EUNDETRAF II Course

1. EUNDETRAF II Course CHAPTER 8 : SITE CHARACTERIZATION

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 2 (Phare contract, 1999-2000) • Assessment of costs to dismantle the Belgian reactors of Doel and Tihange NPPs • Decommissioning project for IgnalinaNPP (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