Measurement Of Fission Cross Section Research Articles

A multi-cell fission chamber for fission cross-section measurements at the Back-n white neutron beam of CSNS

Abstract A multi-cell fast ionization chamber (FIC) has been designed and built as an initial detector for fission cross section measurement on the Back-n neutron beam at the China Spallation Neutron Source (CSNS). The current FIC has been carefully optimized to match the features of the neutron beam with capability of mounting up to 10 cells. The basic principle and characters of this detector and the front-end electronics are presented, together with the performance test during the first measurement campaign on the Back-n beam line. Some signal processing methods have been applied to deal with the poor γ -flash signals induced by the low intensity neutron beam. The current FIC shows good performances in high timing precision and clear background discrimination.

Neutron-induced fission cross-section measurement of 234U with monoenergetic beams in the keV and MeV range using MicroMegas detectors

The fission cross-section of 234U was measured at incident neutron energies between 300 and 500 keV and 4 and 5 MeV with a setup based on “microbulk” MicroMegas detectors. The standard 235,238U fission cross-sections were used as reference. The neutron beams were produced via the 7Li(p,n) and the 2H(d,n) reactions at the neutron beam facility of the 5.5 MeV Tandem accelerator laboratory at NCSR “Demokritos”. The mass of the actinide content of the targets used and of their impurities was quantitatively determined via α – spectroscopy. The developed methodology and preliminary results are presented.

Fission cross section measurements with the MicroMegas detector

A new MicroMegas detector, based on the innovative Microbulk technology, especially developed within the context of the n TOF collaboration, was used for the first time for fission cross section measurements with monoenergetic neutron beams at the Institute of Nuclear and Particle Physics of the NCSR “Demokritos”. The detector assembly was successfully used in the case of the measurement of the 237Np(n,f) cross section while the first results on the performance of the detector as far as the gain and resolution function are concerned are reported.

Fission cross-section measurement for 238U(n,f)134 ZX reactions induced by D-T neutrons

The D-T neutron induced fission fragments from 238U, i.e., 134Te, 134 mI and 134 gI, were measured by low-background HPGe-γ detector. The fission cross-sections of the 238U(n, f) 134Te, 238U(n, f))134 mI and 238U(n, f)134 gI reactions induced by D-T neutrons, i.e., 14.7 ± 0.3, 14.5 ± 0.3 and 14.1 ± 0.3 MeV, were measured precisely with the neutron activation technique, where the neutron flux was monitored by accompanying α-particle and the neutron energy was determined by the cross-section ratio of 90Zr(n, 2 n)89Zr to 93Nb(n, 2 n)92 mNb reaction. The fission cross-sections of the 238U(n, f)134Te reaction are 54.3 ± 3.5 mb, 52.6 ± 3.4 mb and 50.0 ± 3.2 mb, corresponding to the above energy points. However, the cross-sections of 238U(n, f)134 gI reactions were estimated to be below a value about 8.2 × 10−3 mb, since almost all the 134 gI nuclide, of which characteristic γ-rays were detected in the spectrum, originated from the 134Te decay. Meanwhile, the characteristic γ-rays of 134 mI were not found on any spectrum, and beyond the lower limit of measurement.

A multi-layered fast ionization chamber prototype for fission cross section measurements

Aiming at measuring the fission cross sections of various actinides in a wide neutron energy range at the newly built China Spallation Neutron Source (CSNS)—White Neutron Source (WNS), a dedicated multi-layered fast fission ionization chamber prototype was constructed. The physical design of the prototype and the front-end electronics which are optimized to the features of CSNS are here described. The performance of the prototype with uranium targets at a D-T neutron source is presented, focusing in particular on the good ability in fission fragment discrimination, high detection efficiency of 95.5% and time resolution of 20.3 ns.

Pu242 neutron-induced fission cross-section measurement from 1 to 2 MeV neutron energy

Relative values of the neutron-induced fission cross section $\ensuremath{\sigma}(n,f)$ of $^{242}\mathrm{Pu}$ have been measured with respect to the standard $^{1}\mathrm{H}(n,p)$ elastic scattering cross section, at average energies of 1.0, 1.4, and $1.9\phantom{\rule{4pt}{0ex}}\mathrm{MeV}$. The measurements are part of an international effort to reduce uncertainties and provide independent nuclear data relevant for fast-spectrum reactors. The shape of the measured cross section is in good agreement with data from Tovesson et al. [Phys. Rev. C 79, 014613 (2009)] and with the most recent data from Matei et al. [Phys. Rev. C 95, 024606 (2017)], but disagrees with the shapes of ENDF/B-VII.1 and JEFF3.2 evaluations. Absolute values of $\ensuremath{\sigma}(n,f)$, obtained under some assumptions, indicate an overestimation of $\ensuremath{\sigma}(n,f)$ in the evaluated libraries at 1.0 and $1.4\phantom{\rule{4pt}{0ex}}\mathrm{MeV}$, while a good agreement is found with ENDF/B-VII.1 at $1.9\phantom{\rule{4pt}{0ex}}\mathrm{MeV}$. A careful analysis of the impact of scattered neutrons and anisotropy of the fission fragment angular distribution has been performed. The measurement of the neutron flux by means of a proton-recoil detector is discussed. A comprehensive study of corrections applied to the data, of associated uncertainties, and of correlations between the measurements at the different energies is presented.

