HTGR

Materials for nuclear RPVs developed to meet the advances in RPV technology and attain the safety and reliability are discussed in this section. The designation of the materials has been standardized in the pressure vessel codes and regulations of many countries. The evolution of the ASME Code is described as a typical example. 

As mentioned earlier, the first LWR in Japan was the JAERIJPDR which started operating in 1963. The reactor is of the plate construction type and the material was SA302B modified. The first commercial nuclear power plant was JAPCO s Tokai 1 which went into operation in 1966. The RPV was the Calder Hall-type made of JIS SB46 modified (Coltuf 26 equivalent) plate steel. After that, JAPCO constructed the Tsuruga 1 BWR plant in 1965. The Tokyo Electric Power Company (TEPCO) and Kansai Electric Power Company (KEPCO) also decided to construct the Fukushima 1 

BWR plant and Mihama 1 PWR plant. Next, the BWR, ABWR and PWR plants were constructed and operated in Japan. 

ASME Code Section III at first used the fracture analysis diagram (FAD) for the prevention of brittle fracture. Linear elastic fracture mechanics was introduced in 1972 Summer Addenda, Appendix G. ASME Code Section XI Appendix A and NRC Federal Register lOCFR Part 50 were issued in 1973. In these codes and regulations, RTndt was introduced as an important index temperature to characterize the transition curve of fracture toughness. 

In addition, the codes specify that the toughness degradation of RPV steels during service due to neutron irradiation embrittlement must be considered. In 1975, NRC issued Regulatory Guide 1.99 in which the prediction procedure of toughness degradation based on neutron fluence and P and Cu content of materials was introduced. In 1984, the ASME Code decreased the permissible content of P, Cu and S contents in the materials. 

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The metal is a source of nuclear power. There is probably more energy available for use from thorium in the minerals of the earth s crust than from both uranium and fossil fuels. Any sizable demand from thorium as a nuclear fuel is still several years in the future. Work has been done in developing thorium cycle converter-reactor systems. Several prototypes, including the HTGR (high-temperature gas-cooled reactor) and MSRE (molten salt converter reactor experiment), have operated. While the HTGR reactors are efficient, they are not expected to become important commercially for many years because of certain operating difficulties. 

The coolant for the HTGR is helium. The helium is not corrosive has good heat properties, having a specific heat that is much greater than that of CO2 does not condense and can operate at any temperature has a negligible neutron absorption cross section and can be used in a direct cycle, driving a gas turbine with high efficiency. 

The highest power of a reactor of the HTGR type was 330 MWe in Fort St. Vrain, Colorado. The reactor, started in 1979, had many technical problems, including helium leaks, and did not perform up to expectations. It was shut down in 1989. 

Bulk graphites are also used in the HTGR concept to support and surround the active fuel core. These components tend to be large, complex-shaped blocks and have been produced from commercial grades of molded or extmded graphites. 

The HTGR designed by the General Atomic Company and constructed at Peach Bottom, Pennsylvania, U.S.A., was a 40 MW(e) experimental power plant which was similar in many respects to the Dragon reactor. Peach Bottom started commercial operation on June 1, 1967, and ceased operation on October 31, 1974 [36]. The major performance parameters of the Peach Bottom Reactor are shown in Table 8. 

Bottom 1 prototype HTGR operated successfully in all respects under the auspices of the U.S. Atomic Energy Commission s Power Reactor Demonstration Program. However, its si2e (only 40 MW) was insufficient to justify continued commercial operation. 

Table 9. The major performance parameters of the Fort St. Vrain HTGR [29,38-40]...

The Fort St. Vrain HTGR was permanently shutdown in August 1988 [41]. The nuclear island performed well dining the reactors 15-year life, although significant problems were encoxmtered with some of the reactors non-nuclear support systems [29,38]. 

Fig. 14. HTGR fuel elements (a) prismatic core HTGR fuel element (b) cross section of a spherical fuel element for the pebble bed HTGR. Reprinted from [88], 1977 Ameriean Nuelear Soeiety, La Grange Park, Illinois.

Saito, S., Status of the HTGR Program in Japan. In Proceedings of the IAEA Specialists meeting on the Status of Graphite Development for Gas Cooled Reactors, IAEA-TECHDOC-690, IAEA, Vienna, 1993. pp. 29 36. 

Ishiyama, S. and Eto, M., Recent R D of C/C Composite Control Rod for HTGRs In Proceedings of 22nd Biennial Conf. on Carbon Pub. American Carbon Society, 1995, pp 161-162. 

Fleming, K. N. et al., 1979, A Methodology for Risk Assessment of Major Fires and Its Application to an HTGR Plant, General Atomic GA-A15401. 

The problem with all such ideas is that they are only ideas. Before they can be tuned into real possibilities, considerable R. D. would have to be undertaken, perhaps lasting 10+ years. If successful, which of course cannot be guaranteed, commercially sized demonstration units would have to be built and operated satisfactorily before the generating industry would be Willing to take the risk of investing in these developments. All this implies that with the possible exception of the modular HTGR, it may be 20.years or more before the first commercial plants come into operation We may well have the time, but ways will have to be found to find the money. 

Nishihara, T. et al., Development of control technology for the HTGR hydrogen production system, New Orleans, in Proc. of the Global 2003, New Orleans, Louisiana, November 16-20, p. 320, 2003. 

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