STUK-A130 FEBRUARY 1996 9-

Monitoring of Airborne Contamination Using Mobile Equipment

T. Honkamaa, H. Toivonen, M. Nikkinen

1 0 r 1 v. f STUK-A130 FEBRUARY 1996

Monitoring of Airborne Contamination Using Mobile Equipment

T. Honkamaa, H. Toivonen, M. Nikkinen

FINNISH CENTRE FOR RADIATION AND NUCLEAR SAFETY P.O.BOX 14 FIN-00881 HELSINKI Finland Tel. +358 0 759881 NEXT PAGE(S) left BLANK FINNISH CENTRE FOR RADIATION STUK-A130 AND NUCLEAR SAFETY

T. Honkamaa, H. Toivonen, M. Nikkinen. Monitoring of Airborne Contamination Using Mobile Equipment. Helsinki 1996, 65 pp.

ISBN 951-712-094-X ISSN 0781-1705 Key words Radioactive particles, in-field gamma-ray spectrometry, emergency preparedness

ABSTRACT

Rapid and accurate measurements must be carried out in a nuclear disaster that releases radioactive material into the atmosphere. To evaluate the risk to the people, .external dose rate and nuclide-specific activity concentrations in air must be determined. The Finnish Centre for Radiation and Nuclear Safety (STUK) has built an emergency vehicle to accomplish these tasks in the field. The present paper describes the systems and methods to determine the activity concentration in air.

Airborne particulate sampling and gamma-ray spectrometric analysis are quantitative and sensitive tools to evaluate the nuclide-specific concentrations in air. The vehicle is equipped with a pump sampler which sucks air through a glassfiber filter (flow about 10 1 s"1). Representative sampling of airborne particles from a moving vehicle is a demanding task. According to the calculations the total collection efficiency in calm air is 60 -140% for the particle size range of 0.1 - 30 um. Side wind increases differences in collection efficiency between different particle sizes. The activity concentration in air is obtained by dividing the activity of the filter by the amount of air sampled. The nuclide-specific detection limits for some typical release nuclides are about 0.1 Bq m'3 when the sampling time and collection time of the spectrum are both 10 minutes.

The filter can also be analysed with a beta counter. A simple method and model to evaluate the approximate beta activity concentration in air are presented.

In the field conditions airborne activity concentration measurements in-situ are possible. This facilitates the detection of hazardous activity concentrations immediately. The measuring routines and alarm controls can be automated. The nuclide-specific detection limits are 1 - 10 Bq m'3 (spectrum collection time 10 minutes) when the detector is shielded against the unscattered gammaflux from the ground. FINNISH CENTRE FOR RADIATION AND NUCLEAR SAFETY STUK-A130

PREFACE

Inhalation is an important dose pathway at the early stages of a nuclear accident, hi many scenarios inhalation exposure is dominating over the external dose rate during the first two days after the accident. To protect people it is important to find out the concentration of radionuclides during the plume arrival to the target areas. This, however, is a very demanding measuring problem.

All Nordic countries have stationary and mobile air samplers. However, only in a few places the sampler is located in the vicinity of a laboratory that has the analysing facili- ties. Often it takes a long time before the filters are transported from the field to the laboratory. The time delay is much longer if also the samplers have to be transported to the site in danger.

Air sampling and high-resolution gamma-ray spectrometry on-line are one way to find out the nuclide-specific concentrations. Such devices are commercially available but they are extremely expensive. Another approach is to use mobile units - either cars, boats or helicopters and aeroplanes.

The authors are grateful to Jens Hovgaard, Danish Emergency Management Agency, for initiating this project within the NKS frames (NKS/EKO3/95AVR7). FINNISH CENTRE FOR RADIATION STUK-A130 AND NUCLEAR SAFETY NOMENCLATURE

Symbol description Value Unit First appearance

3 Ac Activity concentration in air variable Bqm" (8) A, Parameter in the model variable - (10) Bg Average background count- rate in beta counter variable cpm (17) Cunningham factor variable (B.6) cc - CR Count rate in the beta-probe variable s"1 (10) CRM Count rate generated by artificial radioactive nuclides variable cpm (13) CR,. Average count rate in beta probe in i:th measurement variable cpm (14) 222Rn-concentration in air variable Bqm"3 (12)

Q Beta-activity concentration oi" sampled air variable Bqm"3 (15) d Height of the detector above ground 1.7 m (Figure 18) D Characteristic dimension (diameter for circle) variable m (6) dae Aerodynamic diameter of a particle variable m (B.3) Do Diameter of the entry of the probe 0.036 m (5) dP Physical diameter of a particle variable m (5) A Diameter of the sampling tube (=diameter of the base ofthediffuser) 0.088 m (B.2) Ea Aspiration efficiency variable - (1)

Ec Total sampling efficiency variable (3) Ef Sampling efficiency of the filter variable (3) EF Calibration factor in gamma-ray spectro- metric measurement of a filter sample variable (Bq s'1) (8) FINNISH CENTRE FOR RADIATION AND NUCLEAR SAFETY STUK-A130

E, Sampling efficiency of the tube variable - (2) E* Transmission efficiency of the tube variable - (B. 1) G Parameter for calculation of aspiration efficiency variable - (A. 3) g Acceleration induced by earth 9.81 ms': (B.3) h, Height (see Figure 18) variable m (27) h, Height (see Figure 18) variable m (27) lv Vena contracta parameter variable - (B.7) L Wall impaction parameter variable - (B.7) L Length of the tube 0.4 m (B.2) L, Deposition loss inside the tube variable - (2) MDA() Minimum detectable activity of given quantity variable Bq m"3 (19) MPA Maximum possible activity concentration of beta active nuclides in air variable Bq m"3 (21) MR Measured count rate ratio (see Figure 13) variable - (13) MRcmI Upper confidence limit for MR, variable - (18) N Net counts in a peak in a gammaspectrum variable (8) Na Number of particles going through the inlet •variable - (1) Na,b Number of background pulses under a peak in a gammaspectrum variable - (9) Nb Number of background pulses under a peak in a gammaspectrum variable - (7) #w Number of background pulses under a peak in a gammaspectrum variable - (9) Nbipeak Number of pulses in a peak in a background gammaspectrum variable - (9) Count rate in beta probe when no artificial activity is present variable S"1 (Figure 13) FINNISH CENTRE FOR RADIATION STUK-A130 AND NUCLEAR SAFETY

1) Number of particles that originally were in the air volume collected variable - (1) 2) Count rate in full energy peak variable s"1 (24) NMDA Detection limit of a nuclide expressed in pulses variable - (7) NR Expected count rate ratio, when no artificial radioactive nuclides are present (see Figure 13) variable - (13) No Full energy peak count- rate for photons arriving from 9=0 variable S"1 (24) R Speed ratio VJV variable - (A. 2) R(6.4>) Relative detector response at the given angles variable - (26) Re Reynolds number variable - (6) Stk Stokes number variable - (5) Source term (activity sv concentration in air) variable Bqnr3 (24) Q Flow rate through the sampler variable mV1 (12) ta measuring time of actual measurement variable s (9) h measuring time of background measurement variable s (9) *, tl, h time variable s (10,13) ^samp sampling time variable min (10) to Parameter in the model (11)-(12) 4.23 min (10) V 1) Velocity of the air flow at the entry of the probe variable ms'1 (A. 2)) 2) Collected air volume in 3 the measuring time variable m (8) Speed of the vehicle variable ms'1 (4) vc Speed of free stream vf near the entry of the probe variable ms'1 (4) 1 Vrs Settling velocity of a particle variable ms' (B.2) Velocity of wind variable ms"1 (4) vw U Flow velocity at the observed point variable ms'1 (6) a Gravity effect angle 0 - (B.8) 1 1 Y gamma yield of a nuclide variable Bq" s" (8) FINNISH CENTRE FOR RADIATION AND NUCLEAR SAFETY STUK-A130

•n Air viscosity 1.8-10"5 N s m"2 (5) e 1) The angle between the axis of the probe and air flow relative to it variable - (A. 1) 2) Angle defined in Figure 15 variable - (25) Kp Effective decay constant of radon progenies in a filter 0.0151 min'1 (10) Attenuation coefficient in air without coherent scattering variable m"1 (27) Density of air 1.2 kg m"3 (6) Pg 3 PP Density of the particle 1000 kg m" (5) oCRj standard deviation of CRS variable - (16) oMR standard deviation of Mi variable - (17) $ Photon fluence rate at the detector variable s"1 m-2 (24) Angle defined in Figure 17 variable - (25) FINNISH CENTRE FOR RADIATION STUK-A130 . AND NUCLEAR SAFETY CONTENTS Page

1 INTRODUCTION 11

2 AIR SAMPLING IN A MOVING VEHICLE 13 2.1 Measurement of airborne contamination 13 2.2 Measuring system in the emergency vehicle of STUK 13 2.3 Theoretical estimation of sampling efficiency 15 2.4 Improving the inlet 24 2.5 Filtration efficiency 26 2.6 Tests 26 2.7 Discussion 26

3 GAMMA-RAY SPECTROMETRIC ANALYSIS OF THE FILTER IN THE FIELD 28 3.1 Gamma-ray spectrometric measuring system of the emergency vehicle of SIUK. 28 3.2 Minimum detectable activities of some nuclides 30 3.3 Discussion 32

4 RAPID CONCENTRATION ESTIMATION OF BETA-ACTIVE NUCLIDES USING SIMPLE DEVICES 35 4.1 Introduction 35 4.2 Measuring devices 35 4.3 Calibrations 36 4.4 Results of calibration measurements 40 4.5 Artificial activity concentration in air 42 4.6 Error prediction and detection limits 42 4.7 Prediction of radon concentration 45 4.8 Computer code for calculation 45 4.9 Detection limits 46 4.10 Reliability of the method 46

5 iAr-S/71/MEASUREMENTS USING HIGH-RESOLUTION GAMMA- RAY SPECTROMETER 48 5.1 Introduction 48 5.2 Methods ' 48 5.3 Theory 50 5.4 Calibration measurements 53 5.5 Detection efficiency and detection limits 54 5.6 Advanced spectrum collection and analysis 57 5.7 Discussion 59 FINNISH CENTRE FOR RADIATION AND NUCLEAR SAFETY STUK-A130

REFERENCES 60

APPENDIXES 62 Appendix A Calculation of aspiration efficiency 62 Appendix B Formulae of the theory of Hangal and Willeke 63 Appendix C Count rates in background measurements 65

10 FINNISH CENTRE FOR RADIATION STUK-A130 AND NUCLEAR SAFETY 1 INTRODUCTION

Real or suspected release of radioactive material to the atmosphere requires rapid and accurate measurements to evaluate the risk to the people. The authorities must have reliable real-time information about the situation before any civil defence countermeasures can be justified and realised. At least two different quantities must be measured before the risk to the people can be evaluated: (1) external dose rate and (2) nuclide-specific activity concentrations in air. In the later stage of the accident the fallout is important, too.

The external dose rate can be measured using relatively cheap and simple devices. An automated network in Finland consists of about 300 stations measuring external dose rate. However, in local accidents the network may be too sparse. Also, a release from a source aboard may cross the country as a narrow beam. Measuring patrols can be used to scan the gaps between the stations and to produce a good overall view about the situation.

In the first stages of an accident the largest risk to the people is often caused by radioactive gases and particles. The external dose-rate measurement is not a sensitive way to evaluate the airborne contamination. Airbome particulate sampling using a filter or a charcoal cartridge and gamma-ray spectrometric analysis of the sample with an HPGe detector are a quantitative and sensitive method to evaluate the nuclide-specific concentrations in air. Usually the gamma-ray spectrometric analysis is performed in a laboratory. However, to get prompt results, some measurements must be carried out on the areas far away from the laboratory facilities. In an emergency, there is not enough time to transfer the sample to the nearest laboratory. The measuring patrol, equipped with an HPGe detector and a small air particulars sampler, must be sent to the field for real-time analysis.

In the field conditions airborne activity concentration measurements in-situ are possible. This facilitates the detection of hazardous activity concentrations immediately.

There are several reasons why a gamma-ray .spectrometric measurement system should be permanently installed in a moving laboratory. The work of the patrol is more efficient when all the instruments are installed on their places in advance. The functioning of the measuring system can be checked instantaneously. The start-up preparations of the patrol are fast. A background shield can be installed around the detector to improve the accuracy of the measurements and to decrease the detection limits. Also, problems with electricity supply are solved in advance using an engine- driven generator.

The mobile laboratory can be equipped with positioning system (GPS) which makes the handling, processing and visualisation of the measuring data easier. The data transfer

11 FINNISH CENTRE FOR RADIATION AND NUCLEAR SAFETY STUK-A130 equipment can be permanently installed in the vehicle. Also, the programs can be tailored for gamma-ray spectrometric measurement, data processing and data transfer. The use of these systems, however, requires special training.

The data transfer to headquarters must be organised well. In an emergency the data communication should be simple, reliable and unambiguous. The best way is to use as automated systems as possible. Computerised mapping systems are not only practical but essential, if large amounts data, labelled with coordinates, is produced.

