Capture Cross Section Of 238U 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.

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.

Nuclear Data Uncertainty Quantification and Propagation for Safety Analysis of Lead-Cooled Fast Reactors

In this study, the Best Estimate Plus Uncertainty (BEPU) approach is developed for the systematic quantification and propagation of uncertainties in the modelling and simulation of lead-cooled fast reactors (LFRs) and applied to the demonstration LFR (DLFR) initially investigated by Westinghouse. The impact of nuclear data uncertainties based on ENDF/B-VII.0 covariances is quantified on lattice level using the generalized perturbation theory implemented with the Monte Carlo code Serpent and the deterministic code PERSENT of the Argonne Reactor Computational (ARC) suite. The quantities of interest are the main eigenvalue and selected reactivity coefficients such as Doppler, radial expansion, and fuel/clad/coolant density coefficients. These uncertainties are then propagated through safety analysis, carried out using the MiniSAS code, following the stochastic sampling approach in DAKOTA. An unprotected transient overpower (UTOP) scenario is considered to assess the effect of input uncertainties on safety parameters such as peak fuel and clad temperatures. It is found that in steady state, the multiplication factor shows the most sensitivity to perturbations in 235U fission, 235U ν, and 238U capture cross sections. The uncertainties of 239Pu and 238U capture cross sections become more significant as the fuel is irradiated. The covariance of various reactivity feedback coefficients is constructed by tracing back to common uncertainty contributors (i.e., nuclide-reaction pairs), including 238U inelastic, 238U capture, and 239Pu capture cross sections. It is also observed that nuclear data uncertainty propagates to uncertainty on peak clad and fuel temperatures of 28.5 K and 70.0 K, respectively. Such uncertainties do not impose per se threat to the integrity of the fuel rod; however, they sum to other sources of uncertainties in verifying the compliance of the assumed safety margins, suggesting the developed BEPU method necessary to provide one of the required insights on the impact of uncertainties on core safety characteristics.

Comparison of ENDF/B-VIII.0 and ENDF/B-VII.1 in criticality, depletion benchmark, and uncertainty analyses by McCARD

The McCARD criticality, depletion, and S/U analyses with the newly published evaluated nuclear data file (ENDF) library ENDF/B-VIII.0 are conducted and their results are compared with those using the previous version of ENDF library, ENDF/B-VII.1. First of all, to verify and validate the newly generated ENDF/B-VIII.0 continuous energy cross section library in the ACE format, 60 cases from International Criticality Safety Benchmark Problems (ICSBEP) and VENUS-1 criticality benchmark analyses are performed using the McCARD code. Overall, the prediction performance of criticality is retained except for the high enriched uranium (HEU) fast and plutonium-based (PU) thermal benchmarks. The large difference in keff for the HEU fast and PU thermal benchmarks comes from the increase of the 238U capture cross sections in the fast energy region and the decrease of 239Pu fission cross section in the thermal energy region, respectively. In the results for the VERA depletion benchmarks, the large differences at the end of the burnup can be attributed to the difference of the one-group 235U absorption cross section. It leads to the increases of the difference in kinf between ENDF/B-VIII.0 and ENDF/B-VII.1 as the burnup proceeds. In the ENDF/B-VIII.0 evaluated nuclear data library, there have been considerable changes in the covariance data. The changes have a strong correlation with the decreases in the total uncertainty of multiplication factor for Godiva and the TMI-1 pin problem.

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.

High precision measurement of the radiative capture cross section of238U at the n_TOF CERN facility

The importance of improving the accuracy on the capture cross-section of 238U has been addressed by the Nuclear Energy Agency, since its uncertainty significantly affects the uncertainties of key design parameters for both fast and thermal nuclear reactors. Within the 7th framework programme ANDES of the European Commission three different measurements have been carried out with the aim of providing the 238U(n, gamma) cross-section with an accuracy which varies from 1 to 5%, depending on the energy range. Hereby the final results of the measurement performed at the n TOF CERN facility in a wide energy range from 1 eV to 700 keV will be presented.

