Consequences of an unrestricted core heatup in an HTGR were assessed as part of the ERDA-funded probabilistic risk assessment study titled ''Accident Initiation and Progression Analysis'' (AIPA). The original objective of the AIPA study was to provide gui...
National Technical Information Service (NTIS)
How much energy is removed from the core and where it is deposited are important considerations in severe accidents. The core heatup rate will affect the timing of the damage progression and the nature of the core debris. Heat transferred from the core to...
The hypothetical accident considered for siting considerations in High Temperature Gas-Cooled Reactors (HTGR) is the so called Unrestricted Core Heatup Accident (UCHA), in which all forced circulation is lost at initiation, and none of the auxillary cooli...
In the most severe hypothetical core heatup accidents in High Temperature Gas Cooled Reactors, the heatup of the concrete reactor vessel can result in gas release from degrading concrete which can ultimately lead to containment building failure. This gas release is largely affected by the moisture migration during ...
DOE Information Bridge
Thermodynamic accident analyses have been performed with computer simulation models to investigate core heatup sequences, sensitivity analyses, power variations, anticipated transients without scram, and core displacement considerations for probabilistic ...
In High Temperature Gas Cooled Reactor (HTGR) siting considerations, the Unrestricted Core Heatup Accidents (UCHA) are considered as accidents of highest consequence, corresponding to core meltdown accidents in light water reactors. Initiation of such ...
This is a followup to an earlier report documenting the MORECA code, an interactive simulation tool for performing independent analyses of postulated MHTGR core transients and heatup accidents. The research was performed at Oak Ridge National Laboratory (...
The formation of combustible atmospheres during unrestricted core heatup accidents in High Temperature Gas-Cooled Reactors is being investigated, considering the effects of only partially mixed atmospheres. It is found that the previously used assumption ...
The decay heat removal by a passive air colling system from a modular high tempeature gas cooled reactor during depressurized core heatup accident scenarios was analyzed. The effects of several design and operating parameters on the peak fuel and vessel temperatures were established. The results indicate that fuel and vessel ...
Energy Citations Database
A computational model has been developed that calculates the thermal degradation of the reactor core of the production reactors at the Savannah River Site (SRS) under postulated severe accident conditions. This model addresses heatup and degradation of th...
Core flow blockage events are a leading contributor to core damage initiation risk in the Advanced Neutron Source (ANS) reactor. During such an accident, insufficient cooling of the fuel could result in core heatup and melting under full coolant flow cond...
Because of the high interest in modular High Temperature Reactor performance and safety, a cooperative project has been established involving the Oak Ridge National Laboratory (ORNL), Arbeitsgemeinschaft Versuchs Reaktor GmbH (AVR), and Kernforschungsanlage Juelich GmbH (KFA) in reactor physics, performance and safety. This paper presents initial results of ORNL's examination of a ...
A description is given of experiments to investigate the behavior of HTGR core materials during hypothetical heatup accidents in which the core temperature is assumed to reach values between 2400/sup 0/C and the graphite sublimation range (>3600/sup 0/C). The work includes BISO coated fuel particle failure, ...
The design features of the modular high-temperature gas-cooled reactor (HTGR) have the potential to make it essentially invulnerable to damage from postulated core heatup accidents. Simulations of long-term loss-of-forced-convection (LOFC) accidents, both with and without depressurization of the primary coolant and ...
The hypothetical accident considered for siting considerations in High Temperature Gas-Cooled Reactors (HTGR) is the so called Unrestricted Core Heatup Accident (UCHA), in which all forced circulation is lost at initiation, and none of the auxillary cooling loops can be started. The result is a gradual slow ...
The design features of the modular high-temperature gas-cooled reactor (MHTGR) have the potential to make it essentially invulnerable to damage from postulated core heatup accidents. This report describes the ORNL MORECA code, which was developed for analyzing postulated long-term core heatup ...
Several safety evaluations for large High Temperature Gas Cooled Reactors (HTGR), using a Prestressed Concrete Reactor Vessel (PCRV) design, have concluded that Unrestricted Core Heatup Accidents (UCHA) present the most important severe accidents, resulting in the dominant source term. While the ...
On September 22, 1970, for the first time an accident simulation experiment with complete failure of the forced core cooling and the nuclear shut-down system was performed in the AVR-reactor: due to a small heat-up of the fuel the nuclear chain-reaction w...
Work continued on high-temperature gas-cooled reactor safety research directed towards both the Fort St. Vrain and 2240-MW(t) lead plant reactors. Code development and verification activities addressed simulations of unrestricted core heatup accidents, st...
This paper summarizes research performed at Oak Ridge National Laboratory (ORNL) to assist the Nuclear Regulatory Commission (NRC) in preliminary determinations of licensability of the US Department of Energy (DOE) reference design of a standard modular high-temperature gas-cooled reactor (MHTGR). The work described includes independent analyses of core ...
This paper provides an overview of high-temperature gas-cooled reactor (HTR) fission product chemistry and its influence on source terms in core heatup accidents. These accidents are risk-dominating for medium-sized HTRs and are characterized by maximum core temperatures of ...
Consequences of an unrestricted core heatup in an HTGR were assessed as part of the ERDA-funded probabilistic risk assessment study titled ''Accident Initiation and Progression Analysis'' (AIPA). The original objective of the AIPA study was to provide guidance for safety research and development ...
Aerosol samples consisting of fission products and elements of light water reactor structural materials were collected during simulating in a laboratory scale the heat-up phase of a core melt accident. The aerosol particles were formed in a steam atmosphe...
Five accident conditions are considered in an analysis of their radiological consequences. The five accident conditions are core heatup resulting from loss of offsite power and earthquake; reheater tube leak; slow depressurization; rapid depressurization; and steam ingress from steam generator main bundle tube ...
An analysis of a loss-of-pumping accident (LOPA) has been performed using a RELAP5 Model of the Savannah River L-Reactor plant. The analysis showed that the LOPA transient was characterized by an early process system cooldown resulting from reactor trip, followed by a heatup and rapid expulsion of process system coolant once pump availability was lost. ...