Dispersion models can be used to forecast the transport of particles and gases. The models are based on measured and predicted meteorological data. During the last years the accuracy of these models is improved because of increased capacity of computers and enhanced method development. The models are very useful in a nuclear accident. The forecasts have great importance when the route of the patrol is chosen.

The present study deals with monitoring of airborne contamination using the experiences obtained from the emergency vehicle of STUK. The following measuring systems of the vehicle are presented: (1) an air sampler for collecting paniculate and iodine samples on a filter, (2) an HPGe gamma-ray spectrometry system, which can be used for sample analysis and in-situ measurements, (3) a beta counter, which can be used for analysing the total beta activity of a filtered sample. In addition, the vehicle has dose-rate measuring systems including a pressurized ionisation chamber and Geiger tubes. The measurements are controlled with PC computers and the data is transferred to headquarters via mobile telephones (NMT 450) and modems. The systems produce measuring data in a specified format. Thus, the data can be fed directly to the databases of a surveillance program (SVO+), which displays the measurements colour-coded on the screen of the computer. The vehicle is kept at constant preparedness for unexpected accidents and incidents.

12 FINNISH CENTRE FOR RADIATION STUK-AI30 AND NUCLEAR SAFETY 2 AIR SAMPLING IN A MOVING VEHICLE

2.1 Measurement of airborne contamination

The radionuclide concentrations in air can be measured by taking a sample on a filter and analysing it with a gamma-ray spectrometer. The average activity concentration during the sampling period is obtained by dividing the activity of the filter by the amount of air sampled. A fast analysis of the sample can be carried out in the field with a gamma-ray spectrometer or beta monitor. More sensitive and accurate analysis will be carried out later in a laboratory.

The gamma-ray spectrometric analysis of a sample is quite accurate. The systematic errors in activity measurement are less than 5% if the calibration is made well. A more critical question is the representativeness of a particulate sample. The particles, which have different properties (size, density, e.g.) behave differently in a sampling system and their sampling efficiency can differ remarkably. The inertial effects are the most important biasing phenomena when air is sampled from a moving vehicle. Thus, the sampling efficiency depends strongly on the aerodynamic diameter1 of the particles. For calculation of settling velocity, see Formulae (B.3 - B.6).

2. 2 Measuring system in the emergency vehicle of STUK

Particulate sampling is a demanding process when performed in a moving vehicle. STUK has developed during the last few years methods to overcome the problems faced in the field. If the sampling system is not properly designed, it may be possible that a large fraction of particles (and activity) is lost and consequently, incorrect concentration estimates are produced.

The particulate sampling system in the vehicle consists of a sampler (DWARF, Senya Ltd, Helsinki), collection tube and glassfiber filter (Figure 1). The filter can be changed inside the vehicle. Thus, the whole sampling procedure can be carried out when the car is moving, In the conical diffusor the speed of free airstream (induced by the wind and vehicle velocity) is decreased to the appropriate level for the filter (approximately 2 ms"1).

Technical details of DWARF-sampler are given in Table I. The sampler is permanently installed in the car and it is equipped with a sound silencer. The control panel is separated from the sampler and is placed near the user. The sampler is easy to use, because it calculates automatically the amount of sampled air and sampling time.

lAerodynamic diameter is the diameter of the unit density (p=l g cm'3) sphere that has the same settling velocity as the particle.

13 FINNISH CENTRE FOR RADIATION AND NUCLEAR SAFETY STUK-A130

Outer tube

O-nng seal

Inner tube

Exhaust tube

Sound silencer

Figure 1. Sampling system of the emergency vehicle ofSTUK. The filter is attached at the front end of the inner tube and it can be changed inside the moving car. The collection tube is short to minimize deposition losses.

The volumetric flow through the sampler is adjustable. The power (1 kW) for the sampler is taken from an electricity generator (capacity 3.5 kW), which is run by the engine of the vehicle. Alternatively, an auxiliary petrol engine can be used.

14 FINNISH CENTRE FOR RADIATION STUK-A130 AND NUCLEAR SAFETY

Table I. Tt chnical information about DWARF-sampler (Senya Ltd, Helsinki)

Size of the filter spherical, 100 mm

Volumetric air flow through max 131s"1 the MN 85/90 glassfiber filter adjustable

Volumetric air flow through max 4.5 1 s'1 the MN 100 charcoal filter adjustable

Displays -pressure difference over calibrated flange -sampling time -amount of sampled air -flow rate

Power consumption max. 1000 W

Weight 10.5 kg

Size 37 x 20 x 20 cm3

2. 3 Theoretical estimation of sampling efficiency

The following factors must be taken into account to estimate the sampling efficiency: (1) aspiration efficiency at the entry of the probe, (2) deposition losses in the tube and (3) sampling efficiency of the filter.

Definitions of sampling efficiency

The aspiration efficiency at the inlet is defined as (Vincent, 1989): N (1) N.7

Na is the number of particles that go through the inlet, and Nf is the number of particles that originally were in the air volume collected. Number concentrations can also be used. The aspiration efficiency depends on the Stokes number of the particle at the inlet

15 FINNISH CENTRE FOR RADIATION AND NUCLEAR SAFETY STUK-A130

(Formula 5), on free stream air velocity and on relative angle between the probe and free stream air velocity. Bouncing and resuspension can also affect the aspiration efficiency of dry, large and gritty particles.

The sampling efficiency from the inlet entry to the filter is

Er(l~Q> (2)

where Lt describes deposition losses inside the tube. The total sampling efficiency is obtained by multiplying Formulae (I) and (2) with the sampling efficiency of the filter Ef:

Ee * EJSE,* Efl-L&. (3)

All these sampling efficiencies vary as a function of the particle aerodynamic diameter. The sample is representative if the total sampling efficiency £,.is one.

Flow field near the entry of the probe

It can be assumed, that the free stream velocity near the entry of the probe is a vector sum of the opposite vehicle velocity Fcand the wind velocity Vw:

Vf = -Ve + V* (4)

The entry of the probe is quite far (60 cm) from the closest surface of the vehicle. So its effect on the stream is not significant. Other cars may occasionally generate excessive turbulence in the coming flow.

Aspiration efficiency

If the velocity is not changed at the entry of the probe, all the particles that originally were in the sampled air volume go through the inlet (Figure 2 a). In principle, this is the most ideal situation, because the sample is fully representative at the entry of the tube. However, some sample preconcentration might be useful to compensate the losses in the tube. This is achieved, if the free stream velocity of air is greater than the velocity of air passing through the inlet. In that case the limiting stream surface is divergent (Figure 2 b). Large particles which have large inertia do not follow the streamlines and

16 FINNISH CENTRE FOR RADIATION STUK-A130 AND NUCLEAR SAFETY the particle concentration at the entry of the probe is increased. On the contrary, if the free stream velocity of air is smaller than the velocity passing through the inlet the concentration of large particles is diluted (Figure 2 c).

Figure 2. a) Isokinetic sampling (R=l, see Formula A.2). The velocity of air is not changed at the inlet. No sample biasing occurs, b) Sub-isokinetic sampling (R>1). The velocity of air decreases at the inlet and the limiting stream surface is divergent. The sample contains particles that originally were not in the collected air volume, c) Superisokinetic sampling (R<1). The velocity of air increases at the inlet. Some particles that originally were in the sampled air volume are lost.

17 FINNISH CENTRE FOR RADIATION AND NUCLEAR SAFETY STUK-A130

The aspiration efficiency is calculated from a simple theory (Vincent, 1989, Vincent et al., 1986). The theory is applicable even for non-isoaxial conditions, where the free stream velocity and the axis of the probe are not parallel. However, the theory is developed for a thin-walled cylindrical probe, while in the vehicle the shape of the probe is conical. The conical probe is "blunter", since it has more aerodynamic blockage against the airstream than a thin-walled probe. A theory is developed for blunt probes, too (Vincent, 1989, Vincent, 1987). It predicts, that there is difference in aspiration efficiency between thin-walled and blunt probes in the intermediate region of Stokes number 0.1

V 18 ti Do' ' where T| is viscosity of air, pp density of the particle, dp physical diameter of the particle and Do diameter of the entry of the probe. However, the theory for blunt samplers cannot be applied, because all the parameters are not known. Besides, using the thin- walled model seems reasonable, since the thickness of the wall is small compared with. the diameter of the entry and the shape of the diffuser is quite streamlined. The Formulae of Vincent (1989) are presented in Appendix A.

Flow in the diffuser

The flow in the tube is characterised by the dimensionless Reynolds number Re:

Re=^i . (6)

pg is the density of air, U is the flow velocity at the observed point and D is the characteristic dimension (diameter of the tube). If the sampler is used at full power, the value of Re is approximately 30,000 at the entry of the probe and 13,000 in the tube after the diffuser. These figures indicate turbulent flow.

In the diffuser the velocity of the flow decreases, because the cross section of the tube increases. However, in the radial direction the velocity of the flow in the diffuser is not constant. The velocity of the fluid is zero at the wall surface and, because of viscosity, a boundary layer appears next to the wall where sheared stresses are present.

The pressure gradient in the diffuser is adverse, since the velocity of the flow is decreased. If the pressure gradient is strong enough, the flow separates from the boundary layer and may even turn backwards (see Figure 3).

The area ratio of the diffuser (ratio of base area and entry area of the diffuser) is 6.25 and the expansion angle (the angle between the diffuser axis and internal wall) is 7.7 degrees. According to the classification given by White (1986) it is likely, that the

18 FINNISH CENTRE FOR RADIATION STUK-A130 AND NUCLEAR SAFETY

P) ioundary layer

centre line

b) Circulating Reattachment backflow point Boundary layer separation point

centre line >f igure. ^Boundary layer separation in adverse pressure gradient inside the diffuser (adopted f'om Seebaugh, 1991). a) An ideal flow inside the diffuser and in the following tube. The velocity of the flow is inversely propotional to the cross section of the tube, b) Circular backflow inside the diffuser. There is an adverse pressure gradient inside the diffuser, since the velocity of the stream is decreasing. If this pressure gradient is strong enough, it separates the flow from the surface of the tube and a reversed flow region appears near the surface.

difluser operates at the transitory stall region. It means that a separated region wanders around inside the diffuser in an unsteady manner.

Another type of flow separation occurs (so called leading edge separation), when the sharp-edged diffuser is sufficiently misaligned with the flow, because the flow cannot instantaneously change its direction around the sharp edge. As a result a separated and reverse flow region appears in the diffuser. The flow may reattach to the surfaces later in the diffuser. A leading edge separation may also couple with a transitory stall

19 FINNISH CENTRE FOR RADIATION AND NUCLEAR SAFETY STUK-A130 separation to generate a separated region that fluctuates with time. The rounding of the edges would significantly reduce the leading edge separation.

The flow separation and circulating backflow are undesirable, since large particles are lost in the deposition process.

Dej. osition in the diffuser and in the tube

If a particle collides with the wall of the probe, it most likely sticks and is permanently removed from the air stream, since adhesion forces are generally much stronger than forces that tend to resuspend the particle. However, it is possible that the particle bounces from the wall. The air stream may also blow the particle off the surface. These effects may be significant for gritty and dry particles. Air goes around the walls of the probe; so to deposit, the particles must turn away from the streamlines. Important deposition mechanisms are gravitation, electromagnetic forces, impaction and diffusion.

Impaction and gravitation near the entry of the tube

Okazaki, Wiener and Willeke (1987) and Okazaki and Willeke (1987) have examined deposition in a cylindrical thin-walled probe, which was placed in the aerosol stream. Their measuring arrangement was isoaxial, i.e. the stream was parallel with the axis of the probe. They observed that significant deposition occurs just downstream of the inlet entry. The phenomenon cannot be explained by the model of fully developed pipe flow. Strong dependence between deposition and Stokes number was found. Deposition is associated with inertial and gravitational effects, because particles cannot move along distorted air flow next to the probe entry. Deposition may occur by impaction or gravity when particles deviate from air streamlines and penetrate to the slowly-flowing boundary layer. Hangal and Willeke (1990) extended the research to non-isoaxial conditions and developed a depos ition model for non-zero angle between the probe axis and aerosol stream.

Because relevant literature seems not to be available, proper estimation of deposition by impaction and gravitation in the diffuser could not be made in the present analysis. Thus, the calculations were carried out using the models given by Hangal and Willeke. Porter et. al. (1991) applied the isoaxial model of Okazaki and Willeke in their sampling system, which was equipped with a diffuser. They calculated particle losses using the diffuser base diameter as a tube diameter in the model. Their experimental data were not in good agreement with the model in highly subisokinetic (i?=5-10) sampling conditions (for definition of R, see Formula (A. 2)). However, quite good agreement was obtained when R*2. The formulae of Hangal and Willeke are in Appendix 2.

20 FINNISH CENTRE FOR RADIATION STUK-A130 AND NUCLEAR SAFETY

Gravitation may produce deposition downstream from the inlet, too. This effect is not significant, since the tube is quite short and flow velocities are high.

