ML20073G390

Forwards Suppl Info to Decommissioning Plan,Per 910409 Request.W/One Oversize Drawing

ML20073G390
Person / Time
Site:	05000054
Issue date:	04/24/1991
From:	Mcgovern J CINTICHEM, INC.
To:	Michaels T NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20073G393	List: ML20073G390
References
NUDOCS 9105030288
Download: ML20073G390 (24)

Text

___ _ _ _ _

v 4 '.

CiNTICHEM, INC.

p.o oox sie TUXEDO, NEW YORK 10D07

[D14) 3b1-2131 April 24, 1991 Document Control Desk U.

S.

Nuclear Regulatory Commission 1 White Flint North 11555 Rockville Pike Rockville, MD 20852

SUBJECT:

USNRC Letter, Docket 50-54, T.

S. Michaels dated April 9, 1991 Thic letter contains the supplemental information to the Cintiche;. Decommissioning Plan that you requested in the above referenced letter.

Very truly yours, i

fc $

sts j J.

c.

McGovern President / Plant Manager JJMcG/bjc cc:

Dr. Paul J.

Merges, Director Bureau of Radiation, D!lSR New York State Department of Environmental Conservation 50 Wolf Road Albany, NY 12233 Dr. Francis J.

Bradley, Principal Radiophysicist Radiological Health Unit New York State Department of Labor One Main Street, Room 813 Brooklyn, NY 11201 Director, Technical Development Programs State of New York Energy Office Agency Building 2, Empire State Plaza Albany, NY 12223 Annette Dorozynski, Supervisor Town of Tuxedo Box 725 Tuxedo Park, NY 10987 Ms. Ava Gartner Berle, Koss and Case 145 Rockefeller Plaza New York, NY 10111 JJM/101.91B ON 9105030288 910424 l

l PDR ADOCK 05000054 P

PDR i

l l

1.'

Tank 1 of the Cintichem Decommissioning and Dirmantling Plan involves the removal of activated and/or contaminated reactor core components.

The major items to be removed included the reactor core support

tower, the core grid plate, the cooling water plenum and flapper valve assembly, the core outlet assembly at the stall operating position, the shim-rod, regulating

rod, and fission chamber drive mechanisms and accessories, the pneumatic rabbit tubes, the thermal column shield assembly, the reactor bridge, and the beam tube liners.

These items were originally to be performed without fuel in the reactor pool.

As a result of the March 8, 1991 revision to the Plan, these items will now be removed with opent fuel in the reactor pool.

In Item 1 of Attachment D

in the March 8,

1993 revision to the Decommissioning and Dismantling Plan indicates that for this Task the presence of the spent fuel will not result in any additional exposure to onsite workers.

It appears to the staff that the potential exists for exposures to increase since many of the activities must be planned around the fact that fuel is now in the fuel pool and that a dam will be inserted between the pool and the stall to perform various activities.

Indicate the basis for your determination that onsite exposures of workers will not be increased as a result of the performance of Task 1 with the fuel onsite.

Cintichem Comments The on site exposure to workers as a result of performing Task #1 with fuel on site would increase if either or both of the following conditions are met:

- Dose rate increase

- Workers are exposed for longer periods of time The basis for our conclusion that the overall dose to workers will not increase is that neither condition will exist as a result of the fuel being stored in the pool.

There are two source terms to consider when characterizing the dose rate and exposure time for Task #1.

They are the spent fuel and the following activated components:

- Reactor core support tower;

- Core grid plate;

- Cooling water plenum and flapper valve assembly;

- Core outlet assembly; Shim rod, regulating rod and fission chamber;

- Drive mechanisms and accessories;

- Pneumatic rabbit tube;

- Thermal column shield assembly;

- Beam tube liners.

I 1

JJM/101.91B

q q

Since the current pool water radioactive contamination is < 5%

E

MPC, there is no effective contribution to worker exposure f rom it.

Cor.t ribution f rom Activated Comraonents Table I shows a comparison of Task 1 activities with and without stored fuel.

The-basis-for the chart is that all of the activities in Task #1 will be performed remotely under water and

)

that the fuel stored in the reactor pool will be under 30 feet of water at all times.

In both cases the initial work of disconnecting the reactor support assembly from the bridge will be done f rom the shelf (Level B) with the water in the stall at 15 feet, (Level B) (see Figure 1).

If the fuel is off site each q

activated piece will be taken to the gamma pit and cut up for packaging.

With fuel on site each activated piece will be placed on the stall shelf (Level B) with the water in the stall at 30 feet (Level A),

and cut, packaged and stored using equipment operated remotely from Lovel A.

The integrated dose f rom the cutting and packaging operations in the gamma pit was expected to be 32 person-rem.

The integrated i

dose f rom the cutting and packaging operation in the stall plus

(

the activated pieces stored on the floor of the -stall, is expected to be less than 0? person-rem, taking into account that there is more water as shie.'. ding in the case of the stall work than in the gamma pit, and as shown in the next section the contribution from the stored fuel is negligible.

. Contribution From the Stored Puel We _know f rom f resh core history that with 10 feet of water from the centerline of the core, the whole body dose rate is 100 mR/hr above the surface of the water.

For f resh fission products we know that 16" of water is one tenth value layer.

