ML20064M100

Public Version of Radiation Shielding Analysis Brunswick Technical Support Ctr/Emergency Operating Facility

ML20064M100
Person / Time
Site:	Brunswick
Issue date:	06/30/1982
From:	Brode R, Nathan S NUS CORP.
To:
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ML20064L981	List: ML20064L993 ML20064M100
References
NUS-4148, NUDOCS 8208260350
Download: ML20064M100 (62)

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1 RADIATION SHIELDING ANALYSIS FOR THE

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BRUNSWICK TECHNICAL SUPPORT CENTER / EMERGENCY OPERATING FACILITY a

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t.J TABLE OF CONTENTS j

Section and Title Page No.

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LIST OF TABLES 11 i! If j

c,i LIST OF FIGURES iii W

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1.0 INTRODUCTION

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SUMMARY

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'j 2.0 ATMOSPHERIC DISPERSION ANALYSES 2-1

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2.1 Meteorological Data 2-1

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U 2.2 Calculations 2-2 l

2.3 References 2-4

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3.0 RADIOLOGICAL ANALYSIS 3-1 D

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.- l 3.1 Methods 3-1 I

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3.2 Assumptions 3-2 IU 3.3 Results 3-3

]t, 3.4 References 3-4

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5 APPENDICES L-L,h APPENDIX A JOINT FREQUENCY DISTRIBUTIONS A-1 OF WIND SPEED AND WIND l

DIRECTION BY ATMOSPHERIC 1.

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STABILITY CLASS 8

APPENDIX B ONSITE METEOROLOGICAL PROGRAM B-1 APPENDIX C METHODS USED IN RADIOLOGICAL C-1 ANALYSIS f

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LIST OF TABLES 1

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Title Page No.

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,1 2-1 TSC/ EOF X/Q Values for 100 m Stack 2-3

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,.1 Releases 3

3-1 Assumptions in Radiological Analysis 3-6 of Brunswick TSC/ EOF

~j 3-2 Results of Radiological Analysis of the 3-7 j

Brunswick TSC/ EOF B-1 Operating Conditions B-7 j

B-2 Major Components B-8

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B-3 Operational Sensor Elevations B-9 n

B-4 Component Accuracy B-10

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C-1 Nuclide Decay Constants and Fission C-18 Yields F'}

(1 C-2 Average Beta and Gamma Energies and C-19 Iodine Inhalation Dose Conversion Factors l-I'!

Ij C-3 Isotopic Gamma Energies and Decay C-20 Fractions 4

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3-1 Schematic HVAC Flow Diagram 3-8 j.

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3-2 Brunswick Steam Electric Plant TSC/ EOF 3-9 Integrated Dose Contribution C-1 Dose Model Activity Flow Schematic C-23

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1.0 INTRODUCTION

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SUMMARY

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~1 j j NUS determined the shielding requirements for the Technical Support Center (TSC) and Emergency Operating Facility (EOF) to

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be built at the Brunswick Steam Electric Plant.

The location 1

of this facility is shown on Figure 1-1.

The TSC/ EOF is de-l signed to meet the criteria given in GDC-19 of 10 CFR 50 whie.h specifies an exposure limit of 5 rem whole body or its equiva-lent to any part of the body for personnel within the facility for the duration of the accident.

The accident duration used is 30 days.

In order to achieve this dose, a design basis

,]h limit of 4.5 rem while occupying the facility was established I

by Carolina Power & Light Company (CP&L).

4 9 l0 Sources of radioactivity that were considered in determining

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the shielding requirements included airborne radioactivity in 1

the secondary containment (reactor building), radioactivity contained in pipes and other equipment both inside and outside q

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body dose due to the passing plume, consisted of two source components, which are: 1) gaseous radioactivity surrounding L

t the facility itself and 2) gaseous radioactivity that accumu-L lates within the facility due to the operation of the ventila-7 p tion system.

NUS determined the whole body dose contribution d

from both components, and the thyroid dose contribution from the activity in the facility.

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NUS also reviewed the plant design to determine the location h

of radioactivity sources in pipes and equipment during an J

accident.

This was done in conjuction with CP&L personnel.

L The KAP-VI computer code was used to determine the shielding requirements for direct radiation from the secondary contain-ment (reactor building) and other sources, and the AXIDENT code q

E3 was used to determine the dose due to the passing plume.

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ll The methods used to estimate atmospheric dispersion of

'I radiological releases and the resulting dispersion f actors are o

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The methods and assumptions used to perform radiological dose analyses, and

' }5..f.,l the resulting integrated doses are presented in Section 3.0.

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r l 6i Based on these analyses, the proposed design for the TSC/ EOF,

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14 inches thick, will result in a whole body dose of 2.4 rem j

integrated over the 30 day duration of the accident.

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~ 1 ~, p"; O er [g ~, i .l u. I 2.0 ATMOSIRERIC DISPERSION ANALYSES t ~ ' i '9 3, Atmospheria dispersion estimates of radiological releases were caldu' lated for the propored Technical Support Cen-i ter/ Emergency Operating Facility at the" Brunswick plant. Cal-1 / culations of relative concentrations (X/Q) were based on 0tj; appropriate conservative models and methodology. Values of X/Q were computed using guidance input information from the following: w i e, j o Regulatory Guide 1.145, " Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants." (Ref. 1) ]g o NUS-3697, " Control Room Habitability Evaluation k) Brunswick Steam Electric Plant." (Ref. 2) h.J i 1 o Partial Plot Plan Brunswick'. Plant (Draft), Drawing j No. D-2386. i j l l. l 2.1 Meteorological Data l i.., u t.! 7 r. j; U Meteorological data for the steaspheric dispersion analyses f.g were collected at the site during* the 4-year period January 1, L l 1976 through December 31, 1979. The data used for each j I aource release type are listed belok r T [# Atmospheric Wind Speed / Combined Data Analysis Stability Wind Direction Recovery % m b L ll Radiological T (105-11m) 105-m level 98 w, A / 4, j,-

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a !qd ( ..} The joint frequency distributions of wind speed and wind di-i [] rection, by atmospheric stability class for the 105-meter d level of wind data, are provided in Appendix A. A brief dis-cription of the onsite meteorological system is provided in N Appendix B. i!Lj 2.2 Calculations 1 The methodologies of Reference 1 were used to compute X/Q ~ values for radiological releases from the 100-meter stack lo- [ cated approximately 183 meters from the proposed TSC/ EOF. L Because the stack is freestanding, a 100 percent elevated re-1 ease was assumed. Winds from the E, ESE, and SE were deter-q 'j mined to affect the TSC/ EOF. The highest 0.5 percent X/Q value from these sectors was determined for the 0- to 2-hour e,, y i, ; time period. Values for time periods greater than 2 hours were determined by logarithmic interpolation between the 2- ,] [i hour and the sector annual average values. In addition, a X/Q i '; value applicable to the first one-half hour of the accident i was calculated representing the fumigation case for an elevat-l' ed release at an inland site (greater than 3200 meters from a l coastline). The X/Q values are given in Table 2-1. '? lE L n i1 l4 \\ 1 lN 1 [c3 I 1u LI p L 2-2 ,p-1. L. er NUS COAPCAATION

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,3li 0-% hrs 2.6 x 10~4 (fumigation) %-2 hrs 3.1 x 10~6 T 2-8 hrs 5.6 x 10~7 h2 8-24 hrs 2.4 x 10~7 1-4 days 3.8 x 10-8 !{j tU 4-30 days 2.7 x 10~I 6 ,e 1 'il a ths 9 ? h1 ,i t_. ] 4 r] !) a F, a L I G TJ r em b i l'

