235U Cross Section Research Articles

Fission Fragment Angular Distribution measurements of235U and238U at CERN n_TOF facility

Neutron-induced fission cross sections of 238U and 235U are used as standards in the fast neutron region up to 200 MeV. A high accuracy of the standards is relevant to experimentally determine other neutron reaction cross sections. Therefore, the detection effciency should be corrected by using the angular distribution of the fission fragments (FFAD), which are barely known above 20 MeV. In addition, the angular distribution of the fragments produced in the fission of highly excited and deformed nuclei is an important observable to investigate the nuclear fission process.In order to measure the FFAD of neutron-induced reactions, a fission detection setup based on parallel-plate avalanche counters (PPACs) has been developed and successfully used at the CERN-n_TOF facility. In this work, we present the preliminary results on the analysis of new 235 U(n,f) and 238 U(n,f) data in the extended energy range up to 200 MeV compared to the existing experimental data.

On Some Fundamental Peculiarities of the Traveling Wave Reactor

On the basis of the condition for nuclear burning wave existence in the neutron-multiplying media (U-Pu and Th-U cycles) we show the possibility of surmounting the so-called dpa-parameter problem and suggest an algorithm of the optimal nuclear burning wave mode adjustment, which is supposed to yield the wave parameters (fluence/neutron flux, width and speed of nuclear burning wave) that satisfy the dpa-condition associated with the tolerable level of the reactor materials radioactive stability, in particular that of the cladding materials. It is shown for the first time that the capture and fission cross sections of238U and239Pu increase with temperature within 1000–3000 K range, which under certain conditions may lead to a global loss of the nuclear burning wave stability. Some variants of the possible stability loss due to the so-called blow-up modes (anomalous nuclear fuel temperature and neutron flow evolution) are discussed and are found to possibly become a reason for a trivial violation of the traveling wave reactor internal safety.

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.

Photo-neutron cross-section measurement in the 8 and 10 MeV bremsstrahlung induced reaction of 238U

The 238U(γ, n)237U reaction cross-section at the end point bremsstrahlung energies of 8 and 10 MeV was measured by using an activation technique. Induced gamma ray activities were measured by high resolution gamma-ray spectrometer with high-purity germanium detector. The photo neutron cross section on 238U is also calculated theoretically using TALYS 1.2 computer code. The experimentally obtained reaction cross sections were compared with the flux weighted average values from the literature data based on mono-energetic photon as well as the value from TALYS. It was found that the cross section of 238U(γ, n)237U reaction increases with increase of bremsstrahlung energy and were closer to the flux-weighted experimental data from literature and the values from TALYS based on mono-energetic photons.

Measurement of the neutron-induced fission cross-section of 241Am at the time-of-flight facility n_TOF

The neutron-induced fission cross-section of 241Am has been measured relative to the standard fission cross-section of 235U between 0.5 and 20MeV. The experiment was performed at the CERN n_TOF facility. Fission fragments were detected by a fast ionization chamber by discriminating against the \(\alpha\)-particles from the high radioactivity of the samples. The high instantaneous neutron flux and the low background of the n_TOF facility enabled us to obtain uncertainties of \( \approx\) 5%. With the present results it was possible to resolve discrepancies between previous data sets and to confirm current evaluations, thus providing important information for design studies of future reactors with improved fuel burn-up.

MEASUREMENT OF PHOTOFISSION CROSS-SECTION OF238UUSING MICROTRON FACILITY

Photofission cross-section of238U was measured using bremsstrahlung radiation energy 7.4–9.0 MeV with energy step of 0.4 MeV by employing Lexan polycarbonate film as solid state nuclear track detector (SSNTD). The photon intensity from the Microtron accelerator at a distance of 15 cm from the bremsstrahlung converter (tantalum target) facility was estimated to be 1010photons/sec using the code EGS-4. In this paper, details of the fission fragment angular distribution measurements of238U target using Lexan polycarbonate have been discussed. The photofission cross-section was calculated using the angular distribution of fission fragments and the results were compared with those obtained using the code EMPIRE-II and various barrier parameters of the RIPL-1, RIPL-2 libraries and with the new analytical fission barrier formula based on the Hugenholtz–Van Hove theorem. The present experimental measurements were in good agreement with the results obtained from the Empire-II code predictions for potential parameter taken from RIPL-1 and a newly developed analytical fission barrier formula.

