Meaningful qualification testing of nuclear reactor components requires a knowledge of the radiation fields expected in a loss-of-coolant accident (LOCA). The overall objective of this program is to define the LOCA source terms and compare these with the ...
National Technical Information Service (NTIS)
Deformation behavior of Zircaloy-4 cladding under simulated loss-of-coolant accident (LOCA) conditions is being investigated in the Multirod Burst Test (MRBT) program in single rod and multirod tests. In these tests, internally-pressurized unirradiated Zi...
Experiments were conducted to assess the effects of dielectric withstand voltage testing of cables and to assess the survivability of aged and damaged cables under loss-of-coolant accident (LOCA) conditions. High potential testing at 240 Vdc/mil on undama...
This paper describes the preliminary analysis of LOFA (loss of flow accident) and LOCA (loss of coolant accident) in the TIBER-II engineering test reactor breeding shield. TIBER-II is a compact reactor with a major radius of 3 m and thus requires a thin, ...
Loss-of-coolant accident (LOCA) tests performed in the Loss-of-Fluid Test (LOFT) Facility provide an important data base for verification of thermal-hydraulic computer models. An important parameter which must be known for LOCA model evaluation is the ini...
Double failures on the emergency-core-cooling systems (ECCSs) can be resulted in a case of loss-of-coolant accident (LOCA) of a boiling water reactor (BWR) by assuming an ECCS line break and the single failure criterion on another ECCS. In the Rig-of-Safe...
A series of tests has been performed in the RD-12 loop to study the behavior of a CANDU-type, primary heat transport system (PHTS) during the blowdown and injection phases of a loss-of-coolant accident (LOCA). Specifically, the tests were used to investig...
Loss-of-coolant is one of the most severe accidents for a nuclear power plant. To demonstrate inherent safety characteristics incorporated into small High-Temperature Gas Cooled Reactor (HTGR) designs, loss-of-coolant accident (LOCA) simulation tests have been conducted with the German pebble-bed High-Temperature ...
Energy Citations Database
Experiments were performed to assess the aging degradation and loss-of-coolant accident (LOCA) behavior of electrical cables subjected to long-term aging exposures. Four different cable types were tested in both the U.S. and France. (1) U.S. 2 conductor w...
Thermal testing of cadmium safety rods was conducted as part of a program to define the response of Savannah River Site (SRS) production reactor core components to a hypothetical loss-of-coolant accident (LOCA) leading to a drained reactor tank. The safety rods are present in the reactor core only during shutdown and are not used as a ...
This paper is the final report of the ROSA-II experimental program, in which summary of the integral test results on thermal hydraulic behavior in a loss-of-coolant accident (LOCA) of pressurized water reactor (PWR) and on the effect of emergency core coo...
In this paper the experimental response of ex-core neutron detectors during both actual and simulated loss-of-coolant accidents (LOCAs) at a pressurized water reactor are analyzed to determine their cause. Various analytical techniques are used to reproduce the ex-core detector response during large-break LOCAs. These techniques ...
The report presents experimental data and calculated steady-state and transient instrument uncertainties from the Small Break Loss of Coolant Accident (LOCA) Heat Transfer Test Series I. The subject test series was composed of six high-pressure, low-flow,...
This report presents the results of Loss-of-Coolant (LOC) Test LOC-11, the first test of the Loss-of-Coolant Accident (LOCA) Test Series conducted by the Thermal Fuels Behavior Program of EG and G Idaho, Inc., in the Power Burst Facility (PBF) at the Idah...
A common understanding and interpretation of BWR system response and the controlling phenomena in LOCA transients has been achieved through the evaluation and comparison of counterpart tests performed in the ROSA-III and FIST test facilities. These facili...
Thermal testing of cadmium safety rods was conducted as part of a program to define the response of Savannah River Site (SRS) production reactor core components to a hypothetical LOCA leading to a drained reactor tank. The safety rods are present in the reactor core only during shutdown and are not used as a control mechanism during operation; thus, their ...
DOE Information Bridge
The PBF--LOCA fuel behavior program is one of several test programs being conducted to provide experimental information on the behavior of nuclear reactor fuels under normal, off-normal and accident conditions in the Power Burst Facility at the Idaho National Engineering Laboratory. Specifically, the PBF--LOCA ...
The fuels behavior research in PBF is directed towards providing a detailed understanding of the response of nuclear fuel assemblies to off-normal and hypothetical accident conditions. Single fuel rods and clusters of highly instrumented fuel rods are installed within a central test space of the PBF core for testing. The core can be ...
A loss-of-coolant accident (LOCA) is one of the most severe accidents for a nuclear power plant. To demonstrate inherent safety characteristics incorporated into small high-temperature gas-cooled reactor (HTGR) design, LOCA simulation tests have been conducted at the Arbeitsgemeinschaft ...
The purpose of the LOFT fuel is to provide a pressurized water reactor core that has (1) test instrumentation for measurement of core conditions and (2) materials and geometric features to ensure heat transfer, hydraulic, mechanical, chemical, metallurgical and nuclear behaviors are typical of large pressurized water reactors (LPWRS) during the loss-of-coolant ...
Data from the Loss-of-Fluid Test (LOFT) Program help quantify the margin of safety inherent in pressurized water reactors during postulated loss-of-coolant accidents (LOCAs). The report analyzes how well externally-mounted fuel rod cladding surface thermo...
Hypothetical instantaneous pipe rupture is now considered to be one of the design basis accidents during the operation of the light water reactor. If a pipe rupture accidnet occurs, the pipe will start moving with the sudden discharge of internal fluid. S...
Thermal shock studies on test specimens fabricated from PWR pressure vessel material are described. The studies are aimed at understanding crack propagation and growth during ECCS quenching following a loss-of-coolant accident.
The concern raised by Generic Safety Issue (GSI-191), 'Assessment of Debris Accumulation on PWR Sump Performance,' is the transport of debris to pressurized-water-reactor (PWR) sump screens following a loss-of-coolant accident (LOCA) and subsequent impact...
Progress is reported in studies to delineate the deformation behavior of unirradiated Zircaloy cladding under conditions postulated for the LWR loss-of-coolant accident (LOCA) and to provide a data base to facilitate assessment of the magnitude and distri...
The report describes the results of aging, condition monitoring, and accident testing of miscellaneous cable types. Three sets of cables were aged for up to 9 months under simultaneous thermal and radiation conditions. A sequential accident consisting of ...
The report describes the results of aging, condition monitoring and accident testing of ethylene propylene rubber (EPR) cables. Three sets of cables were aged for up to 9 months under simultaneous thermal and radiation conditions. A sequential accident co...