SCDAP/RELAP5/MOD3.1, an integrated thermal hydraulic analysis code developed primarily to simulate severe accidents in nuclear power plants, was used to predict the progression of core damage during the TMI-2 accident. The version of the code used for the TMI-2 analysis described in this paper includes models to predict ...
Diffusion rates of molybdenum through graphite were determined in the temperature range of 2250/sup 0/C to 3300/sup 0/C. This work was conducted to characterize the behavior of fission products under an unrestricted core heat-up accident condition in a HTGR (High Temperature Gas Cooled Reactor).
One of the central design features of the 250 MWT modular HTGR is the ability to withstand uncontrolled heatup accidents without severe consequences. This paper describes calculational studies, conducted to test this design feature. A multi-node thermal-hydraulic model of the 250 MWT modular HTGR reactor core was developed and ...
For certain postulated accidents beyond the present design basis for LWRs it has been shown that the potential exists to fail the containment building as a result of extensive core debris-concrete interactions. The MARCH code was developed at BCL to analyse the response of a LWR to core meltdown accidents. MARCH ...
The safety potential of the Modular High-Temperature Gas Reactor (MHTGR) was evaluated, based on the Preliminary Safety Information Document (PSID), as submitted by the US Department of Energy to the US Nuclear Regulatory Commission. The relevant reactor safety codes were extended for this purpose and applied to this new reactor concept, searching primarily for potential ...
A computational model has been developed that calculates the thermal degradation of the reactor core of the production reactors at the Savannah River Site (SRS) under postulated severe accident conditions. This model addresses heatup and degradation of the U-Al fuel and Li-Al or U-metal target assemblies and neighboring structures. ...
The amount of fission product release during a core heatup accident in a medium-sized high-temperature gas reactor depends on the size of the inadvertent opening in the primary circuit; this dependence is assessed. The opening triggers a depressurization event that is assumed to be coupled with the failure of the forced circulation in ...
The high-temperature gas-cooled reactor (HTGR) program will be attractive to a broad range of owner/operators and meet public acceptance if the future HTGRs would be completely free from accidents, which cause a significant release of radioactivity into the environment. An advanced vessel cooling system concept, in which there is no heat loss in normal operation and the decay ...
The report investigates the effects of fuel behavior and control rod motion of an LMFBR core in a postulated accident scenario, in which the heatup of the fuel is caused mainly by decay heat. The sequence of this class of accidents is usually characterized by the loss of sodium, cladding, and can wall before major ...
An analysis of spent fuel heatup following a hypothetical accident involving drainage of the storage pool is presented. Computations based upon a new computer code called SFUEL have been performed to assess the effect of decay time, fuel element design, s...
A model has been developed to predict the thermal hydraulics in the uncovered part of a pressurized water reactor core. The core is considered to be a heterogeneous porous medium with different permeabilities and effective thermal conductivities in the radial and axial directions. The flow in the core is modeled by the ...
The vaporization of core materials other than fission products during a postulated severe light water reactor accident is treated by chemical thermodynamics. The core materials considered were (a) the control rod materials, silver, cadmium, and indium; (b) the structural materials, iron, chromium, nickel, and manganese; (c) cladding ...
The direct-cycle supercritical-pressure (250-bar) light-water-cooled reactors have the advantages of high thermal efficiency, simplicity, and breeding capability. The low-density fluid steam is generated in the core and fed directly to the turbines. A reactor large-break loss-of-coolant accident (LBLOCA) has been analyzed and a computer code was developed ...
During a severe core damage accident in a nuclear light water reactor, the process of melting and solidification of core material results in a cohesive debris bed. This paper determines (a) the initial equilibrium thickness of the lower crust of the bed, which serves as a receptacle for subsequently melted material, (b) the formation ...
This paper presents the results of a detailed comparison of the Source Term Code Package (STCP) and the SCDAP computer codes for simulation of the Power Burst Facility (PBF) Severe Fuel Damage (SFD) test 1-1. The SCDAP code is mechanistic, and has been benchmarked against a wide range of severe accident data. The SFD 1-1 test was designed to simulate the ...
In this paper the SCDAP/RELAPS severe accident analysis computer code, developed at the Idaho National Engineering Laboratory, is used to analyze the fourth in a series of debris formation experiments. The debris formation-four (DF-4) experiment deals with heatup and meltdown of a boiling water reactor (BWR)-representative fuel and control blade assembly ...
Thermodynamic accident analyses have been performed with computer simulation models to investigate core heatup sequences, sensitivity analyses, power variations, anticipated transients without scram, and core displacement considerations for probabilistic safety analyses (PSA) of small gas-cooled high-temperature ...
Core liquid level depression can result in partial core dryout and heatup early in a small break loss-of-coolant accident (SBLOCA) transient. Such behavior occurs when steam, trapped in the upper regions of the reactor primary system (between the loop seal and the core inventory), moves coolant ...
The AYERM code is a computer program which has been developed for the high-temperature gas-cooled reactor (HTGR) safety research program. It is a conjunction of the heat conduction code, AYER, and a set of special subroutines. This modified AYER code can predict the time-dependent release of volatile fission products from a reactor core during a hypothetical ...
... The computer codes RELAP-3 for reactor system blowdown and THETAl-B for fuel element heatup were used to perform these calculations at a ...
DTIC Science & Technology
As a result of the Three Mile Island Unit-2 (TMI-2) accident, the Nuclear Regulatory Commission has initiated a severe fuel damage test program to evaluate fuel rod and core response during severe accidents similar to TMI-2. The first test of Phase I of this series has been successfully completed in the Power Burst Facility at the ...
The ultimate decay heat removal system for the current Modular High Temperature Gas-Cooled reactors is a completely passive natural convection air cooling loop. This paper considers an extremely remote accident scenario, where even this passive system fails, and heat rejection is only via a layer of thermal insulation to the reactor silo structure and the surrounding soil. The ...