Measurement of the 238U(n,γ)239U and 238U(n,2n)237U Reaction Cross Sections Using a Neutron Activation Technique at Neutron Energies of 8.04 and 11.90 MeV

The (n,γ) and (n,2n) capture cross sections of 238U have been measured at neutron energies of 8.04 ± 0.30 and 11.90 ± 0.35 MeV from the 7Li(p,n) reaction using an activation and off-line gamma-ray spectrometric technique. The experimentally determined 238U(n,γ) and 238U(n,2n) reaction cross sections were compared with the evaluated data of ENDF/B-VII.0, JENDL-4.0, JEFF-3.1/A, and CENDL-3.1. The experimental values were found to be in agreement with the evaluated value based on ENDF/B-VII.0, JENDL-4.0, and JEFF-3.1/A but not with CENDL-3.1. The present measurement has been compared with literature data in a wide range of neutron energies. The 238U(n,γ)239U and 238U(n,2n)237U reaction cross sections were also calculated theoretically using the TALYS 1.4 computer code and compared with the experimental data.

High-precision Measurement of the 238U(n,γ) Cross Section with the Total Absorption Calorimeter (TAC) at n_TOF, CERN

The neutron capture cross section of 238U is fundamental to the design and operation of current reactors and future fast nuclear reactors, and thus must be measured to a high level of accuracy. An experiment has been performed at the CERN n_TOF facility using a 4π Total Absorption Calorimeter (TAC) to measure the capture cross section in the resolved resonance region between 1 eV and 25 keV. A preliminary analysis of the TAC data is presented with particular emphasis to the experimental background in this energy region of interest.

ANDES Measurements for Advanced Reactor Systems

A significant number of new measurements was undertaken by the ANDES “Measurements for advanced reactor systems” initiative. These new measurements include neutron inelastic scattering from 23Na, Mo, Zr, and 238U, neutron capture cross sections of 238U, 241Am, neutron induced fission cross sections of 240Pu, 242Pu, 241Am, 243Am and 245Cm, and measurements that explore the limits of the surrogate technique. The latter study the feasibility of inferring neutron capture cross sections for Cm isotopes, the neutron-induced fission cross section of 238Pu and fission yields and fission probabilities through full Z and A identification in inverse kinematics for isotopes of Pu, Am, Cm and Cf. Finally, four isotopes are studied which are important to improve predictions for delayed neutron precursors and decay heat by total absorption gamma-ray spectrometry (88Br, 94Rb, 95Rb, 137I). The measurements which are performed at state-of-the-art European facilities have the ambition to achieve the lowest possible uncertainty, and to come as close as is reasonably achievable to the target uncertainties established by sensitivity studies. An overview is presented of the activities and achievements, leaving detailed expositions to the various parties contributing to the conference.

238U(n,γ) reaction cross section measurement with C6D6detectors at the n_TOF CERN facility.

The radiative capture cross section of 238U is very important for the developing of new reactor technologies and the safety of existing ones. Here the preliminary results of the 238U(n,γ) cross section measurement performed at n_TOF with C6D6 scintillation detectors are presented, paying particular attention to data reduction and background subtraction.

Neutron cross section sensitivity analysis of UO2 and MOX core physics experiments on light water reactors

The perturbation theory based on the transport calculation has been applied to study sensitivity of neutron multiplication factors (keff's) to neutron cross sections used for the reactivity analysis of UO2 and MOX core physics experiments on light water reactors. The studied cross sections were neutron capture, fission and elastic scattering cross sections, and a number of fission neutrons, ν. The obtained sensitivities were multiplied to relative differences in the cross sections between JENDL-4.0 and JENDL-3.3 in order to estimate the reactivity effects. The results show that the increase in keff, 0.3%Δk/kk′, from JENDL-3.3 to JENDL-4.0 for the UO2 core is mainly attributed to the decreases in the capture cross sections of 238U. On the other hand, there are various contributions from the differences in the cross sections of U, Pu, and Am isotopes for the MOX cores. The major contributions to increase in keff are decreases in the capture cross sections of 238U,238Pu, 239Pu, and those to decrease in keff are decreases in ν of 239Pu and increases in the capture cross sections of241Am. They compensate each other, and the difference in keff between JENDL-3.3 and JENDL-4.0 is less than 0.1%Δk/kk′ and relatively small.

Revisiting the 238U Thermal Capture Cross Section and Gamma-Ray Emission Probabilities from 239Np Decay

The precise value of the thermal capture cross section of 238U is uncertain, and evaluated cross sections from various sources differ by more than their assigned uncertainties. A number of the original publications have been reviewed to assess the discrepant data, corrections were made for more recent standard cross sections and other constants, and one new measurement was analyzed. Because of the strong correlations in activation measurements, the gamma-ray emission probabilities from the β– decay of 239Np were also analyzed. As a result of the analysis, a value of 2.683 ± 0.012 b was derived for the thermal capture cross section of 238U. A new evaluation of the gamma-ray emission probabilities from 239Np decay was also undertaken.

