Characterisation of radioactive waste from nuclear decommissioning Peter Ivanov Acoustics & Ionising Radiation National Physical Laboratory, UK IAEA WS RER9106/9009/01, Visaginas, Lithuania, 24-28 August 2015
Contents Introduction Characterization objectives Characterisation process The radionuclide inventory Methods and techniques for characterization Clearance and release from regulatory control Conclusions References
Introduction Radiological characterisation is a key stage of the nuclear decommissioning Starts with the transition from operation to decommissioning Provides vital information on radionuclide inventory, activities and distribution Characterisation stage Allows assessment of the radiological status of the facility Enables decision making related to further decommissioning steps Stages of decommissioning
Transition from operation to decommissioning Decommissioning related activities during the life cycle of an NPP
Transition from operation to decommissioning Major changes during the transition period OPERATIONAL Production oriented management objectives; Repetitive activities; Predominant nuclear and radiological risk; Access to high radiation areas unlikely or for a short time; DECOMMISSIONIONG Project completion oriented management objectives; One-off activities; Reduction of nuclear risk, significantly increased industrial risk; Access to high radiation areas for extended periods;
Transition from operation to decommissioning Major changes during the transition period OPERATIONAL Focus on functioning of systems; Low radiation/contamination levels relatively unimportant; Relatively stable isotopic composition; Routine radioanalytical measurements; DECOMMISSIONIONG Focus on management of material and radioactivity inventory; Low radiation/contamination levels important Isotopic composition changing with time; Larger number of analysis, broader radionuclide variety.
Radiological characterisation Characterisation objectives Initial characterisation Collect sufficient information Assess the radiological status of the facility Identify any problem areas Detailed characterisation Physical, chemical radiological conditions Activity calculations, sample analysis Filling information gaps Decommissioning strategy Dismantling procedures Radiological protection Waste classification
Characterisation objectives Application to decommissioning operations Radiological characterisation Radionuclide inventory RN activities and distribution Selecting decommissioning scenario Immediate dismantling Delayed dismantling Decommissioning operations Shielding Removal of equipment RAW Management Waste classification Clearance Free release
Characterisation Characterisation objectives Health and safety considerations Attention to H&S of the staff Helps dose and risk assessments Compliance with ALARA principle Identifies required protection Initial characterisation: conducted in cooperation with operational staff performed soon after shutdown helps identifying external and ingestion hazards provides basis for categorising of equipment and structures splitting the facility into radiation zones
Characterisation process Review of historical information Validation of calculated results and measured data Calculation methods Characterisation process Review and evaluation of the data In situ measurement, sampling and analysis Sampling and analysis plan
Characterisation process Review of historical information Provides valuable data of possible radiological conditions of the facility and can optimise the characterisation efforts May consist of records of contamination spills or other unusual events, and/or previous surveys and measurements as well as information provided by operational staff members Records of occupational exposures incurred during inspection, survey, maintenance and repair activities are relevant Records about the facility structural condition is important, particularly in the case of deferred dismantling, and structural surveys must be carried out in addition to radiological characterization Although historical information is valuable for preparing a characterization programme, some of the characterization effort should aim at confirming the validity and completeness of the historical data
Characterisation process Implementation of calculation methods Calculation of the induced activity in the reactor Estimation of radionuclide distribution At normal operation As a result of an accident By mobile contamination Validation by sampling and measurements
Characterisation process Preparation of sampling and analysis plan The sampling and analysis plan defines the quality of data necessary to achieve the characterization objectives. The plan should define the following: Types, numbers, sizes, locations and analyses of samples required; Instrument requirements; The radiation protection aspects or controls of the activity; Data reduction, validation and reporting requirements; Quality assurance (QA) requirements; Methodology to be employed when taking the samples and performing the analyses; Requirements for disposal of waste generated during sampling.
Characterisation process Data management A large quantity of data may be generated during characterization activities, roughly divided into three categories: Calculated data- particularly refer to radioactive content and related dose rates in components and structural materials within the influence of the neutron flux. The results of calculations are available in the form of outputs from the appropriate computer codes and can easily be processed. Data obtained from in situ measurements at various locations of the reactor. These data are usually collected in a monitoring programme, using manual or remote measurements of dose rates and/or contamination levels. Data may subsequently be recorded in a computer database. Data resulting from a sampling and analysis programme- detailed information on types and amounts of radionuclides present in the form of activation and/or contamination. The collection and analysis of samples is an expensive and dose intensive task. However, it is the most precise means of verifying the theoretical calculations, predicting future exposures and facilitating the selection of the most appropriate decommissioning actions.
