| 1. |
- Nayak, Arun Kumar, et al.
(författare)
-
A numerical and experimental study of water ingression phenomena in melt pool coolability
- 2009
-
Ingår i: Nuclear Engineering and Design. - : Elsevier BV. - 0029-5493 .- 1872-759X. ; 239:7, s. 1285-1293
-
Tidskriftsartikel (refereegranskat)abstract
- During a postulated severe accident, the core can melt and the melt can fail the reactor vessel. Subsequently, the molten corium can be relocated in the containment cavity forming a melt pool. The melt pool can be flooded with water at the top for quenching it. However, the question that arises is to what extent the water can ingress in the corium melt pool to cool and quench it. To reveal that, a numerical study has been carried out using the computer code MELCOOL The code considers the heat transfer behaviour in axial and radial directions from the molten pool to the overlaying water, crust generation and growth, thermal stresses built-in the crust, disintegration of crust into debris, natural convection heat transfer in debris and water ingression into the debris bed. To validate the computer code, experiments were conducted in a facility named as core melt coolability (COMECO). The facility consists of a test section (200 mm x 200 mm square cross-section) and with a height of 300 mm. About 14 L of melt comprising of 30% CaO + 70% B2O3 (by wt.) was poured into the test section. The melt was heated by four heaters from outside the test section to simulate the decay heat of corium. The melt was water flooded from the top, and the depth of water pool was kept constant at around 700 mm throughout the experiment. The transient temperature behaviour in the melt pool at different axial and radial locations was measured with 24 K-type thermocouples and the steam flow rate was measured using a vortex flow meter. The melt temperature measurements indicated that water could ingress only up to a certain depth into the melt pool. The MELCOOL predictions were compared with the test data for the temperature distribution inside the molten pool. The code was found to simulate the quenching behaviour and depth of water ingression quite well.
|
|
| 2. |
|
|
| 3. |
- Albiol, T., et al.
(författare)
-
SARNET : Severe accident research network of excellence
- 2010
-
Ingår i: PROG NUCL ENERGY. - : Elsevier BV. ; , s. 2-10
-
Konferensbidrag (refereegranskat)abstract
- Fifty-one organisations network in SARNET (Severe Accident Research NETwork of Excellence) their research capacities in order to resolve the most important pending issues for enhancing, with regard to Severe Accidents (SA). the safety of existing and future Nuclear Power Plants (NPPs). This project. co-funded by the European Commission (EC) under the 6th Framework Programme, has been defined in order to optimise the use of the available means and to constitute sustainable research groups in the European Union. SARNET tackles the fragmentation that may exist between the different national R&D programmes, in defining common research programmes and developing common computer tools and methodologies for safety assessment. SARNET comprises most of the organisations involved in SA research in Europe, plus Canada. To reach these objectives, all the organisations networked in SARNET contributed to a joint Programme of Activities, which consisted of: Implementation of an advanced communication tool for accessing all project information, fostering exchange of information, and managing documents: Harmonization and re-orientation of the research programmes, and definition of new ones; Analysis of the experimental results provided by research programmes in order to elaborate a common understanding of relevant phenomena; Development of the ASTEC code (integral computer code used to predict the NPP behaviour during a postulated SA), which capitalizes in terms of physical models the knowledge produced within SARNET; Development of Scientific Databases in which all the results of research programmes are stored in a common format (DATANET); Development of a common methodology for Probabilistic Safety Assessment of NPPs; Development of short courses and writing a textbook on Severe Accidents for students and researchers; Promotion of personnel mobility amongst various European organisations. This paper presents the major achievements after four and a half years of operation of the network, in terms of knowledge gained, of improvement of the ASTEC reference code, of dissemination of results and of integration of the research programmes conducted by the various partners. After this first period (2004-2008), co-funded by the EC, a further contract SARNET2 with the EC for the next four years started in April 2009 as part of the 7th Framework Programme. During this period, the networking activities will focus mainly on the remaining pending issues as determined during the first period, experimental activities will be directly included in the common work and the network will evolve toward complete self-sustainability. The bases for such an evolution are presented in the last part of the paper.
|
|
| 4. |
- Bechta, Sevostian, et al.
