A study of a spectrum of pipe break accidents was made for N-Reactor as part of the accident analysis for the N-Reactor Updated Safety Analysis Report (NUSAR). From this selection of accidents, a list was made of those pipe breaks which would represent th...

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 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 ...

The assumptions, methods of analysis and results of the analysis of single tube accidents which have significant offsite radiological consequences are described. Those accidents described include: pressure tube flow blockage; fuel ejection; and pressure t...

This paper describes the technical approach used in the Modal Study performed by the Lawrence Livermore National Laboratory to evaluate the level of safety provided during the shipment of spent fuel from nuclear power reactors. The evaluation is performed by calculating the responses of representative spent fuel casks to severe highway and railway ...

Simple methods are described for bounding the passive response of a metal fueled liquid-metal cooled reactor to the chilled inlet accident. Calculation of these bounds for a prototype of the Integral Fast Reactor concept shows that failure limits --- eute...

This report describes a study performed by the Lawrence Livermore National Laboratory to evaluate the level of safety provided under severe accident conditions during the shipment of spent fuel from nuclear power reactors. The evaluation is performed using data from real accident histories and using representative truck and rail cask models that likely ...

emergency preparedness responsibility. Browse these EPA Emergency Response subtopics Accidents Accident Preparedness, Accident Prevention, Chemical Accidents, Radiation Accidents...

Team's roles and responsibilities. Browse these EPA Emergency Response subtopics Accidents Accident Preparedness, Accident Prevention, Chemical Accidents, Radiation Accidents...

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 ...

The response of fuel in a boiling water reactor to the rod drop accident (RDA) was studied using the RAMONA-4B computer code. Calculations of this design-basis event has been done conservatively because there was margin to the fuel failure criterion of 17...

This paper describes recent analyses performed by the BWR Severe Accident Technology (BWRSAT) Program at Oak Ridge National Laboratory to estimate the release of debris from the reactor vessel for the unmitigated short-term station blackout accident sequence. Calculations were performed with the BWR Severe Accident ...

Office of Inspector General recommends. Browse these EPA Emergency Response subtopics Accidents Accident Preparedness, Accident Prevention, Chemical Accidents, Radiation Accidents...

about the National Response Plan. Browse these EPA Emergency Response subtopics Accidents Accident Preparedness, Accident Prevention, Chemical Accidents, Radiation Accidents...

The purpose of this calculation is to determine the thermal response of the 44-boiling water reactor (BWR) waste package (WP) to the hypothetical regulatory fire accident. The objective is to calculate the temperature response of the waste package materials to the hypothetical short-term fire ...

... Descriptors : *MENTAL HEALTH, *MOTOR VEHICLE ACCIDENTS, TRAFFIC, RESPONSE, ACCIDENTS, RECORDS, QUESTIONNAIRES ...

This report presents a set of precalculated doses based on a set of postulated accident releases and intended for use in emergency planning and emergency response. Doses were calculated for the PWR (Pressurized Water Reactor) accident categories of the Reactor Safety Study (WASH-1400) using the CRAC ...

about responding to oil spills. Browse these EPA Emergency Response subtopics Accidents Accident Preparedness, Accident Prevention, Chemical Accidents, Radiation Accidents...

about homeland security research. Browse these EPA Emergency Response subtopics Accidents Accident Preparedness, Accident Prevention, Chemical Accidents, Radiation Accidents...

The Modular Accident Analysis Program (MAAP) simulates LWR system response to a severe core accident. Overall, calculations performed with the PWR version of MAAP have compared well with a wide variety of other data. These results have proven MAAP an acceptable tool to support individual plant examinations and ...

Offsite response decision-making methods based on in-plant conditions are developed for use during severe reactor-accident situations. Dose projections are used to eliminate all LWR plant systems except the reactor core and the spent-fuel storage pool from consideration for immediate offsite emergency response during ...

The objective of this calculation is to evaluate the thermal response of the 21-PWR WP (pressurized water reactor waste package) to the regulatory fire event. The scope of this calculation is limited to the two-dimensional waste package temperature calculations to support the waste package design. The information ...

An experimental program to determine the response of LMFBR-type subassemblies to local subassembly accidents caused by pressure loadings is described. Some results are presented and compared with computer calculations.

