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Thematische Bibliographien / In-vessel melt retention / Zeitschriftenartikel

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Zeitschriftenartikel zum Thema „In-vessel melt retention“

Autor: Grafiati

Veröffentlicht am 24. Juni 2021

Zuletzt aktualisiert am 1. Februar 2022

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1

Almyashev, V. I., V. S. Granovsky, V. B. Khabensky, et al. "Oxidation effects during corium melt in-vessel retention." Nuclear Engineering and Design 305 (August 2016): 389–99. http://dx.doi.org/10.1016/j.nucengdes.2016.05.024.

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2

Kang, Kyoung-Ho, Rae-Joon Park, Sang-Baik Kim, Hee-Dong Kim, and Soon-Heung Chang. "Simulant Melt Experiments on In-Vessel Retention Through External Reactor Vessel Cooling." Nuclear Technology 155, no. 3 (2006): 324–39. http://dx.doi.org/10.13182/nt06-a3765.

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3

Theofanous, T. G., C. Liu, S. Additon, S. Angelini, O. Kymäläinen, and T. Salmassi. "In-vessel coolability and retention of a core melt." Nuclear Engineering and Design 169, no. 1-3 (1997): 1–48. http://dx.doi.org/10.1016/s0029-5493(97)00009-5.

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4

Asmolov, V., N. N. Ponomarev-Stepnoy, V. Strizhov, and B. R. Sehgal. "Challenges left in the area of in-vessel melt retention." Nuclear Engineering and Design 209, no. 1-3 (2001): 87–96. http://dx.doi.org/10.1016/s0029-5493(01)00391-0.

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5

Jiang, Nan, Tenglong Cong, and Minjun Peng. "Margin evaluation of in-vessel melt retention for small IPWR." Progress in Nuclear Energy 110 (January 2019): 224–35. http://dx.doi.org/10.1016/j.pnucene.2018.10.003.

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6

Abendroth, M., H. G. Willschütz, and E. Altstadt. "Fracture mechanical evaluation of an in-vessel melt retention scenario." Annals of Nuclear Energy 35, no. 4 (2008): 627–35. http://dx.doi.org/10.1016/j.anucene.2007.08.007.

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7

Zvonarev, Yu A., A. M. Volchek, V. L. Kobzar, and M. A. Budaev. "ASTEC application for in-vessel melt retention modelling in VVER plants." Nuclear Engineering and Design 272 (June 2014): 224–36. http://dx.doi.org/10.1016/j.nucengdes.2013.06.044.

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8

Gencheva, R., A. Stefanova, P. Groudev, B. Chatterjee, and D. Mukhopadhyay. "Study of in-vessel melt retention for VVER-1000/v320 reactor." Nuclear Engineering and Design 298 (March 2016): 208–17. http://dx.doi.org/10.1016/j.nucengdes.2015.12.031.

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9

Valinčius, Mindaugas, Tadas Kaliatka, Algirdas Kaliatka, and Eugenijus Ušpuras. "Modelling of Severe Accident and In-Vessel Melt Retention Possibilities in BWR Type Reactor." Science and Technology of Nuclear Installations 2018 (August 1, 2018): 1–14. http://dx.doi.org/10.1155/2018/7162387.

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One of the severe accident management strategies for nuclear reactors is the melted corium retention inside the reactor pressure vessel. The work presented in this article investigates the application of in-vessel retention (IVR) severe accident management strategy in a BWR reactor. The investigations were performed assuming a scenario with the large break LOCA without injection of cooling water. A computer code RELAP/SCDAPSIM MOD 3.4 was used for the numerical simulation of the accident. Using a model of the entire reactor, a full accident sequence from the large break to core uncover and hea

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10

Granovsky, V. S., V. B. Khabensky, E. V. Krushinov, et al. "Oxidation effect on steel corrosion and thermal loads during corium melt in-vessel retention." Nuclear Engineering and Design 278 (October 2014): 310–16. http://dx.doi.org/10.1016/j.nucengdes.2014.07.034.

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11

Park, Hae-Kyun, and Bum-Jin Chung. "Mass Transfer Experiments for the Heat Load During In-Vessel Retention of Core Melt." Nuclear Engineering and Technology 48, no. 4 (2016): 906–14. http://dx.doi.org/10.1016/j.net.2016.02.015.

