Reactor Safety Research Research Articles

International assessment of application of the Code of Conduct on the Safety of Research Reactors

AbstractThe self-assessments performed by thirty-eight countries on application of the Code of Conduct on the Safety of Research Reactors were analyzed and discussed. The results of this analysis were used to identify areas of satisfactory application of the Code and area needing improvements, and therefore require more attention worldwide. The results showed improvement in application of the Code provisions; ntformat otably in aging management, regulatory supervision, and consideration of human factors. However, there is a continuing need for further improvement in these areas, as well as in operational radiation protection, emergency preparedness and decommissioning planning. Additionally, increased attention needs to be given to periodic safety reviews, evaluation of site-specific hazards, and assessment of extreme external events. The results showed consistency with the feedback from other sources of information on generic safety issues for research reactors.

Investigation of fragmentation phenomena and debris bed formation during core meltdown accident in SFR using simulated experiments

The event of a severe core melt down accident, resulting in the relocation of the active core is analyzed as a part of the nuclear reactor safety research in order to ensure safe removal of decay heat. Molten Fuel Coolant Interaction (MFCI) and debris bed configuration on the core catcher plate assumes importance in assessing the post accident heat removal capability. The key factors affecting the coolability of the debris bed are the bed porosity, morphology of the fragmented particles, degree of spreading/heaping of the debris on the core catcher and the fraction of lump formed. A well defined debris bed is helpful in fixing a prototypical source term for the PAHR studies. Towards this, a series of experiments on fragmentation kinetics and subsequent debris bed formation is conducted with molten Wood's metal (an alloy of Bi 50%, Pb 25%, Sn 12.5% and Cd 12.5% with melting point of 346K) in water simulant system. The experiments are carried out using 2kg, 5kg and 20kg melt inventories. The particle size distribution obtained for the fragmented debris is fit using an Upper Limit Log Normal (ULLN) distribution. The dependence of particle size distribution on initial melt temperature and interaction height is quantified by correlating them to the key parameters i.e. shape factor and location factor in the ULLN expression. Morphology of the debris particles is investigated to understand the fragmentation mechanisms involved. Three major mechanisms of fragmentation are identified namely melt entrainment mechanism, boundary layer stripping and hydrodynamic breakup due to capillary forces. Finally, an approach to quantify the stratification and spreading behaviour of debris on the collector tray is presented by analysing the particle size and mass percentage in different radial zones of the collector tray.

Application of best estimate plus uncertainty in review of research reactor safety analysis

To construct and operate a nuclear research reactor, the licensee is required to obtain the authorization from the regulatory body. One of the tasks of the regulatory authority is to verify that the safety analysis fulfils safety requirements. Historically, the compliance with safety requirements was assessed using a deterministic approach and conservative assumptions. This provides sufficient safety margins with respect to the licensing limits on boundary and operational conditions. Conservative assumptions were introduced into safety analysis to account for the uncertainty associated with lack of knowledge. With the introduction of best estimate computational tools, safety analyses are usually carried out using the best estimate approach. Results of such analyses can be accepted by the regulatory authority only if appropriate uncertainty evaluation is carried out. Best estimate computer codes are capable of providing more realistic information on the status of the plant, allowing the prediction of real safety margins. The best estimate plus uncertainty approach has proven to be reliable and viable of supplying realistic results if all conditions are carefully followed. This paper, therefore, presents this concept and its possible application to research reactor safety analysis. The aim of the paper is to investigate the unprotected loss-of-flow transients "core blockage" of a miniature neutron source research reactor by applying best estimate plus uncertainty methodology. The results of our calculations show that the temperatures in the core are within the safety limits and do not pose any significant threat to the reactor, as far as the melting of the cladding is concerned. The work also discusses the methodology of the best estimate plus uncertainty approach when applied to the safety analysis of research reactors for licensing purposes.