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~. ^^' -...~..,.l._.... m I '} d j 3.0 RADIOLOGICAL ANALYSIS ElG This section summarizes the methods and results of the radia-t 1 tion shielding analysis for the Technical Support Center /Emer- )] l] v. gency Operating Facility during postulated radiological accidents at the Brunswick plant. .i The TSC/ EOF is designed to meet the criteria given in GDC-19

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of 10 CFR 50 which specifies a whole body exposure limit of 5 j rem whole body or its equivalent to any part of the body for q personnel within the facility for the duration of the acci-j l.} dent. The accident duration used is 30 days. o t'i a jt Sources of radioactivity that were considered in determining the shielding requirements included direct shine from airborne {. radioactivity in the secondary containment (re'ac tor o building), radioactivity contained in pipes and other equip-11 p ment both inside and outside the secondary containment as well d as radioactivity f rom the passing plume resulting from second-j ary containment leakage. The whole body dose due to the pass- [ i;j ing plume consists of two source components, which are: 1) a gaseous radioactivity surrounding the f acility itself and 2) ] gaseous radioactivity that accumulates within the facility due [ ~~ to the operation of the ventilation system. The thyroid dose li i2 is due to iodine activity which accumulates in the facility. ll Id (i 3.1 Methods 1 b The methods used to calculate the beta and gamma whole body doses and the thyroid dose to personnel in the TSC/ EOF are standard calculational techniques for modeling the genera-tion, release, transport, buildup, and removal of radio-l nuclides. The equations used to model these phenomena are l,l r. well known, and the specific equations incorporated into the t ( '. ) computer program used in this study to calculate the TSC/ EOF l6 doses are presented in Appendix C of this report. The methods j used to compute the whole body dose contributed by sources of j-r 3-1 NUS COAPOAATION l .._-.....,,.y..,.,.. ,m

-u . ~.... e O r" b m direct radiation outside the TSC/ EOF are based on the work of ~ 1 p Jaeger, Chapter 6 (Ref.1). The shine dose from liquid source j U terms is calculated using the NUS computer code CYLDOSE, which ] is described in Appendix C. Li 3.2 Assumptions .t! [! u' i The assumptions used in this analysis of TSC/ EOF radiation ex-posures are described below and in Table 3-1: j: 1 's 1 o Radionuclides released from the reactor core are un-i iformly distributed throughout the primary contain-i ment. Radionuclides released to the secondary rj !,j containment (reactor building) are assumed to be j uniformly distributed throughout the secondary con-tainment. r o The primary containment leaks at a constant rate of 0.5 percent per day for the duration of the accident. io!J o The primary containment is assumed to consist of a 'l single volume with no washout of radionuclides by containment spray. U I! l: i 4. o The secondary containment (reactor building) ex- .i haust rate is assumed to be one secondary contain-a

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j tj ment volume per day. L p o There is no direct leakage from the primary contain-g ment to the environment. All exhaust from the j [~ secondary containment (reactor building) is filter-i ~ ed by the standby gas treatment system. I r]

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t Ii ~ tJ o The accident duration is assumed to be 30 days. r, o Radionuclides in the TSC/ EOF are assumed to be n l! uniformly distributed throughout that volume. I' o The breathing rate in the TSC/ EOF is assumed to be -4 3.47 x 10 cubic meters per second for the duration of the accident. o X/Q values are not adjusted for the occupancy f actors given in NRC Standard Review Plan 6.4, since the TSC/ EOF is continuously occupied. I o The TSC/ EOF ventilation and cleanup system Iodine l removal filters are sufficiently shielded to have a negligible dose contribution. F 3.3 Results The radiation dose to individuals within the TSC/ EOF during a postulated design basis accident at the Brunswick station is computed using the assumptions above and those presented in r: Table 3-1 and Appendix C. The meteorological data are based on the information presented in Section 2.0. i As described in the Brunswick FSAR, the maximum calculated l [ dose to individuals at the site boundary or within the control room occurs during a postulated loss of coolant accident j (LOCA). This is because the magnitude and duration of the l radionuclide release during a LOCA is much greater than that l i' for any other accident. This is discussed further in Refer- "I ences 2 and 3. Based on this information, the LOCA was se-lected as the basis for the shielding design of the TSC/ EOF. p W l 3-3 NUS COAPCAATION d 8 "' g 9, spo ,p e easgn.. ,go, ,upe my age..y gain awe 9 +p . g,w e g g

-.w.. s I: qU !l i The dose to TSC/ EOF personnel from radioactivity buildup with-in the facility is calculated using the HVAC system model and q the data shown in Figure 3-1. This figure is based on infor-mation supplied by J. E. Sirrine I. The 30-day integrated j dose due to airborne radioactivity within the TSC/ EOF is summarized in Table 3-2. The dose due to various sources of radioactivity outside the TSC/ EOF is also listed in Table 3-2 "I for the proposed shielding thicknesses. ql3 The dose contribution due to secondary containment (reactor building) shine was calculated using the KAP-VI computer code ' g[j (described in Appendix C). The features of the secondary con-tainment (reactor building) and TSC/ EOF building essential to the shielding analysis were inputs to the KAP-VI code. The sources used in the KAP-VI code were based on the method de-j [' scribed in Appendix C. These results are based on a secondary containment (reactor building) concrete wall thickness of 2.0 feet up to the refueling floor and sheet metal (ignored for i G shielding calculations) from there to the roof. The dose due to the passing plume outside the building was calculated j [j assuming an infinite disc source using the CYLDOSE computer 1 l code (described in Appendix C). The integrated dose as a j l function of shielding thickness is shown in Figure 3-2 for t l the secondary containment (reactor building) shine. i q fd As shown in Table 3-2, the calculated doses are well within the current NRC criteria of 5 rem whole body for a TSC/ EOF with 18 inches of concrete on the north wall and 14 inches of 3 concrete on the roof and remaining walls. ll q t .i ( 3.4 References Ob 1.

Jaeger, R.

G., et al., Engineering Compendium on l Radiation Shielding, Volume 1, Springer-Verlag, Inc. New 7 U York (1968). i. p {, 3-4 I-ff NUS CORPORATION

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~.... ^ .h ( - c, 1 S TABLE 3-1 ASSUMPTIONS IN RADIOLOGICAL ANALYSIS OF ~ BRUNSWICK TSC/ EOF D Ii Power level = 2,350 MWt Operating time = 1,000 days Fraction of core radionuclide inventory released to containment i Noble gases = 100 percent Halogens = 25 percent 3 "I Drywell free volume = 164,000 Ft 3 Maximum / minimum wetwell free volume = 2,000,000 ft [j i i Standby gas treatment system flow rate = 3,000 cfm Standby gas treatment system filter efficiencies for iodine lq Elemental = 95 percent i 6 Organic = 95 percent Particulate = 99 percent Primary containment leak rate = 0.5 percent / day Secondary containment air exchange rate = 100 percent / day J I 3 TSC/ EOF volume = 313,600 ft F-TSC/ EOF intake flow rate = 1800 cfm 1 TSC/ EOF ventilation system filter efficiencies for iodine l, r ll Lj ll Elemental = 99 percent lj Organic = 99 percent } Particulate = 99 percent 9 f Stack height = 100 meters b 't r, l i l,

