Fission Neutron Spectrum Research Articles

Measurement of prompt neutron spectra from the239Pu(n,f) fission reaction for incident neutron energies from 1 to 200 MeV

Prompt fission neutron spectra in the neutron-induced fission of ${}^{239}$Pu have been measured for incident neutron energies from 1 to 200 MeV at the Los Alamos Neutron Science Center. Mean energies deduced from the prompt fission neutron spectra (PFNS) lead to the observation of the opening of the second chance fission at 7 MeV and to indications for the openings of fission channels of third and fourth chances. Moreover, the general trend of the measured PFNS is well reproduced by the different models. The comparison between data and models presents, however, two discrepancies. First, the prompt neutron mean energy seems constant for neutron energy, at least up to 7 MeV, whereas in the theoretical calculations it is continuously increasing. Second, data disagree with models on the shape of the high energy part of the PFNS, where our data suggest a softer spectrum than the predictions.

Measurement of the Neutron Spectrum Using the Activation Method

A new method for the measurement of the neutron spectra has been developed (Wieslander et al., 2010, Lövestam et al., 2009). The technique is based on the measurement of γ-rays from the neutron-induced activity in a series of different metal disks followed by spectrum unfolding procedures. In this study, we measure for the first time with this technique a prompt fission neutron spectrum. Special attention is paid to the unfolding procedures and the estimation of the lowest measurable neutron fluence rate, as well as the estimation of sources of the uncertainties.

Prompt Fission Neutron Spectrum Uncertainty Propagation Using Polynomial Chaos Expansion

The polynomial chaos expansion-stochastic collocation method (PCE-SCM) is demonstrated to be a computationally efficient approach for propagating nuclear data uncertainties evaluated for the prompt fission neutron spectra (PFNS) of n + 235U and n + 239Pu fission reactions through two fast neutron critical benchmark experiments. A principal component decomposition of the PFNS covariance matrices yields an efficient representation of the uncertainty in terms of two to four random variables. Both normal and uniform distributions are considered for these random variables, and the random output variables (angular flux and k-eigenvalue) are expressed in terms of Hermite and Legendre chaos expansions, respectively. Tensor product Hermite and Legendre Gauss quadrature sets, respectively, are used to relate the deterministic chaos expansion coefficients to solutions of independent transport k-eigenvalue problems, and the resulting polynomial chaos expansion provides a complete statistical characterization of the uncertainty in the output variables. Direct random sampling of the PFNS followed by repeated solution of the transport problem to create an ensemble of solutions is used to benchmark results obtained from the PCE-SCM implementation. Both direct random sampling and the PCE-SCM implementation yield comparable results where, for the Jezebel and Lady Godiva critical assemblies, the calculated uncertainties in keff resulting from the PFNS propagated uncertainties are found to be of the same order or larger than reported experimental measurement uncertainties, respectively. The PCE-SCM implementation results obtained require orders of magnitude less computational resources compared with the direct random sampling approach.

Evaluation and Uncertainty Quantification of Prompt Fission Neutron Spectra of Uranium and Plutonium Isotopes

The prompt fission neutron spectra (PFNS) of the low-incident-energy neutron-induced fission reactions n + 229-238U and n + 235-242Pu have been systematically evaluated using differential experimental data and the Los Alamos model (LA model). Using the first-order, linear Kalman filter, the LA model parameters are constrained using the experimental data and an evaluation of the PFNS and its uncertainties across a suite of isotopes’ results. Correlations between isotopes of each actinide are presented through the model parameter correlations, and the resulting evaluations can be used to fill in inconsistencies within the ENDF/B-VII.1 library where PFNS data are scarce or in need of an update.

Prompt neutron and gamma-ray emission calculations for 232Th(n,f)

