The international program Phebus FP (fission
|
|
- Bruce Charles
- 6 years ago
- Views:
Transcription
1 1The safety of nuclear reactors 1 6 Results of initial Phebus FP tests FPT-0 and FPT-1 S. BOURDON (IRSN) D. JACQUEMAIN (IRSN) R. ZEYEN (JRC/PETTEN) The international program Phebus FP (fission products) involves performance of in-pile experiments for the purpose of studying key physical phenomena associated with a severe PWR accident. This specifically encompasses fuel rod degradation, release to and transport of radioactive materials in the primary system and the containment, as well as their physico-chemical behavior. Three tests simulating typical low-pressure loss-of-coolant accident (LOCA) phenomena have been successfully completed, two of them in a steam-rich atmosphere (FPT-0 and FPT-1) and a third with a significant reducing phase (FPT-2). FPT-0 was performed in December 1993 with trace-irradiated fuel, FPT-1 in July 1996 and FPT-2 in October 2000, using irradiated fuel. Objectives set for the first two tests were achievement of advanced bundle degradation about 20% of the fuel by mass and release of 70 to 80% volatile FPs. Their results provide an experimental basis of vital importance to understanding fuel degradation and FP behavior. Analysis of FPT-0 and FPT-1 results has now ended, with that of FPT-2 data still underway. Findings from the first two tests are described on the following pages. Phebus program. Design of test bundles and experimental circuits The experimental fuel bundles used in FPT-0 and FPT-1 tests were of similar design. Each bundle contained twenty PWR fuel rods and a control rod made up of a neutron absorber alloy (silverindium-cadmium) with stainless steel cladding and a zircaloy guide tube (figure 1, page 46). While the FPT-1 bundle incorporated 18 PWR fuel rods pre-irradiated to an average burnup of 23.4 GWD/tU, the FPT-0 bundle comprised fresh fuel. Fuel rods measured 1 m in length and contained a fuel mass of 11 kg with a zircaloy cladding. The bundles were surrounded by an insulating shroud and placed in a tube cooled by a pressurized water circuit. The test fuel occupied a cell at the center of the Phebus reactor core. Various elements installed downstream of the bundle simulated the components of a standard PWR: RCS hot leg segment (700 C), steam generator U- tube, RCS cold leg segment (150 C) and containment (figure 2, page 46). SCIENTIFIC AND TECHNICAL REPORT
2 Figure 1 Structure of experimental fuel bundles. Figure 2 Diagram of the Phebus experimental facility. Experimental sequence The bundle degradation phase lasted five to six hours. During this phase, bundle power and steam flow were increased by increments (figure 3) to gradually raise fuel rod temperature to the point where the fuel cladding failed. This was followed by control rod degradation and relocation of the absorber materials, oxidation of the zircaloy cladding, hydrogen production, fuel liquefaction and buildup of liquefied mixtures, molten pool formation, and release of fission products, structural and fuel rod materials. In the degradation process, steam was injected into the bundle at a pressure of approximately 2 bar and a flow of 0.5 to 2.2 g/s, thus generating an oxidizing or steam-rich environment. Following degradation, in-containment test phases lasted for five days, to allow investigation of aerosol deposition and the radiochemical behavior of iodine in sump water and the atmosphere. Bundle degradation results Figure 3 Principal events identified during the FPT-1 test. At the start of degradation phases in both FPT-0 and FPT-1experiments, zircaloy clad rupture due to increase in pressure inside the rods took place at about 800 C (maximum temperature measured at bundle mid-height). Such rupture was clearly confirmed by detection of FP in the containment and aerosols in the circuit. Fuel rod rupture was observed in both tests immediately prior to runaway of the zircaloy oxidation reaction by detection of In 116m in the circuit for a maximum bundle temperature between 1160 and 1330 C. These results indicate that control rod rupture took place at a temperature below the stainless steel melting point. Premature degradation of the control rod was probably caused by chemical interaction of its stainless steel cladding with the zircaloy guide tube. The liquefied mixture (of silver-indium-cadmium, stainless steel and zircaloy) then probably relocated to the lower grid during the oxidation period and solidified in the lower part of the bundle, forming a metallic mass rich in Ag and Zr. Hydrogen from zircaloy oxidation was detected in the FP circuit when fuel temperatures exceeded 1100 C in the upper bundle. An increase in fuel rod heatup rate was measured in 46 INSTITUT DE RADIOPROTECTION ET DE SÛRETÉ NUCLÉAIRE
3 1 both tests at a maximum bundle temperature of C. Temperature peaks of 2500 C and 2200 C were observed in FPT-0 and FPT-1, with heatup rates of 10 to 15 C/s. Most hydrogen production took place during the oxidation runaway. Total steam starvation did not occur at bundle outlet in either of the tests. In test FPT- 1, a second lower hydrogen production spike was observed in the last twenty minutes of the transient. This delayed oxidation probably resulted from downward displacement of the molten materials to a colder zone where cladding had not yet been fully oxidized. The total mass of released hydrogen represented some 115 g in FPT-0 and 96 g in FPT-1; this corresponded to 77% and 64% oxidation of the bundle s zircaloy inventory. Analysis of these results shows that, in both tests, there was significant displacement of materials at the start of heatup, with maximum fuel temperature at 2000 to 2300 C. Low temperature relocation of fuel (i.e. at about 500 C below the melting point of pure UO 2 ) could be explained by formation, in the upper half of the bundle, of eutectic mixtures made up of absorber rod materials (with iron, nickel and chromium oxides) and partially oxidized cladding. During heatup, the resulting mixtures progressed downward, through the upper grid. At bundle center, in hotter zones, these mixtures may have interacted with absorber materials and fuel rods. This process probably led to fuel rod degradation in the upper half of the bundle with partial accumulation of the interacting ( Figure 4 Radiography and tomography of test bundles. materials near the lower grid. Analysis of measured bundle and shroud temperatures showed that fuel rod degradation and relocation took place gradually throughout the heatup period, which led to formation of a homogeneous molten pool at the lower grid. Tests ended with reactor trip when rapid, significant temperature increases were detected in the lower portion of the shroud. The safety of nuclear reactors In both tests, there was significant displacement of materials at the start of heatup, with maximum fuel temperature at 2000 to 2300 C. SCIENTIFIC AND TECHNICAL REPORT
