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.

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.

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 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.