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Calculation of fuel pin failure timing under LOCA conditions

Description: The objective of this research was to develop and demonstrate a methodology for calculation of the time interval between receipt of the containment isolation signals and the first fuel pin failure for loss-of-coolant accidents (LOCAs). Demonstration calculations were performed for a Babcock and Wilcox (B W) design (Oconee) and a Westinghouse (W) 4-loop design (Seabrook). Sensitivity studies were performed to assess the impacts of fuel pin burnup, axial peaking factor, break size, emergency core cooling system (ECCS) availability, and main coolant pump trip on these items. The analysis was performed using a four-code approach, comprised of FRAPCON-2, SCDAP/RELAP5/MOD3, TRAC-PF1/MOD1, and FRAP-T6. In addition to the calculation of timing results, this analysis provided a comparison of the capabilities of SCDAP/RELAP5/MOD3 with TRAC-PF1/MOD1 for large-break LOCA analysis. This paper discusses the methodology employed and the code development efforts required to implement the methodology. The shortest time intervals calculated between initiation of containment isolation and fuel pin failure were 11.4 s and 19.1 for the B W and W plants, respectively. The FRAP-T6 fuel pin failure times calculated using thermal-hydraulic data generated by SCDAP/RELAP5/MOD3 were more conservative than those calculated using data generated by TRAC-PF1/MOD1. 18 refs., 7 figs., 4 tabs.
Date: October 1, 1991
Creator: Jones, K.R.; Wade, N.L.; Siefken, L.J.; Straka, M. & Katsma, K.R.
Partner: UNT Libraries Government Documents Department

Scoping analysis of fission gas behavior in UO/sub 2/ fuel for an in-core thermionic reactor

Description: This paper gives results from a preliminary evaluation of swelling, and, in particular, swelling caused by the retained fission gases (Xe, Kr) in the fuel. Design parameters of the thermionic unit cell and its operating conditions were provided by Space Power, Inc. The analysis tool used is a mechanistic fission-gas behavior code, FASTGRASS, developed by J. Rest at ANL (Rest, 1984). In the FASTGRASS analysis, the UO/sub 2/ fuel column is taken as a solid cylinder. No external restraint is imposed on the UO/sub 2/ fuel other than the ambient pressure. The maximum fuel burnup is assumed to be 5 at%. The operating conditions specified for the UO/sub 2/ fuel fall into two categories: (a) low linear power (q' = 5-8 kW/ft) and high fuel surface temperature (T/sub s/ = 1700, 1800, 1900 K) and (b) high linear power (q' = 8-12 kW/ft) and low fuel surface temperature (T/sub s/ = 1300, 1400, 1500, 1600 K). Taking the upper and lower linear powers in each category in combination with the specified fuel surface temperatures resulted in a total of 14 cases for the scoping analysis. The total irradiation time (based on linear power and 5 at% burnup) for these cases varies from 2.5 to 6 years.
Date: October 1, 1984
Creator: Liu, Y.Y. & Rest, J.
Partner: UNT Libraries Government Documents Department

SPARC-90: A code for calculating fission product capture in suppression pools

Description: This report describes the technical bases and use of two updated versions of a computer code initially developed to serve as a tool for calculating aerosol particle retention in boiling water reactor (BWR) pressure suppression pools during severe accidents, SPARC-87 and SPARC-90. The most recent version is SPARC-90. The initial or prototype version (Owczarski, Postma, and Schreck 1985) was improved to include the following: rigorous treatment of local particle deposition velocities on the surface of oblate spherical bubbles, new correlations for hydrodynamic behavior of bubble swarms, models for aerosol particle growth, both mechanistic and empirical models for vent exit region scrubbing, specific models for hydrodynamics of bubble breakup at various vent types, and models for capture of vapor iodine species. A complete user's guide is provided for SPARC-90 (along with SPARC-87). A code description, code operating instructions, partial code listing, examples of the use of SPARC-90, and summaries of experimental data comparison studies also support the use of SPARC-90. 29 refs., 4 figs., 11 tabs.
Date: October 1, 1991
Creator: Owczarski, P.C. & Burk, K.W. (Pacific Northwest Lab., Richland, WA (United States))
Partner: UNT Libraries Government Documents Department

