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A scoping study of fission product transport from failed fuel during N Reactor postulated accidents

Description: This report presents a scoping study of cesium, iodine, and tellurium behavior during a cold leg manifold break in the N Reactor. More detail about fission product behavior than has previously been available is provided and key parameters that control this behavior are identified. The LACE LA1 test and evidence from the Power Burst Facility Severe Fuel Damage tests are used to test the key model applied to determine aerosol behavior. Recommendations for future analysis are also provided. The primary result is that most of the cesium, iodine, and tellurium remains in the molten uranium fuel. Only 0.0035 of the total inventory is calculated to be released. Condensation of the most of the species of cesium and iodine that are released is calculated, with 0.998 of the released cesium and iodine condensing in the spacers and upstream end of the connector tubes. Most of the tellurium that is released condenses, but the chemical reaction of tellurium vapor with surfaces is also a major factor in the behavior of this element.
Date: January 1, 1988
Creator: Hagrman, D.L.
Partner: UNT Libraries Government Documents Department

Code Development and Analysis Program: cladding mechanical limits (CMLIMT)

Description: Revised models are described for cladding mechanical limits. The update incorporates important new data from several Nuclear Regulatory Commission and German experimental programs and defines a single physically reasonable failure criterion for cladding under tensile stress. Alternate simplified expressions are also derived for use in obtaining estimates of typical cladding shape after burst.
Date: May 1, 1979
Creator: Hagrman, D. L.
Partner: UNT Libraries Government Documents Department

Fuel thermal conductivity (FTHCON). Status report. [PWR; BWR]

Description: An improvement of the fuel thermal conductivity subcode is described which is part of the fuel rod behavior modeling task performed at EG and G Idaho, Inc. The original version was published in the Materials Properties (MATPRO) Handbook, Section A-2 (Fuel Thermal Conductivity). The improved version incorporates data which were not included in the previous work and omits some previously used data which are believed to come from cracked specimens. The models for the effect of porosity on thermal conductivity and for the electronic contribution to thermal coductivity have been completely revised in order to place these models on a more mechanistic basis. As a result of modeling improvements the standard error of the model with respect to its data base has been significantly reduced.
Date: February 1, 1979
Creator: Hagrman, D. L.
Partner: UNT Libraries Government Documents Department

MATPRO-Version 11: a handbook of materials properties for use in the analysis of light water reactor fuel rod behavior

Description: This handbook describes the materials properties correlations and computer subcodes (MATPRO-Version 11) developed for use with various LWR fuel rod behavior analytical programs at the Idaho National Engineering Laboratory. Formulations of fuel rod material properties, which are generally semiempirical in nature, are presented for uranium dioxide and mixed uranium--plutonium dioxide fuel, zircaloy cladding, and fill gas mixtures.
Date: February 1, 1979
Creator: Hagrman, D.L. & Reymann, G.A. (comps.)
Partner: UNT Libraries Government Documents Department

MELCOR aerosol transport module modification for NSSR-1

Description: This report describes modifications of the MELCOR computer code aerosol transport module that will increase the accuracy of calculations for safety analysis of the International Thermonuclear Experimental Reactor (ITER). The modifications generalize aerosol deposition models to consider gases other than air, add specialized models for aerosol deposition during high speed gas flows in ducts, and add models for resuspension of aerosols that are entrained in coolants when these coolants flash. Particular attention has been paid to the adhesion of aerosol particles once they are transported to duct walls. The results of calculations with the modified models have been successfully compared to data from Light Water Reactor Aerosol Containment Experiments (LACE) conducted by an international consortium at Hanford, Washington.
Date: March 1, 1996
Creator: Merrill, B.J. & Hagrman, D.L.
Partner: UNT Libraries Government Documents Department

Behavior of four PWR rods subjected to a simulated loss-of-coolant accient in the power burst facility