The report describes the results of aging, condition monitoring, and accident testing of crosslinked polyolefin (XLPO) cables. Three sets of cables were aged for up to 9 months under simultaneous thermal and radiation conditions. A sequential accident con...
Meaningful qualification testing of nuclear reactor components requires a knowledge of the radiation fields expected in a loss-of-coolant accident (LOCA). The overall objective of this program is to define the LOCA source terms and compare these with the output of various simulators employed for radiation ...
The RELAP5YA computer code was developed to analyze postulated accidents and transients in light water reactor systems. The code has been assessed against many separate-effects and integral test results that address relevant thermal-hydraulic phenomena. The assessment results have established the validity of the code in predicting small- and large-break ...
The Transient Reactor Analysis Code (TRAC) is an advanced systems code for light-water-reactor accident analysis. The code was developed originally to analyze large-break loss-of-coolant accidents (LOCAs) and running time was not a primary development criterion. TRAC-PF1 was developed because increased application of the code to long ...
This report presents the results of an experimental program to determine the aging and loss-of-coolant accident (LOCA) behavior of electrical connections in order to obtain an initial scoping of their performance. Ten types of connections commonly used in nuclear power plants were tested. These included 3 types of conduit seals, 2 ...
This paper describes the preliminary analysis of LOFA (loss of flow accident) and LOCA (loss of coolant accident) in the TIBER-II engineering test reactor breeding shield. TIBER-II is a compact reactor with a major radius of 3 m and thus requires a thin, high efficiency shield on the inboard side. The use of ...
Loss-of-coolant is one of the most severe accidents for a nuclear power plant. To demonstrate inherent safety characteristics incorporated into small High-Temperature Gas Cooled Reactor (HTGR) designs, loss-of-coolant accident (LOCA) simulation tests have...
This report describes the results of an assessment of the CONTAIN and MELCOR computer codes for ITER LOCA and LOVA applications. As part of the assessment, the results of running a test problem that describes an ITER LOCA are presented. It is concluded that the MELCOR code should be the preferred code for ITER severe ...
Loss-of-coolant accident (LOCA) tests performed in the Loss-of-Fluid Test (LOFT) Facility provide an important data base for verification of thermal-hydraulic computer models. An important parameter which must be known for LOCA model evaluation is the initial fuel rod stored energy prior to ...
The analytical support in 1985 for Cylindrical Core Test Facility (CCTF), Slab Core Test Facility (SCTF), and Upper Plenum Test Facility (UPTF) tests involves the posttest analysis of 16 tests that have already been run in the CCTF and the SCTF and the pretest analysis of 3 ...
To date, there has been substantial analytical and experimental effort to define the margins between design basis loss-of-coolant accident (LOCA) behavior and regulatory limits on maximum fuel rod cladding temperature and deformation. As a result, there is extensive documentation on the modeling of fuel rod behavior in test reactors ...
The effect of fuel burnup on the embrittlement of various cladding alloys was examined with laboratory tests conducted under conditions relevant to loss-of-coolant accidents (LOCAs). The cladding materials tested were Zircaloy-4, Zircaloy-2, ZIRLO, M5, an...
Addendum C to the 30 Degree Sector Test Facility (30 SSTF) Experimental Task Plan defines the objectives, specific test conditions, and summarized test operating procedures for the transient loss of coolant accident (LOCA) simulation tests performed in th...
An investigation was performed to assess the effects of scale in a reduced-scale integral test facility designed to simulate the response of a commercial four-loop pressurized water reactor (PWR) during a hypothesized loss-of-coolant accident (LOCA). The facility considered in the investigation was the Loss-of-Fluid ...
Research has been conducted within the Component Assessment Program to evaluate the failure and degradation modes of unaged instruments exposed to environments exposed to environments within and beyond the design basis. Current qualification test requirements as they relate to electronic components in containment were also evaluated. This paper summarizes the salient findings ...
The RELAP4/MOD6 computer code was used to predict the thermal-hydraulic transient for Loss-of-Fluid Test (LOFT) Loss-of-Coolant Accident (LOCA) experiments L2-2, L2-3, and L2-4. This analysis will aid in the development and assessment of analytical models used to analyze the LOCA performance of commercial power ...
The general context of the work reported is a postulated Loss-of-Coolant Accident (LOCA) in a Pressurized Water Reactor (PWR), although many of the basic processes being studied may also apply to Boiling Water Reactors (BWRs). The program is a continuing effort to develop analytical and empricial tools which will contribute to best-estimate and licensing ...
The Rig of Safety Assessment (ROSA) III facility is a volumetrically scaled (1/424) boiling water reactor (BWR) system with an electrically heated core designed for integral loss-of-coolant accident (LOCA) and emergency-core-cooling-system (ECCS) tests. Experimental results obtained so far confirm that the severest single failure ...
This paper presents a review of safety studies for accidental sequences in the European solid breeder test blanket module (TBM) system. These studies are the starting point for the Preliminary Safety Analysis Report of ITER, under preparation to get the construction permit first and then later the operation licence. In general the reduced inventory of activation products and ...
NASA Astrophysics Data System (ADS)
The effects on water properties of boric acid and the variation in boric acid concentration in the reactor vessel during a postulated loss-of-coolant accident (LOCA) transient were determined. The concentration variation during a LOCA resulted in small wa...
This paper reports on the RELAP5 computer code used to simulate four small-scale loss-of-coolant accident (LOCA) experiments. The purpose of the study is to help assess RELAP5 under conditions similar to those expected during a large-break LOCA at an Advanced Test Reactor (ATR). During an ATR large-break ...
Experience from irradiation in PWRs has confirmed the M5{sup R} possesses all the properties required for upgraded operation including new fuel management approaches and high duty reactor operation. In this paper accident behavior is demonstrated through a comparison of M5{sup R} and Zircaloy-4 cladding behavior under RIA (Reactivity Insertion Accident) ...
This paper reports on an unheated, integral thermal-hydraulic facility scaled to the Advanced Test Reactor (ATR) designed, constructed, and operated to gather simulated large-break loss-of-coolant accident (LOCA) data for use in assessing codes used in ATR analysis. Eighteen experiments were performed in the facility to establish a ...
The thermal stored energy in a fuel rod is the driving function for the severest postulated nuclear energy-related accident, the loss-of-coolant-accident (LOCA). Because of this, the final acceptance criteria for emergency core cooling systems requires ca...
