ML20129E273

Characterization of Aerosol Sample from TMI Reactor Auxiliary Bldg

ML20129E273
Person / Time
Site:	Three Mile Island
Issue date:	01/31/1981
From:	Kanapilly G, Newton G, John Stanley INHALATION TOXICOLOGY RESEARCH INSTITUTE
To:
Shared Package
ML20117P504	List: ML20117P501 ML20129E262 ML20129E268 ML20129E273 ML20129E281 ML20129E332 ML20129E343 ML20129E355 ML20129E367 ML20129E382 ML20129E397 ML20129E406
References
REF-10CFR9.7 LMF-70, NUDOCS 8506060416
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3 LNF Category: UC-48 CHARACTERIZATION OF AN AEROSOL SAMPLE FROM THREE MILE ISLAND REACTOR AUXILIARY BUILDING t

G. M. Kanapilly J. A. Stanley G J. Newton B. A. Wong P. B. DeNee t'

Inhalation Toxicology Research Institute Lovelace Biomedical and Environmental Research Institute P. O. Box 5890 Albuquerque,FM 87115 January 1981 8506060416 850522 PDR 10CFR PT9.7 PDR Prepared for the Office of the Assistant Secretary for the Environment of the United States Department of Energy under Contract Number DE-AC04-76EV01013.

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• - - TABLE OF CONTENTS fage,

-i ACKNOWLEDGMENT ............................................................................ .

~ EXECUTIVE SUPMARY ......................................................................... 1

. INTRODUCTION .............................................................................. 1 2

MATERIALS AND METHODS .....................................................................

- Aerosol Sample ....................................................................... 2 2

Dissolution ..........................................................................

Analys i s of Di s sol ved Ac ti v i ty . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . '3 3

Detemi na tion of Gamma Acti vi ti es . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

Scanning Electron Microscopy Energy Dispersive X-Ray Analysis . . . . . . . . . . . . . . . . . . . . . . . . 3

. Alpha Spectroscopy ................................................................... 4 4

RESULTS ................... ...............................................................

7 DISCUSSION ................................................................................

9 REFERENCES .................................................................................

LIST OF TABLES fage.

Table 1. Radionuclide Activity (nCi) in the TMI Filter Samples as of 79305 .............. 2 Table 2. Elemental Composition of the TMI Aerosol Sample as Determined by Energy Di spersi ve X-Ray Analys i s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 LIST OF FIGURES a age,

.P_

Figure 1. A schemati c of the flow-through dissolution system . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 Figure 2. Gamma spectrum of Segment 1 of the TMI aerosol as determined in a coaxial Ge-Li detector .................................................... 4 Figure 3. Dissolution of radionuclides in a flow-through system expressed as fraction dissolved versus elapsed time (hours) ............................. ...,........ 5 Figure 4. Dissolution of radionuclides in a static system expressed as fraction dissolved versus elapsed time (hours) .......................................... 5 Figure 5. Scanning electron micrographs of original aerosol sample from Segment 3 ........ 6 Figure 6. Scanning :lectron micrograph of aerosol particles from Segment 1 after dissolution experiment in the flow-through system . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 Figure 7. An alpha spectrum of the aerosol sample determined on an electroplated sample by using a surface barrier detector and pulse height analyzer ........... 7 1

V. . .

ACKNOWLEDGMENT We would like to thank Drs. R. O. McClellan and B..B. Boecker for valuable discussions and

' suggestions during.the course of this study. We are indebted to Drs. M. ~ B. Snipes, R. G. Cuddihy and S. H. Weissmap.for their critical review of the manuscript. We thank Mr.' Donald A. Nitti of General Public Utilities (GPU) Services Corporation for providing the air filter samples. We acknowledge Ms. Sally Burt's contributions in the in vitro dissolution studies and the contribu-

. tions of Mrs. Ancilla Bay, Mr. Ken Ahlert and Dr. R. A._ Guilmette in the alpha spectroscopic analysis of the samples.

