Data from top injection ECCS tests conducted at Columbia University have been analyzed as part of an effort to qualify the FLASH-6 computer program for performing post-blowdown heat transfer calculations for the LWBR Safety Analysis. These experiments, which employed a full-scale fuel assembly with electrical heater rods to simulate an inlet rupture for a pressurized water reactor, provided test conditions and rod cooling mechanisms quite similar to those encountered in the postulated LWBR cold leg break loss-of-coolant accident. Clad temperature predictions were obtained using both the modified Dittus-Boelter and the Dougall-Rohsenow correlations to evaluate beyond CHF heat transfer coefficients. Overall comparisons using the FLASH calculated flow rates indicated that the rod temperature calculations were conservative using either of the heat transfer correlations because virtually none of the coolant was calculated to penetrate the heated test assembly. Heat transfer model comparisons were also performed by adjusting the calculation so that coolant was injected directly into the top of the rod bundle to simulate the experimentally observed flow conditions. Once this downflow was forced, conservative temperature predictions were obtained using the Dougall-Rohsenow correlation, whereas the modified Dittus-Boelter beyond CHF option yielded non-conservative results.

A model was developed to estimate the radiation dose commitments received by people in the vicinity of a facility that releases radionuclides into the atmosphere. This model considers dose commitments resulting from immersion in the plume, ingestion of contaminated food, inhalation of gaseous and suspended radioactivity, and exposure to ground deposits. The dose commitments from each of these pathways is explicitly considered for each radionuclide released into the atmosphere and for each daughter of each released nuclide. Using the release rate of only the parent radionuclide, the air and ground concentrations of each daughter are calculated for each position of interest. This is considered to be a significant improvement over other models in which the concentrations of daughter radionuclides must be approximated by separate releases.

The shape of the thorium absorption cross section near thermal energies was investigated. This shape is dominated by one or more negative energy resonances whose parameters are not directly known, but must be inferred from higher energy data. Since the integral quantity most conveniently describing the thermal cross section shape is the Westcottg-factor, effort was directed toward establishing this quantity to high precision. Three nearly independent g-factor estimates were obtained from measurements on a variety of foils in three different neutron spectra provided by polyethylene-moderated neutrons from a /sup 252/Cf source and from irradiations in the National Bureau of Standards ''Standard Thermal Neutron Density.'' The weighted average of the three measurements was 0.993 +- 0.004. This is in good agreement with two recent evaluations and supports the adequacy of the current cross section descriptions.

METER is a Monte Carlo computer program which can be used to simulate the interaction between independent random variables and their effects on one or more dependent random variables. The program is easy to use for simple simulations but is capable of accommodating complex simulations. METER processes input, generates random numbers from several common frequency distributions under user control, performs the simulation which the user has coded in FORTRAN, and displays results.

Pressure pulse tests were conducted with both solid and flexible test sections installed in a test vessel filled with room temperature water. A surge tank whose volume was approximately equal to that of the test vessel with the test section installed was connected to the test vessel by a /sup 1///sub 8/ inch I.D., 8 inch long surge line. Pressure pulses of magnitude up to 1275 psid and durations from 4.6 to 55.8 msec were generated in the test vessel with a drop hammer and piston pulse generator. FLASH-34 calculations show good agreement with the test data. In particular, FLASH-34 accurately predicts (a) the decrease in peak pressure and the increase in pulse duration due to the presence of a flexible test section, (b) the time delay between the occurrence of the pressure pulse in the test vessel and its arrival in the surge tank and (c) the magnitudes of the transient pressure differences between the test vessel and surge tank caused by the time delay. All of the structural responses were in the elastic range and were approximately quasi-static for the pulss tested. The test data versus calculation comparisons presented here provide preliminary qualification for FLASH-34 calculations of transient …

Critical heat flux experiments were conducted for upflow of water in a vertical 84 inch annular flow channel, 0.303 inch heated I.D. and 0.500 inch unheated O.D. Test data were obtained at pressures from 1200 to 2000 psia, mass velocities from 0.25 x 10/sup 6/ to 2.8 x 10/sup 6/ lb/hr-ft/sup 2/ and inlet temperatures ranging from 200 to 600/sup 0/F. Three different test sections were employed with (1) axially uniform heat flux over the 84 inch length to serve as a no-hot-patch data base, (2) axially uniform heat flux over 82 inches with a 1.5 heat flux ratio hot patch over the last two inches, and (3) axially uniform heat flux over 82 inches with a 2.25 heat flux ratio hot patch over the last two inches.

