebra II; Runge-Kutta integrators; and three text I/O families of packages. In two cases, a family consists of a single non-generic package. In all other cases, a family comprises a generic package and its instances for a selected group of scalar types. All generic packages are designed to be easily instantiated for the types declared in the user facility. The linear algebra package is LINRAG2. This package includes subprograms supplementing those in NPO-17985, An Ada Linear Algebra Package Modeled After HAL/S (LINRAG). Please note that LINRAG2 cannot be compiled without LINRAG. Most packages have widespread applicability, although some are oriented for avionics applications. All are designed to facilitate writing new software in Ada. Several of the packages use conventions introduced by other programming languages. A package of string subprograms is based on HAL/S (a language designed for the avionics software in the Space Shuttle) and PL/I. Packages of scalar and array subprograms are taken from HAL/S or generalized current Ada subprograms. A package of Runge-Kutta integrators is patterned after a built-in MAC (MIT Algebraic Compiler) integrator. Those packages modeled after HAL/S make it easy to translate existing HAL/S software to Ada. The General-Purpose Ada Packages program source code is available on two 360K 5.25" MS-DOS format diskettes. The software was developed using VAX Ada v1.5 under DEC VMS v4.5. It should be portable to any validated Ada compiler and it should execute either interactively or in batch. The largest package requires 205K of main memory on a DEC VAX running VMS. The software was developed in 1989, and is a copyrighted work with all copyright vested in NASA.

Reliability of High I/O High Density CCGA Interconnect Electronic Packages under Extreme Thermal Environment

NASA Technical Reports Server (NTRS)

Ramesham, Rajeshuni

2012-01-01

This paper provides the experimental test results of advanced CCGA packages tested in extreme temperature thermal environments. Standard optical inspection and x-ray non-destructive inspection tools were used to assess the reliability of high density CCGA packages for deep space extreme temperature missions. Ceramic column grid array (CCGA) packages have been increasing in use based on their advantages such as high interconnect density, very good thermal and electrical performances, compatibility with standard surface-mount packaging assembly processes, and so on. CCGA packages are used in space applications such as in logic and microprocessor functions, telecommunications, payload electronics, and flight avionics. As these packages tend to have less solder joint strain relief than leaded packages or more strain relief over lead-less chip carrier packages, the reliability of CCGA packages is very important for short-term and long-term deep space missions. We have employed high density CCGA 1152 and 1272 daisy chained electronic packages in this preliminary reliability study. Each package is divided into several daisy-chained sections. The physical dimensions of CCGA1152 package is 35 mm x 35 mm with a 34 x 34 array of columns with a 1 mm pitch. The dimension of the CCGA1272 package is 37.5 mm x 37.5 mm with a 36 x 36 array with a 1 mm pitch. The columns are made up of 80% Pb/20%Sn material. CCGA interconnect electronic package printed wiring polyimide boards have been assembled and inspected using non-destructive x-ray imaging techniques. The assembled CCGA boards were subjected to extreme temperature thermal atmospheric cycling to assess their reliability for future deep space missions. The resistance of daisy-chained interconnect sections were monitored continuously during thermal cycling. This paper provides the experimental test results of advanced CCGA packages tested in extreme temperature thermal environments. Standard optical inspection and x-ray non-destructive inspection tools were used to assess the reliability of high density CCGA packages for deep space extreme temperature missions. Keywords: Extreme temperatures, High density CCGA qualification, CCGA reliability, solder joint failures, optical inspection, and x-ray inspection.

Packaging Technologies for High Temperature Electronics and Sensors

NASA Technical Reports Server (NTRS)

Chen, Liang-Yu; Hunter, Gary W.; Neudeck, Philip G.; Beheim, Glenn M.; Spry, David J.; Meredith, Roger D.

2013-01-01

This paper reviews ceramic substrates and thick-film metallization based packaging technologies in development for 500 C silicon carbide (SiC) electronics and sensors. Prototype high temperature ceramic chip-level packages and printed circuit boards (PCBs) based on ceramic substrates of aluminum oxide (Al2O3) and aluminum nitride (AlN) have been designed and fabricated. These ceramic substrate-based chip-level packages with gold (Au) thick-film metallization have been electrically characterized at temperatures up to 550 C. A 96% alumina based edge connector for a PCB level subsystem interconnection has also been demonstrated recently. The 96% alumina packaging system composed of chip-level packages and PCBs has been tested with high temperature SiC devices at 500 C for over 10,000 hours. In addition to tests in a laboratory environment, a SiC JFET with a packaging system composed of a 96% alumina chip-level package and an alumina printed circuit board mounted on a data acquisition circuit board was launched as a part of the MISSE-7 suite to the International Space Station via a Shuttle mission. This packaged SiC transistor was successfully tested in orbit for eighteen months. A spark-plug type sensor package designed for high temperature SiC capacitive pressure sensors was developed. This sensor package combines the high temperature interconnection system with a commercial high temperature high pressure stainless steel seal gland (electrical feed-through). Test results of a packaged high temperature capacitive pressure sensor at 500 C are also discussed. In addition to the pressure sensor package, efforts for packaging high temperature SiC diode-based gas chemical sensors are in process.

Packaging Technologies for High Temperature Electronics and Sensors

NASA Technical Reports Server (NTRS)

Chen, Liangyu; Hunter, Gary W.; Neudeck, Philip G.; Beheim, Glenn M.; Spry, David J.; Meredith, Roger D.

2013-01-01

This paper reviews ceramic substrates and thick-film metallization based packaging technologies in development for 500degC silicon carbide (SiC) electronics and sensors. Prototype high temperature ceramic chip-level packages and printed circuit boards (PCBs) based on ceramic substrates of aluminum oxide (Al2O3) and aluminum nitride (AlN) have been designed and fabricated. These ceramic substrate-based chiplevel packages with gold (Au) thick-film metallization have been electrically characterized at temperatures up to 550degC. A 96% alumina based edge connector for a PCB level subsystem interconnection has also been demonstrated recently. The 96% alumina packaging system composed of chip-level packages and PCBs has been tested with high temperature SiC devices at 500degC for over 10,000 hours. In addition to tests in a laboratory environment, a SiC JFET with a packaging system composed of a 96% alumina chip-level package and an alumina printed circuit board mounted on a data acquisition circuit board was launched as a part of the MISSE-7 suite to the International Space Station via a Shuttle mission. This packaged SiC transistor was successfully tested in orbit for eighteen months. A spark-plug type sensor package designed for high temperature SiC capacitive pressure sensors was developed. This sensor package combines the high temperature interconnection system with a commercial high temperature high pressure stainless steel seal gland (electrical feed-through). Test results of a packaged high temperature capacitive pressure sensor at 500degC are also discussed. In addition to the pressure sensor package, efforts for packaging high temperature SiC diode-based gas chemical sensors are in process.

Multilayered microelectronic device package with an integral window

DOEpatents

Peterson, Kenneth A.; Watson, Robert D.

2003-01-01

An apparatus for packaging of microelectronic devices is disclosed, wherein the package includes an integral window. The microelectronic device can be a semiconductor chip, a CCD chip, a CMOS chip, a VCSEL chip, a laser diode, a MEMS device, or a IMEMS device. The package can comprise, for example, a cofired ceramic frame or body. The package has an internal stepped structure made of a plurality of plates, with apertures, which are patterned with metallized conductive circuit traces. The microelectronic device can be flip-chip bonded on the plate to these traces, and oriented so that the light-sensitive side is optically accessible through the window. A cover lid can be attached to the opposite side of the package. The result is a compact, low-profile package, having an integral window that can be hermetically-sealed. The package body can be formed by low-temperature cofired ceramic (LTCC) or high-temperature cofired ceramic (HTCC) multilayer processes with the window being simultaneously joined (e.g. cofired) to the package body during LTCC or HTCC processing. Multiple chips can be located within a single package, according to some embodiments. The cover lid can include a window. The apparatus is particularly suited for packaging of MEMS devices, since the number of handling steps is greatly reduced, thereby reducing the potential for contamination. The integral window can further include a lens for optically transforming light passing through the window. The package can include an array of binary optic lenslets made integral with the window. The package can include an electrically-switched optical modulator, such as a lithium niobate window attached to the package, for providing a very fast electrically-operated shutter.

Short communication: Effect of active food packaging materials on fluid milk quality and shelf life.

