ML19309A558

Chapter 5 of VA Polytechnic Inst & State Univ Research & Training Reactor PSAR, Radiation Safety.

ML19309A558
Person / Time
Site:	05000124
Issue date:	11/01/1979
From:	VIRGINIA POLYTECHNIC INSTITUTE & STATE UNIV., BLACKSB
To:
References
NUDOCS 8003310432
Download: ML19309A558 (16)

Text

.

5. RADIATION SAFETY 5.1 Shielding Analysis The shielding analysis was performed using a combination of observed data and calculations. Radiation levels with the reactor operating at 100 W ,

(fast neutron, thermal neutron, and gamma) were taken from routine radiation survey logs performed by the staff at those points where there was no auxiliary shielding adjacent to the original (10 W) biological shield. The results are shown in Figures 5.1, 5.2, and 5.3. At points where there was shielding, it was removed for radiation survey purposes and then was reinstalled. I l

The procedure was to take the levels of radiation which escape from the reactor when operated at 100 W and to extrapolate to.the levels which would l

be expected at 500 W. From the expected radiation levels for 500 W , the l dose on the outer side of the modified shield was calculated.

4

'5.1.1 Fast Neutrons Following the method of Lamarsh [5.1], the fast neutron flux which pene-trates a shield is given by R* (5.1) 4f (x) = 4f(o) e where 4 (o) = incident flux, n/cm -see f

E = removal Cross Section = 0.089 cm-1 x = thickness of shield The highest dose rate due to fast neutrons is found on top of the reactor directly above the central stringer of the central graphite reflector. The l

5.1 l 1

8003810 Q R

I f

• j +

i i ,

t___________ a

, I l W , ~s j f \ ',s ,m x

/ r-------

' WEST g \

/ f- ~~~ ~'FY{~1 k f ~l~2

./

1. --h- - - -- r - __ _

STORAGE FUEL g j CHEMISTRY / / PIT

/

LABORATORY g f HOT l l C LL l m l l

}- : _

-  ::: ._g,, / [  %. -

/ g

- c (, / \ / E

/ I 8 CHEM. RABBIT NTUBING g l[ / j '

I -

g -%. NAA l TOP l j j- $

g S q RABBIT CLOSURES l  ! -

N REACTOR ' TUBING \ /

\,

s;- - - -- - -

h= = = == = = , = = = . = = =.

) , l[

• \ h / - ~ - ~

/

L ROOM ELECTRONIC

\ \ NIO 'f- /

/

-- /

6A SHOP N, #

7As \ N /

N5 '

\- ' '

,5 s , f

/2.(5 N e' '

\ l \ N ,/ /

aST -*

\ '  !

\ \ l FUEL

/ IOA \ ,

N _, /

fl j STORAGE PIT ANTE ROOM $ i l /

e ROOM 6 'N. , __ _

y/

o o a o PROCESS PIT o E lolo o 1

i Fig. 5.1 Jarna Isodose Curves (Old Shielding)

i .

j - _ _ _ _ _ . _ . . . _ _

I l__] I______ __- .I

- s _ _ _ _ _ _ _ _ _ _ _ ,

N

[ ,-______

L______p_ , _J- - - - - - \

WEST FUEL f -

! STORAGE CHEMISTRY /

PqT j LABORATORY / __. _10~' ~

/ HOT / \

/ \ \

l h=== [

C "L

/ / $:: 15 N \

,,g- j  :: . s g

]8 CHEM. RABBIT TUBING r

' \

(

\

'j'~% -J I \

\

- I '

qs / NAAf TOP j

========x= = = = = = = = = - _ = = = = . = = = = :::=

= = -! _d= / { {

ROOM ELECTRONIC l 6A SHOP  ! \ - - / __

g 7A -

I \, /

{

> SHIELD /

~ b TANK / i 'e' l s

\ 2.5 %A /

4

\ N' -/ - /

/

EAST L*

' I FUEL l \

/ IOA fl STORAGE)

PITp l

A NTE' ROOM \ l j ROOM \

s %_

e i -

/

_' *W m., ~~~

O O O ~

~~ . _ _ ... -

1

' PROCESS PIT

! o E lol0 o i

5.2 Fast Neutron Isodose Curves (Old Shielding) 4

l I

e .

