Canadian Nuclear Laboratories (CNL) has an on-going Research & Development programme to support the development of a scaled-down 300 MWe version of the Canadian Super-Critical Water Reactor (SCWR) concept. The 300 MWe and 170-channel reactor core concept uses low enriched uranium fuel and features a maximum cladding temperature of 500°C. Our goal is to test surfacemodified zirconium alloys for use as fuel cladding. Zirconium alloys are attractive as they offer low neutron cross section thereby allowing the use of low enriched fuel. In this paper, we report on the results of general corrosion experiments used to evaluate chromiumcoated zirconium-based alloys in the two chemistries (630 ug/kg O2 in both deaerated and lithiated supercritical water). These experiments were conducted in a refreshed autoclave at 500°C and 23.5 MPa. After exposure, the weight gain and the hydrogen absorption were examined. At adequate coating thickness, longitudinal and transverse coupons show similar corrosion behaviour with improved corrosion resistance compared to uncoated coupons. The measured concentrations of hydrogen absorption are higher for the transverse coupons. Alkaline treatment resulted in higher weight gains than was found in pure oxygenated supercritical water.

The critical heat flux performance of an advanced plutonium-based mixed oxide fuel for potential use in a pressure tube heavy water reactor has been studied experimentally at Canadian Nuclear Laboratories with an electrically-heated string simulator of 43-element fuel bundles. The fuel simulator has a uniform axial power profile and a radial power profile representative of the plutonium-based MOX fuel. The CHF measurements were made in the MR-3 heat transfer loop facility using R-134a refrigerant as the working fluid. The test matrix included system pressures from 1.47 to 2.11 MPa, mass flow rates from 12.7 to 14.7 kg/s and inlet temperatures from 31 to 59°C, which are representative of the water-equivalent reactor operating conditions of 9 to 12.5 MPa pressure, 13.5 to 21.3 kg/s mass flow rate and the desired inlet subcoolings. Compared to conventional natural uranium fuel, the radial power profile of a MOX fuel exhibits a steeper and uneven distribution across the fuel element rings, with a higher value in the outer ring. It was found that CHF values of the MOX fuel are significantly lower than those of the natural uranium fuel. Based on the experimental data, a correlation has been derived to account for the effect of radial power profile on CHF. This correlation can be used to evaluate the relative CHF values of advanced/non-conventional fuel designs with radial power profiles deviating from that of natural uranium fuel.

The Atomic Energy Act, as amended, authorizes the United States (U.S.) Department of Energy (DOE) and its predecessor agencies to distribute nuclear materials to public or private institutions for the purposes of education as well as research and development. Significant transformations throughout the nuclear industry have led to changes in programmatic responsibility for loaned nuclear materials. DOE has established several programs to catalog, transfer ownership, retrieve, and/or dispose of these loaned nuclear materials. The variety of loaned nuclear material types as well as operational and regulatory variations between facility licensees have created unique challenges for the retrieval and dispositioning processes. These include packaging and transportation, confirmation of regulatory jurisdiction, property transfer, and disposal of sources with no remaining economic value. This paper explores the methods and actions taken by DOE to address these challenges. Lessons learned and the best practices identified from these programs are also presented.

Naturally occurring radioactive radon gas is produced from radioactive decay of naturally radioactive element radium and emits alpha particles. The radon gas is the main source of natural background radiation that contributes to about more than 60% of the annual radiation dose to humans on Earth. In this research, radon concentrations ( 222 Rn) in brick samples that are available in Iraqi markets were measured using solid state nuclear track ( CR-39 ). Also, uranium-238 ( 238 U), radium-226 ( 226 Ra) were calculated in all samples under study together with some radiation parameters such as exhalation of radon gas rate ( F O ), annual effective dose ( D Rn ), and excess lifetime cancer risk ( ELCR ). It was found that, the average value of 222 Rn, 238 U, and 226 Ra concentrations in studied samples were 48.75 ± 10.61 Bq/m 3 , 0.084 ± 0.02 ppm, and 115.57 ± 29.60 mBq/kg, respectively. Also, it was found that the average of each of F O , D Rn , and ELCR were 27.54 ± 5.99 μBq/m 2 .h, 0.88 ± 0.17 nSv/y, and (3.01 ± 0.65)×10 −3 , respectively. After data analysis, the resulted data were examined and compared with the global average and the permissible limits which recommended by the international scientific agencies such as International Commission on Radiological Protection ( ICRP 2010), United Nations Scientific Committee on the Effects of Atomic Radiation ( UNSCEAR 2017), Organization for Economic Cooperation and Development (OECD 2009). It was found that radiation levels from brick samples used in local markets for most models fell within the permissible limits and may not cause any danger to human beings.

