ML18031B269

Forwards Info Addressing Criterion 6 of NUREG-0737,Item II.B.3 Re post-accident Sampling Sys.Total Mission Doses for Sampling & Analysis of post-accident Fluids Estimated to Be 1.3 Rem Whole Body & 11.4 Rem Extremities to Individual

ML18031B269
Person / Time
Site:	Browns Ferry
Issue date:	04/01/1987
From:	Gridley R TENNESSEE VALLEY AUTHORITY
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
RTR-NUREG-0737, RTR-NUREG-737, TASK-2.B.3, TASK-TM NUDOCS 8704070480
Download: ML18031B269 (15)

Text

C REGULATO f INFORMATION DISTRIBUTION YSTEM (RIDS)

ACCESSION NBR:8704070480 DOC ~ DATE! 87/04/01 NOTARIZED'O DOCKET FACIL:50-259 Browns Ferry Nuclear>> Power, Stations Unit 1> Tennessee 05000259 50 260 Br owns Ferry Nuclear>> Power Stations Unit 2i Tennessee 05000260 50-296 Browns Ferry Nuclear Power. Stations Unit 3< Tennessee 05000296 AUTH,NAME>>

AUTHOR AFFILIATION GRIDLEYcR+

Tennessee Valley Authority RECIP ~ NAMEl RECIPIENT AFFILIATION Document Control Branch (Document Contr ol Desk)

SUBJECT:

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TENNESSEE VALLEYAUTHORITY 5N 157 Lookout Place AIR 0 I >Ssg U.S. Nuclear Regulatory Commission ATTN:

Document Control Desk Washington, D.C.

20555 Gentlemen:

In the Matter of Tennessee Valley Authority Docket Nos. 50-259 50-260 50-296 BROWNS FERRY NUCLEAR PLANT NUREG-0737, ITEM II.B.3 POSTACCIDENT SAMPLING SYSTEM In reference to our letter dated December 19,

1986, concerning information submitted to the NRC addressing NUREG-0737, Item II.B.3 criteria for the redesigned BFN Postaccident Sampling System (PASS),

we stated that the response to Criterion 6 of NUREG-0737, Item II.B.3 would be submitted by March 31, 1987.

Accordingly, enclosed is our submittal addressing Criterion 6.

If additional information is required, please get in touch with T. Roland Phillips at (205) 729-2858.

Very truly yours, TENNESSEE VALLEY AUTHORITY R. Gridley Director Nuclear Sa ety and Licensing Enclosure cc:

See page 2

8704070480 870PQj PDR ADOCK 0$0p0pg9 P

PDR An Equal Opportunity Employer

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U.S. Nuclear Regulatory Commission cc (Enclosure):

Mr. G.

G. Zech, Assistant Director Regional Inspections Division of TVA Projects Office of Special Projects U.S. Nuclear Regulatory Commission 101 Marietta St.,

NW, Suite 2900 Atlanta, Georgia 30323 Browns Ferry Resident Inspector Browns Ferry Nuclear Plant P.O.

Box 311

Athens, Alabama 35611

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ENCLOSURE BROWSES FERRY. RESPONSE IN REFERENCE TO NUREG 0737$

ITEM II. B.37 CRITERION 6 (POST ACCIDENT SAMPLING)

Criterion:

(6)

L The design basis for plant equipment for reactor coolant and containment atmosphere sampling and analysis must assume that it is possible to obtain and analyze a sample without radiation exposures to any individual exceeding the criteria of GDC 19 (Appendix A, 10 CFR Part

50) (i.e.,

5 rem whole

body, 75 rem extremities).

(Note that the design and operational review criterion was changed from the operational limits of 10 CFR Part 20 Clari fication:

(HUREG-0578) to the GDC 19 criterion (October 30, 1979 letter from H.

R.

Denton to all licensees).

Consistent with Regulatory Guide 1.3 or 1.4 source terms, provide information on the predicted personnel exposures based on person-motion for sampling, transport and analysis of all required parameters.

