Monte Carlo simulation using the meter system with application related to LWBR (open access)

Monte Carlo simulation using the meter system with application related to LWBR

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.
Date: February 1, 1977
Creator: Beaudoin, B. R.
System: The UNT Digital Library
ASBLT: a system of DATATRAN MODULES which process core fuel loading for use in as-built calculations (open access)

ASBLT: a system of DATATRAN MODULES which process core fuel loading for use in as-built calculations

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.
Date: February 1, 1979
Creator: Beaudoin, B.R.; Beggs, W.J.; Case, C.R. & Wilczynski, R.
System: The UNT Digital Library
Properties of thoria and thoria-urania: a review (open access)

Properties of thoria and thoria-urania: a review

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 Data were taken from a review of the literature and from both published and unpublished data derived from the Light Water Breeder Reactor (LWBR) Program at the Bettis Atomic Power Laboratory. 30 figs., 23 tables, 163 refs.
Date: June 1, 1978
Creator: Belle, J. & Berman, R. M.
System: The UNT Digital Library
Methods for assessing homogeneity in ThO/sub 2/--UO/sub 2/ fuels (open access)

Methods for assessing homogeneity in ThO/sub 2/--UO/sub 2/ fuels

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.
Date: June 1, 1978
Creator: Berman, R. M.
System: The UNT Digital Library
Shippingport operations with the Light Water Breeder Reactor core. (open access)

Shippingport operations with the Light Water Breeder Reactor core.

This report describes the operation of the Shippingport Atomic Power Station during the LWBR (Light Water Breeder Reactor) Core lifetime. It also summarizes the plant-oriented operations during the period preceding LWBR startup, which include the defueling of The Pressurized Water Reactor Core 2 (PWR-2) and the installation of the LWBR Core, and the operations associated with the defueling of LWBR. The intent of this report is to examine LWBR experience in retrospect and present pertinent and significant aspects of LWBR operations that relate primarily to the nuclear portion of the Station. The nonnuclear portion of the Station is discussed only as it relates to overall plant operation or to unusual problems which result from the use of conventional equipment in radioactive environments. 30 refs., 69 figs., 27 tabs.
Date: March 1, 1986
Creator: Budd, W. A.
System: The UNT Digital Library
Internal hydriding in irradiated defected Zircaloy fuel rods: A review (open access)

Internal hydriding in irradiated defected Zircaloy fuel rods: A review

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
System: The UNT Digital Library
Cladding corrosion and hydriding in irradiated defected zircaloy fuel rods (open access)

Cladding corrosion and hydriding in irradiated defected zircaloy fuel rods

Twenty-one LWBR irradiation test rods containing ThO/sub 2/-UO/sub 2/ fuel and Zircaloy cladding with holes or cracks operated successfully. Zircaloy cladding corrosion on the inside and outside diameter surfaces and hydrogen pickup in the cladding were measured. The observed outer surface Zircaloy cladding corrosion oxide thicknesses of the test rods were similar to thicknesses measured for nondefected irradiation test rods. An analysis model, which was developed to calculate outer surface oxide thickness of non-defected rods, gave results which were in reasonable agreement with the outer surface oxide thicknesses of defected rods. When the analysis procedure was modified to account for additional corrosion proportional to fission rate and to time, the calculated values agreed well with measured inner oxide corrosion film values. Hydrogen pickup in the defected rods was not directly proportional to local corrosion oxide weight gain as was the case for non-defected rods. 16 refs., 6 figs., 8 tabs.
Date: August 1985
Creator: Clayton, J. C.
System: The UNT Digital Library
Out-of-pile accelerated hydriding of Zircaloy fasteners (open access)

Out-of-pile accelerated hydriding of Zircaloy fasteners

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.
Date: October 1, 1979
Creator: Clayton, J. C.
System: The UNT Digital Library
Sources of internal hydriding in unirradiated thoria-fueled Zircaloy rods (open access)

Sources of internal hydriding in unirradiated thoria-fueled Zircaloy rods

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.
Date: February 1, 1979
Creator: Clayton, J. C.
System: The UNT Digital Library
Forces in bolted joints: analysis methods and test results utilized for nuclear core applications (open access)

Forces in bolted joints: analysis methods and test results utilized for nuclear core applications

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.
Date: March 1, 1981
Creator: Crescimanno, P. J. & Keller, K. L.
System: The UNT Digital Library
Out-of-pile hydriding and thermal relaxation of reactor fasteners using Zircaloy components: the REM-120 test. (open access)

Out-of-pile hydriding and thermal relaxation of reactor fasteners using Zircaloy components: the REM-120 test.

