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Future Aspects of HTR Development

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Modular High-temperature Gas-cooled Reactor Power Plant
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Abstract

For the modular HTR in the future, different developments are possible, to improve the energy conversion, the safety, the economical conditions of the fuel supply, and the waste management. One important topic is the realization of higher thermal power obeying the concept of limitation of the maximal fuel temperature during accidents to values of below 1600 °C. A well-suited possibility is the application of annular core. For reactor pressure vessels consisting of forged steel values of around 400 MWth will be possible, prestressed vessels allow much higher thermal power till around 1500 MWth, using an annular core too. Very high helium temperatures, which can be important for future processes of electricity generation or for process heat applications, can be produced by the special fuel cycle OTTO, in which spherical fuel elements run just one time through the core. The temperature differences between fuel and helium then become very small at the core exit. Furthermore, improvements of the fuel elements can be realized: better coatings and corrosion protection reduce fission product release and possible damage by corrosion. For the safety of the plant and for special sites, prestressed reactor pressure vessels can be advantageous, because they can be transported in pieces and are evaluated as burst safe. Oriented on rising safety requirements of the future, innovative containments, and the possibility to arrange this component underground become important. Some measures are improved filter concepts and storage volumes for off-gas. Thorium can be used as fertile material and will broaden the fuel basis in the future. Breeding will be possible in HTR plants too, if the burnup is reduced and reprocessing will be developed. Even partitioning and transmutation can be applied to spent HTR fuel too in a far future. Intermediate storage systems, however, will be the most advantageous solution for the next decade, because these systems can be designed with extreme safety standards. Even extreme earthquake loads can be tolerated. Further improvements are possible like underground siting or enhanced physical protection against very large airplanes or weapons. Final storage offers some further improvements using additional ceramic layers for the spent fuel elements or special canisters. Much work has been done already for the development of components and materials for intermediate cycle. These solutions can help to realize different cycles for the generation of electricity or for cogeneration in the future.

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References

  1. International Event Scale (INES), Scale for the Evaluation of Nuclear Accidents, IAEA, Vienna, 1990

    Google Scholar 

  2. Special Issue: A technology roadmap for Generation IV, Nuclear Energy Systems, issued by the US-DOE Nuclear Energy Research, Advisory Committee and the Generation IV International Forum, Dec. 2002. IAEA, Objectives for the development of advanced nuclear plants, IAEA-TECDOC-682, Jan. 1993

    Google Scholar 

  3. Spiegelberg-Plauer R., Stern W., International reporting of nuclear and radiological events at the international atomic energy agency, Atomwirtschaft 52., Jg., Heft 2, 2007

    Google Scholar 

  4. Fricke U., Analysis on the possibility to rise the thermal power of inherently safe high-temperature reactors by optimization of the core design, Diss. University of Duisburg, 1987

    Google Scholar 

  5. Dazhong Wang, Analysis of a high-temperature reactor with an inner region of graphite balls, JŰL-1809, July 1982

    Google Scholar 

  6. Kupitz J., Neutron physical and thermo-hydraulic layout of an annular rector with special fuel elements and proposal for a shutdown system, JŰL-1644, Feb. 1980

    Google Scholar 

  7. Sun Yuliang, Analysis to transfer the safety features of the modular reactor on a large power reactor, JŰL-2585, Feb. 1992

    Google Scholar 

  8. Koster A., Status of the PBMR concept, Presentation in FZ Jülich, March 2007

    Google Scholar 

  9. Schulten R., et al., Industrial nuclear power plant with high-temperature reactor PR 500– OTTO-principle – for the production of process steam, JŰL-941-RG, April 1973

    Google Scholar 

  10. Hansen U., Schulten R., Teuchert E., Physical properties of the “once through then out” pebble-bed reactor, KFA Jülich, Aug. 1971