~Res oose The Postaccident Sampling System (PASS) is designed to permit sampling operations following an accident with radioactive releases of the magnitude of a design basis loss of coolant accident.

The laboratory analysts are required to go to the sample

station, take gas and liquid samples, place the samples in casks, transport the

, et samples to the laboratory and complete the required analyses.

Dose calculations were made to obtain reasonable estimates of the expected dose received by the laboratory analysts while obtaining, transporting, and analyzing the samples.

The times required to carry out the sampling and analysis are those listed in the response to criterion l.

Radiation calculations consistent with Regulatory Guide 1.3 source terms have been made to determine the dose rate conditions for I

each of the identified time steps.

The exposure to personnel comes from shine through the secondary containment wall and from the samples during sampling, transport, and analysis.

All dose rates are for one hour after reactor shutdown.

The dose rates in the turbine building from shine through the secondary containment wall were determined using HICROSKIELD.

The assumptions for the calculations were a containment atmosphere source term of 100 percent noble gases and 50 percent iodine, and a drywell leak rate of 2 percent per day.

To determine the dose rates from the postaccident

samples, the core radionuclide inventory after three years of operation was calculated (Tables 1, 2, and 3) and the following release terms were used:

Sam le Containment Atmosphere Depressurized

Coolant, Dissolved Gas Hob'le Gas 100 Kalo en X

25 50 Others A

Using these release

terms, General Electric has calculated the sample and sample panel dose rates that are us'ed to determine mission doses.

The total mission doses for the sampling and analysis of postaccident fluids is estimated to be 1.3 Rem whole body and 11.4 Rem extremities to an individual.

This total dose may be distributed among several laboratory analysts and, therefore, radiation exposures to any individual will not exceed 5 rem whole body and 75 rem extremities.

CORE IHYEHTORY OF HALOGEHS

'CINDER 8/18/80

{ 0.5-COnTRIBVTORS)

KILO-CVRIES PER N<t DECAY PEPIOD:

SHUTDOHH I

1.0 HR 8.0 HR 24.0 HR 35.4h G. lm 2.40h 31.8 m

6.0 m

2.87m 55.

s 4.5 s

55.7 s

15.9 s

4.5 s

1.6 s

0.6 s

0.3 s

Br-82 Br-82M Br-83 Br-84 Br-84M Br-85 Br-86 Br-86M Br-87 Br-88 Br-89 Br-90 Br-91 BI-92 TOTAl BROMIHES 0.25 0.22 3.12 5.54 0.20 6.64 4.79 4.82 11.22 12.07 8.61 5.54 1.82 0.20 65.04 0.25 2.49

1. 67

4. 41 0.21 0.33 0.54 0.16 0.16 12.4 h

8.9 m

8. 04(j 2.29h 20.8 h

9.

s 52.6 m

3.6 m

6.56h 85.

s 46.

24.6 s

6.5 s

2.4 s

0.86s 0.4 s

TOTAL 1-130 I-130M 1-131 1-132 1-133 I-133M 1-134 I-134M 1-135 1-136 I-136M 1-137 1-138 1-139 1-140 1-141 IODINES

l. 05 0.77 26.31 38.45 55.02 1.59 60.56 5.61 51.95 25.05 14.31 24.87 12.69 5.88 1.80 0.32 326.23 1.00

26. 28 38.27 54.00 42.52 46.77 208.84

0. 67 25.74 36.33 43.12 0.37 22.38 128.61 0.28 24.45

31. 56 25.30 4.15 85.74

CORE IHYEHTORY OF tlOBLE GASES CItiDER 8/18/80

('>0.5K COhTRIBUTIOH)

KILO-CURIES PER Aft DiCAY PERIOD:

SHUTDOlfN 1.0 flR 8.0 HR 24.0 HR 1.86h 10.73y 4.48h 76.

