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.

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 ...

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...

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 bases for loss-of-coolant accident (LOCA) safety analyses required by reactor licensing regulations in the United States of America (USA), Federal Republic of Germany (FRG), and Japan are investigated and related to new data obtained since the regulations were established. The licensing approaches used in the three countries are ...

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...

A methodology is presented to determine the transient temperature distributions in fuel bundles under loss-of-coolant accident (LOCA) conditions using a recently developed variational technique for the solution of radial-azimuthal heat conduction in the fuel rods and the modified view factor concept proposed by Uchida and Nakamure to model the radiative ...

Severe accidents have been the subject of a great deal of analysis and research, particularly in the light water reactor community. Although severe accident analysis in Canada deuterium-uranium (CANDU) reactors has not been published abundantly, a significant body of research and analysis has been accumulated. This has occurred because CANDU has directly ...

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...

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 initiation of a ...

The behavior of the source term in a VVER-1000 type reactor is calculated using the Source Term Code Package (STCP). The input data are based on the Russian plant Zaporozhye-5. The selected accident sequence is a small break LOCA (Loss Of Coolant Accident...

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) fo...

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.

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.

In light of the TMI-2 nuclear reactor accident, transient and small LOCA events have been identified as areas of some of the most urgent research needs in light water reactor safety. Computer codes which have or are being developed for predicting the ther...

The LOCA (loss of coolant accident) is a hypothesized, low-probability accident used as a licensing basis for nuclear power plants. Computer codes which have been under development for at least a decade have been the principal tools used to assess the consequences of the hypothesized LOCA. Models exist in two ...

Arguments are presented that support the proposal that a separate burnout risk analysis, for the Flow Instability (FI) phase of a LOCA, not be required for reactor restart. With expected reactor power limits, flow instability will occur before critical heat flux (CHF). Since FI power limits preclude the occurrence of flow instability in a bounding ...

The present work continues our effort to perform an integrated safety analysis for the HYLIFE-II inertial fusion energy (IFE) power plant design. Recently we developed a base case for a severe accident scenario in order to calculate accident doses for HYLIFE-II. It consisted of a total loss of coolant accident ...

The Water Reactor Analysis Package (WRAP) has been expanded to provide the capability to analyze loss-of-coolant accidents (LOCAs) in both pressurized water reactors (PWRs) and boiling water reactors (BWRs) by using evaluation models (EMs). The input specifications for modules in the WRAP-EM system are presented in this document along with the JOSHUA input ...

Assembly power limits are prescribed for each reactor charge so that the Emergency Cooling System (ECS) will prevent core damage from exceeding specified damage limits during a postulated loss-of-coolant (LOCA) or loss-of-pumping (LOPA) accident. Generic assembly power limits which include a 10% uncertainty factor have been determined for the Mark 16B-31 ...

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 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 data ...

The MELCOR code, developed by Sandia National Laboratories, is capable of simulating the severe accident phenomena of light water reactor nuclear power plants (NPPs). A specific large-break loss-of-coolant accident (LOCA) for Kuosheng NPP is simulated with the use of the MELCOR 1.8.4 code. This accident is induced ...

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...

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 ...

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 documentation includes a delineation of the required operator response and the ...

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 ...

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 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 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...

A methodology has been developed for probability-based standards for low-pressure piping systems that are attached to the reactor coolant loops of advanced light water reactors (ALWRs) which could experience reactor coolant loop temperatures and pressures because of multiple isolation valve failures. This accident condition is called an intersystem ...

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 ...

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 ...

According to Maanshan Unit 1 plant probabilistic risk assessment (PRA), the small loss-of-coolant accident (LOCA) S{sub 2}WH is ranked third in frequency among postulated accidents. Maanshan-1 is a 2775-MW (thermal), three-loop Westinghouse pressurized water reactor with large dry containment. The sequence S{sub 2}WH is defined as a ...

A comparison has been made of SCDAP/RELAP5/MOD3- and TRAC-PF1/MOD1- based calculations of the fuel pin failure timing (time from containment isolation signal to first fuel pin failure) in a loss-of-coolant accident (LOCA). The two codes were used to calcu...

