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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 the average, 8 percent below the light water CHF values.
Date: May 1980
Creator: Williams, C. L. & Beus, S. G.

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 third test section had a 2.15 inch (5.46 cm) section with a peak-to-average heat flux ratio of 2.19 (hot patch) superimposed at the exit end of the alternating high and low heat flux profile.
Date: February 1981
Creator: Beus, S. G. & Seebold, O. P.

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 test sections were employed both with 72 inch uniform and 12 inch alternating heat flux sections. The second test section had a 0.44 inch hot patch with a peak-to-average heat flux ratio of 2.7 superimposed on the alternating flux profile at the exit end. Critical heat flux results with the alternating heat flux profile and with the superimposed hot patch were shown to be equivalent to those obtained in previous tests with a uniform heat flux profile except for several data points at low mass velocity and high enthalpy for which there is an apparent experimental bias in the uniform heat flux results.
Date: September 1979
Creator: Beus, S. G. & Humphreys, D. A.

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 found. A comparison between the model and three other sets of gas release data is also shown, again with good agreement.
Date: February 1, 1981
Creator: Dollins, C.C.

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 calculations using techniques which were verified by more detailed calculations of several core concepts. It is concluded that the maximum fuel utilization which could be achieved by practical once-through pressurized light water reactor cores with either uranium or thorium is about 17 MWYth/ST U/sub 3/O/sub 8/ (Megawatt Years Thermal per Short Ton of U/sub 3/O/sub 8/). This is about 50% higher than that of current commercial light water reactor cores. Achievement of this increased fuel utilization would require average fuel burnup beyond 50,000 MWD/MT and incorporation of the following design features to reduce parasitic losses of neutrons: reflector blankets to utilize neutrons that would otherwise leak out of the core; fuel management practices in which a smaller fraction of the core is replaced at each refueling; and neutron economic reactivity control, such as movable fuel control rather than soluble boron control. For a hypothetical situation in which all neutron leakage and parasitic losses are eliminated and fuel depletion is not limited by design considerations, a maximum fuel utilization of about 20 MWYth/ST U/sub 3/O/sub 8/ is calculated for either uranium or thorium. It is concluded that fuel utilization for comparable reactor designs is better with uranium fuel than with thorium fuel for average fuel depletions of 30,000 to 35,000 MWD/MT which are characteristic of present light water reactor cores.
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.

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 detailed Monte Carlo analyses, ENDF/B-IV data are found to perform reasonably well. Adequate (though partly fortuitous) agreement is found with integral measurements of thorium resonance capture in lattices. A new, harder fission spectrum for /sup 233/U can correct the principal discrepancy observed with ENDF/B-IV, a bias trend in K/sub eff/ attributed to an underprediction of leakage.
Date: June 1, 1979
Creator: Hardy, J., Jr.; Ullo, J.J. & Steen, N.M.

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 rods--both unpressurized and uncoated and one intentionally defected. These comparisons indicate that both pre-pressurization and graphite coating can substantially improve fuel element performance capability.
Date: November 1, 1978
Creator: Hoffman, R.C. & Sherman, J.

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 the grains and on the grain boundaries to the edge tunnels. Intergranular and intragranular densifications are considered. Densification takes place by vacancy boil-off due to thermal excitation and vacancy knockout by the passage of fission fragments through the pores. The migration rates of both vacancies and interstitials to pores are also calculated. Comparisons are made between the model and experimental data for swelling, gas release, and densification and found to be in reasonable agreement in most cases.
Date: June 1, 1978
Creator: Dollins, C.C.

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 aqueous critical assemblies. Harder prompt fission spectra for U233 and U235 can correct the principal discrepancy observed with ENDF/B-IV, a bias trend in K/sub eff/ attributed to an underprediction of leakage.
Date: September 1, 1980
Creator: Hardy, J. Jr. & Ullo, J.J.

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 no clear trends of K/sub eff/ were evident and data sensitivities were insignificant. The UO/sub 2/-PuO/sub 2/ lattices showed just one trend of K/sub eff/, which indicated an overprediction of U238 capture consistent with that observed for uranium-H/sub 2/O experiments. There was however a approx. 1% discrepancy in calculated K/sub eff/ between the two sets of UO/sub 2/-PuO/sub 2/ lattices studied.
Date: April 1, 1981
Creator: Hardy, J. Jr. & Ullo, J.J.

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/sup 232/ and the differential data, which underpredicts these parameters. Sensitivities to other nuclear data components, notably the fission neutron spectrum, were determined. A new harder U233 spectrum significantly reduces a bias trend in K/sub eff/ vs leakage.
Date: March 1, 1980
Creator: Ullo, J.J.; Hardy, J. Jr. & Steen, N.M.

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 the analysis of the nuclear performance of LWBR includes this resonance absorption procedure. Test comparisons of results with RCPO1 give very good agreement.
Date: July 1, 1980
Creator: Schmidt, E. & Eisenhart, L.D.

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 order to ascertain the net vacancy flow, it is necessary to model the point defect behaviour in the fuel. This model is reported. Knowing the net flow of vacancies to the bubble, the bubble size, the diffusivity can be determined and the long range migration of the gas out of the fuel can be calculated. The model has also been modified to allow release of all the gas on the grain boundaries during a fast temperature transient.
Date: February 1, 1981
Creator: Dollins, C. C.

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 in the form of the Clausius-Duhem inequality. Observer invariant equations are derived by use of an invariance principle which is a generalization of the principle of material frame indifference.
Date: November 1, 1980
Creator: Martin, S.E. & Newman, J.B.

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 19 by 19 array of fuel rods instead of the reference 17 by 17 array. An instrument well and 28 guide tubes for control rods have been allocated to each prebreeder fuel assembly in a pattern similar to that for the reference fuel assemblies. Backfit of these prebreeder concepts into the reference reactor would require changes only to the upper core support structure while providing flexibility for alternatives in the type of fuel used.
Date: December 1, 1979
Creator: Conley, G.H.; Cowell, G.K.; Detrick, C.A.; Kusenko, J.; Johnson, E.G.; Dunyak, J. et al.