In France, two options can be considered to handle the Very Low Level Waste (VLLW) and the Low Level Waste (LLW). The first one is the incineration at CENTRACO facility and the second one is the disposal at ANDRA sites. The waste acceptance in these two channels is dependent upon the adequacy between the waste characteristics (physical chemistry and radiological) and the channel specifications. If the waste characteristics and the channel specifications (presence of significant quantities of halogens, complexants agents, organic components… or/and high activity limits) are incompatible, an alternative solution have to be identify. It consists of a waste pre-treatment process. For Cadarache LOR (Liquides Organiques Radioactifs) waste streams, two radioactive scintillation cocktails have to be treated. They are composed of a mix of organic liquids and water: for the first one, 19 % of organic compounds (xylene, mesitylene, diphenyloxazole, TBP…) and 86.9 % of water, and for the second one, 23 % of organic compounds (TBP…) and 77 % of water. They contain halogens (chlorine and fluorine), complexants agents (nitrate, sulphate, oxalate and formate) and have got αβγ spectra with mass activities equal to some 100 Bq/g. Therefore, tritium is also present. As a consequence, in order for storage acceptance at the ANDRA site, it is necessary to pre-treat the waste. An adequate solution seems to be a solidification process using NOCHAR polymers. Indeed, NOCHAR polymers correspond to an important variety of products applied to the treatment of radioactive aqueous and organic liquids (solvent, oil, solvent/oil mixing …) and sludge through a mechanical and chemical solidification process. For Cadarache LOR, N910 and N960 respectively dedicated to the organic and aqueous liquids solidification are considered. With the N910, the organic waste solidification occurs in two steps. As the organic liquid travels moves through the polymer strands, the strands swell and immobilize the liquid. Then as the polymer-organic cure, over time, the polymer continues to collapse on the organic to create a permanent chemical bond. The N960 has the ability to absorb aqueous waste up to 100 times its own weight. It creates a strong mechanical bond which permanently traps the contamination imbedded in the aqueous liquids. Therefore, these two NOCHAR polymers seem to be able to constitute a suitable solidification matrix for a final acceptance in storage on ANDRA sites. In order to validate the solidification process using NOCHAR polymers as an acceptable solution for Cadarache LOR treatment, some solidification tests realised with N910 and N960, have been carried out for different Waste/Polymer ratios. To determine the best Waste/Polymer ratio and the optimal experimental parameters, exudation tests have been made. Indeed, the process prevents leaching and it results in the absence residual free organic or aqueous liquid which is forbidden in storage by ANDRA specifications. With these results, the obtained scientific data constitute a fundamental basis of an ANDRA agreement. As a conclusion, the aim of this study is to demonstrate that the pre-treatment by solidification using NOCHAR polymers can constitute a solution for Cadarache LOR handling and more generally, for various organic and mixed organic/aqueous waste which can not be directly acceptable at CENTRACO facility or at ANDRA storage sites. So then, this study is a solid background to demonstrate the feasibility of the waste pre-treatment by solidification with NOCHAR polymer and to encourage this process development.

Based on the decision of the State Office for Nuclear Safety, the Dukovany NPP has been obliged to secure the efficient capacities for the disposal of spent ion exchange resins. Therefore, in September 2010, based on the contract with supplier company AMEC Nuclear Slovakia s.r.o. has begun with pumping and treatment of ion exchange resins from the storage tank 0TW30B02, situated in the auxiliary building. The SIAL ® technology, developed in AMEC Nuclear Slovakia, has been used for the solidification purposes. This technology allows an on-site treatment of various special radioactive waste streams (resins, sludge, sludge/resins and borates) at the room temperature. The SIAL ® matrix and technology were licensed by the Czech State Office for Nuclear Safety in 2007. On-site treatment and solidification of spent ion exchange resins at Dukovany NPP involves process of resin removal from tank using remotely operated manipulator, resin transportation, resin separation from free water, resin filling into 200 dm 3 drums and solidification into SIAL ® matrix in 200 dm 3 drums using the FIZA S 200 facility. The final product is observed for compressive strength, leachability, radionuclide composition, dose rate, solids and total weight. After meeting the requirements for final disposal and consolidation, the drums are being transported for the final disposal to the Repository at Dukovany site. During the 3 month’s trial operation in 2010, and the normal operation in 2011 and 2012, 189 tons of dewatered resins have been treated into 1960 drums, with total activity higher than 920 GBq. At the end of trial run (2010), 22 tons of dewatered resins were treated into 235 drums. During standard operation approximately 91 tons in 960 drums (2011) and 76 tons in 765 drums (2012) were treated. The weights of resins in the drum ware in the range from 89–106 kg and compressive strength limit (10 MPa) has already been achieved 24 hours after fixation. The final measured strength values ranged from 19.0–34.7 MPa and real leachability values ranged from 0.03–0.65%, far below the 4% limit value. Collective effective dose of all workers in 2012 was 7.7 mSv (12.6 mSv in 2011, 6.2 mSv in 2010). Average individual effective dose in 2012 was 0.55 mSv (14 workers), and maximal individual effective dose was 2.25 mSv. This approach allows fast, safe and cost effective immobilization and transformation of dangerous radioactive waste such as sludge and resins into the solid form, which is suitable for long term storage or disposal.