i s e I~~~l L________._ _j w -

l

' g t

c------ WEST I  ; - _ L______ _ _g ------ FUEL STORAGE CHEMISTRY PIT LABORATORY -

l HOT hr: C LL /6 ' ~~ N

_- === %,

!!/ N -

1. M \

e 8 CHEM. RABBIT TUBING e%+c[* /

/ s~

O.5 \

- ~ D. x / / \ O.25 }

h /lT

' TOP

~

\ [NAA j g

~

/ /,R A B Bl T l CLOSURESi l N

REACTOR \

- = = = =M := :gBlyG\_p

=_= j j

{  ::== ====

ROOM

= = = = = = = = = = = _ _ _ . = = = :: : = = = =

ELECTRONIC

~

\- l --

SHOP / _ _ _

6A 7A - \ /

xg

\ SHIELD /

TANK / , e' qg ,

~ ~ -- ,s -*

N / EAST i '

FUEL

/ IOA fII STORAGE i PIT ANTE ROOM I ROOM 6 I i o o o o PROCESS PIT o E lolo a

' 5.3 Thermal Neutron Isodose Curves (Old Shielding)

measured dose rate is 285 mrem /hr and the proposed shielding addition is 1 ft.

of concrete with additional lead and concrete directly above the central stringer.

Only the 1 ft. of concrete was treated in the calculation.

For 1 ft. of concrete shielding:

Dose Rate = 285 exp (-0.089 x 30.5)

= 18.9 mrem /hr (fast neutron)

The other two places where fast neutron dose rates were predicted to be significant were the side of the top blocks (streaming radiation) and the outer surface of the south beam port plug where the dose levels were 155 and

. 125 mram/hr., respectively. Neglecting absorption in air and noting that the shield thickness for the south beam port is 2 ft., the calculated dose rates with the additional shielding are 10.3 mrem /hr for the top of the core and 0.6 mrem /hr for the south beam port.

5.1.2 Thermal Neutrons Thermal neutron exposure has never been a problem for reactor personnel at VPI&SU. Typical neutron doses received are at the lower limit of detection.

Neutron dose rates predicted for 500 KW are acceptable without the additional shielding and will be even lower with the modified configuration. Nevertheless, thermal neutron penetration calculations were performed for three locations in the reactor room. The method used was to solve the steady-state diffusion equation for an infinite slab with an infinite plane source of both fast and thermal neutrons. It can be shown that the solution of such a problem with known neutron fluxes is:

-I

• R 4th(*} " #th(

~

R *f e"*+ R *f *

(5.2)

D h (K "ZR) th( ~R

( .

5.5

where $g(o) = incident thermal flux D = thermal diffusion length = 0.484 cm th Ia = thermal neutron absorption cross section, and K = Ia/Dth "

• Results of the above thermal neutron calculations are included in Table 5.1.

5.1.3 Gammas Dose rates due to gammas arise from two interactions, those gammas emitted from tN reactor face and those produced in neutron absorption reactions within the shield itself. First the gamma radiation emitted from the reactor surface will be treated. According to Lamarsh [5.1] the dose rate on the outer surface of the shield can be expressed by the relation:

~

x = x, B(px) e "* (5.3) where i, = dose rate incident on the shield B (px) = build-up factor

-1 p = attenuation coefficient, cm The build-up factor, which accounts for radiation scattered in the shield, can be expressed in a number of ways, one of which is the sum of exponentials. This is known as the Taylor form of the build-up factor [5.1], namely, B (px) = A ye "N*"1 + A 2

e" 2 (5.4) where A y , A2 , a 2are functions of the incident gamma energy and Ay+A2 ~ l*

For conservatism all gammas are assumed to have an energy of 5 MeV. and the above constants in concrete are, 5.o

A = 11.17 A = -10.17 7 2 ay = -010206 a = 0.0269 2

For 1 ft. of concrete: B e "* = 0.269 P

For 2 ft. of concrete: B e #* = 0.0539 P

For 4 ft. of concrete: B e "* = 0.00160 P

The highest dose rate due to gammas is found at the thermal column door where a rate of 300 mr/hr is expected at 500 KW. At this location there will be 4 ft. of concrete shielding and the dose rate on the outer surface of the shield will be, x = (300) (0.0016) = 0.4o mr/hr Although this exposure rate on the outer surface of the shield is quite small, due to the large surface area of the thermal column door there is a large amount of scatter from the primary radiation and shielding to the south side of the thermal column. To prevent this radiation from streaming out some additional

, shielding will be needed.

The highest radiation level on top of the shield is predicted to be 50 mr/hr with the shield thickness of 1 ft. Therefore the dose rate on the outer surface will be 13.4 mr/hr.* The south beam port has the next highest dose rate with a reading of 300 mr/hr expected at the plug and a shield thick-ness of 2 ft. to give a dose rate on the outer surface of 16.2 mr/hr.