This report presents the work performed to verify the CEFLASH digital computer code modeling of the hydro-dynamic loads in a steam generator tube during a loss-of-coolant accident (LOCA). The test loop simulated the primary side thermal-hydraulic conditio...
A loss-of-coolant accident (LOCA) simulation program is evaluating the thermal-hydraulic and mechanical effects on pressurized water reactor (PWR) test fuel bundles. This Experiment Operation Plan (EOP) Addendum 2, together with the referenced EOP, descri...
In light water reactors, particularly the pressurized water reactor (PWR), the severity of a loss-of-coolant accident (LOCA) would limit how high the reactor power can operate. Although the best-estimate LOCA licensing methodology can provide the greatest margin on the peak cladding temperature (PCT) evaluation during a ...
Conservative assumptions about the effectiveness of emergency core cooling during a loss-of-coolant accident have restricted operating limits for many nuclear plants. The behavior of reactors is now better understood for a variety of transient conditions, and safety margins can be calculated more accurately. Ten years of research and testing have produced ...
The objective of the LOFT fuel design and fabrication effort was to provide a pressurized water reactor core that has (1) materials and geometric features to ensure that heat transfer, hydraulic, mechanical, chemical, metallurgical and nuclear behaviors are typical of large pressurized water reactors (PWR) during the loss-of-coolant accident (LOCA) ...
This report describes the water-flow-test of 64-rod PWR fuel assembly simulation which was tested under loss-of-coolant-accident (LOCA) conditions. The test, involving cladding deformation and rupture in the temperature region of the Zircaloy alpha phase,...
Results of the ROSA-II tests simulating a loss-of-coolant accident (LOCA) and effects of an emergency core cooling system (ECCS) in a pressurized water reactor (PWR) are presented including the test conditions and interpretations of the data in test runs ...
Fuel rod cladding surface temperatures have been estimated in Loss-of-Fluid Test (LOFT) Facility and in Power Burst Facility loss-of-coolant accident (LOCA) tests using data obtained with thermocouples welded to the cladding outer surface. These cladding ...
Transient Loss of Coolant Accident (LOCA) experiments conducted in the Steam Sector Test Facility have addressed multidimensional system refill-reflood phenomena. Tests conducted over a pressure domain from 150 psia to ambient have investigated Emergency ...
The thermo-hydraulic behavior of the coolant in the primary system of a nuclear reactor is important in the core heat transfer analysis during a hypothetical loss-of-coolant accident (LOCA). The heat transfer correlations are strongly dependent on local thermo-hydraulic conditions of the coolant. The present work allows to calculate such ...
This is a report on one of a series of experiments to simulates a loss-of-coolant accident (LOCA) using full-length fuel rods for pressurized water reactors (PWR). The experiments were conducted by Pacific Northwest Laboratory (PNL) under the LOCA simulation Program sponsored by the US Nuclear Regulatory Commission (NRC). The major ...
The work presented in this paper was performed by Brookhaven National Laboratory (BNL) in support of Nuclear Regulatory Commission's (NRC) effort towards the resolution of Generic Issue 105 ''Interfacing System Loss of Coolant Accidents (LOCAs) at Boiling Water Reactors (BWRs).'' For BWRs, intersystem ...
The accumulator in a Pressurized Water Reactor (PWR) is generally pressurized with inert nitrogen cover gas, and the accumulator water will be saturated with nitrogen. Nitrogen released due to system depressurization during a Loss-of-Coolant Accident (LOCA) transient, consists of the nitrogen that is in the gas phase as well as nitrogen coming out of the ...
Deformation behavior of Zircaloy-4 cladding under simulated loss-of-coolant accident (LOCA) conditions is being investigated in the Multirod Burst Test (MRBT) program in single rod and multirod tests. In these tests, internally-pressurized unirradiated Zircaloy-4 tubes containing internal ...
The ROSA-IV Program's large scale test facility (LSTF) is a test facility for integral simulation of thermal-hydraulic response of a pressurized water reactor (PWR) during a small break loss-of-coolant accident (LOCA) or an operational transient. This doc...
Multirod burst test No. 7807 was performed with a view to estimating the quantity of coolant flow channel restriction caused by the ballooning of claddings in a fuel assembly during a postulated LOCA. The test was conducted on a condition that the initial...
The Subprogram for Cooperation in the AVR Test Program is being carried out within the United States/FRG High-Temperature Reactor Umbrella Agreement. The AVR Test Program is investigating performance and safety features pertinent to modular gas-cooled rea...
The High Burnup Cladding Performance program is being conducted at ANL to provide data in support of Loss-of-Coolant Accident (LOCA) and Reactivity-Initiated Accident (RIA) licensing criteria assessments for fuels at high burnup, as well as licensing crit...
For Zion, the following accident sequences were examined: station blackout with failure of auxiliary feedwater (TMLB'), loss of collant accidents (LOCA's) of various sizes inside containment, and the V-sequence interacing system LOCA which involves direct...
Excessive oxidation, hydriding, and extensive irradiation damage occur in high-burnup fuel cladding, and as result, mechanical properties of high-burnup fuels are degraded significantly. This may influence the current fuel cladding failure limits for loss-of- coolant-accident (LOCA) situations, which are based on fuel cladding behavior for zero burnup. To ...
This analysis suggests that the most cost-effective method to reduce the risk due to Interfacing System Loss of coolant accidents (ISLs) would be to establish a minimum testing frequency for pressure isolation valves. The suggested minimum frequency would be to perform leak testing of the pressure isolation valves at each refueling and ...
Within the scope of an order placed by the Federal Ministry for Research and Technology with the Battelle Institute e.V./Frankfurt, tests for the determination of thermalhydraulic processes during a loss-of-coolant accident (LOCA) were carried out in a te...
The Water Reactor Research Test Facilities Division has completed a series of experiments to investigate the effects of upper head injection during hypothetical small break loss-of-coolant accidents (LOCAs). Indications from the test series are that none ...
Capabilities of the RETRAN-02 MOD2 computer program for small-break loss-of-coolant accident (SBLOCA) analysis are evaluated through a posttest analysis of a simulated BWR SBLOCA experiment on General Electric Company's two-loop test apparatus (Test No. 6...
Experiment LP-02-6 was conducted on October 3, 1983. It was the first large-break loss-of-coolant accident (LOCA) simulation and the fourth experiment at all conducted in the Loss-Of-Fluid-Test (LOFT) facility at the Idaho National Engineering Laboratory ...