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 ab...

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 electrical heaters are heated to failure in a low-pressure, ...

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 ...

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 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...

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...

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...

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...

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, ...

Framatome-ANP and EDF have defined a generic approach for using a best-estimate code in design basis accident studies called Deterministic Realistic Method (DRM). It has been applied to elaborate a LB LOCA ECCS evaluation model based on the CATHARE code. From a prior statistical analysis of uncertainties, the DRM derives a conservative ...

Procedures in accordance with Appendix A of 10 CFR 50, GDC 2, call for an appropriate combination of the effects of the accident loads and loads caused by natural phenomena (such as earthquakes) to be reflected in the design bases of safety equipment. This requirement of interaction of loads has been implemented in various ways both within the NRC and the ...

This paper describes a novel diagnostic method based on inverse models that could be applied to identification of transients and accidents in nuclear power plants. In particular, it is shown that such models could be successfully applied to identification of loss-of-coolant accidents (LOCAs). This is demonstrated ...

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...

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...

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 ...

This paper describes the analysis of surge line break transients for the soviet designed, water cooled, light water moderated, power reactors referred to as VVERS. These events represent an intermediate size loss of coolant accident (LOCA) for these plants and provide a severe challenge to the safety system design. The pressurizer surge line represents the ...

This paper describes the analysis of surge line break transients for the soviet designed, water cooled, light water moderated, power reactors referred to as VVERS. These events represent an intermediate size loss of coolant accident (LOCA) for these plants and provide a severe challenge to the safety system design. The pressurizer surge line represents the ...

If a loss-of-coolant accident occurs in a fusion reactor, the temperature in the vacuum vessel will rise. If the decay heat is not removed, then the plasma vacuum boundary may melt. In this paper, the effects of the decay heat in a LOCA are analyzed numerically based on the Fusion Experimental Reactor (FER). In the case of a ...

A beryllium dust oxidation model has been developed at the Idaho National Laboratory (INL) by the Fusion Safety Program (FSP) for the MELCOR safety computer code. The purpose of this model is to investigate hydrogen production from beryllium dust layers on hot surfaces inside a fusion reactor vacuum vessel (VV) during in-vessel loss-of-cooling accidents ...

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 ...

Probabilistic risk assessment methodology is applied to generate an evaluation of the relative likelihood of safe recovery following selected pressurized water reactor (PWR) design basis accidents for a Russian V213 nuclear power reactor. US-designed PWRs similar to the V213 are used for reference and comparison. This V213 risk assessment is based on ...

In Westinghouse-designed reactors, the reactor coolant pump (RCP) seals constantly require a modest amount of cooling. This cooling function depends on the service water (SW) system. Upon the loss of the cooling function due to the unavailability of the SW, component cooling water system or electrical power (station blackout), the RCP seals may degrade, resulting in a loss-of-coolant ...

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 ...

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 ...

An T{sub p}S{sub L}B accident analysis for Taipower's Maanshan Unit 1 plant is reported. The plant is a 2775-MW(thermal) pressurized water reactor with large dry containment. Based on Maanhshan level-1 probabilistic risk assessment, the T{sub p}S{sub L}B sequence ranks first in accident frequency. The basic definition of T{sub ...

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 5MW low temperature nuclear heating reactor (NHR-5) is a new and advanced type of nuclear reactor developed by Institute of Nuclear Energy Technology (INET) of Tsinghua University of China in 1989. Its main loop is a thermal-hydraulic system with natural circulation. This paper studies the safety of NHR under the condition of loss-of-coolant accidents ...

Loss-of-coolant-accident (LOCA) and anticipated transient without scram (ATWS) calculations have been performed for the two Kernforschungszentrum Karlsruhe advanced pressurized water reactor reference designs (a homogeneous reactor with p/d = 1.2 and a heterogeneous reactor), for a homogeneous reactor with a tighter fuel rod lattice (p/d = 1.123), and for ...