Correct mathematical models of a reactor core in benchmarking of reactor physics experiments require detailed knowledge of exact isotopic composition of fuel pellets, especially the presence of neutron absorbers. Among them, knowledge of 234 U and 236 U abundance in fuel is important, because their absorption is considerably stronger than that of the most common 238 U and their concentration is significantly higher than any other absorbers. Thus, the fuel pellets used in the nuclear research reactor LR-0 have been experimentally investigated using the procedure that involves a method of secondary ion mass spectrometry (SIMS), which allowed for analyses of miniature amount of the pellet material without disturbing its integrity. The pellets were swiped with cotton cloths and the microscopic particles collected on the cloth were analyzed. The isotopic composition of individual uranium particles then provided information required for the reactor core modeling and in addition to that some hints about used source materials, history of nuclear material, and the manufacturing process of the pellets.

In the present work, Radon Excess Lung Cancer (ELC) is estimated for fourteen soil samples from Babylon cement plant in Iraq. CR-39 nuclear track detector is utilized to measure the uranium content and radon 222Rn concentrations in soil samples. Uranium concentrations in soil samples varied from 0.008 to 0.05 ppm with mean value 0.025±0.013. Radon concentrations is founded between 31 and 92 Bq/m 3 with mean value 56.72 and standard division, SD, 17.29. Radon Excess Lung Cancer per Million Persons per Year is determined with mean value 863 and standard division, SD, 261.65 (463.81-12082.8). Also, Annual effective dose, E (msv y-1), has a range from 0.77 to 2.32 with mean 1.44 and SD 0.44.

Melting of the terminal debris bed at the bottom of a Canada deuterium uranium (CANDU) 6 calandria vessel during a severe accident is studied using a three-dimensional transient finite element analysis. Settling of the debris is modeled, as are the effects of changing temperature and porosity on debris material properties. The time and magnitude of maximum heat rejection from the debris bed, and the maximum mass of molten debris, are calculated as functions of the accident timing. The results are compared with those from the modular accident analysis program ( MAAP )- CANDU severe accident code.

During burnup of nuclear fuel, fission products accumulate. Post-irradiation examination of burned up nuclear fuel has revealed the presence of several phases, namely, the fuel matrix of UO 2 , with dissolved oxides present; a white metallic phase consisting of the so-called “noble metals” (i.e., Mo–Ru–Pd–Rh–Tc); a gray oxide phase consisting of alkali or alkaline earth oxides (e.g., BaZrO 3 or Cs 2 UO 4 ); and an another metallic inclusion containing a mixture of UPd 3 –URh 3 –URu 3 , which is not completely assessed due to the lack of phase diagrams of the UPd 3 –URh 3 , URh 3 –URu 3 , and UPd 3 –URu 3 . Understanding how these phases behave becomes especially important from a safety perspective, if one considers a potential accident scenario. The quaternary system U–Pd–Rh–Ru has been evaluated and a thermodynamic model has been developed by first considering the six binary subsystems and the four ternary subsystems. A critical examination of the U–Pd, U–Rh, and U–Ru experimental phase diagrams has been made, with attention placed on both the solution phases, generally present on the uranium side of the diagrams and the UPd 3 –URh 3 –URu 3 compounds. Finally, the implications of this new model and its potential refinements of the Royal Military College of Canada nuclear fuel treatment developed by previous authors (notably the RMCC group under Thompson and Lewis) will be explored.

The value of β eff for Training Research Isotopes of General Atomics (TRIGA) Mark III reactor, belonging to the National Institute of Nuclear Researches (ININ), is reported. The TRIGA Mark III reactor core was simulated with MCNP6 to deduce the effective multiplication factor (k eff ) for critical state and after a small insertion of positive reactivity (∼0.20 $). To perform more realistic simulations, we had to incorporate in the composition of the low-enrichment uranium (LEU) fuel element the produced poisons in a time period of six years, considering the operation time in days, during which the reactor was operating at maximum power (1 MW th ). The calculation of the β eff value was obtained with the k eff results, calculated with the code, and the reactor periods measured experimentally. We also obtained directly the β eff value, through the card of MCNP6 to calculate k eff (KOPTS) card of the MCNP6 code in order to compare both values.

