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Application of Entry-Time Processes in Asset Management for Nuclear Power Plants (Final Report)

Description: A mathematical model of entry-time processes was developed, and a computational method for solving that model was verified. This methodology was demonstrated via application to a succession of increasingly more complex subsystems of nuclear power plants. The effort culminated in the application to main generators that constituted the PhD dissertation of Shuwen (“Eric”) Wang. Dr. Wang is now employed by ABS Consulting, in Anaheim, CA. ABS is a principal provider to the nuclear industry of technical services related to reliability and safety.
Date: January 23, 2008
Creator: Nelson, Paul
Partner: UNT Libraries Government Documents Department

Flooding Experiments and Modeling for Improved Reactor Safety

Description: Countercurrent two-phase flow and “flooding” phenomena in light water reactor systems are being investigated experimentally and analytically to improve reactor safety of current and future reactors. The aspects that will be better clarified are the effects of condensation and tube inclination on flooding in large diameter tubes. The current project aims to improve the level of understanding of flooding mechanisms and to develop an analysis model for more accurate evaluations of flooding in the pressurizer surge line of a Pressurized Water Reactor (PWR). Interest in flooding has recently increased because Countercurrent Flow Limitation (CCFL) in the AP600 pressurizer surge line can affect the vessel refill rate following a small break LOCA and because analysis of hypothetical severe accidents with the current flooding models in reactor safety codes shows that these models represent the largest uncertainty in analysis of steam generator tube creep rupture. During a hypothetical station blackout without auxiliary feedwater recovery, should the hot leg become voided, the pressurizer liquid will drain to the hot leg and flooding may occur in the surge line. The flooding model heavily influences the pressurizer emptying rate and the potential for surge line structural failure due to overheating and creep rupture. The air-water test results in vertical tubes are presented in this paper along with a semi-empirical correlation for the onset of flooding. The unique aspects of the study include careful experimentation on large-diameter tubes and an integrated program in which air-water testing provides benchmark knowledge and visualization data from which to conduct steam-water testing.
Date: September 14, 2008
Creator: Solmos, M., Hogan, K.J., VIerow, K.
Partner: UNT Libraries Government Documents Department

Nonlinear Seismic Correlation Analysis of the JNES/NUPEC Large-Scale Piping System Tests.

Description: The Japan Nuclear Energy Safety Organization/Nuclear Power Engineering Corporation (JNES/NUPEC) large-scale piping test program has provided valuable new test data on high level seismic elasto-plastic behavior and failure modes for typical nuclear power plant piping systems. The component and piping system tests demonstrated the strain ratcheting behavior that is expected to occur when a pressurized pipe is subjected to cyclic seismic loading. Under a collaboration agreement between the US and Japan on seismic issues, the US Nuclear Regulatory Commission (NRC)/Brookhaven National Laboratory (BNL) performed a correlation analysis of the large-scale piping system tests using derailed state-of-the-art nonlinear finite element models. Techniques are introduced to develop material models that can closely match the test data. The shaking table motions are examined. The analytical results are assessed in terms of the overall system responses and the strain ratcheting behavior at an elbow. The paper concludes with the insights about the accuracy of the analytical methods for use in performance assessments of highly nonlinear piping systems under large seismic motions.
Date: June 1, 2008
Creator: Nie,J.; DeGrassi, G.; Hofmayer, C. & Ali, S.
Partner: UNT Libraries Government Documents Department

MARKOV Model Application to Proliferation Risk Reduction of an Advanced Nuclear System

