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Some Experimental Results on Safety Aspects of Modular HTR

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Modular High-temperature Gas-cooled Reactor Power Plant

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Abstract

In this chapter, an overview on some important experiments regarding the safety of modular HTR is given. A major topic is the proof tests on the concept of self-acting decay heat removal. Relevant results for the transport of the heat in the pebble bed, in the reactor structure, and from the surface of the reactor pressure vessel are explained and show that today the different steps are well known and show good overestimation with calculations and experimental data gained from integral AVR experiments. In case of extreme assumptions on accidents, the decay heat is stored in concrete structures and therefore large concrete blocks have been tested at very high temperature with good success. The reactivity behavior of HTR core has been tested in practical operation and in accompanying tests on special facilities. The strong negative temperature coefficient has been verified by these, and especially the experiment in AVR (simulation of total loss of active shutdown system) and the ATWS experiment in HTR-10 have demonstrated this important characteristic of modular HTR. Many experiments regarding the behavior of water in the primary system and in the core of HTR plants have been carried out. These were the measurements of corrosion rates of graphite from the important parameters like temperature, pressure, and type of graphite. The transport of water, forming of steam and gases, and changes of materials have been measured in detail. Similar activities have been carried out for the accidents with ingress of air. Solutions to exclude the occurrence and consequence of large air ingress have been developed and qualified too, for example burst-safe vessels, which would allow just small openings in case of breaks. An important activity was the possible transport of air in primary systems too and thereby to qualify computer programs. A major experimental work of HTR development was directed toward the behavior of fission products in the reactor. The intention was to define source terms for the normal operation and for accidents. A central importance has the VAMPYR experiments in AVR, which gave information on the contamination of the circuit by solid fission products. Irradiated fuel elements have been heated up in hot cells (KŰFA tests), and these experiments delivered release rates for all relevant fission products dependent from burnup, operation temperatures, heating temperature, and time. The requirement to limit the maximal temperature of modular HTR with LEU TRISO fuel to a value of 1600 °C in all accidents is a result of these activities. Further experiments were done to study the transport, deposition, and remobilization of fission products from different reactor components. A broad basis of data today is available, to describe these processes. The earthquake behavior of reactor systems was a further topic of experimental research. The large test facility SAMSON has been used to test relevant components and especially pebble beds at large acceleration values. It was stated that reactor designs should be carried out with high safety factors for strength of earthquakes, which were assumed until now.

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References

  1. Breitbach G., Heat transport process in pebble bed regarding radiation, Diss. RWTH Aachen, Dec. 1978

    Google Scholar 

  2. Robold K., Heat transport in the inside and on the boundary of boundary of pebble beds, JÜL-1893, 1984

    Google Scholar 

  3. Barthels H., Schürenkremer M., The effective heat conductivity λeff in pebble bed of high-temperature reactor, JÜL-1893, 1984

    Google Scholar 

  4. Zehner P., Schlünder H., Heat conductivity of pebble bed at medium temperature, Technik 42 (1970)

    Google Scholar 

  5. Fricke U., Analysis of possibilities to enlarge the power of inherent safe high-temperature reactors by optimization of the core layout, Diss. University of Duisburg, 1987

    Google Scholar 

  6. Stöcker B., Analysis of the self-acting decay heat removal in case of HTR including natural convection, diss. RWTH Aachen, Dec. 1997

    Google Scholar 

  7. Nieβen H.F., Gerwin H., Scherer W., Numerical simulation of SANA1-expermients with the TINTE code in: Heat transport and afterhead removal for gas cooled reactors under accident conditions, IAEA-TECDOC-1163, Vienna, Jan. 2001

    Google Scholar 

  8. IAEA, Decay heat removal and heat transfer under normal and accident conditions in gas cooled reactors, IAEA-TECDOC-757, Aug. 1994

    Google Scholar 

  9. IAEA, Heat transport and after heat removal for gas cooled reactors under accident conditions, IAEA-TECDOC-1163, Jan. 2001