PubMed

Wong, Dana E; Goddard, Julie M

2014-01-01

Active packaging, in which active agents are embedded into or on the surface of food packaging materials, can enhance the nutritive value, economics, and stability of food, as well as enable in-package processing. In one embodiment of active food packaging, lactase was covalently immobilized onto packaging films for in-package lactose hydrolysis. In prior work, lactase was covalently bound to low-density polyethylene using polyethyleneimine and glutaraldehyde cross-linkers to form the packaging film. Because of the potential contaminants of proteases, lipases, and spoilage organisms in typical enzyme preparations, the goal of the current work was to determine the effect of immobilized-lactase active packaging technology on unanticipated side effects, such as shortened shelf-life and reduced product quality. Results suggested no evidence of lipase or protease activity on the active packaging films, indicating that such active packaging films could enable in-package lactose hydrolysis without adversely affecting product quality in terms of dairy protein or lipid stability. Storage stability studies indicated that lactase did not migrate from the film over a 49-d period, and that dry storage resulted in 13.41% retained activity, whereas wet storage conditions enabled retention of 62.52% activity. Results of a standard plate count indicated that the film modification reagents introduced minor microbial contamination; however, the microbial population remained under the 20,000 cfu/mL limit through the manufacturer's suggested 14-d storage period for all film samples. This suggests that commercially produced immobilized lactase active packaging should use purified cross-linkers and enzymes. Characterization of unanticipated effects of active packaging on food quality reported here is important in demonstrating the commercial potential of such technologies. Copyright © 2014 American Dairy Science Association. Published by Elsevier Inc. All rights reserved.

Common Ada (tradename) Missile Package (CAMP) Project. Missile Software Parts. Volume 8. Detail Design Document

DTIC Science & Technology

1988-03-01

PACKAGE BODY ) TLCSC P661 (CATALOG #P106-0) This package contains the CAMP parts required to do the vaypoint steering portion of navigation. The...3.3.4.1.6 PROCESSING The following describes the processing performed by this part: package body WaypointSteering is package body ...Steering_Vector_Operations is separate; package body Steering_Vector_Operations_with_Arcsin is separate; procedure Compute Turn_Angle_and Direction (UnitNormal C

Function Package for Computing Quantum Resource Measures

NASA Astrophysics Data System (ADS)

Huang, Zhiming

2018-05-01

In this paper, we present a function package for to calculate quantum resource measures and dynamics of open systems. Our package includes common operators and operator lists, frequently-used functions for computing quantum entanglement, quantum correlation, quantum coherence, quantum Fisher information and dynamics in noisy environments. We briefly explain the functions of the package and illustrate how to use the package with several typical examples. We expect that this package is a useful tool for future research and education.

19 CFR 11.3 - Package and notice requirements for cigars and cigarettes; package requirements for cigarette...

Code of Federal Regulations, 2013 CFR

2013-04-01

... cigarettes; package requirements for cigarette papers and tubes. 11.3 Section 11.3 Customs Duties U.S... STAMPING; MARKING Packing and Stamping § 11.3 Package and notice requirements for cigars and cigarettes; package requirements for cigarette papers and tubes. Exemptions from tax on cigars, cigarettes, and...

19 CFR 11.3 - Package and notice requirements for cigars and cigarettes; package requirements for cigarette...

Code of Federal Regulations, 2011 CFR

2011-04-01

... cigarettes; package requirements for cigarette papers and tubes. 11.3 Section 11.3 Customs Duties U.S... STAMPING; MARKING Packing and Stamping § 11.3 Package and notice requirements for cigars and cigarettes; package requirements for cigarette papers and tubes. Exemptions from tax on cigars, cigarettes, and...

49 CFR 173.474 - Quality control for construction of packaging.

Code of Federal Regulations, 2011 CFR

2011-10-01

... 49 Transportation 2 2011-10-01 2011-10-01 false Quality control for construction of packaging. 173...-GENERAL REQUIREMENTS FOR SHIPMENTS AND PACKAGINGS Class 7 (Radioactive) Materials § 173.474 Quality control for construction of packaging. Prior to the first use of any packaging for the shipment of Class 7...

49 CFR 173.474 - Quality control for construction of packaging.

Code of Federal Regulations, 2010 CFR

2010-10-01

... 49 Transportation 2 2010-10-01 2010-10-01 false Quality control for construction of packaging. 173...-GENERAL REQUIREMENTS FOR SHIPMENTS AND PACKAGINGS Class 7 (Radioactive) Materials § 173.474 Quality control for construction of packaging. Prior to the first use of any packaging for the shipment of Class 7...

49 CFR 178.935 - Standards for wooden Large Packagings.

Code of Federal Regulations, 2010 CFR

2010-10-01

... Packagings. (i) Natural wood used in the construction of Large Packagings must be well-seasoned, commercially...) Reconstituted wood used in the construction of Large Packagings must be water resistant reconstituted wood such... Packaging types are designated: (1) 50C natural wood. (2) 50D plywood. (3) 50F reconstituted wood. (b...

9 CFR 381.144 - Packaging materials.

Code of Federal Regulations, 2011 CFR

2011-01-01

... 9 Animals and Animal Products 2 2011-01-01 2011-01-01 false Packaging materials. 381.144 Section... Packaging materials. (a) Edible products may not be packaged in a container which is composed in whole or in... to health. All packaging materials must be safe for the intended use within the meaning of section...

49 CFR 178.523 - Standards for composite packagings with inner glass, porcelain, or stoneware receptacles.

Code of Federal Regulations, 2010 CFR

2010-10-01

... 49 Transportation 2 2010-10-01 2010-10-01 false Standards for composite packagings with inner... Packaging Standards § 178.523 Standards for composite packagings with inner glass, porcelain, or stoneware receptacles. (a) The following are identification codes for composite packagings with inner receptacles of...

19 CFR 11.3 - Package and notice requirements for cigars and cigarettes; package requirements for cigarette...

Code of Federal Regulations, 2014 CFR

2014-04-01

... cigarettes; package requirements for cigarette papers and tubes. 11.3 Section 11.3 Customs Duties U.S... STAMPING; MARKING Packing and Stamping § 11.3 Package and notice requirements for cigars and cigarettes; package requirements for cigarette papers and tubes. Exemptions from tax on cigars, cigarettes, and...

19 CFR 11.3 - Package and notice requirements for cigars and cigarettes; package requirements for cigarette...

Code of Federal Regulations, 2010 CFR

2010-04-01

... cigarettes; package requirements for cigarette papers and tubes. 11.3 Section 11.3 Customs Duties U.S... STAMPING; MARKING Packing and Stamping § 11.3 Package and notice requirements for cigars and cigarettes; package requirements for cigarette papers and tubes. Exemptions from tax on cigars, cigarettes, and...

19 CFR 11.3 - Package and notice requirements for cigars and cigarettes; package requirements for cigarette...

Code of Federal Regulations, 2012 CFR

2012-04-01

... cigarettes; package requirements for cigarette papers and tubes. 11.3 Section 11.3 Customs Duties U.S... STAMPING; MARKING Packing and Stamping § 11.3 Package and notice requirements for cigars and cigarettes; package requirements for cigarette papers and tubes. Exemptions from tax on cigars, cigarettes, and...

19 CFR 10.922 - Retail packaging materials and containers.

Code of Federal Regulations, 2014 CFR

2014-04-01

... 19 Customs Duties 1 2014-04-01 2014-04-01 false Retail packaging materials and containers. 10.922... Trade Promotion Agreement Rules of Origin § 10.922 Retail packaging materials and containers. (a) Effect on tariff shift rule. Packaging materials and containers in which a good is packaged for retail sale...

19 CFR 10.922 - Retail packaging materials and containers.

Code of Federal Regulations, 2013 CFR

2013-04-01

... 19 Customs Duties 1 2013-04-01 2013-04-01 false Retail packaging materials and containers. 10.922... Trade Promotion Agreement Rules of Origin § 10.922 Retail packaging materials and containers. (a) Effect on tariff shift rule. Packaging materials and containers in which a good is packaged for retail sale...

19 CFR 10.461 - Retail packaging materials and containers.

Code of Federal Regulations, 2011 CFR

2011-04-01

... 19 Customs Duties 1 2011-04-01 2011-04-01 false Retail packaging materials and containers. 10.461... Free Trade Agreement Rules of Origin § 10.461 Retail packaging materials and containers. Packaging materials and containers in which a good is packaged for retail sale, if classified with the good for which...

19 CFR 10.922 - Retail packaging materials and containers.

Code of Federal Regulations, 2012 CFR

2012-04-01

... 19 Customs Duties 1 2012-04-01 2012-04-01 false Retail packaging materials and containers. 10.922... Trade Promotion Agreement Rules of Origin § 10.922 Retail packaging materials and containers. (a) Effect on tariff shift rule. Packaging materials and containers in which a good is packaged for retail sale...

19 CFR 10.1022 - Retail packaging materials and containers.

Code of Federal Regulations, 2012 CFR

2012-04-01

... 19 Customs Duties 1 2012-04-01 2012-04-01 false Retail packaging materials and containers. 10.1022... Free Trade Agreement Rules of Origin § 10.1022 Retail packaging materials and containers. (a) Effect on tariff shift rule. Packaging materials and containers in which a good is packaged for retail sale, if...

19 CFR 10.3022 - Retail packaging materials and containers.

Code of Federal Regulations, 2013 CFR

2013-04-01

... 19 Customs Duties 1 2013-04-01 2013-04-01 false Retail packaging materials and containers. 10.3022...-Colombia Trade Promotion Agreement Rules of Origin § 10.3022 Retail packaging materials and containers. (a) Effect on tariff shift rule. Packaging materials and containers in which a good is packaged for retail...