Description: The Generation IV International Forum (GIF) emphasizes proliferation resistance and physical protection (PR&PP) as a main goal for future nuclear energy systems. The GIF PR&PP Working Group has developed a methodology for the evaluation of these systems. As an application of the methodology, Markov model has been developed for the evaluation of proliferation resistance and is demonstrated for a hypothetical Example Sodium Fast Reactor (ESFR) system. This paper presents the case of diversion by the facility owner/operator to obtain material that could be used in a nuclear weapon. The Markov model is applied to evaluate material diversion strategies. The following features of the Markov model are presented here: (1) An effective detection rate has been introduced to account for the implementation of multiple safeguards approaches at a given strategic point; (2) Technical failure to divert material is modeled as intrinsic barriers related to the design of the facility or the properties of the material in the facility; and (3) Concealment to defeat or degrade the performance of safeguards is recognized in the Markov model. Three proliferation risk measures are calculated directly by the Markov model: the detection probability, technical failure probability, and proliferation time. The material type is indicated by an index that is based on the quality of material diverted. Sensitivity cases have been done to demonstrate the effects of different modeling features on the measures of proliferation resistance.
Date: July 13, 2008
Creator: Bari,R.A.
Partner: UNT Libraries Government Documents Department

Human Factors Considerations in New Nuclear Power Plants: Detailed Analysis.

Description: This Nuclear Regulatory Commission (NRC) sponsored study has identified human-performance issues in new and advanced nuclear power plants. To identify the issues, current industry developments and trends were evaluated in the areas of reactor technology, instrumentation and control technology, human-system integration technology, and human factors engineering (HFE) methods and tools. The issues were organized into seven high-level HFE topic areas: Role of Personnel and Automation, Staffing and Training, Normal Operations Management, Disturbance and Emergency Management, Maintenance and Change Management, Plant Design and Construction, and HFE Methods and Tools. The issues where then prioritized into four categories using a 'Phenomena Identification and Ranking Table' methodology based on evaluations provided by 14 independent subject matter experts. The subject matter experts were knowledgeable in a variety of disciplines. Vendors, utilities, research organizations and regulators all participated. Twenty issues were categorized into the top priority category. This Brookhaven National Laboratory (BNL) technical report provides the detailed methodology, issue analysis, and results. A summary of the results of this study can be found in NUREG/CR-6947. The research performed for this project has identified a large number of human-performance issues for new control stations and new nuclear power plant designs. The information gathered in this project can serve as input to the development of a long-term strategy and plan for addressing human performance in these areas through regulatory research. Addressing human-performance issues will provide the technical basis from which regulatory review guidance can be developed to meet these challenges. The availability of this review guidance will help set clear expectations for how the NRC staff will evaluate new designs, reduce regulatory uncertainty, and provide a well-defined path to new nuclear power plant licensing.
Date: February 14, 2008
Creator: OHara,J.; Higgins, J.; Brown, W. & Fink, R.
Partner: UNT Libraries Government Documents Department


Description: This project is investigating countercurrent flow and “flooding” phenomena in light water reactor systems to improve reactor safety of current and future reactors. To better understand the occurrence of flooding in the surge line geometry of a PWR, two experimental programs were performed. In the first, a test facility with an acrylic test section provided visual data on flooding for air-water systems in large diameter tubes. This test section also allowed for development of techniques to form an annular liquid film along the inner surface of the “surge line” and other techniques which would be difficult to verify in an opaque test section. Based on experiences in the air-water testing and the improved understanding of flooding phenomena, two series of tests were conducted in a large-diameter, stainless steel test section. Air-water test results and steam-water test results were directly compared to note the effect of condensation. Results indicate that, as for smaller diameter tubes, the flooding phenomena is predominantly driven by the hydrodynamics. Tests with the test sections inclined were attempted but the annular film was easily disrupted. A theoretical model for steam venting from inclined tubes is proposed herein and validated against air-water data. Empirical correlations were proposed for air-water and steam-water data. Methods for developing analytical models of the air-water and steam-water systems are discussed, as is the applicability of the current data to the surge line conditions. This report documents the project results from July 1, 2005 through June 30, 2008.
Date: September 26, 2008
Creator: Vierow, Karen
Partner: UNT Libraries Government Documents Department


