ML20117L459

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Responds to Re Response of micro-R Meter to Spectrum of Gamma Rays Monitored in Course of Final Survey & Initial Survey Measurements in Exposure Room,Resolution of Geometrical Impact on Observations
ML20117L459
Person / Time
Site: 05000187
Issue date: 05/09/1986
From: Benveniste J
NORTHROP CORP.
To: Berkow H
Office of Nuclear Reactor Regulation
References
NUDOCS 9609130189
Download: ML20117L459 (49)
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. .. ra m e 9 May 1986 ,

Dr. Herbert N. Berkow, Director Standardization and Special Projects Directorate Division of PWR L1 ansing-8, NRR U.S. Nuclear Regulatory Commission Washington, DC 20555

Dear Dr. Berkow:

In response to the questions posed in your letter of February 28, 1986, we have reviewed our original data, assembled it into a fonn you will find easier to digest, and provided some explanatory text.

To insure that we are being completely responsive, we have organized our answers into two sections.Section I addresses the question regarding the response of the micro R meter to the spectrum of gamma rays monitored in the course of our final survey. We show by means of analysis and supporting measurements that the response of the l'ilcro-R meter is accurate to well within the desired +20%.

Section 11 addresses the initial survey measurements In the exposure room, resolution of the geometrical impact on our observations, and a comparison of the radiation levels from the concrete rubble with background. We find that the radiation froni the concrete is the same as background within the precision of our measurements, or approximately 10% of background.

This easily meets or surpasses the 5 pr/hr staff position.

We trust we have answered your questions satisfactorily. Early approval of our request for unrestricted use of our facility and our plan for disposal of concrete rubble at a local landfill are requested.

Sincerely yours, I dbmwa,5f J Benveniste, Chainnan '

' Corporate Radiation Coninittee

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$' JUSTIFICATION FOR USE OF SCINTILLATION DETECTOR NER QUESTION:

Radiation measurements were obtained by Northrop Corporation using a (1) sodium iodide crystal which is calibrated for a certain source type and strength. The methods and standards used for calibration of this Please survey instrument were not specified in the final report.

provide a detailed description of your calibration procedure and justif y use of a scintillation detector, which is an energy dependent j instrument, to measure dose rate with reasonkble accuracy (e.g. 120%) l over the wide range of energies expected in your surveys.

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I NORTHROP RESPONSE:

The roentgen is a unit of radiation exposure whose definition is based on It the ef f ect of X or gamma radiation on the air through which they pass.

is defined as the' quantity of X or gamma radiation which produces 1 electrostatic unit of electricity of either sign per cubic centimeter of dry air at 0 deg C and 760mm Hg.

The Ludlum micro-R meter is a sodium iodide crystal - photomultiplier assembly which measures the pulse-rate generated by the interaction of a photon flux with the crystal. DursCertificates are customarily sent to Ludlum l' Weasurements Inc. for calibration. of calibration for two recent episodes are provided in Attachments 1 and 2. Note that a Cs--137 source is used as the standard and the calibration is extended to the lower ranges by means of a pulse generator. The meter reading (M) may be related to the rate of charge generation in air (and hence to roentgens per unit time) through the relation l

(1) f , M = r(e)E l!

where E is the radiation exposure in roentgens, and r(e) is an energy-dependent meter response f actor. The scale of the micro-R meter is m

calibrated so that for the 662 kev gamma rays from cesium-137 r(662)=1 and (2) 3 Mc = Ec z.

{ see ref,' 1.

Smy,

k The f actor r(e) is plotted as a function of photon energy in Figure 1 (ref.2) 7 For a flux of photons of energy ei, Equation 1 is written i..

Mi Wi = r(ei) Ei or Ei =

(3) r(ei)

Thus, for a complex spectrum, the total exposure may be written EEi = E (4)

_r(ei) ,

where the Mi are the individual meter readings which would be contributed by

[ the components of the complex spectrum.

i.

If we divide Equation (4) by Equation, (2) we write y

, ~Mi - (5) 2 1 E 17

  • Q ,r(ei),

and if we consider the case where the composite meter reading contributed by the complex spectrum is the same as that given by the same flux of Cs-137 gamma rays (i .e. , E Mi=We ), Equation (5) becomes

~

r

~Mi

~

EEi _r(ei)_

=

(6)

Ee EMi I

Primordial Radioacti<e Source in Concrete Normal concrete contains primordial sources of radiation (ref. 3). The gamma spectrum from potassium-40, uranium-238 series, and thorium-232 series are extremely complex. But the measurements of the attenuation of this gamma field, utilizing lead attenuators, ' indicate the spectrum can be 3 represented by a two component photon field. The meter reading for this two component spectrum may be represented by 6

, M(x) = M)exp

+ M2exPl -

3 e o

where x is the thickness of the attenuator. The experimental data corres-ponding to this expression appears in Table III and Figure 2. After normal-p izing so that W1+M2 = 1, we find from Figures 1, 2 and 3:

^

W1 = .667 W2 = .333 x1 = 1 gm/cm2 x2 = 19 gms/cm2 e1 = 200 kev I

e2 = 1500 kev r(el) = 6.00 r(eg) = 0.35 Returning to Equation (6) yields E E; .667 + .333 = 1.06 EC

_- 6.00 .35 That is, a given meter reading corresponds to a 6% greater exposure from the primordial radioactive source in concrete than from a Cs-137 source---the 4

meter is underestimating the exposure by 6%.

? Neutron Activated Concrete Source The gamma exposure from the neutron activated concrete in the facility is treated in a related, but somewhat different, way. This is because the emitted gamma flux is dominated by the natural, primordial radioactivity to  ;

the extent that the neutron induced gamma rays could not be measured I directly. Thus, we rely on a calculation of the spectral composition of the emitted gamma rays, relate this to the exposures contributed by each element of the spectrum, and finally make the connection to the meter readings as in the previous example.

p ,

k.,

r I

i The isotopic composition of the neutron activation induced radioactivity .t e-depth in the concrete is known to be (ref. 3).

