ML20151J105

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Rev 2 to Procedure TDR 494, TMI-1,Post-Accident Sampling Radiological Analysis
ML20151J105
Person / Time
Site: Three Mile Island Constellation icon.png
Issue date: 05/10/1984
From:
GENERAL PUBLIC UTILITIES CORP.
To:
Shared Package
ML20151J101 List:
References
RTR-NUREG-0737, RTR-NUREG-737, TASK-2.B.3, TASK-TM TDR-494, NUDOCS 8406260021
Download: ML20151J105 (44)
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g TDR NO. 494 REVCION NO. A SUDGET TECHNICAL DATA REPORT ACTIVITY NO. _1219Ql.__. PAGE / OF __M DEPARTMENT /SECTION NAF/TMI Fuel Pro.iects -

T!G-1 RELEASE DATE A-x 83 REVISION DATE 3'-M-F/

DOCUMENT TITLE: "TMI-1, POST-ACCIDENT SAMPLING RADIOLOGICAL ANALYSIS" ORIGINATOR SIGNATURE DATE APPROVALIS) SIGNATURE DATE 0$r > $ $W S/9/S Y~ k'  ;

S~ W h

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APPROVAL FOR EXTERNAL DISTRIBUTION ,DATE y M hh foe-D Does this TDRinclude recommendation (s)?[5Yes ONo If yes.TFWR/TR #- 300661 ,

o DISTRIBUTION A88 TRACT:

G. R. ' Bond STATEMENT OF PROBLEM T. G. Broughton T. Y. Byoun

  1. D. K. Croneberger This report evaluates the TMI-1 on-site radiological conditions L. W. Harding in terms of personnel radiation exposures during the U. W. Langenbach post-accident primary coolant sampling and associated spectral analyses.

J. D. Luoma M. Mahgerefteh

SUMMARY

AND CONCLUSIONS J. D. Snell J. A. Tangen

  • P. S. Walsh
1. With lead shielding currently installed in the Sa=pling Room, the radiation exposures to plant personnel during J. S. Williams (T!G.) the post-accident sampling are well within the radiation
  • R. F. Wilson .

limits set by NUREG-0737, II.B.3.

J. Wetmore R. Shaw (T!C) 2. The radiation exposures to personnel are:

H. D. Hukill (TMI)

M. R. Knight (TIC) TECHNICIAN NATURE OF RADIATION E. C. Fuhrer (T!G) I.D. WORK STAY TIMI EXPOSURE R. J. Toot.E C 7*Mi)

Tech. I Prepare Sample 2 min.

Room & Valve' 0.89 Rem {

Lineups I Tech. II Take Liquid 3 min. 2.21 Rem Sample I Tech. III Take Gas Sample 3.5 min. 1.28 Rem ,

Tech. IV Cleanup and 15 min. 0.10 Rem Decommission

3. With four (4) inch thick lead shielding (or equivalent thickness for other shiolding material) around the GeLi detector in the Counting Room, the background radiation will be reduced to sufficiently low level to assuro'(cont'd) tCOVER PAGE ONLY aoooooso 4.s3 8406260021 840608 PDR ADOCK 05000289

._G_--.-.____ _

PDR _ _- . _ . . -.

TDR 494 Rev. 2 Page 2 of 44

SUMMARY

AND CONCLUSIONS an accuracy within a factor of two (2) for the measurement of the primary coolant activity.

(4) If an accident involves depressurized condition, the sampling can be performed through the Decay Heat Removal System 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />,:

after the accident and radiation exposure will be less than 0.2 Rom.

(5) The existing Containment Atmosphere Sampling (CAS) System with some piping rearrangements, is capable of taking representative gas samples even without providing heat trace (see Recommendations below).

The total iodine sample loss in the Sample Bomb will be less than 3%.

(6) Radiation exposures to a technician who performs sample analysisss/tr [Jb=.IE4-are 1.0 Rem whole body dose and approximately 3.0 Rem of extremity dose.

4 RECOMMENDATION To eliminate the possibility of the Containment Atmosphere Sample condensation in the Sample Bomb, the following design modifications are recommended:

1. Remove the Sample Bomb or reroute the CAS piping to install the Sample Bomb in a vertical configuration.
2. Install a small cask at a low point of the piping to collect condensed water.
3. Provide insulation to a non-heat-traced portion of the piping.

4 4

9

D*GE 3 of 4f DOCUMENT NO.

GUCl88r roa 494 TITLE "TMI-1; Post-Accident Sampling Radiological Analysis" REV

SUMMARY

OF CHANGE APPROVAL DATE e

1 . Resolution of N3C's comments.

. Reflecting updated time - and - motion data for the primary coolant sampling.

. Added analysis for the iodine plate-out and condensation in the CAS Systen.

,...-...7 2 . Add a paragraph on the plant pr'ocedure (RCP-1613) for the protective respirator requirements during accident period. ,

f gg A0000034 7 02 t

l TDR No. 494 Rev. 2 Page 4 of 44 TABLE OF CONTENTS 4

1 I. INIRODUCTION II. METHODS AND DATA l

f l A. Source Terms

  • B. Dose Acceptance Criteria C. Post-Accident Sampling Scenario D. Radiation Source Geometry

', E. Airborne Activities in the Counting Room F. Computer Code Used C. Dose Rate Results i

III.

