Cores fast reactors

The fifth component is the stmcture, a material selected for weak absorption for neutrons, and having adequate strength and resistance to corrosion. In thermal reactors, uranium oxide pellets are held and supported by metal tubes, called the cladding. The cladding is composed of zirconium, in the form of an alloy called Zircaloy. Some early reactors used aluminum fast reactors use stainless steel. Additional hardware is required to hold the bundles of fuel rods within a fuel assembly and to support the assembhes that are inserted and removed from the reactor core. Stainless steel is commonly used for such hardware. If the reactor is operated at high temperature and pressure, a thick-walled steel reactor vessel is needed. 

Most fast reactors that use Na or NaK as coolant utilize an intermediate heat exchanger (IHX) that transfers heat from the radioactive core coolant to a nonradio active Hquid-metal coolant loop, which has the reactor s steam generator. This helps minimize the spread of contamination in the event of a leak or fire. 

H. Ninokata et al., Long life Multipurpose Small Size Fast Reactor with Liquid Metallic-fueled Core , Progress in Nuclear Energy, 37, No. 1-4, 299-306 (2000). 

The Integral Fast Reactor would also be capable of breeding plutonium which could be used as nuclear fuel. This type of reactor was seen as the key to a nuclear future. Liquid sodium is a volatile substance that can burst into flames if it comes into contact with either air or water. An early liquid sodium-cooled breeder reactor, the Fermi I, had a melting accident when 2% of the core melted after a few days of operation. Four years later when the reactor was about to be put into operation again a small liquid sodium explosion occurred in the piping. 

M. Dalle Donne, Heat Transfer in Gas Cooled Fast Reactor Cores, Ann. Nucl. Energy (5) 439-... 

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... 

However, the choice of liquid sodium as a coolant and principal design features of fast reactors were mainly determined in the 1960s, as already mentioned, by the requirement of high power densities in the reactor core (about 500 kW(th)/l for MOX fuel), and the need of a weakly moderating material with good heat transfer properties. The important fact was also that sodium is practically non-corrosive to stainless steal. 

Considerable experience has been gained in the Russian Federation with lead-bismuth (PbBi) eutectic alloy application as reactor coolant. Since Bi is sufficiently rare and expensive metal, and also it is a source of volatile a-active Vo, the proposal to use lead as a coolant in power fast reactors is now under consideration in several countries. Lead based alloys are currently being considered for hybrid systems (accelerator driven fast reactors) in which the coolant could double as the spallation source for driving the core. 

High power rating and temperature impose special requirements to fast reactor thermohydraulics. Analysis of thermohydraulic issues assumes reliable hydraulic and heat transfer relationships to be worked out. As a result, distributions of the coolant flow rate, its velocity, and finally, fuel and core structure temperatures would be obtained. All mentioned parameters are required for evaluation of core integrity and mechanical behaviour. 

One of the attractive features of the fast reactor is its hard neutron spectrum. To expand this feature, a metallic fuel core is employed in the 4S. However, it is more difficult to reduce void reactivity for a core with a harder spectrum. It is very important to design the void reactivity to be negative in order to prevent a severe nuclear accident in the event of sudden loss of coolant, sudden loss of coolant flow or a large gas bubble entrainment in the core. 

PSA studies were carried out based on available data from various fast reactors to establish the probability of plugging in SA. Based on these studies, lowering of scram thresholds for core AT and core mean temperature (0m) from the fuel SA thermocouple was done and reactor operation continued with CCPM at 80 mm position. 

FBTR has been fully commissioned with small core up to a power level of 13.4 MWt and the performance of all the safety related systems has been satisfactory. Large number of modifications was carried out based on experience feed back and analysis of various incidents to improve system performance. Construction, commissioning and operation of FBTR have given considerable amount of experience and confidence, which will help in its smooth and sustained operation at nominal power and will also give useful feedback for the design and commissioning of large fast reactors. 

The experimental fast reactor JOYO at the Japan Nuclear Cycle Development Institute s Oarai Engineering Center attained initial criticality in April 1977 and was the first liquid metal cooled fast reactor in Japan. From 1983 to 2000, JOYO operated with the MK-II core as an irradiation test bed to develop the fuels and materials for future Japanese fast reactors. Thirty-five duty cycle operations and thirteen special tests with the MK-II core were completed by June 2000 without any fuel pin failures or serious plant trouble. The reactor is currently being upgraded to the MK-III core. This paper provides a review of the operational experiences obtained through the JOYO s operation. 

JOYO is a sodium cooled fast reactor with mixed oxide (MOX) fuel. The main reactor parameters of the MK-II irradiation bed core are shown in Table 1, which compares the MK-II with the future MK-III core. 

The successful operations of JOYO provide a wealth of experiences with core management, impurity control, reactor engineering tests, innovative instrumentation techniques, operation and maintenance support systems, and component modifications. These experiences and accumulated data are to be used for the design of future fast reactors. They are also useful for upgrading the JOYO core and plant to the MK-III configuration and are essential to secure steady and safe reactor operation and enhance the irradiation capability of JOYO in the future. 

MAEDA, S., et al.. Fast Reactor Core Management in Japan Twenty Years Evolution at JOYO, Proc. 5th International Topical Meeting on Research Reactor Fuel Management (RRFM 2001) Eurogress Aachen, Germany, 1-4 April 2001, European Nuclear Society (ENS) (2001) pp. 56-60. 

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