Measurement of spectral averaged cross sections of capture reactions in the LR-0 reference neutron field.

Nuclear data evaluations are periodically updated to render state of the art of knowledge, and the independent validation experiments are still of interest of the community. Current work describes validation of selected capture reactions used for monitoring of thermal neutrons in mixed fields, as well as reactions responsible for activation of components exposed to neutron flux. The validation was performed also for reaction 50Cr (n,γ), which is very important reaction directly affecting criticality safety, because chromium is essential component of stainless steel used as structural component of the core baffle closely adjoining to nuclear fuel. Described experiments were performed in reference neutron field of the LR-0 reactor. The activation rate was derived by gamma spectrometry using well-characterized HPGe detector. It was found that the capture cross sections for the dosimetry reactions 23Na (n,γ), 58Fe (n,γ), and 59Co (n,γ) agreed well within 10 % with calculations in thermal and epithermal region, but larger discrepancy was found for the isotopes of tin.

The spectrum-averaged cross section investigation of 117Sn(n,n')117mSn and 67Zn(n,p)67Cu reactions.

The spectrum averaged cross sections (SACS) in standard neutron field, e.g. 252Cf(s.f.), is a preferable tool for cross section evaluation and validation. A set of reaction measurements with high energy thresholds was previously performed. The presented work focuses on lower energy threshold reactions, namely on the inelastic scattering of the tin foil, more specifically the reaction 117Sn(n,n')117mSn, and the zinc foil reaction, namely 67Zn(n,p)67Cu. These reactions are of special interest due to their intermediate energy range, which is essential in classical reactor dosimetry and fast reactor dosimetry. The experiments were carried out in a standard neutron field formed by 252Cf(s.f.) source in Rez. The experimental results were compared with calculations using MCNP6.2, ENDF/B-VII.1 transport library, and ENDF/B-VIII.0 and IRDFF-II cross section data library. Additionally, the calculations using CEA code DARWIN/PEPIN2 using JEFF-3.0/A were executed. The obtained experimental SACS of previously measured reactions were in good agreement with the SACS calculated using the IRDFF-II library. Additionally, the calculational reaction rate of 67Zn(n,p)67Cu was in accordance with the experimental data in case of ENDF/B-VIII.0 nuclear data library. Moreover, the calculational results of 117Sn(n,n')117mSn obtained by DARWIN/PEPIN2 code (using JEFF-3.0/A nuclear data library) are closest to the experimental results.

Characterization of the secondary neutron field inside a cyclotron for production of radiopharmaceuticals.

During production of radiopharmaceuticals, the radiation situation in cyclotron pit is an important parameter, which is being monitored to ensure fulfilment of the limits and conditions of safe operation. The neutron flux in the structural components of the accelerator is also an important parameter, because the secondary neutrons are responsible for activation of cyclotron structural components and may even affect structural changes in it. This paper aims to characterize the neutron field in inner positions of medical accelerator IBA 18/9 by activation detectors and by means of scintillation spectrometry. The backward angle measurement was realized also in special liquid water target (H218O) at U120M cyclotron to confirm the data obtained in IBA 18/9 cyclotron.

MEASUREMENT OF PROMPT GAMMAS OF INELASTIC SCATTERING IN D2O SPHERE USING AMBE SOURCE.

The inelastic neutron scattering is often followed by the emission of gamma photon. As the prompt gammas have a discrete level character they can be used for the identification of nuclides. Because of this fact, a good knowledge of photon production from inelastic scattering is important. Described research deals with the measurement of gamma originated from inelastic scattering of neutrons on 16O. The 241Am-Be was used as a neutron source because of its high average neutron energy. The oxygen in form of heavy water was used for maximization of neutron flux on oxygen and minimization of background gammas' production, namely 2223 keV gammas accompanying capture on hydrogen 1H. The gamma spectrum was measured by HPGe and the stilbene detector. The HPGe measured quantities are comparted with calculation and discrepancies between measured and calculated gamma fluxes are reported. Stilbene measurement shows indistinguishability of gamma peaks above 6 MeV.