In the frame of the decommissioning of nuclear power plants or laboratories, the penetration depth of the contamination in concrete surfaces: walls, floor, is unknown. Its knowledge requires sample analysis, that is time consuming and expensive. The main goal of the work is to define and evaluate a non-destructive measurement technique for the evaluation of the contamination depth in concrete. Estimation of accuracy of measurements for different contamination levels, time of measurement, value of natural radionuclides (NRN) concentration in concrete and background radiation dose were carried out. The type of relevant detector depends of selected limiting sensitivity, the weight and sizes of the device.

When out of normal material or very high activity is encountered in a Deactivation & Decommissioning (D&D) project, special considerations are needed. As the source term increases, the level of attention to planning, project cost and risk management must increase as well. WorleyParsons (WP) is engaged in identifying, screening and documenting (with follow on project design expected) the various remediation methods available to remove highly contaminated soils discovered under a hot cell. Soil dose rates have been measured in the 70 to 100 Sieverts/hour range, with radionuclide mix to be defined in mid to late summer 2011 from samples collected in June 2011. The scope of this project is to identify remedial techniques and removal methods (means and methods) and develop a set of technically implementable remedial alternatives (alternatives). These remedial alternatives will be screened by a Technical Panel consisting of WP and client experts using a WP defined screening process to narrow the list of alternatives. Transparency and un-biased choice of actions is key to owner and public confidence in the process and the choice of the ultimate path forward. The poster will show Phase 1 — how potential remedial techniques and removal methods (means and methods) were first identified, catalogued and initial screened. Following the initial screening, the remaining means and methods are combined to form an alternative — a stabilization means with a removal method are matched with a waste container to from an alternative. A rough order of magnitude (ROM) cost estimate and high level schedule are also developed for each alternative. In parallel with the above, screening criteria with associated design attributes were identified and given a weight factor and attribute value, which are then multiplied together and used to develop a decision matrix for the developed alternatives. Two parallel processes — screening criteria development and alternative development — are being pursued to allow better focus on each while not prejudicing the final alternative selection. Phase 2 actions — the expert panel - grade each alternative against the pre-approved screening criteria using the decision matrix that compiles results — real time but blind to the panel. The top two to four scores from this last effort will proceed to a pre-conceptual, conceptual design phase and the expert panel reconvened to make the final decision on the remediation methods to be used for removing the soil / contamination.

