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in subjects: "Meltdown"

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Accident source terms for light-water nuclear power plants using high-burnup or MOX fuel.

Representative accident source terms patterned after the NUREG-1465 Source Term have been developed for high burnup fuel in BWRs and PWRs and for MOX fuel in a PWR with an ice-condenser containment. These source terms have been derived using nonparametric order statistics to develop distributions for the timing of radionuclide release during four accident phases and for release fractions of nine chemical classes of radionuclides as calculated with the MELCOR 1.8.5 accident analysis computer cod… more

Date: January 1, 2011

Creator: Salay, Michael (U.S. Nuclear Regulatory Commission, Washington, D.C.); Gauntt, Randall O.; Lee, Richard Y. (U.S. Nuclear Regulatory Commission, Washington, D.C.); Powers, Dana Auburn & Leonard, Mark Thomas

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Iodine transport analysis in the ESBWR.

A simplified ESBWR MELCOR model was developed to track the transport of iodine released from damaged reactor fuel in a hypothesized core damage accident. To account for the effects of iodine pool chemistry, radiolysis of air and cable insulation, and surface coatings (i.e., paint) the iodine pool model in MELCOR was activated. Modifications were made to MELCOR to add sodium pentaborate as a buffer in the iodine pool chemistry model. An issue of specific interest was whether iodine vapor removed… more

Date: March 1, 2009

Creator: Kalinich, Donald A.; Gauntt, Randall O.; Young, Michael Francis & Longmire, Pamela

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

HISTORICAL AMERICAN ENGINEERING RECORD - IDAHO NATIONAL ENGINEERING AND ENVIRONMENTAL LABORATORY, TEST AREA NORTH, HAER NO. ID-33-E

Test Area North (TAN) was a site of the Aircraft Nuclear Propulsion (ANP) Project of the U.S. Air Force and the Atomic Energy Commission. Its Cold War mission was to develop a turbojet bomber propelled by nuclear power. The project was part of an arms race. Test activities took place in five areas at TAN. The Assembly & Maintenance area was a shop and hot cell complex. Nuclear tests ran at the Initial Engine Test area. Low-power test reactors operated at a third cluster. The fourth area was for… more

Date: February 1, 2005

Creator: Stacy, Susan & Gilbert, Hollie K.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Transient Temperature and Pressure in the Reactor Room During a Core Meltdown Accident

The purpose of this numerical model is to determine the optimum ventilation exhaust flow rate for the reactor room. The influence of steam produced in the reactor vessel, on the reactor room pressures, is included in the model. A parametric study of the affect of various steam mass flow rates is included in this document. The affect of steam on the conditions in the reactor room is significant at modest flow rates.

Date: July 17, 2001

Creator: Shadday, M.A.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Models for the Configuration and Integrity of Partially Oxidized Fuel Rod Cladding at High Temperatures

Models were designed to resolve deficiencies in the SCDAP/RELAP5/MOD3.2 calculations of the configuration and integrity of hot, partially oxidized cladding. These models are expected to improve the calculations of several important aspects of fuel rod behavior. First, an improved mapping was established from a compilation of PIE results from severe fuel damage tests of the configuration of melted metallic cladding that is retained by an oxide layer. The improved mapping accounts for the relocat… more

Date: January 1, 1999

Creator: Siefken, L. J.

Item Type: Report

Partner: UNT Libraries Government Documents Department

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Creep failure of a reactor pressure vessel lower head under severe accident conditions

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 heat-up of the lower head. Sandia National Laboratories has performed seven experiments at 1:5th scale simulating creep failure of a RPV lower head. This paper describe… more

Date: August 1998

Creator: Pilch, M. M.; Ludwigsen, J. S.; Chu, T. Y. & Rashid, Y. R.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

A review of experiments and results from the transient reactor test (TREAT) facility.

The TREAT Facility was designed and built in the late 1950s at Argonne National Laboratory to provide a transient reactor for safety experiments on samples of reactor fuels. It first operated in 1959. Throughout its history, experiments conducted in TREAT have been important in establishing the behavior of a wide variety of reactor fuel elements under conditions predicted to occur in reactor accidents ranging from mild off normal transients to hypothetical core disruptive accidents. For much of… more

Date: July 28, 1998

Creator: Deitrich, L. W.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Quick look data report for COMET Test U2

Investigations are underway at Forschungszentrum Karlsruhe (FZK) addressing methods to terminate and stabilize a core melt accident situation ex-vessel. In this approach, the molten core-concrete interaction (MCCI) begins erosion of the concrete, and after erosion proceeds to some modest depth, it exposes and unseals an array of tubes. The tubes are connected to a water reservoir pressurized by static water head. Upon unsealing, the tubes direct a flow of water into the bottom of the corium lay… more

Date: January 8, 1998

Creator: Farmer, M. T.; Spencer, B. W.; Kilsdonk, D. J. & Aeschlimann, R. W.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Developmental assessment of the SCDAP/RELAP5 code