This study reviews the influence of certain important variables, namely coolng fluid temperature, flaw shape and size, wall thickness, vessel radius, cladding, heat transfer coefficient, fluence, and time following a loss of coolant accident (LOCA) on the tendency of preexistent flaws in a reactor vessel to propagate due tothe activation of the emergency ...

RETRAN represents a new computer code approach for analyzing the thermal-hydraulic response of Nuclear Steam Supply Systems (NSSS) to hypothetical Loss of Coolant Accidents (LOCA) and Operational Transients. In contrast to the ''conservative'' approach, RETRAN provides ''best ...

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...

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...

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....

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...

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...

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...

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 control mechanism ...

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...

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 Reactor AVR. The AVR ...

A study on post-accident core coolability of LWR is discussed based on the practical fuel failure behavior experienced in NSRR, PBF, PNS and others. The fuel failure behavior at LOCA, RIA and PCM conditions are reviewed, and seven types of fuel failure modes are extracted as the basic failure mechanism at accident ...

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 ...

The models of the mechanistic code MFPR (Module for Fission Product Release) developed by IBRAE in collaboration with IRSN are described briefly in the first part of the paper. The influence of microscopic defects in the UO2 crystal structure on fission-gas transport out of grains and release from fuel pellets is described. These defects include point defects such as vacancies, interstitials and ...

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 ...

Models for cesium and iodine release from light-water reactor (LWR) fuel rods failed in steam were formulated based on experimental fission product release data from several types of failed LWR fuel rods. The models were applied to a pressurized water reactor (PWR) undergoing a hypothetical loss-of-coolant accident (LOCA) temperature ...

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 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 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 ...

The objective of this project was to perform stress analysis for graphite support structures of the General Atomics� 600 MWth GT-MHR prismatic core design using ABAQUS ® (ver. 6.75) to assess their structural integrity in air-ingress accident conditions where the structure weakens over time due to oxidation damages. The graphite support structures of prismatic type GT-MHR ...

The objective of this project was to perform stress analysis for graphite support structures of the General Atomics� 600 MWth GT-MHR prismatic core design using ABAQUS � (ver. 6.75) to assess their structural integrity in air-ingress accident conditions where the structure weakens over time due to oxidation damages. The graphite support structures of prismatic type GT-MHR ...

The Nuclear Regulatory Commission (NRC) Commissioners approved the proposed change to 10 CFR Part 50 Appendix K in July 1988. The change allows reactor vendors to use either the previous Appendix K requirements or a best-estimate analysis with defined uncertainties. This change to Appendix K has led NRC to investigate and the nuclear industry to develop an acceptable method to evaluate the ...

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 ...

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 temperature and pressure. Leakage, ...

The TRAC-PF1/MOD1 code (TRAC) 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). For this ...

During a hypothetical loss-of-coolant-accident (LOCA) in a Mark I boiling water reactor filled with water. Injection occurs below the water surface in a vertically downward direction through 0.6-m-diam open-end pipes called downcomers. A 1/5-scale facility was designed to experimentally verify the predicted loading function and to study the ensuing fluid ...

The simplified boiling water reactor (SBWR) is the latest design in the family of boiling water reactors (BWRs) from General Electric. The concept is based on many innovative, passive, safety systems that rely on naturally occurring phenomena, such as natural circulation, gravity flows, and condensation. Reliability has been improved by eliminating active systems such as pumps ...

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 ...

Countercurrent flow (CCF) is one of the most important phenomena occurring during a loss-of-coolant accident (LOCA) in a light water reactor because CCF affects the mass flow of core cooling water by limiting the water flow from the upper plenum to the core, from the downcomer to the lower plenum, and from the tubes in the steam generator to the steam ...

Analyses of two large-break loss-of-coolant-accident (LOCA) experiments, namely, LOFT L2-5 and Semiscale S-06-3, were performed with RELAP5/MOD2/cy 36.04. Excessive cooling, which occurred right before final quench, has been found in both calculations. The causes of the excessive cooling may be quite complex during large-break LOCA ...