INVESTIGATION ON EFFECTS OF ENLARGED PIPE RUPTURE SIZE AND AIR PENETRATION TIMING IN REAL-SCALE EXPERIMENT OF SIPHON BREAKER

To ensure the safety of research reactors, the water level must be maintained above the required height. When a pipe ruptures, the siphon phenomenon causes continuous loss of coolant until the hydraulic head is removed. To protect the reactor core from this kind of accident, a siphon breaker has been suggested as a passive safety device. This study mainly focused on two variables: the size of the pipe rupture and the timing of air entrainment. In this study, the size of the pipe rupture was increased to the guillotine break case. There was a region in which a larger pipe rupture did not need a larger siphon breaker, and the water flow rate was related to the size of the pipe rupture and affected the residual water quantity. The timing of air entrainment was predicted to influence residual water level. However, the residual water level was not affected by the timing of air entrainment. The experimental cases, which showed the characteristic of partical sweep-out mode in the separation of siphon breaking phenomenon [2], showed almost same trend of physical properties.

Zinc corrosion after loss-of-coolant accidents in pressurized water reactors – Physicochemical effects

Within the framework of the reactor safety research, generic experimental investigations were carried out aiming at the physicochemical background of possible zinc corrosion product formation, which may occur inside the reactor pressure vessel during the sump circulation operation after loss-of-coolant accidents in pressurized water reactors. The contact of the boric acid containing coolant with hot-dip galvanized steel containment internals causes corrosion of the corresponding materials resulting in dissolution of the zinc coat. A retrograde solubility of zinc corrosion products with increasing temperature was observed during batch experiments of zinc corrosion in boric acid containing coolants. Thus, the formation and deposition of solid corrosion products cannot be ruled out if the coolant containing dissolved zinc is heated up during its recirculation into hot regions within the emergency cooling circuit (e.g. hot-spots in the core). Corrosion experiments at a lab-scale test facility, which included formation of corrosion products at a single heated cladding tube, proved that dissolved zinc, formed at low temperatures in boric acid solution by zinc corrosion, turns into solid deposits of zinc borates when contacting heated zircaloy surfaces during the heating of the coolant. Moreover, the temperature of formation influences the chemical composition of the zinc borates and thus the deposition and mobilization behavior of the products.

Estimation of Control Rod Worth in a VVER-1000 Reactor using DRAGON4 and DONJON4

Abstract One of the main issues in safety and control systems design of power and research reactors is to prevent accidents or reduce the imposed hazard. Control rod worth plays an important role in safety and control of reactors. In this paper, we developed a justifiable approach called D4D4 to estimate the control rod worth of a VVER-1000 reactor that enables to perform the best estimate analysis and reduce the conservatism that utilize DRAGON4 and DONJON4. The results are compared with WIMS-D4/CITATION to show the effectiveness and superiority of the developed package in predicting reactivity worth of the rod and also other reactor physics parameters of the VVER-1000 reactor. The results of this study are in good agreement with the plant's FSAR.

Measurement of neutronic safety parameters of the 3 MW TRIGA Mark-II research reactor

Measurement of neutronic safety parameters of the nuclear reactor is very important for reactor safety, efficient operation and experimental research. In this study, some neutronic safety parameters such as control rod worth, core excess reactivity, loss of reactivity with power increases, power defect, reactivity coefficients, cooling effect on fuel temperature and xenon poisoning were measured of the 3 MW TRIGA Mark-II research reactor of Bangladesh Atomic Energy Commission (BAEC). All of these safety parameters have significant effect on the reactor control system. Many of the reactor parameters which determine the reactivity of reactor are dependent on the temperature of the fuel, moderator and coolant. The TRIGA fuel elements have been designed to provide a significant prompt negative temperature coefficient of reactivity that allows safe reactor operation. The neutronic safety parameters were measured with digital instrumentation & control (I&C) system of the 3 MW TRIGA research reactor. The experimental results of neutronic parameters were found within the safety limit as mentioned in the Safety Analysis Report (SAR) of the BAEC 3 MW TRIGA research reactor. The present study will contribute to analyze the behavior of neutronic safety parameters that occur in the reactor core and to provide an experimental database for validation and improvement of the models of the reactor.

