Neutronic Characteristics of ENDF/B-VIII.0 Compared to ENDF/B-VII.1 for Light-Water Reactor Analysis

The Evaluated Nuclear Data File (ENDF)/B-VIII.0 data library was released in 2018. To assess the new data during development and shortly after release, many validation calculations were performed with static, room-temperature benchmarks. Recently, when performing validation of ENDF/B-VIII.0 for pressurized water reactor depletion calculations, a regression in performance compared to ENDF/B-VII.1 was observed. This paper documents extensive benchmark calculations for light-water reactors performed using continuous energy Monte Carlo code with ENDF/B-VII.1 and -VIII.0 and neutronic characteristics of ENDF/B-VIII.0 are discussed and compared to those of ENDF/B-VII.1. It is our recommendation that ENDF/B data library assessment should include reactor-specific benchmark assessments, including depletion cases, such that these types of regressions may be caught earlier in the data development cycle.

Calculation of neutronic characteristics in different reflector materials with a 15-MWt reactor core using VVR-KN fuel type

VVR-KN is one of the low enriched fuel types intended for a research reactor of a newCentre for Nuclear Energy Science and Technology (CNEST) of Viet Nam. As a part of design orientation for the new research reactor, the calculations of neutronic characteristics in a reactor core reflector using different materials were carried out. The investigated core configuration is a 15-MWt power loaded with VVR-KN fuel assemblies and surrounded by a reflector using beryllium, heavy water or graphite respectively. MCNP5 code together with up-to-date nuclear data libraries were used for these calculations. This paper presents the calculation results of neutron energy spectrum, neutron spatial distribution in the reflector using the above-mentioned materials. Besides, neutronic characteristics calculated for silicon doping irradiation holes in the reflector are also presented and the utilization capabilities of different reflector materials are discussed.

Feasibility Study for the Fast Periodic Pulsed Reactor with UO2 Fuel

In order to study the feasibility of the fast periodic pulsed reactor with UO2 as fuel (abbreviated as FPPRU), the core models with different load schemes are designed. Neutronic characteristics of two typical design schemes are compared, and the better design scheme is determined. The critical search method is established for analyzing the reactor dynamics. Furthermore, the theoretical estimation formulas are derived to study the factors affecting the reactor dynamics clearly and intuitively. The reactor dynamics of the fast periodic pulsed reactor with UO2 and PuO2 as fuel are compared. The thermal hydraulic characteristic of FPPRU is studied with the sub-channel model. The results show that the design scheme of the FPPRU meets the demand of neutronics and thermal hydraulics safety. Meanwhile, the pulse parameter quality of the FPPRU with UO2 as fuel is not as good as that of IBR-2 with PuO2 as fuel.

The Effects of the Number of Coated Fuel Particles on the Neutronic Aspects of 25 MWt Pebble Bed Reactor with Thorium Fuel

High-Temperature Gas Reactor (HTGR) is a type of reactor that continues to be developed because of its advantages in terms of economic aspects, proliferation resistance, and safety aspects. One of the safety aspect improvements is due to the use of the Coated Fuel Particle (CFP). A coated fuel particle is a fuel with a diameter smaller than 1 mm and is protected by several carbon layers. In the Pebble Bed Reactor (PBR) type of HTGR design, the CFP is placed in a 6 cm fuel ball. How much CFP is put into the fuel ball will determine the neutronic characteristics of the reactor. In this study, the effect of the amount of CFP in the fuel ball on the 25 MWt PBR design using Thorium fuel and its impact on several important neutronic aspects, such as the effective multiplication factor, the amount of fuel enrichment, the utilization of fissile material, and the density of the fissile material formed. The calculation was performed by the Monte Carlo MVP / MVP-BURN code. This study found that the coated fuel particle fraction of 15% was the optimum value for the studied neutronic parameters.

