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Potential for Hepa filter damage from water spray systems in filter plenums

Description: The water spray systems in high efficiency particulate air (HEPA) filter plenums that are used in nearly all Department of Energy (DOE) facilities for protection against fire was designed under the assumption that the HEPA filters would not be damaged by the water sprays. The most likely scenario for filter damage involves filter plugging by the water spray, followed by the fan blowing out the filter medium. A number of controlled laboratory tests that were previously conducted in the late 1980s are reviewed in this paper to provide a technical basis for the potential HEPA filter damage by the water spray system in HEPA filter plenums. In addition to the laboratory tests, the scenario for HEPA filter damage during fires has also occurred in the field. Afire in a four-stage, HEPA filter plenum at Rocky Flats in 1980 caused the first three stages of HEPA filters to blow out of their housing and the fourth stage to severely bow. Details of this recently declassified fire are presented in this paper. Although these previous findings suggest serious potential problems exist with the current water spray system in filter plenum , additional studies are required to confirm unequivocally that DOE`s critical facilities are at risk.
Date: January 1, 1997
Creator: Bergman, W.; Fretthold, J.K. & Slawsld, J.W.
Partner: UNT Libraries Government Documents Department

Removal of Sarin Aerosol and Vapor by Water Sprays

Description: Falling water drops can collect particles and soluble or reactive vapor from the gas through which they fall. Rain is known to remove particles and vapors by the process of rainout. Water sprays can be used to remove radioactive aerosol from the atmosphere of a nuclear reactor containment building. There is a potential for water sprays to be used as a mitigation technique to remove chemical or bio- logical agents from the air. This paper is a quick-look at water spray removal. It is not definitive but rather provides a reasonable basic model for particle and gas removal and presents an example calcu- lation of sarin removal from a BART station. This work ~ a starting point and the results indicate that further modeling and exploration of additional mechanisms for particle and vapor removal may prove beneficial.
Date: September 1, 1998
Creator: Brockmann, John E.
Partner: UNT Libraries Government Documents Department

Iodine Revolatilization in a Grand Gulf Loca

Description: The TRENDS models are applied at each time step to each control volume. Significant amounts of water occur only in the wetwell and drywell sump (the refueling pool is not a factor, as discussed earlier). In Fig. 2, we show the radiolytic acid production feeding into each of these pools. Since the water is initially neutral and no chemical additives are present, the acid additions are the major factors affecting pH. In Fig. 3, we see the downward trend of pH resulting from these acid additions. The conversion of iodide (I{sup {minus}}) to molecular iodine (I{sub 2}) is most noticeable in the wetwell, since this is the repository of most iodide and HCl. Gradually, during the transient small amounts of more volatile iodine are formed. While iodide remains the dominant form, noticeable amounts of I{sub 2} and intermediate species are created. Once produced in water, some I{sub 2} is free to evaporate into airspace. Fig. 4 indicates the increase in all airborne iodine throughout the transient. This is compared to the MELCOR result for CsI aerosol, which decreases dramatically due to containment sprays. The I{sub 2} in the airspace can be vented to the enclosure building or the environment. In the present accident sequence, the only path to the environment was through the SGTS, which was assumed to operate as in MELCOR. However, both are dwarfed by the MELCOR gaseous release during the first 12 h because MELCOR does not model spray washout of gaseous iodine. Steadily increasing throughout the transient, the revolatilization release is eventually more than an order-or-magnitude higher than the MELCOR aerosol release. Also, 99% of iodine flowing directly through the SGTS was retained in filters. The remaining 1% was released to the environment. In addition, a small flow bypassing the SGTS filters vented directly into the ...
Date: January 1, 1999
Creator: Beahm, E.C. & Weber, C.F.
Partner: UNT Libraries Government Documents Department

Mitigation of unconfined releases of hazardous gases via liquid spraying

Description: The capability of water sprays in mitigating clouds of hydrofluoric acid (HF) has been demonstrated in the large-scale field experiments of Goldfish and Hawk, which took place at the DOE Nevada Test Site. The effectiveness of water sprays and fire water monitors to remove HF from vapor plume, has also been studied theoretically using the model HGSPRAY5 with the near-field and far-field dispersion described by the HGSYSTEM models. This paper presents options to select and evaluate liquid spraying systems, based on the industry experience and mathematical modeling.
Date: February 1, 1997
Creator: Fthenakis, V.M.
Partner: UNT Libraries Government Documents Department

