Fission Cross Section Of 235U Research Articles

EXFOR-based simultaneous evaluation for neutron-induced fission cross section of plutonium-242

ABSTRACT The 242Pu neutron-induced fission cross section was evaluated from 100 keV to 200 MeV. The experimental 242Pu and 235U fission cross sections and their ratios in the EXFOR library were reviewed and analysed by the least-squares method. Additional simultaneous evaluation was performed by including the experimental database of the 233,238U and 239,240,241Pu fission cross sections and their ratios developed for JENDL-5 evaluation. The 242Pu fission cross sections from our evaluation and JENDL-5 evaluation are close to each other below 1 MeV while they systematically differ from each other above 10 MeV. The cross section from our evaluation is systematically lower than the JENDL-4.0 cross section in the prompt fission neutron spectrum peak region (∼5% lower around 1 MeV). The newly evaluated 242Pu fission cross section was verified against the cross section measured in the 252Cf spontaneous fission neutron field and criticalities of small-sized LANL fast systems and demonstrated better performance than the JENDL-4.0 cross section on the same level with the JENDL-5 cross section.

Sensitivity and uncertainty analysis of effective multiplication factor of KRITZ-1 benchmark using ENDF/B-VIII.0 and JEFF3-3 nuclear data libraries

Sensitivity and uncertainty analysis have been performed to estimate the uncertainty in the effective neutron multiplication factor, keff, of the KRITZ-1 benchmark and to identify which nuclear reactions and isotopes of interest could be significantly influencing it. First, the adjoint-weighted perturbation method using the KSEN card of the MCNP6 code is applied to compute the sensitivity vectors for 235U, 238U, 1H, 16O, and 10B isotopes with the JEFF-3.3 and ENDF/B-VIII.0 nuclear data libraries. Then, the sensitivity vectors and covariance matrices were combined to calculate the uncertainty in keff due to uncertainties in nuclear data. The 44 group covariance matrices have been processed with the NJOY21 system.It was found that the uncertainty in keff is mostly due to uncertainties on ν‾p , fission cross section of 235U, the capture cross section of the two isotopes 238U and 1H in the aforementioned data libraries

Progress in measurements of fission cross sections and total cross sections at CSNS Back-n white neutron source

The neutron-induced cross sections are of great significance for the design of nuclear devices and advanced reactors for the nuclear energy production. At CSNS Back-n white neutron source, new measurements of the fission cross sections and total cross sections are performed with two sets of Day-one spectrometers based on the multi-cell fast fission ionization chamber (FIC). The neutron-induced 236,238U fission cross sections relative to 235U from the fission threshold energy to 200 MeV were measured by using the TOF method and the Fast Ionization Chamber Spectrometer for Fission Cross Section Measurement (FIXM) in the single/double bunch mode of Back-n. The experimental uncertainties are analyzed in detail, and the results from the two modes are consistent. The measured 236,238U/235U fission cross section ratios are compared with previous experiments and evaluations. The 236,238U(n,f) cross sections are obtained based on the standard 235U fission cross section. The neutron total cross sections of carbon and aluminum in the energy region from 1 eV to 20 MeV have been measured by using the TOF method and transmission method based on the Neutron Total Cross Section Spectrometer (NTOX) in the double bunch mode. The total cross section results after unfolding are in good agreement with the previous measurements as well as the broadening of the ENDF/B-Ⅷ.0 evaluation with Gaussian function within the experimental uncertainty. The present results provide the experimental data for further measurements, relevant evaluations and the design of nuclear system.

Comments on the Physics of the Frisch-Peierls Memorandum

This technical note offers comments and suggestions regarding four issues involved with the Frisch-Peierls memorandum of 1940: (1) Propagation of transcription errors in subsequent publishings of the memorandum; (2) Data bearing on F&P’s adoption of 10 b as the fission cross section of 235U; (3) The origin of their assertion that the critical radius is about 0.8 times the mean free path for fission if scattering is disregarded; and (4) The origin of a multiplicative factor of 0.2 in their yield formula and the consistency of their calculations.

