235U Cross Section Research Articles

Calculation And Uncertainty Analysis of Core Parameters of ALMANAR Reactor Using New Nuclear Data Libraries

Abstract A small modular lead-cooled fast spectrum core concept called ALMANAR designed to produce 45 MWth power for 22 years operating without refuelling was proposed in a previous study. The neutronics investigation showed its excellent inherent safety features. It could be considered as a candidate for future electricity source for the near future. It is noteworthy that the target accuracy for eigenvalue calculation for keff regardless of spectrum is set to 300 pcm. However, findings in this analysis revealed that the keff uncertainty was larger for the recently released nuclear data libraries (about 800 pcm), mostly from 235U capture cross section (624 pcm) in the case of ENDF/B-VIII.0 and 238U inelastic scattering cross section (437 pcm) in the case of JENDL-5. Selected kinetic parameters of the ALMANAR core and their uncertainty were also evaluated and analysed. No major impact on the total ßeff, leffeff and λeff simulation results was found. In order to improve the reliability of criticality calculations of the lead-cooled small fast reactor, the accuracy of capture and fission cross section of 235,238U, the capture cross section of 10B and the elastic scattering cross section of 208Pb at the fast energy range of ENDF/B-VIII.0 should be improved. Furthermore, the inelastic scattering and capture cross section of 238U, fission and capture cross section of 235U and the capture cross section of 10B of JENDL-5 should also be improved.

Fission cross-section measurements for the 238U(n,f)97m+gNb, 238U(n,f)133gTe and 238U(n,f)130gSb reactions induced by D-T neutrons

In this study, we conducted measurements of the independent fission cross-sections of 238U(n, f)97m+gNb, 238U(n, f)133gTe reactions and the cumulative cross section of 238U(n, f)130gSb reactions induced by neutron at energies around 14 MeV, i.e., 14.1 ± 0.3, 14.5 ± 0.3 and 14.7 ± 0.3 MeV. The measurement results were obtained by the neutron activation method in combination with off-line γ-ray spectrometry techniques. The neutron flux was monitored on line by the accompanying α-particle from T(d, n)4He reaction, and the neutron energies were determined by the cross-section ratio of 90Zr(n, 2n)8+gZr to 93Nb(n, 2n)92mNb reactions. The independent fission cross-sections of the fission reactions were obtained by subtracting the influence of precursor nuclei or excited states. The obtained results are as follows: for 238U(n, f)97m+gNb, the independent cross sections are 1.0 ± 0.89, 0.98 ± 0.85 and 0.78 ± 0.70 mb at the specified neutron energy points. For 238U(n, f)133gTe, the independent fission cross-sections are 26.8 ± 2.8, 27.7 ± 2.9 and 20.5 ± 2.3 mb, respectively, at the same neutron energy points. As for 238U(n, f)130gSb, the obtained cumulative fission cross-sections are 5.35 ± 0.58, 5.05 ± 0.53 and 4.03 ± 0.44 mb, respectively, at the specified neutron energy points.

A comparative study on k-infinity uncertainty due to cross-section covariance data in nuclear data libraries

ABSTRACT The k-infinity uncertainty due to the cross-section covariance data (COVs) in a PWR-UO2 fuel assembly geometry was compared between the nuclear data libraries (NDLs) and the individual effects of nuclide and reaction-pair COVs were estimated. The k-infinity uncertainty was evaluated through the random sampling of CASMO5 calculations with the COVs in JENDL-4.0, JENDL-5, ENDF/B-VII.1, ENDF/B-VIII.0, JEFF-3.3, TENDL-2019, and TENDL-2021, neglecting some reaction-pair COVs, which had mathematically incorrect correlation elements. The evaluation confirmed that the major reasons for the discrepancy of the k-infinity uncertainty were the differences in the of 235U and the (n,γ) cross section of 238U. Additionally, the individual effects of the discrepancy of COVs were specified from the findings in the sensitivity coefficients of the and the (n,γ) cross section on the k-infinity for the neutron energy. Consequently, this study provided useful knowledge base for evaluating the discrepancy of k-infinity uncertainty, which is the crucial perspectives for the uncertainty analysis in the best estimate plus uncertainty approach.