Six different fastened joints representing the various combinations of materials in contact with Zircaloy in the Light Water Breeder Reactor (LWBR) were tested in 520/sup 0/F water for 60 days to determine the out-of-pile hydriding tendencies of the Zircaloy components in the joints. No evidence of massive, accelerated hydriding was found in this test although other testing has shown that local hydriding can occur in one of the fasteners. The same six fastener designs were tested in 600/sup 0/F water for 60 days to assess their load retention capability (thermal relaxation). The measured relaxation of these fasteners confirmed the predicted values for the conditions of the test.
Date: January 1, 1980
Creator: Duenkel, D.A.
System: The UNT Digital Library
Nondestructive assay of UO/sub 2/--ThO/sub 2/ fuel pellets using the delayed neutron pellet assay gage (open access)

Nondestructive assay of UO/sub 2/--ThO/sub 2/ fuel pellets using the delayed neutron pellet assay gage

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.
Date: June 1, 1979
Creator: Emert, C.J.; Milani, S. & Beggs, W.J.
System: The UNT Digital Library
Development and control of the process for the manufacture of zircaloy-4 tubing for LWBR fuel rods (open access)

Development and control of the process for the manufacture of zircaloy-4 tubing for LWBR fuel rods

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.
Date: January 1, 1981
Creator: Eyler, J.H.
System: The UNT Digital Library
Calculational model used in the analysis of nuclear performance of the Light Water Breeder Reactor (LWBR) (open access)

Calculational model used in the analysis of nuclear performance of the Light Water Breeder Reactor (LWBR)

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.
Date: August 1978
Creator: Freeman, L. B.
System: The UNT Digital Library
Determination of statistically based design limits associated with engineering models. (open access)

Determination of statistically based design limits associated with engineering models.

This report provides a usable reference of methods and procedures for the construction of both one-sided and two-sided ..gamma../P statistical tolerance limits for design application to both linear and nonlinear models in any number of variables.
Date: February 1, 1980
Creator: Ginsburg, H.
System: The UNT Digital Library
A Procedure for Calculation of Steady-State Temperature in Zircaloy-Clad, Bulk-Oxide Fuel Elements Using the Figro Computer Program (open access)
Fission gas release from ThO/sub 2/ and ThO/sub 2/--UO/sub 2/ fuels (open access)

Fission gas release from ThO/sub 2/ and ThO/sub 2/--UO/sub 2/ fuels

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.
Date: August 1, 1978
Creator: Goldberg, I.; Spahr, G. L.; White, L. S.; Waldman, L. A.; Giovengo, J. F.; Pfennigwerth, P. L. et al.
System: The UNT Digital Library
Measurement of the thorium absorption cross section shape near thermal energy (open access)

Measurement of the thorium absorption cross section shape near thermal energy

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.
Date: November 1, 1976
Creator: Green, L.
System: The UNT Digital Library
FLASH6 simulation of semiscale blowdown data, NRC Standard Problems 2 and 3 (open access)

FLASH6 simulation of semiscale blowdown data, NRC Standard Problems 2 and 3

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.
Date: September 1, 1979
Creator: Harris, B.D.; Prelewicz, D.A. & Beus, S.G.
System: The UNT Digital Library
Nuclear analysis and performance of the Light Water Breeder Reactor (LWBR) core power operation at Shippingport (open access)

Nuclear analysis and performance of the Light Water Breeder Reactor (LWBR) core power operation at Shippingport

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.
Date: April 1, 1984
Creator: Hecker, H. C.
System: The UNT Digital Library
Summary of the nuclear design and performance of the Light Water Breeder Reactor (LWBR) (open access)

Summary of the nuclear design and performance of the Light Water Breeder Reactor (LWBR)

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.
Date: June 1, 1979
Creator: Hecker, H. C.
System: The UNT Digital Library
Design features of the Light Water Breeder Reactor (LWBR) which improve fuel utilization in light water reactors (open access)

Design features of the Light Water Breeder Reactor (LWBR) which improve fuel utilization in light water reactors

This report surveys reactor core design features of the Light Water Breeder Reactor which make possible improved fuel utilization in light water reactor systems and breeding with the uranium-thorium fuel cycle. The impact of developing the uranium-thorium fuel cycle on utilization of nuclear fuel resources is discussed. The specific core design features related to improved fuel utilization and breeding which have been implemented in the Shippingport LWBR core are presented. These design features include a seed-blanket module with movable fuel for reactivity control, radial and axial reflcetor regions, low hafnium Zircaloy for fuel element cladding and structurals, and a closely spaced fuel rod lattice. Also included is a discussion of several design modifications which could further improve fuel utilization in future light water reactor systems. These include further development of movable fuel control, use of Zircaloy fuel rod support grids, and fuel element design modifications.
Date: August 1981
Creator: Hecker, H. C. & Freeman, L .B.
System: The UNT Digital Library
Thoria powder process development (open access)

Thoria powder process development

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.
Date: October 1, 1979
Creator: Hutchison, C.R. & Lloyd, R.
System: The UNT Digital Library
Iodine and cesium in oxide fuel pellets and zircaloy-4 cladding of irradiated fuel rods (open access)

Iodine and cesium in oxide fuel pellets and zircaloy-4 cladding of irradiated fuel rods

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.
Date: March 1, 1979
Creator: Ivak, D. M. & Waldman, L. A.
System: The UNT Digital Library