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in subjects: "Fission Product Release"

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Source term experiment STEP-3 simulating a PWR severe station blackout

For a severe PWR accident that leads to a loss of feedwater to the steam generators, such as might occur in a station blackout, fission product decay heating will cause a water boiloff. Without effective cooling of the core, steam will begin to oxidize the Zircaloy cladding. The noble gases and volatile fission products, such as Cs and I, that are major contributors to the radiological source term, will be released from the damaged fuel shortly after cladding failure. The accident environment w… more

Date: May 21, 1987

Creator: Simms, R.; Baker, L., Jr. & Ritzman, R. L.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Integral fast reactor safety features

The Integral Fast Reactor (IFR) is an advanced liquid-metal-cooled reactor concept being developed at Argonne National Laboratory. The two major goals of the IFR development effort are improved economics and enhanced safety. In addition to liquid metal cooling, the principal design features that distinguish the IFR are: (1) a pool-type primary system, (2) an advanced ternary alloy metallic fuel, and (3) an integral fuel cycle with on-site fuel reprocessing and fabrication. This paper focuses on… more

Date: January 1, 1988

Creator: Cahalan, J. E.; Kramer, J. M.; Marchaterre, J. F.; Mueller, C. J.; Pedersen, D. R.; Sevy, R. H. et al.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Core cooling under accident conditions at the high flux beam reactor (HFBR)

In certain accident scenarios, e.g. loss of coolant accidents (LOCA) all forced flow cooling is lost. Decay heating causes a temperature increase in the core coolant and the resulting thermal buoyancy causes a reversal of the flow direction to a natural circulation mode. Although there was experimental evidence during the reactor design period (1958--1963) that the heat removal capacity in the fully developed natural circulation cooling mode was relatively high, it was not possible to make a co… more

Date: January 1, 1991

Creator: Tichler, P.; Cheng, L. (Brookhaven National Lab., Upton, NY (USA)) & Fauske, H. (Fauske and Associates, Inc., Burr Ridge, IL (USA))

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

EBR-II: twenty years of operating experience

Experimental Breeder Reactor No. 2 (EBR-II) is an unmoderated, sodium-cooled reactor with a design power of 62.5 MWt. For the last 20 years EBR-II has operated safely, has demonstrated stable operating characteristics, has shown excellent performance of its sodium components, and has had an excellent plant factor. These years of operating experience provide a valuable resource to the nuclear community for the development and design of future liquid metal fast reactors. This report provides a br… more

Date: January 1, 1985

Creator: Lentz, G. L.; Buschman, H. W. & Smith, R. N.

Item Type: Article

Partner: UNT Libraries Government Documents Department

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Status of RBCB testing of LMR oxide fuel in EBR-II

The status is given of the the American-Japanese collaborative program in Experimental Breeder Reactor 2 to determine the run-beyond-cladding-breach performance of (UPu)O{sub 2} fuel pins for liquid-metal cooled reactors. Phase 1 of the collaboration involved eighteen irradiation tests over 1981--86 with 5.84-mm pins in 316 or D9 stainless steel. Emphasis in Phase 2 tests from 1989 onwards is with larger diameter (7.5mm) pins in advanced claddings. Results include delayed neutron and fission ga… more

Date: January 1, 1991

Creator: Strain, R.V.; Bottcher, J.H.; Gross, K.C.; Lambert, J.D.B. (Argonne National Lab., IL (United States)); Ukai, S.; Nomura, S. et al.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Modeling and analysis of the unprotected loss-of-flow accident in the Clinch River Breeder Reactor

The influence of fission-gas-driven fuel compaction on the energetics resulting from a loss-of-flow accident was estimated with the aid of the SAS3D accident analysis code. The analysis was carried out as part of the Clinch River Breeder Reactor licensing process. The TREAT tests L6, L7, and R8 were analyzed to assist in the modeling of fuel motion and the effects of plenum fission-gas release on coolant and clad dynamics. Special, conservative modeling was introduced to evaluate the effect of … more

Date: January 1, 1985

Creator: Morris, E. E.; Dunn, F. E.; Simms, R. & Gruber, E. E.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Experiment data report IFA-226 postirradiation examination. [PWR, BWR]

IFA-226 contained twelve, mixed plutonium-uranium oxide fuel rods arranged in two, six-rod clusters. The assembly was designed to study fuel-cladding mechanical interaction, fuel thermal response, and fission gas release as a function of fuel density, initial fuel-to-cladding gap, rod power, and burnup. Data were obtained from fuel rod centerline thermocouples, fission gas pressure transducers, and cladding elongation sensors. Results of both nondestructive and destructive examinations are pres… more

Date: September 1, 1977

Creator: Bagger, C.; Carlsen, H.; Domanus, J.; Hougaard, H.; Larsen, E. & Larsen, N.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Overview of experimental support for fission-product transport analyses at Oak Ridge National Laboratory

The program was designed to determine fission product and aerosol release rates from irradiated fuel under accident conditions, to identify the chemical forms of the released material, and to correlate the results with experimental and specimen conditions with the data from related experiments. These tests of PWR fuel were conducted and fuel specimen and test operating data are presented. The nature and rate of fission product vapor interaction with aerosols were studied. Aerosol deposition rat… more

Date: January 1, 1983

Creator: Wichner, R. P.

