• Nuclear Engineering and Technology
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• v.51 no.1
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• pp.125-137
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• 2019

The purpose of this paper is to study the source term behavior after severe accidents by using a semi-kinetic model for simulation and calculation of in-containment activity. The reactor containment specification and the safety features of the containment under different accident conditions play a great role in evaluating the in-containment activity. Assuming in-vessel and instantaneous release of radioactivity into the containment, the behavior of in-containment isotopic activity is studied for noble gasses (Kr and Xe) and the more volatile elements of iodine, cesium, and aerosols such as Te, Rb and Sr as illustrative examples of source term release under LOCA conditions. The results of the activity removal mechanisms indicates that the impact of volumetric leakage rate for noble gasses is important during the accident, while the influence of deposition on the containment surfaces for cesium, mainly iodine isotopes and aerosol has the largest contribution in removal of activity during evolution of the accident.

• Nuclear Engineering and Technology
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• v.44 no.3
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• pp.249-260
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• 2012

Main focus of the combined nuclear research activities at Aachen University (RWTH) and the Research Center J$\ddot{u}$lich (J$\ddot{U}$LICH) is the experimental and analytical investigation of containment phenomena and processes. We are deeply convinced that reliable simulations for operation, design basis and beyond-design basis accidents of nuclear power plants need the application of so-called lumped-parameter (LP) based codes as well as computational fluid dynamics (CFD) codes in an indispensable manner. The LP code being used at our institutions is the GRS code COCOSYS and the CFD tool is ANSYS CFX mostly used in German nuclear research. Both codes are applied for safety analyses especially of beyond design accidents. Focal point of the work is containment thermal-hydraulics, but source term relevant investigations for aerosol and iodine behavior are performed as well. To increase the capability of COCOSYS and CFX detailed models for specific features, e.g. recombiner behavior including chimney effect, building condenser, and wall condensation are developed and validated against facilities at different scales. The close connection between analytical and experimental activities is notable and identifying feature of the RWTH/J$\ddot{U}$LICH activities.

• Nuclear Engineering and Technology
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• v.41 no.8
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• pp.995-1004
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• 2009

The Advanced Power Reactor 1400 (APR1400) is an evolutionary advanced light water reactor (ALWR) based on the Optimized Power Reactor 1000 (OPR1000), which is in operation in Korea. The APR1400 incorporates a variety of engineering improvements and operational experience to enhance safety, economics, and reliability. The advanced design features and improvements of the APR1400 design include a pilot operated safety relief valve (POSRV), a four-train safety injection system with direct vessel injection (DVI), a fluidic device (FD) in the safety injection tank, an in-containment refueling water storage tank (IRWST), an external reactor vessel cooling system, and an integrated head assembly (IHA). Development of the APR1400 started in 1992 and continued for ten years. The APR1400 design received design certification from the Korean nuclear regulatory body in May of2002. Currently, two construction projects for the APR1400 are in progress in Korea.

• Nuclear Engineering and Technology
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• v.46 no.2
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• pp.207-216
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• 2014

This paper revisits the Fukushima accident to draw lessons in the aspect of nuclear safety considering the fact that the Fukushima accident resulted in core damage for three nuclear power plants simultaneously and that there is a high possibility of a failure of the integrity of reactor vessel and primary containment vessel. A brief review on the accident progression at Fukushima nuclear power plants is discussed to highlight the nature and characteristic of the event. As the severe accident management measures at the Fukushima Daiich nuclear power plants seem to be not fully effective, limitations of current severe accident management strategy are discussed to identify the areas for the potential improvements including core cooling strategy, containment venting, hydrogen control, depressurization of primary system, and proper indication of event progression. The gap between the Fukushima accident event progression and current understanding of severe accident phenomenology including the core damage, reactor vessel failure, containment failure, and hydrogen explosion are discussed. Adequacy of current safety goals are also discussed in view of the socio-economic impact of the Fukushima accident. As a conclusion, it is suggested that an investigation on a coherent integrated safety principle for the severe accident and development of innovative mitigation features is necessary for robust and resilient nuclear power system.

