Accident analysis computer code for nuclear reactors

Originally presented as the first author's thesis, (Nucl. E.)--in the M.I.T. Dept. of Nuclear Engineering, 1977Includes bibliographical references (pages 122-123

Design of central irradiation facilities for the MITR-II research reactor

"September 1976."Also issued as a Ph. D. thesis by the first author and supervised by the second author, MIT Dept. of Nuclear Engineering, 1977Includes bibliographical references (pages 94-95)Design analysis studies have been made for various in-core irradiation facility designs which are presently used, or proposed for future use in the MITR-II. The information obtained includes reactivity effects, core flux and power distributions, and estimates of the safety limits and limiting conditions for operation. A finite-difference, diffusion theory computer code was employed in two and three dimensions, and with three and fifteen group energy schemes. The facilities investigated include the single-element molybdenum sample holder, a proposed double-element irradiation facility and a proposed central irradiation facility design encompassing most of the area of the three central core positions. In addition, a comparison of the effects of various absorber materials has been made for a core configuration which includes three solid dummies. Flux levels in the molybdenum holder facility and in the beam ports were calculated for both three and five dummy cores. Flooding the sample tube in these cases was found to increase the safety and operating limits, but not to unacceptable levels for an 8 inch blade height. For the five dummy case, the operating limit in the C-ring was predicted to reach its maximum allowed value at a blade bank height of 13.6 inches. The reactivity effect of flooding was calculated to be 0.19%AK for the five dummy case, in direct agreement with the measured value. Flooging the large sample channel in the double element facility was found to increase the reactivity by 1.5 6%AK ff and also to cause an unacceptable power-peaking. The proposed central irradiation facility is a thermal flux-trap which could produce thermal flux values of up to 2.0 x 1014 n/cm 2 sec. Computer estimates show that flooding this facility's central sample tube would increase this value to 2.5 x 1014 n/cm2 sec, without resulting in an unacceptable power peak

Beginning-of-Life Data Report for the Instrumented Fuel Assembly (IFA)-527

This report presents beginning-of-life (BOL) data from the first four months of operation of the six-rod instrumented fuel assembly (IFA)-527 in the Halden Boiling Water Reactor (HBWR), Halden, Norway. This assembly is the last in a series of U.S. Nuclear Regulatory Commission (NRC)-sponsored tests to verify steady-state fuel performance computer codes. IFA-527 contains five identical rods with high-density stable fuel pellets and 0.23-mm diametral gaps and one rod with similar fuel pellets but with a 0.06-mm diametral gap. All six rods were xenon-filled to provide simulation of the effects of fission gas and to enhance the observable effects of fuel cracking and relocation on fuel temperatures. The assembly operated successfully from July 1, 1980, to August 15, 1980; and then the reactor was shut down until September 10, 1980. Sometime during the shutdown, four of the six rods suffered pressure boundary failure. The decision was made to restart the reactor to collect operating data with failed rods. This report presents both pre- and postfailure data for IFA-527

Observation of Porosity Reduction in a Densification-Prone Test Fuel Rod: Data and Analysis

Instrumented fuel assembly (IFA)-431 was irradiated in the Halden Boiling Water Reactor (HBWR) for the purpose of extending the steady-state data base. Rod 6 of this assembly began irradiation with UO{sub 2} fuel of 92% theoretical density (TD) that was unstable with respect to in-reactor densification. Thermal resintering tests resulted in a final density of 95.3% TD while post-irradiation examination (PIE) indicated a final density of 96.5% TD. Observed microstructural changes were consistent with published densification studies; there was a marked depletion of submicrometer diameter pores and total pore volume. However, grain size increased only slightly, indicating that internal pellet temperatures did not reach the 1875K applied in resintering tests. Oensification was observed to increase the temperatures in rod 6, but temperatures did not become as high as for a sibling rod that simulated instantaneous densification. Temperatures calculated with U.S. Nuclear Regulatory Commission (NRC) fuel performance computer codes were generally higher than observed temperatures