Probabilistic risk assessment techniques have been applied to obtain guidance in choosing nuclear safety research and development that is most worthwhile for high-temperature gas-cooled reactor (HTGR) nuclear power plants. The probabilistic techniques used are similar to those employed in the Reactor Safety Study for light water reactors (LWRs), WASH-1400, directed by Dr. N. C. Rasmussen. The ...
The accident at the Three Mile Island unit 2 coincided with a shift in perspective on commercial light water reactor (LWR) safety from one that emphasized design-basis accidents to one that emphasized an increasing concern for low-probability severe accidents. This shift in emphasis was reflected in the research programs initiated by ...
Recently, the capabilities of the CORMLT code, which was designed to predict heatup, degradation, and meltdown of core and Reactor Pressure VEssel (RPV) internals during postulated severe accidents, were enhanced to enable tracking of individual fission product species during core meltdown. In addition, a ...
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 natural circulation boiling water reactor (BWR) with a rated capacity of 600 MW (electric) has been conceptually designed for small- and medium-sized light water reactors. The components and systems in the reactor are simplified by eliminating pumped recirculation systems and pumped emergency core cooling systems. Consequently, the volume of the reactor building is -- 50% of ...
This document identifies data that can be used for assessing various models embodied in severe accident analysis codes. Phenomena considered in this document, which were limited to those anticipated to be of interest in assessing severe accidents in the USEPR developed by AREVA, include: � Fuel Heatup and Melt Progression � Reactor ...
This paper describes an economic model which was developed to evaluate the net costs incurred by a utility due to an accident induced outage at a nuclear power plant. During such an outage the portion of the plant operating costs associated with power production are saved; however, the owning utility faces a sizable expense as fossil fuels are burned as a substitute for the ...
The report reviews briefly the HTGR core analytical methods that were developed during the course of the program. The features of these analytical methods are compared with methods used to perform similar analyses, and examples of the use of these methods are cited. Included are discussions of HEATUP (a computer code for the thermal analysis of an LOFC ...
Progress is reported in the following areas: systems and safety analysis; fission product technology; primary coolant technology; seismic and vibration technology; confinement components; primary system materials technology; safety instrumentation; loss of flow accident analysis using HEATUP code; use of coupled-conduction-convection model for ...
Critical heat flax onset and propagation (frequently referred to as burnout'' or DNB'' (Departure from Nucleate Boiling)) is one of the most investigated heat transfer processes in the past fifteen years, because of its implications on core design of watercooled reactors. In the last two or three years, studies in this field ...
An interactive simulation code for studying postulated heatup accidents in modular high-temperature gas-cooled reactors (MHTGRs) has been adapted to assist with parametric design studies of the US Department of Energy's (DOE's) direct-cycle gas-turbine MH...
Research by the nuclear industry and regulating agencies has shown that the oxidation and melting of the Zircaloy cladding of nuclear fuel rods has a dominant influence upon the behavior of a reactor core during a severe accident like TMI-2. The work presented in this dissertation addresses three important aspects of that research. First, the influence of ...
A model is described for the movement of melted metallic material through a ceramic porous debris bed. The model is designed for the analysis of severe accidents in LWRs, wherein melted core plate material may slump onto the top of a porous bed of relocated core material supported by the lower head. The permeation of the melted ...
This paper presents an analytical and numerical analysis which evaluates the core-structure heat-up and subsequent relocation of molten core materials during a NWR short-term station blackout accident with ADS. A simplified one-dimensional approach coupled with bounding arguments is first presented to establish an ...
Fuel performance models based on empirical evidence are used to predict particle failure and fission product release in the design of high-temperature gas-cooled reactors (HTGRs). Advances in HTGR fuel performance models have improved the agreement between observed and predicted performance and contributed to an enhanced position of the HTGR with regard to investment risk and passive safety. Heavy ...
The Three Mile Island (TMI-2) accident in 1979 resulted in approximately 45% of the fuel collapsing into an irregularly-shaped debris bed near the center of the core, while some of the molten material flowed into the lower dome of the reactor vessel where it solidified. The immediate cause of this severely degraded geometry was loss of coolant and ...
MELCOR is a fully integrated, engineering-level computer code that models the progression of severe accidents in light water reactor nuclear power plants. MELCOR is being developed at Sandia National Laboratories for the US Nuclear Regulatory Commission as a second-generation plant risk assessment tool and the successor to the Source Term Code Package. The entire spectrum of ...
MELCOR is a fully integrated, engineering-level computer code that models the progression of severe accidents in light water reactor (LWR) nuclear power plants, which is being developed at Sandia National Laboratories for the US Nuclear Regulatory Commission (US-NRC). The entire spectrum of severe accident phenomena, including reactor coolant system and ...
All BWR degraded core experiments performed prior to CORA-33 were conducted under ``wet`` core degradation conditions for which water remains within the core and continuous steaming feeds metal/steam oxidation reactions on the in-core metallic surfaces. However, one dominant set of accident ...
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 ...
NASA Astrophysics Data System (ADS)
A series of small breaks in the pressurizer vent piping and primary coolant system were analyzed. In the larger vent break case, a four-inch break, the ECCS was actuated on high confinement pressure and low primary system pressure. The smaller vent break, a two-inch break, did not cause ECCS actuation. Breaks in the primary coolant system were studied to determine if there was a break size or ...
An analysis of spent fuel heatup following a hypothetical accident involving drainage of the storage pool is presented. Computations based upon a new computer code called SFUEL have been performed to assess the effect of decay time, fuel element design, storage rack design, packing density, room ventilation, drainage level, and other variables on the ...
The Accident Diagnostic, Analysis and Management (ADAM) computer code has been developed as a tool for on-line applications to accident diagnostics, simulation, management and training. ADAM's severe accident simulation capabilities incorporate a balance of mechanistic, phenomenologically based models with simple parametric ...
An interactive simulation code for studying postulated heatup accidents in modular high-temperature gas-cooled reactors (MHTGRs) has been adapted to assist with parametric design studies of the US Department of Energy`s (DOE`s) direct-cycle gas-turbine MHTGR concept. The studies show that the proposed MHTGR designs are very robust and can generally ...