MELCOR 1.8.3 assessment: CSE containment spray experiments

Description: MELCOR is a fully integrated, engineering-level computer code, being developed at Sandia National Laboratories for the USNRC, that models the entire spectrum of severe accident phenomena in a unified framework for both BWRs and PWRS. As part, of an ongoing assessment program, the MELCOR computer code has been used to analyze a series of containment spray tests performed in the Containment Systems Experiment (CSE) vessel to evaluate the performance of aqueous sprays as a means of decontaminating containment atmospheres. Basecase MELCOR results are compared with test data, and a number of sensitivity studies on input modelling parameters and options in both the spray package and the associated aerosol washout and atmosphere decontamination by sprays modelled in the radionuclide package have been done. Time-step and machine-dependency calculations were done to identify whether any numeric effects exist in these CSE assessment analyses. A significant time-step dependency due to an error in the spray package coding was identified and eliminated. A number of other code deficiencies and inconveniences also are noted.
Date: December 1, 1994
Creator: Kmetyk, L.N.
Partner: UNT Libraries Government Documents Department

Guidelines for nuclear power plant safety issue prioritization information development. Supplement 5

Description: This is the sixth in a series of reports to document the development and use of a methodology developed by the Pacific Northwest Laboratory (PNL) to calculate, for prioritization purposes, the risk, dose, and cost impacts of implementing potential resolutions to reactor safety issues (see NUREG/CR-2800, Andrews, et al., 1983). This report contains the results of issue-specific analyses for 34 generic issues. Each issue was considered within the constraints of available information at the time the issues were examined and approximately 2 staff-weeks of labor. The results are referenced as one consideration in NUREG-0933, A Prioritization of Generic Safety Issues (Emrit, et al., 1983).
Date: July 1, 1996
Creator: Daling, P. M. & Lavender, J. C.
Partner: UNT Libraries Government Documents Department

Degradation and Failure Characteristics of NPP Containment Protective Coating Systems

Description: Nuclear power plants (NPPs) must ensure that the emergency core cooling system (ECCS) or safety-related containment spray system (CSS) remains capable of performing its design safety function throughout the life of the plant. This requires ensuring that long-term core cooling can be maintained following a postulated loss-of-coolant accident (LOCA). Adequate safety operation can be impaired if the protective coatings which have been applied to the concrete and steel structures within the primary containment fail, producing transportable debris which could then accumulate on BWR ECCS suction strainers or PWR ECCS sump debris screens located within the containment. This document will present the data collected during the investigation of coating specimens from plants.
Date: April 10, 2001
Creator: Sindelar, R.L.
Partner: UNT Libraries Government Documents Department

The effect of uncertainties in nuclear reactor plant-specific failure data on core damage frequency

Description: It is sometimes the case in PRA applications that reported plant-specific failure data are, in fact, only estimates which are uncertain. Even for detailed plant-specific data, the reported exposure time or number of demands is often only an estimate of the actual exposure time or number of demands. Likewise the reported number of failure events or incidents is sometimes also uncertain because incident or malfunction reports may be ambiguous. In this report we determine the corresponding uncertainty in core damage frequency which can b attributed to such uncertainties in plant-specific data using a simple but typical nuclear power reactor example.
Date: May 1, 1995
Creator: Martz, H.F.
Partner: UNT Libraries Government Documents Department

CONTAIN assessment of the NUPEC mixing experiments

Description: The ability of the CONTAIN code to predict the thermal hydraulics of five experiments performed in the NUPEC 1/4-scale model containment was assessed. These experiments simulated severe accident conditions in a nuclear power plant in which helium (as a nonflammable substitute for hydrogen) and steam were coinjected at different locations in the facility with and without the concurrent injection of water sprays in the dome. Helium concentrations, gas temperatures and pressures, and wall temperatures were predicted and compared with the data. The use of different flow solvers, nodalization schemes, and analysis methods for the treatment of water sprays was emphasized. As a result, a general procedure was suggested for lumped-parameter code analyses of problems in which the thermal hydraulics are dominated by water sprays.
Date: August 1, 1995
Creator: Stamps, D.W.
Partner: UNT Libraries Government Documents Department

Ex-vessel melt-coolant interactions in deep water pool: Studies and accident management for Swedish BWRs