Review of tube support plate analysis for steam generators of Millstone Unit II Nuclear Power Plant

Description: Magnetite growth in steam generator tube support plates was observed in the Millstone Unit II Nuclear Power Plant. If growth is allowed to continue, the tube may eventually fail resulting from plate shifting and the squeezing action of the growing magnetite. The corrective actions undertaken by the Northeast Nuclear Energy Company (NNECO) for this effect have been summarized in a report submitted to the U.S. Nuclear Regulatory Commission (NRC) entitled, Millstone Unit No. II Steam Generator Repairs and Corrective Actions, Docket No. 50-336. The analytical study part of this report is reviewed here, and conclusions and recommendations for further research are given.
Date: October 1, 1978
Creator: Ma, S.M. & Lu, S.C.H.
Partner: UNT Libraries Government Documents Department

Primary system boron dilution analysis

Description: The results are presented for an analysis conducted to determine the potential paths through which nonborated water or water with insufficient boron concentration might enter the LOFT primary coolant piping system or reactor vessel to cause dilution of the borated primary coolant water. No attempt was made in the course of this analysis to identify possible design modifications nor to suggest changes in administrative procedures or controls.
Date: October 10, 1978
Creator: Crump, R.J.; Naretto, C.J.; Borgen, R.A. & Rockhold, H.C.
Partner: UNT Libraries Government Documents Department

Fast-mixed spectrum reactor. Progress report for 1980

Description: Reactor physics, fuel cycle, thermal-hydraulics and fuel cycle cost studies have been performed for this concept and are reported. The most serious drawback of previous FMSR designs, namely the level of irradiation damage to the stainless steel of the cladding and duct materials, has been greatly reduced by the new design. The peak fuel burnup level is also reduced. Work continued on earlier FMSR designs, and in particular, the centrally-moderated FMSR. Emphasis was placed on defining the first core and then the total sequence of core histories over the 30-year life of the reactor. It was found possible to define a two-year fuel cycle with limited reactivity swing over the cycle. Fuel cycle cost studies were begun. The results indicate a modest fuel cycle cost advantage for the FMSR, but the basic cost assumptions must be improved for metal fuel. Improved thermal-hydraulic analysis capabilities have greatly improved the understanding of heat transfer behavior.
Date: October 1, 1980
Creator: Fischer, G.J.; Galperin, A.; Shenoy, S. & Atefi, B.
Partner: UNT Libraries Government Documents Department

Category I structures program. [PWR; BWR]

Description: The objective of the Category I Structure Program is to supply experimental and analytical information needed to assess the structural capacity of Category I structures (excluding the reactor cntainment building). Because the shear wall is a principal element of a Category I structure, and because relatively little experimental information is available on the shear walls, it was selected as the test element for the experimental program. The large load capacities of shear walls in Category I structures dictates that the experimental tests be conducted on small size shear wall structures that incorporates the general construction details and characteristics of as-built shear walls.
Date: October 27, 1981
Creator: Endebrock, E.G. & Dove, R.C.
Partner: UNT Libraries Government Documents Department

Internal hydriding in irradiated defected Zircaloy fuel rods: A review (LWBR Development Program)

Description: Although not a problem in recent commercial power reactors, including the Shippingport Light Water Breeder Reactor, internal hydriding of Zircaloy cladding was a persistent cause of gross cladding failures during the 1960s. It occurred in the fuel rods of water-cooled nuclear power reactors that had a small cladding defect. This report summarizes the experimental findings, causes, mechanisms, and methods of minimizing internal hydriding in defected Zircaloy-clad fuel rods. Irradiation test data on the different types of defected fuel rods, intentionally fabricated defected and in-pile operationally defected rods, are compared. Significant factors affecting internal hydriding in defected Zircaloy-clad fuel rods (defect hole size, internal and external sources of hydrogen, Zircaloy cladding surface properties, nickel alloy contamination of Zircaloy, the effect of heat flux and fluence) are discussed. Pertinent in-pile and out-of-pile test results from Bettis and other laboratories are used as a data base in constructing a qualitative model which explains hydrogen generation and distribution in Zircaloy cladding of defected water-cooled reactor fuel rods. Techniques for minimizing internal hydride failures in Zircaloy-clad fuel rods are evaluated.
Date: October 1, 1987
Creator: Clayton, J C
Partner: UNT Libraries Government Documents Department