The development and assessment of the late-phase damage progression models in the current version (designated MOD3.2) of the SCDAP/RELAP5 code are described. The SCDAP/RELAP5 code is being developed at the Idaho National Engineering and Environmental Laboratory under the primary sponsorship of the US Nuclear Regulatory Commission (NRC) to provide best-estimate transient simulations of light water reactor coolant systems (RCS) during severe accident conditions. Recent modeling improvements made … more

Date: December 31, 1997

Creator: Harvego, E. A.; Slefken, L. J. & Coryell, E. W.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Interpretation of the XR2-1 experiment and characteristics of the BWR lower plenum debris bed

The Ex-Reactor (XR) experiments have been conducted to advance the understanding of BWR severe accident melt progression events. The XR2-1 experiment addresses the fate of the initial large (code-predicted) movements of molten metals from the upper core to the lower core and core plate region. For this question, which has ramifications for blockage formation in the core region, the XR2-1 test results provide significant and perhaps definitive insights. Nevertheless, some events that occurred du… more

Date: November 1997

Creator: Hodge, S. A. & Ott, L. J.

Item Type: Article

Partner: UNT Libraries Government Documents Department

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Models and correlations of the DEBRIS Late-Phase Melt Progression Model

The DEBRIS Late Phase Melt Progression Model is an assembly of models, embodied in a computer code, which is designed to treat late-phase melt progression in dry rubble (or debris) regions that can form as a consequence of a severe core uncover accident in a commercial light water nuclear reactor. The approach is fully two-dimensional, and incorporates a porous medium modeling framework together with conservation and constitutive relationships to simulate the time-dependent evolution of such re… more

Date: September 1, 1997

Creator: Schmidt, R.C. & Gasser, R.D.

Item Type: Report

Partner: UNT Libraries Government Documents Department

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Development potential for thermal reactors and their fuel cycles

The advantages of molten salt reactors (MSRs) for power production are very briefly described in this paper. The MSRs considered are those with on-line fuel processing, external cooling, and fluoride salt separation. Characteristics noted include lack of meltdown potential, small radioactive source terms, and complete burnup of fissile material. The burnup capability of MSRs would allow them to be used to dispose of plutonium while producing energy. 8 refs.

Date: August 1, 1997

Creator: Dodds, H. L. & Gat, U.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Experiments on interactions between zirconium-containing melt and water (ZREX): Hydrogen generation and chemical augmentation of energetics

The results of the first data series of experiments on interactions between zirconium-containing melt and water are described. These experiments involved dropping 1-kg batches of pure zirconium or zirconium-zirconium dioxide mixture melt into a column of water. A total of nine tests were conducted, including four with pure zirconium melt and five with Zr-ZrO{sub 2} mixture melt. Explosions took place only in those tests which were externally triggered. While the extent of zirconium oxidation in… more

Date: August 1, 1997

Creator: Cho, D. H.; Armstrong, D. R.; Gunther, W. H. & Basu, S.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Final results of the XR2-1 BWR metallic melt relocation experiment

This report documents the final results of the XR2-1 boiling water reactor (BWR) metallic melt relocation experiment, conducted at Sandia National Laboratories for the U.S. Nuclear Regulatory Commission. The objective of this experiment was to investigate the material relocation processes and relocation pathways in a dry BWR core following a severe nuclear reactor accident such as an unrecovered station blackout accident. The imposed test conditions (initial thermal state and the melt generatio… more

Date: August 1, 1997

Creator: Gauntt, R.O. & Humphries, L.L.

Item Type: Report

Partner: UNT Libraries Government Documents Department

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Late-phase melt progression experiment: MP-2. Results and analysis

In-pile experiments addressing late-phase processes in Light Water Reactors (LWRs) were performed in the Annular Core Research Reactor (ACRR) at Sandia National Laboratories. Melt Progression (MP) experiments were designed to provide information to develop and verify computer models for analysis of LWR core damage in severe accidents. Experiments examine the formation and motion of ceramic molten pools in disrupted reactor core regions. The MP-2 experiment assembly consisted of: (1) a rubble be… more

Date: May 1997

Creator: Gasser, R. D.; Gauntt, R. O. & Bourcier, S. C.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

COMSORS: A light water reactor chemical core catcher

The Core-Melt Source Reduction System (COMSORS) is a new approach to terminate lightwater reactor (LWR) core-melt accidents and ensure containment integrity. A special dissolution glass made of lead oxide (PbO) and boron oxide (B{sub 2}O{sub 3}) is placed under the reactor vessel. If molten core debris is released onto the glass, the following sequence happens: (1) the glass absorbs decay heat as its temperature increases and the glass softens; (2) the core debris dissolves into the molten glas… more

Date: February 24, 1997

Creator: Forsberg, C. W.; Parker, G. W.; Rudolph, J. C.; Osborne-Lee, I. W. & Kenton, M. A.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Experiments to investigate direct containment heating phenomena with scaled models of the Calvert Cliffs Nuclear Power Plant