Description: This Performance Assessment for the Savannah River Site E-Area Low-Level Waste Facility was prepared to meet requirements of Chapter IV of the Department of Energy Order 435.1-1. The Order specifies that a Performance Assessment should provide reasonable assurance that a low-level waste disposal facility will comply with the performance objectives of the Order. The Order also requires assessments of impacts to water resources and to hypothetical inadvertent intruders for purposes of establishing limits on radionuclides that may be disposed near-surface. According to the Order, calculations of potential doses and releases from the facility should address a 1,000-year period after facility closure. The point of compliance for the performance measures relevant to the all pathways and air pathway performance objective, as well as to the impact on water resources assessment requirement, must correspond to the point of highest projected dose or concentration beyond a 100-m buffer zone surrounding the disposed waste following the assumed end of active institutional controls 100 years after facility closure. During the operational and institutional control periods, the point of compliance for the all pathways and air pathway performance measures is the SRS boundary. However, for the water resources impact assessment, the point of compliance remains the point of highest projected dose or concentration beyond a 100-m buffer zone surrounding the disposed waste during the operational and institutional control periods. For performance measures relevant to radon and inadvertent intruders, the points of compliance are the disposal facility surface for all time periods and the disposal facility after the assumed loss of active institutional controls 100 years after facility closure, respectively. The E-Area Low-Level Waste Facility is located in the central region of the SRS known as the General Separations Area. It is an elbow-shaped, cleared area, which curves to the northwest, situated immediately north of the Mixed ...
Date: March 31, 2008
Creator: Wilhite, E
Partner: UNT Libraries Government Documents Department


Description: The Department of Energy (DOE) and Tennessee Valley Authority (TVA) entered into an Interagency Agreement to transfer approximately 40 metric tons of highly enriched uranium (HEU) to TVA for conversion to fuel for the Browns Ferry Nuclear Power Plant. Savannah River Site (SRS) inventories included a significant amount of this material, which resulted from processing spent fuel and surplus materials. The HEU is blended with natural uranium (NU) to low enriched uranium (LEU) with a 4.95% 235U isotopic content and shipped as solution to the TVA vendor. The HEU Blend Down Project provided the upgrades needed to achieve the product throughput and purity required and provided loading facilities. The first blending to low enriched uranium (LEU) took place in March 2003 with the initial shipment to the TVA vendor in July 2003. The SRS Shipments have continued on a regular schedule without any major issues for the past 5 years and are due to complete in September 2008. The HEU Blend program is now looking to continue its success by dispositioning an additional approximately 21 MTU of HEU material as part of the SRS Enriched Uranium Disposition Project.
Date: June 5, 2008
Creator: Magoulas, V; Charles Goergen, C & Ronald Oprea, R
Partner: UNT Libraries Government Documents Department

Further Dosimetry Studies at Rhode Island Nuclear Science Center.

Description: The RINSC is a 2 mega-watt, light water and graphite moderated and cooled reactor that has a graphite thermal column built as a user facility for sample irradiation. Over the past decade, after the reactor conversion from a highly-enriched uranium core to a low-enriched one, flux and dose measurements and calculations had been performed in the thermal column to update the ex-core parameters and to predict the effect from in-core fuel burn-up and rearrangement. The most recent data from measurements and calculations that have been made at the RINSC thermal column since October of 2005 are reported.
Date: May 5, 2008
Creator: Reciniello,R.N.; Holden, N.E.; Hu, J.-P.; Johnson, D.G.; Meddleton, M. & Tehan, T.N.
Partner: UNT Libraries Government Documents Department

Materials Degradation in Light Water Reactors: Life After 60,???

Description: Nuclear reactors present a very harsh environment for components service. Components within a reactor core must tolerate high temperature water, stress, vibration, and an intense neutron field. Degradation of materials in this environment can lead to reduced performance, and in some cases, sudden failure. A recent EPRI-led study interviewed 47 US nuclear utility executives to gauge perspectives on long-term operation of nuclear reactors. Nearly 90% indicated that extensions of reactor lifetimes to beyond 60 years were likely. When polled on the most challenging issues facing further life extension, two-thirds cited plant reliability as the key issue with materials aging and cable/piping as the top concerns for plant reliability. Materials degradation within a nuclear power plant is very complex. There are many different types of materials within the reactor itself: over 25 different metal alloys can be found with can be found within the primary and secondary systems, not to mention the concrete containment vessel, instrumentation and control, and other support facilities. When this diverse set of materials is placed in the complex and harsh environment coupled with load, degradation over an extended life is indeed quite complicated. To address this issue, the USNRC has developed a Progressive Materials Degradation Approach (NUREG/CR-6923). This approach is intended to develop a foundation for appropriate actions to keep materials degradation from adversely impacting component integrity and safety and identify materials and locations where degradation can reasonably be expected in the future. Clearly, materials degradation will impact reactor reliability, availability, and potentially, safe operation. Routine surveillance and component replacement can mitigate these factors, although failures still occur. With reactor life extensions to 60 years or beyond or power uprates, many components must tolerate the reactor environment for even longer times. This may increase susceptibility for most components and may introduce new degradation modes. While all ...
Date: April 1, 2008
Creator: Busby, Jeremy T; Nanstad, Randy K; Stoller, Roger E; Feng, Zhili & Naus, Dan J
Partner: UNT Libraries Government Documents Department