Accurate determination of cross-sections for 238U (n, 2n) 237U induced by neutrons around 14 MeV

The cross-sections of 238U (n, 2n) 237U induced by neutrons with energies around 14MeV were determined by measuring the emission rate of 208.00keV γ-rays from 237U decay with an accurately calibrated HPGe γ-ray spectrometer at CIAE. In order to correct for the self-attenuation of 208.0keV γ-ray in U sample, a 177Lu-spiking method was introduced. As a result, a total relative standard uncertainty of 2.5% was achieved for the measured cross-section values.

Influence of the uncertainty of the microscopic cross sections of 238U and plutonium isotopes on the neutron multiplication coefficient in VVÉR

The influence of the uncertainties in the microscopic cross sections of 238U and 239–241Pu on the neutron multiplication coefficient in VVER-1000 is studied. Data on the uncertainty of the cross sections were obtained by analyzing the ENDF/B files of the JENDL 3.3 system using the ERRORJ program. The characteristics of the reactor are calculated using the TRIFON 2.1 and SHERHAN programs. An approach associated with the calculation of the coefficients of sensitivity of the fuel assemblies and the reactor to standard samples in a small-group approximation is used.

Evaluation of the 238U Neutron Cross Section in the Unresolved Resonance Range

This paper presents a new analysis of the 238U cross sections in the unresolved resonance range, from 20 to 150 keV. Statistical analysis of the resonance parameters in the resolved resonance range with random-matrix theory provides accurate experimental values of strength function, average radiative width and average level spacing for s- and p-wave resonances. Above 20 keV, the simultaneous fit of selected experimental data (average transmission and capture) is performed with a statistical model of nuclear reactions as implemented in the SAMMY code.Compared to previous evaluations, such as those described by Fröhner or by Maslov et al., this work benefits from the accurate transmission data measured by Harvey et al. at Oak Ridge Electron Linear Accelerator, which have never been studied before. This new evaluation was written into the current ENDF format for use in practical applications. This work stresses the need for an improved ENDF format to store average resonance parameters and cross sections in the unresolved resonance range.

Isotopic and Energy Groupwise Dependence of Fuel Temperature Coefficient of Reactivity in Natural Uranium-Fueled Pressurized Heavy Water Reactors

We have studied the individual effect of the temperature dependence of the multigroup cross sections of 238U, 235U, 239Pu, 240Pu, and 16O on the calculated fuel temperature coefficient (FTC) by performing detailed sensitivity studies. The thermal contribution and the Doppler contribution of the FTC have been estimated for the above isotopes for the 19-element UO2-fueled heavy water lattice of the pressurized heavy water reactor (PHWR). The groupwise breakdown of the FTC due to 238U resonances has also been obtained. The FTC of Canada deuterium uranium reactor (CANDU)-type pressurized heavy water moderated lattices using UO2 fuel becomes less negative with burnup and changes sign at high burnups. Our studies clearly demonstrate that the positive component of the FTC in natural UO2-fueled PHWRs arises primarily because of the temperature dependence of scattering cross sections of 16O in agreement with the earlier findings of Stammler. In this paper, we have calculated the reactivity due to the change in fuel temperature, and all our discussions are based on this fuel temperature reactivity rather than the FTC itself.

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.