Turbulent deposition

Since the flow inside the tube is turbulent, the particles are subjected to turbulent diffusion. Because of their inertia, the particles cannot follow the flow; thus, they may hit the wall and deposit. The turbulent deposition velocity increases as the intensity of ihc turbulence increases. Turbulent deposition is most significant for large particles (>10 urn).

A theory is presented for the turbulent deposition in the cylindrical tube, which is in good agreement with experimental results (Agarval and Liu, 1974). The theory assumes, that the flow is fully developed. Flow separation in the diffuser generates excessive turbulence, but no theory for calculating turbulent deposition in the expanding diffuser was found in the literature. Besides, the direction of coming flow affects the degree of flow separation in the diffuser. The theory of Agarval and Liu, applied to the cylindrical tube behind the diffuser, suggests that the deposition is negligible, since the diameter of the tube is quite large.

Browman diffusion

For particles below 0.1 um in diameter the gravitation and inertial effects are negligible. The most important deposition mechanism is Brownian diffusion. It has no significance since the diameter of the tube is large, the tube itself is short and flow velocity is high.

Electromagnetic forces

Electromagnetic forces between the tube and the particle may be significant. Radioactive particles are generally very charged. It has been shown, that the importance of electromagnetic forces between the sampling tube and the micrometer-class particles are insignificant, if the tube is made from conductive material (Liu et al., 1985). In the vehicle of STUK the tube is made from stainless steel and it is connected to ground.

Total sampling efficiency

The total sampling efficiency of the tube is the aspiration efficiency of the inlet times the penetration efficiency of the tube. The aspiration efficiency can be calculated with a valid and applicable theory, but the deposition losses cannot be properly estimated. To get some idea of the losses, the theory of Hangal and "Willeke (1990) is used to get estimates of inertial and gravitational losses near the entry of the probe. The losses are calculated using the diameter of the diffuser base as a tube diameter. No method is available to estimate the turbulent losses in the diffuser. Probably, the losses are not large, if the probe is aligned with the coming flow. For large yaw orientations in respect

21 FINNISH CENTRE FOR RADIATION AND NUCLEAR SAFETY STUK-A130 to the coming flow the effects of flow separation may be significant. The calculated sampling efficiencies are in Figures 4-5.

3-j 100 km h-1 o 2.5- •80kmK1

1 effi c 60 km h" C I-O1 N 1 O 1 — ^•' ** — "~ 40 km h" ra 1 - ,- Ä — — — £ ' • 30 km h-1 "5. ' 20 km h"1 3 0.5- 0- (3 10 20 30 40 50 Particle aerodynamic diameter, urn

3 .100 km IT1 2.5 • .80 km h"1

ienc y • — -• — —" *"~ "~ 2 * — — -"*" 60 km h"1 1.5 40 km IV1 1 30 km h"1 20 km h"1 0.5 Aspiratio n e f

0 0 10 20 30 40 50 Particle aerodynamic diameter, urn

Figure. 4. a, Aspiration ejpciency at the entry of the diffuser, no wind, sampling flow 10 I s~'. The isokinetic condition is achieved (Formula A. 1) when the velocity of the vehicle is 35 km h~'. If the velocity is greater the sampling is sub-isokinetic and the collection efrciency of large particles is more than 100%. b) Aspiration efficiency at the entry of the diffuser, side wind 10 m s'1, sampling flow 101 s'1. The isokinetic condition is again achieved when the velocity of the vehicle is 35 km h'1. The curves are much steeper than in the Figure 4a.

22 FINNISH CENTRE FOR RADIATION STUK-A130 AND NUCLEAR SAFETY

2.5 -7

^- —• . s* -——_ 2- 100 km h'1

— .^, ^—_- 1 1.5- 80 km h" _--— 60 km h'1 g efficienc y 1 - — ••— —" .— 1 ^=: 40 km h"

— ^. —-—, 1 0.5- •—=— 30 km h" Sampli n 20 km h"1 n

() 10 20 30 40 50 60 Particle aerodynamic diameter, um

Particle aerodynamic diameter, um

Figure 5. a) The "total" sampling efficiency of the tube, no wind, sampling flow 101 s'1. Turbulent deposition and particle bouncing are not taken into account. The curves are not very steep, indicating that the sampling is quite representative at the di iving speed 40-60 km h'1. b) The "total" sampling efficiency of the tube, side wind 10 m s'1, sampling flow 101 s~'. Turbulent deposition and particle bouncing are not taken into account. The harmful effects of strong side wind can clearly be seen. Particles below 20 jim may be highly oversampled and particles above 30 fim are undersampled. Large particles above 50 fitn are totally lost because ofimpaction.

23 FINNISH CENTRE FOR RADIATION AND NUCLEAR SAFETY STUK-A130

2. 4 Improving the inlet

The Figures 4 and 5 show, that the orientation of the coming flow affects sampling efficiency. The effect of strong side wind is inconvenient, since the large particles tend to be lost and consequently, the representativeness of the sample is poor. The losses of large particles may exceed the present estimates, since the effect of turbulence is not taken into account at yaw angles.

Effect of side wind is reduced, if a large-diameter shroud is installed around the inlet (Figure 6). Anand et al. (1991) have presented and tested this type of sampling inlet, hi the shroud the velocity of air is adjusted to 50% of the free stream velocity of air. In principle, the flow at the entry of the shroud is sub-isokinetic, but because the diameter of the shroud is large, the differences in aspiration efficiency remain moderate for particles up to 30 |im. The main objective of the shroud is to align the streamlines along the diffuser and to compensate the harmful effects of the side wind. The flow field should not be disturbed at the entry of ths diffuser. The round edges of the shroud prevent the flow to separate from the surfaces.

The total sampling efficiency of the shrouded probe is given in Figures 7a) and 7b). The aspiration efficiencies are calculated as if all the tubes would be thin-walled. The inlet losses in the shroud are calculated from the theory of Hangal and Willeke (1990) and the losses in the diffuser from the theory of Okazaki and Willeke (1987). Turbulent losses are again neglected due to lack of relevant literature.

Figure 6. Shrouded probe. The dimensions of the diffuser are the same as in Fig 1. The free stream velocity of air is decreased inside the shroud. When the edges of the shroud are rounded the flow separation inside the shroud is decreased. The ef'.cts of side wind are reduced, because the shroud aligns the coming flow along the diffiser axis.

24 FINNISH CENTRE FOR RADIATION STUK-A130 AND NUCLEAR SAFETY

a) 1.6 JOJHTI II1 •• ..— 1.4 SOkmrf ._____ 60 km h'1 ~ .

| 0.8 ^•* c

0.2

10 20 30 40 50 60 Particle aerodynamic diameter, urn

b) 1100 km h"' 1.6 - 80 km h"1 ^\ "S 1.2 - ^—' N s ^40 km h"1 N ^ N k 0.4 - \ 0.2 - 0- ____ 10 20 30 40 SO 60 Particle aerodynamic diameter, urn

Figure 7 a) The "total" sampling efficiency of the shrouded probe, no wind, sampling flow 101 s'1. It is assumed that the velocity inside the shroud is 1.5 times lower than the free stream velocity of air. The sampling seems to be quite representative at the velocity of 80 km h'1. b) The "total" sampling efficiency of the shrouded probe, side wind 10 m s'1, sampling flow 101 s''. It is assumed, that the velocity inside the shroud is 1.5 times lower than the free stream velocity component parallel to the shroud axis. When compared with the Figure 5b, the sampling is more representative

25 FINNISH CENTRE FOR RADIATION AND NUCLEAR SAFETY STUK-A130

2. 5 Filtration efficiency

The sampling efficiency of a glassfiber filter is practically 1 if the particle diameter is above 0.2-0.5 um. There is a penetration window between the particle sizes of 0.01 - 0.2 |im. The filtration efficiency can be calculated from the theory presented by Hinds (1982), but the results are not presented here. They predict that the minimum sampling efficiency is about 30% at the particle diameter of 0.05 um.

2. 6 Tests

The experience to use an air sampler in a moving vehicle is good. However, the sampling system has not been throughly tested for its sampling efficiency. Some tests have been made, where radioactive radon progenies are collected on a filter that was examined with the HPGe spectrometer installed in the vehicle. The test gave reasonable estimates for outdoor radon progeny concentration. After field trials, dust is observed on the inner wall of the diffuser. Clearly, some large particles are lost in the sampling procedure.

A test was performed to ensure that the flow field through trie filter is uniform. An aerosol sample was collected from indoor air, which contained about 5 Bq m'3 radon (isotope 222Rn) and its radioactive daughters. The filter was divided radially in two parts. Both parts were examined with a gamma-ray spectrometer. The activity collected on both parts was proportional to the filtration surface. The test shows that the flow field above the filter is uniform, when the vehicle is not moving. The radioactive daughters of radon are often bound in small particles around 0.01-0.1 urn.

2. 7 Discussion

Factors affscting sampling efficiency and representativeness of the sample were examined. Sampling efficiency was described with theoretical models. Test arrangements covering all sampling conditions would be difficult. Laboratory tests should be made to get better estimates for sampling efficiencies.

The aspiration efficiency is above 100% at the normal driving speed (60-80 km h"1), i.e. the sampling is subisokinetic. However, enhanced concentration of large particles at the entry of the probe is not necessarily a drawback, because they are more likely deposited on the walls of the tube. Also, this error is conservative improving detection limits. Proper estimate of deposition on the walls could not be made. The most appropriate model found in the literature predicts, that sampling is quite representative for the particles smaller than 30 um in aerodynamic diameter. It was also shown that the sampling efficiency is greatly affected by strong side wind. Lots of deposition

26 FINNISH CENTRE FOR RADIATION STUK-A130 AND NUCLEAR SAFETY occurs on the inner walls of the diffuser. Thus, for more detailed analysis this may be a good place to take swipe samples in a fallout situation.

The sampler can be improved and the harmful effects of side wind can be reduced by installing a large-diameter (200 mm) shroud around the diffuser (Figure 6). Its function is to align the streamlines along the diffuser axis. However, the long shroud may introduce practical problems for the driving safety.

After the Chernobyl accident the largest hot particles found in Finland were 7 urn in diameter. The densities of these particles were around 10 g cm'J, giving 20-30 um for the aerodynamic diameter. For these particles the assumption of 100% sampling efficiency seems reasonable in calm wind conditions.

27 FINNISH CENTRE FOR RADIATION AND NUCLEAR SAFETY STUK-A130

3. GAMMA-RAY SPECTROMETRIC ANALYSIS OF THE FILTER IN THE FIELD

Air sampling and gamma-ray spectrometric analysis with an HPGe detector are an accurate and rapid way to obtain nuclide-specific information about the sample. The sampling site, however, may be far away from the nearest laboratory. To get prompt results, the analysis must be carried out on the spot.

3. 1 Gamma-ray spectrometric measuring system of the emergency vehicle of STUK

The gamma-ray spectrometric measuring system of the emergency vehicle of STUK is presented in Figure 8. The operation of the detector has given some problems, but generally the experiences are good. The detector is insulated from the body of the

Lead Notebook shielding PC

Detector

Portable Liquid spectroscopy nitrogen unit dewar Figure 8.77; e gamma-ray spectrometric measuring system of the emergency vehicle ojSTUK. Detector and preamplifier (ORTEC POP-TOP; efficiency 37%) are inside a lead shield with a lid. The detector operates at low temperatures; it is cooled in a liquid nitrogen dewar. ORTEC Nomad 92 X-P gamma-ray spectrometric unit contains high voltage supply, linear amplifier, digital spectrum stabilizer and multichannel analyzator. Nomad feeds the high voltage (3000 V) for the detector and power fir the preamplifier and receives the signal from preamplifier. Nomad is controlled with a PC via parallel port or via bus card and cable. Nomad is driven with a 230 VAC or 12 VDC. It has also a battery that lasts for 8 hours.

28 FINNISH CENTRE FOR RADIATION STUK-A130 AND NUCLEAR SAFETY vehicle with plastic foam to attenuate vibration. The support structure is satisfactory, because the vibration does not decrease the resolution of the measurement and the detector dead-time is not increased too much. The widths of the photopeaks are broadened 10 - 20%. The dead-time of the detector is increased approximately 1%, when driving on a smooth asphalt road. If the road is bumpy, the dead-time is occasionally increased as much as 10%. The detector has broken down twice during an operation of two years. At the first time the vacuum escaped inside the detector capsule, which is a typical fault of ORTEC POP-TOP detectors also in laboratory use. The second failure was due to a bad contact in the preamplifier. This was most likely caused by excessive vibrations.

The thickness of the lead shield is 5 cm and the thickness of the lid is 3 cm. The shield attenuates the background radiation by a factor of 3 to 40 depending on the energy.

The activity of a filter from the DWARF-sampler can be measured on a holder (see Figure 9), which is placed on the top of the detector. The holder ensures that the measuring geometry is the same for all filters and the sample is not moved during the measurement. Calibration factors are calculated numerically with DECCA-program (Aaltonen et ai., 1994). It calculates detection efficiencies for arbitrary cylinder geometries. Also special beakers (cylindrical. 42 mm in diameter, so-called williams- beakers) can be employed. The detector is calibrated for this geometry with solutions, but williams-beakers are cumbersome to use, since the filter must be folded before it goes in to the beaker. However, the williams-geometry is approximately 80% more efficient than the filter holder-geometry. The detection efficiencies for williams- and filter holder-geometries are given in Table II.