｀２３８´Ｕの中性子捕獲断面積非分離共鳴領域における評価

It has been pointed out for a long time that the 238U capture cross sections in the unresolved resonance region evaluated on the basis of available experiments are larger than the ones expected from the reactor physics analysis. A new analysis of capture and transmission data by the shape analysis method results in the correction factors to reduce the values obtained in the early measurement which is one of the major experiments having larger values. Theoretical model fitting to available experiments and recent measurements give similar values. These are the basis of the new version of the major evaluated nuclear data files, ENDF/B-VI, JEF-2 and JENDL-3.

Space, energy and anisotropy effects on 238U effective capture cross sections in the resonance region

Agreement between calculations and measurements within prescribed limits of error is always the test of engineering design analysis. Large and puzzling discrepancies do exist between several measured and calculated important integral reactor parameters. A thorough and exhaustive investigation of the methods used in reactor analysis revealed that in the generation of effective resonance cross sections no anisotropy effects are considered in the resonances. This is true in the integral transport and fundamental-mode codes. The neglect of anisotropy introduces errors at two levels: (1) the effective group cross sections such as σ c , σ f and σ s ; and (2) the diffusion coefficients and P 1 and higher components of the scattering cross sections. The study showed that the inclusion of linear scattering anisotropy increases, in general, the cell effective capture cross section of 238U in both ZPR-6/5 and TRX-3 reactors. The increase was up to 2% in TRX-3 and 0.5% in ZPR-6/5. The effect on the multiplication factor was −0.003% Δk k for ZPR-6/5 and −0.05% Δk k for TRX-3.

Measurements of the 238U(n, γ) Cross Section at Thermal and Fast Neutron Energies

The capture cross section of 238U was measured using the activation technique and 235U(n, f) and 197Au(n, γ) as reference cross sections. Capture events were measured by detection of two prominent ...

Absolute measurement of the radiative capture cross section of238U for 30 keV neutrons

Absolute measurement of the radiative capture cross section of238U for 30 keV neutrons

Measurement of the Uranium-238 Capture Cross Section for Incident Neutron Energies up to 100 keV

AbstractThe neutron capture cross section of 238U was measured for incident neutron energies between 5 eV and 100 keV using a pulsed electron Linac neutron source and the time-of-flight technique. Capture gamma rays were detected by a large liquid scintillator located on a 40-m flight path. The incident neutron flux was monitored by a 10BF3 ionization chamber. The cross section was normalized by the saturated resonance technique.The data have uncertainties which increase from ∼5% at 1 keV to 10% at 100 keV. These data are compared with results from other measurements and with various evaluations.

An Analysis of the Neutron Capture Cross Section of 238U Between 1 keV and 15 MeV

The best available values of the 238U neutron capture cross section in the range of a few keV to 15 MeV are derived by evaluation of the experimental techniques and reference cross sections and, where necessary, revising and rejecting experimental values. A principal objective is to obtain 238U capture cross sections which are consistent with fission cross sections previously selected by the author and which can, therefore, be used for consistent analysis of fast reactors. All available experiments are discussed and a detailed tabulation of the individual experimental data and selected best cross sections is given.

A measurement of the thermal neutron radiative capture cross section of 238U

The 228U ( n, γ) 238U cross section has been measured using the standard thermal neutron flux facility available at the National Physical Laboratory. The 239U activity induced in the reaction was determined after irradiation by absolute counting techniques. Values of 2.69 ± 0·03 b for the 2200 m/sec capture cross section and 23.40 ± 0.05 m for the half life of the 231U ground state were obtained.

Absolute Radiative Capture Cross Section for Fast Neutrons in 238U

The capture cross section of 238U has been measured absolutely at a neutron energy of 30 keV using kinematically collimated neutrons from the 7Li(p, n) 7Be reaction near threshold. Activation techniques were used to determine both the number of capture events and the number of neutrons that occurred during the irradiation. The result of the 238U capture cross section measurement is 479 ± 14 mb at 30 keV. In addition, the shape of the 238 U capture cross section has been measured for neutron energies from 25 to 500 keV using neutrons from the 7Li(p, n)7 Be reaction. The capture reactions in the 238 U target were detected using a large liquid scintillator tank and time-of-flight techniques. The relative neutron flux was measured using a flat response neutron detector. The cross-section shape measurement was normalized to the present absolute measurement at 30 keV. The present measurement has been compared with several measured values, theoretical calculations, and compiled values of the 238U capture cross section as given by other authors.