    Google Scholar 

  11. Tcuchert E., Maly V., Haas K.A., Basic study for the pebble-bed reactor with OTTO fuel cycle, JŰL-858-RG, May 1972

    Google Scholar 

  12. Teuchert E., Bohl L., Rütten H.J., Haas K.A., The pebble-bed high-temperature reactor as a source of nuclear process heat, Vol. 2, Core Physics Studies, JŰL-1114-RG, Oct. 1974

    Google Scholar 

  13. Teuchert E., Rütten H.J., Werner H., Haas K.A., Schulten R., Closed thorium cycles in the pebble-bed HTR, JŰL-1569, Jan. 1979

    Google Scholar 

  14. Teuchert E., Rütten H.J., Core physics and fuel cycles of the pebble-bed reactor, Nuclear Engineering and Design, 34, 1975

    Article  Google Scholar 

  15. Teuchert E., et al., Physics features of the HTR for process heat, Nuclear Engineering and Design, 78, 1984

    Article  Google Scholar 

  16. Nabielek H., et al., Fuel for pebble-bed HTRs, Nuclear Engineering and Design, 78, 1984

    Article  Google Scholar 

  17. IAEA, Fuel performance and fission product behavior in gas-cooled reactor, IAEA-TECDOC-978, Nov. 1997

    Google Scholar 

  18. Nickel H., HTR coated particles and fuel elements, HTR/ECS 2002, Cadarache, 2002

    Google Scholar 

  19. Petti D., Coated particle fuel behavior under irradiation, HTR ECS 2002, High-Temperature Reactor School, Cadarache, France, Nov. 2002

    Google Scholar 

  20. Williams D.F., Fuels for VHTR, FZ/OH 2004, The 2004 Frederic Joliot + OTTO Hahn Summer School, Cadarache, France, Aug./Sept., 2004

    Google Scholar 

  21. Wolf L., Ballensiefen G., Fröhling W., Fuel elements of the high-temperature pebble-bed reactor, Nuclear Engineering and Design, Vol. 34, No. 1, 1975

    Article  Google Scholar 

  22. Teuchert E., Once through cycles in the pebble-bed HTR, JŰL-1470, Dec. 1977

    Google Scholar 

  23. Teuchert E., Fuel cycles of the pebble-bed reactor in the computer simulation, JŰL-3632, Jan. 1999

    Google Scholar 

  24. Mulder E.J., Pebble-bed reactors with equalized core power distribution, inherently safe and simple, JŰL-3632, 1999

    Google Scholar 

  25. Zgliczynski J.B., Silady F.A., Neylan A.J., The gas turbine –modular helium reactor (GT-MHR), high efficiency, cost competitive nuclear energy for the next century, GA-A21610, International Topical Meeting on Advanced Reactor Safety, Pittsburgh PA, USA, April 1994

    Google Scholar 

  26. IAEA, 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-TECDOC-1163, Jan. 2001

    Google Scholar 

  27. Kugeler K., Sappock M., Beine B., Wolf W., Development of an inactive heat removal system for high-temperature reactors, IAEA-TECDOC-757, Vienna, Aug. 1994

    Google Scholar 

  28. Design of reactor containment systems for nuclear power plants, Safety, No. NS-G-1.10, Vienna 2004

    Google Scholar 

  29. Dräger H., On the possibilities of reactor containments to withstand terroristic attacks with large airplanes, Münohen, 2001

    Google Scholar 

  30. IAEA, External events excluding earthquakes in the design of nuclear power plants, IAEA Safety Guide No. NS-G-1.5, 2003

    Google Scholar 

  31. Hauck M., Design and layout of storage vessel systems for radioactive waste under extreme conditions of accidents, Dipl. Arbeit, RWTH Aachen, June 2003

    Google Scholar 

  32. NN, Special issue: Supplementary report vol. 1, Prestressed concrete pressure vessels under hypothetical accident conditions, JŰL-Report, RS 447, Jan. 1984