0100

open/ O-ring

filter

Figure 9. A filter holder made from plastic. The shortest distance between the end of the HPGe capsule and the filter is 2 mm.

29 FINNISH CENTRE FOR RADIATION AND NUCLEAR SAFETY STUK-A130

Table II. Detection efficiency of the HPGedetector in the emergency vehicle of STVKfor different geometries.

Energy, Efficiency Efficiency keV williams- filter holder, beaker, (Bqs)1 (Bqs)1

50 5.47E-02 9.50E-03 60 7.58E-02 2.58E-02 80 1.28E-01 7.00E-02 100 1.75E-01 1.05E-01 110 1.89E-01 1.19E-01 150 1.88E-01 1.14E-01 200 1.53E-01 9.33E-02 300 1.10E-01 6.57E-02 500 7.03E-02 4.35E-02 700 5.32E-02 3.31E-02 1200 3.37E-02 2.08E-02 2000 2.22E-02 1.36E-02 3000 1.58E-02 9.81E-03

3. 2. Minimum detectable activities of some nuclides

Tlieory

In a laboratory the background count rate is normally known with very little systematic error in long background measurements. However, a long background measurement is not useful in a moving vehicle, since the background level may drastically differ with time and place. The background measurements must be short, which may lead to significant statistical uncertainties. It affects the accuracy and increases the detection limits of the measurements.

Normally, the photopeaks associated to nuclides released in an accident cannot be seen in the background measurement (in long measurements 137Cs may be detected because of global fallout). Photopeaks generated by natural radioactive nuclides usually do not interfere with the detection of release nuclides, since HPGe-detector has high resolution. Thus, the minimum detectable activity (MDA) of a nuclide being monitored can be calculated directly from the background level of the channels corresponding to the energy of gamma transition of the nuclide (region of interest). Only one measurement is needed to determine the activity of the nuclides in the sample or their MDA.

30 FINNISH CENTRE FOR RADIATION STUK-A130 AND NUCLEAR SAFETY

In this report the following formula is used for MDA:

(7) Formula (7) is equal to the critical level (cc=O.Ol) determined by Currie (1968) assuming normal distribution of background counts. Assumption of normal distribution is valid if number of counts is high enough, hi equation (7) the MDA is given in number of counts. Number of counts (N) can be converted to activity concentration (Ae) using amount of sampled air (V), counting time (t), gamma yields of photons (y) and calibrated efficiency factors from Table II (EF).

A <8> Because of fallout, not only the environment but also the measuring vehicle may be contaminated. Several photopeaks associated to release nuclides might be detected in the background measurement. In case of contamination, or if it is suspected, a separate background measurement must be carried out and it must be subtracted from the results of the actual sample measurement. Consequently, the analysis takes a longer time and the MDA's are higher. Local and chronological changes in the background spectrum may be large. Thus, the background measurement must be carried out in adequately short intervals. Every second or third spectrum should be collected for background only.

If the number of pulses in photopeak and in background are assumed to be normally distributed, MDA in twofold measurement is obtained from:

r2

\

where Nbb is the number of pulses in background below photopeak in background measurement, Nbpeak is number of pulses in photopeak in background measurement, Nab is number of pulses in background below photopeak in actual measurement, ta is the measuring time of actual measurement and tb is measuring time of background measurement. Note, that Equation (9) is reduced to (7) if ta « t b and N b?cak = 0. Equation (9) is derived by making a null hypothesis Naptak-NbtPealr0 and calculating for it a one-directional 99% confidence limit.

Estimating the background

To calculate the MDA's, information about background count rates is needed. In

principle, Nb in the Equation (7) is obtained directly from any measured spectrum. However, the calculated background count rates in the region of interests refer only to the situation and place, where the measurement was carried out. In other conditions

31 FINNISH CENTRE FOR RADIATION AND NUCLEAR SAFETY STUK-A130

(different activities of samples, changes in environmental radioactivity) the measured count rates are not fully representative.

For the background calculations a spectrum was measured outdoors in the Helsinki area in summer 1994. The measuring time was 60 min. The vehicle was not moving during the measurement and the engine was not running. A particulate sample (filter), which was taken with the sampling system of the vehicle, was placed on the detector. In this way the effect of radon daughters was taken into account. During the sampling the 222Rn concentration in air was about 2 Bq m"3. The results of background measurement are in Appendix C.

Estimating the MDA's beforehand from Equation (9) is difficult, since the count rates in a photopeak and in the background below it are generally not known. The following assumptions were made: (1) the background spectrum is similar to that presented in Appendix C and (2) no photopeaks associated to release nuclides are detected. Thus, in reality the MDA's may be somewhat higher.

Results

The calculated MDA's of some release nuclides in onefold measurement (no separate background measurement) are given in Table III and in twofold measurement (separate background measurement) in Table IV, respectively.

3. 3. Discussion

In sampling and gamma-ray spectrometric analysis of the filter the MDA's are typically 0.05 - 0.5 Bq m'3 in a 10 minute analysis of a filter through which 6 m3 air has passed (ampling time about 10 minutes). Quantitative nuclide-specific data are obtained almost in real-time. Clearly, this method is very useful in an emergency.

Tables III and IV give good estimates for MDA's when no release nuclides are detected. Some variation in background occurs due to natural radioactivity, but it does not have a large affect on MDA's. If the vehicle and the environment are badly contaminated, the MDA's are higher, of course. Then some nuclides, which have lower gamma yields or lower concentrations in the environment may be hidden below the elevated background generated by the overall contamination.

32 FINNISH CENTRE FOR RADIATION STUK-A130 AND NUCLEAR SAFETY

Table III. M.nimum detectable activities for some release nuclides in gamma-ray measurement of a filter. Tlie sampling time is 10 minutes andfmv rate through the sampler is 0.010 m3 s'1. Sampling efficiency is assumed to be 100% for all nuclides. MDA according to Formula (7), Bq m3 energy. gamma- measurins time (min) nuclide keV yield 10 30 60 Ce-144 133.5 0.108 4.8E-01 2.8E-01 2.0E-01 Ce-141 145.5 0.484 1.1E-01 6.4E-02 4.5E-02 Te-132 228.2 0.882 7.7E-02 4.5E-02 3.2E-02 La-140 328.8 0.207 4.2E-01 2.4E-01 1.7E-01 Cs-136 340.6 0.468 1.9E-01 1.1E-01 7.7E-02 1-131 364.5 0.812 1.1E-01 6.1E-02 4.3E-02 Sb-125 428.0 0.296 3.3E-01 1.9E-01 1.4E-01 La-140 487.0 0.459 2.3E-0I 1.3E-01 9.3E-02 Ru-103 497.1 0.864 1.2E-01 7.2E-02 5.1E-02 1-133 529.9 0.863 1.3E-01 7.5E-02 5.3E-02 Ba-140 537.4 0.199 5.7E-01 3.3E-01 2.3E-01 1-132 552.6 0.161 7.2E-01 4.1E-01 2.9E-01 Cs-134 604.7 0.976 1.1E-01 6.6E-02 4.6E-02 Ru-106 621.8 0.098 1.0E+00 5.8E-01 4.1E-01 1-131 637.0 0.0727 1.4E+00 8.0E-01 5.6E-01 1-132 667.7 0.987 1.1E-01 6.5E-02 4.6E-02 Sb-127 685.5 0.357 3.2E-01 1.8E-01 1.3E-01 Te-129m 696.0 0.029 4.0E+00 2.3E+00 1.6E+00 Zr-95 756.7 0.548 1.9E-01 1.1E-01 7.7E-02 Nb-95 762.2 0.99 1.0E-01 6.1E-02 4.3E-02 1-132 772.6 0.762 1.4E-01 7.9E-02 5.6E-02 Cs-136 818.5 1 1.1E-01 6.3E-02 4.4E-02 1-135 1132.0 0.225 6.0E-01 3.5E-01 2.5E-01 1-135 1260.0 0.286 5.2E-01 3.0E-01 2.1E-01 La-140 1596.2 0.955 1.6E-01 9.0E-02 6.4E-02

33 FINNISH CENTRE FOR RADIATION AND NUCLEAR SAFETY STUK-A130

Table IV. Minimum detectable activities for some release miclides in gamma-ray measurement of a filter. T\\e background is taken into account (Formula (9). Background measuring time is 5 min. The sampling time is 10 minutes and flow rate through the sampler is 0.010 ms s~'. Sampling efficiency is assumed to be 100% for all miclides.

MDA according 10 formula (9), Bq trf3 energy. gamma- measuring time ("mini nuclide keV yield 10 30 60 Ce-144 133.5 0.108 7.4E-01 6.2E-01 5.9E-01 Ce-141 145.5 0.484 1.5E-01 1.3E-01 1.2E-01 Te-132 228.2 0.882 1.1E-01 9.6E-02 9.0E-02 La-140 328.8 0.207 5.9E-01 4.9E-01 4.6E-01 Cs-136 340.6 0.468 2.7E-01 2.2E-01 2.1E-01 1-131 364.5 0.812 1.6E-01 1.3E-01 1.2E-01 Sb-125 428.0 0.296 4.7E-01 3.8E-01 3.6E-01 La-140 487.0 0.459 3.2E-01 2.6E-01 2.4E-01 Ru-103 497.1 0.864 1.7E-01 1.4E-01 1.3E-01 1-133 529.9 0.863 1.8E-01 1.5E-01 1.4E-01 Ba-140 537.4 0.199 8.0E-01 6.5E-01 6.0E-01 1-132 552.6 0.161 1.0E+00 8.2E-01 7.6E-01 Cs-134 604.7 0.976 1.6E-01 1.3E-01 1.2E-01 Ru-106 621.8 0.098 1.5E+00 1.2E+00 1.2E+00 1-131 637.0 0.0727 2.1E+00 1.7E+00 1.6E+00 1-132 667.7 0.987 1.7E-01 1.4E-01 1.3E-01 Sb-127 685.5 0.357 4.7E-01 3.9E-01 3.7E-01 Te-129m 696.0 0.029 5.9E+00 4.9E+00 4.7E+00 Zr-95 756.7 0.548 3.2E-01 2.8E-01 2.6E-01 Nb-95 762.2 0.99 1.8E-01 1.5E-01 1.5E-01 1-132 772.6 0.762 2.3E-01 2.0E-01 1.9E-01 Cs-136 818.5 1 1.7E-01 1.5E-01 1.4E-01 1-135 1132.0 0.225 9.8E-01 8.4E-01 8.0E-01 1-135 1260.0 0.286 7.8E-01 6.5E-01 6.2E-01 La-140 1596.2 0.955 2.2E-01 1.8E-01 1.6E-01

34 FINNISH CENTRE FOR RADIATION STUK-A130 AND NUCLEAR SAFETY

4. RAPID CONCENTRATION ESTIMATION OF BETA-ACTIVE NUCLIDES USING SIMPLE DEVICES

4. 1 Introduction

Gamma-ray spectrometry is a complicated measuring method in the field. The use of a spectrometer requires trained personnel and the devices must be handled gently. Because of the limited resources in the aftermath of an accident the use of gamma-ray spectrometers is possible only for a few teams.

Besides the nuclide-specific data, also the total beta activity concentration in air is a relevant quantity in emergency situations. The beta activity concentration can be measured with simple devices. In principle this method gives sufficiently sensitive and accurate estimates for airborne activity. The largest problem is to differentiate the artificial beta activity from the natural activity (daughters of 222Rn and 220Rn). Beta activity concentration of these nuclides in air is normally 0.1 to 10 Bq m'3. The half- lives of radon daughters are quite short, some tens of minutes. This makes it possible to separate their count rate from that caused by the relatively long-lived artificial nuclides.

The present approach describes a measuring procedure which can be used to estimate the total beta activity concentration in air with simple, robust and relatively cheap devices. The use of the system is easy to leam. A computer code facilitates the analyses of the measured results.

4. 2 Measuring devices

To measure the total beta activity concentration in air the following devices are needed: a) an air sampler and air filters b) a beta counter c) a stand for the beta counter and for the filter to guarantee, that all the measurements are carried out in similar geometry. Two possible measuring procedures are presented in Figure 10 a and 10 b. The assembly in Figure 10 b is more useful, since it allows to perform real-time measurements. Sampling and simultaneous beta counting are easy to realise in portable equipment.

The technical details of DWARF-sampler are given in Table I (p. 15). The beta probe GMP-11 is connected to RDS-120 universal survey meter (Rados Technology, Turku, Finland). Details of the probe are presented in Table Y. The beta counter is sensitive also to X- and gamma rays. The end window of the probe is so thin that also highly

35 FINNISH CENTRE FOR RADIATION AND NUCLEAR SAFETY STUK-A130

Table V. Technical information about GMP-11 beta-probe (Rados Technology, Turku).

