Universal Reactor Research Articles

Estimate of Neutron Fluence Using Isotope Ratio Method for ZR-2.5NB Materials Irradiated in the National Research Universal Reactor

Abstract The isotope ratio method (IRM) is a technique used for estimating the energy production in a fission reactor by measuring isotope ratios in nonfuel reactor components. This method has been successfully demonstrated on the estimation of cumulative energy production as well as plutonium production in graphite-moderated and light-water-moderated reactors for nonproliferation purposes. In this paper, IRM was used to estimate neutron fluence in Zr-2.5Nb materials irradiated in fast neutron (FN) irradiation facilities in the National Research Universal (NRU) reactor. Neutron fluence has been shown to be an important parameter for studying irradiation effects on the performance and properties of critical reactor components. Selected isotope ratios of hafnium and iron were used as indicators of the neutron fluence in Zr-2.5Nb sample materials. Correlations between neutron fluence and the indicator element isotope ratios were generated using the reactor physics simulation codes Winfrith improved multigroup scheme (WIMS)-AECL and SCALE/Oak Ridge Isotope GENeration code (ORIGEN). Inductively coupled plasma-mass spectrometry (ICP-MS) was used to obtain accurate measurements of the isotope ratios. Neutron fluence values estimated using IRM, were in good agreement with the values based on measured irradiation power histories of the NRU reactor. This study proposes a potential application of IRM to the estimation of neutron fluence for critical reactor components in heavy-water-moderated reactors such as pressure and calandria tubes in CANDU® reactors.

Weldability of irradiated Stainless Steel 304 materials harvested from the National Research Universal (NRU) reactor

The weldability of the neutron-irradiated Stainless Steel 304 (SS 304) materials containing estimated helium levels of 12 to 44 appm has been studied through a Collaborative Research and Development Agreement between the Canadian Nuclear Laboratories and Oak Ridge National Laboratory. The work utilizes a SS 304 component harvested from the National Research Universal reactor. Laser beam welding was used to explore suitable welding conditions and associated parameters, to determine helium concentration limits for successful welding of irradiated stainless-steels. The experimental results show that maximum fusion zone void sizes, fusion zone void quantity and total heat affected zone helium-induced crack length generally increased with increasing He levels and effective weld heat input. A proper combination of welding parameters with low effective heat input, such as, laser power, weld travel speed and wire feed speed, were shown to improve the welding quality, and potentially reduce the length and probability of the formation of grain boundary helium-induced cracking and voids of irradiated SS 304 even with He concentrations as high as about 44 appm.

The effect of cold work on irradiation creep of Zircaloy-2 alloys at low temperature

Internally pressurized capsules made from Zircaloy-2 tubing were irradiated in the National Research Universal (NRU) reactor at Chalk River Laboratories (CRL) at a temperature of approximately 57°C (330 K) in a fast neutron flux of about 3.37 × 1017 n·m−2·s−1 (E>1 MeV). The purpose of this experiment was to study the effect of cold work and crystallographic texture on biaxial stress irradiation creep. The creep capsules were cold worked in both transverse and axial directions. The applied hoop stress during irradiation ranged from 0 to approximately 200 MPa. Two irradiation cycles were completed with a fast neutron fluence of 3.18 × 1025 n·m−2 (E >1 MeV). Diametral and axial strains were measured after each irradiation cycle.This experiment has provided the following insights on Zircaloy-2 alloys irradiated at low temperature. The axial strain increases with fast neutron fluence, while the diametral strain either increases or decreases with fluence depending on the value of hoop stress and the amount of cold work. Both the diametral and axial strain rates increase non-linearly with the hoop stress. There is a weak dependence of the strain rate on cold work at low hoop stresses (50-150 MPa) suggesting that dislocation slip is not a dominant strain producing mechanism at irradiation temperature of about 57°C. The non-linear dependence of the strain rate on cold work that is observed at 200 MPa indicates that dislocation slip mechanisms start to dominate at this stress level.Dislocation densities of the creep capsules were measured by X-ray line broadening, and their structures were further analyzed using transmission electron microscopy (TEM). From these results, the weak dependence of the creep rate on cold work at low stresses can be explained using rate theory model formalism.