An experimental program to determine the response of LMFBR-type subassemblies to local subassembly accidents caused by pressure loadings is described. Some results are presented and compared with computer calculations. (JWR)

the Office of Inspector General recommends. Browse these EPA Disasters subtopics Accidents Accident Preparedness, Accident Prevention, Chemical Accidents, Radiation Accidents...

... and recovering from a nuclear accident or incident ... Descriptors : * EMERGENCIES, *NUCLEAR RADIATION, *RESPONSE, *RADIATION ...

Simple methods are described for bounding the passive response of a metal fueled liquid-metal cooled reactor to the chilled inlet accident. Calculation of these bounds for a prototype of the Integral Fast Reactor concept shows that failure limits --- eutectic melting, sodium boiling and fuel pin failure --- are not exceeded. 2 refs., 1 ...

The report describes the results of a series of calculations conducted to investigate the response of BWR Mark III containments to short-term station blackout severe accident sequences. The BWR-LTAS, BWRSAR, and MELCOR codes were employed to conduct quant...

The report describes the results of a series of calculations conducted to investigate the response of BWR Mark II containments to short-term and long-term station blackout severe accident sequences. The BWR-LTAS, BWRSAR, and MELCOR codes were employed to ...

The Three Mile Island nuclear plant (TMI-2) was modeled using the Transient Reactor Analysis Code (TRAC-P1A) and a preliminary calculation, which simulated the initial part of the accident that occurred on March 28, 1979, was performed. The purpose of this calculation was to provide a better understanding of the system ...

The risks associated with the transport of spent nuclear fuel by truck and rail have been reexamined and compared to results published in NUREG-O170 and the Modal Study. The full reexamination considered transport of PWR and BWR spent fuel by truck and rail in four generic Type B spent fuel casks. Because they are typical, this paper presents results only for transport of PWR spent fuel in ...

A scoping study of spent fuel cask transportation accidents was performed to provide the NRC with an assessment of existing information and to recommend, on a priority basis, the additional information that should be obtained to allow specification of increasingly realistic source terms. The scope was limited to the escape of radionuclides from the cask to the environment ...

The MAAP Accident Response System (MARS) is a userfriendly computer software developed to provide management and engineering staff with the most needed insights, during actual or simulated accidents, of the current and future conditions of the plant based on current plant data and its trends. To demonstrate the reliability of the MARS ...

... Radiation on Behavior and the Brain; Nuclear Weapons Accident Response Procedure; Management of Radiation Accidents; Medical Operations ...

This document represents Phase I of a two-phase project. The entire project consists of determining a series of minimum accidents of concern and their associated neutron and photon leakage spectra that may be used to determine Criticality Accident Alarm compliance with ANSI/ANS-8.3. The inadvertent assembly of a critical mass of material presents a ...

This paper describes the methodology used in a quantitative hazard assessment of a nuclear weapon disassembly process. Potential accident sequences were identified using an accident-sequence fault tree based on operational history, weapon safety studies, a hazard analysis team composed of weapons experts, and walkthroughs of the process. The experts ...

Proceedings of the First Symposium on the Nuclear Accident Dosimetry Programi Techniques and Uses are presented. Topics covered include justification of the general system, descriptions of gamma and neutron systems, energy-response calculations of threshold-detector unlts, radiation-leakage calculations for ...

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 ...

Spatial disorientation (SD) is a real risk and one of the main causes of aircraft mishaps. Recent studies about accidents in military aviation (1,2,3,4,5) have calculated that approximately in 32% of the most serious accidents, SD is involved. Not in vain...

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 ...

The ANSI/ANS 8.3 standard defines the minimum accident of concern for criticality alarm systems (CAS) as the accident that can deliver a dose of 20 rads in 60 s at a point 2 m from the surface of the system in free air. A definition this broad presents the criticality specialist with a dilemma when attempting to define critical systems in terms of power ...

This report summarizes the assessments provided by ARAC during the first two weeks after the Chernobyl reactor accident began. Results of this work and measurements made by European countries during that same period show that no major short-term acute health effects would be expected in Europe as a result of this accident. Statistical long-term health ...