Photofission and photoneutron cross sections for 238U and 232Th

A specific objective of the recent IAEA Coordinated Research Project on Photonuclear Data and Photon Strength Functions (Code F41032; Duration 2016-2019) has been to measure photonuclear cross-section data where needed, for unexplored nuclei and cases of discrepant existing data. A dedicated experimental campaign has been conducted at the laser Compton-scattering γ-ray source of the NewSUBARU synchrotron radiation facility of SPring8, Japan, where photoneutron reactions for 11 nuclei from 9Be to 209Bi have been investigated in the Giant Dipole Resonance energy region. The measurements followed the development of a flat-effi ciency moderated neutron detection array and the associated neutron-multiplicity sorting techniques. The IAEA CRP campaign has been followed at NewSUBARU by new measurements of photofission and photoneutron reactions on 238U and 232Th in the energy range of 5.87 MeV – 20.14 MeV. The neutron-multiplicity sorting of high-multiplicity fission neutron coincidence events has been performed using a dedicated energy dependent, multiple firing statistical treatment. The photoneutron (γ, xn) and photofission (γ, fxn) reactions have been discriminated by considering a Gaussian distribution of prompt-fission-neutrons (PFN) multiplicities predicted by the evaporation theory. We here provide preliminary experimental (γ, n), (γ, 2n) and (γ, F) cross sections, average energies of PFN and of photoneutrons emitted in (γ, n) and (γ, 2n) reactions, as well as the mean number of PFN per fission act. The new 238U cross sections are compared with recent statistical-model calculations performed with the EMPIRE code on existing data.

Comparison of the ENDF/B-VII.0, ENDF/B-VII.1, ENDF/B-VIII.0, and JEFF-3.3 Libraries for the Nuclear Design Calculations of the Nuclear Power Plant Krško With the CORD-2 System

Abstract Recently, two new nuclear reaction data evaluations have been released: ENDF/B-VIII.0 and JEFF-3.3. Since the neutron nuclear data profoundly influence predictions of the nuclear systems behavior, many researchers have been investigating new data striving for more accurate predictions. The purpose of this study is to examine the effects of the neutron data libraries on the nuclear design calculations of the nuclear power plant (NPP) Krško core. ENDF/B-VII.0, ENDF/B-VII.1, ENDF/B-VIII.0, and JEFF-3.3 libraries are considered. In the first part of this work the effect on the depletion of the typical NPP Krško fuel assembly in infinite geometry is investigated. In the second part, analysis of all 30 completed NPP Krško operating cycles is performed. Performed analysis has indicated differences of a few 100 pcm in multiplication factor for a fresh fuel due to differences in 235U and 238U cross sections. For a burned fuel assemblies, differences are even higher due to different rate of fission products formation, 235U burnout, and Pu production. Observed differences in libraries resulted in differences of several tens of ppm in critical Boron concentration on the core level. Differences in control rods worth and Boron coefficients were inside 1%. Some differences in isothermal temperature coefficient were observed, however, they only marginally affect core power defect going from zero to full power.

STUDY ON CRITICALITY AND NEUTRONIC SAFETY PARAMETERS OF NUSCALE FUEL ASSEMBLY

STUDY ON CRITICALITY AND NEUTRONIC SAFETY PARAMETERS OF NUSCALE FUEL ASSEMBLY. NuScale, a typical Pressurized Water Reactor (PWR) Small Modular Reactors (SMRs), offers a new opportunity for the future of nuclear industry. With 160 MW thermal power, NuScale has several advantages such as flexibility due to its modularity in construction. This work is focused on the study of criticality and neutronic safety parameters of NuScale fuel assembly using MCNP6 code and ENDF/B-VII library. The calculation results shows that criticality of fuel assembly type D is the highest among other assembly types because it has a fuel pin with pure UO2 without Gd2O3 concentration. The Doppler temperature coefficient (DTC) of fuel assembly type C is the most negative among other assemblies due to Doppler broadening effect on resonance region of capture cross section of 238U which is the highest concentration. The moderator temperature coefficient (MTC) of fuel assembly type D is the most negative among the other assembly types. The effective delayed neutron fraction (βeff) does not reflect a consistent trend among fuel assembly types. Fuel assembly type D shows the highest prompt neutron lifetime (ℓ) while the highest neutron generation time (Ʌ) is shown in assembly type C. It can be concluded that this study provides adequate results that can be used as a first step to carry out the neutronic computation and analysis of the NuScale full core.Keywords: Criticality, safety parameters, NuScale fuel assembly, MCNP6, ENDF/B-VII.

