atw - International Journal for Nuclear Power | 05.2019

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atw Vol. 64 (2019) | Issue 5 ı May DECOMMISSIONING AND WASTE MANAGEMENT 274 climate, radioactive material release data etc., and then to establish countermeasures to minimize the impact. Even if counter measures are properly established, the radiation surveillance of the surrounding area should be continuously carried out to confirm the validity and sustainability of the impact assessment to be performed [3, 4]. The radiological impact assessment plan for population needs a rough description of the following points: pp Exposure scenarios with potential exposure pathways. pp Population exposure evaluation method during the normal operation. pp Population exposure evaluation method during the accident. pp Plan to minimize radiological impacts. 3.12 Fire protection Since there are combustible materials such as electric wires and PVC pipes in nuclear facilities, there is a possibility of a potential fire accident when decommissioning as follows [6, 7]; pp Fire risk due to electric short circuit. pp Fire hazard when cutting with oxygen-acetylene torch. To minimize the likelihood of fire accident during the decommissioning phase, flammable materials should not be stored in decommissioning facilities unless stored in a refractory installation. In order to prevent the possibility of fire accident due to short circuit, all existing power supply is cut off and a separate external power supply is used. Since additional flammable and ignitable materials can be used in the decommissioning of nuclear facilities, the fire protection plan applied to the operating nuclear facility must be modified in consideration of decommissioning charac teristics. Therefore, the preliminary DP should outline the fire protection plan for prevention, detection, and evolution of fires that may occur during the decommissioning process, taking into consideration the characteristics of the expected decommissioning activities. 4 Conclusion remarks The decommissioning plan (DP) is divided into the preliminary DP and the final DP according to the construction/operation phase of the nuclear facility in Korea. It is described in detail from the preliminary DP to the final DP. The most important factor in preparing the DP is to make full use of the design/construction/operation data of the nuclear facilities. Therefore, in this study, it is reviewed the major safety considerations such as safe dismantling activities of nuclear facilities, dismantling procedures and dismantling methods, which is necessary for the preparation of the DP through the review of local and oversea decommissioning lessons learned experience. Since the preliminary DP must be submitted at the time of applying for the construction phase of the new nuclear facilities, there is a limit to the depth of the technical contents of each item in comparison with the final DP. Nonetheless, the preliminary DP should include the expected decommissioning strategy, the appropriateness of decommissioning, securing decommissioning resources, decommissioning safety and radiation protection plans, and the amount of decommissioning waste generation. Since the most important input to prepare the DP is the design/construction/ operation data, these data are carefully maintained over their lifetime in accordance with the relevant quality assurance procedures. Therefore, it is necessary to describe the management programs of these data in the preliminary DP. The safety consideration for the preparation of the preliminary DP is reviewed and its preparation guideline is established. However, in order to prepare a preliminary DP for existing nuclear facilities in Korea, it is necessary to draw up important factors for enhancing decommis sioning safety and efficiency by conducting a conceptual decommis sioning design, when prepare the DP. And it needs to know in advance what the key design and operational data related to these factors are. For example, the database showing the contamination information inside the nuclear facility is a basic data for evaluating the facility characteristics, managing the radiation of the workers, and evaluating the amount of waste disposal. Therefore, it is necessary to understand and supplement the current status on the database in the nuclear facilities. For all important factors it is necessary to construct and operate related database system to manage them for decommissioning activity from the design stage of nuclear facilities. Acknowledgments This work was supported by the Nuclear Safety Research Program through the Korea Foundation of Nuclear Safety (KoFONS), granted financial resource from the Nuclear Safety and Security Commission ( NSSC) (No. 1605008-0318-SB110), and by the National Research Foundation of Korea (NRF), granted financial resource from the Ministry of Science, ICT and Future Planning (No. 2017M2A8A5015148 and No. 2016M2B2B1945086), Republic of Korea. References 1. NSSC Notice 2015-8 (2015), Standard Format and Content of Decommissioning Plan for Nuclear Facilities, Korea Nuclear Safety and Security Commission. 2. IAEA (2005), Standard Format and Content for Safety Related Decommissioning Documents. IAEA Safety Reports Series No. 45 3. IAEA (2014), Decommissioning of Facilities “General Safety Requirements”. IAEA Safety Standard Series No. GSR Part 6 4. IAEA (1999), Safety Guide on Decommissioning of Nuclear Power and Research Reactors. Safety Standard Series No. WS-G-2.1 5. EPRI (2001), Decommissioning Pre-Planning Manual. 1003025, Final Report 6. EPRI (2006), Decommissioning Planning, Experiences from U.S. Utilities. 1013510, Final Report 7. GRS (2009), Guide to the Decommissioning, the Safe Enclosure and the Dismantling of facilities or parts thereof as defined in Section 7 of the Atomic Energy Act. Authors Byung-Sik Lee 1 Dankook University 119, Dandae-ro, Dongnam-gu, Cheonan-si Chungnam, 31116 Republic of Korea Kyung-Woo Choi 2 Korea Institute of Nuclear Safety 62 Gwahak-ro, Yuseong-gu Daejeon, 34142 Republic of Korea Decommissioning and Waste Management Guideline to Prepare a Preliminary Decommissioning Plan for Nuclear Facilities in Republic of Korea ı Byung-Sik Lee and Kyung-Woo Choi
