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.

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.

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 remained that could not be removed by adjustment of the spectrum mean energy alone.

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.

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*), and the ratio of U238 fission to U235 fission (delta/sup 28/). Full-core Monte Carlo calculations for two lattices showed good agreement with cell Monte Carlo-plus-multigroup P/sub l/ leakage corrections. Newly measured parameters for the low energy resonances of U238 significantly improved rho/sup 28/. In comparison with other CSEWG analyses, the strong correlation between K/sub eff/ and rho/sup 28/ suggests that U238 resonance capture is the major problem encountered in analyzing these lattices.

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.

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.

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.

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.

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.

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.

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.

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 interstitials because they have a larger capture radius and, since they contain the same number of vacancies as the smaller, higher temperature zones, they heal faster resulting in less total damage. The resulting theoretical predictions are compared with experimental data and found to be in good agreement.

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.

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.

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.

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.

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.

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 …