Prototype reactor operation

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. 

The Dounreay site was established as the site of the UK Fast Breeder Nuclear programme in 1955 and became operational in 1958. It accommodated three reactors, the Materials Test Reactor, (DMTR, 1958-1969), the Dounreay Fast Reactor (DFR, 1959-1977) and the Prototype Fast Reactor (PFR, 1974-1994). With all reactor operations now finished and the reactors already de-fuelled the site is undergoing active decommissioning which is planned to be completed by 2032. The Dounreay site has been cited by UKAEA as being the second biggest nuclear decommissioning challenge in the UK with similar liabilities to those at Sellafield but with smaller waste volumes. 

Stable operation of the demonstration reactor BN-600 in Russia with a nominal power output of 600MW(e) for 20 years and an average load factor of 72%, successful operation of the prototype reactors BN-350 in Kazakshstan and Phenix in France as well as the reliable operation of MOX fuel at high bumup (20% witti an irradiation dose in excess of 160 displacement per atom (dpa) in the cladding) in PFR (UK) and Phenix, are milestone in the implementation of LMFR technology. 

Three sodium cooled fast reactors are currently in operation in Russia, namely BR-10 and BOR-60 experimental reactors and BN-600 demonstration reactor NPP. NPP with the BN-350 prototype reactor is now on the territory of Kazakhstan Republic. However Russian institutions and enterprises which participated in the design development and construction of the BN-350 reactor are now involved in its operation. 

It should be noted that in other South and East Asia countries with few indigenous fossil fuel and little uranium ore reserves there is the same situation concerning effective nuclear fuel breeding by LMFR. Republic of Korea s LMFR program consists of development, design and construction of a prototype reactor of 150-350 MW(e) power. The first fully-proven reactor is planned to be in operation by 2025. In China, experimental fast reactor CEFR-25 is planned to become critical in 2005. 

The present status with respect to the fast breeder test reactor operating experience and prototype fast breeder reactor design are presented below. 

Several prototype reactor projects are going ahead, e g. in China, India, Japan, the Russian Federation (these reactors are likely to be commissioned by 2010). However, large scale commercial reactor construction is not expected before 2020. There is a major interest for all countries to preserve the operational experience for both the ongoing and future long term projects. 

The total number of immediate deaths attributable directly to these incidents over 35 or more years of nuclear reactor operation is less than 35—three at a military prototype reactor in 1981 in the USA and 31 at Chernobyl. However, three subsequent deaths have been reported at Chernobyl (see Appendix 7). Of the fuel meltdown incidents (excluding Chernobyl-4), eight relatively serious incidents have been selected and subjected to some analysis in the following subsection. It is noted that, of these fuel meltdown incidents, only one (Three Mile Island-2) was at an operating, fully developed power plant. All of the other incidents involved research reactors or developmental or prototype plant. Three relatively minor incidents are also reviewed where single channel fuel overheating occurred in graphite-moderated plant. 

The purpose of this section is to compare the features of the RBMK reactor operated at Chernobyl with reactor types pertinent to the UK. It will be recollected that the RBMK covers a large number of reactors and the comparisons made are indeed with Chernobyl No. 4. The UK reactors covered are in three classes the commercial reactors now built and operated or in commission (Magnox and Advanced Gas-cooled Reactor (AGR)) the prototype Steam Generating Heavy Water Reactor (SGHWR) and Prototype Fast Reactor (PFR) that have comparable performance to commercial reactors and the proposed Pressurised Water Reactor (PWR) or Sizewell B design which, it... 

The radiation protection problems which emerge during the operation of a fusion prototype reactor, either in normal or in accident conditions, are essentially connected with the presence of tritium, with the generation of neutrons with energies of 2.45 and 14.1 MeV (derived from the D-D and D-T reactions) and with the delayed radiation (and related thermal decay power), for the activation of the structures of the machine. 