If we assume a worst case, that is fresh fuel stored in the pool, we would have 5

times ' one core worth of fuel (176 elements) stored in the bottom of the pool with 24 feet of water shielding to the closest point where people will be working on the shelf (Point 1, Pigure 1).

Under these conditions the calculated radiation dosq to the worker-from the stored fuel source term would be 5 x 10-o mr/hr.

Therefore the stored fuel source term does not contribute to the

-on site exposure to workers.

2.

Justification _ should be provided as_to why the change.in-the Decommissioning and Dismantling Plan to have spent fuel on site does not necessitate the Radiation Protection Program of Section 2 of the October 1990 submittal to be amended.

JJM/101.91B

=.

~

T/SIE 1 T/& #1 - WIM1 NJD WIMKUP UIOluD Pull 1% SIS:

PUIL U! Dim 30 ITl?r OF WMUt AT Id1 TIMES l

Time ikquired i

Itacation l

l w/o fuel I w/ fuel I w/o fuel I w/ fuel i

l i

I I

I I

l i

I I

i i

l l

I

- Disconnect reactor support 1

i i

I, I

st;ructure fran t> ridge i 10 days I care I

stall I

see i

I I

I I

I I

I I

I I

- Cut up activated pieces 1 50 days I came Igame pit I stall l

I i

lapprox, 115 feet l

I i

110 feet oflof water I

I l

l water Ishielding i I

I Ichielding i l

I I

I I

I I

I i

1 1

- Package activated pieces 1 20 days I sanc igamma pit I ctall l

I I

lapprox.

130 feet l

1 1

110 feet oflof water I

I I

Iwater Ichielding I I

I ishielding i I

I I

I I

I I

I I

I I

- Store packagen i

N/A luntil fuel l l

stall l

I lic gone i

I I

I I

I I

I JJM/106.91B

POOL cA,r E 5 TALL

, f.i VA r te.- m...

g g, c - ga - - - -._

//

./,

/</./,,

639'

• )8

{ R [) ((R

/-

$(18' 3'

~ -

, - _ -... ~.

</z.

..n

• L A.c an/,s',i / r.' :

,,//',,

y l

l

/,',,

/.

/,

o

' p,7/

l

/. //

/

4 j

i

,! N h,

l l

, /,,i,

,. [/f.I

,p

////

/ /.

i t

///

.i y/

CANAL %

i r t L:

/.

s i

4 ',

i w,-n

/ PClIO i

,' ;>'a 797' Lb' ;' ;

y,

,-w...,w-B VAtcR

/

,S 32'- 3' m,

.. E.L.F

-...$. $.3, i

~

, -/J,,

.' [ V E '

...o' SH i

c,

/,,

fz

)

~m... ~. _ ~

._ ~s.-

/

jp

,s,sss,wun.y c,.

i

.w,q

f,w.,+/:w.swsye i

q l

vu,sss,w

u...

u

.... ssswe,.

s,w.w.ss,e;j.w,v

y/

I

,w g /,v y w i

//

{

.sWN.NWg.W

.s%WN,'?s'hN*

l

//

1 w), -.J.,

d, I

/

l

.NW.sswy "sp' s

.sy/NNr. 9..

//

t

'//

j

}

A.'. W.Ns % w(?

I S T F. L R-A n

()

av.v,e^sNc v sssw<

t.

o

.sw<s qhv "

/-

i s

u.w...s w X<

//

.ss

+ Ms w.w,t,. / y,w.x NM:--

5

!,7 9, 7.,

,,/.

t-e > -y

swe,

,Quas

-u u.:w.u u u m o

A i.._._

i 30 FT L E V Ei_ A

=

l EVEL B-15 rT y

x D

2

.i p

hM t

y f

I

kii, v-j T

q

/.

's s

g l

y9r ivj y

\\.___./'

i s

y s

s l

c-uuMus s rauct SHELF N, s I

s un go x g x

/ 31 HIGH QTY - 00)

..__._./

\\

l l

l

/

i CDRE

.-f -

i n

S' T AL L AREA l

i

_a. _-

1.

i i

- POINT 1

.~.l. -.- - -

i e

c",

m.__..

i L'~

1 l FLODR EL.

/

/

778' 7.5' l

,/

f __

C, l _ i_..__ _[

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

!.a/. -.. -. _ _..

.J'.y*

. _._._. 1 9, l

i

(

}/8' -

l' ;

I

/

gygggg44g

a if 4

=

are processed until it is determined that they are i.

acceptable for release.

Cintichem also indicated that they are limited by the State of New York's Department of Environmental Conservation to a release rate of 10 mci /yr.

i It appears from this discussion that the latter is the limiting criterion for release of liquid effluents.

Cjntichem should provide its criteria for determining that a batch of liquid effluent is acceptable for release.

Such criteria would probably be found in procedures.

The staff is-attempting to determine how they can be assured that Cintichem has noe underestimated their 11guld effluents and to -- what degree the NRC can be assured that the liquid effluents and their associated doses have not been underestimated.

Therefore, Cintichem should provide the details of the determination of liquid effluents resulting from the decommissioning and dismantling. effort.

Cintichem Response i

As'atated, Cintichem does discharge liquid process effluents only by batch release from hold up tanks.

The attached procedure, ilP-M-08, describes how each tan < is sampled, and how t1e samples are assayed with a High Purity; Germanium detector and multichannel system.

After reviewing the results of these analyses tanks are released as per-the procedure.

