McMaster Nuclear Reactor Research Articles

CFD Simulations of a Loss of Heat Sink Experiment in the McMaster Nuclear Reactor

The McMaster nuclear reactor (MNR) is an important research facility that not only provides researchers with neutrons for fundamental science, but also supplies the medical industry with isotopes used for cancer treatment. To ensure the safety and performance of the MNR, modeling of the thermal hydraulics during nominal and accidental conditions is required. For such a task, system codes are customarily used. While system codes can assess the safety aspects of complex thermal–hydraulic systems, the question arises whether such systems can be modeled appropriately in a three-dimensional manner, such as computation fluid dynamics (CFD), while capturing the thermal mixing of the coolant. Modeling the thermal hydraulics of nuclear research reactors using CFD is relatively new. Therefore, validating such methods against experimental data is of utmost importance. The validation of CFD is the main focus of this work. More specifically, the influence of the loss of heat sink on the pool temperature is assessed using the CFD code Ansys CFX. The validation basis was provided by an experiment performed at the MNR in March 2023. In this experiment, the secondary heat removal system was intentionally shut down, and the pool temperature was measured in a few positions. The results obtained by modeling the loss of heat sink in the MNR using Ansys CFX agree well with the experiment.

A Study of Gamma Heating, Neutron Flux, and Gamma Flux Under Several Reactor Core Conditions in the McMaster Nuclear Reactor

Changes in the thermal power of a nuclear research reactor will lead to changes in experimental, irradiation, and testing conditions. Consequently, reactor core parameters are inevitably susceptible to changes. One such parameter is gamma heating (GH), which results from gamma interaction with materials. In this work, a gamma thermometer was used to measure GH over the course of 7 operational days and nights. In addition, the Monte Carlo reactor physics code Serpent-2 was used to evaluate the sensitivity of common detection methods for monitoring reactor core parameters such as neutron fluxes, GH, and gamma flux under the following conditions: reactor core power variation, reactor core fuel shuffling, and detector vicinity fuel assembly shuffling. The GH values obtained through measurements and calculations were linearly proportional to the reactor power. In addition, the Serpent-2 code for the McMaster nuclear reactor showed that despite maintaining the reactor power core at the same level, the fuel burnup distribution could alter the studied parameters.

Feasibility of gadolinium oxide paint as neutron shielding

A robust Gd2O3 based thermal neutron shielding paint has been produced to act as the neutron attenuation medium to augment radiation shielding and to produce simple low resolution neutron imaging resolution standards. The neutron attenuation of this paint has been studied using a neutron radiographical imaging beamline at the McMaster Nuclear Reactor, and the McMaster Alignment Diffractometer neutron beam instrument. The data indicates a thermal neutron transmission factor of 0.457±0.002 and a mass attenuation coefficient of 89±2gcm2 for a single coat of paint.

On-line measurement and simulation of the in-core gamma energy deposition in the McMaster nuclear reactor

In a nuclear reactor, gamma radiation is the dominant energy deposition in non-fuel regions. Heat is generated upon gamma deposition and consequently affects the mechanical and thermal structure of the material. Therefore, the safety of samples should be carefully considered so that their integrity and quality can be retained. To evaluate relevant parameters, an in-core gamma thermometer (GT) was used to measure gamma heating (GH) throughout the operation of the McMaster nuclear reactor (MNR) at four irradiation sites. Additionally, a Monte Carlo reactor physics code (Serpent-2) was utilized to model the MNR with the GT located in the same irradiation sites used in the measurement to verify its predictions against measured GH. This research aids in the development of modeling, calculation, and prediction of the GH utilizing Serpent-2 as well as implementing a new GH measurement at the MNR core. After all uncertainties were quantified for both approaches, comparable GH profiles were observed between the measurements and calculations. In addition, the GH values found in the four sites represent a strong level of radiation based on the distance of the sample from the core. In this study, the maximum and minimum GH values were found at 0.32 ± 0.05 W/g and 0.15 ± 0.02 W/g, respectively, corresponding to 320 Sv/s and 150 Sv/s. These values are crucial to be considered whenever sample is planned to be irradiated inside the MNR core.

