Prompt Fission Neutron Research Articles

Fusion neutron source and array of particle detectors for nondestructive interrogation of special nuclear materials

Presented herein are the outcomes of an experimental test involving a pioneering portable-active interrogation system designed for the nondestructive detection of special nuclear materials (SNMs). The system relies on the threshold energy neutron analysis concept and incorporates a portable deuterium–deuterium (DD) neutron generator producing a particle intensity of 5 × 107 n/s, coupled with three arrays of tensioned metastable fluid detectors (TMFDs) to detect secondary neutrons from the fissile material. In the presence of the fissile material, prompt fission neutrons are emitted, with an average energy of approximately 2 MeV, and around 30% of these neutrons have energies above that of the DD neutron source (2.45 MeV). The detection of a statistically significant neutron population exceeding this threshold firmly indicates the presence of SNM. TMFDs exhibit high sensitivity in efficiently detecting neutrons above the threshold while adeptly discriminating against neutrons below the threshold as well as gamma rays. This unique feature allows the interrogation system to maintain a lightweight profile without necessitating substantial shielding materials. The validation experiments involved the placement of 70 or 140 g masses of U-235 within a 1 m3 inspection volume. Measurements were carried out over 30 min intervals, repeated numerous times, both with and without U-235, at a DD neutron source intensity of 8 × 105 n/sec. Experimental count rates with natural uranium (NU) are consistently above those without NU. The probability of detection (PD) and probability of false alarm (PFA) were assessed utilizing these count rates. The DD neutron source intensity and inspection time were normalized at 5 × 107 n/sec and 90 s, respectively. The results indicated a PD of approximately 74% and 98% for detecting 70 and 140 g of U-235, respectively, with a PFA of <5%. These promising outcomes align with the specified PD (>90%) and PFA (<5%) targets outlined in ANSI standards.

Design and performance of an epithermal neutron detector based on PFN

In this paper, an epithermal neutron detector suitable for the Prompt Fission Neutron (PFN) uranium logging method is designed by Monte Carlo simulation. According to the simulation results, the epithermal neutron detector composed of a 1 mm cadmium (Cd) layer, a 5 mm high-density polyethylene (HDPE) layer, and a 3He tube is sensitive to epithermal neutrons. Through the simulation test of Monte Carlo in formations with different uranium content, the results show that the uranium content in the formation has an obvious linear relationship with E/T, and the detection limit can meet the minimum standard for uranium ore detection. It meets the design expectation and provides a basis for the optimal design of an epithermal neutron detector.

Matrix effect correction in active neutron interrogation for residual fissile mass assessment in radioactive waste drums

Metallic waste in the form of shells and nozzles remains following the reprocessing of spent fuel at La Hague plant (France). Before the final disposal, drums containing the waste are transported to the compaction facility where its volume is reduced by a factor of 5. With the objective of controlling the criticality/safety levels, an active neutron interrogation system at the entrance of the facility is used to assess the residual fissile mass remaining in the waste, which relies on prompt fission neutron detection (known as the differential die-away technique, DDA). The measured signal is proportionally linked to the fissile mass by a calibration coefficient. However, two effects can produce an inaccurate prediction. The number of induced fissions for the same fissile mass varies according to the waste matrix composition, particularly the neutron slow-down and absorption ratio. Secondly, the presence of fissile clusters can impact the neutron interrogation due to an increase in self-shielding effect. This work presents a matrix effect correction method based on the use of internal flux monitors (sensitive to the neutron absorption ratio) and the transmission signal (sensitive to the slow-down ratio). It relies on a numerical model of the whole measurement cell developed with the Monte-Carlo N-Particle transport (MCNP) code. The calibration coefficient of 72 different matrices, representative of the waste produced at La Hague, were simulated and a statistical model using a multilinear regression was then established. This simulated-based approach provides a more robust and comprehensive estimation of the residual fissile mass without impacting accuracy thanks to the flux monitors and the transmission signals.