Effect of collective enhancement in level density in the fission of pre-actinides

Fission fragment angular distributions for three reactions, $^{19}\mathrm{F}+^{182}\mathrm{W},^{19}\mathrm{F}+^{187}\mathrm{Re}$, and $^{19}\mathrm{F}+^{193}\mathrm{Ir}$, are measured in the laboratory energy range of 82--120 MeV. Extracted fission cross sections of the present systems as well as those of three others from literature $(^{19}\mathrm{F}+^{192}\mathrm{Os},^{19}\mathrm{F}+^{194}\mathrm{Pt}$, and $^{19}\mathrm{F}+^{197}\mathrm{Au})$ are compared with the predictions of a statistical model which takes into account the effects of shell, orientation degree of freedom, and collective enhancement in level density (CELD). In all the cases, the standard statistical model predictions overestimate the measured fission cross section, indicating the presence of some amount of dynamical effects in the exit channel. A dissipation strength of $2\ifmmode\times\else\texttimes\fi{}{10}^{21}\phantom{\rule{4pt}{0ex}}{\mathrm{s}}^{\ensuremath{-}1}$ is found to be sufficient to reproduce the data of all the reactions. No scaling of fission barrier height to fit the data is required.

Measurement of fission cross-section for 134Te and 134I in 232Th(n,f) reaction induced by 14 MeV neutrons

The fission cross-sections of the 232Th(n,f)134Te and 232Th(n,f)134I reactions induced by neutrons around 14MeV were measured by the neutron activation and off-line gamma ray technique. The neutron fluence was monitored on line using the accompanying α-particles from the T(d,n)α reaction, meanwhile, neutron energies were obtained from the cross-section ratio method of 90Zr(n,2n)89(m+g)Zr to 93Nb(n,2n)92mNb reactions. The experimentally determined 232Th(n,f)134Te reaction cross-sections are 9.57±0.47mb at En=14.1±0.3MeV, 10.0±0.5mb at En=14.5±0.3MeV and 10.7±0.6mb at En=14.7±0.3MeV, respectively. Whereas, the isotope 134I of which characteristic γ-rays were detected in the γ-spectrum was concluded to be originated from 134Te decay. According to the cascade deduction formulism, the cross-section of 232Th(n,f)134I was estimated to be below a value (~5.7×10−3mb) refered to the Minimal Detectable Activity (MDA) with the gamma measurements.

Measurement of fission cross section for 232Th (n,f) 131 ZX ( Z = 50 , 51, 52, 53) reaction induced by neutrons around 14 MeV

The fission cross sections of 232Th (n,f) 131m, gSn , 232Th (n,f) 131Sb , 232Th (n,f) 131m, gTe , 232Th (n,f) 131I fission reactions induced by 14MeV neutrons were measured precisely with the neutron activation technique. The neutron flux was monitored by accompanying $ \alpha$ particle in the irradiation and the neutron energies were determined by the cross section ratio of 90Zr (n,2n) 89Zr to 93Nb (n,2n) 92mNb reaction. The values of the cross sections of 232Th (n,f) 131m, gSn were analyzed, and the cross sections of 232Th (n,f) 131Sb were deduced to be $ 6.5\pm 0.7$ , $ 6.3\pm 0.6$ , $ 6.1\pm 0.6$ mb at $ 14.1\pm 0.3$ , $ 14.5\pm 0.3$ and $ 14.8\pm 0.3$ MeV, respectively. The values of the cross sections of 232Th (n,f) 131gTe were deduced to be $ 1.8\pm 0.1$ , $ 1.5\pm 0.1$ and $ 1.4\pm 0.1$ mb at $ 14.1\pm 0.3$ , $ 14.5\pm 0.3$ and $ 14.8\pm 0.3$ MeV, respectively. The values of the cross sections of 232Th (n,f) 131I were given as $ 1.8\pm 0.2$ , $ 1.6\pm 0.2$ , $ 1.5\pm 0.1$ mb at $ 14.1\pm 0.3$ , $ 14.5\pm 0.3$ and $ 14.8\pm 0.3$ MeV, respectively.

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.