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CHARACTERIZATON OF AN AEROSOL SAMPLE FROM THREE MILE ISLAND REACTOR AUXILIARY BUIL G. M. Kanapilly, J. A. Stanley, G. J. Newton, B. A. Wong and P. B. DeNee .

EXECUTIVE

SUMMARY

The accident which occurred at the Three Mile Island Unit Two Nuclear Generating Plant on March 28, 1979 resulted in contamination of the containment and auxiliary buildings. Analyses were performed at this laboratory for: (1) radioisotopic composition and (2) dissolution charac-teristics of material collected on a filter during a week of continuous air sampling in the aux-7 iliary building. This work was done to help characterize the environment in which cleanup opera-tions are being performed. Gamma and alpha spectroscopy along with scanning electron microscopy '

were employed to determine the elemental and isotopic composition of the aerosol. The major I37 Cs. Dissolution behavior of the radioisotopes found on the filter were 095r, 90Sr. 1340s and ,

aerosol in a synthetic serum ultrafiltrate containing DTPA was determined using both static and 89-90 Sr and flow-through systems. In both systems, rapid dissolution of greater than 90". of both 134-137 Cs was found. Only slight differences in the dissolution rates of Sr and Cs were ob-served. Scanning electron microscopic analysis showed the presence of respirable size particles ,

as well as larger particles ranging up,to 10 pm. The major matrix components were Fe, Ca, S Mg, Al and St. Except for Al and Si, all matrix materials dissolved in the in uitro dissolution _

systems. Although the radionuclides were present in a heterogeneous matrix, their observed rapid dissolution behavior justified their classification as Class O compounds.I Knowledge of such a classification enables bettes evaluation of bioassay data and predictions of dose distribution ,

af ter inhalation exposure to this aerosol. The techniques used in this study may be applicable to other aerosols of unknown composition.  ;

INTRODUCTION The Three Mile Island nuclear reactor accident in March 1979, resulted in the contamination of the containment and auxiliary buildings. During the cleanup operations of the auxiliary building, airborne radioactive fission products were encountered. The potential inhalation of these aerosols by workers and others involved in the cleanup prompted the characterization of the aerosol. In addition, knowledge of the characteristics of this aerosol could provide some infor-mation on the nature of the release. Thus, the quantities, distributions and the physicochemical forms of the radioactive isotopes were the primary data needed. Data on elemental composition, size characteristics ard other relevant physicochemical properties of the aerosols were also desir.ble.

The dissolution behavior of an inhaled aerosol is onc of the major factors that determine its retention, translocation and resulting dose distribution.I Since the nature of the aerosol from the auxiliary building of the Three Mlle Island (TMI) was not known, there was no basis for predicting its dissolution in the lung af ter inhalation deposition. An important variable af-fecting the dissolution rates is the solvent. A solvent which has been used extensively for in viero dissolution studies of several radionuclides in a variety of chemical forms is a synthetic serum 2 '4 ultrafiltrate. The in viero dissolution of many aerosols in this solvent were quali-tatively comparable to their in ulvo dissolution in the lungs of animals.3'4 Therefore, in viero dissolution studies on this aerosol were conducted using this solvent. Other auxiliary studies for characterizing the aerosol included scanning electron microscopic (SEM) analysis, energy dispersive x-ray analysis (EDXA) and alpha and gamma spectroscopy.

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.' MATERIALS AND METHODS Aerosot Sample 9

- The aerosol sample, which TMI personnel supplied, was obtained by filtering about 10 cc of air from the auxiliary building on a 2-inch diameter, glass fiber filter during a period of about-8 days. According to the information supplied by personnel from the GPU Service Corporation, the 137 Cs, 10% 134 Cs, 60% 89 90 Sr 90 , 2-4%

~ isotopic concentrations were approximately 20% Sr. 2-4% Y 58 60 Co. _The sample appeared to be a thick, black deposit. Pre-140Ba 140La and 1% of C0 and sumably, this material contained normal room aerosol, combustion aerosol and other unknown aero-sols.

