Assessment of Neutronic Characteristics of Accident-Tolerant Fuel and Claddings for CANDU Reactors

The objective of this study was to evaluate accident-tolerant fuel (ATF) concepts being considered for CANDU reactors. Several concepts, including uranium dioxide/silicon carbide (UO2-SiC) composite fuel, dense fuels, microencapsulated fuels, and ATF cladding, were modelled in Serpent 2 to obtain reactor physics parameters, including important feedback parameters such as coolant void reactivity and fuel temperature coefficient. In addition, fuel heat transfer was modelled, and a simple accident model was tested on several ATF cases to compare with UO2. Overall, several concepts would require enrichment of uranium to avoid significant burnup penalties, particularly uranium-molybdenum (U-Mo) and fully ceramic microencapsulated (FCM) fuels. In addition, none of the fuel types have a significant advantage over UO2 in terms of overall accident response or coping time, though U-9Mo fuel melts significantly sooner due to its low melting point. Instead, the different ATF concepts appear to have more modest advantages, such as reduced fission product release upon cladding failure, or reduced hydrogen generation, though a proper risk assessment would be required to determine the magnitude of these advantages to weigh against economic disadvantages. The use of uranium nitride (UN) enriched in N15 would increase exit burnup for natural uranium, providing a possible economic advantage depending on fuel manufacturing costs

Analysis of Control Rod Drop Accidents for the Canadian SCWR Using Coupled 3-Dimensional Neutron Kinetics and Thermal Hydraulics

The Canadian Supercritical Water-cooled Reactor (SCWR), a GEN IV reactor design, is a hybrid design of the well-established CANDU™ and Boiling Water Reactor with water above its thermodynamic critical point. Given the batch fueled design, control rods are used to manage the reactivity throughout the fuel cycle. This paper examines the consequences of a control rod drop accident (CRDA) for the Canadian SCWR. The asymmetry generated by the dropped rod requires an accurate 3-dimensional neutron kinetics calculation coupled to a detailed thermal-hydraulic model. Before simulating the CRDAs, the proper implementation of the 3D reactivity feedback was verified and various sensitivity studies were performed. This work demonstrates that the proposed safety systems for the SCWR core are capable of terminating the CRDA sequence prior to exceeding maximum sheath and centerline temperatures. In one instance involving a rod on the periphery of the core, the proposed trip setpoint (115% FP) was not exceeded and a new steady state was reached. Therefore it is recommended that the design also include provisions for a high-log rate and/or local Neutron Overpower Protection (NOP) trips, similar to existing CANDU designs such that reactor shutdown can be assured for such spatial anomalies

Selected papers from OECD-NEA PSBT benchmark

Historically, the prediction of safety margins has been based on system level thermal-hydraulic calculations employing suitable empirical formulations for assembly specific geometries and fuel-element grid spacers. These works have assessed response, margins, and consequences for the system based on one-dimensional two-fluid or drift-flux type thermalhydraulics formulations with fuel-vendor specific hydraulic losses and heat transfer characteristics for various fuel assemblies, including the so-called hot channel. Analysis of the hot channel gives important information on flow rates, fuel element centerline temperature, fuel sheath temperature, and margin to the departure from nucleate boiling. Given the reliance of the above approaches on empirical formulations obtained from complex and often difficult experiments, there is significant interest in obtaining reliable and accurate results from computation tools which employ more fundamental empirical relationships which can be obtained from subsets of the domain or from other scaled experiments

VALIDATION OF THE POLARIS CANDU EXTENSION FOR LATTICE PHYSICS

Polaris is a new lattice physics package, introduced in version 6.2 of the SCALE package. It uses a method of characteristics transport solver and the embedded self-shielding method. It is able to model light water reactor systems with a minimal amount of input. The goal of this project is to include support for CANDU models in Polaris for the next version of SCALE. So far, the model has been implemented and shown to give results with reasonable agreement to other SCALE sequences. This study extends the model to a reflector model, and shows that most quantities agree well with other codes. Some quantities, such as keff and assembly discontinuity factors, are sensitive to meshing. This study also performs a correlation between the TRITON and Polaris sequences using Sampler to perturb the nuclear data. Overall, there is good agreement between the two codes, though coolant void reactivity is only moderately correlated, likely due to the differences in resonance self-shielding methods. Additionally, this work shows that a coarser mesh can be used to speed up uncertainty calculations compared to the mesh used for a best estimate. Finally, this work shows that the mass lumping feature in CENTRM significantly affects heavy water moderated calculations, whether using TRITON or calculating self-shielding factors, and thus should be disabled for heavy water calculations

Evaluation of ASSERT-PV V3R1 against the PSBT Benchmark

Void fraction and DNB calculations conducted using ASSERT-PV V3R1 are evaluated against data from the NUPEC database as part of the OECD/NEA Pressurized Water Reactor Subchannel Benchmark Tests (PSBT). Void fraction measurements were well represented in the isolated single subchannel cases, with 77.0% of all predicted values falling within ±2σexp =0.06 of the experimental value. In the B5 type bundle, an average void fraction error of ϵ-α=-0.0540 was reported at the lower elevation, while this value was ϵ-α=-0.0405 at the upper measurement location. ASSERT was able to predict the steady state DNB power of the bundles to within ±10% of the measured value for a total of 344 times out of 432. Sensitivity studies conducted indicate that the Ahmad correlation with the Groeneveld 1995 CHF lookup table yielded the most accurate results, although some data points fell within the limiting quality region where the accuracy was reduced