Spectrum averaged cross section measurements of lutetium using standard 252Cf neutron source.

The spectrum averaged cross section (SACS) in a standard neutron field is a preferable tool for cross section validation. The presented work uses only neutron standard, i.e., 252Cf(sf) reaction neutron field, for validation of lutetium threshold cross sections. SACS were inferred from gamma spectrometry derived reaction rates. The SACS which were derived include 175Lu (n,2n)174Lu, 175Lu (n,3n)173Lu, 175Lu (n,p)175Yb, and 176Lu (n,n')176m1Lu reactions. All these reactions SACS were measured for the first time. MCNP6.2 calculations using JEFF-3.3 or ENDF/B-VIII.0 libraries for lutetium cross sections were compared with experimental data. The agreement was found very poor for all reactions under study. Thus there is a need for their improvement. The presented data can be also used for validation of the various theoretical models.

Testing of various neutron filters in reference neutron field in LR-0 reactor for nuclear data validation and verification.

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.

Constraining high energy tail of 235U(nth,f) prompt fission neutron spectrum.

There is a lack of reliable experiments aiming at the prompt fission neutron spectrum of 235U for energies higher than 10 MeV. The presented experiment performed at the LVR-15 light water reactor aimed at the measurement of very high threshold reactions spectral averaged cross sections such as 55Mn (n,2n)54Mn, 197Au (n,2n)196Au, 197Au (n,3n)195Au, 209Bi(n,3n)207Bi, 209Bi(n,4n)206Bi. 209Bi(n,3n)207Bi and 209Bi(n,4n)206Bi reactions were measured for the first time. 58Ni(n,p)58Co reaction was used as a monitor reaction. The experimental spectral averaged cross sections are derived from reaction rates measured by means of high purity Germanium spectroscopy at the well defined detector. The experimental spectral averaged cross sections are compared with calculations using either ENDF/B-VIII.0 or JEFF-3.3 235U prompt fission neutron spectrum and IRDFF-II cross sections. The discrepancy is higher with higher mean response energy for JEFF-3.3 unlike ENDF/B-VIII.0, where the agreement is good within broad range of mean response energy. Moreover, due to the high thermal neutron flux in the reactor, the experimental reaction rate is compared with calculated 198Au (n,g)199Au reaction rate. The difference of -55.3% for double capture reaction was found in comparison with ROSFOND-2010 calculations.

The influence of core power distribution on neutron flux density behind a pressure vessel of a VVER-1000 Mock Up in LR-0 reactor.

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.

Validation of nickel isotopes neutron cross sections using nickel spherical benchmark.

The fast leakage neutron spectra have been measured on spherical nickel benchmark assembly of diameter 50 cm. The 252Cf neutron source with approximate emission of 5.0·108 n/s was placed into the centre of the sphere. Fast neutron spectrum in the range of 1-10 MeV was measured in the distance of 1 m from the sphere centre by means of proton recoil method using scintillation stilbene crystal. The experimental data were compared to transport calculations based on several evaluated nuclear data libraries using MCNP6. MCNP6 was compared with SCALE/MONACO program using the same ENDF/B-VII.1 library which leads to different differential neutron flux results. Best experimental agreement with calculation in MCNP6 is achieved with ENDF/B-VII.1 library. Contrary, worst agreement is achieved with JENDL-4.0 and CENDL-3.1 libraries. Furthermore, cross section sensitivity analysis for elastic and inelastic scattering for both main nickel isotopes (58Ni, 60Ni) was performed. It was shown that 58Ni isotope has higher influence on the result than 60Ni isotope in the entire energy range under study. The highest influence has the elastic XS of 58Ni around energy of 1.5 MeV. The inelastic cross section (XS) of 58Ni dominates in the energies above 2 MeV where two percent rise due to the inelastic XS leads up to 3.3% decrease in the neutron flux.