Cleaning Up Sites Contaminated with Radioactive Materials: International Workshop Proceedings (2009)

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15 Remediation of Contaminated Facilities at the Kurchatov Institute V. G. Volkov, Yu. A. Zverkov, S. G. Semenov, A. V. Chesnokov, and A. D. Shisha, Russian Research Center—Kurchatov Institute INTRODUCTION During many years of research activities to develop nuclear technologies for military and civil applications, the Russian Research Center—Kurchatov Institute accumulated at its site considerable amounts of solid radioactive waste and spent nuclear fuel. Solid radioactive waste produced before the mid-1960s, some of it having high specific activity, had been placed in interim storage at a special site within the Kurchatov Institute. According to initial estimates, 1,200 m 3 of radioactive waste with a total activity of about 3.7 × 1015 Bq (100,000 Ci) at the time of disposal were placed in temporary storage facilities at the site. At pres- ent, because of intensive construction in Moscow, the institute is surrounded by a densely populated district of the city, and the radioactive waste disposal site adjoins an urban residential area.   Ponomarev-Stepnoi, N. N., V. G. Volkov, N. Ye. Kukharkin, et al. 2002. Rehabilitation of radio- actively contaminated facilities and the site of the Russian Research Center—Kurchatov Institute. Conference Handbook of IBC’s Eighth International Conference and Exhibition on Decommission- ing of Nuclear Facilities—Managing the Legacy, London, United Kingdom, November 11-12, 2002, EA1141, IBC Global Conferences. 99 

100 CLEANING UP SITES CONTAMINATED WITH RADIOACTIVE MATERIALS MIGRATION OF RADIOACTIVE WASTE CONTAMINATION INTO GROUNDWATER To evaluate the environmental impact of old radioactive waste sites and to predict the spread of radioactive contamination through groundwater, monitor- ing boreholes were drilled at the waste disposal site and areas adjoining it on the south and west. Equipped with filter columns, these boreholes were designed to permit observation of the level, chemical composition, and radionuclide content of the groundwater. Available data on the geological structure of the soil, permeability coef- ficients of subsurface horizons, groundwater level, and volume of radioactivity served as a basis for calculation of the groundwater flow structure and strontium- 90 dispersal range. Apparently due to a sharp increase in the groundwater level in the early 1990s, the groundwater flow structure changed, resulting in a risk of contamination beyond the Kurchatov Institute. In addition, according to a model created to predict strontium-90 dispersion, if remediation work is done on the temporary radioactive waste sites, the area of groundwater contamination with strontium-90 content exceeding the action level (5 Bq/L) will remain within the Kurchatov Institute buffer zone. If remediation activities are not performed, the contamination may spread further with groundwater beyond the disposal site, with the strontium-90 content exceeding the action level. PROPOSED MEANS OF REMEDIATION OF THE RADIOACTIVE WASTE DISPOSAL SITE Initially, two alternative approaches to remediation of the radioactive waste disposal site were considered: 1. Creation of geophysically engineered barriers in the radionuclide migra- tion pathways 2. Complete disposition of the waste sites and decontamination of radioac- tive soil To estimate the cost of the first option, evaluations were made of the possibil- ity of building engineered barriers to reduce substantially the spread of contami- nation by infiltration with simultaneous use of sorbents to extract strontium-90 from the groundwater. Zeolites and apatites were considered as sorbents, and underground leaching technologies were proposed to extract cesium-137 from soils.   Rastorguev, A. V., K. Bukharin, V. G. Volkov, et al. 2005. Prognosis of radionuclide contamina- tion spreading on the site of temporary waste storage of RRC Kurchatov Institute. Proceedings of the International Congress ECORAD 2004: The Scientific Basis for Environment Protection Against Ra- dioactivity, Aix-en-Provence, France, September 8-10, 2004. Radioprotection 40(Supp.1):367-370.   Op. cit. 