Abstract Model calculations of prompt neutron and gamma-ray emission of 232 Th(n,f) up to the incident energy of 20 MeV are reported. For the first time the experimental average prompt gamma-ray energy data of this fissioning nucleus are very well described by a model result. The calculated total average prompt neutron multiplicity gives an excellent description of experimental data over the entire fast neutron energy range, including the variations around the energies of fission cross-section resonances and the sudden increase at the incident energy where the second fission chance is opening. The ν ¯ ( A ) behavior consisting in the multiplicity increase with the incident energy for heavy fragments only, experimentally observed for 235 U(n,f) and 237 Np(n,f) and confirmed by model calculations, is supported by the Point-by-Point model results in the case of 232 Th(n,f), too. The correlation between the sub-barrier resonances in the fission cross-section and the non-statistical variations of fragment and prompt emission data around the resonance energies, already discussed for the fertile actinides 238 U and 234 U, is confirmed by the present results of 232 Th(n,f). The average prompt gamma multiplicity result, almost constant with the incident neutron energy, is in agreement with the experimental observation of Frehaut. The calculated prompt fission neutron spectra describe well the existing experimental data. The lump around 8 MeV, exhibited by the experimental spectrum data measured at 14.6 MeV, is well described by the present calculations by including the contribution of the (n,xn) spectra of neutrons emitted prior to the scission provided by nuclear reaction calculations.

Sensitivity and uncertainty analysis for an accelerator-driven system with JENDL-4.0

A sensitivity and uncertainty analysis was performed for the accelerator-driven system (ADS) proposed by the Japan Atomic Energy Agency (JAEA) with the latest version of the Japanese Evaluated Nuclear Data Library (JENDL-4.0). Significant discrepancies have been found between the reactor physics parameters of JENDL-4.0 and those of JENDL-3.3. An analysis with the sensitivity coefficients showed that the major contributors to these discrepancies are the differences in the inelastic scattering cross sections of 206Pb and 207Pb, and the capture and inelastic scattering cross sections and ν value of 241Am. The uncertainty analysis with the JENDL-4.0 covariance data found that the covariances of the fission neutron spectrum of minor actinides (MAs) have a considerable impact on the uncertainties of the reactor physics parameters.

Statistical and evaporation models for the neutron emission energy spectrum in the center-of-mass system from fission fragments

Abstract The neutron emission energy spectra in the CMS (center-of-mass) frame from two compound nuclei produced by fission are studied. The neutron spectra calculated with the Hauser–Feshbach statistical model are compared with the evaporation theory, and the definition of the temperature is revisited. Using the Monte Carlo technique we average the CMS neutron spectra from many fission fragments to construct the representative CMS spectrum from both the light and heavy fragments. The CMS spectra for each fission fragment pair are also converted into the laboratory frame to calculate the total prompt fission neutron spectrum that can be observed experimentally. This is compared to measured laboratory data for thermal neutron induced fission on 235 U. We show that the Hauser–Feshbach calculation gives a different spectrum shape than the Madland–Nix model calculation.

Pulse shape discrimination in the plastic scintillator EJ-299-33

New advances in plastic scintillation compositions have opened the field to new, exciting instruments capable of neutron-gamma ray pulse shape discrimination (PSD). We present PSD figure of merit parameters and neutron time-of-flight from Cf-252 using a 5.08-cm diameter by 5.08-cm thick sample of PSD-capable plastic scintillator EJ-299-33 and compare these results to those from a same-sized EJ-309 liquid scintillator detector. An offline, digital PSD method was applied to both detectors. The results show that EJ-299-33 plastic PSD is very good, having a figure of merit of approximately 0.9 for 120keVee threshold; however the EJ-309 liquid scintillator PSD is superior to the EJ-299-33 plastic scintillator PSD, with a figure of merit of 1.5 at the same measurement threshold. We also found that the EJ-299-33 has reduced neutron detection efficiency compared to the EJ-309. For the fission neutron spectrum measured here, the ratio of the plastic to liquid total number of measured neutrons was approximately 0.63.

Source test of the prototype neutron detector for the large-acceptance multipurpose spectrometer at RAON

A neutron detector array will be essential for the study of the nuclear symmetry energy in the large-acceptance multipurpose spectrometer (LAMPS) at the planned rare-isotope beam facility RAON in Korea. We have built the prototype neutron detector for LAMPS and examined its performance by using radiation sources. For data taking, we tested the voltage-threshold discriminator (VTD) and the constant-fraction discriminator (CFD) modules for the pulse process. The intrinsic time resolution of the prototype detector is estimated to be 723 ps for VTD and 488 ps for CFD. The fission neutrons and gammas emitted from 252Cf can be clearly separated in the time distribution. We reconstruct the energy spectrum of the spontaneous fission neutrons from 252Cf, which can be described well by using the empirical Watt spectrum.