    Google Scholar 

  10. Hennies H.H., Kessler G., Eibel J., Improved containment concept for future pressurized water reactors, International Workshop on Safety of Nuclear Installations of the Next Generation and Beyond, Chicago, IL., USA, August 1989, IAEA-TECDOC-550, 1990

    Google Scholar 

  11. Scholtyssek W., Alsmeyer H., Erbacher F.J., Decay heat removal after a PWR core meltdown accident, International Conference on Design and Safety of Advanced Nuclear Power Plants (ANP 92), Tokyo, Japan, Oct. 1992, Vol. III

    Google Scholar 

  12. Kuczera B., Erbacher F.J., Scholtyssek W., Investigation on ex-vessel core melt cooling in future PWR-containment, IAEA Technical Meeting on Thermohydraulic of Cooling Systems in Advances Water-Cooled Reactors, Villingen Switzerland, May 1993

    Google Scholar 

  13. Cheng X., Erbacher F.J., Neitzel H.J., Passive containment cooling for next generation water cooled reactors, Proc. ICONE-4 Conf., New Orleans, USA, March 1996

    Google Scholar 

  14. Spilker H., Natural conception cooling of heat producing radioactive waste in transport and storage vessels, Kerntechnik, 54, No. 4, 1989

    Google Scholar 

  15. Cheng X., Development of experimental validated procedures for the design of containment cooling systems with air and natural convection, Wissensch. Beudite FZ Karlsruke, FZKA, 6056, 1998

    Google Scholar 

  16. Schartmann F., Heat release from transport and intermediate storage vessels for highly radioactive substances in normal operation and during accidents, Diss. RWTH Aachen, Oct. 2000

    Google Scholar 

  17. Von Heesen W., et al., Heat transfer from transport cask storage facilities for spent fuel elements, Nuclear Technology, Vol. 62, 1983

    Google Scholar 

  18. Delle W., Nickel H., AVR graphite structures, in: AVR-Experimental High Temperature Reactors—21 Years of Successful Operation for a Future Energy Technology, VDI-Verlag Gmbh, Düsseldorf, 1990

    Google Scholar 

  19. Haag G., Properties of ATR-2E graphite and property changes due to fast neutron irradiation, JÜL-4183, Oct. 2003

    Google Scholar 

  20. Kugeler K., Sappock M., Wolf L., Development on an inactive heat removal system for high temperature reactors, Jahrestagung Kerntechnik, KTG, 1991

    Google Scholar 

  21. Fröhling W. et al., Prestressed cast pressure vessels (VGD) as burst-safe pressure vessel for innovative applications in the nuclear technology, Monographic of Forsdung Sentrum Jülich, Energy Technology, Vol. 14, 2000

    Google Scholar 

  22. Geiβ M. et al., Experimental and analytical work to qualify the pressure vessel (prestressed) and the primary cell of the HTR-Modul INNA project, Batelle rep., April 1993

    Google Scholar 

  23. Kugeler K., Sappok E.P., Wolf L., and Kneer A., Qualification of a passive heat removal system for high temperature reactors, Jahrestagung Kerntechnik, 1993

    Google Scholar 

  24. Wolf L., Knner R., Schulz R., Giannices A., Häfner W., Passive heat removal experiments for an advanced HTR-Module reactor pressure vessel and cavity design, IAEA-TECDOC-757, Vienna, Aug. 1994

    Google Scholar 

  25. Beine B., Large-scale test setup for the passive heat removal system and the prestressed cast iron pressure vessel of a 200 MWth modular high-temperature reactor, 3rd Int. Seminar on Small and Medium-Sized Nuclear Reactors, New Delhi, India, Aug. 1991

    Google Scholar 

  26. Takeda S., et al., Mockup experiment of vessel cooling system, KFA-JAERI Cooperation on Exchange of Information in the Field of Research and Development, JAERI, Oarai, 1996

    Google Scholar 

  27. IAEA-TECDOC-1163, Heat transport and after heat removal for gas cooled reactors under accident conditions results of simulation of the HTTR-RCCS mockup with the THANPA CSTZ code, IAEA, Jan. 2001