This document supports the development and presentation of the following accident scenario in the TWRS Final Safety Analysis Report: Subsurface Leak Remaining Subsurface. The calculations needed to quantify the risk associated with this accident scenario are included within.

This document supports the development and presentation of the following accident scenario in the TWRS Final Safety Analysis Report: Subsurface Leak Remaining Subsurface. The calculations needed to quantify the risk associated with this accident scenario are included within.

This document supports the development and presentation of the following accident scenario in the TWRS Final Safety Analysis Report: Steam Intrusion from Interfacing Systems. The calculations needed to quantify the risk associated with this accident scenario are included within.

This document provides the toxicological dose caculations related to the toxic chemical releases from spill accidents at T Plant Facility.

This document provides the toxicological dose caculations related to the toxic chemical releases from spill accidents at T Plant Facility.

... Accession Number : ADA468875. Title : Response to a Chemical Incident or Accident- Who Is In Charge? Descriptive Note : Master's thesis. ...

The DEFORM-4 module is the segment of the SAS4A Accident Analysis Code System that calculates the fuel pin characterization in response to a steady state irradiation history, thereby providing the initial conditions for the transient calculation. The various phenomena considered include fuel porosity migration, ...

The purpose of this calculation is to determine the thermal response of the 5-defense high level waste (DHLW)/Department of Energy (DOE) codisposal waste package (WP) to the hypothetical fire accident. The objective is to calculate the temperature response of the DHLW glass to the hypothetical ...

A statistical study is presented of the blowdown phase of a design basis accident (double-ended cold leg guillotine break) in the Zion pressurized water reactor. The response surface method was employed to generate a polynomial approximation of the peak clad temperatures calculated by RELAP4/MOD6. The nodalization was a modification of ...

The thermal behavior of fuel rods during simulated accident conditions is extremely sensitive to the heat transfer coefficient which is, in turn, very sensitive to the cladding surface temperature and the fluid conditions. The development of a semianalytical, lumped-parameter fuel rod model which is intended to provide accurate calculations, in a ...

The FRAP-T5 transient fuel rod behavior code, with its automated uncertainty analysis subcode, was used to estimate the uncertainty of best estimate calculations simulating five selected reactor accident and transient events. Code input was supplied by the four US reactor vendors. Response surface methodology was used to estimate ...

A preliminary study of the dynamic response of a large (1200 MWe) pool-type deck structure to an energetic core disassembly accident was performed. The analysis proceeded as two independent, decoupled calculations: containment calculations and deck response calculations. ...

An assessment of many potential HFIR severe accident phenomena was conducted during the HFIR design effort, and many severe accident mitigating features were designed into the plant. These evaluation typically incorporated a bounding'' or highly conservative analysis approach and employed tools and techniques representative of the state ...

The system response code RELAP/MOD2 Idaho National Engineering Laboratory cycle 36.02, with modifications developed by Advanced Nuclear Fuels Corporation (ANF), was used to perform small-break loss-of-coolant accident (SBLOCA) calculations for the Comanche Peak steam electric station (CPSES) unit 1. The ability of the ANF-RELAP code to ...

The Supercritical Carbon Dioxide (S-CO{sub 2}) Brayton Cycle is a promising advanced alternative to the Rankine saturated steam cycle and recuperated gas Brayton cycle for the energy converters of specific reactor concepts belonging to the U.S. Department of Energy Generation IV Nuclear Energy Systems Initiative. A new plant dynamics analysis computer code has been developed for simulation of the ...

This paper describes fluid-structure-interaction and structure response analyses of a reactor vessel subjected to loadings associated with postulated accidents, using the improved hybrid Lagrangian-Eulerian code ALICE-II. The objective of the present analyses is to study the cover response and potential for missile generation in ...

This paper presents the results and insights gained from MELCOR analyses of two severe accident scenarios, a Loss of Coolant Accident (LOCA) and a Station Blackout (TMLB) in Oconee, a Babcock & Wilcox (B&W) designed PWR with a large dry containment, and comparisons with Source Term Code Package (STCP) calculations of the same ...

This paper presents the results and insights gained from MELCOR analyses of two severe accident scenarios, a Loss of Coolant Accident (LOCA) and a Station Blackout (TMLB) in Oconee, a Babcock Wilcox (B W) designed PWR with a large dry containment, and comparisons with Source Term Code Package (STCP) calculations of the same ...