Within emerging best-estimate-plus-uncertainty (BEPU) approaches, code output uncertainties can be inferred from the propagation of fundamental or microscopic uncertainties. This paper examines the propagation of fundamental nuclear data uncertainties though the entire analysis framework to predict macroscopic reactor physics phenomena, which can be measured in Canada Deuterium Uranium (CANDU) reactors. In this work, 151 perturbed multigroup cross sections libraries, each based on a set of perturbed microscopic nuclear data, were generated. Subsequently, these data were processed into few-group cross sections and used to generate full-core diffusion models in PARCS. The impact of these nuclear data perturbations leads to changes in core reactivity for a fixed set of fuel compositions of 4.5 mk. The impact of online fueling operations was simulated using a series of fueling rules, which attempted to mimic operator actions during CANDU operations such as studying the assembly powers and selecting fueling sites, which would minimize the deviation in power from some desirable reference condition or increasing or decreasing fueling frequency to manage reactivity. An important feature of this analysis was to perform long-transients (1–3 years) starting with each one of the 151 perturbed full core models. It was found that the operational feedback reduced the standard deviation in core reactivity by 99% from 0.0045 to 2.8 × 10 −5 . Overall, the conclusions demonstrate that while microscopic nuclear data uncertainties may give rise to large macroscopic variability during simple propagation, when important macrolevel feedback are considered the variability is significantly reduced.

Uranium tetrafluoride was synthesized using a novel method, which consists of a combination of carbochlorination reaction between a mixture of U 3 O 8 and sucrose carbon with chlorine, and a solid-state halogen exchange reaction between the products of the carbochlorination reaction and sodium fluoride. The thermodynamic feasibility to produce the halogen exchange reaction between UCl 4 and NaF was analyzed. Reactions are favorable in standard conditions, even at low temperature. We have prepared a mixture of UCl 4 and UCl 2 O 2 by U 3 O 8 treatment in Cl 2 atmosphere with presence of sucrose carbon at 900 °C. UCl 4 and UCl 2 O 2 were obtained as a condensed product, which was collected in a quartz capsule containing NaF. The capsules were sealed after several repeated stages of argon purges and mechanic vacuum. Subsequently, they were treated at 300–350 °C for 2 h. We obtained that when NaF is the limiting reagent, the solid product of the thermal treatment of the capsules consists in a mixture of UF 4 and NaCl. Solid products were characterized by X-ray diffraction (XRD), scanning electron microscopy (SEM), fluorescence spectroscopy, and energy dispersive X-ray spectroscopy. Gaseous products were identified by Fourier-transform infrared spectroscopy.

Thorium-based fuel cycles can improve fuel sustainability within the nuclear power industry. The Canadian supercritical water-cooled reactor (SCWR) concept uses this path to achieve the sustainability requirement of the Gen-IV Forum. The study of thorium dioxide/thoria ThO 2 -based fuel irradiation behavior is significantly less advanced than that of uranium dioxide (UO 2 ) fuel, although ThO 2 possesses superior thermal conductivity, thermal expansion, higher melting temperature, and oxidation resistance that may improve both fuel performance and safety. The fuel and sheath modeling tool (FAST), a fuel performance model for UO 2 fuel, was developed at the Royal Military College of Canada (RMCC). FAST capability has been extended to include thoria (ThO 2 ), thorium uranium dioxide (Th,U)O 2 , and thorium plutonium dioxide (Th,Pu)O 2 as fuel pellet materials, to aid in designing and performance assessment of Th-based fuels, including SCWR (Th,Pu)O 2 fuel. The development and integration of ThO 2 and (Th,U)O 2 models into the existing FAST model led to the multipellet material FAST (MPM-FAST). Model development was performed in collaboration between RMCC and Canadian Nuclear Laboratories (CNL). This paper presents an outline of the ThO 2 and (Th,U)O 2 MPM-FAST model, a comparison between modeling results with postirradiation examination (PIE) data from a test conducted at CNL, and an account of the knowledge gap between our ability to model ThO 2 and (Th,U)O 2 fuel compared to UO 2 . Results are encouraging when compared to PIE data.