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12

Ma, Weimin, Yidan Yuan, and Bal Raj Sehgal. "In-Vessel Melt Retention of Pressurized Water Reactors: Historical Review and Future Research Needs." Engineering 2, no. 1 (2016): 103–11. http://dx.doi.org/10.1016/j.eng.2016.01.019.

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13

Doan, Manh Long, Van Thai Nguyen, and Chi Thanh Tran. "An analysis of In-Vessel Melt Retention strategy for VVER-1000 considering the effect of torospherical lower head vessel." Nuclear Engineering and Design 371 (January 2021): 110972. http://dx.doi.org/10.1016/j.nucengdes.2020.110972.

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14

Knudson, D. L., J. L. Rempe, K. G. Condie, K. Y. Suh, F. B. Cheung, and S. B. Kim. "Late-phase melt conditions affecting the potential for in-vessel retention in high power reactors." Nuclear Engineering and Design 230, no. 1-3 (2004): 133–50. http://dx.doi.org/10.1016/j.nucengdes.2003.11.029.

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15

Wang, Hongdi, Walter Villanueva, Yangli Chen, Artem Kulachenko, and Sevostian Bechta. "Thermo-mechanical behavior of an ablated reactor pressure vessel wall in a Nordic BWR under in-vessel core melt retention." Nuclear Engineering and Design 379 (August 2021): 111196. http://dx.doi.org/10.1016/j.nucengdes.2021.111196.

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16

Pivano, Adrien, Pascal Piluso, Nourdine Chikhi, Jules Delacroix, Pascal Fouquart, and Romain Le Tellier. "Experiments on interactions of molten steel with suboxidized corium crust for in-vessel melt retention." Nuclear Engineering and Design 355 (December 2019): 110271. http://dx.doi.org/10.1016/j.nucengdes.2019.110271.

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17

Tusheva, P., E. Altstadt, H. G. Willschütz, E. Fridman, and F. P. Weiß. "Investigations on in-vessel melt retention by external cooling for a generic VVER-1000 reactor." Annals of Nuclear Energy 75 (January 2015): 249–60. http://dx.doi.org/10.1016/j.anucene.2014.07.044.

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18

Lo Frano, Rosa, Riccardo Ciolini, and Alessio Pesetti. "Analysis of feasibility of a new core catcher for the in-vessel core melt retention strategy." Progress in Nuclear Energy 123 (May 2020): 103321. http://dx.doi.org/10.1016/j.pnucene.2020.103321.

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19

Park, R. J., S. B. Kim, K. Y. Suh, J. L. Rempe, and F. B. Cheung. "Detailed Analysis of Late-Phase Core-Melt Progression for the Evaluation of In-Vessel Corium Retention." Nuclear Technology 156, no. 3 (2006): 270–81. http://dx.doi.org/10.13182/nt06-a3790.

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20

Knudson, D. L., J. L. Rempe, K. G. Condie, K. Y. Suh, F. B. Cheung, and S. B. Kim. "ICONE11-36542 LATE-PHASE MELT CONDITIONS AFFECTING THE POTENTIAL FOR IN-VESSEL RETENTION IN HIGH POWER REACTORS." Proceedings of the International Conference on Nuclear Engineering (ICONE) 2003 (2003): 321. http://dx.doi.org/10.1299/jsmeicone.2003.321.

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21

Skakov, Mazhyn K., Nurzhan Ye Mukhamedov, Alexander D. Vurim, and Ilya I. Deryavko. "Temperature Dependence of Thermophysical Properties of Full-Scale Corium of Fast Energy Reactor." Science and Technology of Nuclear Installations 2017 (2017): 1–7. http://dx.doi.org/10.1155/2017/8294653.

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For the first time the paper determines thermophysical properties (specific heat capacity, thermal diffusivity, and heat conductivity) of the full-scale corium of the fast energy nuclear reactor within the temperature range from ~30°С to ~400°С. Obtained data are to be used in temperature fields calculations during modeling the processes of corium melt retention inside of the fast reactor vessel.