Oak Ridge Naional Laboratory has experimentally investigated heat transfer and high-pressure reflood under conditions similar to those expected in a small-break loss-of-coolant accident (SBLOCA). This report addresses the results of a series of six high-pressure reflood tests that were run in January of 1980.
The Westinghouse steam/water mixing program was initiated to obtain test data and to develop thermal/hydraulic models which will represent the steam/water interactions expected in a PWR cold leg during a postulated Loss-Of-Coolant Accident (LOCA). The Wes...
Historically, thermal hydraulics analyses on Large Break Loss of Coolant Accidents (LOCA) have been focused on the transients within the reactor or steam generator. Few have studied the effects of steam blowdown on the containment building. This paper discusses some theoretical issues as well as presenting numerical and experimental results of the blowdown ...
Simulated LOCA (loss of coolant accident) tests and subsequent mechanical tests on Zircaloy-4 cladding were carried out to evaluate the failure behavior of the cladding. Zircaloy-4 claddings were oxidized in a steam environment from 900 to 1250 �C for a given time period followed by a flooding of cool water to ...
This study is part of the preliminary safety analysis for the new power expansion project on the University of Missouri Research Reactor (MURR). The loss of coolant accident (LOCA), which is initiated by hypothetical pipe ruptures at the most adverse positions (V507 A B) in both the hot and cold legs of the primary coolant loop, is analyzed with the ...
The response of the pressure-suspension containment system of Mark I boiling-water reactors to a loss-of-coolant accident (LOCA) is being studied. This response is a design basis for light-water nuclear reactors. Part of the study is being carried out on a /sup 1///sub 5/-scale experimental facility that models the pressure-suppression containment system ...
A program was implemented as Savannah River site (SRS) to use the system code RELAP5 for K15.1 loss-of-coolant accident (LOCA) and loss-of-pumping accident (LOPA) reactor power limits calculations. The RELAP5 improvement program consolidated one system code and upgraded existing models. The existing SRS RELAP5 model was modified to ...
The objectives of fuel safety research program at Japan Atomic Energy Agency (JAEA) are; to evaluate adequacy of present safety criteria and safety margins; to provide a database for future regulation on higher burnup UO{sub 2} and MOX fuels, new cladding and pellets; and to provide reasonably mechanistic computer codes for regulatory application. The JAEA program is comprised of ...
This paper presents results of the first nuclear blowdown tests (LOC-11A, LOC-11B, LOC-11C) ever conducted. The Loss-of-Coolant Accident (LOCA) Test Series is being conducted in the Power Burst Facility (PBF) reactor at the Idaho National Engineering Laboratory, near Idaho Falls, Idaho, for the Nuclear Regulatory ...
The present study consists of the simulation of two loss of coolant accidents, LOCA 6' and LOCA 2', in one of the residual heat removal system (RHR) lines outside the containment, using the thermal-hydraulic code RELAP/MOD3.2. Both transients have been si...
The Loss of Coolant Accident (LOCA) Test Series being conducted in the Power Burst Facility (PBF) at the Idaho National Engineering Laboratory has been designed to provide data for the development and the assessment of fuel behavior computer codes used to predict the response of a pressurized light water reactor (PWR) during a ...
The RELAP5/MOD2.5 code (RELAP5) is used to perform best-estimate analyses of certain postulated Design Basis Accidents (DBAs) in SRS production reactors. Currently, the most limiting DBA in terms of reactor power level is an instantaneous double-ended guillotine break (DEGB) loss of coolant accident (LOCA). A six-loop RELAP5 K Reactor ...
A loss-of-coolant accident (LOCA) can cause a loss-of-offsite power (LOOP) wherein the LOOP is usually delayed by few seconds or longer. Such an accident is called LOCA with consequential LOOP, or LOCA with delayed LOOP (here, abbreviated as LOCA/LOOP). This paper analyzes ...
The mass flowrate and steam quality measuring of two phase flowrate is an essential issue in the tests of loss-of-coolant accident (LOCA). The spatial stochastic distribution of phase concentration would cause a differential pressure noise when two phase ...
A computer code (TRAC-PF1/MOD1) developed for predicting transient thermal and hydraulic integral nuclear steam supply system (NSSS) response was benchmarked. Post-small break loss-of-coolant accident (LOCA) data from a scaled, experimental facility, desi...
The Loss-of-Fluid Test (LOFT) Program is providing data to evaluate analytical models used to predict the thermal-hydraulic and fuel rod response of a pressurized water reactor (PWR) under loss-of-coolant accident (LOCA) conditions. The fuel rod response for the first nuclear loss-of-coolant experiment (LOCE), LOCE L2-2, is summarized.
This paper presents the accuracy and uncertainty of fuel rod behavior calculations performed by the transient Fuel Rod Analysis Program (FRAP-T6) during large break loss-of-coolant accidents. The accuracy of the code was determined primarily through compa...
RELAP5 MOD3.1.1 is being used to simulate Loss of Coolant Accidents (LOCA) for the Simplified Boiling Water Reactor (SBWR) being proposed by General Electric (GE). One of the major components associated with the SBWR is the Passive Containment Cooling Sys...
A series of four, large-break loss-of-coolant accident fuel behavior experiments have been performed in the Power Burst Facility (PBF) at the Idaho Engineering Laboratory. These experiments have been analyzed by using out-of-pile data to understand the ph...
The LOCA-ECC Thermal Shock Program was established to investigate the potential for flaw propagation in pressurized-water reactor (PWR) vessels during injection of emergency core coolant following a loss-of-coolant accident. Studies thus far have included fracture mechanics analyses of typical PWRs, the design and construction of a thermal shock ...
The Super SARA Test Program (SSTP) is a major European Economic Community effort to study light-water reactor safety during large- and small-break loss-of-coolant accident (LOCA) events in the ESSOR reactor. The SSTP will simulate small-break LOCA's by producing slow temperature ramps to high fuel rod ...
New initiatives were outlined in two areas. Suggested under Task 1 was an approach to provide data related to accident testing methodology for safety-related electrical cable. This approach has two main objectives: to provide experimental data and analyses of current acceptance criteria for loss-of-coolant accident ...
The postulated loss-of-coolant accident (LOCA) of a pressurized water reactor has been the subject of intensive experimental and analytical studies in light water reactor safety analysis. Many efforts are devoted to the investigation of the thermodynamic behavior of the reactor core and the effectiveness of the emergency-core-cooling system during reflood ...
Fission product release tests were performed with light water reactor (LWR) fuel rod segments containing large amounts of cesium and iodine in the pellet-to-cladding gap space in order to check the validity of the previously published Source Term Model for this type of fuel. The model describes the release of fission product cesium and iodine from LWR fuel rods for controlled ...
