ML20033E327

Forwards Response to 891222 Request for Addl Info Re Power Increases,Consisting of Tables Listing Constants Used in Whole Body Dose Estimates & Proposed Plan for Determination of Ar-41 Concentrations

ML20033E327
Person / Time
Site:	Ohio State University
Issue date:	02/28/1990
From:	Redmond R OHIO STATE UNIV., COLUMBUS, OH
To:	Michaels T Office of Nuclear Reactor Regulation
References
NUDOCS 9003120332
Download: ML20033E327 (14)

Text

.

T

11. - E Engineering Experiment Station 142 Hitchcock Hall 2070Neil Avenue O~ -

~

Columbus, OH 43210-1275 4'Am Phone 614-292-4903 H

FAX # 614-292-9021

..a UNIVERSITY February 28, 1990 Theodore S. Michaels. Senior Project Manager Non-Power Reactor, Decommissioning and Environmental Project Directorate Office of Nuclear Reactor Regulation US Nuclear Regulatory Commission Washington, D.C.

20555

Dear Mr. Michaels:

The staff of The Ohio State University Research Reactor, License R-75 Docket No. 50-150, have completed the answers to your request for additional information with regard to the power increase which was dated December 22, 1989. The answers to the six questions are enclosed. They have been reviewed and approved by the Reactor Operations Committee.

Questions on the technical content of this submittal should be directed to Mr. Richard Myser or Mr. Joseph Talnagi of the reactor staff (Phone 614-292-6755).

Thank you for your prompt attention to this matter.

Sincerely, fdmY

/

v Robert F. Redmond, Director Engineering Experiment Statjon RFR:sl c: with enclosure D. Miller J. Ta3nagi l

R. Myser II W g#5 9003120332 900228

\\\\\\

PDR ADOCK 05000150

\\'

P PDC Collegeof Engineering

.O Answers to the " Request for Additional Information with Regard to the Power Increase Regarding The Ohio State University Non-Power Reactor"

1. Ar-41

a. If one assumes that the Ar-41 inside the reactor building is modeled as in a sphere with a radius of 26 feet then

~M N)]/p g where

[K(E)ES (1-e D

8

=

rem /hr

-6 2

K(E) = 1.84x10 rem /hr/MeV/cm /ecc E

= 1.2330 MeV S

.2527 photons /cm /sec

=

R

= 792.5 cm y,

= 0.722x10~

~

cm D

= 0.000464 rem /hr D

= 0.464 mrem /hr Assuming 600 full poser hours of operation per year and thereft're 600 hours0.00694 days <br />0.167 hours <br />9.920635e-4 weeks <br />2.283e-4 months <br /> of immersion in the Ar-41 cloud provides an estimate of 278.4 millirem / year.

Even if this dose occurred in one calendar quarter, it 4

would be well below the 10CFR20.101 limit of 1250 milliress,

b. If one assumes an infinite radius cloud then the term e #8R

~

becomes zero and the equatton becomes

[K(E)ES ]/p,

D

=

rem /hr where K(E) = 1,84x10~ rem /hr/McV/cm /sec E

= 1.203G MeV

-8 S

= 4.83x10 C1/ml y

3

= 0.001787 photons /cm /sec p,

= 6.722x10~ /cm D

= 0.3x10~ rem /hr

= 0.003 millirem /hr

.,i.

~i Assuming 600 full power hours of operation per year and therefore 600 hours0.00694 days <br />0.167 hours <br />9.920635e-4 weeks <br />2.283e-4 months <br /> of immersion ~1n the Ar-41 cloud provides an estimate of 37.8 millirem / year.

This is a conservative estimate because the assumption of an infinite cloud maximizes the term (3-e'#8R).

The dose to.an individual in the unrestricted area is well below the limits of 20CTR20.105.