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j Due to Activity in Containment 2.34 Due to Activity in Pipes and Equipment 0.008 1

l .- i i Total 2.4 j J y a ) ,1 f-] Thyroid u i Due to Activity in TSC/ EOF .01 i 11 l1 1' l' l l} ':1 li .] lf i

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m.i . :.1 m ....i..._.._.. . ~. _. _ l ii : 1 la r1 .b APPENDIX B I Bl.0 ONSITE METEOROLOGICAL PROGRAM i e, 3 il A 360-foot, guyed, open-latticed tower supports the lower and [ upper levels of meteorological instrumentation. Wind di-r e<: tion, wind speed, wind variance (sigma theta), and dew point temperatures are recorded at both levels. Ambient tem-(-b perature is measured at the lower level. The differential temperature between the upper and lower levels is measured by twin, redundant delta temperature systems operating simultan- ) eously. Solar radiation and precipitation are collected near ground level. The wind sensors are mounted on 12-foot booms oriented perpendicular to the general northeast-a.outhwest prevailing wind flow to minimize tower shadow effects. The temperature probes and lithium chloride dew point sensor are housed in Climet aspirated shields mounted on 8-foot booms. A i q l' complete specification of major system component operating conditions is presented in Table B-1; component manuf acturer l l and manuf acturer model numbers may be found in Table B-2. a Operational sensor elevations are displayed in Table B-3 and component accuracies are shown in Table B-4. li The meteorological tower is located 0.3 miles north-northeast U of the reactor complex, with the base of the tower at 21 feet ], above mean sea level. An environmentally controlled shelter, i! 4 which houses recording instruments, signal conditioning de-1; s l; vices, and remote data access equipment, is located adjacent l to the tower. i, '! ~ The Westinghouse Environmental Monitoring System is the I ll primary data collection system. This system converts sensor outputs to a proportional number of discrete pulses that are lj y, i ll.= l! l; B-1 i j! NUS COAPOAATION i, .a t

• g er

==w ww. s g =, m.=, e -m* w ,-q+ee-g,ap= p pgy - gm. g ,e e p.-.g

aw q

yp geg

... =.: :.. ^^ ~ ^

')

m i electronically integrated and recorded on magnetic tape in 15-minute averaging periods. Also, direct readout of any para-meter is possible with this system. A test jack for each ,Il parameter is provided so that a pulse test counter may be plugged into it. The counter sums the pulses produced in a

'~

specific time interval, and the subsequent pulse total can then be converted to engineering units by use of a formula of the form y = mx + b. r Esterline Angus Twin Strip Chart Recorders are used for providing an analog record of both the upper and lower level I wind directions and speeds to back up the Westinghouse system. l In addition, 15-minute averaged upper and lower level wind apeeds and directions, both differential temperatures, and I ambient temperature parameters are telemetered to the CP&L l' general offices on an hourly basis via voice grade telephone lines to the site, giving CPEL the capability of detecting malfunctions of these parameters within 24 hours. a i t l3 I' 1 j !I l -- ir lI l! l B-2 i NUS CORPOAATION ...._-,,...J,.n,.j7,, ..,,,,,,r..

l Il L} !l r' l il

. J

l B2.0 DATA REDUCTION

] a The Westinghouse system magnetic tape cassettes are changed and brought back to the general office approximately once per j month for translating. Computer programs convert all para-P meter pulse totals into engineering units. The data is then r i. i reviewed and checked for consistency with the onsite strip i. i charts and the Wilmington, North Carolina, Weather Service data. The edited 15-minute averaged data is then compiled into hour-ly averages and stored on magnetic-history tapes. m

p. !td Routine computer outputs from the Westinghouse pulse data collection system include the following:

b:3 l il a. Monthly Data summaries listing maximum temperature, a p 1 P, average temperature, barometric pressure, precipi-i; t-tation, solar radiation, and upper level and lower fi, level dew point temperatures as a daily average and Li 1j monthly average. i j 9, t 'j U b Hourly averages of precipitation, barometric press-ure, ambient temperature, differential temperature, { .vrJ upper and lower level dew points, upper and lower .m j level wind directions and wind speeds, upper and ,! Il lower level wind direction variance (sigma theta), 6 j '", Pasquill stability classes (as outlined in Regula-al dr" tory Guide 1.23) computed from the average of the lA;a two delta temperature

systems, and accumulated l

j solar radiation (langleys/ minute) l I'j gt (9 ll % ]O ll.

l., '"

B-3 ',] L, u hj NUS COAPCAATION 1u - T " m ~ " " - ' ~ ~...

I J e j. s.a ? c. The 15-minute averages of both upper and lower level wind directions, speeds, and sigma theta; baro-4, [l metric pressure; and accumulated solar radiation j 'l d. Joint wind frequency distributions by direction (as p, [ outlined in Regulatory Guida 1.23) for both upper and lower levels, showing average wind speeds and i number of unrecovered data hours I1 The analog strip charts are changed twice per month. They are used as backup data to provide checks on the other systems and to provide consistency of data. .g I 64 .J t {j i L l i; q ! L3 i

• j'l i

!a

• o L.

I!p !L a e + t 4 4 L. f4 B-4 NUS CORPORATION i. _ _. ~~.

it i h l t Li d,,.. O r{ B3.0 MAINTENANCE AND CALIBRATION 9 @U An onsite maintenance and calibration program was initiated in 1976. Regulatory Guide 1.23 data recovery requirementa are m met by performing scheduled calibrations carried out on a d semiannual basis such that } n [] a. All wind systems are changed and replaced with National Bureau of Standards (NBS) traceable cali-i [, brated wind sensors, per Regulatory Guide 1.23 - lp 1 b. All ambient and differential temperature systems 4 l1 a are changed and replaced with NBS traceable cali-

)

brated systems, per Regulatory Guide 1.23 ? l l c. The lithium chloride dew point sensor bobbin is -l changed ~f d. The Cambridge dew point systems are changed .l !,[> e. Calibrations of the barometric

pressure, solar J

i radiation, and precipitation systems are verified a (sensors are changed on an annual basis) c{ a f. All other onsite equipment is calibrated or its l} calibration is verified ll " H] l4 In addition to the scheduled calibrations, interim calibra-L tions are performed at 6-week intervals. A further enhance-q ment of data recovery is achieved by operating twin, redund-l 1,!-! delta temperature systems simultaneously. Comparison of

ant, q.

j d i. B-5 l 9 NUS COAPORATION ' ' ~ * ^

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r. - ~ - -,-r g.,,.. y y y z n -.~~,-.n,

I i-e,,

i j U j

. j the two systems on a real-time basis through the hourly data

~' (received at the CP&L general offices) gives CP&L the capa-bility to detect discrepancies in either system, usually with-in 24 hours (except on weekends). l r'l a n } m i:1 lJ J q u f 4 ..i .t I 6 l' k ll t! I' I ~ f. 1 i 1; I l' ' Im l !. f f I f'i i i u ., 9 B-6 L' I i r) '.] NUS COAPORATION 1 b ~

.ig D R a r; 4' ..) .o ? U TABLE B-1 m

\\

,dj OPERATING CONDITIONS l; Component Conditions h Wind sensor -40 F to+120 F, up to 100 percent d relative

humidity, up to i

125 mph wind speed Temperature sensors -50 F to +130 F Aspirated temperature shields -60 F to +150 F G 1 62 Honeywell dew point sensor -40 F to +160 F, 11 percent relative t i,. humidity and above l Cambridg6 dew point system Transmitter Unit -80 F to +16G F g l control unit -80 F to +120 F r,; j Total precipitation sensor No limitations d Solar radiation sensor No limitations y 1 Barometric pressure sensor -30 F to +170 F, 0 percent to l Il 90 percent relative humidity

s. /

I Magnetic tape recording -20 F to +140 F 0 I,j packages c Strip chart recorder +20 F to +120 F }I' Signal converter (transmuter) -40 F to +120 F, 5 percent to 95 percent relative humidity R h Telecoder (encoder) 0 F to +120 F, 0 percent to J 100gercent relativg humidity at +77 F to +104 F without

ji

O condensation b -%.~ m ~1 l* ^ I, l: a l :- !j B-7

I I$

r NUS CORPORATION u 11 -._.._n..