• As was the case with the fast neutron dose rates on the reactor top, the reading represents locations where there is some radiation streaming. The modi-fied shield is much larger and covers all the cracks in the original shield which allowed radiation streaming.

l l

5.7 l

l l

~

1

• l l

5.1.4 Capture Gammas To calculate the dose rate due to capture gammas produced within the shield, the distribution of thermal neutrons within the shield must be evaluated in order to find the distribution of sources. A method used by Andrews [5.2], when power was increased to 100 KW, yielded results in good 1 l

agreen. ant with actual measurements. According to Andrews the energy flux due I to gammas on the surface of the shield is given by: I

-kx <

4 (x) = 4,(o) - R f e ye e (ux) + Et (ux(v-1)+1u)"l+

Ds (K -I ) 2k "1"

- R* x

+ Ri f( cY e Ey (px) + E1 (px) (u2-1) + Lp "2+ (5.5)

Ds (k -E ) R "2"

-1 where I = thermal neutron capture cross section = 0.0094cm E

y = gamma energy = SMev

-1 v = k/u, v ~ r /u, p = 0.066cm = the gamma capture 2 R cross section.

The two locations where capture gamma radiation will be most prominent are the top of the core and the south beam port. On top of the core the capture gamma dose rate will be 1.9 mr/hr.

A summary of all measured and calculated dose rates is given in Table 5.1.

, 1 I i 5.8 1 1

TABLE 5.1 i

Measured and Calculated Dose Rates l I

100 KW (Old Shield) 500 KW (New Shield)

Gamma Fast Thermal Gamma Fast Thermal Location (mr/hr) (mrem /hr) (mrem /hr) (mr/hr) (mrem /hr) (arem/hr) i Above CSTR 14.0 56.3 1.87 15.3 18.9 3.0  :

South Beam Port 5.0 2.5 0.22 16.4 0.6 0.3 Thermal Column 0.5 1.25 0.1 0.5 0.03 0.03 e

From the above tabulation it is clear that the modified shielding will have little effect on gamma dose rates, but will drastically reduce the fast neutron dose rates.

Af ter the shield modifications were completed, a thorough survey was made of the dose rates due to gamma radiation, fast neutrons and thermal neutrons. Results are shown in Figures 5.4, 5.5 and 5.6 which should be compared with Figures 5.1, 5.2 and 5.3. In qualitative agreement with the above calculations, the dose rates for fast neutrons are significantly less with the modified shield in place while the gamma dose rates are changed very little.

The cumulative dose rates Gaan-reml at 500 KW operation depend on the annual energy generation , E. From Figure 4.1 it is seen that for the period 1985-1990, 180 Mw hrs /yr will be generated compared to 110 Mw hrs /yr for 1979. Thus D (1990) = D (1979) x E (1980) x DR (New Shield)

E (1979) DR (Old Shield) 8 Therefore, D (1990) a D (1979) x x = 0.55 D (1979).

0 In conclusion, the cumulative dose in 1990 is estimated to be 55% of the cumulative dose in 1979.

5.9

^

N / -

. /

D N f -, m -- 1

\

/

M pr n.

g

. . _ _ _  ! I

a

\

/ . - m

{ _

?:. i_ OkN  !

e

7. 7 A \ '

'm

[

N'N 'qx

/ \ '

/

^

;~ -_ s

.- n / \ \

N f, '] \* ,

/ xs ' k:.:g 3 w. ,-

2 i

~ ^

{ ]

T  ?:

~

\ i  ?; '

s t y i i'

'* I,,.* '

Nb

} . :i fl gy; l

- .- - .: y j

?p - ,

/ /

lli B gr'

. 6'. ~sy ii'

• I I

mh '::;;? .

/

/.

%: ~,% -

% y ,

d.2

\( %N ' -

/ * /l i 5 .; f...:nj N

! l'F ?!N9l i

_ 1  %  :: .1.- /

L 75 ^'! 'dpi , j

/ p;

+  ;  ;  ;.,

( -

x

. - ./ / t

\ _%.__ / i

/

s N .

s s e -

"/- ~~- - -

y .___ . _ _ .