Since 1979, Uranium enrichment technology has been researched through the gas centrifuge method, at Ningyo-toge Environmental Engineering Center of Japan Atomic Energy Agency (JAEA). In addition, the Demonstration Plant, that is final stage test facilities, was operating continuously from 1988 to 2001. As a result, a lot of residues accumulated in the plant. Most of this accumulation was found be uranium intermediate fluoride. The basic decommission policy of JAEA is that equipments of gas centrifuge will be decontaminated by sulfuric acid immersion method for clearance and reuse. In our plan, approximately 90% of metals will be cleared and reused, and then the remaining 10% will be disposed of radioactive waste. We propose a combination of sulfuric acid immersion method and the systematic chemical decontamination as an efficient method for decontamination of uranium enrichment facilities. This paper focuses on the method and performance of systematic chemical decontamination using IF 7 gas. The following (Figure 1) shows our decommission policy and position of systematic chemical decontamination by IF 7 gas for uranium enrichment plant. The IF 7 treatment technique belongs to the systematic decontamination technology. It has the high performance decontamination technique for the plant that accumulates the uranium intermediate fluoride, such as UF 4 , UF 5 , U 2 F 9 , and U 4 F 17 , which exist in the uranium enrichment plant through the Gas Centrifuge, called GCF. The one of characteristics of the IF 7 treatment, the secondary waste is just an IF 5 and little residues. In addition, this IF 5 can be reused as materials for making new IF 7 gas. The IF 7 treatment can also be performed in the room temperature and very low pressure like a 10–45hPa. Furthermore, the IF 7 treatment is a simple method using chemical reaction. For this reason, we hardly need to care about secondary reaction with the exception of the reaction with IF 7 gas and the uranium intermediate fluoride. This is a very important feature when applying to a large-scale plant. In order to carry out the IF 7 treatment, we only set up a few equipments in GCF uranium enrichment plant, which were IF 7 feeding equipment and two circulating pumps. IF 7 gas cylinders are seated in IF 7 feeding equipment. This is the only equipment. Figure 2 shows the IF 7 treatment system. We carried out the IF 7 treatment for the four cascades in the uranium enrichment Demonstration Plant. The weights of uranium residue in the cascades were approximately 700kgU per cascade prior to the IF 7 treatment. In the IF 7 treatment, we were able to find the near-optimal processing condition. As a result, we could confirm the IF 7 treatment period for one cascade which was 60 days. The main factor to determine the IF 7 treatment period is the pressure and the flow rate of reaction product gas (UF 6 and IF 5 mixture gas) exhausted from the cascade. Although we carried out the IF 7 treatment with the maximum value of the flow rate, which our facility has, it is possible to further shorten the IF 7 treatment period by setting a higher gas flow rate. Moreover, after the IF 7 treatment, we evaluated the uranium recovery rate for cascades and the residues’ uranium weight in the main equipment of GCF. In addition, in the evaluation of the uranium recovery rate, we enable to confirm the uranium recovery rate of all cascades achieved more tan 98%. Furthermore, the average of uranium recovery rate more than 99% in the cascade that has been processed at the end. As a result, radioactive concentration of uranium in the main parts of the GCF fell to 1.0B q/g and below.

Several different types of detectors, software, and hardware designs are employed in instruments used to monitor radioactive content of freight shipments in order to detect illicit trafficking of nuclear materials. The instruments can be container monitors, hand-held radiation detectors, mobile analysis systems, or fixed radiation portal monitors. However, within the various groupings (e.g., portal monitors), all instruments are expected to solve the same problem, that is, to detect and identify any radioactive material present according to the prescribed investigative procedure or CONOPS. The best way to compare the performance of different instruments is with a numerical score or Figure of Merit (FOM). The FOM must quantify the performance of the instrument with respect to true positives (TP), false positives (FP), and false negatives (FN). The minimization of FN for certain radionuclides (e.g., uranium and plutonium or other Special Nuclear Material (SNM)) is more important than the minimization of FN for non-threat nuclides (e.g., low level NORM). In a similar way, the minimization of FP for SNM is more important than falsely reporting the common NORM nuclides which represent no threat. System performance depends on the measurement detail (e.g., transit speed through a portal, measurement time in general, amount and distribution of NORM), therefore the test conditions should also be included in the statement of the FOM. The FOM is expected to vary significantly with the above measurement details. A FOM has been developed based on the number of true positives (TP), the number of false important positives (FIP), the number of FP, the number of true positives for SNM (TPSNM), and the number of false positives for SNM (FPSNM). This formula rates the overall performance with extra weight given to FP and FN for SNM. Examples will be shown for testing of portal monitors.

