• Nuclear Engineering and Technology
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• 제47권2호
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• pp.148-156
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• 2015

Several vendors have recently been actively pursuing the development of integral pressurized water reactors (iPWRs) that range in power levels from small to large reactors. Integral reactors have the features of minimum vessel penetrations, passive heat removal after reactor shutdown, and modular construction that allow fast plant integration and a secure fuel cycle. The features of an integral reactor limit the options for placing control and safety system instruments. The development of instrumentation and control (I&C) strategies for a large 1,000 MWe iPWR is described. Reactor system modeling-which includes reactor core dynamics, primary heat exchanger, and the steam flashing drum-is an important part of I&C development and validation, and thereby consolidates the overall implementation for a large iPWR. The results of simulation models, control development, and instrumentation features illustrate the systematic approach that is applicable to integral light water reactors.

• Nuclear Engineering and Technology
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• 제39권4호
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• pp.249-256
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• 2007

Light water reactors operated under supercritical pressure conditions have been selected as one of the promising future reactor concepts to be studied by the Generation IV International Forum. Whereas the steam cycle of such reactors can be derived from modem fossil fired power plants, the reactor itself, and in particular the reactor core, still need to be developed. Different core design concepts shall be described here to outline the strategy. A first option for near future applications is a pressurized water reactor with $380^{\circ}C$ core exit temperature, having a closed primary loop and achieving 2% pts. higher net efficiency and 24% higher specific turbine power than latest pressurized water reactors. More efficiency and turbine power can be gained from core exit temperatures around $500^{\circ}C$, which require a multi step heat up process in the core with intermediate coolant mixing, achieving up to 44% net efficiency. The paper summarizes different core and assembly design approaches which have been studied recently for such High Performance Light Water Reactors.

• 방사선산업학회지
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• 제17권3호
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• pp.239-256
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• 2023

The classification of the radiation work performed in Korean nuclear power plants (NPPs) must be understood to provide workers with more comprehensive radiation protection. This study used annual reports on occupational exposure to investigate and analyze the similarities and differences in the radiation work performed in Korean NPPs with pressurized water reactors (PWRs) and pressurized heavy water reactors (PHWRs). The results showed that the radiation work performed in Korean NPPs could be classified into three categories. Category 1 contains work at the highest level. This work can be divided into individual tasks belonging to Category 2, which enables the evaluation of the radiation dose during the work. The work in Category 2 consists of tasks from Category 3, which contains basic detailed tasks that are not further subdivided. This study emphasized the need for the systematic management of the radiation work performed in both Korean PWRs and PHWRs, such as the tasks in Category 3, which are similar, with similar working conditions, for PWRs and PHWRs. It also suggested the need to establish a list of radiation work for decommissioning because Kori Unit 1 and Wolsong Unit 1 are currently in permanent shutdown and preparations are being made for their decommissioning.

• Nuclear Engineering and Technology
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• 제41권8호
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• pp.1005-1014
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• 2009

The Dual Unit Optimizer 2000 MWe (DUO2000) is put forward as a new design concept for large power nuclear plants to cope with economic and safety challenges facing the $21^{st}$ century green and sustainable energy industry. DUO2000 is home to two nuclear steam supply systems (NSSSs) of the Optimized Power Reactor 1000 MWe (OPR1000)-like pressurized water reactor (PWR) in single containment so as to double the capacity of the plant. The idea behind DUO may as well be extended to combining any number of NSSSs of PWRs or pressurized heavy water reactors (PHWRs), or even boiling water reactors (BWRs). Once proven in water reactors, the technology may even be expanded to gas cooled, liquid metal cooled, and molten salt cooled reactors. With its in-vessel retention external reactor vessel cooling (IVR-ERVC) as severe accident management strategy, DUO can not only put the single most querulous PWR safety issue to an end, but also pave the way to very promising large power capacity while dispensing with the huge redesigning cost for Generation III+ nuclear systems. Five prototypes are presented for the DUO2000, and their respective advantages and drawbacks are considered. The strengths include, but are not necessarily limited to, reducing the cost of construction by decreasing the number of containment buildings from two to one, minimizing the cost of NSSS and control systems by sharing between the dual units, and lessening the maintenance cost by uniting the NSSS, just to name the few. The latent threats are discussed as well.

• 방사선산업학회지
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• 제18권1호
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• pp.23-33
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• 2024

This study was conducted to analyze the characteristics of three high-exposure tasks performed by radiation workers in Korean pressurized water reactors (PWRs) and to identify factors that reduce their exposure during work. Three high-exposure tasks were selected based on a previous study. In this previous study, nozzle dam installation and removal, eddy current testing, and manway opening and closing were determined as high-exposure tasks through normalization (radiation dose per unit time). Based on the analysis of the characteristics of the high-exposure tasks in this study, the high-exposure tasks were steam generator-related tasks performed inside and outside the water chamber. This study analyzed the reduction factors for high-exposure tasks and suggested improvements in terms of time, distance, and shielding. The use of the characteristics of high-exposure tasks and their dose reduction factors enables Korean PWRs to optimize radiation protection for workers who receive relatively high doses.