Experimental study of the influence of borehole parameters on prompt fission neutron uranium logging and its corrections

In prompt fission neutron uranium logging, borehole environmental parameters affect the measured results and must be corrected. In order to explore the influence of borehole parameters on the interpretation of logging results, this paper builds a sandstone type uranium ore block model to simulate the field production drilling device based on the “Epithermal/Thermal neutron counting rate ratio” (E/T) theory. The effects of borehole diameter, thickness of iron tube and well fluid on the decay rate of epithermal and thermal neutrons and their uncertainty correction methods were investigated. The results show that the effect of borehole diameter on E/T is negligible. The iron tube thickness has a certain effect on the moderation and absorption of epithermal and thermal neutrons, and its E/T increases slightly with increasing thickness. The influence of iron tube thickness on E/T is corrected and the relative uncertainty is less than 5%. The well fluid thickness also affects the decay rate of epithermal and thermal neutrons, and its E/T follows the law of negative exponential attenuation. The influence of well fluid thickness on E/T is corrected and the relative uncertainty is less than 5%. This study provides technical guidance for field well survey of uranium deposit.

Prompt-fission observable and fission yield calculations for actinides by TALYS

The nuclear reaction code TALYS adopts the Hauser-Feshbach statistical decay theory, to de-excite fission fragments. This involves for instance the evaporation of prompt fission neutrons and γ-rays. TALYS incorporates databases of primary fission fragment distribution which consists of primary fission fragment yield and data for excitation energy distribution of fission fragments. We conducted a sensitivity study on three parameters in TALYS and fitted them in order to reproduce experimental and evaluated data, in thermal neutron-induced fission of 235U. Moreover, we demonstrate a large-scale calculation of average prompt neutron and γ-ray multiplicities for 243 selected actinide isotopes.

Measurements of 235U(nth,f) Spectrum Averaged Cross Sections (SACS) in zero power reactors

This paper deals with the measurement of Spectrum Averaged Cross Sections in two different neutron fields formed in zero power reactors. The first was Benchmark Neutron Reference Field in the LR-0 reactor, and the second field was in the center of the vertical channel touching the fuel in the VR-1 reactor. The spectrum averaged cross section differs for both cases as the spectra differ, but after normalization to 235U PFNS using calculated correction, both results are in good agreement, thus confirming the spectra in both cases are similar in the 1 – 14 MeV region. A good agreement between lower threshold reactions averaged in actual reactor spectra and prompt fission neutron spectrum of 235U is reported as well.

Multiplicity of prompt fission neutron in the 239Pu(n,f) reaction and its energy dependence

Multiplicity of prompt fission neutron in the ²³⁹Pu(n,f) reaction and its energy dependence

Calculation of the fission products for neutron-induced fission of 235U

The fission model, G4ParaFissionModel, was enhanced in this study, mainly focusing on refining the energy dependence of the peak-to-valley ratio in the mass distribution and the energy dependence of the average total kinetic energy (TKE). The enhanced model was employed to investigate the characteristics of fission products from 235U(n, f) reaction. The calculated results, including fission yield, TKE distribution, prompt fission neutron and gamma spectra, were compared with both evaluated and experimental data. The comparison shows that these physical observables related nuclear data, which are of importance for developments of the nuclear power and physics, can be reasonably well reproduced.

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.

Development of a neutron probe to perform a combined measurement of uranium concentration and hydrogen porosity for mining applications

This work reports the development of a new neutron probe for mining prospection and exploitation. It allows a combined measurement of hydrogen porosity, based on the detection of neutrons backscattered by the geological formation, and of uranium concentration based on induced-fission prompt neutrons, with the Differential Die-away Technique. The probe includes a pulsed neutron generator and a single neutron counter embedded in a polyethylene block wrapped with cadmium, saving the place for a gamma-ray detector compared to usual porosity probes with two 3He counters. We first report design and feasibility studies using MCNP 6.2 simulations, showing the significant effect of hydrogen porosity on the uranium signal and the need for a combined interpretation. Then, we report laboratory experiments that validate the simulation methods of both the porosity and uranium useful signals, but also the active background due to 17O activation. Taking into account neutron absorbers in the sand of our testing system, like boron, lithium and gadolinium measured by ICP-MS, the simulated and experimental signals of prompt fission neutrons agree within +20 %. The relative effect of water, introduced in the central hole of our experimental system, is about 30 % on the backscattered neutron signal, in both simulation and experiment. Finally, the spatial sensitivity of the tool, which is maximal in front of the detector and extends to approximately 20 cm (FWHM) in the vertical axis, is well reproduced with MCNP.