Review of Methods and Results for Reactor Physics Analysis of Thorium-Based Fuels From Irradiation Experiments Conducted in the National Research Universal Reactor

Abstract The development of advanced fuels and fuel cycles for conventional Generation III+ reactors, advanced reactors (such as Generation IV), and small modular reactors will help ensure the long-term sustainability and safety of the use of nuclear energy. Thorium-based fuels are an example of an advanced fuel that could augment and extend uranium resources. As part of previous efforts in Canada to investigate the use of thorium-based fuels for potential use in pressure-tube heavy-water reactors and other technologies, irradiation experiments with various thorium-containing fuels had been conducted in the National Research Universal (NRU) reactor at the Chalk River Laboratories. The NRU reactor physics computational tools and methods, developed initially for conventional enriched uranium fuel, had also been extended for analysis of thorium-based fuel experiments, with most satisfactory results. In this study, a number of physics and operation factors have been analyzed to quantify their effects on the production and burnup of 233U in thorium-based fuel, as well as to improve both the analysis and the use of thorium fuels and fuel cycles.

The impact of 40 years of radiation on the integrity of copper

Cu piping previously used to carry process air was harvested from the now shutdown National Research Universal reactor at the Canadian Nuclear Laboratories site in Chalk River, Ontario. Metallurgical and microscopy examinations found no quantifiable differences between irradiated and non-irradiated sections of the pipe. This provides evidence for the safety case for the use of the Cu-coated used fuel container proposed by the Nuclear Waste Management Organization for the permanent disposal of high-level nuclear waste.

High-Accuracy Relative Biological Effectiveness Values Following Low-Dose Thermal Neutron Exposures Support Bimodal Quality Factor Response with Neutron Energy.

Theoretical evaluations indicate the radiation weighting factor for thermal neutrons differs from the current International Commission on Radiological Protection (ICRP) recommended value of 2.5, which has radiation protection implications for high-energy radiotherapy, inside spacecraft, on the lunar or Martian surface, and in nuclear reactor workplaces. We examined the relative biological effectiveness (RBE) of DNA damage generated by thermal neutrons compared to gamma radiation. Whole blood was irradiated by 64 meV thermal neutrons from the National Research Universal reactor. DNA damage and erroneous DNA double-strand break repair was evaluated by dicentric chromosome assay (DCA) and cytokinesis-block micronucleus (CBMN) assay with low doses ranging 6–85 mGy. Linear dose responses were observed. Significant DNA aberration clustering was found indicative of high ionizing density radiation. When the dose contribution of both the 14N(n,p)14C and 1H(n,γ)2H capture reactions were considered, the DCA and the CBMN assays generated similar maximum RBE values of 11.3 ± 1.6 and 9.0 ± 1.1, respectively. Consequently, thermal neutron RBE is approximately four times higher than the current ICRP radiation weighting factor value of 2.5. This lends support to bimodal peaks in the quality factor for RBE neutron energy response, underlining the importance of radiological protection against thermal neutron exposures.

Post-irradiation examination of U-7Mo/Mg and U-10Mo/Mg dispersion fuels irradiated in the NRU reactor

As part of a global initiative to reduce the usage of highly enriched uranium (HEU) fuel for research and test reactors, Al-clad U-7Mo/Mg and U-10Mo/Mg pin-type mini-elements with a low-enriched uranium (U-235 enrichment of 19.75 wt%) loading of 4.5 g U/cm3 were fabricated at the Canadian Nuclear Laboratories (CNL). The mini-elements were irradiated in the National Research Universal (NRU) reactor at maximum linear power ratings of 100 kW/m up to 80 at% U-235 burnup. Hot cell macroscopic examination of the discharged fuel elements at 30, 60 and 80 at% U-235 burnup indicated that they were intact without breakaway swelling or fuel cladding failure. The results of interim post-irradiation examination (PIE) that was conducted on mini-elements discharged at 30 at% U-235 burnup are reported in this paper. A follow-up paper will present the PIE results of the mini-elements that achieved burnups of 60 and 80 at% U-235. In the present study, the microstructure and phase composition of the irradiated U-7Mo/Mg and U-10Mo/Mg fuel cores were investigated using optical microscopy and neutron diffraction analysis. The weight fraction of different crystalline phases in U-7Mo/Mg and U-10Mo/Mg fuel cores at low burnup was determined through Rietveld refinement. This paper includes a description of the irradiation tests along with preliminary PIE results from the irradiated U-Mo/Mg mini-elements and describes the results of optical microscopy characterization and neutron diffraction analysis.

A prototype compact accelerator-based neutron source (CANS) for Canada

Canada’s access to neutron beams for neutron scattering was significantly curtailed in 2018 with the closure of the National Research Universal (NRU) reactor in Chalk River, Ontario, Canada. New sources are needed for the long-term; otherwise, access will only become harder as the global supply shrinks. Compact Accelerator-based Neutron Sources (CANS) offer the possibility of an intense source of neutrons with a capital cost significantly lower than spallation sources. In this paper, we propose a CANS for Canada. The proposal is staged with the first stage offering a medium neutron flux, linear accelerator-based approach for neutron scattering that is also coupled with a boron neutron capture therapy (BNCT) station and a positron emission tomography (PET) isotope production station. The first stage will serve as a prototype for a second stage: a higher brightness, higher cost facility that could be viewed as a national centre for neutron applications.

