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1 History of Nuclear Power Reactor Development

1 History of Nuclear Power Reactor Development

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1 History of Nuclear Power Reactor Development

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  1. Some Issues in R&D of Nuclear GraphiteInstitute of Nuclear Energy Technology , 100084, Peking, ChinaShijiangXuJune 2014

  2. 1 History of Nuclear Power Reactor Development 1.1 history of nuclear power reactor

  3. 1 History of Nuclear Power Reactor Development 1.2 Risk in Nuclear Power Human error: Windscale Three mile island Chernobyl Natural disaster Tsunami (Japan)

  4. 1 History of Nuclear Power Reactor Development 1.3 Generation Four Reactor Criteria of G-4 Nuclear Reactor: 1 Improved safety: Inherent safety, passive residual heat remove ,…; 2 Improved Economy: Effect use of resouces, high thermal efficiency, short construction period,…; 3 Nuclear proliferation resistance: low Pu quality, difficult to reprocess of spent fuel, enhanced physical protection,…; 4 Low production of radioactive waste.

  5. 1 History of Nuclear Power Reactor Development 1.4 Candidates of the Fourth Generation Reactor System

  6. 2 Very High Temperature Reactor 2.1 Pebble bed type 2.2 Prismatic type

  7. 2 Very High Temperature Reactor 2.1 Pebble Bed High Temperature Gas Cooled Reactor

  8. 2 Very High Temperature Reactor 2.1.1 Structure of Pebble Bed HTGR Reflector Top Reflector Side Reflector Bottom Reflector

  9. 2 Very High Temperature Reactor 2.1.2 Reflector Brocks

  10. 2 Very High Temperature Reactor 2.1 .3 spherical fuel element Half Section Coated Particle 0.92 mm Uranium Dioxide kernel 0.5mm 5mm Graphite layer Coated particles imbedded in Graphite Matrix Pyrolytic Carbon Silicon Carbite Barrier Coating Inner Pyrolytic Carbon Porous Carbon buffer layer Fuel Sphere Dia. 60mm

  11. 2 Very High Temperature Reactor 2.2.1 Schematical Diagram of Prismatic HTGR

  12. 2 Very High Temperature Reactor 2.2.2 Prismatical Fuel Element

  13. 3 Molten Salt Reactor 3.1 Solid fuel type Molten salt without fuel is used as coolant , it also has some moderation function 3.2 Liquid fuel type Molten salt with fuel is used as fuel and coolant, it also has some moderation function. Graphite is the structure, refractor and moderator material of both type reactors

  14. 3 Molten Salt Reactor 3.1 Solid fuel molten salt reactor Solid fuel molten salt reactor is the first step of R&D of liquid fuel molten salt reactor. It uses molten LiBeF salt as coolant and graphite as moderator, reflector and structure materials. It uses the fuel element of high temperature gas cooled reactor as its fuel element. The reactor design is more or less similar to HTR.

  15. 3 Molten Salt Reactor 3.2.1 Schematic Diagram of Liquid Molten Salt Reactor

  16. 3 Molten Salt Reactor 3.2.2 Main features of liquid fuel molten salt reactor. 1 Liquid fuel could be treated on line to separate the new produced nuclear fuel and fission products with high neutron absorption cross section, therefore to improve the conversion or breed ratio. 2 First loop can work in the condition of high temperature with low pressure. 3 The pressure boundary of first loop is the only barrier of fission products. The fission products, esp. gaseous fission products, are likely more easy to release to the environment under accident conditions.

  17. 4 Nuclear Graphite and Fourth Generation Generation Reactor 4.1 Nuclear graphite is the reflector, moderator and structure material of HTGR and MSR. It is the essnetial material for building these reactors. Graphite provides the potential for these reactors to realize their special ability and is also the limiting factor for the performance of the reactors.

  18. 4 Nuclear Graphite and Fourth Generation Reactor 4.2 Does the HTGR meet the criteria of fourth generation reactor? Yes: Inherent safty, Nuclear proliferation resistance Economy (high utilization of resource; short construction period, i.e. low investment risk; high thermal efficiency, No: Economy (shorter reactor service life), Large volume of radioactive waste ,esp. if it has to be treated as radioactive waste for the reason of C-14. Even though the radio activity is not so high (roughly about 25.7Ci/t, according to the experience of AVR), the public may not accept.