Description: In Swedish BWRs having an annular suppression pool, the lower drywell beneath the reactor vessel is flooded with water to mitigate against the effects of melt release into the drywell during a severe accident. The THIRMAL code has been used to analyze the effectiveness of the water pool to protect lower drywell penetrations by fragmenting and quenching the melt as it relocates downward through the water. Experiments have also been performed to investigate the benefits of adding surfactants to the water to reduce the likelihood of fine-scale debris formation from steam explosions. This paper presents an overview of the accident management approach and surfactant investigations together with results from the THIRMAL analyses.
Date: January 1, 1993
Creator: Sienicki, J. J.; Chu, C. C.; Spencer, B. W.; Frid, W. & Loewenhielm, G.
Partner: UNT Libraries Government Documents Department

Progress in MELCOR development and assessment

Description: MELCOR models the progression of severe accidents in light water reactor nuclear power plants. Recent efforts in MELCOR development to incorporate CORCON-Mod3 models for core-concrete interactions, new models for advanced reactors, and improvements to several other existing models have resulted in release of MELCOR 1.8.3. In addition, continuing efforts to expand the code assessment database have filled in many of the gaps in phenomenological coverage. Efforts are now under way to develop models for chemical interactions of fission products with structural surfaces and for reactions of iodine in the presence of water, and work is also in progress to improve models for the scrubbing of fission products by water pools, the chemical reactions of boron carbide with steam, and the coupling of flow blockages with the hydrodynamics. Several code assessment analyses are in progress, and more are planned.
Date: April 1, 1995
Creator: Summers, R. M.; Kmetyk, L. N.; Cole, R. K., Jr.; Smith, R. C.; Elsbernd, A. E.; Stuart, D. S. et al.
Partner: UNT Libraries Government Documents Department

LOFT suppression tank spray system piping: heat exchanger BS-H-31 piping modifications

Description: A stress analysis of the piping modification, resulting from relocation of heat exchanger BS-H-31 of the LOFT Blowdown Suppressing Tank Spray System, was performed. The piping, fittings, and supports were found to comply with the criteria of Section III of the ASME Boiler and Pressure Vessel Code, 1974.
Date: November 7, 1977
Creator: Blandford, R.K.
Partner: UNT Libraries Government Documents Department

Sodium spray and jet fire model development within the CONTAIN-LMR code

Description: An assessment was made of the sodium spray fire model implemented in the CONTAIN code. The original droplet burn model, which was based on the NACOM code, was improved in several aspects, especially concerning evaluation of the droplet burning rate, reaction chemistry and heat balance, spray geometry and droplet motion, and consistency with CONTAIN standards of gas property evaluation. An additional droplet burning model based on a proposal by Krolikowski was made available to include the effect of the chemical equilibrium conditions at the flame temperature. The models were validated against single-droplet burn experiments as well as spray and jet fire experiments. Reasonable agreement was found between the two burn models and experimental data. When the gas temperature in the burning compartment reaches high values, the Krolikowski model seems to be preferable. Critical parameters for spray fire evaluation were found to be the spray characterization, especially the droplet size, which largely determines the burning efficiency, and heat transfer conditions at the interface between the atmosphere and structures, which controls the thermal hydraulic behavior in the burn compartment.
Date: December 31, 1993
Creator: Scholtyssek, W. & Murata, K. K.
Partner: UNT Libraries Government Documents Department

System 80+{trademark} Standard Design: CESSAR design certification. Volume 7: Amendment I

Description: This report, entitled Combustion Engineering Standard Safety Analysis Report - Design Certification (CESSAR-DC), has been prepared in support of the industry effort to standardize nuclear plant designs. These documents describe the Combustion Engineering, Inc. System 80+{sup TM} Standard Design. This report, Volume 7, in conjunction with Volume 6, provides a description of engineered safety features.
Date: December 21, 1990
Partner: UNT Libraries Government Documents Department

Review of Hanford production reactor confinement (Project CGI-791)

Description: The Project CGI-791, Reactor Confinement for Hanford production reactors, vas conceived as a result of deliberations concerning the release of the Wahluke Slope secondary exclusion zone for private use, and initiated by an authorization and directive from the Atomic Energy Commission to proceed with such a project on a high priority basis. This authorization was coupled with the establishment of limits on reactor heat-generation rates and fission product inventories. These limits were to be re-evaluated after significant progress has been made toward providing some undefined degree of control of the ultimate release to the environs of any air-borne radioactive fission products which might be released into the building housing any Hanford production reactor. It is the purpose of this report to summarize the philosophical and technical bases for the nature and degree of confinement provided under project CGI-791, to describe the project, and, further, to indicate the expected effectiveness of this confinement system. In discussing the expected effectiveness of the project this report is responsive to a request from the AEC-HOO.
Date: October 14, 1960
Creator: Trumble, R. E.
Partner: UNT Libraries Government Documents Department