The First Two Decades of Neutron Scattering at the Chalk River Laboratories

The early advances in neutron scattering at the Chalk River Laboratories of Atomic Energy of Canada are recorded. From initial nuclear physics measurements at the National Research Experimental (NRX) reactor came the realization that, with the flux available and improvements in monochromator technology, direct measurements of the normal modes of vibrations of solids and the structure and dynamics of liquids would be feasible. With further flux increases at the National Research Universal (NRU) reactor, the development of the triple-axis crystal spectrometer, and the invention of the constant-Q technique, the fields of lattice dynamics and magnetism and their interpretation in terms of the long-range forces between atoms and exchange interactions between spins took a major step forward. Experiments were performed over a seven-year period on simple metals such as potassium, complex metals such as lead, transition metals, semiconductors, and alkali halides. These were analyzed in terms of the atomic forces and demonstrated the long-range nature of the forces. The first measurements of spin wave excitations, in magnetite and in the 3D metal alloy CoFe, also came in this period. The first numerical estimates of the superfluid fraction of liquid helium II came from extensive measurements of the phonon–roton and multiphonon parts of the inelastic scattering. After the first two decades, neutron experiments continued at Chalk River until the shut-down of the NRU reactor in 2018 and the disbanding of the neutron effort in 2019, seventy years after the first experiments.

Investigation of in-beam prompt and delayed neutron counting techniques for detection and characterization of special nuclear material

A neutron counting (NC) technique for in-beam active neutron interrogation has been developed and deployed at the National Research Universal (NRU) reactor of the Canadian Nuclear Laboratories (CNL). A distinguishing feature of the method is that the sample to be interrogated remains stationary during the irradiation and counting periods for the detection of prompt and delayed neutrons. The NC system permits measurements of prompt and delayed neutrons using a mono-energetic or broad-spectrum (white) thermal neutron beam. An MCNPX computational study of an NC system employing nine 3He tubes showed that a delayed neutron detection efficiency of 22% could be achieved. The presence of a sub-milligram mass of 235U could be revealed in less than ten minutes in the in-beam delayed neutron experiment incorporating a white thermal neutron beam. The technique readily differentiated between 235U and 233U isotopes by analysis of delayed neutron count rates. The lower flux of the in-beam DN experiment does not permit the trace analysis that is possible with in-core irradiation, but does permit non-destructive analysis of large samples and could prove invaluable for the initial survey of materials of unknown content and origin. The study also demonstrated that in-beam prompt neutron analysis could be the ultimate solution when the neutron source is weak, the sample is shielded, the fissile mass is small, or interrogation time is limited. Detection time was reduced to a few seconds in the examination of prompt neutrons either in single or coincidence method. This paper presents the application of an in-beam NC system using a mono-energetic and white thermal neutron beam at the NRU reactor and assesses the performance of a newly constructed portable NC system with a large neutron detection array.

THERMAL BEHAVIOUR CORRELATION FOR THERMAL COLUMN IN THE NATIONAL RESEARCH UNIVERSAL (NRU) REACTOR

The National Research Universal (NRU) reactor is a major research facility that provides a beam of slow neutrons with a minimum of gamma rays and other types of radiation for experimental purposes. The thermal column consists of 5 graphite radial sections separated with an air gap for cooling. The graphite components require continuous monitoring to ascertain that temperatures are controlled within safe margins. Wall temperatures of the graphite sections are obtained via thermocouples affixed to the column walls. The safety margins for operation of the thermal column are driven by the temperatures of the closest radial section to the reactor core (HG1). Although, most of the thermocouples in HG1 are no longer functional, the thermocouples are functional in the adjacent graphite section (HG2). This study relied on the historical data of the graphite temperatures over a few years to develop an empirical correlation that relates temperatures in HG1 to those of HG2. The correlation sets limits on the functional thermocouples in HG2 to ensure HG1 remains within the prescribed limits (149–232 °C). Correlations were developed using statistical analysis of the historical data. A control band of approximately 40 °C for HG2 with confidence levels of 68% and 95%, respectively, were established.

