### Abstract

Original language | English |
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Title of host publication | SAFIR, The Finnish Research Programme on Nuclear Power Plant Safety 2003 - 2006, Interim Report |

Publisher | VTT Technical Research Centre of Finland |

Chapter | 2.2 |

Pages | 30-37 |

Publication status | Published - 2004 |

MoE publication type | D2 Article in professional manuals or guides or professional information systems or text book material |

### Publication series

Series | VTT Tiedotteita - Research Notes |
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Number | 2272 |

ISSN | 1235-0605 |

### Fingerprint

### Cite this

*SAFIR, The Finnish Research Programme on Nuclear Power Plant Safety 2003 - 2006, Interim Report*(pp. 30-37). VTT Technical Research Centre of Finland. VTT Tiedotteita - Research Notes, No. 2272

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*SAFIR, The Finnish Research Programme on Nuclear Power Plant Safety 2003 - 2006, Interim Report.*VTT Technical Research Centre of Finland, VTT Tiedotteita - Research Notes, no. 2272, pp. 30-37.

**A systematic study of cross section library based discrepancies in LWR criticality calculations.** / Leppänen, Jaakko.

Research output: Chapter in Book/Report/Conference proceeding › Chapter or book article › Professional

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 -