    Google Scholar 

  33. Schartmann F., Heat removal from transport and intermediate storage vessels for high radioactive materials in normal operation and in case of accidents, Diss. RWTH Aachen, Oct. 2000

    Google Scholar 

  34. VDI statement, The safety relevant design of nuclear facilities in Germany against terroristic attack, VDI, Düsseldorf, Nov. 2001

    Google Scholar 

  35. IAEA, Seismic evaluation of existing nuclear power plants, IAEA Safety Reports Series No. 28, Vienna, 2003

    Google Scholar 

  36. Ahorner L., Seismologic expertise for the site Jülich regarding the planned spallation source, Jülich, Oct. 1984

    Google Scholar 

  37. NN, Information about the burst protection for the PWR of the BASF nuclear plant for steam supply, 1980

    Google Scholar 

  38. Burrow E.D., Willaims A.J., Reactor pressure vessel, Nuclear Engineering International, 1969

    Google Scholar 

  39. Schöning J., The prestressed concrete reactor pressure vessel of THTR, construction and pressure test, Atom + Strom, Heft 6, 1982

    Google Scholar 

  40. Schäfer J.K., Schwiers H.G., Brunton J.D., Crowder R., Experience of design and construction of PCPVs for AGRs and HTRs in Europe, European Nuclear Conference, April 1975

    Google Scholar 

  41. Neylan. J., The multicavity PCRV, Nuclear Engineering International, Aug 1974

    Google Scholar 

  42. Fröhling W., Prestressed cast pressure vessels (VGD) as a burst safe pressure enclosure for innovative application in the nuclear technology, Monograph of FZ Jülich, Energy Technology, Vol. 14, 2000

    Google Scholar 

  43. Bounin D., The presstressed cast iron pressure vessel VGD – a new concept for large vessels, Konstruieren + Gieβen, 12, Vol. 4, 1987

    Google Scholar 

  44. Warnke P., Elter C., Mitterbacher P., Results of the pressure tests on the VGD –vessel for the control gas storage in THTR 300, Firmensdirift Siempelkomp, 1982

    Google Scholar 

  45. Schilling F.E., Beine B., The prestressed cast iron reactor pressure vessel (PC/PV), Nuclear Engineering and Design, Vol. 25, No. 2, July 1973

    Article  Google Scholar 

  46. 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 Reactor, New Delhi, India, Aug. 1991

    Google Scholar 

  47. Wolf L., Kneer 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 

  48. Hammelmann K.H., Kugeler M., Fröhling W., Experiment for a prestressed cast steel vessel, Final report on experiments at elevated temperature of VGGD, KFA-IRE-OB-1/87, Mar. 1987

    Google Scholar 

  49. Fröhling W., Kugeler M., Hammelenann K. H., Phlippen P.W., Design aspects and development goals of prestressed cast steel vessels, 18. MPA-Seminar, Stuttgart, 1992

    Google Scholar 

  50. Stoltz A., Analysis of the structural mechanical behavior of the top and bottom design of prestressed cast steel vessel using finite element methods, Dipl. Arbeit, Univ. Duisburg, April 1988

    Google Scholar 

  51. Wagner U., Prestressed cast steel vessel as primary enclosure for high-temperature reactors of small and medium power, comparison of systems, structural mechanics, design and safety, Diss. Univ. Duisburg, 1989

    Google Scholar 

  52. Helbig G., Structural mechanical analysis of different variants of a prestressed wall of a cast steel vessel using finite elements, Dipl. Arbeit, Univ. Duisburg, 1988

    Google Scholar 

  53. Huber G., Specific civil engineering questions during the construction of prestressed concrete reactor pressure vessels; comparison of French, British and German technologies, Techn. Mitteilung Krupp Forsch. Ber. Bd31, Heft 2, 1973

    Google Scholar 

  54. Jühe S.H., release of carbon dust in case of depressurization accidents of high-temperature reactors, Diss. RWTH Aachen, Dec. 2011