    Google Scholar 

  28. Shina Y., Hishida M., Heat transfer in the upper part of the HTTR-pressure vessel, during loss of forced cooling, IAEA-TECDOC-757, Vienna, Aug. 1994

    Google Scholar 

  29. Takeda S., et al., Test apparatus of cooling panel system for MHGR, IAEA, TECDOC-757, Vienna, Aug. 1994

    Google Scholar 

  30. Hishida M., et al., Studies on the primary pipe rupture accident of the high temperature reactor, NURETH: Proc. 4th Int. Topical Meeting on Nuclear Reactor Thermal Hydraulics, Vol. 1, Karsluke, Oct. 1989

    Google Scholar 

  31. Altes J., Escherich K.H., Nickel M., Wolters J., Experimental study of the behavior of the prestressed concrete pressure vessel of the THTR-300 at severe accident temperatures, Trans. 11 SmiRT, Tokyo, 1989, Vol. H

    Google Scholar 

  32. Altes J., Escherich K.H., Nickel M., Wolters J., Behavior of prestressed concrete pressure vessel of the HTR-500 at severe accident temperatures, Trans. 10 SmiRT Arnheim, 1988, Vol. H

    Google Scholar 

  33. Altes J., Escherich K.H., Nickel M., Wolters J., Experimental study of the behavior of prestressed concrete pressure vessels of high temperature reactors at accident temperatures, Trans. 9 SmiRT Lausanne, Volume H., 1987

    Google Scholar 

  34. Kim D., Reflooding of the liner cooling of an HTR with medium power and prestressed concrete reactor pressure vessel in case of a core heat up accident after failure of the liner cooling, JÜL-2543, 1991

    Google Scholar 

  35. Altes J., Schneider U., Schimmepfennig K., Behavior of a PCRV for HTR under high core temperature loading. Trans. 6. SmiRT Paris, Vol. H, Paper 2/5, 1981

    Google Scholar 

  36. Schimmelpfennig K., Altes J., Sepcial aspects on the behavior of PCRV under extremely high core temperature loading, Trans. 7 SmiRT Chicago, Vol. H, 1983

    Google Scholar 

  37. Schneider U, Diederichs U., Physical properties of concrete from 20°C up to melting, Betonwerk und Fertigteil-technik, Part I in Vol. 3, Part II in Vol. 4, 1981

    Google Scholar 

  38. Krüger K., Experimental simulation of a loss of cooling accident with the AVR Reactor, JÜL-2297, Aug. 1989

    Google Scholar 

  39. Iyoku T., Rehm W., Jahn W, Analytical investigation of the AVR loss of coolant accident simulation test LOCA, KFA-ISR, 9/92, Sept. 1991

    Google Scholar 

  40. Petersen K., To the safety concept of the high-temperature reactor with natural heat transport from the core during accidents, JÜL-1872, Oct. 1983

    Google Scholar 

  41. Ivens G., Krüger K., Experiences with the AVR on safety, Conference for Safety of HTR, KFK Jülich, JÜL-Conf 53, June 1985

    Google Scholar 

  42. Kugeler K., et al., Simulation of the loss of coolant accident with the AVR reactor, in: AVR Experimental High-Temperature Reactor, VDI Verlag, Düsseldorf, 1990

    Google Scholar 

  43. Krüger V., Bergerfurth A., Burge S., Simulation of AVR—loss of coolant—accidents, Jahestagung Kerntechnik (Germany), 1989

    Google Scholar 

  44. Wimmers M., Berger Furth A., The physics of AVR reactor, in: AVR—Experimental High Temperature Reactor—21 Years of Successful Operation for a Future Energy Technology, VDI-Verlag Gmbh, Düsseldorf, 1990

    Google Scholar 

  45. Pohl W., Wimmiens M., Schmidt H., Jnug D., Determination of reactivity in the AVR-core, Nuclear Science and Engineering, April 1987

