Super-Phenix Benchmark used for Comparison of PNC and CEA Calculation Methods,and of JENDL-3.2 and CARNAVAL IV Nuclear Data

Hunter

PNC TN9410 98-015, 81 Pages, 1998/02

PNC-TN9410-98-015.pdf:3.15MB

The study was carried out within the framework of the PNC-CEA collaboration agreement. Data were provided, by CEA, for an experimental loading of a start-up core in Super-Phenix. This data was used at PNC to produce core flux snapshot calculations. CEA undertook a comparison of the PNC results with the equivalent calculations carried out by CEA, and also with experimental measurements from SPX. The resu1ts revealed a systematic radial flux tilt between the calculations and the reactor measurements, with the PNC tilts only 30-401 of those from CEA. CEA carried out an analysis of the component causes of the radial tilt. It was concluded that a major cause of radia1 tilt differences between the PNC and CEA calculations lay in the nuclear datasets used: JENDL-3.2 and CARNAVAL IV. For the final stage of the study, PNC undertook a sensitivity analysis, to examine the detailed differences between the two sets of nuclear data. The PNC flux calculations modelled SPX in both 2D (RZ) and 3D (hex-Z) geometries, using the diffusion programs CITATION and MOSES. The sensitivity analysis of the differences between the JENDL-3.2 and CARNAVAL IV nuclear datasets used the SAGEP calculational route. Both datasets were condensed to a single, non-standard, set of energy group boundaries. There were some incompatibilities in the cross-section formats of the two datasets. The sensitivity analysis showed that a relatively small number of nuclear data items contributed the bulk of the radial tilt difference between calculations with JENDL-3.2 and with CARNAVAL IV. A direct comparison between JENDL-3.2 and CARNAVAL IV data revealed the following. The Nu values showed little difference (<5|%). The only large fission cross-section differences were at low energy (<30% otherwise, with <10% typical). Although down-scattering reactions showed some large fractional differences, absolute differences were negligible compared with in-group scattering; for in-group scattering fractional ...

Development of sodium columnar combustion code SOFIRE-M3

Ohno, Shuji;

PNC TN9410 92-370, 54 Pages, 1992/11

PNC-TN9410-92-370.pdf:1.19MB

Conputational code SOFIRE-M3 has been developed to evaluate the thermal consequences which would be brought by the sodium columnar leak and fire accident postulated in the FBRs. New code is the improved version of SOFIRE-MII and has following features. (1)Sodium columnar combustion rate is calculated from the empirical formula which has been derived from the results of Run-E3 test series and has the terms of sodium leak rate, leak height, and oxygen concentration in the atmosphere. (2)Calculational parameters of heat distribution fraction, which determine the reaction heat transferred to both sodium and gas, have been optimized by the post-test calculations of Run-E3 using the SOFIRE-M3 code. The code predicts heat transfer phenomena within 30% of accuracy when the optimized parameters are used. Calculations of large-scale sodium leak and fire tests Run-E2 and Run-D2, which had been performed at the SAPFIRE facility in 1985 and 1986 respectively, showed good agreement between code calculated and measured data. This SOFIRE-M3 code can contribute in the near future to the more optimized designing and safety evaluation of FBR p1ant. Refinement of the empirical formula and improvement of the code remain as future subjects related to the effect of sodium leak rate on the columnar combustion rate.

Kinetic study on incineration of ion exchange resins by fluidized bed

Yahata, Taneaki; *; Hirata, Masaru; Abe, Masayoshi

Proc. of the 3rd Int. Conf. on Nuclear Fuel Reprocessing and Waste Management; RECOD91,Vol. 2, p.790 - 795, 1991/00

no abstracts in English