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Markov Model of Accident Progression at Fukushima Daiichi

Description: On March 11, 2011, a magnitude 9.0 earthquake followed by a tsunami caused loss of offsite power and disabled the emergency diesel generators, leading to a prolonged station blackout at the Fukushima Daiichi site. After successful reactor trip for all operating reactors, the inability to remove decay heat over an extended period led to boil-off of the water inventory and fuel uncovery in Units 1-3. A significant amount of metal-water reaction occurred, as evidenced by the quantities of hydrogen generated that led to hydrogen explosions in the auxiliary buildings of the Units 1 & 3, and in the de-fuelled Unit 4. Although it was assumed that extensive fuel damage, including fuel melting, slumping, and relocation was likely to have occurred in the core of the affected reactors, the status of the fuel, vessel, and drywell was uncertain. To understand the possible evolution of the accident conditions at Fukushima Daiichi, a Markov model of the likely state of one of the reactors was constructed and executed under different assumptions regarding system performance and reliability. The Markov approach was selected for several reasons: It is a probabilistic model that provides flexibility in scenario construction and incorporates time dependence of different model states. It also readily allows for sensitivity and uncertainty analyses of different failure and repair rates of cooling systems. While the analysis was motivated by a need to gain insight on the course of events for the damaged units at Fukushima Daiichi, the work reported here provides a more general analytical basis for studying and evaluating severe accident evolution over extended periods of time. This work was performed at the request of the U.S. Department of Energy to explore 'what-if' scenarios in the immediate aftermath of the accidents.
Date: November 11, 2012
Creator: A., Cuadra; R., Bari; Cheng, L-Y; Ginsberg, T.; Lehner, J.; Martinez-Guridi, G. et al.
Partner: UNT Libraries Government Documents Department

NERI FINAL TECHNICAL REPORT, DE-FC07-O5ID14647, OPTIMIZATION OF OXIDE COMPOUNDS FOR ADVANCED INERT MATRIX MATERIALS

Description: In order to reduce the current excesses of plutonium (both weapon grade and reactor grade) and other transuranium elements, a concept of inert matrix fuel (IMF) has been proposed for an uranium free transmutation of fissile actinides which excludes continuous uranium-plutonium conversion in thermal reactors and advanced systems. Magnesium oxide (MgO) is a promising candidate for inert matrix (IM) materials due to its high melting point (2827 C), high thermal conductivity (13 W/K {center_dot} m at 1000 C), good neutronic properties, and irradiation stability However, MgO reacts with water and hydrates easily, which prevents it from being used in light water reactors (LWRs) as an IM. To improve the hydration resistance of MgO-based inert matrix materials, Medvedev and coworkers have recently investigated the introduction of a secondary phase that acts as a hydration barrier. An MgO-ZrO{sub 2} composite was specifically studied and the results showed that the composite exhibited improved hydration resistance than pure MgO. However, ZrO{sub 2} is insoluble in most acids except HF, which is undesirable for fuel reprocessing. Moreover, the thermal conductivity of ZrO{sub 2} is low and typically less than 3 W {center_dot} m{sup -1} {center_dot} K{sup -1} at 1000 C. In search for an alternative composite strategy, Nd{sub 2}Zr{sub 2}O{sub 7}, an oxide compound with pyrochlore structure, has been proposed recently as a corrosion resistant phase, and MgO-Nd{sub 2}Zr{sub 2}O{sub 7} composites have been investigated as potential IM materials. An adequate thermal conductivity of 6 W {center_dot} m{sup -} 1 {center_dot} K{sup -1} at 1000 C for the MgO-Nd{sub 2}Zr{sub 2}O{sub 7} composite with 90 vol% MgO was recently calculated and reported. Other simulations proposed that the MgO-pyrochlore composites could exhibit higher radiation stability than previously reported. Final optimization of the composite microstructure was performed on the 70 vol% MgO-Nd{sub 2}Zr{sub 2}O{sub 7} composite that ...
Date: January 11, 2009
Creator: PI: JUAN C. NINO, ASSOCIATE PROFESSOR
Partner: UNT Libraries Government Documents Department

Status report on high fidelity reactor simulation.