Description: Cladding deformation characteristics resulting from the first nuclear blowdown tests (LOC-11) conducted in the Power Burst Facility (PBF) are emphasized in this paper. The objective of the LOC-11 tests was to obtain data on the thermal, mechanical, and materials behavior of pressurized and unpressurized fuel rods when exposed to a blowdown similiar to that expected in a pressurized water reactor (PWR) during a hypothesized double-ended cold-leg break. The test hardware consisted of four separately shrouded fresh fuel rods of PWR 15 x 15 design. Initial plenum pressures ranged from atmospheric to 4.8 MPa (representative of end-of-life). During LOC-11C, the four fuel rods were subjected to 6.5 hours of nuclear operation at approximately 67 kW/m average rod power to cause decay heat build-up. Just before the start of blowdown, cladding surface temperatures were about 620 K and fuel centerline temperatures were in the 2500 to 2600 K range. During the 30-second blowdown transient, CHF occurred 2 seconds after initiation. Fuel centerline temperature dropped continuously, while cladding surface temperatures increased. Maximum cladding temperatures of 1030 to 1050 K occurred 15 seconds into the transient. Posttest destructive examination revealed cladding microstructures and oxide thicknesses consistent with the measured cladding temperatures. The cladding surface thermocouples did not appreciably affect cladding temperature distributuion (fin cooling effect) in the vicinity of the thermocouples.
Date: January 1, 1978
Creator: Cook, T.F.; Hagrman, D.L. & Sepold, L.K.
Partner: UNT Libraries Government Documents Department

Volatilization from PCA steel alloy

Description: The mobilizations of key components from Primary Candidate Alloy (PCA) steel alloy have been measured with laboratory-scale experiments. The experiments indicate most of the mobilization from PCA steel is due to oxide formation and spalling but that the spalled particles are large enough to settle rapidly. Based on the experiments, models for the volatization of iron, manganese, and cobalt from PCA steel in steam and molybdenum from PCA steel in air have been derived.
Date: August 1, 1996
Creator: Hagrman, D.L.; Smolik, G.R.; McCarthy, K.A. & Petti, D.A.
Partner: UNT Libraries Government Documents Department

FRAP-GCFR: a code for the transient analysis of gas-cooled fast reactors

Description: The fuel rod analysis computer program (FRAP-T) has been modified to analyze transient fuel rod behavior in the gas cooled fast reactor (GCFR). Major features modeled include the effects of helium coolant, ribbed cladding, pressure equalization system, and 316 stainless-steel material properties. The code capabilities include models of elastic-plastic-creep deformation behavior, two-dimensional heat conduction, rod ballooning, and transient axial gas flow. The models are particularly relevant to GCFR transient analysis in that partial or full blockage of the pressure equalization system could cause a significant pressure differential to exist during a helium depressurization, the GCFR design basis accident. An analysis of the design basis accident illustrating the predictive capabilities of the code is presented.
Date: January 1, 1979
Creator: Peck, S O; Hagrman, D L & Bohn, M P
Partner: UNT Libraries Government Documents Department

In situ vitrification model development and implementation plan

Description: This document describes the In Situ Vitrification (ISV) Analysis Package being developed at the INEL to provide analytical support for (ISV) safety analysis and treatment performance predictions. Mathematical models and features which comprise this analysis package are presented and the proposed approach to model development and implementation is outlined. The objective of this document is two fold: to define preliminary design information and modeling objectives so that ISV modeling personnel can effectively modify existing models and formulate new models which are consistent with the objectives of the ISV treatability study and to provide sufficient technical information for internal and external reviewers to detect any shortcomings in model development and implementation plans. 27 refs., 17 figs., 3 tabs.
Date: August 1, 1990
Creator: MacKinnon, R.J.; Murray, P.E.; Johnson, R.W.; Hagrman, D.L.; Slater, C.E. & Marwil, E.S.
Partner: UNT Libraries Government Documents Department

Cobalt release from PCA steel during possible fusion reactor accidents

Description: Possible accident scenarios for a fusion reactor include breaches in the vacuum or cooling system. Intruding air or steam could react with structural or plasma facing materials, possibly mobilizing radioactive isotopes. Safety assessments must consider the early dose at the site boundary from the release of these activated materials. Previous calculations have indicated that cobalt isotopes dominate dose calculations for designs using stainless steel. Values used in these calculations, however, had been largely determined by the measurement limits of the chemical analysis methodology instead of measured releases. The purpose of the current study was to refine the analytical method to reduce the limit for detecting cobalt, and then test PCA steel in air and steam between 973 and 1473 K. Goals were to obtain more accurate measurements of cobalt mobilization in terms of g/m{sup 2}{center_dot}h and insight into the mobilization mechanisms.
Date: January 1, 1995
Creator: Smolik, G. R.; McCarthy, K. A.; Hagrman, D. L. & McNew, E. B.
Partner: UNT Libraries Government Documents Department

