ML20246P113

Revised, Effluent & Waste Disposal Semiannual Rept for First Half 1985

ML20246P113
Person / Time
Site:	Prairie Island
Issue date:	06/30/1985
From:	NORTHERN STATES POWER CO.
To:
Shared Package
ML20246K241	List: ML20246K236 ML20246P061 ML20246P074 ML20246P098 ML20246P109 ML20246P113 ML20246P131 ML20246Q233
References
NUDOCS 8909110045
Download: ML20246P113 (27)

Text

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l AMENOMENT to the PRAIRIE ISLAND Effluent and Waste Disposal Semiannual Report (for the 1st Half year of 1985)

This report is being resubmitted to include an abnormal release for this period resulting from an inadvertent release of some gas from Unit 2 Volume Control Tank.

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K2 19852 Effluent Sc'minnnual Report REV. 1

'Page11'of.9 Retention: Lifetime L'

p EFFLUENT SEMIANNUAL REPORT v

01-JAN-85 THROUGH 30-JUN-85' t

SUPPLEMENTAL:INFORMATION Facility:

Prairie Island Nuclear Generating Plant Licensee:

Northern States Power Company

. License Numbers: DPR-42 & DPR-60 A. Regulatory Limits 1.

Liquid Effluents:.

a.

The~ dose or dose commitment to an individua3 from radioactive materials in liquid-effluents-released from the site shall be limited to:

for'the quarter 3.0 arem to the total body 10.0 mrem to any organ for the year 6.0 mrem to the total body 20.0 mrem to.any organ 2.

. Gaseous Effluents:

a.

The dose rate due to radioactive materials released in gaseous effluents from the site shall be limited to:

noble gases 4500 mrem / year total body (3000 arem/ year skin-I-131, H-3, LLP (1500 mrem / year to any organ b.

The dose due to radioactive gaseous effluents shall be limited to:

noble gases (10 mrad / quarter gamma (20 mrad / quarter beta (20 mrad / year gamma 640 mrad / year beta I-131, H-3, LLP (15 mrad / quarter (30 mrad / year

_ _ _ _ _ -_ _ D

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[gf 1985: ETFLUENT SEMIANNUAL REPORT-REV. ' 1 PAGE 2 B.

Maximum Permissible Concentration L

1.

rission and activation gases in gaseous releases:

-10.CFR=20. Appendix B, Table 2, Column 1 2.

Iodine'and particulate with halflives greater than 8 days in gaseous releases:

10 CFR 20, Appendix B, Table 2, Column 1 3.

Liquid' effluents for radionuclides other than dissolved or entrained gases:

10 CFR 20,. Appendix B, Table 2, Column 2 4.

Liquid' effluent dissolved and entrained' gases:

2.0E-04 uCi/ml Total Activity

(

.C.

Average Energy Not applicable to Prairie Island regulatory limits.

D.

Measurements and approximations of total activity 1.

rission and activation gases Total GeLi 25%

in gaseous releases:

N':clide GeLi 2.

Iodines in gaseous releases:

Total GeLi-125%

Nuclide GeLi 3.

Particulate in gaseous releases:

Total GeLi i25%

Nuclide GeLi 4.

Liquid effluents Total GeLi 251 Nuclide GeLi

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1985LEFFLUENT~ SEMIANNUAL ~ REPORT.

REV.'1" PAGE 3' i

11.0. BATCH RELEASES'(LIQUID)

QTR: 01 QTR: '02 1.1' ' NUMBER.0F BATCH-RELEASES 7.00E+01 4.40E+01

.1. 2 TOTAL TIME-PERIOD (HRS) 1.14E+02 6.83E+01

1. 3 : MAXIMUM TIME 1 PERIOD- (HRS) 2.91E+00~

2.82E+00 1.4 AvthAGE TIME PERIOD (HRS) 1.63E+00 1.55E+00 1.5 MINIMUM' TIME PERIOD -(HRS) 1.17E+00 1.17E+00 1.6 AVERAGE MISSISSIPPI RIVER FLOW (CFS) 1.86E+04 3.97E+04 it 2.0 BATCH RELEASES'(GASEOUS)