Diameter of the window 27.8 mm Thickness of the window (mice) 1.5 -2.0 mg cm"2 Inner diameter of geiger tube 30.9 mm The sensitive length of cathode 37mm Fill gas neon + argon; halogen quenched Recommended high voltage 595 V Dead time (HV 595V) 190 us

energetic alpha particles may be detected. However, its sensitivity to high energy gamma rays is poor since the dimensions of the Geiger tube are small.

4. 3 Calibrations

The equipment must be calibrated to get a conversion factor between the count rate and the beta activity within the filter. The calibration factor differs from nuclide to nuclide, because beta and gamma yields and transition energies are different. In the present analysis the calibration was made only for one artificial nuclide 144Ce. A calibration was made also for radon progenies to understand their effect on the count rates.

a) b) *V Belta probe (thin \ window Geiger) Beta probe / / Fitter Stand Stand sampled ' / / separately ^\ —n -Elter LJ h : ,1 Aluminiunl plate The inlet of DWARF-sampler Figure 10. a) A stand for the filter and beta counter (assembly A). In this system the sampling and beta counting are separated, b) On-line measurement of the filter activity (assembly B). A stand for beta counter is placed on the air sampler so that the beta counter hangs over the filter. Tlie air sampling and beta counting are simultaneous processes. The results and the feedback for decisions (=e.g. how long sampling should be continued for good results) are obtained almost immediately.

36 FINNISH CENTRE FOR RADIATION STUK-A130 AND NUCLEAR SAFETY

Table VI. The most significant photon and beta transitions of 1A4Ce and 144Pr.

"Te- tvge vield (YBq sV1) max. enerev averaee enerev P- 0.196 184.7 keV 49.35 keV P- 0.0460 238.1 keV 65.26 keV P" 0.758 318.2 keV 90.23 keV Y 0.0348 38.13 keV Y 0.108 133.5 keV Y 0.0533 91.54 keV X 0.540 36.03 keV X 0.296 35.55 keV 144Pr tvpe vield (YBq s) -1) max. enerev averaee enerev P" 0.0108 810.3 keV 266.9 keV P- 0.0117 2300 keV 894.7 keV P- 0.977 2996 keV 1222 keV Y 0.0148 696.5 keV

Calibration based on luCe

The detection efficiencies of the measuring assemblies were obtained by measuring a f lter impregnated with 144Ce solution. I44Ce is a beta active isotope (half-life 285 d). It has a short-lived daughter '""Pr (half-life 17 min). Thus, 144Ce and 144Pr were in equilibrium in the filter. The most significant beta, gamma and X-transitions of 144Ce and 144Pr are listed in Table VI. The filter was covered with a thin plastic layer. The energies of the beta particles emitted by UACe are quite low, so most of these beta particles are absorbed in the filter, plastic foil or window of the Geiger tube. They are not taken into account in the calculations. However, the detector is sensitive to brehmsstrahlung generated by beta particles in the materials around the beta counter. Backscattering of beta particles can contribute to the overall counting efficiency, too.

Calibration based on radon progenies

The most important natural radioactive nuclides in air are 222Rn and its short-lived daughters. Radon progenies are solid and they are typically attached to 0.05 - 0.25 um airborne particles. There are also small amounts of another radon isotope (220Rn) and its progenies in air. The simplified decay chains of 222Rn and 220Rn are presented in Figure 11. The most significant beta and gammatransitions of progenies 214Bi, 214Pb, 212Pb and 208Tl are in Table VII.

37 FINNISH CENTRE FOR RADIATION AND NUCLEAR SAFETY STUK-A130

Table VII. The most significant photon and betatransitions of radon progenies. The beta particles penetrate through a thin window of a geiger tube without significant attenuation. Also the most energetic alpha-particles are likely to be detected.

Progenies of 222Rn:

2UPb: tvne vield fCBa sV'l max. energv average energv CO . 0.480 672.1 keV (207 keV) CO . 0.425 728.1 keV (227 keV) CO . 0.0630 1024 keV (337 keV) X 0.0621 74.82 keV X 0.105 77.11 keV Y 0.0749 242.0 keV Y 0.192 295.2 keV Y 0.372 351.9 keV

tvpe vield (YBa sV'l max. energv average energy CO . 0.0480 1065 keV (352 keV) P'1 0.0828 1423 keV (492 keV) 0.176 1506 keV (525 keV) PCO . - 0.179 1540 keV (539 keV) 0.0752 1893 keV (684 keV) pCO . " 0.177 3270 keV (1270 keV) Y 0.463 609.3 keV V 0.151 1120keV Y 0.158 1765 keV Progenies of 220Rn:

suPb: tvpe vield (YBa si "'I max. energv average energv P-1 0.0522 157.5 keV (41.9 keV) P-1 0.851 334.2 (94.4 keV) p-1 0.0990 572.8 (173 keV) X 0.107 74.82 keV X 0.180 77.11 keV Y 0.446 238.6 keV

t\T>e vield (YBa sV1) max. energv average energv CO . 0.0800 1519keV (531 keV) CO . 0.484 2246 keV (832 keV) X 0.0804 87.30 keV Y 0.118 727.2 keV

MtTL tvne vield aBasV'l max. energv average energv CO . 0.228 1284 keV (438.7 keV) CO . 0.220 1517 keV (532.5 keV) p- 0.509 1794keV (646.5 keV) Y 0.858 583.1 keV Y 0.216 510.8 keV y 0.120 860.4 keV

38 FINNISH CENTRE FOR RADIATION STUK-A130 AND NUCLEAR SAFETY a) b)

1600 a 3.67 d

Alfa decay Alfa decay

222Rn 220Rn Beta decay Beta decay 3.8 d 55.6 s

218Po 2i4Po 216PO 212po 4 3.05 min 1.6-10" s 0.15 s 2/3 310"7s

214Bi 212Bi ,19.7 min i i ,60.6 min 208 214Pb 210pb 1/3 Pb 26.8 min 22.3 a 10.6 h Stable

208Tl i 3.1 min Figure 11. Simplified decay chains oj 222Rn and 220Rn.

Two test samples were collected and the count rate was measured in the assemblies shown in Figures 10 a and 10 b. The tests were carried out in a large garage where 222Rn concentration was about 4.5 Bq m'3. Flow rate through the sampler was 0.01 mV. Sampling times were 30 min for both tests. The background measurements were carried out in the same place before the test.

The activity of progenies in the filter was obtained from gamma-ray spectrometric measurement of a comparison sample. It was collected with another air sampler in the same room at the same time as the actual sample was taken for measurement in stand A. The gamma-ray spectrometric measurement was carried out in several short stages. This allows to detect the time behaviour of the progenies. The 222Rn concentration of the air in the garage was also roughly estimated from the gamma-ray spectrometric results.

The radon concentration in the garage was quite low. Thus, for better statistical accuracy, one measurement was carried out in the premises of STUK. The

39 FINNISH CENTRE FOR RADIATION AND NUCLEAR SAFETY STUK-A130 measurement was carried out using assembly B (Figure 10 b). The sampling time was 31.5 min and the beta counting was started at the same time as the sampling. The radon concentration in the office was about seven times higher than in the garage (30 Bq m'3).

4. 4 Results of calibration measurements

The count rate in the background measurements varied between 40 and 46 cpm. The background was measured for a long time to get the statistical error below 1 cpm.

The results show that the detection efficiency for beta activity of the filter impregnated with ""Ce is 1.401 ± 0.002 cpm Bq"1 for geometry A and 1.383 ± 0.002 cpm Bq'1 for geometry B. Only beta particles emitted by H4Pr are taken into account. The error in activity determination of the filter is 7%. The differences between these two measuring geometries are so small that in practise the detection efficiency for hard beta radiation is 1.4 ±0.1 cpm Bq'1. The error limits refer to the confidence limit of one o.

According to the gamma-ray spectrometric results, the beta activity of the filter 6-32 min after the sampling was 83 ± 4 Bq in average (41 Bq 214Bi, 28 Bq 214Pb. 6.6 Bq 212Pb, 5.1 Bq 212Bi and 2.8 Bq 208Tl). 212Bi was no. detected in gamma- ray spectrometry; its activity is estimated from other nuclides. During this period the average count rate in p measurement was 161.8 ± 2.6 cpm. giving for the detection efficiency 1.95 ± 0.10 cpm Bq'1.

The results of the measurement carried out in the office are presented in Figure 12. For the pulse rate the following simple model is used:

^^\ t < t0, (10)

e'^'''i t>t0. (11)

Ax is a parameter which depends on sampling rate, sampling efficiency and radon concentration in air. Parameter tsamp is sampling time and Xef, effi ctive decay constant. The values of Au Xefi and 10 were determined using GIGAFIT-program (A. Birchall, NRPB, UK). The error of effective decay constant was calculated by SAS GLM- 1 procedure. The calculated values are: Xef,= 0.0151±0.0003 min' in Formula (10), Xefr 1 = 0.0147±0.0003min' in Formula (11), ,4, = 1506 and to = 4.23 min. The estimated errors represent one standard deviation. Because the radon concentration CRn 3 in the office was about 30 Bq m' the value of Ax in the formula is approximately

40 FINNISH CENTRE FOR RADIATION STUK-A130 AND NUCLEAR SAFETY

700 • 600 • measured model 500 /. u • i a. u 400

•g i I 300 / o O 200

100 /

-40 -20 20 40 60 Time, min

Figure 12. Test sampling in the premises ofSTUK. The sampling time was 31.5 min. The model 10-11 is fitted in the data. Parameter values are A, = 1506,

Aefr=0.0147min andto=4.23.

(12)

The model 10 -11 is based on the decay chain of 222Rn. Effective half-life (45 min) is

approximately the sum of the half-lives of the nuclides in the chain. Time shift (f0) is due to 218Po which is almost a pure alfa emitter.

Normally 222Rn progenies are the dominating beta active nuclides in the filter when the sampling is in progress and during the first hour after the sampling. The 220Rn progenies, 212Pb and212Bi, are relatively long-lived (half-lives 10 h and 1 h); so some hours after the sampling they are the most significant beta active nuclides in the filter. Thus, the model 10 can be utilized only during the first hour during the sampling and the model 11 only during the first hour after the sampling. Otherwise, 212Pb and212Bi may be associated to artificial nuclides. In any case their contribution is small and below the detection limit. Besides radon daughters, othernuclides were not detected in the tests (filter in gamma-ray spectroscopy). However, it is possible that dust and dirt, attached to the filter, contain long-lived natural radioactive nuclides (such as 226Ra), and thus they may interfere with the use of the model.

41 FINNISH CENTRE FOR RADIATION AND NUCLEAR SAFETY STUK-A130

4. 5 Artificial activity concentration in air

The model 10-11 can be used to differentiate long-lived artificial nuclides from radon daughters. If the time-behaviour differs from the model (especially if A, is higher than normally or Xer is smaller than normally) the presence of artificial nuclides is suspected.

Let us assume that there are both natural and artificial beta active nuclides in air. The count rate generated by natural nuclides is similar to the model 10-11 and half-lives of artificial nuclides are assumed to be long. During the sampling artificial activity is detected, if the count rate rises faster than predicted by the model (see Figure 13 a). Using notations of Figure 13 a the count rate generated by artificial activity in the filter is calculated as:

After the sampling the situation is analogous; if the count rate decreases slower tiian predicted by the model the count rate generated by artificial activity in the filter is calculated as (see Figure 13 b):

MR-NRru = CRV art X_NR 1

Count rate CR^ is converted to beta activity concentration Cp from the formula:

CR n .[cpm] CJBq m "3]= artV \ (15) l.4[cpm

4. 6 Error prediction and detection limits

The error of the ratio of count rates is calculated as (Knoll, 1979)

QMR L) +( -) • (16) V J MR \\ CR, CR2

Relative errors of count rates are calculated from the formula:

42 FINNISH CENTRE FOR RADIATION STUK-A130 AND NUCLEAR SAFETY

a) 200 -T Radon conc. 2.6 Bq m-3 MR=CR:/CR, Radon conc. 1 Bq m-3 + NR=NC2/NC, artificial 1 Bq m-3 CR, *.— rate , cp i NC: — —• CR, co u Ne t

NC,

20 40 ll 60 80 100 Timc,min 0 -

b) 100 Radon conc. 2.6 Bq m-3 • MR=CR2/CR, NR=NC:/NC, Radon conc. 1 Bq m-3 + CR ":— , artificial 1 Bq m-3 2 — —_ CRl «^ — I v. ^. ^»

•~"--^,

NC2

20 40 60 80 100 Time.min Figure 13. Behaviour of count rate as a function of time (a) during the sampling and (b) aper the sampling. Sampling rate is 0.010 m3 s'1 and sampling time 50 min. The count rates are calculated from Formulae 10 (Figure a), 11 (Figure b), 12 and 15. Also notations used in Formulae (13) and (14) are shown.

