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EUNDETRAF II Course
CHAPTER 8 : SITE CHARACTERIZATION
Table of contents
General Objectives of Site Characterization
Site Characterization before Dismantling
Methods and Techniques for Site Characterization
Site Characterization after Dismantling
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
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
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
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
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
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
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
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
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)
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
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
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
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
Methods and Techniques for Site Characterization
Obtain representative In situ measurements
Sampling/analyses
Computer Code Calculations
to understand radiological conditions encountered during decommissioning
In Situ Measurements
Three types of measurements: Dose rate measurements
Radioactive contamination measurements
Measurement of individual radionuclide activities by spectrometry
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
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
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
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
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)
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
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
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
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
LLWAA-DECOM Input Window
LLWAA-DECOM Output Values
Particle deposition and release rates
Deposited activities (Bq/m2) and scaling factors at any given time (shutdown, decontamination or decommissioning)
LLWAA-DECOM output window
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
Projects Performed with LLWAA-DECOM
Technical design for decommissioning Kozloduy NPP units 1 and 2 (Phare contract, 1999-2000)
Assessment of costs to dismantle the Belgian reactors of Doel and Tihange NPPs
Decommissioning project for Ignalina NPP (BERD, from 2002, on-going)
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
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)
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
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
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]
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
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