Supervision retains all release records for use in compiling the monthly water release report 4

submitted to the NYSOEC that is required by the order on consent,

1. D200059005.

Cintichem will continue to use decontamination and-surveillance procedures-before batches of waste water are released so that the MPC limit per 10 CFR part 20.106 and the NYSDCC 6 NYCRR Part 380.4 and the 10 mci annual limit per the SPDES permit # NY0004464 will not be: exceeded.

(See Attachment e

HP-M-0 8).

4.

We have reviewed the calculation-of the accident X/0 values i

for the Decommissioning and Dismantling Plan and have noted what we believe to be several inconsistencies in the calculations.

. We would request that the Cintichem generation of the accident values of X/0 be reviewed to determine if they were calculated - in an appropriate manner and if they are internally consistent.

The following background information is provided to 11'lustrate these inconsistencies.

In the Cintichem decommissioning and dismantling

plan, App'endix -E calculated the accident dose resulting from reactor dismantling and Appendix F calculated the accident dose resulting from hot cell dismantling.

Some specifics arc given below.

JJM/101.918 v-gy-vs---

wW -grr,

• • e---

y

>----a,,=-sv,,iis4 M#-#w"'*w

"'r14'U""-'Tw'mu'w"fv'--W'aWWw=*

am-'=-w- - -. - - -

• a---

A-

• • TNTT**-*r#

Cintichem Response The Radiation Protection program that is described in Section 2.1 of the October 1990 submittal of the Cintichem Decommissioning Plan is appropriate for decommissioning tasks with spent fuel on site because:

a.

Spent fuel was stored in the same location before reactor

shutdown, b.

Spent fuel in its present storage location does not noticeably contribute to the background' exposure rates in the reactor building.

(See Cintichem response to question 1).

c.

The described program is stronger than that existing during reactor operations because:

1.

Health Physics technician staf f has increased from 6 to 12 people.

2.

Each decommissioning task will be governed by detailed work procedures and will have a

Health Physics technician monitoting the task.

3.

Radiation and contamination curvey schedules have not decreased since the shut down of the reactor.

In effect, they have increased because of the constant survei'1ance by the 3ealth Physics staff during decommissioning tasks.

d.

The proposed changes to the Decommissioning Plan dated March 8, 1991, included adding the requirement to keep area monitors and criticality monitors operational as Engineered Safety Features while fuel is on site.

e.

The Emergency Plan, that includes responses to criticality related incidents, remain in effect throughout the decommissioning project.

3.

In the January 11, 1991 response to the staff's December 21, 1990 RAI Cintichem provided the quantity of liquid effluents to be released during the decommissioning and dismantling period.

Following staff review of this response, the staf; requested, in a February 13, 1991 RAI that Cintichem provide the values for the decontamination factors (DF) for the various radionuclides associated with the evaporator and the ion exchange bed.

The February 19, 1991 Cintichem response indicated that the DF applied depends upon the waste stream and the anticipated contaminant.

As a result of an April 2,

1991 conference call with Cintichem, Cintichem indicated that liquid wastes JJM/101.91B

r bppendix E Stability Class F Receptor:

Child q

i Ground Level release Receptor located at the nearest residence, 457 meters.

Appendix F Stability Class F Receptor:

Teenager Ground Level release in

Receptor located at site boundary, 183 meters.

Response to 12/21/90 RAI, (01/11/91 RAI response)

Stability Class A Receptor:

Teenager Elevated release a-60 meters Receptor located at site boundary, 259 meters.

Response to 2/13/91 RAI, (2/19/91 RAI response)

Stability Class A Elevated Release = 60 meters Receptor located at site boundary, 259 meters.

L Stability Class P-Elevated release = 60 meters Receptor located at site boundary, 259 meters.

JJM/101.91B

i 4

Cintichem comments We have re-written Section 5.2.1 to make the accident scenarios clearer and have recalculated the X/0 values for Appendix E to increase the clarity and apparent internal consistency of these calculations.

Comments are included below for each of the documents specified in your most recent correspondence followed by the revised Appendix E.

5.2.3

SUMMARY

OF RESULTS Accidental Cutting of Activated Reactor Component in the Pool

}ii t hou t Controls __bv Plasma Torch A maximum credible accident for a plasma torch cut into the most highly _ activated reactor core component hypothetically results in a lung; dose to the negligent operator and offsite individuals.

The radioactive material is presumed vaporized within the reactor building and the operator is exposed for a short interval'without

-benefit of local containment filtration or respiratory protection devices.

'The airborne radioactive material is filtered through one HEPA filtration unit before being discharged through an elevated' stack.

-The critical offsite individual is a teen hypothetically located at the site _ boundary on Hogback Mountain at a height equal to the plume centerline.

The potential dose is calculated to be 1.0 x 10-3 R(m ef f ective dose equivalent and represents 0.20 percent of the applicable annual limit of 0.5 Rem.

A teen located at the nearest residential site could potentially receive a dose of 4.6 x 10-6 Rem or 9.2 x 10-4 percent of the 0.5 Rem annual limit.

NOTE: There are no proposed changes to the remainder of Section 5.2.1..

APPENDIX E Calculations are provided for both Class F and h_ conditions.

For offsite individuals which' are located below the stack plume centerline, the class A condition yields the highest dose.-

For an individual-located at the same height as_ the stack-plume centerline, the Class P condition yields the highest X/O and,

-therefore, the highest' dose.

