Sodium-cooled fast reactors

Lee, D. H., 1970, Studies of Heat Transfer and Pressure Drop Relevant to Subcritical Once-through Evaporator, Paper IAEA-SM-130/56, Symp. on Progress in Sodium-Cooled Fast Reactor Engineering, Monte Carlo, Monaco. (4)... 

This cycle uses solid reactants. Small dendritic copper particles are used to carry out the last reaction to make the transformation of all the solid copper to CuCl, thereby maximizing hydrogen yield. The reported efficiency of this cycle is 49% [66]. This low temperature cycle is believed to eliminate many of the engineering and materials issues associated with the other two previously discussed cycles, however this cycle is also in the initial stages of development [111]. The temperature ranges are such that lower temperature nuclear reactors, e.g. sodium-cooled fast reactors, could be used with this cycle [69]. A hybrid version of this cycle is under investigation in Argonne National Laboratory [66,112]. 

In considering the operational safety and accident analyses of sodium-cooled fast reactors, similar information on the release of fission products from sodium is needed. Although the extent of vaporization can often be calculated from thermodynamic considerations (3, 4), appropriate transport models are required to describe the rate phenomena. In this chapter the results of an analytical and experimental investigation of cesium transport from sodium into flowing inert gases are presented. The limiting case of maximum release is also considered. 

A new thermochemical and electrolytic hybrid hydrogen production system in lower temperature range has been developed by the Japan Nuclear Cycle Development Institute (JNC) to achieve the hydrogen production from water by using the heat from a sodium cooled fast reactor (SFR) [7]. 

A concept for nuclear production of hydrogen, FR-MR , which combines sodium cooled fast reactors (SFR) with the membrane reformer technology, has been studied jointly by MHI, ARTEC, TGC and NSA[15]. 

Because the thorium atom density is higher in thorium metal than in any thorium compound, metal is the preferred form of thorium where the hipest nuclear reactivity or hipest density is wanted. One likely nuclear application is in a sodium-cooled fast reactor where thorium would capture a neutron and be converted to... 

Shirin, V. M., et al. Use of Lead in Unloading Systems of Sodium-Cooled Facilities, in IAEA Symposium on Progress in Sodium-Cooled Fast Reactor Engineering, Monaco, Mar. 1970. 

I HE Reactor Engineering Division of the Argonne National Laboratory is currently working on the design of a sodium-cooled fast reactor. Figure 1 shows the complexity of equipment that must remain compatible with, and operate in sodium at temperatures ranging from 580° to 900 °F. 

At the present state of the art, a corner of an article about evaluation of population hazards is hardly an appropriate place in which to attempt an exposition of reactor safety. Nevertheless, we may contrive a brief description of these types of reactor accidents which, it is thought, could lead to fission product release. The intention is to illustrate ways in which fuel could be damaged and then release fission products ultimately to the atmosphere. Though gas-cooled reactors, water-cooled reactors, and sodium-cooled fast reactors will be discussed, no comparisons, invidious or otherwise, are intended between the safety of these systems. 

Some Fission Product Activities in the Fuel of a Sodium-Cooled Fast Reactor ... 

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. 

Considerable experience has been gained by the Russian specialists on tests and operation of sodium cooled fast reactors (over 100 reactor-years). Based on this experience, modifications were made of systems and components of the reactors in operation, as well as of the BN-800 reactor design. 

Fast Breeder Test Reactor (FBTR) is a 40 MWt/ 13.2 MWe sodium cooled, mixed carbide fuelled, loop type reactor. It has two primary and secondary sodium loops and a common steam water circuit, which supplies high pressure, high temperature superheated steam to turbine generator (TG). Heat is rejected in cooling tower (Fig 1). A 100% capacity dump condenser is provided for reactor operation even when the TG is not in service. The mmn aim of the reactor is to generate experience in the design, construction and operation of sodium cooled fast reactors and to serve as an irradiation facility for the development of fuels and structural material for fast reactors. It achieved first criticality in Oct 85 with Mark I core... 

International Topical meeting on sodium cooled fast reactor safety, 3-7 Obninsk, Russian Federation. 

Large experience has been already gained with sodium cooled fast reactor operation. The use of sodium as a coolant poses fire danger in case of its leakage and interaction with air or water. Operating experience testifies the possibility of coping with the mentioned problem, but the quest for excellence calls for future improvement in LMFRs technology. 

In some experimental and prototype NPPs with sodium-cooled fast reactors water/steam-sodium contacts took place in SG. The requirements to sodium-water steam generator from the standpoint of its tightness should be much higher than those to any other heat exchanger. When considering specific problems, on the basis of experience obtained on some NPPs the following conclusions can be drawn ... 

In conclusion, although the CEA has now decided to focus its R D on gas cooled reactor concepts, with the prospect of perfecting a gas cooled fast reactor in the long term (4th generation), it nevertheless shall preserve activities and expertise it has acquired on sodium cooled fast reactors over the years. CEA would like to pursue further exchanges with other countries who are also engaged in liquid metal fast reactor research. 

Session 1 Sodium cooled fast reactor operational experience... 

The papers presented a comprehensive overview of the accumulated experience with the operation of sodium cooled fast reactors. The worldwide 40+ years of fast reactor development represent a total of 300 years of operation. Based on this figure, it was concluded that the sodium cooled fast reactor technology has reached a mature stage. The advantages of this type of reactor were pointed out by the various presenters ... 

Session 4 Sodium cooled fast reactor knowledge preservation... 

France, Japan and the Russian Federation presented the status on sodium cooled fast reactor experience preservation made in these coimtries. The reports underlined these countries large experience with design, construction and operation of sodium cooled fast reactors. The discussions underlined the importance of the IAEA support for knowledge preservation of fast reactor experience. 

Recognizing the importance of sodium cooled fast reactor knowledge preservation, there is the need to pursue an internationally coordinated activity aiming at the analysis of operational experience with fast reactor equipment and systems, as well as at the generalization of the lessons learned. The objectives of this activity are to ... 

Safeguard the feedback from commissioning, operation, and decommissioning experience of experimental and power sodium cooled fast reactors ... 

As a next step, the need to establish an international sodium cooled fast reactor database was put forward. In an ad-hoc working group, the session elaborated the general structure of such a database. Table 1 gives an example of the structure. 

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