The radionuclide inventory Contamination Following shutdown and discharge of irradiated fuel, the residual radionuclide inventory of a nuclear reactor falls into two categories- activated and contaminated materials. Neutron activated materials. These materials are located near the core and have been irradiated by neutrons. The reactor core is the most activated part of the reactor structure. The portion of the reactor exposed to relatively low neutron fluxes is essentially the biological shield, usually made of concrete and steel reinforcements. Contaminated materials. Contamination arises from the activation of the corrosion and erosion products conveyed by the coolant and from the dispersion of the irradiated fuel and fission products through cladding breaches. Activation products Fission products Actinides
The radionuclide inventory Major activation products 3 H Half life: 12.33 y Production mode: neutron capture in deuterium in D 2 O moderators or in the concrete bio-shield in the 6 Li(n,a) 3 H. Decay: Pure emitter (E max =18.59 kev) Properties: extremely mobile and readily exchanges with water in human tissue. Measurement: Distillation and LS counting
The radionuclide inventory Major activation products 14 C Half life: 5730 y Production mode: activation of trace nitrogen in the 14 N(n,p) 14 C reaction with a crosssection of 1.81 b. Additional minor routes are via neutron capture of 13 C with a crosssection of 0.9 mb and 12 C (98.89%, 3.4 mb) indirectly via 13 C Decay: Pure emitter (E max = 156 kev) Measurement: LS counting 36 Cl Half life: 3.01 10 5 y Production mode: neutron capture and activation of trace chlorine in reactor construction material- in stainlesssteel and aluminium reactor components with a crosssection of 0.04 mb Decay: - 98.1% (Emax=709 kev) and EC 1.9% (weak X rays) Properties: It is important from the viewpoint of disposal, because of its long half-life, the solubility of chloride salts and potential pathways to the biosphere Measurement: Chemical separation and LSC
The radionuclide inventory Major activation products 41 Ca Half life: 1.03 10 5 y Production mode: by the 40 Ca(n, ) 41 Ca reaction in the 96.9% abundant 40 Ca, which is present in the concrete and appears as an impurity, e.g. in graphite. Decay: EC (weak X ray emissions, ) to stable 41 K Properties: Because of its long half-life and its chemical and biological properties, 41 Ca is important for the long term safety assessment of the final disposal. Measurement: Chemical separation and LS counting 55 Fe Half life: 2.73 y Production mode: by the 54 Fe(n, ) 55 Fe reaction in the 5.9% abundant stable iron isotope 54 Fe with a cross-section of 2.25 b Decay: EC 100% Properties: After production of 55 Fe in a reactor core, translocation of this and other radionuclides from the reactor vessel through the coolant system will be a function of corrosion and deposition rates. For GCRs and other reactors, 55 Fe is the major short term component of the radioactive inventory following shutdown Measurement: Chemical separation followed by LS counting
The radionuclide inventory Major activation products 63 Ni Half life: 100.1y Production mode: neutron capture of the 3.6% abundant 62 Ni with cross section 14.2 b Decay: 100% - (E max =67 kev) to stable 63 Cu Properties: 63 Ni is the most abundant activation product expected to be present in a light water reactor (LWR) on the time-scale of deferred dismantlement. As a weak emitter it only represents an inhalation hazard Measurement: Chemical separation and LS counting 60 Co Half life: 5.27 y Production mode: by the 59 Co(n, ) 60 Co reaction on the 100% abundant 59 Co with a cross-section of 18.7 b. Cobalt is a trace constituent in both carbon and stainless steels. Decay: 100% - to stable 60 Ni Properties: 60 Co is the dominant dose producing radionuclide in the reactor interior on a 10 to some 50 year timescale. The production rate of 60 Co is sufficiently high in the high flux region near the core that a substantial portion of the stable cobalt (up to one third) may be transmuted over the life of the reactor. Measurement: Easily measured by -spectrometry