REMEDIATION OF CONTAMINATED FACILITIES 101 III IV 7 8 6 I I 1 2 II II 3 5 4 III IV FIGURE 15-1  Basic layout of boreholes for the “double envelope” design. NOTE: Boreholes 1 and 2 are for exhaust (optimum flow rate Q1,2 = 0.3 m3/d), and bore- holes 3-8 are for injection (optimum flow rate15-1.eps Figure Q3-8 = 0.1 m3/d). Circles need redrawing These steps would involve a large volume of boring operations and high ma- terial costs. In the meantime, to evaluate the efficiency of such countermeasures, specialists from the Radon enterprise performed laboratory studies on conditions that would permit efficient chemical reactions involved in underground leaching. The studies were performed using near-bottom soils from Temporary Storage Fa- cility No. 3 (the contaminated area was 10 × 8 m2 and 3 m thick) as an example. On the basis of geological and hydrological information and information on the morphology of the area contaminated with cesium-137 within Storage Facility No. 3 at the radioactive waste disposal site, an optimum design for a “double envelope” underground leaching approach involving two exhaust and six injec- tion process boreholes was selected. The basic layout of the boreholes is shown in Figure 15-1. Based on laboratory studies, a mixture of sulphuric and phosphoric acids (0.25M H2SO4 + 0.25M H3PO4) was selected as a process solution. The process solution was injected through boreholes 3-8 and pumped out through boreholes 1-2. The optimum flow rate values for this scheme were estimated as follows: for exhaust boreholes, Qe = 0.3 m3/d, and for injection boreholes, Qi = 0.1 m3/d. Because of clogging, this design resulted in flows that were tens of times less than the flow under ideal conditions, when the soil contains no organic inclu- sions. Calculations were then made to estimate the time required to clean up the near-bottom area of the cesium-137–contaminated Storage Facility No. 3. They totaled t = 4.55 yr at efficiency E = 84.6 percent and t = 2.14 yr at E = 60 percent   Lunev, L. I. 1982. Application Conditions and Physicochemical Basics of Underground Uranium Leaching. Moscow: MGRI Publishing House. 

102 CLEANING UP SITES CONTAMINATED WITH RADIOACTIVE MATERIALS (attainment of the residual specific activity of sand Ct = 10 Bq/g). The foregoing calculations assumed that it would be possible to solve the problem of carbon dioxide release associated with the leaching process. If there is no acceptable technical option for solving this problem, the above calculations will have no practical use. In conclusion, the studies demonstrated that cleaning cesium-137 from rock enclosures under conditions prevalent at Facility No. 3 at the radioactive waste disposal site presents major technical and technological difficulties. Therefore, underground leaching cannot yet be proposed for immediate implementation. According to estimates, the work required to stabilize the waste sites entails high material costs. Therefore, it was decided to pursue the second option, which in- volves complete disposition of the sites. After remediation, the radioactive waste disposal site will still belong to the Kurchatov Institute, so standards established for the institute and its personnel were accepted as the remediation standard. Table 15-1 presents reference levels for background and soil contamination at the disposal site and technogenic gamma background dose rates (above the natural background). These levels are officially established by Kurchatov Institute Order No. 74, dated February 29, 2000, and by sanitary-epidemiological findings of the Russian Federation Ministry of Health Agency for Sanitary-Epidemiological Inspection. As specified by the groundwater remediation criterion, the action level for possible population exposure was 5 Bq/L for strontium-90. TABLE 15-1  Reference Levels of Technogenic Background and Soil Contamination at the Radioactive Waste Disposal Site Monitored Parameter Value Note Equivalent technogenic gamma 2.5 µSv/hr Mean integral dose at Kurchatov Institute background dose rate at the (250 µrem/hr) perimeter: 0.8 mSv/yr (in 2005) radioactive waste disposal site Gamma background on the soil up to 3.0 µSv/hr Soil may be used for filling pits, holes, surface at the disposal site (up to 300 µR/hr) trenches, and so forth at disposal site above 3.0 µSv/hr Soil must be decontaminated or removed for (above 300 µR/hr) disposal TECHNOLOGIES FOR DISPOSITION OF THE OLD WASTE SITES Given the lack of accurate data on design features of the old sites and com- position of the radioactive waste they held, disposition of the sites was performed in accordance with the following standard sequence of steps:   Volkov, V. G., G. G. Gorodetsky, Yu. A. Zverkov, et al. 2004. Radwaste management tech- nologies used in remediation of radioactively contaminated facilities and areas of RRC Kurchatov Institute. Pp. 141-156 in Proceedings of the Seventh International Conference “Nuclear Technology Safety: RW Management,” September 27-October 1, 2004, St. Petersburg, Russia. St. Petersburg: ProAtom Publishing House. 