19 CFR 10.461 - Retail packaging materials and containers.

Code of Federal Regulations, 2013 CFR

2013-04-01

... 19 Customs Duties 1 2013-04-01 2013-04-01 false Retail packaging materials and containers. 10.461... Free Trade Agreement Rules of Origin § 10.461 Retail packaging materials and containers. Packaging materials and containers in which a good is packaged for retail sale, if classified with the good for which...

19 CFR 10.3022 - Retail packaging materials and containers.

Code of Federal Regulations, 2014 CFR

2014-04-01

... 19 Customs Duties 1 2014-04-01 2014-04-01 false Retail packaging materials and containers. 10.3022...-Colombia Trade Promotion Agreement Rules of Origin § 10.3022 Retail packaging materials and containers. (a) Effect on tariff shift rule. Packaging materials and containers in which a good is packaged for retail...

19 CFR 10.461 - Retail packaging materials and containers.

Code of Federal Regulations, 2014 CFR

2014-04-01

... 19 Customs Duties 1 2014-04-01 2014-04-01 false Retail packaging materials and containers. 10.461... Free Trade Agreement Rules of Origin § 10.461 Retail packaging materials and containers. Packaging materials and containers in which a good is packaged for retail sale, if classified with the good for which...

19 CFR 10.539 - Retail packaging materials and containers.

Code of Federal Regulations, 2013 CFR

2013-04-01

... 19 Customs Duties 1 2013-04-01 2013-04-01 false Retail packaging materials and containers. 10.539...-Singapore Free Trade Agreement Rules of Origin § 10.539 Retail packaging materials and containers. Packaging materials and containers in which a good is packaged for retail sale, if classified with the good for which...

19 CFR 10.1022 - Retail packaging materials and containers.

Code of Federal Regulations, 2014 CFR

2014-04-01

... 19 Customs Duties 1 2014-04-01 2014-04-01 false Retail packaging materials and containers. 10.1022... Free Trade Agreement Rules of Origin § 10.1022 Retail packaging materials and containers. (a) Effect on tariff shift rule. Packaging materials and containers in which a good is packaged for retail sale, if...

19 CFR 10.461 - Retail packaging materials and containers.

Code of Federal Regulations, 2012 CFR

2012-04-01

... 19 Customs Duties 1 2012-04-01 2012-04-01 false Retail packaging materials and containers. 10.461... Free Trade Agreement Rules of Origin § 10.461 Retail packaging materials and containers. Packaging materials and containers in which a good is packaged for retail sale, if classified with the good for which...

19 CFR 10.539 - Retail packaging materials and containers.

Code of Federal Regulations, 2014 CFR

2014-04-01

... 19 Customs Duties 1 2014-04-01 2014-04-01 false Retail packaging materials and containers. 10.539...-Singapore Free Trade Agreement Rules of Origin § 10.539 Retail packaging materials and containers. Packaging materials and containers in which a good is packaged for retail sale, if classified with the good for which...

19 CFR 10.539 - Retail packaging materials and containers.

Code of Federal Regulations, 2011 CFR

2011-04-01

... 19 Customs Duties 1 2011-04-01 2011-04-01 false Retail packaging materials and containers. 10.539...-Singapore Free Trade Agreement Rules of Origin § 10.539 Retail packaging materials and containers. Packaging materials and containers in which a good is packaged for retail sale, if classified with the good for which...

19 CFR 10.539 - Retail packaging materials and containers.

Code of Federal Regulations, 2012 CFR

2012-04-01

... 19 Customs Duties 1 2012-04-01 2012-04-01 false Retail packaging materials and containers. 10.539...-Singapore Free Trade Agreement Rules of Origin § 10.539 Retail packaging materials and containers. Packaging materials and containers in which a good is packaged for retail sale, if classified with the good for which...

19 CFR 10.1022 - Retail packaging materials and containers.

Code of Federal Regulations, 2013 CFR

2013-04-01

... 19 Customs Duties 1 2013-04-01 2013-04-01 false Retail packaging materials and containers. 10.1022... Free Trade Agreement Rules of Origin § 10.1022 Retail packaging materials and containers. (a) Effect on tariff shift rule. Packaging materials and containers in which a good is packaged for retail sale, if...

49 CFR 373.105 - Low value packages.

Code of Federal Regulations, 2010 CFR

2010-10-01

... 49 Transportation 5 2010-10-01 2010-10-01 false Low value packages. 373.105 Section 373.105... Carrier Receipts and Bills § 373.105 Low value packages. The carrier and shipper may elect to waive the... “low value” packages. This includes the option of shipping such packages under the provisions of 49 U.S...

10 CFR 60.135 - Criteria for the waste package and its components.

Code of Federal Regulations, 2011 CFR

2011-01-01

... IN GEOLOGIC REPOSITORIES Technical Criteria Design Criteria for the Waste Package § 60.135 Criteria for the waste package and its components. (a) High-level-waste package design in general. (1) Packages for HLW shall be designed so that the in situ chemical, physical, and nuclear properties of the waste...

10 CFR 60.135 - Criteria for the waste package and its components.

Code of Federal Regulations, 2013 CFR

2013-01-01

... IN GEOLOGIC REPOSITORIES Technical Criteria Design Criteria for the Waste Package § 60.135 Criteria for the waste package and its components. (a) High-level-waste package design in general. (1) Packages for HLW shall be designed so that the in situ chemical, physical, and nuclear properties of the waste...

10 CFR 60.135 - Criteria for the waste package and its components.

Code of Federal Regulations, 2014 CFR

2014-01-01

... IN GEOLOGIC REPOSITORIES Technical Criteria Design Criteria for the Waste Package § 60.135 Criteria for the waste package and its components. (a) High-level-waste package design in general. (1) Packages for HLW shall be designed so that the in situ chemical, physical, and nuclear properties of the waste...

10 CFR 60.135 - Criteria for the waste package and its components.

Code of Federal Regulations, 2012 CFR

2012-01-01

... IN GEOLOGIC REPOSITORIES Technical Criteria Design Criteria for the Waste Package § 60.135 Criteria for the waste package and its components. (a) High-level-waste package design in general. (1) Packages for HLW shall be designed so that the in situ chemical, physical, and nuclear properties of the waste...

7 CFR 993.501 - Consumer package of prunes.

Code of Federal Regulations, 2011 CFR

2011-01-01

... 7 Agriculture 8 2011-01-01 2011-01-01 false Consumer package of prunes. 993.501 Section 993.501... CALIFORNIA Pack Specification as to Size Definitions § 993.501 Consumer package of prunes. Consumer package of prunes means consumer package as defined in § 993.22. Effective Date Note: At 70 FR 30613, May 27...

Neutron fluxes in test reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Youinou, Gilles Jean-Michel

Communicate the fact that high-power water-cooled test reactors such as the Advanced Test Reactor (ATR), the High Flux Isotope Reactor (HFIR) or the Jules Horowitz Reactor (JHR) cannot provide fast flux levels as high as sodium-cooled fast test reactors. The memo first presents some basics physics considerations about neutron fluxes in test reactors and then uses ATR, HFIR and JHR as an illustration of the performance of modern high-power water-cooled test reactors.

"Plain packaging" regulations for tobacco products: the impact of standardizing the color and design of cigarette packs.

PubMed

Hammond, David

2010-01-01

Tobacco packaging and labeling policies have emerged as prominent and cost-effective tobacco control measures. Although packaging policies have primarily focused on health warnings, there is growing recognition of the importance of packaging as a marketing tool for the tobacco industry. The current paper reviews evidence on the potential impact of standardizing the color and design of tobacco packages -so called "plain" packaging. The evidence indicates three primary benefits of plain packaging: increasing the effectiveness of health warnings, reducing false health beliefs about cigarettes, and reducing brand appeal especially among youth and young adults. Overall, the research to date suggests that "plain" packaging regulations would be an effective tobacco control measure, particularly in jurisdictions with comprehensive restrictions on other forms of marketing.

Quality and safety aspects of meat products as affected by various physical manipulations of packaging materials.

PubMed

Lee, Keun Taik

2010-09-01

This article explores the effects of physically manipulated packaging materials on the quality and safety of meat products. Recently, innovative measures for improving quality and extending the shelf-life of packaged meat products have been developed, utilizing technologies including barrier film, active packaging, nanotechnology, microperforation, irradiation, plasma and far-infrared ray (FIR) treatments. Despite these developments, each technology has peculiar drawbacks which will need to be addressed by meat scientists in the future. To develop successful meat packaging systems, key product characteristics affecting stability, environmental conditions during storage until consumption, and consumers' packaging expectations must all be taken into consideration. Furthermore, the safety issues related to packaging materials must also be taken into account when processing, packaging and storing meat products.

Packaging films for electronic and space-related hardware

NASA Astrophysics Data System (ADS)

Shon, E. M.; Hamberg, O.