The radionuclide inventory Major activation products 93 Zr Half life: 1.5 10 6 y Production mode: neutron capture of the 3.6% abundant 62 Ni with cross section 14.2 b Decay: 93 Zr decays by emission (E max =60 kev) to 93m Nb (T 1/2 =15.8 y) Properties: 93 Zr in irradiated fuel cladding or in control rods may be the most important activation radionuclide in long term scale and is considered one of the critical radionuclides for long term safety of the final disposal. Measurement: Chemical separation and LS counting 99 Tc Half life: 2.11 10 5 y Production mode: 99 Tc (T 1/2 =1.5 10 6 y) is produced by neutron activation of 98 Mo (24.13% abundance). Decay: 100% - to stable 99 Ru Properties: The ability if 99 Tc to form anions makes it very mobile in the environment, which together with the long half-life and high production rate is behind the reason 99 Tc is of key importance for the long term safety. Both 93 Zr and 99 Tc are also major fission products with significant contribution to the radioactivity of the SNF Measurement: Chemical separation and LSC or proportional counting
The radionuclide inventory Activities and radionuclide inventory of a GC Magnox reactor, UK
The radionuclide inventory Major fission products 90 Sr Half life: 28.7 y Decay: 90 Sr decays (T 1/2 = 28.7 a), pure β- emitter (E max = 546 kev), in equilibrium with its daughter 90 Y (also a pure β- emitter). Properties: 90 Sr is one of the most abundant fission products. As one of the major fission products, there is a potential for large contamination inventories of this radionuclide. Measurement: Assessment of the activity of 90 Sr in samples requires radiochemical analysis and β- spectroscopy/lsc. 93 Zr and 99 Tc 93 Zr and 99 Tc are not only produced in a nuclear activation reactions but are also among the longest lived fission products with a cumulative fission yield >6 % each for both 235 U and 239 Pu fission
The radionuclide inventory Major fission products 106 Ru 106 Ru is produced by fission and decays (T 1/2 =374 d) by emission (E max =39 kev) to 106 Rh, which decays (half-life: 30 s) by emission (E max =3.54 MeV) to stable 106 Pd. Due to its short life, 106 Ru is not a radionuclide critical for disposal. 106 Ru can cause some radiation hazards by formation of hot spots, mainly in reprocessing or high level waste treatment facilities. 106 Ru can easily be measured by spectrometry. 129 I 129 I is a fission product with T 1/2 = 1.6 10 7 y, β- emitter (E max = 154 kev). The long life of this radionuclide and its nature as a volatile b emitter are considered very important for waste disposal. 129 I can be correlated with the easily measured 137 Cs and can also be quantified in laboratories by ƴ- spectrometry or ICP-MS.
The radionuclide inventory Major fission products 137 Cs is produced by fission and is one of the most abundant FPs, has a half-life of 30y and decays by emission (maximum energy: 1.17 MeV) to 137 Ba. Approximately 85% of the decays go through 137m Ba and thus are accompanied by the emission of its 662 kev ƴ- rays and 137 Cs is easily measured by ƴ- spectrometry. Because of its high water solubility, 137 Cs is easily transported in most LWR circuits. The design life of disposal facilities (300 y) is based on the 137 Cs half-life. 144 Ce is produced by fission and decays (T 1/2 = 285 d) by emission (E max = 318 kev). Approximately 11% of the β- decays are accompanied by the emission of a 133 kev ƴ- photon. Because of its short life, 144 Ce is not a radionuclide critical for disposal.
The radionuclide inventory Major actinides 238, 239, 241 Pu- produced by the decay of radionuclides produced by multiple neutron capture in 235 U and 238 U. 239 Pu (T 1/2 =24110y) is produced by decay of 239 Np, which is the daughter of another emitter, 239 U. 238 Pu (T 1/2 =87.7y) is an α-emitter produced by the decay of 238 Np and by α decay of 242 Cm, which are decay products of radionuclides produced by multiple neutron capture in 235 U and 238 U. 241 Pu (T 1/2 =14.35) is produced by multiple neutron capture in 238 U, 239 Pu and related isotopes. It decays primarily by emission to 241 Am. Following chemical separation, 238 Pu and 239 U are determined via α-spectrometry, 241 Pu by LSC.
