RU2846581C2 - Method of processing irradiated reactor graphite - Google Patents

Abstract

FIELD: physics.

SUBSTANCE: invention relates to nuclear power engineering and environmental protection, specifically to the reactor graphite wastes processing technology, which excludes the ingress of radionuclides into the environment, and can be used mainly in the uranium-graphite reactors decommissioning, as well as during processing of wastes after repair, restoration and dismantling works of graphite masonry. For this, graphite is preliminarily crushed and ground to a powder state, mixed with bentonite in a powdered state to obtain a homogeneous mixture and then directed to form pellets by pelletizing the mixture in apparatus: pelletizers to produce carbon-bentonite pellets, which undergo additional pelletizing before sintering process, in pure bentonite powder with subsequent hardening of granules to give the pellets the strength required for storage by sintering in an inert gas atmosphere at a temperature higher than 150 °C.

EFFECT: providing safety during disposal due to resistance of carbon-bentonite pellets to leaching of radionuclides.

1 cl

Description

The invention relates to the field of nuclear energy and environmental protection, in particular to the technology of processing reactor graphite waste, eliminating the release of radionuclides into the environment, and can be used primarily in the decommissioning of uranium-graphite reactors, as well as in the processing of waste after repair, restoration and dismantling work on graphite stacks.

In uranium-graphite reactors (UGR), which include reactors of the PUGR, AMB, EGP, RBMK, Magnox and other types, graphite is used as a moderator, which is exposed to irradiation during operation.

By 2038, the nuclear industry in the Russian Federation plans to finally shut down installations with uranium-graphite reactors (UGR). Only 11 RBMK units contain about 25,000 tons of GR-280 graphite, which is subject to safe disposal.

Irradiated graphite after repair, restoration and dismantling works consists of: dust, crumbs, fragments of graphite products, rings, bushings, whole and fragmented blocks.

The following main radionuclides are present in the irradiated graphite UGR: 13C, 36Cl, 238Pu, 239Pu, 242Cm, 244Cm, 241Am, U, 134Cs, 137Cs, 58Co, 60Co, 154Eu, 155Eu, 141Ce, 144Ce, 103Ru, 106Ru, 95Zr, 95Nb, 65Zn, 125Sb, 46Sc, 59Fe, 110Ag, 54Mn, etc.

During neutron activation of the isotopes 14N and 13C in reactor graphite, the radionuclide 14C is formed.

The distribution characteristics of radionuclides in graphite are determined by neutron activation of atoms of chemical elements contained in the media circulating through the graphite masonry and corrosion products of the reactor facility's structural materials, which are fixed in the graphite structure during operation, as well as the consequences of incidents involving damage to the integrity of the fuel matrix.

The half-life of radionuclides is T1/2=5730 years for 14C and T1/2=301000 years for 36Cl. Considering that the duration of the potential hazard period of such RAW reaches tens of thousands of years, it seems technically impossible to ensure the preservation of the hermeticity of the package. Graphite fragments will most likely be subject to destruction.

Under burial conditions, after the loss of the hermeticity of the packaging, the intensity of the release of radionuclides from the radioactive waste will be determined by the resistance of the radioactive waste form itself to the leaching of long-lived radionuclides (LLRN).

In graphite for UGR of Russian production (GR-76, GR-220, GR-280) the surface area of open pores is 225 ^cm2 /g (with a graphite density of about 1.7 g/ ^cm3 ). Consequently, the surface area of open pores of reactor graphite is more than two orders of magnitude greater than the surface area of graphite products and fragments. Thus, the main contribution to the intensity of leaching of DZRN from graphite RW will be made by leaching from the surfaces of pore walls.

In order to reduce the potentially hazardous impact of graphite radioactive waste on the environment to an acceptable level, it is proposed to store graphite in the form of carbon-bentonite pellets, which will ensure: maximum reduction in the leaching rate, elimination of destruction, reduction in the specific radioactivity of the waste.

A known method for recycling solid organ-containing waste contaminated with radioactive components is [RU 2335700, IPC F23G 5/027, G21F 9/32, published 10.10.2008], selected as an analogue. According to the specified method, solid organ-containing radioactive waste contaminated with radioactive components is thermally decomposed without oxygen access. The resulting dusty pyrolysis gases are burned to form flue gases. The flue gases are passed through a catalytic afterburner cartridge. The gas temperature is reduced in a heat exchanger. The flue gases are cleaned of dust using a cyclone and washed with liquid in a scrubber. The sludge captured in the scrubber is sent for repeated pyrolysis in the thermal decomposition chamber. The purified flue gases are released into the atmosphere through a smoke exhauster. The coke residue formed in the thermal decomposition chamber during waste pyrolysis is burned using air on a grate. The grate is located under the thermal decomposition chamber. The resulting ash is discharged into a sealed container.