Description: This report presents the effort under way at Argonne National Laboratory toward a comprehensive, integrated computational tool intended mainly for the high-fidelity simulation of sodium-cooled fast reactors. The main activities carried out involved neutronics, thermal hydraulics, coupling strategies, software architecture, and high-performance computing. A new neutronics code, UNIC, is being developed. The first phase involves the application of a spherical harmonics method to a general, unstructured three-dimensional mesh. The method also has been interfaced with a method of characteristics. The spherical harmonics equations were implemented in a stand-alone code that was then used to solve several benchmark problems. For thermal hydraulics, a computational fluid dynamics code called Nek5000, developed in the Mathematics and Computer Science Division for coupled hydrodynamics and heat transfer, has been applied to a single-pin, periodic cell in the wire-wrap geometry typical of advanced burner reactors. Numerical strategies for multiphysics coupling have been considered and higher-accuracy efficient methods proposed to finely simulate coupled neutronic/thermal-hydraulic reactor transients. Initial steps have been taken in order to couple UNIC and Nek5000, and simplified problems have been defined and solved for testing. Furthermore, we have begun developing a lightweight computational framework, based in part on carefully selected open source tools, to nonobtrusively and efficiently integrate the individual physics modules into a unified simulation tool.
Date: December 11, 2006
Creator: Palmiotti, G.; Smith, M.; Rabiti, C.; Lewis, E.; Yang, W.; Leclere,M. et al.
Partner: UNT Libraries Government Documents Department

Godiva IV and Juliet Diagnostics CED-1, Rev. 1 (IER-176)

Description: The Juliet experiment is currently in preliminary design (IER-128). This experiment will utilize a suite of diagnostics to measure the physical state of the device (temperature, surface motion, stress, etc.) and the total and time rate of change of neutron and gamma fluxes. A variety of potential diagnostics has been proposed in this CED-1 report. Based on schedule and funding, a subset of diagnostics will be selected for testing using the Godiva IV pulsed reactor as a source of neutrons and gammas. The diagnostics development and testing will occur over a two year period (FY12-13) culminating in a final set of diagnostics to be fielded for he Juliet experiment currently proposed for execution in FY15.
Date: April 11, 2012
Creator: Scorby, J C & Myers, W L
Partner: UNT Libraries Government Documents Department

Requirements for Reactor Physics Design

Description: It has been recognized that there is a need for requirements and guidance for design and operation of nuclear power plants. This is becoming more important as more reactors are being proposed to be built. In parallel with activities in individual countries are norms established by international organizations. This paper discusses requirements/guidance for neutronic design and operation as promulgated by the U.S. Nuclear Regulatory Commission (NRC). As an example, details are given for one reactor physics parameter, namely, the moderator temperature reactivity coefficient. The requirements/guidance from the NRC are discussed in the context of those generated for the International Atomic Energy Agency. The requirements/guidance are not identical from the two sources although they are compatible.
Date: April 11, 2008
Creator: Diamond,D.J.
Partner: UNT Libraries Government Documents Department

FIFTH INTERIM STATUS REPORT: MODEL 9975 PCV O-RING FIXTURE LONG-TERM LEAK PERFORMANCE

Description: A series of experiments to monitor the aging performance of Viton{reg_sign} GLT O-rings used in the Model 9975 package has been ongoing for six years at the Savannah River National Laboratory. Sixty-seven mock-ups of 9975 Primary Containment Vessels (PCVs) were assembled and heated to temperatures ranging from 200 to 450 F. They were leak-tested initially and have been tested at nominal six month intervals to determine if they meet the criterion of leaktightness defined in ANSI standard N14.5-97. Fourteen additional tests were initiated in 2008 with GLT-S O-rings heated to temperatures ranging from 200 to 400 F. High temperature aging continues for 36 GLT O-ring fixtures at 200-350 F. Room temperature leak test failures have been experienced in 6 of the GLT O-ring fixtures aging at 300 and 350 F, and in all 3 of the GLT O-ring fixtures aging at higher temperatures. No failures have yet been observed in GLT O-ring fixtures aging at 200 F for 30-48 months, which is still bounding to O-ring temperatures during storage in KAMS. High temperature aging continues for 6 GLT-S O-ring fixtures at 200-300 F. Room temperature leak test failures have been experienced in all 8 of the GLT-S O-ring fixtures aging at 350 and 400 F. No failures have yet been observed in GLT-S O-ring fixtures aging at 200 or 300 F for 19 months. For O-ring fixtures that have failed the room temperature leak test and been disassembled, the Orings displayed a compression set ranging from 51-95%. This is significantly greater than seen to date for packages inspected during KAMS field surveillance (23% average). For GLT O-rings, service life based on the room temperature leak rate criterion is comparable to that predicted by compression stress relaxation (CSR) data at higher temperatures (350-400 F). While there are no comparable failure data ...
Date: April 11, 2011
Creator: Daugherty, W. & Hoffman, E.
Partner: UNT Libraries Government Documents Department