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22

Yu, Peng, and Weimin Ma. "Development of a lumped-parameter code for efficient assessment of in-vessel melt retention strategy of LWRs." Progress in Nuclear Energy 139 (September 2021): 103874. http://dx.doi.org/10.1016/j.pnucene.2021.103874.

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23

Gencheva, R., A. Stefanova, and P. Groudev. "Plant application of ICARE/ASTECv2.0r3 computer code for investigation of in-vessel melt retention in VVER-1000 reactor design." Annals of Nuclear Energy 81 (July 2015): 207–12. http://dx.doi.org/10.1016/j.anucene.2015.02.039.

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24

Nakata, Alexandre Ezzidi, Masanori Naitoh, and Chris Allison. "NEED OF A NEXT GENERATION SEVERE ACCIDENT CODE." JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA 21, no. 3 (2019): 119. http://dx.doi.org/10.17146/tdm.2019.21.3.5630.

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Two international severe accident benchmark problems have been performed recently by using several existing parametric severe accident codes: The Benchmark Study of the Accident at the Fukushima Daiichi Nuclear Power Plant (BSAF) and the Benchmark of the In-Vessel Melt Retention (IVMR) Analysis of a VVER-1000 Nuclear Power Plant (NPP). The BSAF project was organized by the Nuclear Power Engineering Center (NUPEC) of the Institute of Applied Energy (IAE) in Japan for the three Boiling Water Reactors (BWRs) of the Fukushima NPP. The IVMR Project was organized by the Joint Research Center (JRC) o

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25

Agrawal, Animesh, Bae Hoon Lee, Scott A. Irvine, et al. "Smooth Muscle Cell Alignment and Phenotype Control by Melt Spun Polycaprolactone Fibers for Seeding of Tissue Engineered Blood Vessels." International Journal of Biomaterials 2015 (2015): 1–8. http://dx.doi.org/10.1155/2015/434876.

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A method has been developed to induce and retain a contractile phenotype for vascular smooth muscle cells, as the first step towards the development of a biomimetic blood vessel construct with minimal compliance mismatch. Melt spun PCL fibers were deposited on a mandrel to form aligned fibers of 10 μm in diameter. The fibers were bonded into aligned arrangement through dip coating in chitosan solution. This formed a surface of parallel grooves, 10 μm deep by 10 μm across, presenting a surface layer of chitosan to promote cell surface interactions. The aligned fiber surface was used to culture

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26

Balashevska, Yu, D. Gumenyuk, Iu Ovdiienko, et al. "Strengthening the SSTC NRS Scientific and Technical Potential through Participation in the IAEA Coordinated Research Projects." Nuclear and Radiation Safety, no. 1(89) (March 19, 2021): 5–13. http://dx.doi.org/10.32918/nrs.2021.1(89).01.

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The State Scientific and Technical Center for Nuclear and Radiation Safety (SSTC NRS), a Ukrainian enterprise with a 29-year experience in the area of scientific and technical support to the national nuclear regulator (SNRIU), has been actively involved in international research activities. Participation in the IAEA coordinated research activities is among the SSTC NRS priorities. In the period of 2018–2020, the IAEA accepted four SSTC NRS proposals for participation in respective Coordinated Research Projects (CRPs). These CRPs address scientific and technical issues in different areas such a

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27

Mao, Jianfeng, Yunkai Liu, Shiyi Bao, Lijia Luo, Zhiming Lu, and Zengliang Gao. "Structural integrity investigation for RPV with various cooling water levels under pressurized melting pool." Mechanical Sciences 9, no. 1 (2018): 147–60. http://dx.doi.org/10.5194/ms-9-147-2018.

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Abstract. The strategy denoted as “in-vessel retention (IVR)” is widely used in severe accident (SA) management by most advanced nuclear power plants. The essence of IVR mitigation is to provide long-term external water cooling in maintaining the reactor pressure vessel (RPV) integrity. Actually, the traditional IVR concept assumed that RPV was fully submerged into the water flooding, and the melting pool was depressurized during the SA. The above assumptions weren't seriously challenged until the occurrence of Fukushima accident on 2011, suggesting the structural behavior had not been appropr

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28

Xie, Zhi Gang, Yan Ming He, Jian Guo Yang, and Zeng Liang Gao. "Microstructural Evolution of Nuclear Power Steel A508-III in the Creep Process at 800°C." Applied Mechanics and Materials 853 (September 2016): 153–57. http://dx.doi.org/10.4028/www.scientific.net/amm.853.153.

