• Journal of Radiation Industry
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• v.17 no.3
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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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• v.56 no.10
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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.

• Journal of Radiation Protection and Research
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• v.35 no.4
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• pp.172-177
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• 2010

The radioactive effluents from pressurized heavy water reactors (PHWRs) are relatively larger than those from pressurized water reactors (PWRs). Futhermore, radioactive effluents from PHWRs are released continuously. Thus, the discharge of radioactive effluents is strictly controlled. To do this, radiation detectors are installed at stacks of reactor buildings to monitor the concentration of radioactive effluents in real-time. Derived release limits (DRLs) of annual discharge are also set up for each radionuclide and effluents are rigidly controlled not to exceed those limits. In this paper, the discharge process of radioactive effluents, the standard for establishment of DRL and its methodology, and currents status for PHWRs were reviewed.

• Nuclear Engineering and Technology
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• v.41 no.8
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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.

• Nuclear Engineering and Technology
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• v.45 no.5
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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.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.38 no.2
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• pp.167-175
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• 2014

If volumetric flaws such as bearing pad fretting flaws and debris fretting flaws are detected in the pressure tubes of pressurized heavy water reactors during in-service inspection, the initiation of fatigue cracks and delayed hydrogen cracking from the detected volumetric flaws shall be assessed by using elastic stress concentration factors in accordance with CSA N285.8-05. The CSA N285.8-05 presents only an approximate formula based on linear elastic fracture mechanics for the debris fretting flaw. In this study, an engineering formula considering the geometric characteristics of the debris fretting flaw in detail was derived using two-dimensional finite element analysis and Kinectrics, Inc.'s engineering procedure with slight modifications. Comparing the application results obtained using the derived formula with the three-dimensional finite element analysis results, it is found that the results obtained using the derived formula agree well with the results of the finite element analysis.

• Nuclear Engineering and Technology
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• v.49 no.3
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• pp.638-644
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• 2017

The importance of developing a source-term assessment technology has been emphasized owing to the decommissioning of Kori nuclear power plant (NPP) Unit 1 and the increase of deteriorated NPPs. We analyzed the behavioral mechanism of corrosion products in the primary system of a pressurized heavy-water reactor-type NPP. In addition, to check the possibility of applying the CRUDTRAN code to a Canadian Deuterium Uranium Reactor (CANDU)-type NPP, the type was assessed using collected domestic onsite data. With the assessment results, it was possible to predict trends according to operating cycles. Values estimated using the code were similar to the measured values. The results of this study are expected to be used to manage the radiation exposures of operators in high-radiation areas and to predict decommissioning processes in the primary system.

• Nuclear Engineering and Technology
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• v.44 no.3
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• pp.311-322
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• 2012

Since the crisis at the Fukushima plants, severe accident progression during a station blackout accident in nuclear power plants is recognized as a very important area for accident management and emergency planning. The purpose of this study is to investigate the comparative characteristics of anticipated severe accident progression among the three typical types of nuclear reactors. A station blackout scenario, where all off-site power is lost and the diesel generators fail, is simulated as an initiating event of a severe accident sequence. In this study a comparative analysis was performed for typical pressurized water reactor (PWR), boiling water reactor (BWR), and pressurized heavy water reactor (PHWR). The study includes the summarization of design differences that would impact severe accident progressions, thermal hydraulic/severe accident phenomenological analysis during a station blackout initiated-severe accident; and an investigation of the core damage process, both within the reactor vessel before it fails and in the containment afterwards, and the resultant impact on the containment.

• Nuclear Engineering and Technology
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• v.53 no.1
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• pp.351-356
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• 2021

Korean nuclear power plants (NPPs) have various radiation protection programs to attain radiation exposure as low as reasonably achievable (ALARA). In terms of ALARA, this paper provides a comprehensive overview of administrative dose control for occupationally-exposed workers in Korean NPPs. In addition to dose limits, administrative dose constraints are implemented to resolve an inequity of radiation exposure in which some individuals in NPPs receive relatively higher doses than others. Occupational dose constraints in Korean NPPs are presented in this paper with the background of how those values were determined. For pressurized water reactors, 80% and 90% of the annual average limit for an effective dose, 20 mSv/y, are set as the primary and secondary dose constraints, respectively. Pressurized heavy water reactors (PHWRs) have also established the primary and secondary dose constraints corresponding to 70% and 80% of the effective dose limit, and additional constraints for tritium concentration are provided to control internal exposure in PHWRs. Follow-up measures for exceeding these administrative dose constraints are also introduced compared to exceeding the dose limits. Finally, analysis results of dose distributions show how the implementation of administrative dose constraints impacted the occupational dose distributions in Korean NPPs during the years 2009-2018.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.7 no.2
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• pp.14-20
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• 2011

The end of design lifetime for Wolsong unit 1 will be reached on 20th November in 2012. So the license renewal documents for the continuous operation of Wolsong unit 1 is under reviewing now. Major components of primary system such as pressure tubes, feeder pipes including delayed neutron monitoring system tubing are being replaced and many components of secondary system are also being repaired. In this paper, the assessment on the wear degradation of delayed neutron monitoring system tubing(on the other hand, DN tube was called) was performed for the ageing management of the same component. The wear defects of this component was one of causes that resulted in heavy water leakage accidents. Therefore design specifications of Wolsong uint 1 and heavy water leakage accidents of pressurized heavy water reactors were reviewed and causes of wear defect for DN tubes were analyzed. Wear propagation equations based on the heavy water leakage history were made and the proper repairing time was possible to be expected if the continued operation was considered. Finally design change items of DN tubes that were conducted for the long term operation of Wolsong unit 1 are introduced.