.. _ ~... ]; r!d ITU b TABLE B-2 a MAJOR COMPONENTS F Is: ~ Component Manufacturer Model Number I'I U Sensors Wind sensor Meteorology Research, Inc. 1074-22 Single-element Rosemount 104ABG-1 temperature sensor i Dual-element temperature Rosemount 104ABG-2 w' sensor U Dew point sensor Honeywell SSPO29DO21 ~, Total precipitation Weathermeasure Corp. P-511E ,j sensor t ll Solar radiation sensor Eppley Laboratory, Inc. 8-48 i Barometric pressure Rosemount 1105A9Al j sensor f]l I Cambridge dew point EG&G International, Inc. 110 I sensor (transmitter unit) l 9 Sensor support equipment Cambridge dew point EG&G International, Inc. 110-C1 g {, control V Strip chart recorders for Esterline Angus E1102R wind speed and direction 1 Ild Aspirated temperature Climet 016-1 shield for single-y element temperature h sensor i h Aspirated temperature Climet 016-2 q shield for dual- .,d L element temperature sensor and Honeywell ] (Q dew point sensor L l L! p B-8 r,j u l NUS COAPORATION i u,.....

L.... ' :~' ~ .(..-...,,a. .:..' l.......... a,... x.,. I U .'J i']b p g,j TABLE B-3 OPERATIONAL SENSOR ELEVATIONS ul e Operational Elevations

j If Sensor Above Tower Base (m) d we Wind 11.5 and 104.6 54 Honeywell dew point 10.3 F}

Cambridge dew point 11.5 and 104.6

)

i.. Solar radiation l '. 5 Differential temperature 10.2 to 103.2 Precipitation 1.5 e I 1 J Barometric pressure 1.5 r I

j 1

u .1 !I I. !, l n 1 L) et b= J r1 0 n- 0g B-9 q. t 9. d

,~4 NUS CORPORATION u

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• P a
• NW*
• 4 ***

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..-.~.~..:.a........ .....,.w. .L f4 fG d A 1 L b TABLE B-4 .J COMPONENT ACCURACY p id Component Accuracy l Wind sensor 6) Wind speed 10.4 mph or 1 percent, whichever is i greater = 1.0 mph 7 Wind direction, O to 540 1 5.4 degrees F" Honeywell dew point sensor +2 F at or.above 11 percent relative Eumidity 3u Cambridge dew point system 10.5 F (errgr extreme) above a dew r point of -20 F (excluding readout instrumentation). Error extreme 4 increases in approximately ligear ~' fashion to 12 degrees at -80 F. Solar radiation sensor 10.04 calories / square centimeter / minute (pyranometer) (langleys) Differential temperature system +0.186 F over ambient temperature range c Trom -50 F to +130 F v Ambient temperature system 10.498 F q d Magnetic tape recorder il pulse per interval fl Strip chart recorder il percent of full scale, direction = p d 15.4 degrees, speed = 1 1.0 2jn u Total precipitation sensor 10.5 percent calibrated at 0.5 inch ( lj per hour) 'l l4 Barometric pressure sensor 10.006 inch of mercury (temperature !i Q effect: 10.1 inch of mercury per 100 l Oi degrees of Fahrenheit operating 1 temperature span) 1 n

]

b.- ') B-10 1-c, l iLi NUS CORPORATION ..-. l[ .-..-.,.y..,. . --.,,... ~.

,i .._ i

1...... ~ _ - i. -

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• 4

] APPENDIX C !2 U METHODS USED IN RADIOLOGICAL ANALYSIS

{

l The dose calculation computer program (AXIDENT) which consists - f(1 of a release pathway model and a dose evaluation model was l! used to evaluate dose contributions from the passing plume. j c.$ The release model computes activity inventories and releases f in the containment and TSC/ EOF based on TID-14844 (Ref. 1) releases and prespecified flow ratec. filter efficiencies, - } g halogen non-removal factors, and meteorological data. The j program computes individual doses within the TSC/ EOF. I p C1.0 RELEASE MODEL J l i j f The activity release pathway model is shown in Figure C-1.

9 Four activity nodes are represented:

two primary containment 4 volumes (sprayed and unsprayed), the secondary containment volume, and the TSC/ EOF. The equations for nodal activities, containment release and integrated TSC/ EOF activity are deriv- , '( ed from first order activity balances in the following para-graphs. The definitions of all variables used are presented t Il in Section C3.0. U C1.1 Primary Activity

n..

': Ui The primary containment activity is the sum of the activity'in ( the sprayed and unsprayed regions. l A, ^1 + ^2 m v2 dA ~A A ~AA ~A# ~AA A1+ A I de sp 1 1 1 r1 p l ~ 2 n -1A ~AA -A A = 1 2 r 2 p 2 ^2 + ^1 (3) ~ C-1 NUS CCAPCAATICN 3 4 4 g_s .-2 ":L ' ~

' ~ : :' "' ~P ' ~ ~ M' ~~' V * ' '~~

--.n- ~ .=.a-.._ .....-...-..:. a. -.

l..

il O Ru The simultaneous solutions of Equations 2 and S when combined ) with Equation 1 gives the primary containment activity as t n 4 l

j

(.; ,i {, i A "C e 2-C e'* 1 * (4) p 2 4 r l 't

i A

I 10 1 ~ *1 +A20 ( ~ "I) l'i C (5) = 2 ()

• 2 ~ "I

p A

I ~"2)#A20 2 ~ *2 I I 10 1 (6) C = I i m2 ~ "I (A

• A

') I7) m m p2 s 1 2 i (A 1+A 2 + A + E )2 + Il V V 1 2 l i' 7., .g S 7+fA l -4(y A +AA) g 2 1 r^g b Ag+A*A

• A (8)

A- = 7 r p sp ! !I A +A+A (9) A = 2 1 r p j;:

• J

~m2t ~mit -C e (10) C e r A = 1 4 3 i e t 'I C-2 lj m, L5 j J,I NUS CCAPCAATICN .,.,..7,7..,,... 3-- vn-..

a _. m. .___ u _ _... __-.._. 4 ,') s u n .J I! + E e1 A10 ( A - m1 V V 20 T A O 1 r. 3 2 -i '1 C (11) = L-4 m 2 ~ "I A10 ( 1 ~ *2