Investigation of the coolability of a continuous mass of relocated debris to a water-filled lower plenum. Technical report

Description: This report documents work performed to support the development of an analytical and experimental program to investigate the coolability of a continuous mass of debris that relocates to a water-filled lower plenum. The objective of this program is to provide an adequate data base for developing and validating a model to predict the coolability of a continuous mass of debris relocating to a water-filled lower plenum. The model must address higher pressure scenarios, such as the TMI-2 accident, and lower pressure scenarios, which recent calculations indicate are more likely for most operating LWR plants. The model must also address a range of possible debris compositions.
Date: September 1, 1994
Creator: Rempe, J.L.; Wolf, J.R.; Chavez, S.A.; Condie, K.G.; Hagrman, D.L. & Carmack, W.J.
Partner: UNT Libraries Government Documents Department

Extension of SCDAP/RELAP5 severe accident models to non-LWR reactor designs. [Non-Light Water Reactors]

Description: The SCDAP/RELAP5 code has been extended to calculate the core melt progression and fission product transport that may occur in non-LWR reactors during severe accidents. The code's approach of connecting together according to user instructions all of the parts that constitute a reactor system give the code the capability to model a wide range of reactor designs. The models added to the code for analyses of non-LWR reactors include: (a) oxidation and melt progression in cores with U-Al based fuel elements, (b) movement of liquefied material from its original place in the core to other parts of the reactor systems, such as the outlet piping, (c) fission product release from U-Al based fuel and zinc release from aluminum, and (d) fission product release from a pool of molten core material. 9 refs., 5 figs.
Date: January 1, 1990
Creator: Allison, C.M.; Siefken, L.J.; Hagrman, D.L. (EG and G Idaho, Inc., Idaho Falls, ID (USA)) & Cheng, T.C. (Los Alamos National Lab., NM (USA))
Partner: UNT Libraries Government Documents Department

Modeling requirements for in situ vitrification

Description: This document outlines the requirements for the model being developed at the INEL which will provide analytical support for the ISV technology assessment program. The model includes representations of the electric potential field, thermal transport with melting, gas and particulate release, vapor migration, off-gas combustion and process chemistry. The modeling objectives are to (1) help determine the safety of the process by assessing the air and surrounding soil radionuclide and chemical pollution hazards, the nuclear criticality hazard, and the explosion and fire hazards, (2) help determine the suitability of the ISV process for stabilizing the buried wastes involved, and (3) help design laboratory and field tests and interpret results therefrom.
Date: November 1, 1991
Creator: MacKinnon, R.J.; Mecham, D.C.; Hagrman, D.L.; Johnson, R.W.; Murray, P.E.; Slater, C.E. et al.
Partner: UNT Libraries Government Documents Department

Power Burst Facility (PBF) severe fuel damage test 1-4 test results report

Description: A comprehensive evaluation of the Severe Fuel Damage (SFD) Test 1-4 performed in the Power Burst Facility (PBF) at the Idaho National Engineering Laboratory is presented. Test SFD 1-4 was the fourth and final test in an internationally sponsored light water reactor severe accident research program, initiated by the US Nuclear Regulatory Commission. The overall technical objective of the test was to contribute to the understanding of fuel and control rod behavior, aerosol and hydrogen generation, and fission product release and transport during a high-temperature, severe fuel damage transient. A test bundle, comprised of 26 previously irradiated (36,000 MWd/MtU) pressurized water-reactor-type fuel rods, 2 fresh instrumented fuel rods, and 4 silver-indium-cadmium control rods, was surrounded by an insulating shroud and contained in a pressurized in-pile tube. The experiment consisted of a 1.3-h transient at a coolant pressure of 6.95 MPa in which the inlet coolant flow to the bundle was reduced to 0.6 g/s while the bundle fission power was gradually increased until dryout, heatup, cladding rupture, and oxidation occurred. With sustained fission power and heat from oxidation, temperatures continued to rise rapidly, resulting in zircaloy and control rod absorber alloy melting, fuel liquefaction, material relocation, and the release of hydrogen, aerosols, and fission products. The transient was terminated over a 2100-s time span by slowly reducing the reactor power and cooling the damaged bundle with argon gas. A description and evaluation of the major phenomena, based upon the response of on-line instrumentation, analysis of fission product and aerosol data, postirradiation examination of the fuel bundle, and calculations using the SCDAP/RELAP5 computer code, are presented. 40 refs., 160 figs., 31 tabs.
Date: April 1, 1989
Creator: Petti, D.A.; Martinson, Z.R.; Hobbins, R.R.; Allison, C.M.; Carlson, E.R.; Hagrman, D.L. et al.
Partner: UNT Libraries Government Documents Department