Fe - 55 81 percent of disintegrations l Co - 60 4 Eu - 152 13.9 i

Eu - 154 1.1 l l

I L

Since Fe-55 emits very low energy gamma rays (less than 10 kev) that, to a large extent, are self-absorbed in the concrete and are below the cut-off energy of the detector, Fe-55 need not be included in the analysis of the l gamma rays emitted from the concrete. The remaining isotopic source terms, '

Sj after normalization, are:

S60 = 0.21 S152 = .73 S354 = .06 There are approximately eighty different photon energies emitted by these sources. All of these photon types were arranged into five energy groups, each with a calculated emission probability, Ni , and a radiation length in concrete, xi. The probability of each photon energy group being represented in the emitted flux is given by I'

$,(eg) = T Sj Njx; (7) 1 There is an implicit assumption in using this functional representation that the neutron activation is rather uniform in the concrete, and that the neutron relaxation length is large compared to the gamma radiation length.

The f unction po {ez ) is tabulated for each energy group in Table IV. This information is normalized and summarized in Table V, column (2), and plotted

in Figure 4. However, this is not quite the whole story for equation (7) s' is derived from an assumption that the gamma rays are absorbed exponentially without producing secondary photons prior to emerging from the concrete. In f act, however, some of the interactions are Compton scatters in which the photon is not absorbed, but only degraded in energy. This feature is estimate this build-up by adding 10% of the flux in each energy group to each of the lower energy groups as in Table V, Columns 3 and 4, and in Figure 4. p(ei) designates the flux with the build-up included.

e

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l The flux in each of these energy groups is related to the corresponding exposure through the relation

$(ej) = f(e;)E; (8) where the function f(e), flux-exposure factor, is plotted in Figure 1, and normalized to 1 for the Cesium gamma ray. Using this expression the values for exposure associated with each of the photon groups were calculated in ,

Table V, column (7) . Finally, using Equation (3) the meter readings corresponding to these exposures were calculated and listed in column (8).

Summing the meter readings from each energy group J

e I M; = I r(e;)E; (9) yields the meter reading for the entire complex spectrum. Dividing Equation (9) by Equation (2) gives

{

\

IM; __ T r(e:) E; I Mc Ec and if we consider the case where the total exposure due to the complex spectrum is the same as that from a flux of Cs-137 gamma rays (ie. Ec =IEi)

Equation (10) becomes 1

I M: = Ir (ei)E; , 3,33 (33)

Mc IEj That is, a given exposure yields a 13% higher meter reading by the complex spectrum from the neutron activated concrete than from a Cs-137 source.

i In summary then, we have seen that with respect to the Cs-137 source for which the micro-R meter has been calibrated, the micro-R meter under-estimates the exposure from primordial concrete gamma rays by 6% and overestimates the exposure from neutron activated concrete by 13%. Our measurements in the exposure room indicate that the radiation exposure due E

to neutron activation is less than 107, of the total (primordial plus  ;

activati*on gamma rays) since this is the pret.ision of the background  !

4 1. measurements. Thus, the weighted meter response underestimates the exposure from the entire spectrum of gamma rays by about 4%. We may conclude, therefore, that the micro-R meter readings *: took in performing our final  ;

survey were reliable to well within 120%.

This result has been confirmed by representatives of hTC-Region V who conducted an independent survey of the residual radioactivity in the facility using a Reuter-Stokes pressurized ion chamber. i

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't SBCTION II

, SURVEY OF CONCRETE NRR QUESTION:

, (2) Your report iadicates that your survey of the Exposure room was ,

hampered by thi small dimensions of the room, which resulted in your

. providing 4 pi readir.gs versus measurements which would directly i

demonstrate satiefaction of the requirements of 5 uR/hr, at one meter 4

from any planar surface. Your calculations indicated that the readings were close to background. However, because of the geometry problem, you dissantled the room and placed the resulting biological shield concrete ubble in a pile. Your survey of this pile also indicated readingr close to background. However, during its final survey of your f acility. the Region V staf f found that the configuration of ti e con. crete rubble pile precluded their obtaining the required final survey measurements with the available instrumentation. '

Accordingly, we requtst that Northrop Corporation: (a) provide all i the readings that were taken of the Exposure Room walls prior to 4

dismantling, with and without shielding, (b) discuss the method used to determine ' background' radiation and provide the respective data, (c) show f rom your measurement techniques and data that the contribu-tion of dose rate from the surrounding walls could be subtracted from the relevant planar wall to provide the 'close-to-background" dose rate from the planar wal2, s(d) physically redistribute the concrete

rubble pile into a homoger.eous configuration with as uniform and even a surf ace as possible and not more than about I foot deep. Measure the radioactive contamination levels at the surf ace and the radiation
lev e l s at one meter f rom the surf ace at a suitable number of locations.

NORTHROP RESPONSE:

Since the exposure room was classified as a "high potential' area the walls, floor, and ceiling were marked off in one-meter grids and 757,of the squares s were selected f or measurement by means of a random number generator.

Figures Sa-e show schematic represent stions of the manner in which all surf aces were sectioned for the survey. Data on the west wall is missing because this is the wall which is shared by the pool and it was essentially eliminated when the window was removed.

1 .

Tables Vla-e contain the recorded readings from a Ludlum Model 19 pR-meter at 1 meter and 1 centimeter above the surface. Background measurements were taken at 30 locations in 16 different areas in and around the Northrop r

compl e x . These locations are identified in Figure 6 and the data is summarized in Table VII. It shows that the average of 30 readings is 11.6 -

0.8 pR/hr at 1 meter. .

