Multi-parametric prompt neutron and fission fragment experimental data described by the “Point by Point” model

The experimental multi-parametric data concerning the prompt neutron multiplicity as a function of fission fragment mass and total kinetic energy as well as other experimental data like the total excitation energy of the fission fragment pair versus the fragment mass, the prompt neutron average center-of-mass energy as a function of the fragment mass, the total average multiplicity as a function of the fission fragment total kinetic energy and the average prompt γ-ray energy versus the fragment mass, when they exist, allow much more detailed verification of the models and parameter determination used to evaluate the prompt neutron multiplicity and spectrum data.     

Systematic behaviour of the average parameters required for the Los Alamos model of prompt neutron emission

A large number of fissioning systems have been studied in the frame of the Los Alamos model which showed interesting regular behaviours of the average model parameters (the energy release in fission 〈E_r〉, the total kinetic energy of the fission fragments 〈TKE〉, the average neutron separation energy 〈S_n〉 from the fission fragments, the prompt gamma-ray energy 〈E_γ〉 and the level density parameter 〈a〉 parameterized as 〈C〉 = A/〈a〉, where A is the mass number of the fissioning nucleus)) as well as of other quantities in connection with the prompt fission neutron emission (such as the total average prompt neutron multiplicity at thermal incident energy, the total average fission fragment excitation energy leading to prompt neutron emission, the total average prompt fission energy deposition, the average excitation energy given to the fragments, and the average center-of-mass energy of prompt neutrons and so on) and to elaborate systematics of model parameters.     

Experimental prompt fission neutron “sawtooth” data described by the “point by point” model

The experimental data concerning the prompt fission neutron number as function of the fission fragment mass (currently named “sawtooth” data) when they exist, allow a more refined verification of the “point by point” model and in the same time the validation of the methods used to determine the model parameters corresponding to the fission fragment pairs covering the entire fission fragment range.The prompt neutron multiplicities of each fission fragment pair provided by the “point by point” model are compared with the experimental data concerning the sum of neutrons emitted by the light and heavy fragments forming a pair obtained from experimental sawtooth data.The available fission fragment pair multiplicity and sawtooth experimental data for the ^233,235,238U(n, f) and ²³⁷Np(n, f) reactions as well as for the spontaneous fission of ²⁵²Cf are well described by the “point by point” model results, proving again that this treatment is a powerful tool for prompt fission neutron multiplicity and spectrum evaluation purposes.     

Prompt fission neutron data calculation using experimental fission fragment charge distributions in the frame of the Point by Point treatment

The experimental fission fragment (FF) mass distributions are important ingredients for the Point by Point (PbP) model calculation of multiplicity, spectra and other quantities related to the prompt neutron emission. For many fissioning systems experimental FF mass distributions are missing and in these cases an original procedure, exploiting available experimental FF charge distributions in the frame of PbP treatment, is proposed.     

Point Scattered Function (PScF) for fast neutron radiography

Fast neutron radiography opened up a new range of possibilities to image extremely dense objects. The removal of the scattering effect is one of the most challenging problems in neutron imaging. Neutron scattering in fast neutron radiography did not receive much attention compared with X-ray and thermal neutron radiography. The purpose of this work is to investigate the behavior of the Point Scattered Function (PScF) as applied in fast neutron radiography.The PScF was calculated using MCNP as a spatial distribution of scattered neutrons over the detector surface for one emitting source element. Armament and explosives materials, namely, Rifle steel, brass, aluminum and trinitrotoluene (TNT) were simulated. Effect of various sample thickness and sample-to-detector distance were considered. Simulated sample geometries included a slab with varying thickness, a sphere with varying radii, and a cylinder with varying base radii. Different neutron sources, namely, Cf-252, DT as well as DD neutron sources were considered. Neutron beams with zero degree divergence angle; and beams with varying angles related to the normal to the source plane were simulated.Curve fitting of the obtained PScF, in the form of Gaussian function, were given to be used in future work using image restoration codes. Analytical representation of the height as well as the Full Width at Half Maximum (FWHM) of the obtained Gaussian functions eliminates the need to calculate the PScF for sample parameters that were not investigated in this study.     

Prompt fission neutron spectrum evaluation for 252Cf(SF) in the frame of the multi-modal fission model

In the present work, an attempt to improve the evaluation of the prompt fission neutron spectrum of ²⁵²Cf(SF) is made. The multi-modal fission concept is included into the Los Alamos model. A more generalized form of the fission fragment residual nuclear temperature distribution and a possible anisotropy effect of the prompt neutron emission in the center-of-mass system are taken into account, too. The multi-modal fission parameters entering the prompt fission neutron spectrum model are determined on the basis of the experimental data concerning the fission fragment total kinetic energy TKE(A) and mass distribution Y(A) measured at IRMM. The calculated prompt neutron spectrum is obtained in better agreement with the standard point-wise evaluation of Mannhart and compared to other evaluations made with different models.     

