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Synergistic Failure of BWR Internals

Description: Boiling Water Reactor (BWR) core shrouds and other reactor internals important to safety are experiencing intergranular stress corrosion cracking (IGSCC). The United States Nuclear Regulatory Commission has followed the problem, and as part of its investigations, contracted with the Idaho National Engineering and Environmental Laboratory to conduct a risk assessment. The overall project objective is to assess the potential consequences and risks associated with the failure of IGSCC-susceptible BWR vessel internals, with specific consideration given to potential cascading and common mode effects. An initial phase has been completed in which background material was gathered and evaluated, and potential accident sequences were identified. A second phase is underway to perform a simplified, quantitative probabilistic risk assessment on a representative high-power BWR/4. Results of the initial study conducted on the jet pumps show that any cascading failures would not result in a significant increase in the core damage frequency. The methodology is currently being extended to other major reactor internals components.
Date: October 1, 1999
Creator: Ware, Arthur Gates & Chang, T-Y
Partner: UNT Libraries Government Documents Department

REGIONAL BINNING FOR CONTINUED STORAGE OF SPENT NUCLEAR FUEL AND HIGH-LEVEL WASTES

Description: In the Continued Storage Analysis Report (CSAR) (Reference 1), DOE decided to analyze the environmental consequences of continuing to store the commercial spent nuclear fuel (SNF) at 72 commercial nuclear power sites and DOE-owned spent nuclear fuel and high-level waste at five Department of Energy sites by region rather than by individual site. This analysis assumes that three commercial facilities pairs--Salem and Hope Creek, Fitzpatrick and Nine-Mile Point, and Dresden and Moms--share common storage due to their proximity to each other. The five regions selected for this analysis are shown on Figure 1. Regions 1, 2, and 3 are the same as those used by the Nuclear Regulatory Commission in their regulatory oversight of commercial power reactors. NRC Region 4 was subdivided into two regions to more appropriately define the two different climates that exist in NRC Region 4. A single hypothetical site in each region was assumed to store all the SNF and HLW in that region. Such a site does not exist and has no geographic location but is a mathematical construct for analytical purposes. To ensure that the calculated results for the regional analyses reflect appropriate inventory, facility and material degradation, and radionuclide transport, the waste inventories, engineered barriers, and environmental conditions for the hypothetical sites were developed from data for each of the existing sites within the given region. Weighting criteria to account for the amount and types of SNF and HLW at each site were used in the development of the environmental data for the regional site, such that the results of the analyses for the hypothetical site were representative of the sum of the results of each actual site if they had been modeled independently. This report defines the actual site data used in development of this hypothetical site, shows how the individual site ...
Date: October 1, 1998
Creator: W. Lee Poe, Jr
Partner: UNT Libraries Government Documents Department

Volumes, Masses, and Surface Areas for Shippingport LWBR Spent Nuclear Fuel in a DOE SNF Canister

Description: The purpose of this calculation is to estimate volumes, masses, and surface areas associated with (a) an empty Department of Energy (DOE) 18-inch diameter, 15-ft long spent nuclear fuel (SNF) canister, (b) an empty DOE 24-inch diameter, 15-ft long SNF canister, (c) Shippingport Light Water Breeder Reactor (LWBR) SNF, and (d) the internal basket structure for the 18-in. canister that has been designed specifically to accommodate Seed fuel from the Shippingport LWBR. Estimates of volumes, masses, and surface areas are needed as input to structural, thermal, geochemical, nuclear criticality, and radiation shielding calculations to ensure the viability of the proposed disposal configuration.
Date: October 22, 1999
Creator: Davis, J.W.
Partner: UNT Libraries Government Documents Department