BURFEL BIAS AND MCNP BENCHMARKING BURFEL CALCULATIONS OF NRU LOOP FUEL

The Burnup of Fuel Elements (BURFEL) code system has been used to calculate powers and burnups for experimental fuel irradiated in the National Research Universal Reactor (NRU) loops. BURFEL-calculated burnups, based on the calorimetrically measured loop powers, have been observed to exhibit biases with respect to their chemically measured counterparts. A high-fidelity Monte Carlo N-Particle method involving the NRU full-core model has been used for benchmarking BURFEL calculations, resulting in similar biases attributed to uncertainties in the loop powers. This study provides quantitative insights into the observed BURFEL biases for the purpose of possibly correcting existing loop fuel irradiation data for such biases.

A universal reactor platform for batch and flow: application to homogeneous and heterogeneous hydrogenation

Micro-CSTRs have been developed and used to determine optimal pressure hydrogenation conditions in batch, before being reconfigured for continuous flow.

Application of the universal ultrasonic reactor in the processing of rare earth metal ores concentrates

In recent years, heavy industry has rapidly increased interest in rare earth metals (REE). At the same time, new tasks on completeness of extraction and quality (purity) of REE are set. Providing new requirements for the quality of rare-earth metals can be achieved by two modern methods of ore processing. The first method is traditional leaching, but with the use of modern ultrasonic reactors of a through passage type of domestic production. The second method is leaching with the use of expensive imported impregnated sorbents that require special disposal after the deposition process of the desired fraction of material. The disadvantage of ultrasonic devices for processing of rare-earth metals is that the assigned parameters of the working chamber (length and diameter) are calculated for a specific type of ore being processed. Therefore, ultrasonic reactors operating in the metallurgical industry cannot be used to process all types of REE ores. The aim of the work is to study the efficiency of processing concentrates of ores containing rare earth elements by leaching using a universal ultrasonic reactor suitable for processing various concentrates containing rare earth elements. In this work, alkaline ore processing is carried out in an ultrasonic reactor of a special design, which allows regulation of the dimensions of the reactor working space this makes it possible to configure the reactor for highly efficient ore processing at different initial concentrations of valuable components. As shown by the results of the experiments, the extraction of rareearth metals and other valuable components of the ore in the ultrasonic reactor of this design is not less than 98.3%.

Application of a Graded Approach to Support the National Research Universal Reactor U-2 Experimental Loop Return to Service

Abstract The National Research Universal (NRU) reactor at Canadian Nuclear Laboratories (CNL) operated safely for over 60 years and supported a wide range of applications including, testing of fuels and materials under typical power reactor conditions in two experimental loops (U-1 and U-2). Both experimental loops had been taken out of service to address seismic deficiencies. CNL applied a graded approach to successfully return one of these loops, the U-2 Loop to service. The graded approach, without compromising safety, applied a risk informed methodology commensurate to the potential risk posed by the operation of the U-2 Loop. The work enabled the U-2 Loop to resume operation until the NRU reactor was permanently shut down in Mar. 31, 2018, generating valuable data that will be used in the development of advanced nuclear fuels and materials. This paper describes the graded approach employed by CNL that supported U-2 loop return to service (RTS). The use of graded approach is articulated to support development of safety and licensing cases for small modular reactor projects.

Validation of ENDF/B-VIII thermal neutron scattering data of heavy water by differential cross section measurements at various temperatures

Temperature-dependent thermal neutron differential scattering cross sections for heavy water (D2O) were measured using a triple-axis spectrometer at the National Research Universal reactor located in Chalk River, Canada. The measurements were performed at temperatures from 22 to 85 °C and at atmospheric pressure, with incident neutron energies from 16 to 44 meV and scattering angles from 10 to 110°. The temperature range in the measurements covered the common D2O moderator conditions (65–70 °C and 1 atm pressure) in pressurized heavy-water reactors. The measured results were compared with the D2O data libraries from the ENDF/B-VII.1 evaluation and the recently released ENDF/B-VIII.0 evaluation. The experimental results indicate significant improvements in the new release compared to the previous ENDF/B-VII.1 release.

Seventy years of scientific impact using neutron beams at the Chalk River Laboratories

The 31 March 2018 closure of the National Research Universal reactor marked the end of over 70 years of materials research using neutron beams from major neutron sources at the Chalk River Laboratories in Chalk River, Ontario, Canada. This closure will have a major impact on the Canadian materials research community, including researchers in the physics, chemistry, and engineering of materials. After a brief review of the history of neutron beams at the Chalk River Laboratories, we present the results of a bibliometric study of the scientific output of the research with neutron beams. In this study, we compiled a complete bibliographic record of the research papers beginning with the first neutron scattering experiments at the National Research Experimental reactor in 1947, analyzed the citations from 1980 onward, and benchmarked the results against major neutron beam facilities in other countries and against other major research facilities in Canada. We also conducted a broader bibliometric analysis of the use of neutron scattering data among all Canadians, regardless of where the data were taken. The results provide a useful metric of the size of the Canadian neutron scattering community and places into context the importance of access to this research tool.