Fission cross-section measurement for the 238U(n,f)89Rb reaction induced by D-T neutrons

The fission reactions of 238U induced by D-T neutrons were investigated by neutron activation technique and low-background HPGe-γ spectrometer. The half-life was used to identify the possible daughter nuclide. The fission cross-sections of 238U(n,f)89Rb reaction induced at the neutron energy around 14 MeV, i.e., 14.1 ± 0.3, 14.5 ± 0.3 and 14.7 ± 0.3 MeV, were measured precisely. The neutron flux was monitored by accompanying α-particle of the T(d,n)4He reaction and the neutron energies were determined by the cross-section ratio of 90Zr(n,2n)89Zr to 93Nb(n,2n)92mNb reaction. The fission cross-sections of the 238U(n,f)89Rb reaction are 33.2 ± 2.1 mb, 32.2 ± 2.0 mb, and 32.6 ± 2.1 mb, corresponding to the above neutron energy points.

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.

Accurate measurement of the standard 235U(n,f) cross section from thermal to 170 keV neutron energy

An accurate measurement of the 235U(n,f) cross section from thermal to 170 keV of neutron energy has recently been performed at n_TOF facility at CERN using 6Li(n,t)4He and 10B(n,α)7Li as references. This measurement has been carried out in order to investigate a possible overestimation of the 235U fission cross section evaluation provided by most recent libraries between 10 and 30 keV. A custom experimental apparatus based on in-beam silicon detectors has been used, and a Monte Carlo simulation in GEANT4 has been employed to characterize the setup and calculate detectors efficiency. The results evidenced the presence of an overestimation in the interval between 9 and 18 keV and the new data may be used to decrease the uncertainty of 235U(n,f) cross section in the keV region.

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.

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 ...

Determination of the Radiochemical Yield of Fission-Fragment 99Mo In VVR-Ts

The parameters of the neutron flux in the experimental channels of VVR-Ts are determined. It is shown that to evaluate the 99Mo yield it is necessary to take account of, aside from the neutron flux, the dependence of the fission cross section of 235U on the temperature of the neutron gas. The blocking coefficient is calculated for an experimental uranium target. The computed radiochemical yield of 99Mo is compared with the experimentally measured value. It is shown that the average yield equals 0.75. A correlation is found between the neutron flux density, which is directly proportional to the power of the VVR-Ts nuclear reactor, at the target location and the radiochemical yield of 99Mo.

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.

Development of Ionisation Chambers for the Simultaneous Measurement of the Neutron-induced Capture and Fission Cross Section of 233U

A new simultaneous measurement of σ(n,f) and σ(n, γ) will be performed at the neutron time-of-flight facility GELINA in Geel (Belgium). The fission events will be detected by a multi-plate high-efficiency ionisation chamber (IC). An efficient array of C6D6 scintillators will be used for the detection of gamma-rays. The disentanglement between fission and capture gamma-rays can be achieved by using anticoincidence events between the IC and the C6D6 detectors. Given the difference in the fission and capture cross sections, the assignment of a gamma-ray to one or the other reaction type has to be very efficient and reliable. The IC efficiency is not 100 % and a correction has to be applied to take into account the undetected fission events. To keep this correction factor low and reliable, the efficiency parameter of the IC should be high and known with a high degree of accuracy. The IC efficiency towards fission can be defined as a ratio between the number of detected neutrons in coincidence or not with fission fragments. It is therefore a value directly extractable from the experimental data. Results from test experiments of the IC will be presented and discussed, along with IC MCNPX simulations.

High-accuracy determination of the neutron flux at n_TOF

The neutron flux of the n_TOF facility at CERN was measured, after installation of the new spallation target, with four different systems based on three neutron-converting reactions, which represent accepted cross sections standards in different energy regions. A careful comparison and combination of the different measurements allowed us to reach an unprecedented accuracy on the energy dependence of the neutron flux in the very wide range (thermal to 1 GeV) that characterizes the n_TOF neutron beam. This is a pre-requisite for the high accuracy of cross section measurements at n_TOF. An unexpected anomaly in the neutron-induced fission cross section of 235U is observed in the energy region between 10 and 30keV, hinting at a possible overestimation of this important cross section, well above currently assigned uncertainties.