SUMMARY

OF RESULTS AND DISCUSSIONS j

A. Overall Results

~

B. Radiation f rom Sink Drain Trap and Drain Line C. Radiation f rom the Undiluted Coolant in the Sink-D. Scattered Radiation l E. Residual Contamination during Subsequent Sampling Attempts i

F. Airborne Activity Originating irom the Sink

G. Airborne Activity in the Auxiliary Building f

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j . ..' ..

TDR No. 494 Rev. 2 4 Page 5 of 44 H. Iodine Plate-Out and Condensation in the Sample Bomb of the CAS l

System I. Primary Coolant Sampling by Using Decay Heat Removal System l

J. Radiation Exposure During Gas and Liquid Sample Analysis 1

IV. REFERENCES i

APPENDIX (A) Descriptions of the ISOSHLD-II Code i

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TDR Ns. 494 Rev. 2 Page 6 of 44 LIST OF TABLES TABLE 1. Radiation Source Tera Data LIST OF FIGURES Fig. 1. General Arrangement Drawing for Nuclear Sampling Room Fig. 2. Primary Coolant Sample Line Pipings Fig. 3. Simplified Shielding Geometry in the Sample Room Fig. 4. Shielding Geometry for S123 Fig. 5. CAS System Piping Arrangements Fig. 6. Iodine Removal by Containment Spray i

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TDR Ns. 494 Rev. 2 Page 7 of 44 I. INIRODUCTION The purpose of this report is to evaluate the TMI-1 onsite radiological conditions during the post-accident sampling and associated spectral anayses in terms of personnel radiation exposures and shielding requirements for compliance with the radiation criteria set by NUREG-0737, II.B.3.

Specifically:

a i 1. Prediction of personnel radiation exposures based on person-motion for sampling, transport and analysis of all i required parameters.

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2. Analysis of the background levels in the Counting Room and I recommendation of shielding requirements to reduce the background radiation low enough to assure an accuracy within a factor of 2 in measuring the primary coolant activity.

NRC requires the licensee to have the capability to promptly obtain reactor coolant samples (within 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> f rom the time a decision is made) without exceeding limits of the radiation criteria of GDC 19, App. A,10CFR Part 50 (i.e. , 5 Rem whole body, 75 Rea extremities).

The TMI-1 Plant Procedure No. 1004.15, Rev. 4 further reduces these radiation limits to 3 Rea whole body and 18 3/4 Rem to the extremities. The post-accident sampling activities consist of the

-- q l

TDR Ns. 494 Rev. 2 ,

Page 8 of 44  !

opening / closing valves and taking liquid and gas samples in the sample room, dilution of liquid sample in the radio-chemistry laboratory, and analyzing the diluted samples in the counting roon  :

and the primary chemistry laboratory. These tasks will be performed by 5 or 6 technicians to avoid the excessive radiation exposure to -

an individual.

The radiation source term data used in this report are same as those stipulated in Reg. Guides 1.3 and 1.4 (Table 1). Radiation dose f rates are analysed by using computer code ISOSHLD-II (ref.1), which employs a point kernel integration method.

II. METHODS AND DATA P

i A. Source Term Data

1. The source term data listed in Table 1 of this report were taken f rom Table 3.1-2 of Ref. 2, which is consistent with R.G.1.3 and R.G.1.4 source terms: "
a. Liquid Containing Systems: 100% of the core equilibrius t

noble gas inventory 50E of the core equilibrium halogen inventory and 1% of all others will be assumed to be mixed in the reactor coolant and liquide recirculated by

- _ . .- ~~ . _ _ _ _ _ _ . - - - - . . - . .

TDR N2. 494 Rev. 2 Page 9 of 44 i .

i I high pressure injection (HPI) and low pressure injection I

(LPI). In determining the source term for recirculated. l l depressurized cooling water, it was assumed that the l water contains no noble gases.

h. Gas Containing Systems: 100% of the core equilibrium l

l noble gas inventory and 25% of the core equilibrium i

l halogen activity will be assumed to be mixed in the containment atmosphere.

2. All the sample line piping source were conservatively assumed to contain sero-time decayed reactor coolant activities (Table 1).
3. The source term for the liquid sample bottle was assumed to be depressurized (No noble gases) and 30-minute decayed.
4. Liquid sample cource term in the Counting Room - 2 cm3 of diluted liquid sample with dilution factors (DF) of 10 I

and 10 (Ref. 4).

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r TDR Ns. 494 Rev. 2 Page 10 of 44

5. Extremity dose rate f rom Syringe containing I cc of 30-min decayed noble gas sources

(

j A syringe is used to draw 1 cc of noble gases f rom a l

l container, with volume of 300 cc, holding noble gases which l

are transferred f rom 65 cc PC (Primary Coolant) liquid l

sample bottle.

B. Dose Acceptance Criteria r

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1. Criterion of CDC19, Appendix A,10CFR Part 50:

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5 Rea Whole Body l

75 Rea Extremities

2. TM1-1 Plant Procedure 1004.15, Rev. 4 (used in this report) 3 Rea Whole Body 18 3/4 Res to Extremities C. Post-Accident Samplina Activity Scenario (Fla. 21
1. Tech. I spends 25 minutes in the sample Room establishing valve lineup (all the sample lines with no primary coolant yet). Then Tech. I spends 2 minutes opening the valve I

p i

TDR Ns. 494 Rev. 2 l Page 11 of 44 i CAV-110 (Fig. 2) around sample lines which contain primary coolant. Tech. I will also check both temperature (T11023) and pressure (PI 1103) indicators within 2 minute time period.