Measurement, simulation and uncertainty quantification of the neutron flux at the McMaster Nuclear Reactor

Neutron flux measurements in research reactors can be used for code validation and optimizing in-core activation procedures. Since the fuel adjacent to an irradiation site undergoes burnup, and may be shuffled, local flux measurements may be subject to an additional source of burnup-dependent uncertainty. It is unfeasible to perform these measurements for all core conditions; therefore, reactor physics codes may provide supplemental flux information.This work includes a validation study of the MCNP6 model of the McMaster Nuclear Reactor (MNR). Irradiations were performed over several months, with an emphasis on uncertainty quantification during data processing. No change in the local flux was measured over this period of operation, indicating that burnup effects may be insignificant compared to other sources of uncertainty. These results were validated by five sets of computational data from historical MNR cores. Burnup effects do not need to be accounted for in determining neutron flux uncertainties.

Simulation approach and code functionality assessment using deterministic and Monte Carlo codes system for U-235 core-follow depletion calculation

Improving the performance of reactor simulation codes and adequately predicting the fuel composition are crucial to increase the accuracy of simulations and the quality of the analysis. McMaster Nuclear Reactor (MNR) follows a strict daily operational and shutdown schedule, except for Sundays. This results in a buildup of neutron poison during the reactor shutdown and hence the Control Rods (CRs) are extracted in the time of the reactor startup. Therefore, tracking MNR core composition using faithful Control Rods (CRs) positions data to estimate the axial fuel inventory profile is not yet established. In this study, nodal diffusion (OSCAR-4) and Monte Carlo (Serpent-2) simulations codes were employed. This study investigates: (1) code-to-code differences in the multiplication factor (keff) and in the axial fuel inventory considering the same burnup step sizes and total energy release with all CRs out; (2) for long cycles of core calculation, Monte Carlo stochastic code is computationally expensive, therefore an adequate temporal discretization steps are studied and identified. The keff values for the fresh reactor when all rods are out is: 1.1339 and 1.1283, and the results at the end of the burnup calculation is: 1.0101 and 1.0049, for Serpent-2 and OSCAR-4, respectively. In addition, the findings showed that averaging the CRs travel distance, for calculation the fuel inventory, can provide results similar to those with most tracked CRs motion.

OSCAR-4 CODE SYSTEM COMPARISON AND ANALYSIS WITH A FIRST-ORDER SEMI-EMPIRICAL METHOD FOR CORE-FOLLOW DEPLETION CALCULATION IN MNR

Knowledge of the isotopic composition of a nuclear reactor core is important for accurate core-follow and reload analysis. In the McMaster Nuclear Reactor, fuel depletion estimates are based upon a semi-empirical calculation using flux-wire measurements. These estimates are used to plan and guide fuelling operations. To further support operations, an OSCAR-4 model is being developed. To evaluate the performance of the OSCAR-4 code for this application, 2 points of comparison, considering the period between 2007 and 2010, are presented: (i) the multiplication factor keff and (ii) U-235 fuel inventory. The latter is compared with a simple first-order semi-empirical calculation. The calculation of keff for the last operational 3 months yields 0.997 ± 0.002 (vs. 1.000 for an operating reactor), and differences in both core-average inventory and the maximum standard fuel assembly inventories estimates are found to be 5.7% and 7.5%, respectively.

Serpent-2 and OSCAR-4 computational tools compared against McMaster nuclear reactor improved operational data history for U-235 fuel inventory tracking, local power tracking and validation of multiplication factor

The tracking of fuel inventory is important for safety and for knowledge of the flux distribution. In a research reactor, tracking of the reactor core fuel composition using simulation code is crucial for fuel management strategy, such as by optimizing the core pattern for sample irradiation purposes and for providing an economical fuel cycle length. Operational data history for the McMaster Nuclear Reactor (MNR) is a valuable asset for comparing and validating simulation code. A previous study examined U-235 content using the nodal diffusion code Overall System for CAlculation of Reactors (OSCAR-4) against standard MNR operational data and recommended considering the Pu-239 content in the energy production in the standard operational data. In this work, a new method was implemented by introducing a Fuel Inventory Correction (FIC) factor for improving the U-235 operational data records. U-238 and Pu-239 code-to-code tracking comparisons were also employed. The maximum difference between U-235 at end of life (EOL) and the updated operational data history was 2.43% and 6.97% for Serpent 2 and OSCAR-4, respectively. The FA power profile tracking showed a similar response whenever fuel shuffling and/or refueling occurred, with a maximum different of 16% (30 kW) in one cycle. The multiplication factors for the two sets of code differed systematically, averaging 0.9982 and 1.0051 for Serpent 2 and OSCAR-4, respectively.