• V V

20 A ~ II I C l' r. 3 m -m t 2 1 ) (C - C ) e "2* - (C -C ( 'I, A ~ = 2 2 4 g 3 i' l I i. l ./ Note that the above solution for A degenerates to a one-vol-P l ume problem if A,p =0. C1.2 Secondary Activity t f1 l The rate of change of secondary containment activity is the c fraction of the primary activity that goes to the secondary d containmeht less the removal by decay, cleanup, and leakage (or exhaust) to the environment. 8 dA I a s f AA -AA -AA -AA (14) = dt s 1 p 3 s r s s s d'$ g I f AA -AA (15) lf = a 1 p 4 s A3 + A +A (16) A = 4 e s i i* ll ') ~ f AC f kC e "I + C' e' 4 (17) 8 ~ 8 ~ A = e 2 s A -m lj .i 4 2 4 ~ *1 l{ '. * } p I f A C f A C 7 g A C 5 so A -m 2 4 ~*1 C-3 [I NUS CCAACAATCN l_. o ,r. n --.s -~

~ -.... . -.: w ...s.-... d d p l~.i I l C1.3 Containment Activity Release Rate ll C i !, (I td i' The containment activity releace rate has two components: the {-] secondary containment release after filtration, and the frac-r-1; U tion of the primary containment leakage that bypasses the secondary containmerit. g l(j R, F A A, + (1. - f,) A A (19) = 3 l p i .!O 1 C C e "2

• A - m e"* 1*

(20) I ~ FAf X R = r 3 s 1 A -m j {J .4 2 4 1

i l

+ FA C

• At

+ j.. 3 5 1 2* - C e "l* ~ ~ (1 - f,) At 2 g C e i R-C6* 7* 8* II = ~ r IlJ FA f ,.a 3 s +1-f A C (22) a C = 6 A -m 8 1 2 ll r 4 2 14 l 14 rT 11 yx g C +1-f A C (23) = H 7 A -m s 1 2 4 1I u C (24) 8 3 5

u.,

.4 d f* ll ll C-4 .a j l* s., l NUS OCAPCAAECN l} ] u . -.. n r. y. .,.....;m...,.....y..

.n ~.. ... =. n 0; u 1 O C1.4 Integrated Release from Containment ] The integrated release from the containment is obtained by 'l integrating the release rate, Equation 21, over the time it period of interest. 1 l "{= i R dt (25) R = r D D -4) (26) (1-e "2 ) (1-e "l ) ~ R = + (1-e ] "2 "1 A 4 LI .s C1.5 TSC/ EOF Activity 3, (2 i !'l] The rate of change of activity in the TSC/ EOF is the differ-

j pi ence between the rate at which activity is drawn in from the h

ou tside air and the rate at which it is removed by decay, d 1 cleanup, and leakage (or exhaust). g, e A -1A F2 "cc I -AA 4 g e r r e y c ec (27) =

l (28)

C R-AA l jp g = g 7 l4 j dt x+ cc +x x = t'l 7 c V l, cc 1: (30) WQ)c F q C = {:"J 9 2 cc

• l; i.)

8* C C 9 6* g 7* 9 dA_c = ~ ,l - t. dt-f',ai (31) -A.A i c 'i D 9 6 o 7 o 8 ,- A t A ,-m2t, ,-qt 4 ~~ = c A -m -m x7 x4 ll 7 2 7 , ll- - 7* u +C e = 10 b}l'? m C-5 ( r.' NUS CORPCAADON l 'l a u

.._..-....._a r J I \\ pI C CC f C'C 9y 9g + A - "1 C10 = Aco A-m 7 2 7 7~4 (33) r. C1.6 Integrated Activity in TSC/ EOF r, The integrated activity in the TSC/ EOF is obtained by inte-grating Equation 32 over the time period of interest. l ..} =[Adt R CC CCg7

j g6 (1 ~ *" 2 ) ~

g (h - m) (1 ~ '" I ~ 1 c (17 - m )"2 m 2 C C C !j i 9 g _ x y (1 -

• A t) +

( i,,- A t) ~4 10 7 + x (A 4 7 m G I h) u Implicit in the above derivations is the assumption of con-(~ stant coefficients. In the actual transient simulation, l solutions are broken into a sequence of discrete time inter- ,3 l q vals over which the input parameters that make up the co- ! b2 efficients are prospecified constants. The input parameters consist of flow

rates, X/Qs, decay and iodine removal n

fb constants, provided as stepwise constant functions of time. \\i I.I Initial secondary containment and TSC/ EOF activity inven-tories are assumed to be zero. Initial primary activity may be based on the analysis of TID-14844 (Ref. 1) using the jU fractional iodine release assumptions of Regulatory Guide 1.3 (Ref. 2) or 1.4 (Re f. 3 ). The source term eq'uation is 7 d -AT 3 r A = 8.65 x 10 p 7 gg ( 1_e ) (curies) (36) p li t2 lP ll 1-C-6 l! I ' NUS COAPCAATION il1 a i! ~

~ ' - ~ ' . ~ Elif I .i C2.0 DOSE MODEL l f,f jl o At the end of each tin interval, TSC/ EOF individual thyroid i r (~ and whole body doses are determined using the containment release rate, integrated TCC/ EOF activity, and input values of l X/Q at the TSC/ EOF intake. 1 o Thyroid inhalation dose in the TSC/ EOF is given by the follow-1 m lij ing equation: .i I! Sl i L)

I

.~/ j DT= D (rem) (37) T .1 i i m. j k.h. BR R DCF g g i m Y ,i ec i m where p. l f '] breathing rate BR = p d 3.47 x 10-4 3m /sec (Ref. 4) = >4 m 'l { Beta dose in the TSC/ EOF is given by: 'l ]{ J D, D, (rem) (38) I = i i ? ti H 0.23 R E W = c d i i (39) V i cc p i) where average beta energy (MeV/ dis) t E = j-(See Table C-2. ) ] c: 1b C-7 NUS CCAPCAATION

.~. 3 J l C]T

1 i

i r_ Gama dose in the TSC/ EOF is given by I i " '4 =[D (rem) D p; y y (40) i I }j 'i,j l'* 1 1 b -* r cc g g,j j gi) C 7 ~ i c-Gama energies and fractions are presented in Table C-1. .l Absorption coefficients divided by the density of air are l~ listed in Table C-2. l) 1 Ea n

u i

4 ,o i 'O ,i [1 i ,,,1 4 4 e 4 i G{J f*} Ilb i1 Pi jt f c* 5 C-8 b i NUS COAPCAATION i ; I! i>

~ l u c. c.) I'

i. ' D L}

C3.0 NOMENCLATURE L o (j (j A = Primary containment activity p [ Ay = Activity in sprayed volume u j A2 = Activity in unsprayed volume Primary containment leak rate A e y Radiological decay constant (Sec~1) (See i rj A = r Table C-1) Claanup rate in primary containment A = q p j lj f = Fraction of activity released to sprayed y volume g

}}

f = Fra tion of activity released to unsprayed 2 volume /. V = Sprayed volume y ] V = Unsprayed volume 2 '4 3 A = Sec ndary leak rate .J A,p = Spray removal rate 2jq f = Fraction of primary leakage which enters secondary s !j L> containment F = Filter non-removal factor for secondary building i e. j, ( exhaust system F = Filter non-removal factor for TSC/ EOF (center) 2 { intake system j (X/Q) c = Atmospheric dispersion to TSC/ EOF ] q = TSC/ EOF intake flow cc ] V = Combined TSC/ EOF volume cc hq Byg = Average gamma energy (MeV/ dis) See Table C-2) LJ Egg = Average beta energy (MeV/ dis) See Table C-2) li R = Integrated release f rom both containments (Ci) l' P 3 Lj V = TSC or EOF free volume (m ) gg E = Energy of jth gamma of ith isotopes (MeV/y) yg 3 (See Table C-3) (3 f i,j = Fraction of jth gamma of ith isotope (Y/ dis) h,I ll1 u q. ~ C-9 A l-h NUS COAPCAATION l} " .=