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2. Tech. II spends 4 min. In the Sample Room taking the liquid sample (Ref. 4):
  • Verify that temperature of 150*F or less is indicated on CA4 TI.
  • Raise hood door and open CA-V107 to allow the desin, water to flush sink.
  • Opan CA-V16 (PC sample valve) and allow CA-V16 to purge (open) for 15 seconds and close the valve.
  • Open valve CA-V16 and fill the 300 at sample bottle to l

preserked fill line level (30 al).

Move sample bottle to lead pts (1 inch lead shielding),

l and cap it.

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TDR No. 494 Rev. 2 Page 12 of 44 l

3. Tech. III takes the gas sample for 3.5 minutes. Tech. III l 1s assumed to have a 30 second extremity contact exposure to the gaseous sample of I cc. A syringe is used to extract l 1 cc of gaseous sample f rom a 300 cc container which l contains noble gases drawn f rom the liquid sample.

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4. Tech. IV spends 15 minutes in the Sample Room cleaning up the gas sample train and decommissioning the system within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. ,

l

5. Technicians are expected to follow plant procedure for protective respirator requirements (RCP-1613) to reduce inhalation doses during the sampling.

D. Radiation Source Geometry 1

1. Sample line pipings c'o ntaining primary coolant are given in Figures 2, 3, and 4 l

( 2. All ssapling line pipings have 3/8" - DIA and 0.065" - thick pipe wall thicknesses.

3. Individual Source Description (Actual PC saeple pipings were converted to ISOSHt.D code geometry) a) 81 ,2,3 = 3 of 30"-long pipes (see Figures 3 & 4) with 4"-thick lead shielding s

TDR No. 494 Rev. 2 Page 13 of 44 b) S 4,5 - 2 of 7 f t-long pipes with 2"-thick lead shielding c) S 6,7 = Same as S 4,5 except 10 f t-long

= 2 of 3 f t-long pipes with 2"-thick d) S 8,9 lead shielding

= 5 f t-long pipe with 4"-thick lead shielding e) S 10 ,

f) S 11 = Cooler source with 2" lead shielding (Ref. 3) g) S 12,13 = 2 of 16 f3-long pipes with 4"-thick lead shielding 4

h) S 14 = Unshielded portio of both upstream (10")

and down' stream '(6") of the valve CA-V16.

= 2"-diameter and 5 f t-long drain line this

1) S 15 drain line carries both primary coolant (0.25 gpm) and 'demin water (7.5 gpm) during the liquid sample taking period.

j) S 16 = 30-min. decayed an > degassed primary coolant s i sample (30 cc)' contained in the liquid sample j 1

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- " * - bottle (2"-dia. and ,415"-high' cylindrical).

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TDR No. 494 Rev. 2 Page 14 of 44 k) S 17 = 1.5"-thick lead pig which contains the PC sample bottle (S16)

1) S 18 = 1 cc f the gaseous samples in a syringe i

m) S yg = Airborne radioactivity originsting f rom sink n) S 20 = Undiluted primary coolant in the sink which is assumed to be 2 f t-diameter and 1/32-inch thick film. This is a very conaervative model beesuse, in actual case, the primary coolant f rom the sample piping (0.25 gpm) will be diluted with demineral-ized water (7.5 gpm) with mixing ratio of approximately 30.

o) S 21 = Unshielded piping (about 3"-long) in the flow indicator (CA-7-FI)

K. Airborne Activities in the Counting Room

1. Atmospheric Dilution Factor (X/Q) was based on the diffused-source point receptor configuration (Ref. 3)
  • 3 X/Q = 3.44 x 10-3 sec/M

,e ,,

TDR No. 494 Rev. 2

'; age 15 of 44

2. Wind Speed of 0.782 m/sec and Pasquill Type of "F" l
3. Finite cloud model in the Counting Room F. Computer Code Used - IOSOSHLD II (Ref. 1) which employs a point kernel integration method. Further descriptions on the ISOSHLD II code are given in Appendix (A) of this report.

i

G. Dora Rata Retulta

-1. RADIATION EXPOSURE TO TECH I .

RAD EXPOSURE STAY RADIATION SOURCE DURING THE WORK WORK DESCRIPTION TIME & RADIATION LEVEL (Rem /Hr) (Rem / ACTIVITY Prepare Sample Room 25 min. Airborne Released = 0.3 0.125 WB

- Valve Lineups (No PC from containment in the pipings) Containment = 0.005 0.003 WB Direct Shine Open Valve CA-V-110 2 min. S1 ,2,3 (Pipings) =

0.895 0.76 WB

& Check TI 1023 S4 5 (Pipings) = 16.44 S 8 ,,9 (Pipings) =

2.42 Si g (Pipings) = 0.17 S12,13 (Pipings) = 1.75 S11 (Pipings) = 1.13 SUBTOTAL = 22.8 (Rem /hr)

TOTAL RADIATION EXPOSURE TO TECH. I IS:

m TDI'= 0.125 + 0.76 + 0.003 $yH< E (D *

-10.888Remi  ;".

E $

/

2. RADIATION EXPOSURE TO TECH II ,

RAD EXPOSURE STAY RADIATION SOURCE DURING THE WORK ..