43 FINNISH CENTRE FOR RADIATION AND NUCLEAR SAFETY STUK-A130

°CR, CR. t/CR,

where Bg is the background count rate and tb is the counting time of the background. If the number of pulses is high enough (at least 50 net pulses per measurement period) the ratio MR can be assumed to be normally distributed. Then, the value of the ratio MR is less than MRconf at the confidence level of 95%:

MR.^f=NR + 1.64 oMR. (18) com

The detection limit MDA(C/?aM) is then obtained by substituting MRcmf for MR in the Formula (13) or (14) MR ,-NR MDA(CR ,)= ^£ CR , during the sampling,

h

MR.-NR MDA(CR )= 222L CR , after (he sampling. (20) ar \-NR

The detection limits are here in count rates, but they can be converted to activity concentrations using the Formula (15). Here the detection limits refer to the 5% probability of false detection.

If the measuring time is short, the MDA's calculated from the preceding formulae are high. Then, a more efficient way to calculate the detection limit is to use maximum possible activity (MPA):

MPA= — . lA[cpm Bq'l]V[m*]

In other words, all the activity in the filter is assumed to be artificial.

44 FINNISH CENTRE FOR RADIATION STUK-A130 AND NUCLEAR SAvETY

4. 7 Prediction of radon concentration

Radon concentration in outdoor air is typically about 1 Bq m"3, hi some weather types (inversion) the concentration can rise up to 30 Bq m"3. Therefore, the situation, where activity concentration of artificial beta active nuclides in air is higher than 5 Bq m'3, can be diagnosed directly from the total net count rate except during inversion weather condition.

Radon concentration can be estimated from Formula (12) if the parameter^ , is known. It can be obtained from the count rate measurement in some period during or after the sampling (Formulae (10) and (11)). If the measurement is carried out during the sampling the parameter A! is obtained from (see Figure 13 a):

A Ai= (22)

\r (Exp(-Xefr t2)-Exp(-Xer tj)

Similarly, if the measurement is carried out after the sampling, the integral is calculated from Formula (11), and (see Figure 13 b)

A = CR

(23 \rt )) > ef samp''

4. 8 Computer code for the calculation

The present model is too labourious to be used by hand in the field conditions. Thus, a computer code BEACON (BEta Activity CONcentration) was written to cope with these equations and to automate the calculations. The computer is also convenient in data handling and data storage. The BEACON is written using Visual Basic (Microsoft Corp.); thus, running it requires Microsoft Windows (Microsoft Corp.)-

The code divides the measurements carried out during the sampling into two sections. It calculates then the artificial activity concentration, its MDA and 222Rn activity concentration from count rate ratios between these two sections. The two sections are selected so that the MDA is the lowest possible. The measurements after the sampling are handled in the similar way. The MPA is calculated only from the last measurement period.

45 FINNISH CENTRE FOR RADIATION AND NUCLEAR SAFETY STUK-A130

The data transmission from a beta counter to computer can be automated. However, in the present work it was not possible, because some hardware and software development should be made by the manufacturer before the instrument (GMP-11) can be used automatically. So the data points must be fed manually.

4. 9 Detection limits

The detection limits of the method depend on the 222Rn and background count rate and measuring and sampling time. The detection limits (MDA:s and MPA:s) are given as a function of sampling time in the Figures 14 a and 14 b.

4.10 Reliability of the method

In this chapter a method was described for rapid determination of artificial beta activity concentration in air. The method utilizes simple devices which can be used with little training. The detection limits are low enough to cope with the requirements of the emergency preparedness. A hazardous contamination in air can be detected almost immediately. The computer code BEACON is essential for simplifying calculations. The use of the simple Formula (11) decreases detection limits if the sampling time is longer than 15 min or if the count rates are high enough. Formula (11) can be used after the sampling to ensure that the detection is real.

The method is calibrated for 144Ce. The results will, however, differ from nuclide to nuclide. Also the 222Rn concentration, calculated by this method, is not strictly correct, since ~Rn itself is not detected. So the result depends on the ratios between 222Rn and its daughters in air, which can vary considerably.

The variations between 222Rn and its daughters will also affect the parameters in the Formulae (10-11). However, all the tests show that the model describes satisfactorily the time-behaviour of the count rates when only natural radioactivity is present. A test, where artificial radioactivity is present, could not be arranged. The model cannot be used for long sampling or measuring times. This method, however, is designed for emergency purposes. The results must be produced fast. Another limitation of this model is that the assumptions are not valid in the presence of short-lived artificial radionuclides. A significant contribution of these nuclides is possible only if the emission source is close to the sampling site. The serious situation is most probably diagnosed directly from the tot,".5 net count rate and substraction of natural radioactivity is not needed.

46 FINNISH CENTRE FOR RADIATION STUK-A130 AND NUCLEAR SAFETY

a) 1 10 20 20 -i l ft °0 \ \ \ 1 MDA 7 - *~—— 16 - MPA 14 - —•—=: • ————__^ 12 - x £ s \ 10 - \ Q 10 __ 1 •s. o 6 - 1 ~ | 4 - ~ —— __ 1 1 - i — 0 - () 10 20 30 40 50 60 Time, min

b) 20 1 1 MDA . 7 li \\ Ifi - Tl - MPA 14 - \\\ ~~~- . 12 - \ ~~- 10 - 1 > s —-= —=== 8 - _LQ_ \ r- * v •-• 6 - != • \ \ 4 - — , r~ - -.

MD A o r MPA , B q n —~=== 2 - 1 — — ~ —i- ======0 10 20 30 40 50 60 Time, min Figure 14. a) Detection limits as a function of measuring time (during the sampling). MDA (minimum detectable activity) is calculated from Formula (20) and MPA (maximum possible activity) from Formula (22). The values of parameters were as follows: sampling rate 13.5 I s'1 and background count rate 45 cpm ± 3 cpm. No artifcial nuclides were assumed, but the radon concentration was varied (the numbers in the graph). The Formula (20) is useless, if the radon concentration is low (below J Bq m'3) since a lower estimate for artificial activity concentration is obtained from MPA. However, if the concentration is higher (10 - 20 Bq m'3) the MDA decreases below MPA within half an hour, b) Detection limits as a function of measuring time after the sampling. MDA is calculated from Formula (21) and MPA fiom Formida (22). The values of parameters were as follows: sampling rate 13.5 Is'1, background count rate 45 cpm ± 3 cpm and the sampling time 30 min. The beta measurement was started 2 minutes after finishing the sampling. Tlie MDA is again useless, if radon concentration is low. The MDA decreases much faster as a function of time than in Figure 14a). FINNISH CENTRE FOR RADIATION AND NUCLEAR SAFETY STUK-A130

5. IN-SITU MEASUREMENTS USING HIGH- RESOLUTION GAMMA-RAY SPECTROMETER 5.1 Introduction

Environmental radioactivity can be monitored in-situ with a gamma-ray spectrometer. This is a well-known method and it has been widely used after the Chernobyl accident. However, the applications have mainly emphasized the measurement of fallout. The use of in-situ gamma-ray spectrometry for the measurement of airborne radioactivity provides many advantages that are most useful at the early stages of an accident. The results are readily available and the problems in sampling are completely avoided. Also the nobel gases can be detected. The measurement itself is easy to perform and it gives opportunities for highly automated measuring routines.

The in-situ measurements are generally much less efficient than sample measurements. However, in serious emergency situations, the count rates can be so high, that very short (tens of seconds) measuring periods give adequate statistical accuracy. Using normal routines the analysis of a spectrum and report generation will take five minutes at minimum. Thus, this stage is the bottleneck, that limits the rate of information collection. For in-situ measurements the data collection can be totally automated. The data analysis and reporting can also be automated with modern software, although there is a possibility for false identifications. The automatic combination of the results with a location label, received from GPS-navigator, makes the system very powerful in an emergency situation.

5. 2 Methods

The detector in the emergency vehicle of STUK can be shielded in three different ways:

1 Unshielded (Fig 15 a). The lead shield around the detector is totally removed. 2 Partially shielded (Fig 15 b). The lead shield is around the detector, but the lid of the shield is open. 3 Shielded (Fig 15 c). The lead shield is around the detector and the lid is closed.

In the first option the detector monitors the whole environment around it. The attenuation created by the body of the vehicle is not very significant. This shielding option is not practical for detection of airborne radioactivity, since usually the gammafiux coming from the ground is much larger than the gammaflux from airborne activity (large background). The airborne radioactivity produces often in a few hours

48 FINNISH CENTRE FOR RADIATION STUK-A130 AND NUCLEAR SAFETY

Figure 15. Difirent measuring options in the emergency vehicle ofSTUK.

fallout that gives stronger contribution to the unshielded detector than the radioactive cloud itself. The unshielded geometry is useful only in fallout measurements and in searching lost point sources.

The option in Figure 15 b is optimal for detecting airborne radioactivity. The disturbing gammaflux from fallout and natural radioactive nuclides on the ground is strongly attenuated, while the photons coming from the airborne radionuclides are efficiently detected.

The third shielding option (Figure 15 c) is useful in air sampling applications and as a background measurement for the actual in-situ measurement carried out in the geometry of Figure 15 b. If the vehicle and the environment are seriously contaminated, the photopeaks generated by the release nuclides are seen in the background measurement. Thus, these photopeaks are subtracted from the actual measurement. The percentage of unscattered photons that penetrate through the lid without scattering must be determined.

49 FINNISH CENTRE FOR RADIATION AND NUCLEAR SAFETY STUK-A130

5. 3 Theory

The theory used here is adopted from Sowa (1990), since it is directly applicable for the measuring principle of the vehicle. Originally the theory of in-situ gamma-ray spectrometry was presented by Beck et al.(1972).

The basic problem in in-situ gamma-ray spectrometry is to find out the ratio between the full energy peak count rate Nj and the source term S v. (in this case the activity concentration in air):

Nf N0Nf^ T rh M e ' (24)

where AV$ is the photopeak efficiency for photons arriving from 0=0 (see Figure 16).

NJN0 is the angle correction factor, which takes into account the angular distribution of gammaflux and detector efficiency. It is calculated from a convolution:

2JI (25) 0 0

w U.UI :—i__i_. -4-U- ""•*"—"•'"•"• -j-i-i-•i-f- : :—4_..|_4...L:..;.. 'JZIX 7 : 7 E ••-• —|—M~ Li... T " -j— —I-F " f" "TTT • • 1 ; ; • >i ; ; : ; i ! T o • rrr ! i "T"FT c i ! i '• s j ...Li 0) 'jo | ~- •— i T -i—• • —«*—f—T S 0.001 - ...... —^^ — f .... — "•/ " i l-r 1 -i~n I i (Q I i —]•"• O> I zziz _J.ft IllD Q. ••••- 4—— O -j- i"H ••-» ! if i L* • O i 1 1 1 ! 1 | III Q. n rmrH - • i i i U.UUU1 10 100 1000 10000 Energy, keV

Figure 16. The photopeak efficiency of the detector for photons arriving from angle 8=0.

50 FINNISH CENTRE FOR RADIATION STUK-A130 AND NUCLEAR SAFETY

) is the relative detector efficiency at angles Q and (j> and d$(6,({))/(d0 dcj>) the relative portion of the unscattered gammaflux at angles 6 and 4> respectively. The angle

Nf/N0 on angle (j) must be taken into account in the calculations, since the attenuation pattern of the vehicle is complex. The flax O does not depend on angle . if contamination is laterally homogenously distributed in air. When N/No is calculated, the flux must be normalized:

Tt/2 dQ=l. // dQdfy (26) o o

Finally, the term O/Sv is the ratio between the gammaflux 0 and activity concentration in air. This term can be calculated from geometrical considerations. The only geometry discussed here is the so-called infinite slab geometry (Figure 18). If the detector height d is assumed to be 0 the following formula is valid:

(27)

where E, (x) is the exponential integral of second order:

E2(x)=x[—dt. J t2 (28)

Figure 17. Defnition of angle (p within the STUK emergency vehicle. 51 FINNISH CENTRE FOR RADIATION AND NUCLEAR SAFETY STUK-A130

Clean air

Clean air

Figure 18. Infnite slab geometry.

The angular distribution of the flux is obtained from:

(29)

The angular distribution does not depend on angle (j) since the geometry is laterally symmetric. If h,=0 and /;,=«> the equations are simplified to:

(30)

and

(31)

52 FINNISH CENTRE FOR RADIATION STUK-A130 AND NUCLEAR SAFETY

5. 4. Calibration measurements

Calibration measurements were carried out using standard point sources. The following nuclides were used (gammatransition energies in brackets): ^'Am (59.5 keV), 133Ba (81.0 keV, 302.8 keV and 356.0 keV), 137Cs (661.6 keV) and 60Co(l 173.2 keV and 1332.5 keV). The point sources were manufactured by Amersham and had a nominal activity of 37 kBq (10 uCi).

Calibration measurements were carried out in the geometry shown in Figure 15 b and Figure 15 c at angles 0 < 90. The sources were placed at the known points outside the vehicle and the photopeak count rates were measured. The measurements were carried out from 0=0 to 90 degrees at 10 degrees intervals; two horizontal direction angles were selected: = 90 and 270 degrees. At (j>= 90 degrees the attenuation is stronger, since at that direction there is a box for gasmasks. The relative efficiencies are shown in Figure 19.

The detection efficiency for airborne activity can be calculated from these measurements. The attenuation of the lid was in average the same as the attenuation caused by 32 mm thick lead plate (Figure 20).