Monte Carlo simulations and n-p differential scattering data measured with Proton Recoil Telescopes

The neutron-induced fission cross section of 235U, a standard at thermal energy and between 0.15 MeV and 200 MeV, plays a crucial role in nuclear technology applications. The long-standing need of improving cross section data above 20 MeV and the lack of experimental data above 200 MeV motivated a new experimental campaign at the n_TOF facility at CERN. The measurement has been performed in 2018 at the experimental area 1 (EAR1), located at 185 m from the neutron-producing target (the experiment is presented by A. Manna et al. in a contribution to this conference). The 235U(n,f) cross section from 20 MeV up to about 1 GeV has been measured relative to the 1H(n,n)1H reaction, which is considered the primary reference in this energy region. The neutron flux impinging on the 235U sample (a key quantity for determining the fission events) has been obtained by detecting recoil protons originating from n-p scattering in a C2H4 sample. Two Proton Recoil Telescopes (PRT), consisting of several layers of solid-state detectors and fast plastic scintillators, have been located at proton scattering angles of 25.07° and 20.32°, out of the neutron beam. The PRTs exploit the ΔE-E technique for particle identification, a basic requirement for the rejection of charged particles from neutron-induced reactions in carbon. Extensive Monte Carlo simulations were performed to characterize proton transport through the different slabs of silicon and scintillation detectors, to optimize the experimental set-up and to deduce the efficiency of the whole PRT detector. In this work we compare measured data collected with the PRTs with a full Monte Carlo simulation based on the Geant-4 toolkit.

Investigation of the chance structure of the fission cross section of 238U by neutrons in energy range from 14 to 18 Mev using the delayed neutrons

In the present work, delayed neutrons are used for the investigation of the chance structure of the fission cross section. For this, time parameters of delayed neutron precursors emitted from the fission of 238U by neutrons with the energies from 14 to 18 MeV have been measured. The T(d,n) reaction was used as a neutron source. The experimental data have been processed to account the effect of degradation of detector counting rate in high intensity neutron flux and the effect of concomitant neutron source D(d,n) caused by implanted deuterium ions in neutron target. The data obtained based on the time parameters of delayed neutron precursors have been used in the analysis of the chance structure of the fission cross section of the 238U.

Absolute cross section measurements of 238U(n,f) and 237Np(n,f) in the neutron energy range 1-2.4 MeV

New standard (n,f) cross sections other than 235U are important to study the relevant cross sections for Generation-IV power plants. A specific need for such standards is for performing new experiments with quasimonoenergetic neutron beams, such as those produced by Van de Graaf accelerators. Neutrons down-scattered to low energies in the experimental environment, so called room-return, become relevant for this type of measurements. Hence, a standard (n,f) cross section with a fission threshold is of great interest, in order to suppress the contribution from room-return background. For this reason we have performed two experiments at the VDG of the National Physical Laboratory to measure absolutely the (n,f) cross sections of 235U, 238U and 237Np in the fast neutron energy region. Our preliminary results are in agreement with the most up-to-date evaluations.

Thermal neutron scattering cross sections of 238U and 235U in the γ phase

The development of metallic, low-enrichment uranium fuels requires accurate prediction of their neutron transport properties and reactivity parameters, which in turn require thermal neutron scattering data. Accurate prediction of thermal neutron scattering data, including thermal cross sections, requires knowledge of the phonon scattering properties of the medium, but such matrix binding effects in next-generation fuels such as U–Mo, U–Zr, and U–Si are typically neglected because these effects are often difficult to measure or calculate. Using molecular dynamics simulations with previously published interatomic potentials, we calculate the phonon dispersion relations and phonon densities of states for 235U and 238U in the α and γ phases. The performance of these potentials was evaluated using published ab initio simulation data and inelastic neutron scattering data. The phonon densities of states obtained by each potential were then utilized to calculate the thermal neutron scattering cross sections of 235U and 238U at 1113 K using the NJOY program. The resulting thermal neutron scattering cross sections are assessed by comparison to data obtained from available experimental densities of states. The cross sections generated show how the addition of binding effects decreases the cross section by up to a factor of six over the free-atom model. A definite effect on reactivity is also demonstrated by the use of these thermal libraries on a simple core model. As a consequence, the cross sections generated in this work provide a better description of the true cross section than the free-atom data currently available. We also discuss the sensitivity of the thermal scattering cross sections to the phonon density of states.