Development of Neutron Detector Arrays for Neutron-Induced Reaction Measurements

The outgoing neutron energy spectra from neutron-induced fission of various actinides are important for basic understanding of the fission process near the scission point as well as playing a large role in neutron transport codes, which are heavily relied upon in the design of advanced nuclear reactors and simulations of critical assemblies. The reliability of the results of neutron transport models is a strong function of the quality of the nuclear data used as input. Currently, the world's experimental database of fission neutron spectra is severely incomplete (especially for higher incident neutron energies) with large uncertainties in key portions of the outgoing energy spectra. Many transport codes use evaluated data libraries, which are based on the approach of the Los Alamos model. Other theoretical models have been developed, but the available data cannot distinguish the results of different models (as is the case for <formula formulatype="inline" xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink"><tex Notation="TeX">$^{239}$</tex> </formula> Pu). Better measurements are needed for all incident and outgoing neutron energies, but most urgently in the low-energy (below 1 MeV) and high-energy (above 6 MeV) portions of the outgoing spectra where theoretical model results differ greatly. We present the design considerations (and some characterization results) of the two Chi-Nu neutron detector arrays: one array of <formula formulatype="inline" xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink"><tex Notation="TeX">$^{6}$</tex></formula> Li-glass detectors and one array of liquid-scintillator detectors. These detector arrays are being constructed to meet the challenge of measuring the prompt fission neutron spectra (for a few common actinides) to a higher accuracy and precision than achieved previously and over a larger incident energy range than has been covered by previous experimenters. We see a significant reduction in neutron-scattering backgrounds with our new array designs.

Sensitivity and uncertainty in the effective delayed neutron fraction ([formula omitted

Precise knowledge of the effective delayed neutron fraction (βeff) and the corresponding uncertainty is important for nuclear reactor safety analysis. The interest in developing the methodology for estimating the uncertainty in βeff was expressed in the scope of the UAM project of the OECD/NEA. The sensitivity and uncertainty analysis of βeff performed using the standard first-order perturbation code SUSD3D is presented. The sensitivity coefficients of βeff with respect to the basic nuclear data were calculated by deriving Bretscher's k-ratio formula. The procedure was applied to several fast neutron benchmark experiments selected from the ICSBEP and IRPhE databases. According to the JENDL-4.0m covariance matrices and taking into account the uncertainties in the cross-sections and in the prompt and delayed fission spectra the total uncertainty in βeff was found to be in general around 3%, and up to ∼7% for the 233U benchmarks. An approximation was applied to investigate the uncertainty due to the delayed fission neutron spectra. The βeff sensitivity and uncertainty analyses are furthermore demonstrated to be useful for the better understanding and interpretation of the physical phenomena involved. Due to their specific sensitivity profiles the βeff measurements are shown to provide valuable complementary information which could be used in combination with the criticality (keff) measurements for the evaluation and validation of certain nuclear reaction data, such as for example the delayed (and prompt) fission neutron yields and interestingly also the 238U inelastic and elastic scattering cross-sections.

Angular resolution study of a combined gamma-neutron coded aperture imager for standoff detection

Nuclear threat source observables at standoff distances of tens of meters from mCi class sources include both gamma-rays and neutrons. This work uses simulations to investigate the effects of the angular resolution of a mobile gamma-ray and neutron coded aperture imaging system upon orphan source detection significance and specificity. The design requires maintaining high sensitivity and specificity while keeping the system size as compact as possible to reduce weight, footprint, and cost. A mixture of inorganic and organic scintillators was considered in the detector plane for high sensitivity to both gamma-rays and fast neutrons. For gamma-rays (100 to 2500keV) and fission spectrum neutrons, angular resolutions of 1–9° and radiation angles of incidence appropriate for mobile search were evaluated. Detection significance for gamma-rays considers those events that contribute to the photopeak of the image pixel corresponding the orphan source location. For detection of fission spectrum neutrons, energy depositions above a set pulse shape discrimination threshold were tallied. The results show that the expected detection significance for the system at an angular resolution of 1° is significantly lower compared to its detection significance an angular resolution of ∼3–4°. An angular resolution of ∼3–4° is recommended both for better detection significance and improved false alarm rate, considering that finer angular resolution does not result in improved background rejection when the coded aperture method is used. Instead, over-pixelating the search space may result in an unacceptably high false alarm rate.