• Nuclear Engineering and Technology
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• v.45 no.6
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• pp.795-802
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• 2013

For the improvement of APR1400 Diverse Protection System (DPS) design, the Advanced DPS (ADPS) has recently been developed to enhance the fault tolerance capability of the system. Major fault masking features of the ADPS compared with the APR1400 DPS are the changes to the channel configuration and reactor trip actuation equipment. To minimize the fault occurrences within the ADPS, and to mitigate the consequences of common-cause failures (CCF) within the safety I&C systems, several fault avoidance design features have been applied in the ADPS. The fault avoidance design features include the changes to the system software classification, communication methods, equipment platform, MMI equipment, etc. In addition, the fault detection, location, containment, and recovery processes have been incorporated in the ADPS design. Therefore, it is expected that the ADPS can provide an enhanced fault tolerance capability against the possible faults within the system and its input/output equipment, and the CCF of safety systems.

• Nuclear Engineering and Technology
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• v.30 no.1
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• pp.58-74
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• 1998

A PC window-based computer code, CONPAS (CONtainment Performance Analysis System), has been developed to integrate the numerical, graphical, and results-operation aspects of Level 2 probabilistic safety assessments (PSA) for nuclear power plants automatically. As a main logic for accident progression analysis, it employs a concept of the small containment phenomenological event tree (CPET) helpful to trace out visually individual accident progressions and of the detailed supporting event tree (DSET) for its detailed quantification. For the integrated analysis of Level 2 PSA, the code utilizes five distinct, but closely related modules. Its computational feasibility to real PSAs has been assessed through an application to the UCN 3&4 full scope Level 2 PSA. Compared with other existing computer codes for Level 2 PSA, the CONPAS code provides several advanced features: (1) systematic uncertainty analysis / importance analysis / sensitivity analysis, (2) table / graphical display & print, (3) employment of the recent Level 2 PSA technologies, and (4) highly effective user interface. The main purpose of this paper is to introduce the key features of CONPAS code and results of its feasibility study.

• Nuclear Engineering and Technology
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• v.45 no.5
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• pp.625-636
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• 2013

Pressing demands of economic competitiveness, the need for large-scale deployment, minimizing the need of human intervention, and experience from the past events and incidents at operating reactors have guided the evolution and innovations in reactor technologies. Indian innovative reactor 'AHWR' is a pressure-tube type natural circulation based boiling water reactor that is designed to meet such requirements, which essentially reflect the needs of next generation reactors. The reactor employs various passive features to prevent and mitigate accidental conditions, like a slightly negative void reactivity coefficient, passive poison injection to scram the reactor in event of failure of the wired shutdown systems, a large elevated pool of water as a heat sink inside the containment, passive decay heat removal based on natural circulation and passive valves, passive ECC injection, etc. It is designed to meet the fundamental safety requirements of safe shutdown, safe decay heat removal and confinement of activity with no impact in public domain, and hence, no need for emergency planning under all conceivable scenarios. This paper examines the role of the various passive safety systems in prevention and mitigation of severe plant conditions that may arise in event of multiple failures. For the purpose of demonstration of the effectiveness of its passive features, postulated scenarios on the lines of three major severe accidents in the history of nuclear power reactors are considered, namely; the Three Mile Island (TMI), Chernobyl and Fukushima accidents. Severe plant conditions along the lines of these scenarios are postulated to the extent conceivable in the reactor under consideration and analyzed using best estimate system thermal-hydraulics code RELAP5/Mod3.2. It is found that the various passive systems incorporated enable the reactor to tolerate the postulated accident conditions without causing severe plant conditions and core degradation.

