• Nuclear Engineering and Technology
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• v.47 no.4
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• pp.434-442
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• 2015

The advanced passive pressurized water reactor (PWR) is being constructed in China and the passive residual heat removal (PRHR) system was designed to remove the decay heat. During accident scenarios with increase of heat removal from the primary coolant system, the actuation of the PRHR will enhance the cooldown of the primary coolant system. There is a risk of power excursion during the cooldown of the primary coolant system. Therefore, it is necessary to analyze the thermal hydraulic behavior of the reactor coolant system (RCS) at this condition. The advanced passive PWR model, including major components in the RCS, is built by SCDAP/RELAP5 code. The thermal hydraulic behavior of the core is studied for two typical accident sequences with PRHR actuation to investigate the core cooling capability with conservative assumptions, a main steam line break (MSLB) event and inadvertent opening of a steam generator (SG) safety valve event. The results show that the core is ultimately shut down by the boric acid solution delivered by Core Makeup Tank (CMT) injections. The effects of CMT boric acid concentration and the activation delay time on accident consequences are analyzed for MSLB, which shows that there is no consequential damage to the fuel or reactor coolant system in the selected conditions.

• Nuclear Engineering and Technology
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• v.51 no.1
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• pp.45-53
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• 2019

Thermal-hydraulic passive safety systems (PSSs) are incorporated into many advanced reactor designs on the bases of simplicity, economics and inherent safety nature. Several factors among which are the critical parameters (CPs) that influence failure and reliability of thermal-hydraulic (t-h) passive systems are now being explored. For simplicity, it is assumed in most reliability analyses that the CPs are independent whereas in practice this assumption is not always valid. There is need to critically examine the dependency influence of the CPs on reliability of the t-h passive systems at design stage and in operation to guarantee safety/better performance. In this paper, two multivariate analysis methods (covariance and conditional subjective probability density function) were presented and applied to a simple PSS. The methods followed a generalized procedure for evaluating t-h reliability based on dependency consideration. A passively water-cooled steam generator was used to demonstrate the dependency of the identified key CPs using the methods. The results obtained from the methods are in agreement and justified the need to consider the dependency of CPs in t-h reliability. For dependable t-h reliability, it is advisable to adopt all possible CPs and apply suitable multivariate method in dependency consideration of CPs among other factors.

• Nuclear Engineering and Technology
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• v.51 no.1
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• pp.60-72
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• 2019

The air-cooled passive decay heat removal system (APDHR) was proposed to provide the ultimate heat sink for non-LOCA accidents. The APDHR is a modified one of Passive Auxiliary Feed-water system (PAFS) installed in APR+. The PAFS has a heat exchanger in the Passive Condensate Cooling Tank (PCCT) and can remove decay heat for 8 h. After that, the heat transfer rate through the PAFS drastically decreases because the heat transfer condition changes from water to air. The APDHR with a vertical heat exchanger in PCCT will be able to remove the decay heat by air if it has sufficient natural convection in PCCT. We conducted the thermal-hydraulic simulation by the MARS code to investigate the behavior of the APR + selected as a reference plant for the simulation. The simulation contains two phases based on water depletion: the early phase and the late phase. In the early phase, the volume of water in PCCT was determined to avoid the water depletion in three days after shutdown. In the late phase, when the number of the HXs is greater than 4089 per PCCT, the MARS simulation confirmed the long-term cooling by air is possible under extended Station Blackout (SBO).

• Nuclear Engineering and Technology
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• v.56 no.6
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• pp.2375-2387
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• 2024