Fig. 5.4 Gamma Isodose Curves (New Shielding) 5.10 .

l 1

1

- m - %

"n ,

.s.' - 1... u ,_ c

_m  %

, / a \ \.' A \

' ' ~

k j

I -

g} \ '

/

. ;(. <

.m . > ,i-3-

l- g ,

g. ;g, gpp >

I

- N  %" s ,

l

~~~~=, i

<'! <3- , ,

) I' s#': ' fkQ:l '

( )

s_ l g

g E.

, (4

,/

, I {j,"' d ..

.E-I  ;,s .. . ,- / f Q, W

\ / {  :

~ *

'[

n s s --v7 f ,] . . ,

M G n @l M t%# R M / _ _,# 5 e

Ng ,

I

. . N

, g

)

l Fig. 5.5 Fast Neutron Isodose Curves G

(New Shielding) t ).11 -

i , .,.

- . . . , . . ., . . . s . .....,

)

l b,

p w#ser " Lni -%1 ir A

m m ___ -__ A s h

u j

.g,._ m.._. ._

,q. -4: C 1* Ih .  :?vi'. s r"

'g , ,-l~s gs

\ .,

wh

.-- [ ,

p 'Nfh s.::s; w:

.x j' 5

\

) =,

, h, ' g.[

s^-

g -

i ,

/

j , , s,...-  : a \ >

s.

ag;.

se ' l i

8 N

• I 4 '

, I 1

I 'd' . :':/

$[k

')

( )[

q.;c 5 l' .giMI l l  % \

• i

/. I

(

- \ f .

4 m , -

i s

h g ,

J'l jXe.

.q- < .

;

I,

. . ,e 8-9

'I -

e-N'jgt ,

/ _ .

.5,

i. ~ ;: < j y. , ,

--fa l,] 2., i

/ _, [ .

[',

\

\ '

l ,

.g - - s l l

) - ' .

l l

~

l

~ , .. . - - - - - - -

• ==u. ames' p Fig. 5.6 Thermal Neutron Isodose Curves l (New Shielding) ,

5.12 .

4 4 6 9 O G .$ 8 4 # # $ 4 b

- . . , - - ---,,,a

1 I

l l

5.2 Airborne Emissions j 1

The predominant airborne emission from the reactor is Argon-41. The current technical specifications limit the A-41 emission to 100 pCi/second, not to exceed 315 Ci/ calendar year. During .* recent day of typical operation (7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> at 100 KW), the actual emission was 1.2 Ci or 48 pCi/sec.

In order to operata at 500 KW, provision must be made to reduce the A-41 i

emission. This reduction will be accomplished by the use of an inert gas blan-ket (N ) around the reactor core. A demonstration of such a system is described 2

in Reference 5.3 which showed that A-41 production can be reduced by as much as a factor of 5.

One problem that was encountered in the demonstration of the inert gas system was a pressure transient which caused the coolant to overflow the core tanks and spill out over the graphite moderator-reflector. To avoid this type of incident, the inert gas system will be equipped with a reservoir tank. The pressure in the reservoir tank will be continuously monitored and an automatic pressure-relief valve will be installed.

5.13

l 5.3 Liquid and Solid Effluents Solid radioactive wastes are generated from reactor operations.

l The sources of such wastes include:

(1) Samples from activation analysis l

(2) Ion exchange resins (3) Contaminated tools, clothes, etc.

(4) Primary coolant waste Solid wastes are stored in appropriate shielding until the shorter-lived radionuclides decay. Periodically the wastes are packaged and shipped off site in accordance with the VPI & SU Radiation Safety Manual.

A summary of solid and liquid radioactive wastes from reactor opera-tions which were shipped of f-site is given in Table 5.2. With the increase in operating power from 100 KW to 500 KW, a =edest increase in solid radio-active wastes is anticipated. From the projected reactor operating schedule one can estimate this increase to be from 10 mci /yr in 1979 to 18 mci /yr in 1985.

5.14

Table 5.2

- Waste Disposal from Reactor Operations (Nov. 1975 - Dec. 1979)

A. Solid Waste Isotope Quantity (mci)

Co-60 28.2 Cr-51 0.0001 Th-232 0.0011 Zn-65 12.7 Eu-154 0.0010 Total 40.9022 B. Liquid Waste Co-60 0.002 Th-232 0.023 Total 0.025 I

N i

l i

e l

l l

l. 5.15 i

r

References 5.1 Lamarsh, John R., Introduction to Nuclear Engineeri,ug, Addison-Wesley (1975) 5.2 Andrews, J. Barclay,II, " Shielding Calculations for the VPI UTR-10 Reactor at 10 KW", M.S. Thesis, VPI, Dec. 1964 5.3 Holland, Thomas E., "A Study of Argon-41 Production by the VPI & SU Research Reactor under Various N Blanketed Conditions", M.S. Thesis, VPI & SU, June f972.

4 4

=

5.16 l

, , .- ,