RESUMEN
Neutron activation analysis is the reference method used for offline determination of the neutron flux density in defined positions. It can be used in the nuclear energy industry-as well as in medical- or space applications. For accurate neutron flux evaluation, well-known and reliable cross sections are needed. In the thermal and fast energy region, many reliable monitoring reactions exists, however, in case of the epithermal and intermediate energy region, there are practically no dosimetry nuclear reactions sensitive specifically in this energy range. Due to this fact, both new data are being measured and methodologies are under development to describe and test this energy region. It was found that various neutron filters can be used to cut parts of neutron spectra and thus methodology based on spectrum filtering could potentially be employed to survey cross sections of interest. It this paper, the use of 3 different filters - B4C, Cd, and In is studied, on the case of the 55Mn(n,γ) reaction. Measured values of that cross section in the given filtered reference spectra are reported.
RESUMEN
Only neutron spectrum standard is 252Cf spontaneous fission neutron spectrum. However, the high energy tail of this spectrum is loaded with high uncertainty. To reduce this uncertainty, it is crucial to use validated cross sections with low uncertainty. The explored set of reactions covers 58Ni(n,X)57Co, 169Tm(n,3n)167Tm, 197Au(n,3n)195Au, 209Bi(n,3n)207Bi and 209Bi(n,4n)206Bi threshold reactions. Measurement of dosimetric 169Tm(n,3n)167Tm, 209Bi(n,3n)207Bi and 209Bi(n,4n)206Bi reactions spectral averaged cross sections (SACS) is included in NEA's High Priority Nuclear Data Request List. All these reactions SACS were measured for the first time. All SACS were derived from experimentally determined reaction rates by gamma spectrometry using the same high-purity germanium detector. In the case of 169Tm(n,3n)167Tm reaction, the difference between experimental and calculated value using the IRDFF-II library is only 1.45%. Concerning 197Au(n,3n)195Au reaction, the reasonable agreement is achieved only using the TENDL-2017 library. In the case of 209Bi(n,3n)207Bi reaction, agreement within uncertainty is not achieved with any library unlike 209Bi(n,4n)206Bi reaction where the agreement within uncertainty is achieved with IRDFF-II library. The best agreement for 58Ni(n,X)57Co reaction is achieved using ENDF/B-VIII library.
RESUMEN
The neutron flux distribution behind a reactor pressure vessel (RPV) is an important parameter that is monitored to determine neutron fluence in the RPV. Together with mechanical testing of surveillance specimens, these are the most important parts of in-service inspection programs that are essential for a realistic and reliable assessment of the RPV residual lifetime. The fast neutron fluence values are determined by a calculation. These calculation results are accompanied by measurements of induced activities of the activation foils placed in the capsules behind the RPV at selected locations, namely in azimuthal profile. In case of discrepancies between the measured and calculated activities of the activation foils placed behind the pressure vessel, it is difficult to determine the source of the deviation. During such analysis, there arises a question on the influence of power peaking near core boundary on neutron profile behind the RPV. This paper compares the calculated and measured increase of the neutron flux density distribution behind the reactor pressure vessel in the azimuthal profile that has arisen from the replacement of 164 fuel pins located close to reactor internals by pins with the higher enrichment. This work can be understood as the first step in the characterization of the effect of incorrectly calculated pin power or burn-up in the fuel assembly at the core boundary relative to the neutron flux distribution behind reactor pressure vessel. Based on a good agreement between the calculated and experimental values, it can be concluded that the mathematical model used to evaluate the power increase is correct.