This paper describes the design and operation of the Combined Technology Automated Waste Characterisation System (CTA-WCS) at JRC Ispra. The WCS was designed for the measurement of fission products and uranium and plutonium containing waste arising from nuclear fuel and nuclear materials processing and reactor operations. The WCS covers a range of activity including Low and Intermediate Level Waste (LLW and ILW). The system is designed to measure the waste in 200 and 400 (440) litre drums with a maximum drum weight of 1500 kg. Gamma-ray measurements of radio-nuclide content are performed by a gamma ray measurement station which functions as either a Segmented Gamma Scanner (SGS) or Tomographic Segmented Gamma-ray Scanner (TSGS). Either of these two techniques may be employed to perform the functions of drum screening, non-destructive examination (NDE) and, where appropriate, final drum assay. Coupled to the gamma ray station is a surface dose-rate measurement station, which employs 6-shielded Geiger-Muller detectors. Active and passive neutron measurements are performed by an advanced, graphite lined Differential Die-away (DDA) system, which comprises the neutron measurement station. The DDA performs conventional passive neutron totals, coincidence, and multiplicity counting and active DDA total neutron counting. Data analysis is based on the use of a range of matrix calibrations, some determined by Monte Carlo analysis. Linking the gamma ray and neutron measurement stations is an automated roller conveyor with a 20 drum buffer capability and a weight measurement station. Drums are identified by bar code reading technology. Once loaded, the system performs automatic assay of up to 20 drums and then returns the drums to the buffer position on the conveyor. The first WCS of this type was supplied to the European Commission at the Joint Research Centre (JRC) Ispra in northern Italy and it was commissioned at the end of 2007. It is now in a phase of pilot operation. Results will be presented from the first drum measurement campaign.

The shotclay method is construction method of the high density bentonite engineered barrier by spraying method. Using this method, the dry density of 1.6 Mg/m 3 , which was considered impossible with the spray method, is achieved. In this study, the applicability of the shotclay method to HLW bentonite-engineered barriers was confirmed experimentally. In the tests, an actual scale vertical-type HLW bentonite-engineered barrier was constructed. This was a bentonite-engineered barrier with a diameter of 2.22 m and a height of 3.13 m. The material used was bentonite with 30% silica sand, and water content was adjusted by mixing chilled bentonite with powdered ice before thawing. Work progress was 11.2 m 3 and the weight was 21.7 Mg. The dry density of the entire buffer was 1.62 Mg/m 3 , and construction time was approximately 8 hours per unit. After the formworks were removed, the core and block of the actual scale HLW bentonite-engineered barrier were sampled to confirm homogeneity. As a result, homogeneity was confirmed, and no gaps were observed between the formwork and the buffer material and between the simulated waste and the buffer material. The applicability to HLW of the shotclay method has been confirmed through this examination.

As one of the repository operation technologies for high-level radioactive waste (HLW), the pre-fabricated engineered barrier system module (PEM) was carried out the examination of handling and emplacement technique for engineered barrier system (EBS). The PEM technology was examined to confirm technological applicability. The PEM is concept of the integration of EBS as the module in the surface facilities, and transporting the module underground facilities. This concept is the one of the candidate concepts of horizontal emplacement techniques for EBS in Japan. Therefore, PEM is the same level large size and heavy weight as EBS, and it is necessary to examine the applicability of handling and emplacement techniques. Full-scale level tests were performed to confirm the applicability of these techniques with the air bearing/air jack devices. In the tests, we prepared the testing devices of full-scale level size/weight and confirmed the applicability of these technologies as an elemental technology on the condition of considering the environment of an underground tunnel. The air bearing test that produced the surface-roughness of the tunnel environment was carried out the evaluation concerning the transportation performance of the air bearing. And, the air jack test was carried out the holding and emplacement of PEM. The repository operation technology with the air bearing/jack device was confirmed to execute the examination, and to apply to handling and emplacement technique for PEM.

In the Fugen nuclear power plant (FUGEN), the dismantling of equipments in the turbine building has started in 2008, and the dismantling of equipments around the reactor is scheduled around in 2015. To evaluate the management data on this dismantling of equipments around reactor appropriately, it is very important to study whether the conventional evaluation models have the applicability for FUGEN or not. Thus, the management data on the dismantling of equipments in 3rd/4th feedwater heater room conducted in 2008 was calculated with the conventional evaluation models. The conventional evaluation models were made by data obtained from Japan Power Demonstration Reactor (JPDR) decommissioning program. It was found that there were large differences between the calculated values and the actual data. For finding the cause in the difference between them, the dismantling of equipments in 3rd/4th feedwater heater room was divided into three processes: i) the preparation process, ii) the dismantling process, and iii) the clean-up process. In the both process of preparation and clean-up, the calculated values were smaller than the actual data. These were mostly caused by the plant scale difference between JPDR and FUGEN, because the conventional evaluation models were built by analyzing the actual data on the decommissioning of JPDR which is smaller than FUGEN. The primary expression depending on the area of working space was built as new evaluation models for the preparation and clean-up processes. In the dismantling process, on the other hand, it was found that there were characteristic differences in the dismantling of feedwater heater as follows: 1) the calculated values were significantly larger than the actual data, 2) the actual data for the dismantling of 3rd feedwater heater was larger than that of 4th one, though these equipments were almost same weight. It was found that these were brought by the difference in the descriptions of dismantling of feedwater heaters, and the new evaluation models reflecting the descriptions of dismantling were built for the appropriate evaluation of the management data. The calculated values with the new evaluation models for each process showed the good agreement with the actual data. In this report, study on evaluation models of management data for dismantling of equipments in the feedwater heater room will be described.