Safety Evaluation Report Restart of K-Reactor Savannah River Site

Description: In April 1991, the Department of Energy (DOE) issued DOE/DP-0084T, Safety Evaluation Report Restart of K-Reactor Savannah River Site.'' The Safety Evaluation Report (SER) documents the results of DOE reviews and evaluations of the programmatic aspects of a large number of issues necessary to be satisfactorily addressed before restart. The issues were evaluated for compliance with the restart criteria included in the SER. The results of those evaluations determined that the restart criteria had been satisfied for some of the issues. However, for most of the issues at least part of the applicable restart criteria had not been found to be satisfied at the time the evaluations were prepared. For those issues, open or confirmatory items were identified that required resolution. In August 1991, DOE issued DOE/DP-0090T, Safety Evaluation Report Restart of K-Reactor Savannah River Site Supplement 1.'' That document was the first Supplement to the April 1991 SER, and documented the resolution of 62 of the open items identified in the SER. This document is the second Supplement to the April 1991 SER. This second SER Supplement documents the resolution of additional open times identified in the SER, and includes a complete list of all remaining SER open items. The resolution of those remaining open items will be documented in future SER Supplements. Resolution of all open items for an issue indicates that its associated restart criteria have been satisfied, and that DOE concludes that the programmatic aspects of the issue have been satisfactorily addressed.
Date: October 1, 1991
Partner: UNT Libraries Government Documents Department

Chemical and Radiochemical Constituents in Water from Wells in the Vicinity of the Naval Reactors Facility, Idaho National Engineering and Environmental Laboratory, Idaho, 1996

Description: The U.S. Geological Survey, in response to a request from the U.S. Department of Energy's Pittsburgh Naval Reactors Office, Idaho Branch Office (IBO), samples water from 13 wells during 1996 as part of a long-term project to monitor water quality to the Snake River Plain aquifer in the vicinity of the Naval Reactors Facility (NRF), Idaho National Engineering and Environmental Laboratory, Idaho. The IBO requires information about the mobility of radionuclide- and chemical-waste constituents in the Snake River Plain aquifer. Waste-constituent mobility is determined principally by (1) the rate and direction of ground-water flow; (2) the locations, quantities, and methods of waste disposal; (3) waste-constituents chemistry; and (4) the geochemical processes taking place in the aquifer. The purpose of the data-collection program is to provide IBO with water-chemistry data to evaluate the effect of NRF activities on the water quality of the Snake River Plain aquifer. Water samples were analyzed for naturally occurring constituents and man-made contaminants.
Date: October 1, 1999
Creator: Knobel, L. L.; Bartholomay, R. C.; Tucker, B. J. & Williams, L. M.
Partner: UNT Libraries Government Documents Department

ABWR (advanced boiling water reactor) Design Verification Program

Description: The ABWR Design Verification Program is aimed at restoring confidence in the US licensing process by demonstrating its workability by obtaining USNRC preapproval of GE's ABWR Standard Plant. The purpose of this work is to achieve full NRC approval of the ABWR through the award of an NRC Staff final design approval (FDA) and design certification. The approach is to (1) establish a licensing basis with the NRC Staff for the ABWR, (2) prepare and submit, for NRC Staff review, an SSAR to obtain an FDA, and (3) participate in a rulemaking process to obtain certification of the ABWR design. This program was initiated August 27, 1986. This report, the fourth annual progress report, summarizes progress on this program from October 1, 1989 through September 30, 1990. 9 refs., 5 tabs.
Date: October 1, 1990
Creator: Fox, J.N.
Partner: UNT Libraries Government Documents Department

Transactions of the nineteenth water reactor safety information meeting

Description: This report contains summaries of papers on reactor safety research to be presented at the 19th Water Reactor Safety Information Meeting at the Bethesda Marriott Hotel in Bethesda, Maryland, October 28--30, 1991. The summaries briefly describe the programs and results of nuclear safety research sponsored by the Office of Nuclear Regulatory Research, USNRC. Summaries of invited papers concerning nuclear safety issues from US government laboratories, the electric utilities, the Electric Power Research Institute (EPRI), the nuclear industry, and from the governments and industry in Europe and Japan are also included. The summaries have been compiled in one report to provide a basis for meaningful discussion and information exchange during the course of the meeting, and are given in the order of their presentation in each session. The individual summaries have been cataloged separately.
Date: October 1, 1991
Creator: Weiss, A. J.
Partner: UNT Libraries Government Documents Department

Transformer failure and common-mode loss of instrument power at Nine Mile Point Unit 2 on August 13, 1991

