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Normal Operation (NO) of APT Blanket System and its Components Based on Initial Conceptual Design

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

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 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

APT Blanket Detailed Bin Model Based on Initial Plate-Type Design -3D FLOWTRAN-TF Model

Description: This report provides background information for 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 for the APT. This report gives a brief description of the FLOWTRAN-TF code which was used for detailed blanket bin modeling.
Date: October 7, 1998
Creator: Hamm, L.L.
Partner: UNT Libraries Government Documents Department

APT Blanket System Loss-of-Flow Accident (LOFA) Analysis Based on Initial Conceptual Design - Case 1: with Beam Shutdown and Active RHR

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 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 2: with Beam Shutdown Only

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. This report documents the results of simulations of a Loss-of-Flow Accident (LOFA) where power is lost to all of the pumps that circulate water in the blanket region, the accelerator beam is shut off and neither the residual heat removal nor cavity flood systems operate.
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

ACHIEVING THE REQUIRED COOLANT FLOW DISTRIBUTION FOR THE ACCELERATOR PRODUCTION OF TRITIUM (APT) TUNGSTEN NEUTRON SOURCE

Description: The Accelerator Production of Tritium neutron source consists of clad tungsten targets, which are concentric cylinders with a center rod. These targets are arranged in a matrix of tubes, producing a large number of parallel coolant paths. The coolant flow required to meet thermal-hydraulic design criteria varies with location. This paper describes the work performed to ensure an adequate coolant flow for each target for normal operation and residual heat-removal conditions.
Date: November 1, 2000
Creator: SIEBE, D. & PASAMEHMETOGLU, K.
Partner: UNT Libraries Government Documents Department

An evaluation of the effects of valve body erosion on motor-operated valve operability

Description: INEL engineers evaluated effects of erosion-induced valve wall thinning on motor-operated valve operability. The authors reviewed reports that identified the extent and location of erosion damage in nuclear plant valves and chose a globe valve with severe erosion damage to assess the potential for loss of operability. They developed a finite element model of the selected valve and performed structural analyses with valve closing forces, seismic effects, and increased erosion areas to analyze effects of erosion on structural integrity. Results indicate that while some local stresses at the points of maximum erosion exceeded yield, the general stresses were well below yield. Therefore, displacements will be small and bending will not occur. It is concluded that erosion-related wall thinning is not likely to create an operability problem for motor-operated valves.
Date: December 1, 1995
Creator: Hunt, T. H.; Nitzel, M. E. & Weidenhamer, G. H.
Partner: UNT Libraries Government Documents Department

Natural Circulation in the Blanket Heat Removal System During a Loss-of-Pumping Accident (LOFA) Based on Initial Conceptual Design

Description: A transient natural convection model of the APT blanket primary heat removal (HR) system was developed to demonstrate that the blanket could be cooled for a sufficient period of time for long term cooling to be established following a loss-of-flow accident (LOFA). The particular case of interest in this report is a complete loss-of-pumping accident. For the accident scenario in which pumps are lost in both the target and blanket HR systems, natural convection provides effective cooling of the blanket for approximately 68 hours, and, if only the blanket HR systems are involved, natural convection is effective for approximately 210 hours. The heat sink for both of these accident scenarios is the assumed stagnant fluid and metal on the secondary sides of the heat exchangers.
Date: October 7, 1998
Creator: Hamm, L.L.
Partner: UNT Libraries Government Documents Department

An analysis of the proposed MITR-III core to establish thermal-hydraulic limits at 10 MW. Final report

Description: The 5 MW Massachusetts Institute of Technology Research Reactor (MITR-II) is expected to operate under a new license beginning in 1999. Among the options being considered is an upgrade in the heat removal system to allow operation at 10 MW. The purpose of this study is to predict the Limiting Safety System Settings and Safety Limits for the upgraded reactor (MITR-III). The MITR Multi-Channel Analysis Code was written to analyze the response of the MITR system to a series of anticipated transients in order to determine the Limiting Safety System Settings and Safety Limits under various operating conditions. The MIT Multi-Channel Analysis Code models the primary and secondary systems, with special emphasis placed on analyzing the thermal-hydraulic conditions in the core. The code models each MITR fuel element explicitly in order to predict the behavior of the system during flow instabilities. The results of the code are compared to experimental data from MITR-II and other sources. New definitions are suggested for the Limiting Safety System Settings and Safety Limits. MITR Limit Diagrams are included for three different heat removal system configurations. It is concluded that safe, year-round operating at 10 MW is possible, given that the primary and secondary flow rates are both increased by approximately 40%.
Date: June 1, 1997
Creator: Harling, O.K.; Lanning, D.D.; Bernard, J.A.; Meyer, J.E. & Henry, A.F.
Partner: UNT Libraries Government Documents Department

