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.

EQUILIBRIUM MOLECULAR DYNAMICS CALCULATIONS OF THERMAL CONDUCTIVITY: A “HOW-TO” FOR THE BEGINNERS

The ready availability of codes such as LAMMPS (Large-scale Atomic/Molecular Massively Parallel Simulator) for molecular dynamics simulations has opened up the realm of atomistic modelling to novice code users with an interest in computational materials modelling but who lack the appropriate theoretical or computational background. As such, there is significant risk of the “user effect” having a negative impact on the quality of results obtained using such codes. Here, we present a “how-to” procedure for equilibrium molecular dynamics-based nuclear fuel thermal conductivity calculations using the Green–Kubo method with an interatomic potential developed by Cooper et al. [ 1 ]. The various steps of the simulation are identified and explained, along with criteria to assess the quality of the intermediate and final results, discussion of some problems that can arise during a simulation, and some inherent limitations of the method. Calculated thermal conductivities for UO2 and ThO2 will be compared with the available experimental data and also with similar thermal conductivity calculations using nonequilibrium molecular dynamics, reported in the open literature.

DIFFERENCES IN EXIT BURNUP, DEPLETION, AND FUEL REACTIVITY CALCULATIONS ARISING FROM THE CHOICE OF MODELLING CODE FOR BUNDLES WITH THORIUM AND URANIUM

A computational benchmark, using the deterministic codes WIMS-AECL and WOBI, and the stochastic code SERPENT, is made for burnup calculations of advanced thorium fuels in heavy water moderated reactors. Exit burnups and the concentration of the longer-lived actinides from the deterministic code set of WIMS-AECL and WOBI, which are 2-D, were compared to those from a full 3-D calculation in SERPENT. Results for reactivity vs. time are in general agreement to within a few mk (<1% in overall neutron multiplication) and appear to be systematic. Results for exit burnup were larger, in the 3%–6% range, because small reactivity effects can be amplified here.

CONCEPTUAL DESIGN OF AN ORGANIC-COOLED SMALL NUCLEAR REACTOR TO SUPPORT ENERGY DEMANDS IN REMOTE LOCATIONS IN NORTHERN CANADA

Small nuclear reactors can offer safe, reliable, and long-lasting district heating and electrical power generation to remote locations in northern Canada. A conceptual design of an organic-cooled and moderated reactor based upon the SLOWPOKE-2 research reactor is proposed for potential employment in northern Canada. For viability, this design extends the SLOWPOKE-2’s power to 1 MWth. An added pump circulates the organic coolant, a partially hydrogenated terphenyl mixture known as HB-40, to facilitate greater heat transfer. The reactor incorporates the same low-enriched uranium dioxide fuel as the SLOWPOKE-2. Reactor control is accomplished through hafnium absorber rods and a movable beryllium reflector. The reactor neutronics are simulated using the deterministic code, WIMS-AECL, and the probabilistic code, MCNP 6. The service life of fuel in this reactor operating at full power exceeds 11 years. The conceptual design has demonstrated negative reactivity coefficients indicating strong potential for inherent safety.

THE CANADIAN NUCLEAR SAFETY COMMISSION’S READINESS TO REGULATE SMALL MODULAR REACTORS

Over the course of the last several years the Canadian Nuclear Safety Commission (CNSC) has engaged with numerous vendors and potential licenses of small modular reactor (SMR) technology. This paper discusses what an SMR is and what potentially makes them different from standard nuclear power plants (NPP). Readiness activities for the potential licensing of SMRs are described as well as modifications being made to the CNSC’s existing regulatory framework to facilitate the same, without reducing safety. The role of the CNSC’s discussion paper DIS-16-04, Small Modular Reactors: Regulatory Strategy, Approaches and Challenges (DIS 16-04) and how feedback received on it helped confirm the CNSC’s modifications to be undertaken to the regulatory framework, as well as areas requiring further clarity, are highlighted. Finally, the role of the CNSC Vendor Design Review process is described as well as other readiness activities undertaken by the CNSC that are helping to ensure that the CNSC will be ready to accept and evaluate a license application for an SMR.

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.

THE EFFECTS OF TYPICAL THORIUM IMPURITIES ON THORIUM-BASED NUCLEAR FUELS

Natural thorium contains impurities of numerous isotopes. A study was performed to examine the influence of naturally occurring impurities in thorium-based fuels on a few parameters of interest such as: exit burnup, coolant void reactivity (CVR), fuel temperature coefficients (FTC), and the radiotoxicity of the used fuel. Two different fuel bundle designs were modeled: a 43-element bundle containing 2.25% U-233, and a 35-element bundle containing 1.45% U-233. Naturally occurring thorium fuel impurities were applied to both fuel bundle models at various concentrations, from 0% to 100% of the expected maximum. For burnup-averaged k-infinity (k∞) values of 1.050 and 1.030, exit burnup, burnup-weighted CVR, and burnup-weighted FTC were calculated using the neutron transport code WIMS-AECL, and plotted against fraction of full impurity concentration to determine how the impurity levels affect these reactor physics parameters of interest. For the most-realistic (for CANDU) burnup-averaged k∞ of 1.050, both the inhalation radiotoxicity and the production of U-232 were calculated using the fuel depletion code WOBI. Up to the maximum impurity concentrations considered, no effects on the investigated fuel performance parameters were found to be greater than a few percent.

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.

ISSUES AND CHALLENGES IN ASSESSING ECOLOGICAL AND HUMAN HEALTH RISK FROM THE SITING OF SMRS IN CANADA

There are many issues and challenges in assessing ecological and human health risk from siting small modular reactors (SMRs) in northern or Arctic regions. Environmental guidance for Canadian nuclear facilities is largely derived from data and models relevant to temperate regions, with no explicit guidance or parameters for northern regions. International Atomic Energy Agency guidance provides some data and parameters for northern regions, but there remains a paucity of data and models. Although wildlife often comprise a major part of northern and Arctic diets, there are few data or transfer parameters for these ecosystems. Data and transport models are available for weapon test and Sellafield/La Hague fission products in northern oceans, but very little is known about circulation or fate and transport in estuarine and coastal areas typical of northern Canada. Baseline data, parameters, and models are needed for key processes and pathways to accurately assess ecological and human health risk.

RESULTS OF A PHENOMENA IDENTIFICATION AND RANKING TABLE (PIRT) EXERCISE FOR A SEVERE ACCIDENT IN A SMALL MODULAR HIGH-TEMPERATURE GAS-COOLED REACTOR

Canada has attracted specific interest from developers of nonwater-cooled small modular reactor (SMR) technologies, including concepts based on high-temperature gas-cooled reactors (HTGRs). It is anticipated that some research and development (R&D) will be necessary to support safety analysis and licensing of these reactors in Canada. The Phenomena Identification and Ranking Table (PIRT) process is a formalized method in which a panel of experts identifies which physical phenomena are most relevant to the reactor safety analysis and how well understood these phenomena are. The PIRT process is thus a tool to assess current knowledge levels and (or) predictive capabilities of models, thus providing direction to a focused R&D program. This paper summarizes the results of a PIRT process performed by a panel of experts at Canadian Nuclear Laboratories for a limiting or “worst-case” accident scenario at a generic HTGR-type SMR. Suggestions are given regarding the highest priority R&D items to support severe accidents analysis of these reactors.