Description: On August 13, 1991, at Nine Mile Point Unit 2 nuclear power plant, located near Scriba, New York, on Lake Ontario, the main transformer experienced an internal failure that resulted in degraded voltage which caused the simultaneous loss of five uninterruptible power supplies, which in turn caused the loss of several nonsafety systems, including reactor control rod position indication, some reactor power and water indication, control room annunciators, the plant communications system, the plant process computer, and lighting at some locations. The reactor was subsequently brought to a safe shutdown. Following this event, the US Nuclear Regulatory Commission dispatched an Incident Investigation Team to the site to determine what happened, to identify the probable causes, and to make appropriate findings and conclusions. This report describes the incident, the methodology used by the team in its investigation, and presents and the team's findings and conclusions. 59 figs., 14 tabs.
Date: October 1, 1991
Partner: UNT Libraries Government Documents Department

Comparisons of HELIOS Calculated Isotope Concentrations to Measured Values for Several Reactor Systems

Description: Heavy metal and fission product noble gas concentrations in spent fuel from two different PWR'S were calculated using HELIOS and compared to measured results from the literature. It was found that for the U-235/U-238 and Pu-240/Pu-239 isotopic ratios, the HELIOS calculation agreed to within the experimental uncertainty. For the Xe-131/Xe-134 isotopic ratios, HELIOS tended to overestimate the result by up to 4%. Conversely for the Xe-132/Xe-134 ratios, HELIOS underestimated the result by a slight amount ({approximately}1%). This suggests that either the fission product yields for Xe-131 and Xe-132 should be slightly altered or that the absorption cross-section for Xe-131 should be slightly increased. More analysis is necessary to determine which of these two alternatives is more appropriate. This work has shown that the accuracy of HELIOS (within 2% for heavy metals and within 4% for fission noble gases) is sufficient for most analyses.
Date: October 21, 1998
Creator: Charlton, W.S.; Perry, R.T.; Fearey, B.L. & Parish, T.A.
Partner: UNT Libraries Government Documents Department

Synergistic failure of BWR internals

Description: Boiling Water Reactor (BWR) core shrouds and other reactor internals important to safety are experiencing intergranular stress corrosion cracking (IGSCC). The United States Nuclear Regulatory Commission has followed the problem, and as part of its investigations, contracted with the Idaho National Engineering and Environmental Laboratory to conduct a risk assessment. The overall project objective is to assess the potential consequences and risks associated with the failure of IGSCC-susceptible BWR vessel internals, with specific consideration given to potential cascading and common mode effects. An initial phase has been completed in which background material was gathered and evaluated, and potential accident sequences were identified. A second phase is underway to perform a simplified, quantitative probabilistic risk assessment on a representative high-power BWR/4. Results of the initial study conducted on the jet pumps show that any cascading failures would not result in a significant increase in the core damage frequency. The methodology is currently being extended to other major reactor internals components.
Date: October 25, 1999
Creator: Ware, A. G. & Chang, T. Y.
Partner: UNT Libraries Government Documents Department

The corrosion behavior of hafnium in high-temperature-water environments

Description: The high-temperature-water corrosion performance of hafnium is evaluated. Corrosion kinetic data are used to develop correlations that are a function of time and temperature. The evaluation is based on corrosion tests conducted in out-of-pile autoclaves and in out-of-flux locations of the Advanced Test Reactor (ATR) at temperatures ranging from 288 to 360 C. Similar to the corrosion behavior of unalloyed zirconium, the high-temperature-water corrosion response of hafnium exhibits three corrosion regimes: pretransition, posttransition, and spalling. In the pretransition regime, cubic corrosion kinetics are exhibited, whereas in the posttransition regime, linear corrosion kinetics are exhibited. Because of the scatter in the spalling regime data, it is not reasonable to use a best fit of the data to describe spalling regime corrosion. Data also show that neutron irradiation does not alter the corrosion performance of hafnium. Finally, the data illustrate that the corrosion rate of hafnium is significantly less than that of Zircaloy-2 and Zircaloy-4.
Date: October 1, 1999
Creator: Rishel, D.M.; Smee, J.D. & Kammenzind, B.F.
Partner: UNT Libraries Government Documents Department

