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Presentation on theme: "CHAPTER 8 : SITE CHARACTERIZATION"— Presentation transcript:


2 Table of contents General Objectives of Site Characterization
Site Characterization before Dismantling Methods and Techniques for Site Characterization Site Characterization after Dismantling

3 General Objectives of Site Characterization
Database of information about: Quantity and type of radionuclides Distribution of radionuclides throughout the site Physical and chemical states of radionuclides To gather on the basis of: Survey of existing data Calculations In-situ measurements Sampling and analyses

4 Aim of Site Characterization
Assessment of various decommissioning options and their consequences to select a strategy (immediate or deferred dismantling): Operating techniques (decontamination processes, dismantling procedures – hands on, semi-remote or fully remote – tools required) Radiological protection of workers, general public and environment Waste classification Resulting costs

5 Stepwise Characterization
In the beginning of the decommissioning planning stage: Collection of sufficient information Assessment of radiological status of the facility Nature and extent of any problems areas Basis: available information including historical operations documentation Planning of the overall decommissioning programme Prioritizing and sequencing major decommissioning activities

6 Stepwise Characterization
Progress of the planning: More detailed data concerning physical, chemical and radiological conditions of the nuclear installation Basis: performing calculations of induced activity, taking samples and conduct inspections to fill information gaps Set-up of scheduling and work force requirements Assessment of exposures in radioactive areas Project decisions to select a preferred decommissioning scenario

7 Decisions taken on the basis of Site Characterization
Decisions on decommissioning operations: Partial of full decontamination Provisions for shielding Partial removal of equipment Waste classification Initial estimates of project costs and schedules taken on the basis of the actual distribution of the radioactive inventory and on associated radiation exposures

8 Extent of Site Characterization
Characterization is time and money consuming Specific objectives are needed and minimum necessary to define exposures and meet the requirements of waste transport and disposal regulation Immediate dismantling: extensive survey to support decisions on waste disposal and radiological protection Differed dismantling: less extensive initial survey and detailed characterization of short lived radionuclides may be less important

9 Site Characterization before Dismantling
Successive steps of the characterization programme: Review of historical information Implementation of calculation methods Preparation of the sampling and analysis plan based on an appropriate statistical approach Performance of in situ measurements, sampling and analyses Review and evaluation of data obtained Comparison of calculated results and measured data

10 Review of historical information
Gather valuable information about radiological conditions of the site Records and recollection of accidents and incidents Previous surveys and measurements: occupational exposures incurred during inspection, maintenance and repair activities or replacements of contaminated equipment List of possible contaminants from a review of installation history Structural surveys on the basis of as-built drawings and modifications brought to structures or equipment

11 Implementation of Calculation Methods
Use of computer codes to calculate: The induced activity in a nuclear installation and its immediate surroundings (radioactive inventory) The radionuclide distribution as a result of normal operation, accident and transport of mobile contamination Are the theoretical calculations sufficient for the subsequent planning of decommissioning activities? Supplement calculations by a detailed sampling and measurement plan (irradiated foils)

12 Preparation of the Sampling and Analysis Plan
Sampling and analysis plan defines: Types, numbers, sizes, locations and analyses of samples required Instrument requirements Radiation protection aspects and controls of the activity Data reduction, validation and reporting requirements Quality assurance requirements Methodology to take samples and perform analyses Requirements for disposal of waste generated during sampling

13 Performance of In-Situ Measurements, Samplings and Analyses
In-situ measurements and/or samples should be taken on various components that can be reasonably accessed Samples of irradiated and/or contaminated materials such that laboratory analyses enable to determine individual radionuclide activities and concentrations Expensive process Difficult for highly activated components and structures

14 Review and Evaluation of Data Obtained
Continuous assessment of data to determine if requirements are met: Departure from original plan when contamination is more important than expected and a greater number of samples is needed Different possibilities: Altering the sampling technique Changing the frequency Redefining the regions

15 Comparison of Calculated Results with Measured Data
Comparison of calculated results with measured data to: Obtain a validation of the accuracy of the calculations Guide adjustments of the theoretical models used Increasing confidence in application of codes Cost effective method of obtaining characterization information

16 Methods and Techniques for Site Characterization
Obtain representative In situ measurements Sampling/analyses Computer Code Calculations to understand radiological conditions encountered during decommissioning

17 In Situ Measurements Three types of measurements:
Dose rate measurements Radioactive contamination measurements Measurement of individual radionuclide activities by spectrometry

18 Sampling and Analyses Objectives
Sampling and laboratory analytical programme: Verification of theoretical calculations for material activation Estimation of surface contamination fields Development of correlation factors for difficult-to-measure radionuclides Programme providing an actual database containing information on the range of compositions, quantities and locations of radionuclides for activated components and contaminated interior and exterior surfaces

19 Sampling and Analyses Representative samples are needed for accurate characterization Total activity per unit weight can be deduced if sample is representative of the entire component Analysis with sophisticated equipment: germanium detectors, α spectroscopy equipment, liquid scintillation

20 Sampling schemes Unbiased sampling schemes:
For areas expected to have little or no surface contamination or expected to be homogeneous in the degree and characteristics of the contamination Discrete sampling areas and survey units for the measurements Comparison with a background population to determine whether it has been affected by the facility’s operation Biased sampling schemes (accessibility): Finding or defining contamination or activation in areas where it is known to exist or likely to occur