Xenon poisoning calculation code for miniature neutron source reactor（MNSR）

In line with the actual requirents and based upon the specific characteristics of MNSR,a revised point-reactor model was adopted to model MNSR‘s xenon poisoning.The corresponding calculation code.MNSRXPCC(Xenon Poisoning Calculation Code for MNSR),was developed and tested by the Shanghai MNSR data.     

The optimization of gamma spectra processing in prompt gamma neutron activation analysis (PGNAA)

The uncertainty of the elemental analysis is one of the major factors governing the utility of on-line Prompt Gamma Neutron Activation Analysis (PGNAA) in the blending and sorting of bulk materials. In this paper, a general method applicable to Gamma spectra processing is presented and applied to PGNAA in mineral industry. Based on the Fourier transform of spectra and their de-correlation in the Fourier space (the improvement of the conditioning of the correlation matrix), processing of overlapping of characteristic peaks minimizes the propagation of random errors, which optimizes the accuracy and decreases the detection limits of elemental analyses. In comparison with classical methods based on the linear combinations of relevant regions of spectra the improvement may be considerable, especially when several elements are interfering. The method is applied to four case stories covering both borehole logging and on-line analysis on conveyor belt of raw materials.     

Status of delayed-neutron precursor data: half-lives and neutron emission probabilities

We present in this paper a compilation of the present status of experimental delayed-neutron precursor data; i.e. β-decay half-lives (T_1/2) and neutron emission probabilities (P_n) in the fission-product region (27 ≤ Z ≤ 57). These data are compared to two model predictions of substantially different sophistication: (i) an update of the empirical Kratz-Herrmann formula (KHF), and (ii) a unified macroscopic-microscopic model within the quasi-particle random-phase approximation (QRPA). Both models are also used to calculate so far unknown T_1/2 and P_n values up to Z = 63. A number of possible refinements in the microscopic calculations are suggested to further improve the nuclear-physics foundation of these data for reactor and astrophysical applications.     

A benchmarked MCNP model of the in vivo detection of gadolinium by prompt gamma neutron activation analysis

Gadolinium (Gd)-based contrast agents are a valuable diagnostic aid for magnetic resonance imaging (MRI). The amount of free Gd deposited in tissues following contrast enhanced MRI is of toxicological concern. The McMaster University in vivo prompt gamma neutron activation analysis facility has been adapted for the detection of Gd in the kidney, liver, and the leg muscle. A simple model of the HPGe detector used for detection of the prompt γ-rays following Gd neutron capture has been created using Monte Carlo simulation. A separate simulation describing the neutron collimation and shielding apparatus has been modified to determine the neutron capture rate in the Gd phantoms. The MCNP simulation results have been confirmed by experimental measurement. The deviations between MCNP and the experiment were between 1% and 18%, with an average deviation of 3.8 ± 6.7%. The validated MCNP model is to be used to improve the Gd in vivo measurement sensitivity by determining the best neutron moderator/reflector arrangement.     

Phantom studies of Cd, Hg and Cl by prompt gamma neutron activation analysis using a Pu-Be neutron source

Prompt gamma neutron activation analysis is a means of non-invasive monitoring for occupational exposure to toxic heavy metals such as Cd and Hg. Preliminary kidney detection limits from previous phantom studies at McMaster were 13.6 ± 0.2 ppm for Cd (125 mL phantom) and 315 ± 24 ppm for Hg (125 mL phantom) using the ²³⁸Pu-Be neutron source and 0.88 ± 0.01 ppm for Cd (125 mL phantom) and 16.91 ± 0.05 ppm for Hg (30 mL phantom) using the thermal neutron beam port at the McMaster Nuclear Reactor. The detection limits vary greatly between the two methods due to differences in experimental set-up, neutron energy spectra and a difference in dose by more than a factor of 100. The Hg detection limit from preliminary data is much higher than expected for both neutron source types. In order to explain the apparent detection limit discrepancy, measurements of Hg and Cd phantoms were performed using the ²³⁸Pu-Be neutron source. The results were compared to phantom measurements of Cl, a well-known neutron activation element.     

Approach for the fission fragment total kinetic energy TKE(A) calculation: Application to prompt neutron emission models

A simple approach for the calculation of the fission fragment total kinetic energy, TKE(A), based on the electrostatic repulsion between the fragments connected by a neck in the pre-scission configuration is described. The calculated TKE(A) is obtained in good agreement with the experimental data for many fissioning systems, such as ^233,235U(n_th, f), ²³⁹Pu(n_th, f), ²³⁷Np(n, f), ²⁴²Pu(SF), with minor adjustment of only one parameter. Due to the fact that the present approach can provide with enough trust TKE(A) distributions for fissioning systems for which experimental TKE(A) data do not exist, the possibilities to use the refined Point by Point model of prompt neutron emission can be considerably extended.     

