111 Matching Results

Search Results

Advanced search parameters have been applied.

Implementation of nodal equivalence parameters in DIF3D-VARIANT for core analysis of prismatic Very High Temperature Reactor (VHTR).

Description: The VARIANT module of the DIF3D code has been upgraded to utilize surface-dependent discontinuity factors. The performance of the new capability is verified using two-dimensional core cases with control rods in reflector and fuel blocks. Cross sections for VHTR components were generated using the DRAGON and HELIOS codes. For rodded block cross sections, the DRAGON calculations used a single-block model or the multi-block models combined with MCNP4C flux solutions, whereas the HELIOS calculations utilized multi-block models. Results from core calculations indicate that multiplication factor, block power, and control rod worth are significantly improved by using surface-dependent discontinuity factors.
Date: March 15, 2007
Creator: Lee, C. H.; Joo, H. K.; Yang, W. S. & Taiwo, T. A.
Partner: UNT Libraries Government Documents Department

Application of ENDF data to the AVR reactor with highly enriched uranium fuel and thorium feed

Description: Calculations were done applying ENDF data to the German AVR pebble bed reactor at KFA. Several core models were used, and the results obtained with ORNL methods for the multiplication, reaction rates, temperature coefficient of reactivity, and fuel temperature distributions are reported and compared. Only a small difference in multiplication is found for this core in going from ENDF/B-IV to ENDF/B-V cross-section data. The temperature coefficients calculated with the ENDF/B-V are somewhat smaller in magnitude. The worth of control rods was obtained and only a small difference was found with the data, but the calculated results are high compared with experiment. Neutron reaction rates with the key actinides are reported for three-dimensional core calculations.
Date: September 1, 1986
Creator: Vondy, D.R.
Partner: UNT Libraries Government Documents Department

Simplified modeling of the EBR-II control rods

Description: Simplified models of EBR-II control and safety rods have been developed for core modeling under various operational and shutdown conditions. A parametric study was performed on normal worth, high worth, and safety rod type control rods. A summary of worth changes due to individual modeling approximations is tabulated. Worth effects due to structural modeling simplification are negligible. Fuel region homogenization and burnup compression contributes more than any other factor. Reference case C/E values (ratio of calculated worth from detailed model to measured worth) of 1.072 and 1.142 for safety and normal worth rods indicate acceptable errors when the approximations are used. Fuel burnup effect illustrates rod worth sensitivity to the modeling approximation. Aggregate effects are calculated under a reduced mesh.
Date: June 25, 1995
Creator: Angelo, P.L.
Partner: UNT Libraries Government Documents Department

Enhancement of REBUS-3/DIF3D for whole-core neutronic analysis of prismatic very high temperature reactor (VHTR).

Description: Enhancements have been made to the REBUS-3/DIF3D code suite to facilitate its use for the design and analysis of prismatic Very High Temperature Reactors (VHTRs). A new cross section structure, using table-lookup, has been incorporated to account for cross section changes with burnup and fuel and moderator temperatures. For representing these cross section dependencies, three new modules have been developed using FORTRAN 90/95 object-oriented data structures and implemented within the REBUS-3 code system. These modules provide a cross section storage procedure, construct microscopic cross section data for all isotopes, and contain a single block of banded scattering data for efficient data management. Fission products other than I, Xe, Pm, and Sm, can be merged into a single lumped fission product to save storage space, memory, and computing time without sacrificing the REBUS-3 solution accuracy. A simple thermal-hydraulic (thermal-fluid) feedback model has been developed for prismatic VHTR cores and implemented in REBUS-3 for temperature feedback calculations. Axial conduction was neglected in the formulation because of its small magnitude compared to radial (planar) conduction. With the simple model, the average fuel and graphite temperatures are accurately estimated compared to reference STAR-CD results. The feedback module is currently operational for the non-equilibrium fuel cycle analysis option of REBUS-3. Future work should include the extension of this capability to the equilibrium cycle option of the code and additional verification of the feedback module. For the simulation of control rods in VHTR cores, macroscopic cross section deviations (deltas) have been defined to account for the effect of control rod insertion. The REBUS-3 code has been modified to use the appropriately revised cross sections when control rods are inserted in a calculation node. In order to represent asymmetric core blocks (e.g., fuel blocks or reflector blocks containing asymmetric absorber rods), surface-dependent discontinuity factors based on nodal ...
Date: October 13, 2006
Creator: Lee, C. H.; Zhong, Z.; Taiwo, T.A.; Yang, W.S.; Khalil, H.S.; Smith, M.A. et al.
Partner: UNT Libraries Government Documents Department