    Google Scholar 

  46. Kirch N., Ivens G., Results of AVR experiments,, in: AVR—Experimental High Temperature Reactor—21 Years of Successful Operation for a Future Energy Technology, VDI-Verlag Gmbh, Düsseldorf, 1990

    Google Scholar 

  47. Zhang Z., Sun Y., Current status of nuclear power and HTR development in China, Atomwirtschaft, 51. Jg, Heft 12, Dec. 2006

    Google Scholar 

  48. Zuying G., Lei S., Thermal hydraulic transient analysis of the HTR-10, Nuclear Engineering and Design 218, 2002

    Article  Google Scholar 

  49. Drüke V., Filges D., Kirch N., Neef R.D., experimental and theoretical studies of criticality safety by ingress of water in systems with pebble-bed high-temperature gas-cooler reactor fuel, Nuclear Science and Technology 57, 1975

    Article  Google Scholar 

  50. Nabbi R., Jahn W., Meister G., Rchm R., Safety analysis of the reactivity transient, resulting from water ingress into a high-temperature pebble-bed reactor, Nuclear Technology, Vol. 62, Aug. 1983

    Google Scholar 

  51. Mathews D., Brogli R., Chawla R., Stiller P., LEU HTR Experiments for the PROTEUS critical faciligy, Jahrestagung Kerntecknik (Germany), 1991

    Google Scholar 

  52. Wallerbos E.J.M., Reactivity effects in a pebble bed type nuclear reactor—an experimental and calculational study, Diss. TUDeft, 1998

    Google Scholar 

  53. Ashworth F.P.O., Kinsey D.V., Wilkinson V.J., A review of HTR graphite corrosion, Reports of Dragon Project, 858, 1972

    Google Scholar 

  54. Kubaschewski P., Heinrich B., Influence of accidents with water ingress on the behavior of fuel elements of THTR, Jahrestagung Kerntechnik (Germany), 1978

    Google Scholar 

  55. Lönnissen K.J., Analysis of the dependence of the graphite/steam reaction from pressure in the region of porous diffusion in connection with water ingress accidents in high temperature reactors, Diss. RWTH Aachen, JÜL-2159, 1987

    Google Scholar 

  56. Hinssen K.H., Katscher W., Lönnissen K.J., Experimental investigation of corrosion of IG 110 graphite by steam, JÜL-Spez-578, July 1999

    Google Scholar 

  57. Katschev W., Experiments for graphite corrosion during accidents with ingress of air and water in HTR, in: Sicherheit von Hochtemperatur reactor, Jülich, 1985

    Google Scholar 

  58. Stauch B., SUPERNOVA, an experiment for analysis of graphite corrosion in case of severe accidents with ingress of air and water into a pebble-bed HTR, KFA Jülich, 1984

    Google Scholar 

  59. Steinbrink W., Analysis of the application of an emergency cooling measure by water injection into the core of a pebble-bed high-temperature reactor as a diversary decay heat removal system after extreme loss of cooling accidents, Diss. Univ. GH Duisburg, 1986

    Google Scholar 

  60. Stulgies A., The behavior of water in the core of HTR-plants—corrosion of fuel elements with special emphasis on the Leidenfrost effect, Diss. University of Duisburg, 1986

    Google Scholar 

  61. Leber A., Transport and separation of droplets in the primary circuit of high temperature reactor in case of water ingress accident, JÜL-4050, April 2003

    Google Scholar 

  62. Kugeler K., Epping Ch., Schmidtleiu P., Forming of aerosols by graphite corrosion in case of an accident of water ingress into the core of a high temperature reactor, Jahrestagung Kerntechnik (Germany), 1989

    Google Scholar 

  63. Katscher W., Moormann R., Graphite corrosion in HTR pebble-beds in case of air ingress accidents and their interaction with water ingress, JÜL-Conf-43, Nov. 1981