When the pR-meter was set up in the exposure room at 1 meter from the walls

- we read much higher levels (e.g. an average of 18.4 pR/hr for 71 readings, with a maximum of 24 pR/hr) . This was easily understood to be due to the '4r" geometry to which the detector was exposed; that is, the detector was completely surrounded by activated wall as opposed to being above a planar surf ace as for the background measurements. To illustrate this f act, we perf ormed a couple of simple experiments. A 2 inch thick by 8 inch wide by 16 inch long layer of lead was placed below the detector, with the

center of the detector at 1.375 inches above the center of the surface. The meter read 14 R/hr. The detector was placed below the layer of lead at 1.875 inches below the surface. The meter read 15.5 R/hr. For the detector completely surrounded by 2 inches of lead, the meter read zero. See Figures 7a,b & c f or illustrations of the shielding arrangements. The solid angle for this geometry was calculated from ,

S = 4 [ 77 - (arctan A + arctan B - 2 arctan C)] (12) where, .5 27 A= _y+Z 2 2 2 .5 0

L Z2 + X 2,,

t i

2 q. 5 7

g C=JX2 + y2 J i

x and y are the distances f rom the meter detector to the outside edges of the 1.ayer of lead brick, and z is the height of the detector above or below 3

the brick layer.

.t a.

k 4

4 e m l

The data is summarized below:

EASUREENT x y z S EXPOSURE E/S

.. ANGLE ( inches ) (st' rad) ( R/hr) (R/hr/sr) i kNhlRE SPHbRb - - -

$2Sh 21 3 k.69 i UPPER SPERE 4 8 1.375 7.62 14.0 1.84

]

L0nER SPEP2 4 8 1.875 8.15 15.5 1.90 AVERAGE 1.81 BACKGROUND - - -

6.28 11.6,- 0.8 1.85; 0.13 By calculating the exposure per steradian we are able to eliminate the solid angle ef f ect and combine the results of the three observations in the b

exposure room. We see that the average is 1.81 R/hr/sr. The correspond-

ing number for the background is 1.85 pR/hr/sr and the difference is 0.04 pR/hr/sr. This means that if we were able to remove the surrounding walls and ceiling to infinity, the exposure measured at 1 meter from a plane surface in the exposure room would be essentially background, and therefore, well within the 5 #R/hr staff position.

When the NRC was reluctant to accept this kind of evidence in fulfillment of the requirements, we decided to eliminate the '4n' geometry by eliminating the exposure room. This was done by cutting a s&ath through the concrete shield from the hot cell on the north through the south wall of the exposure room. This left large, planer surfaces for us to work with. Most of the concrete rubble was taken to an enclosed area in the middle of the

, Johnson property parking lot.

The rubble which came from immediately around the exposure room was stored separately in anticipation of further interest on the part of the NRC. This rubble was eventually moved to the Johnson property storage yard and i distributed homogeneously over an area such that the depth of material n-nowhere more than about one foot. Prior to this move, however, backgsound 9 background measurements were made in the storage yard in the vicinity of the 1

proposed laydown area. The storage yard was marked of f in a grid with 3-meter grid size and readings were taken at I meter above the center of each square. The squares to be read were selected by means of a random

, number generator. Figure 8 shows the results of these measurements graphically and Figure 9 shows the distribution of the readings and an analysis of the data.

x Figures 10 and 11 show the outline of the area over which the rubble from the exposure room was distributed and the exposure meter readings obtained at I meter above the rubble and at the surf ace of the rubble, respectively.

j. Comparison of the exposure readings at 1 meter above the wurface yields:

DESCRIPTION AVERACE NUMBER OF SURVEY R/hr MEASURRMRNTS RATE

. RUBBLE 8.63 + 0.38 30 MARCH 86 (fromexposure room)

RUBBLE 8.77 + 0.26 39 JANUARY 86 l (from exterior walls)

BACKCROUND 9.08 + 0.36 30 JAN & MAR 86 (of storage yard)

The difference is zero within the precision of the measurements.

Thus, we have demonstrated that:

1

1. Our technique for treating the data obtained in a confined space like the exposure room so that they might be directly compared with readings obtained above a planar surface is correct and reliable.
2. Radiation emanating from the concrete rubble is essentially at background levels and easily meets the 5 R/hr criterion.

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! TABLES 1

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l TABLE I MEASUREMENT OF ATTENUATION OF CS-137 RADIATION BY LEAD ATTENUATORS LEAD VICTOR LUDLUM NORMALIZED NORMALIZED ATTENUATOR METER METER METER CALCULATED THICKNESS READING READING READINGS ATTENUATION (gms/cm2) (mR/hr) (mR/hr) FACTOR

[1] [2] [3]

0.00 ----- ------

0.97500 1.00000 1.80 67.00 ------

0.83750 0.83615 l 3.21 56.00 ------

0.70000 0.72081 i 4.68 50.00 ------

0.62500 0.61749

6.18 43.00 ------

0.53750 0.52729 8.03 35.00 ------

0.43750 0.43399 14.98 17.00 ------

0.21250 0.20881 24.52 8.00 ------

0.10000 0.07650 47.76 ~0.60 0.600 0.00750 0.00663

53.98 ~0.25 0.250 0.00313 0.00344 j 60.93 ----

0.140 0.00175 0.00166 70.47 ----

0.060 0.00075 0.00061 76.70 ----

0.025 0.00031 0.00031 83.65 ----

0.013 0.00016 0.00015 89.88 ----

0.008 0.00010 0.00008 Note:

1] Lead attenuators were inserted at the source position.

i 2] The exposure from the Cs-137 source at 1 meter was

r 31,400 mR/hr. At 20 meters, the lead equ thickness of the intervening air was 1.8 grms/cm(valent

~

. The calculated exposure, without air attenuation, at 20 meters is 80 mR/hr.

