Program - Brookhaven National Laboratory

PR 66
Processing and Validation of JEFF3.1.2 Neutron Cross-Section Library into Various
Formats: ACE, PENDF, GENDF, MATXSR and BOXER.
O. Cabellos, Department of Nuclear Engineering, Universidad Politécnica de Madrid, Spain.
Following the processing and validation of JEFF-3.1 performed in 2006 and presented in ND2007, and as
a consequence of the latest updated of this library (JEFF-3.1.2) in February 2012, a new processing and
validation of JEFF-3.1.2 neutron cross-section library is presented in this paper. The nuclides processed
are all the evaluations of the General Purpose Library and Thermal Scattering JEFF-3.1.2 Library, including
the important light nuclei, structural materials, fission products, control rod materials and burnable
poisons, all major and minor actinides. The processed library was generated with NJOY-99.364 nuclear
data processing system plus some specific updates. The processed library contains continuous energy neutron
cross-section data files at ten different temperatures in ACE format. In addition, NJOY inputs are
provided to generate PENDF, GENDF, MATXSR and BOXER formats. The processed library has undergone
strict Q & A procedures, being compared with other available libraries (JEFF-3.1.1, ENDF-B/VII.0
or VII.1) using JANIS-3.4 and PREPRO-2000 codes. MCNP5 is used for validation purposes with a set
of 119 criticality benchmark experiments taken from ICSBEP-2010. This work has been done with the
support of the OECD/NEA Data Bank.
PR 67
Data Adjustment for Coarse Group Structures used in Neutronics Calculations.
R.B. Thom, AWE.Plc.
Normal practice in multi-dimensional neutronics calculations is to use coarse group structures due to computational
time constraints. Condensing data to coarse group structures results in a loss of accuracy and
this is exacerbated by the transport approximations used, eg Sn order and Legendre scatter order. At
AWE, for fissile systems, compensation for this loss of accuracy is made by adjusting principal cross sections
of the most important nuclides, specifically Pu239, U235 and U238. Adjustments are deduced by
application of the AWE code NDxadj and may be made to the fission cross section, the total inelastic
cross section and nubar (the average number of neutrons emitted in fission); compensating adjustment
is made to the elastic cross section to preserve the total cross section. The principle of the adjustment
method is to vary a set of cross sections, often in defined energy regions, until a close match is found with
experimental data, usually k-effective for standard critical assemblies; the minimisation technique used is
based on Powell’s method[1]. The paper will demonstrate the effectiveness of the adjustment method by
comparing calculations of critical systems (taken from ICSBEP[2]), using unadjusted and adjusted data.
British Crown Owned Copyright 2012/AWE Published with the permission of the Controller of Her Britannic
Majesty’s Stationary Office. [1] ”Direct Search Algorithms for Optimisation Calculations”, MJ Powell,
Acta Numerica, pp 287-336, (1998). [2] ”International handbook of evaluated criticality saftey benchmark
experiments”, J.B Briggs et al., Tech Report NEA/NSC/DOC(95)04/I, (2004).
PR 68
296