Multiplicity counting using organic scintillators to distinguish neutron sources: An advanced teaching laboratory

In this advanced instructional laboratory, students explore complex detection systems and nondestructive assay techniques used in the field of nuclear physics. After setting up and calibrating a neutron detection system, students carry out timing and energy deposition analyses of radiation signals. Through the timing of prompt fission neutron signals, multiplicity counting is used to carry out a special nuclear material (SNM) nondestructive assay. Our experimental setup is comprised of eight trans-stilbene organic scintillation detectors in a well-counter configuration, and measurements are taken on a spontaneous fission source as well as two (α,n) sources. By comparing each source's measured multiplicity distribution, the resulting measurements of the (α,n) sources can be distinguished from that of the spontaneous fission source. Such comparisons prevent the spoofing, i.e., intentional imitation, of a fission source by an (α,n) neutron source. This instructional laboratory is designed for nuclear engineering and physics students interested in organic scintillators, neutron sources, and nonproliferation radiation measurement techniques.

Experimental Study of the influence of PVC casing thickness and well fluid density on prompt fission neutron uranium logging and its corrections

Prompt fission neutron uranium logging is a direct method for measuring uranium content.This method is not influenced by uranium-radium disequilibrium, and provides an accurate and direct way to obtain the uranium content of the formation. However, in field logging, the results of logging are susceptible to the influence of lithology, well fluid and casing. Thus, it must be corrected. Based on the“epithermal /thermal neutron counting rate ratio”(i.e., NE/T ) theory of uranium quantification, a uranium block model is built to simulate field production boreholes. Meanwhile, the influences of PVC casing thickness, well fluid composition and well fluid density on the decay rate of epithermal neutrons and thermal neutrons and the error correction method are investigated. The experimental results show that NE/T increases linearly with increasing thickness of the PVC casing, and the correction equation is μ PVC = -0.0501 dPVC + 1, with a goodness of fit R 2 = 0.9931. The effect of different well fluid compositions and densities on the decay rate of double neutrons is kaolinite ≥ montmorillonite < chlorite < illite. μ_mud follows a quadratic decay law with increasing well fluid density and the fit R 2 < 0.95. Finally, theaverage corrected uranium quantification error is within 5% of the actual uranium content.

Prompt fission neutron multiplicity in the 239Pu(n, f) reaction and its energy dependence

Measuring prompt fission neutrons to high precision is an experimental challenge, especially for radioactive fissioning nuclides. However, accurate average multiplicities, , and kinetic energy distributions of prompt fission neutrons are essential for fundamental and applied nuclear physics. We present here a recent measurement of the 239Pu (n,f) as a function of the incident-neutron energy, over the range 1-700 MeV. The measurement was performed with a cutting-edge setup and an innovative technique, which allowed to minimize and account for the main sources of bias. An unprecedented precision was therefore achieved. Our data are compared to GEF predictions as well as to evaluated libraries. For the first time, at low energies, the ENDF/B-VIII.0 nuclear data evaluation is validated with an independent measurement and the evaluated uncertainty reduced by up to 60%. This work paves the way to precisely measure prompt fission neutron multiplicities on highly radioactive nuclei.

TALYS calculations of prompt fission observables and independent fission product yields for the neutron-induced fission of ^{235}U

The TALYS nuclear reaction code’s Hauser-Feshbach statistical decay model has been adapted in order to calculate prompt fission neutron and gamma -ray observables by iterating over deexciting fission fragments. Several fission fragment generators such as GEF, HF^{3}D, and SPY were employed to provide TALYS with databases. These databases contain standardized tables with fission fragment yields, mean excitation energies and their widths, and average total kinetic energy, as a function of charge and mass number of primary fission fragments. The resulting calculations, including prompt particle multiplicities, spectra, average energies, and independent fission product yields, were compared with experimental and evaluated data. This work first outlines the new methodology implemented in TALYS and examines the effects of three important parameters on the final evaporation data. Furthermore, the neutron-induced fission of ^{235}U is investigated in detail as a function of incident energy. The results from TALYS, with input from GEF and HF^{3}D, were compared with available experimental data and the results of the stand-alone GEF code. The proposed methodology contributes to an improved capability to model the fission process.