Experimental measurement and Monte Carlo code simulation of the gamma heating at different irradiation sites in a nuclear research reactor

Interaction of radiation with matter results in heat generation and in effect, changes in the irradiation sample’s quality. Knowledge about heat deposition within the irradiation sites can go a long way in enhancing the safety and quality of the irradiation condition. To measure such results, SCK-CEN Gamma Thermometer (GT) is used for experimentally measuring Gamma Heating (GH) during the operation of McMaster Nuclear Reactor (MNR) at three irradiation sites. In addition, Monte Carlo reactor physics code (Serpent-2) is also used for modeling the MNR in order to compare as well as validate it against experimental and operational measured data such as the multiplication factor (keff) during the operational day against the actual values (unity), the experimentally measured GH values. After finding a difference of 0.45% at the beginning of the day and 0.32% at the end of the day in the keff, a comparable GH profile was observed between the measurement and the calculation. Additionally, the average axial differences of GH results found in the beryllium irradiation site (2A), and the graphite irradiation sites (8B and 8E) are 4.71%, 10.71%, and 11.72%.

Quantification of system uncertainties in activation experiments at nuclear research reactors

Neutron flux measurements in reactor cores are often used for code validation studies, however reported experimental flux uncertainties remain limited in scope. This work presents both a detailed assessment of several flux measurement uncertainties and a guideline for reporting these quantities. The effects of the sample position, the presence of 135Xe, the control rod positions, reactor power, and nuclear data during activation procedures were individually quantified at the McMaster Nuclear Reactor. NiCr flux wires were used to study the 51Cr (n,γ) activation reaction experimentally and through a series of MCNP6 simulations, and the SCALE-6.2.2 252-group library was used to study cross-section data uncertainty for use in translating activation measurements into neutron flux values. The placement of the sample and the reactor power measurements were found to be the most significant contributors to the overall measurement uncertainty. A procedure for reporting the results and uncertainties of an irradiation procedure is provided.

Practical aspects of 153Gd as a radioactive source for use in brachytherapy

The goal of this study was to investigate the production, purification and immobilization techniques for a 153Gd brachytherapy source. We have investigated the maximum attainable specific activity of 153Gd through the irradiation of Gd2O3 enriched to 30.6% 152Gd at McMaster Nuclear Reactor. The advantage of producing 153Gd through this production pathway is the possibility to irradiate pre-sealed pellets of 152Gd enriched Gd2O3, thereby removing the need to perform chemical separation with large quantities of radio-impurities. However, small amounts of long-lived impurities are produced from the irradiation of enriched 152Gd targets due to traces of Eu in the sample. If the amount of impurities produced is deemed unacceptable, 153Gd can be isolated as an aqueous solution, chemically separated from impurities and loaded onto a sorbent with a high affinity for Gd before encapsulation.

Ultra-Violet Light Emission from HPV-G Cells Irradiated with Low Let Radiation from 90Y; Consequences for Radiation Induced Bystander Effects