4 Figure 5a Post-test examinations and final bundle status Gamma scanning, radiography, tomography and destructive testing subsequently confirmed the advanced stage of FPT-0 and FPT-1 fuel bundle degradation achieved during the experiments (figure 4, page 47). Analysis of bundle post-irradiation examinations (PIEs) showed accumulation in the lower bundle of a (U-Zr)O 2 corium containing 1 to 2% iron and chromium oxides (by weight). PIE also identified a solidified, Ag-rich metallic mass just below the UO 2 -rich molten pool. Interpretation of tomograms enabled evaluation of total UO 2 mass in this pool (about 2.6 kg in FPT-0 and 2 kg in FPT-1). Moreover, corium melting temperature in the lower bundle, as measured during PIE, was around 2500 C. Transport of iodine-131 (volatile FP) in the cold leg (150 C) during the FPT-1 transient. Figure 5b Transport of ruthenium-103 (low volatility FP) in the hot leg (700 C) and cold leg (150 C) during the FPT-1 transient. Based on PIE results for the FPT-1 bundle, it was clear that fresh fuel rods had sustained significantly less damage than the pre-irradiated rods. FP and material behavior in the test circuits RELEASE AND TRANSPORT OF FISSION PRODUCTS, FUEL, BUNDLE STRUCTURES AND CONTROL ROD MATERIALS Released bundle fractions were comparable for FPT-0 and FPT-1, with slightly lower volatile FP values in FPT-1. These lower values can be attributed to a lesser degree of bundle degradation in the second test. In FPT-1, fuel material (U) release was similar to that measured in FPT-0. The difference in type of fuel used in the two tests (trace-irradiated for FPT-0, irradiated for FPT-1) did not seem to sharply impact U release. FP elements can be classified as follows, based on the fractions released: high release (approx. 90% of inventory): rare gases Kr and Xe and volatile FPs I, Cs and Te; significant release (20-60% of inventory): Mo, Tc and Sb, and Sn originating from bundle structures; low release (5-15%): Ag and In from the control rod and Re from the thermocouples; very low release (less than 1% of inventory): low volatility FPs Ba(La), Ru, Sr, Zr-Nb, Nd and Sm (in FPT-1), fuel material U and Zr from bundle structures. In short, results of tests FPT-0 and FPT-1 confirmed the release fractions obtained in analytical experiments, except for Ba (lower release in the Phebus series). In both tests, the release and transport of materials in the primary circuit correlated closely with bundle degradation events (figures 5a and 5b). Volatile FP release generally reached a maximum during the early and late zircaloy oxidation phases. Their rate of release decreased in the corium propagation phase. Release of fuel and structural materials, as well as silver (from the silver-indium-cadmium control rod) peaked during the late oxidation phase. Except for iodine and cadmium, most elements were already in condensed form in the circuit hot leg (700 C). Cs behavior was more complex than expected, since a significant fraction of this element was probably transported as vapor and another fraction in condensed form. Materials transported as vapor in the hot leg condensed in 48 INSTITUT DE RADIOPROTECTION ET DE SÛRETÉ NUCLÉAIRE
5 1 For FPT-1, the aerosol mass transiting through the circuit was 150 g in the hot leg and 130 g in the cold leg (for an initial fuel inventory of 11 kg). ( The safety of nuclear reactors the steam generator, so that all of them, other than a small fraction of iodine, were in condensed form in the circuit cold leg. Condensed materials were transported as multicomponent aerosols in both hot and cold legs. The aerosol mass predominantly comprised control rod, structural and fuel materials. For FPT-1, the aerosol mass transiting through the circuit was 150 g in the hot leg and 130 g in the cold leg (for an initial fuel inventory of 11 kg). Aerosol mass concentrations were highest in the late oxidation phase. Aerosol composition (by mass) was determined by degradation events, with large Ag, In, Cd and Sn contributions during the first oxidation phase, large Ag and Re input and significant Cs and Mo contributions at start of meltdown, and predominantly Ag, Re and U content in the late oxidation phase. CIRCUIT RETENTION OF FISSION PRODUCTS, AND BUNDLE/STRUCTURAL/CONTROL ROD MATERIALS The quantity of material deposited at bundle outlet seems to correlate with the degree of volatility of the various elements involved. The vapors of low volatility elements were probably the first to condense. This was confirmed for Ru and Zr; in both cases, some 45% of the freed fraction was deposited in the upper plenum (at degraded bundle outlet), with only low deposits in the vertical line. Similar behavior was observed for Ag. There was little deposition of volatile elements in the upper plenum and more in the vertical line. In the steam generator tube, iodine and cadmium deposits, due essentially to steam condensation, were more significant than aerosol deposits, and represented some 25% of the mass injected into the tube. Aerosol deposits resulting mostly from thermophoresis represented only 14% of the mass entering the tube. Most deposition (85%) took place in the SG hot leg. AEROSOL BEHAVIOR IN THE CONTAINMENT Settling was the main mechanism of aerosol deposition in the containment: some 65 to 70% of total inventory was deposited by gravity on the containment floor. Diffusiophoresis (aerosol entrainment in condensed steam) was the second most important deposition mechanism. As a result, 25% to 28% of containment inventory was entrained by condensation towards the painted surfaces of the condenser. Containment wall deposits were of minor significance: only 2% of the inventory was detected on walls at the end of the aerosol deposition phase. During containment floor washing, most of the containment inventory was entrained into the sump, where it appeared either as soluble matter in sump water or as surface deposits (having primarily settled by gravity to the sump floor). Closer analysis showed that Cs was present in soluble form, while Mo, Tc, Ba, Cd and Re were only partially dissolved. Volatile FPs I, Te and Sb, as well as low volatility FPs Ru, La, Sm and Nd, control rod elements Ag and In, actinides and elements from the fuel and its structures remained mostly insoluble and were present as particulate matter in the sump. After analysis, a complete overview of data for all of these elements was possible, as shown for iodine in figure 6 (page 50). SPECIFIC BEHAVIOR OF IODINE In both tests, nearly all of the fuel bundle iodine (about 87% of total inventory) was released in gaseous form. This gaseous iodine probably reacted partially with Ag, In, Cs or Rb in the vertical line above the bundle, to form a metal iodide vapor. In the circuit hot leg, most of the iodine flow was in gaseous or metal vapor form. Iodine was then transported to the SG tube, where it formed more deposits than any other element (23.5% of bundle inventory in FPT-0 and 19.2% in FPT-1). Deposition was due either to metal iodide condensation in the fluid and on SCIENTIFIC AND TECHNICAL REPORT