Prompt Photofission Neutron Detection in Depleted Uranium

The detection of prompt photofission neutrons during active interrogation is a strong indication of the presence of special nuclear materials. However, the high-energy photons used for interrogation create a very challenging radiation environment for the detection of prompt fission signatures. These challenges include detector saturation and pulse pile-up. Additionally, there is an elevated neutron background that further challenges the detection of prompt fission neutrons. This background is produced because of (\ensuremath{\gamma}, Xn) photonuclear reactions in the surrounding high-Z materials. Here, we demonstrate the detection of prompt photofission neutrons in these challenging environments. Depleted uranium (DU) and lead targets are interrogated with bremsstrahlung photons produced by a 9-MV electron linear accelerator. Fast neutrons are detected with trans-stilbene organic detectors, and scintillation pulses are analyzed using a previously developed and demonstrated artificial neural network system. We observe a 5 times higher photoneutron count rate when the lead target is replaced with the DU target. Additionally, we observe a difference in the photoneutron light-output distributions of lead and DU. This difference in the measured distributions is due to the difference in the photoneutron-energy spectra; DU photoneutrons are emitted with (\ensuremath{\gamma}, n) and watt-energy spectra, whereas lead photoneutrons are emitted with only the (\ensuremath{\gamma}, n) spectrum.

Manhattan Project 1940s research on the prompt fission neutron spectrum

We describe how the prompt fission neutron spectrum (PFNS) was determined for the Manhattan Project at Los Alamos. Early work before World War II at American and British universities is described, together with theoretical work by Feather at Cambridge and Bethe at Los Alamos. As the Manhattan Project was being planned in 1942, two experiments on natural uranium were commissioned that proved to be influential: 1) An integral experiment at Chicago by Christy and Manley that accurately determined the average PFNS spectrum energy, 2.2 ± 0.2 MeV; 2) Bloch and Staub’s Stanford cyclotron measurement of the PFNS spectrum, which obtained an average energy of 1.70 ± 0.34 MeV. These two papers, previously unavailable outside of Los Alamos, are reproduced in the Supplementary Appendix. From these data, at the beginning of the project in 1943 Serber estimated an average 235U PFNS energy of 2 MeV, and indeed this agrees with today’s best estimate. The challenges facing the scientists involved both the availability of only very small samples of enriched uranium and plutonium targets, and fast neutron detection technologies. During the project, 235U and 239Pu PFNS were measured by Nicodemus and Staub. These also proved to be quite accurate and gave an average spectrum energy of 2 MeV for 235U. [This is not reproduced in the Appendix because it was published after the war in Physical Review 89, 1288 (1953)]. New methods were developed to enable more accurate measurements, and this paper describes how the PFNS was determined surprisingly well by 1945. We end by describing the post-war measurements in the 50s, including the PFNS data used by Ford and Wheeler in their simulations in 1951, the Bonner 1952 data, the seminal 1952 Watt paper with a new empirical parametrization of the PFNS, and the accurate PFNS measurement undertaken at Los Alamos by Cranberg et al. in 1956. We compare the measurements with our best understanding today as embodied in the Evaluated Nuclear Data File ENDF/B-VIII.0. Some images from historical documents in our Los Alamos National Security Research Center (NSRC) archives are shown.

Study of 239Pu(n,f) prompt fission neutron spectrum

The experimental data concerning 239Pu prompt fission neutron spectrum above 10 MeV are sparse, and no spectrum averaged cross section has been measured in this neutron field yet. The first measurement was performed at the IBR-2 reactor, which uses plutonium dioxide fuel. The present work aimed at the measurement of spectral averaged cross sections for high threshold reactions such as 46Ti(n,p)46Sc, 54Fe(n,p)54Mn, 58Ni(n,X)57Co, 89Y(n,2n)88Y, and 209Bi(n,3n)207Bi to explore the high energy region of the 239Pu prompt fission neutron spectrum. Measurement of spectrum averaged cross section for the 89Y(n,2n)88Y reaction is included in NEA high priority request list. The 58Ni(n,p)58Co reaction was employed as a monitor reaction. The experimental spectral averaged cross sections were derived from reaction rates measured by means of high purity Germanium spectrometry. The experimental values were compared with spectrum averaged cross sections compiled from either ENDF/B-VIII.0, GMA or JEFF-3.3 239Pu prompt fission neutron spectrum and IRDFF-II cross sections. The best agreement, i.e. within uncertainties, was achieved using GMA 239Pu prompt fission neutron spectrum for all reactions under study. Calculation results using JEFF-3.3 and ENDF/B-VIII.0 evaluations of the 239Pu prompt fission neutron spectrum agree well with the experimental data except for the high energy region reflected in the disagreement with the 209Bi(n,3n)207Bi reaction.