A Comparison of Proliferation Resistance Measures of Misuse Scenarios Using a Markov Approach

Description: Misuse of declared nuclear facilities is one of the important proliferation threats. The robustness of a facility against these threats is characterized by a number of proliferation resistance (PR) measures. This paper evaluates and compares PR measures for several misuse scenarios using a Markov model approach to implement the pathway analysis methodology being developed by the PR&PP (Proliferation Resistance and Physical Protection) Expert Group. Different misue strategies can be adopted by a proliferator and each strategy is expected to have different impacts on the proliferator's success. Selected as the probabilistic measure to represent proliferation resistance, the probabilities of the proliferator's success of misusing a hypothetical ESFR (Example Sodium Fast Reactor) facility system are calculated using the Markov model based on the pathways constructed for individual misuse scenarios. Insights from a comparison of strategies that are likely to be adopted by the proliferator are discussed in this paper.
Date: May 11, 2008
Creator: Yue,M.; Cheng, L.-Y. & Bari, R.
Partner: UNT Libraries Government Documents Department

LFR "Lead-Cooled Fast Reactor"

Description: The main purpose of this paper is to present the current status of development of the Lead-cooled Fast Reactor (LFR) in Generation IV (GEN IV), including the European contribution, to identify needed R&D and to present the corresponding GEN IV International Forum (GIF) R&D plan [1] to support the future development and deployment of lead-cooled fast reactors. The approach of the GIF plan is to consider the research priorities of each member country in proposing an integrated, coordinated R&D program to achieve common objectives, while avoiding duplication of effort. The integrated plan recognizes two principal technology tracks: (1) a small, transportable system of 10-100 MWe size that features a very long refuelling interval, and (2) a larger-sized system rated at about 600 MWe, intended for central station power generation. This paper provides some details of the important European contributions to the development of the LFR. Sixteen European organizations have, in fact, taken the initiative to present to the European Commission the proposal for a Specific Targeted Research and Training Project (STREP) devoted to the development of a European Lead-cooled System, known as the ELSY project; two additional organizations from the US and Korea have joined the project. Consequently, ELSY will constitute the reference system for the large lead-cooled reactor of GEN IV. The ELSY project aims to demonstrate the feasibility of designing a competitive and safe fast power reactor based on simple technical engineered features that achieves all of the GEN IV goals and gives assurance of investment protection. As far as new technology development is concerned, only a limited amount of R&D will be conducted in the initial phase of the ELSY project since the first priority is to define the design guidelines before launching a larger and expensive specific R&D program. In addition, the ELSY project is expected ...
Date: May 11, 2006
Creator: Cinotti, L; Fazio, C; Knebel, J; Monti, S; Abderrahim, H A; Smith, C et al.
Partner: UNT Libraries Government Documents Department

ENERGY EFFICIENCY LIMITS FOR A RECUPERATIVE BAYONET SULFURIC ACID DECOMPOSITION REACTOR FOR SULFUR CYCLE THERMOCHEMICAL HYDROGEN PRODUCTION

