Measurement of very high threshold reactions using 252Cf source.

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.

Validation of IRDFF-II library by means of 252Cf spectral averaged cross sections.

Spectrum-averaged cross sections (SACS) is an important quantity usable in validation of nuclear cross sections. Especially in case of dosimetrical reactions there is a request on precise validation. This paper presents SACS measured in the reference 252Cf(sf) neutron field for neutron dosimetry reactions to validate recently updated IRDFF-II library intended mainly for neutron metrology applications. It covers 46Ti(n,p)46Sc, 47Ti(n,p)47Sc, 48Ti(n,p)48Sc, 58Ni(n,p)58Co, 60Ni(n,p)60Co, 58Ni(n,2n)57Ni, 24Mg(n,p)24Na, 27Al(n,α)24Na and 63Cu(n,α)60Co threshold reactions. Measurement of 60Ni(n,p)60Co SACS is included in NEA's High Priority Nuclear Data Request List since existing data are discrepant. Spectral averaged cross sections were derived from experimentally determined reaction rates by gamma spectrometry using a same high-purity germanium detector to measure all irradiated samples. Measured spectrum-averaged cross sections agree very well within quoted uncertainties with those calculated using the IRDFF-II library. No other library achieves such good performance. Thus, the presented data support use of the cross sections of the mentioned reactions from IRDFF-II library.

Validation of differential cross sections by means of 252Cf spectral averaged cross sections.

The results of systematic evaluations of the spectrum-averaged cross section measurements performed in the spontaneous fission 252Cf neutron field are presented. The Following threshold reactions were investigated: 23Na(n,2n)22Na, 54Fe(n,p)54Mn, 54Fe(n,α) 51Cr, 27Al(n,p)27Mg, 27Al(n,α)24Na, 19F(n,2n)18F, 90Zr(n,2n)89Zr and 89Y(n,2n)88Y. The spectrum-averaged cross sections for 23Na(n,2n)22Na, 54Fe(n,α)51Cr and 89Y(n,2n)88Y reactions were measured for the first time. This quantity is compared with calculations carried with the IRDFF-v1.05 library. There is a notable disagreement exceeding uncertainties only for 54Fe(n,p)54Mn and 54Fe(n,α) 51Cr reactions. The spectrum-averaged cross sections were inferred from experimentally determined reaction rates. The experimental reaction rates were derived for irradiated samples from the Net Peak Areas measured using the semiconductor high purity germanium spectroscopy. The presented experimental data can be used to validate nuclear data libraries and reactions used in the practical reactor dosimetry and to specify high energy tail of the 252Cf neutron spectrum.

Validation of zirconium isotopes (n,g) and (n,2n) cross sections in a comprehensive LR-0 reactor operative parameters set.

Zirconium is an important material used in most of reactor concepts for fuel cladding. Thus the knowledge of its cross section is important for reliable prediction of fuel operation. Also 90Zr(n,2n) reaction, is included in IRDFF files as dosimetry cross section standard. Due to its very high threshold, 12.1MeV, it is suitable for measurement of high energy neutrons. One of possible interesting applications is also evaluation of prompt fission neutron spectra in 235U and 238U what is under auspices of the International Atomic Energy Agency in CIELO project. The experimental values - obtained with the LR-0 nuclear reactor - of various zirconium cross sections were compared with calculations with the MCNP6 code using IAEA CIELO, ENDF/B-VII.0, JEFF-3.1, JEFF-3.2, JENDL-3.3, JENDL-4, ROSFOND- 2010, and CENDL-3.1 transport libraries combined with the dosimetry cross sections extracted from the IRDFF library. Generally, the best C/E agreement for 90Zr(n,2n) cross section, was found with the IAEA CIELO 235U evaluation that includes an updated prompt fission neutron spectra in the evaluated data file. The cross section of this reaction averaged over LR-0 spectra was determined being 28.9 ± 1.2 µb, corrected to spectral shift, spectral averaged cross section in 235U was determined to be 0.107 ± 0.005mb. Notable discrepancies were reported in both 94Zr(n,g) and 96Zr(n,g).

Measurement of (23)Na(n,2n) cross section in well-defined reactor spectra.

The present paper aims to compare the calculated and experimental reaction rates of (23)Na(n,2n)(22)Na in a well-defined reactor spectra of a special core assembled in the LR-0 reactor. The experimentally determined reaction rate, derived using gamma spectroscopy of irradiated NaF sample, is used for average cross section determination. The resulting value averaged in spectra is 0.91±0.02µb. This cross-section is important as it is included in International Reactor Dosimetry and Fusion File and is also relevant to the correct estimation of long-term activity of Na coolant in Sodium Fast Reactors. The calculations were performed with the MCNP6 code using ENDF/B-VII.0, JEFF-3.1, JEFF-3.2, JENDL-3.3, JENDL-4, ROSFOND-2010 and CENDL-3.1 nuclear data libraries. Generally the best C/E agreement, within 2%, was found using the ROSFOND-2010 data set, whereas the worst, as high as 40%, was found using the ENDF/B-VII.0.