A mathematical study of a class of two-phase fluid flow equations is presented for RELAP, a safety code for the computer simulation of loss-of-coolant accidents (LOCA's) in light water nuclear reactors. Mathematically, these equations are a system of non-linear stiff ordinary differential equations. Our results on the Jacobian matrix of this ...

Based on level 2 analyses in IPE (Individual Plant Examination) submittals accident progression, perspectives were obtained for all containment types. These perspectives consisted of insights on containment failure modes, releases therein, and factors responsible for the results. To illustrate the types of perspectives acquired on severe ...

The Water Reactor Analysis Package (WRAP) has been expanded to provide the capability to analyze loss-of-coolant accidents (LOCAs) in both pressurized water reactors (PWRs) and boiling water reactors (BWRs) by using evaluation models (EMs). The input spec...

THYDE-P2, being characterized by the new thermal-hydraulic network model, is applicable to analysis of RCS behaviors in response to various disturbances including LB (large break)-LOCA(loss-of-coolant accident). In LB-LOCA analysis, THYDE-P2 is capable of...

Core temperature and containment pressure have been calculated, for LEU core of PARR-1, following a large break Loss of Coolant Accident (LOCA). Heat transfer from bare reactor core to containment air has been calculated using analytical methods. The anal...

The present report is the user's manual for the computer code RODSWELL developed at the JRC-Ispra for the thermomechanical analysis of LWR fuel rods under simulated loss-of-coolant accident (LOCA) conditions. The code calculates the variation in space and...

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,...

The recently acquired MARCH computer program has been used to characterize the threat of hydrogen (H/sub 2/) combustion to the Westinghouse (W) ice condenser containment. The rates of hydrogen release, transport and combustion within the ice condenser containment for hypothetical accidents are studied. The small break LOCA (S/sub 2/D) and the large break ...

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 accident ...

This paper deals with the evaluation of tungsten temperature of divertor during external loss of coolant accident (Ex-LOCA) in connection with the safety design of nuclear fusion experimental reactor. The present result and the results of U.S.A. and EC ar...

A detailed dynamic analysis of the LOFT reactor core support structures was performed to determine the ability of the flow skirt/core filler and hold-down springs to withstand Loss-of-Coolant Accident (LOCA) plus Safe Shutdown Earthquake (SSE) loadings. A...

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...

An in-vessel severe accident progression has been analyzed to generate the basic data for an evaluation of the in-vessel severe accident management strategies and to identify the thermal hydraulic condition of the reactor vessel and the damage state of the in-vessel materials at a reactor vessel failure by using the SCDAP/RELAP5/MOD3.3 computer code during ...

An analytical model for evaluating the reactor containment pressure transient following a loss-of-coolant accident (LOCA) is presented. The model uses the CONTEMPT-LT computer program developed by Aerojet Nuclear Company. The sample problem studied is the containment response following the most severe postulated LOCA at the ...

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 response to the ...

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 response to the ...

This document describes the procedure used by UNI to develop and use the RELAP4 code for analysis of the N-Reactor system during a LOCA. The code modifications required are described as are the responses to the 10CFR50 App. K requirements. Each of the models used in the analysis is described and discussed. This document is the primary reference for the ...

THYDE-B1 is a computer code for predicting the thermohydraulic response of the primary system of a BWR during a loss-of-coolant accident (LOCA) aiming at the evaluation of the performance of the emergency core cooling system (ECCS). This code is mainly ap...

Vertical slip flow and flooding models, which have been incorporated in a version of the RELAP4 computer code by Aerojet Nuclear Company have led to significant improvements in modeling nuclear reactor coolant system phenomena during postulated large and small break loss-of-coolant accidents. The vertical slip flow model computes the separated fluid component ...

A probabilistic risk assessment (PRA) was conducted for a loss of coolant accident, (LOCA) in the McMaster Nuclear Reactor (MNR). A level 1 PRA was completed including event sequence modeling, system modeling, and quantification. To support the quantification of the accident sequence identified, data analysis using the Bayesian method ...

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 Program will obtain data ...

