Ex-Vessel Coolability Analysis for a Belgian NPP

2009 ◽  
Author(s):  
Jarne R. Verpoorten ◽  
Miche`le Auglaire ◽  
Frank Bertels
Keyword(s):  
State Of The Art ◽  
The State ◽  
Severe Accident ◽  
Research Projects ◽  
Specific Data ◽  
The Core ◽  
Core Damage ◽  
Accident Management ◽  
Lower Head ◽  
Core Cooling

During a hypothetical Severe Accident (SA), core damage is to be expected due to insufficient core cooling. If the lack of core cooling persists, the degradation of the core can continue and could lead to the presence of corium in the lower plenum. There, the thermo-mechanical attack of the lower head by the corium could eventually lead to vessel failure and corium release to the reactor cavity pit. In this paper, it is described how the international state-of-the-art knowledge has been applied in combination with plant-specific data in order to obtain a custom Severe Accident Management (SAM) approach and hardware adaptations for existing NPPs. Also the interest of Tractebel Engineering in future SA research projects related to this topic will be addressed from the viewpoint of keeping the analysis up-to-date with the state-of-the art knowledge.

scholarly journals Simulation of a severe accident at a typical PWR due to break of a hot leg ECCS injection line using MELCOR code.

2019 ◽  
Vol 7 (2B) ◽  
Author(s):  
Seung Min Lee ◽  
Nelbia Da Silva Lapa ◽  
Gaianê Sabundjian
Keyword(s):  
Cooling System ◽  
Severe Accident ◽  
Time Intervals ◽  
Severe Accidents ◽  
Independent Analysis ◽  
Accident Management ◽  
Injection Line ◽  
Core Cooling ◽  
Control Volumes ◽  
Physical Phenomena

The aim of this work was to simulate a severe accident at a typical PWR, initiated with a break in Emergency Core Cooling System line of a hot leg, using the MELCOR code. The model of this typical PWR was elaborated by the Global Research for Safety and provided to the CNEN for independent analysis of the severe accidents at Angra 2, which is similar to this typical PWR. Although both of them are not identical, the results obtained of that typical PWR may be valuable because of the lack of officially published simulation of severe accident at Angra 2. Relevant parameters such as pressure, temperature and water level in various control volumes, after the break at the hot leg, were calculated as well as degree of core degradation and hydrogen production within the containment. The result obtained in this work could be considered satisfactory in the sense that the physical phenomena reproduced by the simulation were in general very reasonable, and most of the events occurred within acceptable time intervals. However, the uncertainty analysis was not carried out in this work. Furthermore, this scenario could be used as a base for the study of the effectiveness of some preventive or/and mitigating measures of Severe Accident Management by implementing each measure in this model.

scholarly journals Assessment of Severe Accident Management for Small IPWR under an ESBO Scenario

2019 ◽  
Vol 2019 ◽  
pp. 1-10 ◽  
Author(s):  
Hao Yu ◽  
Minjun Peng
Keyword(s):  
High Pressure ◽  
Low Pressure ◽  
Lower Pressure ◽  
Severe Accident ◽  
Maximum Temperature ◽  
Pressurized Water ◽  
The Core ◽  
Severe Accidents ◽  
Accident Management ◽  
Degradation Studies

Interest in evaluation of severe accidents induced by extended station blackout (ESBO) has significantly increased after Fukushima. In this paper, the severe accident process under the high and low pressure induced by an ESBO for a small integrated pressurized water reactor (IPWR)-IP200 is simulated with the SCDAP/RELAP5 code. For both types of selected scenarios, the IP200 thermal hydraulic behavior and core meltdown are analyzed without operator actions. Core degradation studies firstly focus on the changes in the core water level and temperature. Then, the inhibition of natural circulation in the reactor pressure vessel (RPV) on core temperature rise is studied. In addition, the phenomena of core oxidation and hydrogen generation and the reaction mechanism of zirconium with the water and steam during core degradation are analyzed. The temperature distribution and time point of the core melting process are obtained. And the IP200 severe accident management guideline (SAMG) entry condition is determined. Finally, it is compared with other core degradation studies of large distributed reactors to discuss the influence of the inherent design characteristics of IP200. Furthermore, through the comparison of four sets of scenarios, the effects of the passive safety system (PSS) on the mitigation of severe accidents are evaluated. Detailed results show that, for the quantitative conclusions, the low coolant storage of IP200 makes the core degradation very fast. The duration from core oxidation to corium relocation in the lower-pressure scenario is 53% faster than that of in the high-pressure scenario. The maximum temperature of liquid corium in the lower-pressure scenario is 134 K higher than that of the high-pressure scenario. Besides, the core forms a molten pool 2.8 h earlier in the lower-pressure scenario. The hydrogen generated in the high-pressure scenario is higher when compared to the low-pressure scenario due to the slower degradation of the core. After the reactor reaches the SAMG entry conditions, the PSS input can effectively alleviate the accident and prevent the core from being damaged and melted. There is more time to alleviate the accident. This study is aimed at providing a reference to improve the existing IPWR SAMGs.