The principal hydraulic performance features of the Light Water Breeder Reactor are summarized in this report. The calculational models and procedures used for prediction of reactor flow and pressure distributions under steady-state and transient operating conditions are described. Likewise, the analysis models for evaluation of the static and dynamic performance characteristics of the hydraulically-balanced and hydraulically-buffered movable-fuel reactivity-control system are outlined. An extensive test program was conducted for qualification of the subject LWBR hydraulic evaluation models. The projected LWBR hydraulic performance is shown to fulfill design objectives and functional requirements.

This report describes the thermal evaluation of the core for the Shippingport Light Water Breeder Reactor. This core contains unique thermal-hydraulic features such as (1) close rod-to-rod proximity, (2) an open-lattice array of fuel rods with two different diameters and rod-to-rod spacings in the same flow region, (3) triplate orifices located at both the entrance and exit of fuel modules and (4) a hydraulically-balanced movable-fuel system coupled with (5) axial-and-radial fuel zoning for reactivity control. Performance studies used reactor thermal principles such as the hot-and-nominal channel concept and related nuclear/engineering design allowances. These were applied to models of three-dimensional rodded arrays comprising the core fuel regions.

In-reactor densification of ThO/sub 2/ and ThO/sub 2/--UO/sub 2/ fuel of low burnup and low power operation (hence low temperature) was assessed by measuring fuel pellet diameter changes. Pellet diameter changes ranged from nil in a large grain, low temperature thoria pellet (98.9 percent theoretical density) to -1.06 percent in a small grain, moderate temperature ThO/sub 2/-30 w/o UO/sub 2/ pellet (93.8 percent theoretical density). A correlation was established between quantity of small pores (<2.3 ..mu..m diameter) and as-fabricated fuel grain size. An empirical equation, based on densification (pore closure) plus fuel swelling, was formulated for pellet diameter change as a function of initial grain size and fuel burnup.

The PDQ-8 program is designed to solve the neutron diffusion, depletion problem in one, two, or three dimensions on the CDC-6600 and CDC-7600 computers. The three dimensional spatial calculation may be either explicit or discontinuous trial function synthesis. Up to five lethargy groups are permitted. The fast group treatment may be simplified P(3), and the thermal neutrons may be represented by a single group or a pair of overlapping groups. Adjoint, fixed source, one iteration, additive fixed source, eigenvalue, and boundary value calculations may be performed. The HARMONY system is used for cross section variation and generalized depletion chain solutions. The depletion is a combination gross block depletion for all nuclides as well as a fine block depletion for a specified subset of the nuclides. The geometries available include rectangular, cylindrical, spherical, hexagonal, and a very general quadrilateral geometry with diagonal interfaces. All geometries allow variable mesh in all dimensions. Various control searches as well as temperature and xenon feedbacks are provided. The synthesis spatial solution time is dependent on the number of trial functions used and the number of gross blocks. The PDQ-8 program is used at Bettis on a production basis for solving diffusion--depletion problems. The report describes …

Fission gas release data are presented from 51 fuel rods irradiated as part of the LWBR irradiations test program. The fuel rods were Zircaloy-4 clad and contained ThO/sub 2/ or ThO/sub 2/-UO/sub 2/ fuel pellets, with UO/sub 2/ compositions ranging from 2.0 to 24.7 weight percent and fuel densities ranging from 77.8 to 98.7 percent of theoretical. Rod diameters ranged from 0.25 to 0.71 inches and fuel active lengths ranged from 3 to 84 inches. Peak linear power outputs ranged from 2 to 22 kw/ft for peak fuel burnups up to 56,000 MWD/MTM. Measured fission gas release was quite low, ranging from 0.1 to 5.2 percent. Fission gas release was higher at higher temperature and burnup and was lower at higher initial fuel density. No sensitivity to UO/sub 2/ composition was evidenced.