6 Figure 6 Iodine-131 balance in the test circuits. tube walls, to chemisorption of gaseous iodine onto Cd deposits, or to iodine condensation on aerosols, which are subsequently deposited by thermophoresis. In the circuit cold leg, most of the iodine was apparently transported by aerosols. More than 60% of bundle iodine inventory was injected into the containment (63% in FPT-0 and 64% in FPT-1). Most of it had been transported by aerosols that either settled on the containment bottom or deposited on the containment surfaces (essentially by diffusiophoresis on the painted condenser surfaces). In both tests, gaseous iodine fractions measured in the containment reached their highest levels during zircaloy oxidation phases (for FPT-1, about 0.2% of bundle inventory, during the first phase of oxidation, which represented 4% of containment inventory at that point in time; and, for FPT-0, some 3% of bundle inventory, or 33% of containment inventory at the same point in time). In FPT-1, following the two oxidation phases, gaseous iodine fraction decreased by a factor of 2 in less than an hour. This is linked to diffusiophoresis of gaseous iodine at the painted condensing surfaces and/or to a fast chemical transformation of the iodine species considered. The volatile fraction of the iodine originating from the primary circuit also disappeared in a few hours in FPT-0, probably due to deposition on the containment surfaces (painted condensers). Evolution of the gaseous iodine fraction during the aerosol phase was different in the two tests (figure 7). In FPT-1, the gaseous iodine fraction increased significantly from 0.07% to an average fraction of 0.14% of bundle inventory immediately after containment isolation. It then remained more or less constant at this value for about 5 hours. The increase (from 0.07% to 0.14%) can be attributed to a measured desorption of gaseous iodine from the painted condensing surfaces. Measured data was consistent with 50 INSTITUT DE RADIOPROTECTION ET DE SÛRETÉ NUCLÉAIRE
7 1 release of organic iodides. In FPT-0, the gaseous iodine fraction decreased exponentially during the aerosol phase, dropping from 2.6% to 0.32% of bundle inventory. Unlike what was observed in FPT-1, there was no significant decrease during FPT-0 in iodine activity on the painted condenser surfaces. In both tests, before the washing phase, the gaseous iodine fraction was divided between I 2 and the organic iodides with, in the case of FPT-1, an increase in organic iodide fractions during the aerosol phase that followed containment isolation. During the washing phase, iodine in the sump behaved mainly like particulate matter and settled by gravity to the bottom of the sump, meaning that this iodine was transported to the sump and formed an insoluble species (e.g. particulate Agl). After washing, the gaseous iodine fraction was low in both cases, representing only 0.063% of bundle inventory in FPT-0 and 0.094% of this inventory in FPT-1 (figure 7). Speciation of gaseous iodine differed from one test to the other. In FPT-1, measured data was homogeneous and showed that I 2 was the predominant gaseous iodine species at the time of washing. After this phase, the I 2 contribution to the gaseous fraction diminished, while that of organic iodides increased. In FPT-0, data was not precise enough to detect any impact of washing on the gaseous iodine fraction; however, organic iodides were the major gaseous iodine species at that time. In both tests, measured data showed that in the long term, organic iodides predominated over other gaseous iodine species. Results indicated a small transfer of gaseous iodine from the sump to the atmosphere due to iodine trapping by Ag and the ensuing inhibition of volatile iodine formation by radiolysis. ( Figure 7 General behavior of the in-containment gaseous iodine fraction during Phebus tests FPT-0 and FPT-1. Conclusion Several of the key phenomena potentially occurring in a severe PWR accident were reproduced by Phebus FPT-0 and FPT-1 experiments. These included clad burst, zircaloy oxidation and hydrogen production, control rod failure and relocation of absorber materials, relocation of fuel rods, UO 2 buildup and molten pool formation and FP and aerosol release, and transport and deposition in the primary circuit and the containment. Findings derived from experimental results can now be used to address important questions about the source term in a severe accident situation. The control rod failure process and early accumulation of metallic materials in the lower fuel bundle (below the UO 2 pool) were clearly identified. Measurement of total hydrogen mass The safety of nuclear reactors Measured data showed that, over the long term, organic iodides were the predominating gaseous iodine species. SCIENTIFIC AND TECHNICAL REPORT
8 1 Leaktight housing for remote handling in the Phebus facility. produced in both tests provided interesting insight into zircaloy oxidation kinetics. The relatively low temperature at which fuel rod relocation took place is a point worth considering in severe PWR accident assessments. The FPT experiments showed that relocation of materials and their buildup leads to formation of a UO 2 - rich molten pool and its gradual propagation through the lower part of the bundle. Post-irradiation examinations plainly demonstrated that, under identical conditions, irradiated fuel sustained more damage than fresh fuel. For fission products, the points of ongoing interest to severe PWR accident assessment include: the main phases of material transport through the circuit, which are correlated with bundle degradation events (zircaloy oxidation, dislocation of materials, etc.); deposition in the circuit, which in FPT experiments, affected mainly the hot leg (in piping above the degraded fuel bundle and at the SG inlet); injection of a significant quantity of gaseous iodine during zircaloy oxidation phases; aerosols that transported FPs in the circuit, which were the multicomponent type, mostly made up of fuel bundle structural and control rod materials; long-term predominance of organic iodides in the gaseous iodine fraction, in the containment model; containment sump chemistry, marked by trapping of iodine by silver (released on degradation of the silver-indium-cadmium control rod), which precludes formation of volatile iodine by radiolysis. Experiment FPT-2 was geared to the same types of phenomena as FPT-1, for steam-poor conditions in the primary circuit and different in-containment conditions impacting iodine volatility. The last test in the program, FPT-3, will study the effect of control rod absorber material (B 4 C) on fuel bundle degradation and the impact of its oxidation products on FP behavior. ( Relocation of materials and their buildup leads to formation of a UO 2 -rich molten pool and its gradual propagation through the lower part of the bundle. 52 INSTITUT DE RADIOPROTECTION ET DE SÛRETÉ NUCLÉAIRE
Steam Flow-rate Effect on the Transient Behaviour in Phebus Experiment FPT-1
Steam Flow-rate Effect on the Transient Behaviour in Phebus Experiment FPT-1 Salwa Helmy 1, Basma. Foad 1 N. Eng. Safety, Dept. of NRRA, Cairo, Egypt 1 Nuclear and Radiological Regulatory Authority (NRRA)
More informationSUMMARY OF THE RESULTS FROM THE PHEBUS FPT-1 TEST FOR A SEVERE ACCIDENT AND THE LESSONS LEARNED WITH MELCOR
SUMMARY OF THE RESULTS FROM THE PHEBUS FPT-1 TEST FOR A SEVERE ACCIDENT AND THE LESSONS LEARNED WITH MELCOR JONG-HWA PARK *, DONG-HA KIM and HEE-DONG KIM Korea Atomic Energy Research Institute, 150 Deokjin-dong,
More informationSIMULATION OF CONTAINMENT PHENOMENA DURING THE PHEBUS FPT1 TEST WITH THE CONTAIN CODE
International Conference Nuclear Energy for New Europe 2002 Kranjska Gora, Slovenia, September 9-12, 2002 www.drustvo-js.si/gora2002 SIMULATION OF CONTAINMENT PHENOMENA DURING THE PHEBUS FPT1 TEST WITH
More informationEvaluation of FPTRAN module of RELAP/SCDAPSIM Code Using PHEBUS FPT-01 Experiment