• Nuclear Engineering and Technology
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• v.56 no.3
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• pp.803-811
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• 2024

Nuclear accidents such as Fukushima Daiichi have highlighted the potential of passive safety systems to replace or complement active safety systems as part of the overall prevention and/or mitigation strategies. In addition, passive systems are key features of Small Modular Reactors (SMRs), for which they are becoming almost unavoidable and are part of the basic design of many reactors available in today's nuclear market. Nevertheless, their potential to significantly increase the safety of nuclear power plants still needs to be strengthened, in particular the ability of computer codes to determine their performance and reliability in industrial applications and support the safety demonstration. The PASTELS project (September 2020-February 2024), funded by the European Commission "Euratom H2020" programme, is devoted to the study of passive systems relying on natural circulation. The project focuses on two types, namely the SAfety COndenser (SACO) for the evacuation of the core residual power and the Containment Wall Condenser (CWC) for the reduction of heat and pressure in the containment vessel in case of accident. A specific design for each of these systems is being investigated in the project. Firstly, a straight vertical pool type of SACO has been implemented on the Framatome's PKL loop at Erlangen. It represents a tube bundle type heat exchanger that transfers heat from the secondary circuit to the water pool in which it is immersed by condensing the vapour generated in the steam generator. Secondly, the project relies on the CWC installed on the PASI test loop at LUT University in Finland. This facility reproduces the thermal-hydraulic behaviour of a Passive Containment Cooling System (PCCS) mainly composed of a CWC, a heat exchanger in the containment vessel connected to a water tank at atmospheric pressure outside the vessel which represents the ultimate heat sink. Several activities are carried out within the framework of the project. Different tests are conducted on these integral test facilities to produce new and relevant experimental data allowing to better characterize the physical behaviours and the performances of these systems for various thermo-hydraulic conditions. These test programmes are simulated by different codes acting at different scales, mainly system and CFD codes. New "system/CFD" coupling approaches are also considered to evaluate their potential to benefit both from the accuracy of CFD in regions where local 3D effects are dominant and system codes whose computational speed, robustness and general level of physical validation are particularly appreciated in industrial studies. In parallel, the project includes the study of single and two-phase natural circulation loops through a bibliographical study and the simulations of the PERSEO and HERO-2 experimental facilities. After a synthetic presentation of the project and its objectives, this article provides the reader with findings related to the physical analysis of the test results obtained on the PKL and PASI installations as well an overall evaluation of the capability of the different numerical tools to simulate passive systems.

• Nuclear Engineering and Technology
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• v.49 no.8
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• pp.1575-1588
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• 2017

In order to develop a domestic research roadmap for severe accidents, a special committee was established by the Korean Nuclear Society. One of the subcommittees discussed the characteristics and the relevant technical issues in the stages of fission product release and physical forms of radionuclide release and transport. The group members developed a tree to identify fission product release phenomena by tracing failures of individual defense-in-depth barriers and added possible countermeasures against failure. For each elemental issue, they searched for technical problems by examining the phenomena, accident management actions, and regulatory aspects relevant to the mitigation features for containment, including mitigation strategies against containment bypass accidents. Regulatory concerns, including the source term and the acceptance criteria for radionuclide release, were also considered. They identified further research needs regarding important technical issues based on the degree of the current knowledge level in Korea and in foreign countries, looking at the significance and urgency of issues and the expected research period required to reach an advanced level of knowledge. As a result, the group identified the 12 most important and urgent issues, most of which were expected to require mid-term and long-term research periods.

• Nuclear Engineering and Technology
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• v.26 no.4
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• pp.516-525
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• 1994

Since the TMI and Chernobyl accidents, many passive safety features are suggested in advanced reactors in order to enhance the safety in future nuclear power plants. In order to verify the effectiveness and provide the data for detailed design of passive cooling system, in the present work, the effects of air inlet position and external condition on the natural circulated air flow rate and the natural and forced convective heat transfer coefficient have been investigated for the one-side heated closed path such as the passive containment cooling system of the Westinghouse's AP-600. A series of experiments have been peformed with the 1/26th scaled segment type test facility of the AP-600 passive containment. Under natural and forced convection, the air velocities and temperatures are measured at several points of the air flow path. The experimental result are compared with a simple one-dimensional model and it shows a good agreement.