Boiling water reactors (BWRs) incorporate many unique structural features that make their expected response under severe accident conditions very different from that predicted in the case of pressurized water reactor accident sequences. The effect of the BWR procedural and structural differences upon the progression of a severe ...

Methods are described for determining the fully coupled neutronic/hydrodynamic response of fast burst reactors (FBR) under disruptive accident conditions. Two code systems, PAD (1 -D Lagrangian) and NIKE-PAGOSA (3-D Eulerian) were used to accomplish this. This is in contrast to the typical methodology that computes these responses by ...

... Response Plan (FRERP), and DoD Accident Response Group (ARG ... MILITARY PUBLICATIONS, *MANUALS, NUCLEAR RADIATION, RESPONSE. ...

The results are presented of Phase 1 of the APRICOT (Analysis of PRImary COntainment Transients) Program. APRICOT is an international cooperative activity for comparison and benchmarking of computational methods used to analyze LMFBR structural response to pressure loads from CDA's (Core Disruptive Accidents). Phase 1 involved parallel ...

This paper presents the thermal-hydraulic response of the Oconee-1 plant to several small-break loss-of-coolant accidents and a description of the TRAC-PF1 plant model of Oconee-1 used in support of the Nuclear Regulatory Commission directed pressurized thermal shock study. The small-break transients investigated included a stuck-open ...

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...

Three generic sheltering/relocation strategies are identified and discussed. They are: population relocation only (no specific sheltering response initiated); sheltering at location following by relocation; and preferential sheltering followed by relocation. Shielding factors representative of these strategies are calculated, and the adequacy of using ...

APRICOT (Analysis of Primary Containment Transients) is a cooperative activity for comparison and benchmarking of computational methods used to analyze LMFBR (Liquid Metal Fast Breeder Reactor) structural response to pressure loads from HCDA's (Hypothetical Core Disruptive Accidents). Independent experts review the ...

A quantitative analysis was made of the response of plate fuel elements in the beyond DNB (direct nucleate boiling)'' heat transfer region. Two solutions were derived, one with no scram and the other with a scram at0.8 sec after start of accident, and a forced DNB 0.5 sec after accident is assumed in both ...

One of the most important uses of dose assessment models in response to accidents at nuclear facilities is to help provide guidance to emergency response managers for identifying, and mitigating, the consequences of an accident once the accident has been ...

This paper describes and evaluates operational experiences with the Accident Response Mobile Manipulation System (ARMMS) during simulated accident site salvage operations which might involve nuclear weapons. The ARMMS is based upon a teleoperated mobility...

Feb 10, 2003 ... So as it effects everything from launch pads to analyses that are ... in a specific activity with the Columbia Accident Investigation Board. ...

... The chief executive is responsible to ensure that each manager is provided with a periodic summary of his unit's safety performance. The Accident ...

This document includes the calculations performed to quantify the risk associated with the unmitigated and mitigated accident scenarios described in the TWRS FSAR for the accident analysis titled: Surface Leaks Resulting in Pool.

This document includes the calculations performed to quantify the risk associated with the unmitigated and mitigated accident scenarios described in the TWRS FSAR for the accident analysis titled: Surface Leaks Resulting in Pool.

This document includes the calculations performed to quantify the risk associated with the unmitigated and mitigated accident scenarios described in the TWRS FSAR for the accident analysis titled: Subsurface Leaks Resulting in Pool.

A methodology of transient and accident analyses able to carried out calculations for all transients and accidents required to support operation and operation licensing of Angra-1 reactor reload, is presented. (Atomindex citation 18:076910)

A number of conceivable reactivity accidents were analyzed, using conservatively pessimistic assumptions and approximations, to permit evaluation of reactor safety. Most of the calculations, which are described in detail, were performed by a digital kinetics program, MURGATROYD. Some analog analyses were also made. None of the ...

A system has been developed to assess radiation dose distribution inside the body of exposed persons in a radiological accident by utilising radiation transport calculation codes-MCNP and MCNPX. The system consists mainly of two parts, pre-processor and post-processor of the radiation transport calculation. Programs for the ...