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The A508-III steel is widely used to manufacture the lower heads of commercial reactor pressure vessels (RPV). In severe accident, the reactor core in the RPV begins to melt and meanwhile the technology of in-vessel retention (IVR) exerts its role. In this case the inner surface of RPV will expose to temperatures over a phase transition temperature. However, the significant nonlinear feature of creep curve of A508-III steel suffered heterogeneous damage was not studied. In this work, the creep tests were performed for the steel at the phase transition temperature of 800°C. The microstructural

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29

Zhan, Dekui, Xinhai Zhao, Shaoxiong Xia, Peng Chen, and Huandong Chen. "Numerical Simulation and Validation for Early Core Degradation Phase under Severe Accidents." Science and Technology of Nuclear Installations 2020 (August 3, 2020): 1–12. http://dx.doi.org/10.1155/2020/6798738.

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Early core degradation determines the amount of hydrogen generated by cladding oxidation as well as the temperature, the mass, and the composition of corium that further relocates into the lower head of reactor pressure vessel (RPV), which is essential for the effectiveness analysis of in-vessel retention (IVR) and hydrogen recombiners. In this paper, the mechanisms of controlling phenomena in the early phase of core degradation are analysed at first. Then, numerical models adopted to calculate (1) core heating up, (2) cladding oxidation, (3) dissolution between molten zirconium and fuel pelle

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30

Kim, Su-Hyeon, and Bum-Jin Chung. "Heat load imposed on reactor vessels during in-vessel retention of core melts." Nuclear Engineering and Design 308 (November 2016): 1–8. http://dx.doi.org/10.1016/j.nucengdes.2016.08.010.

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31

Kim, Su-Hyeon, Hae-Kyun Park, and Bum-Jin Chung. "Two- and three-dimensional experiments for oxide pool in in-vessel retention of core melts." Nuclear Engineering and Technology 49, no. 7 (2017): 1405–13. http://dx.doi.org/10.1016/j.net.2017.05.008.

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32

Armstrong, Cheryl M., Andrew G. Gehring, George C. Paoli, Chin-Yi Chen, Yiping He, and Joseph A. Capobianco. "Impacts of Clarification Techniques on Sample Constituents and Pathogen Retention." Foods 8, no. 12 (2019): 636. http://dx.doi.org/10.3390/foods8120636.

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Determination of the microbial content in foods is important, not only for safe consumption, but also for food quality, value, and yield. A variety of molecular techniques are currently available for both identification and quantification of microbial content within samples; however, their success is often contingent upon proper sample preparation when the subject of investigation is a complex mixture of components such as foods. Because of the importance of sample preparation, the present study employs a systematic approach to compare the effects of four different separation techniques (glass

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33

Ferguson, Tracy, CAPT Anthony Lloyd, and Jon Turban. "Enhancing Preparedness and Response ≈ Transition Management Architecture Improvements." International Oil Spill Conference Proceedings 2017, no. 1 (2017): 2017100. http://dx.doi.org/10.7901/2169-3358-2017.1.000100.

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Experts continue to debate about the range of threats that could realistically occur in America today. Disagreements range through the prevention, preemption, and response strategies with advocates continuing to argue for robust “whole-of-government” capabilities to muster and effective response. The debate is complicated by the increased societal churn driven by the changing popular culture, intense effects of technology change and impacts from social media and the 24 hour news cycle. Whether you can hear it, or see it, or not, the truth remains regarding an underlying latency of increased ri

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34

Herd, Oliver, Maria Abril Arredondo Garcia, James Hewitson, et al. "An Adapting Bone Marrow Niche Creates a Nurturing Environment for Hematopoiesis during Immune Thrombocytopenia Progression." Blood 134, Supplement_1 (2019): 222. http://dx.doi.org/10.1182/blood-2019-129795.