This paper describes the modeling and benchmarking of the Savannah River Site K-Reactor emergency core coolant system (ECCS), using the Transient Reactor Analysis Code (TRAC). The ECCS model was benchmarked against plant data obtained from various ECCS configurations. Next, the benchmarked model was used to simulate various loss-of-coolant accidents ...

The principal objective of this overall study is to compare alternative containment concepts for PWRs with regard to their potential for reducing public risk and with regard to their construction costs. This study is divided into two parts; the first part, which was commenced some months before TMI-2 accident, was done using the assumptions of the design basis ...

Nowadays regulatory rules and code models development are progressing on the goal of using best-estimate approximations in applications of license. Inside this framework, IBERDROLA is developing a PWR LOCA Analysis Methodology with one double slope, by a side the development of an Evaluation Model (upper-bounding model) that covers with conservative form the different aspects ...

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 operated in various modes to provide test ...

A modular computational system known as the Water Reactor Analysis Package (WRAP) is being developed at the Savannah River Laboratory for analysis of loss of coolant accidents (LOCA's) and other transients in water reactor systems. At this time, WRAP is e...

The VEERA facility was built in 1987 for experiments that simulate soluble neutron poison (boric acid) behaviour in a pressurized water reactor (PWR) during the long-term cooling period of loss-of-coolant accidents (LOCAs). The experiments provided insigh...

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...

The numerical methods and physical models used in the Transient Reactor Analysis Code (TRAC) versions PD2 and PF1 are discussed. Particular emphasis is placed on TRAC-PF1, the version specifically designed to analyze small-break loss-of-coolant accidents (LOCAs).

In the event of a loss of coolant accident (LOCA) in a nuclear power plant, it is possible that insulation for pipes or other items inside the containment building could be dislodged by the high energy break jet. This insulation debris could affect the re...

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...

One of the ways to increase the passive safety of a CANDU reactor is to decrease the contact conductance between the pressure and calandria tubes during loss of coolant/loss of emergency core cooling (LOCA/LOECC) situations. For severe accident scenarios,...

This is a Quarterly Progress Report on the Creare Downcomer Effects Program. The general context of this work is a postulated Loss-of-Coolant Accident (LOCA) in a Pressurized Water Reactor (PWR), although many of the basic processes being studied may also...

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...

Operation of K Reactor with a cooling tower requires that 186 Basin loss of inventory transients be considered during Design Basis Accident analyses requiring ECS injection, such as the LOCA and LOPA. Since the cooling tower systems are not considered saf...

This paper presents the application of RELAP5 to the calculation of a Large Break (200% doubled-ended rupture) Loss-of-Colant-Accident (LBLOCA) at the reactor vessel inlet for the proposed Westinghouse AP600 design. A parametric calculation was also perfo...

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 tungsten in the inboard ...

This paper documents a study of the consequences of loss of coolant accidents in the Advanced Neutron Source reactor, and it introduces the concept of an inertial flow diodes to mitigate the effect of large cold leg breaks. 2 refs., 1 fig.

In the unlikely event of a loss of coolant accident (LOCA) in a nuclear power plant, it would be necessary to avoid the complication of excessive hydrogen build-up in the containment. The size of recombiners needed to assist in controlling hydrogen accumu...

Generic Safety Issue (GSI)-191 Assessment of Debris Accumulation on PWR Sump Performance raised the concern of debris transport to pressurized-water-reactor (PWR) sump screens following a loss-of-coolant accident (LOCA) and subsequent impact to emergency ...

In the event of a postulated loss of coolant accident (LOCA) the cladding surface of PHWR fuel element will be exposed to high temperature steam environment which may result in extensive oxidation and embrittlement of the cladding tube. High temperature s...

Stationary two-phase flow experiments with a convergent nozzle are performed. The experimental results are appropriate to validate advanced computer codes, which are applied to the blowdown-phase of a loss-of-coolant accident (LOCA). The steam-water exper...

A heat removal system for nuclear reactors serving as a supplement to an Emergency Core Cooling System (ECCS) during a Loss of Coolant Accident (LOCA) comprises a plurality of heat pipes having one end in heat transfer relationship with either the reactor...

