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Successive collision calculation of resonance absorption

Description: The successive collision method for calculating resonance absorption solves numerically the neutron slowing down problem in reactor lattices. A discrete energy mesh is used with cross sections taken from a Monte Carlo library. The major physical approximations used are isotropic scattering in both the laboratory and center-of-mass systems. This procedure is intended for day-to-day analysis calculations and has been incorporated into the current version of MUFT. The calculational model used for … more
Date: July 1, 1980
Creator: Schmidt, E. & Eisenhart, L.D.
Partner: UNT Libraries Government Documents Department
open access

Critical heat flux experiments in a circular tube with heavy water and light water.

Description: Experiments were performed to establish the critical heat flux (CHF) characteristics of heavy water and light water. Testing was performed with the up-flow of heavy and of light water within a 0.3744 inch inside diameter circular tube with 72.3 inches of heated length. Comparisons were made between heavy water and light water critical heat flux levels for the same local equilibrium quality at CHF, operating pressure, and nominal mass velocity. Results showed that heavy water CHF values were, on… more
Date: May 1980
Creator: Williams, C. L. & Beus, S. G.
Partner: UNT Libraries Government Documents Department
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Spatial distribution measurements of fission neutrons in water as an oxygen data test

Description: The spatial distribution of the total neutron density from a /sup 252/Cf source in pure water was measured to high statistical precision at distances from 11 to 80 cm from the source. Assuming the adequacy of the ENDF/B-IV hydrogen, and reasonable constraints on the fission spectrum mean energy, good agreement between experiment and a one-dimensional transport calculation was obtained for both ENDF/B-III and IV oxygen, with Version III slightly better. However, small residual differences remain… more
Date: February 1978
Creator: Green, L. & Ullo, J.J.
Partner: UNT Libraries Government Documents Department
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Post-irradiation recovery of growth in Zircaloy. [Fast neutron irradiation]

Description: The previously reported model for fast neutron induced growth in Zircaloy has been modified to predict post-irradiation recovery of the growth. The predicted results are compared to growth recovery data obtained by Adamson. Agreement is good if the binding energy between vacancies and depleted zones is reduced from the in-pile value of 1.1 eV to a post-irradiation value of 1.0 eV. Such a reduction is reasonable.
Date: December 1977
Creator: Dollins, C. C.
Partner: UNT Libraries Government Documents Department
open access

Monte Carlo analyses of TRX slightly enriched uranium-H/sub 2/O critical experiments with ENDF/B-IV and related data sets

Description: Four H/sub 2/O-moderated, slightly-enriched-uranium critical experiments were analyzed by Monte Carlo methods with ENDF/B-IV data. These were simple metal-rod lattices comprising Cross Section Evaluation Working Group thermal reactor benchmarks TRX-1 through TRX-4. Generally good agreement with experiment was obtained for calculated integral parameters: the epi-thermal/thermal ratio of U238 capture (rho/sup 28/) and of U235 fission (delta/sup 25/), the ratio of U238 capture to U235 fission (CR*… more
Date: December 1977
Creator: Hardy, J. Jr.
Partner: UNT Libraries Government Documents Department
open access

Review of thorium-U233 cycle thermal reactor benchmark studies

Description: A survey is made of existing integral experiments for U233 systems and thorium-uranium based fuel systems. The aim is to understand to what extent they give a consistent test of ENDF/B-IV nuclear data. A principal result is that ENDF/B-IV leads to an underprediction of neutron leakage. Results from testing alternate thorium data sets are presented. For one evaluation due to Leonard, the results depict a possible growing discrepancy between measured integral parameters such as rho/sup 02/ and I/… more
Date: March 1, 1980
Creator: Ullo, J.J.; Hardy, J. Jr. & Steen, N.M.
Partner: UNT Libraries Government Documents Department
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Integral testing of thorium and U233 data for thermal reactors

Description: A survey is made of integral experiments useful for testing thorium and /sup 233/U nuclear data in thermal reactor applications. Emphasis is on homogeneous /sup 233/U--H/sub 2/O criticals and simple, water-moderated /sup 233/U--thorium and /sup 235/U--thorium lattice experiments. Thorium--/sup 233/U-graphite experiments are also discussed briefly. Although the available experiments provide a fairly consistent test of important nuclear data, their accuracy and scope leave much to be desired. In … more
Date: June 1, 1979
Creator: Hardy, J., Jr.; Ullo, J.J. & Steen, N.M.
Partner: UNT Libraries Government Documents Department
open access

