Michal Košťál

Measurement and calculation of fast neutron and gamma spectra in well defined cores in LR-0 reactor

A well-defined neutron spectrum is essential for many types of experimental topics and is also important for both calibration and testing of spectrometric and dosimetric detectors. Provided it is well described, such a spectrum can also be employed as a reference neutron field that is suitable for validating selected cross sections. The present paper aims to compare calculations and measurements of such a well-defined spectra in geometrically similar cores of the LR-0 reactor with fuel containing slightly different enrichments (2%, 3.3% and 3.6%). The common feature to all cores is a centrally located dry channel which can be used for the insertion of studied materials. The calculation of neutron and gamma spectra was realized with the MCNP6 code using ENDF/B-VII.0, JEFF-3.1, JENDL-3.3, ROSFOND-2010 and CENDL-3.1 nuclear data libraries. Only minor differences in neutron and gamma spectra were found in the comparison of the presented reactor cores with different fuel enrichments. One exception is the gamma spectrum in the higher energy region (above 8MeV), where more pronounced variations could be observed.

Determination of critical assembly absolute power using post-irradiation activation measurement of week-lived fission products

The work presents a detailed comparison of calculated and experimentally determined net peak areas of longer-living fission products after 100 h irradiation on a reactor with power of ~630 W and several days cooling. Specifically the nuclides studied are (140)Ba, (103)Ru, (131)I, (141)Ce, (95)Zr. The good agreement between the calculated and measured net peak areas, which is better than in determination using short lived (92)Sr, is reported. The experiment was conducted on the VVER-1000 mock-up installed on the LR-0 reactor. The Monte Carlo approach has been used for calculations. The influence of different data libraries on results of calculation is discussed as well.

Measurement of 23Na(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.

Measurement of neutron spectra in a silicon filtered neutron beam using stilbene detectors at the LVR-15 research reactor

A well-defined neutron spectrum is an essential tool for calibration and tests of spectrometry and dosimetry detectors, and evaluation methods for spectra processing. Many of the nowadays used neutron standards are calibrated against a fission spectrum which has a rather smooth energy dependence. In recent time, at the LVR-15 research reactor in Rez, an alternative approach was tested for the needs of fast neutron spectrometry detector calibration. This process comprises detector tests in a neutron beam, filtered by one meter of single-crystalline silicon, which contains several significant peaks in the fast neutron energy range. Tests in such neutron field can possibly reveal specific problems in the deconvolution matrix of the detection system, which may stay hidden in fields with a smooth structure and can provide a tool for a proper energy calibration. Test with several stilbene scintillator crystals in two different beam configurations supplemented by Monte-Carlo transport calculations have been carried out. The results have shown a high level of agreement between the experimental data and simulation, proving thus the accuracy of used deconvolution matrix. The chosen approach can, thus, provide a well-defined neutron reference field with a peaked structure for further tests of spectra evaluation methods and scintillation detector energy calibration.

On 54Fe neutron cross section importance in iron

The presented paper aims to evaluate the importance of 54Fe XS in iron by means of measuring the reaction rates of the selected reactions on 54Fe and measuring a fast neutron leakage spectra from the iron sphere of 100cm in diameter by a stilbene scintillation detector with subsequent XS sensitivity analysis. The reactions involved in the study were 54Fe(n,p) and 54Fe(n,α). Measured neutron induced reaction rates in 54Fe are compared with calculated ones in different nuclear data libraries. We show that there are notable discrepancies in 54Fe(n,α) reaction. The results of the leakage spectra differ significantly in various libraries, library ENDF/B-VII.1 in region 3.5-7.0MeV gives relatively good agreement. CIELO library underestimates the result; however JEFF-3.2 overestimates results., 252Cf with the emission rate of 9.53E8 n/s was used as a neutron source for all experiments involved.

Irradiation capabilities of LR-0 reactor with VVER-1000 Mock-Up core.

Even low power reactors, such as zero power reactors, are sufficient for semiconductor radiation hardness effect investigation. This reflects the fact that fluxes necessary for affecting semiconductor electrical resistance are much lower than fluxes necessary to affect material parameters. The paper aims to describe the irradiation possibilities of the LR-0 reactor with a special core arrangement corresponding to VVER-1000 dosimetry Mock-Up.