Electrical and mechanical properties of several commercial ethylene-propylene rubber (EPR) materials, typically used as electrical cable insulation, have been monitored during three simulations of nuclear power plant aging and accident stresses. The measu...

Electrical and mechanical properties of three commercial cross-linked polyolefin (XLPO) materials, typically used as electrical cable insulation, have been monitored during three simulations of nuclear power plant aging and accident stresses. For one XLPO...

If critical heat flux (CHF) occurs in a reactor core, it is very likely to happen during accident and then transient conditions. In the case, for instance, of a hypothetical LOCA event, we would have severe pressure, flow rate and power transients such as...

A guide for implementing regulations concerning the control of combustible gases in containment atmospheres following a LOCA in LWR type reactors with cylindrical, zircaloy-clad, oxide fuel elements is presented.

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...

CONTEMPT4/MOD2 describes the long-term behavior of multicompartment pressurized water reactor (PWR) containment systems and experimental containment systems subjected to postulated loss-of-coolant accident (LOCA) conditions. The program calculates the tim...

In the event of a loss of coolant accident (LOCA) in a nuclear power plant, it is possible that insulation for pipes or other items inside the containment building could be dislodged by the high energy break jet. This insulation debris could affect the re...

The design basis loss of coolant accident (LOCA) for light water nuclear reactors postulates a major break in a coolant line. Both the response of the reactor vessel and its mechanical system as well as the response of the pressure suppression containment...

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 report discusses ...

Because the containment structure is the last barrier against the release of radioactivity, an assessment was undertaken to identify the design weaknesses and estimate the margins of safety for the SEP containments under the postulated, combined loading conditions of a safe shutdown earthquake (SSE) and a design basis accident (DBA). The design basis ...

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 and design basis ...

A general analysis approach which includes the dynamic time-history analysis and the static analysis is presented. The secondary system piping is that piping which excludes the primary coolant loop in the Pressurized Water Reactor (PWR) plant, and the main steam, feedwater and reactor recirculation piping in the Boiling Water Reactor (BWR) plant. Piping response to the induced motion from a Loss ...

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 ...

Currently in the conceptual design stage, the Advanced Neutron Source Reactor (ANSR) will operate at a high heat flux, a high mass flux, an a high degree of coolant subcooling. Loss-of-coolant accident (LOCA) analyses using RELAP5 have been performed as part of an early evaluation of ANSR safety issues. This paper discusses the RELAP5 ANSR conceptual ...

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 qualification testing. The ...

This amendment to the General Electric Company Licensing Topical Report, ''BWR/6 Fuel Assembly--Evaluation of Combined Safe Shutdown Earthquake (SSE) and Loss-of-Coolant Accident (LOCA) Loadings'', is in response to the requests for additional information enclosed with a letter from Mr. Olan D. Parr, Division of Project ...

RELAP4/MOD7, a best estimate computer code for the calculation of thermal and hydraulic phenomena in a nuclear reactor or related system, is the latest version in the RELAP4 code development series. This paper evaluates the capability of RELAP4/MOD7 to calculate refill/reflood phenomena. This evaluation uses the data of International Standard Problem 10, which is based on West ...

A mechanistic fog inerting model has been developed to account for the effects of fog on the upward lean flammability limits of a combustible mixture based on the thermal theory of flame propagation. Benchmarking of this model with test data shows reasonably good agreement between the theory and the experiment. Applications of the model and available fog data to determine the ...

The Full-Scale Mark II CRT (Containment Response Test) Program is in progress at the Tokai-Mura Establishment of the Japan Atomic Energy Research Institute (JAERI). The primary objective of the on-going CRT Program is to provide a data base for evaluation of the pressure suppression pool (wetwell) hydrodynamic loads associated with a postulated loss-of-coolant ...

A postulated accident scenario for the Savannah River Site (SRS) K-reactor is a Double Ended Guillotine Break Loss of Coolant Accident (DEGB/LOCA) due to a coolant pipe break at the plenum inlet. The DEGB/LOCA consists of two parts, the first of which applies to the first few seconds of the transient. The first ...