Energy of

u the photon 0X1 CJ •a o CS

>> o .S2 'S3 1 CU I

-80 -60 -40 -20 0 20 40 60 80 Angle, degrees

Figure 19. Relative efficiency of the HPGe detector as a function of angle 6. The efciency was measured outside the vehicle. The detector is in the lead shield and the lid is open. For comparison purposes, the angle d is chosen to positive at the direction oj

53 FINNISH CENTRE FOR RADIATION AND NUCLEAR SAFETY STUK-A130

0 03 i ; 1 ; 0 025 - f s, d open \. 1 "E 0.02 - Licl closed > lii; c 0 015 - MM -f-Ul ! i i | £ 0.01 - i 1 i 1 lii MM ! i ! i MM 0.005 - M i i IM1 n . . •»• •— i : ; i () 500 1000 1500 2000 2500 Energy, keV

Figure 20. Detection efficiency (Nj/ySJ of the measuring system for airborne homogenous activity. For131 I contamination in air, for example, the count rate at 365 keVpeak (y = 0.85) would be 0.019 cps (=0.022 x 0.85) per unit activity concentration (Bq m'3).

The background measurements were carried out on a field in Helsinki area. The background spectra were measured in the geometries shown in Figures 15 b and 15 c. The results are in appendix A. Both measurements lasted about one hour

5. 5. Detection efficiency and detection limits

The detection efficiency for airborne activity of the measuring system in Figure 15 b was calculated from the calibration measurements. The dependency on angle =270 degrees represents the efficiency at all other angles. The detection efficiency when the lid of the shield is closed is also calculated. The results are in Figure 20. The actual detection efficiency in the twofold measurement is the difference between these two efficiencies. The detection limit is defined in Formula (7) and Formula (8) for onefold measurement and in (10) for twofold measurement. Since the detection efficiency of the system is known, the detection limits can be calculated from the background level. The results are in Table VIII and Table IX.

54 FINNISH CENTRE FOR RADIATION STUK-A130 AND NUCLEAR SAFETY

Table VW. Minimum detectable activities for some release miclides in in-situ measurement (Bq m'3). Homogenous activity distribution in air is assumed. Background measurement is not performed. enersv. aarama- measuring measuring nuclide keV yield time 10 min time 30 min Xe-133 81.0 0.371 3.7E+01 2.2E+01 Ce-144 133.5 0.108 6.1E+01 3.5E+O1 Ce-141 145.5 0.484 1.3E+01 7.3E+O0 Kr-88 196.3 0.260 1.7E+01 1.0E+01 Te-132 228.2 0.882 4.3E+00 2.5E+00 Xe-133m 233.2 0.102 3.7E+01 2.2E+01 La-140 328.8 0.207 1.2E+01 7.1E+00 Cs-136 340.6 0.468 5.1E+00 2.9E+00 1-131 364.5 0.812 2.8E+00 1.6E+00 Sb-125 428.0 0.296 7.0E+00 4.1E+00 La-140 487.0 0.459 4.2E+00 2.4E+00 Ru-103 497.1 0.864 2.1E+00 1.2E+00 1-133 529.9 0.863 1.9E+00 1.1E+00 Ba-140 537.4 0.199 8.1E+00 4.7E+00 1-132 552.6 0.161 9.3E+00 5.4E+00 Cs-134 604.7 0.976 1.6E+00 9.3E-01 Ru-106 621.8 0.098 1.5E+01 8.6E+00 1-131 637.0 0.0727 2.0E+01 1.2E+01 1-132 667.7 0.987 1.5E+00 8.4E-01 Sb-127 685.5 0.357 4.1E+00 2.4E+00 Te-129m 696.0 0.029 5.1E+01 2.9E+01 Zr-95 756.7 0.548 3.0E+00 1.7E+00 Nb-95 762.2 0.990 1.7E+00 9.6E-01 1-132 772.6 0.762 2.2E+00 1.2E+00 Cs-136 818.5 1.00 1.7E+00 9.8E-01 1-135 1132.0 0.225 6.3E+00 3.6E+00 1-135 1260.0 0.286 4.6E+00 2.6E+00 La-140 1596.2 0.955 1.4E+00 8.1E-01 Kr-88 2392.1 0.35 3.1E+00 1.8E+00 FINNISH CENTRE FOR RADIATION AND NUCLEAR SAFETY STUK-A130

Table IX.M.nimum detectable activities for some release nuclides in in-situ measurement (Bq m'3). Homogenous activity distribution in air is assumed. Background is taken into account. Background measuring time is 5 min. energy. gamma- measuring measuring measuring nuclide keV yield time 5 min time 10 min time 30 min Xe-133 81.0 0.371 5.4E+01 3.8E+01 2.4E+01 Ce-144 133.5 0.108 8.8E+01 6.4E+01 4.0E+01 Ce-141 145.5 0.484 1.8E+01 1.3E+01 8.1E+00 Kr-88 196.3 0.26 2.5E+01 1.9E+01 1.2E+01 Te-132 228.2 0.882 6.4E+00 4.8E+00 3.2E+00 Xe-133m 233.2 0.102 5.6E+01 4.1E+01 2.8E+01 La-140 328.8 0.207 2.0E+01 1.5E+01 1.1E+01 Cs-136 340.6 0.468 8.2E+00 6.4E+00 4.9E+00 1-131 364.5 0.812 4.6E+00 3.7E+00 2.9E+00 Sb-125 428.0 0.296 1.2E+01 1.0E+01 8.3E+00 La-140 487.0 0.459 7.3E+00 5.9E+00 4.9E+00 Ru-103 497.1 0.864 3.7E+00 3.1E+00 2.5E+00 1-133 529.9 0.863 3.5E+OO 2.9E+00 2.5E+00 Ba-140 537.4 0.199 1.5E+01 1.3E+01 1.1E+01 1-132 552.6 0.161 1.8E+01 1.5E+01 1.3E+01 Cs-134 604.7 0.976 2.9E+00 2.4E+00 2.1E+00 Ru-106 621.8 0.098 2.8E+01 2.4E+01 2.1E+01 1-131 637.0 0.0727 3.9E+01 3.3E+01 2.8E+01 1-132 667.7 0.987 2.9E+00 2.5E+00 2.2E+00 Sb-127 685.5 0.357 8.2E+00 7.1E+00 6.2E+00 Te-129m 696.0 0.029 1.0E+02 8.9E+01 7.8E+01 Zr-95 756.7 0.548 6.0E+00 5.2E+00 4.5E+00 Nb-95 762.2 0.99 3.4E+00 2.9E+00 2.5E+00 1-132 772.6 0.76 4.4E+00 3.8E+00 3.3E+OO Cs-136 818.5 1 3.3E+00 2.8E+00 2.4E+00 1-135 1132.0 0.225 1.5E+01 1.3E+01 1.1E+01 1-135 1260.0 0.286 1.1E+01 9.6E+00 8.5E+00 La-140 1596.2 0.955 3.4E+00 2.9E+00 2.5E+00 Kr-88 2392.1 0.35 9.4E+00 7.9E+00 6.8E+00

56 FINNISH CENTRE FOR RADIATION STUK-A130 AND NUCLEAR SAFETY

5.6 Advanced spectrum collection and analysis

STUK measurement vehicle has a folly automated gamma-ray spectroscopy system for various kinds of analysis needs. Gamma spectrum analysis and collection are controlled by SAMPO-family gamma spectrum analysis software. Several methods and procedures have been developed to perform calibration generation and verification, background and sample collection and analysis and constant environmental monitoring. The software can be applied to fallout mapping and to search of lost radiation sources.

The latest version of SAMPO, SAMPO 90+ (version 3.70) can handle several analysers directly. There are calibrations for energy,efficiency and gamma peak shape. These calibrations can be generated semi-automatically: only a calibration spectrum, a calibration source description library and a 2-point energy calibration are needed. Maintaining existing calibrations can be performed automatically.

SAMPO has very powerful peak analysis tools. The peak searching uses the method of smoothed second differences. Interactive peak adding is also possible. Precalculated shape-calibration is used to discriminate Compton edges and backscatter peaks. Peak area determination uses the same shape calibration. Peak fitting resolves overlapping peaks accurately, up to 32 peaks can be added to a multiplet. The user can interactively add and drop peaks during the fitting procedure or the residual analysis can be activated to perform the job automatically. Identification is carried out using candidate matrix, interfering nuclides are resolved using weighed least squares method. Activities are calculated and decay corrected. Detection limits are calculated for nuclides in a separate library and background peaks can be subtracted from the peak table.

Spectrum collection and analysis can be automated using special macro language. Reports are generated using end user programmable report generator language (RGL). RGL is a full featured programming language including most essential flow control, basic mathematical functions and many other sophisticated functions needed in gamma- ray spectroscopy. RGL can see inside the data structure of SAMPO.

In addition to its algorithms, SAMPO relies on graphical visualization and interaction. In addition to the graphical fitting and spectrum display, there is a display for monitoring the spectra collected in a series. Currently, spectrum, ROIs and searched photopeak significances are presented. Integration times below 0.5 s per spectrum can be used.

The measurement vehicle has a 36.9% HPGe detector. A 5x5" Nal scintillator detector is also available, if needed. The standard startup procedure contains calibration measurement, calibration verification and possibly calibration correction. These operations can be performed using minimal user intervention. Automated macro procedures are run after the installation of a calibration source set. hi the calibration

57 FINNISH CENTRE FOR RADIATION AND NUCLEAR SAFETY STUK-A130 verification, all the calibrations are checked against the standard values. Energy calibration correction has two different modes: rough and fine. Background is measured using an automated macroprocedure, too.

Nal-system HPGe-system GPS O

Nal-PC HPGe-PC -Nal spectra collection -Spectra/GPS collection -extra analysis -online analysis

HUB

Data-PC Dose-PC

-data communication - DGPS collection - dose rate collection

MODEM (D PIC Z3 Ö DGPS NMT-phone

Figure 21. Data managing and measuring systems of the vehicle. The computers are networked using Windows for Workgroups. Sampo is run in HPGe-PC and in Nal- PC.

58 FINNISH CENTRE FOR RADIATION STUK-A13 0 AND NUCLEAR SAFETY

The user can set alarm levels on some ROIs and total counts. Normally, a full peak analysis is performed during the next spectrum collection. Position information is saved using GPS navigator, which is integrated in the measurement computer. SAMPO saves the position information within the spectrum file. Suspicious spectra can be scanned in real time and a more detailed analysis can be performed in another computer located in the same WfW (Microsoft corp. USA) network inside the vehicle (Figure 21).

The default analysis results are saved in 1 min - 1 h intervals and an automated script is sending the results to the headquarters via NMT450 or NMT900.

5. 7 Discussion

The calculations reveal, that the in-sitii spectrometry is the fastest method to detect and to identify airborne radioactivity at the concentration level of 10 - 100 Bq m"3. Hazardous concentrations are detected immediately. Thus, mobile units, having gamma-ray spectrometric measuring equipment, should be sent out at the very early stages of the accident to the areas at risk.

The result of in-situ measurement may contain some errors. If the activity distribution in air is not homogenous one must be careful when analysing the results. Typically, at the ground level the activity concentration is depleted, because particles deposit on surfaces. Then, the obtained result is conservative, because the measuring result represents an air layer of few hundred metres, although the people are breathing ground-level air.

If the roof or the body of the vehicle is contaminated, the obtained results are no longer reliable. Thus, the in-situ measurement should be applied only on the first stages of the accident. Also high trees and buildings or steep hills and rocks, may be in the view of the detector.

59 FINNISH CENTRE FOR RADIATION AND NUCLEAR SAFETY STUK-A130

REFERENCES

Aaltonen H, Klemola S, Ugletveit F. Validation of a Method for Computer Calculation of Germanium Detector Efficiences. Nuclear Instruments & Methods in Physics Research 1994; A 339:87.

Agarval JK, Liu BYH. Experimental Observation of Aerosol Deposition in Turbulent Flow. Journal of Aerosol Science 1974; 5:145.

Anand AR, Fan B, Ortiz CA, McFarland AR. Design and Testing of Aerosol Inlets and Transport Systems. In: Meeting Review: Airborne Aerosol Inlet Workshop 1991; NCAR/TN-362+1A.

Beck HL, DeCampo J, Gogolak C. In-situ Ge(Li) and Nal(Tl) Gamma-Ray Spectrometry. US Atomic Energy Commission, New York, 1972, Report HASL-258.

Belyaev SP, Levin LM. Techniques for Collection of Representative Aerosol Samples. Journal of Aerosol Science 1974; 5: 325.

Currie LA. Limits for Qualitative Detection and Quantitative Determination. Analytical Chemistry 1968; 40: 586.

Flanagan RC, Seinfield JH. Fundamentals of air pollution engineering, Prentice Hall Inc. 1988.

Friedlander SK, Johnstone HF. Deposition of Suspended Particles from Turbulent Gas Streams. Ind. Eng. Chem. 1957; 49: 1151.

Hangal S, Willeke K. Atmospheric environment 1990; 24A: 2379.

Hinds WC. Aerosol Technology. USA: John Wiley & Sons, 1982.

