ALEPH2 - A general purpose Monte Carlo depletion code
Conference
·
OSTI ID:22105706
- SCK CEN, Boeretang 200, B-2400 Mol (Belgium)
- AREVA NP, Tour AREVA, Pl. J. Millier, 92084 Paris La Defense (France)
The Monte-Carlo burn-up code ALEPH is being developed at SCK-CEN since 2004. A previous version of the code implemented the coupling between the Monte Carlo transport (any version of MCNP or MCNPX) and the ' deterministic' depletion code ORIGEN-2.2 but had important deficiencies in nuclear data treatment and limitations inherent to ORIGEN-2.2. A new version of the code, ALEPH2, has several unique features making it outstanding among other depletion codes. The most important feature is full data consistency between steady-state Monte Carlo and time-dependent depletion calculations. The last generation general-purpose nuclear data libraries (JEFF-3.1.1, ENDF/B-VII and JENDL-4) are fully implemented, including special purpose activation, spontaneous fission, fission product yield and radioactive decay data. The built-in depletion algorithm allows to eliminate the uncertainties associated with obtaining the time-dependent nuclide concentrations. A predictor-corrector mechanism, calculation of nuclear heating, calculation of decay heat, decay neutron sources are available as well. The validation of the code on the results of REBUS experimental program has been performed. The ALEPH2 has shown better agreement with measured data than other depletion codes. (authors)
- Research Organization:
- American Nuclear Society, Inc., 555 N. Kensington Avenue, La Grange Park, Illinois 60526 (United States)
- OSTI ID:
- 22105706
- Country of Publication:
- United States
- Language:
- English
Similar Records
Analysis of fission-product energy release after thermal-neutron fission of {sup 235}U and {sup 239}Pu by scale-6.1.3 with recently available fission product yield data - 14419
Analysis of Reactor Physics Experiment for the Irradiated LWR MOX Fuels
Enhanced Monte-Carlo-Linked Depletion Capabilities in MCNPX
Conference
·
Tue Sep 15 00:00:00 EDT 2015
·
OSTI ID:23100874
Analysis of Reactor Physics Experiment for the Irradiated LWR MOX Fuels
Conference
·
Sat Jul 01 00:00:00 EDT 2006
·
OSTI ID:21016408
Enhanced Monte-Carlo-Linked Depletion Capabilities in MCNPX
Conference
·
Sat Jul 01 00:00:00 EDT 2006
·
OSTI ID:21021173