The upper plenum structures of Three Mile Island Unit 2 (TMI-2) have been of interest because of both the limited damage locations and the low structure temperatures in the upper part of the plenum. Damage to the core upper grid was localized to the two quadrants away from the hot-leg connections and was essentially limited to the bottom of the grid. Examination of two ...
Data obtained from the CORA bundle heatup and melting experiments, performed at Kernforschungszentrum, Karlsruhe, Germany, are being analyzed at the Idaho National Engineering Laboratory. The analysis is being performed as part of a systematic review of core melt progression experiments for the United States Nuclear Regulatory Commission to (a) develop an ...
A series of four experiments has been conducted at Argonne National Laboratory's TREAT Reactor. These experiments, which are sponsored by an international consortium organized by the Electric Power Research Institute, are designed to investigate the source term, i.e., the type, quantity and timing of release of radioactive fission products from a light water reactor to the environment in ...
MELCOR is a fully integrated computer code that models all phases of the progression of severe accidents in nuclear power plants. It is being developed for the US Nuclear Regulatory Commission (NRC) by Sandia National Laboratories and is designed to provide an improved severe-accident/source term analysis capability relative to the older source term code ...
Thermal hydraulic simulation capability for accident conditions is needed at present in VTT in several programs. Traditional thermal hydraulic models are too heavy for simulation in the analysis tasks, where the main emphasis is the rapid neutron dynamics or the core melting. The GENFLO thermal hydraulic model has been developed at VTT for special ...
The stringent safety demands for advanced small pebble bed high-temperature reactors (HTRs) are outlined. Main results of German studies on source term estimation are discussed. Core heatup events are no longer dominant for modern fuel, but fission product transport during water ingress accidents (steam cycle plants) and He-circuit ...
Two sets of MELCOR calculations have been completed studying the effects of air ingression on the consequences of various severe accident scenarios. One set of calculations analyzed a station blackout with surge line failure prior to vessel breach, starting from nominal operating conditions; the other set of calculations analyzed a station blackout occurring during shutdown ...
This paper presents the results from MELCOR (Version 1.8BC) calculations of the Long-Term Station Blackout Accident Sequence, with failure to depressurize the reactor vessel, at the Peach Bottom (BWR Mark I) plant, and presents comparisons with Source Term Code Package (STCP) calculations of the same sequence. This sequence assumes that batteries are available for six hours ...
An interactive workstation-based simulator has been developed for performing analyses of modular high-temperature gas-cooled reactor (MHTGR) core transients and accidents. It was originally developed at Oak Ridge National Laboratory for the US Nuclear Regulatory Commission to assess the licensability of the US Department of Energy (DOE) steam cycle design ...
GE latest evolution of the Boiling Water Reactor, the ESBWR, is an advanced, 4500 MWth nuclear power plant design, submitted to the NRC for design certification in 2005. This paper presents the key results of performance analyses of ESBWR ECCS and containment systems. The ESBWR is designed to take full advantage of passive features to improve the plant performance and economics. The key features ...
A study was performed to provide a description of the release of fission products from failed fuel particles during a core heatup event in an HTGR. The need for this study was established in the Accident Initiation and Progression Analysis program. The release of fission products was measured from laser-failed BISO ThO/sub 2/, TRISO ...
The paper presents the results of an experimental investigation on the formation mechanism and physical characterization of simulated nuclear aerosols that could likely be released during an HTGR core heat-up accident. Experiments were carried out in a high-temperature flow system consisting essentially of an inductively heated release ...
The release of fission products from failed fuel particles was measured under simulated accident (core heatup) conditions. A generic model and specific model parameters that describe delayed fission product release from the kernels of failed HTGR fuel particles were developed from the experimental results. The release of fission ...
Analyses were performed to determine the water retention capabilities of the primary coolant system (PCS) of pressurized water reactors during severe accidents. The objectives were to (1) determine the locations and magnitudes of retained water during core uncovering, (2) predict the PCS initial conditions at core uncovery, and (3) ...
The Idaho National Laboratory (INL), under the auspices of the U.S. Department of Energy, is performing research and development that focuses on key phenomena important during potential scenarios that may occur in the Next Generation Nuclear Plant (NGNP)/Gen-IV very high temperature reactor (VHTR). Phenomena Identification and Ranking Studies to date have identified that an air ingress event ...
Air-ingress following the pipe rupture is considered to be the most serious accident in the VHTRs due to its potential problems such as core heat-up, structural integrity and toxic gas release. Previously, it has been believed that the main air-ingress mechanism of this accident is the molecular diffusion process ...
Natural circulation flows can develop within a reactor coolant system (RCS) during certain severe reactor accidents, transferring decay energy from the core to other parts of the RCS. The associated heatup of RCS structures can lead to pressure boundary failures; with notable vulnerabilities in the pressurizer surge line, the hot leg ...
The DC experiment series investigates the heatup and melt of dry reactor core debris through nuclear heating of actual reactor materials in order to obtain the thermal properties of dry debris, the nature of the transition from a debris bed to a molten po...
The hypothetical core disruptive accident in an LMFBR is discussed under the following main headings: reactor dynamics; mechanical consequences; and post- accident heat removal. 79 references. (DCC)
Core flow blockage events are a leading contributor to core damage initiation risk in the Advanced Neutron Source (ANS) reactor. During such an accident, insufficient cooling of the fuel could result in core heatup and melting under full coolant flow condition. Coolant inertia forces acting on ...
The DF-4 in-pile fuel damage experiment investigated the behavior of boiling water reactor (BWR) fuel canisters and control blades in the high temperature environment of an unrecovered reactor accident. This experiment, which was carried out in the Annular Core Research Reactor (ACRR) at Sandia National Laboratories, was performed under the ...
The moderator temperature coefficient of reactivity (MTC) is an important parameter in safety analyses for thermal reactors. A positive MTC can exacerbate the severity of heatup transients, while a negative MTC can worsen the severity of cooldown transients. In particular, a strongly negative MTC is the major determinate of the severity of the steamline-break ...