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Immune thrombocytopenia (ITP) is an acquired autoimmune disease characterised by low platelet counts (&lt;100 x 109/L) and manifests as a bleeding tendency. The demand on hematopoiesis is elevated in chronic ITP, where sustained platelet destruction mediated by an activated immune system is likely to cause considerable stress on progenitor populations. Intriguingly, this increased stress does not appear to result in functional exhaustion, as chronic ITP patients do not present with pancytopenia. By using a novel murine model of chronic ITP, generated by injecting mice with anti-CD41 antibo

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35

Andriolo, Lena, Clément Meriot, and Nikolai Bakouta. "Preliminary Investigations of the Feasibility of In-Vessel Melt Retention Strategies for a Small Modular Reactor Concept." Journal of Nuclear Engineering and Radiation Science 5, no. 2 (2019). http://dx.doi.org/10.1115/1.4042360.

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The study presented in this paper is part of the technological surveillance performed at the Electricité De France (EDF) Research and Development (R&D) Center, in the Pericles department, and investigates the feasibility of modeling in-vessel melt retention (IVMR) phenomena for small modular reactors (SMR) with the modular accident analysis program version 5 in its EDF proprietary version (MAAP5_EDF), applying conservative hypotheses, such as constant decay heat after corium relocation to the lower head. The study takes advantage of a corium stratification model in the lower head of the ve

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36

Verma, P. K., P. P. Kulkarni, P. Pandey, S. V. Prasad, and A. K. Nayak. "Critical Heat Flux on Curved Calandria Vessel of Indian PHWRs During Severe Accident Condition." Journal of Heat Transfer 143, no. 2 (2020). http://dx.doi.org/10.1115/1.4048823.

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Abstract In pressurized heavy water reactors (PHWRs), during an unmitigated severe accident, the absence of adequate cooling arising from multiple failures of the cooling system leads to the collapse of pressure tubes and calandria tubes, which may ultimately relocate to the lower portion of the calandria vessel (CV) forming a debris bed. Due to the continuous generation of decay heat in the debris, it will melt and form a molten pool at the bottom of the CV. The CV is surrounded by calandria vault water, which acts as a heat sink at this scenario. In-vessel corium retention (IVR) through the

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37

Mao, Jianfeng, Shiyi Bao, Zhiming Lu, Lijia Luo, and Zengliang Gao. "The Influence of Crust Layer on Reactor Pressure Vessel Failure Under Pressurized Core Meltdown Accident." Journal of Nuclear Engineering and Radiation Science 4, no. 4 (2018). http://dx.doi.org/10.1115/1.4040494.

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The so-called in-vessel retention (IVR) was considered as a severe accident management strategy and had been certified by Nuclear Regulatory Commission (NRC) in U.S. as a standard measure for severe accident management since 1996. In the core meltdown accident, the reactor pressure vessel (RPV) integrity should be ensured during the prescribed time of 72 h. However, in traditional concept of IVR, several factors that affect the RPV failure were not considered in the structural safety assessment, including the effect of corium crust on the RPV failure. Actually, the crust strength is of signifi

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38

Madokoro, Hiroshi, Alexei Miassoedov, and Thomas Schulenberg. "Coupling of a Reactor Analysis Code and a Lower Head Thermal Analysis Solver." Journal of Nuclear Engineering and Radiation Science 5, no. 1 (2019). http://dx.doi.org/10.1115/1.4041278.

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Due to the recent high interest on in-vessel melt retention (IVR), development of detailed thermal and structural analysis tool, which can be used in a core-melt severe accident, is inevitable. Although RELAP/SCDAPSIM is a reactor analysis code, originally developed for U.S. NRC, which is still widely used for severe accident analysis, the modeling of the lower head is rather simple, considering only a homogeneous pool. PECM/S, a thermal structural analysis solver for the reactor pressure vessel (RPV) lower head, has a capability of predicting molten pool heat transfer as well as detailed mech

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39

Bachrata, Andrea, Fréderic Bertrand, Nathalie Marie, and Fréderic Serre. "A Comparative Study on Severe Accident Phenomena Related to Melt Progression in Sodium Fast Reactors and Pressurized Water Reactors." Journal of Nuclear Engineering and Radiation Science 7, no. 3 (2020). http://dx.doi.org/10.1115/1.4047921.