Disentangling the 16O cross section using light water and heavy water benchmark assemblies

Fast neutron leakage spectra from the light and heavy water sphere of 30cm in diameter with neutron source in its centre were measured by a stilbene scintillation detector in the region of 1-10MeV in the distance of 85cm from the spheres surface. We use the light and heavy water to eliminate the effect of hydrogen. 252Cf with the approximate emission rate of 5.5E8 n/s was used as a neutron source for all measurements involved and was placed in the centres of the spheres. The measured neutron spectra are compared with MCNP transport code calculations in ENDF/B-VII.0, ENDF/B-VIII.b4 and JENDL-4 nuclear data libraries. Experimental results for both cases follows similar trend. The best agreement is achieved with ENDF/B-VIII.b4 library in both cases. All libraries underestimate experimental measurement in the region of 3-4MeV. Furthermore, JENDL-4 library overestimates experiment in the region of 4-6.5MeV. In addition, we performed cross section sensitivity analysis for elastic, inelastic and (n,α) reaction in JENDL-4 and ENDF/B-VIII.b4 libraries since they have almost independent evaluations of 16O.

Comparison of various hours living fission products for absolute power density determination in VVER-1000 mock up in LR-0 reactor

Measuring power level of zero power reactor is a quite difficult task. Due to the absence of measurable cooling media heating, it is necessary to employ a different method. The gamma-ray spectroscopy of fission products induced within reactor operation is one of possible ways of power determination. The method is based on the proportionality between fission product buildup and released power. The (92)Sr fission product was previously preferred as nuclide for LR-0 power determination for short-time irradiation experiments. This work aims to find more appropriate candidates, because the (92)Sr, however suitable, has a short half-life, which limits the maximal measurable amount of fuel pins within a single irradiation batch. The comparison of various isotopes is realized for (92)Sr, (97)Zr, (135)I, (91)Sr, and (88)Kr. The comparison between calculated and experimentally determined (C/E-1 values) net peak areas is assessed for these fission products. Experimental results show that studied fission products, except (88)Kr, are in comparable agreement with (92)Sr results. Since (91)Sr has notably higher half-life than (92)Sr, (91)Sr seems to be more appropriate marker in experiments with a large number of measured fuel pins.

The effect of biological shielding on fast neutron and photon transport in the VVER-1000 mock-up model placed in the LR-0 reactor

The paper is intended to show the effect of a biological shielding simulator on fast neutron and photon transport in its vicinity. The fast neutron and photon fluxes were measured by means of scintillation spectroscopy using a 45×45 mm(2) and a 10×10 mm(2) cylindrical stilbene detector. The neutron spectrum was measured in the range of 0.6-10 MeV and the photon spectrum in 0.2-9 MeV. The results of the experiment are compared with calculations. The calculations were performed with various nuclear data libraries.

On similarity of various reactor spectra and 235 U prompt fission neutron spectrum

A well-defined neutron spectrum is an essential tool not only for calibration and testing of neutron detectors used in dosimetry and spectroscopy but also for validation and verification of evaluated cross sections. A new evaluation of thermal-neutron induced 235U PFNS was performed by the International Atomic Energy Agency (IAEA) in the CIELO (Collaborative International Evaluated Library Organisation Project) project; new measurements of Spectral Averaged Cross sections averaged in the evaluated spectrum are to be obtained. In general, a neutron spectrum in the core is not identical to the pure fission one because fission neutrons undergo many scattering reactions, but it can be shown that PFNS and reactor spectra become undistinguishable from a certain energy boundary. This limit is important for experiments, because when the studied reaction threshold is over this limit, the spectral averaged cross sections in PFNS can be derived from the measured reactions in the reactor core. The evaluation of the neutron spectrum measurements in three different thermal-reactor cores shows that this lower limit is around the energy of 5.5 - 6 MeV. Above this energy the reactor spectra becomes identical with the 235U PFNS. IAEA CIELO PFNS is within 5% of the measured PFNS from 10 to 14 MeV in a LR-0 reactor, while ENDF/B-VII evaluated PFNS underestimated measured neutron spectra.