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EUNDETRAF II Course CHAPTER 8 : SITE CHARACTERIZATION.

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Presentation on theme: "EUNDETRAF II Course CHAPTER 8 : SITE CHARACTERIZATION."— Presentation transcript:

1 EUNDETRAF II Course CHAPTER 8 : SITE CHARACTERIZATION

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

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

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

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

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

7 Decisions taken on the basis of Site Characterization n Decisions on decommissioning operations: z Partial of full decontamination z Provisions for shielding z Partial removal of equipment z Waste classification z 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 n Characterization is time and money consuming n Specific objectives are needed and minimum necessary to define exposures and meet the requirements of waste transport and disposal regulation n Immediate dismantling: extensive survey to support decisions on waste disposal and radiological protection n Differed dismantling: less extensive initial survey and detailed characterization of short lived radionuclides may be less important

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

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

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

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

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

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

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

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

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

18 Sampling and Analyses Objectives n Sampling and laboratory analytical programme: z Verification of theoretical calculations for material activation z Estimation of surface contamination fields z Development of correlation factors for difficult-to-measure radionuclides n 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 n Representative samples are needed for accurate characterization n Total activity per unit weight can be deduced if sample is representative of the entire component n Analysis with sophisticated equipment: germanium detectors, α spectroscopy equipment, liquid scintillation

20 Sampling schemes n Unbiased sampling schemes: z For areas expected to have little or no surface contamination or expected to be homogeneous in the degree and characteristics of the contamination z Discrete sampling areas and survey units for the measurements z Comparison with a background population to determine whether it has been affected by the facilitys operation n Biased sampling schemes (accessibility): z 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 n Floors: potential spills, heavy traffic n Walls: dust settling, sprays or steam leaks n Horizontal surfaces: dust settling (surfaces of pipes, railings …) n Ceilings: dust leaks, contaminated air circulation n Reactor pressure vessel n Reactor internals n Bioshield

22 Analysis Techniques n Initial analysis by gamma spectrometry n Comprehensive radiochemical analyses [Bq/cm 2 or Bq/g] at off-site laboratory to measure all important radionuclides z C-14 (β - ): liquid scintillation (1 Bq/g) z Co-60 (β -, γ): gamma spectrometry (0.5 Bq/g) z Ni-59 (X): X ray spectrometry (10 Bq/g) z Sr-90 (β - ): beta counting or liquid scintillation (1Bq/g) z Nb-94 (β -, γ): gamma spectrometry or ICPMS (0.5 Bq/g) z Pu-238 (α, X): alpha spectrometry (0.02 Bq/g)

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

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

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

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

27 LLWAA-DECOM Input Window

28 LLWAA-DECOM Output Values n Particle deposition and release rates n Deposited activities (Bq/m 2 ) and scaling factors at any given time (shutdown, decontamination or decommissioning)

29 LLWAA-DECOM output window

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

31 Projects Performed with LLWAA-DECOM n Technical design for decommissioning Kozloduy NPP units 1 and 2 (Phare contract, 1999-2000) n Assessment of costs to dismantle the Belgian reactors of Doel and Tihange NPPs n Decommissioning project for Ignalina NPP (BERD, from 2002, on-going)

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

33 Site Characterization after Dismantling n Survey of remaining buildings and infrastructures z Non dismantled elements are surveyed for radiological conditions by an independent organization z Final survey report to regulatory body for appraisal z Final survey is less extensive than before dismantling but targeted z Comparison of external dose rate with natural background z 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 n Soil sampling and analysis z National and international standard requirements z Investigation strategy for exploratory survey z Methods of drilling and sampling apparatus for soil, sediments and groundwater z Analysis of soil contaminated with radioactive materials z Strategy differs by the type of soil and the already available knowledge about the site

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

36 Site Characterization after Dismantling n Evaluation of site clearance z Maximum radiation and contamination levels n Dose rate [mSv/h] n Contamination [counts/m 2 ] or [Bq/m 2 ] n Nuclide specific activity [Bq/m 2 ] or [Bq/kg] z Remaining health physics risks for the public after decommissioning [Sv/year]

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

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


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