Kelly A. Jordan

Analysis of the scintillation mechanism in a pressurized 4He fast neutron detector using pulse shape fitting

An empirical investigation of the scintillation mechanism in a pressurized 4He gas fast neutron detector was conducted using pulse shape fitting. Scintillation signals from neutron interactions were measured and averaged to produce a single generic neutron pulse shape from both a 252Cf spontaneous fission source and a (d,d) neutron generator. An expression for light output over time was then developed by treating the decay of helium excited states in the same manner as the decay of radioactive isotopes. This pulse shape expression was fitted to the measured neutron pulse shape using a least-squares optimization algorithm, allowing an empirical analysis of the mechanism of scintillation inside the 4He detector. A further understanding of this mechanism in the 4He detector will advance the use of this system as a neutron spectrometer. For 252Cf neutrons, the triplet and singlet time constants were found to be 970 ns and 686 ns, respectively. For neutrons from the (d,d) generator, the time constants were found to be 884 ns and 636 ns. Differences were noted in the magnitude of these parameters compared to previously published data, however the general relationships were noted to be the same and checked with expected trends from theory. Of the excited helium states produced from a 252Cf neutron interaction, 76% were found to be born as triplet states, similar to the result from the neutron generator of 71%. The two sources yielded similar pulse shapes despite having very different neutron energy spectra, validating the robustness of the fits across various neutron energies.

On the combination of delayed neutron and delayed gamma techniques for fission rate measurement in nuclear fuel

Novel techniques to measure newly induced fissions in spent fuel after re-irradiation at low power have been developed and tested at the Proteus zero-power research reactor. The two techniques are based on the detection of high energy gamma-rays emitted by short-lived fission products and delayed neutrons. The two techniques relate the measured signals to the total fission rate, the isotopic composition of the fuel, and nuclear data. They can be combined to derive better estimates on each of these parameters. This has potential for improvement in many areas. Spent fuel characterisation and safeguard applications can benefit from these techniques for non-destructive assay of plutonium content. Another application of choice is the reduction of uncertainties on nuclear data. As a first application of the combination of the delayed neutron and gamma measurement techniques, this paper shows how to reduce the uncertainties on the relative abundances of the longest delayed neutron group for thermal fissions in 235U, 239Pu and fast fissions in 238U. The proposed experiments are easily achievable in zero-power research reactors using fresh UO2 and MOX fuel and do not require fast extraction systems. The relative uncertainties (1σ) on the relative abundances are expected to be reduced from 13% to 4%, 16% to 5%, and 38% to 12% for 235U, 238U and 239Pu, respectively.

Analysis for in-situ fission rate measurements using 4 He gas scintillation detectors

Active neutron interrogation is a powerful NDA technique that relies on detecting and analyzing fission neutrons produced in a fuel sample by an interrogating high neutron flux. 4He scintillation gas fast neutron detectors are investigated in this paper for use in a novel fission rate measurement technique The He-4 detectors have excellent gamma rejection, a fast response time, and give significant information on incident neutron energy allowing for energy cuts to be applied to the detected signal. These features are shown in this work to allow for the detection of prompt fission neutrons in-situ during active neutron interrogation of a 238U sample. The energy spectrum from three different neutrons sources (252Cf, AmBe, AmLi) is measured using the 4He detection system and analyzed. An initial response matrix for the detector is determined using these measurements and the kinematic interaction properties of the elastic scattering with the 4He.

Modification of SCWR Assembly Designs with Coupled MCNP5/SCA Methods to Promote Safer Operation

Abstract This work concerns the comparison of supercritical water reactor (SCWR) assembly designs using coupled reactor physics and thermal-hydraulic methods. In the SCWR, large density gradients in the supercritical water (used as coolant and moderator) will require detailed multiphysics analysis. The Super Light Water Reactor (SLWR) was analyzed previously [Hughes et al., Nucl. Eng. Des., Vol. 270 (2014)], where MCNP5 was coupled with density and temperature results from a single-channel code. MCNP5 then provided the single-channel code with a linear heat profile. In the present work, that proposed assembly design is determined to have a negative density coefficient of reactivity. Two alternate designs with different geometries and water-to-fuel ratios are presently considered to address this issue. It is found that adding an additional row of pins is more effective at producing a positive density coefficient than is reducing the size of the moderator boxes.

Coupled Computational Heat Transfer and Reactor Physics for SCWR

To fully model the physics present within the proposed supercritical water reactor (SCWR), the thermal hydraulics calculations (yielding temperatures and densities in each material as a function of space) must be coupled to the neutronic calculations (yielding reactivity and neutron flux shapes). To enable this full coupling, a 3D model of a supercritical water reactor is being implemented in the CFD software OpenFOAM with the same geometry as a 3D MCNP (neutronics) model. Coupling will performed through result exchange between the two codes — densities and temperatures from OpenFOAM to MCNP, with heat generation returned from the neutronics calculations. Use of a reduced-geometry model is advisable due to the high computational cost of each OpenFOAM/MCNP coupling iteration. In the present work, a 1.5D thermal model of a single fuel pin was coupled with a 3D MCNP model. The thermal model includes single channel analysis (cladding/coolant heat transfer) as well as heat transfer within the cladding, helium gas gap, and uranium dioxide fuel itself. These heat transfer zones provide specific data points of the radial temperature profile. Because no radial mesh is considered, full radial dependence is not possible. The model provides limited radial dependence unlike what a 1D code could provide; thus, 1.5D is used to indicate the incomplete radial dependence that is included in the model. Iteration between the two codes is performed until heat generation is converged to within 10% between successful MCNP results. A discussion of these 3D/1.5D coupled results and the path forward to fully 3D coupling is provided.Copyright