Evaluation of the235U resonance parameters to fit the standard recommended values

A great deal of effort has been dedicated to the revision of the standard values in connection with the neutron interaction for some actinides. While standard data compilation are available for decades nuclear data evaluations included in existing nuclear data libraries (ENDF, JEFF, JENDL, etc.) do not follow the standard recommended values. Indeed, the majority of evaluations for major actinides do not conform to the standards whatsoever. In particular, for the n + 235 U interaction the only value in agreement with the standard is the thermal fission cross section. A resonance re-evaluation of the n + 235 U interaction has been performed to address the issues regarding standard values in the energy range from 10−5 eV to 2250 eV. Recently, 235 U fission cross-section measurements have been performed at the CERN Neutron Time-of-Flight facility (TOF), known as n_TOF, in the energy range from 0.7 eV to 10 keV. The data were normalized according to the recommended standard of the fission integral in the energy range 7.8 eV to 11 eV. As a result, the n_TOF averaged fission cross sections above 100 eV are in good agreement with the standard recommended values. The n_TOF data were included in the 235 U resonance analysis that was performed with the code SAMMY. In addition to the average standard values related to the fission cross section, standard thermal values for fission, capture, and elastic cross sections were also included in the evaluation. This paper presents the procedure used for re-evaluating the 235 U resonance parameters including the recommended standard values as well as new cross section measurements.

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 ...

Neutron induced fission cross section measurements of240Pu and242Pu

Accurate neutron induced fission cross section of 240 Pu and 242 Pu are required in view of making nuclear technology safer and more efficient to meet the upcoming needs for the future generation of nuclear power plants (GEN-IV). The probability for a neutron to induce such reactions figures in the NEA Nuclear Data High Priority Request List [1]. A measurement campaign to determine neutron induced fission cross sections of 240 Pu and 242 Pu at 2.51 MeV and 14.83 MeV has been carried out at the 3.7 MV Van De Graaff linear accelerator at Physikalisch-Technische Bundesanstalt (PTB) in Braunschweig. Two identical Frisch Grid fission chambers, housing back to back a 238 U and a A Pu target (A = 240 or A = 242), were employed to detect the total fission yield. The targets were molecular plated on 0.25 mm aluminium foils kept at ground potential and the employed gas was P10. The neutron fluence was measured with the proton recoil telescope (T1), which is the German primary standard for neutron fluence measurements. The two measurements were related using a De Pangher long counter and the charge as monitors. The experimental results have an average uncertainty of 3–4% at 2.51 MeV and for 6–8% at 14.81 MeV and have been compared to the data available in literature.

Neutron-induced fission cross-section measurement of 234U with quasi-monoenergetic beams in the keV and MeV range using micromegas detectors

Accurate data on neutron-induced fission cross-sections of actinides are essential for the design of advanced nuclear reactors based either on fast neutron spectra or alternative fuel cycles, as well as for the reduction of safety margins of existing and future conventional facilities. The fission cross-section of $^{234}$U was measured at incident neutron energies of 560 and 660 keV and 7.5 MeV with a setup based on ‘microbulk’ Micromegas detectors and the same samples previously used for the measurement performed at the CERN n_TOF facility (Karadimos et al., 2014). The $^{235}$U fission cross-section was used as reference. The (quasi-)monoenergetic neutron beams were produced via the 7Li(p,n) and the 2H(d,n) reactions at the neutron beam facility of the Institute of Nuclear and Particle Physics at the ‘Demokritos’ National Centre for Scientific Research. A detailed study of the neutron spectra produced in the targets and intercepted by the samples was performed coupling the NeuSDesc and MCNPX codes, taking into account the energy spread, energy loss and angular straggling of the beam ions in the target assemblies, as well as contributions from competing reactions and neutron scattering in the experimental setup. Auxiliary Monte-Carlo simulations were performed with the FLUKA code to study the behaviour of the detectors, focusing particularly on the reproduction of the pulse height spectra of α-particles and fission fragments (using distributions produced with the GEF code) for the evaluation of the detector efficiency. An overview of the developed methodology and preliminary results are presented.

Use of integral experiments for the assessment of a new235U IRSN-CEA evaluation

The Working Party on International Nuclear Data Evaluation Co-operation (WPEC) subgroup 29 (SG 29) was established to investigate an issue with the 235 U capture cross-section in the energy range from 0.1 to 2.25 keV, due to a possible overestimation of 10% or more. To improve the 235 U capture crosssection, a new 235 U evaluation has been proposed by the Institut de Radioprotection et de Surete Nucleaire (IRSN) and the CEA, mainly based on new time-of-flight 235 U capture cross-section measurements and recent fission cross-section measurements performed at the n_TOF facility from CERN. IRSN and CEA Cadarache were in charge of the thermal to 2.25 keV energy range, whereas the CEA DIF was responsible of the high energy region. Integral experiments showing a strong 235 U sensitivity are used to assess the new evaluation, using Monte-Carlo methods. The k eff calculations were performed with the 5.D.1 beta version of the MORET 5 code, using the JEFF-3.2 library and the new 235 U evaluation, as well as the JEFF-3.3T1 library in which the new 235 U has been included. The benchmark selection allowed highlighting a significant improvement on k eff due to the new 235 U evaluation. The results of this data testing are presented here.