2005 International Nuclear Atlantic Conference - INAC 2005 Santos, SP, Brazil, August 28 to September 2, 2005 ASSOCIAÇÃO BRASILEIRA DE ENERGIA NUCLEAR - ABEN ISBN: 85-99141-01-5 Evaluation of FPTRAN module
More informationSource Term Prediction History and Current Practices
Photos placed in horizontal position with even amount of white space between photos and header Used by permission from TEPCO Used by permission from TEPCO All materials from UUR Open Source Reports SAND27-7697
More informationPHEBUS REACTOR: THE DRIVING OF A SEVERE ACCIDENT
PHEBUS REACTOR: THE DRIVING OF A SEVERE ACCIDENT M.-C. ANSELMET, J. BONNIN, F. SERRE, G. AUGIER, S. BAYLE, J.-C. CABRILLAT, G. REPETTO Institut de Protection et de Sûreté Nucléaire, Département de Recherche
More informationMartin Steinbrück Karlsruhe Institute for Technology Postfach 1, D Karlsruhe, Germany
Study of Boron Behaviour in the Primary Circuit of Water Reactors under Severe Accident Conditions: a Comparison of Recent Integral and Separate-Effects Data Tim Haste, Frédéric Payot, Cristina Dominguez,
More informationSource terms designate typical environmental releases of radioactive substances,
1 2 Source term evaluation studies for PWRs J. FLEUROT (IRSN) J.-M. EVRARD (IRSN) B. CHAUMONT (IRSN) Source terms designate typical environmental releases of radioactive substances, which are also the
More informationDENSITY STRATIFICATION AND FISSION PRODUCT PARTITIONING IN MOLTEN CORIUM PHASES. D.A. Powers Sandia National Laboratories Albuquerque, NM USA
DENSITY STRATIFICATION AND FISSION PRODUCT PARTITIONING IN MOLTEN CORIUM PHASES D.A. Powers Sandia National Laboratories Albuquerque, NM USA A. Behbahani U.S. Nuclear Regulatory Commission Washington,
More informationDRAFT: SEVERE FUEL DAMAGE EXPERIMENTS WITH ADVANCED CLADDING MATERIALS TO BE PERFORMED IN THE QUENCH FACILITY (QUENCH-ACM)
Proceedings of the 16th International Conference on Nuclear Engineering ICONE16 May 11-15, 2008, Orlando, Florida, USA ICONE16-48074 DRAFT: SEVERE FUEL DAMAGE EXPERIMENTS WITH ADVANCED CLADDING MATERIALS
More information5.5. Release of fission products during a core melt accident
Development of the core melt accident 255 5.5. Release of fission products during a core melt accident This section deals with releases of fission products (FPs) from degraded fuel or corium during an
More informationPost-Test Analysis of the QUENCH-13 Experiment
Post-Test Analysis of the QUENCH-13 Experiment Jon Birchley 1, Henrique Austregesilo 2, Christine Bals 2, Roland Dubourg 3, Tim Haste 1, Jean-Sylvestre Lamy 4, Terttaliisa Lind 1, Bernard Maliverney 4,
More informationRecent progress in source term research and evaluations with the ASTEC code
Enhancing nuclear safety Recent progress in source term research and evaluations with the ASTEC code Jacquemain D., Vola D., Cantrel L., Chevalier-Jabet K., Mun C. IRSN Nuclear Safety Division 8 th International
More informationNRC Source Term Research Outstanding Issues and Future Directions
S4-1 invited NRC Source Term Research Outstanding Issues and Future Directions Farouk Eltawila Director Office of Nuclear Regulatory Research Dana A. Powers Advisory Committee on Reactor Safeguards U.S.
More informationPWR and BWR plant analyses by Severe Accident Analysis Code SAMPSON for IMPACT Project
GENES4/ANP2003, Sep. 15-19, 2003, Kyoto, JAPAN Paper 1074 PWR and BWR plant analyses by Severe Accident Analysis Code SAMPSON for IMPACT Project Hiroshi Ujita 1*, Yoshinori Nakadai 2, Takashi Ikeda 3,
More information2017 Water Reactor Fuel Performance Meeting September 10 (Sun) ~ 14 (Thu), 2017 Ramada Plaza Jeju Jeju Island, Korea
HALDEN S IN-PILE TEST TECHNOLOGY FOR DEMONSTRATING THE ENHANCED SAFETY OF WATER REACTOR FUELS Margaret A. McGrath 1 1 OECD Halden Reactor Project, IFE: Os Alle 5/P.O. Box 173, 1751 Halden, Norway, Margaret.mcgrath@ife.no
More informationASTEC Model Development for the Severe Accident Progression in a Generic AP1000-Like
ASTEC Model Development for the Severe Accident Progression in a Generic AP1000-Like Lucas Albright a,b, Dr. Polina Wilhelm b, Dr. Tatjana Jevremovic a,c a Nuclear Engineering Program b Helmholtz-ZentrumDresden-Rossendorf
More informationNURETH Progress on Severe Accident Code Benchmarking in the Current OECD TMI-2 Exercise
NURETH-15 544 Progress on Severe Accident Code Benchmarking in the Current OECD TMI-2 Exercise G. Bandini (ENEA), S. Weber, H. Austregesilo (GRS), P. Drai (IRSN), M. Buck (IKE), M. Barnak, P. Matejovic
More informationVVER-440/213 - The reactor core
VVER-440/213 - The reactor core The fuel of the reactor is uranium dioxide (UO2), which is compacted to cylindrical pellets of about 9 height and 7.6 mm diameter. In the centreline of the pellets there
More informationSource Term modeling for CANDU reactors
Source Term modeling for CANDU reactors IAEA Technical Meeting on Source term Evaluation for Severe Accidents October 21-23, 2013 Objectives of presentation To provide overview of the current state in
More informationJ. Stuckert, M. Große, M. Steinbrück
Bundle reflood tests QUENCH-14 and QUENCH-15 with advanced cladding materials: comparable overview J. Stuckert, M. Große, M. Steinbrück Institute for Materials Research KIT University of of the State of
More informationcompounds thermal expansion using high temperature X-ray diffraction, as well as calorimetric measurements (Differential Scanning Calorimetry,
Our group is engaged in research activities in the field of nuclear reactor technology, including the materials chemistry and chemical thermodynamics of nuclear materials. Our research areas encompass
More informationCorium Retention Strategy on VVER under Severe Accident Conditions
NATIONAL RESEARCH CENTRE «KURCHATOV INSTITUTE» Corium Retention Strategy on VVER under Severe Accident Conditions Yu. Zvonarev, I. Melnikov National Research Center «Kurchatov Institute», Russia, Moscow
More informationSMR/1848-T21b. Course on Natural Circulation Phenomena and Modelling in Water-Cooled Nuclear Reactors June 2007
SMR/1848-T21b Course on Natural Circulation Phenomena and Modelling in Water-Cooled Nuclear Reactors 25-29 June 2007 T21b - Selected Examples of Natural Circulation for Small Break LOCA and Som Severe
More informationSpecification for Phase VII Benchmark
Specification for Phase VII Benchmark UO 2 Fuel: Study of spent fuel compositions for long-term disposal John C. Wagner and Georgeta Radulescu (ORNL, USA) November, 2008 1. Introduction The concept of
More informationIn Vessel Retention Strategy VVER 1000/320 VVER 2013 Conference
ÚJV Řež, a. s. In Vessel Retention Strategy VVER 1000/320 VVER 2013 Conference J. Zdarek Presentation content Background of SA issues VVER 1000/320 Containment and RPV Cavity Configuration IVR Strategy
More informationAccident Progression & Source Term Analysis
IAEA Training in Level 2 PSA MODULE 4: Accident Progression & Source Term Analysis Outline of Discussion Overview of severe accident progression and source term analysis Type of calculations typically
More informationDESIGN AND SAFETY- SUPPORT ANALYSES OF AN IN-PILE MOLTEN SALT LOOP IN THE HFR
DESIGN AND SAFETY- SUPPORT ANALYSES OF AN IN-PILE MOLTEN SALT LOOP IN THE HFR stempniewicz@nrg.eu M.M. Stempniewicz, E.A.R. de Geus, F. Alcaro, P.R. Hania, K. Nagy, N.L. Asquith, J. de Jong, L. Pool, S.