3] The Victoreen 440 ion chamber was used for the higher flux readings.

i

+

TABLE II PARAMETERS USED IN THE CALIBRATION OF THE MICRO-R-METER PHOTON l--PHOTON RELAXATION LENGTH--l METER FIELD ENERGY l LEAD CONCRETg AIR. NaI l RESPONSE REACTION (kev) l---------(gms/cm )----------l FACTOR FACTOR

[e] l-----------[x]--------------l [r(e)] [f(e)]

[1] [2] [3]

15 0.010 0.121 0.645 0.021 0.10 1.00

27 0.022 0.500 2.128 0.127 1.00 0.53

! 50 0.128 2.564 4.762 0.093 10.00 10.00 75 0.339 4.348 5.882 0.250 15.00 10.67 150 0.508 6.944 7.353 1.600 10.00 5.33 300 2.475 9.259 9.346 5.988 3.20 2.40 600 8.000 12.346 12.500 12.048 1.20 1.13 662 8.929 12.821 12.821 12.987 1.00 1.00 1200 13.699 16.949 16.949 18.182 0.50 0.63 1500 19.231 19.231 19.231 21.277 0.40 0.53 2400 22.727 24.390 24.390 25.641 0.20 0.39 Note:

1] Photon relaxation length data from Radiological Health Handbook (ref.3).

2] Meter-Exposure Response Factor data, r(e), from

Eberline Micro-R-meter specifications (ref.2) (see figure 1). The factor, r(e), has the dimensions 7

of (meter units)/(uR/hr) .

3] Photon Field-Exposure Factor data, f(e), from Radiological Health Handbook - (ref.3) (see figure

, 1). The factor, g(e), has the dimensions of 0.77

((Cs-137 photons)/(cm /sec))/(uR/hr) .

' "z' .

t' .

e, a

e j L

e 4 r TABLE III I{

MEASUREMENTS OF ATTENUATION OF THE CONCRETE PRIMORDIAL RADIOACTIVITY BY LEAD ATTEUUATORS LEAD METER METER METER NORMALIZED ATTENUATOR READING ZERO NET READING THICKNESS OFFSET READING (gms/cm2) (uR/hr) (uR/hr) (uR/hr) (Mp(x))

0.000 5.83 0.20 5.63 1.000

! 0.432 3.20 0.20 3.00 0.800

< t 2.191 1.15 0.10 1.05 0.281 3.748 1.10 0.10 1.00 0.267 5.247 1.05 0.10 0.95 0.254

. . 6.717 0.90 0.10 0.80 0.213 8.130 0.85 0.10 0.75 0.200 15.625 0.70 0.10 0.60 0.161 23.265 0.65 0.10 0.55 0.147 26.840 0.65 0.10 0.55 0.147 58.090 0.20 0.10 0.10 0.027 Note: The source was the primordial radioactivity

- in the concrete floor of the Reactor Facility. The micro-R-meter was at one meter above the floor surface, surrounded on four sides and the top by 2 inches of lead. The rectangular aperture between the detector and the floor was(4 x 8) square The detector recessed 1.375 inches from i inches.

the aperture. Lead attenuators were added below t the aperture.

4 b

I I

o s,

se

{

gq -TABLE IV ,

'{l GAMMA FIELD ,

FOR NEUTRON ACTIVATED CONCRETE ISOTOPE ISO GAMMA GAM. RAD. GAMMA NORMALIZED SOURCE FRACT ENERGY FRACT LENGTH FIELD FRACTION OF GAMMA (Sj) (Ni ) (xi) (S$iN xi) FIELD t Average Energy of Group = 50 kev k* EU152 0.73 40 0.730 3.16 1.68396 0.08939 ,

EU154 0.06 45 0.260 3.35 0.05226 0.00277 EU152 0.73 50 0.002 3.54 0.00517 0.00027 Energy Group e db == 1. 74139 == 0. 0 92 4 3

______________i__________________________________________

Subtotals Average Energy of Group = 150 kev ,

EU152 0.73 122'O.280 5.52 1.12829 0.05989 l EU154 0.06 123 0.405 5.55 0.13486 0.00716 EU154 0.06 188 0.003 6.86 0.00123 0.00007 (g = 1.26438 Energy Group e $ == 0. 06712

_______________2__ Subtotals Average Energy of Group = 300 kev EU152 0.73 245 0.075 7.83 0.42869 0.02276 BU154. 0.06 248 0.066 7.87 0.03117 0.00165 l EU152 0.73 296 0.004 8.60 0.02511 0.00133 i

. EU152 0.73 329 0.'001 9.07 0.00662 0.00035 '

EU152 0.73 344 0.365 9.27 1.79328 0.09519 EU152 0.73 368 0.009 9.59 0.06301 0.00334 Energy Group e db == 2. 34788 == O .124 6 2

_______________3__ Subtotals Average Energy of Group = 600 kev

. EU154 0.06 401 0.002 10.01 0.00120 0.00006 EU152 0.73 411 0.022 10.14 0.16285 0.00864 EU152 0.73 416 0.001 10.20 0.00745 0.00040 3

EU152 0.73 444 0.028 10.54 0.21544 0.01144 EU152 0.73 444 0.003 10.54 0.02308 0.00123

-EU154 0.06 445 0.005 10.55 0.00316 0.00017 I_ EU15'4 0.06 478 0.002 10.93 0.00131 0.00007 E EU152 0.73 489 0.004 11.06 0.03230 0.00171 EU152 0.73. 503 0.002 11.21 0.01637 0.00087 EU154 0.06 558 0.003 11.81 0.00213 0.00011 g EU152 0.73 564 0.005 11.87 0.04333 0.00230 EU152 0.73 566 0.001 11.90 0.00869 0.00046 EU154 0.06 582 0.008 12.06 0.00579 0.00031 i EU152 0.73 586 0.005 12.10 0.04416 0.00234 EU154 0.06 592 0.048 12.17 0.03505 - 0.00186 i LLfI l

TABLE IV (CONTINUED)

