The purpose of this activity is to determine the structural performance of waste packages (WP) subject to rock fall design basis event (DBE) dynamic loads and document the calculation results that describe the threshold rock sizes for crack-initiation and through cracks in waste package shells. This activity is associated with the waste package design. AP-3.12Q, Revision 0, ICN 0, Calculations, is used to develop the calculation.

This report describes important future research directions in nanoscale science, engineering and technology. It was prepared in connection with an anticipated national research initiative on nanotechnology for the twenty-first century. The research directions described are not expected to be inclusive but illustrate the wide range of research opportunities and challenges that could be undertaken through the national laboratories and their major national scientific user facilities with the support of universities and industry.

The purpose of this calculation is to document (1) the Waste Package Degradation (WAPDEG) version 3.09 (CRWMS M&O 1998b. ''Software Routine Report for WAPDEG'' (Version 3.09)) simulations used to analyze degradation and failure of 2-cm thick titanium grade 7 corrosion resistant material (CRM) drip shields (that are placed over single-layer waste packages composed of a 30-cm thick A-5 16 carbon steel corrosion allowance material (CAM)) as well as degradation and failure of the waste packages themselves, and (2) post-processing of these results into tables of drip shield/waste package degradation time histories suitable for use as input into the Integrated Probabilistic Simulator for Environmental Systems (RIP) version 5.19.01 (Golder Associates 1998) computer code. This calculation supports Performance Assessment analysis of the License Application Design Selection (LADS) Enhanced Design Alternative IV.

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The purpose of this calculation is to provide a bounding dose consequence analysis of the immobilized plutonium (can-in-canister) waste form to be handled at the Monitored Geologic Repository (MGR) at Yucca Mountain. The current concept for the Plutonium Can-in-Canister waste form is provided in Attachment III. A typical design basis event (DBE) defines a scenario that generally includes an initiating event and the sequences of events that follow. This analysis will provide (1) radiological releases and dose consequences for a postulated, bounding DBE and (2) design-related assumptions on which the calculated dose consequences are based. This analysis is part of the safety design basis for the repository. Results will be used in other analyses to determine or modify the safety classification and quality assurance level of repository structures, systems, and components (SSCs). The Quality Assurance (QA) program applies to this calculation. The work reported in this document is part of the analysis of MGR DBEs and is performed using AP-3.12Q, Calculations. The work done for this analysis was evaluated according to QAP-2-0, Control of Activities. This evaluation determined that such activities are subject to DOE/RW/0333PY Quality Assurance Requirements and Description (DOE 1998), requirements. This calculation is quality affecting because the …

This report is one of six reports developed under the U.S. Department of Energy (DOE) program in Power System Integration and Reliability (PSIR). The objective of this report is to review, analyze, and evaluate critical reliability issues demonstrated by recent disturbance events in the North America power system. Eleven major disturbances are examined, most occurring in this decade. The strategic challenge is that the pattern of technical need has persisted for a long period of time. For more than a decade, anticipation of market deregulation has been a major disincentive to new investments in system capacity. It has also inspired reduced maintenance of existing assets. A massive infusion of better technology is emerging as the final option to continue reliable electrical services. If an investment in better technology will not be made in a timely manner, then North America should plan its adjustments to a very different level of electrical service. It is apparent that technical operations staff among the utilities can be very effective at marshaling their forces in the immediate aftermath of a system emergency, and that serious disturbances often lead to improved mechanisms for coordinated operation. It is not at all apparent that such efforts can be …

Designed to help define new scientific directions related to complex systems in order to create new understanding about the nano world and complicated, multicomponent structures.

Thermal analysis of the 9975 package with three configurations of the BNFL 3013 outer container (with Rocky Flats, SRS, and BNFL inner containers) have been performed for Normal Conditions of Transport (NCT) of plutonium oxide and metal. The NCT is defined in 10 CFR 71.71(c)(1) s an ambient of 100 F (38 C) in still air with 800 W/m{sup 2} and 400 W/m{sup 2} of solar heating on the drum top and sides, respectively. The 9975 drum package is considered to be in an upright position, and the drum bottom is adiabatic. The Rocky and SRS 3013 configurations with Pu metal contents (19 watts) result in acceptable (similar) packaging temperatures, however the plutonium metal temperatures are lower for the SRS design (SRS has helium fill gas whereas Rocky is essentially air filled). The BNFL configuration for Pu oxide contents (19 watts) result in acceptable temperatures and pressures based on limits in the 9975 Safety Analysis Report (SARP). However, for 30 watts of Pu oxide, the fiberboard peak temperatures are very near the SARP allowable. The pressure in the 3013 container is 688.4 psig for the 30 watt Pu oxide content and 569.5 psig for the 19 watt Pu oxide content. …

National technical Information Service, Springfield, Virginia. "Exteranl Radiation Surveillance, Section 4.7" Pacific Northwest National Laboratory, richland, Washington.