More informationSEVERE ACCIDENT PHENOMENA part 1: In-vessel
SEVERE ACCIDENT PHENOMENA part 1: In-vessel Workshop on Severe Accident Management Guidelines 11-15 December 2017, Vienna, Austria presented by Randall Gauntt (Sandia National Laboratories) Outline Severe
More informationLFW-SG ACCIDENT SEQUENCE IN A PWR 900: CONSIDERATIONS CONCERNING RECENT MELCOR / CALCULATIONS
LFW-SG ACCIDENT SEQUENCE IN A PWR 900: CONSIDERATIONS CONCERNING RECENT MELCOR 1.8.5 / 1.8.6 CALCULATIONS F. DE ROSA ENEA FIS NUC - Bologna 1 st EUROPEAN MELCOR USERS GROUP Villigen, Switzerland 15-16
More informationFukushima-Daiichi - a radiochemical view of the evolving situation in Summer 2011.
Fukushima-Daiichi - a radiochemical view of the evolving situation in Summer 2011. Kath Morris. Research Centre for Radwaste and Decommissioning The University of Manchester With thanks to Dr Edward Blandford,
More informationUKEPR Issue 05
Title: PCER Sub-Chapter 6.1 Sources of radioactive materials Total number of pages: 16 Page No.: I / III Chapter Pilot: S. BOUHRIZI Name/Initials Date 06-08-2012 Approved for EDF by: T. MARECHAL Approved
More informationModeling and Analysis of In-Vessel Melt Retention and Ex-Vessel Corium Cooling in the U. S.
Modeling and Analysis of In-Vessel Melt Retention and Ex-Vessel Corium Cooling in the U. S. E. L. Fuller, S. Basu, and H. Esmaili Office of Nuclear Regulatory Research United States Nuclear Regulatory
More informationRadionuclide Release at Fukushima
Radionuclide Release at Fukushima Peter F. Caracappa, Ph.D., CHP American Nuclear Society Connecticut Section November 16, 2011 Overview Accident Review Radioactive Material Releases Transport and Deposition
More informationThe Nuclear Crisis in Japan
The Nuclear Crisis in Japan March 21, 2011 Daniel Okimoto Alan Hanson Kate Marvel The Fukushima Daiichi Incident 1. Plant Design 2. Accident Progression 3. Radiological releases 4. Spent fuel pools " Fukushima
More informationMaterial Selection According to ALARA during Design Stages of EPR. P. Jolivet, A. Tamba, F. Chahma AREVA
Material Selection According to ALARA during Design Stages of EPR P. Jolivet, A. Tamba, F. Chahma AREVA Tour AREVA 1 Place Jean Millier 92084 Paris La Defense France E-mail: patrick.jolivet@areva.com,
More informationUnderstanding the effects of reflooding in a reactor core beyond LOCA conditions
Understanding the effects of reflooding in a reactor core beyond LOCA conditions F. Fichot 1, O. Coindreau 1, G. Repetto 1, M. Steinbrück 2, W. Hering 2, M. Buck 3, M. Bürger 3 1 - IRSN, Cadarache (FR)
More informationSevere Accident Progression Without Operator Action
DAA Technical Assessment Review of the Moderator Subcooling Requirements Model Severe Accident Progression Without Operator Action Facility: Darlington Classification: October 2015 Executive summary After
More informationSimulation of thermal hydraulics accidental transients: evaluation of MAAP5.02 versus CATHAREv2.5
1/12 Simulation of thermal hydraulics accidental transients: evaluation of MAAP5.02 versus CATHAREv2.5 J. Bittan¹ 1) EDF R&D, Clamart (F) Summary MAAP is a deterministic code developed by EPRI that can
More informationSIMULATION OF LIVE-L4 WITH ATHLET-CD
SIMULATION OF LIVE-L4 WITH ATHLET-CD T. Hollands, C. Bals Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) ggmbh Forschungszentrum, Boltzmannstraße 14, 85748 Garching, Germany thorsten.hollands@grs.de;
More informationControlled management of a severe accident
July 2015 Considerations concerning the strategy of corium retention in the reactor vessel Foreword Third-generation nuclear reactors are characterised by consideration during design of core meltdown accidents.