Description: A recuperative bayonet reactor design for the high-temperature sulfuric acid decomposition step in sulfur-based thermochemical hydrogen cycles was evaluated using pinch analysis in conjunction with statistical methods. The objective was to establish the minimum energy requirement. Taking hydrogen production via alkaline electrolysis with nuclear power as the benchmark, the acid decomposition step can consume no more than 450 kJ/mol SO{sub 2} for sulfur cycles to be competitive. The lowest value of the minimum heating target, 320.9 kJ/mol SO{sub 2}, was found at the highest pressure (90 bar) and peak process temperature (900 C) considered, and at a feed concentration of 42.5 mol% H{sub 2}SO{sub 4}. This should be low enough for a practical water-splitting process, even including the additional energy required to concentrate the acid feed. Lower temperatures consistently gave higher minimum heating targets. The lowest peak process temperature that could meet the 450-kJ/mol SO{sub 2} benchmark was 750 C. If the decomposition reactor were to be heated indirectly by an advanced gas-cooled reactor heat source (50 C temperature difference between primary and secondary coolants, 25 C minimum temperature difference between the secondary coolant and the process), then sulfur cycles using this concept could be competitive with alkaline electrolysis provided the primary heat source temperature is at least 825 C. The bayonet design will not be practical if the (primary heat source) reactor outlet temperature is below 825 C.
Date: June 11, 2009
Creator: Gorensek, M. & Edwards, T.
Partner: UNT Libraries Government Documents Department

Review of environmental effects on fatigue crack growth of austenitic stainless steels.

Description: Fatigue and environmentally assisted cracking of piping, pressure vessel cladding, and core components in light water reactors are potential concerns to the nuclear industry and regulatory agencies. The degradation processes include intergranular stress corrosion cracking of austenitic stainless steel (SS) piping in boiling water reactors (BWRs), and propagation of fatigue or stress corrosion cracks (which initiate in sensitized SS cladding) into low-alloy ferritic steels in BWR pressure vessels. Crack growth data for wrought and cast austenitic SSs in simulated BWR water, developed at Argonne National Laboratory under US Nuclear Regulatory Commission sponsorship over the past 10 years, have been compiled into a data base along with similar data obtained from the open literature. The data were analyzed to develop corrosion-fatigue curves for austenitic SSs in aqueous environments corresponding to normal BWR water chemistries, for BWRs that add hydrogen to the feedwater, and for pressurized water reactor primary-system-coolant chemistry. The corrosion-fatigue data and curves in water were compared with the air line in Section XI of the ASME Code.
Date: July 11, 1994
Creator: Shack, W. J.; Kassner, T. F. & Technology, Energy
Partner: UNT Libraries Government Documents Department

Supercritical carbon dioxide cycle control analysis.

Description: This report documents work carried out during FY 2008 on further investigation of control strategies for supercritical carbon dioxide (S-CO{sub 2}) Brayton cycle energy converters. The main focus of the present work has been on investigation of the S-CO{sub 2} cycle control and behavior under conditions not covered by previous work. An important scenario which has not been previously calculated involves cycle operation for a Sodium-Cooled Fast Reactor (SFR) following a reactor scram event and the transition to the primary coolant natural circulation and decay heat removal. The Argonne National Laboratory (ANL) Plant Dynamics Code has been applied to investigate the dynamic behavior of the 96 MWe (250 MWt) Advanced Burner Test Reactor (ABTR) S-CO{sub 2} Brayton cycle following scram. The timescale for the primary sodium flowrate to coast down and the transition to natural circulation to occur was calculated with the SAS4A/SASSYS-1 computer code and found to be about 400 seconds. It is assumed that after this time, decay heat is removed by the normal ABTR shutdown heat removal system incorporating a dedicated shutdown heat removal S-CO{sub 2} pump and cooler. The ANL Plant Dynamics Code configured for the Small Secure Transportable Autonomous Reactor (SSTAR) Lead-Cooled Fast Reactor (LFR) was utilized to model the S-CO{sub 2} Brayton cycle with a decaying liquid metal coolant flow to the Pb-to-CO{sub 2} heat exchangers and temperatures reflecting the decaying core power and heat removal by the cycle. The results obtained in this manner are approximate but indicative of the cycle transient performance. The ANL Plant Dynamics Code calculations show that the S-CO{sub 2} cycle can operate for about 400 seconds following the reactor scram driven by the thermal energy stored in the reactor structures and coolant such that heat removal from the reactor exceeds the decay heat generation. Based on the results, ...
Date: April 11, 2011
Creator: Moisseytsev, A. & Sienicki, J. J. (Nuclear Engineering Division)
Partner: UNT Libraries Government Documents Department