Measurement of control rod reactivity and shut down margin of 3 MW TRIGA Mark-II research reactor using analogue and digital I&C system

Measurement of reactor safety parameters is essential for reactor safety, operation and experimental research. Reactor control rod reactivity has been measured by the positive period method using analogue and digital instrumentation & control (I&C) system of the 3MW TRIGA Mark-II research reactor of Bangladesh Atomic Energy Commission (BAEC). The BAEC research reactor had been operating with the analogue I&C system since September 1986–July 2011. The analogue I&C system has been replaced by a new digital I&C system on June 2012 and the reactor is now operating with the PC based digital I&C system. In the newly installed digital I&C system several modifications, upgradations and replacement tasks were carried out including control rod drive systems. After the installation of the digital I&C system, some nuclear safety parameters (e.g., control rod worth, core excess reactivity and shut down margin) were measured to ensure the safe operation of the reactor. In this study, the measured control rod reactivity, core excess reactivity and shut down margin were also compared with the previous measured (analogue system) values for the validation of the digital I&C system of the reactor. The measured safety parameters were found within the safety limit as mentioned in the safety analysis report (SAR) of the BAEC research reactor.

Experimental database on steam–water flow with phase transfer in a vertical pipe

The evolution of steam–water flow with an initial sub-cooling along an 8m long pipe with an inner diameter of about 200mm was measured to obtain a database suitable for the qualification of CFD-codes on poly-dispersed flows with phase transfer. Phase transfer has to be considered in Nuclear Reactor Safety Research, e.g. in case of sub-cooled boiling or bubble entrainment caused by Emergency Core Cooling (ECC) injection. Bubble size distributions are important, since the phase transfer rate is proportional to the interfacial area density. To develop and validate closure models for CFD codes experimental data with high resolution in space and time are required. The experiments were conducted at the TOPFLOW facility of the Helmholtz–Zentrum Dresden–Rossendorf. Steam was injected into upwards flowing sub-cooled water via orifices in the pipe wall located at different distances from measuring plane. 1mm and 4mm injection orifices are used to vary the initial bubble size distribution. The variation of the distance between the location of the gas injection and the measuring plane allows investigating the evolution of the flow along the pipe. Pressure, steam and water flow rates and the sub-cooling were also varied. Measurements are done using wire-mesh sensors and thermocouples. Data on averaged void fraction, radial gas volume fraction profiles, profiles of the gas velocity and bubble size distributions in dependency of the L/D ratio are available.

Generation IV Reactor Safety and Materials Research by the Institute for Energy and Transport at the European Commission's Joint Research Centre

To support the drafting, development, implementation and monitoring of European energy and transport policy, the Institute for Energy and Transport of the European Commissions’ Joint Research Centre conducts pre-competitive research in the areas of experimental qualification of advanced fuels and materials as well as simulation and modelling of reactor safety and material performance. The work covers assessments, design optimisation and improvements to the safety and performance of new, innovative reactor systems, materials and instrumentation, in order to meet the EU's long-term energy needs while respecting enhanced safety, sustainability, and economic aspects.The research is linked, and contributes, to related EURATOM Framework Programme projects, Generation IV International Forum (GIF), International Atomic Energy Agency as well as OECD's Nuclear Energy Agency (OECD NEA) activities. The current paper gives an overview and examples of past, current, and upcoming activities in the areas of reactor safety assessments, advanced fuel irradiation and materials research.