More informationZRO 2 AND UO 2 DISSOLUTION BY MOLTEN ZIRCALLOY
International Conference Nuclear Energy for New Europe 2002 Kranjska Gora, Slovenia, September 9-12, 2002 www.drustvo-js.si/gora2002 ZRO 2 AND UO 2 DISSOLUTION BY MOLTEN ZIRCALLOY J. Stuckert, A. Miassoedov,
More informationEffects of Source Term on Off-site Consequence in LOCA Sequence in a Typical PWR
Effects of Source Term on Off-site Consequence in LOCA Sequence in a Typical PWR Seok-Jung HAN a, Tae-Woon KIM, and Kwang-Il AHN a Korea Atomic Energy Research Institute, P.O. Box 105, Yuseong, Daejeon,
More informationAccidents in nuclear facilities
2 Accidents in nuclear facilities 62 Scientific and Technical Report 2007 - IRSN 2 Accidents in nuclear facilities... 64 2.1 FIRST results of the Phebus FPT3 test... 66 2.2 Study of ruthenium chemistry
More informationMultiphase Flow Dynamics 4
Multiphase Flow Dynamics 4 Nuclear Thermal Hydraulics von Nikolay I Kolev 1. Auflage Multiphase Flow Dynamics 4 Kolev schnell und portofrei erhältlich bei beck-shop.de DIE FACHBUCHHANDLUNG Thematische
More informationS. Gupta - G. Poss - M. Sonnenkalb. OECD/NEA THAI Program for Containment Safety Research: main Insights and Perspectives
S. Gupta - G. Poss - M. Sonnenkalb OECD/NEA THAI Program for Containment Safety Research: main Insights and Perspectives. Introduction Overall objectives of OECD/NEA THAI projects: To provide containment
More informationCEA ACTIVITIES SUPPORTING THE OPERATING FLEET OF NPPS
CEA ACTIVITIES SUPPORTING THE OPERATING FLEET OF NPPS Colloque SFEN Atoms for the future Christophe Béhar 24 OCTOBRE 2012 Christophe Béhar - October 24th, 2012 PAGE 1 DEN ASSIGNMENTS Nuclear Energy Support
More informationNuclear Safety. Lecture 3. Beyond Design Basis Accidents Severe Accidents
Nuclear safety Lecture 3. Beyond Design Basis Accidents Severe Accidents Ildikó Boros Prof. Dr. Attila Aszódi Budapest University of Technology and Economics Institute of Nuclear Techniques (BME NTI) 1
More informationCANDU Safety #12: Large Loss of Coolant Accident F. J. Doria Atomic Energy of Canada Limited
CANDU Safety #12: Large Loss of Coolant Accident F. J. Doria Atomic Energy of Canada Limited 24-May-01 CANDU Safety - #12 - Large LOCA.ppt Rev. 0 1 Overview Event sequence for a large break loss-of of-coolant
More informationCONTRIBUTION OF RESEARCH REACTORS TO THE PROGRAMMES FOR RESEARCH AND TECHNOLOGICAL DEVELOPMENT ON SAFETY
CONTRIBUTION OF RESEARCH REACTORS TO THE PROGRAMMES FOR RESEARCH AND TECHNOLOGICAL DEVELOPMENT ON SAFETY J. Couturier, F. Pichereau, C. Getrey, J. Papin, B. Clément INSTITUT DE RADIOPROTECTION ET DE SURETE
More informationSpecification for Phase IID Benchmark. A. BARREAU (CEA, France) J. GULLIFORD (BNFL, UK) J.C. WAGNER (ORNL, USA)
Specification for Phase IID Benchmark PWR-UO 2 Assembly: Study of control rods effects on spent fuel composition A. BARREAU (CEA, France) J. GULLIFORD (BNFL, UK) J.C. WAGNER (ORNL, USA) 1. Introduction
More informationA New Method Taking into Account Physical Phenomena Related to Fuel Behaviour During LOCA
S. BOUTIN S. GRAFF A. BUIRON A New Method Taking into Account Physical Phenomena Related to Fuel Behaviour During LOCA Seminar 1a - Nuclear Installation Safety - Assessment AGENDA 1. Context 2. Development
More informationThe Fukushima Daiichi Incident Dr. Matthias Braun - 16 November p.1
Dr. Matthias Braun - 16 November 2012 - p.1 The Fukushima Daiichi Incident 1. Plant Design 2. Accident Progression 3. Radiological releases 4. Spent fuel pools 5. Sources of Information Matthias Braun
More information4.2 DEVELOPMENT OF FUEL TEST LOOP IN HANARO
4.2 DEVELOPMENT OF FUEL TEST LOOP IN HANARO Sungho Ahn a, Jongmin Lee a, Suki Park a, Daeyoung Chi a, Bongsik Sim a, Chungyoung Lee a, Youngki Kim a and Kyehong Lee b a Research Reactor Engineering Division,
More informationNew Reactors Programme. GDA Close-out for the AP1000 Reactor. GDA Issue GI-AP1000-RC-01 Revision 0 Accident Source Terms
New Reactors Programme GDA Close-out for the AP1000 Reactor GDA Issue GI-AP1000-RC-01 Revision 0 Accident Source Terms Assessment Report: ONR-NR-AR-16-044 Revision 0 March 2017 Page 1 of 38 , 2017 If you
More informationFundamental Research Program for Removal of Fuel Debris
International Symposium on the Decommissioning of TEPCO s Fukushima Daiichi Nuclear Power Plant Unit 1-4 1 Fundamental Research Program for Removal of Fuel Debris March 14, 2012 Tadahiro Washiya Japan
More informationPassive Autocatalytic Recombiner. Hydrogen Control and Mitigation for Combustible Gas Control
Passive Autocatalytic Recombiner Hydrogen Control and Mitigation for Combustible Gas Control Passive Autocatalytic Recombiner (PAR) Severe Accident-Qualified PAR for Combustible Gas Control Passive safety
More informationUniversity of Zagreb Faculty of Electrical Engineering and Computing NPP KRŠKO CONTAINMENT MODELLING WITH THE ASTEC CODE
Journal of Energy VOLUME 64 2015 journal homepage: http://journalofenergy.com/ SINIŠA ŠADEK sinisa.sadek@fer.hr DAVOR GRGIĆ davor.grgic@fer.hr University of Zagreb Faculty of Electrical Engineering and
More informationThe Fukushima Daiichi Incident Dr. Matthias Braun - 19 May p.1
Dr. Matthias Braun - 19 May 2011 - p.1 The Fukushima Daiichi Incident 1. Plant Design 2. Accident Progression 3. Radiological releases 4. Spent fuel pools 5. Sources of Information Matthias Braun PEPA4-G,
More informationChapter 4 Safety Principles for France s Pressurised Water Reactors
Chapter 4 Safety Principles for France s Pressurised Water Reactors 4.1. Concept of severe accident A severe accident or core melt accident at a PWR is an accident during which the reactor fuel is significantly
More informationThe Fukushima Daiichi Incident
The data and information contained herein are provided solely for informational purposes. None of the information or data is intended by AREVA to be a representation or a warranty of any kind, expressed
More informationIn-core measurements of fuel-clad interactions in the Halden reactor
In-core measurements of fuel-clad interactions in the Halden reactor Peter Bennett Halden Project IAEA Technical Meeting on Fuel Rod Instrumentation and In-Pile Measurement Techniques Halden, Norway 3
More informationSIMULATION OF THE QUENCH-06 EXPERIMENT WITH MELCOR 1.8.5
International Conference Nuclear Energy in Central Europe 2001 Hoteli Bernardin, Portorož, Slovenia, September 10-13, 2001 www: http://www.drustvo-js.si/port2001/ e-mail: PORT2001@ijs.si tel.:+ 386 1 588
More informationJoint ICTP-IAEA Course on Natural Circulation Phenomena and Passive Safety Systems in Advanced Water Cooled Reactors
2152-5 Joint ICTP-IAEA Course on Natural Circulation Phenomena and Passive Safety Systems in Advanced Water Cooled Reactors 17-21 May 2010 SELECTED EXAMPLES OF NATURAL CIRCULATION FOR SMALL BREAK LOCA
More informationA study of the revaporisation behaviour of deposits from the metallic vertical line of Phébus FPT3
A study of the revaporisation behaviour of deposits from the metallic vertical line of Phébus FPT3 P. D. W. BOTTOMLEY 1, E. FONTANA 1, D. PAPAIOANNOU 1, G. MONTAGNIER 1, E. TEIXEIRA 1, C. DIEBOLD 1, S.