Analytical methods and test data employed in the core design of bolted joints for the LWBR core are presented. The effects of external working loads, thermal expansion, and material stress relaxation are considered in the formulation developed to analyze joint performance. Extensions of these methods are also provided for bolted joints having both axial and bending flexibilities, and for the effect of plastic deformation on internal forces developed in a bolted joint. Design applications are illustrated by examples.

High purity ceramic grade urania (/sup 233/UO/sub 2/) used in manufacturing the fuel for the Light Water Breeder Reactor (LWBR) core was made from uranium-233 that was obtained by irradiating thoria under special conditions to result in not more than 10 ppM of uranium-232 in the recovered uranium-233 product. A developmental study established the operating parameters of the conversion process for transforming the uranium-233 into urania powder with the appropriate chemical and physical attributes for use in fabricating the LWBR core fuel. This developmental study included the following: (a) design of an ion exchange purification process for removing the gamma-emitting alpha-decay daughters of uranium-232, to reduce the gamma-radiation field of the uranium-233 during LWBR fuel manufacture; (b) definition of the parameters for precipitating the uranium-233 as ammonium uranate (ADU) and for reducing the ADU with hydrogen to yield a urania conversion product of the proper particle size, surface area and sinterability for use in manufacturing the LWBR fuel; (c) establishment of parameters and design of equipment for stabilizing the urania conversion product to prevent it from undergoing excessive oxidation on exposure to the air during LWBR fuel manufacturing operations; and (d) development of a procedure and a facility to reprocess …

Information on the physical, chemical, and mechanical properties of thoria and thoria-urania is reviewed and assessed. The properties discussed are those judged to be important for an understanding of the behavior of these oxides as nuclear fuel materials. Evaluation was made, where possible, of the effects of composition, material variables, temperature, and irradiation exposure. Data were taken from a review of the literature and from both published and unpublished data derived from the Light Water Breeder Reactor (LWBR) Program at the Bettis Atomic Power Laboratory. 30 figs., 23 tables, 163 refs.

ThO/sub 2/-UO/sub 2/ solid solutions fabricated as LWBR fuel pellets are examined for uniform uranium distribution by means of autoradiography. Kodak NTA plates are used. Images of inhomogeneities are 29 +- 10 microns larger in diameter than the high-urania segregations that caused them, due to the range of alpha particles in the emulsion, and an appropriate correction must be made. Photographic density is approximately linear with urania content in the region between underexposure and overexposure, but the slope of the calibration curve varies with aging and growth of alpha activity from the parasitic /sup 232/U and its decomposition products. A calibration must therefore be performed using two known points--the average photographic density (corresponding to the average composition) and an extrapolated background (corresponding to zero urania). As part of production pellet inspection, plates are evaluated by inspectors, who count segregations by size classes. This is supplemented by microdensitometer scans of the autoradiograph and by electron probe studies of the original sample if apparent homogeneity is marginal.

The purpose of the Department of Energy Water Cooled Breeder Program was to demonstrate pratical breeding in a uranium-233/thorium fueled core while producing electrical energy in a commercial water reactor generating station. A demonstration Light Water Breeder Reactor (LWBR) was successfully operated for more than 29,000 effective full power hours in the Shippingport Atomic Power Station. The reactor operated with an availability factor of 76% and had a gross electrical output of 2,128,943,470 kilowatt hours. Following operation, the expended core was examined and no evidence of any fuel element defects was found. Nondestructive assay of 524 fuel rods determined that 1.39 percent more fissile fuel was present at the end of core life than at the beginning, proving that breeding had occurred. This demonstrates the existence of a vast source of electrical energy using plentiful domestic thorium potentially capable of supplying the entire national need for many centuries. To build on the successful design and operation of the Shippingport Breeder Core and to provide the technology to implement this concept, several reactor designs of large breeders and prebreeders were developed for commercial-sized plants of 900--1000 Mw(e) net. This report summarizes the Water Cooled Breeder Program from its inception in 1965 …

The technical requirements for the Light Water Breeder Reactor (LWBR) fuel elements (fuel rods) imposed certain unique requirements for the low hafnium Zircaloy-4 tubing used as fuel rod cladding. This report describes, in detail, the tube manufacturing process, the product and process controls used, the inspections and tests performed, and the efforts involved in refining a commercial tube reducing process to produce tubes that would satisfy the requirements for LWBR fuel rod cladding.