Fuji Electric had developed the low pressure oxygen plasma technology for mild decomposition and mineralization of an organic material such as ion exchange resin. This method is suitable for radioactive spent resin volume/weight reduction and stabilization for final disposal. On this process, the ion-exchange resins are vaporized and decomposed into gas-phase with pyrolysis, and then, they are decomposed and oxidized with low-pressure plasma activity based on oxygen. And this process is achieved under moderate condition for radio active waste. • incinerate temperature: 400–700 deg C; • low-pressure (low-temperature) plasma condition: 10–50 Pa. From the result of this process, named of LPOP(low pressure oxidation process) by the inductively coupled plasma, we have confirmed that the process is applicable for organic fireproof waste including ion-exchange resin, and found that the used resin treatment performance is the same as cold test (using imitate spent resin) [1] [2] [3]. In this paper, the outline of the LPOP technology, and two research results on the possibility of solidification with cement of LPOP residue for geological disposes are reported. (1)Study of the residue chemical form after LPOP process (2)Study of the solidification character with cement.

NUKEM Technologies was contracted to supply a dry, automated drum belt (tumbling) Abrasive Blasting Unit (ABU) to the Joint Research Centre of the European Commission in Ispra, Italy. The ABU was installed in the centralised radioactive waste management area of the JRC-Ispra site in Italy. The unit is to be employed for the decontamination to clearance levels of slightly contaminated metal components and, where practical, concrete or heavy concrete (density ∼3200 kg/m 3 ) blocks arising from the dismantling of nuclear facilities. The presentation is based on the successful construction and installation of the ABU at the JRC Ispra site. Among the several possibilities of adapting conventional abrasive units to nuclear applications, an automatic tumbling machine was preferred, due to the larger output and (mainly) for the ease of operation, with minimum direct handling of contaminated material by operators, thus satisfying the ALARA principle. Consideration was also given to Belgoprocess’ successful experience with a predecessor, similar unit. After adequate size reduction batches of up to about 800 kg of material to be decontaminated are automatically introduced into the blasting chamber. Pieces between 100 mm and 800 mm long, between 100 mm and 500 mm wide and between 5 mm and 300 mm high can be effectively treated in the unit, the maximum weight of a single piece being limited to 100 kg. Short lengths of pipe may be included; the final dimensions of pipe to be decontaminated will be established during the nuclear commissioning tests. Other components with hard-to-reach surfaces may also be included. The content of the chamber is tumbled by two bladed drums, while sharp steel grit is sprayed onto the contaminated components, thus removing the surface layer including any contamination. From experience, 30 minutes of treatment is sufficient to remove contamination to levels below expected clearance levels for most materials. The decontaminated components are removed from the blasting chamber automatically and collected in skips. Dust and grit are led to a series of separators; the grit gets recycled to the blasting chamber, cleaned off contaminants such as paint are fed to collection bins, and the dust is bagged into waste drums. Airflow through the whole system cleans the decontaminated components, transports the dust to the collecting area, and acts as a dynamic barrier to limit risks of contamination of the surrounding areas. Prior to release back into the room, the air is filtered in a series of automatically cleaned filters, followed by HEPA filters. The whole facility is operated in an automatic mode: the operators are only required to place drums or pallets of contaminated material onto the feeder, and remove skips of decontaminated material and drums of secondary waste such as dust. The presentation will describe the system and potential applications in the nuclear industry in detail.