The total S and y activity on the filter was about 3.5 u01. The filter was divided into four segments (Table 1) with approximately 40% activity on each of segments 1 and 2 and 10% activity

-on each of segments 3 and 4. Segments 1 and 2 were used in the dissolution studies. Segment 3 was used for electron microscopic analysis and the fourth segment for determining isotopic com-position by radiochemical analysis.

Table 1 Radionuclide Activity (nC1) in the TMI Aerosol Filter Samples as of 79305 (11/1/79) 137 134 9037.90y 89 Sample Cs Cs 37 Segment 1 657 157 124 440 Segment 2 634 164 120 424 Segment 3 157 37 23 105 Segment 4 218 54 41 146 Total 1666 412 308 1112 Dissolution Two different dissolution systems, a flow-through system and a static system, were used in this study.2 In the flow-through system, the solvent was directly in contact with the particles.

This dynamic dissolution system provides estimates of early dissolution rates. The static disso-lution system was a simpler system better adapted for long-term dissolution studies.

In both dissolution systems, the samples were sandwiched between two Nuclepore* filters (100

.nm pore diameter) and secured in a filter holder. The flow-through filter holder was a 47 nei polypropylene filter holder (Millipore, Inc.). A Delrin* filter holders with both filter faces open to the solvent, was used in the static system.

ThesolventusedinthisstudywasasyntheticultraffitratecontainingDTPA(SUF+DTPA).

The pH, tonic concentration, precipitating and chelating capacity of this solvent were designed to be similar to that of serum ultrafiltrate.2'3 The solvent pH was maintained between 7.3 and 7.4 by exposing it to 5% CO 2 in air. Both dissolution studies were conducted at 37'c by main-taining the dissolution systems in a water bath.

A schematic of the flow-through system is shown in Figure 1. The solvent in a lucite reser-voir was pressurized with 5% CO 2 in air. The maintenance of a 0.5 ml/ min flow rate required a pressure of 1.5 to 2.5 psig. A Gelman* filter cartridge with pore diameter of 0.2 um was used as a prefilter for the solvent. The eluant fractions were collected at 20-minute intervals during the first day and then at 30-minute intervals by using a fraction collector.

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- - Pressurs Gsuge 53 C2 2 18888 N l l Lucito Reservoir Selvent Filter Helderl f Filter Capsule I i' i

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t_ _ _ _ _ _ _ " J Water Bath to Fraction CeHector t

Figure 1. A schematic design of the flow-through dissolution system used to study the in vitro dissolution of radionuclides present on an air filter sample collected in the auxiliary building at the TMI facility.

In the static system the filter assembly was placed in 100 ml of solvent in a 500 ml con-tainer, sealed with a lid and then placed in the water bath. Solvent changes were made at 2. 4 and IB hours and then every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The solvent was changed by removing the filter assembly from the solvent and placing it in fresh solvent.

Analysia of Dissolved Activity Gamma activity dissolved in the flow-through system was determined by counting the fractions directly in a Beckman Autogamma Counter (Model 8000) with a Na! crystal detector. Ten milliliter ,

aliquots of the static system samples were counted in the same detector. Total dissolved beta activities were determined by counting 100 ul aliquots prepared on planchets in a Beckman Wide Beta 11 gas proportional counter.

Undissolved activity remaining on the filter or total sample on segment four of the filter sample were analyzed by dissolving the activity in 7 M hcl at 80*C. After dilution, aliquots were counted in the Wide Beta !! counter and the Autogamma counter. The filter retained < 1%

activity, although some black residue was still present.

The 90Sr 90Y activities in the original aerosol sample and in the dissolution fractions were 9

determined by precipitating 90 Y phosphate at pH 5 and counting the associated beta emissions.

