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Journal articles on the topic 'In-vessel melt retention'

Author: Grafiati

Published: 24 June 2021

Last updated: 1 February 2022

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1

Almyashev, V. I., V. S. Granovsky, V. B. Khabensky, E. V. Krushinov, A. A. Sulatsky, S. A. Vitol, V. V. Gusarov, 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 (September 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 (June 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 (November 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 (April 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 heat-up as well as corium relocation to the lower head is presented. The ex-vessel cooling was modelled in order to evaluate the applicability of RELAP/SCDAPSIM code for predicting the heat fluxes and reactor pressure vessel wall temperatures. The results of different ex-vessel heat transfer modes were compared and it was concluded that the implemented heat transfer correlations of COUPLE module in RELAP/SCDAPSIM should be applied for IVR analysis. To investigate the influence of debris separation into oxidic and metallic layers in the molten pool on the heat transfer through the wall of the lower head the analytical study was conducted. The results of this study showed that the focusing effect is significant and under some extreme conditions local heat flux from reactor vessel could exceed the critical heat flux. It was recommended that the existing RELAP/SCDAPSIM models of the processes in the debris should be updated in order to consider more complex phenomena and at least oxide and metal phase separation, allowing evaluating local distribution of the heat fluxes.

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10

Granovsky, V. S., V. B. Khabensky, E. V. Krushinov, S. A. Vitol, A. A. Sulatsky, V. I. Almjashev, S. V. Bechta, 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 (August 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 (March 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 (May 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 (December 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 (November 12, 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) of the European Commission (EC) in Holland (Europe) for a Pressurized Water Reactor (PWR). The obtained results of both projects have shown very large discrepancies between the used severe accident codes for both reactor types BWR and PWR. Consequently, the results for a real plant analysis by these integral codes, may not be correct after the beginning of core melt. Discrepancies of results of ex-vessel phenomena in the containment between the codes are in general larger. Therefore, there is a strong need for a reliable new generation mechanistic severe accident code which can simulate severe accident scenarios from an initiating event till containment failure with better accuracy not only for existing light water reactors but also for new generation IV reactor types. SAMPSON mechanistic ex-vessel modules coupled with SCDAPSIM and a new thermal-hydraulic module ASYST-ISA with particularly newly developed options for the reactor coolant system (RCS) and material properties applicable to new reactor deigns, is proposed as a best etimate new generation severe accident code for several reasons which are described in this paper.Keywords: Severe accident, SAMPSON, SCDAPSIM, ASYST-ISA, Steam explosion, Hydrogen detonation

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25

Agrawal, Animesh, Bae Hoon Lee, Scott A. Irvine, Jia An, Ramya Bhuthalingam, Vaishali Singh, Kok Yao Low, Chee Kai Chua, and Subbu S. Venkatraman. "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 cells present in the vascular wall, in particular fibroblasts and smooth muscle cells. This topography induced “surface guidance” over the orientation of the cells, which adopted an elongated spindle-like morphology, whereas cells on the unpatterned control surface did not show such orientation, assuming more rhomboid shapes. The preservation of VSMC contractile phenotype on the aligned scaffold was demonstrated by the retention ofα-SMA expression after several days of culture. The effect was assessed on a prototype vascular graft prosthesis fabricated from polylactide caprolactone; VSMCs aligned longitudinally along a fiberless tube, whereas, for the aligned fiber coated tubes, the VSMCs aligned in the required circumferential orientation.

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26

Balashevska, Yu, D. Gumenyuk, Iu Ovdiienko, O. Pecherytsia, I. Shevchenko, Yu Yesypenko, and O. Zhabin. "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 as: 1) performance of probabilistic safety assessment for multi-unit/multi-reactor sites; 2) use of dose projection tools to ensure preparedness and response to nuclear and radiological emergencies; 3) phenomena related to in-vessel melt retention; 4) spent fuel characterization. This article presents a brief overview of the abovementioned projects with definition of scientific contributions by the SSTC NRS (participation in benchmarks, development of methodological documents on implementing research stages and of IAEA technical documents (TECDOC) for demonstration of best practices and results of research carried out by international teams).

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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 (March 2, 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 appropriately assessed. Therefore, the paper tries to address the structure-related issue on determining whether RPV safety can be maintained or not with the effect of various water levels and internal pressures created from core meltdown accident. In achieving it, the RPV structural behaviors are numerically investigated in terms of several field parameters, such as temperature, deformation, stress, plastic strain, creep strain, and total damage. Due to the presence of high temperature melt on the inside and water cooling on the outside, the RPV failure is governed by the failure mechanisms of creep, thermal-plasticity and plasticity. The creep and plastic damages are interacted with each other, which further accelerate the failure process. Through detailed investigation, it is found that the internal pressure as well as water levels plays an important role in determining the RPV failure time, mode and site.