Nuclear Regulatory Commission issuances. Volume 47, Number 3

Description: This report includes the issuances received during the specified period from the Commission (CLI), the Atomic Safety and Licensing Boards (LBP), the Administrative Law Judges (ALJ), the Directors` Decisions (DD), and the Decisions on Petitions for Rulemaking (DPRM). The two issuances included here are: (1) the Commission issuance to the US Enrichment Corporation and (2) the Director`s Decision to the North Atlantic Energy Service Corporation.
Date: March 1, 1998
Partner: UNT Libraries Government Documents Department

Condensation enhancement on a pool surface caused by a submerged liquid jet

Description: One advanced nuclear reactor design has a residual heat removal (RHR) pipe connected to the bottom of a steam generator outlet plenum. The water in the plenum can become thermally stratified during postulated loss of coolant accidents. Cold water injected through the RHR pipe has the potential effect of increasing the steam condensation on the pool surface due to the stirring action of the jet. The amount of increase depends on a number of factors, including the jet velocity and the pool height above the jet injection point. Prediction of steam condensation rates, before and after the jet breaks the pool surface, is the topic of this paper. Data and correlations exist for pre surface breakthrough and a method has been developed for post breakthrough. The models have been incorporated into the reactor safety analysis computer software known as RELAP5. Comparisons of predictions against data are presented.
Date: May 1997
Creator: Shumway, R. W.
Partner: UNT Libraries Government Documents Department

TRAC analysis of design basis events for the accelerator production of tritium target/blanket

Description: A two-loop primary cooling system with a residual heat removal system was designed to mitigate the heat generated in the tungsten neutron source rods inside the rungs of the ladders and the shell of the rungs. The Transient Reactor Analysis Code (TRAC) was used to analyze the thermal-hydraulic behavior of the primary cooling system during a pump coastdown transient; a cold-leg, large-break loss-of-coolant accident (LBLOCA); a hot-leg LBLOCA; and a target downcomer LBLOCA. The TRAC analysis results showed that the heat generated in the tungsten neutron source rods can be mitigated by the primary cooling system for the pump coastdown transient and all the LBLOCAs except the target downcomer LBLOCA. For the target downcomer LBLOCA, a cavity flood system is required to fill the cavity with water at a level above the large fixed headers.
Date: August 1, 1997
Creator: Lin, J.C. & Elson, J.
Partner: UNT Libraries Government Documents Department

A study of thermal stratification in the cold legs during the subcooled blowdown phase of a loss of coolant accident in the OSU APEX thermal hydraulic testing facility.

Description: Thermal stratification, which has been linked to the occurrence of pressurized thermal shock (PTS), is observed to occur during the early stages of simulated loss of coolant accidents (LOCAS) in the Oregon State University Advanced Plant Experiment (OSU APEX) Thermal Hydraulic Test Facility. The OSU APEX Test Facility is a scaled model of the Westinghouse AP600 nuclear power plant. Analysis of the OSU APEX facility data has allowed the determination of an onset criteria for thermal stratification and has provided support for the postulated mechanisms leading to thermal stratification. CFX 4.1, a computational fluid dynamics code, was used to generate a model of the cold legs and the downcomer that described the phenomena occurring within them. Some mixing phenomena were predicted that lead to non-uniformity between the two cold legs attached to the steam generator on the side of the facility containing the Passive Residual Heat Removal (PRHR) injection system. The stratification was found to be two phase and unlikely to be a factor in PTS.
Date: November 4, 1998
Creator: Wachs, D. M.
Partner: UNT Libraries Government Documents Department

Decay heat removal by natural convection - the RVACS system.

