Certain postulated transient conditions in a PWR may cause the flow rate and the temperature of the coolant entering the core from one of the cold leg loops to differ significantly from those in the other loops. When the flow rate and/or temperature of the loops are not balanced, the resulting thermal mixing at the core inlet will become nonuniform, and the assumption of perfect mixing is not only unwarranted, but may lead to an erroneous prediction of radial power distribution. The objective of this paper is to describe the results of the three-dimensional steady-state calculations for the analysis of thermal mixing in a PWR. The specific case chosen for the present study is one of the tests conducted in the Oconee-1 PWR. Those tests were conducted to determine the extent of downcomer and lower plenum mixing in response to various temperature differences or flow imbalances between the two coolant loops. These tests were performed specifically to obtain coolant temperature profiles at the core inlet, core outlet, and various axial locations in the core under imposed offset loop conditions. The calculations were performed using the COMMIX-1A computer code.

The response of a four-loop Westinghouse pressurized water reactor to SBLOCAs initiated as a result of an anticipated transient without scram (ATWS) has been analyzed using the RELAP5 computer code. The ATWS is initiated by a loss-of-feedwater, and the small breaks were due to either one or three stuck-open safety valves or reactor coolant pump seal failure. For the cases analyzed, the results show that a LOF-ATWS followed by a SBLOCA does not have more safety significance than that found when each accident is analyzed independently of one another.

Space power systems must have both high power density (kW/m/sup 3/) and high specific power output (kW/kg), while maintaining high reliability. Several power source and conversion options are available. A sketch of the Rankine system under consideration is shown. The design for the 100 kWe system analyzed here uses the same basic design concept as described by Jones for a 5 MWe system. In brief, it consists of a uranium nitride fuel pin reactor core with a tungsten/lithium hydride shield. Potassium at approx. 25% quality exits the reactor core. A separator is used to provide potassium vapor (at approx. 99%+ quality) to a full impulse turbine. Liquid from the separator is recirculated back to the core inlet using jet pumps. A heat pipe radiator condenses the vapor exiting the turbine while a separate turbine driven boiler feed pump returns the condensed liquid to the core inlet. Sizes and weights for the various system components were determined using design algorithms or were scaled from previous design studies.

The population balance methodology is described and applied to the transport and capture of polydispersed colloids in packed columns. The transient model includes particle growth, capture, convective transport, and dispersion. We also follow the dynamic accumulation of captured colloids on the solids. The multidimensional parabolic partial differential equation was solved by a recently enhanced method of characteristics technique. This computational technique minimized numerical dispersion and is computationally very fast. The FORTRAN 77 code ran on a VAX-780 in less than a minute and also runs on an IBM-AT using the Professional FORTRAN compiler. The code was extensively tested against various simplified cases and against analytical models. The packed column experiments by Saltelli et al. were re-analyzed incorporating the experimentally reported size distribution of the colloid feed material. Colloid capture was modeled using a linear size dependent filtration function. The effects of a colloid size dependent filtration factor and various initial colloid size distributions on colloid migration and capture were investigated. Also, we followed the changing colloid size distribution as a function of position in the column. Some simple arguments are made to assess the likelihood of colloid migration at a potential NTS Yucca Mountain waste disposal site. 10 refs., 3 …

Analysis of drawdown data from short-term tests of the L.R. Sweezy No. 1 well indicates a formation permeability of around 126 md, and a flow-rate dependent skin. Conventional analysis techniques, however, are inadequate for analyzing the buildup data. A computer simulation of the production history of the Sweezy well, using formation properties inferred from the drawdown tests, displays good agreement with the drawdown data; such agreement was, however, not obtained for the buildup phase of the tests. Parametric calculations to investigate the anomalous reservoir response suggest that this behavior is, primarily, the result of stress-induced hysteresis in formation permeability and compressibility. The pressure response observed during long-term flow tests may involve additional drive mechanisms such as leaky boundaries, shale recharge and/or long-term formation creep.