The removal of surface contamination is a major challenge for nearly all nuclear facilities undergoing, or awaiting, decommissioning. Conventional means of surface decontamination can expose workers to unnecessary hazards, and are often not fit-for-purpose due to size constraints or weight restrictions. Additionally, conventional methods are not always easily deployed remotely due to their complexity or required services. The use of ultra high pressure water for surface decontamination, known as hydrolasing, is recognized as a technology which can be used in various applications requiring surface removal. Hydrolasing is an advantageous technology for many reasons including its versatility, overall simplicity and relative ease of remote deployment. For the nuclear industry, one of the largest challenges with regards to the use of hydrolasing is the requirement for the full recovery of the injected water and removed solids. For non-nuclear applications, there is often no requirement for recovery of the liquid and solid waste, which has led to few system designs which will recover the waste in full. S.A. Robotics’ experience with the deployment of ultra high pressure water systems for nuclear applications has shown that full recovery of injected water and removed solids is achievable in both underwater and in-air applications. Innovative equipment and system design have allowed S.A. Robotics’ hydrolasing systems to achieve near 100% solid and liquid recovery during concrete hydrolasing. This technology has been deployed for Fluor Hanford at Hanford’s K-Basins, as well as for UKAEA as part of the Windscale Piles decommissioning project. The purpose of this paper is to provide a short description of the hydrolasing process and the associated waste issues, describe the unique design features of S.A. Robotics’ hydrolasing systems which combat these issues, and provide an overview of two of the hydrolasing projects that S.A. Robotics has completed.

Radioactive sources are used for a variety of purposes, e.g. in medical treatment and diagnosis, research applications, measurement, testing, detection and calibration in industry, educational activities in colleges and universities etc. As part of its mission, ONDRAF/NIRAS, the Belgian Radioactive Waste Management Agency, draws up an inventory of all radioactive substances and nuclear installations on the Belgian territory. In recent years this inventory has been used to launch specific campaigns for the collection of different types of radioactive sources. In addition to this, the Royal Decree of 23 May 2006 concerning the transposition into Belgian law of the Spent High Activity Sealed Sources and the Management of Orphan Sources Directive of the EU (2003/122/EURATOM) has led to an increase in the number of requests addressed to ONDRAF/NIRAS for the collection of these types of radioactive waste and to an intensified collaboration between ONDRAF/NIRAS and the Belgian Safety Authority FANC/AFCN towards an effective management of orphan sources. Specific properties of these spent sources such as their activity, external dose rate, weight, size and/or their invalid special form certificate may complicate the transport and final treatment and conditioning of this type of waste and that is why these operations require careful attention. An overview of the radioactive sources already collected as radioactive waste or still present in the nuclear installations, different cases and problems encountered are presented in this paper, as well as the waste management options adopted by ONDRAF/NIRAS to deal with this type of waste.

The 15-MWel prototype pilot reactor AVR is a pebble bed HTGR. It was designed in the late 50s and was connected to the grid end of 1967. After 21y of successful operation the reactor was shut down end of 1988. In 1994 the first decommissioning license was granted and work with defueling, dismantling and preparation of a Safe Enclosure started. The primary system is contaminated with the fission products Sr90 and Cs137 and the activation products are Co60, C14 and H3. Due to the large amounts of Sr and Cs bound to graphite dust, the dismantling of systems connected to the pressure vessel is very tedious. In 2003 the AVR company was restructured and the strategy of the decommissioning was changed from safe enclosure to green field, i.e. the complete direct dismantling of all facilities and clean up of the site. The highlight during the dismantling is the removal of the reactor vessel (diameter ca. 7.6m and length ca. 26m) in one piece. Before handling the reactor vessel it will be filled with low density cellular concrete. Subsequently the reactor building will be cut open and the reactor vessel (total weight ca. 2100Mg) lifted out and transported to an interim store.

Spent ion exchanger media are considered to be problematic waste that, in many cases, requires special approaches and precautions during its immobilization to meet the acceptance criteria for disposal. The waste acceptance criteria define, among others, the quality of waste forms for disposal, and therefore will sometimes define appropriate treatment options. The selection of treatment options for spent ion exchange materials must consider their physical and chemical characteristics. Basically, the main methods for the treatment of spent organic ion exchange materials, following to pretreatment methods are: - Direct immobilization, producing a stable end product by using Cement, Bitumen, Polymer or High Integrity Containers, - The destruction of the organic compounds by using Thermochemical processes or Oxidation to produce an inorganic intermediate product that may or may not be further conditioned for storage and/or disposal, - The complete removal of the resin inner structural water by a thermal process. After a thorough technical economical analysis, Tractebel Engineering selected the Resin Hot Compaction Process to be installed at Tihange Nuclear Power Plant. The Resin Hot Compaction Process is used to make dense homogenous organic blocks from a wide range of particulate waste. In this process spent resins are first dewatered and dried to remove the inner structural water content. The drying takes place in a drying vessel that holds the contents of two 200 L drums (Figure). In the oil heated drying and mixing unit, the resins are heated to the necessary process temperature for the hot pressing step and then placed into special metal drums, which are automatically lidded and immediately transferred to a high force compactor. After high force compaction the pellets are transferred to a measuring unit, where the dose rate, height and weight are automatically measured and recorded. A volume reduction factor of approximately up to four (depending on the type of resins) is achievable using hot compaction techniques. This paper will describe the application of the Resin Hot Compaction Process, at Tihange NPP.