2. NHA in the Unrestricted Area

a. Transport fission gases'to the outside via the fan (in SAR Table'8.13 i

the purge rate is given as-0.857/ hour or 60,000 cuft/ hour)

Calculate'the following:

For a one hour exposure at some time after the release calculate i

the maximum potential (1) Thyroid dose from lodines (2) Immersion Dose for all gases including fodines.

b. Transport fission gases to the outside via leakage (use appropriate l

assumptions regarding diffusion of fission-products to the outside; in i

SAR Table 8.12 the purge rate for leakage is given as 0.0042/ hour: or 294 cuft/ hour or 10% per day)

Calculate the following:

1 For a one hour exposure at some time.after the release calculate the maximum potential (1) Thyroid dose for lodines-(2) Immersion Dose for all gases including lodines To answer 2a(1) and 2b(1) we proceeded as follows:

The source terms (in microcuries produced) are given under Isotope Activity in Table 8.10.

Whatoneneedstoknowistheconcentrationofeachfsotopeinsido'the.

j building. There are 70,000 cuft, which is 3.9822x10 al.

If one assumes complete mixing then the concentration of each may,be determined from Table 8.10.

1.e, 1-131 3x10 pC1/1.9822x10 ml

=

-6

'1.5x10 C1/ml

=

Using $ (0) 1/[(0.5)(s)(u)] from p. 150 of SAR

=

(dilution factor)

-3 3

D 921x10 sec/m

=

The building exhaust. rate is 0.47195m /sec (1000 CFM) from p.-

150 SAR 2

I

'f i

i

Using A Aq $ (0) where $ (0) is dilution factor estimated

=

(effectkve above, A = discharge concentration and exposure conc.)

q = building exhaust rate 3

Then % =

1.5x10' pC1/ml x 0.47195m /secx9.921x10-3,,cj,3 (effecIlve exposure conc.)

7.02x10' C1/ml.for 1 - 131

=

Therefore the effective exposure concentration In a factor of-213.67 times less than the concentraton inside the building.

This should be the same for each: isotope.

If'we assume that the total' dose will be reduced by the same factor as the concentration in air is reduced, one can simply correct the results found in Table.8.9 by this factor.

For example, the dose for a two hour exposure'would be 31.73 mrens instead of 6.78 Hems to the thyroid.

Similar reasoning may be applied to the thyroid dose to an-individual outside the building for two hours while the building fan is off. The purge rate is given as'10% per day (p. 222 of the SAR)3 This is about 294 curt / hour or 0.082 cuft/sec.

(0,0023m /sec).

Using the equation 1.Sx10~ pCi/ml x 0.0023m /sec x 9.921x10' sec/m

%3

=

eTrective exposure conc.

- 3.44x10' IyC1/ml 4

This is a factor of 4.36x10 less than the original concentration. Reducing the dose by this factor yields a two hour thyroid dose of 0.156 area.

One hour doses would be less than these by;about a factor of two since the main contributors to the total dose, la131 and 1-133,'

both have relatively long half lives in relation.to the time involved.

To answer 2a(2) and 2b(2) we proceeded as follows:

From p. 47 of the document " Answers to NRC, Questions Concerning.

500 Kw operation", the following equation to estimate the dose rate assuming a source-distributed in air la provided.

R (K(E)ESv(1-e~Ms ))j D

=

g assuming R 4

• the equation becomes

[K(E)ES ]/y,where D

=

y 2

rem /hr.per Mev/cm /sec (Constant 1.84x10-6)

. K(E)

=

3

J "e

3

'Sj source strength in photons /cm 7,,g_

=

E average photon energy.in NeV (Table 8.11 SAR)'

=

linear absorption coefficient in air (Table 8.11~SAR) p,

.=

Since the average photon energy in MeV from Table 8'.11 of,the-SAR is given per disintegration, one must assume this is based-

- on.one photon of average energy per disintegration. Therefore to find S 'wo will assume.the same.