Survivable pulse power space radiator

Description: A thermal radiator system is described for use on an outer space vehicle, which must survive a long period of nonuse and then radiate large amounts of heat for a limited period of time. The radiator includes groups of radiator panels that are pivotally connected in tandem, so that they can be moved to deployed configuration wherein the panels lie largely coplanar, and to a stowed configuration wherein the panels lie in a stack to resist micrometerorite damage. The panels are mounted on a boom which separates a hot power source from a payload. While the panels are stowed, warm fluid passes through their arteries to keep them warm enough to maintain the coolant in a liquid state and avoid embrittlement of material. The panels can be stored in a largely cylindrical shell, with panels progressively further from the boom being of progressively shorter length. 5 figs.
Date: March 11, 1988
Creator: Mims, J.; Buden, D. & Williams, K.
Partner: UNT Libraries Government Documents Department

Development of small, fast reactor core designs using lead-based coolant.

Description: A variety of small (100 MWe) fast reactor core designs are developed, these include compact configurations, long-lived (15-year fuel lifetime) cores, and derated, natural circulation designs. Trade studies are described which identify key core design issues for lead-based coolant systems. Performance parameters and reactivity feedback coefficients are compared for lead-bismuth eutectic (LBE) and sodium-cooled cores of consistent design. The results of these studies indicate that the superior neutron reflection capability of lead alloys reduces the enrichment and burnup swing compared to conventional sodium-cooled systems; however, the discharge fluence is significantly increased. The size requirement for long-lived systems is constrained by reactivity loss considerations, not fuel burnup or fluence limits. The derated lead-alloy cooled natural circulation cores require a core volume roughly eight times greater than conventional compact systems. In general, reactivity coefficients important for passive safety performance are less favorable for the larger, derated configurations.
Date: June 11, 1999
Creator: Cahalan, J. E.; Hill, R. N.; Khalil, H. S. & Wade, D. C.
Partner: UNT Libraries Government Documents Department

Examination of spent PWR fuel rods after 15 years in dry storage.

Description: Virginia Power Surry Nuclear Station Pressurized Water Reactor (PWR) fuel was stored in a dry inert atmosphere Castor V/21 cask at the Idaho National Environmental and Engineering Laboratory (INEEL) for 15 years at peak cladding temperatures decreasing from about 350 to 150 C. Prior to the storage, the loaded cask was subjected to extensive thermal benchmark tests. The cask was opened to examine the fuel for degradation and to determine if it was suitable for extended storage. No rod breaches had occurred and no visible degradation or crud/oxide spallation were observed. Twelve rods were removed from the center of the T11 assembly and shipped from INEEL to the Argonne-West HFEF for profilometric scans. Four of these rods were punctured to determine the fission gas release from the fuel matrix and internal pressure in the rods. Three of the four rods were cut into five segments each, then shipped to the Argonne-East AGHCF for detailed examination. The test plan calls for metallographic examination of six samples from two of the rods, microhardness and hydrogen content measurements at or near the six metallographic sample locations, tensile testing of six samples from the two rods, and thermal creep testing of eight samples from the two rods to determine the extent of residual creep life. The results from the profilometry (12 rods), gas release measurements (4 rods), metallographic examinations (2 samples from 1 rod), and microhardness and hydrogen content characterization (2 samples from 1 rod) are reported here. The tensile and creep studies are just starting and will be reported at a later date, along with the additional characterization work to be performed. Although only limited prestorage characterization is available, a number of preliminary conclusions can be drawn based on comparison with characterization of Florida Power Turkey Point rods of a similar vintage. Based ...
Date: February 11, 2002
Creator: Einziger, R. E.; Tsai, H. C.; Billone, M. C. & Hilton, B. A.
Partner: UNT Libraries Government Documents Department

Commercial Light Water Reactor Tritium Extraction Facility Geotechnical Summary Report