The Idaho National Engineering and Environmental Laboratory (INEEL), Nexant Inc. and the Oregon State University (OSU) developed an innovative Multi-Application Small Light Water Reactor (MASLWR) concept. The MASLWR is a small, modular, safe, and economic natural circulation light water reactor developed with the primary goal of producing electric power, but with the flexibility to be used for ...
The FLECHT-SEASET correlation was evaluated to determine its ability to support a conservative prediction of peak cladding temperature (PCT) for the hottest fuel rod during reflooding of a reactor core following a loss-of-coolant accident (LOCA). A computer program, FLECSET, was written to facilitate the evaluation. Using FLECSET, 15 FLECHT-SEASET tests ...
Small break loss-of-coolant accident (SBLOCA) data were recorded during tests 9.lb and 6.2 TC in the Boucle d'Etudes Thermohydrouliques Systeme (BETHSY) facility at the Centre d'Etudes Nucleares de Grenoble (CENG) complex in Grenoble, France. The data from test 9.lb form the basis for the International Standard Problem number 27 (ISP-27). For each test the ...
Small break loss-of-coolant accident (SBLOCA) data were recorded during tests 9.lb and 6.2 TC in the Boucle d`Etudes Thermohydrouliques Systeme (BETHSY) facility at the Centre d`Etudes Nucleares de Grenoble (CENG) complex in Grenoble, France. The data from test 9.lb form the basis for the International Standard Problem number 27 (ISP-27). For each test the primary system ...
This paper reports on Run SB-CL-05, a test similar to Semiscale Run S-UT-8. The test results show that the core was uncovered briefly during the accident and that the rods overheated at certain core locations. Liquid holdup on the upflow side of the steam-generator tubes was observed. After the loop seal cleared, the ...
MELCOR is a fully integrated, engineering-level computer code that models the progression of severe accidents in light water reactor nuclear power plants. MELCOR is being developed at Sandia National Laboratories for the US Nuclear Regulatory Commission as a second-generation plant risk assessment tool and the successor to the Source Term Code Package. A broad spectrum of ...
MELCOR is a fully integrated, engineering-level computer code that models the progression of severe accidents in light water reactor nuclear power plants. MELCOR is being developed at Sandia National Laboratories for the U.S. Nuclear Regulatory Commission as a second-generation plant risk assessment tool and the successor to the Source Term Code Package. A broad spectrum of ...
The TRAC-BD1 (transient reactor analysis code) provides a consistent and unified analysis capability of the entire loss-of-coolant accident (LOCA) sequence, beginning with the blowdown phase, through heatup, reflood with quenching, and finally the refill ...
Modeling and code development work on the modular High-Temperature Gas-Cooled Reactor (HTGR) continued with the development and testing of a thermal model of the upper reflector. The longer-term heatup accident scenario in which cavity wall cooling is los...
MELCOR is a fully integrated, engineering-level computer code that models the progression of severe accidents in light water reactor (LWR) nuclear power plants, being developed at Sandia National Laboratories for the US Nuclear Regulatory Commission (USNRC). The entire spectrum of severe accident phenomena, including reactor coolant system and containment ...
Information is presented concerning the scenario for core degradation; early core degradation in small break loss-of-coolant accidents; and containing a degraded core accident. (ERA citation 08:026288)
This report studies the thermal behavior of debris beds impermeable to cooling water. It considers these debris beds as idealizations of the slumped TMI-2 core as conjectured between 2 h, 54 min and 3 h, 46 min into the accident. Section 2 calculates the decay heat levels as functions of time after shutdown, assuming various fractions of the more volatile ...
Core flow blockage events have been identified as a leading contributor to core damage initiation risk in the Advanced Neutron Source (ANS) reactor. During such an accident, insufficient cooling of the fuel in a few adjacent blocked coolant channels out of several hundred channels, could also result in core ...
The very high-temperature gas-cooled reactor (VHTR) is envisioned as a single- or dual-purpose reactor for electricity and hydrogen generation. The concept has average coolant temperatures above 9000C and operational fuel temperatures above 12500C. The concept provides the potential for increased energy conversion efficiency and for high-temperature process heat application in addition to power ...
The very high temperature gas-cooled reactors (VHTGRs) are those concepts that have average coolant temperatures above 9000C or operational fuel temperatures above 12500C. These concepts provide the potential for increased energy conversion efficiency and for high-temperature process heat application in addition to power generation and nuclear hydrogen generation. While all the High Temperature ...
The very high temperature gas-cooled reactors (VHTGRs) are those concepts that have average coolant temperatures above 900 degrees C or operational fuel temperatures above 1250 degrees C. These concepts provide the potential for increased energy conversion efficiency and for high-temperature process heat application in addition to power generation and nuclear hydrogen generation. While all the ...
The US Department of Energy is performing research and development (R&D) that focuses on key phenomena that are important during challenging scenarios that may occur in the Next Generation Nuclear Plant (NGNP) Program / GEN-IV Very High Temperature Reactor (VHTR). Phenomena identification and ranking studies (PIRT) to date have identified the air ingress event, following on the heels of a VHTR ...
The HEATUP code, a modification of the general, time-dependent, one-, two-, and three-dimensional program HEATING5, was designed for the thermal analysis of a Loss of Forced Circulation accident in a High Temperature Gas-Cooled Reactor. This report contains a description of the computational model which includes: a description of the basic problem; a short ...
VICTORIA is a mechanistic computer code that treats fission product behavior in the reactor coolant system during a severe accident. During an accident, fission products that deposit on structural surfaces produce heat loads that can cause fission products to revaporize and possibly cause structures, such as a pipe, to fail. This mechanism had been lacking ...
VELCOR is an integrated, engineering-level computer code that models the progression of severe accidents in light water reactor (LWR) nuclear power plants. The entire spectrum of severe accident phenomena, including reactor coolant system and containment thermal-hydraulic response, core heatup, degradation and ...