252Cf fission-neutron spectrum using a simplified time-of-flight setup: An advanced teaching laboratory experiment

The removal of PuBe and AmBe neutron sources from many university teaching laboratories (due to heightened security issues) has often left a void in teaching various aspects of neutron physics. We have recently replaced such sources with sealed 252Cf oil-well logging sources (nominal 10–100 μCi), and developed several experiments using them as neutron sources. This includes a fission-neutron time-of-flight experiment using plastic scintillators, which utilizes the prompt γ rays emitted in 252Cf spontaneous fission as a fast timing start signal. The experiment can be performed with conventional nuclear instrumentation and a 1-D multi-channel pulse-height analyzer, available in most advanced teaching laboratories. Alternatively, a more sophisticated experiment using liquid scintillators and n/γ pulse-shape discrimination can be performed. Several other experiments using these neutron sources are also feasible. The experiments can introduce students to the problem of detecting the dark matter thought to dominate the universe and to the techniques used to detect contraband fissionable nuclear materials.

Experimental Studies of Prompt Fission Neutron Energy Spectra

Abstract Prompt fission neutron spectra were measured in the reactions 238U(n,f), 235U(n,f) and 237Np(n,f) at different incident neutron energies. The neutrons were detected using a coaxial doped p-terphenyl scintillation detector in coincidence with fission fragments and their time-of-flight was recorded. The properties of the neutron detector were determined and the results are presented in this work. A preliminary neutron detection efficiency was applied to data from the neutron-induced fission of 238U at En = 5.2 MeV, leading to encouraging results.

Statistical Theory of Fission Chains and Generalized Poisson Neutron Counting Distributions

The neutron counting probability distribution for a multiplying medium was shown by Hage and Cifarelli to be a generalized Poisson distribution that depends on the fission chain number distribution. An analytic formula is obtained for this number distribution, the probability to produce a number of neutrons in a fission chain. The formula depends on the probability that a fission spectrum neutron induces a subsequent fission and depends on the probability distribution for a specific number of neutrons to be produced in an individual induced fission. The formula is an exact solution to a functional equation due to Böhnel for the probability generating function. The Böhnel equation is derived as the t →∞ limit of a rate equation for a neutron population generating function, related to a rate equation studied by Feynman. The Böhnel equation is also shown to be a fixed point of an iteration problem, related to one studied by Hawkins and Ulam, where the iteration generates the chain a generation at a time. The discrete iteration problem is shown to be connected to the continuous time evolution of the chain. An explicit solution for the time evolution of the chain is given in the simplified approximation where at most two neutrons are created by an induced fission. The t →∞ limit of this equation gives a simple analytic expression for the solution to the Böhnel equation in this approximation. A generalized Poisson counting distribution constructed from the theoretical fission chain probability number distribution is compared to experimental data for a multiplying Pu sample.

Fission Neutron Spectrum Sensitivity Study for the Case of Advanced Heavy Water Reactor

In the WIMSD-IAEA multigroup nuclear data library, the isotopes and weights adopted for WLUP libraries to calculate the average fission spectra for 235U, 238U, and 239Pu are in the ratio of 54%, 8%, and 38%, respectively. The average fission neutron spectrum in the existing multigroup WIMSD-IAEA library applicable for the U-Pu cycle is not rigorously applicable for systems that are based on the thorium fuel cycle because of two aspects. First, the weightage of the fission neutron spectrum of 232Th and 233U nuclides, which are important isotopes in the thorium fuel cycle, are not considered in obtaining the average multigroup fission spectrum in the conventional WIMSD-IAEA library. Second, the 232Th/233U system spectrum is required for condensation of the fission spectrum as done in generating other multigroup cross sections and parameters for the thorium fuel cycle. In this work, we have processed the fission neutron spectrum data from the basic evaluated nuclear data file (ENDF/B-VI.8) for each important isotope in the thorium fuel cycle using the Th/233U spectrum and using a FORTRAN program developed and validated by us for this purpose. The final average fission spectrum to be fed into the WIMSD-IAEA library is prepared by mixing the isotopic multigroup fission spectrum of individual isotopes 233U, 239Pu, and 241Pu with appropriate weights corresponding to their respective power fractions in the advanced heavy water reactor (AHWR) lattice. Using the WIMSD library with modified effective fission spectra, the lattice k-infinity calculations of AHWR are performed as a function of burnup. The difference in the infinite multiplication factor, which is expressed in terms of reactivity in mk, ranges from 0.48 to 0.94 mk as burnup in the AHWR proceeds from 0 to 55 GWd/tonne.