The Surtsey Test Facility is used to perform scaled experiments simulating High Pressure Melt Ejection accidents in a nuclear power plant (NPP). The experiments investigate the effects of direct containment heating (DCH) on the containment load. The results from Zion and Surry experiments can be extrapolated to other Westinghouse plants, but predicted containment loads cannot be generalized to all Combustion Engineering (CE) plants. Five CE plants have melt dispersal flow paths which circumvent… more

Date: February 1997

Creator: Blanchat, T. K.; Pilch, M. M. & Allen, M. D.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

An assessment of the CORCON-MOD3 code. Part 1: Thermal-hydraulic calculations

This report deals with the subject of CORCON-Mod3 code validation (thermal-hydraulic modeling capability only) based on MCCI (molten core concrete interaction) experiments conducted under different programs in the past decade. Thermal-hydraulic calculations (i.e., concrete ablation, melt temperature, melt energy, concrete temperature, and condensible and non-condensible gas generation) were performed with the code, and compared with the data from 15 experiments, conducted at different scales us… more

Date: September 1, 1996

Creator: Strizhov, V.; Kanukova, V.; Vinogradova, T.; Askenov, E. & Nikulshin, V.

Item Type: Report

Partner: UNT Libraries Government Documents Department

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Termination of light-water reactor core-melt accidents with a chemical core catcher: the core-melt source reduction system (COMSORS)

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 glass, the glass melts and the debris dissolves into the molten glass, thus creating a homogeneous molten glass. The molten glass, with dissolved core debris, spreads into a wide pool, distributing the heat for removal by radiation to the reactor… more

Date: September 1, 1996

Creator: Forsberg, C. W.; Parker, G. W.; Rudolph, J. C.; Osborne-Lee, I. W. & Kenton, M. A.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Consequences of the Chernobyl accident for the natural and human environments

In the ten years since the Chernobyl accident, an enormous amount of work has been done to assess the consequences to the natural and human environment. Although it is difficult to summarize such a large and varied field, some general conclusions can be drawn. This background paper includes the main findings concerning the direct impacts of radiation on the flora and fauna; the general advances of knowledge in the cycling of radionuclides in natural, seminatural and agricultural environments; s… more

Date: July 1, 1996

Creator: Dreicer, M.; Aarkog, A.; Alexakhin, R.; Anspaugh, L.; Arkhipov, N. P. & Johansson, K. -J.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

The characterization and risk assessment of the `Red Forest` radioactive waste burial site at Chernobyl Nuclear Power Plant

The `Red Forest` radioactive waste burials created during emergency clean-up activities at Chernobyl Nuclear Power Plant represent a serious source of radioactive contamination of the local ground water system with 9OSr concentration in ground water exceeding the drinking water standard by 3-4 orders of magnitude. In this paper we present results of our hydrogeological and radiological `Red Forest` site characterization studies, which allow us to estimate 9OSr subsurface migration parameters. W… more

Date: June 1, 1996

Creator: Bungai, D.A.; Skalskij, A.S.; Dzhepo, S.P. & Waters, R.D.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Natural convection in a uniformly heated pool

In the event of a core meltdown accident, to prevent reactor vessel failure from molten corium relocation to the reactor vessel lower head, the establishment of a coolable configuration has been proposed by flooding with water the reactor cavity. In Reference 3, it was shown that for the heavy-water new production reactor (NPW-HWR) design, this strategy, e.g., the rejection of decay heat to a containment decay heat removal system by boiling of water in the reactor cavity, could keep the reactor… more

Date: May 1, 1996

Creator: Tzanos, C.P.

Item Type: Article

Partner: UNT Libraries Government Documents Department

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Eliminating the possibility at Chernobyl 4 of recriticality with positive feedback

We have recently published an article in which we discuss means by which plutonium and other fissile material stored underground could reach criticality with positive feedback and therefore explosive potential. The Chernobyl rubble involving hundreds of tons of material is similar in some respects to the systems analyzed in the paper, and the practices there to control criticality may well increase the probability of a second event at Chernobyl 4. This paper explores the Chernobyl situation and… more

Date: April 29, 1996

Creator: Bowman, C. D.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Resolution of the direct containment heating issue for all Westinghouse plants with large dry containments or subatmospheric containments

This report uses the scenarios described in NUREG/CR-6075 and NUREG/CR-6075, Supplement 1, to address the direct containment heating (DCH) issue for all Westinghouse plants with large dry or subatmospheric containments. DCH is considered resolved if the conditional containment failure probability (CCFP) is less than 0.1. Loads versus strength evaluations of the CCFP were performed for each plant using plant-specific information. The DCH issue is considered resolved for a plant if a screening ph… more

Date: February 1996

Creator: Pilch, M. M.; Allen, M. D. & Klamerus, E. W.

Item Type: Report

Partner: UNT Libraries Government Documents Department