This method has its disadvantages:

- there is no system for capturing reaction products containing radioactive carbon;

- low productivity of the installation and long duration of chemical processes.

A method is known (Patent RU 2169230, published 27.10.2002, BI No. 30), which includes grinding reactor graphite waste, introducing powdered aluminum, titanium dioxide or chromium oxide and a modifier into the ground waste. The prepared mixture is placed in a container and a layer of igniter with a combustion temperature of at least 2500 K is placed on top of the mixture. Thermal treatment of the mixture is carried out in an inert atmosphere, in the self-propagating high-temperature synthesis (SHS) mode with the formation of carbide oxide material.

The disadvantages of the method are: high heat losses in the contact zone of the reacting mixture with the container walls. The reaction in this zone does not proceed completely, and not all the graphite is bound in the carbide oxide material, which causes partial shedding of graphite from the block surface. The method is not intended for processing reactor graphite waste containing fragments of reactor structures.

The closest to the claimed invention is the method for processing radioactively contaminated metal and graphite waste from uranium-graphite nuclear reactors [RU 2435241, IPC G21F 9/30, published 27.11.2011], selected as an analogue. According to the specified method, a layer of contaminated graphite is loaded into the furnace. The graphite is ignited in an oxidizing plasma environment generated by the furnace plasma torch. After which the furnace is switched off. Radioactively contaminated metal waste and flux are loaded into the furnace from top to bottom, alternately and layer by layer. The metal is melted. The molten metal and the resulting slag flux are removed from the furnace. This method is closest to the claimed method in its technical essence and achieved effect and was selected as a prototype.

This method has its disadvantages:

- none of the known slag fluxes is capable of selectively cleaning the outgoing aerosol-containing gas from radioactive reaction products;

- the complexity of creating and maintaining a discharge to obtain plasma in the proposed system;

- it is difficult to obtain pure molten metal for removal from the shaft furnace, since radioactive metal carbonates will predominantly form in this system.

The objective of the proposed invention is to eliminate the above-mentioned disadvantages and create a method for processing irradiated reactor graphite.

The technical result of the proposed method is as follows:

- Ensuring safety during disposal due to the resistance of carbon-bentonite pellets to the leaching of radionuclides.

- Ensuring the safety (integrity) of radioactive waste by converting it into a monolithic matrix.

- Improving the characteristics of the RAO form for further handling.

- We provide processing of any type of graphite radioactive waste (dust, crumbs, fragments of graphite products, rings, bushings, blocks).

- Low generation of secondary waste.

The method is carried out as follows: graphite is pre-crushed and ground to a powder state, mixed with bentonite, also in a powder state, to obtain a homogeneous batch and then sent to form pellets by pelletizing the batch in pelletizing apparatus to obtain carbon-bentonite pellets, which undergo additional pelletizing before the sintering process, in pure bentonite powder, followed by hardening of the granules to give the pellets the strength necessary for storage, by sintering in an inert gas atmosphere at a temperature above 150°C.

Processing of graphite radioactive waste into carbon-bentonite pellets will ensure:

• reduction of the specific activity of radioactive waste and a significant reduction (virtually elimination) of open graphite surfaces and, consequently, leaching of long-lived radionuclides during further disposal;

• compliance of the form of radioactive waste with the criteria for acceptability for disposal - pellets obtained as a result of sintering have a structurally stable, strong monolithic form;

• optimal filling of the RAW packaging due to the ability to select the required pellet sizes.

Claims (1)

A method for processing irradiated reactor graphite, differing from the known method in that the graphite is first crushed and ground to a powder state, mixed with bentonite in a powder state, to obtain a homogeneous batch, and then sent to form pellets by pelletizing the batch in pelletizing apparatus to obtain carbon-bentonite pellets, which undergo additional pelletizing before the sintering process, in pure bentonite powder, followed by hardening of the granules to impart to the pellets the strength necessary for storage, by sintering in an inert gas atmosphere at a temperature above 150°C.