University Reactor Conversion Lessons Learned Workshop for Purdue University Reactor

Description: The Department of Energy’s Idaho National Laboratory, under its programmatic responsibility for managing the University Research Reactor Conversions, has completed the conversion of the reactor at Purdue University Reactor. With this work completed and in anticipation of other impending conversion projects, the INL convened and engaged the project participants in a structured discussion to capture the lessons learned. The lessons learned process has allowed us to capture gaps, opportunities, and good practices, drawing from the project team’s experiences. These lessons will be used to raise the standard of excellence, effectiveness, and efficiency in all future conversion projects.
Date: September 1, 2008
Creator: Woolstenhulme, Eric C. & Hewit, Dana M.
Partner: UNT Libraries Government Documents Department

TMIST-2 Transport Test

Description: In anticipation of the TMIST-2 experiment in the Advanced Test Reactor, there was a need to determine if the tritium that is expected to be observed at the outlet of the experiment would be seen or if it may be lost on its way from the experiment in the core to the measurement station. To assist in resolving that issue, a bench-scale experiment was conducted in the Idaho National Laboratory’s Safety and Tritium Applied Research (STAR) facility using deuterium and a mass spectrometer in lieu of tritium with ion chambers, bubblers, and scintillation counting. The experiment replicated the concentration of the hydrogen isotope, the flow rates anticipated, and the residence times. It was found that there was initial uptake on tubing walls, presumably due to oxidation of the hydrogen isotopes to water and adsorption or isotopic exchange, but that saturates relatively quickly, and once saturated, the concentration of deuterium at the outlet of the tubing system was essentially the same as it was at the experiment inlet under the conditions modeled in the experiment.
Date: February 1, 2008
Creator: Longhurst, Glen R.
Partner: UNT Libraries Government Documents Department

Next Generation Nuclear Plant Phenomena Identification and Ranking Tables (PIRTs) Volume 1: Main Report

Description: A phenomena identification and ranking table (PIRT) process was conducted for the Next Generation Nuclear Plant (NGNP) design. This design (in the conceptual stage) is a modular high-temperature gas-cooled reactor (HTGR) that generates both electricity and process heat for hydrogen production. Expert panels identified safety-relevant phenomena, ranked their importance, and assessed the knowledge levels in the areas of accidents and thermal fluids, fission-product transport and dose, high-temperature materials, graphite, and process heat for hydrogen production. This main report summarizes and documents the process and scope of the reviews, noting the major activities and conclusions. The identified phenomena, analyses, rationales, and associated ratings of the phenomena, plus a summary of each panel's findings, are presented. Individual panel reports for these areas are provided as attached volumes to this main report and provide considerably more detail about each panel's deliberations as well as a more complete listing of the phenomena that were evaluated.
Date: March 1, 2008
Creator: Ball, Sydney J
Partner: UNT Libraries Government Documents Department

Next Generation Nuclear Plant Phenomena Identification and Ranking Tables (PIRTs) Volume 2: Accident and Thermal Fluids Analysis PIRTs