P GAMMA FIELD

} ,^ FOR NEUTRON ACTIVATED CONCRETE ISOTOPE ISO GAMMA GAM. RAD. GAMMA NORMALIZED SOURCE FRACT ENERGY FRACT LENGTH- FIELD FRACTION OF GAMMA (Sj) (Ni ) (xi) (S3i N x1) FIELD Average Energy of Group = 600 kev (continued)

EU154 0.06 625 0.003 12.50 0.00225 0.00012 EU152 0.73 656 0.001 12.81 0.00935 0.00050

, EU152 0.73 674 0.001 12.98 0.00948 0.00050 1 EU154 0.06 677 0.001 13.01 0.00078 0.00004 5- EU152 0.73 679 0.005 13.03 0.04756 0.00252 EU152 0.73 689 0.008 13.12 0.07662 0.00407 EU154 0.06 692 0.017 13.15 0.01341 0.00071 EU154 0.06 716 0.002 13.38 0.00161 0.00009 EU152 0.73 719 0.003 13.41 0.02937 0.00156 EU154 0.06 723 0.197 13.44 0.15886 0.00843 EU154 0.06 757 0.043 13.76 0.03550 0.00188 EU152 0.73 765 0.002 13.83 0.02019 0.00107 EU152 0.73 779 0.127 13.96 1.29423 0.06870 Energy Group e = 0.12216

_ __ _ _ _ _ ___ _ _ __ _4_ _S u b t o t aol ==2.30152 s

Average Energy of Group = 1200 kev EU152 0.73 810 0.003 14.23 0.03116 0.00165 EU154 0.06 816 0.005 14.28 0.00428 0.00023

. EU152 0.73 842 0.002 14.51 0.02118 0.00112

EU154 0.06 845 0.006 14.53 0.00523 0.00028
  • r EU154 0.06 851 0.002 14.59 0.00175 0.00009 EU152 0.73 867 0.042 14.72 0.45132 0.02396 3 EU154 0.06 873 0.115 14.77 0.10191 0.00541

. EU154 0.06 893 0.005 14.94 0.00448 0.00024 EU154 0.06 904 0.008 15.03 0.00721 0.00038 EU152 0.73 919 0.004 15.16 0.04427 0.00235 EU152 0.73 926 0.003 15.22 0.03333 0.00177 EU152 0.73 963 0.001 15.52 LO3130 0.00060 EU152 0.73 964 0.144 15.52 1.63146 0.08660 EU154 0.06 996 0.103 15.78 0.09752 0.00518 1

EU152 0.73 1005 0.007 15.85 0.08099 0.00430 EU154 0.06 1005 0.179 15.85 0.17023 0.00904

[ EUl&4 0.06 1047 0.001 16.18 0.00097 0.00005

g. EU152 0.73 1085 0.002 16.47 0.02405 0.00128 EU152 0.73 1086 0.100 16.48 1.20304 0.06386 EU152 0.73 1090 0.017 16.51 0.20489 0.01088 EU152 0.73 1109 0.002 16.65 0.02431 0.00129

~

EU152 0.73 1112 0.133 16.67 1.61849 0.08591 EU154 0.06 1118 0.001 16.72 0.00100 0.00005 EU154 0.06 1128 0.003 16.79 0.00302 0.00016

, +

TABLE IV (CONTINUED)

GAMMA FIELD FOR NEUTRON ACTIVA?ED CONCRETE ISOTOPE ISO GAMMA GAM. RAD. GAMMA NORMALIZED SOURCE FRACT ENERGY FRACT LENGTH FIELD FRACTION OF GAMMA (S$ ) (Ni ) (xi) (S$iN xi) FIELD Averge Energy of Group = 1200 kev (continued)

EU154 0.06 1141 0.002 16.89 0.00203 0.00011 CO60 0.21 1173 0.210 17.12 0.75499 0.04008

% EU152 0.73 1213 0.014 17.41 0.17793 0.00944 EU154 0.06 1242 0.001 17.62 0.00106 0.00006 EU154 0.06 1246 0.009 17.65 0.00953 0.00051 EU152 0.73 1250 0.002 17.68 0.02581 0.00137 EU154 0.06 1274 0.355 17.85 0.38021 0.02018 EU152 0.73 1293 0.001 17.98 0.01313 0.00070 EU152 0.73 1299 0.016 18.02 0.21047 0.01117 CO60 0.21 1332 0.220 18.25 0.84315 0.04476 EU152 0.73 1408 0.207 18.76 2.83482 0.15048

] EU152 EU154 0.73 0.06 1458 0.005 1494 0.006 19.09 19.33 0.06968 0.00696 0.00370 0.00037 EU152 0.73 1528 0.003 19.54 0.04279 0.00227 EU154 0.06 1593 0.010 19.96 0.01198 0.00064

.. EU154 0.06 1597 0.018 19.98 0.02158 0.00115 Energy Group e5 Subtotals $o = 11.18354 = 0.59367

========================================================x i Totals for All Groups o =18. 83 871 p =1.00000

(

4 l

1

{.

J

. . - e us%&anMnet t MM%%A% .a < . -

TABLE V r

PHOTON FLUX DISTRIBUTION AND EXPOSURE BASED ON THE NEUTRON ACTIVATION SOURCE DISTRIBUTION IN CONCRETE AVERAGE l-PHOTON FLUX FRACTION---l METER FLUX EXPOSURE METER PHOTON l NO BUILDUP WITH l RESP. EXPOSURE READING ENERGY l BUILDUP FRACTION BUILDUP l FACTOR FACTOR

[1] l [2] [3] [4] l[5] [6] [7] [8]

ei l p(ei) g $ (ei) lf(ei) r(ei) Ei Mi l l 50 0.092 0.091 0.140 10.00 10.00 0.015 0.150

, 150 0.067 0.084 0.116 10.00 5.33 0.023 0.230 300 0.125 0.072 0.151 3.20 2.40 0.067 0.214 600 0.122 0.059 0.139 1.20 1.13 0.130 0.157 1200 0.594 0.000 0.454 0.50 0.63 0.764 0.382 Totals: 1.000 1.000 1.000 1.133

=====================================================

Note:

1] Photon energy of group, ei, from table IV.