    Google Scholar 

  64. Pershagen B., Light water reactor safety, Pergamon Press, Oxford, New York, Sydney, Tokyo, Toronto, 1985

    Google Scholar 

  65. Ziermann E., Ivens G., Final report on the power operation of the AVR—experimental nuclear power plant, JÜL-3448, Oct. 1997

    Google Scholar 

  66. Kröger W., Benefit of safety assessments, in: AVR—Experimental High Temperature Reactor—21 Years of Successful Operation for a Future Energy Technology, VDI-Verlag Gmbh, Düsseldorf, 1990

    Google Scholar 

  67. Fröhling W., et al., Aspects of chemical stability in innovative nuclear reactors, JÜL-2960, Aug. 1994

    Google Scholar 

  68. Kubaschewski P., Heinrich B., Corrosion of graphitic reactor components in normal operation and in accidents, Jahrestagung Kerntechnik (Germany), 1984

    Google Scholar 

  69. Hinssen K.H., Katscher W., Moormann R., Kinetics of graphite/oxygen reaction in the region of porous diffusion, JÜL-2052, April 1986/JÜL-1875, Nov. 1983

    Google Scholar 

  70. Barthels H., Experiments for the transport of heat and mass in the core and in regions near the core, in: Siclearheit von Hochtemperatur Reaktor, Jülich, 1985

    Google Scholar 

  71. Epping C., The air ingress into the core of a pebble bed high temperature reactor, Diss. University of Duisburg, Aug. 1989

    Google Scholar 

  72. Hurtado Gutierrez A.M., Analysis of massive air ingress into high temperature reactor, Diss. RWTA Aachen, Dec. 1990

    Google Scholar 

  73. Roes. J., Experimental analysis of the graphite corrosion and forming of aerosols in the core of a pebble-bed high temperature reactor, JÜL-2956, 1994

    Google Scholar 

  74. Schaaf Th., Experiments for the transfer of heat and material because of natural convection in case of air ingress accident in a high temperature reactor, JÜL-3620, Jan. 1999

    Google Scholar 

  75. Kuhlmann M.B., Experiments on the transport of gas and graphite corrosion in air ingress accidents in high temperature reactors, Diss. RWTH Aachen, 2002

    Google Scholar 

  76. Ziermann E., Ivens G., Description of experiments VAMPYR I, II in AVR, in: Final Report on the Power Operation in the AVR-experimental Nuclear Power Plant, JÜL-3448, Oct. 1997

    Google Scholar 

  77. Von der Deckeu C.B., Wawrzik U., Dust and activity behavior, in: AVR—Experimental High Temperature Reactor—21 Years of Successful Operation for a Future Energy Technology, VDI-Verlag Gmbh, Düsseldorf, 1990

    Google Scholar 

  78. Schenk W., Analysis of the behavior of coated particle and spherical fuel elements at accident temperature, JÜL-1490, May 1978

    Google Scholar 

  79. Schenk W., Naoumidis A., Nickel H., The behavior of spherical HTR-fuel elements under accidents conditions, Journal of Nuclear Materials 124, 1984

    Article  Google Scholar 

  80. Goodin D. T., Schenk W., Nabielek H., Accident condition testing of U.S. and FRG high temperature gas-cooled reactor fuels, Kernforschungsanlage Jülich, JÜL-Spez-286, Jan. 1985

    Google Scholar 

  81. Schenk W., Pitzer P., Nabielek H., Fission product release from pebble bed fuel elements at accidental temperatures, JÜL-2091, Oct. 1986

    Google Scholar 

  82. Schenk W., Pitzer D., Nabielek H., Fission product release profiles from spherical HTR-fuel elements at accident temperatures, JÜL-2234, Sept. 1988

    Google Scholar 

  83. Schenk W., Nabielek H., Spherical fuel elements with TRISO-coated particles at accident conditions, JÜL-Spez-487, Jan. 1989

    Google Scholar 

  84. Schenk W., Gontard R., Nabielek H., Performance of HTR samples under high irradiation and accident simulation conditions, with emphasis on test capsules HTR-p 4 and Sl-p1; FZJ Report Juel-3373, April 1997