FRAPCON-2 Developmental Assessment

FRAPCON-2 calculations using all mechanical and gas release options are compared with well-characterized experimental data and with calculations of generic fuel rod response by FRAPCON-1. These comparisons indicate that FRAPCON-2 is capable of analyzing the fuel rod response for the wide range of cases for which the code was designed and compares well with experimental data

Shuttle ascent and shock impingement aerodynamic heating studies

The collection and analysis of aerodynamic heating data obtained from shock impingement experimental investigation were completed. The data were categorized into four interference areas; fin leading edge, wing/fuselage fin/plate corners, and space shuttle configurations. The effects of shock impingement were found to increase the heating rates 10 to 40 times the undisturbed values. A test program was completed at NASA/Langley Research Center to investigate the magnitudes and surface patterns of the mated shock interference flowfield. A 0.0065 scale thin-skin model of the MDAC 256-20 space shuttle booster mated with a Stycast model of the MDAC Internal tank orbiter was tested in the 20-inch M=6 tunnel, the 31-inch M=10 tunnel, and the 48-inch Unitary Plan Tunnel. The gap region of the ascent configuration was the principal area of interest where both thermocouple and phase-change paint data were obtained. Pressure and heat transfer distributions data on the leeward surface of a 75-degree sweep slab delta wing are presented. The effects of surface roughness on boundary layer transition and aerodynamic heating were investigated

An evaluation of tight-pitch PWR cores

Originally presented as the author's thesis, Ph.D. in the M.I.T. Dept. of Nuclear Engineering, 1979.The impact of tight pitch cores on the consumption of natural uranium ore has been evaluated for two systems of coupled PWR's namely one particular type of thorium system-U-235/U02: Pu/Th02: U-233/ThO2--and the conventional recycle-mode uranium system- U-235/U02: Pu/UO . The basic parameter varied was the fuel-to-moderator volume ratio (F/M) o the (uniform) lattice for the last core in each sequence. Although methods and data verification in the range of present interest, 0.5 (current lattices)< F/M < 4.0 are limited by the scarcity of experiments with F/M > l.0,the EPRI-LEOPARD and LASER programs used for the thorium and uranium calculations, respectively, were successfully benchmarked against several of the more pertinent experiments. It was found that by increasing F/M to "3 the uranium ore usage for the uranium system can be decreased by as much as 60% compared to the same system with conventional recycle (at F/M 0.5). Equivalent savings for the thorium system of the type examined here are much smaller (10%) because of the poor performance of the intermediate Pu/ThO2 core--which is not substantially improved by increasing F/M. Although fuel cycle costs (calculated at the indifference value of bred fissile species) are rather insensitive to the characteristics of the tight pitch cores, system energy production costs do not favor the low discharge burnups which might other- wise allow even greater ore savings (80%). Temperature and void coefficients of reactivity for the tight pitch cores were calculated to be negative. Means for implementing tight lattice use were investigated, such as the use of stainless steel clad in place of zircaloy; and alternatives achieving the same objective were briefly examined, such as the use of D20/H20 mixtures as coolant. Major items identified requiring further work are system redesign to accommodate higher core pressure drop, and transient and accident thermal-hydraulics.DOE Contract no. EN-77-S O2-4570

Design and fuel management of PWR cores to optimize the once-through fuel cycle

Originally presented as the first author's thesis, (Sc.D.) in the M.I.T. Dept. of Nuclear Engineering, 1978.The once-through fuel cycle has been analyzed to see if there are substantial prospects for improved uranium ore utilization in current light water reactors, with a specific focus on pressurized water reactors. The types of changes which have been examined are: (1) re-optimization of fuel pin diameter and lattice pitch, (2) Axial power shaping by enrichment gradation in fresh fuel, (3) Use of 6-batch cores with semi-annual refueling, (4) Use of 6-batch cores with annual refueling, hence greater extended (.doubled) burnup, (5) Use of radial reflector assemblies, (6) Use of internally heterogeneous cores (simple seed/blanket configurations), (7) Use of power/temperature coastdown at the end of life to extend burnup, (8) Use of metal or diluted oxide fuel, (9) Use of thorium, and (10) Use of isotopically separated low a cladding material. a State-of-the-art LWR computational methods, LEOPARD/PDQ-7/FLARE-G, were used to investigate these modifications. The most effective way found to improve uranium ore utilization is to increase the discharge burnup. Ore savings on the order of 20% can be realized if greatly extended burnup (- double that of current practice) is combined with an increase in the number of batches in the core from 3 to 6. The major conclusion of this study is that cumulative reductions in ore usage of on the order of 30% are fore- seeable relative to a current PWR operating on the once-through fuel cycle, which is comparable to that expected for the same cores operated in the recycle mode.DOE Contract no. EN-77-S-02-4570