SCDAP/RELAP5/MOD2 code manual

Description: The SCDAP/RELAP5 code has been developed for best-estimate transient simulation of light water reactor coolant systems during a severe accident. The code models the coupled behavior of the reactor coolant system, the core, and the fission products and aerosols in the system during a severe accident transient as well as large and small break loss-of-coolant accidents, operational transients such as anticipated transient without SCRAM, loss of offsite power, loss of feedwater, and loss of flow. A generic modeling approach is used that permits as much of a particular system to be modeled as necessary. Control system and secondary system components are included to permit modeling of plant controls, turbines, condensers, and secondary feedwater conditioning systems. The modeling theory and associated numerical schemes are documented in Volumes I and II to acquaint the user with the modeling base and thus aid in effective use of the code.
Date: September 1, 1989
Creator: Allison, C. M.; Johnson, E. C.; Berna, G.A.; Cheng, T.C.; Hagrman, D.L.; Johnsen, G.W. et al.
Partner: UNT Libraries Government Documents Department

SCDAP/RELAP5/MOD2 code manual

Description: The SCDAP/RELAP5 code has been developed for best-estimate transient simulation of light water reactor coolant systems during a severe accident. The code models the coupled behavior of the reactor coolant system, the core, and the fission products and aerosols in the system during a severe accident transient as well as large and small break loss-of-coolant accidents, operational transients such as anticipated transient without SCRAM, loss of offsite power, loss of feedwater, and loss of flow. A generic modeling approach is used that permits as much of a particular system to be modeled as necessary. Control system and secondary system components are included to permit modeling of plant controls, turbines, condensers, and secondary feedwater conditioning systems. The modeling theory and associated numerical schemes are documented in Volumes I and in this document, Volume II, to acquaint the user with the modeling base and thus aid in effective use of the code. 135 refs., 48 figs., 8 tabs.
Date: September 1, 1989
Creator: Allison, C. M.; Johnson, E .C.; Berna, G.A.; Cheng, T.C.; Hagrman, D.L.; Johnsen, G.W. et al.
Partner: UNT Libraries Government Documents Department

SCDAP/RELAP5/MOD2 code manual

Description: The SCDAP/RELAP5 code has been developed for best-estimate transient simulation of light water reactor coolant systems during a severe accident. The code models the coupled behavior of the reactor coolant system, the core, and the fission products and aerosols in the system during a severe accident transient as well as large and small break loss-of-coolant accidents, operational transients such as anticipated transient without SCRAM, loss of offsite power, loss of feedwater, and loss of flow. A generic modeling approach is used that permits as much of a particular system to be modeled as necessary. Control system and secondary system components are included to permit modeling of plant controls, turbines, condensers, and secondary feedwater conditioning systems. This document, Volume III, contains detailed instructions for code application and input data preparation. In addition, Volume III contains user guidelines that have evolved over the past several years from application of the RELAP5 and SCDAP codes at the Idaho National Engineering Laboratory, at other national laboratories, and by users throughout the world. 2 refs., 32 figs., 9 tabs.
Date: September 1, 1989
Creator: Allison, C. M.; Johnson, E .C.; Berna, G. A.; Cheng, T. C.; Hagrman, D. L.; Johnsen, G. W. et al.
Partner: UNT Libraries Government Documents Department