Progress in evaluation and improvement in nondestructive examination reliability for inservice inspection of Light Water Reactors (LWRs) and characterize fabrication flaws in reactor pressure vessels

Description: This paper is a review of the work conducted under two programs. One (NDE Reliability Program) is a multi-year program addressing the reliability of nondestructive evaluation (NDE) for the inservice inspection (ISI) of light water reactor components. This program examines the reliability of current NDE, the effectiveness of evolving technologies, and provides assessments and recommendations to ensure that the NDE is applied at the right time, in the right place with sufficient effectiveness that defects of importance to structural integrity will be reliably detected and accurately characterized. The second program (Characterizing Fabrication Flaws in Reactor Pressure Vessels) is assembling a data base to quantify the distribution of fabrication flaws that exist in US nuclear reactor pressure vessels with respect to density, size, type, and location. These programs will be discussed as two separate sections in this report. 4 refs., 7 figs.
Date: October 1, 1991
Creator: Doctor, S.R.; Bowey, R.E.; Good, M.S.; Friley, J.R.; Kurtz, R.J.; Simonen, F.A. et al.
Partner: UNT Libraries Government Documents Department

A summary of lessons learned at the Shippingport Station Decommissioning Project (SSDP)

Description: This paper describes the lessons learned from a management perspective during decommissioning. The lessons learned are presented in a chronological sequence during the life of the project up to the present time. The careful analysis of the lessons learned and the implementation of corresponding actions have contributed toward improving the effectiveness of decommissioning as time progresses. The lessons learned should be helpful in planning future decommissioning projects.
Date: October 1, 1987
Creator: Crimi, F.P. & Mullee, G.R.
Partner: UNT Libraries Government Documents Department

LOFT primary coolant addition and Control Piping System stress analysis

Description: A stress analysis was performed on the Primary Coolant Addition and Control Piping System to determine if it met the conditions of the ASME Code, Section III, for Class 2 components. Results indicate that the Addition and Control System does not meet Section III criteria as the system is now installed. Only hanger (support) modifications are required to bring the stresses within the limits set forth in the Code. A design temperature of 459/sup 0/F was assumed for the analysis. The specified design temperature of 650/sup 0/F has been revised by ECRA's L-5713 and L-5714.
Date: October 31, 1978
Creator: Murdock, S.M.
Partner: UNT Libraries Government Documents Department

Ground-based testing of space nuclear power plants

Description: Small nuclear power plants for space applications are evaluated according to their testability in this two part report. The first part introduces the issues involved in testing these power plants. Some of the concerns include oxygen embrittlement of critical components, the test environment, the effects of a vacuum environment on materials, the practically of racing an activated test chamber, and possible testing alternative the SEHPTR, king develop at the Idaho National Engineering Laboratory. 10 refs., 6 figs., 1 tab.
Date: October 22, 1990
Creator: McDonald, T.G.
Partner: UNT Libraries Government Documents Department

Boilup threshold for the bottled-up transition phase pool. [LMFBR]

Description: Since the inception of the hypothesized transition phase, for the late stages of a postulated LMFBR accident, there has been a continual effort to characterize the anticipated conditions of such a hypothetical state. To date, several techniques and methods have been employed to analyze the potential for energetic criticality. As part of this effort, an arbitrary criterian of monotonical dispersiveness has been employed as the measure of diminished recriticality potential. The various attempts to demonstrate monotonic dispersiveness have included experimental demonstrations, theoretical approaches, and integrated analysis using both. As part of this treatment, flow regime maps have been devised as a convenient method for inferring the state of dispersiveness. They included bubbly, churn turbulent, foam and drop fluidized regimes. Of these, foam and drop fluidized regimes were considered the most dispersive. The main thrust of the analysis to date, including flow regime maps, relates primarily to the open pool configuration. However, the bottled configuration may be the pertinent geometry. To date, no reliable escape path has been demonstrated for the advanced stages of core disruption, although strong potential escape mechanisms have been identified and are currently being analyzed. The bottled pool is examined in this paper.
Date: October 1, 1978
Creator: Martin, F. J.
Partner: UNT Libraries Government Documents Department