Total Sr was precipitated from the remaining solution at pH 10.

t Decemination of Garrta Activities Gamma activities in the aerosol samples were determined by counting for 17 hours1.967593e-4 days <br />0.00472 hours <br />2.810847e-5 weeks <br />6.4685e-6 months <br /> in a co-2 axial Ge-Li detector with an active surface area of 16 cm and a resolution of 1.4 kev at 622 kev. After the dissolution studies, the filter samples were also counted in this Ge-Li detector  ;

system to determine the remaining gama activity.

Scanning Electron Microscopy Energy Dieperalve X-ray Analysie Samples were transferred onto polished carbon stubs by pressing the carbon stubs against the sample on the filter. Replicate carbon stub samples were obtained from the same aerosol filter and were observed in a JEOL 11odel JSM-35 scanning electron microscope (SEM). Energy dispersive -

x-ray analysis (EDXA) of selected areas was performed by using a Li-drifted Si detector (Kevex r'

4

,Model 78) and Kevex Model 5100 multichannel analyz:r. Micrographs were also obtained by using

- the secondary Glectron imaging mode of th2 SEM. All nicrographs were takin using 25 kV accal-erating voltage and about i na beam current.

Alpha Spectroscopy

> An aliquot of the original aerosol sample in 7 M hcl solution was evaporated to dryness, re-dissolved in 3 M hcl and the iron was removed by ethyl ether extraction. The actinides were then electroplated on stainless steel discs from a NaHSO4 Na2SO4 buffer at pH 2.10 The alpha activ-ities were determined in a 128-channel pulse height analyzer with a resolution of 14.8 kev per 2

channel. The detector was a 300 m surface barrier silicon detector (Princeton Gamma Tech).

The duration of alpha counting and analysis was 19 hours2.199074e-4 days <br />0.00528 hours <br />3.141534e-5 weeks <br />7.2295e-6 months <br />.

RESULTS A gama spectrum of the aerosol sample is shown in Figure 2. All four segments had iden-tical gamma peaks with the same ratios of 134 Cs to 137 Cs. The amounts of the major radionuclides present in the different segments and the total aerosol filter are shown in Table 1. The activi-134 137 ties of Cs and Cs were determined directly from gama spectroscopy and the activites of 9037,90Y and 89 5r on Segment' 4 by radiochemical methods. The 09 5r activity was calculated by subtracting 90Sr 90Y activity from the total observed O'Sr and 90Sr 90Y activity. The activities of 90Sr 90Y and 89 Sr in the other three segments were calculated from the 137 Cs activity in each segment and the ratios of 90Sr 90yj 137 Cs and 89 5r/ 137 Cs observed in Segment 4.

137Cs F- .

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O O

Lu d

.J IAJ K -

134Cs g34 ,

134Ce L i I I I I I f f - t 10 219 428 637 846 1055 1264 1473 1682 18 91 K.E.V.

Figure 2. Gamma spectrum of Segment 1 of the TMI aerosol as determined with a coaxial Ge-Li detector.

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, , The dissoluti;n data obtained in the flow-through and static systems are summarized in Fig-ures 3 and 4, respectiv21y, as fractions of undissolv;d activity remaining on the filttr versus elapsed time. Total gamma activity was directly determined on the filters before the samples were subjected to dissolution. The ini+ial total beta activities on these filters were calcu-lated from the ratio of total beta coun.s in the Wide Beta Il counter to 137 Cs in segment 4 and the 137 Cs count on segment 1 (flow-through system) and segment 2 (static system). The undis-solved fractions were calculated on the basis of beta counts. The 90Sr 90Y activity, both ini-tial and those in the solutions were determined radiochemically. Summation of all dissolved radioactivites and those remaining on the filter provided estimates of initial amounts of. radio-activities within a few percent of those activities shown in Table 1.

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Figure 3. Dissolution of radionuclides in 7

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o e i _- T 7 7 777 a flow-through system expressed as frac- 0 5 to 15 20 25 30 35 40 45 50 tion dissolved versus elapsed time. TIME HOURS 09-O di tu 0 7 '-

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O 10 0 200 300 400 500 600 700 TIME, HOURS Figure 4. Dissolution of radionuclides in a static system expressed as fraction dissolved versus elapsed time.

