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[en] The Dukovany nuclear power plant was put into operation in 1985-1987. These are 4 units of WWER-440 reactor type. The specific feature of the WWER-440 design is six reactor cooling loops, that means each reactor is connected to six circulating pumps and six steam generators (SGs). Equipment DEKOZ PG, as shown, was designed for the chemical decontamination of the SGs of the plant. It separates the primary part of SG from the remaining part of the reactor cooling loop, serves for filling and draining the decontamination solutions into and from SG separated section and ensures also their circulation. Used decontamination solutions are drained by compressed air from the SG into the liquid waste draining lines. The device does not allow the recovery of the used decontamination solutions.
[en] After the Fukushima accident, one of the highest priorities for CERNAVODA NPP was to investigate events that can lead to Spent Fuel Bay (SFB) loss of cooling and loss of coolant inventory. In CANDU plants, fuelling is performed on-power. Daily, fresh fuel bundles are loaded in core and spent fuel bundles are discharged from the core, transferred and stored in SFB. Due to the SFB limited storage capacity, bundles having 6 years or more of cooling time are transferred to the Dry Storage Facility. Thus, as per design, a maximum number of around 38 000 fuel bundles can be stored, at any time, in SFB. Following a loss of class III and class IV power sources (e.g. Station Blackout), the cooling and purification systems for SFB water become unavailable. Consequently, the bay water temperature increases up to the boiling conditions and, due to boiling and vaporization, the water inventory and level will decrease in time. The decrease of coolant level can leave uncovered a number of fuel bundles, degrading their cooling. The present paper reviews the analysis methodology and results for a typical event of Spent Fuel Bay loss of cooling. Methodologies used in the analysis and results presented are focused upon the CANDU fuel thermal-hydraulic behaviour during the event and upon its potential radiological hazard. (author)
[en] The Fukushima accident raised a concern on severe accident risk of a spent fuel pool (SFP), since the earthquake may breach the pool boundary and/or stop cooling the pool due to loss of AC power. Since then, a substantial effort has been made to analyse severe accidents which may occur in SFPs. In this context, the present study was intended to assess the severe accident risk of the SFP in a Nordic boiling water reactor (BWR). Two accident scenarios of risk importance, namely loss of cooling flow accident (LOF) and loss of coolant accident (LOCA) due to 0.01 m2 breach at the bottom of the pool, were simulated by two different MELCOR versions 1.8.6 and 2.2. The results show that the fuel degradation occurs at ~55 h and ~3 h for the LOF scenario and the LOCA scenario, respectively. Larger amount of H2 generation was predicted in the LOF (c.a. 1500 kg) than the LOCA (c.a. 400~450 kg). A sensitivity study on the breach size showed that a larger breach size led to earlier fuel degradation but less H2 generation. The comparison of the simulation results from the two MELCOR versions indicated that the transients of water level in the pool were similar, but the fuel degradation began earlier and more H2 was produced in MELCOR 2.2 simulation. (author)
[en] The accident at the Japanese nuclear power plant (NPP) Fukushima-1 in March 2011 showed that possibility of accidents with potentially serious radiation consequences could not be excluded with large-scale measures for improvement of safety level. For spent nuclear fuel storage facilities, one of such accidents may be the interruption of heat removal from spent nuclear fuel (SNF) due to the failure of the cooling system as a result of disruption of the power supply system with the failure of backup power sources or rapid full dehydration of the wet SNF storage as a result of the destruction of building structures and its depressurization. The decision to take preventive measures in advance to minimize exposure to personnel and the public is based on conservative estimates of possible radioactive discharges. To perform such assessments, the operating organizations carry out a calculated justification of the thermal and hydraulic characteristics of the SNF system in the accident scenarios with long-term blackout and a violation of heat removal. APROS is one of the software tools that are used in SEC NRS for calculating the thermal-hydraulic characteristics of systems in transient modes by solving the equations of heat and mass transfer in a steam-water mixture. For more detailed calculations of the structural elements of spent fuel assemblies (SFA) temperature, the ANSYS software is used, which implements the finite element method. The results obtained with the help of the above simulation tools are used by specialists of SEC NRS to assess the protective measures developed by operating organizations.
[en] The International Atomic Energy Agency (IAEA) and the Generation IV International Forum (GIF) have jointly committed to collaboration between their respective programmes, and to share information in selected areas of mutual interest. One of the key areas of emphasis in both the GIF and the IAEA programmes is the safety of liquid metal cooled fast reactors (LMFRs) including sodium cooled fast reactors (SFRs) and lead or lead-bismuth eutectic (LBE) cooled fast reactors (LFRs). A particularly important area of mutual interest is the harmonization of safety approaches, safety requirements, Safety Design Criteria (SDC), and Safety Design Guidelines (SDG) for the next-generation advanced LMFRs under development worldwide. This topic has gained increased importance in the aftermath of the accident that occurred in 2011 at the Fukushima Daiichi nuclear power plant, which drew renewed attention to nuclear safety and to the importance of an international safety framework for reactors currently in operation as well as for new designs.