>._m.._ . P

l

! T ~. l: Ma3 = Energy absorption coefficient for air (m-1) (See Table C-4) = Total absorption coefficient for air (m-1) 3 (See Table C-4) Y

i r

= Radius of hemisphere with same volume as TSC or EOF (m) U; 7 A, = Cleanup rate in secondary containment i! A = Cleanup rata in TSC/ EOF c-g 3 b V = Combined TSC/ EOF free volume (m ) gg !j ', R,i = Integrated TSC/W acWty (Ci-sec) !l O If f'] DCF = Dose conversion factor (rem / curie) (See Table C-2) g U ( P = Base loaded core power (Mwt) o = Fission yield (percent) (See Table C-1) (j m yg lf d T = 1000 days (assumed) o f = Fraction of core inventory available for release h = 0.25 (for iodines) (Ref. 2) .s = 1.0 (for noble gases) f = 0.91 (for elemental iodine) (Ref. 2) g it - = 0.05 (for particulate iodine) II = 0.04 (for organic iodine) l f.3 1 = 1.0 (for noble gases) ms il Q = Mixing flow rate between sprayed and unsprayed

• ]

volumes r .U ] C-10

t. a

'I ! *j p l' 1 t:1 ! [ ~, NUS CCAPCAATION t : - v .7 n --

~.. .. = ) ',J

j

'l [~ l ~' C4.0 CALCULATION OF DOSE DUE TO DIRECT RADIATION FROM THE BRUNSWICK REACTOR BUILDING The KAP-VI Code (Ref. 10) is used to compute the integrated j h) dose to different points within the TSC/ EOF building from di-rect radiation from the reactor building af ter a postulated f. loss-of-coolant accident. KAP-VI is a derivative of the QAD (Ref.5) series of point-kernel computer programs designed for C estimating the effects of gamma rays and neutrons that origin-lj 1 ate in a volume-distributed source. y i n 1 \\ {1 KAP-VI is a point-kernel code designed to calculate the radia- ] l ,q tion level at detector points located within or outside a com-q 'j plex radiation source geometry describable by a combination of l quadratic surfaces. The code evaluates the material thick-j,! nesses intercepted along the line-of-sight from the source point to the detector point. These material thicknesses (or ~# j ' rj path lengths) then are employed in attenuation functions to calculate the flux, dose rate, or heating rate at the de-f tector. The attenuation function for gamma rays employs ex-i ponential attenuation with a buildup factor. Three optional neutron attenuation functions are included: (1) a modified Albert-Welton function for calculating fast neutron dose rate using removal cross sections; (2) a bivariant polynomial expression for computing neutron spectra using infinite media ,d moments data; and (3) a monovariant polynomial for computing

g; neutron spectra using infinite media moments data.

1) The code also handles either cylindrical, spherical, slab, ' l 3 g disc, line, or point sources. Different source distributions may be employed for neutrons and gamma rays. A variety of [a options are available for describing the source distribution. I Q a tj C-ll .i f e, {9 NUS CO APOAATION w

)

... ~.-. n a ni! The source distributions are assumed separable along the g !j geometric axes. An option is provided to describe azimuthal source density variation by specifying input data for discrete point sources. U] Specific desirable features which have been incorporated in the KAP-VI code are: g I) (1) Input data preparation has been simplified to allow minimum input for running " stacked" cases. 1 1J t (2) The code uses the " point-in-region" concept to cal-h culate the boundary surface-zone relationship ] (ambiguity index) which is required as input in ] '7 other point-kernel codes. Ti (j (3) A routine is included in the code to calculate gamma nC ray mass absorption coefficients for up to twenty elements as a function of input gamma ray energy v, i from either internal calculations or from magnetic o j tape data and internal calculations. 1[ I (4) A routine is included in the program to calculate the cubic polynomial coefficients for buildup f actors as a function of input gamma ray energy from a library of bivariant polynomial data.

U l

(5) A routine is included which will interpolate a i! closely-spaced source distribution (obtained from a ,i.i discrete ordinate transport source calculation) to a source mesh description more economic and amenable ) to point-kernel calculations. 1 ', s i 1.". lj C-12 r s [1 NUS COAPCAATION / p,,,_ _,,,,___.7

Z ..Z--. 21.._ _ --. ]R 11N O (d (6) A routine is included which calculates and normal-izes point source strengths for a variety of source geometries and functional variations of source distributions. bD (7) Input data are checked for consistency to eliminate many erroneous calculations that can occur if input 2 data for a problem are incomplete. (8) The program has the capability to calculate fluxes and/or other radiation responses such as heating ( rates at multiple detector points for each source b region. [i (9) The program has no set limit on the number of source regions which can be run in a single problem. This g b feature is handled as a set of stacked source region problems. The program computes the summation at f'j each detector point of the rieutron and/or photon c radiation from each source region O .I ! t 4 (10) The program allows the user to input separate source distributions for neutrons and gamma rays within the p 63 same source region. t l a I Q (11) The program eliminates unnecessary response func-tion computations by accumulating flux data as a I? function of detector point and energy group during l the calculation for each source region. Calcula- { tions for up to ten response functions are performed I only at the completion of each source region calcu-lation and/or at the completion of source region g li.. U problems. ~ j.. C-13 cJ 'l _1 NUS COAPORATION y3

=

= x = -

- n

_...._ _..__. _ _..._ m.. 1 m c} L ~ g (12) An option is included for calculating the flux at a ld detector located within a gamma ray source region. This option circumvents the numerical difficulties introduced by the " inverse square law" when a source point is too close to the detector. The radiation source streng th in the secondary containment ~ (reactor building) is obtained from the activity in the 1 secondary containment using the equations derived in Section n C1.2. For conservatism, the effects of leakage from the b secondary containment to the atmosphere are ignored. The con-version f rom isotopic inventory to source strength is based on t the radioactive decay schemes of Lederer et. al. (Ref. 7) ,1 Lj Integrated source strengths are obtained by trapezoided in- ]} tegration of the calculated time dependent source strength. j' l 1a i f1 l i ( a i i J k u r) Il lJ lb 5 1* :., l l tl C-14 J i. NUS COAPCAATION

~l. u-i u Xa ]Ed C5.0 CALCULATION OF DOSE DUE TO DIRECT RADIATION FROM r, l! l') PIPING SOURCES t1 ll hl r* (! The NUS developed CYLDOSE computer code (Ref. 11) is used to f} calculate the dose due to radioactive fluids contained in the i[ pipes in the plant. CYLDOSE calculates the linear attenu-I! ation, scatter buildup, and resulting tissue dose rate from a

'g cylindrical gamma radiation source. Multiple source materials i !J and shield materials may be specified.