Analyses of a loss-of-coolant experiment carried out at the PSB-VVER test facility with the RELAP5/MOD3.2 code have been performed independently by analysts at the Electrogorsk Research and Engineering Center (EREC) and the Idaho National Engineering and Environmental Laboratory (INEEL). The PSB-VVER facility is a full-height scale model of a VVER 1000 reactor that is approximately 1/300 scale in ...
The environmental consequences of an LMFBR accident involving breach of containment are so severe that such accidents must not be allowed to happen. Present methods for analyzing hypothetical core disruptive accidents like a loss of flow with failure to s...
This study evaluates the values (benefits) and impacts (costs) associated with potential resolutions to Generic Issue 143, ``Availability of HVAC and Chilled Water Systems.`` The study identifies vulnerabilities related to failures of HVAC, chilled water, and room cooling systems; develops estimates of room heatup rates and safety-related equipment vulnerabilities following ...
Experiment-specific models have been employed since 1986 by Oak Ridge National Laboratory (ORNL) severe accident analysis programs for the purpose of boiling water reactor experimental planning and optimum interpretation of experimental results. The large integral tests performed to date, which start from an initial undamaged core state, have involved ...
Severe accident management can be defined as the use of existing and/or alternative resources, systems and actors to prevent or mitigate a core-melt accident. For each accident sequence and each combination of severe accident management strategies, there ...
An extensive study was performed for the prediction of hydrogen production during a BWR6 core heatup transient. Hydrogen production was found to be strongly affected by the transient core heat transfer and vessel hydrodynamics. Various sources important for steam generation and hydrogen production were identified. The results of the ...
An overview of the analysis of core-disruptive accidents is given. These analyses are for the purpose of understanding and predicting fast reactor behavior in severe low probability accident conditions, to establish the consequences of such conditions and to provide a basis for evaluating consequence limiting design features. The ...
This document describes the following aspects: Core meltdown prevention guidelines; severe accidents with core meltdown mitigation; diagnosis and prognosis methods. Figs. (Atomindex citation 24:067128)
The most relevant open questions combined with activity release during hypothetical core meltdown accidents refer to the chemical behavior of the highly reactive elements iodine, cesium, and tellurium, to the release characteristics of the medium-volatile fission and activation products, to the properties of the resulting aerosol particles, and to various ...
A brief historical background and a description of short- and long-term task plan development for effective closure of this important safety issue for the HFIR are given. Short-term aspects deal with Fuel-Coolant-Interaction (FCI) issues experimentation, modeling, and analysis for the flow-blockage-induced steam explosion events in direct support of the SAR. Long-term aspects deal with addressing ...
A rapid water ingress transient, resulting from steam generator tube or tube-sheet failures, could lead to a reactivity insertion and core heatup in the Modular High Temperature Gas-Cooled Reactors. This paper considers the effect of hypothetical rapid an...
This study addresses the type of spent fuel storage accident that has been hypothesized to be the most severe, i.e., one that leads to a complete drainage of the water from the pool. The objective is to analyze the thermal-hydraulics phenomena involved when the storage racks and their contents become exposed to air, and to determine the conditions that could lead to cladding ...
The analysis of spent fuel heatup following a hypothetical accident such as drainage of pool water and the integrity evaluation of spent fuel stored in inert or air gases during long-term storage of spent fuel were performed. The computation based upon a ...
In-vessel core melt progression describes the progression of the state of a reactor core from core uncovery up to reactor vessel melt through in uncovered accidents or through temperature stabilization in accidents recovered by core reflooding. Melt progr...
A severe accident in a nuclear power plant could result in the relocation of large quantities of molten core material onto the lower head of he reactor pressure vessel (RPV). In the absence of inherent cooling mechanisms, failure of the RPV ultimately becomes possible under the combined effects of system pressure and the thermal ...
Four experiments are being conducted in the TREAT facility to investigate the behavior of fission products released from typical LWR fuel overheated to the point of catastrophic cladding degradation. Heatup and steam flow transients are used that simulate the conditions expected in operating power reactors undergoing various types of hypothetical severe ...
The moderator temperature coefficient (MTC) of reactivity is an important parameter in safety analyses for thermal reactors. A positive MTC can exacerbate the severity of heatup transients, while a negative MTC can worsen the severity of cooldown transients. In particular, a strongly negative MTC is the major determinant of the severity of the steamline-break ...
Information is presented concerning accident definition and use of event trees, event tree methodology, potential accidents covered by the reactor safety study, analysis of potential accidents involving the reactor core, and analysis of potential accidents not involving the ...
MOXY is used for the thermal analysis of a planar section of a boiling water reactor (BWR) fuel element during a loss-of-coolant accident (LOCA). The code employs models that describe heat transfer by conduction, convection, and thermal radiation, and heat generation by metal-water reaction and fission product decay. Models are included for considering fuel-rod swelling and ...
The NEPTUN data discussed in this report are from core uncovery (boil-off) experiments designed to investigate the mixture level decrease and the heat up of the fuel rod simulators above the mixture level for conditions simulating core boil-off for a nuclear reactor under small break loss-of-coolant accident conditions. The first ...
The TMI-2 accident resulted in extensive core damage and recent data confirms that the reactor vessel was challenged from molten core materials. A hypothesized TMI accident sencario is presented that consistently explains the TMI data and is also consiste...
The MARCH and HECTR computer codes are used in this study to examine hydrogen production, transport, and combustion in an ice-condenser containment for a number of hypothesized severe accidents. Both degraded-core and core-meltdown accidents are treated. ...
After code damage during a severe accident in a nuclear reactor, the degraded core has to be cooled down and the decay heat should be removed in order to cease the accident progression and maintain a stable state. The cooling of core melt is divided into ...
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 ...
In this paper, two 600 MW{sub e} reactors are compared regarding safety relevant reactivity coefficients, waste-burning capabilities and reactivity swings during burn-up. Furthermore, comparisons of unprotected Loss-of-Flow and Loss-of-Heat Sink calculations are presented. In the first part of this paper, oxide fuels with an inert {sup 92}Mo matrix (occupying a 50% volume fraction) are ...