[en] We demonstrate a simple, low-cost, and passive radiative cooler based on a monolithic design consisting of thin nanoporous anodic alumina (NAA) films grown on aluminium sheets. The NAA/Al structure maintains a high broadband reflectivity close to 98% within the solar spectrum (0.4–2.2μm) and simultaneously exhibits a high average emissivity of 88% within the atmospheric infrared (IR) transmission window of 8–13μm with the peak IR emission approaching 99% at a wavelength of 10μm. Optical modelling of the system using optical parameters of the materials confirms that the high solar reflectance arises due to the transparent nature of NAA and high reflectivity of bottom Al, while the large thermal IR emissivity arises from the interference effects of the NAA film and the high absorption of IR light due to phonon resonances in alumina at wavelength larger than 10μm. Further, we estimate the average cooling power of NAA/Al to be about 136 W m−2 at ambient temperature even after including the contribution to heat input from external non-radiative processes. This robust and light weight NAA/Al can be projected as an excellent alternative to optical solar reflectors used in spacecraft for thermal heat management and rooftop cooling green technologies. (author)
[en] Since the Fukushima Daiichi accident, increased attention has been paid to the vulnerability of the Spent Fuel Pools (SFP). This vulnerability is a concern for SFPs safety because generally the fuel clad is the sole barrier against fission product release in case of dewatering. Also, the potential source term is several times the one present in the reactor vessel. For example, French SFPs can harbour up to 2.5 times the number of fuel assemblies present in the core of a 900 MW(e) reactor. The IAEA “International Experts Meeting on Strengthening Research and Development Effectiveness in the Light of the Accident at the Fukushima Daiichi Nuclear Power Plant”, held in Vienna in 2015, concluded that one priority is to investigate SFP loss of coolant and loss of cooling accidents. The OECD/NEA edited a Status Report on Spent Fuel Pools under Loss-of-Cooling and Loss-of-Coolant Accident Conditions and afterward gathered and expert group to establish a Phenomena Identification and Ranking Table on Spent Fuel Pools under Loss-of-Cooling and Loss-of-Coolant Accident Conditions. Several R&D programs dealing with SFP accidents were carried out. This PIRT concluded to the needs of further R&D concerning SFP accidental conditions and prioritized the topics to be investigated.
[en] Contents: National status of SFR development in Japan. Prospects of Japan Sodium-cooled Fast Reactor -Introduction: JAEA has developed conceptual design of an advanced loop-type SFR, named JSFR.; R&Ds for innovative technologies adopted in JSFR have been conducted as well as design study for improving maintainability and repairability and safety measures based on lessons learned from the Fukushima Daiichi Nuclear Power Plants accident.; JAEA is developing the design concept of a pool-type SFR based on the technology obtained from the above.; The development of this concept will broaden not only options for reactor types in Japan but also the range and depth of international cooperation. Design concept of a pool-type SFR; Study of reactor structure - Structural design and seismic evaluation, - Thermal hydraulic evaluation; Study of safety design - Safety design concept, - Applicability evaluation of SASS. Concluding Remarks: JAEA is developing the design concept of pool-type sodium-cooled fast reactor (SFR) that addresses Japan's specific siting conditions such as earthquakes and meets safety design criteria (SDC) and safety design guidelines (SDGs) for Generation IV SFRs.; The development of this concept will broaden not only options for reactor types in Japan but also the range and depth of international cooperation.
[en] Fuel pool cooling is an essential task in the scope of nuclear power applications. During the first years of commercial nuclear power implementation robust fuel pool cooling systems have been developed and used for several decades. Two decades ago the development of a new cooling technology/concept was initiated to ensure prevention of accidents, including fuel damage. The so-called advanced cooling technology offers a modular design system which enables tailor-made robust and cost efficient cooling solutions. However, all the advanced cooling systems feature an indispensable and distinctive fall back option of a passive heat removal in case of a station blackout as most important feature. In contradiction to conventional cooling systems the advanced cooling solutions use immersed heat exchangers to establish an additional safety barrier inside the heat removal chain. This results in the necessity of a free convective heat transfer on pool water side. This in turn requires a special design approach and methodology. Because of the huge nominal heat load and the size of the heat removal systems itself full size test are under economical aspects nearly impossible. In this paper a purpose-built simulation and design methodology is presented, which has been developed and proved in the scope of several first-of-a-kind projects during the last years.
[en] It is well-known that Loss of Coolant Accident (LOCA) has a significant contribution to fuel damage. Not only LOCA in the reactor cooling system but also LOCA in the spent fuel pool (SFP) need to be evaluated since it stores a lot of spent fuels which contain significant amount of radionuclide. In our previous studies, ART Mod 2 were modified and validated to ensure accurate calculation of fission product behaviour. The objective of this study is to assess fission product behaviour in the SFP during LOCA (or complete draining) using ART Mod 2. The geometry and conditions of the Robert Emmett Ginna Nuclear Power Plant is used as the reference. Caesium iodide (CsI) in gas form and caesium hydroxide (CsOH) in aerosol form are used to represent caesium compounds. It is found that modified ART Mod 2 code can capture the trend of CsI gas release into environment, but not that of CsOH aerosols. Total caesium compounds release and retention in all forms can be more accurately estimated if source term ratio of gas and aerosol, difference between wall and ambient temperatures, and chemical reactions are appropriately considered. (author)