Dose points may be selected anywhere along the side of the source, or at its end e on the axis or outside the outer radius. A line source h approximation is used for dose points at the side of the 'l i 2 lt i} source and at the end outside the outer radius whereas a f~ truncated cone source approximation is used for dose points on l '(', the axis at the end of the source. For convenience of calcu- ,1 lation, the gamma energy emitted by the source (s) is divided p into groups and each group is designated by a number and an 1!M average energy for that group. calculations may be done con-sidering one or a combination of these groups. Source ,i j 'j strengths associated with these energy groups may be read into the code as data or calculated by the code. The code will, at (;.f the user 's option, increment the initial thickness value of ,lI " the last shield material by a specified amount until a pre-i [ specific dose rate limit is reached. I,

.j dO 3

m it (b u b i; i M 'i 'll C-15 j-a 6 5

• 4

.{ NUS CCAPOAATION m I -. _,,.--m--- ,w- .,_m,:. ~

2,-.- .....~~..n . +. j "A J -t LJ Ilr; a C6.0 REFERFNCES Il ~j ~ l.; 1. DiNunno,J. J., et al., " Calculation of Distance Factors C for Power and Test Reactor Sites," TID-14844 (1962). l1 b 't 2. U. S. Atomic Energy Commission, Regulatory Guide 1.3, ] j., " Assumptions Used for Evaluating the Potential Radio-1 logical Consequences of a Loss of Coolant Accident for If ] [] Boiling Water Reactors," Rev. 1, Directorate of Regu-i latory Standards (1973). 1 c. H 1 3. U. S. Atomic Energy Commission, Regulatory Guide 1.4, " Assumptions Used for Evaluating the Potential Radio-j >j logical Consequences of a Loss-of-Coolant Accident for !l .4 Pressurized Water Reactors," Rev. 2, Directorate of ll !j Regulatory Standards (1974). r-- ((! d 4. International Commission on Radiological Protection, ~ L1 Report of Committee II on Permissible Dose for Internal j {' Radiation, Pergamon Press. (1959). J l{ q S. CA-3573, QAD-Point-Kernel General Purpose Shielding lI L3 Codes, Oak Ridge National Laboratory. 4 f.9

. ) Q 6.

Meak, M. E. and Rider, B. F., " Summary of Fission Pro-239 241 I duct Yields for U 5, Pu and Pu at Thermal, {} Fission Spectrum and 14 MeV Neutron Energies," APED-5398 (1968). d 7. Lederer, C. M.,et al., Table of Isotopes, 6th

Edition, y

j John Wiley and Sons, New York (1968). R

3 a

f4. C-16 ii l' '*c l' y1 l

• ]

i n NL'S CORPORATION , 9j 4 j e nv y== ~g - e+ - ep, e g y-qg a ~w-w e - -. ee. e e'ese+--%---== , = w_ , e e y =y ei - ,-=ee- = e e e-... e

, v.... -.... . ~. -.. - - - -. -. - - - ~ ~ ~ - j p )b j g 8. " Final Environmental Statement Concerning Proposed Rule Jl Making Action: Numerical Guides for Design Objectives l and Limiting Conditions for Operation to Meet the Cri- -r ,.j ' i, 21 terion ' As Low as Practicable' for Radioactive Material s in Light-Water-Cooled Nuclear Power Reactor Effluent," .j WASH-1258, Volume 2, Directorate of Regulatory Stand-ards, U.S.A.E.C. (July 1973). b G] .a 9.

Hubbell, J.

H., " Photon Cross Sections, Attenuation Co-l r. efficients, and Energy Absorption Coefficients from 10 j ( kev to 100 GeV," NSRDS-NBS 29 (1969). i !f 10.

Disney, R.

K. and

Ziegler, S.

L., " Nuclear Rocket i ~ Shielding Methods Modification, Updating, and Input Data 1 Preparation, Volume 6, Point-Kernel Methods," WANL-PR-l'# (LL)-034 (August 1970). 11. Arcieri, W. C. and Nathan, S. J., "CLYDOSE Program Des-ct cription and User's Manual," NUS-3780 (April 1981). !! !-l li

2 lj 12.

Nathan, S. J., " AXIDENT, A Digital Computer Dose Calcula-tion Model," NUS-1954, Revision 2 (August 1979). .;,.i O .t ..J 9

1.,

7

a j

r. ,1 i T .s j C-17 l\\ ~ l: } l. ') l! NUS COAPCAATION ,e ---.- - ~ - - - ~ .a- -,:,. n v w m. m.

r .w -

---

w..........-..-.

.a.~... ..-~....-: LJ i TABLE C-1 ' rt NUCLIDE DECAY CONSTANTS AND FISSION YIELDS (Ref.6) t i m. l 'l Decay Congtant Fission Yield Nuclide (sec ) (percent) i ! If3f 9.97 (-7)* 2.91 m I 8.37 (-5) 4.33 l33 ~j I 9.17 (-6 ) 6.69 i y34 j I 2.22 (-4) 7.8 135 + I 2.87 (-5) 6.2 {_,}8 83m Kr 1.03 (-4 ) 0.52 85m Kr 4.38 (-5) 1.3 85 Kr 2.04 (-9) 0.27 [j 0 Kr 1.52 (-4) 2.5

I Kr 6.88 (-5) 3.56 il

[ l31m Xe 6.79 (-7) 0.022 Xef33m 3.55 (-6 ) 0.17 l Xe 1.52 (-6) 6.69 l Xef35m 7.40 (-4) 1.8 Xe 2.11 (-5) 6.3 l38 Xe 6.60 (-4) 5.9 itl fi l f .. i Read as 9.97 x 10-7 a

F lL q

L, l'I r I. I L i.! C-18 m. I, I ll l^ ' :1 I! l I ',* NUS COAPORATION l . _ _. -. _.,.......,,.m...._. ~.. _ - -

. = ..'..:.... ~.. ...c.-w-.....- ..w.. f'- a t.. 'lI~ 1 .4 TABLE C-2 F["f AVERAGE BETA AND GAMMA ENERGIES AN IODINE INHALATION DOSE CONVERSION FACTORS d Nuclide (MeV dis) (Ref.7) (;MeV is) (Ref. 7) DCF (rem / curie) (Ref. 8) i J l3 l I 0.371 0.197 1.48 (+6) I,3 2.40 0.448 5.35 (.4) 1 I 0.477 0.423 4.00 (+5) 134 (7 I 1.939 0.455 2.50 (+ 4 ) l35 I 1.779 0.308 1.24 (+5) 83m r.r 0.005 0.034 f. Kr 0.156 0.233 85 Kr 0.0021 0.223 87 Kr 1.375 1.050 00 Kr 1.743 0.341 l3 Xe 0.022 0.135 3 Xe 0.033 0.155 q l33 Xe 0.030 0.146 Xef35m 0.422 0.097 l,' j Xe 0.246 0.322 l38 j Xe 2.870 0.800 ? t F-. I J .M l f 10 Il ll U ij 8 C-19

!l l.

ll m '. i 4 5 NUS CORPOAATION i ,i _..-.m.._.._..

i C-_ ,~'^ ] 7 IJ I'. C O L C EU Ud r-1 -m u-s TABLE C-3 g ISOTOPIC GAMMA ENERGIES AND DECAY FRACTIONS (Ref.6) I-838 1 882 I-833 1 134 1 435 BE=l3tM