In case of a severe accident of a light water reactor caused by loss of core cooling capacity, as was observed in the TMI-2 accident, a debris bed consisting of the degraded core materials may be formed as the result of the interaction between the melting...
CORCON is a computer code for modelling the interactions between molten core materials and concrete, such as might occur following a core meltdown accident in a Light Water Reactor. It may also be applied to experiments which simulate such accident condit...
An assessment of accident energetics in LMFBR core-disruptive accidents is given with emphasis on the generic issues of energetic recriticality and energetic fuel-coolant interaction events. Application of a few general behavior principles to the oxide-fueled system suggests that such events are highly unlikely following a postulated ...
For severe accident assessment in a light water reactor (LWR), heat transfer models in a narrow annular gap between the overheated core debris and the reactor pressure vessel (RPV) are important for evaluating RPV integrity and emergency procedures. Some heat transfer models have been proposed as gap cooling CHF (critical heat flux) but local heat fluxes ...
This report evaluates both the positive and negative consequences of adding water to a degraded reactor core during a severe accident. The evaluation discusses the earliest possible stage at which an accident can be terminated and how plant personnel can best respond to undesired results. Specifically discussed are (a) the potential ...
A brief reflection on the current understanding of the unprotected transient overpower accident is presented and an attempt to place it into perspective with regard to the overall question of the FFTF core energetics.
The Core-Melt Source Reduction System (COMSORS) is a new approach to terminate light-water reactor core melt accidents and ensure containment integrity. A special dissolution glass is placed under the reactor vessel. If core debris is released onto the gl...
In this paper an alternate pump trip criterion is described that meets the intent of the U.S. Nuclear Regulatory Commission pump trip requirement (i.e., to minimize primary system mass loss during a small-break loss- of-coolant accident (SBLOCA)) while providing the operators with a valuable tool to differentiate between various generic types of off-nominal transient ...
A postulated high pressure boiloff in a boiling water reactor was analyzed using the SCDAP/RELAP5 computer code. Calculated damage to the core consisted of cladding failure, heavy oxidation, and relocation of molten core material. The slow cladding heatup rate allowed a thick zirconium dioxide shell to develop on the rod and channel ...
A series of four experiments is being conducted at Argonne National Laboratory's TREAT Reactor. They were designed to provide some of the necessary data regarding magnitude and release rates of fission products from degraded fuel pins, physical and chemical characteristics of released fission products, and aerosol formation and transport phenomena. These are in pile experiments, whereby the test ...
A parameter study was carried out for the FRM and the KKE7 core; in this study the effect of the important parameters on the chronological accident sequence and the max. power that was reached was investigated for the startup accident as the limiting case...
French studies have early considered the fact that, despite all the precautions taken, the possibility of severe accidents cannot be absolutely excluded; these accidents include core meltdown and a more or less significant loss, at an early or later stage...
The MARCH code, written at Battelle's Columbus Laboratories for the U.S. Nuclear Regulatory Commission, describes the response of LWR systems to accidents which can result in core meltdown. The calculations are performed from the start of the accident thr...
This paper summarizes experiments on creep rupture of reactor pressure vessel (RPV) lower heads under the thermal and pressure loads of a core meltdown accident. Lower head failure (LHF) is of importance to accident assessment and accident management.
Separate abstracts are included for each of the 22 papers presented concerning the mechanical effects of core accidents, analysis of hypothetical accidents, sodium-air reactions, and accident activity release. (DCC)
One aspect of fast reactor safety analysis consists of calculating the strongly coupled system of physical phenomena which contribute to the reactivity balance in hypothetical whole-core accidents: these phenomena are neutronics, fuel behaviour and heat t...
Five areas of potential accidents have been evaluated for the Argonaut-UTR reactors. They are: insertion of excess reactivity, catastrophic rearrangement of the core, explosive chemical reaction, graphite fire, and a fuel-handling accident.
Mechanistic analyses of transient-under-cooling accidents have led in some cases to a mild initiating phase instead of a direct hydrodynamic disassembly of the core. The fuel is then trapped in the core by the strong mechanical surroundings and blockages ...
The techniques being developed in the United States for analyzing postulated whole-core accidents in LMFBRs are briefly reviewed. The key mechanistic analysis methods are discussed in detail. Important research projects in the area of fission product effe...
The present status of the core-disruptive phase of an accident in a Liquid Metal Fast Breeder Reactor (LMFBR) is reviewed. In addition, a critical survey of methods used in analyzing this phase is presented. Some general conclusions made in the review are...
The various operations required for receipt, handling, defueling and storage of spent Shippingport PWR Core 2 fuel assemblies have been evaluated to determine the potential accidents and their consequences. These operations will introduce approximately 16...
Fission product source terms for loss-of-coolant and core meltdown accidents in light water reactors are reviewed. The results presented in the Reactor Safety Study are summarized, and modifications of these results, due to more recent experimental studies, are described.
For the investigation of the sequences of a hypothetical core meltdown accident, the primary interest is directed towards the integrity of the containment. In case of insecurity, the valid interests become then the probable time period of failure and the ...
Evacuation, sheltering followed by population relocation, and iodine prophylaxis are evaluated as offsite public protective measures in response to nuclear reactor accidents involving core-melt. Evaluations were conducted using a modified version of the R...
Engineering calculations are performed to predict the temperatures of the gas and structures along the primary-coolant system during postulated severe reactor accidents. The calculations cover the time span between the beginning of core uncovery and core ...
Design-basis depressurization accident analyses for the Fort St. Vrain reactor were performed using the FLODIS (Ref. 4) code. The FLODIS code models the active core, side reflector, gas annulus between the core barrel and the PCRV liner, and the PCRV cool...
During a severe accident of a light water reactor, as was observed in the TMI-2 accident, a debris bed which consists of particulate debris would be formed in the core region and/or lower head region. There exsits potential for the debris remelting follow...