Description: A geotechnical investigation program has been completed for the Circulating Light Water Reactor - Tritium Extraction Facility (CLWR-TEF) at the Savannah River Site (SRS). The program consisted of reviewing previous geotechnical and geologic data and reports, performing subsurface field exploration, field and laboratory testing and geologic and engineering analyses. The purpose of this investigation was to characterize the subsurface conditions for the CLWR-TEF in terms of subsurface stratigraphy and engineering properties for design and to perform selected engineering analyses. The objectives of the evaluation were to establish site-specific geologic conditions, obtain representative engineering properties of the subsurface and potential fill materials, evaluate the lateral and vertical extent of any soft zones encountered, and perform engineering analyses for slope stability, bearing capacity and settlement, and liquefaction potential. In addition, provide general recommendations for construction and earthwork.
Date: January 11, 2000
Creator: Lewis, M.R.
Partner: UNT Libraries Government Documents Department

Design Studies of ``100% Pu'' Mox Lead Test Assembly

Description: In this document the results of neutronics studies of <<100%Pu>> MOX LTA design are presented. The parametric studies of infinite MOX-UOX grids, MOX-UOX core fragments and of VVER-1000 core with 3 MOX LTAs are performed. The neutronics parameters of MOX fueled core have been performed for the chosen design MOX LTA using the Russian 3D code BIPR-7A and 2D code PERMAK-A with the constants prepared by the cell spectrum code TVS-M.
Date: January 11, 2001
Creator: Pavlovichev, A.M.
Partner: UNT Libraries Government Documents Department

Developing ``SMART'' Equipment and Systems through Collaborative NERI Research and Development: A First Year of Progress

Description: The US Department of Energy (DOE) created the Nuclear Energy Research Initiative (NERI) in 1999 to conduct research and development with the objectives of: (1) overcoming the principal technical obstacles to expanded nuclear energy use, (2) advancing the state of nuclear technology to maintain its competitive position in domestic and world markets, and (3) improving the performance, efficiency, reliability, and economics of nuclear energy. The NERI program is now beginning its second year with increased funding and an emphasis on international participation. Among the programs selected for funding was the ``Smart Equipment and Systems to Improve Reliability and Safety in Future Nuclear Power Plant Operations''. This program is a 36 month collaborative effort bringing together the technical capabilities of Westinghouse Nuclear Automation, Sandia National Laboratories, Duke Engineering and Services (DE and S), Massachusetts Institute of Technology (MIT) and Pennsylvania State University (PSU). The goal of the program is to design, develop, and evaluate an integrated set of tools and methodologies that can improve the reliability and safety of advanced nuclear power plants through the introduction of smart equipment and predictive maintenance technology. The results have implications for reduced construction costs. This paper discusses: (1) the goals and significance of the program, (2) the significant achievements of the program's first year and the current direction for its continuing efforts and (3) potential cooperation with the domestic nuclear and component manufacturing industries, and with international organizations.
Date: September 11, 2000
Creator: HARMON,DARYL L.; GOLAY,MICHAEL W.; CHAPMAN,LEON D. & MAYNARD,KENNETH P.
Partner: UNT Libraries Government Documents Department

Fuel-Cycle energy and emission impacts of ethanol-diesel blends in urban buses and farming tractors.

Description: About 2.1 billion gallons of fuel ethanol was used in the United States in 2002, mainly in the form of gasoline blends containing up to 10% ethanol (E10). Ethanol use has the potential to increase in the U.S. blended gasoline market because methyl tertiary butyl ether (MTBE), formerly the most popular oxygenate blendstock, may be phased out owing to concerns about MTBE contamination of the water supply. Ethanol would remain the only viable near-term option as an oxygenate in reformulated gasoline production and to meet a potential federal renewable fuels standard (RFS) for transportation fuels. Ethanol may also be blended with additives (co-solvents) into diesel fuels for applications in which oxygenation may improve diesel engine emission performance. Numerous studies have been conducted to evaluate the fuel-cycle energy and greenhouse gas (GHG) emission effects of ethanol-gasoline blends relative to those of gasoline for applications in spark-ignition engine vehicles (see Wang et al. 1997; Wang et al. 1999; Levelton Engineering et al. 1999; Shapouri et al. 2002; Graboski 2002). Those studies did not address the energy and emission effects of ethanol-diesel (E-diesel or ED) blends relative to those of petroleum diesel fuel in diesel engine vehicles. The energy and emission effects of E-diesel could be very different from those of ethanol-gasoline blends because (1) the energy use and emissions generated during diesel production (so-called ''upstream'' effects) are different from those generated during gasoline production; and (2) the energy and emission performance of E-diesel and petroleum diesel fuel in diesel compression-ignition engines differs from that of ethanol-gasoline blends in spark-ignition (Otto-cycle-type) engine vehicles. The Illinois Department of Commerce and Community Affairs (DCCA) commissioned Argonne National Laboratory to conduct a full fuel-cycle analysis of the energy and emission effects of E-diesel blends relative to those of petroleum diesel when used in the types of ...
Date: September 11, 2003
Creator: Wang, M.; Saricks, C. & Lee, H.
Partner: UNT Libraries Government Documents Department