Conditions for the manifestation of transient effects in fission

The conditions for the manifestation of transient effects in fission are carefully examined by analysing experimental data where fission is induced by peripheral heavy-ion collisions at relativistic energies. Experimental total nuclear fission cross sections of 238U at 1 A GeV on gold and uranium targets are compared with the predictions of a nuclear-reaction code, where transient effects in fission are modeled using different approximations to the numerical time-dependent fission-decay width: a new analytical description based on the solution of the Fokker–Planck equation and two widely used but less realistic descriptions, a step function and an exponential-like function. The experimental data are only reproduced when dissipation is considered. The influence of transient effects on the fission process, as well as the deduced value of the dissipation strength β, depends strongly on the approximation applied for the time-dependent fission-decay width. In particular, a meticulous analysis sheds severe doubts on the use of the exponential-like in-growth function. Finally, we investigate which should be the characteristics of experimental observables to be most sensitive to transient effects in fission. The pertinence of observables related to the excitation energy at saddle is discussed.

Errors in evaluated neutron cross sections

The underestimation of the errors in the recommended neutron cross sections calculated on the basis of an analysis of measurement results using strict statistical methods is discussed. The basic reasons for the underestimation of the errors in the cross sections are described. A statistical model is proposed — a constant bias model — for taking into account effectively the component of the measurement error that is unknown to the experimentor. The results of a statistical analysis of the measurements of the ratio of the fission cross sections of 238U and 235U on the basis of the constant bias model are presented. It is noted that the experimental errors are much higher (on the average by a factor of 2) than the declared errors. The consequences of using the generally accepted method for correcting the errors in the evaluated cross sections are examined for an exactly solvable model example. It is shown that such a procedure does reconstruct the real errors in the evaluated cross sections.

Uranium symmetric/asymmetric neutron-induced fission up to 200 MeV

The symmetric SL-mode and asymmetric lumped (S1 + S2)-mode fission cross-sections of 235U(n,F) and 233U(n,F) reactions are calculated up to $E_{\rm {n}} = 200$ MeV within a statistical model. For each U nuclide, emerging in $(\rm {n},x\rm {nf})$ reactions a separate triaxial outer fission barrier is assumed for the SL-mode. To reproduce the measured branching ratio of symmetric and asymmetric fission events for the 238U(n,F) reaction, more fissions coming from neutron-deficient nuclei were assumed. The damping of the triaxial collective modes contribution to the level density at the SL-mode outer saddle was essential for the branching ratio description. These assumptions allow to reproduce observed fission cross-sections of 235U(n,F) and 233U(n,F) reactions. The calculated branching ratio sensitivity to the target nuclide fissility is investigated.

Global coupled-channel optical potential for nucleon–actinide interaction from 1 keV to 200 MeV

A coupled-channel method was used for the estimation of nucleon optical potential parameters for the actinide region based on the axial rigid-rotor model built on a saturated coupling scheme. A new global form of optical potential was proposed to incorporate energy dependences guided by physical principles. Best-fit potential parameters were searched for to reproduce experimental nucleon scattering angular distributions, low-energy observables (strength functions and scattering radii) and neutron total cross sections of 238U and 232Th. It was found that the present form of the optical potential could reproduce both the neutron- and proton-data for a wide energy range, namely from 1 keV to 200 MeV. Available experimental data for other actinides were also described to a good accuracy with the same set of parameters only by adjusting the deformation parameters for each nucleus.

Measurement of cross sections of threshold detectors with spectrum average technique

Summary Cross sections of the reactions 103Rh(n, n′)103mRh, 115In(n, n′)115mIn, 232Th(n, f), 47Ti(n, p)47Sc, 64Zn(n, p)64Cu, 58Ni(n, p)58Co, 54Fe(n, p)54Mn, 46Ti(n, p)46Sc, 27Al(n, p)27Mg, 56Fe(n, p)56Mn, 24Mg(n, p)24Na, 59Co(n, α)56Mn, 27Al(n, α)24Na and 48Ti(n, p)48Sc were measured with average neutron energies above effective threshold by using the activation method through usage of spectrum average technique in an irradiation system where there are three equivalent Am/Be sources, each of which has 592 GBq activity. The cross sections were determined with reference to the fast neutron fission cross section of 238U. The measured values and published values are generally in agreement.