WORK DESCRIPTION TIME & RADIATION LEVEL (Rem /Hr) (Rem / ACTIVITY Open CA-V16, Purge 20 see S14 (Unshielded Pipe) = 28.1 0.629 the RC for 15 seconds, S15 (Drain Line) = 68.5 and close CA-V16 Si g (NG from sink) =

43.0 (5 Sec)

S20 (Diluted RC Film) = 8.10 in the sink floor)

SUBTOTAL = 147.7 (Rem /hr)

Obtain Grab Sample (5 sec) 22 see Radiation Level Same - 147.7 0.763 WB place sample bottle (2 sec) As above 4.28 Extremity fill the bottle (5 sec) to 30 m1 mark, close the S16 (Unshielded =

25.7 (10 see) sampic bottle (3 sec) Sample bottle) and close CA-T16 (5 sec) SAMPLE BOTTLE EXTREMITY =1540 EXTREMITY (10 Sec)

SUBTOTAL = 173.4 (Rem /hr) WB Move sample bottle to lead 11 See S16 (Sample Bottle) =

25.7 (3 Sec) 0.11 Pig on Cart (2 Sec) and Push Cart to containment barrier S14 (Residual RC =

0.86 (11 Sec)

(7 Sec) Close Hood in upstream of CA-V16)

S15 (Trapped RC =

7.07 (5 Sec) in the drain line)

S17 (Sample in the =

11.8 (7 Sec) lead pig - 20 ft distance)

Other pipings including =

18.6 (11 Sec) cooler, and scattered D'E'E!

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2 RADIATION EXPOSURE TO TECH II (continu:d)

RAD EXPOSURE  :.

STAY RADIATION SOURCE DURING THE WORK WORK DESCRIPTION TIME & RADIATION LEVEL (Rem /Hr) (Rem / ACTIVITY ~

r

  • i Other Tech. II's Movement 2 min.
  • Airborne f rom containment = 0.16 0.71 in the general sample
  • S4,5 (Pipings) = 9.67 room area
  • 36,7 (Pipings) = 5.09
  • S8,9 (Pipings) = 2.42
  • S10 (Pipings) = 0.17
  • S11 (10 ft. f rom cooler) = 1.13
  • S15 (Trapped draw line - = 1.8 10 f t. away)
  • S41 (Scatt. f rom residual = 0.86 RC the CA-T16 Upstream)

SUBTOTAL =

21.3 (Rem /hr)

TOTAL RADIATION EXPOSURE TO TECH. II IS:

  • TD II = 0.629 + 0.763 + 0.11 + 0.71

=l2.21 REM WBJ

  • EXTREMITYDOSETOTECH.II=(4.28 REM (EXTREMITY))

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3. RADIATION EXPOSURE TO TECH III -

RAD EXPOSURE -

STAY RADIATION SOURCE DURING THE WORK WORK DESCRIPTION TIME & RADIATION LEVEL (Rem /Hr) (Rem / ACTIVITY Prepare for the gas sample 1 min.

  • S1 ,2,3 (Pipings) = 0.895
  • 0.413 taking - valve manipula- & 20 sec.
  • S4 ,5 (Pipings) = 9.67 tions (verify, open, and
  • S8,9 (Pipings) = 2.42 close various valves) *Sg i (Pipings) = 0.514
  • S11 (Cooler) = 1.13
  • S12,13 (Pipings) = 0.58
  • S14 (Shadowed RC piping - -

1.33 unshielded)

  • SIS (Shadowed drain line = 0.176 trap)
  • S16 (Lead pig on the =

1.89 the cart-5 f t away)

  • SUBTOTAL =

18.6 (Rem /hr)

Withdraw 1 al of gas sample 30 S.ec

  • Same as above = 18.6
  • 0.31 WB f rom 300 m1 expansion *S18 (Syringe-1.5 f t) = 0.069
  • 1.2 Extremity cylinder using the locking syringe (20 Sec) and place syringe in lead pig on cart. *S18 (Syringe-contact) = 144 (Extremity)

Open hood door and close 10 See *S45 (Piping) = 9.67

  • 0.052 valve CA-V107 in the *S89 (Piping) = 2.42 sample sink *S10 (Piping) = 0.514
  • S11 (Cooler) = 1.13 0.37
  • S14 (Residual) = 4.58
  • S15 (Diluted) = 0.69 m

SUBTOTAL = 19.0 {ps<E G " .E

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3. RADIATION EXPOSURE TO TECH III (centinuad)
  • RAD EXPOSURE -

STAY RADIATION SOURCE DURING THE WORK WORK DESCRIPTION TIME & RADIATION LEVEL (Rem /Hr) (Rem / ACTIVITY Take samples on cart to 30 Sec.

  • Liquid sample'in the lead pig = 11.8
  • 0.145

. Chem. Laboratory

  • Gas sample in the lead pig = 0.013
  • Pipings scattered, and general = 16.7 (10 Sec)

Other Tech. III's movement 1 Min.

  • General area (piping, shadowed, = 21.3
  • 0.355

'in the general sample room and unshielded)

TOTAL RADIATION EXPOSURE TO TECH. III IS:

  • TD III = 0. 413 + 0. 31 + 0. 052 + 0.145 + 0. 355

-(1.275 Rem WB}

  • EXTREMITIES =(_1.2 Rem) 2 .< "

& " .E as*

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I TDR No. 494 Rev. 2 Page al of 44

4. Radiation Exposure to Tech. IV Cleanup of the gas sample train should be performed within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> (Ref. 4) af ter gas sample has been taken or prior to taking the next sample, which will occur first. Tech. IV will perfo rm: (1) various valve manipulations, (2) system flushing, drain, purge and (3) decommission of the system. The technician will spend approximately 8.5 minutes, which was actually timed in the plant. The radiation level in the sample room will be reduced by approximately a factor of 15 in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> ranging f rom 0.2 Rem /hr to 0.4 Rem /hr in the room.