Estimating the Uncertainty in Reactivity Accident Neutronic Calculations

Description: A study of the uncertainty in calculations of the rod ejection accident in a pressurized water reactor is being carried out for the US Nuclear Regulatory Commission. This paper is a progress report on that study. Results are presented for the sensitivity of core energy deposition to the key parameters: ejected rod worth, delayed neutron fraction, Doppler reactivity coefficient, and fuel specific heat. These results can be used in the future to estimate the uncertainty in local fuel enthalpy given some assumptions about the uncertainty in the key parameters. This study is also concerned with the effect of the intra-assembly representation in calculations. The issue is the error that might be present if assembly-average power is calculated, and pin peaking factors from a static calculation are then used to determine local fuel enthalpy. This is being studied with the help of a collaborative effort with Russian and French analysts who are using codes with different intra-assembly representations. The US code being used is PARCS which calculates power on an assembly-average basis. The Russian code being used is BARS which calculates power for individual fuel pins using a heterogeneous representation based on a Green's Function method.
Date: October 26, 1998
Creator: Diamond, D. J.; Yang, C. Y. & Aronson, A. L.
Partner: UNT Libraries Government Documents Department

Closed ThUOX Fuel Cycle for LWRs with ADTT (ATW) Backend for the 21st Century

Description: A future nuclear energy scenario with a closed, thorium-uranium-oxide (ThUOX) fuel cycle and new light water reactors (TULWRs) supported by Accelerator Transmutation of Waste (ATW) systems could provide several improvements beyond today's once-through, UO{sub 2}-fueled nuclear technology. A deployment scenario with TULWRs plus ATWs to burn the actinides produced by these LWRs and to close the back-end of the ThUOX fuel cycle was modeled to satisfy a US demand that increases linearly from 80 GWe in 2020 to 200 GWe by 2100. During the first 20 years of the scenario (2000-2020), nuclear energy production in the US declines from today's 100 GWe to about 80 GWe, in accordance with forecasts of the US DOE's Energy Information Administration. No new nuclear systems are added during this declining nuclear energy period, and all existing LWRs are shut down by 2045. Beginning in 2020, ATWs that transmute the actinides from existing LWRs are deployed, along with TULWRs and additional ATWs with a support ratio of 1 ATW to 7 TULWRs to meet the energy demand scenario. A final mix of 174 GWe from TULWRs and 26 GWe from ATWs provides the 200 GWe demand in 2100. Compared to a once-through LWR scenario that meets the same energy demand, the TULWR/ATW concept could result in the following improvements: depletion of natural uranium resources would be reduced by 50%; inventories of Pu which may result in weapons proliferation will be reduced in quantity by more than 98% and in quality because of higher neutron emissions and 50 times the alpha-decay heating of weapons-grade plutonium; actinides (and possibly fission products) for final disposal in nuclear waste would be substantially reduced; and the cost of fuel and the fuel cycle may be 20-30% less than the once-through UO{sub 2} fuel cycle.
Date: October 6, 1998
Creator: Beller, D.E.; Sailor, W.C. & Venneri, F.
Partner: UNT Libraries Government Documents Department

Theory and application of deterministic multidimensional pointwise energy lattice physics method

Description: The theory and application of deterministic, multidimensional, pointwise energy lattice physics methods are discussed. These methods may be used to solve the neutron transport equation in multidimensional geometries using near-continuous energy detail to calculate equivalent few-group diffusion theory constants that rigorously account for spatial and spectral self-shielding effects. A dual energy resolution slowing down algorithm is described which reduces the computer memory and disk storage requirements for the slowing down calculation. Results are presented for a 2D BWR pin cell depletion benchmark problem.
Date: October 5, 1999
Creator: Zerkle, M.L.
Partner: UNT Libraries Government Documents Department

Economic Study of Spent Nuclear Fuel Storage and Reprocessing Practices in Russia