This report describes the results of an assessment of the CONTAIN and MELCOR computer codes for ITER LOCA and LOVA applications. As part of the assessment, the results of running a test problem that describes an ITER LOCA are presented. It is concluded th...
Following a hypothetical Loss of Coolant Accident (LOCA) the moderator level in the reactor tank would decrease. The current operating procedure with the new Type Q Septifoil is to maintain Septifoil cooling during a LOCA. With the Type Q Septifoil the co...
A computer code called TRAC is being developed by the Los Alamos Scientific Laboratory for analysis of loss-of-coolant accidents (LOCA's) and other transients in light water reactors. This code differs from existing codes and other codes under development...
LOFT is designed to monitor and survive Loss-Of-Coolant-Accidents (LOCAs). This report presents the primary design difference from LPWRs that were required to accomplish this. These design differences may be of interest to the nuclear power generator indu...
The report presents the analytical bases used to establish the adequacy of the long term cooling of the reactor following a Loss of Coolant Accident. The results of a typical post-LOCA long term analysis are also presented.
The objective of this work was to determine hydrogen concentration variations with position and time in a closed containment compartment with radiolytic hydrogen generation in the water on the compartment floor following a Loss-of-Coolant-Accident (LOCA)....
Ex-vessel Loss-of-Coolant-Accident (LOCA) simulations for the International Thermonuclear Experimental Reactor (ITER) were performed using the MELCOR code. The main goals of this work were to estimate the ultimate pressurization of the heat transport syst...
Performance of high-burnup fuel and fuel cladding fabricated from new types of alloys (such as Zirlo, M5, MDA, and duplex alloys) under loss-of-coolant-accident (LOCA) situations is not well understood at this time. To correctly interpret the results of i...
The process of emergency core cooling in a LOCA of a pressurized water reactor is summarized. The thermohydraulics in the reactor core and the loading of the fuel rod claddings during a LOCA are covered in more detail. Some recent experimental results on ...
The objective of this research was to develop and demonstrate a methodology for calculation of the time interval between receipt of the containment isolation signals and the first fuel pin failure for loss-of-coolant accidents (LOCAs). Demonstration calcu...
This paper describes the results of an analysis of loss of coolant accidents (LOCA`s) for the Soviet designed, light water cooled and moderated reactors referred to as VVERS. The VVER unit selected for this analysis is designated as VVER-440 Model 213. This plant generates 440 MWe and is of current interest since fifteen are now operating and additional ...
A computer code (TRAC-PF1/MOD1) developed for predicting transient thermal and hydraulic integral nuclear steam supply system (NSSS) response was benchmarked. Post-small break loss-of-coolant accident (LOCA) data from a scaled, experimental facility, designated the One-Through Integral System (OTIS), were obtained for the Babcock and Wilcox NSSS and ...
Experiments were conducted to assess the effects of high potential testing of cables and to assess the survivability of aged and damaged cables under Loss-of-Coolant Accident (LOCA) conditions. High potential testing at 240 Vdc/mil on undamaged cables suggested that no damage was incurred on the selected virgin ...
The purpose of this paper is to show the importance of considering thermal nonequilibrium effects in computer simulations of the refill and reflood portions of pressurized water reactor (PWR) loss-of-coolnat accident (LOCA) transients. Although RELAP4 assumes thermodynamic equilibrium between phases, models that account for the nonequilibrium phenomena ...
Downcomer fluid phenomena which affect the delivery of emergency core coolant (ECC) to the core area and influence the thermal-hydraulic behavior of a nuclear reactor during a loss-of-coolant accident (LOCA) are identified and discussed. The experimental results from the Loss-of-Fluid Test (LOFT) facility are compared with results from ...
This paper describes analyses and experiments that have been performed for a low-pressure heavy-water reactor design to help characterize shutdown heat removal through condensation heat transfer in the primary heat exchangers (PHXs) under loss-of-coolant accident (LOCA) conditions. This is one of many passive safety features that are dedicated to the ...
A landmark safety test has been conducted at the AVR-reactor, a high-temperature gas-cooled reactor (HTGR) in the Federal Republic of Germany owned by the Arbeitsgemeinschaft Versuchsreaktor, AVR in Juelich. The 46-MW(t), 15-MW(e) AVR reactor was subjected to a simulated loss-of-coolant accident (LOCA), a very severe occurrence in ...
The Loss-of-Fluid Test (LOFT) facility is a 50 MW(t), volumetrically scaled, pressurized water reactor (PWR) system. The LOFT facility was designed to study the engineered safety features (ESF) in commercial PWR systems as to their response to the postulated loss-of-coolant accident (LOCA). With recognition of the differences in ...
A series of experiments was conducted with highly irradiated light-water reactor fuel rod segments to investigate fission products released in steam in the temperature range 500 to 1200/sup 0/C. (Two additional release tests were conducted in dry air.) The primary objectives were to quantify and characterize fission product release under conditions postulated for a spent-fuel ...
The US Department of Energy's High Flux Isotope Reactor (HFIR) was analyzed to evaluate it's response to a spectrum of loss-of-coolant accidents (LOCAs) with potential for leading to core damage. The MELCOR severe accident analysis code (version 1.7.1) was used to evaluate the overall dynamic response of HFIR. Before ...
A Passive Safety Injection System (PSIS) and a Safety Injection System (SIS) with reference to a typical pressurized water reactor have been studied. The performance of the PSIS has been analyzed for a large break Loss of Coolant Accident (LOCA) in one of the cold leg of reactor coolant system. The SIS is a huge system consisting of many active components ...
Emergency Core Cooling Systems (ECCS) are required on all light water reactors (LWRs) in the United States to provide cooling of the reactor core in the event of a break in the reactor piping. These accidents are called loss-of-coolant accidents (LOCA), a...
This paper presents the results of TRAC-PF1/MOD3 benchmarks of the Rig FA experiments performed at the Savannah River Laboratory to simulate prototypic reactor fuel assembly behavior over a range of fluid conditions typical of the emergency cooling system (ECS) phase of a loss-of-coolant accident (LOCA). The primary purpose of this work was to use the SRL ...
The TRAC-P1 program was designed primarily for the analysis of large-break loss-of-coolant accidents (LOCAs) in pressurized water reactors (PWRs). Because of its versatility, however, it can be applied directly to many analyses ranging from blowdowns in simple pipes to integral LOCA tests in multiloop facilities. A ...