Public Service Company of Colorado 330-MW(e) high-temperature gas-cooled reactor research and development program. Quarterly progress report for the period ending December 31, 1973

Description: The research and development program for the Public Service Company of Colorado 330-MW(e) High-temperature Gas-cooled Reactor is reviewed. Studies of nuclear analysis of the fuel cycle, fuel zoning, and power shaping are included. The fuel development and fuel element proof testing programs include demonstration testing of the performance of test fuel elements by means of irradiation at expected operating temperatures and capsule irradiation tests of production fuel. (auth)
Date: January 31, 1974
Partner: UNT Libraries Government Documents Department

Tory II-C reactor cold critical assembly program

Description: Declassified 27 Nov 1973. The cold critical experiments with the Tory II-C reactor were performed in the critical facility of the Lawrence Radiation Laboratory at Livermore. The reactor is described briefly; the experimental equipment and the arrangement for the experiments is described in detail. The core poison was removed without difficulty, and criticality was achieved at 1624 hours July 14, 1965. With the design core configuration the shutdown margin was less than had been intended; criticality was attained at a shim rod bank position of 38.5 inches insertion rather than the expected 35 inches All Hastelloy tie tubes were replaced by Rene'-41. A heated core experiment with DELTA T = 129 plus or minus 7 deg F above room temperature provided a measure of the temperature coefficient which was found to be --3.9 x 10/sup -3/ plus or minus 7% inch of shim rod bank travel/ deg F. Other experiments included flux mapping, using fission foils, fission counters and copper flux wires, transients to determine control element worths, studies of flux spectrum by means of threshold detectors, and measurements of multiplication for various shutdown configurations for purposes of operational safety. Transient measurements, critical rod configurations, and flux distributions were also measured with the core poisoned in various stages to simulate elevated temperatures. 12 references. (auth)
Date: January 17, 1964
Creator: Morton, J.R. III; Varljen, T.C.; Petruzzi, J. Jr.; Piowaty, J.M. & Gardner, L.L.
Partner: UNT Libraries Government Documents Department

Neutronic safety parameters and transient analyses for potential LEU conversion of the Budapest Research Reactor.

Description: An initial safety study for potential LEU conversion of the Budapest Research Reactor was completed. The study compares safety parameters and example transients for reactor cores with HEU and LEU fuels. Reactivity coefficients, kinetic parameters and control rod worths were calculated for cores with HEU(36%) UAl alloy fuel and UO{sub 2}-Al dispersion fuel, and with LEU (19.75%)UO{sub 2}-Al dispersion fuel that has a uranium density of about 2.5 g/cm{sup 3}. A preliminary fuel conversion plan was developed for transition cores that would convert the BRR from HEU to LEU fuel after the process is begun.
Date: September 27, 1999
Creator: Pond, R. B.; Hanan, N. A.; Matos, J. E. & Maraczy, C.
Partner: UNT Libraries Government Documents Department

Reactor control system upgrade for the McClellan Nuclear Radiation Center Sacramento, CA.

Description: Argonne National Laboratory is currently developing a new reactor control system for the McClellan Nuclear Radiation Facility. This new control system not only provides the same functionality as the existing control system in terms of graphic displays of reactor process variables, data archival capability, and manual, automatic, pulse and square-wave modes of operation, but adds to the functionality of the previous control system by incorporating signal processing algorithms for the validation of sensors and automatic calibration and verification of control rod worth curves. With the inclusion of these automated features, the intent of this control system is not to replace the operator but to make the process of controlling the reactor easier and safer for the operator. For instance, an automatic control rod calibration method reduces the amount of time to calibrate control rods from days to minutes, increasing overall reactor utilization. The control rod calibration curve, determined using the automatic calibration system, can be validated anytime after the calibration, as long as the reactor power is between 50W and 500W. This is done by banking all of the rods simultaneously and comparing the tabulated rod worth curves with a reactivity computer estimate. As long as the deviation between the tabulated values and the reactivity estimate is within a prescribed error band, then the system is in calibration. In order to minimize the amount of information displayed, only the essential flux-related data are displayed in graphical format on the control screen. Information from the sensor validation methods is communicated to the operators via messages, which appear in a message window. The messages inform the operators that the actual process variables do not correlate within the allowed uncertainty in the reactor system. These warnings, however, cannot cause the reactor to shutdown automatically. The reactor operator has the ultimate responsibility of using ...
Date: March 10, 1999
Creator: Power, M. A.
Partner: UNT Libraries Government Documents Department