Description: This report describes a study of nuclear power economics in Russia. It addresses political and institutional background factors which constrain Russia's energy choices in the short and intermediate run. In the approach developed here, political and institutional factors might dominate short-term decisions, but the comparative costs of Russia's fuel-cycle options are likely to constrain her long-term energy strategy. To this end, the authors have also formulated a set of policy questions which should be addressed using a quantitative decision modeling which analyzes economic costs for all major components of different fuel cycle options, including the evolution of uranium prices.
Date: October 1, 1997
Creator: Singer, C. E. & Miley, G. H.
Partner: UNT Libraries Government Documents Department

Intercomparison of Results for a Pwr Rod Ejection Accident

Description: This study is part of an overall program to understand the uncertainty in best-estimate calculations of the local fuel enthalpy during the rod ejection accident. Local fuel enthalpy is used as the acceptance criterion for this design-basis event and can also be used to estimate fuel damage for the purpose of determining radiological consequences. The study used results from neutron kinetics models in PARCS, BARS, and CRONOS2, codes developed in the US, the Russian Federation, and France, respectively. Since BARS uses a heterogeneous representation of the fuel assembly as opposed to the homogeneous representations in PARCS and CRONOS, the effect of the intercomparison was primarily to compare different intra-assembly models. Quantitative comparisons for core power, reactivity, assembly fuel enthalpy and pin power were carried out. In general the agreement between methods was very good providing additional confidence in the codes and providing a starting point for a quantitative assessment of the uncertainty in calculated fuel enthalpy using best-estimate methods.
Date: October 1, 1999
Creator: Diamond, D. J.; Aronson, A.; Jo, J.; Avvakumov, A.; Malofeev, V.; Sidorov, V. et al.
Partner: UNT Libraries Government Documents Department

NRC Support for the Kalinin (Vver) Probabilistic Risk Assessment

Description: The US Nuclear Regulatory Commission (NRC) and the Federal Nuclear and Radiation Safety Authority of the Russian Federation have been working together since 1994 to carry out a probabilistic risk assessment (PRA) of a VVER-1000 in the Russian Federation. This was a recognition by both parties that this technology has had a profound effect on the discipline of nuclear reactor safety in the West and that the technology should be transferred to others so that it can be applied to Soviet-designed plants. The NRC provided funds from the Agency for International Development and technical support primarily through Brookhaven National Laboratory and its subcontractors. The latter support was carried out through workshops, by documenting the methodology to be used in a set of guides, and through periodic review of the technical activity. The result of this effort to date includes a set of procedure guides, a draft final report on the Level 1 PRA for internal events (excluding internal fires and floods), and progress reports on the fire, flood, and seismic analysis. It is the authors belief that the type of assistance provided by the NRC has been instrumental in assuring a quality product and transferring important technology for use by regulators and operators of Soviet-designed reactors. After a thorough review, the report will be finalized, lessons learned will be applied in the regulatory and operational regimes in the Russian Federation, and consideration will be given to supporting a containment analysis in order to complete a simplified Level 2 PRA.
Date: October 26, 1998
Creator: Bley, D.; Diamond, D. J.; Chu, T. L.; Azarm, A.; Pratt, W. T.; Johnson, D. et al.
Partner: UNT Libraries Government Documents Department

Laser ablation of concrete.

Description: Laser ablation is effective both as an analytical tool and as a means of removing surface coatings. The elemental composition of surfaces can be determined by either mass spectrometry or atomic emission spectroscopy of the atomized effluent. Paint can be removed from aircraft without damage to the underlying aluminum substrate, and environmentally damaged buildings and sculptures can be restored by ablating away deposited grime. A recent application of laser ablation is the removal of radioactive contaminants from the surface and near-surface regions of concrete. We present the results of ablation tests on concrete samples using a high power pulsed Nd:YAG laser with fiber optic beam delivery. The laser-surface interaction was studied on various model systems consisting of Type I Portland cement with varying amounts of either fine silica or sand in an effort to understand the effect of substrate composition on ablation rates and mechanisms. A sample of non-contaminated concrete from a nuclear power plant was also studied. In addition, cement and concrete samples were doped with non-radioactive isotopes of elements representative of cooling waterspills, such as cesium and strontium, and analyzed by laser-resorption mass spectrometry to determine the contamination pathways. These samples were also ablated at high power to determine the efficiency with which surface contaminants are removed and captured. The results show that the neat cement matrix melts and vaporizes when little or no sand or aggregate is present. Surface flows of liquid material are readily apparent on the ablated surface and the captured aerosol takes the form of glassy beads up to a few tens of microns in diameter. The presence of sand and aggregate particles causes the material to disaggregate on ablation, with intact particles on the millimeter size scale leaving the surface. Laser resorption mass spectrometric analysis showed that cesium and potassium have similar chemical ...
Date: October 5, 1998
Creator: Savina, M.
Partner: UNT Libraries Government Documents Department