Shedding light on the 239Pu fission source term with new high-precision experiments and advanced fission modeling

In the last decade, there has been a renaissance of fission research resulting in new high-precision experiments and advanced fission modeling. For instance, the Chi-Nu and CEA teams supplied, for the first time, the 239Pu prompt fission neutron spectrum (PFNS) for broad ranges of incident and outgoing neutron energies. The CEA team also measured 239Pu average prompt neutron multiplicities, ν̄p, with lower statistical uncertainties and a technique significantly different than the one used in the past. The NIFFTE collaboration provided 239Pu(n,f)/235U(n,f) cross section shape ratios with uncertainties below 1% utilizing a novel detector type. Advanced fission event generators were developed, among them CGMF, FIFRELIN, FREYA, and GEF, which calculate post-scission fission observables in a correlated manner. These new experimental data and more consistent fission models change the evaluated PFNS, ν̄p, and (n,f) cross sections, some only modestly, compared to ENDF/B-VIII.0. In turn, the individual new nuclear data distinctly change simulated effective neutron multiplication factors of fast critical assemblies, but their combined impact is small, while affecting the prediction of LLNL pulsed sphere neutron leakage spectra and reaction rates only within experimental uncertainties. Also, the parameters obtained from fitting to ν̄p reproduce various post-scission fission observables within the uncertainties of experimental data. This indicates that new differential experiments and consistent fission modeling reduce compensating errors present in ENDF/B-VIII.0.

Templates of expected measurement uncertainties

The covariance committee of CSEWG (Cross Section Evaluation Working Group) established templates of expected measurement uncertainties for neutron-induced total, (n,γ), neutron-induced charged-particle, and (n,xn) reaction cross sections as well as prompt fission neutron spectra, average prompt and total fission neutron multiplicities, and fission yields. Templates provide a list of what uncertainty sources are expected for each measurement type and observable, and suggest typical ranges of these uncertainties and correlations based on a survey of experimental data, associated literature, and feedback from experimenters. Information needed to faithfully include the experimental data in the nuclear-data evaluation process is also provided. These templates could assist (a) experimenters and EXFOR compilers in delivering more complete uncertainties and measurement information relevant for evaluations of new experimental data, and (b) evaluators in achieving a more comprehensive uncertainty quantification for evaluation purposes. This effort might ultimately lead to more realistic evaluated covariances for nuclear-data applications. In this topical issue, we cover the templates coming out of this CSEWG effort–typically, one observable per paper. This paper here prefaces this topical issue by introducing the concept and mathematical framework of templates, discussing potential use cases, and giving an example of how they can be applied (estimating missing experimental uncertainties of 235U(n,f) average prompt fission neutron multiplicities), and their impact on nuclear-data evaluations.

Producing Evaluation-quality 239Pu Average Prompt Fission Neutron Multiplicities using a Correlated Fission Model

An evaluation of the average prompt fission neutron multiplicity, ν¯ p, of 239Pu(n,f) is shown. This evaluation includes (a) the correlated fission model CGMF, and (b) a detailed analysis of past and recently published experimental data. Using CGMF-calculated ν¯ pas prior enables to link, through the use of evaluated model input parameters, ν¯ p to other fission observables such as the prompt fission neutron spectrum (PFNS), pre-neutron emission fission yields as a function of mass, and the average total kinetic energy of the fragments. These evaluated parameters produce realistic predictions of many fission observables, while the evaluated ν¯ p agrees well (χ2 ≈ 1) with data. Moreover, with the new evaluated ν¯ p, the effective neutron multiplication factor of fast Pu ICSBEP critical assemblies are predicted with a mean bias of 58 pcm compared to 18 pcm with ENDF/B-VIII.0, when paired with a new 239Pu PFNS and fission cross section. Due to these encouraging validation results, the evaluated ν¯ p is currently part of a release candidate for the 239Pu ENDF/B-VIII.1 file. Hence, a correlated fission model was used for the first time for evaluating ν¯ p that is of evaluation quality. This is an important step towards consistent evaluations of prompt fission observables.