Analysis of Pin Removal Experiments Conducted in a Supercritical Light Water Reactor-Like Test Lattice

A comprehensive program of integral experiments, largely based on the measurement of reaction rate distributions, was carried out recently in the PROTEUS zero-power research reactor at the Paul Scherrer Institute in Switzerland, employing a fuel lattice resembling that of a supercritical light water reactor. The present paper reports on the analysis of a complementary set of measurements, in which the reactivity effects of removing individual pins from the unperturbed, heterogeneously moderated reference lattice were investigated. It has been found that the detailed Monte Carlo modeling of the whole reactor using MCNPX is able - as in the case of the reaction rate distributions - to reproduce the experimental results for the pin removal worths within the achievable statistical accuracy. A comparison of reduced-geometry calculations between MCNPX and the deterministic light water reactor assembly code CASMO-4E has revealed certain discrepancies. On the basis of a reactivity decomposition analysis of the differences between the codes, it has been suggested that these could be due at least partly to CASMO-4E deficiencies in calculating the effect, upon pin removal, of the extra moderation in the neighboring fuel pins.

SU-C-209-05: Monte Carlo Model of a Prototype Backscatter X-Ray (BSX) Imager for Projective and Selective Object-Plane Imaging

PURPOSE To develop a Monte Carlo N-Particle (MCNP) model for the validation of a prototype backscatter x-ray (BSX) imager, and optimization of BSX technology for medical applications, including selective object-plane imaging. METHODS BSX is an emerging technology that represents an alternative to conventional computed tomography (CT) and projective digital radiography (DR). It employs detectors located on the same side as the incident x-ray source, making use of backscatter and avoiding ring geometry to enclose the imaging object. Current BSX imagers suffer from low spatial resolution. A MCNP model was designed to replicate a BSX prototype used for flaw detection in industrial materials. This prototype consisted of a 1.5mm diameter 60kVp pencil beam surrounded by a ring of four 5.0cm diameter NaI scintillation detectors. The imaging phantom consisted of a 2.9cm thick aluminum plate with five 0.6cm diameter holes drilled halfway. The experimental image was created using a raster scanning motion (in 1.5mm increments). RESULTS A qualitative comparison between the physical and simulated images showed very good agreement with 1.5mm spatial resolution in plane perpendicular to incident x-ray beam. The MCNP model developed the concept of radiography by selective plane detection (RSPD) for BSX, whereby specific object planes can be imaged by varying kVp. 10keV increments in mean x-ray energy yielded 4mm thick slice resolution in the phantom. Image resolution in the MCNP model can be further increased by increasing the number of detectors, and decreasing raster step size. CONCLUSION MCNP modelling was used to validate a prototype BSX imager and introduce the RSPD concept, allowing for selective object-plane imaging. There was very good visual agreement between the experimental and MCNP imaging. Beyond optimizing system parameters for the existing prototype, new geometries can be investigated for volumetric image acquisition in medical applications. This material is based upon work supported under an Integrated University Program Graduate Fellowship sponsored by the Department of Energy Office of Nuclear Energy.

Measurement of the Fast Neutron Response for

Time-of-flight (TOF) and coincident scattering measurements were conducted to measure the light response of a pressurized 4He fast neutron scintillation detector as a function of deposited energy up to 5 MeV. The energy deposited in the detector by a neutron was measured by its angle of scatter and compared to the resulting light output. Whereas previous research has exclusively focussed on the energy information contained in the slow component, this work demonstrates that the fast component is also sensitive to neutron energy, and the entire scintillation signal can therefore be used. The gamma rejection capability of the detector was also measured for a variety of gamma sources. The detector demonstrated a inherent gamma rejection rate of 97.31%, which was increased to 99.89% after the application of pulse shape discrimination (PSD) algorithms. The characterization of gamma rejection and light response parameters will enable implementation of these detectors for neutron spectroscopy in mixed radiation fields.

^4 {\rm He}

The spatial response of pressurized helium-4 fast neutron scintillation detectors is characterized using collimated neutron source measurements and MCNPX-PoliMi simulations. A method for localizing the position of each detected event is also demonstrated using the two-sided photomultiplier readout. Results show that the position of particle interaction along the axis of the active volume has a measurable effect on the scintillation light response of the detector. An algorithm is presented that uses the probability distribution of relative interaction positions to perform source localization, further demonstrating the applicability of these detectors as tools for the detector of hidden shielded nuclear material.

Scintillation Detectors Using a Coincidence Scattering Method

Novel techniques to measure induced fissions in spent fuel after re-irradiation at low power have been developed and tested at the Proteus zero-power research reactor. The two techniques are based on the detection of high energy gamma-rays emitted by short-lived fission products and delayed neutrons.