QTR: 01 QTR: 02 2.1 NUMBER OF' BATCH RELEASES 8.00E+00 1.00E+00 2.2 TOTAL TIME. PERIOD (HRS) 1.45E+02 4.64E-01 2.3 MAXIMUM TIME PERIOD (HRS) 3.68E+01 4 64E-01 2.4 AVERAGE TIME PERIOD (HRS) 1.81E+01 4.64E-01 2.5 MINIMUM TIME PERIOD (HRS) 8.43E-02 4.64E-01 3.0 ABNORMAL RELEASES (LIQUID)

QTR: 01 QTR:'02 3.1 NUMBER OF BATCH RELEASES 0.00E+00 0.00E+00 3.2 TOTAL ACTIVITY RELEASED (CI) 0.00E+00 0.00E+00 3.3 TOTAL TRITIUM RELEASED (CI) 0.00E+00 0.00E+00 4.0 ABNORMAL RELEASES (GASEOUS)

QTR: 01 QTR: 02 4.1, NUMBER OF BATCH RELEASES 1.00E+00 0.00E+00 4.2 TOTAL ACTIVITY RELEASED (CI) 2.49E-02 0.00E+00 SEE ATTACHMENT #1 l

1985 EFFLUENT SEMIANNUAL REPORT REV. 1 PAGE 4 TABLE 1A GASEOUS. EFFLUENTS - SUMMATION'OF ALL RELEASES l

QTR: 01 QTR: 02 5.0 FISSION AND ACTIVATION GASES 5.1 TOTAL RELEASE (CI) 1.25E+01 4.67E+00 5.2 AVERAGE RELEASE RATE (UCI/SEC) 1.59E+00 5.94E-01 5.3 GAMMA DOSE (MRAD) 7.85E-03 1.37E-03 5.4 BETA DOSE (MRAD) 2.25E-02 6.32E-03 5.5 PERCENT OF GAMMA TECH SPEC (%)

7.85E-02 1.37E-02 5.6 PERCENT OF BETA TECH SPEC (%)

1.13E-01 3.16E-02 6.0 IODINES 6.1 TOTAL I-131 (CI) 3.43E-03 1.80E-06 6.2 AVERAGE RELEASE RATE (UCI/SEC) 4.36E-04 2.29E-07 7.0 PARTICULATE 7.1 TOTAL RELEASE (CI) 1.08E-05 0.00E+00 7.2 AVERAGE RELEASE RATE (UCI/SEC) 1.37E-06 0.00E+00 8.0 TRITIUM 8.1 TOTAL RELEASE (CI) 1.17E+01 1.47E+01 8.2 AVERAGE RELEASE RATE (UCI/SEC) 1.49E+00 1.87E+00 9.0 TOTAL IODINE, PARTICULATE AND TRITIUM (UCI/SEC) 1.49E+00 1.87E+00 10.0 DOSE (MREM) 2.24E-01 2.67E-02 11.0 PERCENT OF TECH SPEC (%)

1.49E+00 1.78E-01 12.0 GROSS ALPHA (CI) 0.00E+00 9.82E-07

4 '

4 1985 EFFLUENT SEMIANNUAL REPORT REV. 1 PAGE 5 TABLE 1C GASEOUS EFFLUENTS - GROUND LEVEL RELEASES

- 13.0 FISSION AND ACTIVATION GASES

/

CONTINUOUS MODE BATCH MODE NUCLIDE UNITS QTR: 01 QTR: 02 QTR: 01 QTR: 02 AR-41 CI 4.49E-03 KR-85 CI 1.12E+00 KR-85M CI 8.04E-04 KR-87 CI 2.96E-04 KR-88 CI 1.32E-03 XE-133 CI 9.91E+00 3.55E+00 2.54E+00 XE-135 CI 1.15E-02 XE-135M CI 1.60E-04 TOTAL CI 9.91E+00 4.67E+00 2.56E+00 0.00E+00 14.0 IODINES CONTINUOUS MODE BATCH MODE l

NUCLIDE UNITS QTR: 01 QTR: 02 QTR: 01 QTR: 02 I-131 CI 3.43E-03 1.80E-06 I-133 CI 5.24E-05 TOTAL CI 3.48E-03 1.80E-06 0.00E+00 0.00E+00 l

1 l

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1985 EFFLUENT SEMIANNUAL REPORT REV. 1 PAGE 6 TABLE 1C (CONTINUED)