The coolant in the Savannah River Site (SRS) production nuclear reactor assemblies is circulated as a subcooled liquid under normal operating conditions. This coolant is evenly distributed throughout multiple annular flow channels with a uniform pressure profile across each coolant flow channel. During the postulated Loss of Coolant Accident (LOCA), which ...
This paper summarizes the results of an assessment of our TRAC-PF1/MOD3 Mark-22 prototype fuel assembly model against single-assembly data obtained from the A'' Tank single-assembly tests that were performed at the Savannah River Laboratory. We felt the data characterize prototypic assembly behavior over a range of air-water flow conditions of interest for ...
This report describes the results of aging, condition monitoring, and accident testing of miscellaneous cable types. Three sets of cables were aged for up to 9 months under simultaneous thermal ({approx_equal}100{degrees}C) and radiation ({approx_equal}0.10 kGy/hr) conditions. A sequential accident consisting of high dose rate ...
This report describes the results of aging, condition monitoring, and accident testing of miscellaneous cable types. Three sets of cables were aged for up to 9 months under simultaneous thermal ([approx equal]100[degrees]C) and radiation ([approx equal]0.10 kGy/hr) conditions. A sequential accident consisting of high dose rate ...
This report describes the results of tests conducted on three different designs of full-size electrical penetration assemblies (EPAs) that are used in the containment buildings of nuclear power plants. The objective of the tests was to evaluate the behavior of the EPAs under simulated severe accident conditions using steam at elevated ...
The current water reactor safety activities of ANC are accomplished in four programs. The Semiscale Program consists of small-scale nonnuclear thermal- hydraulic experiments for the generation of experimental data that can be applied to analytical models describing loss-of-coolant accident (LOCA) phenomena in water-cooled nuclear power plants. ...
This report assesses the effect on safety and cost of the requirements to combine loss-of-coolant-accident (LOCA) and safety-shutdown earthquakes (SSE) loads in the design of nuclear power plants. Analysis is limited mainly to plants recently completed or near completion, where current definitions of LOCA and SSE loading phenomena ...
Testing of the LOFT fuel rod cladding surface thermocouples has been performed to evaluate how accurately the LOFT thermocouples measure the cladding surface temperature during a loss-of-coolant accident (LOCA) sequence and what effect, if any, the thermocouple would have on core performance. Extensive testing has ...
The analysis of the Design Basis Loss of Coolant Acident (LOCA) for Savannah River Site (SRS) reactors involves the best estimate reactor system thermal-hydraulics code TRAC-PFI/MOD1. Power levels for the L-3.1 and P-10.2 subcycles were determined based, in part, on TRAC analyses of the first few seconds of a plenum inlet break LOCA. The TRAC code is ...
This report describes the results of aging, condition monitoring, and accident testing of ethylene propylene rubber (EPR) cables. Three sets of cables were aged for up to 9 months under simultaneous thermal ([approx equal]100[degrees]C) and radiation ([approx equal]0.10 kGy/hr) conditions. A sequential accident consisting of high dose ...
This report describes the results of aging, condition monitoring, and accident testing of ethylene propylene rubber (EPR) cables. Three sets of cables were aged for up to 9 months under simultaneous thermal ({approx_equal}100{degrees}C) and radiation ({approx_equal}0.10 kGy/hr) conditions. A sequential accident consisting of high dose ...
Experiments were performed to assess the aging degradation and loss-of-coolant accident (LOCA) behavior of electrical cables subjected to long-term aging exposures. Four different cable types were tested in both the U.S. and France: (1) U.S. 2 conductor with ethylene propylene rubber (EPR) insulation and a Hypalon jacket. (2) U.S. 3 ...
A second series of thermocouple effects tests (TC-3) is being conducted in the Power Burst Facility (PBF) at the Idaho National Engineering Laboratory (INEL). These tests are designed to evaluate the influence of external surface thermocouples on the behavior of nuclear fuel rods during a large break loss-of-coolant accident ...
This report presents the work performed to verify the CEFLASH digital computer code modeling of the hydro-dynamic loads in a steam generator tube during a loss-of-coolant accident (LOCA). The test loop simulated the primary side thermal-hydraulic conditions in an operational nuclear steam generator. The loop consisted of 5 full size ...
A small-scale nonnuclear integral test facility designed to simulate a pressurized water reactor (PWR) system was used to evaluate the effects of a small break loss-of-coolant accident (LOCA) on the system thermal-hydraulic response. The experiment approximated a 2.5% (11-cm diameter) communicative break in the cold leg of a PWR, and ...
The reactor safety rods may be subjected to high temperatures due to gamma heating after the core coolant level has dropped during the ECS phase of hypothetical LOCA event. Accordingly, an experimental safety rod testing subtask was established as part of a task to address the response of reactor core components to this accident. This ...
One of the primary objectives of the LOFT (Loss-of-Fluid Test) Program is to provide data required to evaluate and improve the analytical methods currently used to predict the LOCA (Loss-of-Coolant Accident) response of large pressurized water reactors. The purpose of the paper is to describe the computer modeling methods used in ...
Transient Loss of Coolant Accident (LOCA) experiments conducted in the Steam Sector Test Facility have addressed multidimensional system refill-reflood phenomena. Tests conducted over a pressure domain from 150 psia to ambient have investigated Emergency Core Cooling Systems (ECCS) combinations and injection ...
The PBF/LOFT Lead Rod (LLR) Test Program is being conducted to provide experimental information on the behavior of nuclear fuel under normal and accident conditions in the Power Burst Facility (PBF) at the Idaho National Engineering Laboratory. The PBF/LLR tests are designed to simulate the test conditions for the ...
When overaged from thermal or radiation environments, composite insulation composed of a layer of ethylene propylene rubber (EPR) covered with a bonded layer of chlorosulfonated polyethylene (CSPE[Hypalon]) can crack if subjected to steam environments associated with loss-of-coolant accidents (LOCAs). The work described in this report evaluated the effects ...
Instrumentation for making two-phase measurements in experimental refill-reflood test facilities was developed through the Advanced Instrumentation for Reflood Studies (AIRS) program. These unique instrumentation systems were designed to survive the severe in-vessel environmental conditions that exist during a simulated pressurized water reactor loss-of-coolant ...
Postulated long-term loss of coolant accidents (LOCA) for the International Thermonuclear Experimental Reactor (ITER) may involve the ingress of air or steam into the plasma chamber. Reactions of these gases with the hot plasma facing components will cause oxidation, transport, and release of activated species. To predict radioactivity releases, we ...