Item Type: Article

Partner: UNT Libraries Government Documents Department

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Effects of lateral separation of oxidic and metallic core debris on the BWR MK I containment drywell floor

In evaluating core debris/concrete interactions for a BWR MK I containment design, it is common practice to assume that at reactor vessel breach, the core debris is homogeneous and of low viscosity, so that it flows through the pedestal doorway and spreads in a radially uniform fashion throughout the drywell floor. In a recent study performed by the NRC-sponsored Severe Accident Sequence Analysis (SASA) program at Oak Ridge National Laboratory, calculations indicate that at reactor vessel botto… more

Date: January 1, 1986

Creator: Hyman, C.R.; Weber, C.F. & Hodge, S.A.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Fission product Pd-SiC interaction in irradiated coated particle fuels

Silicon carbide is the main barrier to fission product release from coated particle fuels. Consequently, degradation of the SiC must be minimized. Electron microprobe analysis has identified that palladium causes corrosion of the SiC in irradiated coated particles. Further ceramographic and electron microprobe examinations on irradiated particles with kernels ranging in composition from UO/sub 2/ to UC/sub 2/, including PuO/sub 2 -x/ and mixed (Th, Pu) oxides, and in enrichment from 0.7 to 93.0… more

Date: April 1, 1980

Creator: Tiegs, T. N.

Item Type: Report

Partner: UNT Libraries Government Documents Department

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Estimated amount of /sup 85/Kr available for release from intact fuel rods in the Three Mile Island Unit 2 Nuclear Power Station

Reactor core dismantling operations planned for the Three Mile Island Unit 2 Power Station could cause rupture of any fuel rods that might have survived the accident in a leak-tight condition. Calculations were performed in an attempt to determine the amount of /sup 85/Kr that might be present in the free gaseous form in the plenum and void spaces of such fuel rods. Estimates were made of the number of fuel rods surviving intect, the temperature transient to which they were exposed, the amount … more

Date: October 1, 1983

Creator: Lorenz, R. A.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Characterization of sodium fires and fission products. Progress report, October-December 1975. [LMFBR]

The objectives of the research program are to develop a computer program for calculating two-dimensional, transient, natural-convection phenomena, such as those arising from various postulated sodium spill accidents in LMFBR heat transfer vaults, head compartments, containment buildings, and secondary heat transfer systems; to develop experimental programs and conduct tests that will characterize the behavior of sodium, sodium oxide, fuel, fission product, and other aerosols as they might be ge… more

Date: January 1, 1975

Creator: Heisler, M. P.; Johnson, R. P.; Otter, J. M.; Nelson, C. T.; Specht, E. R.; Guderjahn, C. et al.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Assessment of improved fission-product transport models in VICTORIA against the ORNL HI and VI tests

New models for the release of fission-products from fuel have recently been incorporated into a comprehensive code, VICTORIA, for the prediction of radionuclide behavior under severe reactor accident conditions as an alternative to a simple Booth diffusion calculation. It is widely known that the Booth model has severe limitations when used under conditions of changing temperature and power. A new transport model based on a two node diffusive flow formulation has been implemented into VICTORIA.… more

Date: January 1, 1991

Creator: Domagala, P.; Rest, J. & Zawadzki, S.A.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Aerosols released during large-scale integral MCCI tests in the ACE Program

As part of the internationally sponsored Advanced Containment Experiments (ACE) program, seven large-scale experiments on molten core concrete interactions (MCCIs) have been performed at Argonne National Laboratory. One of the objectives of these experiments is to collect and characterize all the aerosols released from the MCCIs. Aerosols released from experiments using four types of concrete (siliceous, limestone/common sand, serpentine, and limestone/limestone) and a range of metal oxidation … more

Date: January 1, 1992

Creator: Fink, J. K.; Thompson, D. H.; Spencer, B. W. & Sehgal, B. R.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Evaluation of strontium-90 radial concentration profiles in Peach Bottom HTGR Core 2 fuel elements. HTGR base technology program, HTGR chemistry studies (189a 01329). [483 2000]

Radial concentration profiles of /sup 90/Sr were evaluated for four fuel elements of the Peach Bottom High-Temperature Gas-Cooled Reactor (HTGR) Core 2. A transport model utilizing a single effective diffusion coefficient was able to reproduce the strontium concentraion profile in graphite at temperatures above about 960/sup 0/C. Below this temperature, ''tails'' were observed in the experimental concentration profile and the profiles could not be completely reproduced. Diffusion coefficients d… more

Date: February 1, 1979

Creator: Haire, M.J.