Three core concepts for producing uranium-233 in commercial pressurized light water reactors for possible use in water-cooled breeder reactors

Description: Selected prebreeder core concepts are described which could be backfit into a reference light water reactor similar to current commercial reactors, and produce uranium-233 for use in water-cooled breeder reactors. The prebreeder concepts were selected on the basis of minimizing fuel system development and reactor changes required to permit a backfit. The fuel assemblies for the prebreeder core concepts discussed would occupy the same space envelope as those in the reference core but contain a 1… more
Date: December 1, 1979
Creator: Conley, G. H.; Cowell, G. K.; Detrick, C. A.; Kusenko, J.; Johnson, E. G.; Dunyak, J. et al.
Partner: UNT Libraries Government Documents Department
open access

Critical heat flux tests with high pressure water in an internally heated annulus with alternating axial heat flux distribution

Description: Critical heat flux experiments were performed with an alternating heat flux profile in an internally heated annulus. The heated length was 84 inches with a square wave alternating heat flux profile over the last 12 inches having a maximum-to-average heat flux ratio of 1.76. Test data were obtained at pressures from 800 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 400 to 600/sup 0/F. Two different electrically heated t… more
Date: September 1979
Creator: Beus, S. G. & Humphreys, D. A.
Partner: UNT Libraries Government Documents Department
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Fission gas release from oxide fuels at high burnups

Description: The steady state gas release, swelling and densification model previously developed for oxide fuels has been modified to accommodate the slow transients in temperature, temperature gradient, fission rate and pressure that are encountered in normal reactor operation. The gas release predictions made by the model were then compared to gas release data on LMFBR-EBRII fuels obtained by Dutt and Baker and reported by Meyer, Beyer, and Voglewede. Good agreement between the model and the data was foun… more
Date: February 1, 1981
Creator: Dollins, C. C.
Partner: UNT Libraries Government Documents Department
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Swelling and gas release in oxide fuels during fast transients

Description: The previously reported swelling and gas release model for oxide fuels has been modified to predict fission gas bahavior during fast temperature transients. Under steady state or slowly varying conditions it has been assumed in the previous model that the pressure caused by the fission gas within the gas bubbles is in equilibrium with the surface tension of the bubbles. During a fast transient, however, net vacancy migration to the bubbles may be insufficient to maintain this equilibrium. In or… more
Date: February 1, 1981
Creator: Dollins, C. C.
Partner: UNT Libraries Government Documents Department
open access

Monte Carlo analysis of Pu-H/sub 2/O and UO/sub 2/-PuO/sub 2/-H/sub 2/O critical assemblies with ENDF/B-IV data

Description: A set of critical experiments, comprising thirteen homogeneous Pu-H/sub 2/O assemblies and twelve UO/sub 2/-PuO/sub 2/ lattices, was analyzed with ENDF/B-IV data and the RCPO1 Monte Carlo program, which modeled the experiments explicitly. Some major data sensitivities were also evaluated. For the Pu-H/sub 2/O assemblies, calculated K/sub eff/ averaged 1.011. The large (2.7%) scatter of K/sub eff/ values for these assemblies was attributed mostly to uncertainties in physical specifications since… more
Date: April 1, 1981
Creator: Hardy, J. Jr. & Ullo, J.J.
Partner: UNT Libraries Government Documents Department
open access

In-pile temperature dependence of the yield strength and growth of Zircaloy

Description: A previously reported growth model is modified and low temperature growth predictions in Zircaloy are made to support the proposal that the smaller irradiation damage observed at lower temperature results from the damage sites, or depleted zones, being larger at lower temperatures. This proposal holds that the vacancies making up the zones cannot migrate at the lower temperature and, therefore, cannot condense into small voids or dislocation loops. The larger zones serve as better sinks for int… more
Date: March 1978
Creator: Dollins, C. C.
Partner: UNT Libraries Government Documents Department
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Irradiation testing of internally pressurized and/or graphite coated Zircaloy-4 clad fuel rods in the NRX Reactor (AWBA Development Program). [LWBR]