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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 evolution at different creep stages was characterized by scanning electron microscopy and transmission electron microscopy. The results show that, at the second creep stage, more coarsening second phase particles occur in the steel. With the creep processing, the grain size and diameter of second phase particles increase. At the tertiary creep stage, the grain size increases significantly, and the second phase particles coarsen during the process of atom migration. In addition, Micro-cracks and voids also come into being in the situation and they can become larger by combing each other during the creep process. At this stage, the growth of cavities and second phase particles coarsening become the main mechanism of creep damage. The trend of microstructural evolution is consistent with the creep constitutive equation obtained for the A508-III steel at the phase transition temperature of 800°C. The results obtained provide indispensable foundation to establish the relationship between the macroscopic creep and microscopic damage.

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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 pellets, and (4) formation of a molten pool in the core active section are presented. Compared with integral codes for severe accident analysis (such as MAAP and MELCOR), the models in this paper are established at the fuel pin level and the calculation is performed in 3D, which can capture the detail local phenomena during the core degradation and eliminate the average effect due to equivalent rings used in integral codes. In addition, most of the control equations in this paper are calculated by implicit schemes, which can improve the accuracy and stability of the calculation. In the simulation, the calculation oxidation is calculated by using the oxygen diffusion model, while the dissolution is calculated with Kim, Hayward, Hofmann, and IBRAE models to perform uncertainty analysis. For the validation, the cladding oxidation model is verified by Olander theoretical cases in the conditions of both steam-rich and steam-starved. The dissolution models are validated by the RIAR experiment. The code is overall verified by Phebus FPT0 on the integral phase of core early degradation. According to the simulation results, it can be inferred that the dissolution reaction between the molten zirconium and fuel pellets is the main reason for the melting of UO2 at low temperature. In the case of starved steam, part of the fuel pellets can melt down even at 2248 K and relocate to the bottom of the core, which is much lower than the melting point of UO2 (3113 K).

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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 (October 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 (December 3, 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 wool, 50 μm polypropylene filters, graphite felt, and continuous flow centrifugation (CFC)) on sample preparation. To define the physical effects associated with the use of these separation methods, a multifactorial analysis was performed where particle size and composition, both pre- and post- processing, were analyzed for four different food matrices including lean ground beef, ground pork, ground turkey and spinach. Retention of three important foodborne bacterial pathogens (Escherichia coli O157:H7, Salmonella enterica, and Listeria monocytogenes) was also examined to evaluate the feasibility of the aforementioned methods to be utilized within the context of foodborne pathogen detection. Data from the multifactorial analysis not only delineated the particle size ranges but also defined the unique compositional profiles and quantified the bacterial retention. The three filtration membranes allowed for the passage of bacteria with minimal loss while CFC concentrated the inoculated bacteria. In addition, the deposition and therefore concentration of food matrix observed with CFC was considerably higher for meat samples relative to spinach. However, filtration with glass wool prior to CFC helped clarify meat samples, which led to considerably lower amounts of solids in the CFC vessel post processing and an increase in the recovery of the bacteria. Overall, by laying a framework for the deductive selection of sample preparation techniques, the results of the study can be applied to a range of applications where it would be beneficial to scientifically guide the pairing of the criteria associated with a downstream detection method with the most advantageous sample preparation techniques for complex matrices such as foods.

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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 (May 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 risk in our society. Further compounding this is the change in the oil economy. Latent risk has risen there as well, challenging current preparedness efforts. Increased flexibility, transitional success, better data sharing methods, and deeper situational awareness is needed for planning, preparedness, and response success. Coast Guard legal authorities are foundational in this regard especially as it relates to the proper apportionment of National Contingency Plan resources. The Coast Guard Vessel Response and Facility Response Plan regulations reflect an appropriate effort to assure the retention and allocation of those resources to meet preparedness and response requirements. How can we be sure, however, that this “force lay down” is effective? Can those resources be better accessed to support NCP requirements? This poster will depict a way to envision better transition of VRP/FRP resources. It will also explain a capability and architecture developed to ease the rapid shifts from day-to-day operations to a rapidly expanding crisis.