The radionuclide inventory Major actinides 241 Am- produced by the β-decay of 241 Pu, decays itself (T 1/2 = 432 y) by α- emission (E α = 5.486 MeV) to 237 Np and has a low energy ƴ- line at 59.5 kev. For measuring purposes, Am is chemically separated and measured by α- spectrometry. 242 Cm (T 1/2 =162.8d) is a short-lived α-emitting member of the uranium decay series, produced by decay of 242 Am, which is produced by neutron capture from 241 Am. 244 Cm (T 1/2 =18.1d) is an α-emitter, member of the thorium decay series, produced by multiple neutron captures. Both Cm isotopes can be measured by α-spectrometry after chemical separation and electrodeposition.
The radionuclide inventory Uranium isotopes 234 U (T 1/2 =2.5 10 5 y) is part of the natural uranium isotopic mixture (0.005%), also produced by α decay of 238 Pu and neutron capture of 233 U. 234 U decays by α- emission (Eα = 4.722MeV (28.4%) and 4.774 MeV (71.4%)). 235 U (T 1/2 =7 10 8 y) is the fissile part of the natural uranium isotopic mixture (0.7%) used in most nuclear reactors, but is also produced by a decay of 239 Pu. 235 U decays by α emission (Eα = 4.215MeV (6.0%), 4.366 MeV (18.8%) and 4.398 MeV (57.2%)). 236 U (T 1/2 =2.3 10 7 y) is produced by neutron capture in 235 U or α decay of 240 Pu and decays by α-emission (Eα = 4.445MeV (26.1%) and 4.494 MeV (73.8%)). 238 U is the major part of natural uranium (99.27%) and also of the fuel. 238 U decays (T 1/2 =4.5 10 9 y) by α-emission (Eα = 4.151MeV (20.9%) and 4.198 MeV (79.0%)).
The principal activation products present in reactor materials at shutdown are: Relative importance of radionuclides with time In terms of radiation levels, 60 Co is the most predominant radionuclide In steel: 55 Fe, 60 Co, 59 Ni, 63 Ni, 39 Ar, 94 Nb In reinforced concrete: 3 H, 14 C, 41 Ca, 55 Fe, 60 Co, 152,154 Eu In graphite: 3 H, 14 C, 152 Eu and 154 Eu For steels, 55 Fe and 60 Co account for the major part of the inventory in the first ten years after shutdown The most abundant radionuclides in contamination residues still present 10 20 years after the reactor shutdown generally include 3 H, 60 Co, 55 Fe and 137 Cs After about 20 30 years, the most abundant radionuclides generally include 63 Ni, 137 Cs, 60 Co and 90 Sr The long lived transuranic actinides 241 Am, 238, 239, 240 Pu and 244 Cm do not become significant parts of the radionuclide inventory until after 100 200 years
Relative importance of radionuclides with time A B A. Calculated decay of principal radionuclides in PWR reactor (Trino NPP, Italy) B. Calculated decay of principal radionuclides in Magnox reactor (Latina NPP, Italy) Source: Radiological Characterization of Shut Down Nuclear Reactors for Decommissioning Purposes, IAEA TECHNICAL REPORTS SERIES No. 389, 1998
Characterisation Methods and techniques for characterisation Calculation of neutron induced activity In situ measurements Characterization programme objectives- to obtain representative calculations, in situ measurements and samples/analyses which provide an understanding of the radiological conditions that will be encountered during decommissioning The induced activity, and the radionuclide concentrations present in neutron irradiated components (e.g. reactor pressure vessel, reactor vessel internal components, biological shield) and the associated ƴ- dose rates are usually estimated by neutron activation calculations Sampling and analysis Scaling factors Internal and external contamination of plant systems and surfaces can be determined from direct in situ measurements on the systems and surfaces of interest In the case of removable surface contamination, wipe samples may be taken, which can be counted for either total activity or individual radionuclide concentrations For more precise determination samples may be further analysed by radiochemical separation followed by spectrometric analysis
Computer codes for calculating the induced activity. Input data: Methods and techniques for characterisation Calculation of neutron induced activities Plant operational history (i.e. time power histograms); Input cross-section data set for given neutron spectra and temperatures; Nuclear fuel characteristics (e.g. fuel geometry, enrichment, burnup level); Geometry and masses of the components subjected to the neutron flux, such as the reactor pressure vessel, the reactor internals and the biological shield; Material composition characteristics of each item potentially irradiated, including trace element composition; and The length of the decay period following final shutdown.