(författare)
-
On the EU-Japan roadmap for experimental research on corium behavior
- 2019
-
Ingår i: Annals of Nuclear Energy. - : PERGAMON-ELSEVIER SCIENCE LTD. - 0306-4549 .- 1873-2100. ; 124, s. 541-547
-
Tidskriftsartikel (refereegranskat)abstract
- A joint research roadmap between Europe and Japan has been developed in severe accident field of light water reactors, focusing particularly on reactor core melt (corium) behavior. The development of this roadmap is one of the main targets of the ongoing EU project SAFEST. This paper presents information about ongoing severe accident studies in the area of corium behavior, rationales and comparison of research priorities identified in different projects and documents, expert ranking of safety issues, and finally the research areas and topics and their priorities suggested for the EU Japan roadmap and future bilateral collaborations. These results provide useful guidelines for (i) assessment of long-term goals and proposals for experimental support needed for proper understanding, interpretation and learning lessons of the Fukushima accident; (ii) analysis of severe accident phenomena; (iii) development of accident prevention and mitigation strategies, and corresponding technical measures; (iv) study of corium samples in European and Japanese laboratories; and (v) preparation of Fukushima site decommissioning.
|
|
| 5. |
- Journeau, C., et al.
(författare)
-
Safest roadmap for corium experimental research in Europe
- 2018
-
Ingår i: ASCE-ASME J of Risk & Uncertainty in Engineering Systems Part B. - : ASME Press. - 2332-9017 .- 2332-9025. ; 4:3
-
Tidskriftsartikel (refereegranskat)abstract
- Severe accident facilities for European safety targets (SAFEST) is a European project networking the European experimental laboratories focused on the investigation of a nuclear power plant (NPP) severe accident (SA) with reactor core melting and formation of hazardous material system known as corium. The main objective of the project is to establish coordinated activities, enabling the development of a common vision and severe accident research roadmaps for the next years, and of the management structure to achieve these goals. In this frame, a European roadmap on severe accident experimental research has been developed to define research challenges to contribute to further reinforcement of Gen II and III NPP safety. The roadmap takes into account different SA phenomena and issues identified and prioritized in the analyses of severe accidents at commercial NPPs and in the results of the recent European stress tests carried out after the Fukushima accident. Nineteen relevant issues related to reactor core meltdown accidents have been selected during these efforts. These issues have been compared to a survey of the European SA research experimental facilities and corium analysis laboratories. Finally, the coherence between European infrastructures and R&D needs has been assessed and a table linking issues and infrastructures has been derived. The comparison shows certain important lacks in SA research infrastructures in Europe, especially in the domains of core late reflooding impact on source term, reactor pressure vessel failure and molten core release modes, spent fuel pool (SFP) accidents, as well as the need for a large-scale experimental facility operating with up to 500 kg of chemically prototypic corium melt.
|
|
| 6. |
- Ma, Weimin, et al.
(författare)
-
Experimental study on natural circulation and its stability in a heavy liquid metal loop
- 2007
-
Ingår i: Nuclear Engineering and Design. - : Elsevier BV. - 0029-5493 .- 1872-759X. ; 237:15-17, s. 1838-1847
-
Tidskriftsartikel (refereegranskat)abstract
- Motivated by the increasing interest in heavy liquid metal (HLM) cooled fast reactors and accelerator driven system (ADS), the TALL test facility was designed and constructed at KTH to investigate the thermal-hydraulic characteristics of HLM. In this paper, the HLM natural circulation characteristics in a HLM loop were investigated with experiments in the TALL test facility. The study includes measurements on (1) start-up of natural circulation from different initial conditions; (2) stability of natural circulation; (3) effects of influencing parameters and (4) capability of natural circulation. The experimental data are compared to predictions with a relevant code (RELAP5). Significant natural convection flow was observed in the experiments. It was found that the natural circulation was easily established and stabilized. It took only a few minutes to have a stable natural circulation prevailing from cold conditions. The natural circulation flowrate depends on the loop resistance, and the temperature difference between the hot leg and the cold leg, as determined by the power level and the heat sink capacity. The experiments show that the maximum flowrate for the natural circulation is similar to 0.5 kg/s (corresponding to similar to 0.5 m/s in the heat exchanger), resulting in heat removal of similar to 15 kW from the core tank, which is comparable to the capacity of similar to 100 W/cm of the electric heater elements. The preliminary analysis performed with the RELAP5 code is in reasonable agreement with the experimental data.
|
|
| 7. |
- Ma, Weimin, et al.
(författare)
-
In-Vessel Melt Retention of Pressurized Water Reactors : Historical Review and Future Research Needs
- 2016
-
Ingår i: Engineering. - : Elsevier. - 2095-8099. ; 2:1, s. 103-111
-
Forskningsöversikt (refereegranskat)abstract
- A historical review of in-vessel melt retention (IVR) is given, which is a severe accident mitigation measure extensively applied in Generation III pressurized water reactors (PWRs). The idea of IVR actually originated from the back-fitting of the Generation II reactor Loviisa VVER-440 in order to cope with the core-melt risk. It was then employed in the new deigns such as Westinghouse AP1000, the Korean APR1400 as well as Chinese advanced PWR designs HPR1000 and CAP1400. The most influential phenomena on the IVR strategy are in-vessel core melt evolution, the heat fluxes imposed on the vessel by the molten core, and the external cooling of the reactor pressure vessel (RPV). For in-vessel melt evolution, past focus has only been placed on the melt pool convection in the lower plenum of the RPV; however, through our review and analysis, we believe that other in-vessel phenomena, including core degradation and relocation, debris formation, and coolability and melt pool formation, may all contribute to the final state of the melt pool and its thermal loads on the lower head. By looking into previous research on relevant topics, we aim to identify the missing pieces in the picture. Based on the state of the art, we conclude by proposing future research needs.
|
|
| 8. |
- Ma, Weimin, et al.