Stress biaxiality in high-burnup PWR fuel cladding under reactivity-initiated accident conditions

In order to promote a better understanding of failure mechanisms of high-burnup pressurized water reactor (PWR) fuels under reactivity-initiated accident (RIA) conditions, stress biaxiality in cladding has been estimated for the pellet-cladding (PC) mechanical interaction (PCMI) phase. The estimation was based on an analysis of the transient elongations of a pellet stack and a cladding tube measured in RIA-simulating experiments in the nuclear safety research reactor (NSRR) using the RANNS code. Stress biaxiality in the high-burnup PWR fuel cladding during the PCMI phase has been estimated to be 0.7–0.8, on average, at the mid-wall of the cladding. A comparison with fresh fuel test results and a sensitivity analysis showed that the effects of burnup and pulse width on cladding stress biaxiality are less than 10% for the investigated range. The present analysis also indicated that PC friction is strong, and that the cladding constraint on pellet stack elongation is significant irrespective of burnup. Therefore, it is recommended that strong PC friction be assumed, which is similar to the mechanical bonding condition, and that fuel pellets be treated as deformable materials in models of fuel behavior during the PCMI phase.

Actual status of the research alliances “Condensation Induced Water Hammer” and “Boiling processes in Pressurized Water Reactors”

The German Nuclear Society (KTG) is a non-profit scientific technical association with more than 2500 members. KTG strives for the development and dissemination of the stateof-the-art in nuclear science and technology to promote the peaceful use of nuclear energy and related disciplines. The technical work of the KTG members is performed within the framework of currently 9 topical sections, which, for instance, jointly organize the KTG Annual Meetings, topical meetings devoted to selected subjects and several other activities. In recent years, the topical sections Reactor Safety, Thermal and Fluid Dynamics, as well as Reactor Physics and Computational Methods, have successfully established a series of meetings on \Topical Issues of Reactor Safety Research in Germany. These meetings represent a forum for authorities, operators, technical support organizations, universities and vendors. A Technical Program Committee (TPC) consisting from members of the participating topical sections is responsible for selecting the issues to be addressed at the meeting. The acceptance and success of this series of meetings is mainly due to the current subjects dealt with. In addition to the presentation of scientific and technical research results, this forum offers the chance for a free exchange of opinions outside licensing and supervisory procedures. A balance is presented between survey papers on the background and the state of the art in the related field, specific technical presentations, and outlook papers. The participants are provided with a general overview on the area addressed and with new ideas, and impetus for their future work. In order to further promote these conference series, the TPC has initiated co-operation with the journal KERNTECHNIK. The cooperation started in 2011 with a very successful edition of the topic \Requirements for the demonstration of effective emergency core cooling during loss-ofcoolant accidents involving the release of insulation material and other substances. Two contributions of this special issue were among the top five downloads in the period 2010 up to 2011 of KERNTECHNIK. The 2012 meeting on \Topical Issues of Reactor Safety Research in was held at the Helmholtz-Zentrum Dresden-Rossendorf (HZDR) on 11–12 October 2012, hosted by the Institute of Fluid Dynamics of HZDR and the TUV NORD SysTec GmbH & Co. KG and the section Thermal and Fluid Dynamics of the German Nuclear Society (KTG). At the meeting selected results of the two research alliances \Condensation induced water hammer – CIWA and \Boiling processes in PWR were presented. Both research alliances are sponsored by the Federal Ministry of Education and Research within the Energy Research 2020+ concept. Even though Germany is phasing out of domestic power generation from nuclear energy, it should be noted that the Federal Ministry of Education and Research continues to fund further scientific and technological R&D in support of safe operation of existing nuclear reactors. One of the most effective paths for ensuring high quality basic and applied research programs is to provide funding of multi-organizational, multi-disciplinary research alliances. In addition, special attention should be paid to the support of young scientists and researchers, and to the dissemination and transfer of research findings to industry. The selection of the main topics \Condensation Induced Water Hammer and \Boiling processes in PWR at the actual topical meeting was made by the following background: Pressure surges are still a cause of serious damages in both nuclear and conventional power generation plants. Understanding the mechanisms and underlying physics, as well as developing measures to prevent and/or mitigate their initiations are still challenging tasks for utilities and technical support organizations. Boiling processes play an important role in energy generating facilities were energy transfer is essential. The deeper understanding of the mechanisms not only increases the safety of the facility but also enables to enhance the effective operation. In the past the development of promising Computational Fluid Dynamics (CFD) models stagnated not least due to missing of CFD grade experiments. The research alliance \Boiling processes in PWR contributed in both directions: Small and large scale experiments and model development. For enhanced dissemination of the state-of-theart, selected papers of the meeting are published in this special issue of KERNTECHNIK. Both, the topical meeting as well as the now presented special issue of KERNTECHNIK is certainly an impressive proof that the strategic objectives of the Basic Energy Research 2020+ concept can be achieved. EDITORIAL