Ex-vessel melt-coolant interactions in deep water pool: Studies and accident management for Swedish BWRs

Description: In Swedish BWRs having an annular suppression pool, the lower drywell beneath the reactor vessel is flooded with water to mitigate against the effects of melt release into the drywell during a severe accident. The THIRMAL code has been used to analyze the effectiveness of the water pool to protect lower drywell penetrations by fragmenting and quenching the melt as it relocates downward through the water. Experiments have also been performed to investigate the benefits of adding surfactants to the water to reduce the likelihood of fine-scale debris formation from steam explosions. This paper presents an overview of the accident management approach and surfactant investigations together with results from the THIRMAL analyses.
Date: January 1, 1993
Creator: Sienicki, J.J.; Chu, C.C.; Spencer, B.W. (Argonne National Lab., IL (United States)); Frid, W. (Swedish Nuclear Power Inspectorate, Stockholm (Sweden)) & Loewenhielm, G. (Vattenfall AB, Vaellingby (Sweden))
Partner: UNT Libraries Government Documents Department

Station blackout calculations for Peach Bottom

Description: A calculational procedure for the Station Blackout Severe Accident Sequence at Browns Ferry Unit One has been repeated with plant-specific application to one of the Peach Bottom Units. The only changes required in code input are with regard to the primary continment concrete, the existence of sprays in the secondary containment, and the size of the refueling bay. Combustible gas mole fractions in the secondary containment of each plant during the accident sequence are determined. It is demonstrated why the current state-of-the-art corium/concrete interaction code is inadequate for application to the study of Severe Accident Sequences in plants with the BWR MK I or MK II containment design.
Date: January 1, 1985
Creator: Hodge, S.A.
Partner: UNT Libraries Government Documents Department

A simplified model of aerosol removal by containment sprays

Description: Spray systems in nuclear reactor containments are described. The scrubbing of aerosols from containment atmospheres by spray droplets is discussed. Uncertainties are identified in the prediction of spray performance when the sprays are used as a means for decontaminating containment atmospheres. A mechanistic model based on current knowledge of the physical phenomena involved in spray performance is developed. With this model, a quantitative uncertainty analysis of spray performance is conducted using a Monte Carlo method to sample 20 uncertain quantities related to phenomena of spray droplet behavior as well as the initial and boundary conditions expected to be associated with severe reactor accidents. Results of the uncertainty analysis are used to construct simplified expressions for spray decontamination coefficients. Two variables that affect aerosol capture by water droplets are not treated as uncertain; they are (1) [open quote]Q[close quote], spray water flux into the containment, and (2) [open quote]H[close quote], the total fall distance of spray droplets. The choice of values of these variables is left to the user since they are plant and accident specific. Also, they can usually be ascertained with some degree of certainty. The spray decontamination coefficients are found to be sufficiently dependent on the extent of decontamination that the fraction of the initial aerosol remaining in the atmosphere, m[sub f], is explicitly treated in the simplified expressions. The simplified expressions for the spray decontamination coefficient are given. Parametric values for these expressions are found for median, 10 percentile, and 90 percentile values in the uncertainty distribution for the spray decontamination coefficient. Examples are given to illustrate the utility of the simplified expressions to predict spray decontamination of an aerosol-laden atmosphere.
Date: June 1, 1993
Creator: Powers, D.A. (Sandia National Labs., Albuquerque, NM (United States)) & Burson, S.B. (Nuclear Regulatory Commission, Washington, DC (United States). Div. of Safety Issue Resolution)
Partner: UNT Libraries Government Documents Department

Reactor operation safety information document

Description: The report contains a reactor facility description which includes K, P, and L reactor sites, structures, operating systems, engineered safety systems, support systems, and process and effluent monitoring systems; an accident analysis section which includes cooling system anomalies, radioactive materials releases, and anticipated transients without scram; a summary of onsite doses from design basis accidents; severe accident analysis (reactor core disruption); a description of operating contractor organization and emergency planning; and a summary of reactor safety evolution. (MB)
Date: January 1, 1990
Partner: UNT Libraries Government Documents Department