    Google Scholar 

  85. Schenk W., Nabielek H., High temperature reactor fuel fission product release and distribution at 1600°C, Nuclear Technology 96 (1991), pp. 323

    Article  Google Scholar 

  86. Schenk W., Verfondern K.L., Nabielek H., Toscana E.H., Limits of LEU TRISO particle performance, Proceedings of HTR-TN International HTR Fuel Seminar, Brussels, Belgium, Feb. 1–2, 2001

    Google Scholar 

  87. Nickel H., Nabielek H., Pott G., Mehner A.W., Long time experience with HTR fuel elements, Proceedings of HTR-TN, International HTR Fuel Seminar, Brussels, Belgium, Feb. 1–2, 2001

    Google Scholar 

  88. Iniotakis N., von der Decken C.B., The retardation effect of structural graphite on the release of fission products in case of hypothetical accidents of HTR’s, Gas Cooled Reactor Today, BNES, London (1982)

    Google Scholar 

  89. IAEA, Fuel performance and fission product behavior in gas cooled reactors, IAEA-TECDOC-978, Vienna, Nov. 1997

    Google Scholar 

  90. Freis D., Accident simulation and analysis of post-irradiation on spherical fuel elements for high temperature reactors, Diss. RWTH Aachen, July 2010

    Google Scholar 

  91. Schenk W., Pott G., Fuel elements with coated particles with high burnup (UO2-TRISO) in air, Jahetagung Kerntechnik (Germany), 1995

    Google Scholar 

  92. Drecker S., Koschmieder R., Meister G., Moormann R., Fission product release in HTR stem ingress events by UO2-oxidation, Jahrestagung Kerntechnik (Germany), 1991

    Google Scholar 

  93. BBC/HRB, SAMSON—a vibration test facility for simulating, HRB-Publication, D1229E, 1985

    Google Scholar 

  94. Jakobs H., Kemter F., Schmidt G., The facility SAMSON in the industrial research, Jahetagung Kerntechnik (Germany), 1985

    Google Scholar 

  95. Kleine Tebbe A., Kemter F., Schmidt G., Analysis of the behavior of a pebble bed reactor in accidents of earthquake using the facility SAMSON, Jahrestagung Kerntechnik (Germany), 1984

    Google Scholar 

  96. Breitbach G., Wolters J., Gas exchange between the primary circuit and the reactor-containment of a high temperature reactor, Report of KFA Jülich, 1980

    Google Scholar 

  97. Breitbach G., David H.P., Nickel M., Wolters J., Gas exchange between a helium containing vessel and the environmental via a downward directed tube and the relevance for the HTR-Modul, JÜL-Spez-273, Sept. 1984

    Google Scholar 

  98. Schmidt U., et al., Neutron physical measurements and calculation during the start of operation of THTR, Jahrestagung Kerntechnik (Germany), 1985

    Google Scholar 

  99. Lange M., Experiments for the self-acting decay heat removal in HTR, Diss. RWTH Aachen, Jan. 1995, JÜL-3012

    Google Scholar 

  100. Esser F., Experimental analysis to the forming and separation of drops in the gas circuit of high-temperature reactors in case of accident with water ingress, JÜL-3635, July 1998; Diss. RWTH Aachen, 1998

    Google Scholar 

  101. Bürktsolz A., Droplet separation, VCH Verlagsgesellschaft, Weinheim, 1989

    Google Scholar 

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  1. Institute for Reactor Safety and Reactor, RWTH Aachen University, Jülich, Germany

    Kurt Kugeler

  2. Institute of Nuclear and New Energy Tech, Tsinghua University, Beijing, China

    Zuoyi Zhang

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  1. Kurt Kugeler
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Kugeler, K., Zhang, Z. (2019). Some Experimental Results on Safety Aspects of Modular HTR. In: Modular High-temperature Gas-cooled Reactor Power Plant. Springer, Berlin, Heidelberg. https://doi.org/10.1007/978-3-662-57712-7_14

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