NAA Algerian Laboratory Evaluation Processed by WEPAL and IAEA during 2011-2012

Participation in inter-comparison programs is an important process to enhance the accuracy and precision of the analytical techniques used in the laboratories. At Es-Salam Research Reactor, our laboratory uses two functional analytical techniques such as INAA and k0-INAA in the research projects and for the completion of the service works. Recently, our laboratory has participated in two inter-comparison tests organized by IAEA within the framework of the AFRA project to assess the analytical performance of eighteen (18) African analytical laboratories participating in the RAF/4/ 022 project, Enhancement of Research Reactor Utilization and Safety by taking part in analytical proficiency testing IAEA in conjunction with WEPAL, the Wageningen Evaluating Programs for Analytical Laboratories. The Proficiency Testing tests related to the determination of major, minor and trace elements in materials of the International Soil and Plant Analytical Exchange (Wepal codes ISE, IPE), were held in the fourth quarter of 2011 and the first quarter of 2012, respectively. Concentrations of thirty (30) elements such as Al, As, Ba, Br, Ca, Cd, Ce, Co, Cr, Cs, Fe, Ga, K, La, Mg, Mo, Mn, Na, Nd, Rb, Sb, Sc, Sr, Th, Ti, U, V, W, Zn, Zr were determined in ISE and IPE samples. All results were evaluated and reported in this paper.

Experimental study of siphon breaking phenomenon in the real-scaled research reactor pool

Pipe rupture is one of the main causes of loss-of-coolant accident (LOCA). A siphon-breaking system would provide a passive mean of preventing LOCA, increasing the safety of research reactors. But despite the need for such a system, previous research on siphon breaking has not been conducted in a systematic manner. In this study, specific lines and holes were selected to act as siphon breakers, and the effect of size and other variables were investigated using an experimental facility similar on the scale of a real reactor. The performance of various siphon breakers was evaluated experimentally for different siphon-breaker sizes, pipe-rupture points, and pipe-rupture sizes. The effect of an orifice was also considered. Visualization of siphon breaking and examination of transient pressure data were used to analyze siphon-breaking phenomena. Filling a horizontal main pipe at the highest point by entrained air had a large effect on triggering siphon breaking; however, the stacked air entrained during the siphon-breaking event alone was insufficient to cause the phenomenon. All of the experimental parameters were investigated by comparing the undershooting height and transient pressure data trends. Experimental investigation and observation could give the possible postulate that all experimental parameters could be described as physical parameters, such as air flow rate, water flow rate and air quantity.

The LOBI Integral System Test Facility Experimental Programme

The LOBI project has been carried out in the framework of the European Commission Reactor Safety Research Programme in close collaboration with institutional and/or industrial research organizations of EC member countries. The primary objective of the research programme was the generation of an experimental data base for the assessment of the predictive capabilities of thermal-hydraulic system codes used in pressurised water reactor safety analysis. Within this context, experiments have been conducted in the LOBI integral system test facility designed, constructed, and operated (1979–1991) at the Ispra Site of the Joint Research Centre. This paper provides a historical perspective and summarizes major achievements of the research programme which has represented an effective approach to international collaboration in the field of reactor safety research and development. Emphasis is also placed on knowledge management aspects of the acquired experimental data base and on related online open access/retrieval user functionalities.

