Lead-cooled fast reactors

Uranium and mixed uranium—plutonium nitrides have a potential use as nuclear fuels for lead cooled fast reactors (136—139). Reactors of this type have been proposed for use ia deep-sea research vehicles (136). However, similar to the oxides, ia order for these materials to be useful as fuels, the nitrides must have an appropriate size and shape, ie, spheres. Microspheres of uranium nitrides have been fabricated by internal gelation and carbothermic reduction (140,141). Another use for uranium nitrides is as a catalyst for the cracking of NH at 550°C, which results ia high yields of H2 (142). 

Actinide nitrides are known for Th through Cm. All of the nitrides are high melting compounds with melting points of 2630 °C, 2560 °C, and 2580 °C for Th, Np, and Pu, respectively. The actinide nitrides can decompose to give N2. Thorium, uranimn, and plutonium nitrides are well known and can be used as nuclear fiiels. Fuels of this type, especially uranium and mixed uranium plutonium nitrides, can be used in lead-cooled fast reactors, which have been proposed as a possible next-generation nuclear reactor and for use in deep-sea research vehicles. 

Techniques to counter the heavy metal coolant disadvantages are being developed, but in spite of this work and the apparent disadvantages of sodium, the consensus in favour of sodium remains strong. This is demonstrated by fact that before lead-cooled fast reactor BREST-300 is built, MINATOM will first build a sodium-cooled LMFR BN-800 (E. Adamov, NW, 23 September 1999). Moreover, in the last few years sodium has been chosen in both China and the Republic of Korea for the respective fast reactor development project. This is a significant endorsement for sodium as a fast reactor coolant. 

Studies of lead-bismuth and lead-cooled fast reactors are being carried out in the Russian Federation (RF) organizations Institute of Physics and Power Engineering (IPPE) and EDO... 

Preliminary design study of lead cooled fast reactor with nitride fuel assemblies has been performed by the Japanese specialists to improve uranium resource utilization and transmutation of HLW nuclides. Plant size limitations caused by seismic resistance... 

Lead cooled fast reactor design assumes the subassembly with fuel pins arranged in square lattice with large pitch-to-diameter ratio (s/d 1.4-1.5). In LMFR a more tight arrangement of fuel pins is adopted (s/d 1.1-1.18) ... 

It should be pointed out that presently only liquid metal coolant-sodium is widely adopted for fast reactors. Mercury was used for a short period ("Clementine" reactor in the USA and BR-2 in the USSR having, respectively, thermal output powers of 30 and 100 kW). A number of lead cooled fast reactors are being studied presently. 

Preliminary studies on lead-bismuth and lead cooled reactors and ADS (accelerator driven systems) have been initiated in France, Japan, the United States of America, Italy, and other countries. Considerable experience has been gained in the Russian Fedaration in the course of development and operation of reactors cooled with lead-bismuth eutectic, in particular, propulsion reactors. Studies on lead cooled fast reactors are also under way in this country. 

Lead-Cooled Fast Reactor with an On-Site Fuel Cycle. 2713... 

Although the lead-cooled fast reactor (LFR) is not dealt with in this part, it is also a promising advanced FR and will be introduced in the second part by Prof. Adamov. In addition, several topics of international importance, such as nonproliferation, are treated in his part. Other advanced reactors, such as the gas-cooled fast reactor (GFR), the super critical water-cooled reactor (SCWR), and the molten-salt reactor (MSR), are briefly introduced in Sects. 58.1.4, 58.1.5, and 58.1.7, respectively. 

This made it possible for Russia to initiate development (Adamov and Orlov 1992) of a fast reactor named lead-cooled fast reactor of natural safety (BREST), i.e., a pilot 300 MWe reactor (for trying out, e.g., equilibrium operation) and the first-of-the-kind 1,200 MWe power unit, with the R D effort to support it, aimed at demonstration of the new fast reactor and its closed fuel cycle within the normal period of 20 years. 

An example of an innovative lead-cooled fast reactor of natural safety is the pilot 300 MWe reactor BREST-OD-300, which is being developed in Russia (ISTC 2001 Adamov et al. 1997). 

In the future, electricity production at large NPPs is likely to remain the main application of nuclear energy. This factor and the reduction of unit costs with increase in the power and number of nuclear units were the reasons for conceptual study of a 1,200 MWe lead-cooled fast reactor as a candidate basic component of a large-scale nuclear power mix. The BREST-OD-300 being developed as a prototype of the BREST-1200 reactor, their design and engineering features are largely similar, as may be seen from the data of Table 58.6 (ISTC 2001 Adamov et al. 1997). 

Lead-cooled fast reactor. (From US EXDE Nuclear Energy Research Advisory Committee and the Generation IV International Forum, A technology roadmap for generation IV nuclear energy systems, GIF-002-00, 2002.)... 

The lead-cooled fast reactor uses either lead or lead-bismuth eutectic in the primary coolant loop. This gives similar advantages as with the SFR in terms of operational safety. Several of these reactor designs were built and operated on Russian submarines. 

WADE, D., DOCTOR, R., PEDDICORD, K.L., STAR-H2 A 400 MW,h lead-cooled fast reactor for hydrogen manufacture in a sustainable hierarchical hub-spoke energy infrastructure, paper 1189 GENES4/ANP2003, Kyoto, Japan (Sept. 15-19, 2003). 

XXIV-3] SIENICKI, J., et al, A small secure transportable autonomous lead cooled fast reactor for deployment at remote sites, Americas Nuclear Energy Symposium 2004 (Paper presented at Symposium, Miami Beach, Florida, Oct. 3-6, 2004). 

Refueiing. The refueling, operation, and maintenance of the AHTR will have many similarities to sodium- and lead-cooled fast reactors. 

An agreement to submit design descriptions for this report was not reached with the designers of BREST-300 lead cooled fast reactor from RDIPE (NIKIET) of the Russian Federation and the designers of CANDU X NC reactor from AECL of Canada (the latter is a Generation IV system with supercritical light water coolant). A description of the BREST-300 can be found in reference [21]. 

Lead and LBE are relatively inert liquids with very good thermodynamic properties. The LFR would have multiple applications including production of electricity, hydrogen, and process heat. System concepts represented in plans of the GIF System Research Plan are based on the European Lead-cooled Fast Reactor, Russia s BREST-OD-300 (fast reactor with lead coolant BbicxpbiH PeaKTop co CBHmtoBbiM TeiiJiOHOCHTeJieM in Russian abbreviations) and the Small Secure Transportable Autonomous Reactor concept designed in the US. 

Lead-cooled fast reactor NPP (Russian design BREST-OD-300 reactor coolant — liquid lead P 0.1 MPa and Pin/Pout = 420/540°C primary power cycle — indirect subcritical pressure Rankine steam cycle Pi 17 MPa (P = 22.064 MPa) and Pin/Pout — 340/505°C (Por = 374°C) high-temperature steam superheat (in one of the previous designs of BREST-300 NPP primary power cycle was indirect supercritical-pressure Rankine steam cycle Pjn 24.5 MPa (P r = 22.064 MPa) and Pin/Pout = 340/520°C (Per = 374°C) also, note that power-conversion cycle in different lead-cooled fast reactor designs from other countries is based on a supercritical pressure carbon-dioxide Brayton gas turbine cycle. 41-43... 

Historical GFR concepts as weU as the Generation IV GFRs represent an alternative to liquid metal—cooled fast reactors (LMFRs). The use of gases leads to a harder neutron spectrum compared with the fast reactor cores of sodium- and lead-cooled fast reactors... 

Overview and motivation for lead-cooled fast reactor systems... 

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