Honkamaa TPS. Radiation Measurements in a Moving Vehicle and Real Time Transfer of Measuring Results. Master's thesis, Helsinki University of Technology, 1994 (in Finnish).

Knoll GF. Radiation Detection and Measurement. USA: John Wiley & Sons, 1979.

Liu BYH, Pui DYH, Rubow KL, Szymanski WW. Electrostatic Effects in Sampling and Filtration. Ann. Occup. Hyg. 1985; 17: 251.

60 FINNISH CENTRE FOR RADIATION STUK-A130 AND NUCLEAR SAFETY

Okazaki K, Willeke K. Transmission and Deposition Behavior of Aerosols in Sampling Inlets. Aerosol Science Technology 1987; 7: 275.

Okazaki K, Wiener RW, Willeke K. The Combined Effect of Aspiration and Transmission on Aerosol Sampling Accuracy for Horizontal Isoaxial Sampling. Atmos. Environment 1987; 21: 1181.

Porter J, Clarke A, Pueschel R. Aircraft Studies of Size-Dependent Aerosol Sampling Through Inlets. In: Meeting Review: Airborne Aerosol Inlet Workshop 1991;NCAR/TN-362+lA.

Seebaugh WR. Applications of Principles of Aerodynamics to Inlet/Diffuser Design. In: Meeting Review: Airborne Aerosol Inlet Workshop 1991; NCAR/TN-362+1A.

Sinkko K. Gammaspektrien tietokoneanalyysi näytemittauksissa. Licensiate thesis, Helsinki University, Department of Physics, 1981.

Sowa W. Direct Measurement of Homogenously Distributed Radioactive Air Contamination widi Germanium Detectors. Radiation Protection Dosimetry 1990; 32: 171.

Vincent JH. Aerosol Sampling, Science and Practice. John Wiley & Sons, Bath, Avon 1989

Vincent JH. Recent Advances in Aspiration Theory for Thin-walled and Blunt Aerosol Sampling Probes. J. of Aerosol Science 1987; 18: 487.

Vincent JH. Stevens DC, Mark D, Marshall M, Smith TA. On the Aspiration Characteristics of Large-diameter. Thin-walled Aerosol Sampling Probes at Yaw Orientations with Respect to the Wind. J. of Aerosol Science 1986; 17: 211.

White FM. Fluid Mechanics, 2nd Edition.New York: McGraw-Hill, 1986.

61 FINNISH CENTRE FOR RADIATION AND NUCLEAR SAFETY STUK-A130

APPENDIXES

APPENDIX A Calculation of aspiration efficiency

According to Vincent (1989) the condition for isokinetic sampling may be expressed for probes at all orientations (0 < 0 < 90) in the form: R cosö=l, (A.1) where R is the sampling velocity ratio for the moving air: V, R=-£. (A.2)

The angle 6 is the angle between the probe axis and the coming flow (0 < 90 degrees). In Figures 2 (p. 17) the angle 0 is zero. For the aspiration efficiency, the following formula is given:

[] l/2 m Nf 1+G(Q,R,DQ) Stk (cosQ+4R sm d)

*(R cos0-l). Gravitation is not taken into account in this formula. Stokes number is calculated from the free stream velocity of air. When Stk is large (> 10) Formula (A.3) can be simplified: N ~R cos6. (A.4)

G{QJi,D0) is a parameter; its value is estimated from empirical experiments. Vincent et. al. (1986) suggest a constant value:

G(Q,R,D0)=2.l ± 0.9. (A.5) Earlier Levin and Belyaev (1974) expressed G as

G(Q,R,D0)=2+0.62± XV They performed the measurements at 0=0.

62 FINNISH CENTRE FOR RADIATION STUK-A130 AND NUCLEAR SAFETY

APPENDIX B Formulae of the theory of Hangal and Willeke (1990)

The gravitational settling

The transmission efficiency Ets is expressed as:

K6 is the gravitational settling parameter:

d] po VfL 7recos6

where dp is particle diameter, pp particle density L is the length of the sampling tube (=40 cm), D, is the diameter of the tube (in our case the diameter of diffuser base), pg is the density of air and T| is the viscosity of air. The settling velocity V-^ can be obtained from the following formulae (Flanagan and Seinfield, 1988):

3T12

log(—)1/3 = -1.387+2.153(logGfl1/3)- CD (B.4) 0.548(log(Ga 1/3))2+0.05665(Iog(Gar1/3))3,

Dp JD /T(.

CD 4£(pp-pg)T]

These formulae are valid if dae is above 20 urn. If dae is below 50 urn, the Stokes law is valid and the settling velocity can be calculated from:

Impaction:

The transmission efficiency of impactor is obtained from

63 FINNISH CENTRE FOR RADIATION AND NUCLEAR SAFETY STUK-A130

2 £fJ=exp(-75(Jw+/v) ), (B.7) where Iw is the wall impaction parameter and 7V is the vena contracta impaction parameter:

Iw=Stk R™ sin(0±a) sin(^), (B.8)

°-3, /?/•/?

Iv=0,fcrR>\.

In the Formula (B.8) angle a is the gravity effect angle. In this system a=0, since the axis of the probe is horizontal.

64 FINNISH CENTRE FOR RADIATION STUK-A130 AND NUCLEAR SAFETY

Appendix C. Count rate (cps) in background continuum at various energies. An experiment performed for a 36.9 % HPGe on a field (see p. 32). background measurement filter measurement resolution, Energy, keV lid open lid closed lid closed channels 81.0 1.9E-01 5.9E-03 1.3E-02 8 133.5 1.2E-01 5.9E-03 8.9E-03 9 145.5 1.1E-01 4.3E-03 8.9E-03 9 196.3 5.4E-02 4.3E-03 8.0E-03 10 228.2 3.7E-02 4.3E-03 7.1E-03 10 233.2 3.7E-02 4.3E-03 7.1E-03 10 328.8 1.2E-02 3.2E-03 6.2E-03 10 340.6 1.0E-02 3.2E-03 6.2E-03 10 364.5 8.3E-O3 3.2E-03 5.3E-03 10 428.0 5.0E-03 2.7E-03 5.3E-03 10 487.0 4.1E-03 2.1E-03 4.4E-03 10 497.1 3.7E-03 2.1E-03 4.4E-03 10 529.9 2.9E-03 2.1E-03 4.4E-03 10 537.4 2.9E-03 2.1E-03 4.4E-03 10 552.6 2.5E-03 2.1E-03 4.4E-03 10 604.7 2.5E-03 1.6E-03 3.6E-03 10 621.8 2.1E-03 1.6E-03 2.7E-03 10 637.0 2.1E-03 1.6E-03 2.7E-03 10 667.7 1.7E-03 1.6E-03 2.7E-03 11 685.5 1.7E-03 1.6E-03 2.7E-03 11 696.0 1.7E-03 1.6E-03 2.7E-03 11 756.7 1.7E-03 1.6E-03 1.8E-03 12 762.2 1.7E-03 1.6E-03 1.8E-03 12 772.6 1.7E-03 1.6E-03 1.8E-03 12 818.5 1.7E-03 1.3E-03 1.8E-03 12 1132.0 8.3E-04 1.1E-03 1.3E-03 14 1260.0 6.2E-04 8.0E-04 1.3E-03 14 1596.2 4.1E-04 4.0E-04 8.9E-04 15 2392.1 2.1E-04 2.7E-04 4.4E-04 16

65 FINNISH CENTRE FOR RADIATION AND NUCLEAR SAFETY STUK-A130

STUK-A reports

STUK-A129 Saxen R, Koskelainen STUK-A121 Ikäheimonen TK, Kle- U. Radioactivity of surface water mola S, Uus E, Sjöblom K-L. Moni- and fresh water fish in Finland in toring of radionuclides in the vi- 1991-1994. cinities of Finnish nuclear power plants in 1991-1992. Helsinki 1995. STUK-A128 Savolainen S, Kairemo K, Liewendahl K, Rannikko S. Radi- STUK-A120 Puranen L, Jokela K, oimmunoterapia. Hoidon radionukli- Hietanen M. Altistumismittaukset dit ja annoslaskenta. suurtaajuuskuumentimien hajasäteily- kentässä. Helsinki 1995. STUK-A127 Arvela H. Asuntojen radonkorjauksen menetelmät. STUK-A119 Voutilainen A, Mäke- läinen I. Huoneilman radonmittauk- STUK-A126 Pöllänen R, Toivonen set Itä-Uudenmaan alueella: Tilanne- H, Lahtinen J. OTUS-reactor inven- katsaus ja radonennuste. Askola, La- tory management system based on pinjärvi, Liljendal, Loviisa, Myrsky- ORIGEN 2. Helsinki 1995. lä, Mäntsälä, Pernaja, Pornainen, Porvoo, Porvoon mlk, Pukkila, STUK-A125 Pöllänen R, Toivonen Ruotsinpyhtää ja Sipoo. Helsinki H, Lahtinen J, Ilander T. Transport 1995. of large particles released in a nu- clear accident. Helsinki 1995. STUK-A118 Reiman L. Expert judgment in analysis of human and STUK-A124 Arvela H. Residential organizational behaviour in nuclear radon in Finland: Sources, variation, power plants. Helsinki 1994. modelling and dose comparisons. Helsinki 1995. STUK-A117 Auvinen A, Castren O, Hyvönen H, Komppa T, Musto- STUK-A123 Aaltonen H, Laaksonen nen R, Paile W, Rytömaa T, Salo- J, Lahtinen J, Mustonen R, Ranta- maa S, Servomaa A, Servomaa K, vaara A, Reponen H, Rytömaa T, Suomela M. Säteilyn lähteet ja Suomela M, Toivonen H, Varjoranta vaikutukset. Helsinki 1994. T. Ydinuhkat ja varautuminen. Hel- sinki 1995. STUK-A116 Säteilyturvakeskuksen tutkimushankkeet 1994-1995. Musto- STUK-A122 Rantavaara A, Saxen nen R, Koponen H (toim.). Helsinki R, Puhakainen M, Harva T, Ahosilta 1994. P, Tenhunen J. Radioaktiivisen las- keuman vaikutukset vesihuoltoon. STUK-A115 Leszczynski K. Assess- Helsinki 1995. ment and comparison of methods for FINNISH CENTRE FOR RADIATION STUK-A130 AND NUCLEAR SAFETY solar ultraviolet radiation measure- STUK-A106 Servomaa A, Komppa ments. Helsinki 1995. T, Servomaa K. Syöpäriski säteilyhaittana. Helsinki 1992. STUK-A114 Arvela H, Castren O. Asuntojen radonkorjauksen kustan- STUK-A105 Mustonen R. Building nukset Suomessa. Helsinki 1994. materials as sources of indoor ex- posure to ionizing radiation. Helsinki STUK-A113 Lahtinen J, Toivonen 1992. H, Pöllänen R, Nordlund G. A hypothetical severe reactor accident STUK-A104 Toivonen H, Klemola in Sosnovyy Bor, Russia: Short-term S, Lahtinen J, Leppänen A, Pöllänen radiological consequences in sout- R, Kansanaho A, Savolainen A.L., hern Finland. Helsinki 1993. Sarkanen A, Valkama I, Jäntti M. Radioactive Release from Sosnovyy STUK-A112 Uus E, Puhakainen M, Bor, St. Petersburg, in March 1992. Saxen R. Gamma-emitting radionuc- Helsinki 1992. lides in the bottom sediments of so- me Finnish lakes. Helsinki 1993. STUK-A103 Uus E, Sjöblom K-L, Ikäheimonen T.K, Saxen R, Klemola STUK-A111 Huurto L, Jokela K, S. Monitoring of radionuclides in the Servomaa A. Magneettikuvauslait- Baltic Sea in 1989-1990. Helsinki teet, niiden käyttö ja turvallisuus 1993. Suomessa. Helsinki 1993. STUK-A102 Uus E, Sjöblom K-L, STUK-A110 Jokela K. Broadband Klemola S, Arvela H. Monitoring of electric and magnetic fields emitted radionuclides in the environs of Fin- by pulsed microwave sources. Hel- nish nuclear power plants in 1989- sinki 1994. 1990. Helsinki 1992.

STUK-A109 Saxen R, Aaltonen H, STUK-A101 Toivonen M. Improved Ikäheimonen TK. Airborne and de- processes in therapy dosimetry with posited radionuclides in Finland in solid LiF thermoluminescent detec- 1988-1990. Supplement 11 to Annual tors. Helsinki 1991. Report 1989. Helsinki 1994.

STUK-A108 Arvela H, Mäkeläinen The full list of publications is avail- I, Castren O. Otantatutkimus asunto- able from jen radonista Suomessa. Helsinki 1993. Finnish Centre for Radiation and Nu- clear Safety STUK-A107 Karppinen J, Parviai- P.O. BOX 14 nen T. Säteilyaltistus sydänangiogra- FIN-00881 HELSINKI fiatutkimuksissa ja kineangiogra- Finland fialaitteiden toimintakunto. Helsinki Tel.+358 0 759 881 1993. SÄTEILYTURVAKESKUS Strälsäkerhetscentralen Finnish Centre for Radiation and Nuclear Safety

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