The Paducah Gaseous Diffusion Plant (PGDP) operated by the Department of Energy is challenged with selecting the appropriate remediation technology to cleanup contaminants at Waste Area Group (WAG) 6. This research utilizes value-focused thinking and multiattribute preference theory concepts to produce a decision analysis model designed to aid the decision makers in their selection process. The model is based on CERCLA's five primary balancing criteria, tailored specifically to WAG 6 and the contaminants of concern, utilizes expert opinion and the best available engineering, cost, and performance data, and accounts for uncertainty in contaminant volume. The model ranks 23 remediation technologies (trains) in their ability to achieve the CERCLA criteria at various contaminant volumes. A sensitivity analysis is performed to examine the effects of changes in expert opinion and uncertainty in volume. Further analysis reveals how volume uncertainty is expected to affect technology cost, time and ability to meet the CERCLA criteria. The model provides the decision makers with a CERCLA-based decision analysis methodology that is objective, traceable, and robust to support the WAG 6 Feasibility Study. In addition, the model can be adjusted to address other DOE contaminated sites.

This project is designed to develop a family of novel NO{sub x} control technologies, called Second Generation Advanced Reburning (SGAR) which has the potential to achieve 90+ NO{sub x} control in coal fired boilers at a significantly lower cost than Selective Catalytic Reduction. The ninth reporting period in Phase II (October 1-December 31, 1999) included preparation of the 10 x 10{sup 6} Btu/hr Tower Furnace for tests and setting the SGAR model to predict process performance under Tower Furnace conditions. Based on results of previous work, a paper has been prepared and submitted for the presentation at the 28 Symposium (International) on Combustion to be held at the University of Edinburgh, Scotland.

This report presents work carried out under contract DE-AC22-95PC95144 ''Combustion 2000 - Phase II.'' The goals of the program are to develop a coal-fired high performance power generation system (HIPPS) that is capable of: {lg_bullet} thermal efficiency (HHV) {ge} 47% {lg_bullet} NOx, SOx, and particulates {le} 10% NSPS (New Source Performance Standard) {lg_bullet} coal providing {ge} 65% of heat input {lg_bullet} all solid wastes benign {lg_bullet} cost of electricity {le} 90% of present plants Phase I, which began in 1992, focused on the analysis of various configurations of indirectly fired cycles and on technical assessments of alternative plant subsystems and components, including performance requirements, developmental status, design options, complexity and reliability, and capital and operating costs. Phase I also included preliminary R&D and the preparation of designs for HIPPS commercial plants approximately 300 MWe in size. Phase II, had as its initial objective the development of a complete design base for the construction and operation of a HIPPS prototype plant to be constructed in Phase III. As part of a descoping initiative, the Phase III program has been eliminated and work related to the commercial plant design has been ended. The rescoped program retained a program of engineering research and …

A model is described for the movement of melted metallic material through a ceramic porous debris bed. The model is designed for the analysis of severe accidents in LWRs, wherein melted core plate material may slump onto the top of a porous bed of relocated core material supported by the lower head. The permeation of the melted core plate material into the porous debris bed influences the heatup of the debris bed and the heatup of the lower head supporting the debris. A model for mass transport of melted metallic material is applied that includes terms for viscosity and turbulence but neglects inertial and capillary terms because of their small value relative to gravity and viscous terms in the momentum equation. The relative permeability and passability of the porous debris are calculated as functions of debris porosity, particle size, and effective saturation. An iterative numerical solution is used to solve the set of nonlinear equations for mass transport. The effective thermal conductivity of the debris is calculated as a function of porosity, particle size, and saturation. The model integrates the equations for mass transport with a model for the two-dimensional conduction of heat through porous debris. The integrated model has …

This project is designed to develop a family of novel NO{sub x} control technologies, called Second Generation Advanced Reburning which has the potential to achieve 90+ NO{sub x} control in coal fired boilers at a significantly lower cost than Selective Catalytic Reduction. The seventh reporting period in Phase II (April 1-June 30, 1999) included experimental activities and combined chemistry-mixing modeling on advanced gas reburning. The goal of combustion tests was to determine the efficiency of advanced reburning using coal as the reburning fuel. Tests were conducted in Boiler Simulator Facility (BSF). Several coals were tested. The modeling effort was focused on the description of N-agent injection along with overfire air. Modeling identified process parameters that can be used to optimize the AR-Lean process.

