• Energy Research

AbstractThe Korean Prototype Gen-IV sodium-cooled fast reactor (PGSFR) is supposed to be loaded with a relatively-costly low-enriched U fuel, while its envisaged transuranic fuels are not available for transmutation. In this work, the U-enrichment reduction by improving the neutron economy is pursued to save the fuel cost. To improve the neutron economy of the core, a new reflector material, PbO, has been introduced to replace the conventional HT9 reflector in the current PGSFR core. Two types of PbO reflectors are considered: one is the conventional pin-type and the other one is an inverted configuration. The inverted PbO reflector design is intended to maximize the PbO volume fraction in the reflector assembly. In addition, the core radial configuration is also modified to maximize the performance of the PbO reflector. For the baseline PGSFR core with several reflector options, the U enrichment requirement has been analyzed and the fuel depletion analysis is performed to derive the equilibrium cycle parameters. The linear reactivity model is used to determine the equilibrium cycle performances of the core. Impacts of the new PbO reflectors are characterized in terms of the cycle length, neutron leakage, radial power distribution, and operational fuel cost.

Abstract The purpose of this work is to investigate the utilization of a set of nanofluids as a coolant/moderator for Pressurized Water Reactor (PWR) APR 1400 PLUS7 assembly design through modeling their effect on essential neutronics parameters and quantities of interest via SERPENT with simplistic thermal-hydraulics feedback. Various scenarios are discussed including fresh and burned conditions as well as the effect of nanoparticles crud deposition. Al2O3, TiO2, and CNT water-based nanofluids with a range of concentrations in the span of 0.001 wt%–0.3 wt% were chosen for the present case study. For each nanofluid/concentration combination; the effect on the assembly’s multiplication factor, Fuel Temperature Coefficient (FTC), Moderator Temperature Coefficient (MTC), and the discharge burnup are studied and highlighted. Moreover, the effect of a CRUD-like deposition of the nanoparticles on the axial power shifts is studied via SERPENT.

AbstractImportant safety parameters such as the fuel temperature coefficient (FTC) and the power coefficient of reactivity (PCR) of the CANada Deuterium Uranium (CANDU-6) reactor have been evaluated using the Monte Carlo method. For accurate analysis of the parameters, the Doppler broadening rejection correction scheme was implemented in the MCNPX code to account for the thermal motion of the heavy uranium-238 nucleus in the neutron-U scattering reactions. In this work, a standard fuel lattice has been modeled and the fuel is depleted using MCNPX. The FTC value is evaluated for several burnup points including the mid-burnup representing a near-equilibrium core. The Doppler effect has been evaluated using several cross-section libraries such as ENDF/B-VI.8, ENDF/B-VII.0, JEFF-3.1.1, and JENDL-4.0. The PCR value is also evaluated at mid-burnup conditions to characterize the safety features of an equilibrium CANDU-6 reactor. To improve the reliability of the Monte Carlo calculations, we considered a huge number of neutron histories in this work and the standard deviation of the k-infinity values is only 0.5–1 pcm.

Abstract The results of criticality, sensitivity and uncertainty (S/U) analyses on the 1st core criticality of the Indonesian 30 MWth Multipurpose Reactor RSG GAS (MPR-30) using the recent nuclear data libraries (ENDF/B-VII.1 and JENDL-4.0) and analytical tools available at present (Whisper-1.1) are presented. Two groups of criticality benchmark cases have been carefully selected from the experiments conducted during the first criticality approach and control rod calibrations. The C/E values of k (effective neutron multiplication factor) for the worst case is found around 1.005. Large negative sensitivities are found in (n,γ) reaction of H-1, U-235, Al-27, U-238, and Be-9 while large positive sensitivities are found in U-235 (total nu and fission), H-1 (elastic), Be-9 (free gas, elastic) and H-1 S(α,β) (lwtr.20t, inelastic). The energy dependency of the sensitivities is also presented and discussed in detail. From the S/U analysis, it is found that the uncertainties of k due to the nuclear data are about 0.6%, which are comparable to the [C/E−1] values. k Differences in the sensitivities amongst the two nuclear data libraries are also identified, and recommendation for improving the nuclear data library is given.

Abstract Sensitivity and uncertainty analyses of the effective neutron multiplication factor (keff) were conducted for the High Temperature Engineering Test Reactor (HTTR) using the ENDF/B-VII.1 nuclear data and the 44-group covariance library provided in the Whisper-1.1. Amongst the criticality benchmark series, the fully-loaded core was selected and analyzed by using a continuous energy Monte Carlo code, MCNP6.2. The top-five positive sensitivity coefficients (+0.1–+0.03) were found to be U-235 (total nu and fission), C-12 (elastic), graphite S(a,b) (elastic and inelastic), while the top-five negative ones (−0.1–−0.003) were U-238 (n,gamma), U-235 (n,gamma), B-10 (n,alpha), C-12 (n,gamma) and Si-28 (n,gamma). The keff uncertainty originated from the nuclear data was calculated using the previously obtained sensitivity coefficients and the 44-group covariance library. In the present S/U analysis, we obtained the keff uncertainty of about 0.5%. The JENDL-4.0 and ENDF/B-VIII.0 nuclear data were also analyzed for comparison.

Abstract This work presents a preliminary study of a novel plate-type fuel concept for a high-performance and ultra-safe research reactor. This new fuel type consists of coated particle fuel (CPF) randomly dispersed in an aluminum matrix with a certain packing fraction that can be adjusted depending on the reactor design requirements. The CPF can also be varied in the fuel kernel material between UC and UO 2 . For the purpose of this study, UO 2 was used as the reference fuel type. Using this novel fuel type, a 20 MWth pool-type research reactor was investigated to determine the preliminary performance and safety characteristics of the new fuel. The core thermal analysis was done using the MATRA-P code. The neutronics analysis was done using the Monte Carlo Serpent code for an equilibrium cycle resulting from a multi-batch fuel management. In this analysis, it was found that the Doppler effect is significantly enhanced through the implementation of CPF, in turn improving the inherent safety of the reactor. In addition to the notable improvement in safety, the new fuel type also promises to be able to achieve a high thermal neutron flux, improving the performance and utility of the reactor. It is concluded that the CPF-based fuel concept presented in this paper can enable new high-flux reactor designs using simple plate-type fuels with improved safety.

Abstract This paper presents the neutronics and transient studies of a supercritical CO2-cooled micro modular reactor (MMR). The suggested MMR is an extremely compact, integrated, and truck-transportable reactor with 36.2 MWth power and a 20-year lifetime without refueling. A salient feature of the MMR is that all the components including the generator are contained in a small reactor vessel. For a minimum volume and maximum lifetime of the MMR core, a fast neutron spectrum is utilized. To improve the fuel-coolant compatibility and to maximize the fuel volume fraction, UC fuel is considered. A unique replaceable fixed absorber (RFA) is also introduced in order to maintain the excess reactivity at less than 1 dollar during the whole operational period. A drum-type reactivity control system is adopted in the MMR as the primary control system to prevent a control rod ejection accident and a single absorber rod located at the core center is used as the secondary ultimate shutdown system. The neutronics analysis is performed by using the continuous energy Monte Carlo code Serpent with the ENDF/B-VII.1 library. In addition, an unprotected loss of flow (ULOF) accident is simulated by the system analysis code GAMMA+ to show the passive safety of the MMR system.