Methods and techniques for characterisation Calculation of neutron induced activities The calculation of neutron induced activities requires, as a first step, knowledge of the spatial and energy distributions of the neutron flux throughout the system. The neutron flux is then used to determine the individual reaction rates of the parent radionuclides whose daughters give rise to the ionizing radiations. These reaction rates are then used to obtain the level of activity per unit weight of parent element according to the reactor irradiation history and the subsequent decay time. The final stage is the calculation of the component activity from the known concentration of the parent elements in the material from which the component is manufactured, together with the mass of the components.
Methods and techniques for characterisation Calculation of neutron induced activities- computer codes Spatial and energy distribution of the neutron flux: For simple geometries one dimensional codes: ANISN, XSDRNPM, SN1D For complex geometries, two dimensional neutron transport codes: DOT/DORT, COROUT, or TWODANT For three dimensional neutron transport calculations: TORT For very complex geometries, codes based on the Monte Carlo method: MCBEND, MORSE, KENO 5, MCNP and TRIPOLI Spatial distribution of neutron induced radioactivity in all materials of the reactor ORIGEN2 tracks a large number of isotopes through specified irradiation and decay times, accounting for the creation and depletion of radionuclides throughout the reactor s operating lifetime
Methods and techniques for characterisation In-situ measurements In situ measurement for characterization: Dose rate measurements Radioactive contamination measurements Measurement of relative individual radionuclide activities by spectrometry The choice of techniques has to take into account following parameters: The nature of the radioactivity (activation, contamination) and the type of radiation emitted; Physical and geometrical conditions; Required accuracy and uncertainties
Methods and techniques for characterisation Dose rate measurements Measurements of radiation fields Provide an acceptable estimate of the activity if the relationship between activity content and radiation field is well established. Should be made at fixed, convenient distances from either internal or external contamination. Gross radiation readings doesn t indicate the nature and quantity of each isotope unless a detailed analysis is performed in order to derive isotopic concentrations comparable with total radiation readings Data needed: radiation (α,, ) dose or exposure rates Specific uses: to identify radiation hazards and access limitations, to specify decommissioning procedures and methods and to estimate waste volumes Collection methods: direct measurements, screening level, air monitoring
Methods and techniques for characterisation Contamination measurements Loose contamination Measured by wipe test- taking a small piece of material such as a filter paper and rubbing it over a specified area of the surface (i.e.100 300 cm 2 ). Total or fixed contamination Stationary detector at a fixed distance from the surface, for a fixed period of time The detector is commonly coupled with an instrument that integrates the counts over the time selected and gives a numerical result. Some of these instruments also have the capability to store a number of results for later computer analysis Scanning a surface The instrument is held close to the surface as described above, but is moved systematically along the surface at a speed that is sufficiently low (3 5 cm/s) to allow detection of changes in the radiation field. The limiting speed is a function of the detector sensitivity, the type and intensity of the radiation and the instrument resolving time
Methods and techniques for characterisation Spectrometry measurements The most detailed analysis for radionuclides can be obtained by using spectrometry. This approach is required if the ratio of emitters changes or is unknown. Spectrometry can be used for α, β or ƴ emitting radionuclides. One important application is the use of in situ ƴ spectrometry to characterize the contamination on the inner surfaces of pipes and other components. By using appropriate algorithms, it is possible to transform the measured -spectrum to a radionuclide specific surface contamination. This is a useful method to validate computer codes. In situ ƴ spectrometry in decommissioning (Magnox Electric, UK).
Methods and techniques for characterisation Scaling factors (nuclide vectors) Based on developing a correlation between easily measurable gamma emitting nuclides (key nuclides) and DTM nuclides (α- and -emitters). Typical key nuclides: 60 Co, 137 Cs The SF method has been developed from operating experience at nuclear power plants for radioactive waste. Can be used to demonstrate a correlation in a situation where DTM nuclides and ETM nuclides are coproduced. Scaling factors are waste stream and case specific Basic flow-chart for SF method development
Methods and techniques for characterisation Scaling factors (nuclide vectors) ETM/DTM correlation is based on Similarity in production mechanism; Similarity in transport behaviour within plant systems A DTM nuclide such as 63 Ni, which is produced by activation, can be expected to correlate with a key nuclide such as 60 Co, which is also produced by activation and has physicochemical characteristics similar to it. An example for a typical nuclide vector in the contamination on metallic items in nuclear power plant with light-water reactors with slight alpha contamination could be the following: ( 60 Co, 137 Cs, 90 Sr, 241 Am, α rest ) = (52%, 39%, 5%, 1%, 3%)
Sampling and analysis The main purposes of the sampling and laboratory analysis: verification of theoretical calculations for materials activation; estimation of surface contamination fields by sample removal and analysis; development of correlation factors for hard-to-detect radionuclides. The programme will provide an actual database containing information on the range of compositions, quantities and locations of radionuclide residues for activated components and contaminated interior and exterior surfaces. Taking concrete samples from the biological Shield, UK graphite reactor To be effective, such analysis generally requires the use of sophisticated equipment such as germanium detectors and multichannel analysers, a spectroscopy equipment or LSC.