Experimental results on the coolability of a debris bed with multidimensional cooling effects

Within the reactor safety research, the removal of decay heat from a debris bed (formed from corium and residual water) is of great importance. In order to investigate experimentally the long term coolability of debris beds, the scaled test facility “DEBRIS” (Fig. 1) has been built at IKE. A large number of experiments had been carried out to investigate the coolability limits for different bed configurations ( Rashid et al., 2008; Groll et al., 2008; Schmidt, 2004). Analyses based on one-dimensional configurations underestimate the coolability in realistic multidimensional configurations, where lateral water access and water inflow via bottom regions are favoured. Following the experiments with top- and bottom-flooding flow conditions this paper presents experimental results of boiling and dryout tests at different system pressures based on top- and bottom-flooding via a down comer configuration. A down comer with an internal diameter of 10 mm has been installed at the centre of the debris bed. The debris bed is built up in a cylindrical crucible with an inner diameter of 125 mm. The bed of height 640 mm is composed of polydispersed particles with particle diameters 2, 3 and 6 mm. Since the long term coolability of such particle bed is limited by the availability of coolant inside the bed and not by heat transfer limitations from the particles to the coolant, the bottom inflow of water improves the coolability of the debris bed and an increase of the dryout heat flux can be observed. With increasing system pressure, the coolability limits are enhanced (increased dryout heat flux).

CFD analyses of fibre transport and fibre deposition at plunging jet conditions

Abstract The investigation of insulation debris generation, transport and sedimentation becomes important with regard to reactor safety research for PWR and BWR, when considering the long-term behaviour of emergency core cooling systems during all types of loss of coolant accidents. A joint research project on such questions is being performed in cooperation between the University of Applied Sciences Zittau/Görlitz (HSZG) and the Helmholtz-Zentrum Dresden-Rossendorf (HZDR). The project deals with the experimental investigation of particle transport phenomena in coolant flow and the development of CFD models for its description (see [10–12]). While the experiments are performed at the University at Zittau/Görlitz, the theoretical modelling efforts are concentrated in Rossendorf. In the current paper, the basic concepts for CFD modelling are described and feasibility studies are presented. The model capabilities are demonstrated via complex flow situations, where a plunging jet agitates insulation debris.

State of the art of controlling loss-of-coolant accidents with release of insulation material