The total radiation exposure to Tech. IV was conservatively estimated by assuming that he stays in the room for 15 minutes:

0.4

  • 15/60 = 0.1 (Rem) 3
5. Contact does rate f rom a 2 cm of liquid sample (DF = 10 )

in the Counting Room = 0.21 mrem /hr

6. The maximum airborne activity in the Counting Room is 160 mrem /hr.

In order to assure the + 50% accuracy of the sample analysis, the background radiation (160 mrem /hr) should be reduced to about one half of the 2 cm liquid sample does rate

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i'DR No. 494 Rev. 2 Page 22 of 44 (0.5 x 0.21 = 0.105 mrem /hr). This may be accomplished by providing a 4-inch thick lead shielding (or equivalent thickness for other shielding material) to the GeLi detector.

III.

SUMMARY

OF RESULTS AND DISCUSSIONS A. Overall Results

1. The radiation exposures to each Technician during the post-accident sampling activities are:

Radiation Radiation Exposure Technician I.D. (Rem) Time Duration (MIN)

I 0.89 2 II 2.21 3 III 1.28 3.5 IV 0.10 15

2. The extremity dose received by a Technician handling the post-accident noble gas syringe for 30 seconds is approximately 1.2 Rem. The extremity dose to Tech. II taking the liquid sample for 10 seconds is approximately 4.3 Rem.

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l TDR No. 494 Rev. 2 Page 23 of 44

3. With 4-inch thick lead shielding (or equivalent thickness for other shielding material) around the GeLi detector in the Counting Room, the background radiation levels would be reduced to low enough to assure an accuracy within a f actor of 2 in measuring the PC activity.

B. Radiation f rom the Sink Drain Trap and Drain Line When a technician (Tech. II) opens valve CA-V16, and lets it purge sample pipings for 15 seconds, the sink drain line will carry a 0.25 gpm of the primary coolant and becomes highly radioactive. Prior to the opening of valve CA-V16, the technician has opened the demineralized water valve CA-V107 (7.5 gpm) which is lef t open until af ter both liquid and gas samples are taken. The radiation level of the unshielded 5 f t. long and 2 inch diameter drain line is 68.5 Rem /hr. at a 2 f t. distance.

It is assumed that the technician purging the sample line (15 sec.) and taking the liquid sample (27 sec.) is exposed to this unshielded drain line for 42 seconds, resulting in the drain line exposure of. approximately 0.8 Rem, i

In regard to the radiation f rom the sink drain trap, as

't mentioned above, the primary coolant (0.25 gpa) should be diluted by a f actor of approximately 30 because of the high flow rate of the demineralized water (7.5 gpm) into the sink.

TDR No. 494 Rev. 2 Pagejbfof44 Furthermore, the demineralized water will continuously flow even af ter the liquid sample is taken and the valve CA-V16 is closed. The accident fluid in the drain trap would be negligible except some plateout on the inside surf ace of the d rain line.

In the analysis the assumptions used were as follows:

(i) Radiation f rom the 2 f t. long drain trap containing 0.25 gpm of primary coolant and 7.5 gpm of the de-mineralized water is 7.07 Rem /hr. at 5 f t. distance. The Tech. II is exposed to this radiation when he moves the liquid sample bottle to lead pig and move the lead pig to

. the containment barrier.

(ii) Another technician '(Tech. III) taking gas sample is exposed to the shadowed or scattered radiation from the drain trap at 10 f t distance (0.18 Rem /hr.) for two (2) minutes.

C. Radiation from the Undiluted Coolant in the Sink I

When both the primary coolant (0.25 gpm) and demineralized water (7.5 gpm) flow into the sink, both will intermix in the sink at approximate ration of one-to-thirty.

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TDR No. 494 Rev. 2 Page AS~of 44 It was conservatively assumed that the primary coolant will be diluted by a f actor of 10 and forms a film of 2 f t, diameter and 1/32 inch thick (43 mils). The radiation is approximately 8.1 Rem /hr. at 3 f t, distance.

D. Scattered Radiation All the primary coolant carrying pipings in the sample room are shielded by either 2 inch or 4 inch thick lead except short length of pipe upstream and downstream of CA-V16. Radiation level in the general area of the sample room ranges from 15' Rem /hr. to 25 Rem /hr. Dose rates due to the radiation streaming f rom either valve reach rod gaps or sample line pipings behind the 4" thick lead shielding wall ranges f rom 0.046 Rem /hr (reach rod gap) or 0.26 Rem /hr. (f rom behind the 4 inch lead shielding), which are less than 3% of total dose rate in the general area (20 Rem /hr).

E. Residual Contamination during Subsequent Sampling Attempts During subsequent sampling attempts (24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> af ter

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the first sampling), the estimated dose rates are approximately 0.13 Rem /hr. at t = 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, and 0.10 Rem /hr. at t = 48 honra due to the primary coolant film plated out on the surface.

TDR No. 494 Rev. 2 Page Jt6 of 44 F. Airborne Radioactivity Originating f rom the Sink Tech. II purges the sample piping for 15 seconds prior to taking liquid sample. When the primary coolant flows into the sink, gaseous activities contained in the coolant will become airborne and exhausted through the hood. With 0.25 gpm of primary j coolaat flow rate, the total volume of the coolant into the sink I

for the 15 second purging will be approximately 240 cc.

l Assuming gaseous activities, from the primary coolant, are uniformly distributed in the 2 f t. diameter and 5 f t. high cylindrical column, the dose rate at 2 f t. distance would range f rom 43 Rem /hr. to 17.2 Rem /hr. depending upon air exhaust rate through the hood. Radiation exposure f rom this was also considered in the analysis.

l G. Airborne Activity in the Auxiliary Building There are two release paths of the airborne activities into the Auxiliary Building:

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1) Radioactive cloud originating f rom a Reactor Containment  !