The revision of Appendix E included a_ calculation of plume rise which had been considered but not included in previous calculations submitted to NRC.

lnclusion of the plume height' changed the maximum offaite dose from the Laurel Ridge housing development to a location at the site boundary on Hogback Mountain.

The complex terrain of the Cintichem site has been fully considered in this revised analysis without any simplifying assumptions and a map has been included with the revised Appendix E for increased clarity.

The release is an elevated one, and the revised calculation is clear on this matter.

The offsite JJM/101.91B z.-

-=

~

critical individual is the teen and an expanded explanation of this important parameter has also'been included.

Four locations; offsite are chesen for analysis which are the nearest site boundary, _ the critical site boundary, _and two -locations in the nearest housing development.

APPENDIX F This is a different accident than that analyzed in Appendix E.

A

-55 gallon drum of waste breaks open on the ground and, therefore, this: accident is a t rue _ ground-release as analyzed in _ the Plan.

The proper receptor is a teen and - Class F. stability does _ yield

_the highest X/O value at the site boundary, 183_ meters from the accident and-at the same elevation.

Response-to 12/21/90 RAI,-

(1/11/91 RAI-responce?

The Appendix E. hypot.hetical accident includes consideration of doses at the site boundary and the nearest housing development as mentioned.

- A more

refined, less conservative model which

-considers plume rise, is used in the revised analysis.

As a-consequence, the conclusion - of highest dose shifts f rom - the housing-development to the southern site boundary location _on Hogback Mountain.-

Response to 2/13/9.' RAI, (2/19/91 RAI response)

The general discussion in our response is correct.

However, the

' Appendix E analysis as revised _herein accounts for a significant plume rise (133 - feet) and, theref ore, the former assumption of '

the plume centerline-being approximately equal to the elevation l

of the housing development - is : not valid under closer scrutiny.

- The consideration that a Class A stability may yield a higher _ X/0 t

at the nearest boundary than :that calculated from Class F has L

also been included in the revised Appendix.E.

hese revisions to Appendix E and Section 5.2.1 of the Plan T

should provide a better understanding of the local meteorological L

and topography that govern potential environmental exposures from the credible postulated accident conditions proposed in the plan.

I JJM/101.91B

4 APPENDIX E ESTIMATION OF RADIATION DOSE DUE TO ACCIDENTAL CUTTING OF ACTIVATED REACTOR COMPONENTS

SUMMARY

Calculation of potential dose to an of f site individual located at the site boundary and in the nearest residential community are shown here for a maximum credible accident regarding a plasma torch cut into the most highly activated reactor core component.

Potential dose is also estimated for a grossly negligent operator wh,o uses this tool to cut through the component and thereby vaporizes radioactive material within the reactor building without benefit of local containment filtration or respiratory protection dcvices.

The airborne radioactive material is

- filtered through one HEPA filtration unit before being discharged through an elevated stack.

The attached calculation shows that the critical offsite individual is a teen hypothetically located at 9,e site boundary on Hogback Mountain at a height equal to the ;1ume centerline.

The potential dose is calculated to be 1.0 x 10-3 Rem effective dose equivalent and represents 0.20 percent of the applicable annual limit of 0.5 Rem.

A teen located at the nearest residential. area could potentially receive a dose of 4.6 x 10-6 Pem or 9.2 x 10-4 percent of the 0.5 Rem annual limit.

The adult torch operator could potentially receive an effective dose equivalent of 0.27 Rem which represents.5.5 percent of the applicable annual limit of 5.0 Rem.

Ppr an Offsite Individual Cobalt-60 contributes 99.2% of the dose.

Zinc-65 and Iron-55 are minor contributors although they are included in the analysis.

The-dose f raction was determined by multiplying the " Activation" by the

" Dose Factor" in the table below and normalizing the results.

The Critical Individual for each of these three isotopes is the teenager, and the lung is the organ irradiated in each case.

Dose Critical Factor-3/

Dose Activation Indi-Organ mran-m Isotore (C1/cm) vidual Irradiated PCi-vr Fraction H-3 2.196 x J-9 Teen Total Body 1.27 x 10-3 0.000 18 Child Bone 3.59 x 10-2 0.000 C-14 4.22 x 10 3 x 10-Teen Lung 1.24 x 10-1 0.001 Fe-55 5.28 Co-60 7.26 x 10-2 Teen Lung 8.72 0.992 Ni-59 8.99 x 10-7 Adult Done 9.60 x 10-2 0.000 Ni-63 2.53 x 10-4 Child Bone 8.21 x 10-1 0.000 Zn-65 3.46 x 10-3 Teen Lung 1.24 0.007 Total 1.000 JJM/101.91B

Fo'r an Onsite Individual:

The Critical Individual is an

adult, since*only adults are radiation workers.

The table is shown again below for the adult worker for the three isotopes of concern.

Dose Critical Factor Activation Indi-Organ mran-nr3/

Dose Isotop:

(Ci/qm) vidual Irradiated pCi-yr Fractio _n Fe-55 5.28 x 20-3 Adult Lung 7.21 x 10-2 0.001 Co-60 7.26 x 10-2 Adult Lung 5.97 0.992 Zn-65 3.46 x 10-3 Adult Lung 8.64 x 10-1 0.007 Total 1.000 Cobalt-60 contributes 99.2% of the lung dose, while Zinc-65 and Iron-55 contribute lesser amounts.