    Google Scholar 

  55. Eckardt B., Containment venting for light water reactor plants, Kerntechnik 53, No 1, 1988

    Google Scholar 

  56. Kuczera B., Wilhelm J., Depressurization systems for LWR containments, Atomwirtsdraft, Mar. 1989

    Google Scholar 

  57. Johansson K., Nilsson L., Persson Å., Filter design considerations for implementing a vent filter system at the Barseback nuclear power plant, International Meeting on Thermal Nuclear Reactor Safety, Chicago, Aug/Sept. 1982

    Google Scholar 

  58. Elkhawaukey S., Comparison of modern concepts for separation of fine dust, Stud. Arbeit, RWTH Aachen, April 2008

    Google Scholar 

  59. Ackermann G., Operation and maintenance of nuclear power plants, VEB Deutsche Verlag für Grundrloff Induschie, Leipnig, 1982

    Google Scholar 

  60. Hake G., The relation of reactor design to siting and containment in Canada, Proceedings IAEA, Containment and Siting of Nuclear Power Plants, April 1967, STI/PUB/154

    Google Scholar 

  61. Allensworth J.A., et al., Underground siting of nuclear power plants: potential benefits and penalties, NUREG-0255, 8/1977

    Google Scholar 

  62. Bender F. (editor), Underground siting of nuclear power plants, E.Schweizerbart’sche Verlagsbuchhandlung, Nägele und Obermiller, Stuttgart, 1982

    Google Scholar 

  63. IAEA, Containment and siting of nuclear power plants, STI/PUB/154, April 1967

    Google Scholar 

  64. Kröger W., Containments for high-temperature reactors regarding special aspects of environmental protection and safety, JŰL-1098RG, 1974

    Google Scholar 

  65. Schulten R., et al., Industrial nuclear power plant with high-temperature reactor PR 500 OTTO principle for the production of process steam, JŰL-941RG, 1973

    Google Scholar 

  66. Kröger W., et al., Evaluation of underground construction of nuclear power plants in the soil of an open hole, JŰL-1478, Jan. 1978

    Google Scholar 

  67. Russel C.R., Reactor safeguards, Pergamon Press, Oxford, London, New York, Paris, 1962

    Google Scholar 

  68. Riera J.D., On the stress analysis of structure subjected to air craft impact forces, Nuclear Engineering and Design, Dec. 8, 1968

    Google Scholar 

  69. RSK, Guidelines of the reactor safety commission for the safety technical design of pressurized water reactors, Ges für Reaktor Sicherhest, Köln, 1977

    Google Scholar 

  70. United States Nuclear Regulatory Commission, Use of probabilistic risk assessments in plan-specific, risk-informed decision making: General Guidance, Revision 1 of Standard Review Plan Chapter 19, NUREG-0800, Washington, D.C., November 2002

    Google Scholar 

  71. Kugeler M., Kugeler K., Romes G., The development of a fast discharge system for pebble-bed reactors, IRE/KFA JŰlich, May 1979

    Google Scholar 

  72. Phlippen P.W., The development of a fast discharge system as a diversitary decay heat removal system for high-temperature reactors with spherical fuel elements, JŰL-1788, July 1982

    Google Scholar 

  73. Buchmann R., experimental analysis for the fast discharge system for pebble-bed high-temperature reactor, Diss. RWTH Aachen, 1983

    Google Scholar 

  74. Schrör H., Constructive and thermohydraulic design of a fast discharge system, JŰL-Spez-53, Aug. 1979

    Google Scholar 

  75. NN, Special issue: Thorium and gas cooled reactors, Annals of Nuclear Energy, Vol. 5, No. 8–10, 1978, Pergamon Press

    Google Scholar 

  76. Lung M., Gremm O., Perspectives of Thorium fuel cycle, Nuclear Engineering and Design, 180, 1998

    Article  Google Scholar 

  77. Teuchert E., Rütten H.J., Werner H., Reducing the world Uranium requirement by the Thorium fuel cycle in high-temperature reactors, Nuclear Technology 58, Sept. 1982