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TDR No. 494 Rev. 2 Page 27 of 44

2) Noble gas activities entrained in the primary coolant sample and introduced to the Auxiliary Building through Sink Drain line - Sump Tank - Auxiliary Building release l path.

2 1

The radiation level in the Auxiliary Building due to the l~ airborne activities f rom the containment release (item 1 above) will be approximately 300 mrem /hr whole body at t = 0 hr. and 160 mrem /hr at 't: = 30 minutes after an accident. The radiation level due to the airborne activities of item (2) above would be less than 0.5 mrem /hr, which is insignificant. This is because most of the noble gas activities are released through the Sample Room Exhaust hood before they are introduced to the Auxiliary Building.

4 The Auxiliary Building is not required to be habitable during a

the post-accident period and it is not a leak-tight building.

Should there be a need to access the building af ter an accident

which releases same activities as stipulated in the R.G.1.3 and i R.G.1.4, technicians must wear complete breathing apparatus i

before entering the building.

l l

s v + --- .. ,e-- ~m w - , ,--e

p.

i . .: ..

TDR No. 494 Rev. 2 Page 25 of 44 H. Iodine Plate-Out and Condensation in the Containment Atmosphere Sample Bomb

With existing IMI-1 Containment Atmosphere Sampling (CAS) system piping arrangement in Figure 5, the gas Sample Bomb is located at the low point of the pipings. Heat tracing is not provided either upstream or downstream of the Sample Bomb. Should a reactor accident condition involve the containment air saturated at approximately 175'F, the existing CAS system might not provide the representative samples because of

(1) sample condensation and possibly accumulation of the condensed water in the Sample Bomb, and (2) iodine plate-out in the Sample Bomb due to large I temperature difference.

The iodine removal through plate-out and condensation is a t

function of temperature difference @hT) and sample residence time in the Sample Bomb. The residence time of gas samples in j approximately 6 feet of the non-heat-traced portion of piping is less than 2 seconds. A 1/4 inch diameter piping and' O.25 CFM gas flow rate is assumed.

In order to eliminate the possibility of the Sample Bomb collecting . condensed water which would lead to a drastic lodine -

. removal, the following design modifications are recommended: .

TDR No. 494 Rev. 2 Page J9 of 44 l

l i

(1) Either rerouting the CAS system piping to install its Sample Bomb in a vertical configuration, or removing the Sample Bomb (2) Installation of a small cask at a low point of the piping to collect condensed sater (3) Provide insulation to reduce temperature difference between gas sample and inside surface of the pipe.

With the above modifications, the iodine removal rate in the Sample Bomb would be equivalent to or could be bounded by the removal rate during the Containment Spray to the saturated atmosphere (Reference 5).

In Reference 5, the iodine removal by using the Containment Spray system is extensively analyzed and key assumptions are:

(1) caustic water (pH=9) sprays into the saturated containment atmosphere (180*F), (2) 8 x 10' f t3 /hr spray rate and (3) water droplet diameter of 0.1 cm. These assumptions are either equivalent to or more conservative than our containment air sample flow through the Sample Bomb because of: (1) preferential iodine reaction with the caustic water, (2) saturated air at 180*F, and (3) approximately same temperature difference as the Sample Bomb condition without heat trace p rovided.

TDR No. 494 Rev. 2 Page 30 of 44 According to the results given in Ref. 5 (shown in Fig. 6 of this report), approximately 99% of the iodine activities are 1

removed in about 6 minutes af ter the Containment Spray is initiated. Using this result and the 2-second sample residence time in the non-heat-traced portion of the CAS piping, the 4

iodine loss would be about 0.6%, which is insignificant.

~

By applying an arbitrary conservatism factor of 5, the maximum iodine loss through the Sample Bomb is approximately 3%.

1 The 3% loss of iodine sample would not cause a "non-representa-tive" post-accident sampling. This is because of following reasons:

(1) The noble gas sample taken through the existing CAS system should be " representative" whether the heat trace is provided or not.

(2) The 3% loss of iodine activities is insignificant compared with the uncertainties invclved in the iodine release path from reactor core to the sample line.

The results of the above bounding analysis indicate that the existing TMI-1 CAS system, without additional heat tracing and l

with some piping rearrangements recommended above, is capable of 1

l 4

L_

TDR No. 494 Rev. 2 Page ay of 44 l

taking representative gas sample. The design modifications for  :

l i the CAS system piping mentioned earlier in this subsection were recommended through Ref. 6.

I. Radiation Exposure considerations for taking primary coolant sample f rom Decay Heat Removal System The existing post-accident sampling system is not capable of

< sampling the primary coolant if the accident accompanies a depresserized condition. The sampling will be carried out by

' ~

drawing the primary coolant f rom the decay heat removal system.

Assuming that the sample .is taken f rom the decay heat removal system 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> af ter an accident, the radiation level in the A

sampling room will be approximately 9.8 Rem /hr or less depending on locations, i

i Using the same sampling cs' enario for each technician given in i

Section II-C of this report, the total radiation exposure to the Tech. II would be about 0.15 Rem. The other technicians would i

receive radiation exposure of less than 0.15 Rem.

