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open access

AIRWAY: a fortran computer program to estimate radiation dose commitments to man from the atmospheric release of radionuclides

Description: 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 fr… more
Date: June 1979
Creator: Rider, J. L.
open access

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

Description: 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.
open access

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

Description: 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 … more
Date: August 1978
Creator: Freeman, L. B.
open access

Cladding corrosion and hydriding in irradiated defected zircaloy fuel rods

Description: 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 oxi… more
Date: August 1985
Creator: Clayton, J. C.
open access

Corrosion of Zircaloy-4 tubing in 68OF water

Description: 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 mod… more
Date: December 1, 1978
Creator: Marino, G. P. & Fischer, R. L.
open access

Critical heat flux experiments with a local hot patch in an internally heated annulus

Description: 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) ax… more
Date: February 1, 1979
Creator: Mortimore, E.P. & Beus, S.G.
open access

Densification related pellet diameter shrinkage in low burnup thoria-base fuels

Description: 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… more
Date: September 1, 1978
Creator: Spahr, G. L.
open access

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

Description: 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 feature… more
Date: August 1981
Creator: Hecker, H. C. & Freeman, L .B.
open access

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

Description: 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.
open access

Effect of fuel chips on cladding stress in zircaloy clad oxide fuel rods

Description: 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 d… more
Date: November 1978
Creator: Yerman, J. F.
open access

Effect of simulated thermal shield motion on nuclear instrument response: measurements and calculations

Description: 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 woul… more
Date: August 1, 1979
Creator: Schick, W. C., Jr.; Emert, C. J.; Shure, K. & Natelson, M.
open access

End-of-life destructive examination of light water breeder reactor fuel rods

Description: Destructive examination of 12 representative Light Water Breeder Reactor fuel rods was performed following successful operation in the Shippingport Atomic Power Station for 29,047 effective full power hours, about five years. Light Water Breeder Reactor fuel rods were unique in that the thorium oxide and uranium-233 oxide fuel was contained within Zircaloy-4 cladding. Destructive examinations included analysis of released fission gas; chemical analysis of the fuel to determine depletion, iodine… more
Date: October 1, 1987
Creator: Richardson, K.D.
open access

Ex-reactor deformation of externally pressurized short lengths of fuel rod cladding.

Description: 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 tes… more
Date: May 1, 1979
Creator: Selsley, I. A.
open access

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

Description: 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… more
Date: August 1, 1978
Creator: Goldberg, I.; Spahr, G. L.; White, L. S.; Waldman, L. A.; Giovengo, J. F.; Pfennigwerth, P. L. et al.
open access

FLASH-6: simulation of top injection emergency core cooling heat transfer tests

Description: 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… more
Date: May 1, 1977
Creator: Lincoln, F. W.
open access

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

Description: 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 … more
Date: September 1, 1979
Creator: Harris, B.D.; Prelewicz, D.A. & Beus, S.G.
open access

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

Description: 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 il… more
Date: March 1, 1981
Creator: Crescimanno, P. J. & Keller, K. L.
open access

Fuel rod-grid interaction wear: in-reactor tests

Description: 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 … more
Date: November 1, 1979
Creator: Stackhouse, R. M.
open access

Installation of the Light-Water Breeder Reactor at the Shippingport Atomic Power Station

Description: This report summarizes the refueling operations performed to install a Light Water Breeder Reactor (LWBR) core into the existing pressurized water reactor vessel at the Shippingport Atomic Power Station. Detailed descriptions of the major installation operations (e.g., primary system preconditioning, fuel installation, pressure boundary seal welding) are included as appendices to this report; these operations are of technical interest to any reactor servicing operation, whether the reactor is a… more
Date: May 1, 1983
Creator: Massimino, R.J. & Williams, D.A.
open access

Internal hydriding in irradiated defected Zircaloy fuel rods: A review

Description: 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… more
Date: October 1, 1987
Creator: Clayton, J C
open access

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

Description: 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 fi… more
Date: March 1, 1979
Creator: Ivak, D. M. & Waldman, L. A.
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