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Scaning electron micrographs of particles from Segment 3, which represents the original aerosol simple, and those from Segment 1 af ter dissolution in the flow-through system are shown in Figures 5 and 6, respectively. Average elemental composition of a large number of particles from Segment 3 and 1 are shown in Table 2.

137 An alpha spectrum of the original aerosol is shown in Figure 7. Sased on Cs counts, this sample represented about 4% of the total sample. The total activity of Pu, Pu or I Am, and Pu in this sample is 0.50 dpm. Cn this basis, an estimate of total actinide alpha acti-vity on the original aerosol was calculated to be 13 dpm.

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Figure 6. Scanning electron micrograph of Figure S. Scanning electron micrographs of aerosol particles from Segment 1 of the TMI original aerosol sample from Segment 3 of the filter af ter dissolution experiment in the flow-IMI filter. Magnification 3000 X. throttgh system. Maanification 3000 X.

Table 2 Elemental Composition of the TMI Aerosol Filter Sample as Determined by Energy Dispersive May Analysis Weight Atomic Standard Element (~ ) (-) Deviation Segment 3 Before Dissolution Mg 7.2 9.6 2.3 Al 16.2 19.3 1.6 Si 27.1 31.1 1.5 5 17.3 17.4 1.1 Ca 18.0 14.4 0.93 Fe 14.3 8.2 1.0 Seament 1 After Dissolution Al 63.8 64.8 7.9 51 36.2 35.3 3.2 7

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. 20t A3 k 5093 Ka f 24 -

239p, U i 5150 Kev g) s 60 - -

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Figure 7. An alpha spectrum O 16 -

of the TMI aerosol sample o determined on an electroplated W sample by using a surface

> 12 23s Pu barrier detector and pulse I- S756 Kew height analyzer.

g 244Cm W O -

5800 Kev J t O'

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4724 4907 5089 5272 5456 5637 5820 6003 K.E.V.

DISCUSSION Analysis of the gamma spectrum (Fig. 2) indicated that 137 Cs and 13'Cs were the primary -

gamma-emitting radionuclides on the filter. The data supplied by GpU suggested the presence of small quantities of 1OBa, 140La and 60 Co. Since the gamma spectrum (Fig. 2) was obtained about a month after the aerosol was collected, it was expected that no 140 Ba or 140 La would be ob-served. The energy peaks at 1168 and 1365 kev from 134 Cs, could have been interpreted by TMI ,

p personnel as that of N Co and 60 Co. It is noteworthy that there was no indication of any lan-thanide isotopes, such as 144 Ce, present in the sample.

l The quantities and proportions of 134Cs, 90$r 90Y and 89 Sr (Table 1) were similar to the estimates provided by GPU. All data indicated that t.1e' major radionuclides were 134Cs. I37Cs, 09 5r and 90Sr 90Y If there were other beta-emitting lanthanides, they would have been notice-able during the repeated counting of the 90 Y fractions obtained during the radiochemical analysis of 90$r 90Y The half-time of decay of the 90 Y fraction indicated no major, long-lived lanth-anide isotopes. The observed actinide radionuclides (Fig. 7) were unexpected. Although the 4 alpha activity was only 13 dpm in the total aerosol sample, it is a significant observation that there were actinide nuclides present in the aerosol. Since the physicochemical behavior of Ianthanide and actinides are similar, there may have been minute but undetectable amounts of gamma- and beta-emitting lanthanide radionc11 des also present in the original aerosol.

The presence of major isotopes of alkali and alkaline earth elements, which are generally water soluble, but not other fission products such as I44 Ce and 9I Y in detectable quantities sug-gest that the major release mechanism was probably one of dissolution in water and then disper-sfon in the auxiliary building. In addition, the presence of the small amount of insoluble actinide suggests other possible mechanisms of particle release as a minor contributing factor.

The elemental analysis (Table 2) suggests that the aerosol consisted of nonradioactive materials.

