Multi-Application Small Light Water Reactor Research Articles

Scaling challenges in small modular reactor

The small modular reactor (SMR) has become a popular reactor design concept in recent years. Many regulatory authorities are evaluating a plethora of license applications related to SMRs. One important step of the reactor design is the prototype computer model assessment, and scaling is indispensable toward developing meaningful assessment data. In SMR scaling, new challenges may rise due to compact layout and unique thermal hydraulic phenomena. These phenomena may be greatly distorted or may not even be reproducible at all in the test facility. Therefore, investigation of these phenomena is needed to ensure figures of merit are adequately preserved with minimum distortion. In this paper, some phenomena were highlighted to understand the complexity. An early SMR design - the Multi-Application Small Light Water Reactor (MASLWR) is used to illustrate the scaling of these phenomena.

The utilization of ThO2 fuel discharged from an ADS in the Multi-Application Small Light Water Reactor (MASLWR)

This study presents a neutronic and thermal–hydraulic estimation to convert a Multi-Application Small Light Water Reactor (MASLWR) with UO2 core to the UO2 + ThO2 core or ThO2 with the minimum possible modifications in the geometry and main parameters of MASLWR core. The thorium considered in this work was already enriched in 233U. MCNPX 2.7.0 (Monte Carlo code) has been used to calculate neutronic parameters such as effective multiplication coefficient (Keff), the nuclear fuel evolution during the burn-up, and the power peaking factor (Pmax/Pav) in the radial direction of the MASLWR different cores. RELAP5 code also has been applied to calculate thermal–hydraulic parameters for fuel rod hot channel, such as the surface heat flux, the coolant channel temperature. The centerline temperature of the fuel was calculated axially and radially, as well as the departure from nucleate boiling (DNB) ratio. The results show that the utilization of thorium (enriched with 233U) fuel improves the overall performance characteristics of the MASLWR. Less or no burnable poisons may be used (especially at the beginning of the reactor cycle); longer reactor cycle and higher fuel burn-up can be achieved, and most attractive feature is its resistance to nuclear proliferation. Also, thermal–hydraulic analysis of fuel rods displayed that the ThO2 fuel rod has a lower axial centerline temperature than UO2, and the DNBR value is about 5.2 for both fuel types and occurred at 0.6 of the fuel height from the inlet side.

Nuclear energy system’s behavior and decision making using machine learning

Early versions of artificial neural networks’ ability to learn from data based on multivariable statistics and optimization demanded high computational performance as multiple training iterations are necessary to find an optimal local minimum. The rapid advancements in computational performance, storage capacity, and big data management have allowed machine-learning techniques to improve in the areas of learning speed, non-linear data handling, and complex features identification. Machine-learning techniques have proven successful and been used in the areas of autonomous machines, speech recognition, and natural language processing. Though the application of artificial intelligence in the nuclear engineering domain has been limited, it has accurately predicted desired outcomes in some instances and has proven to be a worthwhile area of research. The objectives of this study are to create neural networks topologies to use Oregon State University’s Multi-Application Small Light Water Reactor integrated test facility’s data and evaluate its capability of predicting the systems behavior during various core power inputs and a loss of flow accident. This study uses data from multiple sensors, focusing primarily on the reactor pressure vessel and its internal components. As a result, the artificial neural networks are able to predict the behavior of the system with good accuracy in each scenario. Its ability to provide technical data can help decision makers to take actions more rapidly, identify safety issues, or provide an intelligent system with the potential of using pattern recognition for reactor accident identification and classification. Overall, the development and application of neural networks can be promising in the nuclear industry and any product processes that can benefit from utilizing a quick data analysis tool.

MELCOR Analysis of OSU Multi-Application Small Light Water Reactor (MASLWR) Experiment

A Multi-Application Small Light Water Reactor (MASLWR) conceptual design was developed by Oregon State University (OSU) with emphasis on passive safety systems. The passive containment safety system employs condensation and natural circulation to achieve the necessary heat removal from the containment in case of postulated accidents. Containment condensation experiments at the MASLWR test facility at OSU are modeled and analyzed with MELCOR, a system-level reactor accident analysis computer code. The analysis assesses its ability to predict condensation heat transfer in the presence of noncondensable gas for accidents where high-energy steam is released into the containment. This work demonstrates MELCOR’s ability to predict the pressure-temperature response of the scaled containment. Our analysis indicates that the heat removal rates are underestimated in the experiment due to the limited locations of the thermocouples and applies corrections to these measurements by conducting integral energy analyses along with computational fluid dynamics simulation for confirmation. The corrected heat removal rate measurements and the MELCOR predictions on the heat removal rate from the containment show good agreement with the experimental data.

Applying uncertainty and sensitivity on thermal hydraulic subchannel analysis for the multi-application small light water reactor

Small modular reactors (SMRs) are a recent advancement in commercial nuclear reactor design with growing interest worldwide. New SMR concepts, such as the Multi-Application Small Light Water Reactor (MASLWR), must undergo a licensing processes established by the U.S. Nuclear Regulatory Commission (NRC) prior to commercial operation. Given the lack of historical, full scale operating experience, a general uncertainty and sensitivity analysis methodology was developed to help aid SMR designs through this process. Uncertainty was quantified through the empirical cumulative distribution function (ECDF) created from a desired data set. Linear regression techniques were applied to measure sensitivity. This methodology was demonstrated through the thermal hydraulic steady-state subchannel analysis of the MASLWR concept using RELAP5-3D Version 4.0.3 and VIPRE-01 Mod 2.2.1. Twelve uncertain input parameters were selected. System response uncertainty in the minimum departure from nucleate boiling ratio (MDNBR), maximum fuel temperature, and maximum clad temperature was evaluated. These figures were shown to satisfy U.S. NRC regulatory requirements for steady state operation at the 95 percent probability and 95 percent confidence level under the evaluated conditions. Sensitivity studies showed input parameters affecting local power generation within the core had a large influence on MDNBR, maximum fuel temperature, and maximum clad temperature.