y

'S Effective exposure concentration x 3.7x10 dps/uCi

=

y

-in yC1/ml

'(Axqx $ (0))

8 A=

Isotope (#Cl) / bu11 ding (1.9822x10 mli activity volume 3

building exhaust rate either ~0.47195 /sec'(fan on).

q

=

or-0.0023m /sec:+(fan,0ff)

-3 3

$ (0) 9.921x10 sec/m

=

the only variable is isotope ~ activity Constant for fan on =

4 Isotope Activity x (3.7x10 / 1 9822x10 )x(.47195x9.921x107)

~0

8.74x10 Constant for fan off

Isotope Activity x (3.7x10* / 1.9822x1b )x(_.00N3x9[921x10~ )

~

-0

= 4.2Gx10 See Table 1 for S values and other' constants-y i

V s

4 4-

+

W

t x

0 Table 1-Constants Used in Whole-Body Dose Estimates Nuclide Average Gamma

. Absorption S

S Symbol-Energy Coefficient fan # n fan #off-o (MeV)

,jnAir-5)

(cm x 10

-6 131 0.4 3.9 2.6x10

'1.3x10 3

-6

~

132; 0.8.-

3.7 3.8x10 1.9x10 133; 0.55-

' 3. 9 __

5.9x10 2.9x1b~

~

~4

~

134.

-1,3 3.' 4 -

7.1x10 3.4x10 3

z

~4

-6 135

- 1. 5

' 3.3

-5.8x10 2.8x10 3

85m 0.2 3.5 4.2 2x10" Kr

-2_-

87 2.0

- 3.0 8.1 3.9x10 i

gp I~

88

.0 3.0

-1.2x10 5.7x10 Kr 131m 0.10

3. 3' O.4 4.6x16~

Xe I

133m 0.23

'3.6 2.1x10 1x10 y

133 0.08 3.2 2.1x10 1x10

~

Xe 135m 0.52 3.9 2.1x10 1x10 '-

~

Xe 135 0.25 3.6 221x10 '

1x10~

Xe r

W 5

=

c

[K(E) S ]/p, D

=

y Dose in rem per hour'from an effective exposure concentration caused by the release of isotopes through the vent-fan (Case 1)-or via Icakage with the vent fan off (Case 2) in the lee of the building Case 1 case 2

-6

-8 1-131 4.9x10 2.5x10

~

I-132 1.5x10 7.6x10' 1-133 1.5x10~

7'5x10' I-134 5.2x10' 2.4x10

~

I-135 4.8x10~

2.3x10

~

-2

~

Kr-85m 4.4x10 2.1x10

-3 Kr-87 9.9x10~

4.8x10 Kr-88 1.5 7.0x10~

Xe-131m 8.4x10~

~

4.1x10 Xe-133m 2.5x10~

~

1.2x10

-2

~4 Xe-133 9.7x10 4.6x10 I'

~

Xe-135m 5.2x10 2.5E10~

~I Xe-135 2.7x10 1.3x10 3.75 rem / hour 1.79x10' rem / hour

-2 The numbers of 3.75 rem / hour and 1.79x10 res/ hour were determined assuming an infinite cloud.

In Appendix E to the Answers,to NRC Questions Concerning 500 Kw Operation" it was stated that,the most likely occupied unrestricted area was'210 feet away.

The least likely occupied unrestricted area is 28 feet below the exhaust fan.

If the above dose rates are corrected for these distances as the radil of the radioactive cloud, significantly.Iower doses are.the result.

3.75 rem / hour x (1-e #8 )

~

~

assuming p,

3.0x10 for all isotopes

=

1 1

6 4

r

. d q

i..