J. Radiation Exposure During Gas and Sample Analyses Af ter both gas (1 CC) and liquid (30 ml) samples are taken, l

i Tech. III moves the samples in a lead pig to the Chemistry Lab (see page 18 of this report). Another technician will perform i

w -r --- -,,r- -w

TDR No. 494 Rev. 2 Page 32,of 44 sample analyses which consist of: (1) hydrogen analysis for the gas sample, and (2) boron, chloride and gamma scan analysis for the liquid sample. The technician's total radiation exposure during this time will be approximately 1.0 Rem of whole body (WB) dose and 3 Rem to his extremities.

Details of these results are given below:

Radiation Source Exposure Work and Dose Rate per Activity Description Stay Time (Rem /hr) (Rem)

  • Liquid Sample 12 Min.
  • 0.378 WB Analysis (Boron, (1.5 Min. of (5 ft)

Chloride, pH, contact and Gamma exposure)

  • 1 ml of un-
  • 0.28 WB Scan Analysis) shielded source 1.4 (2 ft)
  • 1 m1 source
  • 0.553 contact = Extremity 22.1 (6 in)
  • Gas Sample 10 Mi.n.
  • 0.315 WB (1 Min. (5.ft) contact)
  • Syringe =
  • 0.025 WB

, 0.15 (2 f t)

  • Syringe contact
  • 2.4

= 144 Extremity Total Exposure 0.998 Rem WB and 2.95 Rem -

Extremities

~

W

it TDR No. 494 Rev. 2 Page 33 of 44 V. REFERENCES

1. BNWL-236, UC-34, " User's Manual f or ISOSHLD code", June 1966.
2. GPU TDR No.183, Rev. 3, April 21,1981.
3. GPU Calculation No N1779-5412-005, February 4,1983.
4. TMI-1; EPIP 1004.15, Rev. 4, January 19, 1984.

I f

5. UNI-SA-48,15th DOE Nuclear Air Cleaning Conf erence - Impacts of Sophisticated Fog Spray Models on Accident Analyses.
6. GPU Memo No NF 84-3697, "TMI-1; Iodine Plate-out and Sample Condensation in the Gas Sample Bomb," 3/15/84 and GPU TFWR no. B0061, 5/1/84 i

i e

. . .. .. . zu.. aaa n_.. 7 p 77W490, AW.2

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TABLE :1 SHIELDING SOURCE TERMS (T=0) .

Liquid Gaseous ( ) Containment Source Source Reactor Coolant Airborne Activity Activity Concentration Concentration Isotope (Ci) -

(Ci) (pCi/cc) (pci/cc)

Br-84 7.85 + 6 3.93 + 6 2.43 + 4 6.94 + 1 Kr-83m 9.25 + 6 9.26 + 6 2.87 + 4 1.63 + 7.

Kr-85m 2.19 + 7 2.19 + 7 6.79 + 4 3.87 + 2 Kr-85 5.30 + 5 5.30 + 5 1.64 + 3 9.36 + 0 Kr-87 4.00 + 7 4.00 + 7 1.24 + 5 7.06 + 2 Kr-88 5.60 + 7 5.60 + 7 1.74 + 5- 9.89 + 2 Rb-88 5.64 + 5 - 1.75 + 3 -

Sr-89 7.42 + 5 - 2.30 + 3 -

Sr-90 3.99 + 4 - 1.24 + 2 -

. Sr-91 9.72 + 5 -

3.01 + 3 -

Sr-92 9.50 + 5 -

2.94 + 3 -

Y-90 3.96 + 4 - 1.23 + 2 -

Y-91 9.85 + 5 -

3.05 + 3 -

Mo-99 1.28 + 6 - 3.97 + 3 -

., Ru-106 2.29 + 5 -

7.10 + 2. ' .-

_. Xe-131m 4.38 v 5 4.38 + 5 1.36 + 3 7.73 + 0 Xe-133m -

3.07 + 6 3.07 + 6 9.51 + 3 5.42 + 1 Xe-133 1.27 + 8 1.27 + 8 3.93 + 5 2.24 + 3 Xe-135m 3.26 + 7 3.26 + 7 1.01 + 5 5'.'76 + 2 Xe-135 2.09 + 7 2.09 + 7 6.48 + 4 3.69 + 2 Xe-138 1.17 + 8 1.17 + 8 3.63 + 5 2.07 + 3 I-131 3.68 + 7 1.84 + 7 1.14 + 5 3.25 + 2 1-132 4.31 + 7 2.16 + 7 1.34 + 5 3.81 + 2 I-133 6.40 + 7 3.20 + 7 1.98 + 5 5.65 + 2 1-134 8.00 + 7 4.00 + 7 2.48 + 5 7.06 + 2

. I-135 6.35 + 7 3.18 + 7 1.97 + 5 5.62 + 2 Cs-134 1.27 + 4 -

3.93 + 1 -

Cs-136 8.02 + 3 - 2.48 + 1 -

Cs-137 4.99 4 -

1.55 + 2 -

Cs-138 1.23 + 6 -

3.81 + 3 - -

Ba-137m 4.67 + 4 -

1.45 + 2 -

Ba-140 1.25 + 6 -

3.87 + 3 -

La-140 1.27 + 6 -

3.93 + 3 -

Ce-144 7.50 + 5 -

2.32 + 3 -

Cr-513 - - 5.20 - 3 .

