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Recommendations for Addressing Axial Burnup in the PWR Burnup Credit Analyses

Description: This report presents studies performed to support the development of a technically justifiable approach for addressing the axial-burnup distribution in pressurized-water reactor (PWR) burnup-credit criticality safety analyses. The effect of the axial-burnup distribution on reactivity and proposed approaches for addressing the axial-burnup distribution are briefly reviewed. A publicly available database of profiles is examined in detail to identify profiles that maximize the neutron multiplication factor, k{sub eff}, assess its adequacy for PWR burnup credit analyses, and investigate the existence of trends with fuel type and/or reactor operations. A statistical evaluation of the k{sub eff} values associated with the profiles in the axial-burnup-profile database was performed, and the most reactive (bounding) profiles were identified as statistical outliers. The impact of these bounding profiles on k{sub eff} is quantified for a high-density burnup credit cask. Analyses are also presented to quantify the potential reactivity consequence of loading assemblies with axial-burnup profiles that are not bounded by the database. The report concludes with a discussion on the issues for consideration and recommendations for addressing axial burnup in criticality safety analyses using burnup credit for dry cask storage and transportation.
Date: October 23, 2002
Creator: Wagner, J.C.
Partner: UNT Libraries Government Documents Department


Description: The experiments conducted this past quarter have suggested that the bacterium Pseudomonas fluorescens strain CL0145A is effective at killing zebra mussels throughout the entire range of pH values tested (7.2 to 8.6). Highest mortality was achieved at pH values characteristic of preferred zebra mussel waterbodies, i.e., hard waters with a range of 7.8 to 8.6. In all water types tested, however, ranging from very soft to very hard, considerable mussel kill was achieved (83 to 99% mean mortality), suggesting that regardless of the pH or hardness of the treated water, significant mussel kill can be achieved upon treatment with P. fluorescens strain CL0145A. These results further support the concept that this bacterium has significant potential for use as a zebra mussel control agent in power plant pipes receiving waters with a wide range of physical and chemical characteristics.
Date: October 15, 2002
Creator: Molloy, Daniel P.
Partner: UNT Libraries Government Documents Department

The Effect of pH on Nickel Alloy SCC and Corrosion Performance

Description: Alloy X-750 condition HTH stress corrosion crack growth rate (SCCGR) tests have been conducted at 360 C (680 F) with 50 cc/kg hydrogen as a function of coolant pH. Results indicate no appreciable influence of pH on crack growth in the pH (at 360 C) range of {approx} 6.2 to 8.7, consistent with previous alloy 600 findings. These intermediate pH results suggest that pH is not a key variable which must be accounted for when modeling pressurized water reactor (PWR) primary water SCC. In this study, however, a nearly three fold reduction in X-750 crack growth rate was observed in reduced pH environments (pH 3.8 through HCl addition and pH 4-5.3 through H{sub 2}SO{sub 4} addition). Crack growth rates did not directly correlate with corrosion film thickness. In fact, 10x thicker corrosion films were observed in the reduced pH environments.
Date: October 10, 2002
Creator: Morton, D.S. & Hansen, M.
Partner: UNT Libraries Government Documents Department


Description: The purpose of this document is to evaluate the potential for criticality for the fissile material that could accumulate in the near-field (invert) and in the far-field (host rock) beneath the U.S. Department of Energy (DOE) spent nuclear fuel (SNF) codisposal waste packages (WPs) as they degrade in the proposed monitored geologic repository at Yucca Mountain. The scope of this calculation is limited to the following DOE SNF types: Shippingport Pressurized Water Reactor (PWR), Enrico Fermi, Fast Flux Test Facility (FFTF), Fort St. Vrain, Melt and Dilute, Shippingport Light Water Breeder Reactor (LWBR), N-Reactor, and Training, Research, Isotope, General Atomics reactor (TRIGA). The results of this calculation are intended to be used for estimating the probability of criticality in the near-field and in the far-field. There are no limitations on use of the results of this calculation. The calculation is associated with the waste package design and was developed in accordance with the technical work plan, ''Technical Work Plan for: Department of Energy Spent Nuclear Fuel and Plutonium Disposition Work Packages'' (Bechtel SAIC Company, LLC [BSC], 2002a). This calculation is subject to the Quality Assurance Requirements and Description (QARD) per the activity evaluation under work package number P6212310Ml in the technical work plan TWP-MGR-MD-0000 10 REV 01 (BSC 2002a).
Date: October 18, 2002
Creator: Radulescu, H.
Partner: UNT Libraries Government Documents Department

A particle-bed gas cooled fast reactor core design for waste minimization.