REMEDIATION OF CONTAMINATED FACILITIES 103 • Drilling of exploratory boreholes in the repository boundary areas and radioactive waste mass followed by a radiation survey • Removal of the built-up ground from the storage facilities and destruc- tion and removal of the facility roofs • Extraction of waste from the facilities, waste sorting, and waste loading into certified containers • Inspection and disposition of storage facility structural elements • Sorting of soil and removal of contaminated soil from storage facility pits • Final radiation survey of storage facility pits and their backfilling with clean soil During disposition of the sites containing concrete-encased high-level waste, shadow radiation shielding was built around them. Exploratory boreholes were made in the site boundary areas and radioactive waste mass to determine more precisely the locations of the repositories, their geometric sizes, and their design features, as well as to perform radiation surveys. Equivalent gamma dose rates were measured along the depth of the boreholes using a certified UIM-2-2 instru- ment with a BDMG-100 detector. Collimated detectors were used to measure distribution of specific activities of gamma-emitting radionuclides. Visual inspec- tion of the boreholes was performed with a specially developed compact video camera, and the signal was recorded on a computer. Built-up ground was removed from the facility roofs and roof openings using conventional construction machines equipped with the necessary attachments, de- pending on the type of work. A truck crane was used to lift easily removable roof slabs. Cast–in situ concrete roofs were destroyed using EK-12 and EK-270 ex- cavators equipped with hydraulic hammers. In individual cases where a roof was found to be made of thick monolithic concrete (for example, during disposition of Repository No. 2), the roof was destroyed with a device for electric-discharge demolition of concrete structures. This technique uses electric-discharge energy released in condensed media in a plasma channel as a high-power current pulse passes through the channel. To accomplish this, shot holes were drilled in the monolithic concrete and filled with water, after which an electric discharge was produced that broke the in situ concrete slab into fragments. In addition to this remediation work, dust-suppression techniques were applied and aerosol activity in the atmosphere in the working area was monitored. The presence of high-level radioactive waste in Repository No. 4 required construction of additional radiation shielding around it. Radiation calculations were performed for several shielding designs, taking into account the particular characteristics of the repository location, geometrical parameters of the space to   Smirnov, V. P., Ye. G. Krastelev, V. M. Nistratov, et al. 1999. Development and application of a mobile facility for electro-discharge destruction of rock and structures. Mining Journal 11:56-58. 

104 CLEANING UP SITES CONTAMINATED WITH RADIOACTIVE MATERIALS be opened, and approximate ratios between specific activities of the major dose- producing radionuclides, cobalt-60 and cesium-137. For operations at Storage Facility No. 4, radiation shielding was designed and constructed based on the calculation results. The roof consisted of 6-m-long, 20-cm-thick paving slabs resting on outer support walls built of foundation blocks measuring 2,400 × 600 × 400 mm reinforced with metal trusses. A labyrinth was built to allow robots to pass into the shielding while preventing ionizing radiation from streaming out beyond the shielded area. The radioactive waste was extracted from the old facilities using conven- tional wheeled and crawler construction machines as well as Swedish Brokk-110 and Brokk-330 robots. Low-level waste was extracted using front loaders. The robots removed intermediate-level waste and fragments of high-level waste. To protect operators against ionizing radiation, construction equipment cabs were shielded with lead sheets and provided with protective leaded glass. Both construction equipment and robots were equipped with collimated detectors to measure the activity of the radioactive waste being extracted. During operations at the high-level waste storage facility, color video moni- toring cameras were installed inside the radiation-shielding structure. Their sig- nals were received by monitors located in the excavator cabs. To warn personnel about radiation hazards, working areas were equipped with threshold collimated detectors that produced audible and light alarms when the allowed gamma dose rate level was exceeded. The concrete matrix of the facility was gradually destroyed through the open- ing in the radiation shielding by an excavator with a hydraulic hammer located on the shielding roof. Fragments of the broken waste mass that contained low- and intermediate-level waste and concrete spalls were removed from under the shielding structure by a front loader and placed into metal or reinforced concrete containers. The container type to be used for loading the waste was selected based on measurements taken by a collimated detector mounted on the top edge of the shielding structure. To detect canisters of high-level waste or their fragments in the destroyed concrete mass, operators used a gamma camera that transmitted its signal to a monitor. When canisters or high-level waste fragments were detected, the shield- ing roof was put in place completely, and robots then extracted, measured, and packaged the high-level waste inside the shielding structure without any person- nel being exposed to high radiation fields. The robots destroyed the remains of the waste concrete matrix, gripped the detected high-level waste, and removed it   Volkov, V. G., A. G. Volkovich, A. S. Danilovich, O. P. Ivanov, S. V. Smirnov, and V. Ye. Stepanov. 2005. Application of new instruments for radioactive waste sorting in remediation activi- ties at RRC Kurchatov Institute. Pp. 135-141 in Proceedings of the Eighth International Conference “Nuclear Technology Safety: Economy and Management of Ionizing Radiation Sources,” September 26-30, 2005, St. Petersburg, Russia. 