An analysis has been performed to determine the principal chemical forms for the structural and fission product elements during a postulated severe core damage accident in tritium powered core in the Savannah River Site (SRS) reactors. These reactors are ...
The main purpose of this work is to analyse the termohydraulic behaviour of emergency cooling water, during reflooding of a PWR core submitted to a postulated loss-of-coolant accident, with the scope of giving the boundary conditions needed to verify fuel...
An analysis has been carried out to assess the potential of a melting attack upon the reactor vessel lower head and incore instrument nozzle penetration weldments during the TMI core relocation event at 224 minutes. Calculations were performed to determine the potential for molten corium to undergo breakup into droplets which freeze and form a debris bed versus impinging upon ...
Estimates activity release from coated particles and heavy metal contamination in MHTGR cores due to thermal transients.
A number of numerical experiments were performed to assess the validity of diffusion theory for calculating the reactivity state of various small core LMFBR disrupted geometries. The disrupted configurations correspond, in general, to various configurations predicted by SAS3A for transient undercooling (TUC) and transient overpower (TOP) accidents for ...
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... REACTORS, TEST REACTORS, DETECTION, REACTOR ACCIDENTS, SIMULATION, INSTRUMENTATION, RESPONSE, NOISE, FREQUENCY ...
Comparisons of measured and calculated LWR fuel rod responses during a reactivity initiated accident test are presented. The results indicate that the computer code, FRAP-T5, adequately calculates the fuel rod behavior up to the time at which the gap closes and provides a good thermal solution up to the time of gross fuel and cladding relocation. Three areas have been ...
The techniques being developed in the United States for analyzing postulated whole-core accidents in LMFBRs are briefly reviewed. The key mechanistic analysis methods are discussed in detail. Important research projects in the area of fission product effects are examined. Some typical results on the role of fission products in ...
The capability of the RELAP5/MOD3 code to validate various transients encountered in RBMK reactor postulated accidents has been assessed. The assessment results include a loss of coolant accident at the inlet of the core pressure tube, the blockage of a pressure tube, and the pressure response of the core cavity to ...
This chapter discusses the development of current state-of-the-art computational methods used in the United States for analysis of core meltdown accidents (Hypothetical Core Disruptive Accidents) in LMFBRs. The emphasis is on the phenomenological basis and numerical methods of the codes SAS, VENUS-II, SIMMER, and ...
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 reports documents work performed the NRC/RES Accident Management Guidance Program to evaluate possible strategies for mitigating the consequences of PWR severe accidents. The selection and evaluation of strategies was limited to the in-vessel phase of the severe accident, i.e., after the initiation of core ...
This paper examines the calculation and experimental investigation of heat removal modes from the core of a shutdown BN-600 reactor. The calculation was based on the thermal balance formula of the plant with ''dry'' steam generators of the third circuit. It is essential to know the speed of the heating up of the power plant in order to ...
The tacit assumption in early severe accident studies was that the melting of a reactor core would result in failure of the reactor pressure vessel and eventually failure of the containment building and release of fission products to the environment. This assumption was shown to be wrong by the TMI-2 accident in which 50% of the ...
Supplement 6 is divided into three principal sections, one responding to questions regarding reflood analysis, another presenting the results of a sensitivity study determining the effect of exposure on peak cladding temperatures, and the third detailing and justifying the RELAP4 model proposed for calculating ECC injection rates from end-of-bypass (EOBY) to ...
For many years, study of whole-core transients and accidents in fast reactors has been a major element of safety research and development programs in the US. Numerous models and computer codes have been developed and validated for use in these studies. Historically, emphasis has been placed on describing the core disruptive ...
The adequacy of the containment of fast reactors has been traditionally evaluated by analyzing the response of the containment to a spectrum of core disruptive accidents. The current approach in the U.S. is to consider fast reactor response to accidents in terms of four lines of assurance (LOAs). Thus, LOA-1 is to prevent ...
In support of probabilistic risk assessment (PRA) studies on the high-temperature gas-cooled reactor (HTGR), an experimental program was conducted for iodine plateout on HTGR primary circuit metals during core heatup conditions. Metal iodine formation and adsorption characteristics were measured primarily for mild steel and to a limited extent for Incoloy ...
Physical processes are investigated that lead to the thermal equilibration of a disrupted liquid metal fast breeder reactor (LMFBR) core following a hypothetical core-disruptive accident (HCDA). Their impact is assessed, particularly as relating to the SIMMER code. The turbulent structure in the core region is ...
Los Alamos Scientific Laboratory reactor safety groups have performed a detailed mechanistic analysis of a best-estimate composite sequence of events for the March 28, 1979, accident at the Three Mile Island - Unit 2 (TMI-2) nuclear reactor. One aspect of that study is analyzed: the core response to the calculated thermal hydraulic transient, including an ...
Estimates of the end-state core configuration have been developed from recent inspection of the lower regions of the TMI-2 core, core support assembly, and lower plenum regions. The inspection data have provided a basis for estimating the extent of damage...
Over the last few years, the Core Shroud of Boiling Water Reactors (BWRs) operating in foreign countries, have developed cracks at weld locations. As a first step for assessment of structural safety of Tarapur Atomic Power Station (TAPS) core shroud, its ...
Molten fuel produced in a core overheating accident is caught by a molten core retention assembly consisting of a horizontal baffle plate having a plurality of openings therein, heat exchange tubes having flow holes near the top thereof mounted in the ope...
This report summarizes the results of developmental work at Rensselaer Polytechnic Institute (RPI) of methods of core modeling for use in the analyses of the progression of accidents that involve core damage in BWRs. Accomplishments include the developmen...
Prior to February 1985 it was the accepted technical opinion that little or no fuel melting had occurred in the TMI-2 core during the accident of March 28, 1979. However, at this time a camera was inserted between the core support cylinder and the reactor...
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 document reports the results of a series of experiments performed to study the interactions expected in a hypothetical core meltdown accident between molten core debris, liquid sodium, and reactor materials in the Clinch River Breeder Reactor. The mat...