The Soviet designed VVER-440 model V230 and VVER-440 model V213 reactors do not use full containments to mitigate the effects of accidents. Instead, these VVER-440 units employ a sealed set of interconnected compartments, collectively called the accident localization system (ALS), to reduce the release of radionuclides to the atmosphere during ...

The source range monitor (SRM) data recorded during the first 4 h of the Three Mile Island Unit 2 (TMI-2) accident following reactor shutdown were analyzed. An effort to simulate the actual SRM response was made by performing a series of neutron transport calculations. Primary emphasis was placed on simulating the changes in SRM ...

At 0820 PST on 28 March 1979, the Department of Energy's Emergency Operations Center advised the Atmospheric Release Advisory Capability (ARAC) that the Three Mile Island nuclear power plant in Harrisburg, Pennsylvania, had experienced an accident some four hours earlier, resulting in the atmospheric release of xenon-133 and krypton-88. This report describes ...

For a quarter of a century the Federal Government and the nuclear industry have deliberately deceived the American public about the risks of nuclear power. Facts have been systematically withheld, distorted, and obscured, and calculations have been deliberately biased in order to present nuclear power in an unrealistically favorable light. Most persistent and flagrant have ...

The source range monitor (SRM) data recorded during the first 4 hours of the Three Mile Island Unit No. 2 (TMI-2) accident following reactor shutdown were analyzed. An effort to simulate the actual SRM response was made by performing a series of neutron transport calculations. Primary emphasis was placed on simulating the changes in ...

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 ...

... ANALYSIS OF NASA'S POST-CHALLENGER RESPONSE AND RELATIONSHIP TO THE COLUMBIA ACCIDENT AND INVESTIGATION by ...

This paper describes fluid-structure-interaction and structure response analyses of a reactor vessel subjected to loadings associated with postulated accidents, using the hybrid Lagrangian-Eulerian code ALICE-II. This code has been improved recently to accommodate many features associated with innovative designs of reactor vessels. ...

This work quantifies the expected dose rates at a series of criticality alarm locations due to several postulated criticality accidents at the Westinghouse Environmental MANAGEMENT COMPANY OF OHIO (WEMCO) Fernald site. One- and two-dimensional discrete- ordinates calculations were performed for seven different shielding configurations using leakage spectra ...

Gap conductance is a significant factor affecting the stored energy in a fuel rod at the beginning of a hypothetical accident sequence, as well as the thermal and mechanical response of the fuel rod during the accident. Additional well-characterized experimental results are needed to evaluate and improve the current analytical ...

The thermal effect of a loss of flow accident and afterheat to the MARS blanket are investigated. The temperature response of the first wall, as well as the whole blanket, is calculated with a finite difference method. For a loss of flow accident, the plasma has to be quenched within 10 to 35 seconds, beyond which ...

Selected experimental data and results calculated from experimental data obtained from the Semiscale Mod-1 PWR blowdown heat transfer test series are analyzed. These tests were designed primarily to provide information on the core thermal response to a loss-of-coolant accident. The data are analyzed to determine the effect of core flow ...

We have made Monte Carlo calculations of the scintillation spectrometer response for the photon field from a cloud of contaminated air after selected scenarios of a nuclear power plant accident. Calculations (using MCNP5 code-X-5 Monte Carlo Team, 2005) were performed for 36 main energy lines of the expected ...

In the first few hours following the TMI-2 accident large variations (factors of 10-100) in the source range (SR) detector response were observed. The purpose of this analysis was to quantify the various effects which could contribute to these large variations. The effects evaluated included the transmission of neutrons and photons from the core to ...

The system response in a transportation accident environment is an element to be considered in an overall Transportation System Risk Assessment (TSRA) framework. The system response analysis uses the accident conditions and the subsequent accident progression analysis to develop the ...

This report presents the results of a DOE-sponsored assessment of nuclear accident response resources. It identifies the mobile resources that could be required to respond to different types of nuclear accidents including major ones like TMI-2, identifies the resources currently available and makes recommendations for the design and ...

...safety reporting and incident/accident response. 91.1021 ...Ownership Operations Program Management § 91.1021 Internal safety reporting and incident/accident response. (a) Each...to an aviation...