The Pacific Northwest Loads and Resources Study (White Book) is published annually by BPA and establishes the planning basis for supplying electricity to customers. It serves a dual purpose. First, the White Book presents projections of regional and Federal system load and resource capabilities, along with relevant definitions and explanations. Second, the White Book serves as a benchmark for annual BPA determinations made pursuant to its regional power sales contracts. Specifically, BPA uses the information in the White Book for determining the notice required when customers request to increase or decrease the amount of power purchased from BPA. The White Book will not be used in calculations for the 2002 regional power sales contract subscription process. The White Book compiles information obtained from several formalized resource planning reports and data submittals, including those from the Northwest Power Planning Council (Council) and the Pacific Northwest Utilities Conference Committee (PNUCC). The White Book is not an operational planning guide, nor is it used for determining BPA revenues. Operation of the Federal Columbia River Power System (FCRPS) is based on a set of criteria different from that used for resource planning decisions. Operational planning is dependent upon real-time or near-term knowledge of system …

Pure copper and Alloy 5083 aluminum were processed by equal channel angular extrusion (ECAE); their microstructural evolution and corresponding mechanical properties were investigated. Work also began on the possible use of ECAE to synthesize advanced materials or to consolidate metal powders or powder mixtures. The die tooling used for ECAE is described and selected microstructural and mechanical property results for ECAE-processed copper and cold-rolled (conventionally-processed) copper in the as-processed and annealed condition are compared. Results thus far show that the “pure” metal is prone to low temperature recrystallization after large strain hardening—more beneficial effects are expected in the dispersion-strengthened and precipitation-hardening alloys. The large range of tensile properties and grain sizes from the copper allowed a flow stress analysis to be performed. From this analysis, a new model for flow stress behavior is proposed. An evaluation of ECAE processing of material for spot welding electrodes began. Results to date include electrodes of ECAE-processed commercially pure copper (Alloy 101). Future work involving Glidcop® (Al2O3 oxide dispersionstrengthened copper) and CuCrZr (Cr-Zr precipitation dispersion) materials will be required to fully investigate the benefits of ECAE for electrode life extension. Initial work on Aluminum Alloy 5083 showed that ECAE led to grain refinement as …

Models were designed to resolve deficiencies in the SCDAP/RELAP5/MOD3.2 calculations of the configuration and integrity of hot, partially oxidized cladding. The modeling was improved in five areas. First, the configuration defined for melted metallic cladding retained by an adjacent oxide layer was improved. Second, the empirical model to account for the effect on oxidation of intact cladding is not significantly reduced by the presence of a rather high concentration of relocated material. Third, models for the dissolution of the oxide layer by the metallic layer were implemented into the code. Fourth, a model was added to calculate the thermal stress applied to the oxide layer by the temperature gradient across the oxide layer and to compare this stress to the ultimate strength of the oxide layer. Fifth, a new rule based on theoretical and experimental results was established for identifying the regions of a fuel rod with oxidation of both the inside and outside surfaces of the cladding. The assessment of these models and their integration into SCDAP/RELAP5 showed that the calculated axial distribution in cladding oxidation and relocation are in significantly better agreement with experimental results than is currently the case. The modeling changes account for three aspects of …

The SCDAP/RELAP5 computer code is a best-estimate analysis tool for performing nuclear reactor severe accident simulations. This report describes the justification, theory, implementation, and testing of a new modeling capability which will refine the analysis of the movement of molten material from the core region to the vessel lower head. As molten material moves from the core region through the core support structures it may encounter conditions which will cause it to freeze in the region of the lower core plate, delaying its arrival to the vessel head. The timing of this arrival is significant to reactor safety, because during the time span for material relocation to the lower head, the core may be experiencing steam-limited oxidation. The time at which hot material arrives in a coolant-filled lower vessel head, thereby significantly increasing the steam flow rate through the core region, becomes significant to the progression and timing of a severe accident. This report is a revision of a report INEEL/EXT-00707, entitled "Preliminary Design Report for SCDAP/RELAP5 Lower Core Plate Model".

