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Plutonium Recycle Test Reactor Final Safeguard Analysis: Supplement 2, Consequences of a Primary Coolant Leak

Description: Review of previous work is presented in addition to the results of recent studies concerning loss of primary coolant when the system is cold and pressurized and the problem of supplying adequate cooling following the injection of light water.
Date: November 15, 1960
Creator: Wittenbrock, N. G. & Muraoka, J.
Partner: UNT Libraries Government Documents Department

Estimating Loss-of-Coolant Accident Frequencies for the Standardized Plant Analysis Risk Models

Description: The U.S. Nuclear Regulatory Commission maintains a set of risk models covering the U.S. commercial nuclear power plants. These standardized plant analysis risk (SPAR) models include several loss-of-coolant accident (LOCA) initiating events such as small (SLOCA), medium (MLOCA), and large (LLOCA). All of these events involve a loss of coolant inventory from the reactor coolant system. In order to maintain a level of consistency across these models, initiating event frequencies generally are based on plant-type average performance, where the plant types are boiling water reactors and pressurized water reactors. For certain risk analyses, these plant-type initiating event frequencies may be replaced by plant-specific estimates. Frequencies for SPAR LOCA initiating events previously were based on results presented in NUREG/CR-5750, but the newest models use results documented in NUREG/CR-6928. The estimates in NUREG/CR-6928 are based on historical data from the initiating events database for pressurized water reactor SLOCA or an interpretation of results presented in the draft version of NUREG-1829. The information in NUREG-1829 can be used several ways, resulting in different estimates for the various LOCA frequencies. Various ways NUREG-1829 information can be used to estimate LOCA frequencies were investigated and this paper presents two methods for the SPAR model standard inputs, which differ from the method used in NUREG/CR-6928. In addition, results obtained from NUREG-1829 are compared with actual operating experience as contained in the initiating events database.
Date: September 1, 2008
Creator: Eide, S. A.; Rasmuson, D. M. & Atwood, C. L.
Partner: UNT Libraries Government Documents Department

APT Blanket System Loss-of-Coolant Accident (LOCA) Based on Initial Conceptual Design - Case 1: External HR Break Near Inlet Header

Description: The APT blanket system has about 57 MW of thermal energy deposited within the blanket region under normal operating conditions from the release of neutrons and the interaction of the High energy particles with the blanket materials. This corresponds to about 48 percent of total thermal energy deposited in the APT target/blanket system. The deposited thermal energy under normal operation conditions is an important input parameter used in the thermal-hydraulic design and accident analysis.
Date: October 7, 1998
Creator: Hamm, L.L.
Partner: UNT Libraries Government Documents Department

Blanket Module Boil-Off Times during a Loss-of-Coolant Accident - Case 0: with Beam Shutdown only

Description: This report is one of a series of reports that document LBLOCA analyses for the Accelerator Production of Tritium primary blanket Heat Removal system. This report documents the analysis results of a LBLOCA where the accelerator beam is shut off without primary pump trips and neither the RHR nor the cavity flood systems operation.
Date: October 7, 1998
Creator: Hamm, L.L.
Partner: UNT Libraries Government Documents Department

APT Blanket System Loss-of-Coolant Accident (LOCA) Analysis Based on Initial Conceptual Design - Case 3: External HR Break at Pump Outlet without Pump Trip

Description: This report is one of a series of reports that document normal operation and accident simulations for the Accelerator Production of Tritium (APT) blanket heat removal (HR) system. These simulations were performed for the Preliminary Safety Analysis Report.
Date: October 7, 1998
Creator: Hamm, L.L.
Partner: UNT Libraries Government Documents Department

APT Blanket System Internally Dry Flooded Cavity Accident Based on Initial Plate-Type Design

Description: Typically, heat conduction alone is insufficient for cooling components under decay heat conditions. However, due to various design features associated with the blanket modules heat conduction alone can transfer all of the deposited energy when under flooded cavity conditions.
Date: October 7, 1998
Creator: Hamm, L.L.
Partner: UNT Libraries Government Documents Department