The German Nuclear Society (KTG) is a non-profit scientific technical association with more than 2500 members. KTG strives for the development and dissemination of the stateof-the-art in nuclear science and technology to promote the peaceful use of nuclear energy and related disciplines. In 2009 the KTG has celebrated its 40 anniversary. The technical work of the KTG members is performed in the framework of currently 9 topical sections, which for instance jointly organize the KTG Annual Meetings, topical meetings devoted to selected subjects and several other activities. In recent years, the topical sections Reactor Safety, Thermal and Fluid Dynamics, as well as Reactor Physics and Computational Methods, have successfully established a series of meetings on “Topical Issues of Reactor Safety Research in Germany”. These meetings represent a forum for authorities, operators, technical support organizations, and vendors. A Technical Programm Committee (TPC) consisting from members of the participating topical sections, is responsible for selecting the issues to be addressed at the meeting. The acceptance and success of this series of meetings is mainly due to the current subjects dealt with. Also this forum offers the chance for a free exchange of opinions outside licensing and supervisory procedures. A balance is presented between survey papers on the background and the state of the art in the related field, specific technical presentations, and outlook papers. The participants are provided with a general overview on the area addressed and with new ideas and impetus for their future work. In order to further promote these kind of meetings the TPC has initiated co-operation with the Independent Journal for Nuclear Engineering, Energy Systems, Radiation and Radiological Protection KERNTECHNIK. For enhanced dissemination of the state-of-the-art, the key papers of the meetings are published in special issues of the journal. The 2010 meeting on “Topical Issues of Reactor Safety Research in Germany” was held at the Forschungszentrum Dresden-Rossendorf (FZD) on October 7–8 2010. It was hosted by the Institute of Safety Research (IfS) of FZD. The second day of the 2010 meeting was devoted to the ensurance of emergency core cooling (ECC) at PWRs and BWRs during loss-of-coolant accidents (LOCA) involving the release of thermal insulation material, which may lead to the clogging of the sump or the wet well strainers. The presentations were assigned to three sessions: “Overview and International Situation”, “Experiments and Model Development” and “Ensuring the Emergency Core Cooling for PWRs and BWRs”. A selection of these presentations is published in this special issue of KERNTECHNIK. In 1977, the first precursor of such a strainer clogging event occurred in BWR Gundremmingen (KRB-A), Germany. During an overfeeding transient a safety valve was damaged leading to a LOCA and almost complete release of the thermal insulation material around the break area. Some portions of the released material were collected in front of the strainers. Since KRB-A had an emergency condenser the ECC system was not started. Therefore, no coolant loaded with mineral fibers was sucked in and the fibrous material did neither clog the strainers nor enter the core. Nevertheless, the analysis of the event identified the potential of fiber entry and deposit into the core and its influence on core cooling. Because of the decommissioning of KRB-A the safety case of this event was inadequately assessed and the conclusions drawn were insufficiently distributed nationally, as well as internationally. Further events with release of insulation material and/or the clogging of strainers by mineral wool or other types of material later on occurred, for instance the Swedish NPP Bareback-2 (1992), and the US NPPs Susquehanna-2 (1988), Perry (1993), and Limerick-1 (1995). In German PWRs and BWRs several strainer modifications and backfittings were implemented over the last 10 years. These measures take into account the recommendations of the German Reactor-Safety Commission (RSK) on the robust demonstration of the effectiveness of the ECC during loss-of-coolant accidents involving the release of insulation material or other substances and on the removal of deposits from the PWR strainers by back flushing as required for PWRs in the RSK statements of July 2004 and March 2008. The RSK requirements are based on the following fundamental acceptance criteria:

Generic experiments at the sump model “Zittauer Strömungswanne” (ZSW) for the behaviour of mineral wool in the sump and the reactor core

Abstract The investigation of insulation debris transport, sedimentation, penetration into the reactor core and head loss build up becomes important to reactor safety research for PWR and BWR, when considering the long-term behaviour of emergency core cooling systems during loss of coolant accidents. Research projects are being performed in cooperation between the University of Applied Sciences Zittau/Görlitz and the Helmholtz-Zentrum Dresden-Rossendorf. The projects include experimental investigations of different processes and phenomena of insulation debris in coolant flow and the development of CFD models. Generic complex experiments serve for building up a data base for the validation of models for single effects and their coupling in CFD codes. This paper includes the description of the experimental facility for complex generic experiments (ZSW), an overview about experimental boundary conditions and results for upstream and down-stream phenomena as well as for the long-time behaviour due to corrosive processes.

Behavior of Coated Fuel Particle of High-Temperature Gas-Cooled Reactor under Reactivity-Initiated Accident Conditions

In order to clarify the failure mechanism and determine the failure limit of the High-Temperature Gascooled Reactor (HTGR) fuel under reactivity-initiated accident (RIA) conditions, pulse irradiations were performed with unirradiated coated fuel particles at the Nuclear Safety Research Reactor (NSRR). The energy deposition ranged from 0.578 to 1.869 kJ/gUO2, in the pulse irradiations and the estimated peak temperature at the center of the fuel particle ranged from 1,510 to 3,950 K. Detailed examinations after the pulse irradiations showed that the coated fuel particles failed above 1.40 kJ/gUO2, where the peak fuel temperature reached over the melting point of UO2 fuel. It was concluded that the coated fuel particle was failed by the mechanical interaction between the melted and swelled fuel kernel and the coating layer under RIA conditions.