The Nevada Nuclear Waste Storage Investigations Project of the US Department of Energy provides that agency with data for evaluating volcanic tuff beneath Yucca Mountain, Nevada, to determine its suitability for a potential repository of high-level radioactive waste. Thickness of the unsaturated zone, which consists of fractured, welded and nonwelded tuff, is about 1640 to 2460 feet (500 to 750 meters). One question to be resolved is an estimate of minimum ground-water traveltime from the disturbed zone of the potentail repository to the accessible environment. Another issue is the potential for diffusive or convective gaseous transport of radionuclides from an underground facility in the unsaturated zone to the accessible environment. Gas samples were collected at intervals to a depth of 1200 feet from the unsaturated zone at Yucca Mountain, Nevada. Samples were analyzed for major atmospheric gases; carbon dioxide in the samples was analyzed for carbon-14 activity and for {delta}2!{sup 3}C; water vapor in the samples was analyzed for deuterium and oxygen-18. These data could provide insight into the nature of unsaturated zone transport processes. 15 refs., 4 figs., 4 tabs.

Experimental data suggest that steam produced within a particle bed which is being quenched by flow from an overlying pool of water has a strong potential for being superheated as a result of heat transfer from the unquenched particulate in the dry channels of the bed. A model is presented for the debris bed quench process which considers the effects of steam superheat on both the bed heat flux and on the quench front propagation characteristics. Calculations are presented which demonstrate the effects of steam superheat. A preliminary comparison of the model with experimental data from packed bed quench experiments is also presented. 11 refs., 5 figs.

The operator guidelines were followed for both transients described. Both transients resulted in SG overfill and the tube-rupture flow did not terminate in either transient. The following statements can be deducted from the results of the calculations: the tube-rupture flow could not be stopped for either case during 2600 s (43 min) of transient time; each accident scenario resulted in SG overfill; both SGs overfilled by 1600 s (27 min) and 1800 s (30 min) for Cases 1 and 2, respectively; conditions for isolation of the SGs were not reached; and core subcooling was not lost in either case but the upper head was voided in Case 2. Comparison of the cooldown rates in the two cases after 1200 s (20 min) shows that these rates are equal (i.e., restart of the RCPs did not change the primary-system cooldown rate). However, in Case 2, a steam bubble was formed in the upper head, which did not disappear during the simulated time. One of the immediate actions in the guidelines was to fill both SGs to 95% level. This step was almost unnecessary because the tube-rupture flow was large enough that the 15.4 - mK/s (100 - /sup 0/F h) cooldown-rate …

Procedures for analysis of resonance neutron cross-section data have been implemented in a state-of-the-art computer code SAMMY, developed at the Oak Ridge Electron Linear Accelerator (ORELA) at Oak Ridge National Laboratory. A unique feature of SAMMY is the use of Bayes' equations to determine ''best'' values of parameters, which permits sequential analysis of data sets (or subsets) while giving the same results as would be given by a simultaneous analysis. Another important feature is the inclusion of data-reduction parameters in the fitting procedure. Other features of SAMMY are also described.

The effects of reservoir layering and gravity segregation on nonisothermal injection and falloff tests are investigated. Results show that layering does not affect injection or falloff data if all the layers are permeable and accept fluids from the wellbore. In such cases, the average permeability, skin factor, and distance to the thermal front can be calculated using the techniques developed for homogeneous reservoirs. Special considerations have to be taken for cases where several layers are impermeable or are permeable but do not accept fluids of the well face. In the first case (impermeable layers), knowledge of the total thickness of the permeable layers is required for the existing techniques to be applied successfully. In the second case, the existing techniques cannot be applied, but characteristic responses from injection and falloff test are seen; therefore, this case can be identified easily. 13 refs., 8 figs.

A sophisticated program for isotopic analysis of plutonium gamma-ray spectra using small computers has been developed. It is implemented on a DEC LSI-11/2 configured in a portable unit without a mass storage device for use by IAEA inspectors in the field. Only the positions of the 148-keV /sup 241/Pu and 208-keV /sup 237/U peaks are needed as input. Analysis is completed in 90 seconds by fitting isotopic component response functions to peak multiplets. 9 refs., 2 figs., 1 tab.

This paper presents some of the current capabilities of the three-dimensional piping code SHAPS and demonstrates their usefulness in handling analyses encountered in typical LMFBR studies. Several examples demonstrate the utility of the SHAPS code for problems involving fluid-structure interactions and seismic-related events occurring in three-dimensional piping networks. Results of two studies of pressure wave propagation demonstrate the dynamic coupling of pipes and elbows producing global motion and rigorous treatment of physical quantities such as changes in density, pressure, and strain energy. Results of the seismic analysis demonstrate the capability of SHAPS to handle dynamic structural response within a piping network over an extended transient period of several seconds. Variation in dominant stress frequencies and global translational frequencies were easily handled with the code. 4 refs., 10 figs.

