National Research 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.

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.

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.

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.

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.

PRELIMINARY METHODOLOGY FOR THE ANALYSIS OF THE NATIONAL RESEARCH UNIVERSAL REACTOR USING INTEGRATED SEVERE ACCIDENT MODELLING CODES

The National Research Universal (NRU) Reactor is a multi-purpose research reactor located at Atomic Energy of Canada Limited (AECL) Chalk River Laboratories. The severe accident case for the NRU has been explored through deterministic and probabilistic safety analysis (PSA) including multi-level PSAs that detail the progression and consequences of a severe accident in the NRU. These previous calculations lack the interconnected and comprehensive features of a full severe accident modelling code that is now the standard for severe accident analysis of power reactors. It was of interest within AECL to evaluate modern severe accident modelling codes to the NRU reactor case to enhance the understanding of accident progression and predict the system damage and radiation release consequences of a severe accident, which is a very low probability event. The NRU is smaller and operates at a lower power than the large scale power reactors (e.g., pressurized heavy water reactors, pressurized water reactors, and boiling water reactors) that these codes were designed to analyze. Additionally, the NRU has a unique design different from the power reactors and several features relevant to severe accidents including filtered venting, large passive heat sinks, and a dispersion fuel design of uranium-silicide in an aluminum matrix. The major severe accident analysis codes available to AECL and their applicability to the NRU are explored in this paper. In addition, a preliminary strategy for employing the most applicable codes to the NRU for the purposes of severe accident modelling is proposed.

Replacement of Tritiated Water from Irradiated Fuel Storage Bay

Recently, AECL developed a novel method to reduce tritium emissions (to groundwater) and personnel doses at the NRU (National Research Universal) reactor irradiated fuel storage bay (also known as rod or spent fuel bay) through a water swap process. The light water in the fuel bay had built up tritium that had been transferred from the heavy water moderator through normal fuel transfers. A demonstration of the method was done that involved Computational Fluid Dynamics (CFD) modeling of the swap process and a test program that showed excellent agreement with model prediction for the effective removal of almost all the tritium with a minimal water volume. Building on the successful demonstration, AECL fabricated, installed, commissioned and operated a full-scale system to perform a water swap. This paper describes the results of the successful full-scale water swap operation, which achieved a tritium removal efficiency of ~96%

A Solution-Based Approach for Mo-99 Production: Considerations for Nitrate versus Sulfate Media

Molybdenum-99 is the parent of Technetium-99m, which is used in nearly 80% of all nuclear medicine procedures. The medical community has been plagued by Mo-99 shortages due to aging reactors, such as the NRU (National Research Universal) reactor in Canada. There are currently no US producers of Mo-99, and NRU is scheduled for shutdown in 2016, which means that another Mo-99 shortage is imminent unless a potential domestic Mo-99 producer fills the void. Argonne National Laboratory is assisting two potential domestic suppliers of Mo-99 by examining the effects of a uranyl nitrate versus a uranyl sulfate target solution configuration on Mo-99 production. Uranyl nitrate solutions are easier to prepare and do not generate detectable amounts of peroxide upon irradiation, but a high radiation field can lead to a large increase in pH, which can lead to the precipitation of fission products and uranyl hydroxides. Uranyl sulfate solutions are more difficult to prepare, and enough peroxide is generated during irradiation to cause precipitation of uranyl peroxide, but this can be prevented by adding a catalyst to the solution. A titania sorbent can be used to recover Mo-99 from a highly concentrated uranyl nitrate or uranyl sulfate solution; however, different approaches must be taken to prevent precipitation during Mo-99 production.

Microchemical study of high-burnup CANDU ® fuel by imaging-XPS

An advanced facility for characterization of highly radioactive materials by Imaging X-ray Photoelectron Spectroscopy (XPS) has been developed at the Chalk River Laboratories (CRL), based upon over a decade of prior experience with a prototype system. Auxiliary electron and ion guns provide additional in situ capabilities for scanning electron microscopy (SEM), scanning Auger microscopy (SAM) and composition depth profiling. The application of this facility to the characterization of irradiated fuel materials will be illustrated with selected results taken from a detailed study of the microchemistry at the fuel-sheath interface in a CANDU fuel element that was irradiated to extended burnup in the NRU (National Research Universal) reactor at CRL. Inside surfaces of the end caps and the welds between the sheath and the end caps as well as the thin-walled Zircaloy-4 sheath were analyzed. The in situ SEM capability was essential for selecting different areas on each sample, such as sheath locations with and without a visible retained CANLUB graphite layer, for XPS analysis. Effective infiltration of segregated fission products, especially cesium, into the graphite was demonstrated by depth profiling. A richer chemistry of segregated fission products was found on the end caps than on the sheath with elevated levels of barium, strontium, tellurium, iodine and cadmium as well as cesium. The results are consistent with current understanding of the primary migration route for fission products to the sheath and also indicate that the CANLUB layer functions as a chemical rather than a physical barrier to segregated fission products.

A conceptual model for the fuel oxidation of defective fuel

A mechanistic conceptual model has been developed to predict the fuel oxidation behaviour in operating defective fuel elements for water-cooled nuclear reactors. This theoretical work accounts for gas-phase transport and sheath reactions in the fuel-to-sheath gap to determine the local oxygen potential. An improved thermodynamic analysis has also been incorporated into the model to describe the equilibrium state of the oxidized fuel. The fuel oxidation kinetics treatment accounts for multi-phase transport including normal diffusion and thermodiffusion for interstitial oxygen migration in the solid, as well as gas-phase transport in the fuel pellet cracks. The fuel oxidation treatment is further coupled to a heat conduction equation. A numerical solution of the coupled transport equations is obtained by a finite-element technique with the FEMLAB 3.1 software package. The model is able to provide radial–axial profiles of the oxygen-to-uranium ratio and the fuel temperatures as a function of time in the defective element for a wide range of element powers and defect sizes. The model results are assessed against coulometric titration measurements of the oxygen-to-metal profile for pellet samples taken from ten spent defective elements discharged from the National Research Universal Reactor at the Chalk River Laboratories and commercial reactors.