Fast neutron inspection of sea containers for the presence of “dirty bomb”

The possibility of the detection of “dirty bomb” presence inside sea containers is evaluated. The method proposed for explosive and fissile material detection makes use of two sensors (X-rays and neutrons). A commercial imaging device based on the X-ray radiography performs a fast scan of the container, identifies a “suspect” region and provides its coordinates to the neutron based device for the final “confirmatory” inspection. In this two sensor system a 14 MeV neutron beam defined by the detection of associated alpha particles is used for interrogation of only volume elements marked by X-ray sensor. The object’s nature is determined from passive and neutron induced, gamma energy spectra measurements. Experimental results (time-of-flight and gamma energy spectra) obtained for the irradiation 30 kg of TNT, depleted uranium and other materials hidden inside the container are presented.     

Characterization of atom and ion-induced “internal” electron emission by thin film tunnel junctions

Highly excited charge carriers are released when single or multiply charged ions impinge on metal surfaces. While electron emission into the adjacent vacuum phase is well investigated, one has only limited knowledge about the transport of excited electrons or holes into the bulk of a metal. This shortcoming can be reduced by studying the transport of these excited carriers over buried tunnel barriers in thin film metal-insulator-metal devices. The internal barriers can be tuned by a tunnel voltage which makes the device to a balance for excited electrons and holes. With a simple theoretical model we derive the balance function of different tunnel barriers and show their feasibility for the characterization of particle induced electronic excitations on metal surfaces.     

The investigation of the neutron emission spectra of Th and U for neutron incident energy from 2 to 18 MeV

In this study, experimental neutron-emission spectra produced by (n, xn) reactions on nuclei ²³²Th have been compared with experimental ²³⁸U(n, xn) neutron-emission spectra from 2 to 18 MeV. Angle-integrated cross-sections in neutron induced reactions on targets ²³⁸U have been calculated at the bombarding energies from 2 to 18 MeV. In the calculations, the geometry dependent hybrid model and the cascade exciton model including the effects of pre-equilibrium have been used. In addition, we have described how multiple pre-equilibrium emissions can be included in the Feshbach–Kerman-Koonin (FKK) fully quantum-mechanical theory. By analyzing (n, xn) reaction on ²³⁸U with the incident energy from 2 to 18 MeV, the importance of multiple pre-equilibrium emission can be seen clearly. All calculated results have been compared with experimental data. The obtained results have been discussed and compared with the available experimental data and found agreement with each other.     

Structure of mixed U(IV)-An(III) precursors synthesized by co-conversion methods (where An = Pu, Am or Cm)

Current concepts for future nuclear systems aim at improving the fuel cycle with the main following criteria: economy of resources, minimized volume and lower long-term potential radiotoxicity of ultimate wastes and proliferation risk reduction. Co-management of two (or more) actinides has recently been proposed for recycling reusable energy-producing actinides (mainly U and Pu) together, or for transmuting radiotoxic minor actinides within UO₂-based materials. Co-conversion processes play an important role by closing the actinide separation-purification operations and at the same time producing mixed actinide solid compounds for the fabrication of fresh fuel. Handling of actinides mixtures, from the initial solution up to the solid product, requires innovative synthesis methods and structures, particularly for the minor actinides such as americium and curium. Considering the different designs of future nuclear fuels, various uranium-actinide co-conversion routes are currently investigated in the CEA-ATALANTE facility.     

An evaluation of neutron and gamma-ray-production cross-section data for lead

A survey was made of the available information on neutron and gamma-ray-production cross-section measurements of lead. From these and from relevant nuclear-structure information on the Pb isotopes, we prepared recommended neutron cross-section data sets for lead covering the neutron energy range from 0.00001 eV to 20.0 MeV. The cross sections are derived from experimental results available to February 1972 and from calculations based on optical-model, DWBA, and Hauser-Feshbach theories. Comparisons which show good agreement between theoretical and experimental values are displayed in a number of graphs. Also presented graphically are smoothed total cross sections, Legendre coefficients for angular distributions, and a representative energy distribution of gamma rays from resonance capture.     

The characterization of legacy radioactive materials by gamma spectroscopy and prompt gamma activation analysis (PGAA)

To characterize legacy radioactive materials, it is necessary to determine both the radioactive and, in the case of carrier-based materials, the stable, non-radioactive chemical constituents. Reputable process knowledge may afford some insight but, absent such information, gamma spectroscopy and (non-destructive) prompt gamma activation analysis (PGAA) cover essentially all of the analytical needs, with the former addressing most radionuclides with the exception of the pure β⁻-emitters and the latter addressing the stable chemical constituents. This paper integrates both methods into a general analytical protocol based upon radioanalytical work performed at Lawrence Berkeley National Laboratory (LBNL) and PGAA work performed collaboratively by the various groups. A new LBNL-developed neutron generator is also discussed.