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Abstract The nuclear safety approach has to cover accident sequences involving core degradation in order to develop reliable mitigation strategies for both existing and future reactors. In particular, the long-term stabilization of the degraded core materials and their coolability has to be ensured after a severe accident. This paper focuses on severe accident phenomena in pressurized water reactors (PWR) compared to those potentially occurring in future GenIV-type sodium fast reactors (SFR). First, the two considered reactor concepts are introduced by focusing on safety aspects. The severe ac

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40

Zhu, Jianwei, Jianfeng Mao, Shiyi Bao, Lijia Luo, and Zengliang Gao. "Comparative Study on Reactor Pressure Vessel Failure Behaviors With Various Geometric Discontinuities Under Severe Accident." Journal of Pressure Vessel Technology 139, no. 2 (2017). http://dx.doi.org/10.1115/1.4035697.

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The so-called “in-vessel retention (IVR)” is a basic strategy for severe accident (SA) mitigation of some advanced nuclear power plants (NPPs). The IVR strategy is to keep the reactor pressure vessel (RPV) intact under SA like core meltdown condition. During the IVR, the core melt (∼1327 °C) is collected in the lower head (LH) of the RPV, while the external surface of RPV is submerged in the water. Through external cooling of the RPV, the structural integrity is assumed to be maintained within a prescribed period of time. The maximum thermal loading is referred to critical heat flux (CHF) on t

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Pandey, Pradeep, Parimal P. Kulkarni, Arun Nayak, and Sumit V. Prasad. "Evaluation of Dump Tank Coolability in PHWRs During Late-Phase Severe Accident." Journal of Nuclear Engineering and Radiation Science 5, no. 4 (2019). http://dx.doi.org/10.1115/1.4043108.

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In some of the older design of pressurized heavy water reactors (PHWRs), such as in Rajasthan Atomic Power Station (RAPS) and Madras Atomic Power Station (MAPS), in case of a severe accident, the debris/corium may cause failure of the dump port of calandria and relocate into the dump tank. The sensible and decay heat of debris/corium makes the heavy water in dump tank to boil off leaving the dry debris in dump tank. The dry debris remelt with time and the molten corium, thus, formed has the potential to breach the dump tank and move into the containment cavity, which is highly undesirable. Hen

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42

Mao, Jianfeng, Jianwei Zhu, Shiyi Bao, Lijia Luo, and Zengliang Gao. "Investigation on Structural Behaviors of Reactor Pressure Vessel With the Effects of Critical Heat Flux and Internal Pressure." Journal of Pressure Vessel Technology 139, no. 2 (2016). http://dx.doi.org/10.1115/1.4034582.

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The so-called “in-vessel retention (IVR)” is a severe accident management strategy, which is widely adopted in most advanced nuclear power plants. The IVR mitigation is assumed to be able to arrest the degraded melting core and maintain the structural integrity of reactor pressure vessel (RPV) within a prescribed hour. Essentially, the most dangerous thermal–mechanical loads can be specified as the combination of critical heat flux (CHF) and internal pressure. The CHF is the coolability limits of RPV submerged in water (∼150 °C) and heated internally (∼1327 °C), it results in a sudden transiti

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Prasad, Sumit V., P. P. Kulkarni, D. C. Yadav, P. K. Verma, and A. K. Nayak. "In-Vessel Retention of PHWRs: Experiments at Prototypic Temperatures." Journal of Nuclear Engineering and Radiation Science 6, no. 1 (2019). http://dx.doi.org/10.1115/1.4043999.

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Abstract In pressurized heavy water reactors (PHWRs), multiple failures of engineered safety features may cause a failure of core cooling eventually leading to core collapse. The failed fuel and fuel channels relocate to the bottom of the calandria vessel (CV) and form a terminal debris bed, which generates decay heat. With time, the moderator evaporates and the terminal debris bed ultimately melts and forms a molten pool of corium. If corium breaches the CV and enters the calandria vault, large amounts of hydrogen and other fission gases may be generated due to molten core concrete interactio

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