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[en] Cs-137 is one of the most intensively studied Isotopes used when investigating fission product transport behavior, and much experimental data are available. For LEU fuel, the main sources of information are the measurements of Brown and Faircloth (1976) of the United Kingdom Atomic Energy Authority (UKAEA), Harwell, on 80% dense UO2 prepared using the powder agglomeration technique. The kernel release was determined by cracking irradiated, intact, TRISO coated particles and measuring the Cs-137 content in the kernel and the coating.
[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] Studtite is known to exist at the back-end of the nuclear fuel cycle as an intermediate phase formed in the reprocessing of spent nuclear fuel. In the thermal decomposition of studtite, an amorphous phase is obtained at calcination temperatures between 200 and 500 °C. This amorphous compound, referred to elsewhere in the literature as U2O7, has been characterised by analytical spectroscopic methods. The local structure of the amorphous compound has been found to contain uranyl bonding by X-ray absorption near edge (XANES), Fourier transform infrared and Raman spectroscopy. Changes in bond distances in the uranyl group are discussed with respect to studtite calcination temperature. The reaction of the amorphous compound with water to form metaschoepite is also discussed and compared with the structure of schoepite and metaschoepite by X-ray diffraction. A novel schematic reaction mechanism for the thermal decomposition of studtite is proposed. (author)
[en] This publication summarizes the results of an IAEA technical meeting to review and discuss the analysis, simulation, and modelling of severe accident progression in spent fuel pools. The emphasis was on achieving a better understanding of drivers for improvement to address risks associated with accidents in spent fuel pools, progression to failure of the spent fuel, and the subsequent release of fission products. Discussion sessions enabled the exchange of information regarding the analysis of severe accidents in spent fuel pools, the provision of an overview of current research and development (R&D) activities, and considerations for the planning and execution of further R&D. The meeting served as a forum for Member States to exchange knowledge on current and new code development and methodologies, to identify the gaps for future improvements, and gather information for collaboration on all these aspects.
[en] Acetohydroxamic acid (AHA) is a hydrophilic organic complexing agent with excellent masking or stripping effects on the tetravalent actinides in the extraction operations. Therefore, AHA is a promising reagent for the spent nuclear fuel reprocessing. In the reprosessing of spent uranium-aluminum fuel with a high U-235 enrichment which is used in research reactor, the dissolver solution contains a high concentration of Al(NO3)3, several orders of magnitude higher than that of minor actinides such as plutonium, and dilute tri-n-butyl phosphate(TBP)/n-dodecane is adopted as the extractant. In this research, complexation of Pu(IV) with AHA under the condition of U-Al fuel reprocessing is investigated. Results show that Al(NO3)3 insignificantly affects the mentioned complexation reaction. Further, the possibility of uranium purification against Pu(VI) is also investigated by using AHA in the presence of a high concentration of Al(NO3)3 by simulating some experiments on the separation process. (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] HANARO has been actively utilized since attaining first criticality in 1995. In 2009, the cold neutron source was installed inside the reflector tank. The main utilization fields of HANARO are neutron beam applications, nuclear fuel and material test, radioisotope production, neutron activation analysis and neutron transmutation doping. After the Fukushima accident, HANARO had been requested to evaluate the seismic margin for the reactor’s main components, and the seismic margin assessment led to the reinforcement of HANARO’s wall. In December 2017, HANARO started operating after overcoming many issues for about a 3 year-shutdown. For the period of long-term shutdown, the circumstances surrounding HANARO, such as research reactor regulation and new research reactor construction project, has been drastically changed. To meet the expectations of HANARO to produce world-class science and to respond to the rapidly changing environment, a strategic plan for HANARO was prepared, and 4 missions were re-established. They are 1) advancement of neutron science and technology, 2) not only meeting but also creating the needs of the industry, 3) contributions to the national society issues, and 4) safe and stable operation of the facility. To achieve the missions, all members involved in the HANARO shared their same perception that the stable operation of HANARO is the most important. HANARO is trying hard to give the confidence for its sustainability and excellence through the various activities. (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] This paper presents the results of the modelling of the loss of cooling accident in at-reactor spent fuel pool of VVER-1200. The calculations of the accident were performed with 3 Russian codes: best-estimate severe accident code SOCRAT/V1 (processes in the spent fuel pool), containment lumped-parameters code ANGAR (parameters of the containment atmosphere during the accident) and calculation code GEFEST-ULR (molten corium-concrete interaction). Such combination of codes allowed to perform complex evaluation of the severe accident: from the initial event (loss of cooling) to the melt-through of the spent fuel pool concrete bottom, taking into account change of the containment atmosphere during the accident. (author)