    Google Scholar 

  78. IAEA, Potential of thorium based fuel cycles to constrain plutonium and reduce long-lived waste toxicity, Final report of a coordinated research project, IAEA-TECDOC-1349, Apr. 2003

    Google Scholar 

  79. Baumgärner F. (edit), Chemistry of nuclear waste management, Verlag Karl Thiemig, Vol. 1–3, München, 1980

    Google Scholar 

  80. Merz E., et al., Reprocessing of nuclear fuels, JŰL-Spez-207, May 1983

    Google Scholar 

  81. Merz E., Reprocessing in the Thorium fuel cycle, JŰL-Spez-239, 1984; Status seminar high-temperature reactor—fuel cycle, JŰL-Conf-61, Aug. 1987

    Google Scholar 

  82. Salvatores M., The physics of transmutation for radioactive waste minimization, CEA, FJSS, 1998

    Google Scholar 

  83. Jameson R.A., Venneri F., Bowmann C.D., Accelerator driven transmutation technology—an energy supply bridge to the future without long-lived radioactive waste, AVH-Magazin, 66, 1995

    Google Scholar 

  84. Phlippen P.W., et al., Basic considerations and requests for the assessment of future fuel cycle with special regard to P + T issues, Global 1995, Spet. 1995

    Google Scholar 

  85. Bauer G.S., Accelerator based neutron sources and experiments, FJSS 98, CEA Cadarache, 1998

    Google Scholar 

  86. Rubbia C., et al., Status report on the energy amplifier, CERN 1994, NN, Accelerator driven systems, energy generation and transmutation of nuclear waste, Status Report IAEA-TECDOC-985

    Google Scholar 

  87. Broeders C.H.M., Burning of transuranium isotopes in thermal and fast reactors, Nachrichten FZ-Karlsruhe 29, Jg. 3, 1997

    Google Scholar 

  88. Corafunke E.H.P., et al., Transmutation of nuclear waste, ECN-R-95-025, July 1995

    Google Scholar 

  89. Merz E.R., Walter C.E., Advanced nuclear system consuming excess plutonium, Nato ASI series 1, Disarmament Technologies, Vol. 15, Klumer Academic Publisher, Dordrecht, Boston, London, 1997

    Google Scholar 

  90. Benedict M. Pigford T.H., Levi H.W., Nuclear chemical engineering, McGraw Hill Book Company, New York, 1981

    Google Scholar 

  91. RSK, Special issue: Long time intermediate storage of radioactive waste, Jülich, May 2002

    Google Scholar 

  92. Sailer M., The new RSK guidelines for intermediate storage installations in Germany, Atomwirschaft, 46, Jg., Vol. 5, 2001

    Google Scholar 

  93. KLE, Special issue: Short description of the intermediate storage facility for spent fuel elements on the site of nuclear power plant, Lingen/Kernkraftwerk, Ewsland, July 1999

    Google Scholar 

  94. Soma W., The interim storage spent fuel assemblies, TUG-Meeting, April 2000

    Google Scholar 

  95. Fachinger J., et al., R + D on intermediate storage and final disposal of spent HTR fuel, IAEA-TECDOC-1043, Vienna, 1998

    Google Scholar 

  96. Beiser H., et al., Interim storage facilities in plants with convoy concept, Atomwirtschaft, 46, 2001

    Google Scholar 

  97. IAEA, Special issue: IAEA safety standard series, design of fuel handling and storage systems for nuclear power plant, IAEA, No. Ns-G-1.4, 2003