4 ~

i then For Case 1

~

3.15 x (1-e'#8 ) where R = 210 ft or'6.4x10 cm l

3

.655 rem / hour with vent fan on

=

3.75 x:(1-e #8 ) where R 3 28 ft.or 8.5tt(J cw

~

2

=.095 rem / hour with vent fan on For Case 2-

-2 3

.1.70x10 rem / hour x (1-e'#8 ) where R = 210 feet or 6.4x10 cm' 3.13x10~ rem / hour with vent fan off

=

1.79x10~ ren/ hour x- (1-e # sR)^ where R = 28' feet or 8.53x10 c, 2

~4 4.52x10 rem / hour with vent fan off

=

1

3. Radioactivity in the City Sewer System e

u. The piping for the_ primary system goes to a depth of three. feet into tne' reactor pool.

If there was a break inothe primary piping upstream of the check valve. the maximum amount of water to be potentially.

released.to the floor drain would'be about 870 gallons.

't

b. Two isotopes of interest are: 1. tritium.(H-3) und 2. sodium 24 (Na-

24) which are the two isotopes reasonably expected to be present after reactor operation.

7

1. For tritium we assumed the following:

3

'N=4.36xg0 atoms of Il-2-in the core o = 5x10 barns

& = 6x10 neutrons /cm /sec t = 600 hours0.00694 days <br />0.167 hours <br />9.920635e-4 weeks <br />2.283e-4 months <br /> (.0685y).

A =.603 12.26 y Using the formula A' = No$. (1-e"

) it was determined there would be about'136.6 uCl produced in one year of operation assuming 600 total full power hours. If-the' top three feet of water were released it would contain about 20 uCl of-. tritium.

~

136.6 uC1/21,955,000 ml in the reactor pool

-6

= 6.2x10 uC1/ml which is alrgady below MPC for an unrestricted area (3x10 uC1/ml).

7 Y

,--i s

x

2. For sodium 24 we determined the following:

-6 Concencration = 5x10 uC1/ml for a six hour 7 KW run.

This was measured using a one liter sample in a Marinelli beaker counted on a GeLi detector.

Correctingthistoasaturationactivftyfora500KW run yields a concentration of 1.5x10 uC1/ml

~4 MPC for Na-24 in soluble form is 2x10 uC1/ml The top three feet of water in the pool therefore contains about 4940 uC1. This requires about 24,700 liters or 6525 gallons to be diluted to MPC. Since it is already in 870 gallons the actual amount needed is about 5655 gallons.

The water leaving the Reactor Building enters a 6" line which eventually reaches a 105" line for combined sanitary and storm drainage. These lines are essentially inaccessible. This junction is about one-mile to the East of the Reactor near the Water Resources Building of The Ohio State University. University records indicate that the lowest monthly water-utilizagionfor88-89wasduringJanuary1989. This was 10.2x10 cu.ft. or 76,300,2g0 gal which is about-2,461,400 gal / day (= 9.3x10 ml/ day) 4940 uC1/9.3x10 ml =5.3x10~

uC1/ml which would be the concentration of the sodium 24 in the sewer system after leaving the University.

This is well below MPC for the 4

unrestricted area which is 2x10 C1/ml for soluble form sodium 24.

4. llentinfr, Ventilation and Air Conditionini' System (HVAC) t As a part of the overall startup testing program for operation at 500 KW, a plan to determine Ar-41 concentrations in the restricted aren during various operating conditions shall be implemented.

The measurements will be made using a shielded volume air sampling detection system calibrated for Ar-41.

The ovatem currently in use is calibrated annually and typically indicates MPC = 10 cps.

Figure 6.1 ou page 143 of the SAR indicates that the equilibrium concentration for Ar-41 inside the building is reached in about four hours. Measurements of air exiting the building at the exhaust fan will be made to verify when equilibrium has been reached.

After this, other locations in the building shall be. monitored.

Furnace fans shall be running during all tests.

The testing program to determine Ar-41 concentrations is described below.

8

j Tests shall be completed at least for four power levels 10 KW, 100 KW, 250 KW and 500 KW.

Three different operating conditions shall be investigated.