More informationRadiochemistry Webinars
National Analytical Management Program (NAMP) U.S. Department of Energy Carlsbad Field Office Radiochemistry Webinars Nuclear Fuel Cycle Series Chemistry and Radiochemistry of the Reactor Coolant System
More informationJournal of American Science 2014;10(2) Burn-up credit in criticality safety of PWR spent fuel.
Burn-up credit in criticality safety of PWR spent fuel Rowayda F. Mahmoud 1, Mohamed K.Shaat 2, M. E. Nagy 3, S. A. Agamy 3 and Adel A. Abdelrahman 1 1 Metallurgy Department, Nuclear Research Center, Atomic
More informationKANUPP IAEA Training. Primary Heat Transport System Chemistry Control. Overhead
KANUPP IAEA Training Primary Heat Transport System Chemistry Control Overhead 1 Primary Heat Transport System System Purpose Transport heat from the fission of natural uranium fuel in the reactor fuel
More informationPRESENTED AT THE NUCLEAR
PRESENTED AT THE NUCLEAR INTEGRATION PROJECT WORKSHOP THE BACK-END: HEALING THE ACHILLES HEEL OF THE NUCLEAR RENAISSANCE BY R. G. WYMER VANDERBILT UNIVERSITY MARCH 4, 2008 CONTRIBUTIONS OF SELECTED ACTINIDES
More informationANALYSES OF AN UNMITIGATED STATION BLACKOUT TRANSIENT WITH ASTEC, MAAP AND MELCOR CODE
ANALYSES OF AN UNMITIGATED STATION BLACKOUT TRANSIENT WITH ASTEC, MAAP AND MELCOR CODE Technical Meeting on the Status and Evaluation of Severe Accident Simulation Codes for Water F. Mascari 1, J. C. De
More informationA high temperature heating device for the study of fission product release
Wir schaffen Wissen heute für morgen A high temperature heating device for the study of fission product release Jolanta Švedkauskaitė-Le Gore, Niko Kivel, Ines Günther-Leopold Nuclear Energy and Safety
More informationContribution of Research Reactors to the Programmes for Research and Technological Development on the Safety
CN156/U-30/OR Contribution of Research Reactors to the Programmes for Research and Technological Development on the Safety J. Couturier, F. Pichereau, C. Getrey, J. Papin, B. Clément Institut de Radioprotection
More informationDeviations from the parabolic kinetics during oxidation
Deviations from the parabolic kinetics during oxidation of zirconium alloys Martin Steinbrück, Mirco Große Karlsruhe Institute of Technology,, Germany 17th International ti lsymposium on Zirconium i in
More informationSymposium on Risk Integrated Engineering January 21, 2019, Takeda Hall, The Univ. Tokyo Researches on Severe Accident and Risk Engineering
Symposium on Risk Integrated Engineering January 21, 2019, Takeda Hall, The Univ. Tokyo Researches on Severe Accident and Risk Engineering Koji Okamoto The University of Tokyo okamoto@n.t.u-tokyo.ac.jp
More informationICONE ADAM: AN ACCIDENT DIAGNOSTIC, ANALYSIS AND MANAGEMENT SYSTEM APPLICATIONS TO SEVERE ACCIDENT SIMULATION AND MANAGEMENT
Proceedings of ICONE 10: 10TH International Conference on Nuclear Engineering Arlington, VA, USA, April 14-18, 2002 ICONE10-22195 ADAM: AN ACCIDENT DIAGNOSTIC, ANALYSIS AND MANAGEMENT SYSTEM APPLICATIONS
More informationGerman Experimental Activities for Advanced Modelling and Validation Relating to Containment Thermal Hydraulics and Source Term
German Experimental Activities for Advanced Modelling and Validation Relating to Containment Thermal Hydraulics and Source Term H.-J. Allelein 1,2, S. Gupta 3, G. Poss 3, E.-A. Reinecke 2, F. Funke 4 1
More informationExperiments of the LACOMECO Project at KIT
Experiments of the LACOMECO Project at KIT A. MIASSOEDOV 1, M. KUZNETSOV 1, M. STEINBRÜCK 1, S. KUDRIAKOV 2 Z. HÓZER 3, I. KLJENAK 4, R. MEIGNEN 5, J.M. SEILER 6, A. TEODORCZYK 7 1 KIT, Karlsruhe (DE)
More informationsteam oxidation and post-quench mechanical
Effect of pre-oxide on Zircaloy-4 4high htemperature t steam oxidation and post-quench mechanical properties Guilbert S., Lacote P., Montigny G., Duriez C., Desquines J., Grandjean C. Institut de Radioprotection
More informationUSING NEW VERSIONS OF SEVERE ACCIDENT CODES FOR VVER- 440/213 TYPE NUCLEAR POWER PLANTS
USING NEW VERSIONS OF SEVERE ACCIDENT CODES FOR VVER- 440/213 TYPE NUCLEAR POWER PLANTS András Nemes, Pál Kostka nemes@nubiki.hu kostka@nubiki.hu TM on the Status and Evaluation of Severe Accident Simulation
More informationsevere accident progression in the BWR lower plenum and the modes of vessel failure
1 For Presentation at the ERMSAR Conference held in Marseilles, France, March 24-26, 2015 severe accident progression in the BWR lower plenum and the modes of vessel failure B. R. Sehgal S. Bechta Nuclear
More informationNuclear Engineering and Technology
Nuclear Engineering and Technology 48 (2016) 1174e1183 Available online at ScienceDirect Nuclear Engineering and Technology journal homepage: www.elsevier.com/locate/net Original Article Investigation
More informationANALYSIS ON NON-UNIFORM FLOW IN STEAM GENERATOR DURING STEADY STATE NATURAL CIRCULATION COOLING
ANALYSIS ON NON-UNIFORM FLOW IN STEAM GENERATOR DURING STEADY STATE NATURAL CIRCULATION COOLING Susyadi 1 and T. Yonomoto 2 1 Center for Reactor Technology and Nuclear Safety - BATAN Puspiptek, Tangerang
More informationSevere Accident Natural Circulation Studies at the INEL
Severe Accident Natural Circulation Studies at the INEL Manuscript Completed: November 1994 Date Published: February 1995 Prepared by P. D. Bayless, D. A. Brownson, C. A. Dobbe, K. R. Jones, J. E. O'Brien,