1985-08-01

Flexible packaging films are used to bag and/or wrap precision cleaned electronic or space hardware to protect them from environmental degradation during shipping and storage. Selection of packaging films depends on a knowledge of product requirements and packaging film characteristics. The literature presently available on protective packaging films has been updated to include new materials and to amplify space-related applications. Presently available packaging film materials are compared for their various characteristics: electrostatic discharge (ESD) control, flame retardancy, water vapor transmission rate, particulate shedding, molecular contamination, and transparency. The tradeoff between product requirements and the characteristics of the packaging films available are discussed. Selection considerations are given for the application of specific materials of space hardware-related applications. Applications for intimate, environmental, and electrostatic protective packaging are discussed.

Definition of Small Gram Quantity Contents for Type B Radioactive Material Transportation Packages: Activity-Based Content Limitations

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sitaraman, S; Kim, S; Biswas, D

2010-10-27

Since the 1960's, the Department of Transportation Specification (DOT Spec) 6M packages have been used extensively for transportation of Type B quantities of radioactive materials between Department of Energy (DOE) facilities, laboratories, and productions sites. However, due to the advancement of packaging technology, the aging of the 6M packages, and variability in the quality of the packages, the DOT implemented a phased elimination of the 6M specification packages (and other DOT Spec packages) in favor of packages certified to meet federal performance requirements. DOT issued the final rule in the Federal Register on October 1, 2004 requiring that use ofmore » the DOT Specification 6M be discontinued as of October 1, 2008. A main driver for the change was the fact that the 6M specification packagings were not supported by a Safety Analysis Report for Packaging (SARP) that was compliant with Title 10 of the Code of Federal Regulations part 71 (10 CFR 71). Therefore, materials that would have historically been shipped in 6M packages are being identified as contents in Type B (and sometimes Type A fissile) package applications and addenda that are to be certified under the requirements of 10 CFR 71. The requirements in 10 CFR 71 include that the Safety Analysis Report for Packaging (SARP) must identify the maximum radioactivity of radioactive constituents and maximum quantities of fissile constituents (10 CFR 71.33(b)(1) and 10 CFR 71.33(b)(2)), and that the application (i.e., SARP submittal or SARP addendum) demonstrates that the external dose rate (due to the maximum radioactivity of radioactive constituents and maximum quantities of fissile constituents) on the surface of the packaging (i.e., package and contents) not exceed 200 mrem/hr (10 CFR 71.35(a), 10 CFR 71.47(a)). It has been proposed that a 'Small Gram Quantity' of radioactive material be defined, such that, when loaded in a transportation package, the dose rates at external points of an unshielded packaging not exceed the regulatory limits prescribed by 10 CFR 71 for non-exclusive shipments. The mass of each radioisotope presented in this paper is limited by the radiation dose rate on the external surface of the package, which per the regulatory limit should not exceed 200 mrem/hr. The results presented are a compendium of allowable masses of a variety of different isotopes (with varying impurity levels of beryllium in some of the actinide isotopes) that, when loaded in an unshielded packaging, do not result in an external dose rate on the surface of the package that exceeds 190 mrem/hr (190 mrem/hr was chosen to provide 5% conservatism relative to the regulatory limit). These mass limits define the term 'Small Gram Quantity' (SGQ) contents in the context of radioactive material transportation packages. The term SGQ is isotope-specific and pertains to contents in radioactive material transportation packages that do not require shielding and still satisfy the external dose rate requirements. Since these calculated mass limits are for contents without shielding, they are conservative for packaging materials that provide some limited shielding or if the contents are placed into a shielded package. The isotopes presented in this paper were chosen as the isotopes that Department of Energy (DOE) sites most likely need to ship. Other more rarely shipped isotopes, along with industrial and medical isotopes, are planned to be included in subsequent extensions of this work.« less

Data from fitting Gaussian process models to various data sets using eight Gaussian process software packages.

PubMed

Erickson, Collin B; Ankenman, Bruce E; Sanchez, Susan M

2018-06-01

This data article provides the summary data from tests comparing various Gaussian process software packages. Each spreadsheet represents a single function or type of function using a particular input sample size. In each spreadsheet, a row gives the results for a particular replication using a single package. Within each spreadsheet there are the results from eight Gaussian process model-fitting packages on five replicates of the surface. There is also one spreadsheet comparing the results from two packages performing stochastic kriging. These data enable comparisons between the packages to determine which package will give users the best results.

Packaging Technologies for 500 C SiC Electronics and Sensors: Challenges in Material Science and Technology

NASA Technical Reports Server (NTRS)

Chen, Liang-Yu; Neudeck, Philip G.; Behelm, Glenn M.; Spry, David J.; Meredith, Roger D.; Hunter, Gary W.

2015-01-01

This paper presents ceramic substrates and thick-film metallization based packaging technologies in development for 500C silicon carbide (SiC) electronics and sensors. Prototype high temperature ceramic chip-level packages and printed circuit boards (PCBs) based on ceramic substrates of aluminum oxide (Al2O3) and aluminum nitride (AlN) have been designed and fabricated. These ceramic substrate-based chip-level packages with gold (Au) thick-film metallization have been electrically characterized at temperatures up to 550C. The 96 alumina packaging system composed of chip-level packages and PCBs has been successfully tested with high temperature SiC discrete transistor devices at 500C for over 10,000 hours. In addition to tests in a laboratory environment, a SiC junction field-effect-transistor (JFET) with a packaging system composed of a 96 alumina chip-level package and an alumina printed circuit board was tested on low earth orbit for eighteen months via a NASA International Space Station experiment. In addition to packaging systems for electronics, a spark-plug type sensor package based on this high temperature interconnection system for high temperature SiC capacitive pressure sensors was also developed and tested. In order to further significantly improve the performance of packaging system for higher packaging density, higher operation frequency, power rating, and even higher temperatures, some fundamental material challenges must be addressed. This presentation will discuss previous development and some of the challenges in material science (technology) to improve high temperature dielectrics for packaging applications.

Effect on moisture permeability of typewriting on unit dose package surfaces.

PubMed

Rackson, J T; Zellhofer, M J; Birmingham, P H

1984-10-01

The effects of typewriting on labels of two unit dose packages with respect to moisture permeability were examined. Using an electric typewriter, a standard label format was imprinted on two different types of class A unit dose packages: (1) a heat-sealed paper-backed foil and cellofilm strip pouch, and (2) a copolyester and polyethylene multiple-cup blister with a heat-sealed paper-backed foil and cellofilm cover. The labels were typed at various typing-element impact settings. The official USP test for water permeation was then performed on typed packages and untyped control packages. The original untyped packages were confirmed to be USP class A quality. The packages for which successively harder impact settings were used showed a corresponding increase in moisture permeability. This resulted in a lowering of USP package ratings from class A to class B and D, some of which would be unsuitable for use in any unit dose system under current FDA repackaging standards. Typing directly onto the label of a unit dose package before it is sealed will most likely damage the package and possibly make it unfit for use. Pharmacists who must type labels for the unit dose packages studied should use the lowest possible typewriter impact setting and test for damage using the USP moisture-permeation test.

Hidden costs of antiretroviral treatment: the public health efficiency of drug packaging.

PubMed

Andreu-Crespo, Àngels; Llibre, Josep M; Cardona-Peitx, Glòria; Sala-Piñol, Ferran; Clotet, Bonaventura; Bonafont-Pujol, Xavier

2015-01-01

While the overall percentage of unused antiretroviral medicines returned to the hospital pharmacy is low, their cost is quite high. Adverse events, treatment failure, pharmacokinetic interactions, pregnancy, or treatment simplification are common reasons for unplanned treatment changes. Socially inefficient antiretroviral packages prevent the reuse of drugs returned to the hospital pharmacy. We defined antiretroviral package categories based on the excellence of drug packaging and analyzed the number of pills and costs of drugs returned during a period of 1 year in a hospital-based HIV unit attending to 2,413 treated individuals. A total of 6,090 pills (34% of all returned antiretrovirals) - with a cost of 47,139.91 € - would be totally lost, mainly due to being packed up in the lowest efficiency packages. Newer treatments are packaged in low-excellence categories of packages, thus favoring the maintenance of these hidden costs in the near future. Therefore, costs of this low-efficiency drug packaging, where medication packages are started but not completed, in high-cost medications are substantial and should be properly addressed. Any improvement in the packaging by the manufacturer, and favoring the choice of drugs supplied through efficient packages (when efficacy, toxicity, and convenience are similar), should minimize the treatment expenditures paid by national health budgets.

Hidden costs of antiretroviral treatment: the public health efficiency of drug packaging

PubMed Central

Andreu-Crespo, Àngels; Llibre, Josep M; Cardona-Peitx, Glòria; Sala-Piñol, Ferran; Clotet, Bonaventura; Bonafont-Pujol, Xavier

2015-01-01

While the overall percentage of unused antiretroviral medicines returned to the hospital pharmacy is low, their cost is quite high. Adverse events, treatment failure, pharmacokinetic interactions, pregnancy, or treatment simplification are common reasons for unplanned treatment changes. Socially inefficient antiretroviral packages prevent the reuse of drugs returned to the hospital pharmacy. We defined antiretroviral package categories based on the excellence of drug packaging and analyzed the number of pills and costs of drugs returned during a period of 1 year in a hospital-based HIV unit attending to 2,413 treated individuals. A total of 6,090 pills (34% of all returned antiretrovirals) – with a cost of 47,139.91€ – would be totally lost, mainly due to being packed up in the lowest efficiency packages. Newer treatments are packaged in low-excellence categories of packages, thus favoring the maintenance of these hidden costs in the near future. Therefore, costs of this low-efficiency drug packaging, where medication packages are started but not completed, in high-cost medications are substantial and should be properly addressed. Any improvement in the packaging by the manufacturer, and favoring the choice of drugs supplied through efficient packages (when efficacy, toxicity, and convenience are similar), should minimize the treatment expenditures paid by national health budgets. PMID:26273190

MEMS packaging: state of the art and future trends

NASA Astrophysics Data System (ADS)

Bossche, Andre; Cotofana, Carmen V. B.; Mollinger, Jeff R.