(författare)
-
Thermal-hydraulic performance of heavy liquid metal in straight-tube and U-tube heat exchangers
- 2009
-
Ingår i: Nuclear Engineering and Design. - : Elsevier BV. - 0029-5493 .- 1872-759X. ; 239:7, s. 1323-1330
-
Tidskriftsartikel (refereegranskat)abstract
- Motivated by an increased interest in heavy liquid metal (lead or lead alloy) cooled fast reactors (LFR) and accelerator-driven system (ADS), the present paper presents a study on resistance characteristics and heat transfer performance of liquid lead bismuth eutectic (LBE) flow through a straight-tube heat exchanger and a U-tube heat exchanger. The investigation is performed on the TALL test facility at KTH. The heat exchangers have counter-current flow arrangement, and are made from a pair of 1-m-long concentric ducts, with the LBE flowing in the inner tube of 10mm I.D. and the secondary coolant flowing in the annulus. The inlet temperature of LBE into the heat exchangers is from 200 degrees C to 450 degrees C with temperature drops from 0 degrees C to 100 degrees C within the LBE flow range of Re = 10(4)-10(5). Analysis of the experimental results obtained provides a basic understanding and quantification of the regimes of lead-bismuth flow and heat transfer through a straight tube and a U-shaped tube. The unique data base also serves as benchmark and improvement for system thermal-hydraulic codes (e.g. RELAP, TRAC/AAA) whose development and testing were dominantly driven by applications in water-cooled systems. Lessons and insights learnt from the study and recommendations for the heat exchanger selection are discussed.
|
|
| 9. |
- Nayak, Arun K., et al.
(författare)
-
An experimental study on quenching of a radially stratified heated porous bed
- 2006
-
Ingår i: Nuclear Engineering and Design. - : Elsevier BV. - 0029-5493 .- 1872-759X. ; 236:19-21, s. 2189-2198
-
Tidskriftsartikel (refereegranskat)abstract
- The quenching characteristics of a volumetrically-heated particulate bed composed of radially stratified sand layers were investigated experimentally in the POMECO facility. The sand bed simulates the corium particulate debris bed which is formed when the molten corium released from the vessel fragments in water and deposits on the cavity floor during a postulated severe accident in a light water reactor (LWR). The electrically-heated bed was quenched by water from a water column established over top of it, and later also with water coming from its bottom, which was circulating from the water overlayer through downcomers. A series of experiments were conducted to reveal the effects of the size of downcomers, and their locations in the bed, on the quenching characteristics of the radially stratified debris beds. The downcomers were found to significantly increase the bed quenching rate. To simulate the non-condensable gases generated during the MCCI, air and argon were injected from the bottom of the bed at different flow rates. The effects of gas flow rate and its properties on the quenching behaviour were observed. The results indicate that the non-condensable gas flows reduce the quenching rate significantly. The gas properties also affect the quenching characteristics.
|
|
| 10. |
- Rychkov, Maxym, et al.
(författare)
-
Analysis of VVER-1000 large and small break LOCA experiments with RELAP5/MOD3.2
- 2004
-
Ingår i: Proc. Int. Conf. Nucl. Eng. ICONE. - : ASMEDC. ; , s. 687-696
-
Konferensbidrag (refereegranskat)abstract
- A RELAP5 model for the analysis of the PSB-VVER test facility was developed by the EREC in Russia. The PSB-VVER is a large-scale integral test facility to model the VVER-1000 type NPP. The volume and power scale in this test facility is 1:300 and the elevation scale is 1:1. which corresponds to the elevation mark of the reactor prototype. At the Division of Nuclear Safety, RIT, Sweden, we have modified the PSB-VVER faciltiy's RELAP5 model in order to analyze two of the transient tests performed on the PSB-VVER facility, which serve as the validation matrix described by NEA/CSNI. The objective of the work conducted was to validate the results obtained from RELAP5's calculation with the supplied experimental data from the PSB-VVER test facility. Two accident scenarios have been calculated and analyzed. After being verified against the "11% UP LOCA" test data, the RELAP5/MOD3.2 model was used for a so-called "blind" transient calculation of the test "2×25% HL LOCA" and the results obtained were compared with the experimental data provided after the calculation.
|
|
