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.

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.

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.

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.

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.

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.

Application of 252Cf neutron source for precise nuclear data experiments.

Purpose of this paper is to provide extensive information helpful for anyone performing any experiment involving 252Cf neutron source and aiming for high precision experiments. The paper summarizes basic characteristics and fields of study using 252Cf neutron source. We show the basic characteristics of our source, precise geometry in MCNP6, isotopic content, distribution of the source in the encapsulation and possible use of encapsulation for 27Al(n,2n)26Al reaction estimation and the way of handling of the 252Cf neutron source in our laboratory. Furthermore, we prove that our source is volumetric, i.e. non-point and can be considered as an isotropic in our experimental settings. Influence of the palladium matrix density on the reaction rates is also investigated. Concerning nuclear data, we are measuring fast neutron leakage spectra in various benchmark sets including spheres of different materials. Our work extends benchmark sets to possible use of cubes, in our case graphite cube of side 30 cm. The neutron spectra are measured by stilbene scintilation detector in the energy range of 1-10 MeV in the steps of 100 keV in the distance of 100 cm from the centre of the cube. There is no significant difference between measurements performed using cube or sphere except lower cost for the cube production. The agreement between calculation using ENDF/B-VII.1 library and experimental data is not satisfactory especially in the regions around 3.5 MeV and 8 MeV. The agreement within 4% of calculation to experiment ratio for neutron leakage fluxes of pure 252Cf source validates our MCNP model for studying reaction rates and leakage fluxes in the region above 1 MeV.

Validation of selected (n,2n) dosimetry reactions in IRDFF-1.05 library.

Spectrum-averaged cross sections (SACS) have been measured in the reference 252Cf(sf) neutron field for the following high-threshold (n,2n) neutron dosimetry reactions since they are especially important due to the high threshold which allows validation of upper parts of prompt fission neutron spectrum. This work includes 59Co(n,2n)58Co, 197Au(n,2n)196Au, 169Tm(n,2n)168Tm, 55Mn(n,2n)54Mn, 93Nb(n,2n)92 mNb and 89Y(n,2n)88Y and for the 59Co(n,p)59Fe threshold reactions. SACS were inferred from experimentally determined reaction rates by gamma spectrometry using a semiconductor high-purity germanium detector to measure irradiated samples. Measured reaction rates agree within quoted uncertainties with those calculated from the IRDFF-1.05 library, except for the reaction 55Mn(n,2n)54Mn, for which the measured value is underestimated by 16%. For this reaction the ENDF-B/VII.1 evaluation agrees with measured reaction rate within uncertainties.

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.

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.

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.

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.

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.

Natural iron isotopes influence on the neutron transport.

As an iron is the main structural component of nuclear power plants as well as future fusion power plants, the validation of neutron incident data libraries of iron is a must. Presented paper fits into ongoing validation activities and presents measuring neutron leakage spectra in the 0.1-1.0MeV region from iron sphere of 100cm in diameter by hydrogen proportional detectors. The experimental result is compared with ENDF/B-VII.1, JEFF-3.2 and CIELO nuclear data libraries. No library reasonably well describes whole region under study. Furthermore, elastic and inelastic XS sensitivity analysis for all iron isotopes was carried out. 54Fe isotope elastic XS influence is comparable with 56Fe XS influence up to 0.8MeV. 57Fe isotope elastic XS is significant in the region of 0.14-0.15MeV. Additionally, there are large differences among libraries in both elastic and inelastic XS for 57Fe. Furthermore, it was found that 58Fe isotope XS has negligible influence on the results. As a neutron source, 252Cf with initial emission rate of 9.53E8 n/s was used in this experiment.

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).

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.

Measurement of reaction rates for different neutron induced reactions in 27Al.

The presented paper aims to compare various measured neutron induced reaction rates in Aluminium with computed ones in different nuclear data libraries. A 252Cf neutron source with emission rate of 9.53E8 n/s was used. Reactions involved in the study were 27Al(n,g), 27Al (n,p) and 27Al (n,α).