Description: An accident, thermal fluids, and reactor physics phenomena identification and ranking process was conducted by a panel of experts on the next generation nuclear plant (NGNP) design (consideration given to both pebble-bed and prismatic gas-cooled reactor configurations). Safety-relevant phenomena, importance, and knowledge base were assessed for the following event classes: (1) normal operation (including some reactor physics aspects), (2) general loss of forced circulation (G-LOFC), (3) pressurized loss-of-forced circulation (P-LOFC), (4) depressurized loss-of-forced circulation (D-LOFC), (5) air ingress (following D-LOFC), (6) reactivity transients - including anticipated transients without scram (ATWS), (7) processes coupled via intermediate heat exchanger (IHX) (IHX failure with molten salt), and (8) steam/water ingress. The panel's judgment of the importance ranking of a given phenomenon (or process) was based on the effect it had on one or more figures of merit or evaluation criteria. These included public and worker dose, fuel failure, and primary (and other safety) system integrity. The major phenomena of concern that were identified and categorized as high importance combined with medium to low knowledge follow: (1) core coolant bypass flows (normal operation), (2) power/flux profiles (normal operation), (3) outlet plenum flows (normal operation), (4) reactivity-temperature feedback coefficients for high-plutonium-content cores (normal operation and accidents), (5) fission product release related to the transport of silver (normal operation), (6)emissivity aspects for the vessel and reactor cavity cooling system (G-LOFC), (7) reactor vessel cavity air circulation and heat transfer (G-LOFC), and (8)convection/radiation heating of upper vessel area (P-LOFC).
Date: March 1, 2008
Creator: Ball, Sydney J; Corradini, M.; Fisher, Stephen Eugene; Gauntt, R.; Geffraye, G.; Gehin, Jess C et al.
Partner: UNT Libraries Government Documents Department

Next Generation Nuclear Plant Phenomena Identification and Ranking Tables (PIRTs) Volume 4: High-Temperature Materials PIRTs

Description: The Phenomena Identification and Ranking Table (PIRT) technique was used to identify safety-relevant/safety-significant phenomena and assess the importance and related knowledge base of high-temperature structural materials issues for the Next Generation Nuclear Plant (NGNP), a very high temperature gas-cooled reactor (VHTR). The major aspects of materials degradation phenomena that may give rise to regulatory safety concern for the NGNP were evaluated for major structural components and the materials comprising them, including metallic and nonmetallic materials for control rods, other reactor internals, and primary circuit components; metallic alloys for very high-temperature service for heat exchangers and turbomachinery, metallic alloys for high-temperature service for the reactor pressure vessel (RPV), other pressure vessels and components in the primary and secondary circuits; and metallic alloys for secondary heat transfer circuits and the balance of plant. These materials phenomena were primarily evaluated with regard to their potential for contributing to fission product release at the site boundary under a variety of event scenarios covering normal operation, anticipated transients, and accidents. Of all the high-temperature metallic components, the one most likely to be heavily challenged in the NGNP will be the intermediate heat exchanger (IHX). Its thin, internal sections must be able to withstand the stresses associated with thermal loading and pressure drops between the primary and secondary loops under the environments and temperatures of interest. Several important materials-related phenomena related to the IHX were identified, including crack initiation and propagation; the lack of experience of primary boundary design methodology limitations for new IHX structures; and manufacturing phenomena for new designs. Specific issues were also identified for RPVs that will likely be too large for shop fabrication and transportation. Validated procedures for on-site welding, post-weld heat treatment (PWHT), and inspections will be required for the materials of construction. High-importance phenomena related to the RPV include crack ...
Date: March 1, 2008
Creator: Corwin, William R; Ballinger, R.; Majumdar, S. & Weaver, K. D.
Partner: UNT Libraries Government Documents Department

Novel Processing of Unique Ceramic-Based Nuclear Materials and Fuels

Description: Advances in nuclear reactor technology and the use of gas-cooled fast reactors require the development of new materials that can operate at the higher temperatures expected in these systems. These include refractory alloys base on Nb, Zr, Ta, Mo, W, and Re; ceramics and composites such as those based on silicon carbide (SiCf-SiC); carbon-carbon composites; and advanced coatings. Besides the ability to handle higher expected temperatures, effective heat transfer between reactor componets is necessary for improved efficiency. Improving thermal conductivity of the materials used in nuclear fuels and other temperature critical components can lower the center-line fuel temperature and thereby enhance durability and reduce the risk of premature failure.
Date: November 30, 2008
Creator: Zhang, Hui & Singh, Raman P.
Partner: UNT Libraries Government Documents Department