  19. 4 Nuclear Graphite and Fourth Generation Generation Reactor 4.3 Inherent safty Reactor safty strategy is base on large negative temperature coefficent of reactiviy, not based on active engineering measures; Design of HTGR-modul enables the reactor to use the mechanism of conduction, convection and radiation to remove the residual heat out of the reactor core after lost of coolant accident; The highest temperature of reactor core is in the range of 16000C , all ceramic fuel release almost no fission porduct up to 16000C in 500 hours.

  20. 4 Nuclear Graphite and Fourth Generation Reactor 4.4 Does the Molten Salt Reactor meet the criteria of fourth generation reactor? Yes: nuclear proliferation resistance; economy : high resource utilization (high conversion or breed of nuclear fuel) No: economy: shorter reactor service life large volume of radioactive waste ,esp. if it has to be treated as radioactive waste for the reason of C-14. Even though the radio activity is quite low, the public may not accept. high risk of radioactive materials, esp. gaseous fission products.

  21. 5 Some Issues in R&D of Nuclear Graphite 5.1 Deepening the research on radiation damage 5.1.1 Theoretically, how long could the graphite survived under neutron irradiation? 5.1.2 Is the irradiation damage could be annealed out like the Wigner energy does? 5.1.3 The relationship of Irradiation behavior and nuclear graphite structure, i.e. raw material and fabrication technology. 5.1.4 Effect of recoil fission fragment on Nuclear graphite used in MSR.

  22. 5 Some Issues in R&D of Nuclear Graphite5.2 Study on filler Coke 5.2.1 Coke is the main raw materials of nuclear graphite. Filler coke determines the irradiation behavior of nuclear graphite. There is no general rule for the selection of filler coke up to now. Once the filler coke is changed, the developed nuclear graphite will disappear, new nuclear graphite has to be re-developed from the very beginning, e.g. ASR series to NGB series.

  23. 5 Some Issues in R&D of Nuclear Graphite5.2 Study on filler Coke 5.2.2 Study of filler coke How the Structure develops from the precursor to filler coke and to final graphite? How people could control the fabrication process to produce similar or identical filler structure, therefore peaple could avoid the problem of raw material supply?

  24. 5 Some Issues in R&D of Nuclear Graphite5.2 Study on filler Coke 5.2.3 The relationship of nuclear graphite performance and filler structure.

  25. 5 Some issues in R&D of Nuclear Graphite5.3 Nuclear Graphite for liquid fuel MSR 5.3.1 Bombardment effect of recoil nuclide of fission fragment on nuclear graphite to be used in MSR.

  26. 5 Some issues in R&D of Nuclear Graphite5.3 Nuclear Graphite for liquid fuel MSR 5.3.2 Penetration /infiltration of molten salt in nuclear graphite; effect of environment on the wettability of molten salt –graphite system. Is there any event, e.g. water ingression, surface active element and/or compound produced during reactor service, which could reduce the contact angle to less than 900 ? R&D of molten salt reactor graphite has to be based on the understanding of these information.

  27. 5 Some issues in R&D of Nuclear Graphite5.3 Nuclear Graphite for liquid fuel MSR 5.3.3 Migration of carbon in molten salt reactor; carbon precipitated from molten salt and deposition on the surface of components of coolant system.

  28. 5 Some Issues in R&D of Nuclear Graphite5.4Re-production of used nuclear graphite 5.4.1 Alleviate the raw material supply problem;

  29. Some Issues in R&D of Nuclear Graphite5.4 Re-production of used nuclear graphite 5.4.2 Alleviate the pressure on the treatment of radioactive graphite waste, esp. for the used graphite of molten salt reactor.

  30. Thank you

  31. Radioactive Nuclide Inventory of the graphite samples

  32. Temperature coefficient of reactivity of HTGR

  33. Change of Reactor Power and Temperature after loss of coolant

  34. Decay Heat after Reactor shut down

  35. Accident Temperature of HTGR Modul