Presented is an analysis of the results of the CSTI-1, CSTI-3, and CWTI-11 reactor material experiments in which a jet of molten corium initially at 3080/sup 0/K was directed downward upon a stainless steel plate. The experiments are a continuation of a program of reactor material tests investigating LWR severe accident phenomena. Objective of the present analysis is to determine the existence or nonexistence of a corium crust during impingement from comparison of the measured heatup of the plate (as measured by thermocouples imbedded immediately beneath the steel surface) with model calculations assuming alternately the presence and absence of a stable crust during impingement.

An analysis is presented of the crust stability and heat transfer behavior in the CSTI-1, CSTI-3, and CWTI-11 reactor material experiments in which a jet of molten oxide fuel at approx. 160/sup 0/K above its freezing temperature was impinged normally upon stainless steel plates initially at 300 and 385 K. The major issue is the existence of nonexistence of a stable solidified layer of fuel, or crust, interstitial to the flowing hot fuel and the steel substrate, tending to insulate the steel from the hot molten fuel. A computer model was developed to predict the heatup of thermocouples imbedded immediately beneath the surface of the plate for both of the cases in which a stable crust is assumed to be either present or absent during the impingement phase. Comparison of the model calculations with the measured thermocouple temperatures indicates that a protective crust was present over nearly all of the plate surface area throughout the impingement process precluding major melting of the plate steel. However, the experiments also show evidence for very localized and isolated steel melting as revealed by localized and isolated pitting of the steel surface and the response of thermocouples located within the pitted region.

Extensive study of the Cerro Prieto geothermal field has provided much geologic and thermodynamic data of its structurally-complex, liquid-dominated reservoir. Several of the studies investigated the resource characteristics of fluid and energy flow. An early report by Mercado (1975) showed that the heat source for the part of the reservoir under development, now called Cerro Prieto I (CPI), originated in the eastern part of the field. Subsequent studies confirmed the flow of hot water from the east. A summary of several experimental and numerical studies of fluid and energy transport in the field was given by Lippmann and Bodvarsson (1983). The hydrogeologic model of Halfman et al. (1982) shows hot-water flow from the east divided into a shallow (alpha) aquifer at about 120Om and a deeper (beta) aquifer at about 170Om depth. A cross section along an east-west direction shows a central upflow to the two aquifers and uncertain geology beyond the western border of the field near well M-9. It also shows a fault dividing the line of border wells at M-29 from the inner wells at M-25 to the east. The hydrogeology of the field was described by Sanchez and de la Pena (1981) as an alluvial unit …

A study of the seismic sloshing response of a large pool-type reactor tank with several deck-mounted components is presented. The main objective of the study is to investigate the effects of internal components on the sloshing response and to determine the sloshing loads on the components. The study shows that the presence of internal components can significantly change the dynamic characteristics of the sloshing motion. The sloshing frequencies of a tank with internal components are considerably higher than those of a tank without internals. The higher sloshing frequencies reduce the sloshing wave height on the free surface but the dynamic pressures of the fluid are increased. The effects of seismic isolation on sloshing response are also presented.

The analysis develops a classical theory of how a signal evolves from the initial incoherent spontaneous emission in long undulators. The theory is based on the coupled Klimontovich-Maxwell equations. Formulas for the radiated power, spectral characteristics and electron correlations are derived. The saturation due to nonlinear effects is studied using a quasi-linear extension of the theory. The results agree reasonably well with the recent Livermore experiment in the microwave range. Performance of a possible high-gain free electron laser in a short-wavelength region is evaluated.

The Creare 1/5-scale Phase II experiments which model fluid and thermal mixing of relatively cold high pressure injection (HPI) water into a cold leg of a full-scale pressurized water reactor (PWR) having loop flow are analyzed and found that they cannot achieve complete similarity with respect to characteristic Reynolds and Froude numbers and developing hydrodynamic entry length. Several analyses show that these experiments fall into two distinct regimes of mixing: momentum controlled and gravity controlled (stratification). 18 refs., 9 figs.