Uranium dioxide (UO 2 ) is the typical fuel that is used in the current nuclear power plant; fission gas atoms are produced during and after the nuclear reactor operation; the fission gas atoms have a significant effect on the performance of UO 2 fuel in the nuclear reactor. In this paper, we investigated the diffusion of the fission gas atoms in the UO 2 fuel by using the first-principles calculation method based on the density functional theory (DFT). The results indicate that the volume of the UO 2 cell increased when there is a fission gas atom enters in the UO 2 supercell; the elastic properties of UO 2 are in good agreement with other simulation results and experimental data and the fission gas atoms make the ductility of UO 2 decreased; the fission gas atoms prefer to occupy the octahedral interstitial site (OIS) over the uranium vacancy site and the oxygen vacancy site, and the oxygen vacancy site is the most difficult occupied site due to the formation of an oxygen vacancy is more difficult than that of the uranium vacancy; the diffusion barrier of a He atom in the UO 2 supercell is higher than that of an oxygen atom, that means that the diffusion of the He atom in UO 2 fuel is weaker than that of the oxygen atom. Our works may shed some light on the formation mechanism of the bubbles caused by the fission gas atoms in the UO 2 fuel.

Due to the advantages of small volume, light weight, and long-time running, nuclear reactor can provide an ideal energy source for space crafts. In this paper, two small compact prismatic nuclear reactors with different core block materials are presented, which have a thermal power of 5 MW for 10 years of equivalent full power operation. These two reactors use Mo-14%Re alloy or nuclear grade graphite IG110 as core block material, loaded with 50% and 39.5% enriched uranium nitride (UN) fuel and cooled by helium, whose inlet/outlet temperature of the reactor and operational pressure are 850/1300 K and 2 MPa, respectively. High temperature helium flowing out of the reactor can be used as the working medium for closed Brayton cycle power conversion with high efficiency (more than 20%). Neutronics analyses of reactors for the preliminary design in this paper are performed using reactor Monte Carlo (RMC) code developed by Tsinghua University. Both the reactors have enough initial excess reactivity to ensure 10 years of full power operation without refueling, have safety margin for reactor shutdown with one control drum failed, and remain subcritical in the submersion accident. Finally, the two reactors are compared in aspect of the 235 U mass and the total reactor mass.

The pressurized heavy water reactor (PHWR) technology was conceived in Canada and has moved to several nations for commercial production of electricity. Currently, 49 power reactors operate with PHWR technology producing nearly 25 GWe. The technology is flexible for adopting different fuel cycle options which include natural uranium, different mixed oxide (MOX) fuel, and thorium. The technology has made substantial improvement in materials, construction, and safety since its inception. PHWRs have demonstrated excellent performance historically. Their safety statistics are excellent. Indian PHWRs also have shown economic competitiveness even in small sizes, thus providing an ideal design for new entrants. While the technology features of PHWRs are available even in textbooks, the objective of this paper is to highlight the historical development and salient features, and innovations for further improvement in operation, safety and economics. Thus, this paper shall serve as a curtain raiser for the special issue “Pressurized Heavy Water Reactors (PHWRs) Safety: Post Fukushima.”

Molten salt reactors (MSRs) are promising advanced nuclear reactors for closure of the fuel cycle. This paper discusses the core design of graphite-moderated MSRs, thanks to a parametric study of the fuel and moderator lattice. The study is conducted at equilibrium of the thorium-uranium fuel cycle for several fuel channel radius and moderator block size combinations. The equilibrium composition for each studied configuration is derived with the help of an in-house MATLAB code, EQL0D, which uses the Serpent 2 Monte Carlo neutronics code for the calculation of reaction rates. The results include excess reactivity at equilibrium, mirroring the breeding gain, and the actinide vector composition for each configuration. Moreover, the occurence of an optimum of the excess reactivity per percent uranium-233 was observed. The investigations showed that it is systematically seen at an interchannel distance equal to the neutron slowing-down length in graphite for each configuration and does not depend on the salt channel radius beyond a certain size, which is given by the thermal fission rate reaching the levels of the fast fission rate. In this way, an exotic energy and spatial distribution of the neutrons are attained. The investigations highlight the potential attractiveness, from a neutronics/fuel cycle point of view, of both large fuel channels and moderators with a shorter neutron slowing-down length.