Heat Transfer Boundary Conditions in the RELAP5-3D Code

Description: The heat transfer boundary conditions used in the RELAP5-3D computer program have evolved over the years. Currently, RELAP5-3D has the following options for the heat transfer boundary conditions: (a) heat transfer correlation package option, (b) non-convective option (from radiation/conduction enclosure model or symmetry/insulated conditions), and (c) other options (setting the surface temperature to a volume fraction averaged fluid temperature of the boundary volume, obtaining the surface temperature from a control variable, obtaining the surface temperature from a time-dependent general table, obtaining the heat flux from a time-dependent general table, or obtaining heat transfer coefficients from either a time- or temperature-dependent general table). These options will be discussed, including the more recent ones.
Date: May 1, 2008
Creator: Riemke, Richard A.; Davis, Cliff B. & Schultz, Richard R.
Partner: UNT Libraries Government Documents Department

Field Evaluations of Low-Frequency SAFT-UT on Cast Stainless Steel and Dissimilar Metal Weld Components

Description: This report documents work performed at the Pacific Northwest National Laboratory (PNNL) in Richland, Washington, and at the Electric Power Research Institute's (EPRI) Nondestructive Examination (NDE) Center in Charlotte, North Carolina, on evalutating a low frequency ultrasonic inspection technique used for examination of cast stainless steel (CSS) and dissimilar metal (DMW) reactor piping components. The technique uses a zone-focused, multi-incident angle, low frequency (250-450 kHz) inspection protocol coupled with the synthetic aperture focusing technique (SAFT). The primary focus of this work is to provide information to the United States Nuclear Regulatory Commission on the utility, effectiveness and reliability of ultrasonic testing (UT) inspection techniques as related to the inservice ultrasonic inspection of coarse grained primary piping components in pressurized water reactors (PWRs).
Date: November 1, 2008
Creator: Diaz, Aaron A.; Harris, R. V. & Doctor, Steven R.
Partner: UNT Libraries Government Documents Department

Flexible Conversion Ratio Fast Reactor Systems Evaluation

Description: Conceptual designs of lead-cooled and liquid salt-cooled fast flexible conversion ratio reactors were developed. Both concepts have cores reated at 2400 MWt placed in a large-pool-type vessel with dual-free level, which also contains four intermediate heat exchanges coupling a primary coolant to a compact and efficient supercritical CO2 Brayton cycle power conversion system. Decay heat is removed passively using an enhanced Reactor Vessel Auxiliary Cooling System and a Passive Secondary Auxiliary Cooling System. The most important findings were that (1) it is feasible to design the lead-cooled and salt-cooled reactor with the flexible conversion ratio (CR) in the range of CR=0 and CR=1 n a manner that achieves inherent reactor shutdown in unprotected accidents, (2) the salt-cooled reactor requires Lithium thermal Expansion Modules to overcme the inherent salt coolant's large positive coolant temperature reactivity coefficient, (3) the preferable salt for fast spectrum high power density cores is NaCl-Kcl-MgCl2 as opposed to fluoride salts due to its better themal-hydraulic and neutronic characteristics, and (4) both reactor, but attain power density 3 times smaller than that of the sodium-cooled reactor.
Date: June 30, 2008
Creator: Todreas, Neil & Hejzlar, Pavel
Partner: UNT Libraries Government Documents Department

Final report on LDRD project 105967 : exploring the increase in GaAs photodiode responsivity with increased neutron fluence.