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34

Herd, Oliver, Maria Abril Arredondo Garcia, James Hewitson, Karen Hogg, Saleni Pravin Kumar, Katiuska Pulgar Prieto, Andrew Stone, Paul Genever, and Ian Hitchcock. "An Adapting Bone Marrow Niche Creates a Nurturing Environment for Hematopoiesis during Immune Thrombocytopenia Progression." Blood 134, Supplement_1 (November 13, 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 (<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 antibody (ITP group) or IgG (control group) every 48hrs for 4 weeks, we aimed to define the effect of chronic ITP on hematopoietic progenitors and to elucidate the mechanisms behind the preservation of hematopoiesis. The relative numbers of hematopoietic progenitors in mice with chronic ITP vs controls were analysed by flow cytometry and their fitness was assessed by measuring their relative ability to reconstitute the hematopoietic system of lethally irradiated recipients. There was a significant increase in all hematopoietic progenitors analysed in ITP: 2.96-fold increase in multipotent progenitors, 4.66-fold increase in short-term hematopoietic stem cells (ST-HSCs) and 4.93-fold increase in long-term hematopoietic stem cells (LT-HSCs), which led to an increased ability of ITP donor bone marrow to reconstitute irradiated recipients. The results indicate that chronic ITP drives LT-HSCs out of quiescence and causes increased differentiation into committed progenitors in order to meet the increased demand in platelet production. In support of this, increased megakaryopoiesis was observed in chronic ITP, with a 60.5% increase in the number of megakaryocytes observed in bone marrow sections. Interestingly, similar to the clinical manifestation of ITP, we observed no change in levels of circulating TPO in our ITP model. Next, the effect of chronic ITP on the bone marrow microenvironment was determined due to its essential role in the support and maintenance of hematopoiesis. Histological analysis of bone marrow from mice with chronic ITP (Figure 1) revealed a 66.7% increase in the numbers of LepR+/ Cxcl12-DsRed stromal cells. LepR+/ Cxcl12-DsRed stromal cells are a well characterised stromal cell subset, known to be essential for maintenance and retention of HSCs in the bone marrow microenvironment. During chronic ITP, this stromal cell subset maintained their classically defined perivascular location and retained their ability to produce high levels of hematosupportive cytokines (Cxcl12 and Kitl). Chronic ITP was associated with a significant increase in total bone marrow expression (Cxcl12=2.39-fold increase; Kitl=1.71-fold increase), pointing to perivascular stromal cell expansion as being the source of increased local hematopoietic support. Analysis of the bone marrow vascular network revealed that the average vessel area was increased in chronic ITP (54.3% increase), whilst the number of vessels remained unchanged implying that the marrow sinusoids are vasodilated. We hypothesise that an increase in blood vessel area would aid the extravasation of circulating HSCs back into the bone marrow microenvironment where they would contribute to hematopoiesis. By developing an accurate mouse model of chronic ITP, we have identified key alterations in HSCs and the bone marrow microenvironment. Our data clearly demonstrates that in chronic ITP, HSCs are driven out of quiescence and expand in number in order to contribute to the increased demand for hematopoiesis. Furthermore, the bone marrow microenvironment adapts to this increased differentiation pressure on HSCs by creating a hematosupportive, quiescence promoting environment through the expansion of bone marrow stromal cells, and an increase in blood vessel area. Disclosures No relevant conflicts of interest to declare.

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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 (March 15, 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 vessel, developed by EDF R&D for large-sized prospective pressurized water reactors (PWRs). The analysis is based on a stepwise approach in order to evaluate the impact of various effects during IVMR conditions. First, an analytical calculation is performed in order to establish a reference case to which the MAAP5_EDF code results are compared. In a second step, the impact of the lower head geometry, vessel steel ablation, and subsequent relocation on the heat flux has been analyzed for cases where heat dissipation through radiation is neglected (in first approximation). Finally, the impact of heat losses through radiation as well as the crust formation around the pool has been assessed. The results demonstrate the applicability of the MAAP5_EDF code to SMRs, with heat fluxes lower than 1.1 MW/m2 for relevant cases, and identify modeling improvements.

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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 (November 16, 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 external reactor vessel cooling (ERVC) is conceived as an effective method for maintaining the integrity of a calandria vessel during a severe accident in a nuclear power plant. Under the IVR conditions, it is necessary to ensure that the imparted heat flux due to melt is less than the critical heat flux (CHF) at the bottom of the calandria vessel wall. To evaluate the thermal margin for IVR, experiments are performed in a prototypic curved section of calandria vessel (25o sector) of calandria vessel to determine the CHF, heat transfer coefficient, and its variation along with the curvature of calandria vessel. The effect of moderator drainpipe on CHF and the heat transfer coefficient has also been evaluated. It has been observed that the imparted heat flux is much less than the CHF at the bottom of the calandria vessel.

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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 (September 10, 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 significant importance in the context of a severe reactor accident in which molten core material melts through the reactor vessel and collects on the lower head (LH) of the RPV. Consequently, the RPV integrity is significantly influenced by the crust. A strong, coherent crust anchored to the RPV walls could allow the yet-molten corium to fall away from the crust as it erodes the RPV, therefore thermally decoupling the melt pool from the coolant and sharply reducing the cooling rate. Due to the thermal resistance of the crust layer, it somewhat prevents further attack of melt pool from the RPV. In the present study, the effect of crust on RPV structural behaviors was examined under multilayered crust formation conditions with consideration of detailed thermal characteristics, such as high-temperature gradient across the wall thickness. Thereafter, systematic finite element analyses and subsequent damage evaluation with varying parameters were performed on a representative RPV to figure out the possibility of high temperature induced failures with the effect of crust layer.