Comparison of the composition before and af ter dissolution suggests that only Al and Si were insoluble and the other elements were present as soluble materials. The radionuclides were 8

1

. . present as minor components in a heterogeneous matrix consisting of Fe, Ca, S. Mg, A1, Si and carbonaceous material. The aluminum and silicon represented 43% of the total inorganic mass. It should be noted that the EDXA was incapable of determining C 0 2 H2 , etc. The filter sample was black both before and after dissolution which suggested the presence of soot or other carbonac-eous materials.

The sizes of the aerosol particles were estimated from the electron micrographs (Figs. 5 and 6). Respirable-sized particles as well as larger particles ranging up to 10 t.m were observed.

Although the relative number density of the particles in the electron micrograph after dissolu-tion (Fig. 6) was less than that before dissolution (Fig. 5), there was no apparent size-dependent, preferential dissolution of particles from the filter. Therefore, it is not possible from our study to daduce the size distribution of the soluble fraction which contained the ra-dionuclides. However, it is reasonable to conclude that at least a fraction of the radionuclides were associated with the respirable-size particles.

The dissolution data (Figs. 3 and 4) show rapid dissolution of > 90% of all isotopes. The ts isotopes appear to dissolve almost entirely in both dissolution systems. Less than 2% of the Cs isotopes could be detected on the filters after the Missolution experiments. The data show that within 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />, more than 95% of the Cs isotopes dissolved in both systems. The total beta dissolution rates were somewhat slower than the dissolution rates of the Cs isotopes. It should be remembered that the total beta activity was the sum of 9Sr, 90 37,90y, 134Cs and 137 Cs beta activities. The 90Sr 90Y dissolution rates suggest that a slightly higher fraction, 5 to 10% of the activity, was relatively insoluble compared with Cs. This undissolved fraction of the Sr was present in both dissolution systems.

The International Comission on Raciological Protection (ICRP) has grouped specific chemical forms of radionuclides into three classes (D W, and Y) to describe their expected patterns of retention in different portions of the respiratory tract.8 The three classes, days (C), weeks (W), and years (Y) describe both the expected clearance by mechanical and dissolution / absorption pathways. These classifications have been derived for single radionuclides in known chemical forms based on available in vivo and in vitro data. However, the ICRP solubility classification scheme does not include predictions for heterogenous mixtures such as the one present on the sample analyzed here.

Results from the in vitro dissolution tests performed on this sample demonstrate that most of the radionuclides present dissolved very quickly (within 5 hr) in the SUF plus DTPA solvent.

Although an exact correspondence between in vitro and in vivo dissolution characteristics has not been established for these radionuclides in this form and mixture, it seems reasonable to assume that a similar rapid dissolution would occur in vivo. Thus, based en this assumption and the observed in vitro dissolution characteristics of the present sample, it is appropriate to eval-uate observed air concentrations with respect to derived air concentrations (DAC) by classifying the radionuclides as class D materials.

This study has oemonstrated the applicability of the above te,chniques for characterizing j airborne radionuclides of unknown physical and chemical composition. The data may be useful in  !

determining the body burden of these radionuclides from bioassay data and helping to estimate the dose distributions resulting from inhalation of the aerosol.

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, , REFERENCES

~

1. Morrow, P. E., D. V. Bates, B. R. Fish T. F. Hatch and T. T. Mercar, " Deposition and Retin-tion Models for Internal Dosimetry of Human Respiratory Tract," Health Phys. R: 173-207, 1966.

2. Kanapilly, G. M., O. G. Raabe, C. H. T. Goh and R. A. Chimenti, " Measurement of In vitm Dissolution of Aerosol Particles for Comparison to In vivo Dissolution in the Lower Respi-ratory Tract After Inhalation," Health Phys. H: 497-507, 1973.

3. Kanapilly, G. M., O. G. Raabe and H. A. Boyd, "A Method for Determining the Dissolution Characteristics of Accidentally Released Radioactive Aerosols," in Proceedings of the Third International Congress of IRPA, pp. 1237-1242, Washington, DC, USAEC Document, CONF-73907, 1973.