This study describes the predicted response of the Browns Ferry Nuclear Plant to a postulated complete failure of plant control air. The failure of plant control air cascades to include the loss of drywell control air at Units 1 and 2. Nevertheless, this is a benign accident unless compounded by simultaneous failures in the turbine-driven high pressure ...

This work presents a quantitative assessment of the impact that the emergency operating instructions (EOIs) of operating nuclear power plants may have on the frequency of severe accident sequences. Frequencies of such sequences were calculated based on the operator actions required by the EOIs of a reference BWR plant and compared to those resulting from ...

Severe accident phenomena pertinent to the heavy-water-moderated production reactors of the US Department of Energy are being studied in the Severe Accident Analysis Program (SAAP) at the Savannah River Site. The SAAP has sought to define the behavior of the Savannah River reactors in accident scenarios involving significant fuel ...

The temperature response of three generic surface models in each of three locations in an ice condenser containment building were calculated assuming a hydrogen deflagration event and using the HECTR code. The intent of using the three generic surfaces was to conservatively represent surfaces of various types of safety equipment. Analyses were performed ...

Reduced instrument responses are presented for Thermal-Hyraulic Test Facility (THTF) Test 3.06.6B. This test was conducted by members of the Oak Ridge National Laboratory Pressurized-Water-Reactor (PWR) Blowdown Heat Transfer (BDHT) Separate-Effects Program on August 29, 1980. The objective of the program was to investigate heat transfer phenomena believed to occur in ...

Report gives a detailed analysis of fatal haulage accidents in coal mines during first 6 months of 1951, lists causes of accidents in both bituminous and anthracite mines, points out the responsibility for such accidents, and gives recommendations for the prevention of accidents.

In an attempt to understand the dynamics of extra severe transportation accidents and to evaluate state-of-the-art computational techniques for predicting the dynamic response of shipping casks involved in vehicular system crashes, the Environmental Control Technology Division of ERDA undertook a program with Sandia to investigate these areas. The program ...

In an attempt to understand the dynamics of extra severe transportation accidents and to evaluate state-of-the-art computational techniques for predicting the dynamic response of shipping casks involved in vehicular system crashes, the Environmental Control Technology Division of ERDA undertook a program with Sandia to investigate these areas. This ...

This document supports the development and presentation of the following accident scenario in the TWRS Final Safety Analysis Report (FSAR): Steam Intrusion From Interfacing Systems. The calculations needed to quantify the risk associated with this accident scenario are included in the following sections to aid in the understanding of ...

This volume is the first of a two-volume document that summarizes a joint project conducted by the US Nuclear Regulatory Commission and the European Commission to assess uncertainties in the MACCS and COSYMA probabilistic accident consequence codes. These codes were developed primarily for estimating the risks presented by nuclear reactors based on postulated frequencies and ...

SMART calculates early offsite consequences from nuclear reactor accidents. Once the air and ground concentrations of the radionuclide are estimated, the early dose to an individual is calculated via three pathways: cloudshine, short term groundshine, and inhalation.

Work is underway at Oak Ridge National Laboratory (ORNL) to incorporate certain models of the Boiling Water Reactor Severe Accident Response (BWRSAR) code into a local version of MELCOR. Specifically, the BWR lower plenum debris bed and bottom head response models taken from BWRSAR are being tested within the local MELCOR code ...

The effects of recirculation loops in the ice bed region of an ice-condenser containment are described in this paper. HECTR, a lumped-parameter computer code, was used for this investigation. Although best-estimate calculations are not possible, sensitivity calculations can be performed to bound the effects and provide qualitative insights. Results of ...

Evaluation of the adequacy of the emergency cooling water addition system (ECWA system) requires analysis of the postulated accidents for which the system is required to function. Analysis of these accidents requires a knowledge of the amount of the ECW added that goes to the fuel; both the total amount to the fuel and the amount to the central fuel ...

This document supports the development and presentation of the following accident scenario in the TWRS Final Safety Analysis Report: Fire in Contaminated Area. The calculations needed to quantify the risk associated with this accident scenario are included within.

This document supports the development and presentation of the following accident scenario in the TWRS Final Safety Analysis Report: Fire in Contaminated Area. The calculations needed to quantify the risk associated with this accident scenario are included within.