1998-07-01

Now that the technology for Integrated sensor and MEMS devices has become sufficiently mature to allow mass production, it is expected that the prices of bare chips will drop dramatically. This means that the package prices will become a limiting factor in market penetration, unless low cost packaging solutions become available. This paper will discuss the developments in packaging technology. Both single-chip and multi-chip packaging solutions will be addressed. It first starts with a discussion on the different requirements that have to be met; both from a device point of view (open access paths to the environment, vacuum cavities, etc.) and from the application point of view (e.g. environmental hostility). Subsequently current technologies are judged on their applicability for MEMS and sensor packaging and a forecast is given for future trends. It is expected that the large majority of sensing devices will be applied in relative friendly environments for which plastic packages would suffice. Therefore, on the short term an important role is foreseen for recently developed plastic packaging techniques such as precision molding and precision dispensing. Just like in standard electronic packaging, complete wafer level packaging methods for sensing devices still have a long way to go before they can compete with the highly optimized and automated plastic packaging processes.

Savannah River Site Eastern Transportation Hub: A Concept For a DOE Eastern Packaging, Staging and Maintenance Center - 13143

DOE Office of Scientific and Technical Information (OSTI.GOV)

England, Jeffery L.; Adams, Karen; Maxted, Maxcine

2013-07-01

The Department of Energy (DOE) is working to de-inventory sites and consolidate hazardous materials for processing and disposal. The DOE administers a wide range of certified shipping packages for the transport of hazardous materials to include Special Nuclear Material (SNM), radioactive materials, sealed sources and radioactive wastes. A critical element to successful and safe transportation of these materials is the availability of certified shipping packages. There are over seven thousand certified packagings (i.e., Type B/Type AF) utilized within the DOE for current missions. The synergistic effects of consolidated maintenance, refurbishment, testing, certification, and costing of these services would allow formore » efficient management of the packagings inventory and to support anticipated future in-commerce shipping needs. The Savannah River Site (SRS) receives and ships radioactive materials (including SNM) and waste on a regular basis for critical missions such as consolidated storage, stabilization, purification, or disposition using H-Canyon and HB-Line. The Savannah River National Laboratory (SRNL) has the technical capability and equipment for all aspects of packaging management. SRS has the only active material processing facility in the DOE complex and is one of the sites of choice for nuclear material consolidation. SRS is a logical location to perform maintenance and periodic testing of the DOE fleet of certified packagings. This initiative envisions a DOE Eastern Packaging Staging and Maintenance Center (PSMC) at the SRS and a western hub at the Nevada National Security Site (NNSS), an active DOE Regional Disposal Site. The PSMC's would be the first place DOE would go to meet their radioactive packaging needs and the primary locations projects would go to disposition excess packaging for beneficial reuse. These two hubs would provide the centralized management of a packaging fleet rather than the current approach to design, procure, maintain and dispose of packagings on a project-by-project basis. This initiative provides significant savings in packaging costs and acceleration of project schedules. In addition to certified packaging, the PSMC would be well suited for select designs of 7A Type A packaging and Industrial Packaging. (authors)« less

Youth Work Training Package Review: More of the Same or Radical Rationalisation?

ERIC Educational Resources Information Center

Corney, Tim; Broadbent, Robyn

2007-01-01

The development of a national youth work training package in Australia began over 15 years ago. The current package sits under the umbrella of the general Community Services Industry Training Package. The first stage of a review of this package has been completed and the subsequent report not only confirms the recent trend towards the…

Think INSIDE the Box: Package Engineering

ERIC Educational Resources Information Center

Snyder, Mark; Painter, Donna

2014-01-01

Most products people purchase, keep in their homes, and often discard, are typically packaged in some way. Packaging is so prevalent in daily lives that many of take it for granted. That is by design-the expectation of good packaging is that it exists for the sake of the product. The primary purposes of any package (to contain, inform, display,…

75 FR 33637 - United States v. Bemis Company, Inc., et al.; Public Comments and Response on Proposed Final...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-06-14

... shredded natural cheese packaged for retail sale and flexible-packaging shrink bags for fresh meat in the..., production, and sale of flexible-packaging rollstock for chunk, sliced, and shredded natural cheese packaged..., sliced, and shredded natural cheese packaged for retail sale from two to one, would have eliminated...

19 CFR 158.4 - Liability of carrier for lost or missing packages.

Code of Federal Regulations, 2012 CFR

2012-04-01

... 19 Customs Duties 2 2012-04-01 2012-04-01 false Liability of carrier for lost or missing packages... EXPORTED Lost or Missing Packages and Deficiencies in Contents of Packages § 158.4 Liability of carrier for lost or missing packages. Upon a joint determination or independent determination of quantity as set...

19 CFR 158.4 - Liability of carrier for lost or missing packages.

Code of Federal Regulations, 2013 CFR

2013-04-01

... 19 Customs Duties 2 2013-04-01 2013-04-01 false Liability of carrier for lost or missing packages... EXPORTED Lost or Missing Packages and Deficiencies in Contents of Packages § 158.4 Liability of carrier for lost or missing packages. Upon a joint determination or independent determination of quantity as set...

19 CFR 158.4 - Liability of carrier for lost or missing packages.

Code of Federal Regulations, 2011 CFR

2011-04-01

... 19 Customs Duties 2 2011-04-01 2011-04-01 false Liability of carrier for lost or missing packages... EXPORTED Lost or Missing Packages and Deficiencies in Contents of Packages § 158.4 Liability of carrier for lost or missing packages. Upon a joint determination or independent determination of quantity as set...

19 CFR 158.4 - Liability of carrier for lost or missing packages.

Code of Federal Regulations, 2014 CFR

2014-04-01

... 19 Customs Duties 2 2014-04-01 2014-04-01 false Liability of carrier for lost or missing packages... EXPORTED Lost or Missing Packages and Deficiencies in Contents of Packages § 158.4 Liability of carrier for lost or missing packages. Upon a joint determination or independent determination of quantity as set...

19 CFR 158.4 - Liability of carrier for lost or missing packages.

Code of Federal Regulations, 2010 CFR

2010-04-01

... 19 Customs Duties 2 2010-04-01 2010-04-01 false Liability of carrier for lost or missing packages... EXPORTED Lost or Missing Packages and Deficiencies in Contents of Packages § 158.4 Liability of carrier for lost or missing packages. Upon a joint determination or independent determination of quantity as set...

A Description and Analysis of the German Packaging Take-Back System

ERIC Educational Resources Information Center

Nakajima, Nina; Vanderburg, Willem H.

2006-01-01

The German packaging ordinance is an example of legislated extended producer responsibility (also known as product take-back). Consumers can leave packaging with retailers, and packagers are required to pay for their recycling and disposal. It can be considered to be successful in reducing waste, spurring the redesign of packaging to be more…

49 CFR Appendix B to Part 173 - Procedure for Testing Chemical Compatibility and Rate of Permeation in Plastic Packaging and...

Code of Federal Regulations, 2011 CFR

2011-10-01

... Rate of Permeation in Plastic Packaging and Receptacles B Appendix B to Part 173 Transportation Other... Plastic Packaging and Receptacles 1. The purpose of this procedure is to determine the chemical compatibility and permeability of liquid hazardous materials packaged in plastic packaging and receptacles...

49 CFR 173.28 - Reuse, reconditioning and remanufacture of packagings.

Code of Federal Regulations, 2012 CFR

2012-10-01

... non-bulk packaging. A non-bulk packaging used more than once must conform to the following provisions and limitations: (1) A non-bulk packaging which, upon inspection, shows evidence of a reduction in... thickness of 1.0 mm (0.039 inch). (6) A previously used non-bulk packaging may be reused for the shipment of...

49 CFR 173.28 - Reuse, reconditioning and remanufacture of packagings.

Code of Federal Regulations, 2011 CFR

2011-10-01

... non-bulk packaging. A non-bulk packaging used more than once must conform to the following provisions and limitations: (1) A non-bulk packaging which, upon inspection, shows evidence of a reduction in... thickness of 1.0 mm (0.039 inch). (6) A previously used non-bulk packaging may be reused for the shipment of...

