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.

Measurement of the selected spectral averaged cross sections in a radial channel of the VR-1 reactor.

The spectral averaged cross section is an important quantity used in a validation of nuclear cross section. When the cross sections are averaged over the neutron standard field (252Cf(s,f) or 235U(n,f) neutron spectrum), they can be used for tuning of evaluations. This kind of quantities is very useful because the data in integral measurements can be determined with a significantly smaller uncertainties than the standard differential data. The experiment was aimed at the spectral average cross sections measurement and was performed in a radial channel of VR-1 reactor (with fuel enrichment 19.75 wt %). The results are in a good agreement within the uncertainties with a previous measurements in LR-0 reactor (with fuel enrichment 3.3 wt %), thus it supports the hypothesis that even significant amount of 238U(n,f) neutrons in the LR-0 reactor spectrum does not have a significant influence. The derived spectral averaged cross sections are as follows: 0.1709 ± 0.0115 mb for 89Y(n,2n), 10.738 ± 0.719 mb for 46Ti(n,p), 17.896 ± 1.181 mb for 47Ti(n,p), 0.294 ± 0.02 mb for 48Ti(n,p), 72.994 ± 4.964 mb for 54Fe(n,p), 0.528 ± 0.036 mb for 63Cu(n,α), 0.444 ± 0.029 mb for 93Nb(n,2n)92Nb* and 0.239 ± 0.016 mb for 58Ni(n,x)57Co.

Measurements of neutron transport of well defined silicon filtered beam in lead.

The correct description of neutron transport in lead is an essential task for correct description of tritium production in the DEMO (DEMOnstration Power Station) breeding blanket because some concepts deal with lead as a major component: namely the WCLL (water cooled lithium lead blanket), HCLL (helium cooled lithium lead blanket), and DCLL (dual cooled lithium lead blanket). Concerning the improvement of the knowledge about the transport of fast neutrons in lead, a set of experiments and calculations was carried out to study this problem with a well-defined neutron beam. The neutron flux behind various lead arrangements positioned along the beam axis was measured using a stilbene scintillation crystal (10 mm × 10 mm) with neutron and gamma pulse shape discrimination. The measurement was performed along the beam axis and in the case of the thick target also above the axis, to estimate the neutron angular scatter in lead. The calculations were realized using MCNP6 with various nuclear data libraries. Discrepancies in the angular distribution description in the energy region of about 1 MeV were discovered by these experiments.

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.

On similarity of various reactor spectra and 235U 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.

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 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.

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.

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.

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.

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.