GENERAL ASSUMPTION _S_:

This analysis assumes:

o The most activated section (as determined by TLG's activation analysis) of the most activated component (core support tower) is cut and vaporized by plasma torch.

It is planned to cut this section by hydraulic shearing or other mechanical means; o

The 2"

x 2"

x 1/4" angled aluminum is cut completely through; o

The cut is by plasma torch at its maximum width, 0.25";

o The kerf material is 100% vaporized; o

The cut is done under water.

No activity is assumed to be trapped in the water; o

There is no llEPA filtered containment tent around the pool;.

o The llEPA filters in the building e xhaus t-are only 99%

efficient; o

It takes 1 minute to evacuate the building; o

The most conservative wind type is used (Pasquill-Gifford Type F wind speed 1/2 of site average = 1 m/s);

o The activity is dispersed in only one-half of the building.

This results in a higher airborne concentration; o

The worker is not wearing a respirator; JJM/101.91B

8 e

o The

" effective dose equivalent" weighting factor for stochastic effects is 0.12 for the lung (Ref.

ICRP Publication 26, 1977, p.

21).

The airborne activity is assumed to occupy only one-half of the reactor building (to give a higher concentration).

1/2 building 1 volume = 285,000 ft3 = 142,500 ft3 2

Building 1 exhaust rate = 20,000 ft3/ min 142,500 ft3 7.125 minutes = 428 seconds

=

20,000 ft3/ min This is the minimum discharge time that would result in the maximum concentration in the plume.

Next, the grams of aluminum vaporized are calculated.

Angle is 2" x 2" x 1/4" Assume maximum width of plasma torch (3.1 -> 6.4 x 10-3m f rom NUREG/CR-1756 p.N-12) 6.4 x 10-3m = 1/4" Volume of Aluminum Vaporized (2 In + 2 In) (0.25 In) (0.25 In) = 0.25 In3 length width depth 0.25 In3x (2.54cm)3

= 4.1 cm3 Ind Grams of Aluminum Vaporized 4.1 cm3 x 2.7_q = 11.1 q 3

cm Stack Plume Rise Offsite doses are a function of plume rise above the height of tne stack.

The plume rise for the Cintichem llot Lab and Reactor stack is calculated to be 133 feet (41 m) by using the method of Davidson-Bryant1 fL +Y Ah-d

\\5)

T,,

i St ede, D.H. ed., Mo teorology and Atcznic Ene rgy, USAEC (TID-241901,193 0, p.101.

JJM/101.91B

~..

where,

~6h>

= Plume-rise-(133 ft)

Stack-gas efflux velocity (39.4 ft/sec) (based on w

=

measured-flow in stack of 29680 ft3/ min of combined-exhaust from buildings 1 and 21.

ii

=. Average value of the wind component in the direction of the mean horizontal wind vectors (1 m/sec = 3.20 ft/sec)

Inside stack diameter (4 ft) d

=

AT Stack effluent temperature minus ambient air temperature

=

(680F

.540F = 140F = 7.80K)

NOTE: 540F is year round outdoor-average.

T,

Absolute temperature of stack air (680F = 200C

930K)

I Meteorological Methodology The _ basic equation used to estimate dispersion in an airborne plume as it is blown downwind from a stack is the Gaussian plume equation of Pasquill (1961) as modified by Gifford (1961)2 g

.m t z u /

i a < d'.

m<arfp)=

where X-e (e

a,

-o 2na o,u y

.X concentration in air at x meters downwind, y meter 3

=-

-crosswind, and z meters above ground (C1/m3),

0

= uniform emission rate from the stack (Ci/sec),

u

= mean wind speed'(m/sec),

1 0.r

= horizontal dispersion coefficient (m),

=-vertical ~disiersion coefficient (m),

effective stack H:

A h )ght (physical stack height, h,

plus hei

=

the plume rise

,~

y'

= crosswind distance (m),

'z-vertical distance (m).

=

CASE'#1 Potential Offsite Doses 2

01 f ford, F. A., J r. i 961. Use of Rout 1.ne Metoorelogical Observetians f or Estimeting Atmospheric Di enersion. Nucl. Sof. 2(4): 14-15 es quoted in RBIC Computer Code Cot t ectierr-357 AIRD09-EPA Fetimation of Radle tion Doces Caused by Ai rborne Radionuclides in Areas Surrounding Nuct eer Fe cili ties ORNL-5532 UC-41,11. Jur.e 1979 JJM/101.~91B-

Th'e accident assumes vaporized radioactive material released at an elevated point from the reactor and hot lab building common stack.

The letter designations correspond to positions on the attached map entitled TOPOGRAPHY

(#29473-1).

Four offsite individuals are located as f ollows.

1.

Individual at nearest site boundary (designated A) 2.

Individual at critical site boundary (designated B) 3.

Individual at nearest residence (designated C) 4.

Individual at highest elevation in nearest residential community (designated D)

The stack location is designated E.

An olevation of 800 feet above sea level was chosen as a

reference at which z = 0 and H = 0.

All doses are calculated for 0).

The relationships of the centerline of the plume (y

=

relative height are shown in the drawing below.