49 CFR 173.28 - Reuse, reconditioning and remanufacture of packagings.

Code of Federal Regulations, 2013 CFR

2013-10-01

... non-bulk packaging. A non-bulk packaging used more than once must conform to the following provisions and limitations: (1) A non-bulk packaging which, upon inspection, shows evidence of a reduction in... thickness of 1.0 mm (0.039 inch). (6) A previously used non-bulk packaging may be reused for the shipment of...

49 CFR 173.28 - Reuse, reconditioning and remanufacture of packagings.

Code of Federal Regulations, 2014 CFR

2014-10-01

... non-bulk packaging. A non-bulk packaging used more than once must conform to the following provisions and limitations: (1) A non-bulk packaging which, upon inspection, shows evidence of a reduction in... thickness of 1.0 mm (0.039 inch). (6) A previously used non-bulk packaging may be reused for the shipment of...

49 CFR 173.28 - Reuse, reconditioning and remanufacture of packagings.

Code of Federal Regulations, 2010 CFR

2010-10-01

... non-bulk packaging. A non-bulk packaging used more than once must conform to the following provisions and limitations: (1) A non-bulk packaging which, upon inspection, shows evidence of a reduction in... thickness of 1.0 mm (0.039 inch). (6) A previously used non-bulk packaging may be reused for the shipment of...

49 CFR Appendix B to Part 173 - Procedure for Testing Chemical Compatibility and Rate of Permeation in Plastic Packaging and...

Code of Federal Regulations, 2010 CFR

2010-10-01

... Rate of Permeation in Plastic Packaging and Receptacles B Appendix B to Part 173 Transportation Other... Plastic Packaging and Receptacles 1. The purpose of this procedure is to determine the chemical compatibility and permeability of liquid hazardous materials packaged in plastic packaging and receptacles...

49 CFR Appendix B to Part 173 - Procedure for Testing Chemical Compatibility and Rate of Permeation in Plastic Packaging and...

Code of Federal Regulations, 2012 CFR

2012-10-01

... Rate of Permeation in Plastic Packaging and Receptacles B Appendix B to Part 173 Transportation Other... Plastic Packaging and Receptacles 1. The purpose of this procedure is to determine the chemical compatibility and permeability of liquid hazardous materials packaged in plastic packaging and receptacles...

49 CFR Appendix B to Part 173 - Procedure for Testing Chemical Compatibility and Rate of Permeation in Plastic Packaging and...

Code of Federal Regulations, 2013 CFR

2013-10-01

... Rate of Permeation in Plastic Packaging and Receptacles B Appendix B to Part 173 Transportation Other... Plastic Packaging and Receptacles 1. The purpose of this procedure is to determine the chemical compatibility and permeability of liquid hazardous materials packaged in plastic packaging and receptacles...

21 CFR 800.12 - Contact lens solutions and tablets; tamper-resistant packaging.

Code of Federal Regulations, 2010 CFR

2010-04-01

...-resistant retail packages, there is the opportunity for the malicious adulteration of these products with... confidence in the security of the packages of over-the-counter (OTC) health care products. The Food and Drug... used to make such a solution for retail sale that is not packaged in a tamper-resistant package and...

21 CFR 800.12 - Contact lens solutions and tablets; tamper-resistant packaging.

Code of Federal Regulations, 2011 CFR

2011-04-01

...-resistant retail packages, there is the opportunity for the malicious adulteration of these products with... confidence in the security of the packages of over-the-counter (OTC) health care products. The Food and Drug... used to make such a solution for retail sale that is not packaged in a tamper-resistant package and...

21 CFR 800.12 - Contact lens solutions and tablets; tamper-resistant packaging.

Code of Federal Regulations, 2012 CFR

2012-04-01

...-resistant retail packages, there is the opportunity for the malicious adulteration of these products with... confidence in the security of the packages of over-the-counter (OTC) health care products. The Food and Drug... used to make such a solution for retail sale that is not packaged in a tamper-resistant package and...

21 CFR 800.12 - Contact lens solutions and tablets; tamper-resistant packaging.

Code of Federal Regulations, 2013 CFR

2013-04-01

...-resistant retail packages, there is the opportunity for the malicious adulteration of these products with... confidence in the security of the packages of over-the-counter (OTC) health care products. The Food and Drug... used to make such a solution for retail sale that is not packaged in a tamper-resistant package and...

21 CFR 800.12 - Contact lens solutions and tablets; tamper-resistant packaging.

Code of Federal Regulations, 2014 CFR

2014-04-01

...-resistant retail packages, there is the opportunity for the malicious adulteration of these products with... confidence in the security of the packages of over-the-counter (OTC) health care products. The Food and Drug... used to make such a solution for retail sale that is not packaged in a tamper-resistant package and...

Reactor engineering support of operations at the Davis-Besse nuclear power station

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kelley, D.B.

1995-12-31

Reactor engineering functions differ greatly from unit to unit; however, direct support of the reactor operators during reactor startups and operational transients is common to all units. This paper summarizes the support the reactor engineers provide the reactor operators during reactor startups and power changes through the use of automated computer programs at the Davis-Besse nuclear power station.

Packaging's Contribution for the Effectiveness of the Space Station's Food Service Operation

NASA Technical Reports Server (NTRS)

Rausch, B. A.

1985-01-01

Storage limitations will have a major effect on space station food service. For example: foods with low bulk density such as ice cream, bread, cake, standard type potato chips and other low density snacks, flaked cereals, etc., will exacerbate the problem of space limitations; package containers are inherently volume consuming and refuse creating; and the useful observation that the optimum package is no package at all leads to the tentative conclusion that the least amount of packaging per unit of food, consistent with storage, aesthetics, preservation, cleanliness, cost and disposal criteria, is the most practical food package for the space station. A series of trade offs may have to be made to arrive at the most appropriate package design for a particular type of food taking all the criteria into account. Some of these trade offs are: single serve vs. bulk; conventional oven vs. microwave oven; nonmetallic aseptically vs. non-aseptically packaged foods; and comparison of aseptic vs. nonaseptic food packages. The advantages and disadvantages are discussed.

Practical fundamentals of glass, rubber, and plastic sterile packaging systems.

PubMed

Sacha, Gregory A; Saffell-Clemmer, Wendy; Abram, Karen; Akers, Michael J

2010-01-01

Sterile product packaging systems consist of glass, rubber, and plastic materials that are in intimate contact with the formulation. These materials can significantly affect the stability of the formulation. The interaction between the packaging materials and the formulation can also affect the appropriate delivery of the product. Therefore, a parenteral formulation actually consists of the packaging system as well as the product that it contains. However, the majority of formulation development time only considers the product that is contained in the packaging system. Little time is spent studying the interaction of the packaging materials with the contents. Interaction between the packaging and the contents only becomes a concern when problems are encountered. For this reason, there are few scientific publications that describe the available packaging materials, their advantages and disadvantages, and their important product attributes. This article was created as a reference for product development and describes some of the packaging materials and systems that are available for parenteral products.

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

NASA Astrophysics Data System (ADS)

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

Natural biopolymer-based nanocomposite films for packaging applications.

PubMed

Rhim, Jong-Whan; Ng, Perry K W

2007-01-01

Concerns on environmental waste problems caused by non-biodegradable petrochemical-based plastic packaging materials as well as the consumer's demand for high quality food products has caused an increasing interest in developing biodegradable packaging materials using annually renewable natural biopolymers such as polysaccharides and proteins. Inherent shortcomings of natural polymer-based packaging materials such as low mechanical properties and low water resistance can be recovered by applying a nanocomposite technology. Polymer nanocomposites, especially natural biopolymer-layered silicate nanocomposites, exhibit markedly improved packaging properties due to their nanometer size dispersion. These improvements include increased modulus and strength, decreased gas permeability, and increased water resistance. Additionally, biologically active ingredients can be added to impart the desired functional properties to the resulting packaging materials. Consequently, natural biopolymer-based nanocomposite packaging materials with bio-functional properties have a huge potential for application in the active food packaging industry. In this review, recent advances in the preparation of natural biopolymer-based films and their nanocomposites, and their potential use in packaging applications are addressed.

Signalling product healthiness through symbolic package cues: Effects of package shape and goal congruence on consumer behaviour.

PubMed

van Ooijen, Iris; Fransen, Marieke L; Verlegh, Peeter W J; Smit, Edith G

2017-02-01

Three studies show that product packaging shape serves as a cue that communicates healthiness of food products. Inspired by embodiment accounts, we show that packaging that simulates a slim body shape acts as a symbolic cue for product healthiness (e.g., low in calories), as opposed to packaging that simulates a wide body shape. Furthermore, we show that the effect of slim package shape on consumer behaviour is goal dependent. Whereas simulation of a slim (vs. wide) body shape increases choice likelihood and product attitude when consumers have a health-relevant shopping goal, packaging shape does not affect these outcomes when consumers have a hedonic shopping goal. In Study 3, we adopt a realistic shopping paradigm using a shelf with authentic products, and find that a slim (as opposed to wide) package shape increases on-shelf product recognition and increases product attitude for healthy products. We discuss results and implications regarding product positioning and the packaging design process. Copyright © 2016 Elsevier Ltd. All rights reserved.