APT Blanket System Loss-of-Coolant Accident Based on Initial Conceptual Design - Case 5: External RHR Break Near Inlet Header

Description: This report is one of a series of reports that document normal operation and accident simulations for the Accelerator Production of Tritium (APT) blanket heat removal system. These simulations were performed for the Preliminary Safety Analysis Report.
Date: October 7, 1998
Creator: Hamm, L.L.
Partner: UNT Libraries Government Documents Department

Analysis of an AP600 intermediate-size loss-of-coolant accident

Description: A postulated double-ended guillotine break of an AP600 direct-vessel-injection line has been analyzed. This event is characterized as an intermediate-break loss-of-coolant accident. Most of the insights regarding the response of the AP600 safety systems to the postulated accident are derived from calculations performed with the TRAC-PF1/MOD2 code. However, complementary insights derived from a scaled experiment conducted in the ROSA facility, as well as insights based upon calculations by other codes, are also presented. Based upon the calculated and experimental results, the AP600 will not experience a core heat up and will reach a safe shutdown state using only safety-class equipment. Only the early part of the long-term cooling period initiated by In-containment Refueling Water Storage Tank injection was evaluated. Thus, the observation that the core is continuously cooled should be verified for the later phase of the long-term cooling period when sump injection and containment cooling processes are important.
Date: April 1, 1995
Creator: Boyack, B.E. & Lime, J.F.
Partner: UNT Libraries Government Documents Department

APT Blanket Thermal Analysis of Cavity Flood Cooling with a Beam Window Break

Description: The cavity flood system is designed to be the primary safeguard for the integrity of the blanket modules and target assemblies during loss of coolant accidents, LOCA''s. In the unlikely event that the internal flow passages in a blanket module or a target assembly dryout, decay heat in the metal structures will be dissipated to the cavity flood system through the module or assembly walls. This study supplements the two previous studies by demonstrating that the cavity flood system can adequately cool the blanket modules when the cavity vessel beam window breaks.
Date: November 19, 1999
Creator: Shadday, M.A.
Partner: UNT Libraries Government Documents Department

Mechanical property testing of irradiated zircaloy cladding under reactor transient conditions.

Description: Specimen geometries have been developed to determine the mechanical properties of irradiated Zircaloy cladding subjected to the mechanical conditions and temperatures associated with reactivity-initiated accidents (RIA) and loss-of-coolant accidents (LOCA). Miniature ring-stretch specimens were designed to induce both uniaxial and plane-strain states of stress in the transverse (hoop) direction of the cladding. Also, longitudinal tube specimens were also designed to determine the constitutive properties in the axial direction. Finite-element analysis (FEA) and experimental parameters and results were closely coupled to optimize an accurate determination of the stress-strain response and to induce fracture behavior representative of accident conditions. To determine the constitutive properties, a procedure was utilized to transform measured values of load and displacement to a stress-strain response under complex loading states. Additionally, methods have been developed to measure true plastic strains in the gauge section and the initiation of failure using real-time data analysis software. Strain rates and heating conditions have been selected based on their relevance to the mechanical response and temperatures of the cladding during the accidents.
Date: October 2, 2001
Creator: Daum, R. S.; Majumdar, S.; Tsai, H.; Bray, T. S.; Billone, M. C.; Koss, D. A. et al.
Partner: UNT Libraries Government Documents Department

APT Blanket System Loss-of-Coolant Analysis Based on Initial Conceptual Design - Case 2: External HR Break HR Break at Pump Outlet with Pump Trip

Description: This report is one of a series of reports that document normal operation and accident simulations for the Accelerator Production of Tritium (APT) blanket heat removal system. These simulations were performed for the Preliminary Safety Analysis Report.
Date: October 7, 1998
Creator: Hamm, L.L.
Partner: UNT Libraries Government Documents Department

APT Blanket System Loss-of-Coolant Accident (LOCA) Based on Initial Conceptual Design - Case 4: External Pressurizer Surge Line Break Near Inlet Header