SMALL MODULAR PWR DESIGN FOR TRU RECYCLING WITH McCARD-MASTER TWO-STEP PROCEDURE

In our previous study, a small modular PWR core was designed for TRU (Transuranics) recycling with multi-recycling scheme with a typical two-step procedure using DeCART2D/MASTER code system in which the lattice analysis for producing homogenized group constant was performed by DeCART2D while whole core analysis was conducted by MASTER code. However, the neutron spectrum hardening of the LWR core loaded with TRU requires validating the multi-group cross section library and resonance self-shielding treatment method in lattice calculation. In this study, a new procedure using McCARD/MASTER was used to analyze the SMR core, in which the lattice calculation was performed by a Monte Carlo code called McCARD with a continuous energy library to generate homogenized two-group assembly cross sections. The SMR core analysis was performed to show neutronic characteristics and TRU mass flow in the SMR core with TRU multi-recycling. The result shows that the analyses on the neutronic characteristics and TRU mass flow using the McCARD/MASTER code system showed good agreement with the previous ones using the DeCART2D/MASTER code system. The neutronic characteristics of each cycle of the core satisfied the typical limit of a commercial PWR core and the SMR core consumes effectively TRU with net TRU consumption rates of 8.46~14.33 %.

DEVELOPMENT OF THE MPACT 69-GROUP LIBRARY FOR MAGNOX REACTOR ANALYSIS USING CASL VERA

The Consortium for Advanced Simulation of Light Water Reactors (CASL) has developed the CASL toolset, Virtual Environment for Reactor Analysis (VERA), for pressurized water reactor (PWR) analysis. Recently the CASL VERA was improved for Magnox reactor analysis, which required the development of a new cross section library and new geometrical and thermal feedback capabilities for graphite-moderated Magnox reactors. The MPACT neutronics module of the CASL core simulator is a 3D whole core transport code, which requires a new cross section library with a different energy group structure due to the different neutronic characteristics of Magnox compared with PWR. A new 69-group structure was developed based on the MPACT 51-group structure to have more thermal energy groups and to be a subset of the SCALE 252-group structure. The ENDF/B-VII.1 MPACT 69-group library was developed for Magnox reactor analysis using the SCALE/AMPX and VERA-XSTools for which a super-homogenization method was applied, and transport cross sections were generated for graphite using a neutron leakage conservation method. Benchmark results show that new MPACT 69-group library works reasonably well for Magnox reactor analysis.

Precision neutronic calculations of experiments on the neutron transmission through the reflector layers at the BFS critical facilities for expanding the verification database to justify lead cooled fast reactor designs

The paper presents the results of the efforts concerned with expanding the verification database and estimating the calculation uncertainty of the power density in the steel reflector of lead cooled fast reactor designs based on experiments performed in different years at the BFS critical assemblies by analyzing and revising earlier calculation and experimental studies on the transmission of neutrons through the steel reflector layers. The discussion includes experiments at the BFS-66 critical assembly to model neutron and photon fluxes in the reactor core shielding compositions, as well as experiments at the BFS-64 and BFS-80-2 critical assemblies to model the transmission of neutrons and gamma quanta through the reflector layers of various materials. The information provided in earlier materials with the descriptions of the above experiments has been analyzed and expanded through respective data required to prepare precision calculation models for Monte-Carlo neutronic codes. Precision neutronic models have been developed based on actualized and updated data with a detailed description of the BFS heterogeneous structure and experimental devices, and test calculations have been carried out to confirm their efficiency. The calculations of key neutronic characteristics measured at the BFS-66, -64 and -80-2 assemblies were performed using codes based on the Monte Carlo method (MCU-BR, MCNP, MMK-RF, MMK-ROKOKO) with BNAB-RF and MDBBR50 neutron data and the ROSFOND evaluated neutron data library. The developed precision calculation neutronic models of the experiments discussed can be used to justify lead cooled fast reactor designs, to verify neutronic codes and neutron data, and to evaluate the associated uncertainties.