Description: In conclusion, this work shows that for sodium coolant the reactor vessel auxiliary cooling system (RVACS) is an effective passive heat removal system if the reactor power does not exceed about 1600 MW(th). Its effectiveness is limited by the effective radiative heat transfer coefficient in the inner gap. In a lead cooled system, economic considerations may impose a lower limit.
Date: August 17, 1999
Creator: Tzanos, C. P.
Partner: UNT Libraries Government Documents Department

APT Blanket Thermal Analyses of Top Horizontal Row 1 Modules

Description: The Accelerator Production of Tritium (APT) cavity flood system (CFS) is designed to be the primary safeguard for the integrity of the blanket modules during loss of coolant accidents (LOCAs). For certain large break LOCAs the CFS also provides backup for the residual heat removal systems (RHRs) in cooling the target assemblies. In the unlikely event that the internal flow passages in a blanket module or target assembly dryout, decay heat in the metal structures will be dissipated to the CFS through the module or assembly walls (i.e., rung outer walls). The target assemblies consist of tungsten targets encased in steel conduits, and they can safely sustain high metal temperatures. Under internally dry conditions, the cavity flood fluid will cool the target assemblies with vigorous nucleate boiling on the external surfaces. However, the metal structures in the blanket modules consist of lead cladded in aluminum, and they have a long-term exposure temperature limit currently set to 150 degrees C. Simultaneous LOCAs in both the target and blanket heat removal systems (HRS) could result in dryout of the target ladders, as well as the horizontal blanket modules above the target. The cavity flood coolant would boil on the outside surfaces of the target ladder rungs, and the resultant steam could reduce the effectiveness of convection heat transfer from the blanket modules to the cavity flood coolant. A two-part analysis was conducted to ascertain if the cavity flood system can adequately cool the blanket modules above the targets, even when boiling is occurring on the outer surfaces of the target ladder rungs. The first part of the analysis was to model transient thermal conduction in the front top horizontal row 1 module (i.e. top horizontal modules nearest the incoming beam), while varying parametrically the convection heat transfer coefficient (htc) for the external ...
Date: September 20, 1999
Creator: Shadday, M.A.
Partner: UNT Libraries Government Documents Department

Heat transfer from bubbling pools. Progress report, July 1, 1975--October 1, 1975

Description: It is shown that the heat transfer characteristics of volume-heated boiling pools can be successfully modeled by non-boiling pools with internal gas injection. The strong influence of spatial distribution of bubble sites is relevant to estimates of boiling fuel attack on gas-releasing sacrificial materials. (auth)
Date: January 1, 1975
Creator: Bankoff, S G & Luk, A
Partner: UNT Libraries Government Documents Department

Guidelines for nuclear power plant safety issue prioritization information development. Supplement 5

Description: This is the sixth in a series of reports to document the development and use of a methodology developed by the Pacific Northwest Laboratory (PNL) to calculate, for prioritization purposes, the risk, dose, and cost impacts of implementing potential resolutions to reactor safety issues (see NUREG/CR-2800, Andrews, et al., 1983). This report contains the results of issue-specific analyses for 34 generic issues. Each issue was considered within the constraints of available information at the time the issues were examined and approximately 2 staff-weeks of labor. The results are referenced as one consideration in NUREG-0933, A Prioritization of Generic Safety Issues (Emrit, et al., 1983).
Date: July 1, 1996
Creator: Daling, P. M. & Lavender, J. C.
Partner: UNT Libraries Government Documents Department

Flow reversal power limit for the HFBR

Description: The High Flux Beam Reactor (HFBR) is a pressurized heavy water moderated and cooled research reactor that began operation at 40 MW. The reactor was subsequently upgraded to 60 MW and operated at that level for several years. The reactor undergoes a buoyancy-driven reversal of flow in the reactor core following certain postulated accidents. Questions which were raised about the afterheat removal capability during the flow reversal transition led to a reactor shutdown and subsequent resumption of operation at a reduced power of 30 MW. An experimental and analytical program to address these questions is described in this report. The experiments were single channel flow reversal tests under a range of conditions. The analytical phase involved simulations of the tests to benchmark the physical models and development of a criterion for dryout. The criterion is then used in simulations of reactor accidents to determine a safe operating power level. It is concluded that the limit on the HFBR operating power with respect to the issue of flow reversal is in excess of 60 MW. Direct use of the experimental results and an understanding of the governing phenomenology supports this conclusion.
Date: January 1, 1997
Creator: Cheng, L.Y. & Tichler, P.R.
Partner: UNT Libraries Government Documents Department