France in particular has operational experience with a commercial scale fast reactor, Snperphenix, with 1,300 MWe power ontpnt. In Germany, although a steam-cooled fast reactor had been considered, they have actually developed SFRs, beginning with the experimental reactors, KNK-I and II. Then, a prototype reactor, SNR-300, with 327 MWe power ontpnt was constmcted. It had the same plant concept as the prototype reactor Monju in Japan. However, the provincial government where SNR was constmcted opposed its operation in 1986, and it was dismantled. 

Seventeen LMR prototype reactors and power plants have been built and most of them have accumulated many years of operating experience. Breeding capability and the closed fuel cycle have... 

Germany - AVR - 1966 to 1988 - This prototype helium reactor operated successfully for over 20 years and provided demonstration of 1740°F gas outlet temperature and key safety features, including safe shutdown with total loss of coolant circulation and without control rod insertion. 

The Cirene reactor was a 40-MWe prototype power plant constructed at Latina, 80 km south of Rome. Construction start in 1976 and completion was scheduled for 1984. Commissioning stopped in 1988, before work to reduce the positive void reactivity coefficient was complete, by the general moratorium on nuclear reactor operation imposed by the Italian Government following the Chernobyl accident. 

In Sweden, the first pressure vessel pressurized HWR was constructed at Agesta. This was a project that combined the objectives of two separate concepts one for a district heating reactor and the other for a heat and power reactor. The pressure vessel reactor was conceived as a 65 MWth prototype plant that was to supply district heating and electricity (10 MWe). The reactor was located in an imderground chamber excavated in solid rock and serviced a suburb of Stockholm. The reactor operated with a good degree of reliability. Operation was interrupted over the summer months when district heating was not required. The reactor was shut down in 1975 and decommissioned because it had ceased to be an economical source of power. 

AECL began to study dry storage for spent nuclear fuel in the early 1970s. Silo-like structures called concrete canisters were first developed for the storage of research reactor enriched uranium fuel and then perfected for spent CANDU NUE fuel. By 1987, concrete canisters were being used for the safe and economical storage of all spent fuel accumulated during the operation of AECL s decommissioned prototype reactors. Each canister contains a stack of spent fuel baskets. 

The molten-salt reactor uses molten salt as either the primary coolant or fuel. In either case, both are in motion around the core. One such prototype reactor was built and operated at the Oak Ridge National Laboratory in the 1960 decade. The concept is primarily focused on the thorium/U-233 fuel cycle. The reactor primarily operates near atmospheric pressure, allows for continuous removal of fission products, and offers natural proliferation resistance characteristics. 

NP 300 is a PWR of the French "CAS" type with 9 years of reactor operating experience on the prototype at Cadarache Research Centre. 

DATA FOR AND OPERATING EXPERIENCE WITH REFERENCE/PROTOTYPE PLANT 30 units with 182 reactor operating years. 

Lenz et al. [73] described the development of a 3 kW monolithic steam-supported partial oxidation reactor for jet fuel, which was developed to supply a solid oxide fuel cell (SOFC). The prototype reactor was composed of a ceramic honeycomb monolith (400 cpsi) operated between 950 C at the reactor inlet and 700°C at the reactor outlet [74]. The radial temperature gradient amoimted to 50 K which was attributed to inhomogeneous mixing at the reactor inlet. The feed composition corresponded to S/C ratio of 1.75 and O/C ratio of 1.0 at 50 000 h GHSV. Under these conditions, about 12 vol.% of each carbon monoxide and carbon dioxide were detected in the reformate, while methane was below the detection limit. Later, Lenz et al. [74] described a combination of three monolithic reactors coated with platinum/rhodium catalyst switched in series for jet fuel autothermal reforming. An optimum S/C ratio of 1.5 and an optimum O/C ratio of 0.83 were determined. Under these conditions 78.5% efficiency at 50 000 h GHSV was achieved. The conversion did not exceed 92.5%. In the product of these... 

Many medium and high power research reactors operate closed loops for testing prototype fuel elements for high power research reactors and power reactors. These loops are designed to isolate the test specimen from the reactor and to provide a controlled environment for the... 

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