• Nuclear Engineering and Technology
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• 제50권6호
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• pp.877-885
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• 2018

Various controllers such as proportional-integral-derivative (PID) controllers have been designed and optimized for load-following issues in nuclear reactors. To achieve high performance, gain tuning is of great importance in PID controllers. In this work, gains of a PID controller are optimized for power-level control of a typical pressurized water reactor using particle swarm optimization (PSO) algorithm. The point kinetic is used as a reactor power model. In PSO, the objective (cost) function defined by decision variables including overshoot, settling time, and stabilization time (stability condition) must be minimized (optimized). Stability condition is guaranteed by Lyapunov synthesis. The simulation results demonstrated good stability and high performance of the closed-loop PSO-PID controller to response power demand.

• Nuclear Engineering and Technology
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• 제56권5호
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• pp.1874-1879
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• 2024

Owing to strict radiation safety management in Korean nuclear power plants (NPPs), most radiation workers receive very low radiation doses, even lower than the annual dose limit for the general public. However, the occupational dose distribution indicates that some Korean NPP workers receive a relatively higher dose than the average dose. This inequity in radiation exposure could be reduced by providing customized radiation protection measures, such as dose constraints, to workers receiving relatively higher doses. In this study, dose normalization was performed to identify the highest radiation exposure work in Korean pressurized water reactors (PWRs). The results show that most of the occupational exposure in Korean PWRs occurs during the planned maintenance period. Finally, the three highest radiation exposure tasks in Korean PWRs were identified: nozzle dam installation and removal, eddy current testing, and man-way opening and closing.

• Nuclear Engineering and Technology
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• 제50권1호
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• pp.97-106
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• 2018

Improved load-following capability is one of the most important technical tasks of a pressurized water reactor. Controlling the nuclear reactor core during load-following operation leads to some difficulties. These difficulties mainly arise from nuclear reactor core limitations in local power peaking: the core is subjected to sharp and large variation of local power density during transients. Axial offset (AO) is the parameter usually used to represent the core power peaking. One of the important local power peaking components in nuclear reactors is axial power peaking, which continuously changes. The main challenge of nuclear reactor control during load-following operation is to maintain the AO within acceptable limits, at a certain reference target value. This article proposes a new robust approach to AO control of pressurized water reactors during load-following operation. This method uses robust feedback-linearization control based on the multipoint kinetics reactor model (neutronic and thermal-hydraulic). In this model, the reactor core is divided into four nodes along the reactor axis. Simulation results show that this method improves the reactor load-following capability in the presence of parameter uncertainty and disturbances and can use optimum control rod groups to maneuver with variable overlapping.

• Nuclear Engineering and Technology
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• 제56권10호
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• pp.4280-4288
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• 2024

This research article introduces a study regarding the uncertainties treatment during severe accidents for Pressurized Heavy Water Reactors (PHWRs). The present study is focused upon the unmitigated Station BlackOut (SBO) accident analysis for a CANada Deuterium Uranium (CANDU) type reactor emphasizing the impact of the uncertainties treatment on the relevant key timings of the SBO accident progression through different approaches for the uncertainty parameters' Probabilistic Distribution Functions (PDFs). A comparison between the sensitivity analysis results is provided in the present research study. The uncertainty analysis is performed with the RELAP/SCDAPSIM code with the Integrated Uncertainty Analysis (IUA) package from the code. Results from the research would support the advancements on the best-practices for uncertainty analyses with respect to the parameter's uncertainties distribution functions. Data dispersion is a key element for the realistic quantification of uncertainties in nuclear power plant safety analyses, including severe accidents.

• Nuclear Engineering and Technology
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• 제45권5호
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• pp.589-596
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• 2013

In a pressurized heavy water reactor, following loss of the primary coolant, severe core damage would begin with the depletion of the liquid moderator, exposing the top row of internally-voided fuel channels to steam cooling conditions on the inside and outside. The uncovered fuel channels would heat up, deform and disassemble into core debris. Large inventories of water passively reduce the rate of progression of the accident, prolonging the time for complete loss of engineered heat sinks. The efficacy of available backup and ultimate heat sinks, available in a CANDU 6 reactor, in mitigating the consequences of a prolonged station blackout scenario was analysed using the MAAP4-CANDU code. The analysis indicated that the steam generator secondary side water inventory is the most effective heat sink during the accident. Additional heat sinks such as the primary coolant, moderator, calandria vault water and end shield water are also able to remove decay heat; however, a gradually increasing mismatch between heat generation and heat removal occurs over the course of the postulated event. This mismatch is equivalent to an additional water inventory estimated to be 350,000 kg at the time of calandria vessel failure. In the Enhanced CANDU 6 reactor ~2,040,000 kg of water in the reserve water tank is available for prolonged emergencies requiring heat sinks.