Classification of RAW The costs of radioactive waste management is a significant element of the overall decommissioning costs and may even dominate in some cases. It has been estimated that about 60% of the costs of decommissioning are attributable to waste management, even though only 3% of the materials arising from decommissioning are declared as radioactive waste. Maximize Minimize Clearance Reuse Recycling Treatment Storage Disposal as RAW
Classification of RAW Potentially radioactive material and wastes which arise throughout the lifetime of a nuclear facility Operational wastes in the form of solids, liquids and gases Expired or failed plant components arising as a result of maintenance, modification or life extension worksteam generators, pumps, valves, control rods, spent filters, etc. Structural materials of the facility- steel, concrete, aluminium, graphite, etc.
Classification of RAW RANGE OF DECOMMISSIONING MATERIALS AND WASTES Some of the materials and wastes arising from decommissioning may differ from operational wastes in terms of their mass, volume, chemical, physical, radiological and toxic characteristics. Due to these differences, some of these materials may be considered problematic in that the methods of treatment, conditioning and disposal routinely used at a facility during operation may not be adequate for decommissioning waste. Wastes similar to those produced during operation will often be able to be managed using established treatment facilities, and storage and disposal arrangements. Leaving aside operational wastes, much of the material arising from the decommissioning of a nuclear facility will be either not radioactive or at most only slightly contaminated. Therefore, the majority of decommissioning material can be cleared from regulatory control and only a limited proportion of the materials will have to be disposed of as a radioactive waste.
Clearance and release from regulatory control Radioactive waste routing (road map) Source: Technical solutions for the management of radioactive waste (RAW): overview and methods of selection., Cambridge (2013)
Clearance and release from regulatory control Clearance is the removal of radioactive materials or radioactive objects within authorised practices from any further regulatory control applied for radiation protection purposes. Clearance is based on the concept of triviality of exposure: the radiation risks to individuals caused by the practice or source be sufficiently low as to be considered trivial; the collective radiological impact of the practice or source be sufficiently low as not to warrant regulatory control under the prevailing circumstances; the practices and sources be inherently safe, with no appreciable likelihood of scenarios that could lead to doses above dose limit. In quantitative terms, this is generally related to the stipulation that the effective dose expected to be incurred by any member of the public due to cleared materials is of the order of 10 μsv or less in a year. Once cleared, the materials are subject to no further regulatory restriction or control. Consequently, cleared waste may be treated as normal waste, recycled or reused for any purpose without being considered to be radioactive.
Clearance and release from regulatory control Unconditional clearance of: Solid materials and of liquids for reuse, recycling or disposal (the most general clearance option); Large quantities of building rubble and soil (an option that should be treated separately from the previous one); Buildings for reuse or demolition (no restriction on the future fate of the building). Clearance for a specific purpose: Materials for disposal on landfills or by incineration as well as of liquids for disposal by incineration; Buildings for demolition (i.e. with the restriction that the building needs to be demolished and may not be reused as a workplace); Metal scrap for melting in any conventional foundry that does not need to possess a nuclear license).