These are: Ar-41 production from the reactor paol alone, Ar-41 production from the reactor pool with the rabbit running continuously, and Ar-41 production from the reactor pool including a puff release from the rabbit.

A total of twelve locations may be sampled.

These are:

1 Room 100. Reactor Bay Vent Fan

2. Room 100, Reactor Bay Reactor Pool Top Catwalk

3. Room 100, Reactor Bay BSP Pool Top Catwalk

4. Room 100, Reactor Bay Thermal Column Area

5. Room 100, Reactor Bay Beam Port Area

6. Room 103, Of f ice

7. Room 103A, Office

8. Room 109, Counting Room

9. Room 104, Office

10. Room 201, Conference Room

11. Room 205, Reactor Control Room

12. Room 209, Office It is anticipated that for the puff release, only locations 1, 7, and 11 will be sampled since the concentration of Ar-41 should peak and then decrease back to equilibrium not allowing time to sample all locations.

After measurements are completed for each operating condition. the reactor shall be shut down. Additional measurements at all locations shall be completed after shutdown to determine the purge rate. This will facilitate a determination of how long to monitor after shutdown during normal operations.

From these measurements we shall determine:

1. if any areas need to be posted as airborne radioactivity areas per 20.203(d);

2. what areas reach the highest concentrations of Ar-41, and whether these are typically occupied; and

3. how long we can operate at various powers and conditions while maintaining doses ALARA per 20.103(b)(2).

Table 2 shows the proposed plan for determination of Ar-41 concentrations during start up testing.

After this testing in completed, regular monitoring during operations will be done using the effluent monitor.

9

0 4

e

=.

l

- f

-Table 2. Proposed Plan for Determination of Ar-41 Concentrations.

Power Operating:

Sample Sample.

' Level Condition'

}Iour Locstion i

All (10,100,250,500 KW)

Poo.1 Top-1,2,3,4 Vent

]

All Pool Top 4-5 All i

Shutdown 9 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> Pool Top.

'6,8 All.

All Pool' Top + Rabbit 1,2,3,4

' Vent All Pool Top-+ Rabbit 4-5 All.

Shutdown 9 5 haurs Pool Top + Rabbit 6,8 All' All Pool Top 4 Puff 1,2,3,4 Vent All Pool Top +. Puff 4 '1,7,11 Shutdown 4 5 hour5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />s-Pool Top + Puff 6,8 All'

5. Expanded Environmental Radiation Monitoring Program-In addition to the fenceline monitors (= 20 feet from the' reactor-building) we will add monitors outside of the exclusion area. At least-four of these-will be positioned in the unrestricted area about 100 feet-from-the-reactor building. These areas are essentially unoccupied in all directions except to the South-About 150 feet to the South is=the Van de Graaff Laboratory.

3 To the East js a woods used a few times each year for bird watching and dog i

1 walking. To the North is a bike / jogging path used by a few individuals.-

daily. There is+a gravel access road and open field to the West. A few:

cars per day use the road. The badges monitoring these areas will be changed on a quarterly basis.

4 10 a

=

6. Healt h Physics interface The current Office of Radiation Safety.(ORS) functions and responsibilities at the Nuclear Reactor Laboratory include the following items:

FUNCTION FREQUENCY

1. Conduct area and smear surveys Monthly of NRL

2. Compile summary of results of area Semi-annual and smear surveys conducted by both NRL and ORS

3. Assure proper posting of Radiation Monthly or areas, High Radiation areas, etc.

as needed.

for routine reactor operations

4. Assist in calibration of Area Annual Raulation Monitors (ARM) 4l

5. Assist in calibration of Ar Annual gaseous effluent monitor

6. Assist in calibration of NRL B, 7, Annual neutron survey instruments and liigh Range instruments 7.

Inventory and wipes of sealed Semi-annual radioactive sources assigned to NRL

8. Review records of radioisotope As needed shipments from NRL

9. Special Nuclear Materials Report (SNM)

Semi-annual However, the day to day development and implementation of the health physics program at the reactor has always been the responsibility of the Nuclear Reactor Laboratory (NRL) Staff.