Test Design, Results, and Archived Database for the OECD Lower Head Failure Program Integral Experiments

2006 ◽  
Author(s):  
Larry L. Humphries ◽  
Tze Yao Chu ◽  
John H. Bentz
Keyword(s):  
Wall Temperature ◽  
Pressure Vessel ◽  
Severe Accident ◽  
Core Material ◽  
Material Strength ◽  
Test Design ◽  
Temperature Differential ◽  
Pressurized Water ◽  
Accident Management ◽  
Lower Head

In the event of a severe core meltdown accident, core material can relocate to the lower head of a pressurized water reactor (PWR) vessel resulting in significant thermal and pressure loads to the vessel. The potential for failure of the pressure vessel makes possible the release of core material to the containment. The objective of this experimental/analytical program is to characterize the mode, timing, and size of lower head failure (LHF) under severe accident conditions. The OECD Lower Head Failure (OLHF) project investigates lower head failure for conditions of low reactor coolant system (RCS) pressure (2–5 MPa) and prototypic through-wall temperature differential (ΔTW >200K). Low RCS pressure is motivated by the desire to use the data to develop models for assessing accident management strategies involving reactor pressure vessel (RPV) depressurization. Pressure transient is useful in assessing the effect of water injection as part of accident management strategy. Prototypic through-wall temperature differential, ΔTW, is of importance because of the need to provide data where stress redistribution in the vessel wall occurs (as a result of decreasing material strength with temperature). Test design and results for the four OLHF integral tests are reported and summarized in this paper. A short description of the test conduct and heating history is followed by a description of the vessel failure site, the vessel deformation, temperature profiles, stress state, and rupture dynamics for each test. Key observations and conclusions are summarized for each test. The ∼1/5 scale tests are extensively instrumented to provide temperature, pressure, and displacement data. The vessel surfaces are mapped both before and after the test to provide measurements of pre-test thickness, post-test thickness, and cumulative vessel deformation. Data has been assessed and qualified in data reports for each test. The data has been preserved in MSEXCEL™ spreadsheets with macro utilities to facilitate access and analysis of the data. As a result, there exists a well-archived, well-qualified database for model development and validation.

Study on the Accidental Rupture of Hot Leg or Surge Line in SBO Accident

2006 ◽  
Author(s):  
Kun Zhang ◽  
Xuewu Cao
Keyword(s):  
Pressure Vessel ◽  
Severe Accident ◽  
Base Case ◽  
Management Guidelines ◽  
Accident Management ◽  
Lower Head ◽  
Station Blackout ◽  
Surge Line ◽  
Reactor Pressure ◽  
Relap5 Code

The postulated total station blackout accident (SBO) of PWR NPP with 600 MWe in China is analyzed as the base case using SCDAP/RELAP5 code. Then the hot leg or surge line are assumed to rupture before the lower head of Reactor Pressure Vessel (RPV) ruptures, and the progressions are analyzed in detail comparing with the base case. The results show that the accidental rupture of hot leg or surge line will greatly influence the progression of accident. The probability of hot leg or surge line rupture in intentional depressurization is also studied in this paper, which provides a suggestion to the development of Severe Accident Management Guidelines (SAMG).

scholarly journals Classification of emotions from the recognition of facial expressions applied to the prevention of secondary Alexithymia: an analysis of the state of the art

2020 ◽  
pp. 14-17
Author(s):  
Christian PADILLA-NAVARRO ◽  
Carlos ZARATE-TREJO ◽  
Georges KHALAF ◽  
Pascal FALLAVOLLITA
Keyword(s):  
Artificial Intelligence ◽  
Facial Expressions ◽  
State Of The Art ◽  
The State ◽  
Research Projects ◽  
Emotional Facial Expressions ◽  
Toronto Alexithymia Scale ◽  
Recognition Of Facial Expressions