The passive safety systems (PSSs) within water-cooled reactors are meticulously engineered to function autonomously, requiring no external power source or manual intervention. They depend exclusively on inherent natural forces and the fundamental principles of reactor physics, such as gravity, natural convection, and phase changes, to manage, alleviate, and avert the release of radioactive materials into the environment during accident scenarios like a loss-of-coolant accident (LOCA). PSSs are already integrated into such operating commercial reactors as the Advanced Pressurized Reactor-1000 MWe (AP1000) and the Water-Water Energetic Reactor-1200 MWe (WWER-1200) are adopted in most of the upcoming small modular reactor (SMR) designs. Examples of water-cooled SMR PSSs are the passive emergency core-cooling system (ECCS), passive containment cooling system (PCCS), and passive decay-heat removal system, the designs of which vary based on reactor system-design requirements. However, understanding the accident-event progression and phases of a LOCA is pivotal for adopting a specific PSS for a new SMR design. This study covers the accident-event progression for direct vessel injection (DVI) small-break loss-of-coolant accident (SB-LOCA), associated physics phenomena, knowledge gaps, and important figures of merit (FOMs) that may need to be evaluated and assessed to validate thermal-hydraulics models with an available experimental dataset to support new SMR design and development.

• Nuclear Engineering and Technology
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• v.56 no.8
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• pp.2893-2905
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• 2024

Passive safety systems are integrated into the latest generation of Light Water Reactors (LWRs), including small modular reactors. This paper employs the US-NRC TRACE thermal hydraulic code to examine the performance of a passive safety condenser known as SACO, designed to serve as the ultimate heat sink for dissipating decay heat during accident scenarios. The TRACE model is constructed with reference to the PKL/SACO test facility. The safety condenser (SACO) is interconnected with the PKL facility via the secondary side of steam generator 1, effectively serving as a third natural circulation cooling loop during accident scenarios. In the present research, the thermal-hydraulic behavior of the PKL facility is investigated in the presence of the SACO passive safety system during an extended SBO with Loss of AC Power accident scenario. This SBO can be categorized into three distinct phases depending on the activation of the SACO system and the refilling process of the SACO pool. The first phase is depressurizing using primary and secondary relief valves, the second phase is cooling down using SACO system, and the third phase is the refilling of SACO pool. The findings indicate that the SACO system effectively manages to dissipate all decay heat, even though there is temporary evaporation of the SACO water pool. Furthermore, this study provides sensitivity analysis for the assessments of system codes on the selection of maximum time step.

• Nuclear Engineering and Technology
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• v.45 no.6
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• pp.759-766
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• 2013

As passive safety features for nuclear power plants receive increasing attention, various studies have been conducted to develop safety systems for 3rd-generation (GEN-III) nuclear power plants that are driven by passive systems. The Passive Auxiliary Feedwater System (PAFS) is one of several passive safety systems being designed for the Advanced Power Reactor Plus (APR+), and extensive studies are being conducted to complete its design and to verify its feasibility. Because the PAFS removes decay heat from the reactor core under transient and accident conditions, it is necessary to evaluate the heat removal capability of the PAFS under hypothetical accident conditions. The heat removal capability of the PAFS is strongly dependent on the heat transfer at the condensate tube in Passive Condensation Heat Exchanger (PCHX). To evaluate the model of heat transfer coefficient for condensation, the Multi-dimensional Analysis of Reactor Safety (MARS) code is used to simulate the experimental results from PAFS Condensing Heat Removal Assessment Loop (PASCAL). The Shah model, a default model for condensation heat transfer coefficient in the MARS code, under-predicts the experimental data from the PASCAL. To improve the calculation result, The Thome model and the new version of the Shah model are implemented and compared with the experimental data.

• Journal of Environmental Health Sciences
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• v.35 no.6
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• pp.500-509
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• 2009

Passive samplers have been used for many years for the sampling of organic vapors in work environment atmospheres. Currently, all passive samplers used in domestic occupational monitoring are foreign products. This study was performed to evaluate variable parameters for the development of passive organic samplers, which include the geometry of the device and diffusive length for the sampler design. Four prototype diffusive lengths; A-1(4.5 mm), A-2(7.0 mm), A-3(9.5 mm), A-4(12.0 mm) were tested for adsorption performances to a chemical mixture (benzene, toluene, trichloroethylene, and n-hexane) according to the US-OSHA's evaluation protocol. A dynamic vapor exposure chamber developed and verified by related research was used for this study. The results of study are as follows. The results in terms of sampling rate and recommended sampling time test indicate that the most suitable model was A-3 (9.5 mm diffusive lengths on both sides) for passive sampler design in time weighted average (TWA) assessment. Sampling rates of this A-3 model were 45.8, 41.5, 41.4, and 40.3 ml/min for benzene, toluene, trichloroethylene, and n-hexane, respectively. The A-3 models were tested on reverse diffusion and conditions of low humidity air (35% RH) and low concentrations (0.2 times of TLV). These conditions had no affect on the diffusion capacity of samplers. In conclusion, the most suitable design parameters of passive sampler are: 1) Geometry and structure - 25 mm diameter and 490 $mm^2$ cross sectional area of diffusion face with cylindrical form of two-sided opposite diffusion direction; 2) Diffusive length - 9.5 mm in both faces; 3) Amount of adsorbent - 300 mg of coconut shell charcoal; 4) Wind screen - using nylon net filters (11 ${\mu}m$ pore size).