Description: A previous LDRD studying radiation hardened optoelectronic components for space-based applications led to the result that increased neutron irradiation from a fast-burst reactor caused increased responsivity in GaAs photodiodes up to a total fluence of 4.4 x 10{sup 13} neutrons/cm{sup 2} (1 MeV Eq., Si). The silicon photodiodes experienced significant degradation. Scientific literature shows that neutrons can both cause defects as well as potentially remove defects in an annealing-like process in GaAs. Though there has been some modeling that suggests how fabrication and radiation-induced defects can migrate to surfaces and interfaces in GaAs and lead to an ordering effect, it is important to consider how these processes affect the performance of devices, such as the basic GaAs p-i-n photodiode. In this LDRD, we manufactured GaAs photodiodes at the MESA facility, irradiated them with electrons and neutrons at the White Sands Missile Range Linac and Fast Burst Reactor, and performed measurements to show the effect of irradiation on dark current, responsivity and high-speed bandwidth.
Date: January 1, 2008
Creator: Blansett, Ethan L.; Geib, Kent Martin; Cich, Michael Joseph; Wrobel, Theodore Frank; Peake, Gregory Merwin; Fleming, Robert M. et al.
Partner: UNT Libraries Government Documents Department

Minior Actinide Doppler Coefficient Measurement Assessment

Description: The "Minor Actinide Doppler Coefficient Measurement Assessment" was a Department of Energy (DOE) U-NERI funded project intended to assess the viability of using either the FLATTOP or the COMET critical assembly to measure high temperature Doppler coefficients. The goal of the project was to calculate using the MCNP5 code the gram amounts of Np-237, Pu-238, Pu-239, Pu-241, AM-241, AM-242m, Am-243, and CM-244 needed to produce a 1E-5 in reactivity for a change in operating temperature 800C to 1000C. After determining the viability of using the assemblies and calculating the amounts of each actinide an experiment will be designed to verify the calculated results. The calculations and any doncuted experiments are designed to support the Advanced Fuel Cycle Initiative in conducting safety analysis of advanced fast reactor or acceoerator-driven transmutation systems with fuel containing high minor actinide content.
Date: April 10, 2008
Creator: Hertel, Nolan E. & Blaylock, Dwayne
Partner: UNT Libraries Government Documents Department


Description: Prompt doses from x-rays generated as result of laser beam interaction with target material are calculated at different locations inside the National Ignition Facility (NIF). The maximum dose outside a Target Chamber diagnostic port is {approx} 1 rem for a shot utilizing the 192 laser beams and 1.8 MJ of laser energy. The dose during a single bundle shot (8 laser beams) drops to {approx} 40 mrem. Doses calculated outside the Target Bay doors and inside the Switchyards (except for the 17 ft.-6 in. level) range from a fraction of mrem to about 11 mrem for 192 beams, and scales down proportionally with smaller number of beams. At the 17ft.-6 in. level, two diagnostic ports are directly facing two of the Target Bay doors and the maximum doses outside the doors are 51 and 15.5 mrem, respectively. Shielding each of the two Target Bay doors with 1/4 in. Pb reduces the dose by factor of fifty. One or two bundle shots (8 to 16 laser beams) present a small hazard to personnel in the Switchyards.
Date: March 27, 2008
Creator: Khater, H Y; Brereton, S J & Singh, M S
Partner: UNT Libraries Government Documents Department


Description: A CFD modeling and simulation process for large-scale problems using an arbitrary fast reactor fuel assembly design was evaluated. Three dimensional flow distributions of sodium for several fast reactor fuel assembly pin spacing configurations were simulated on high performance computers using commercial CFD software. This research focused on 19-pin fuel assembly “benchmark” geometry, similar in design to the Advanced Burner Test Reactor, where each pin is separated by helical wire-wrap spacers. Several two-equation turbulence models including the k-e and SST (Menter) k-? were evaluated. Considerable effort was taken to resolve the momentum boundary layer, so as to eliminate the need for wall functions and reduce computational uncertainty. High performance computers were required to generate the hybrid meshes needed to predict secondary flows created by the wire-wrap spacers; computational meshes ranging from 65 to 85 million elements were common. A general validation methodology was followed, including mesh refinement and comparison of numerical results with empirical correlations. Predictions for velocity, temperature, and pressure distribution are shown. The uncertainty of numerical models, importance of high fidelity experimental data, and the challenges associated with simulating and validating large production-type problems are presented.
Date: September 1, 2008
Creator: Hamman, Kurt D. & Berry, Ray A.
Partner: UNT Libraries Government Documents Department