4. Kanapilly, G. M. and C. H. T. Goh, "Some Factors Affecting the In viem Rates of Dissolution of Respirable Particles of Relatively Low Solubility " Health Phys. 25,: 225-237, 1973.

5. Miglio, J. M., B. A. Muggenburg and A. L. Brooks, "A Rapid Method for Determining the Rela-tive Solubility of Plutonium Aerosols," Health Phys. 3_3,: 444-457, 1977.

6. Kanagilly, G M. and G. J. Newton, "A New Method for the Separation of Multicurie Quantities of vuY from $0S r," Int. J. Applied Rad. and Isotopes g: 567-575, 1971.

7. Kressin I. K., " Electrodeposition of Plutonium and Americium for High Resolution Alpha Spectrometry," Anal. Chem. R : 842-846, 1977.,

8. " Limits for Intakes of Radionuclides by Workers," ICRP Publication 30, Part I, pp.63-116, 1978.

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AFFIDAVIT Randall C. Thompson I, Randall C. Thompson, worked at mbree Mile Island " nit II during the month of April, 1979. Before that, I had been a health physics and chemistry technician for PAD Services of Pittsburgh, PA and had held the title of Senior Plant Chemist at the Peach Bottom nuclear plant for the last two years of my nuclear career. For the first ten days to two weeks of my stay at TMI I was the only hp qualified as surveillance tech-nician to change the vent stack monitor filters. I also did containment sampling and provided hp coverage for the task of changing the banks of auxilary building charcoal filters.

While working at T"I, I changed the vent stack nonitor filters twice a day. In every single case, the dose rate from these filters were very high, often constituting a "high radiation area" (more than 100 mr/hr. By way of comparison, consider that at Peach Bottom a four cubic ft./ min filter would yield a dose rate of less than 10 nr. Based on this, I came to the conclusion that the plant was emitting large quantities of radioactive gas through the vent stack.

M On two occasions, once in the GERS counting trailer and once at the HP checkpoint I was able to look at isotopic identi-fication printout sheets from those filters. On both occasions iodine 131 was particularly high, showing concentrations of one to five x10/-5 mc/ml.

Often, because of the high dose rates of these filters, they were set aside so that no one could disturb it before it was decayed to some degree. I was asked specifically on two ]

4 occasions to hide the filters. On both of these instances, when I went to retrieve the hidden filter it had already been taken.

i In all cases, the daily reports of radiation released from I

• hree Mile Island were much lower than I personally observed.

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I, nande4 min "'honsson, From 1976 through 1978, I was the senior health physics and chemical technician, senior plant chemist and surveillance technician (air release specialist) at the Peach Bottom Fuclear Generating facility in Peach Bottom, Pennsylvania.

As a health physics technician, it would he unthinkable not to check for alpha radiation on a daily basis, under normal operating conditions where the possibility of alpha presence is practically nil. However, at mMI, where the conditions present indicated a stronger possibility for alpha contamination than normal, we were not able to check for alpha on a routine or spot basis. *his was because there were no alpha counters available.

I continued to suspect alpha presence at TMI and inquired of many people,. including at least four NRC inspectors with whom I worked. I asked every utility person that I saw for an explanation as to why.there were not alpha counters 1

available. The explanation was that there was no alpha, i and, thus, no reason to check. This was unacceptable cs an l l

explanation since checking is routine. Upon further '

I questioning, it was explained by Meted health physics employees that, indeed, during the first week after the accident, there was an " alpha scare", but after " independent analysis", it was determined that alpha was not alpha after all, but was " piggy-back beta".

i

Following the accident, planned releases were scheduled and monitored. There was ontsituation where the monitoring system f ailed before the release. I tried to set up an alternative sampling system. Meted wanted to go ahead with the release prior to my. setting up the alternative system which would not have taken more than 30 minutes. I pleaded my case with the head MRC person, arguing that legally or morally they could not go ahead. The NRC person stated that they had to go ahead, quipping, "Its the only game in town." The release occurred, unmonitored. I lodged a complaint.