Designs are described for implementing models for calculating the heat transfer and flow losses in porous debris in the lower head of a reactor vessel. The COUPLE model in SCDAP/RELAP5 represents both the porous and nonporous debris that results from core material slumping into the lower head. Currently, the COUPLE model has the capability to model convective and radiative heat transfer from the surfaces of nonporous debris in a detailed manner and to model only in a simplistic manner the heat transfer from porous debris. In order to advance beyond the simplistic modeling for porous debris, designs are developed for detailed calculations of heat transfer and flow losses in porous debris. Correlations are identified for convective heat transfer in porous debris for the following modes of heat transfer; (1) forced convection to liquid, (2) forced convection to gas, (3) nucleate boiling, (4) transition boiling, and (5) film boiling. Interphase heat transfer is modeled in an approximate manner. Designs are described for models to calculate the flow losses and interphase drag of fluid flowing through the interstices of the porous debris, and to apply these variables in the momentum equations in the RELAP5 part of the code. Since the models for …

This report is a collection and review of system operation and failure experiences for air ventilation systems in nuclear facilities. These experiences are applicable for magnetic and inertial fusion facilities since air ventilation systems are support systems that can be considered generic to nuclear facilities. The report contains descriptions of ventilation system components, operating experiences with these systems, component failure rates, and component repair times. Since ventilation systems have a role in mitigating accident releases in nuclear facilities, these data are useful in safety analysis and risk assessment of public safety. An effort has also been given to identifying any safety issues with personnel operating or maintaining ventilation systems. Finally, the recommended failure data were compared to an independent data set to determine the accuracy of individual values. This comparison is useful for the International Energy Agency task on fusion component failure rate data collection.

The spent nuclear fuel dry transfer system (DTS) provides an interface between large and small casks and between storage-only and transportation casks. It permits decommissioning of reactor pools after shutdown and allows the use of large storage-only casks for temporary onsite storage of spent nuclear fuel irrespective of reactor or fuel handling limitations at a reactor site. A cold demonstration of the DTS prototype was initiated in August 1996 at the Idaho National Engineering and Environmental Laboratory (INEEL). The major components demonstrated included the fuel assembly handling subsystem, the shield plug/lid handling subsystem, the cask interface subsystem, the demonstration control subsystem, a support frame, and a closed circuit television and lighting system. The demonstration included a complete series of DTS operations from source cask receipt and opening through fuel transfer and closure of the receiving cask. The demonstration included both normal operations and recovery from off-normal events. It was designed to challenge the system to determine whether there were any activities that could be made to jeopardize the activities of another function or its safety. All known interlocks were challenged. The equipment ran smoothly and functioned as designed. A few "bugs" were corrected. Prior to completion of the demonstration testing, …

The Low-Activity Waste Process Technology Program at the Idaho Nuclear Technology and Engineering Center (INTEC) anticipates that large volumes of low-level/low-activity wastes will need to be grouted prior to near-surface disposal. During fiscal year 1999, grout formulations were studied for transuranic waste derived from INTEC liquid sodium-bearing waste and for projected newly generated low-level liquid waste. Additional studies were completed on radionuclide leaching, microbial degradation, waste neutralization, and a small mockup for grouting the INTEC underground storage tank residual heels.

The purpose of this calculation is to document the WAPDEG version 3.09 (CRWMS M&O 1998b). Software Routine Report for WAPDEG (Version 3.09) simulations used to analyze waste package degradation and failure under the repository exposure conditions characterized by a two-tier thermal loading repository design. Also documented is the post-processing of these results into tables of waste-package-degradation-time histories suitable for use as input into the Integrated Probabilistic Simulator for Environmental Systems (RIP) version 5.19.01 (Golder Associates 1998) computer program. Specifically, the WAPDEG simulations discussed in this calculation correspond to waste package emplacement conditions (repository environment and design) as defined in the Total System Performance Assessment-Viability Assessment (CRWMS M&O 1998a). Total System Performance Assessment-Viability Assessment (TSPA-VA) Analyses Technical Basis Document--Chapter 5, Waste Package Degradation Modeling And Abstraction, pp. 5-27 to 5-29, with the exception that a two-tier thermal loading design feature as specified in the License Application Design Selection (LADS) study was analyzed. The particular design feature evaluated in this report is a modification of the repository horizon elevation and layout within the Topopah Springs Member of Yucca Mountain. Specifically, the modification consists of adding a second level, 50-m above the base case repository layout. Two options were considered, representing two variations …