REMEDIATION OF CONTAMINATED FACILITIES 105 to a special sorting area arranged inside the shielding structure. Inside the shadow shielding, the robots performed all operations required to extract and cut up the waste and then pack it into containers. The gamma camera display was used to direct the robots to high-level radiation sources and monitor the extraction, frag- mentation, and packaging operations. High-level waste was removed from under the shadow shielding and packed in metal containers with concrete inserts that were then grouted in concrete at the top. Before shipping the filled containers to the Radon enterprise, gamma levels near the containers were checked to ensure that they met regulatory requirements. DISPOSITION OF THE OLD REPOSITORIES All radioactive waste management operations involved continuous radiation monitoring, including measurement of activity levels of the waste loaded into containers, as well as continuous monitoring of working areas, personnel, and the volume of activity of radionuclides in the working area. Activity levels of radioactive waste in containers were measured with spectrometric and current- collimated detectors. Measurement results were further processed with custom- developed software, taking into account the container geometry, thickness, and material; the waste-packaging density; and the ratio between activities of the basic radionuclides (cobalt-60 and cesium-137) found in the radioactive waste.  The measurements were made according to approved procedures using certified collimated detectors to monitor specific activity. All dosimetry instruments used for personal and overall radiation monitoring had been registered and entered in the Register of Instruments. For spectrometric and radiometric measurements of waste, soil, groundwater, and air samples per- formed in laboratory conditions, certified instruments and qualified procedures were used. In addition to working areas, the radiation situation was monitored at the entire radioactive waste disposal site. During operations on Storage Facility No. 4, the radiation situation in working areas and at the entire disposal site was monitored with two gamma locators that measured the ionizing radiation photon flux, taking into account its spectral characteristics. One of the gamma locators was used for continuous monitoring of changes   Volkov, V. G., V. N. Potapov, O. P. Ivanov, S. M. Ignatov, N. K. Kononov, and V. Ye. Stepanov. New radiation monitoring instruments and systems and their use in remediation operations at the RRC Kurchatov Institute radioactive waste disposal site. Pp. 371-378 in Proceedings of the Seventh International Conference “Nuclear Technology Safety: RW Management,” September 27-October 1, 2004, St. Petersburg, Russia. St. Petersburg: ProAtom Publishing House.   Volkov, V. G., A. G. Volkovich, A. S. Danilovich, O. P. Ivanov, S. V. Smirnov, and V. Ye. Stepanov. 2005. Application of new instruments for radioactive waste sorting in remediation activi- ties at RRC Kurchatov Institute. Pp. 135-141 in Proceedings of the Eighth International Conference “Nuclear Technology Safety: Economy and Management of Ionizing Radiation Sources,” September 26-30, 2005, St. Petersburg, Russia. 