Description: Irradiation tests on 0.612 inch O.D. by 117-inch long Zircaloy-4 clad fuel rods were performed to assess the effects on fuel rod performance of (1) internal helium pre-pressurization to 500 psi as fabricated, (2) the presence of a graphite barrier coating on the inside cladding surface, and (3) combined pre-pressurization and graphite coating. Periodic dimensional examinations were performed on the test rods, and the results were compared with data obtained from two previously irradiated test r… more
Date: November 1, 1978
Creator: Hoffman, R.C. & Sherman, J.
Partner: UNT Libraries Government Documents Department
open access

Monte Carlo analyses of simple U233 O/sub 2/-ThO/sub 2/ and U235 O/sub 2/-ThO/sub 2/ lattices with ENDF/B-IV data

Description: A number of water-moderated Th-U235 and Th-U233 lattice integral experiments were analyzed in a consistent manner, with ENDF/B-IV data and detailed Monte Carlo methods. These experiments provide a consistent test of the nuclear data. The ENDF/B-IV data are found to perform reasonably well. Adequate agreement is found with integral measurements of thorium capture. Calculated K/sub eff/ values show a generally coherent pattern which is consistent with K/sub eff/ results obtained for homogeneous a… more
Date: September 1, 1980
Creator: Hardy, J. Jr. & Ullo, J.J.
Partner: UNT Libraries Government Documents Department
open access

Model to predict swelling, gas release, and densification in oxide fuels

Description: A model was developed to predict in-pile fission gas swelling, gas release, and densification in oxide fuels. This model considers fission gas behavior at the grain interior, on the grain boundaries, and at grain boundary edges under conditions of total gas bubble destruction by fission fragments and partial gas bubble destruction. When gas bubble swelling on grain edges reaches 5 percent, it is assumed that gas tunnels form along the edges. Gas release takes place by migration of the gas in th… more
Date: June 1, 1978
Creator: Dollins, C. C.
Partner: UNT Libraries Government Documents Department
open access

Critical heat flux experiments in an internally heated annulus with a non-uniform, alternate high and low axial heat flux distribution

Description: Critical heat flux experiments were performed with an alternate high and low heat flux profile in an internally heated annulus. The heated length was 84 inches (213 cm) with a chopped wave heat flux profile over the last 24 inches (61 cm) having a maximum-to-average heat flux ratio of 1.26. Three test sections were employed: one with an axially uniform heat flux profile as a base case and two with 60 inch (152 cm) uniform and 24 inch (61 cm) alternating high and low heat flux sections. The thir… more
Date: February 1981
Creator: Beus, S. G. & Seebold, O. P.
Partner: UNT Libraries Government Documents Department
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Thermomechanical theory of materials undergoing large elastic and viscoplastic deformation

Description: A thermomechanical theory of large deformation elastic-inelastic material behavior is developed which is based on a multiplicative decomposition of the strain. Very general assumptions are made for the elastic and inelastic constitutive relations and effects such as thermally-activated creep, fast-neutron-flux-induced creep and growth, annealing, and strain recovery are compatible with the theory. Reduced forms of the constitutive equations are derived by use of the second law of thermodynamics… more
Date: November 1, 1980
Creator: Martin, S.E. & Newman, J.B.
Partner: UNT Libraries Government Documents Department
open access

Fuel utilization potential in light water reactors with once-through fuel irradiation

Description: Current commercial light water reactor cores operate without recylce of fuel, on a once-through fuel cycle. To help conserve the limited nuclear fuel resources, there is interest in increasing the energy yield and, hence, fuel utilization from once-through fuel irradiation. This report evaluates the potential increase in fuel utilization of light water reactor cores operating on a once-through cycle assuming 0.2% enrichment plant tails assay. This evaluation is based on a large number of survey… more
Date: July 1, 1979
Creator: Rampolla, D. S.; Conley, G. H.; Candelore, N. R.; Cowell, G. K.; Estes, G. P.; Flanery, B. K. et al.
Partner: UNT Libraries Government Documents Department
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