The effect of fuel burnup on the embrittlement of various cladding alloys was examined with laboratory tests conducted under conditions relevant to loss-of-coolant accidents (LOCAs). The cladding materials tested were Zircaloy-4, Zircaloy-2, ZIRLO, M5, and E110. Tests were performed with ...
This report discusses four-foot-long sections of a new production light-water reactor (NP-LWR) generic tritium target rod which were tested to determine if the length of the pellet pencils affects the amount of pellet material relocated during a burst and to characterize the burst. This testing was conducted as a follow-on study of cladding strength and ...
Analyses of gamma and beta dose in selected regions in PWR and BWR containment buildings have been performed for a range of fission product releases from selected severe accidents. The objective of this study was to determine the radiation dose that safety-related equipment could experience during the selected severe accident sequences. The resulting dose ...
Different hypothetical accident sequences have been simulated at Unit 1 of the Juragua nuclear power plant in Cuba, a plant with two VVER-440 V213 units under construction. The computer code MELCOR was employed for these simulations. The sequences simulated are: (1) a design-basis accident (DBA) large loss of coolant accident ...
The paper summarizes the dominant effects which finally ensure the core coolability of a pressurized water reactor (PWR) in a loss-of-coolant accident (LOCA). The results presented relate mainly to research work performed at Karlsruhe Nuclear Research Cen...
Loss-of-coolant accident (LOCA) evaluations for advanced pressurized water breeder reactors have shown that response during a LOCA differs from that of conventional pressurized water reactors due to the size of the primary plant and type of core geometry....
Assembly power limits have been established to prevent bulk boiling of the coolant in SRP reactor assemblies during a design basis loss-of-coolant accident (LOCA). This memorandum provides the methodology for calculating deposited power limits for the P-1...
Continuing research to develop and verify computer models of CANDU-PHW reactor process and safety systems is described. It is focussed on loss-of-coolant accidents (LOCAs) because they are the precursors of more serious accidents. Research topics include:...
Operator Action Event Trees for transient and LOCA initiated accident sequences at the Zion 1 PWR have been developed and documented. These trees logically and systematically portray the role of the operator throughout the progression of the accident. The...
Pressurized water reactor loss-of-coolant accident (LOCA) phenomena are being simulated with a series of experiments in the U-2 loop of the National Research Universal Reactor at Chalk River, Ontario, Canada. The first of these experiments includes up to ...
It is prerequisite for the establishment of nuclear power plant safety and for the maximization of operation efficiency to devlope the accident analysis computer code packages which can predict the results of postulated accidents and evaluate the performa...
An experimental investigation was conducted to characterize the benefits of warm prestress (WPS) in limiting crack extension in the wall of a nuclear vessel during a loss-of-coolant accident (LOCA) followed by introduction of relatively cold water by the ...
In this report the consequences of the LOCA of the Greek Research Reactor on the Greek AEC's personnel are analyzed under conservative assumptions. This accident with very low possibility of appearance has nevertheless no trivial consequences and in order...
When loss of coolant accident (LOCA) occurs, the availability of the emergency core cooling system (ECCS) becomes the most important issue that needs to be analysed in a nuclear reactor. In order to enable the ECCS to remove the released heat from the fue...
This report describes the results of aging, condition monitoring, and accident testing of crosslinked polyolefin (XLPO) cables. Three sets of cables were aged for up to 9 months under simultaneous thermal ({approx_equal}100{degrees}C) and radiation ({approx_equal}0. 10 kGy/hr) conditions. A sequential accident consisting of high dose ...
This report describes the results of aging, condition monitoring, and accident testing of crosslinked polyolefin (XLPO) cables. Three sets of cables were aged for up to 9 months under simultaneous thermal ({approx equal}100{degrees}C) and radiation ({approx equal}0. 10 kGy/hr) conditions. A sequential accident consisting of high dose ...
The purpose of the test program was to determine the effect of cladding surface thermocouples on time-to-critical heat flux (CHF) under blowdown conditions similar to those in the Loss-of-Fluid Test (LOFT) facility during a Loss-of-Coolant Experiment (LOCE). Previous steady state CHF tests indicated that cladding surface thermocouples ...
A reference source of MRBT bundle B-6 test data is presented with minimum interpretation. The primary objective of this 8 x 8 multirod burst test was to investigate cladding deformation in the alpha-plus-beta-Zircaloy temperature range under simulated light-water-reactor (LWR) loss-of-coolant accident (LOCA) ...
A deterministic analysis of the IRIS safety features has been carried out by means of the best-estimate code RELAP (ver. RELAP5 mod3.2). First, the main system components were modeled and tested separately, namely: the Reactor Pressure Vessel (RPV), the modular helical-coil Steam Generators (SG) and the Passive (natural circulation) Emergency Heat Removal System (PEHRS). Then, ...
Two loss-of-coolant accident (LOCA) simulation tests were conducted to investigate the effects of temperature uniformity and radial restraint boundary conditions on Zircaloy cladding deformation with a very low flow of superheated steam. In one test (B-5), boundary conditions typical of a large array were imposed ...
The Upper Head Injection (UHI) System is designed to inject directly into the upper head region of the reactor at 1300 psi during a postulated Loss-of-Coolant Accident (LOCA). Recently, some small break LOCA analyses were performed with specially modified versions of RELAP4/MOD5 created to evaluate large break ...
The purpose of the work is to analyze the scenario of intersystem loss-of-coolant accidents (LOCAs) in the feedwater lines of boiling water reactors (MBWRs), in light of the water hammer event at San Onofre, Unit 1. Such scenario has a potential for high frequency and high consequence. The frequency is potentially high because a similar event has occurred; ...
The emergency core cooling system (ECCS) of the CANDU-PHWR consists of three connected systems, designed to quickly provide sufficient emergency coolant to the core, following a loss of coolant accident (LOCA) in the primary heat transport system (PHTS). These systems are: Moderator system (MS) Emergency transfer line (ETL) Emergency core injection system ...
Two strain-gage-based drag transducers were developed to measure two-phase flow in simulated pressurized water reactor (PWR) test facilities. One transducer, a drag body (DB), was designed to measure the bidirectional average momentum flux passing through an end box. The second drag sensor, a break through detector (BTD), was designed to sense liquid downflow from the upper ...
The Advanced Test Reactor (ATR) is a 250-MW irradiation facility used to test reactor fuels and other materials, and also to produce radioisotopes. The ATR core is divided into five regions, or lobes, that normally operate at different power levels. To support future irradiation programs, it is desired that the maximum lobe power be increased 10% (from 60 ...
