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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.
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

An Evaluation of the PBF LOFT Lead Rod Test Results Concerning Surface Thermocouple Perturbation Effects

Description: The purpose of the Power Burst Facility Loss of Fluid Test (PBF LOFT) Lead Rod (LLR) Test program was to provide experimental data to characterize the mechanical behavior of LOFT type nuclear fuel rods under loss of coolant accident (LOCA) conditions, simulating the test conditions expected for the LOFT Power Ascension (L2) Test series. Although the LLR tests were not explicitly designed to evaluate cladding surface thermocouple perturbation effects, comparison of the Linear Variable Differential Transformer (LVDT) data for rods instrumented with and without cladding thermocouples provided pertinent information concerning the effects of cladding thermocouples on the time to DNB and time to quench data. Documentation and review of this data is presented in the following report. It will be shown that most of the LLR data indicate that the cladding surface thermocouples did not enhance the rewetting characteristics of the rods they are attached to, even though other evidence shows that the surface clad thermocouples did quench early. Finally, in order to accurately interpret and understand the limitations of the LVDT instrumentation, upon which thermocouple perturbation effects were evaluated, an analysis of the LVDT data as well as a review of the atypical response events that occurred during the LLR tests are presented in appendices to this document.
Date: February 8, 1980
Creator: Carboneau, M. L. & Tolman, E. L.
Partner: UNT Libraries Government Documents Department

Analysis of Postulated Core Meltdown of an SRP Reactor - Final Report

Description: An analysis was made to determine the consequences of a postulated accident in which the core of a Savannah River Plant reactor melts down following the loss of coolant. The study was made to determine (1) the potential damage to the reactor building that could impair its integrity for confining activity and (2) the need for additional facilities to prevent the activity confinement system from being overheated by the decay heat in the debris. A preliminary report on this analysis was issued previously. The sequence of events during and following the loss of coolant has now been studied in more detail, and a computer program has been written and used to investigate transient heating effects. This is the final report of the analysis and presents the conclusions.
Date: October 1, 1970
Creator: Durant, W.S. & Brown, R.J.
Partner: UNT Libraries Government Documents Department

Preliminary phenomena identification and ranking tables for simplified boiling water reactor Loss-of-Coolant Accident scenarios

Description: For three potential Loss-of-Coolant Accident (LOCA) scenarios in the General Electric Simplified Boiling Water Reactors (SBWR) a set of Phenomena Identification and Ranking Tables (PIRT) is presented. The selected LOCA scenarios are typical for the class of small and large breaks generally considered in Safety Analysis Reports. The method used to develop the PIRTs is described. Following is a discussion of the transient scenarios, the PIRTs are presented and discussed in detailed and in summarized form. A procedure for future validation of the PIRTs, to enhance their value, is outlined. 26 refs., 25 figs., 44 tabs.
Date: April 1998
Creator: Kroeger, P. G.; Rohatgi, U. S.; Jo, J. H. & Slovik, G. C.
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