Validation of Lead-Cooled Reactors
Development, Calculation and Experimental Validation of the Lead-Cooled Fast Reactor Concept (including the Comparison of Critical and Subcritical Reactor with External Proton Source)
Tech Area / Field
- FIR-REA/Reactor Concept/Fission Reactors
3 Approved without Funding
Federal State Unitary Enterprise Research and Development Institute of Power Engineering named after N.A.Dollezhal, Russia, Moscow
- VNIITF, Russia, Chelyabinsk reg., Snezhinsk\nAll-Russian Scientific Research Institute of Non-Organic Materials named after A. Bochvar, Russia, Moscow
Project summaryThe inevitable growth of the world's energy demands in the next century is bound to meet with difficult problems of resources (oil, gas), with environmental and international challenges. Among the numerous new power technologies under investigation today, nuclear fission power appears to be the only realistic way to arrest the growing mining and burning of conventional fuels. But to be capable of this, nuclear power should increase its generation level by an order of magnitude towards the middle of the next century.
The nuclear power industry of such a scale would require a new nuclear technology which provides a much higher level of safety (inherent safety) and a considerable reduction in specific consumption of uranium (BR і 1, closed fuel cycle), along with unimpaired economic advantages. This technology should also provide safe handling of radioactive wastes and, in conjuction with political measures, prevent proliferation of nuclear weapons.
The philosophy of inherent safety engendered by severe accidents at the nuclear power plants puts forward a new approach where the key safety ensuring measures involve not so much traditional multi-stage engineered barriers and systems, as the technical solutions which combine in the optimal way the fundamental physical and chemical laws and properties inherent in fuel, chain reaction, coolant, fission products and other safety-related reactor components. In this case safety is reached not through complication of the design but in the opposite way, by means of its simplification, thus making it cheaper. In particular, fuel breeding can be used not only for uranium saving, but also as a most important safety factor and a way to simpler design and control. Application of chemically inert coolant without phase transformations within a broad temperature range rules out accidents with boiling, void effects of reactivity and loss of core cooling as well as fires and explosions.
The same target can be reached by using a high-density, heat conductive, heat and radiation resistant fuel which ensures near-optimum negative feedbacks, by "designing" additional threshold negative feedback, by a high level of natural circulation of the coolant, etc.
One of the possible ways to develop the concept of natural safety is a subcritical fast reactor with an external neutron source based on a proton accelerator. Such an approach makes an obvious safety case for the reactor in all the possible and conceivable reactivity accidents.
Fast reactors burn efficiently all actinides as well as long-lived fission products (I, Tc). Utilization of medium-life Cs and Sr and cooling of all other radioactive wastes (RW) upon thorough removal of actinides create the conditions required for radiation-equivalent RW burial, replacing the uranium extracted from the ground with its a-active decay products without disturbing the natural radiation equilibrium. Exclusion of reactivity accidents with fuel failure and great radioactive releases removes the principal objection against such a transmutation-based fuel cycle.
Abandonment of the uranium blanket, the small reactivity margin, the fuel cycle closed at NPPs with circulation of a high-activity mixture of U, Pu and other actinides make it possible to take political and technical measures to prevent Pu thefts and its using for weapons.
Implementation of the above-mentioned options would enable us to meet the main requirements of a large-scale nuclear power industry, including deterministic exclusion of disastrous radioactive releases during severe accidents with reactor power excursions, loss of coolant, destruction of the external barriers (leaktight vessel, reactor building), etc.
An urgent task both for Russia and for a number of other countries is to utilize within the next 2-3 decades the plutonium accumulated in storages which comes from NPPs, extracted from their fuel, and from dismantles nuclear weapons, converted into spent fuel. Pu utilization in reactors which will meet the high safety and economy requirements discussed above and will operate in an open fuel cycle at the first prolonged stage, allows reducing the time for clearing the storages of Pu and meeting the Pu management recommendations made by CISAC of the US National Academy of Sciences.
Such an approach to Pu utilization can be especially attractive for the next stage of nuclear power industry advance, since it dispenses with the necessity for the economically inefficient effort to create special "burners" or to burn plutonium in the operating reactors.
These problems can be solved by the concept of a fast reactor cooled by liquid lead. This concept is being developed in Russia within the framework of the State Program "Environmentally Clean Power Industry" and is based on the vast practical experience obtained at Russian fast reactors with Na cooling and the submarine reactors cooled by heavy liquid metal (Pb-Bi alloy). Therefore, development and demonstration of the proposed technology can be realized in a rather limited time. The initial assumptions have been proved by the design investigations and developments of 300, 600 and 1000 MWe reactors and by the experiments on the key aspects of the concept. A similar reactor could be also employed in a system with an external neutron source (cyclic proton accelerator), proposed by CERN (Carlo Rubbia et al.).
The objective of the proposed project is to elaborate the concept of a lead-cooled reactor for conversion of weapon-grade and extracted power reactor plutonium into spent fuel under operation in the open fuel cycle. To substantiate the concept, a package of computations and experiments are planned to be conducted in physics, thermal hydraulics and safety, two design options for the 300 MWe reactor (BREST-300) are to be developed, and the main charac-teristics of the 600 MWe reactor are to be assessed. The first option is a critical reactor (LCFR) and the second is a reactor with 2% subcriticality, operating from an external proton source with a moderate current (ALCFR). For such a source a cyclic accelerator will be used with a current of ~10 mA, which is to raise particles energies of ~1 GeV.
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