Description: This report is one of a series of reports documenting accident scenario simulations for the Accelerator Production of Tritium (APT) blanket heat removal systems. The simulations were performed in support of the Preliminary Safety Analysis Report (PSAR) for the APT.
Date: October 7, 1998
Creator: Hamm, L.L.
Partner: UNT Libraries Government Documents Department

Buoyancy-driven flow excursions in fuel assemblies

Description: A power limit criterion was developed for a postulated Loss of Pumping Accident (LOPA) in one of the recently shut down heavy water production reactors at the Savannah River Site. These reactors were cooled by recirculating moderator downward through channels in cylindrical fuel tubes. Powers were limited to prevent a flow excursion from occurring in one or more of these parallel channels. During full-power operation, limits prevented a boiling flow excursion from taking place. At low flow rates, during the addition of emergency cooling water, buoyant forces reverse the flow in one of the coolant channels before boiling occurs. As power increases beyond the point of flow reversal, the maximum wall temperature approaches the fluid saturation temperature, and a thermal excursion occurs. The power limit criterion for low flow rates was the onset of flow reversal. To determine conditions for flow reversal, tests were performed in a mock-up of a fuel assembly that contained two electrically heated concentric tubes surrounded by three flow channels. These tests were modeled using a finite difference thermal-hydraulic code. According to code calculations, flow reversed in the outer flow channel before the maximum wall temperature reached the local fluid saturation temperature. Thermal excursions occurred when the maximum wall temperature approximately equaled the saturation temperature. For a postulated LOPA, the flow reversal criterion for emergency cooling water addition was more limiting than the boiling excursion criterion for full power operation. This criterion limited powers to 37% of historical levels.
Date: December 31, 1995
Creator: Laurinat, J.E.; Paul, P.K. & Menna, J.D.
Partner: UNT Libraries Government Documents Department

Degradation and failure characteristics of NPP containment protective coating systems

Description: A research program to investigate the performance and potential for failure of Service Level 1 coating systems used in nuclear power plant containment is in progress. The research activities are aligned to address phenomena important to cause failure as identified by the industry coatings expert panel. The period of interest for performance covers the time from application of the coating through 40 years of service, followed by a medium-to-large break loss-of-coolant accident scenario, which is a design basis accident (DBA) scenario. The interactive program elements are discussed in this report and the application of these elements to the System 5 coating system (polyamide epoxy primer, carbon steel substrate) is used to evaluate performance.
Date: March 30, 2000
Creator: Sindelar, R.L.
Partner: UNT Libraries Government Documents Department

Matching grant program for university nuclear engineering education

Description: The grant augmented funds from Westinghouse Electric Co. to enhance the Nuclear Engineering program at KSU. The program was designed to provide educational opportunities and to train engineers for careers in the nuclear industry. It provided funding and access to Westinghouse proprietary design codes for graduate and undergraduate studies on topics of current industrial importance. Students had the opportunity to use some of the most advanced nuclear design tools in the industry and to work on actual design problems. The WCOBRA/TRAC code was used to simulate loss of coolant accidents (LOCAs).
Date: October 16, 2002
Creator: Bajorek, Stephen M.
Partner: UNT Libraries Government Documents Department

A Thermal Conduction Analysis of Proposed Lateral and Downstream Row 1 Module Plate Designs

Description: The cavity flood system is designed to be the primary safeguard for the integrity of the blanket modules during loss of coolant accidents (LOCAs). In the unlikely event that the internal flow passages in a module dryout, decay heat in the metal structures will be dissipated to the cavity flood system through the module walls. There is a design proposal under consideration to utilize a single 7.5 inch plate design in the APT blanket lateral modules. Currently, the lateral and downstream row 1 modules are designed with 4.0 inch plates, and longer plates are utilized in the lower power outer row modules. There are many benefits (e.g., fabrication costs, assemblage error reductions) to having a single plate design. The purpose of this study is to determine if the longer plate design meets the safety criterion that, when a module is internally dry, it can be adequately cooled by the cavity flood system.
Date: September 20, 1999
Creator: Shadday, M.A.
Partner: UNT Libraries Government Documents Department