Simultaneous measurement of neutron-induced fission and capture cross sections for 241Am at neutron energies below fission threshold

Fission and capture reactions were simultaneously measured in the neutron-induced reactions of 241Am at the spallation neutron facility of the Japan Proton Accelerator Research Complex (J-PARC). Data for the neutron energy range of En=0.1–20eV were taken with the TOF method. The fission events were observed by detecting prompt neutrons accompanied by fission using liquid organic scintillators. The capture reaction was measured by detecting γ rays emitted in the deexcitation of the compound nuclei using the same detectors, where the prompt fission neutrons and capture γ rays were separated by a pulse shape analysis. The cross sections were obtained by normalizing the relative yields at the first resonance to evaluations or other experimental data. The ratio of the fission to capture cross sections at each resonance is compared with those from an evaluated nuclear data library and other experimental data. Some differences were found between the present values and the library/literature values at several resonances.

Measurement of fission cross-section for the 232Th(n,f )141Ba reaction induced by neutrons around 14 MeV

The fission cross-section of the 232Th(n,f)141Ba reaction induced by neutrons around 14MeV was measured precisely with the neutron activation and off-line gamma-ray spectrometric technique. Neutron fluence was monitored on-line using the accompanying $\alpha$ -particles from the 3H(2H,n)4He reaction, whereas the neutron energies were measured by the method of cross-section ratios of 90Zr(n,2n)89Zr to 93Nb(n,2n)92mNb reactions. The experimentally determined 232Th(n,f)141Ba reaction cross-sections were $12.2\pm 0.4$ mb at $E_{n} = 14.1 \pm 0.3$ MeV, 13.0±0.5 mb at $E_{n} = 14.5 \pm 0.3$ MeV and 13.3±0.5 mb at $E_{n} = 14.7 \pm 0.3$ MeV, respectively.

Extended optical model for fission

A comprehensive formalism to calculate fission cross sections based on the extension of the optical model for fission is presented. It can be used for description of nuclear reactions on actinides featuring multi-humped fission barriers with partial absorption in the wells and direct transmission through discrete and continuum fission channels. The formalism describes the gross fluctuations observed in the fission probability due to vibrational resonances, and can be easily implemented in existing statistical reaction model codes. The extended optical model for fission is applied for neutron induced fission cross-section calculations on $^{234,235,238}\mathrm{U}$ and $^{239}\mathrm{Pu}$ targets. A triple-humped fission barrier is used for $^{234,235}\mathrm{U}(n,f)$, while a double-humped fission barrier is used for $^{238}\mathrm{U}(n,f)$ and $^{239}\mathrm{Pu}(n,f)$ reactions as predicted by theoretical barrier calculations. The impact of partial damping of class-II/III states, and of direct transmission through discrete and continuum fission channels, is shown to be critical for a proper description of the measured fission cross sections for $^{234,235,238}\mathrm{U}(n,f)$ reactions. The $^{239}\mathrm{Pu}(n,f)$ reaction can be calculated in the complete damping approximation. Calculated cross sections for $^{235,238}\mathrm{U}(n,f)$ and $^{239}\mathrm{Pu}(n,f)$ reactions agree within 3% with the corresponding cross sections derived within the Neutron Standards least-squares fit of available experimental data. The extended optical model for fission can be used for both theoretical fission studies and nuclear data evaluation.

A Comprehensive Approach to Determination of Nuclear Data of Unstable Nuclei

A comprehensive approach to determine nuclear data of unstable nuclei will be described. It consists of a measurement of fission and capture cross sections, mass distribution of fission fragments (independent fission yields) and number of prompt fission neutrons by the method of surrogate reactions. A multi-dimensional Langevin model is being developed to estimate the independent fission yields theoretically. Furthermore, the β decay properties of the fission fragments, almost all are neutron-rich unstable nuclei, are investigated systematically by improving the gross theory of β decay, which will yield information on the decay heat and delayed-neutron data

Development of a gaseous proton-recoil detector for fission cross section measurements below 1 MeV neutron energy

The elastic H(n,p) reaction is sometimes used to measure neutron flux, in order to produce high precision measurements. The use of this technique is not straightforward to use below incident neutron energy of 1 MeV, due to a high background in the detected proton spectrum. Experiments have been carried out at the AIFIRA facility to investigate such background and determine its origin and components. Based on these investigations, a gaseous proton-recoil detector has been designed, with a reduced low energy background.