A postulated accident scenario for the Savannah River Site (SRS) K-reactor is a Double Ended Guillotine Break Loss of Coolant Accident (DEGB/LOCA) due to a coolant pipe break at the plenum inlet. The DEGB/LOCA consists of two parts, the first of which applies to the first few seconds of the transient. The first ...

A postulated accident scenario for the Savannah River Site (SRS) K reactor is a double-ended guillotine break loss-of-coolant accident (DEGB/LOCA) caused by a coolant pipe break at the plenum inlet. The DEBG/LOCA consists of two parts, the first of which applies to the first few seconds of the transient. The first ...

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 loss-of-coolant ...

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 types of cable-to-device ...

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...

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) wa...

In certain accident scenarios, e.g. loss of coolant accidents (LOCA) all forced flow cooling is lost. Decay heating causes a temperature increase in the core coolant and the resulting thermal buoyancy causes a reversal of the flow direction to a natural c...

The RELAP5/MOD1 computer code was used in the analysis of a loss coolant accident (LOCA), postulated for Angra-2. The power plant was simulated through a division in control volumes and junctions suitable for a small break accident calculation. Initially ...

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...

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 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 accident thermal ...

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 Versuchsreaktor (AVR), the ...

Presented are experiment results of the Large Scale Test Facility (LSTF) conducted at the Japan Atomic Energy Agency (JAEA) with a focus on core exit thermocouple (CET) performance to detect core overheat during a vessel top break loss-of-coolant accident (LOCA) simulation experiment. The CET temperatures are used to start accident ...

Tokamak Cooling Water System (TCWS) drain tanks (DTs) serve two functions: normal operation and safety operation. Normal DTs are used for regular maintenance operations when draining is necessary. Safety DTs are used to receive the water leaked into the Vacuum Vessel (VV) after an in-vessel loss of cooling accident (LOCA) event. The preliminary design of ...

The principal objectives of the fission product release program currently in progress at Oak Ridge National Laboratory are to determine the quantity of radiologically significant fission products released from defected light water reactor (LWR) fuel rods under accident conditions, identify their chemical and physical forms, and interpret the results for use as input to ...

The principal objectives of the fission product release program currently in progress at Oak Ridge National Laboratory are to determine the quantity of radiologically significant fission products released from defected light water reactor (LWR) fuel rods under accident conditions, identify their chemical and physical forms, and interpret the results for use as input to ...

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 terminal-to-ground leakage ...

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 on an inner 4-by-4 square array by two ...

The use of analytical aids by utility technical support teams can enhance the staff's ability to manage accidents. Since instrumentation is exposed to environments beyond design-basis conditions, instruments may provide ambiguous information or may even fail. While it is most likely that many instruments will remain operable, their ability to provide unambiguous ...

In the licensing and validation process of best estimate codes for the analysis of nuclear reactors and postulated accident scenarios, the identification and quantification of the calculational uncertainty is required. One of the most important aspects in this process is the identification and recognition of the crucial contributing phenomena to the overall code uncertainty. ...

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 features of the APR1400 (Korean Next ...

This volume presents the results of the initiating event and accident sequence delineation analyses of the LaSalle Unit II nuclear power plant performed as part of the Level III PRA being performed by Sandia National Laboratories for the Nuclear Regulatory Commission. The initiating event identification included a thorough review of extant data and a detailed plant specific ...

The Generic Letter GL-96-06 issued by the U.S. Nuclear Regulatory Commission (NRC) required the utilities to evaluate the potential for voiding in their Containment Emergency Cooling Units (ECUs) due to a hypothetical Loss Of Coolant Accident (LOCA) or a Main Steam Line Break (MSLB) accompanied by the Loss Of Offsite Power (LOOP). When the offsite power is ...

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 ...

etary surface-based mobile wireless networks. These hybrid mobile networks have been deployed in rugged field loca- tions in the American desert and the ...

The goal of this work is the investigation of the influence of different accident management strategies on the thermal-hydraulics in the containment during a Large Break Loss of Coolant Accident with a large containment leak from the beginning of the accident. The increasing relevance of terrorism suggests a closer look at this kind of ...