Ultrasonic Phased Array Technique for Accurate Flaw Sizing in Dissimilar Metal Welds

Description: Described is a manual,portable non-destructive technique to determine the through wall height of cracks present in dissimilar metal welds used in the primary coolling systems of pressure water and boiler light water reactors. Current manual methods found in industry have proven not to exhibit the sizing accuracy required by ASME inspection requirement. The technique described demonstrated an accuracy approximately three times that required to ASME Section XI, Appendix 8 qualification.
Date: March 11, 2005
Creator: Buttram, Jonathan D
Partner: UNT Libraries Government Documents Department

Geologic Investigation of a Potential Site for a Next-Generation Reactor Neutrino Oscillation Experiment -- Diablo Canyon, San Luis Obispo County, CA

Description: This report provides information on the geology and selected physical and mechanical properties of surface rocks collected at Diablo Canyon, San Luis Obispo County, California as part of the design and engineering studies towards a future reactor neutrino oscillation experiment. The main objective of this neutrino project is to study the process of neutrino flavor transformation or neutrino oscillation by measuring neutrinos produced in the fission reactions of a nuclear power plant. Diablo Canyon was selected as a candidate site because it allows the detectors to be situated underground in a tunnel close to the source of neutrinos (i.e., at a distance of several hundred meters from the nuclear power plant) while having suitable topography for shielding against cosmic rays. The detectors have to be located underground to minimize the cosmic ray-related background noise that can mimic the signal of reactor neutrino interactions in the detector. Three Pliocene-Miocene marine sedimentary units dominate the geology of Diablo Canyon: the Pismo Formation, the Monterey Formation, and the Obispo Formation. The area is tectonically active, located east of the active Hosgri Fault and in the southern limb of the northwest trending Pismo Syncline. Most of the potential tunnel for the neutrino detector lies within the Obispo Formation. Review of previous geologic studies, observations from a field visit, and selected physical and mechanical properties of rock samples collected from the site provided baseline geological information used in developing a preliminary estimate for tunneling construction cost. Gamma-ray spectrometric results indicate low levels of radioactivity for uranium, thorium, and potassium. Grain density, bulk density, and porosity values for these rock samples range from 2.37 to 2.86 g/cc, 1.41 to 2.57 g/cc, and 1.94 to 68.5 percent respectively. Point load, unconfined compressive strength, and ultrasonic velocity tests were conducted to determine rock mechanical and acoustic properties. The ...
Date: June 11, 2004
Creator: Onishi, Celia Tiemi; Dobson, Patrick; Nakagawa, Seiji; Glaser, Steven & Galic, Dom
Partner: UNT Libraries Government Documents Department

Development of a Robust Tri-Carbide Fueled Reactor for Multimegawatt Space Power and Propulsion Applications

Description: An innovative reactor core design based on advanced, mixed carbide fuels was analyzed for nuclear space power applications. Solid solution, mixed carbide fuels such as (U,Zr,Nb)c and (U,Zr, Ta)C offer great promise as an advanced high temperature fuel for space power reactors.
Date: August 11, 2004
Creator: Anghaie, Samim; Knight, Travis W.; Plancher, Johann & Gouw, Reza
Partner: UNT Libraries Government Documents Department