Blackness coefficients, effective diffusion parameters, and control rod worths for thermal reactors - methods

Description: Simple diffusion theory cannot be used to evaluate control rod worths in thermal neutron reactors because of the strongly absorbing character of the control material. However, reliable control rod worths can be obtained within the framework of diffusion theory if the control material is characterized by a set of mesh-dependent effective diffusion parameters. For thin slab absorbers the effective diffusion parameters can be expressed as functions of a suitably-defined pair of blackness coefficients. Methods for calculating these blackness coefficients in the P/sub 1/, P/sub 3/, and P/sub 5/ approximations, with and without scattering, are presented. For control elements whose geometry does not permit a thin slab treatment, other methods are needed for determining the effective diffusion parameters. One such method, based on reaction rate ratios, is discussed.
Date: January 1, 1984
Creator: Bretscher, M.M.
Partner: UNT Libraries Government Documents Department

Critical experiments in support of the CNPS (Compact Nuclear Power Source) program

Description: Zero-power static and kinetic measurements have been made on a mock-up of the Compact Nuclear Power Source (CNPS), a graphite moderated, graphite reflected, U(19.9% /sup 235/U) fueled reactor design. Critical configurations were tracked from a first clean configuration (184 most central fuel channels filled and all control rod and heat pipe channels empty) to a fully loaded configuration (all 492 fuel channels filled, core-length stainless steel pipe in the twelve heat-pipe channels, and approximately half-core-length boron carbide in the outer 4 control rod channels. Reactor physics data such as material worths and neutron lifetime are presented only for the clean and fully loaded configurations.
Date: January 1, 1988
Creator: Hansen, G.E.; Audas, J.H.; Martin, E.R.; Pederson, R.A.; Spriggs, G.D. & White, R.H.
Partner: UNT Libraries Government Documents Department

An overview of reactor physics standards: Past, present and future

Description: This report discusses for determining key static reactor physics parameters which have been developed by groups of experts (working groups) under the aegis of ANS-19, the ANS Reactor Physics Standards Committee. Following a series of sequential reviews, augmented by feedback from potential users, a proposed standard is brought into final form by the working group before it is adopted as a formal standard by the American National Standards Institute (ANSI); Reactor Physics standards are intended to provide guidance in the performance and qualification of complex sequences of reactor calculations and/or measurements and are regularly reviewed for possible updates and/or revisions. The reactor physics standards developed to date are listed and standards now being developed by the respective working groups are also provided.
Date: July 1, 1992
Creator: Cokinos, D.M.
Partner: UNT Libraries Government Documents Department

Calculation of nuclear heating rates in LMFBR control rods

Description: The results of calculations of power densities and worths for control rods which use boron for neutron absorption are presented. The calculations were performed as a function of the control assembly core location, axial position, and /sup 10/B enrichment. (JWR)
Date: January 1, 1973
Creator: Gibson, G. & Dyos, M.W.
Partner: UNT Libraries Government Documents Department

Critical Experiments and Cooperative Evaluations Program seventh quarterly report, April--June 1973

Description: Current activities and technical progress for the period April through June 1973 are reported for two subtasks. Subtask A covers the planning and analysls of critical experiments in ZPPR-3. Subtask B progress is reported for nuclear data evaluation and testing and methods development in support of Subtask A. (auth)
Date: June 1, 1973
Partner: UNT Libraries Government Documents Department

ZPPR progress report, January 1989 through April 1989

Description: Further results are presented from the large, homogeneous assembly ZPPR-18 in the JUPITER-III program. Reaction rate results are given for ZPPR-18B along with measured gamma ray dose results from ZPPR-18A and 18B. Control rod worth results from the ZPPR-18 assemblies are included. Calculation models, measured and calculated k-effective values and measured sodium worth values, are presented for the ZPPR-19 assemblies of the lo program.
Date: April 27, 1989
Creator: Collins, P.J. & Brumbach, S.B.
Partner: UNT Libraries Government Documents Department