• E.133M st.133

{ .else 5.60E-82.3472 2.esE=e3 5340 9.ast-48 .336e 5.eet-e2 22e4 8.sef=st 4eSe,6.eeE=e2 8297 8.etE=el.e)se 3.42E=el 38G2 2.5ef=62.2610 2.etE-92 75ae 2.sef=82.less 1.selae2,2004 3.seE-42 4340 5.9st-el.e354 3.26E=e2 0353 4.6el-42 6 .lII2 2.506-45 2058 5.46E-81 8640 7.eef=42 190s 7.sel-et 0815 3.2eL-42.1604 2.38(.42 232e 4.84E.02. ele 6 6. eel-el 26&3 5.988'-82 504e 1 00E-02 1.0324 1. set-82.eles 6.44L-43 414e e.2eE-el 4 I" 9258 2.5ef-e2 .5840 2.801-82 8.2448 2.stE=e2 4144 3.00E-82 5269 8.ett-el 4488 3.781-03 3 16t5 7.911-01 .5218 8.601-88 8.356e 2.00E.82 51st 9.44E-43 5065 6.241-02 3647 6.6el-48 3254 2. net-06 i* 5934 3.n8L-sl 6246 4.44t-42 5444 4.esE=42 7eFI 5. vet-43 631e 6. eel =82 6130 3.96tast 6800 2.etE-41.,4169 5.440-42 3 ell 5.let.85 I 722e 1.44E 84 *.4547 4.88f-82 4584 1.44E-82 .412e 8.80E-02 3848 2.3eg.ae .6518 4.06E-42 35e8 8.40E-82 1. 4 39 7 9.488.-82 6614 9.141-ek .7144 6.44E-42 1.10 8 7 8 388-82 .t647 6.488-e2 .elas 1.54E-el 1.1243 3.let-02 6785 6.488-82 860s 4.seE=e2 1.1316 8.75L-el .121e 3.2CL-42 8944 7.eef-el 1.1698 7.941 03 .7298 3.2eE-42

• 1688 2.001-82 8.2684 2.58L-el 7129 8.3&L-el 1.e44e 5.8eE.42 1.4575 7.let-s2 n

.5547 1.%48-81 1.0744 8.uSE-48 8.5029 1.2eL-82 e a 1.Iles 2. Gel-82 8.8540 8.2eE=el I.5659 1.48L-42 FJ 1.lett 6.861-02 1.2400 1.e4E-42 8.6705 9.53E-02 C) 3 7234 7.6tf=43 8.3448 2.e4E-82 B.7678 3.000-42 \\ 8 2444 6.005 02 1.4600 e.4ef=82 8.7989 7.6ef=02 3.1634 2.64L-42 1.4904 8.00E-42 8.elle 6.40L=45 y:

8. lies 4.e4L e2 8.6244 5. eel =62 2.4467 e.3eL*=e3 i

3. sees 3. Set-e2 8.79e4 5. tee =42 2.2567 6.3ef=83 8 72ee 3.eet-os 2.0s7,t.est-33

't. flee b.d6L-e3 1,9308 8.341-s2 l.9418 1.3Ste82 2.285e 3.est=03 i 5 2.860s 2.eef-41 2.7244 2.eGE-03 I-i 2.1930 2.461-G3 { a 2.550s 5.041-40 3 2.6048 2.84E=84 i e i. i 1 lo -j t i. 4 % e-

- ~ - ~ - - " - ~ ~~ '} e

i J

LJ Q

t. j cc, a

c2 r:J J P O O O O' m_ 1 a i j t - 4 : !1 TABLE C-3 (continued) t ISOTOPIC GAMMA ENERGIES AND DECAY PRACTIONS (Ref. 6) 3 I Isofortse GAMMA (N[aGIgG se.D F8&CTIONS 4 st-835m sL-il5 st 't se un-s34 an-s5n us,85 au-01 at-ce . sees e.48, se. tite e. m -82 4300 3.04 -02 . set.4.seg482 82.0036 8.5u-se.5ite 8. m.4, 44,8 5.941-88 5 668..,u 02 } .else 3.151-e3..35e5 2.tef-83 45%e 1. Set-et .sevs 6.48 est2e 5.2eL-02 6743 2.581 82.8968 3. elf el 5210's.2eE-et .1999 2.00E-04 2434 3.641-82 412e 8.64t el.4495 7. tee el 436e 4. set-43 3626 3.e;f-32 249e 9.16E-08 2594 3.1st-et 3454 1.3SE el ..sest 8.let-82. u u..cor-53 j 39:a 6 .ne6 2. m -si . m o 1.40c-42 .I,5s t.=8L-42 .m-n .3 fit 8.388-84 4028 2.48[-82 1.3344,1.50t=83 .ela? 3.3lt-en 4342 3.let-43 4344 2.3eE-et 3.38e4 5.5e1-03 .a62 5.est-e3 5fil 5. set-45 1.7788 2.estast i 1.1=l e 2.est -02 9967 8 4;t-e2 i .606& 2.eCL-82 2.stes 1.&etact 64t6 3.28L-ce 2.st24 2.6eL-82 8.le87 8.63I.G2 .739 4.68E=e4 2.556s 9.50L-42 8.tell 9.ctt-43 4126 5.eGE=e8 2.5598 5.tet-42 t.2h93 8. t ci-H 1.4634 3.4GE-45 2.8882 4. set-83 8.5:15 1.5CE-42 i 7 3.3 698 6.44E-4 3 f.52 8 3 8.t.t-4 8 to 2.e295 s.5:i 32 H 2.e1%$ 4.821-62 \\ 2.1959 8 5t[-88 ) 2.2316 3.63t-42 2.3b24 2.eCi-43 i 2.3924 3.42E-01 t fl i l l d 5o l4 4 ..w-.

- ~. l 4 m ') l~2 TABLE C-4 ABSORPTION COEFFICIENTS FOR AIR (Ref.9) l 7. l t E (a) (b) 4/p Ma/p e ,i MeV (cm jg,) gg,2fg,) 2 I m i 0.01 4.99 4.61 J 0.015 1.55 1.27 0.02 0.752 0.511 ['1 0.03 0.349 0.148 ] 0.04 0.248 0.0669 0.05 0.208 0.0406 0.06 0.188 0.0305 ty 0.08 0.167 0.0243 3 l 7 0.1 0.154 0.0234

l 0.15 0.136 0.0250 0.2 0.123 0.0268

]J 0.3 0.107 0.0288

j 0.4 0.0954 0.0295 0.5 0.0870 0.0297

j p.

g'l 0.6 0.0805 0.0290 j O.8 0.0707 0.0289 1.0 0.0636 0.0280 l 1.5 0.0518 0.0257 2.0 0.0445 0.0238 3.0 0.358 0.0212 "1 4.0 0.0308 0.0194 - a ,i l} ]

• From Table 3.-27, NSRDS-NBS 29.

b From Table 1.-7, NSRDS-NBS 29. i l 1, fi .. U '. : o I C-22 ), o NUS CORPORATION "'*" " '~ ~*"s**- ._oe, y

~.- ~ TR _ m L

u. A-T _. _-

- ? Ta'J.a. 1 1 _ i ~~ ~ ZZ..J. ~~ c.,. r '. -_J --J' r~n.- k. 3 7.. --) i..b r 1 l 2 w..> s a ._.a t 1 i l f 1 I i j ft. ail a, l ') .~ t p AE LE ASE N reuea 1=~~~~---------*, essoassateentes ? \\ DO&E %s Ol8 PEAS 4001 SEC0000AAV COesTAltenetteT l N I 43 g I A SPRAYED tesTAltE g I flLTER \\ O RE MOVAL 4 4# RATE F N 2 \\ L Ap ~ W t 2 A TSC/ EOF ~ g \\ __lo_sC.A vl.._ (

• A:
• A s.

Upd8PRAVED Age ' L C \\ d g 288 peu gNg REMOVAL ptMov4L MOW natt MATE gg,,,, PRIAAARY COfeTA49ea4E88T As Dolt B s: i e a, a s seTE 00Ussuany FIGURE C-1 ) DOSE MODEL ACTIVITY FLOW SCHEMATIC i h& 8 - --^- - - -. -}}