Loca tion B (1135 f tl Plume Con te rli ne ze 335 ft (1135 f t) H = 335 f t Top of Stock (1002 ft)

Bona of Stack Loca tion C (952 f t)

(950 ft) 150 ft Locotion C (850 ft) zw 50 ft Ef fective Deee of Stock l""

"^

Long Hondow Rood Reference (EIJO ft)

(800 ft) z

• O ft H=0 i

g(-

p s-JJM/101.91B s

Stability Class-F is normally considered the most conservative for evaluation of offsite dose from a Gaussian effluent plume for which the receptor is at the same elevation as the plume centerline.

Location B (Site Boundary at ilogback Mountain) corresponds to this case.

All other receptors are at a significantly lower elevation than plume centerline, and a Class A assumption yielded the higher dose in these cases.

Regardless, both analyses are included (except for Location B where Class P is clearly more conservative).

A table which outlines significant parameters of the calculations is included below.

Me teorologicot Pareme ters f or Of f si to Dose Calculati on X

l0f f si te l Dowrwind I I

l l

l l

lStebilityl i

U 0

lLoca tionl01 stance (mil r( r) 1 pl m) lu ( rn/sec ) I y (m) l 2 (m) l H (m) i Ctese IX/C( Sec/m3 I I

I I

I I

I I

I I

I I

l l

300 l in 1

0.6 l 1

0 l 102 1 102 l F

[3.7 x 104 l l

l l

l I

I l

l l

1 l

1 I

I I

I I

I I

I I

I l

A I

259 1

0.5 l 5.0 1 1

1 0

1 0

l 102 i F

12.9 x 10-93 l I

I I

I I

I I

I I

I I

l A

l 250 l 04 1

40 I i

1 0

l 0

l 102 l A

17.1 x 10-6 l l

l l

l I

I l

l 1

I I

I I

I I

I I

I I

I I

i l

0 1

777 l 27 l

12 l i

1 0

1 48 l 102 i F

19.2 x 10-9 l l

l I

I I

I I

I I

I I

I D

l 777 l 170 1 280 1 1

1 0

1 46 l 102 l A

16.2 x 10-6 l l

l l

l I

I I

I I

I I

I I

l i

I I

I I

I I

I l

C l

457 l 17 1

0.0 l i

1 0

l 15 l 102 l F

12.5 x 10-20]

I I

I I

I I

I I

I I

I I

C 1

457 l 110 i BC 1

i l

O l

15 l 102 l A

11.7 x 10-5 [

1 I

I I

I I

I I

I 1

1 Calculations of Dose to Offsite Individual Offsite Dose is calcult.ted for a hypothetical individual located at B (flogback Mountain at Site Boundary).

The dose would be higher at this offsite location than at others since the calculated X/0 ratio is greatest here.

Although it is extremely unlikely that an individual would be at this spot due to rocky terrain eri general inaccessibility, we have still made this the basis o:

m.

worst case accident scenario.

I JJM/101.91B

$5.

Calculations of__0 for Each Isotope (7. 26 x 10-2Ci/q) (ll. lg) x 0.01 =

Q(Co-60)

=

T2B sec 1.9 x 10-5 Ci/sec released from the stack A similar calculation for Zn-65 and Fe-55 yields:

