TY - CHAP

T1 - A systematic study of cross section library based discrepancies in LWR criticality calculations

AU - Leppänen, Jaakko

PY - 2004

Y1 - 2004

N2 - Discrepancies in fundamental nuclear data pose a
potential source of uncertainty in neutron transport
calculation. These discrepancies are often overlooked in
reactor physics calculations. There is evidence that
differences in the order of 1% in the multiplication
factor are encountered when criticality calculations are
carried out using different neutron cross section
libraries. Monte Carlo transport calculation codes are
especially prone to such discrepancies, since the base
evaluated data can be used without modifications.
This paper presents some results of a study, in which
cross section library-based discrepancies in light water
reactor criticality calculations were investigated in a
systematic manner. Point-wise cross section libraries
were generated from the ENDF/B-VI.8, JEFF-3.0, JENDL-3.3,
JEF-2.2 and JENDL-3.2 evaluated nuclear data files using
the NJOY-99 nuclear data processing system. The
comparison calculations were carried out using the MCNP4C
Monte Carlo transport calculation code. A systematic
method based on the neutron balance of the system was
developed in order to study the origin of the reactivity
discrepancies. The energy dependence of the cross section
data was taken into account by dividing the neutron flux
spectrum into four energy groups. The comparison
calculations cover the most typical LWR operating
conditions. The basic geometry is an infinite pin-cell
lattice. The main free parameter in the system is the
fuel-to-moderator ratio. Several variations of the basic
geometry were studied, including lattices with burnable
absorber and control pins, a finite lattice with leakage
and lattices with low- and high-burnup fuel pins

AB - Discrepancies in fundamental nuclear data pose a
potential source of uncertainty in neutron transport
calculation. These discrepancies are often overlooked in
reactor physics calculations. There is evidence that
differences in the order of 1% in the multiplication
factor are encountered when criticality calculations are
carried out using different neutron cross section
libraries. Monte Carlo transport calculation codes are
especially prone to such discrepancies, since the base
evaluated data can be used without modifications.
This paper presents some results of a study, in which
cross section library-based discrepancies in light water
reactor criticality calculations were investigated in a
systematic manner. Point-wise cross section libraries
were generated from the ENDF/B-VI.8, JEFF-3.0, JENDL-3.3,
JEF-2.2 and JENDL-3.2 evaluated nuclear data files using
the NJOY-99 nuclear data processing system. The
comparison calculations were carried out using the MCNP4C
Monte Carlo transport calculation code. A systematic
method based on the neutron balance of the system was
developed in order to study the origin of the reactivity
discrepancies. The energy dependence of the cross section
data was taken into account by dividing the neutron flux
spectrum into four energy groups. The comparison
calculations cover the most typical LWR operating
conditions. The basic geometry is an infinite pin-cell
lattice. The main free parameter in the system is the
fuel-to-moderator ratio. Several variations of the basic
geometry were studied, including lattices with burnable
absorber and control pins, a finite lattice with leakage
and lattices with low- and high-burnup fuel pins

M3 - Chapter or book article

T3 - VTT Tiedotteita - Research Notes

SP - 30

EP - 37

BT - SAFIR, The Finnish Research Programme on Nuclear Power Plant Safety 2003 - 2006, Interim Report

PB - VTT Technical Research Centre of Finland

ER -