Alexithymia is a condition that partially or completely deprives you of the ability to identify and describe emotions, and to show affective connotations in the actions of an individual. This problem has been taken to different research projects that seek to study its characteristics, forms of prevention, and implications, and that try to determine a measurement for the experience of an individual with this construct as well as the responses they provide to certain stimuli. Other studies that were reviewed aimed to find a connection between the responses of subjects diagnosed with alexithymia when facing a dynamic of emotional facial expressions to recognize and their assigned grade based on the Toronto Alexithymia Scale (TAS), a metric frequently used to evaluate the presence or absence of alexithymia in an individual. In this work, a review of the different articles that study this connection, as well as articles that describe the state of the art of the implementation of artificial intelligence algorithms applied to the treatment or prevention of secondary alexithymia is presented.

scholarly journals A Biscriptual Morphological Transducer for Crimean Tatar

2019 ◽  
Author(s):  
Francis M. Tyers ◽  
Jonathan N. Washington ◽  
Darya Kavitskaya ◽  
Memduh Gökırmak
Keyword(s):  
Computational Linguistics ◽  
Morphological Analysis ◽  
State Of The Art ◽  
Full Range ◽  
The State ◽  
Loan Words ◽  
The Core ◽  
Finite State ◽  
Morphological Modelling ◽  
Crimean Tatar

This paper describes a weighted finite-state morphological transducer for Crimean Tatar able to analyse and generate in both Latin and Cyrillic orthographies. This transducer was developed by a team including a community member and language expert, a field linguist who works with the community, a Turkologist with computational linguistics expertise, and an experienced computational linguist with Turkic expertise. Dealing with two orthographic systems in the same transducer is challenging as they employ different strategies to deal with the spelling of loan words and encode the full range of the language's phonemes and their interaction. We develop the core transducer using the Latin orthography and then design a separate transliteration transducer to map the surface forms to Cyrillic. To help control the non-determinism in the orthographic mapping, we use weights to prioritise forms seen in the corpus. We perform an evaluation of all components of the system, finding an accuracy above 90% for morphological analysis and near 90% for orthographic conversion. This comprises the state of the art for Crimean Tatar morphological modelling, and, to our knowledge, is the first biscriptual single morphological transducer for any language.

Experimental Study on the Improved In-Vessel Corium Retention Concepts for the Severe Accident Management

2002 ◽  
Author(s):  
K. H. Kang ◽  
R. J. Park ◽  
K. M. Koo ◽  
S. B. Kim ◽  
H. D. Kim
Keyword(s):  
Heat Removal ◽  
Elevated Pressure ◽  
Reactor Vessel ◽  
Experimental Results ◽  
Severe Accident ◽  
Experimental Facility ◽  
Dual Strategy ◽  
Accident Management ◽  
Lower Head ◽  
Inner Diameter

Feasibility experiments were performed for the assessment of improved In-Vessel Corium Retention (IVR) concepts using an internal engineered gap device and also a dual strategy of In/Ex-vessel cooling using the LAVA experimental facility. The internal engineered gap device made of carbon steel was installed inside the LAVA lower head vessel and it made a uniform gap with the vessel by 10 mm. In/Ex-vessel cooling in the dual strategy experiment was performed installing an external guide vessel outside the LAVA lower head vessel at a uniform gap of 25 mm. The LAVA lower head vessel was a hemispherical test vessel simulated with a 1/8 linear scale mock-up of the reactor vessel lower plenum with an inner diameter of 500 mm and thickness of 25 mm. In both of the tests, Al2O3 melt was delivered into about 50K subcooled water inside the lower head vessel under the elevated pressure. Temperatures of the internal engineered gap device and the lower head vessel were measured by K-type thermocouples embedded radially in the 3mm depth of the lower head vessel outer surface and in the 4mm depth of the internal engineered gap device, respectively. In the dual strategy experiment, the Ex-vessel cooling featured pool boiling in the gap between the lower head vessel and the external guide vessel. It could be found from the experimental results that the internal engineered gap device was intact and so the vessel experienced little thermal and mechanical attacks in the internal engineered gap device experiment. And also the vessel was effectively cooled via mutual boiling heat removal in- and ex-vessel in the dual strategy experiment. Compared with the previous LAVA experimental results performed for the investigation of the inherent in-vessel gap cooling, it could be confirmed that the Ex-vessel cooling measure was dominant over the In-vessel cooling measure in this study. It is concluded that the improved cooling measures using a internal engineered gap device and a dual strategy promote the cooling characteristics of the lower head vessel and so enhance the integrity of the vessel in the end.