A new analytical technique based on resonant ionization of krypton with a vacuum ultraviolet (VUV) laser source was used to determine low-level /sup 81/Kr concentrations in groundwater. The long half-life (210,000 years) and low concentration (1.3 x 10/sup 3/ /sup 81/Kr atoms per liter of modern water at 10/sup 0/C) make the detection of /sup 81/Kr by radioactive counting techniques extremely difficult. In this method, krypton gas was removed from water taken from an underground Swiss aquifer using standard cryogenic and chromatographic techniques. Stable krypton isotopes were then reduced by a factor of 10/sup 7/ by a two-stage isotopic enrichment cycle using a commercially available mass spectrometer. The enriched gas containing about 10/sup 8/ stable krypton atoms and about 10/sup 3/ atoms of /sup 81/Kr was implanted into a silicon disc. This disc was then placed in the high vacuum final counting chamber and the krypton was released by laser annealing. This chamber contained a quadrupole mass spectrometer which used a pulsed VUV laser source as the ionizer. The measured signal indicated that the sample contained 1200 (+-300) atoms of /sup 81/Kr.

Intergranular stress corrosion cracking (IGSCC) of sensitized Type 304 stainless steel (SS) has been a recurrent problem in the high-temperature water environment of boiling-water-reactors (BWRs) over the past two decades. The synergistic effects of environmental and material variables on stress corrosion cracking (SCC) of Type 304 SS were investigated at 289/sup 0/C by means of constant-extension-rate-tensile (CERT) tests at a strain rate of 1 x 10/sup -6//s. Correlations among environmental variables (dissolved oxygen and impurity concentrations, viz., H/sub 2/SO/sub 4/, steady-state open-circuit electro-chemical potential) and the SCC susceptibility parameters have been determined. The extensive results over a wide range of open-circuit corrosion potential conditions were analyzed by a model which accounts for the effects of environmental variables, microstructure (e.g., degree of sensitization) and strain rate. The results are consistent with a slip-dissolution mechanism for SCC. Furthermore, representation of the dependence of corrosion potential and average crack growth rate on the dissolved oxygen concentration of the water by a simple mathematical function, in conjunction with the theoretical model, enables predictions of both strain rate and environmental effects on the SCC susceptibility of sensitized Type 304 SS. 12 refs., 7 figs.

The materials CeSn/sub 3/, USn/sub 3/, and NpSn/sub 3/ are all moderately heavy Fermion compounds with electronic specific heat coefficients of 73, 169, and 242 mJ/mole K/sup 2/. CeSn/sub 3/ is known as a mixed valent system and NpSn/sub 3/ is a weak itinerant antiferromagnet. All three are strongly enhanced. Being in the relatively simple Au/sub 3/Cu structure, they form an excellent set of materials to study as representatives of strongly enhanced systems. One would like to ascertain what properties can be determined from band calculations based on density functional theory in the local density approximation. It has already been shown that the Fermi surface topology of CeSn/sub 3/ can be well described in this way even though the experimental masses are much larger than the band results. The enhancement factor for USn/sub 3/ is even larger and NpSN/sub 3/ is indeed predicted to go magnetic. We present here fully relativistic SCF calculations for these materials and discuss the relation between our results and what is known experimentally.

The flow and bottomhole pressure data have been analyzed for the two sands (Nos. 8 and 9) tested by the Gladys McCall No. 1 well. The more productive sand (No. 8) appears to be bounded by two linear faults at distances of approx. 740 feet and approx. 1360 feet from the well and there appears to be a decrease in the formation transmissivity away from the well. The formation properties inferred from the well test analysis have been used with a reservoir simulator to match the bottomhole drawdown/buildup history measured during the Reservoir Limits Test of Sand Zone No. 8. Wellhead pressure data measured during the long-term production testing of Sand Zone No. 8 have been employed to estimate the corresponding downhole pressures. The simulation model based solely on the Reservoir Limits Test is found to be in remarkably good agreement with the estimated bottomhole pressures for the first six months of production testing, but enlargement of the reservoir volume, by moving the boundary most remote from the well outward, is required to adequately match the full production history.

The radiation flux in a concrete block building with windows exposed to irradiation at the ORNL Tower Shielding Facility was measured directly with Bonner balls and the measured flux compared with calculations performed with Oak Ridge's 3-D discrete ordinates code TORT. 6 refs., 2 figs. (ACR)

The TREAT loss-of-flow tests L6 and L7 have been analyzed using the SAS3D accident analysis code. The impetus for the analysis was the need for experimentally supported fuel motion modeling in whole core accident studies performed in support of licensing of the Clinch River Breeder Reactor Project. The input prescription chosen for the SAS3D/SLUMPY fuel motion model gave reasonable agreement with the test results. Tests L6 and L7, each conducted with a cluster of three fuel pins, were planned to simulate key events in the loss-of-flow accident scenario for the Clinch River homogeneous reactor.