Clearance and release from regulatory control Free release measurement facilities Widely used for measurement of metal scrap, building rubble and various other types of material The high throughput makes them ideal for releasing larger quantities of metal scrap, building rubble and other bulk materials arising during the decommissioning phase The containers (usually boxes having a volume of 0.5 m³ or more) are filled with 100s of kg Measured for usually less than 1 minute in the measurement chamber close to 4π geometry The throughput of these devices can reach several tens of tons per work shift As bulk monitors are capable only of measuring the total gamma component, it is crucial to know the appropriate nuclide vector by which the counts delivered by the instruments can be related to volume or mass specific activity values of the radionuclides Source: R&D and Innovation Needs for Decommissioning of Nuclear Facilities, OECD NEA No 7191, 2014
Clearance and release from regulatory control Compliance with regulatory requirements Materials for clearance have to comply not only with the clearance levels, but also with other requirements such as mass restrictions, type of material, destination etc. Metals. Averaging masses for metals are usually chosen in the range of 100 kg up to several 100 kg. This corresponds to the quantity that a waste drum or a box used for bulk monitoring can hold. Building rubble and other bulk material. Averaging masses for building rubble are usually chosen in the range of a few 100 kg up to 1 ton. Averaging areas. The averaging area describes the area of material or building surface over which a measurement may be averaged, i.e. Bq/cm² or Bq/m². Buildings. Averaging areas for buildings are usually chosen in the range of 1 to 10 m², which corresponds well to measurements with in collimated situ - spectrometers.
Clearance and release from regulatory control Alternatives to clearance In many countries, clearance is not considered a viable option for all or part of the material originating from the nuclear sector, e.g. for the large amounts of material arising from decommissioning of nuclear installations. In these cases, other waste management approaches need to be used, instead of or in addition to clearance. Alternatives of clearance Recycling in the nuclear sector Disposal of material as VLLW Interim storage for decay.
Clearance and release from regulatory control VLLW The VLLW - activity equivalent to the lower one or two orders of magnitude of the LLW activity range but above the unrestricted release level. The benefit of designating VLLW separately from LLW is that it can then be segregated and disposed of to dedicated facilities that do not need to meet design criteria as demanding as those for LLW. The conditioning and packaging requirements of VLLW are simpler and cheaper. French VLLW repository in Morvilliers Source: Managing Low Radioactivity Material from the Decommissioning of Nuclear Facilities, IAEA Technical Report Series No. 462, 2008 IAEA Project RER9106 02, Manchester, UK, 1-5 June 2015
Conclusions Radiological characterisation is an essential step in the development of a decommissioning plan Characterisation shall be well planned to avoid extra costs, doses and project delays The characterization objectives need to be clearly defined to avoid unnecessary work The characterization process is iterative in nature and should be reassessed as more data become available Characterization should be performed on a cost to benefit basis, taking into account the need to reduce doses (ALARA principle)
Conclusions Clearance is defined by the IAEA as removal of suspected materials or objects from any further regulatory control Clearance is based on the concept of trivial dose, restricting any individual dose potentially received from the clearance to <10 Sv/y Measurements to confirm compliance with clearance levels can be carried out mainly by in-situ -spectroscopy, bulk monitors, but also dose rate monitors, wipe test and laboratory analysis Determination of correlation factors between the activities of ETM and DTM nuclides for the development of scaling factors is also a valuable characterisation method Alternatives of clearance include disposal as VLLW in dedicated site (same cost as clearance) and disposal in RAW repository site (the cost is an order of magnitude higher compared to clearance and VLLW)
References 1. Transition from Operation to Decommissioning of Nuclear Installations, IAEA Technical Report Series No. 420, 2004 2. Radiological Characterization of Shut Down Nuclear Reactors for Decommissioning Purposes, IAEA TECHNICAL REPORTS SERIES No. 389, 1998. 3. Policies and Strategies for the Decommissioning of Nuclear and Radiological Facilities, IAEA NUCLEAR ENERGY SERIES No. NW-G-2.1, 2011. 4. Application of the Concept of Exclusion, Exemption and Clearance, IAEA Safety Standards Series No. RS-G-1.7, 2004. 5. R&D and Innovation Needs for Decommissioning of Nuclear Facilities, OECD NEA No 7191, 2014.
References 6. Determination and Use of Scaling Factors for Waste Characterization in Nuclear Power Plants, IAEA Nuclear Energy Series No. NW-T-1.18, 2009. 7. Classification of Radioactive Waste, IAEA Safety Standards Series No. GSG-1, 2009. 8. Strategy and Methodology for Radioactive Waste Characterization, IAEA TECDOC No. 1537, (2007). 9. Guidance for the Data Quality Objectives Process, US EPA QA/G (2000). 10. Release of Radioactive Materials and Buildings from Regulatory Control, OECD NEA Status Report No. 6403, 2008. IAEA Project RER9106 02, Manchester, UK, 1-5 June 2015