Mn-54

- - 5.80 - 4 -

Mn-561 - - 1.70 - 2 Fe-59 (3) .

- - 5.80 - 4 -

. Co-58 - - 3.00 - 2 Co-60 - - 4.00 - 3 -

Zr-95, - - 5.00 - 4 -

(1) Based on 100% noble gas core inventory, 50% halogen core inventory, and 1%

of all others core inventory.

(2) Based on 100% noble gas core inventory and 25% halogen core inventory. .

t (3) Crud contribution from Ref. 9.

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TDR No. 494 Rev. 2 Pages$l of 44 APPENDIX (A)

DESCRIPTION OF THE ISOSHLD CODE

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RSIC CODE PACKAGE CCC-T9 m

1. NAME AND TITLE OF CODE ISOSHLD: Kernel Integration Code - General Purpose Isotope
Shielding Analysis.

Two versions are packaged: ISOSHLD I and II. RIBD is used in

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both versions as a subroutine to' calculate fission product inventories. The CCC-31/BREMRAD code package can be used to calculate the bremsstrahlung spectrum mesh, but must be requested separately.

2. C0KIRIBUTOR -

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Battelle Memorial Institute, Pacffic Northwest Laboratories, Richland, Wachington.

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3. Cob 1NG LANGUAGE AND COMPUTER

() ISOSHLD III: FORTRAN IV; UNIVAC 1108 (Update 12/73)

(B) ISOSHLD II: FORTRAN IV; IBM 360 (C). ISOSHLD II: FORTRAN IV; UNIVAN 1108 m

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TDR No. 494 Rev. 2 Page f0L of 44 1 4. NATURE OF PROBLEM SOLVED ISOSHLD calculates the decay gamma-ray and brensstrahlung does at 4

the exterior of a shielded radiation source. The source may be one of a number of common geometric shapes. If the radiation source originated as one or a group of fission products produced under known irradiation conditions, then the strength of the source is also calculated. The code calculates shield region mass attenuation i coefficients, buildup f actors, and other basic data necessary to solve the specific problem.

5. METHOD OF SOLUTION j ISOSHLD performs kernel integration for common geometric shapes.

l The " standard" point attenuation kernel (buildup f actor x expo-t nential attenuation + geometry f actor) is numerically integtated -

f over the source volume for 25' source energy groups. Buildup is 1

considered characteristic of the last shield region-(or a difference specified region) but dependent on the total number of mean f ree paths from source to dose point, and is obtained by interpolation on i effective atomic number f rom atable of point isotopic buildup f actor l data. Mixed mass attenuation coefficients are obtained f rom a i

i library of basic data using code input material density

, specifications. The source strength may be specified 1) as the l

emissions f rom a selection of fission products irradiated under specific conditions, 2) the curies of particular fission and/or j

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bb, i TDR No. 494 l Rev. 2 t l

Page 1R3 of 44 i

activation products, or 3) a number of photons per second of energy

! E specified by input. An exponential source distribution may be i

specified for those source geometries which are applicable. If the source originates in a combination of fission products and their daughters, these are calculated by a fission product inventory procedure which runs through transmutation (decay chain) calculations for each fission product and daughter. The latest modification (ISOSHLD - ISOSHLD II) adds the capability for

! calculating shielded dose rates f rom brensstrahlung sources. This o

addition consists of routines for calculating the brenstrahlung source spectra f rom the beta decay properties of the isotope (s) of

, inte re st . Brenstrahlung photons per group for 25 energy groups (9 j groups below 0.1 MeV have been added) are obtained by interpolation i

from tables of resolved spectra. This spectral mesh, for internal and external bremsstrahlung, is tabulated as a function of the following parameters: beta emitting and stopping nuclides with i

. atomic numbers of 10, 30, 50,' 70, and 90; ratios of photon energy to .

beta end point energy for 25 intervals from 0.00375 to 1.0; beta and point energies at athe intervals 0.1, 0.2, 0.5, 1, 2, and 4 MeV.

Buildup factors for photon energies less than 0.1 MeV are interpolated f rom a table which contains data for 5 values of -

) initial photon energy in the range.0.01 to 0.2 MeV, seven values of

-shield thickness in the range 1 to 20 p, and 6 atomic numbers in the range 13 to 92.

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i-The entire shielding problem is solved for most types of isotope 1

! shielding applications without reference to shielding handbooks for

~. 1 basic data.- l

f k TDR No. 494 l

Rev. 2 Pagepp6of44

6. RESTRICTIONS OR LIMITATIONS These limits apply: 5 source cooling t'mes, 500 radioactive isotopes, 5 shield regions including source region, 25 energy groups, 20 materials in each shield region, choice of 11 source geometries.
7. REFERENCES R. L. Engle, J. Greenborg, and M. M. Hendrickson, "ISOSHLD - A Computer Code for General Purpose Isotope Shielding Analysis",

BNWL-236 (June 1966).

R. O. Gumprecht, "RIBD-Radioisotope Buildup and Decay", Unpublished data.

H. H. Van Tuyl, "BRENRAD - A Computer Code f or External and Internal Brensstrahlung Calculations, HW-83784 (September 1964) (Packaged in CCC-31 only).

G. L. Simmons, J. J. Regimbal, J. Greenborg, E. L. Kelly, Jr. , and H. H. Van Tuyl, "ISOSHLD-II: Code Revision to Include Calculations of Dose Rate from Shielded Brensstrahlung Sources", BNWL-236SUPL (March 1967). ,

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