The RELAP4/MOD6 computer code is used for the analysis of the reactor core heat transfer during the reflooding phase of a postulated loss-of-coolant accident (LOCA) in a pressurized water reactor (PWR). The code requires the user to specify input parameters for the reflood heat transfer models. Results of previous comparisons of code calculations with ...
The tests conducted on the /sup 1///sub 5/-scale BWR Mark I pressure suppression test facility simulate the three-dimensional transient conditions that are encountered in a wetwell pressure suppression system during a hypothetical loss-of-coolant accident (LOCA). Specifically, the nitrogen (N2)-driven air clearing ...
The purpose of this document is to specify the experiment operating procedure for the test series TC-1. The effects of externally mounted cladding thermocouples on the fuel rod thermal behavior during LOCA blowdown and reflood cycles will be investigated in the test. Potential thermocouple effects include: (a) delayed DNB, (b) ...
This paper compares the predictions of the revised 3D-AIRLOCA computer code to those data available from the Aerospace Systems Test Reactor's (ASTR's) loss-of-coolant-accident (LOCA) tests run in 1964. The theoretical and experimental hot-spot temperature responses compare remarkably well. In the ...
One of the primary challenges for Gas-cooled Fast Reactors (GFR) is decay heat removal after a loss of coolant accident (LOCA). Due to the fact that thermal gas cooled reactors currently under design rely on passive mechanisms to dissipate decay heat, there is a strong motivation to accomplish GFR core cooling through natural phenomena. This work ...
This report documents assembly deposited power limits and the corresponding effluent temperature limits recommended for operating the K-14.1 subcycle to ensure sufficient cooling of reactor assemblies during the ECS phase of a Double Ended Guillotine Break (DEGSS) Loss of Coolant Accident (LOCA). The ECS LOCA effluent temperature ...
This paper summarizes the results of aging, condition monitoring, and accident testing of Class 1E cables used in nuclear power generating stations. Three sets of cables were aged for up to 9 months under simultaneous thermal ({approx equal} 100{degrees}C) and radiation ({approx equal}0.10 kGy/hr) conditions. After the aging, the cables were exposed to a ...
This paper summarizes the results of aging, condition monitoring, and accident testing of Class 1E cables used in nuclear power generating stations. Three sets of cables were aged for up to 9 months under simultaneous thermal ({approx_equal} 100{degrees}C) and radiation ({approx_equal}0.10 kGy/hr) conditions. After the aging, the cables were exposed to a ...
The first two loss-of-coolant experiments have been performed in the Loss-of-Fluid Test (LOFT) Facility. The experimental results are compared to analytical model results from the RELAP4 computer code. LOFT is a pressurized water reactor specially designed and instrumented to perform experiments representative of a loss-of-coolant accident ...
ROSA-IV LSTF (Large Scale Test Facility) is designed for integral experiments of a small-break LOCA in a PWR. The small-break LOCA, 2.5% cold-leg break with HPI (High Pressure Injection) system failure, has been analyzed with the computer program RELAP5/M...
ROSA-IV LSTF (Large Scale Test Facility) has been designed for integral experiments of a small-break LOCA in a PWR. The small-break LOCA in LSTF has been analyzed with the computer program RELAP5/MOD0. The following results have been obtained. (1) Thermoh...
Tests of closed-end, internally pressurized, Zircaloy-4 tubing specimens were utilized to develop low strain creep characteristics as a function of time at temperatures in the range of 1475/sup 0/F to 2000/sup 0/F (802/sup 0/C to 1093/sup 0/C) and hoop stresses in the range of 250 to 2500 psi for use in loss-of-coolant accident (LOCA) ...
The TRAC-BD1/MOD1 version of the computer code provides a best-estimate analysis capability for analyzing the full range of postulated accidents in boiling water reactor (BWR) systems and related experimental facilities. This version provides a consistent and unified analysis capability for analyzing all areas of a large or small break loss-of-coolant ...
This article discusses the specificity of RBMK (channel type, boiling water, graphite moderated) reactors and problems of Reactor Cooling System modelling employing computer codes. The article presents, how the RELAP/SCDAPSIM code, which is originally designed for modelling of accidents in vessel type reactors, is fit to simulate the phenomena in the RBMK reactor core and RCS ...
The computer code FLOWTRAN-TF is used to analyze hypothetical hydraulic accidents for the nuclear reactor at the Savannah River Site. During a hypothetical Large Break Loss-of-Coolant Accident (LOCA), reactor assemblies would contain a two-phase mixture of air and water which flows downward. Reactor assemblies consist of nested, ribbed ...
This report describes analyses of the response of a Pressurized-Water Reactor at the Zion Plant to hypothetical core-meltdown sequences. The analyses consider the progression of core meltdown, containment response, and consequences to the public for many specific accident sequences within the categories of Loss of Coolant Accidents ...
... Descriptors : *MENTAL HEALTH, *MOTOR VEHICLE ACCIDENTS, TRAFFIC, RESPONSE, ACCIDENTS, RECORDS, QUESTIONNAIRES ...
DTIC Science & Technology
The ICECON computer code was developed to provide the post-blowdown pressure transient in a Pressurized Water Reactor (PWR) ice condenser containment during a Loss-of-Coolant Accident (LOCA) as required by Appendix K to 10 CFR 50 for ECCS analysis. The calculated containment pressure is used to determine the backpressure for flow from the primary system to ...
A pre-test analysis of a small-break loss-of-coolant accident (SBLOCA) has been performed for the integral effect test loop of Korea Atomic Energy Research Institute (KAERI-ITL), the construction of which will be started soon. The KAERI-ITL is a full-height and 1/310 volume-scaled test facility based on the design ...
Twenty-four terminal blocks were tested in simulated Design Basis Event (DBE), Loss of Coolant Accident (LOCA) environments. The terminal blocks were powered at voltages of 4 Vdc, 45 Vdc, and 125 Vdc. Resulting currents associated with these voltage levels were 1.8 mA, 20 mA, and 1 A, respectively. Terminal-to-terminal and ...
In order to accurately predict reactor hydraulic behavior during a hypothetical Loss-of-Coolant-Accident (LOCA) the performance of reactor coolant pumps under off-design conditions must be understood. The LOCA of primary interest for the Savannah River Site (SRS) production reactors involves the aspiration of air into the recirculated ...
This paper describes quantitatively new mechanisms in the post-CHF regime which provide understanding and predictive capability for several current two-phase forced convective heat transfer problems. These mechanisms are important in predicting rod temperature turnaround and quenching during the reflood phase of either a hypothetical loss-of-coolant accident ...