21 Typical Survey Areas for a NPP
Floors: potential spills, heavy traffic Walls: dust settling, sprays or steam leaks Horizontal surfaces: dust settling (surfaces of pipes, railings …) Ceilings: dust leaks, contaminated air circulation Reactor pressure vessel Reactor internals Bioshield

22 Analysis Techniques Initial analysis by gamma spectrometry
Comprehensive radiochemical analyses [Bq/cm2 or Bq/g] at off-site laboratory to measure all important radionuclides C-14 (β-): liquid scintillation (1 Bq/g) Co-60 (β-, γ): gamma spectrometry (0.5 Bq/g) Ni-59 (X): X ray spectrometry (10 Bq/g) Sr-90 (β-): beta counting or liquid scintillation (1Bq/g) Nb-94 (β-, γ): gamma spectrometry or ICPMS (0.5 Bq/g) Pu-238 (α, X): alpha spectrometry (0.02 Bq/g)

23 Computer Codes Calculation of neutron induced activity
Spatial and energy distribution of the neutron flux Deterministic methods to solve the transport equations by mathematical approximations Simple geometries: ANISN Two-dimensional geometries: DORT Three-dimensional geometries: TORT Stochastic methods such as Monte Carlo Complex geometries: MCBEND and MCNP

24 Computer Codes Spatial distribution of the neutron induced radioactivity in all materials of the reactor Average neutron fluxes in all zones representing the fixed structure of the reactor Material compositions of the zones Time-power histograms for reactor operation Radionuclides specific activities in the zones ORIGEN-2 computer code

25 Computer Codes Calculation of surface contamination
BKM-CRUD model: buildup of activated corrosion and/or erosion products in the primary circuit PACTOLE computer code: ion solubility, release rates of base metals, dissolution of deposits, precipitation of soluble products and deposition rates of solid particles LLWAA-DECOM: Low level Waste Activity Assessment - Decommissioning

26 LLWAA-DECOM Input Parameters
Contamination in the streams of the nuclear systems (calculated by LLWAA) Characteristics of the equipment to be dismantled (piping diameter, pipe rugosity …) Operating conditions in piping systems (fluid velocity, pH, temperature, number of cycles, cycle life …) Particulate diameter distribution of corrosion products Time elapsed between the reactor final shutdown and the decontamination or dismantling

27 LLWAA-DECOM Input Window

28 LLWAA-DECOM Output Values
Particle deposition and release rates Deposited activities (Bq/m2) and scaling factors at any given time (shutdown, decontamination or decommissioning)

29 LLWAA-DECOM output window

30 LLWAA-DECOM validation
Direct measurements of deposited activities present a number of practical difficulties and are not often available Coupling of a dose rate model to the calculated deposited activities and comparison between calculated and measured dose rate Good agreement between calculated and measured dose rates during NPP shutdown and decontamination operations predicted and measured deposited activities during steam generator replacement programmes

31 Projects Performed with LLWAA-DECOM
Technical design for decommissioning Kozloduy NPP units 1 and (Phare contract, ) Assessment of costs to dismantle the Belgian reactors of Doel and Tihange NPPs Decommissioning project for Ignalina NPP (BERD, from 2002, on-going)

32 Site Characterization after Dismantling
Regulatory requirements Radiological protection and waste release criteria Owner/operator/site licensee: statement of end-point objectives Comparison of stated end-points with situation prevailing on site by independent final site survey Decision by regulatory body: discharge of responsibilities imposed by site license conditions – no future requirements Waste: clearance level (municipal sites, reused of recycled) or disposed of under relevant regulatory control

33 Site Characterization after Dismantling
Survey of remaining buildings and infrastructures Non dismantled elements are surveyed for radiological conditions by an independent organization Final survey report to regulatory body for appraisal Final survey is less extensive than before dismantling but targeted Comparison of external dose rate with natural background Radionuclides depend on type of installation/facility (Cs-137, Sr-90 and Co-60 for nuclear reactors, more difficult for H-3 [CANDU reactors] and C-14)

34 Site Characterization after Dismantling
Soil sampling and analysis National and international standard requirements Investigation strategy for exploratory survey Methods of drilling and sampling apparatus for soil, sediments and groundwater Analysis of soil contaminated with radioactive materials Strategy differs by the type of soil and the already available knowledge about the site

35 Site Characterization after Dismantling
Evaluation of site clearance Results of in situ measurements, calculations, statistical data treatment, sampling & analyses, soil analyses are evaluated against defined release criteria Agreement between the involved authorities about release criteria at an early stage of the decommissioning project

36 Site Characterization after Dismantling
Evaluation of site clearance Maximum radiation and contamination levels Dose rate [mSv/h] Contamination [counts/m2] or [Bq/m2] Nuclide specific activity [Bq/m2] or [Bq/kg] Remaining health physics risks for the public after decommissioning [Sv/year]

37 Site Characterization after Dismantling
Options for site reuse Satisfactory final survey report Brown field site Assessed levels of risk from radiological or other hazards is generally acceptable but somewhat higher than what is acceptable for the general public Restrictions on use of the site May be used for industrial purposes such as warehouse, parking facility … but not for housing, schools or agriculture

38 Site Characterization after Dismantling
Options for site reuse Satisfactory final survey report Green field site Radiological risks for the general public is no higher than natural radiation background No restrictions on use of the site


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