Evaluation of EPRI nuclear power division research topics supportive of HTGR technology

Description: For HTGR commercialization studies, an LWR/HTGR Technology Transfer program was devised. Candidate programs were identified out of a total of 208 EPRI NPD (Nuclear Power Division) projects. Of these, 26 project areas presented the highest probability for technology transfer. (DLC)
Date: October 6, 1978
Creator: Available, Not
Partner: UNT Libraries Government Documents Department

Estimation of fracture toughness of cast stainless steels in LWR (light water reactor) systems

Description: A procedure and correlations are presented for predicting fracture toughness J-R curves and impact strength of aged cast stainless steels from known material information. The saturation'' fracture toughness of a specific cast stainless steel, i.e., the minimum fracture toughness that would ever be achieved for the material after long-term service, is estimated from the degree of embrittlement at saturation. Degree of embrittlement is characterized in terms of room-temperature Charpy-impact energy. The variation of the impact energy at saturation for different materials is described in terms of a material parameter {Phi}, which is determined from the chemical composition and ferrite morphology. The fracture toughness J-R curve for the material is then obtained from correlations between room-temperature Charpy-impact energy and fracture toughness. Fracture toughness as a function of time and temperature of reactor service is estimated from the kinetics of embrittlement, which is determined from the chemical composition. Examples for estimating impact strength and fracture toughness of cast stainless steel components during reactor service are described. A common lower-bound'' J-R curve for cast stainless steels with unknown chemical composition is also defined. 15 refs., 19 figs., 3 tabs.
Date: October 1, 1990
Creator: Chopra, O.K.
Partner: UNT Libraries Government Documents Department

Sodium pool fire model for CONACS code. [LMFBR]

Description: The modeling of sodium pool fires constitutes an important ingredient in conducting LMFBR accident analysis. Such modeling capability has recently come under scrutiny at Westinghouse Hanford Company (WHC) within the context of developing CONACS, the Containment Analysis Code System. One of the efforts in the CONACS program is to model various combustion processes anticipated to occur during postulated accident paths. This effort includes the selection or modification of an existing model and development of a new model if it clearly contributes to the program purpose. As part of this effort, a new sodium pool fire model has been developed that is directed at removing some of the deficiencies in the existing models, such as SOFIRE-II and FEUNA.
Date: October 19, 1982
Creator: Yung, S.C.
Partner: UNT Libraries Government Documents Department

TS-1 and TS-2 transient overpower tests on FFTF fuel

Description: The TS-1 and TS-2 TREAT transient experiments subjected a low burnup (2 MWd/kg) and a medium burnup (58 MWd/kg), respectively, FFTF irradiated fuel pin to unprotected 5 cents/s overpower transient conditions. The fuel pin failure response was similar in the two tests, which demonstrated a large margin to failure (P/P/sub 0/ > 3) and a favorable upper level failure location. Thus, for these transient conditions, burnup effects on transient performance appeared to be minimal in the range tested. Pin disruption in the medium burnup TS-2 test was more severe due to the higher fission gas pressurization, but failure occurred at only a 5% lower power level than for the low burnup TS-1 fuel pin. Both tests exhibited axial extrusion of molten fuel to the region above the fuel column several seconds before pin failure, demonstrating a potentially beneficial inherent safety mechanism to delay failure and mitigate accident consequences.
Date: October 3, 1985
Creator: Pitner, A.L.; Ferrell, P.C.; Culley, G.E. & Weber, E.T.
Partner: UNT Libraries Government Documents Department

Power Burst Facility/Boron Neutron Capture Therapy Program for cancer treatment

Description: This bulletin discusses activities during this reporting period in the areas of: supporting technology development; large animal model studies; melanoma project; human studies; stability, pharmacology, and toxicology of drugs; and PBF technical support. (FL)
Date: October 1, 1990
Creator: Ackermann, A.L. (ed.)
Partner: UNT Libraries Government Documents Department

THATCH: A computer code for modelling thermal networks of high- temperature gas-cooled nuclear reactors