To further identify this occurrence, it was in connection with the sampling of reactor water through the Auxiliary Building.

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AFFIDAVI?

Marian F*.- (Joy) *hompson I, Joy Thompson, was employed by RAD Services, Inc. of Pitts-burgh, Pennsylvania as a health physics technician assigned to the Dosimetry department at Three Mile Island Unit II for the last three weeks of April, 1979. My husband, Randall C. Thompson also worked at TMI as a ranking health physics technician and surveillance technician.

Before the accident, my husband and I were incorporated as a publishing company, and published a monthly magazine for chil-dren about skatehoarding. Previous to that, my husband had spent seven years in the nuclear industry as a health physicist.

When Three Mile Island had its accident, we decided to in-vestigate its causes and effects first-hand, as the discrep-ancies in what we were hearing on the news and uhat we were hearing from our friends who were already working at *MI were tremendous.

During my emp5oyment at *MI, our friend and co-worker, David Eloombaum, was accidentally contaminated with radioactive iodine on his face and forehead. Nasal smears were taken before and after decontamination efforts by both David and my husband, and the geli identified the contamination as iodine.

The body scanner, however, found none on the day of the con-tamination. The next day, after navid had developed beta burns on the inside of his nose and the upper part of his throat, the MRC body scanner also found no iodine. It was my.conclu-sion in this' instance that the body scanners were not programmed to detect iodine.

In my job as hp in the dosimetry department, I processed and recorded the daily TLD measurements for all personnel and visitors to TMI. The two TLD machines I used in that proce-ssing were the property of Metropolitan Edison and the NRC.

Each concern was in charge of calibrating their own machine.

I randomly placed groups of badges into one or the other machine to be read, depending on which was busy and which available. Often there would be discrepancies in one or the other machine's average dose measurement, when the average  !

dosages should have been within a few MR of each other. I When I would inform my superiors of the discrepancies, the machines would be re-calibrated during the day shif t, but the differences in average dose measurement would grow each day thereafter until I would again have to complain to my super-iors. It was my belief that both machines were deliberately miscalibrated to underestimate actual dose.

Once in mid-April, I was brought the worker's badges who were

' assigned by my husband to the task of changing the banks of 4

. . . - . . 4 , ,--__,._,,,...m.r,_.. ,,.---.____,__y _,_..r,___,, _ . . , , _ _ . , _ . , _ , _ _ _ . - . _ _ - _ . . , _ _ . . , _ , . . __ ~

2 charcoal filters in the auxilary building. The badges were usually brought over to the trailer from the island as each group of workers emerged from the aux building in order to immediately determine how much dosage they had received so as not to accidentally overexpose them. Most of the time a health physics technician would bring the badges, I would read and record the dosage immediately, then he would take that information back to the island.

t One time, a GPU executive engineer (management-level person) brought the badges for me to read. I was alone in the trailer and it was about two a.m. Despite the fact that the badges were in blue cases (indicating whole-body dose as opposed to yellow badges which were used as extremity measurements), I was given a direct order to intercept the computer documentation and enter the dosages as extremity. The doses ranged from 150 millirem to three hundred and fifty millirem...significant doses. Because those doses were entered as extremity doses, they did not become part of those worker's permanent dose record.

The fact that the computer was programmed to disregard any dose reading under 10 millirem might not be significant in the operation of a normal nuclear power plant, but at TMI the doses were high and the badges were read daily instead of once or twice a month. Thus a person receiving 10 mr or less each day would have no record of any radiation exposure. In some instances, this practice could result in significant underestimation of dose and negation of legal rights if the worker were later to develop diseases caused by exposure to radiation.

The underestimation of dosage and the deliberate cover-up of high doses at Three Mile Island is unethical behavior. I believe that evidence of unethical behavior by the utility should be a consideration in any decision to approve the start-up of TMI Unit I.

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