10 CFR 2.337 - Evidence at a hearing.

Code of Federal Regulations, 2011 CFR

2011-01-01

... chapter by the Director, Office of Nuclear Reactor Regulation, Director, Office of New Reactors, or... the Director, Office of Nuclear Reactor Regulation, Director, Office of New Reactors, or Director... the Director, Office of Nuclear Reactor Regulation, Director, Office of New Reactors, or Director...

Nuclear Reactors. Revised.

ERIC Educational Resources Information Center

Hogerton, John F.

This publication is one of a series of information booklets for the general public published by the United States Atomic Energy Commission. Among the topics discussed are: How Reactors Work; Reactor Design; Research, Teaching, and Materials Testing; Reactors (Research, Teaching and Materials); Production Reactors; Reactors for Electric Power…

10 CFR 2.337 - Evidence at a hearing.

Code of Federal Regulations, 2012 CFR

2012-01-01

... chapter by the Director, Office of Nuclear Reactor Regulation, Director, Office of New Reactors, or... the Director, Office of Nuclear Reactor Regulation, Director, Office of New Reactors, or Director... the Director, Office of Nuclear Reactor Regulation, Director, Office of New Reactors, or Director...

49 CFR 173.36 - Hazardous materials in Large Packagings.

Code of Federal Regulations, 2013 CFR

2013-10-01

... an inner packaging is constructed of paper or flexible plastic, the inner packaging must be replaced...) The Large Packaging is free from corrosion, contamination, cracks, cuts, or other damage which would...

49 CFR 173.36 - Hazardous materials in Large Packagings.

Code of Federal Regulations, 2014 CFR

2014-10-01

... an inner packaging is constructed of paper or flexible plastic, the inner packaging must be replaced...) The Large Packaging is free from corrosion, contamination, cracks, cuts, or other damage which would...

Reviews.

ERIC Educational Resources Information Center

Radcliffe, George; And Others

1988-01-01

Reviews three software packages: 1) a package containing 68 programs covering general topics in chemistry; 2) a package dealing with acid-base titration curves and allows for variables to be changed; 3) a chemistry tutorial and drill package. (MVL)

14 CFR 417.1 - General information.

Code of Federal Regulations, 2014 CFR

2014-01-01

... package, (3) Preliminary and final flight data packages, (4) A tailored version of EWR 127-1, (5) Range...) Missile system pre-launch safety package, (3) Preliminary and final flight data packages, (4) A tailored...

14 CFR 417.1 - General information.

Code of Federal Regulations, 2013 CFR

2013-01-01

... package, (3) Preliminary and final flight data packages, (4) A tailored version of EWR 127-1, (5) Range...) Missile system pre-launch safety package, (3) Preliminary and final flight data packages, (4) A tailored...

14 CFR 417.1 - General information.

Code of Federal Regulations, 2011 CFR

2011-01-01

... package, (3) Preliminary and final flight data packages, (4) A tailored version of EWR 127-1, (5) Range...) Missile system pre-launch safety package, (3) Preliminary and final flight data packages, (4) A tailored...

14 CFR 417.1 - General information.

Code of Federal Regulations, 2012 CFR

2012-01-01

... package, (3) Preliminary and final flight data packages, (4) A tailored version of EWR 127-1, (5) Range...) Missile system pre-launch safety package, (3) Preliminary and final flight data packages, (4) A tailored...

Migration and sorption phenomena in packaged foods.

PubMed

Gnanasekharan, V; Floros, J D

1997-10-01

Rapidly developing analytical capabilities and continuously evolving stringent regulations have made food/package interactions a subject of intense research. This article focuses on: (1) the migration of package components such as oligomers and monomers, processing aids, additives, and residual reactants in to packaged foods, and (2) sorption of food components such as flavors, lipids, and moisture into packages. Principles of diffusion and thermodynamics are utilized to describe the mathematics of migration and sorption. Mathematical models are developed from first principles, and their applicability is illustrated using numerical simulations and published data. Simulations indicate that available models are system (polymer-penetrant) specific. Furthermore, some models best describe the early stages of migration/sorption, whereas others should be used for the late stages of these phenomena. Migration- and/or sorption-related problems with respect to glass, metal, paper-based and polymeric packaging materials are discussed, and their importance is illustrated using published examples. The effects of migrating and absorbed components on food safety, quality, and the environment are presented for various foods and packaging materials. The impact of currently popular packaging techniques such as microwavable, ovenable, and retortable packaging on migration and sorption are discussed with examples. Analytical techniques for investigating migration and sorption phenomena in food packaging are critically reviewed, with special emphasis on the use and characteristics of food-simulating liquids (FSLs). Finally, domestic and international regulations concerning migration in packaged foods, and their impact on food packaging is briefly presented.

Effectiveness of some recent antimicrobial packaging concepts.

PubMed

Vermeiren, L; Devlieghere, F; Debevere, J

2002-01-01

A new type of active packaging is the combination of food-packaging materials with antimicrobial substances to control microbial surface contamination of foods. For both migrating and non-migrating antimicrobial materials, intensive contact between the food product and packaging material is required and therefore potential food applications include especially vacuum or skin-packaged products, e.g. vacuum-packaged meat, fish, poultry or cheese. Several antimicrobial compounds have been combined with different types of carriers (plastic and rubber articles, paper-based materials, textile fibrils and food-packaging materials). Until now, however, few antimicrobial concepts have found applications as a food-packaging material. Antimicrobial packaging materials cannot legally be used in the EU at the moment. The potential use would require amendments of several different legal texts involving areas such as food additives, food packaging, hygiene, etc. The main objective of this paper is to provide a state of the art about the different types of antimicrobial concepts, their experimental development and commercialization, and to present a case study summarizing the results of investigations on the feasibility of a low-density polyethylene (LDPE)-film containing triclosan to inhibit microbial growth on food surfaces and consequently prolong shelf-life or improve microbial food safety. In contrast with the strong antimicrobial effect in in-vitro simulated vacuum-packaged conditions against the psychrotrophic food pathogen L. monocytogenes, the 1000 mg kg(-1) containing triclosan film did not effectively reduce spoilage bacteria and growth of L. monocytogenes on refrigerated vacuum-packaged chicken breasts stored at 7 degrees C.

Nuclear reactor construction with bottom supported reactor vessel

DOEpatents

Sharbaugh, John E.

1987-01-01

An improved liquid metal nuclear reactor construction has a reactor core and a generally cylindrical reactor vessel for holding a large pool of low pressure liquid metal coolant and housing the core within the pool. The reactor vessel has an open top end, a closed flat bottom end wall and a continuous cylindrical closed side wall interconnecting the top end and bottom end wall. The reactor also has a generally cylindrical concrete containment structure surrounding the reactor vessel and being formed by a cylindrical side wall spaced outwardly from the reactor vessel side wall and a flat base mat spaced below the reactor vessel bottom end wall. A central support pedestal is anchored to the containment structure base mat and extends upwardly therefrom to the reactor vessel and upwardly therefrom to the reactor core so as to support the bottom end wall of the reactor vessel and the lower end of the reactor core in spaced apart relationship above the containment structure base mat. Also, an annular reinforced support structure is disposed in the reactor vessel on the bottom end wall thereof and extends about the lower end of the core so as to support the periphery thereof. In addition, an annular support ring having a plurality of inward radially extending linear members is disposed between the containment structure base mat and the bottom end of the reactor vessel wall and is connected to and supports the reactor vessel at its bottom end on the containment structure base mat so as to allow the reactor vessel to expand radially but substantially prevent any lateral motions that might be imposed by the occurrence of a seismic event. The reactor construction also includes a bed of insulating material in sand-like granular form, preferably being high density magnesium oxide particles, disposed between the containment structure base mat and the bottom end wall of the reactor vessel and uniformly supporting the reactor vessel at its bottom end wall on the containment structure base mat so as to insulate the reactor vessel bottom end wall from the containment structure base mat and allow the reactor vessel bottom end wall to freely expand as it heats up while providing continuous support thereof. Further, a deck is supported upon the side wall of the containment structure above the top open end of the reactor vessel, and a plurality of serially connected extendible and retractable annular bellows extend between the deck and the top open end of the reactor vessel and flexibly and sealably interconnect the reactor vessel at its top end to the deck. An annular guide ring is disposed on the containment structure and extends between its side wall and the top open end of the reactor vessel for providing lateral support of the reactor vessel top open end by limiting imposition of lateral loads on the annular bellows by the occurrence of a lateral seismic event.

Request for Naval Reactors Comment on Proposed Prometheus Space Flight Nuclear Reactor High Tier Reactor Safety Requirements and for Naval Reactors Approval to Transmit These Requirements to JPL

DOE Office of Scientific and Technical Information (OSTI.GOV)

D. Kokkinos

2005-04-28

The purpose of this letter is to request Naval Reactors comments on the nuclear reactor high tier requirements for the PROMETHEUS space flight reactor design, pre-launch operations, launch, ascent, operation, and disposal, and to request Naval Reactors approval to transmit these requirements to Jet Propulsion Laboratory to ensure consistency between the reactor safety requirements and the spacecraft safety requirements. The proposed PROMETHEUS nuclear reactor high tier safety requirements are consistent with the long standing safety culture of the Naval Reactors Program and its commitment to protecting the health and safety of the public and the environment. In addition, the philosophymore » on which these requirements are based is consistent with the Nuclear Safety Policy Working Group recommendations on space nuclear propulsion safety (Reference 1), DOE Nuclear Safety Criteria and Specifications for Space Nuclear Reactors (Reference 2), the Nuclear Space Power Safety and Facility Guidelines Study of the Applied Physics Laboratory.« less

Heat transfer analysis of cylindrical anaerobic reactors with different sizes: a heat transfer model.