Thermal and Structural Analysis of Reactor Vessel Lower Head Considering Core Meltdown Accident

2018 ◽  
Author(s):  
Juan Luo ◽  
Jiacheng Luo ◽  
Lei Sun ◽  
Peng Tang
Keyword(s):  
Structural Analysis ◽  
Creep Deformation ◽  
Structural Integrity ◽  
Pressure Vessels ◽  
Reactor Vessel ◽  
Severe Accident ◽  
Creep Failure ◽  
The Core ◽  
Lower Head ◽  
Core Meltdown

In the core meltdown severe accident, in-vessel retention (IVR) of molten core debris by external reactor vessel cooling (ERVC) is an important mitigation strategy. During the IVR strategy, the core debris forming a melt pool in the reactor pressure vessel (RPV) lower head (LH) will produce extremely high thermal and mechanical loadings to the RPV, which may cause the failure of RPV due to over-deformation of plasticity or creep. Therefore, it is necessary to study the thermomechanical behavior of the reactor vessel LH during IVR condition. In this paper, under the assumption of IVR-ERVC, the thermal and structural analysis for the RPV lower head is completed by finite element method. The temperature field and stress field of the RPV wall, and the plastic deformation and creep deformation of the lower head are obtained by calculation. Plasticity and creep failure analysis is conducted as well. Results show that under the assumed conditions, the head will not fail due to excessive creep deformation within 200 hours. The results can provide basis for structural integrity analysis of pressure vessels.

Analysis of Severe Accidents in VVER-1000 Reactors Using the Integral Code ASTEC

2009 ◽  
Author(s):  
Polina Tusheva ◽  
Nils Reinke ◽  
Eberhard Altstadt ◽  
Frank Schaefer ◽  
Frank-Peter Weiss ◽  
...  
Keyword(s):  
Power System ◽  
Failure Time ◽  
Late Phase ◽  
Electric Power System ◽  
Physical Models ◽  
Severe Accident ◽  
Feed Water ◽  
The Core ◽  
Lower Head ◽  
Station Blackout

The studies presented are aiming at a detailed investigation of the behaviour of a VVER-1000/V-320 reactor and the containment structures during a postulated severe accident, including the ways and means by which these accidents may be prevented or mitigated. A hypothetical station blackout scenario (loss of the offsite electric power system concurrent with a turbine trip and unavailability of the emergency AC power system), belonging to the typical beyond design basis accidents, has been investigated. Station blackout results in reactor shut down, loss of feed water and trip of all reactor coolant pumps. Continuous evaporation of the secondary side leads to steam generators’ depletion followed by heating up of the core. In case of unavailability of essential safety systems the core will be severely damaged and finally the reactor pressure vessel (RPV) might fail. The analyses are performed using the integral code ASTEC commonly developed by IRSN (Institut de Radioprotection et de Suˆrete´ Nucle´aire) and GRS (Gesellschaft fu¨r Anlagen- und Reaktorsicherheit mbH). Code-to-code comparative analyses for the early thermal-hydraulic phase have been performed with the GRS code ATHLET. A large number of sensitivity calculations have been done regarding the axial core power distribution, heat losses, and RPV lower head modelling. The analyses have shown that, despite the considerable differences in the codes themselves, the calculation results are similar in terms of thermal hydraulic response. There are discrepancies in timings of phenomena, which are within the limitations of the physical models and the applied nodalizations. It was one objective of this investigation to evaluate the Severe Accident Management (SAM) procedures for VVER-1000 reactors, by for instance estimating the time available for taking appropriate decisions and preparing counter-measures. To evaluate the effect of possible operator actions, a SAM procedure (primary side depressurization) is included into the simulation. Without SAMs, the simulation provides plastic rupture of the RPV after approximately 4.3 h, while with SAMs, a prolongation of the vessel failure time is obtained by approximately 90 minutes. Currently, the late phase of the accident is investigated in more detail by comparing the lower head behaviour as simulated by ASTEC with results from dedicated finite element calculations. The work contributes to the reliability of the ASTEC code by means of plant applications.

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