This report presents the nuclear analysis and discusses the performance of the LWBR core at Shippingport during power operation from initial startup through end-of-life at 28,730 EFPH. Core follow depletion calculations confirmed that the reactivity bias and power distributions were well within the uncertainty allowances used in the design and safety analysis of LWBR. The magnitude of the core follow reactivity bias has shown that the calculational models used can predict the behavior of U/sup 233/-Th systems with closely spaced fuel rod lattices and movable fuel. In addition, the calculated final fissile loading is sufficiently greater than the initial fissile inventory that the measurements to be performed for proof-of-breeding evaluations are expected to confirm that breeding has occurred.

This report describes the use of a delayed neutron pellet assay gage to determine nondestructively the fissile content of fuel pellets during the manufacture of the Light Water Breeder Reactor (LWBR) core. The gage characteristics are described including the nature of the calibration curves and the gage sensitivities to pellet parameters. Statistical methods are derived for analyzing the data to obtain the mean weight percent of total uranium in each blend of fuel material as well as the loading precision of each fuel rod. The fissile loading of each fuel rod was determined to better than 0.25% at the 2 sigma level, and the fissile content of eight fuel compositions in the LWBR core was obtained to better than 0.1%. Use of this gage and the data analysis methods described in this report reduced the need for destructive chemical analysis of fuel pellets by a factor of two.

Seamless Zircaloy-4 tubing is utilized as fuel rod cladding in light water reactors. Water corrosion tests at 68OF have been performed to determine the corrosion and hydriding characteristics of Zircaloy-4 tubing, fabricated by cold reduction and finished in two metallurgical conditions: a stress-relief anneal (SRA) and a recrystallization anneal (RXA). These corrosion tests revealed differences in the post-transition corrosion product weight gains of the two materials. A computer corrosion model, designated CHORT, was developed from the test data and ascribes the observed difference in material weight gain to an assumed difference in the periodicity of a postulated cyclic buckling of the oxide.

ASBLT is a computer program consisting of DATATRAN MODULES which was used during the manufacturing phase of LWBR to collect and evaluate as-built data. The program was part of the LWBR fuel rod inspection process and produced sections of module assembly certification reports. ASBLT used fuel pellet, fuel rod and module assembly data to compute core inventories and to supply input to nuclear design programs for as-built core calculations.

Measurements of fission product iodine and cesium are reported for thoria and binary (ThO/sub 2/--UO/sub 2/) fuels with various irradiation histories. These volatile fission products were measured on the cladding surface or in the fuel by using specially developed radiochemical techniques. The radiochemical iodine measurements are found to be in general agreement with a theoretical iodine release model for irradiated fuel. Microprobe examinations of irradiated fuel rod cladding sections show fission product cesium to be located preferentially at the pellet to pellet interface region. Fission product iodine was detected in the interface region of one sample but generally remained below the microprobe limit of detection. 18 figures, 7 tables.

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.

The low-temperature (less than or equal to 550/sup 0/C), low-pressure (less than or equal to 36 torr) hydrogen absorption characteristics of specific types of Zircaloy-4 internal cladding surfaces (pickled, machined and welded) were investigated. The highest hydrogen contents were found at the machined and abraded surfaces. Although the pickled surface film on Zircaloy-4 retarded hydrogen pickup, especially at lower temperatures (less than or equal to 400/sup 0/C) and very low hydrogen pressures (less than or equal to 3.5 torr), some hydrogen was absorbed through the film even under these conditions. More hydrogen penetrated the pickled surfaces at higher temperatures and pressures. The pickled surfaces absorbed the hydrogen uniformly and without localization even with some film imperfections present. Little hydriding occurred when etched and welded Zircaloy-4 surfaces were exposed to water vapor at corrosion temperatures.