Experience in Operating the BOR-60 Reactor

The results of 30 years of operation of BOR-60 are presented. The operating regimes and parameters for the core and main loops are presented. The operation of the main equipment and various systems is analyzed. A classification of the faults and failures is made. The advantages of currently operating counter-flow steam generators, from the standpoint of safety in the case of the appearance of an interloop leak, are indicated. Advances in sodium technology are briefly examined.     

Safety of Pyroelectrochemical Reprocessing of Spent BOR-60 Fuel

The results of an investigation of the safety of reprocessing 15.5 kg of spent BOR-60 fuel are presented. The absolute amounts of radioactive aerosols entering the ventilation system of the pyroelectrochemical setup during the reprocessing of spent fuel are determined. The transfer of radionuclides from fuel into aerosols is estimated. This makes it possible to obtain an expression for the expected inflow of radioactive substances from the setup, depending on the amount of materials handled in the technological scheme. The data presented on the possibility of obtaining large amounts of uranium and plutonium in the finished product when their content in the solid wastes is negligible are presented. Estimates are made of the irradiation of workers at all stages of the process.__________Translated from Atomnaya Energiya, Vol. 98, No. 4, pp. 280–288, April, 2005.     

Radiation Conditions during the Preparation of Irradiated BOR-60 Fuel Assemblies for Reprocessing

The basic stages in the preparation of irradiated BOR-60 reactor fuel for reprocessing are examined. It is determined that during the separation of the fuel part of the fuel elements the coefficient of transfer of ¹³⁷Cs from the fuel into aerosol is 5·10^–6 and for fragmentation the value is 3·10^–5. It is found that the real catching efficiency for aerosol particles caught by a V-05 filter ranges from 42 to 99%. The specific entry of radioactive aerosols into the ventillation center after the first stage of air purification was 0.3 MBq for -emitters and 7.7 MBq for and emitters per 1 kg of reprocessed fuel. The total collective dose formed at the stages of preparation of a large batch of irradiated fuel (four spent fuel asemblies with average burnup 11.4% and a 10.5 to 23.7 yr holding period) for reprocessing was 11.5·10^–3 persons·Sv.     

Boiling-sodium tests with the BOR-60 reactor

Conclusions These first experiments on the BOR-60 reactor have shown that in principle it is possible to detect the boiling of sodium from the acoustic and neutron fluctuations; useful information has been obtained on the character of the signals and the scope for using various processing methods. However, further measurements and calculations are needed before we can design a reliable real-time system for monitoring for sodium boiling in the core of a fast reactor.USSR, GDR. Translated from Atomnaya Énergiya, Vol. 45, No. 5, pp. 338–342, November, 1978.     

Tests of Models of BREST-OD-300 Reactor Fuel Elements in an Autonomous Lead-Cooled Channel of a BOR-60 Reactor

A special feature of the BREST-OD-300 reactor that is now being designed is that it employs a container-type heat-conducting fuel element with mixed uranium–plutonium mononitride fuel, a lead sublayer, and an expansion volume at the top to collect gaseous products. The fuel elements are arranged in a square array with a wide spacing and are spaced by laminated spacing lattices.The substantiation of the technical solutions adopted for the construction of the reactor fuel elements and fuel assemblies, specifically, the combined effect of the coolant and heat loads on the fuel-element cladding and the spacing lattices, led to the choice of the BOR-60 sodium-cooled fast research reactor as an experimental base and required the development and construction of an autonomous lead-cooled channel loaded into a cell through a passage in the rotatable plugs of the reactor. The channel was tested for two microruns with the BOR-60 reactor operating at 45 MW. The lead temperature at the fuel assembly entrance was 595°C, the working temperature of the cladding was 658°C, the damaging dose was 6.5 displacements/atom, and the fuel burnup was 0.44% h.a. Analysis of the activity of the gas and the lead showed that the fuel elements are sealed. Post-reactor studies have been conducted since August 2002.     

Some Experimental Work Performed on the BOR-60 Reactor

The experimental work performed on a BOR-60 reactor over a period of 30 years of reactor operation is briefly reviewed. The results of investigations of the neutron-physical, thermohydraulic, and dynamical characteristics and the safety parameters of the reactor are presented. The investigations performed and the analysis of transient and emergency regimes made it possible to improve the standard shutdown and cooldown systems in order to soften the temperature conditions on the reactor components.The result of a series of experiments on the safety of fast sodium reactors, among which the introduction of gas into the core, sodium boiling, blocking of the flow in the experimental fuel assembly with destruction of fuel elements, interloop leakage in the steam generators, and so on, are discussed. A complex of diagnostics systems has been developed and tested on the basis of the safety investigations.Analysis of the radiation parameters and characteristics of the reactor made it possible to develop methods and means for monitoring and improving the radiation conditions and the safety of the reactor.Experimental irradiation of various initial materials, using threshold and other reactions, enabled the serial production of radionuclides for medical purposes.     

Investigation of 85Kr Emission and Its Effect on the Radiation Conditions during the Preparation of BOR-60 Fuel for Recovery

The main stages of the BOR-60 closed fuel cycle, implemented on the experimental base at the Scientific-Research Institute of Nuclear Reactors, are examined. The ⁸⁵Kr emission at the stages of preparation of the spent BOR-60 fuel assemblies for recovery is determined experimentally. It is shown that the maximum ⁸⁵Kr emission as a result of destruction of fuel element cladding with oxide uranium fuel is 68%; its contribution to the irradiation dose to the public as a result of mechanical disassembly of the fuel elements in a single BOR-60 fuel assembly with 10% burnup and a 10-yr holding time does not exceed 1·10^–4% of the dose limit (1 mSv/yr).     

ODS cladding fuel pins irradiation tests using the BOR-60 reactor

In order to confirm the irradiation behavior of ODS steels and thus judge their applicability to fuel claddings, fuel pin irradiation tests using 9Cr and 12Cr-ODS claddings developed by JAEA were conducted to burn-up of 11.9 at% and neutron dose of 51 dpa in the BOR-60. Superior properties of the ODS claddings concerning FCCI, dimensional stability under irradiation and so on were confirmed and indicated good application prospects for high burn-up fuel. On the other hand, anomalous irradiation behaviors, fuel pin failure and the microstructure change containing coarse and irregular precipitates, occurred in a part of the fuel pin with 9Cr-ODS cladding. This paper describes evaluation of the obtained irradiation data and the investigation results into the cause of the anomalous irradiation behaviors.