Description: The issue of waste minimization in advanced reactor systems has been investigated using the Particle-Bed Gas-Cooled Fast Reactor (PB-GCFR) design being developed and funded under the U.S. Department of Energy Nuclear Energy Research Initiative (USDOE NERI) Program. Results indicate that for the given core power density and constraint on the maximum TRU enrichment allowable, the lowest amount of radiotoxic transuranics to be processed and hence sent to the repository is obtained for long-life core designs. Calculations were additionally done to investigate long-life core designs using LWR spent fuel TRU and recycle TRU, and different feed, matrix and reflector materials. The recycled TRU and LWR spent TRU fuels give similar core behaviors, because of the fast spectrum environment which does not significantly degrade the TRU composition. Using light elements as reflector material was found to be unattractive because of power peaking problems and large reactivity swings. The application of a lead reflector gave the longest cycle length and lowest TRU processing requirement. Materials compatibility and performance issues require additional investigation.
Date: October 11, 2002
Creator: Hoffman, E. A.; Taiwo, T. A.; Yang, W. S. & Fatone, M.
Partner: UNT Libraries Government Documents Department

TRISO-Coated Fuel Processing to Support High Temperature Gas-Cooled Reactors

Description: The initial objective of the work described herein was to identify potential methods and technologies needed to disassemble and dissolve graphite-encapsulated, ceramic-coated gas-cooled-reactor spent fuels so that the oxide fuel components can be separated by means of chemical processing. The purpose of this processing is to recover (1) unburned fuel for recycle, (2) long-lived actinides and fission products for transmutation, and (3) other fission products for disposal in acceptable waste forms. Follow-on objectives were to identify and select the most promising candidate flow sheets for experimental evaluation and demonstration and to address the needs to reduce technical risks of the selected technologies. High-temperature gas-cooled reactors (HTGRs) may be deployed in the next -20 years to (1) enable the use of highly efficient gas turbines for producing electricity and (2) provide high-temperature process heat for use in chemical processes, such as the production of hydrogen for use as clean-burning transportation fuel. Also, HTGR fuels are capable of significantly higher burn-up than light-water-reactor (LWR) fuels or fast-reactor (FR) fuels; thus, the HTGR fuels can be used efficiently for transmutation of fissile materials and long-lived actinides and fission products, thereby reducing the inventory of such hazardous and proliferation-prone materials. The ''deep-burn'' concept, described in this report, is an example of this capability. Processing of spent graphite-encapsulated, ceramic-coated fuels presents challenges different from those of processing spent LWR fuels. LWR fuels are processed commercially in Europe and Japan; however, similar infrastructure is not available for processing of the HTGR fuels. Laboratory studies on the processing of HTGR fuels were performed in the United States in the 1960s and 1970s, but no engineering-scale processes were demonstrated. Currently, new regulations concerning emissions will impact the technologies used in processing the fuel. Potential processing methods will be identified both by a review of the literature regarding the ...
Date: October 1, 2002
Creator: Del Cul, G.D.
Partner: UNT Libraries Government Documents Department


Description: The purpose of this study is to determine the effect of the initial power level during a rod ejection accident (REA) on the ejected rod worth and the resulting energy deposition in the fuel. The model used is for the hot zero power (HZP) conditions at the end of a typical fuel cycle for the Three Mile Island Unit 1 pressurized water reactor. PARCS, a transient, three-dimensional, two-group neutron nodal diffusion code, coupled with its own thermal-hydraulics model, is used to perform both steady-state and transient simulations. The worth of an ejected control rod is affected by both power level, and the positions of control banks. As the power level is increased, the worth of a single central control rod tends to drop due to thermal-hydraulic feedback and control bank removal, both of which flatten the radial neutron flux and power distributions. Although the peak fuel pellet enthalpy rise during an REA will be greater for a given ejected rod worth at elevated initial power levels, it is more likely the HZP condition will cause a greater net energy deposition because an ejected rod will have the highest worth at HZP. Thus, the HZP condition can be considered the most conservative in a safety evaluation.
Date: October 7, 2002
Partner: UNT Libraries Government Documents Department

The sodium-cooled fast reactor (SFR).

Description: The primary mission for the SFR is the management of high-level wastes, and in particular, management of plutonium and other actinides. The Generation IV Roadmap Fuel Cycle Crosscut Group (FCCG) found that the limiting factor facing an essential role for nuclear energy with the once-through cycle is the availability of repository space worldwide [FCCG Report]. This becomes an important issue, requiring new repository development in only a few decades. Systems that employ a fully closed fuel cycle hold the promise to reduce repository space and performance requirements, although their costs must be held to acceptable levels. Closed fuel cycles, working alone or symbiotically with systems using a once-through cycle, permit partitioning the nuclear waste and management of each partitioned fraction. In the longer term, beyond 50 years, or if major new missions requiring nuclear energy production (such as a major growth in the use of hydrogen as an energy carrier) develop, uranium resource availability also becomes a limiting factor unless breakthroughs occur in mining or extraction technologies. Fast spectrum reactors have the ability to utilize almost all of the energy in the natural uranium versus the 1% utilized in thermal spectrum systems.
Date: October 25, 2002
Creator: Lineberry, M. J. & Allen, T. R.
Partner: UNT Libraries Government Documents Department