Fission yield measurements from highly enriched uranium irradiated inside a boron carbide capsule

A boron carbide capsule was previously designed and tested by Pacific Northwest National Laboratory (PNNL) and Washington State University (WSU) for spectral-tailoring in mixed spectrum reactors. The presented work used this B4C capsule to create a fission product sample from the irradiation of highly enriched uranium (HEU) with a fast fission neutron spectrum. An HEU foil was irradiated inside of the capsule in WSU’s 1 MW TRIGA reactor at full power for 200 min to produce 5.8 × 1013 fissions. After 3 days of cooling, the sample was shipped to PNNL for radiochemical separations and analysis by gamma and beta spectroscopy. Fission yields for products were calculated from the radiometric measurements and compared to measurements from thermal neutron induced fission (analyzed in parallel with the non-thermal sample at PNNL) and published evaluated fast-pooled and thermal nuclear data. Reactor dosimetry measurements were also completed to fully characterize the neutron spectrum and total fluence of the irradiation.

Efficient fission neutron spectrum matrix representation by singular value decomposition technique

In order to treat efficiently a huge fission neutron spectrum (FNS) matrix in a criticality calculation, the singular value decomposition (SVD) technique is introduced to an FNS matrix representation. The required number of SVD components for reconstruction of an FNS matrix is expected to be small since an incident neutron energy dependence of FNS is not so significant. The proposed technique of an SVD-based representation for a fission source term is tested in several fast critical systems. Through an observation of critical eigenvalue dependence on the number of considered SVD components, only six or seven components are required to obtain a critical eigenvalue which agrees with the reference solution within 10−4 dk/kk′. It is also confirmed that a small reactivity effect caused by neutron spectrum shifting can be accurately calculated with the proposed technique.

Evaluations of Fission Chain Yields for 239Pu from Fission-Spectrum Neutrons

Because of the importance of accurate data for fission chain yields (FCYs) for many applications, we present a rigorous “clean sheet” evaluation of all available data to provide an accurate set of pertinent FCYs. Because some nuclear data (e.g., half-lives, branching ratios, etc.) have been refined since the original analyses, where possible we update the data and their associated uncertainties. This evaluation is particularly topical since there are differences in the nuclear data used by radiochemists at different laboratories internationally and since some experiments from the 1970s have been recently reexamined with details published for the first time.The focus of this work is the production of a small set of fission products (95Zr, 99Mo, 144Ce, 147Nd) from plutonium irradiated by fission spectrum neutrons. Because 147Nd is a common isotope used at several laboratories, its production rate is examined critically. We find that most of the interlaboratory discrepancies can be explained by a dependence of its yield on the energy of the neutron causing fission, so we consider in detail the statistical significance of this claim. The potential for neutron energy dependence of 147Nd production from plutonium was first recognized in 1977 by Maeck and recently raised again as a possibility by Chadwick. The data for 95Zr, by contrast, demonstrate no statistically significant energy-dependence trends, but the data at the higher energies demonstrate significant scatter.With the relatively small number of data points, and recognizing that measurement methods and technologies have likely significantly improved in the nearly 30 years since the last measurement, additional measurements to refine the assessment and improve the uncertainties may be warranted.

Calculation of Prompt Fission Neutron from 233U(n, f) Reaction by Multi-Modal Los Alamos Model

An attempt is made to improve the evaluation of the prompt fission neutron emission from 233U(n, f) reaction for incident neutron energies below 6 MeV. The multi-modal fission approach is applied to the improved version of Los Alamos model and the point by point model. The prompt fission neutron spectra and the prompt fission neutron as a function of fragment mass (usually named “sawtooth" data) υ(A) are calculated independently for the three most dominant fission modes (standard I, standard II and superlong), and the total spectra and υ(A) are synthesized. The multi-modal parameters are determined on the basis of experimental data of fission fragment mass distributions. The present calculation results can describe the experimental data very well, and the proposed treatment is thus a useful tool for prompt fission neutron emission prediction.