2] Normalized photon flux, without buildup, from table IV.

3] Buildup factor for each group.

4o (1 + Bu)

4] Normalized photon flux, with buildup: =

Z o(1+ Bu) 7 5] Meter-exposure factor, r(ei), from Figure 1.

6] Flux exposure factor, f(ei), from Figure 1.

7] Normalized Exposure: E; = ( &/f)/I (4/f )

8] Meter reading: M; = r ( ef ) E; ,

)

b i

_ -- -_- -- - i

1 TABLE VI - a EXPOSURE ROOM EAST WALL i

Survey 1 meter Reading 1 centimeter Reading Background Block sR/g PR/m SAL l

f A-2 22.0 22.6 12.0 A-3 22.0 29.0 A-4 20.0 23.4

j. B-1 21.0 21.0 B-4 20.0 20.6 l

C-1 23.0 25.0 1

D-1 22.0 22.4 D-2 22.0 22.6 3-D-3 23.0 23.6 1

e s.

. . TABLE VI - b EXPOSURE ROOM SOUTH WA1.L

$' Survey 1 meter Reading 1 Centimeter Reading Background Block e e/a pd #N F-2 16.0 16.0 12.0 F. 3 16.0 16.0 F-5 15.0 15.3 G-2 20.0 22.4 G-3 20.0 22.8 l G-4 20.0 21.2 H-3 21.0 19.6 H-4 20.0 21.2 H-5 22.0 21.0 E I-2 22.0 24.0 1-3 22.0 20.2 I-4 21.0 20.6 I-5 21.0 20.4 J-1 23.0 19.8

. J-2 21.0 21.0 J-4 20.0 18.2 J-5 20.0 17.8 s.

6.

}

' TABLE VI - c EXPOSURE ROOM NORTH WALL Survey 1 meter Reading 1 centimeter Reading Block Ba:kground jg st/, p%

F-1 21.0 17.0 12.0 F-2 22.0 26.8 F-3 20.0 20.0 F-4 18.0 19.2 F-5 15.0 13.2 C-1 22.0 27.2 G-2 24.0 28.0 C-3 23.0 20.4 G-4 22.0 19.6 G-5 18.0 19.6 H-1 21.0 26.0 H-2 20.0 20.2 H-3 20.0 17.4 H-4 18.0 18.8 H-5 16.0 15.0 I-2 17.0 17.6 I-3 16.0 13.6 I-4 15.0 14.0 I-5 14.0 13.6 J-1 22.0 20.4 J-2 19.0 19.4 J-3 16.0 15.2 J-5 15.0 14.0

~

g i i

I" TABLE VI - d EXPOSURE ROOM FLOOR Survey 1 meter Reading I centimeter Reading Background Block st/u pG p ga, AA-1 22.0 26.0 12.0 AA-2 23.0 26.6 AA-3 23.0 22.6 AA-4 22.0 22.2 BB-1 22.0 22.6 BB-2 23.0 25.4 BB-5 21.0 13.0 CC-1 20.0 20.4 CC-2 21.0 22.6 CC-3 23.0 22.6 i

CC-4 24.0 22.4 i

22.0 CC-5 20.0 1

l 5" .

b.

-2 7-I..

~

s TABLE VI - e EXPOSURE ROOM CEILING Survey 1 meter Reading I centimeter Reading Background block p% p%, pt,La AA-1 17.0 16.8 12.0

AA-2 21.0 18.8 AA-3 18.0 20.6 AA-4 20.0 20.4 AA-5 18.0 21.2 BB-2 22.0 20.4 BB-3 20.0 20.6 BB-4 20.0 20.4 l BB-5 18.0 16.4 CC-1 18.0 16.4 CC-2 20.0 17.4 7

CC-5 20.0 20.2 l

l w

i e,

i l 1 I (b kh 5{ ibb

g. ,a, c-- a p .e, a4+L TABLE VII 9

BACKGROUND SURVEY DATA LOCATION GAMMAS uR[hrat1 meter

1. Cate 15 Area 12 12 747 Building
2. Southwest side 12
3. West side 11 -
4. Center 9

! 5. North-center 11

6. North-east 12
7. Quonset Hut 12 .
8. Quonset Hut 12 l
9. Quonset Hut 11
10. Building 3-61 12 1 11. Building 3-61 11
12. Building 3-61 12
13. Building 2-7 9
14. Building 2-7 11 l 4 15. Building 2-7 11 l
16. ParkingLot747fNESide 12 l
17. Parking tot 747f 12 1 ParkingLot747[NESide
18. NE Side 12
19. Utility Building 1-85 12
20. Utility Building 1-85 12

! 21. Wind Tunnel Parking Lot 12

22. Wind Tunnel Parking Lot 12

- 23. Trans-Wharehouse 12

24. Trans-Wharehouse 12 I ! Trans-Wharehouse 12 i

l Paved City Streets

~; 25. Northrop & Prairie 12 I

26. Prairie & El Segundo 12 27 Northrop & Crenshaw 12 12 i l

i I

v f., Average (30 readings) 11.6 STD Deviation 0.8 ,

i e

1. . __ _ _ ._ . --.

)

! . a 1

4

= r * >

s i

l 1

i s

i j

i

=

3 J

b i <

f d

.I 4

1 f

5 FIGURES I

i 1,

ie d

1 a

4 1

1 i '

k i

1 J

d R

i 1

1 i

1 b

4 g 5

  • i a' ,'),

d 4...

4 i

l 4

i k

r i )

6 r

IRETER AND FIELD FACTORS versus PHOTON ENERGY soo._

t o__

FLUX-EXPOSURE FACTOR f(e) m ~

O H _

O METER RESPONSE FACTOR r(e )

u.