Analyses of the OSU-MASLWR Experimental Test Facility

Today, considering the sustainability of the nuclear technology in the energy mix policy of developing and developed countries, the international community starts the development of new advanced reactor designs. In this framework, Oregon State University (OSU) has constructed, a system level test facility to examine natural circulation phenomena of importance to multi-application small light water reactor (MASLWR) design, a small modular pressurized water reactor (PWR), relying on natural circulation during both steady-state and transient operation. The target of this paper is to give a review of the main characteristics of the experimental facility, to analyse the main phenomena characterizing the tests already performed, the potential transients that could be investigated in the facility, and to describe the current IAEA International Collaborative Standard Problem that is being hosted at OSU and the experimental data will be collected at the OSU-MASLWR test facility. A summary of the best estimate thermal hydraulic system code analyses, already performed, to analyze the codes capability in predicting the phenomena typical of the MASLWR prototype, thermal hydraulically characterized in the OSU-MASLWR facility, is presented as well.

A Five-Year Core for a Small Modular Light Water Reactor

To address the energy needs of developing countries and remote communities, Oregon State University has proposed the Multi-Application Small Light Water Reactor (MASLWR) design. This design uses 8% enriched fuel to achieve five years of operation without refueling. The specific operational conditions (lower pressure and temperature of fuel and coolant), increased enrichment of fuel, and extensive use of gadolinium burnable absorbers lead to significantly different neutron physics compared to conventional pressurized water reactors. In particular, spectrum hardening due to increased thermal neutron absorption, changes in kinetic parameters due to the isotopic content of the fresh and irradiated fuel, and fuel and control rod shadowing by burnable absorbers are consequences of the design requirements. Enhanced neutron leakage from the small MASLWR core also adds complexity. Neutron reflectors and a unique fuel-loading pattern compensate the pronounced axial and radial gradients of the neutron flux and power generation.This paper discusses the neutron physics and thermal-hydraulic issues of the core design for a small reactor with increased fuel enrichment and natural circulation of the coolant. The paper describes three evolutionary steps of the MASLWR core design process and discusses core parameters, advantages, disadvantages, and design limitations as they appeared during the core design feasibility study. The paper demonstrates the feasibility of the core design for five effective years of nonrefueled operation with 8.0% enriched UO2 fuel.

Sensitivity analysis of the MASLWR helical coil steam generator using TRACE

Accurate simulation of transient system behavior of a nuclear power plant is the goal of systems code calculations, and the evaluation of a code's calculation accuracy is accomplished by assessment and validation against appropriate system data. These system data may be developed either from a running system prototype or from a scaled model test facility, and characterize the thermal hydraulic phenomena during both steady state and transient conditions. The identification and characterization of the relevant thermal hydraulic phenomena, and the assessment and validation of thermal hydraulic systems codes, has been the objective of multiple international research programs. The validation and assessment of the best estimate thermal hydraulic system code TRACE against the Multi-Application Small Light-Water Reactor (MASLWR) Natural Circulation (NC), helical coil Steam Generator (SG), Nuclear Steam Supply System (NSSS) design is a novel effort, and is the topic of the present paper. Specifically, the current work relates to the assessment and validation process of TRACE code against the NC database developed in the OSU-MASLWR test facility. This facility was constructed at Oregon State University under a U.S. Department of Energy grant in order to examine the NC phenomena of importance to the MASLWR reactor design, which includes an integrated helical coil SG. Test series have been conducted at this facility in order to assess the behavior of the MASLWR concept in both normal and transient operation and to assess the passive safety systems under transient conditions. In particular the test OSU-MASLWR-002 investigated the primary system flow rates and secondary side steam superheat, used to control the facility, for a variety of core power levels and Feed Water (FW) flow rates. This paper illustrates a preliminary analysis, performed by TRACE code, aiming at the evaluation of the code capability in predicting NC phenomena and heat exchange from primary to secondary side by helical SG in superheated condition and to evaluate the fidelity of various methods to model the OSU-MASLWR SG in TRACE. The analyses of the calculated data show that the phenomena of interest of the OSU-MASLWR-002 test are predicted by the code and that one of the reasons of the instability of the superheat condition of the fluid at the outlet of the SG is the equivalent SG model used to simulate the different group of helical coils. The SNAP animation model capability is used to show a direct visualization of selected calculated data.

Testing of the multi-application small light water reactor (MASLWR) passive safety systems

Experimental thermal hydraulic research has been conducted at Oregon State University for the purpose of assessing the performance of a new reactor design concept, the multi-application small light water reactor (MASLWR). The MASLWR is a pressurized light water reactor design with a net output of 35 MWe that uses natural circulation in both normal and transient operation. Due to its small size, portability and modularity, the MASLWR design is well suited to help fill the potential need for grid appropriate reactor designs for smaller electricity grids as may be found in developing or remote regions. The purpose of the OSU MASLWR test facility is to assess the operation of the MASLWR under normal full operating pressure and full temperature conditions and to assess the passive safety systems under transient conditions. The data generated by the testing program will be used to assess computer code calculations and to provide a better understanding of the thermal-hydraulic phenomena in the design of the MASLWR NSSS. During this testing program, four tests were conducted at the OSU MASLWR test facility. These tests included one design basis accident and one beyond design basis accident. During the performance of these tests, plant operations to include start up, normal operation and shut down evolutions were demonstrated successfully.