m

2. Boh

3. 16m 32 3 s 9.0 s

1.84s 1.27s 0.21s 0.5 s

Kr-83H Kr-85 V,r-85H Kr-87 Kr-88 Kr-89 Kr-90 V,r-91 Kr-92 V,r-93 Kr-94 Kr-95

11. 99d 5.29d 2.23d 0.29s 9.17h 15.3 m

3. 84m 14.2 m

39.7 s

13.6 s

1.72s 1.22s "0.30s l.

s Xe-131H Xe-133 Xe-133M Xe-134H Xe-135 Xe-1 35H Xe-137 Xe-138 Xe-139 Xe-140 Xe-141 Xe-142 Xe-143 Xe-144 TOTAL XENOH TOTAL VRYPTON 3.14 0.30 6.73 12.92 18.30 22.76 22.90 17.19 0.73 3.39 1.23 0.13 117.72 0.16 55.28 2.30 0.40 7.15 10.42 48.52 46.10 36.24 24.04 8.76 3.30 0.59 0.12 243.38

3. 03 0.30 5.84 7.56 14.30 31.03 0.16 55.27 2.30

10. 29 7.32 2.47 77.81 0.88 0.30

l. 98

0. 16 2.53 5.85 0.16 55.05 2.27 19.30 3.43 80.00 0.01 0.30 0.17 0.05 0.53 0.16 53.33

'.09 12.22

0. 64 68.44

CORE. ItNEtfTORY OF ALUMI tKTALS CIttDER 8/18/80

(>0.5X COttTRIBUTIOH)

KILO-CURIES PER tIHt DECAY PERIOD:

SHUTDOWN 1.0 HR 8.0 tlR 24.0 HR

1. 02m 17.7 m

15.2 m

2. / m

4. 28m 58.5 s

4.53s 5.8 s 2.69s 0.36s

0. 21s 0.17s Rb-86 Rb-88 Rb-89

,Rb-90 Rb-90M Rb-91 Rb-92 Rb-93 Rb-94 Rb-95 Rb-96 Rb-97 TOTAL RUBI DIUM 0.05 18.59 24.22 23.16 6'0 29.70 25.96 20.21 10.38 5.09 51 0.23 165.50 0.05 15.80 1.96

17. Bl 0.05 2.83 2.88 O. 045 0.053 0.098

2. 06y 2.90h 13.0 d

30.1 y 32.2 m

2.9 m

9.3 m

63.8 s

25.0 s

1.7 s 1.7 s

1.02s 0.56s 0.19s Cs-134 Cs-134M Cs-136 Cs-137 Cs-138 Cs-138M Cs-139 Cs-140 Cs-141 Cs-142 Cs-143 Cs-144 Cs-145

~

Cs-146 5.36 0.98 1.18 3.32 48.64 2.32 47.90 43.25

31. 89 19.77 10.01 3.06 0.81 0.13 5.36 0.77 1.18

3. 32 21.43

0. 58 5.35 0.15 1.16 3.32 5.35 1.12 3.32 TOTAL DESI U(3 218. 62 32.64 9.98 9.79

TABLE 4 MISSION DOSES FOR POSTACCIOEN1 SAMPLIHG De assed Li uid Sam le (10 ml)

Sample Station Mork Transport

.Ana lysi s Total Li uid Grab Sam les (0,2 ml and 100:1 diluted)

Sample Station Mork Transport Analysis Total Dissolved Gas Sam le (5 cc)

Sample Station Mork Transport Analysis Total Containment Atmos here Sam le (14 cc)

Sample Station Mork Transport Analysis Total Who~le Bod 191 mr 27 ml 218 mr 382 mr 34 mr 106 mr 522 mr 307 mr 17 mr 69 mr 393 mr 125 mr 17 ml 69 mr 211 mr Extremi ties 192 mr 27 ml 219 mr 385 mr 34 mr 964 mr 1383 mr 3820 mr 17 mr 2862 mr 6699 mr 150 mr 17 mf 2960 mr 3127 mr Total Postaccident Sampling and Analysis Dose:

1344 mr 11428 mr

• Analysis for chloride required within 30 days (Refer to clarification of criterion 5).

0016m

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