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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 (January 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 mechanical behavior including creep, plasticity, and material damage. The boundary condition, however, needs to be given manually and thus the application of the stand-alone PECM/S to reactor analyses is limited. By coupling these codes, the strength of both codes can be fully utilized. Coupled analysis is realized through a message passing interface, OpenMPI. The validation simulations have been performed using LIVE test series and the calculation results are compared not only with the measured values but also with the results of stand-alone RELAP/SCDAPSIM simulations.

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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 (November 11, 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 accident scenarios leading to core melting are presented and the initiating events are highlighted. This paper focuses on in-vessel severe accident phenomena, including the chronology of core damage, major changes in the core configuration and molten core progression. Regarding the mitigation means, the in-vessel retention phenomena and the core catcher characteristics are reviewed for these different nuclear generation concepts (II, III, and IV). A comparison between the PWR and SFR severe accident evolution is provided as well as the relation between governing physical parameters and the adopted mitigation provisions for each reactor concept. Finally, it is highlighted how the robustness of the safety demonstration is established by means of a combined probabilistic and deterministic approach.

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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 (February 3, 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 the inside, while the external surface is considered to perform in the environment of the boiling crisis point (∼130 °C). Due to the high temperature gradients, the failure mechanisms of the RPV is found to span a wide range of structural behaviors across the wall thickness, such as melt-through, creep damage, plastic yielding as well as thermal expansion. Besides CHF, the pressurized core meltdown was another evident threat to the RPV integrity, as indicated in the Fukushima accident on 2011. In illustrating the effects of internal pressures and individual CHF on the failure behaviors, three typical RPVs with geometric discontinuity caused by local material melting were adopted for the comparative study. Through finite-element method (FEM), the RPV structural behaviors were investigated in terms of deformation, stress, plastic strain, creep, and damage. Finally, some important conclusions are summarized in the concluding remark. Such comparative study provides insight and better understanding for the RPV safety margin under the IVR condition.

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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 (July 19, 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. Hence, as an accident management strategy, water is being flooded outside the dump tank using fire water hook-up lines to remove the heat from corium to cool and stabilize it and terminate the accident progression, similar to in vessel retention. However, the question is “is the molten corium coolable by this technique.” The coolability of the molten corium in dump tank as in the reactor is assessed by conducting experiments in a scaled facility using a simulant material having comparable thermophysical properties with that of corium. Melting of dry debris resting on dump tank bottom marks the beginning of the experimental investigation for present analysis. Decay heat is simulated by a set of immersed heaters inside the melt. Temperature profiles at different locations in dump tank and in the melt pool are obtained as a function of time to demonstrate the coolability with decay heat. Large temperature gradient is observed within the corium, involving high melt center temperature and low tank wall temperature suggesting formation of crust which insulates the dump tank wall from hot corium.

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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 (September 28, 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 transition of boiling crisis from nucleate to film boiling. Accordingly, from a structural integrity perspective, the RPV failure mechanisms span a wide range of structural behaviors, such as melt-through, creep damage, plastic deformation as well as thermal expansion. Furthermore, the geometric discontinuity of RPV created by the local material melting on the inside aggravates the stress concentration. In addition, the internal pressure effect that usually neglected in the traditional concept of IVR is found to be having a significant impact on the total damage evolution, as indicated in the Fukushima accident that a certain pressure (up to 8.0 MPa) still existed inside the RPV. This paper investigates structural behaviors of RPV with the effects of CHF and internal pressure. In achieving this goal, a continuum damage mechanics (CDM) based on the “ductility exhaustion” is adopted for the in-depth analysis.

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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 (November 29, 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 interaction, which may pressurize the containment leading to containment failure. In addition, the passive catalytic recombiner devices may be incapable of managing such large amounts of hydrogen. Hence, in-vessel retention of corium is the only option to the avert progression of the accident. The heat removal capability of the CV needs to be demonstrated in order to attain the goal of in-vessel retention, to contain the corium during severe accidents. A lot of numerical analysis of heat removal capability of the CV has been done. However, experimental demonstration of in-vessel retention has been rarely presented in the literature, especially for PHWRs. In this paper, in-vessel retention at prototypic temperatures has been presented. Experiments have been carried out in scaled CVs. Different corium simulants have been used at elevated temperatures and corium coolability has been demonstrated.

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