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