J -

9 m _

w 2

3 z 1

- 1.g_ ,

s N

~

s o.1 i i i i iie i iit i i i . ieii; to too sooo PHOTON ENERGY (kev) l'&

Figure 1 n

4 iI li f

METER READING versus LEAD ATTENUATOR t.o -

2 CESIUM SOURCE (862 kev) j (M/Mo) = 1e-ste.se PRIMORlD AL CONCRETE SOURCE

' (sooxev) (1600xev)

~

(M/Mo) = 0.8 8 7 e**# ' + 0.3 3 3e * *#

NEUTRON ACTIVATION SOURCE too nev) (tsowev) (sooxev)

(M/Mo)

  • 0.131 e ~'#*' + 0.20 5e**# F + 0.18 0 e**#8 8 o.1 -- (soosev) (1200 Kev)

.~ + 0.13 8 e#8 8 + 0.337e''

  • 0

~

1

.). _

'! 3 w .

O Z

O -

w

&Y

$ o.o t- J l

$ 0 @0 2 - 4 O

4*O@

~

cn w &A S ~ 1 4 J O O

' < # A'o 0 4

$ ep Z

, O p l Y

" l O

c, o+ ,p  !

4

- o.oot- p O

< -T l

l

-w l

a' 1

e i e i i i g

i i i g

{ o j

. so toc E LEAD ATTENUATOR (GM/CM') l

- Figure 2 k

.C

i 4

E

<1 4

ATTENUATION LENGTH versus PHOTON ENERGY 100 -*

i en

~

i m

I e

i .

iI i .

1 p 10 -

N 5

~

\S 2 - l 4 w c .

, I p

C -

. Z t W

.J  %

Q I

Z 2

o  %. l p 4 i

j , 3 ,_

Z .

W . 6)

' ' >- ~

l H t .

~

9 -

1 i.

1 .

g ' ' ' 'I 3 3 3 g

l ( l

  • I 3 s v 1 i , y 500 i sooo PHOTON ENERGY (kev) t

'l p.

't Figure 3 i.

{-

i I NEUTRON ACTIVATION PHOTON FLUX DISTRIBUT10N 1

i WITHOUT SECONDARY BUILDUP 0.s -

2 S

! E l'

u 0.0. 0.07 0.1: 0.12 0.s e 0 , , ,

100 1000 PHOTON ENERGY (kev) l l

WITH SECONDARY BUILDUP 2 .

r 3 ( -

E

~

0.14 0.12 0.16 0.14 0.46

^

l 0 1 tbo ,

1000 PHOTON ENERGY (kev) 7 .

L

~

Figure 4

- BMdMMfkV!#3MMi&$3&fdRM@tMfl%RftrsEgyr

.s .

1 Figure 5a i

EXPOSURE ROOM- EAST WALL I i

1 I

i- i 44 b4 c4 D4 1

A3 j DS i- PL Lis /

4 WOR TH f ## SouTS*

i A2 D2 i

74 NA

('

A/ b/ C/ D/

1 4

5 u

A 4:

. *1

[ .

am 41 4.-&. .a-- - m - - ..-4 .. , . . _ .

  • A s.4a .- ,,a..h 4F-m g

i 4

e FIGURE Sb EXPOSURE ROOM - SOUTH WALL 4 i G5 pg JS ZS MS 44 s4 p4 J4 Z4 Z3 Hb G3 Fa 3

J3 wasr E4s7 1

J2 Z2 N1 G4 F2

. -- FLOOR L B vfl r

?

M/ at pt Ji Z/

4 2

T- .

u[

n k.x

FIGURE Sc EXPOSURE ROOM - NORTH WALL i

FS G5 HS IS ds l

F4 G+ M4 Z4 J4 l

l F3 Gb H5 ZS JS 1 wssT 24'I' l G2 NZ F1 12 J2

- pt.oOR 'J.2 VE/

,ri ai si l zi s.i I

t.

I i

l

9 9 FIGURE 5d EXPOSURE ROOM - FLOOR AA/ AA1 AAS AA4 9 4

m WAS7* bb/ bb2 boa ob4 Q BAST eq e

CC/ CC2 CCS CC4 0 l U

'h r

i

_m:namenuman

J 4

i ,

i i ,

7-f i

FIGURE 5e i

1 EXPOSURE ROOM - CEILING i 1 4

I 4

1.

i C C/ CC2 CCS h l

CC4 g

! O WBST bo/ bb2 boa BB4  % EAS7*

9 i

t

/ AAS 4A4 h i r '

, e d

?

1 d

h L.

, 1 i..

. f.

L

c_

i .

FIGURE 6 i BACKGROUND SURVEY AREAS LOCATIONS l '

lI" j.i . 3.yp ,

i g 0ie -

l ,4 ign es i ~.,

\-

T 3e 0 '

% a. . -

~

- =Y d S o o\-q" d a  ?- i o i U 4 pil, g-

@ ,V .fL ~'

I' I t-

~

o l'

. , %; i

~

  • ' OF- c ~l l I j'

i T_ Q g 3

i 4 - 1 C2 p_sw 3 ,

b k -

"O \

. a i cn i i

- g 4

L

,- a b

I t- D

'l i

  • L ( _m q f )E

_- )

\

\

c " .I l

L. {

t u / '

emainig aviseuli 1 The construction in the areas surveyed was done during the same time period as the reactor facility. The measured radiation at these locations is consistent with measured natural primordial

- radioactivity in, the concrete and soil, and with the measured radioactivity in the reactor facility.

E m

+

r.

ti 4

4 i

l [JLRTorM576TiG or DRic ks J

(a) 1 ,

4h Sh /L' T4tck-LsAo enicus TYPICAL i .i /-N xN

N / s

%s /

) '/ sl/ 4

/%# s/

pbcrrTom R usvaa. @ # /N g, % /- f

,/

  • or paicas

=

+- (b) p FIGURE 7 a & b r

e P'x.s eLoT svlwoosv

. g..
l, s

l'N

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\

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l I y l I l l l l l l l l l l Ii

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l

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t- .

