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.

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.

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 Data were taken from a review of the literature and from both published and unpublished data derived from the Light Water Breeder Reactor (LWBR) Program at the Bettis Atomic Power Laboratory. 30 figs., 23 tables, 163 refs.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.

Mechanical joints between Zircaloy and nickel-bearing alloys, mainly the Zircaloy-4/Inconel-600 combination, were exposed to water at 450/sup 0/F and 520/sup 0/F to study hydriding of Zircaloy in contact with a dissimilar metal. Accelerated hydriding of the Zircaloy occurred at both temperatures. At 450/sup 0/F the dissolved hydrogen level of the water was over ten times that at 520/sup 0/F. At 520/sup 0/F the initially high hydrogen ingress rate decreased rapidly as exposure time increased and was effectively shut off in about 25 days. Severely hydrided Zircaloy components successfully withstood thermal cycling and mechanical testing. Chromium plating of the nickel-bearing parts was found to be an effective and practical barrier in preventing nickel-alloy smearing and accelerated hydriding of Zircaloy.

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.

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.

The calculational model used in the analysis of LWBR nuclear performance is described. The model was used to analyze the as-built core and predict core nuclear performance prior to core operation. The qualification of the nuclear model using experiments and calculational standards is described. Features of the model include: an automated system of processing manufacturing data; an extensively analyzed nuclear data library; an accurate resonance integral calculation; space-energy corrections to infinite medium cross sections; an explicit three-dimensional diffusion-depletion calculation; a transport calculation for high energy neutrons; explicit accounting for fuel and moderator temperature feedback, clad diameter shrinkage, and fuel pellet growth; and an extensive testing program against experiments and a highly developed analytical standard.

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.

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.

FLASH6 computer program calculations are compared with experimental data from two simulated loss-of-coolant accident blowdown tests which are designated as numbers 2 and 3 in the Standard problem Series sponsored by the Nuclear Regulatory Commission for reactor safety assessment. Both tests are isothermal blowdowns smulating a double-ended, cold-leg break and were conducted in the electrically-heated, 1-1/2 Loop Semiscale System at Idaho National Engineering Laboratory. The blowdown tests were initiated at nominal conditions of 575/sup 0/F, 2250 psia and 17.3 lbm/sec loop flow rate.

This report presents a summary of the nuclear design and expected nuclear performance of the Light Water Breeder Reactor during operation at the Shippingport Atomic Power Station. Performance predictions are presented for core lifetime, breeding margin, power distributions and performance, kinetic and stability parameters, and for core shutdown and reactivity control capability. Also included is a summary of as-built dimensions of core components and of development of breeding parameter equations and sensitivities.

The development program to identify the critical parameters for the process of converting thorium nitrate solution into thoria powder is described. Thorium oxalate hexahydrate is precipitated from the reaction of thorium nitrate solution with oxalic acid. The resulting thorium oxalate hexahydrate slurry is filter pressed into a cake which is air calcined to form thoria powder. Changes in the critical processing parameters such as free nitric acid content of the thorium nitrate solution, precipitation temperature, and calcining temperature altered the thoria powder characteristics, and thus its capability for being fabricated into fuel pellets. The objective of the powder preparation effort was to obtain thoria powders which could be formed by conventional ceramic fabrication techniques into thoria and thoria-urania pellets of high density and high integrity having a nearly uniform large grain structure.

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.

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.

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.

As part of the Light Water Breeder Reactor development program, hydraulic tests of reactor components were performed. This report presents the results of several of those tests performed for components which are somewhat unique in their application to a pressurized water reactor design. The components tested include: triplate orifices used for flow distribution purposes, multiventuri type flowmeters, tight lattice triangular pitch rod support grids, fuel rod end support plates, and the balance piston which is a major component of the movable fuel balancing system. Test results include component pressure loss coefficients, flowmeter coefficients and fuel rod region pressure drop characteristics.

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.

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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 …

Stress corrosion cracking (SCC) in unirradiated recrystallized Zircaloy-4 internally pressurized tubing specimens in atmospheres containing iodine vapor, cesium, or combinations of iodine and cesium is evaluated experimentally in terms of the effects of internal surface flaw morphology, iodine and cesium concentrations, tubing hydrogen content, test temperature, and test atmosphere water vapor content on the time to failure. The iodine vapor SCC data are analyzed in the framework of a fracture mechanics model. Expressions are developed which relate the iodine SCC threshold stress and lifetime for stresses above threshold to temperature, iodine concentration, and surface flaw geometry.

The AIRWAY computer program was developed to estimate the radiation dose commitments accured by all the people affected by the atmospheric release of radionuclides from a nuclear facility. This computer program provides dose commitment estimates for people on the boundary of the facility, in the immediate vicinity (i.e., within 80 to 100 km) and in the portios of the world beyond the immediate vicinity which are affected by the release. The AIRWAY program considers dose commitments resulting from immersion in the atmosphere containing the released radionuclides, ingestion of contaminated food, inhalation of gaseous and suspended radioactivity, and exposure to ground deposits. The dose commitments for each of these pathways is explicitly calculated for each radionuclide released into the atmosphere and for each daughter of each released radionuclide. This is accomplished by calculating the air and ground concentrations of each daughter in each of the regions of interest using the release rate of the parent radionuclide. The AIRWAY computer program is considered to be a significant improvement over other programs in which the effect of daughter radionuclides must be approximated by separate releases.

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 equations developed for use in the LWBR environmental statement to estimate the dose commitment over a given time interval to a given organ of the population in the entire region affected by the atmospheric releases of a radionuclide are presented and may be used for any assessment of dose commitments in these regions. These equations define the dose commitments to the world resulting from a released radionuclide and each of its daughters and the sum of these dose commitments provides the total dose commitment accrued from the release of a given radionuclide. If more than one radionuclide is released from a facility, then the sum of the dose commitments from each released nuclide and from each daughter of each released nuclide is the total dose commitment to the world from that facility. Furthermore, if more than one facility is considered as part of an industry, then the sum of the dose commitments from the individual facilities represents the total world dose commitment associated with that industry. The actual solutions to these equations are carried out by the AIRWAY computer program. The writing of this computer program entailed defining in detail the specific representations of the various parameters such as …