PubMed

Liu, Jiawei; Zhou, Xingqiu; Wu, Jiangdong; Gao, Wen; Qian, Xu

2017-10-01

The temperature is the essential factor that influences the efficiency of anaerobic reactors. During the operation of the anaerobic reactor, the fluctuations of ambient temperature can cause a change in the internal temperature of the reactor. Therefore, insulation and heating measures are often used to maintain anaerobic reactor's internal temperature. In this paper, a simplified heat transfer model was developed to study heat transfer between cylindrical anaerobic reactors and their surroundings. Three cylindrical reactors of different sizes were studied, and the internal relations between ambient temperature, thickness of insulation, and temperature fluctuations of the reactors were obtained at different reactor sizes. The model was calibrated by a sensitivity analysis, and the calibrated model was well able to predict reactor temperature. The Nash-Sutcliffe model efficiency coefficient was used to assess the predictive power of heat transfer models. The Nash coefficients of the three reactors were 0.76, 0.60, and 0.45, respectively. The model can provide reference for the thermal insulation design of cylindrical anaerobic reactors.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bradbury, Andrew M.

The invention relates to a novel phagemid display system for packaging phagemid DNA into phagemid particles which completely avoids the use of helper phage. The system of the invention incorporates the use of bacterial packaging cell lines which have been transformed with helper plasmids containing all required phage proteins but not the packaging signals. The absence of packaging signals in these helper plasmids prevents their DNA from being packaged in the bacterial cell, which provides a number of significant advantages over the use of both standard and modified helper phage. Packaged phagemids expressing a protein or peptide of interest, inmore » fusion with a phage coat protein such as g3p, are generated simply by transfecting phagemid into the packaging cell line.« less

Solvent refined coal reactor quench system

DOEpatents

Thorogood, Robert M.

1983-01-01

There is described an improved SRC reactor quench system using a condensed product which is recycled to the reactor and provides cooling by evaporation. In the process, the second and subsequent reactors of a series of reactors are cooled by the addition of a light oil fraction which provides cooling by evaporation in the reactor. The vaporized quench liquid is recondensed from the reactor outlet vapor stream.

Solvent refined coal reactor quench system

DOEpatents

Thorogood, R.M.

1983-11-08

There is described an improved SRC reactor quench system using a condensed product which is recycled to the reactor and provides cooling by evaporation. In the process, the second and subsequent reactors of a series of reactors are cooled by the addition of a light oil fraction which provides cooling by evaporation in the reactor. The vaporized quench liquid is recondensed from the reactor outlet vapor stream. 1 fig.

19 CFR 158.3 - Allowance for lost or missing packages included in an entry summary.

Code of Federal Regulations, 2012 CFR

2012-04-01

... 19 Customs Duties 2 2012-04-01 2012-04-01 false Allowance for lost or missing packages included in..., DAMAGED, ABANDONED, OR EXPORTED Lost or Missing Packages and Deficiencies in Contents of Packages § 158.3 Allowance for lost or missing packages included in an entry summary. Allowance shall be made in the...

19 CFR 158.3 - Allowance for lost or missing packages included in an entry summary.

Code of Federal Regulations, 2013 CFR

2013-04-01

... 19 Customs Duties 2 2013-04-01 2013-04-01 false Allowance for lost or missing packages included in..., DAMAGED, ABANDONED, OR EXPORTED Lost or Missing Packages and Deficiencies in Contents of Packages § 158.3 Allowance for lost or missing packages included in an entry summary. Allowance shall be made in the...

19 CFR 158.3 - Allowance for lost or missing packages included in an entry summary.

Code of Federal Regulations, 2010 CFR

2010-04-01

... 19 Customs Duties 2 2010-04-01 2010-04-01 false Allowance for lost or missing packages included in..., DAMAGED, ABANDONED, OR EXPORTED Lost or Missing Packages and Deficiencies in Contents of Packages § 158.3 Allowance for lost or missing packages included in an entry summary. Allowance shall be made in the...

19 CFR 158.3 - Allowance for lost or missing packages included in an entry summary.

Code of Federal Regulations, 2011 CFR

2011-04-01

... 19 Customs Duties 2 2011-04-01 2011-04-01 false Allowance for lost or missing packages included in..., DAMAGED, ABANDONED, OR EXPORTED Lost or Missing Packages and Deficiencies in Contents of Packages § 158.3 Allowance for lost or missing packages included in an entry summary. Allowance shall be made in the...

19 CFR 158.3 - Allowance for lost or missing packages included in an entry summary.

Code of Federal Regulations, 2014 CFR

2014-04-01

... 19 Customs Duties 2 2014-04-01 2014-04-01 false Allowance for lost or missing packages included in..., DAMAGED, ABANDONED, OR EXPORTED Lost or Missing Packages and Deficiencies in Contents of Packages § 158.3 Allowance for lost or missing packages included in an entry summary. Allowance shall be made in the...

Options for reducing food waste by quality-controlled logistics using intelligent packaging along the supply chain.

PubMed

Heising, Jenneke K; Claassen, G D H; Dekker, Matthijs

2017-10-01

Optimising supply chain management can help to reduce food waste. This paper describes how intelligent packaging can be used to reduce food waste when used in supply chain management based on quality-controlled logistics (QCL). Intelligent packaging senses compounds in the package that correlate with the critical quality attribute of a food product. The information on the quality of each individual packaged food item that is provided by the intelligent packaging can be used for QCL. In a conceptual approach it is explained that monitoring food quality by intelligent packaging sensors makes it possible to obtain information about the variation in the quality of foods and to use a dynamic expiration date (IP-DED) on a food package. The conceptual approach is supported by quantitative data from simulations on the effect of using the information of intelligent packaging in supply chain management with the goal to reduce food waste. This simulation shows that by using the information on the quality of products that is provided by intelligent packaging, QCL can substantially reduce food waste. When QCL is combined with dynamic pricing based on the predicted expiry dates, a further waste reduction is envisaged.

Implementation and use of direct-flow connections in a coupled ground-water and surface-water model

USGS Publications Warehouse

Swain, Eric D.

1994-01-01

The U.S. Geological Survey's MODFLOW finite-difference ground-water flow model has been coupled with three surface-water packages - the MODBRANCH, River, and Stream packages - to simulate surface water and its interaction with ground water. Prior to the development of the coupling packages, the only interaction between these modeling packages was that leakage values could be passed between MODFLOW and the three surface-water packages. To facilitate wider and more flexible uses of the models, a computer program was developed and added to MODFLOW to allow direct flows or stages to be passed between any of the packages and MODFLOW. The flows or stages calculated in one package can be set as boundary discharges or stages to be used in another package. Several modeling packages can be used in the same simulation depending upon the level of sophistication needed in the various reaches being modeled. This computer program is especially useful when any of the River, Stream, or MODBRANCH packages are used to model a river flowing directly into or out of wetlands in direct connection with the aquifer and represented in the model as an aquifer block. A field case study is shown to illustrate an application.

Study of the effect of post-packaging pasteurization and argon modified atmosphere packaging on the sensory quality and growth of endogenous microflora of a sliced cooked meat product.

PubMed

Pérez-Rodríguez, Fernando; Zamorano, Arturo Rivera; Posada-Izquierdo, Guiomar Denisse; García-Gimeno, Rosa María

2014-01-01

The objective of this work was to study the effect of post-packaging pasteurization on the sensory quality and growth of natural microorganisms during refrigerated storage (6 °C) of a cooked meat product considering two packaging atmospheres based on mixture of typical gases (CO(2)/N(2) (22/78%) and novel gases (CO(2)/Ar (17/83%)). Growth of lactic acid bacteria was significantly different between samples with and without post-packaging pasteurization, showing a growth rate >0.44 and equal to 0.28 log cfu/day, respectively. For samples with post-packaging pasteurization, atmosphere CO(2)/Ar resulted in a lower growth of lactic acid bacteria and a better sensory quality. Overall, samples without post-packaging pasteurization did not show a significant reduction of sensory quality during storage time (121 days) while samples with post-packaging pasteurization showed deterioration in their sensory quality. Further investigation is needed to obtain more definitive conclusions about the effect of post-packaging pasteurization and argon-based packaging atmospheres on cooked meat products.

DOE-EM-45 PACKAGING OPERATIONS AND MAINTENANCE COURSE

DOE Office of Scientific and Technical Information (OSTI.GOV)

Watkins, R.; England, J.

2010-05-28

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 attendedmore » 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.« less