In this study, we aimed to establish the emission of UV photons when HPV-G cells and associated materials (such as the cell substrate and cell growth media) are exposed to low LET radiation. The potential role of UV photons in the secondary triggering of biological processes led us to hypothesize that the emission and absorption of photons at this wavelength explain some radiation induced "bystander effects" that have previously been thought to be chemically mediated. Cells were plated in Petri-dishes of two different sizes, having different thicknesses of polystyrene (PS) substrate, and were exposed to β-radiation from (90)Y produced by the McMaster Nuclear Reactor. UV measurements were performed using a single photon counting system employing an interference-type filter for selection of a narrow wavelength range, 340±5 nm, of photons. Exposure of the cell substrates (which were made of polystyrene) determined that UV photons were being emitted as a consequence of β particle irradiation of the Petri-dishes. For a tightly collimated β-particle beam exposure, we observed 167 photons in the detector per unit μCi in the shielded source for a 1.76 mm thick substrate and 158 photons/μCi for a 0.878 mm thick substrate. A unit μCi source activity was equivalent to an exposure to the substrate of 18 β-particles/cm(2) in this case. The presence of cells and medium in a Petri-dish was found to significantly increase (up to a maximum of 250%) the measured number of photons in a narrow band of wavelengths of 340±5 nm (i.e. UVA) as compared to the signal from an empty control Petri-dish. When coloured growth medium was added to the cells, it reduced the measured count rate, while the addition of transparent medium in equal volume increased the count rate, compared to cells alone. We attribute this to the fact that emission, scattering and absorption of light by cells and media are all variables in the experiment. Under collimated irradiation conditions, it was observed that increasing cell density in medium of fixed volume resulted in a decrease in the observed light output. This followed a roughly exponential decline. We suggest that this may be due to increased scattering at the cell boundary and absorption of the UV in the cells. We conclude that we have measured UVA emitted by cells, cell medium and cell substrates as a consequence of their irradiation by low LET β-particle radiation. We suggest that these secondary UV photons could lead to effects in non-targetted cells. Some effects that had previously been attributed to a chemically mediated "bystander effect" may in fact be due to secondary UV emission. Some radiation bystander effect studies may require re-interpretation as this phenomenon of UV emission is further investigated.

Neutron spectrometry and dosimetry study at two research nuclear reactors using Bonner sphere spectrometer (BSS), rotational spectrometer (ROSPEC) and cylindrical nested neutron spectrometer (NNS)

Neutron spectrometry and subsequent dosimetry measurements were undertaken at the McMaster Nuclear Reactor (MNR) and AECL Chalk River National Research Universal (NRU) Reactor. The instruments used were a Bonner sphere spectrometer (BSS), a cylindrical nested neutron spectrometer (NNS) and a commercially available rotational proton recoil spectrometer. The purposes of these measurements were to: (1) compare the results obtained by three different neutron measuring instruments and (2) quantify neutron fields of interest. The results showed vastly different neutron spectral shapes for the two different reactors. This is not surprising, considering the type of the reactors and the locations where the measurements were performed. MNR is a heavily shielded light water moderated reactor, while NRU is a heavy water moderated reactor. The measurements at MNR were taken at the base of the reactor pool, where a large amount of water and concrete shielding is present, while measurements at NRU were taken at the top of the reactor (TOR) plate, where there is only heavy water and steel between the reactor core and the measuring instrument. As a result, a large component of the thermal neutron fluence was measured at MNR, while a negligible amount of thermal neutrons was measured at NRU. The neutron ambient dose rates at NRU TOR were measured to be between 0.03 and 0.06 mSv h⁻¹, while at MNR, these values were between 0.07 and 2.8 mSv h⁻¹ inside the beam port and <0.2 mSv h⁻¹ between two operating beam ports. The conservative uncertainty of these values is 15 %. The conservative uncertainty of the measured integral neutron fluence is 5 %. It was also found that BSS over-responded slightly due to a non-calibrated response matrix.

The comparison of two MCNP models used for prompt gamma in vivo detection of cadmium and mercury

In vivo detection of trace elements is one of the most important research areas at the Medical Physics and Applied Radiation Sciences Department at McMaster University. Prompt gamma neutron activation analysis (PGNAA) used for detection of cadmium and mercury takes place simultaneously at two different experimental sites; the McMaster Nuclear Reactor (MNR) and the 238Pu/Be neutron source site. This particular study consists of two parts. In the first part the water phantoms (125mL) were used in MCNP simulations. The water phantoms were doped with different concentrations of Cd, Hg and HCl. This is done in order to compare the (n, γ) prompt gamma reaction rate; in fact, the rate of neutron capture by the nuclides of interest; 113Cd, 199Hg and 35Cl. The second part involves, the neutron and photon dosimetry calculations that were performed for both sites using MCNP compatible body builder software developed in Los Alamos. The output of this program is the actual MCNP geometry description for various human anthropomorphic phantoms (different sex and ages). This phantom geometry output is incorporated into the original MCNP geometry and the dosimetry calculations were performed for various organs at risk.