The radioactivity of 14 C of the graphite samples from the dismantled Korea Research Reactor 1&2 (the KRR-1&2) site was analyzed and proposed to be disposed of as a low level radioactive waste rather than self-disposed of. The graphite wastes, with a weight of seven tons, have been generated during the dismantling of a research reactor with a capacity of one MW from 1995 to 2006. The graphite was used as a moderator for the research reactor and so has been radioactivated by thermal neutron. It was thought that the graphite wastes mainly included a radioisotope of stable carbon, 14 C, a pure beta emitter with a half life of 5,730 years and with a maximum decay energy of 156 keV. Therefore, it has been requested to see whether the dismantled graphite radioactive wastes including 14 C can be self-disposed of or not. In the present study, the radioactivity of 14 C in the graphite sample used in the research reactor was analyzed by using a commercialized high temperature furnace and a Liquid Scintillation Counter (LSC). The combustion temperature of the furnace was five hundred degrees centigrade and especially the temperature in the catalyst region was eight hundred degrees centigrade. The recovery from the furnace was 95% for 14 C and the LSC had a quenching efficiency of approximately 66%. Carbosorb was used as a trapping solution for 14 C. The radioactivity of 14 C was measured by a LSC through the procedure of a pre-treatment such as the combustion of a sample in the temperature range of 500–800 degrees centigrade by a high temperature furnace, trapping of 14 C into carbosorb and cocktailing it with a scintillator. The radioactivity was analyzed to have a concentration with a value of much more than a domestic legal limit for a self-disposal. And an individual effective dose rate estimation was also carried out. Finally, it is suggested that the graphite wastes from the dismantled research reactor should be disposed of at a low level radioactive waste disposal site and monitored.

This paper describes Studsvik’s technical concept of LLW-treatment of large, retired components from nuclear installations in operation or in decommissioning. Many turbines, heat exchangers and other LLW components have been treated in Studsvik during the last 20 years. This also includes development of techniques and tools, especially our latest experience gained under the pilot project for treatment of one full size PWR steam generator from Ringhals NPP, Sweden. The ambition of this pilot project was to minimize the waste volumes for disposal and to maximize the material recycling. Another objective, respecting ALARA, was the successful minimization of the dose exposure to the personnel. The treatment concept for large, retired components comprises the whole sequence of preparations from road and sea transports and the management of the metallic LLW by segmentation, decontamination and sorting using specially devised tools and shielded treatment cell, to the decision criteria for recycling of the metals, radiological analyses and conditioning of the residual waste into the final packages suitable for customer-related disposal. For e.g. turbine rotors with their huge number of blades the crucial moments are segmentation techniques, thus cold segmentation is a preferred method to keep focus on minimization of volumes for secondary waste. Also a variety of decontamination techniques using blasting cabinet or blasting tumbling machines keeps secondary waste production to a minimum. The technical challenge of the treatment of more complicated components like steam generators also begins with the segmentation. A first step is the separation of the steam dome in order to dock the rest of the steam generator to a specially built treatment cell. Thereafter, the decontamination of the tube bundle is performed using a remotely controlled manipulator. After decontamination is concluded the cutting of the tubes as well as of the shell is performed in the same cell with remotely controlled tools. Some of the sections of steam dome shell or turbine shafts can be cleared directly for unconditional reuse without melting after decontamination and sampling program. Experience shows that the amount of material possible for clearance for unconditional use is between 95 – 97% for conventional metallic scrap. For components like turbines, heat exchangers or steam generators the recycling ratio can vary to about 80–85% of the initial weight.