OfZn-65) = 9.0 x 10-7 Ci/sec Q(Fe-55) = 1.4 x 10-6 Ci/sec Calculation of Dose at the Hogback Mountain Site Boundary

~~~lbocation B)

X for Co-60 is calculated as follows:

Q(Co-60) x X/O = X (Co-60)

Xsco-60) = 1.9 x 10-5 Ci/sec x 3.7 x 10-3 sec/m3 x 1012 pCi/Ci =

7.0 x 104 pCi/m3 Similarly, X(2n-65) = 3.3 x 103 pC1/m3 Xtre-55) = 5.1 x 103 pCi/m3 Dose Contribution from Each Isotope The. lung dose for the teen is calculated for Co-60, Zn-65 and Fe-55.

Dose Factors are ta' en f rom USNRC Regulatory Guide 1.109 and account for breathing rate.

X x Dose Factor x Exposure Time = Dose For Co-60 0 mrom-m3 x-1 yr.

x 428 sec =

7.0 x 104 Egix8.72x10 m

pCi-yr 3.16 x 10/ sec 8.2 mrem or 8.2 x 10-3 Rem For Zn-65 3.3 x 103 E i x 1.24 x 100 mrem-m3 x 1 yr x 428 sec =

m pCi-yr 3.16 x 10/ sec 5.5 x 10-2 mrem or 5.5 x 10-5 Rem JJM/101.91B

~.

For Fe-55 5.1 x 103 pCi x 1.24 x 10-1 mrem-m3 x 1 yr x 428 sec =

m7-pCi-yr 3.16 x 10/ sec 8.6 x 10-3 mrem or 8.6 x 10-6 Rem Total Lung Dose to Teen from all three isotopes = 8.3 x 10-3 Rem.

Since the "Offective Dose Fquivalent" weighting factor is 0.12 for the lung, the estimated equivalent total body dose is 1.0 x 10-3 Rem.

This value is 0.2 percent of the applicable limit of 0.5 Rem.

Note that potential offsite doses are all lower at the other specified locations included in this analysis.

The next highest dose is just 0.5 percent of that calculated for the Hogback Mountain Site Boundary location.

CASE #2 - A Worker in the Reactor Building It is assumea that the air the worker is breathing has not been filtered by a portable HEPA filter unit.

The airborne activity in the building is:

(7.26 x 10-2 Ci/gm) (11.1 qm) (1012_ pci/Ci)

X (Co-6 0)

=

=

(14 2,~500 f t 3) (0.3 04 8 m/f t) J 2.00 x 108 pCi/m3 (3.46 x 10-3 Ci/qm) (ll.1 qm) (1012 oCi/Ci)

X(2n-65)

=

=

(142,500 f t3) (0.3048 m/f t)3 9.52 x 106 pCi/m3 (5.28 x 10-3 Ci/gm) (ll.1 gm) (1012 _pci/Ci)

X (F e-5 5 )

=

=

(142,500 ftJ ) ( 0. 3 0 4 8 m/ f t ) J 1.45 x 106 pCi/m3 The dose to the worker is:

pCi/m ) (5.97 x 100 mrem-m3) (60 sec) 1 yr_

(2.00x108 3

=

pCi-yr 3.16 x 10' sec 2267 mrem = 2.27 Rem lung dose from Co-60 (9.52x106 pCi/m3) (8.64 x 10-1 mrem-m3) (60 sec) 1 yr

=

pCi-yr 3.16 x 107 sec 15.6 mrem = 0.02 Rem lung dose from Zn-65 (1.45x106 pCi/m3) (7.21 x 10-2 mrem-m3) (60 sec) 1 yr_

=

pCi-yr 3.16 x 10' sec 0.2 mrem = 0.0002 Rem lung dose from Fe-55 JJM/101.91B

Total estimated lung dose to the worker is 2.29 Rem.

Since the " effective doce equivalent" weighting factor is 0.12 for the lung, the equivalent total body dose estimate is 0.27 Rem.

2.29 Rem x 0.12 = 0.27 Rem This value represents 5.5 percent of the annual limit of 5.0 Bem.

9 4

JJM/101.918

y y,

y lPREPARIDHAky~kIdef l

/S N ICIIIGID BY 3 wb l H CINPICH124 HIWR11 lHYSJCS PIOCITX1RIE lAPPI(NAL 3 ([ E Q li,p,M n '

l ISSUfDl_'LuiL._,

HAMP!JfC AIO RMlJW1E PIOCFIXIRES JOR lifVD1tO TAff;S._. IPAGE 1 OP 4 IIh&O8 A.

PUldOSE i

To provide inst. ructions on sample collection, preparation, analysis, and htuiding ol data for all holding tanks.

)

i

.B.

DISCUSSION All water to be released f rom the site that has a credible potential to contah; radioactive contamination must be collected and nonitored before release.

Thi:

water must meet criteria developed to ensure compliance with regulatory limits, te protect the public and the environnent and to ensure adherence to the ALAR:

concept.

Limite exist for each batch teleased and a review and averaging progra' exists on a nonthly and annual basis.

C.

PROCEDURC/DOCUMErn'ATION 1.

Sampling is performed by collecting water at two levels within each tank wit!

a dedicated sampling device.

These two - discrete grab samples are composite < -

to provide a more representative sample.

Two of these representative sample; must be collected f rom each tank.

A minimum volume of one (1) liter pt.

sample is required.

Initiate a " Water Release Permit" for SK and 10K tank:

(HP-MF-25).

2.

Determine the pH and tmperature of each sample and record these values on the Non-Radiological Water Release Data Sheet (HP-MP-01) and on the Release Permi; for Tanks (HP-MF-25).

If pH and ter perature are acceptable, deliver sample and HP-MF-25 release permit to environnental personnel.

If the pil falls outside of the range allowed (6. 0 - 9.0 pH) refer to th<

attached neutralization tables to determine how much NaCll or hcl is needed Once the correct amount of tase or acid has been added, allow 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> to pon:

before resampling for pH.

(For routine samples of the 10,000 gallcn and 5,000 gallon (Mall) storagt tanks, Steps 1 and 2 wil'1 be performed by the Utilities Technicians. ) -

2.

Enter the camplb collection aata in the Sample Collection Data Sheef