106 CLEANING UP SITES CONTAMINATED WITH RADIOACTIVE MATERIALS in the radiation situation in working areas at Storage Facility No. 4, with the measurements displayed on a computer screen via the Internet. The other gamma locator scanned the entire disposal site, measuring gamma spectra from individual areas. Photon flux values measured at the gamma locator position were normal- ized to the gamma exposure dose rate (EDR) value measured by an integral dosimeter mounted on a rotator of one of the gamma locators. Photon flux distri- butions obtained were further used to calculate gamma EDR values in all points of the scanned space, with these data presented as a color palette superimposed on a coordinate image of the scanned object. During remediation operations, the volume activity of aerosols in the work- ing area air was monitored during each shift using UDA-1AB devices and PVP- 4A samplers and at the site perimeter 24 hours a day using Typhoon-type devices. These devices continuously measured the volume of activity of alpha- and beta- active aerosols in the air throughout the operations. During the measurements, the systems were typically located around the working areas about 3-5 m apart, with air samples taken at the height of about 1 m above the ground. The samplers independently monitored the level of air contamination by radioactive aerosol mixtures of various compositions. Air samples were taken while pumping a controlled volume of air through a special filter that was sub- sequently delivered for spectrometric and radiochemical analysis under labora- tory conditions after the sample was taken. The samplers were equipped with a self-contained power supply that allowed air samples to be taken in remote, hard-to-reach locations where the main power supply is inaccessible. The control interval and air sampling points were determined depending on the kind of work being performed, locations of dust-producing activity sources in working areas, and personnel workplaces. The samplers were placed about 1 m from the dust- generating activity sources at an elevation of about 0.5 m above ground level. During operations on Storage Facility No. 4, the samplers were installed not only at personnel locations but also at locations of robotic operations to further evaluate the dust rise and dust-suppression efficiency directly in working areas. In addition to monitoring the overall radiation situation, site managers also moni- tored groundwater activity using a network of observation boreholes located at the radioactive waste disposal site and in the adjoining area. APPLICATION OF DUST-SUPPRESSION TECHNOLOGIES To ensure radiation safety for personnel involved in the work and prevent formation and transfer of radioactive aerosols, all operations involving radioac- tive waste extraction and disposition of the old storage facilities were performed using dust-suppression technologies. Various localizing, isolating, and dust- suppressing protective coatings based on polymeric compounds were used in the operations. These polymeric coatings were capable of preventing the spread of radioactive contamination in the form of dispersion aerosols into the environ- 

REMEDIATION OF CONTAMINATED FACILITIES 107 ment. The protective polymeric coatings were applied to dust-prone surfaces using airless and airstream spraying techniques. The efficiency of using these compounds was controlled by measuring (1) the surface activity of radionuclides in smears taken from the external surface of the coatings applied to the surfaces to be protected and (2) the volume activity of aerosols in the working area air. DECONTAMINATION OF RADIOACTIVE SOIL AND METAL RADIOACTIVE WASTE Disposition of the radioactive waste repositories requires solving problems of how to decontaminate large volumes of radioactive soil and establish a system for post-remediation monitoring of the areas involved. Estimates of radioactive soil volumes suggested that high-performance (2-3 metric tons per hour) soil- cleaning technologies were required. The search for such technologies, which reviewed chemical, electrokinetic, and other soil decontamination methods, indi- cated that they would have to be developed during the remediation operations.10 Two radioactive soil treatment technologies were selected based on performance estimates: dry radiometric separation and water-gravity separation. 11 Further ef- forts were devoted to develop these two technologies and bring the performance of the facilities built around these technologies up to required levels. According to the results of preliminary studies performed by the Bochvar Institute, more than 80 percent of radionuclides in the contaminated soil are accu- mulated in fine sludge or clay fractions or both. Therefore, the technology of wet decontamination of soil consisting of water-gravity separation and classification of the contaminated soil into size fractions followed by segregation and removal of the fine fraction was adopted for soil decontamination. Based on laboratory findings, the optimum approach is to separate contaminated soil into the follow- ing four fractions by size: fraction 1 (lump), greater than 100 mm in size; fraction 2 (coarse), from 3 to 100 mm; fraction 3 (sand), from 0.1 to 3 mm; and fraction 4 (fine or sludge), up to 0.1 mm. A pilot facility for wet decontamination of soil was developed around this technology in cooperation with the Bochvar Institute and the Chemical Technol- ogy Institute. The basic equipment for this pilot facility was fabricated at the Gormashexport enterprise in Novosibirsk. It has a modular design consisting of the three separate components for disintegration, classification, and thicken- ing. This pilot facility for water-gravity separation of contaminated soil was installed at the radioactive waste disposal site. Operation of the pilot facility in 10  Dmitriev, S. A., L. B. Prozorov, M. Yu. Shcheglov, et al. 2001. Electrokinetic method of cleaning radionuclides from soil. Radiation Safety Issues 1:42-49. 11  Volkov, V. G., Yu. A. Zverkov, S. M. Koltyshev, et al. 2005. Main results of start-up and trial operation of the soil decontamination facility. Pp. 120-135 in Proceedings of the Eighth International Conference “Nuclear Technology Safety: Economy and Management of Ionizing Radiation Sources,” September 26-30, 2005, St. Petersburg, Russia. 