The experimental determination and evaluation of the spectral indices of the IPEN/MB-01 reactor for the IRPhE project

New experimental results for the spectral indices of the IPEN/MB-01 reactor are presented in this work. The experimental approach considers a new technique for the 28ρ case which does not require any sort of calculated correction factors. This aspect gave to the IPEN/MB-01 experiment an excellent quality and free of possible bias due to these calculated correction factors. The uncertainty analysis show, even considering the uncertainties of the geometric and material data of the facility, that the final uncertainties are small enough and well understood for a benchmark problem. The experiment was evaluated and included in the IRPhE handbook. The theory/experiment comparison reveals that there is considerable progress in the 238U nuclear data for application in thermal reactors. The theory/experiment comparison employs the nuclear data libraries ENDF/B-VII.0, ENDF/B-VI.8, JEF3.1, and JENDL3.3 and reveals an excellent agreement for the spectral index 28ρ* independently of the nuclear data library considered; aspect never found before in several other comparisons. The long term overprediction of the 238U neutron epithermal capture appears to be eliminated with the improvements in the recent nuclear data libraries. The experimental performed at the IPEN/MB-01 reactor supports the changes in the 238U nuclear data incorporated in ENDF/B-VII.0 as well as in the other libraries studied in this work. The theory/experiment comparison of 25δ* and (C8/F)ept show that these spectral indices are in general slightly overpredicted, thus suggesting that the thermal fission cross section of 235U might be a little bit underestimated. The overall analysis of the theory/experiment comparisons show the excellent applicability of ENDF/B-VII.0 for thermal reactors fuelled with slightly enriched uranium.

Fukushima Plutonium Effect and Blow-Up Regimes in Neutron-Multiplying Media

It is shown that the capture and fission cross-sections of 238U and 239Pu increase with temperature within 1000 K - 3000 K range, in contrast to those of 235U, that under certain conditions may lead to the so-called blow-up modes, stimulating the anomalous neutron flux and nuclear fuel temperature growth. Some features of the blow-up regimes in neutronmultiplying media are discussed.

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.

Neutron induced capture and fission discrimination using calorimetric shape decomposition

The neutron capture and fission cross-sections of 233U have been measured at the neutron time-of-flight facility n_TOF at CERN in the energy range from 1eV to 1keV using a high performance 4π BaF2 Total Absorption Calorimeter (TAC) as a detection device. In order to separate the contributions of neutron capture and neutron induced fission in the TAC, a methodology called Calorimetric Shape Decomposition (CSD) was developed. The CSD methodology is based on the study of the TAC's energy response for all competing reactions, allowing to discriminate between γ s originating from neutron induced fission and those from neutron capture reactions without the need for fission tagging or any additional detection system. In this article, the concept behind the CSD is explained in detail together with the necessary analysis to obtain the TAC's response to neutron capture and neutron induced fission. The discrimination between capture and fission contributions is shown for several neutron energies. A comparison between the 233U neutron capture and fission yield extraction with ENDF/B-VII v1. library data is also provided.

Uncertainty evaluation of calculated and measured kinetics parameters of Tehran Research Reactor

Effective delayed neutron fraction β eff and neutron generation time Λ are important factors in reactor physics calculation and transient analysis. In the first stage of this research, these kinetics parameters have been calculated for two states of Tehran Research Reactor (TRR), i.e. cold (fuel, clad and coolant temperature 20 °C) and hot (fuel, clad and coolant temperature 65, 49 and 44 °C, respectively) states using MTR_PC computer code. The ratio of ( β eff ) i /( β eff ) core plays an important role in reactivity accident analysis codes. This parameter and its contribution to effective delayed neutron fraction from each nucleus have been calculated in cold and hot reactor states. Uncertainty of effective delayed neutron fraction is evaluated in terms of following four quantities; basic delayed neutron constants, delayed neutron spectra, energy dependence of delayed neutron yield ( ν d ) and fission cross-section of 235U and 238U. In the second stage, these parameters have been measured with an experimental method based on Inhour equation. The calculated and measured values are in good agreement. Relative Percent Errors (RPEs) are 2.8% for β eff and 5.7% for Λ in the cold state.

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.