Attached is a list of the current radiation safety procedures in use at the reactor which are revised and maintained by the NRL staff. Whenever there has been a need for increased health physics support, the NRL staff has simply increased its activity in the area to provide a radiologically safe environment and document its effectiveness.

For example, in the 1960's the NRC required the reactor to install a gaseous-effluent monitoring system. The NRL staff accomplished this along with a detailed calibration procedure.

They also developed and provided radiation safety instruction to meet 100FR19 requirements.

In addition, a new radioactive material shipping procedure was recently completed. This required a significant effort on the part of the NRL staff.

In each of these cases.the NRL staff has devoted time and effort to accomplish the health physica goal of a radiologically safe work place.

11

~

4 OPERATING PROCEDURES NUCLEAR REACTOR LABORATORY

.THE OHIO STATE UNIVERSITY RADIATION SAFETY (RS) i Procedure Procedure Latest Revision ROC l

Number Title Number Date Approval i

RS-01 Labeling und Storage of 1

2/2/89 6/10/89 Radioactive Materials RS-02 Radioactive Waste Disposal 1

-2/3/89 6/20/89 RS-03 Calibrating the Gaseous 2

2/3/89 4/13/89 i

Effluent Monitor RS-04 Particulate Air Sampling 2

2/6/89 4/13/89 RS-05 Pool Water Radioactivity 1

2/13/89 4/13/89 RS-06 Quarterly Radiation Monitor 2

2/7/89-4/13/89 RS-07 NRL Weekly Direct Prisk 2

7/21/89 7/6/89 j

RS-08 NRL Smear Survey 5

2/13/89 7/6/89 RS-09 Area Radiation Surveys 2

2/9/89 6/20/89 RS-11 Routine Shipment of Orig.

7/25/88 7/6/89-Radioactive Materials RS-12 Decontamination Procedures 1

2/8/89 6/20/89 RS-13 Sealed Source Wipes 2

2/9/89 6/20/89 j

RS-15 Radiation Safety Instruction 4

2/15/89 6/20/89 l

RS-16 Dosimeter Calibration 1

2/10/89 7/6/89 RS-17 Ar-41 Release Calculation 2

7/28/89 7/6/89 1

12

]

. t The NRL staff realizes there will be increased potential for exposure at 500 KW operation. Evaluation shall occur during the startup testing program.

From this evaluation, reviewed by the Reactor Operations Committee and the NRC, additional health physics programs may be identified as necessary to assure a safe environment. Any such programs shall be implemented by the staff of the Nuclear Reactor Laboratory.

Support, however, shall be provided by the Office of Radiation Safety.

ORS currently provides support for and participates in activities requiring additional health physics emphasis at the Nuclear Reactor Laboratory.

For example, they are currently involved in the review and approval of the fuel shipment procedure, part of this procedure involves trimming the ends of the fuel elements so they fit into the shipping cask properly ORS participated in an Emergency Drill-and critique that was based on a fuel cutting accident.

They will be active participants in the entire fuel element shipping procedure including the element trimming. They will also be active participants in the health physics aspects of the start up testing program.

This will include, in part.icular, aonitoring of the Ar-41 concentration and direct radiation hazards associated with 500 Kw operation.

It is anticipated that an individual from ORS will be required from 25-50%

of the time during start up testing. ORS has agreed to this commitment.

Since the Director of ORS is an ex-officio voting member of the Reactor Operations Committee, the Office of Radiation Safety is always advised of l

reactor activities. Both the Director and Assistant Director of the Office of Radiation Safety have extensive research reactor experience and recognize the need for careful Seulth physica planning and monitoring, before, during, and after the power increase.

They will assure that the health physics program implemented by the NRL is appropriate, s

13 m

-