Phantom studies of Cd, Hg and Cl by prompt gamma neutron activation analysis using a 238Pu–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±0.2ppm for Cd (125mL phantom) and 315±24ppm for Hg (125mL phantom) using the 238Pu–Be neutron source and 0.88±0.01ppm for Cd (125mL phantom) and 16.91±0.05ppm for Hg (30mL 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.

INAA of fluorine in plastics using the medium-flux McMaster Nuclear Reactor

The fluorine contents of plastics, ranging from about 20 μg·g−1 to 66%, may be measured instrumentally using a conventional research nuclear reactor, an automated sample irradiation and counting system, and a set of well-calibrated, in-house, fluorine standards. Plastics with low to medium fluorine contents may be analyzed using 20F by placing the gamma-ray detector at appropriate distances from the irradiated sample. For high-F plastics, samples may be irradiated in a cadmium lined irradiation site, using 19O and 20F. Counting statistics of <3% translate into reproducibility of measurements within ±3% and analytical accuracies of ±1% to ±10%.

Performance Support System for$hbox I^125$Monitoring in the McMaster University Nuclear Reactor

The purpose of this paper is to present a design methodology that can be employed to develop computerized radiation monitoring systems for technology insertion into non-computerized human-machine systems. The methodology seeks to hybridize and exploit the benefits of several diverse disciplines including nuclear engineering, computer science, health physics, real-time system design, human-machine interaction, computer-human interface, and software engineering, in the realization of a Performance Support System (PSS). The focus for combining these disciplines is on the implementation of a specific application: The Iodine Monitoring and Advising System (IMAS), a computerized embedded system for the MNR (McMaster Nuclear Reactor) I <sup xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">125</sup> production facility. IMAS uses radiation monitoring techniques and human factor concepts to provide nuclear reactor operators with a computerized monitoring alarm facility. The migration from a crude and inflexible spill detector to an intelligent performance support system is documented

Hydraulic study of turbulent flow in MTR-type nuclear fuel assembly

For a more accurate safety analysis for the McMaster nuclear reactor (MNR), local velocity measurements in a mock-up of the 18-plate fuel assembly were conducted in the low Reynolds number region with emphasis on edge effects. With modification of the curved plates in the assembly to flat ones, velocity in the narrow channels was measured by laser Doppler velocimetry (LDV). The velocity profile in individual channels and the channel-to-channel flow distribution were obtained. The effect of the sudden transition from the larger rectangular channel to the circular pipe on the velocity profile and the effect of the handle presence in the top end fitting on the flow distribution were investigated. The pressure drop along the assembly was measured and compared to a simple 1D correlation prediction.

A new determination of the structure of water at 25°C

The results of a new neutron diffraction experiment to measure the structure of water are presented. The data, measured at the McMaster Nuclear Reactor, are of a high quality and are analysed to yield the hydrogen-hydrogen pair correlation function using a subtraction procedure which has been used in previous experiments of this kind. This procedure circumvents the necessity of applying inelasticity corrections. The results are in good agreement with earlier work and serve to establish the general correctness of the subtraction procedure when used to determine hydrogen correlations. The data are further analysed to yield separate oxygen-hydrogen and oxygen-oxygen partial structure factors for liquid water. For the second part of the analysis an effective mass model of the dynamic scattering law is used, with the model parameter, the effective mass of the scattering particle, chosen by a least-squares fit to the measured differential cross sections. The final pair correlation functions are obtained using a maximum entropy analysis of the structure functions.

Angular distributions of photoneutrons in206Pb and208Pb using monochromatic photons

The angular distributions of neutrons produced in the (γ, n) reaction are reported for the separated isotopes206Pb and208Pb. The experiment made use of highly chromatic photons generated by neutron capture on nickel and chromium in the core of the McMaster Nuclear Reactor, and a high resolution3He detector. Photoneutron spectra were recorded at 7 angles between 20° and 160°. The results were used to infer the extent of photon multipole mixing between 7 and 11 MeV within the framework of the statistical collective model. The existence of M1 strength could not be confirmed but multipole mixing, most likely E2∶E1 was found at all energies. Estimates of the ratio of E2 to E1 strength is given for all photon energies investigated for both lead isotopes.