(!!P-MP-07).

4.

Measure out 1000 ml of each sample with a clean graduated cylinder an<

transfer to a one liter trarinelli beaker.

Seal tJie marinellis with electricaJ tape and apply a label containing the sample number, description, collectici time, and the jnitials of the collector.

{

5.

Analyze the samples on a High Purity Germnium (HK;e) multichannel analyzer.

The following is a list of release criteria, o

For each individual isotope, the water may not be released unless th<

concentration is below 5% of the MPC.

o The total activit.y in a tank must be < 2n uCi (0.020 trCi ) to b<

releasable.

04/11/91 HP/210E 04/15/91 h

I

- p MTE IPREPAluD Bt20M/J/I4) v/w/9/

CINrIGIIN HFRml PIP / SICS PRTRXUG:S IQIFDID BY 2b Isl 4/u//N lAPPI(NAL DrLL.4 l_dFiu l

Iss11DI ML_ ;

_FMPLIlO NO REL. EASE PROC 12XJRES FOR HOLDI?C_TM&;S lPAGE 2 OP 4 HP-M-08 If the running annual total release exceeds 8 mci, then any release in o

excess of 1 uCi must be approved by supervisory 3rsonnel, o-If more than 1 isotope is present, perform the Unity calculatiorn (Set calculation below).

If the calculation result is < 10%, each individua.

isotope is < 5%, and the total activity is < 20 uCi (0.020 mci), th<

water can be released, o

Any tank sample that does not meet all three of the above critaria shoul(~

be brought to the attention of senior environnental lab staff, Radiatiol:

Saf ety Officer or Staff Health Physicist for follow-up, Any batch of water that does rot treet the appropriate release criteria 11 o

to be held for decay and/or processing.

Inmediately notify the Utility Technician to hold this tank and contact the Radiation Safety Officer o1 Staff Health Physicist (these are the only personnel allowed to authoriz(

water released in excess of these criteria) as soon as possible.

6.

IMP]RTANT:

' 'Lh samples f rom each tank-have to meet the release criterit before sectic s 3.d and e. of form HP-MP-25 may be filled out and delivered to the Utilities Technicians, t

i 7.

tbtify the Utilities Technicians as to the status of each batch of watet l

approximately one hour after receiving the sample 'as this information 10 necessary for them to control water storage.

8.

Place the printout in the "to be reviewed file" for review by senior; personnel.

9.

Store the marinelli beaker as per the sample storage guidelines.

Unity Calculation:

Por isotopes A, B...... Z

% MPC A + % MPC B +...... + % MPC Z = < 10%

Calculation of Total Activity in Tanks Por each sample, add all positive activities.

Multiply this value by 3785 ml/ gal. and the volume of the tank in gallons.

Example:

A SK tank shows the following positives when releasable:

Co-60 3.25 x 10-7 uCi/ml Cs-137 1.02 x 10-8 uCi/ml 3.25 x 10-7 + 1,02 x 10-8 = 3.352 x 10-7 uCi/ml 3.352 x 10-7 uCi/ml x 3785 ml/ gal x 5000 gal = 6.34 uCi (0.006 MCI) 04/11/91 HP/210E 04/15/91-

IPid:PAldD IW;?y.hgghi s.t][j[]y_kdg lOl1GID BY

[44dih.Jyg, I tsj,4 I Cull'IOll24 til%Iml 191YS103 PJCC13XJidS lAPPIO/AL l

TW M Dl_.g $ {jL__

NPL1tO AfD IJJ2ENjij_Ijgq13JRIJL1 RILL 0lp!?O TA?E IPAGE 3&4 IIIH4-08 TABLE 1 l~

I MIDI 11ters of 5 N tholl l

I pil I

ikquired to Neutralize Tank i

l_

l Q>ll of 7) llavins An Acid,pil 1

l_

b. 9_ _

_I 7

i 1

5.8 1

9 l

l

_S. 7 l

11 l

l 5.6 1

15 l

I 5.5_

l 18 I

l_

5.4 1 __

23 1

1 5.3 I

29 I

i 5.2 1

37 l

~

l 5.1 1

46 l

l__

_5. 0 1

57 I

l l

4.8 I

94

___l 4.9 1

73 l

l 4.7 l

114 l

l 4.6_

l,_ _

144

_ l 1

4.5 182 l

l 4.4 i

260 I

l __

4.3 l

280

'l l_

4.2 1

364 I

l ___

4.1 l

460 I

I 4.0 _

1 570 I

i 3.9 I

725 I

l._

3.8 I

920 l

1 3.7 I

lj30 I

lu 20

)

i 3.6 I

4 I

3.5 l

1,80_0 1

I 3 '. 4 l

2 450 l

2 I

3.3 1

2dj20 I

l 3.2 1

3J50 I

l_

3.1 1

4 d_40 l

l 3.0 l

5,700 1

• if pil is less than 3.0 notify ilealth Physics 04/ 11/91 IIP /210E 04/15/91

lPREPARfD IUrM/3/7Al 4hr/9 /

cue 10IIM !! Palmi PirfSICS PICCI1XIRPS 10IKKII) BY heij((ds.(

lAPPIENAL 4 %6lnl 4:n 4 /

l

- If'EA'LD l_un 'II fMPI.I!O Ato RILFASE PI(GIURFS IOR IlO(Dl!Q TNKS lPAGE 4 OF 4 III'H-08

]

TABLE Il 1

l Milliliters of r N llCL l

I pli l

Required to Neutralize Tank i

l I

(pil of 7) Having A Baric pH I

s l

9.1 l

73 I

I 9.2 I

94

.I i

9.3 i

114 l

l 9.4 1

144 I

i 9.5 l

182 I

i 9.6 1

260 l

9.7 1

280 l

I 9.8 I

364 l

l 9.9 l

460 I

I 10.0 1

570 l

I 10.1 1

725 l

I

.10.2 1

920 l

10.3 l

1,130

._ I i

10.4 l

1,420 I

i ~

10.5 1

1,800 l

1 10.6 1

2,450 I

i 10.7 1

2,820 l

i 10.8 1

3,550 I

i 10.9 I

4,440 l

I 11.0 1

5,700 1

• If pH is greater than 11.0 notify llealth Physics 04/11/91 HP/210E 04/15/91

~$

OVERSIZE DOCUMENT PAGE PULLED SEE APERTURE CARDS NUMBER OF. OVERSIZE PAGES FILMED ON APERTURE CARDS

~ APERTURE. CARD /HARD COPY AVAILABLE FROM RECORDS AND REPORTS MANAGEMENT BRANCH