Tests of closed-end, internally pressurized, Zircaloy-4 tubing specimens were utilized to develop low strain creep characteristics as a function of time at temperatures in the range of 1475/sup 0/F to 2000/sup 0/F (802/sup 0/C to 1093/sup 0/C) and hoop stresses in the range of 250 to 2500 psi for use in loss-of-coolant accident (LOCA) analyses. The strain rate above the start of the alpha to beta phase transformation region, approximately 1490/sup 0/F (810/sup 0/C), was found to be sensitive to the test procedure (stress-temperature history). This is believed to result from variations in the metallurgical structure. A prediction model is presented which provides a conservative upper bound to the low strain test data provided in this report and reported in the literature.

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 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.

Wear of the Zircaloy cladding of LWBR irradiation test fuel rods, resulting from relative motion between rod and rod support contacts, is reported. Measured wear depths were small, 0.0 to 2.7 mils, but are important in fuel element behavior assessment because of the local loss of cladding thickness, as well as the effect on grid spring forces that laterally restrain the rods. An empirical wear analysis model, based on out-of-pile tests, is presented. The model was used to calculate the wear on the irradiation test fuel rods attributed to a combination of up-and-down motions resulting from power and pressure/temperature cycling of the test reactor, flow-induced vibrations, and assembly handling scratches. The calculated depths are generally deeper than the measured depths.

Zircaloy clad oxide fuel rods are subjected to a variety of core power transients. One of these, an up-power transient, can place a severe burden on the fuel rod cladding that would potentially lead to rupture if not properly allowed for during the fuel rod design and plant operation. The cladding stress during such a transient can be increased by the presence of fuel chips between the oxide fuel pellet and the cladding. An analysis procedure based on mechanical tests of fuel and cladding was developed that permits calculation of the stress increase due to chips, so that the stress penalty can be accommodated without unnecessary penalties to fuel rod performance. The method of evaluating the maximum cladding bending tensile stress near the chip is described and test data are presented to support the analysis method.

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.

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.

This report presents and discusses the results of physics tests performed at beginning of life on the Light Water Breeder Reactor (LWBR). These tests have confirmed that movable seed assembly critical positions and reactivity worths, temperature coefficients, xenon transient characteristics, core symmetry, and core shutdown are within the range of values used in the design of the LWBR and its reactor protection analysis. Measured core physics parameters were found to be in good agreement with the calculated values.

An experiment has been performed to determine the effect of motion of a thermal shield on the neutron signal expected from ex-core detectors. Using a mockup of the LWBR reactor vessel, thermal shield, and core barrel in conjunction with a /sup 252/Cf neutron source, the change in detector signal with displacement of the various components was investigated. It was found that moving the thermal shield would produce a significant change in detector signal, although the effect was smaller than would be produced by moving the source and core barrel together. The results were substantiated by two-dimensional discrete-ordinate calculations.

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.

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.

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 …

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.

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.

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.

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 …

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.

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 …

The DECAG (Deformation of Cladding into Axial Gaps) ex-reactor test program evaluated deformation of Zircaloy-4 cladding into axial gaps in tubular fuel elements. These axial gaps are the result of cladding elongation and fuel stack shrinkage. The test program consisted of twelve series and subseries of both fully recrystallized and stress-relieved highly cold worked tubing tested under pressure-temperature combinations in autoclaves. The test program also verified the validity of achieving test acceleration through the use of elevated temperatures by correlating both ovality and diameter change at lower temperatures with the Larson--Miller Parameter.