More informationNSSS Design (Ex: PWR) Reactor Coolant System (RCS)
NSSS Design (Ex: PWR) Reactor Coolant System (RCS) Purpose: Remove energy from core Transport energy to S/G to convert to steam of desired pressure (and temperature if superheated) and moisture content
More informationNUMERICAL STUDY OF IN-VESSEL CORIUM RETENTION IN BWR REACTOR
NUMERICAL STUDY OF IN-VESSEL CORIUM RETENTION IN BWR REACTOR M. VALINČIUS Lithuanian Energy Institute Kaunas, Lithuania Email: mindaugas.valincius@lei.lt A. KALIATKA Lithuanian Energy Institute Kaunas,
More informationAP1000 European 15. Accident Analysis Design Control Document
15.2 Decrease in Heat Removal by the Secondary System A number of transients and accidents that could result in a reduction of the capacity of the secondary system to remove heat generated in the reactor
More informationVerification of the MELCOR Code Against SCDAP/RELAP5 for Severe Accident Analysis
Verification of the Code Against SCDAP/RELAP5 for Severe Accident Analysis Jennifer Johnson COLBERT 1* and Karen VIEROW 2 1 School of Nuclear Engineering, Purdue University, West Lafayette, Indiana 47907-2017,
More informationDesign bases and general design criteria for nuclear fuel. 1 General 3. 2 General design criteria 3
GUIDE 1 Nov. 1999 YVL 6.2 Design bases and general design criteria for nuclear fuel 1 General 3 2 General design criteria 3 3 Design criteria for normal operational conditions 4 4 Design criteria for operational
More informationPost-test results of the QUENCH-16 bundle test on air ingress: complex cladding oxidation during reflood and combined hydrogen
Post-test results of the QUENCH-16 bundle test on air ingress: complex cladding oxidation during reflood and combined hydrogen J. Stuckert, M. Steinbrück QWS18, Karlsruhe 2012 Institute for Applied Materials;
More informationEXPERIMENTS ON AIR INGRESS DURING SEVERE ACCIDENTS
13 th International Conference on Nuclear Engineering Beijing, China, May 16-20, 2005 ICONE13-50080 EXPERIMENTS ON AIR INGRESS DURING SEVERE ACCIDENTS Martin Steinbrück *, Alexei Miassoedov **, Gerhard
More informationIRSN views and perspectives on in-vessel melt retention strategy for severe accident mitigation
Florian Fichot Jean-Michel Bonnet Bernard Chaumont IRSN PSN-RES/SAG IRSN views and perspectives on in-vessel melt retention strategy for severe accident mitigation Outline 1. Key points for the feasibility
More informationPhysical Properties. Can increase the strength by cold working but the recrystallization temperature is 400 to 500 C
Zirconium Cladding Why? Physical Properties Corrosion Resistance Radiation Effects ----------------------------------------------- In the early 1950Õs the Navy was looking for a material with low σ a high
More informationR&D activities related to nuclear fuel performance and technology at the DG JRC. Paul VAN UFFELEN
R&D activities related to nuclear fuel performance and technology at the DG JRC Paul VAN UFFELEN 1 Introduction 2 JRC Core Staff (2004) Institute for Reference Materials and Measurements Institute for
More informationSupporting Deterministic T-H Analyses for Level 1 PSA
Supporting Deterministic T-H Analyses for Level 1 PSA ABSTRACT SLAVOMÍR BEBJAK VUJE, a.s. Okružná 5 918 64 Trnava, Slovakia slavomir.bebjak@vuje.sk TOMÁŠ KLIMENT VUJE, a.s. Okružná 5 918 64 Trnava, Slovakia
More informationSAFETY ENHANCEMENT TECHNOLOGY DEVELOPMENT WITH COLLABORATIVE INTERNATIONAL ACTIVITY
SAFETY ENHANCEMENT TECHNOLOGY DEVELOPMENT WITH COLLABORATIVE INTERNATIONAL ACTIVITY KENJI ARAI Toshiba Corporation Yokohama, Japan Email: kenji2.arai@toshiba.co.jp FUMIHIKO ISHIBASHI Toshiba Corporation
More informationBehavior of high burnup fuel during LOCA - Key observations and test plan at JAEA -
Behavior of high burnup fuel during LOCA - Key observations and test plan at JAEA - Fumihisa Nagase Japan Atomic Energy Agency IAEA Technical Meeting on Fuel Behaviour and Modelling under Severe Transient
More informationAcceptance Criteria in DBA
IAEA Safety Assessment Education and Training (SAET) Programme Joint ICTP-IAEA Essential Knowledge Workshop on Deterministic Safety Assessment and Engineering Aspects Important to Safety Acceptance Criteria
More informationELFR The European Lead Fast Reactor DESIGN, SAFETY APPROACH AND SAFETY CHARACTERISTICS. Alessandro Alemberti
ELFR The European Lead Fast Reactor DESIGN, SAFETY APPROACH AND SAFETY CHARACTERISTICS Alessandro Alemberti Alessandro.Alemberti@ann.ansaldo.it TECHNICAL MEETING ON IMPACT OF FUKUSHIMA EVENT ON CURRENT
More informationEPR: Steam Generator Tube Rupture analysis in Finland and in France
EPR: Steam Generator Tube Rupture analysis in Finland and in France S. ISRAEL Institut de Radioprotection et de Sureté Nucléaire BP 17 92262 Fontenay-aux-Roses Cedex, France Abstract: Different requirements
More informationIrradiation Assisted Stress Corrosion Cracking. By Topan Setiadipura [09M51695] (Obara Lab., Nuclear Engineering Dept., Tokyo Tech.
Introduction Short Review on Irradiation Assisted Stress Corrosion Cracking By Topan Setiadipura [09M51695] (Obara Lab., Nuclear Engineering Dept., Tokyo Tech.) Irradiation-assisted stress-corrosion cracking
More information