Measurement of the cross sections for 238U(n, γ)239U reaction in the energy range of 14.1–14.8 MeV using neutron activation method

The neutron capture reaction cross sections of 238U were measured at the neutron energies of 14.1, 14.5 and 14.8 MeV using the activation method and off-line γ-ray spectrometric technique. The induced gamma-ray activities were measured using a low background gamma ray spectrometer equipped with a high resolution HPGe detector. The excitation functions of the 238U(n,γ)239U reaction were also calculated theoretically using the TALYS-1.8 program and are in general agreement with experimental data. The experimentally determined cross sections were compared with the data in the literature, and the evaluated data of ENDF/B-VIII.0, JENDL-4.0, and CENDL-3.1, as well as calculated results from TALYS-1.8.

IAEA CIELO Evaluation of Neutron-induced Reactions on 235U and 238U Targets

Evaluations of nuclear reaction data for the major uranium isotopes 238U and 235U were performed within the scope of the CIELO Project on the initiative of the OECD/NEA Data Bank under Working Party on Evaluation Co-operation (WPEC) Subgroup 40 coordinated by the IAEA Nuclear Data Section. Both the mean values and covariances are evaluated from 10−5 eV up to 30 MeV. The resonance parameters of 238U and 235U were re-evaluated with the addition of newly available data to the existing experimental database. The evaluations in the fast neutron range are based on nuclear model calculations with the code EMPIRE–3.2 Malta above the resonance range up to 30 MeV. 235U(n,f), 238U(n,f), and 238U(n,γ) cross sections and 235U(nth,f) prompt fission neutron spectrum (PFNS) were evaluated within the Neutron Standards project and are representative of the experimental state-of-the-art measurements. The Standards cross sections were matched in model calculations as closely as possible to guarantee a good predictive power for cross sections of competing neutron scattering channels. 235U(n,γ) cross section includes fluctuations observed in recent experiments. 235U(n,f) PFNS for incident neutron energies from 500 keV to 20 MeV were measured at Los Alamos Chi-Nu facility and re-evaluated using all available experimental data. While respecting the measured differential data, several compensating errors in previous evaluations were identified and removed so that the performance in integral benchmarks was restored or improved. Covariance matrices for 235U and 238U cross sections, angular distributions, spectra and neutron multiplicities were evaluated using the GANDR system that combines experimental data with model uncertainties. Unrecognized systematic uncertainties were considered in the uncertainty quantification for fission and capture cross sections above the thermal range, and for neutron multiplicities. Evaluated files were extensively benchmarked to ensure good performance in reactor calculations and fusion-related systems. New comprehensive evaluations show excellent agreement with available differential data and integral performance better than current evaluated data libraries, and represent a step forward in a quest for better nuclear data for applications.

<em>Ab initio</em> calculation of the thermal neutron scattering cross sections of uranium mononitride