The former production site of NUKEM where nuclear fuel-elements were developed and handled from 1958 to 1988 was situated in the centre of an industrial park for various activities of the chemical and metallurgical industry. The size of the industrially used part is about 300.000m 2 . Regulatory routine controls showed elevated CHC (Chlorinated Hydro-Carbons) values of the ground water at the beginning of the 1990’s in an area which represented about 80.000 m 2 down-gradient of locations where CHC compounds were stored and handled. Further investigations until 1998 proved that former activities on the NUKEM site, like the UF 6 conversion process, were of certain relevance. The fact that several measured values were above the threshold values made the remediation of the ground water mandatory. This was addressed in the permission given by the Ministry for Nuclear Installations and Environment of Hesse according to §7 of the German atomic law in October 2000 [1]. Ground water samples taken in an area of about 5.000 m 2 showed elevated values of total Uranium activity up to between 50 and 75 Bq/l in 2002. Furthermore in an area of another 20.000m 2 the samples were above threshold value. In this paper results of the remediation are presented. The actual alpha-activities of the ground waters of the remediation wells show values of 3 to 9Bq/l which are dominated by 80 to 90% U-234 activity. The mass-share of total Uranium for this nuclide amounts to 0,05% on average. The authority responsible for conventional water utilisation defined target values for remediation: 20μg/l for dissolved Uranium and 10μg/l for CHC [2]. Both values have not yet been reached for an area of about 10.000 m 2 . The remediation process by extracting water from four remediation wells has proved its efficiency by reduction of the starting concentrations by a factor of 3 to 6. Further pumping will be necessary especially in that area of the site where the contaminations were found later during soil remediation activities. Only two wells have been in operation since July 2002 when the remediation technique was installed and an apparatus for direct gamma-spectroscopic measurement of the accumulated activities on the adsorbers was qualified. Two further remediation wells have been in operation since August 2006, when the installed remediation technique was about to be doubled from a throughput of 5 m 3 /h to 10 m 3 /h. About 20.000 m 3 of ground water have been extracted since from these two wells and the decrease of their Uranium-concentrations behaves similar to that of the two other wells being extracted since the beginning of remediation. Both, total Uranium-concentrations and the weight-share of the nuclides U-234, U-235 and U-238 are measured by ICP-MS (Inductively Coupled Plasma – Mass Spectrometry) besides measurements of Uranium-Alpha-Activities in addition to the measurement of CHC components of which PCE (Perchlor-Ethene) is dominant in the contaminated area. CHC compounds are measured by GC (Gas Chromatography). Down-gradient naturally attenuated products are detected in various compositions. Overall 183.000m 3 of ground water have been extracted. Using a pump & treat method 11 kg Uranium have been collected on an ion-exchange material based on cellulose, containing almost 100 MBq U-235 activity, and almost 15 kg of CHC, essentially PCE, were collected on GAC (Granules of Activated Carbon). Less than 3% of the extracted Uranium have passed the adsorber-system of the remediation plant and were adsorbed by the sewage sludge of the industrial site’s waste water treatment. The monthly monitoring of 19 monitoring wells shows that an efficient artificial barrier was built up by the water extraction. The Uranium contamination of two ground water plumes has drastically been reduced by the used technique dependent on the amounts of extracted water. The concentration of the CHC contamination has changed depending on the location of temporal pumping. Thereby maximum availability of this contaminant for the remediation process is ensured. If locations with unchanged water quality are detected electrochemical parameters of the water or hydro-geologic data of the aquifer have to be taken into further consideration to improve the process of remediation.

The decommissioning of the Windscale Pile 1 reactor, fifty years after the 1957 fire, is one of the most technically challenging decommissioning projects in the UK, if not the world. This paper presents a summary of the 1957 Windscale Pile 1 accident, its unique challenges and a new technical approach developed to safely and efficiently decommission the two Windscale Pile Reactors. The reactors will be decommissioned using a top down approach that employs an array of light weight, carbon fiber, high payload robotic arms to remove the damaged fuel, the graphite core, activated metals and concrete. This relatively conventional decommissioning approach has been made possible by a recently completed technical assessment of reactor core fire and criticality risk which concluded that these types of events are not credible if relatively simple controls are applied. This paper presents an overview of the design, manufacture and testing of equipment to remove the estimated 15 tons of fire damaged fuel and isotopes from the Pile 1 reactor. The paper also discusses recently conducted characterization activities which have allowed for a refined waste estimate and conditioning strategy. These data and an innovative approach have resulted in a significant reduction in the estimated project cost and schedule.