Effect of neutron fluence on microstructure and mechanical properties of Al 5052 irradiated in the National Research Universal (NRU) reactor

In support of the fitness service assessment on the National Research Universal (NRU) vessel and full evaluation of its properties, an Al 5052 rod made from the same material and irradiated in the NRU reactor, was retrieved and sectioned for microstructure characterization and mechanical testing. Along its axial position, the rod was exposed to various levels of neutron fluence ranging from 0 to 11.9x1026 n/m2 (E < 0.625 eV) and 8.5x1025 n/m2 (E > 0.1 MeV). The estimated peak fluence in the NRU vessel is within this range, i.e., 2.4x1026 n/m2 (E < 0.625 eV) and 8.0x1022 n/m2 (E > 0.1 MeV) after 44 years of service. The hardness, chemical compositions, microstructure, tensile and fracture toughness properties were examined from different sections of the Al 5052 I-rod. The study shows that the hardness increased with continuous production of transmuted Si until it reached a plateau, when all of the dissolved Mg (2.5 wt%) was consumed to form radiation-induced Mg–Si precipitates. Excessive Si produced by transmutation primarily precipitated on grain boundaries. Increase in thermal fluence caused the changes in sizes and chemical compositions of the precipitates in the grain interior and boundaries. The yield stress and ultimate tensile strength increased with increasing thermal fluence, but uniform elongation and fracture toughness decreased. The current results show that the NRU vessel wall is expected to have sufficient ductility, strength and material integrity until the planned shutdown in March 2018, and improve the understanding the effects of neutron irradiation on Al 5000 series alloys, which are widely used as vessel materials in experimental reactors.

Stand-off nuclear reactor monitoring with neutron detectors for safeguards and non-proliferation applications

Safeguards measures are employed at nuclear reactor facilities worldwide, to ensure that nuclear material is not diverted from peaceful uses. Typical safeguards measures involve periodic inspections, off-line verification and video surveillance of fuel cycle activities. Real-time verification of the fissile contents via stand-off monitoring can enhance continuity of knowledge for non-traditional reactor types, including research reactors and small modular reactors. Here we demonstrate the feasibility of using large-area neutron detectors for monitoring nuclear reactors at stand-off distances up to 100 m outside reactor shielding, as a potential reactor safeguards tool. Since the neutron yield per unit reactor power depends upon the isotopic composition of the reactor core, declared changes in fissile composition can be verified without accessing the core. The supporting results of experiments conducted at the National Research Universal reactor in Canada, are presented.

Dosimetric and microdosimetric analyses for blood exposed to reactor-derived thermal neutrons

Thermal neutrons are found in reactor, radiotherapy, aircraft, and space environments. The purpose of this study was to characterise the dosimetry and microdosimetry of thermal neutron exposures, using three simulation codes, as a precursor to quantitative radiobiological studies using blood samples. An irradiation line was designed employing a pyrolytic graphite crystal or—alternatively—a super mirror to expose blood samples to thermal neutrons from the National Research Universal reactor to determine radiobiological parameters. The crystal was used when assessing the relative biological effectiveness for dicentric chromosome aberrations, and other biomarkers, in lymphocytes over a low absorbed dose range of 1.2–14 mGy. Higher exposures using a super mirror will allow the additional quantification of mitochondrial responses. The physical size of the thermal neutron fields and their respective wavelength distribution was determined using the McStas Monte Carlo code. Spinning the blood samples produced a spatially uniform absorbed dose as determined from Monte Carlo N-Particle version 6 simulations. The major part (71%) of the total absorbed dose to blood was determined to be from the 14N(n,p)14C reaction and the remainder from the 1H(n,γ)2H reaction. Previous radiobiological experiments at Canadian Nuclear Laboratories involving thermal neutron irradiation of blood yielded a relative biological effectiveness of 26 ± 7. Using the Particle and Heavy Ion Transport Code System, a similar value of ∼19 for the quality factor of thermal neutrons initiating the 14N(n,p)14C reaction in soft tissue was determined by microdosimetric simulations. This calculated quality factor is of similar high value to the experimentally-derived relative biological effectiveness, and indicates the potential of thermal neutrons to induce deleterious health effects in superficial organs such as cataracts of the eye lens.