Nuclear design and neutronic analysis of thermal neutron reactor need high reliable thermal neutron cross sections. Uranium mononitride (UN) is a candidate fuel material for advanced power reactor with its better thermodynamics and accident tolerance. However, in thermal neutron region, reliable thermal neutron scattering cross sections are lacked for UN, which is disadvantageous to reactor physics simulations. The scattering law of the UN fuel material may impact the thermal neutron spectrum and criticality of the reactor systems. Neutron cross sections in thermal range are correlated with energy, temperature, physical and chemical properties of the scattering medium, reflecting the phonon spectra of material itself. In this paper, based on the ab initio method of quantum mechanics, phonon density of states in UN are calculated by VASP/PHONON code, and used for integral to obtain UN heat capacity at a constant volume. Adopting this new phonon density of states, NJOY/LEAPR code is used to generate S (, ) data by thermal neutron scattering theory and NJOY/THERMR utilizes these data to produce thermal scattering matrix in order to investigate thermal kernel effect of UN. Subsequently, thermal neutron scattering cross sections of UN are generated with NJOY code system. Comparison with uranium dioxide (UO2) in the traditional PWR is done. Results indicate that optimized lattice parameter are in good agreement with the database; the optical modes are well separated from the acoustic modes compared with UO2; heat capacity at a constant volume is consistent with experimental value; the inelastic and elastic cross sections of 238U in UN are lower than those of 238U in UO2. N in UN only deals with incoherent part in elastic cross sections. As the temperature increases, elastic cross sections of UN decrease while inelastic ones increase, and cross sections approach to free atom cross section at high energies. Considering the limitations of 14N, the scattering law and inelastic scattering cross sections are also under investigation using 15N in UN compound. This paper's conclusion fulfill the vacancy of thermal neutron scattering cross sections of UN, which laid a foundation for systematic study on the neutronics properties of UN fuel in the light water reactors as well as for the design of new neutron moderators and neutron filer.

Use of integral data assimilation and differential measurements as a contribution to improve 235U and 238U cross sections evaluations in the fast and epithermal energy range

Critical mass calculations of various HEU-fueled fast reactors result in large discrepancies in C/E values, depending on the nuclear data library used and the configuration modeled. Thus, it seems relevant to use integral experiments to try to reassess cross sections that might be responsible for such a dispersion in critical mass results. This work makes use of the Generalized Least Square method to solve Bayes equation, as implemented in the CONRAD code. Experimental database used includes ICSBEP Uranium based critical experiments and benefits from recent re-analyses of MASURCA and FCA-IX criticality experiments (with Monte-Carlo calculations) and of PROFIL irradiation experiments. These last ones provide very specific information on 235U and 238U capture cross sections. Due to high experimental uncertainties associated to fission spectra, we chose to consider either fitting these data or set them to JEFF-3.1.1 evaluations. The work focused on JEFF-3.1.1 235U and 238U evaluations and results presented in this paper for 235U capture and 238U capture, and inelastic cross sections are compared to recent differential experiment or recent evaluations. Our integral experiment assimilation work notably suggests a 30% decrease for 235U capture around 1–2.25 keV, a 10% increase in the unresolved resonance range when using JEFF-3.1.1 as “a priori” data. These results are in agreement with recent microscopic measurements from Danon et al. [Nucl. Sci. Eng. 187, 291 (2017)] and Jandel et al. [Phys. Rev. Lett. 109, 202506 (2012)]. For 238U cross sections, results are highly dependent on fission spectra.

Simultaneous Measurement of 235U Fission and Capture Cross Sections From 0.01 eV to 3 keV Using a Gamma Multiplicity Detector

The neutron microscopic capture cross section for 235U is a critical parameter for the design and operation of nuclear reactors. The evaluated nuclear data libraries of ENDF/B-VII.1 and JENDL-4.0 have nearly identical values for the neutron capture cross section for neutron energies below 0.5 keV. In the most recent release of the JENDL library the onset of the unresolved resonance region was changed from 2.25 keV to 0.5 keV. In the energy region from 1.5 keV to 2.25 keV the average neutron capture cross section from ENDF/B-VII.1 is about 10% higher than that from JENDL-4.0. In an attempt to address the discrepancies between the libraries, a measurement of the neutron capture cross section of 235U was conducted at the Gaerttner LINAC Center located at Rensselaer Polytechnic Institute. This measurement used a 16-segment -multiplicity NaI(Tl) detector to detect the prompt gammas emitted from neutron interactions with a highly enriched 235U sample. Using the time-of-flight method, detected events were recorded and grouped based on the total gamma energy per interaction and observed multiplicity. A method was developed to separate fission from capture based on total energy deposition and gamma multiplicity. Application of this method in the thermal and resonance region below 0.5 keV for both the fission and capture produced cross sections that are in good agreement with both ENDF/B-VII.1 and JENDL-4.0 evaluations. The measurements support a lower 235U neutron capture cross section in the energy range 0.5 to 2.25 keV, which is closer to JENDL 4.0.