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j..

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li i .. FIGURE 7c

_42-

  • Jo+ asoa Paoesaw l

L FIGU3E 8 STom4*o Wap A - 5, DATA : So JuuAra.y sq st, 60tavan a,Y deemea P.:: . C= z mas. T m s.: 15:00 - l~l:Go M A nr. g. O s s o : 1.um.ow MCP IT,3g k,6P). 3pMM CAL,te,rasno4 va73.; to/j+/es 4 . , e e w x a - u ,a i . , I e e 1 9o 8.s .__, 85

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l

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, .$ l 1

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, !98 . _8.G  ;

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, e n..,. ~.e. .~ e a. - . . . . . . - . . ~-- - - * ~ ~ ~ = - ~ ~ ~ - " ' ' '6 1L I

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. -- 7 -- --

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.i FIGURE 9a i

ABSTAT 4.09 STATISTICAL ANALYSIS OF CONCRITE RUDBLE IN PARXp4G LOT FILES CONCLOT REV6 0 ~PAGE 1 COMMAND; DESC MISSING VALUE TREATMENT: VARWISE THERE ARE 1 VARIABLES AND 38 CASES IN THE DATA SET 38 CASES (100.0%) ARE VALID STD ERROR COEFF OF MEAN STD.DEV. VARIANCE OF MEAN VARIATION (

i VARIABLE 3.02120 i 1 GR 8.77105 0.264991 0.0702205 0.0429873  !

{ I MINIMUM MAXIMUM RANGE TOTAL VARIABLE 1 GR 8.20000 9.30000 1.10000 333.300 l

i MEDIAN MODE SKFWNESS KURTOSIS VARI AB LE 1 GR 8.80000 8.80000 -0.274444 2.22932 l

l I COMMAND: HIST MISSING VALUE TREATMENT: VARWISE i VARI AB LE : 1 GR J

10 15 20 25 30 AT LEAST 8.00000 5 PUT NOT OVER: FREQ  % +-------+--------+-------+-------+--------+------->

8.00000 0 00.0 I

- 8.20000 1 2.6 IXXXX 8.40000 4 10.5 IXXXXXXXXXXXXXXXXXX 8.60000 8 21.1 IXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXX 8.80000 9 23.7 IXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXX 9.00000 10 26.3 IXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXX 9.20000 5 13.2 IXXXXXXXXXXXXXXXXXXXXXX 9.40000 1 2.6 IXXXX 9.60000 0 00.0 I 9.80000 0 00.0 I 10.0000 0 00.0 I --+------ ,

+-------4......--+-------+-------+---..

5 10 15 20 25 30 TOTAL 38 100.0 a

  • =

l

\

l l

  • i FIGURE 9b ABSTAT 4.09 FILE: BXGDLOT STATISTICAL ANALYSIS OF BACXGROUND IN PARXING LOT REV8 0 PAGE 4 COMMAND: DESC ll MISSING VALUE TREATMENT: VARWISE THERE ARE 1 VARIABLES AND 5 CASES IN TFE DATA SET 5 CASES (200.0%) ARE VALID VARIABLE STD ERROR COEFF OF i MEAN STD.DEV. VARIANCE OF MEAN VARIATION f 1 GR 8.44000 0.0894427 0.00000000 l 0.0400000 1.05975 i

VARIABLE MINIMUM MAXIMUM RAtiGC 1 GR TOTAL 8.30000 8.50000 0.200000 42.2000 VARIABLE MEDIAN MODE SE CWNESS 1 GR FURTOSIS  !

8.50000 8.50000 -0.843750 2.07812 i

i COMMAND: DIST '

l MISSING VALUE TREATMENT: VARWISE VARIABLE: 1 GR AT LEAST 8.00000 10 20 30 40 50 60 BUT NOT OVER: FREQ  %

+-------+--------4-~~----+-------+--------+-------+

i 8.00000 0 00.0 1 8.20000 0 00.0 I 8.40000 2

. i 8.60000 3 40.0 IXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXX 60.0 IXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXX.X '

' 8.80000 0 00.0 I 9.00000 0 00.0 I

" 9.20000 0 00.0 I i 9.40000 0 00.0 I  !

I 9.60000 0 00.0 I

' . 9.80000 0 00.0 I l i

10.0000 0 00.0 I TOTAL 5 100.0

+____.._4...___.-+_....4....-_.+.___..._+_..-..+

10 20 30 40 50 60 d

h, .

r i

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I L l A

5 b#

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- - - FIGURE 10 Jonssoa PaoPe n. ry one unru. sunvey GTo***

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FIGURE 11 J Meoa Paoesn. ry SuR.FA u % uva.y OT##* *

  • Y"'" * ~I DATA l ~b MA Rc44 198d;.

Sop.vsy a,y ; Ta u r-: 08:5o MATIE.r2, OSEO! LUDLUM Moo gqp,R,SS S.J4%

CAL tear 2ATiod DATE: to/g/85 A s. c. p a F e W

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REFERENCES

1) Instrument Specification, Ludlum Model - 19 Micro-R-Meter, Ludium Measurements, Inc., Sweatwater, Texas.
2) Instrument Specification. Eberline Micro-R-Meter, Eberline, Santa Fe, New Nexico.
3) TRIGA Reactor Decommissioning Final Report, Northrop Corporation Beverly Hills, California.
4) Radiological Health Handbook, Revised Edition,1970, U.S.

Department of Health and Welfare.

5) David C. Kocher Decay Tables, Technical Information Center, U.S. Department of Energy.

. 1

.e s

_ _ _ _ _ . _ _ . _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ . _ _ _ _ . _ _ _ _ . . _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ . _ __ ___ _ _ _ _ . _ _ _ _ _ _ . _ _ . . _ _ _ _ _

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