• Nuclear Engineering and Technology
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• v.54 no.6
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• pp.2297-2310
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• 2022

A condensation heat transfer model is essential to accurately predict the performance of the passive containment cooling system (PCCS) during an accident in an advanced light water reactor. However, most of existing models tend to predict condensation heat transfer very well for a specific range of thermal-hydraulic conditions. In this study, a new correlation for condensation heat transfer coefficient (HTC) is presented using machine learning technique. To secure sufficient training data, a large number of pseudo data were produced by using ten existing condensation models. Then, a neural network model was developed, consisting of a fully connected layer and a convolutional neural network (CNN) algorithm, DenseNet. Based on the hold-out cross-validation, the neural network was trained and validated against the pseudo data. Thereafter, it was evaluated using the experimental data, which were not used for training. The machine learning model predicted better results than the existing models. It was also confirmed through a parametric study that the machine learning model presents continuous and physical HTCs for various thermal-hydraulic conditions. By reflecting the effects of individual variables obtained from the parametric analysis, a new correlation was proposed. It yielded better results for almost all experimental conditions than the ten existing models.

• Nuclear Engineering and Technology
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• v.53 no.8
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• pp.2499-2508
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• 2021

Passive containment cooling systems (PCCSs) have been actively studied to improve the inherent safety of nuclear power plants. Hered, we present two concepts, open-loop PCCS (OL-PCCS) and closed-loop PCCS (CL-PCCS), applicable to the PWR with a concrete-type containment. We analyzed the heat-removal performance and flow instability of these PCCS concepts using the GOTHIC code. In both cases, PCCS performance improved when a passive containment cooling heat exchanger (PCCX) was installed in the lower part of the containment building. The OL-PCCS was found to be superior in terms of heat-removal performance. However, in terms of flow instability, the OL-PCCS was more vulnerable than the CL-PCCS. In particular, the possibility of flow instability was higher when the PCCX was installed in the upper part of the containment. Therefore, the installation location of the OL-PCCS should be restricted to minimize flow instability. Conversely, a CL-PCCS can be installed without any positional restriction by adjusting the initial system pressure within the loop, which eliminates flow instability. These results could be used as base data for the thermo-hydraulic evaluation of PCCS in PWR with a large dry concrete-type containment.

• Nuclear Engineering and Technology
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• v.55 no.7
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• pp.2438-2446
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• 2023

To cool down a nuclear reactor core and prevent the fuel damage without a pump-driven active component during any anticipated accident, the passive emergency core cooling system (PECCS) was designed and adopted in an advanced light water reactor, i-POWER. In this study, for a validation of the cooling capability of PECCS, thermal-hydraulic integral effect tests were performed with the ATLAS facility by simulating intermediate and small break loss-of-coolant accidents (IBLOCA and SBLOCA). The test result showed that PECCS could effectively depressurize the reactor coolant system by supplying the safety injection water from the safety injection tanks (SITs). The result pointed out that the safety injection from IRWST should have been activated earlier to inhibit the excessive core heat-up. The sequence of the PECCS injection and the major thermal hydraulic transient during the SBLOCA transient was similar to the result of the IBLOCA test with the equivalent PECCS condition. The test data can be used to evaluate the capability of thermal hydraulic safety analysis codes in predicting IBLOCA and SBLOCA transients under an operation of passive safety system.