Description: This report documents the THATCH code, which can be used to model general thermal and flow networks of solids and coolant channels in two-dimensional r-z geometries. The main application of THATCH is to model reactor thermo-hydraulic transients in High-Temperature Gas-Cooled Reactors (HTGRs). The available modules simulate pressurized or depressurized core heatup transients, heat transfer to general exterior sinks or to specific passive Reactor Cavity Cooling Systems, which can be air or water-cooled. Graphite oxidation during air or water ingress can be modelled, including the effects of added combustion products to the gas flow and the additional chemical energy release. A point kinetics model is available for analyzing reactivity excursions; for instance due to water ingress, and also for hypothetical no-scram scenarios. For most HTGR transients, which generally range over hours, a user-selected nodalization of the core in r-z geometry is used. However, a separate model of heat transfer in the symmetry element of each fuel element is also available for very rapid transients. This model can be applied coupled to the traditional coarser r-z nodalization. This report described the mathematical models used in the code and the method of solution. It describes the code and its various sub-elements. Details of the input data and file usage, with file formats, is given for the code, as well as for several preprocessing and postprocessing options. The THATCH model of the currently applicable 350 MW{sub th} reactor is described. Input data for four sample cases are given with output available in fiche form. Installation requirements and code limitations, as well as the most common error indications are listed. 31 refs., 23 figs., 32 tabs.
Date: October 1, 1991
Creator: Kroeger, P.G.; Kennett, R.J.; Colman, J. & Ginsberg, T. (Brookhaven National Lab., Upton, NY (United States))
Partner: UNT Libraries Government Documents Department

Irradiation performance of low-enriched uranium fuel elements

Description: The status of the testing and evaluation of full-sized experimental low- and medium-enriched uranium fuel elements in the Oak Ridge Research Reactor is presented. Medium-enriched elements containing oxide and aluminide have been completely evaluated at burnups up to 75%. A low-enriched U/sub 3/Si/sub 2/ element has been evaluated at 41% burnup. Other silicide and oxide elements have completed irradiation satisfactorily to burnups of 75% and are now being evaluated. All results to date confirm the expected good performance of these elements in the medium power research reactor environment.
Date: October 14, 1984
Creator: Copeland, G.L.; Hofman, G.L. & Snelgrove, J.L.
Partner: UNT Libraries Government Documents Department

A study of the effect of fabrication variables on the void content and quality of fuel plates

Description: The control of void content and quality of dispersion type fuel plates fabricated for research and test reactors are issues of concern to plate fabricators. These two variables were studied by examining the data for various geometries of fuel plates fabricated at ANL. It was found that the porosity of a fuel plate can be increased by: (1) decreasing the fuel particle size, (2) increasing the fuel particle surface roughness, (3) increasing the matrix strength, (4) decreasing the rolling temperature, (5) decreasing the final fuel zone thickness, and (6) increasing the volume percentage of the fuel. Porosity formation is controlled by bulk movement and deformation and/or fracture of particles. The most important factor is the flow stress of the matrix material. Lowering the flow stress will decrease the plate porosity. The percentage of plates with fuel-out-of-zone is a function of the fuel material and the loading. The highest percentage of plates with fuel-out-of-zone were those with U3Si2 which is at this time the most commonly used silicide fuel.
Date: October 1, 1986
Creator: Wiencek, T.C.
Partner: UNT Libraries Government Documents Department

Two-phase natural-circulation experiments in a test facility modeled after Three Mile Island Unit-2. Final report

Description: A series of natural circulation experiments was conducted in a test facility that was configured after the primary and the secondary cooling systems of TMI-2. Results support the feasibility of core residual heat removal by two-phase natural circulation. Tests with noncondensable gas in the primary system indicate that two-phase natural circulation is quite tolerant of the presence of noncondensable gas. The different modes of natural circulation were discovered. Mode 1, during which only saturated steam flows in the hot leg, accomplishes the heat removal via phase changes in the vessel and in the steam generator tubes. Mode 2, during which a percolating flow exists in the hot leg, removes the heat by means of a much faster circulation in the primary loop.
Date: October 1, 1981
Creator: Kiang, R.L.
Partner: UNT Libraries Government Documents Department