Item Type: Report

Partner: UNT Libraries Government Documents Department

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Behavior of fission product tellurium under severe accident conditions

Fission product release tests at Oak Ridge National Laboratory (ORNL) have provided new experimental data that help characterize the behavior of tellurium under severe light-water reactor (LWR) accident conditions. The release of tellurium from the fuel rods is dependent upon the rate and extent of cladding oxidation. Tellurium has been found to be considerably retained by metallic Zircaloy cladding at test temperatures up to 1975/sup 0/C. The results indicate that the tellurium is bound by the… more

Date: January 1, 1986

Creator: Collins, J. L.; Osborne, M. F. & Lorenz, R. A.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Investigation of the chemical explosion of an ion exchange resin column and resulting americium contamination of personnel in the 242-Z building, August 30, 1976

As a result of an explosion in the Waste Treatment Facility, 242-Z Building, 200 West Area of the Hanford Reservation on August 30, 1976, the Manager of the Richland Operations Office (RL), Energy Research and Development Administration (ERDA), appointed an ERDA Committee to conduct a formal investigation and to prepare a report on their findings of this occurrence. The Committee was instructed to conduct the investigation in accordance with ERDAMC 0502, insofar as circumstances would permit, t… more

Date: October 19, 1976

Item Type: Report

Partner: UNT Libraries Government Documents Department

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Accident simulation and consequence analysis in support of MHTGR safety evaluations

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 heatup and steam ingress accidents, and the reviews and analyses of fuel performance and fission product transport technology.

Date: January 1, 1991

Creator: Ball, S. J.; Wichner, R. P.; Smith, O. L.; Conklin, J. C. (Oak Ridge National Lab., TN (United States)) & Barthold, W. P. (Barthold Associates, Inc., Knoxville, TN (United States))

Item Type: Article

Partner: UNT Libraries Government Documents Department

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Impact of Zr metal and coking reactions on the ex-vessel source term predictions of CORCON/VANESA

During a core meltdown accident in a LWR, molten core materials (corium) could leave the reactor vessel and interact with concrete. In this paper, the impact of the zirconium content of the corium pool and the coking reaction on the release of fission products are quantified using CORCON/Mod2 and VANESA computer codes. Detailed calculations show that the total aerosol generation is proportional to the zirconium content of the corium pool. Among the twelve fission product groups treated by the V… more

Date: January 1, 1987

Creator: Lee, M.; Davis, R.E. & Khatib-Rahbar, M.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Independent analysis of selected core-concrete interaction and fission product release experiments with CORCON-MOD2 and VANESA

The discrepancies between experimental findings and the Reactor Safety Study predictions, as well as the rapidly developing data base enabling phenomenological modeling of core-concrete interactions and ex-vessel fission product release, have prompted the development of several new computer models of core-concrete interactions and fission product release during severe accidents. Two such models are the CORCON-MOD2 model of core-concrete interactions and the VANESA model of ex-vessel aerosol and… more

Date: January 1, 1986

Creator: Greene, G.A. & Sanborn, Y.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Status of the TMI-2 core: a review of damage assessments

Assessments of the damage within the core of the Three Mile Island Unit 2 reactor, performed by reconstructing the transient thermal-hydraulic sequence of events, estimating the amount of hydrogen generation, and evaluating the amount of fission products released, are reviewed and summarized. Minimum and maximum bounds of damage to the core are identified.

Date: January 1, 1981

Creator: Croucher, D.W.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Release of volatile fission products from uranium dioxide

Post-irradiation anneal experiments have been used to determine the release of iodine and tellurium from lightly irradiated UO/sub 2/ samples maintained at stoichiometry. The applicability of the equivalent-sphere model of diffusion to release of fission gases has been tested. Diffusion coefficients and activation energies have been evaluated. The diffusion coefficient of /sup 132/Te at 1400/sup 0/C was found to be of an order-of-magnitude larger than that of /sup 131/I. This result may be of i… more

Date: March 1, 1983

Creator: Bayen, D.

Item Type: Report

Partner: UNT Libraries Government Documents Department

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Radioactive materials released from nuclear power plants. Annual report, 1982. Volume 3

Releases of radioactive materials in airborne and liquid effluents from commercial light water reactors during 1982 have been compiled and reported. Data on solid waste shipments as well as selected operating information have been included. This report supplements earlier annual reports issued by the former Atomic Energy Commission and the Nuclear Regulatory Commission. The 1982 release data are summarized in tabular form. Data covering specific radionuclides are summarized.

Date: February 1, 1986

Creator: Tichler, J. & Norden, K.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Accident Management Information Needs: Methodology Development and Application to a Pressurized Water Reactor (PWR) with a Large, Dry Containment

In support of the US Nuclear Regulatory Commission (NRC) Accident Management Research Program, a methodology has been developed for identifying the plant information needs necessary for personnel involved in the management of an accident to diagnose that an accident is in progress, select and implement strategies to prevent or mitigate the accident, and monitor the effectiveness of these strategies. This report describes the methodology and presents an application of this methodology to a Press… more

Date: April 1990

Creator: Hanson, D. J.; Ward, L. W.; Nelson, W. R. & Meyer, O. R.

Item Type: Report

Partner: UNT Libraries Government Documents Department