A Radiological Survey Approach to Use Prior to Decommissioning: Results from a Technology Scanning and Assessment Project Focused on the Chornobyl NPP

Description: The primary objectives of this project are to learn how to plan and execute the Technology Scanning and Assessment (TSA) approach by conducting a project and to be able to provide the approach as a capability to the Chernobyl Nuclear Power Plant (ChNPP) and potentially elsewhere. A secondary objective is to learn specifics about decommissioning and in particular about radiological surveying to be performed prior to decommissioning to help ChNPP decision makers. TSA is a multi-faceted capability that monitors and analyzes scientific, technical, regulatory, and business factors and trends for decision makers and company leaders. It is a management tool where information is systematically gathered, analyzed, and used in business planning and decision making. It helps managers by organizing the flow of critical information and provides managers with information they can act upon. The focus of this TSA project is on radiological surveying with the target being ChNPP's Unit 1. This reactor was stopped on November 30, 1996. At this time, Ukraine failed to have a regulatory basis to provide guidelines for nuclear site decommissioning. This situation has not changed as of today. A number of documents have been prepared to become a basis for a combined study of the ChNPP Unit 1 from the engineering and radiological perspectives. The results of such a study are expected to be used when a detailed decommissioning plan is created.
Date: October 20, 1999
Creator: Milchikov, A.; Hund, G. & Davidko, M.
Partner: UNT Libraries Government Documents Department

K-14 shutdown reactivity

Description: SRS reactor charges are designed to ensure the reactor remains subcritical during chargeback and shutdown conditions. Calculations have recently been performed to determine the shutdown k{sub eff} for the K-14 charge. This document discusses the results and uncertainties.
Date: October 31, 1990
Creator: Chandler, J.R.
Partner: UNT Libraries Government Documents Department

Bounding burnout risk power limits for the K-14 cycle

Description: This document discusses burnout risk (BOR) power limits which are designed to protect the reactor from a significant release of fission products, due to critical heat flux (CHF) burnout of fuel and target assemblies. At expected operating power levels for the reactor restart, approximately 50% of historical full power, the risk of CHF and attendant burnout is negligible. Flow instability power limits will restrict reactor operation, and flow instability will always occur before CHF. BOR power limits must nevertheless be calculated because they are required by the reactor control computer, (2) Bounding BOR limits have been calculated for the K-14 cycle, to fulfill this requirement, and they are presented in this document. Two sets of BOR limits have been calculated: one applicable for the first subcycle, zero to 30% fuel burnup, and the other for the second subcycle, 30% to 55% fuel burnup.
Date: October 1, 1990
Creator: Shadday, M.A. Jr.
Partner: UNT Libraries Government Documents Department

A summary of high-temperature electronics research and development

Description: Current and future needs in automative, aircraft, space, military, and well logging industries require operation of electronics at higher temperatures than today's accepted limit of 395 K. Without the availability of high-temperature electronics, many systems must operate under derated conditions or must accept severe mass penalties required by coolant systems to maintain electronic temperatures below critical levels. This paper presents ongoing research and development in the electronics community to bring high-temperature electronics to commercial realization. Much of this work was recently reviewed at the First International High-Temperature Electronics Conference held 16--20 June 1991 in Albuquerque, New Mexico. 4 refs., 1 tab.
Date: October 18, 1991
Creator: Thome, F. V. & King, D. B.
Partner: UNT Libraries Government Documents Department