Analysis of strategies for improving uranium utilization in pressurized water reactors

Systematic procedures have been devised and applied to evaluate core design and fuel management strategies for improving uranium utilization in Pressurized Water Reactors operated on a once-through fuel cycle. A principal objective has been the evaluation of suggested improvements on a self-consistent basis, allowing for concurrent changes in dependent variables such as core leakage and batch power histories, which might otherwise obscure the sometimes subtle effects of interest. Two levels of evaluation have been devised: a simple but accurate analytic model based on the observed linear variations in assembly reactivity as a function of burnup; and a numerical approach, embodied in a computer program, which relaxes this assumption and combines it with empirical prescriptions for assembly (or batch) power as a function of reactivity, and core leakage as a function of peripheral assembly power. State-of-the-art physics methods, such as PDQ-7, were used to verify and supplement these techniques.These methods have been applied to evaluate several suggested improvements: (1) axial blankets of low-enriched or depleted uranium, and of beryllium metal, (2) radial natural uranium blankets, (3) lowleakage radial fuel management, (4) high burnup fuels, (5) optimized H/U atom ratio, (6) annular fuel, and (7) mechanical spectral shift (i.e. variable fuel-to-moderator ratio) concepts such as those involving pin pulling and bundle reconstitution.The potential savings in uranium requirements compared to current practice were found to be as follows: (1) O0-3%, (2) negative, (3) 2-3%; possibly 5%, (4) "15%, (5) 0-2.5%, (6) no inherent advantage, (7) 10%. Total savings should not be assumed to be additive; and thermal/hydraulic or mechanical design restrictions may preclude full realization of some of the potential improvements

An evaluation of the fast-mixed spectrum reactor

"February 1980."Also issued as an M.S. thesis written by the first author and supervised by the second and third authors, MIT Dept. of Nuclear Engineering, 1980Includes bibliographical references (pages 145-147)An independent evaluation of the neutronic characteristics of a gas-cooled fast-mixed spectrum reactor (FMSR) core design has been performed. A benchmark core configuration for an early FMSR design was provided by Brookhaven National Laboratory, the originators of the concept. The results of the evaluation were compared with those of BNL. Points of comparison included system reactivity and breeding ratio, and region-wise power densities and isotopic compositions as a function of burnup. The results are in sufficiently good agreement to conclude that the neutronic feasibility of the FMSR concept has been independently validated. Significant differences, primarily in higher plutonium isotope concentrations, occur only in regions of low neutronic importance, and plausible reasons for the differences are advanced based on sensitivity studies and comparison of spectral indices. While both M.I.T. and BNL calculations tend to predict that the benchmark design is slightly subcritical, at the beginning of equilibrium cycle, the margin to k = 1.0 is close enough (Ak < 0.03) that the situation can be remedied. Establishment of a consensus fission product cross section set was identified as an objective of merit, since non-negligible differences were found in results computed using various extant sets (BNL, LIB-IV, Japanese). Non-fission heating by gamma and neutron interactions was evaluated for the reference core design using a coupled neutron/gamma cross section set and SN calculations. In the unfueled regions of the core, moderator elements in particular, the non-fission heating rate was found to be significant (averaging about 6 kw/liter), but posed no obvious problems. In fueled regions the common assumption of local deposition of all energy at the point of fission was verified to be a good approximation for most engineering purposes.Engineering and Advanced Reactor Safety Division of the U.S. Department of Energy at Brookhaven National Laboratory contract 472241-