atw Vol. 64 (2019) | Issue 5 ı May PARCS-Subchanflow-TRANSURANUS Multiphysics Coupling for High Fidelity PWR Reactor Core Simulation: Preliminary Results Joaquín R. Basualdo, Victor H. Sánchez, Robert Stieglitz and Rafael Macián-Juan 1 Introduction Traditionally, reactor core simulators use simplified models to predict the fuel temperature and thermal-hydraulic conditions in the core. To achieve better accuracy detailed models should be used to describe all different physical processes (Multiphysics approach). Simplified solvers for the fuel temperature don’t capture the material behavior under irradiation such as, swelling, cracking, pellet-clad interaction, etc. These phenomena affect properties such as fuel thermal conductivity, the fuel rod gap conductance which has an impact in the calculation of the fuel temperature. It is known that the gap conductance during reactor lifetime, depends strongly on the irradiation and power history as shown for instance in [Bielen 2015] thermal-hydraulic, and fuel thermo-mechanical behavior of the core components. Typically in current generation reactor physics analysis these three component areas are given separate consideration or are at best loosely coupled. Within this work, a methodology for tightly coupling the core neutronics code PARCS, thermalhydraulics code PATHS, and fuel rod simulator code FRAPCON was developed. This coupled code package, referred to as FRAPARCS, was applied to two fuel depletion problems: a pin cell and a 5x5 assembly mini-core. The results of the depletion calculations indicate that standalone PARCS does not adequately capture the evolution of fuel rod behavior which influences the Doppler fuel temperature used in cross section evaluation, and as a result significant differences in computed core performance can be seen. In particular, the behavior of the fuel-cladding gap and associated temperature drop was found to be important. FRAPARCS was then applied to the pin cell calculation to evaluate the uncertainty and sensitivity of the nuclear performance of the core due to the influence of fuel thermo-mechanical models available for manipulation in FRAPCON. A sensitivity study was conducted to determine which fuel models were influential on the neutronics outputs; we determined that fuel thermal conductivity, fuel thermal expansion, cladding creep, and fuel swelling had an important influence on the core Doppler temperature and reactivity. Additionally, the heat transfer coefficient was found to be important. Then, FRAPARCS was integrated within the DAKOTA uncertainty package. Two varieties of sampling-based methods (Random and Latin Hypercube Sampling). There are only few publications about mutiphysics simulations in the area of fuel behavior studies [e.g. Magedanz et al. 2015; Hales et al. 2014] and, even less containing studies of reactor core simulations [Holt et al. 2016; Holt et al. 2014]. In an evolutionary approach, at the Karlsruhe Institute for Technology (KIT), the NRC’s neutronics core simulator PARCS [Downar et al. 2012] is being integrated with KIT’s subchannel code SUBCHANFLOW (SCF) [Imke, Sanchez, and Gomez-Torres 2010] and ITU’s fuel behavior code TRANSURANUS (TU) [Lassman et al. 1992] into a single code, PARCS- SCF- TU. For the SCF model, each fuel assembly is represented as a single channel and, analogously, a fuel assembly in the TU model is represented as an average fuel rod. One of the objectives of this coupling is to study the impact that high fidelity solvers have on reactor core simulations. Moreover, a main objective of this coupling is the modeling of the RIA transient scenario for high burnup conditions. For this scenario, the fuel properties and fuel temperature modeling are of great importance since current simulations don’t account for details of the fuel rod thermos-mechanics and subchannel thermal hydraulics. The need of this kind of calculations is an issue brought up in recent years by the CSNI ( Committee on the Safety of Nuclear Installations) Working Group on Fuel Safety [OECD/NEA 2010] and a topic under discussion for regulatory authorities in many countries in Europe. In this paper, results for the OECD/ NEA and U.S. NRC PWR MOX/UO2 core transient benchmark core are used to compare PARCS-SCF and PARCS-SCF-TU with the PARCS standalone solution. Preliminary results are given, which show the impact of modeling the fuel temperature with a fuel behavior code considering burnup. 2 Methodology The neutronics core simulator PARCS, the sub-channel solver SCF and the fuel behavior solver TU have been merged together into a single executable PARCS-SCF-TU. In this Multiphysics coupling, SCF replaces the simple thermal hydraulic solver of PARCS and TU replaces the fuel rod solver of SCF to compute the fuel and cladding temperature distributions. The involved codes are written in FORTRAN using different programming styles and FORTRAN versions. The internal coupling has been developed in Microsoft Visual Studio following its convention for solutions and projects management. To maintain an organized coding and avoid undesired callings to duplicate subroutines or variable names a modularized approach is used. The original codes are encapsulated in projects and they only interact with each other via a main project. Only in special circumstances this rule isn’t followed. New coding necessary for the communication of the codes is modularized in a project dedicated to the coupling. All modifications to the original source code were implemented with pre-compiler directives. This allows the user to compile either only PARCS, or only PARCS-SCF or PARCS-SCF-TU depending on the used keywords. 275 RESEARCH AND INNOVATION Research and Innovation PARCS-Subchanflow-TRANSURANUS Multiphysics Coupling for High Fidelity PWR Reactor Core Simulation: Preliminary Results ı Joaquín R. Basualdo, Victor H. Sánchez, Robert Stieglitz and Rafael Macián-Juan
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