    Google Scholar 

  98. NN, Special issue: intermediate storage facilities on the site of nuclear power plants, Inform Verlag, 2001

    Google Scholar 

  99. Röthemeyer H. (edit), Final storage of radioactive waste, VCH, Weinheim, New York, Basel, Cambridge, 1991

    Google Scholar 

  100. Barre B., Nuclear fuel and fuel cycle, Landolt Börnstein, Energie Technologies, VIII/38, Springer, Berlin, 2005

    Google Scholar 

  101. NN, Special issue: MAW and HTR fuel element: experimental storage in bore holes, JŰL-Conf-60, July 1987

    Google Scholar 

  102. Merz E., Conditioning of radioactive wastes in the nuclear fuel cycle, JŰL-Spez-394, March 1987

    Google Scholar 

  103. Barre B., Radioactive waste management in France – with special emphasis on HLW, Atomwirtschaft, Heft 7, 47. Jg. 2002

    Google Scholar 

  104. Bodansky D., Nuclear energy: principles, practices and prospects, Springer, Science + Baisiners Media, Inc., 2004

    Google Scholar 

  105. Niephaus D., Reference concept for the direct final storage of spent HTR fuel elements in CASTOR THTR/AVR transport and storage vessels, JŰL-3734, Jan. 2000

    Google Scholar 

  106. IAEA, Design and evaluation of heat utilization systems for the high-temperature engineering test reactor, IAEA-TECDOC-1236, Vienna, Aug. 2001

    Google Scholar 

  107. Jansing W., Testing of high-temperature components in the KVK, Specialists Meeting on Heat Exchanging Components of Gas-Cooled Reactors, Düsseldorf, April 16–19, 1984

    Google Scholar 

  108. Harth R., et al., Experience gained from the EVA II and KVK operation, Nuclear Engineering and Design, 121, 1990

    Article  Google Scholar 

  109. Nieβen H.F., et al., Experiments and results of the PNP test steam reformers in the EVA II plant, JŰL-2213, 1988

    Google Scholar 

  110. NN, Special issue: status of development in the field of heat transporting and heat transferring components, Sderifteureihe: Juergiepolitik in Nordrhein Lvestfalen, Baud 16 I and 16 II, 1984

    Google Scholar 

  111. Project NFE (Nuclear long distance energy transport) final report, JŰL-Spez-303, 1985

    Google Scholar 

  112. Ohashi H., et al., Current status of research and development on system integration technology for connection between HTGR and hydrogen production system at JAEA, OECD/NEA, 3rd Information Exchange Meeting on the Nuclear Production of Hydrogen, Oarai Japan, Oct. 2005

    Google Scholar 

  113. Nickel H., Schubert F., Schuster H., Very high temperature design criteria for nuclear heat exchanger in advanced high temperature reactors, Nuclear Engineering and Design, 94, 1986

    Article  Google Scholar 

  114. Kugeler K., Kugeler M., Hohn H., Hurtads A., Very high temperature reactors (VHTR) for process heat generation, Publication prepared

    Google Scholar 

  115. Kugeler K., Schulten R., Considerations on the safety principles of nuclear technology, JŰL-2720, Jan. 1993

    Google Scholar 

  116. IAEA, Safety of nuclear power plants, Design IAEA Safety Standards Series, Requirements, No. NS-R-1, Vienna, 2000

    Google Scholar 

  117. Kugeler K., Barnert H., Phlippen P.W., Scherer W., Sustainable nuclear energy with the HTR for future application, HTGR Application and Development, Beijing, March 2001

    Google Scholar 

  118. Scherer W., et al., Self-acting limitation of nuclear power and of fuel temperature in innovative nuclear reactors, JŰL-2960, Aug. 1994

    Google Scholar 

  119. Fröhling W., et al., Chemical stability of innovative nuclear reactors, JŰL-3118, Sept. 1995

    Google Scholar 

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

    Google Scholar 

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Kugeler, K., Zhang, Z. (2019). Future Aspects of HTR Development. In: Modular High-temperature Gas-cooled Reactor Power Plant. Springer, Berlin, Heidelberg. https://doi.org/10.1007/978-3-662-57712-7_15

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