108 CLEANING UP SITES CONTAMINATED WITH RADIOACTIVE MATERIALS the start-up and adjustment modes has demonstrated its rather high efficiency. The specific activity of 70-80 percent of the initial soil is reduced by four to five times. On average, from 180 to 200 kg of each metric ton of initial soil processed are removed for long-term storage. Recovered water used in processing remains virtually uncontaminated throughout several facility operation cycles. At present, more than 2,000 m3 of radioactive soil have been processed and about 400 m3 have been shipped to the Radon enterprise for long-term storage. Of the radioactive waste extracted from the repositories, up to 30 percent is composed of contaminated metal and metal structures, with these materials primarily classified as low-level waste. Hydroabrasive technology is used to decontaminate the metal waste. A hydroabrasive cutting and decontamination facility was deployed in a specially equipped area. The objects to be processed are placed on reinforced concrete plates resting directly on the metal floor. The hydroabrasive decontamination equipment consists of a high-pressure pump, high-pressure piping, a bin, and a feeder supplying loose abrasive to a hydroabra- sive nozzle. The nozzle may be used either as a cutter or as a sprayer to remove an upper contaminated layer from metal surfaces. The decontamination is performed both manually and using automated tools. In the manual mode, cleaning is accomplished using an operating water pressure of 1,500 atm at an abrasive sand flow rate of about 1 kg per minute. In the au- tomated mode, decontamination is performed by scanning metal surfaces with a water-abrasive jet with a linear velocity of 500 mm per minute using a remotely controlled cart. The operating water pressure used for cleaning is 2,000 atm with the nozzle diameter being 0.6 mm. The hydroabrasive equipment demonstrated good efficiency for low-level metal waste decontamination. However, as the level of surface contamination increases, decontamination costs increase because of higher consumption of abrasive material and, consequently, increased quantities of secondary waste pro- duced in the decontamination process (abrasive, wastewater, and so forth). Cost estimates allowed the most cost-effective decontamination levels to be selected in order to make the decontamination process economically sound. About 250 met- ric tons of metal radioactive waste was decontaminated using the hydroabrasive equipment and then shipped to the Ecomet-S enterprise for remelting. MAIN RESULTS OF THE WORK All 10 old radioactive waste repositories subject to remediation at the waste disposal site have been decommissioned and eliminated. About 3,400 m3 of ra- dioactive waste with a total activity of more than 1.4 × 1013 Bq (approximately 380 Ci) was extracted from the repositories, and more than 3,000 m3 was moved to Radon for long-term storage. During 2005 the gamma dose rate at the radioactive waste disposal site pe- rimeter varied between 0.21 and 0.67 µSv per hour and at the institute perimeter, 

REMEDIATION OF CONTAMINATED FACILITIES 109 TABLE 15-2  Personnel Exposure During Remediation Operations (2003-2006) Parameter 2003 2004 2005 2006 Mean individual dose, mSv 2.0 1.44 1.8 1.9 Collective dose, Man × Sv 0.041 0.080 0.090 0.079 between 0.08 and 0.14 µSv per hour. The volume of activity in the air at the disposal site and within the overall Kurchatov Institute site was by three to five orders of magnitude less than allowable levels for urban populations. External radiation exposure of personnel by years is presented in Table 15-2. Thus, the effective organization of work, the equipment used for operations, and radiation-monitoring instruments allowed this repository containing high- level and other waste to be decommissioned and eliminated rather quickly while meeting radiation safety requirements for onsite personnel and the local popula- tion within a major city. Results of the 3-year remediation effort have demon- strated that the choice of low-, intermediate-, and high-level waste management technologies was correct. During 2005, Repository No. 4, where canisters of high-level waste were encased in strong concrete, was successfully decommis- sioned. Organizational and technical measures that were applied made it possible to complete the work in the shortest possible time in full compliance with all rules