Impact of 235U Resonance Parameter Evaluation in the Reactivity Prediction

Resonance parameter evaluation of the 235U cross sections using the Reich-Moore formalism was done with the computer code SAMMY from 0 to 2.25 keV to address issues with capture cross-section and standard fission cross-section values. The evaluation includes recent capture and fission cross-section measurements as well as high-resolution data used in previous 235U evaluation. Moreover the new 235U resonance parameter evaluation has been used in the calculation of a new benchmark experiment performed at the IPEN/MB-01 research reactor. The experiment, named the inversion point of the isothermal reactivity coefficient, is used to test temperature effects at low temperature. The results demonstrate that the new 235U evaluation has greatly improved the prediction of reactivity temperature coefficient in contrast to previous evaluations. This paper is outlined in two parts, namely the first part deals with the description of the 235U resonance analysis and evaluation up to 2.25 keV, and the second part presents the results of the isothermal reactivity coefficient calculations performed on the IPEN/MB-01 reactor.

Characterization of the Medley setup for measurements of neutron-induced fission cross sections at the GANIL-NFS facility

Neutron-induced fission cross sections of 235U and 238U are widely used as standards for monitoring of neutron beams and fields. An absolute measurement of these cross sections at an absolute scale ...

235U Resonance Structure Effect on Computed Characteristics of Critical Assemblies in Large Physical Testing Facility with Softened Neutron Spectrum

An unusual change of the contributed reactivity of uranium samples with 90% enrichment depending on the size was observed during experiments on the critical assembly of the large physical testing facility with a softened neutron spectrum: the reactivity became negative upon insertion of thin samples. The agreement between the calculations and experimental data in the description of this unusual change of reactivity was obtained by making changes in the radiative capture cross section of 235U and the factors of resonance self-screening in the group approximation and by taking account of the possible missed narrow resonances of 235U in the point representation description of the experiment.

The Use of Nuclear Data as Nuisance Parameters in the Integral Data Assimilation of the PROFIL Experiments

The PROFIL and PROFIL-2 experiments were carried out in the fast reactor PHENIX. They were designed to provide integral information on neutron cross sections [(n,γ), (n,2n), and (n,f)] of several fission products and actinides. Previous interpretations report integral results with unrealistic small uncertainties that only take into account the statistical contribution. This work presents an uncertainty propagation technique able to include systematic uncertainties due to neutron fluence scaling. Such a technique consists of marginalizing analytically the uncertainties of the nuclear data (nuisance parameters) involved in the fluence scaling procedure. For the capture cross sections of 235U, 238U, and 239Pu, the interpretation of the PROFIL and PROFIL-2 experiments with the international library JEFF-3.1.1 provides excellent <C/E> results equal to 1.000, 1.019, and 0.982, respectively, with a relative uncertainty close to 1.5% (1σ).

Study of on-the-fly Doppler broadening in JMCT program

In this paper, we introduce the on-the-fly (OTF) Doppler method which is used to calculate the temperature-dependent cross section. After substituting Adler-Adler multilevel representation into Doppler formulation, the theoretical formulation of temperature dependent cross section is obtained. This theoretical formulation can be approximated by a Taylor series expansion and asymptotic series expansion, which is the base of OTF method. The OTF method can be used to calculate the cross section of any nuclide at any temperature in a range of 300-3000 K based on cross section library of 300 K. For the OTF method, firstly, a series of temperature dependent cross section libraries is produced by NJOY. Secondly, a uniform energy grid is evaluated by the temperature dependent cross section libraries. Thirdly, a polynomial is used to fit the temperature dependent cross section on each energy grid. The coefficient of the polynomial is obtained by single value decomposition algorithm. Finally, the coefficients of the polynomial in all energy grids and the energy grids themselves are written in a text file. To test the cross section polynomial produced by OTF method, we compare the total and absorption cross sections of 238U and 235U calculated by the polynomial with those produced by NJOY. The errors of these cross sections obtained by these two programs are presented in the paper. The text file produced by OTF method can be read by the Monte Carlo code JMCT, which is a coupled neutron/photon transport code developed by IAPCM. After providing the temperature and energy of the particle, the temperature dependent cross sections in two adjacent energy grids are calculated by the polynomial respectively. The cross section of target energy is obtained by linear interpolation. Two benchmarks including a pin-cell model and an assembly model are used to verify the applications of OTF method in JMCT. The results are presented in the paper.