GASEOUS EFFLUENTS - GROUND LEVEL RELEASES 15.0 PARTICULATE CONTINUOUS MODE BATCH MODE l

NUCLIDE UNITS QTR: 01 QTR: 02 QTR: 01 QTR: 02 CS-134 CI 1.28E-06 CS-137 CI 1.38E-06 i

CO-58 CI 4.80E-06 CO-60 CI 1.27E-06 GA-72 CI 8.06E-07 MN-54 CI 7.61E-07 SE-75 CI 5.00E-07 TOTAL CI 5.00E-07 0.00E+00 1.03E-05 0.00E+00 1

l 4

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1985 EFFLUENTfSEMIANNUAL REPORT REV. 1 PAGE 7 TABLE 2A LIQUID EFFLUENTS - SUMMATION OF ALL RELEASES QTR: 01 QTR: 02 16.0 VOLUME OF WASTE PRIOR TO DILUTION (LITERS) 6.43E+07 5.39E+07 17.0 VOLUME OF DILUTION WATER (LITERS) 1.03E+11 8.14E+10 18.0 FISSION AND ACTIVATION PRODUCTS 18.1 TOTAL RELEASE W/O H-3, RADGAS, ALPHA (CI) 6.20E-03 2.19E-03 18.2 AVERAGE DILUTED CONCENTRATION (UCI/ML) 6.02E-11 2.69E-11 19.0 TRITIUM 19.1 TOTAL RELEASE (CI) 1.48E+02 1.79E+02 19.2 AVERAGE DILUTED CONCENTRATION (UCI/ML) 1.44E-06 2.20E-06 20.0 DISSOLVED AND ENTRAINED GASES 20.1 TOTAL RELEASE (CI) 4.46E-03 1.84E-03 20.2 AVERAGE DILUTED CONCENTRATION (UCI/ML) 4.33E-11 2.26E-11 21.0 GROSS ALPHA (CI) 0.00E+00 0.00E+00 22.0 TOTAL TRITIUM, FISSION AND ACTIVATION PRODUCTS (UCI/ML) 1.46E-06 2.20E-06 23.0 TOTAL BODY DOSE (MREM) 1.21E-03 4.34E-04 24.0 CRITICAL ORGAN 24.1 DOSE (MREM) 1.62E-03 4.50E-04 24.2 ORGAN LIVER LIVER 25.0 PERCENT OF TOTAL BODY TECH SPEC LIMIT (%)

4.03E-02 1.45E-02 26.0 PERCENT OF CRITICAL ORGAN TECH SPEC LIMIT (%)

1.62E-02 4.50E-03

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if' 1985fEFFLUENT' SEMIANNUAL REPORT

'REV. 1 PAGE 8 TABLE'2A LLIQUID. EFFLUENTS - SUMMATION OF ALL RELEASES 27.0; IND'IVIDUAL' LIQUID EFFLUENT CONTINUOUS MODE BATCH MODE NUCLIDE UNITS QTR: 01 QTR: 02 QTR: 01 QTR: 02 AG-110M-CI.

9.03E-05 8.62E-04 CS-134 CI 2.84E-05 6.10E-07 CS-136 CI 5.17E CS-137 CI 1.90E-04 2.50E-05 CO-58 CI 8.62E-04 1.59E-04 1.73E-03 1.65E-04 CO-60 CI 4.48E-04 3.51E-03 3.29E-04 GA-72 CI 8.83E-05 MN-54 CI 8.34E-05 6.76E-06 NB-97 CI 7.10E-07 SB-124 CI 1.43E-03 1.20E-04 l-SB-125 CI 3.75E-04 4.61E-04 1.90E-04 SC-47 CI 1.50E-06 SR-92 CI 2.66E-04 ZR-NB-95 CI 4.07E-05 ZR-97 CI-3.66E-05 TOTAL CI 1.88E-03 1.59E-04 4.32E-03 2.03E-03 1

28.0 DISSOLVED AND ENTRAINED GASES CONTINUOUS MODE BATCH MODE NUCLIDE UNITS QTR: 01 QTR: 02 QTR: 01 QTR: 02 ER-85 CI 9.64E-04 XE-133 CI 1.71E-04 3.30E-03 1.84E-03 XE-135 CI 2.99E-05 3.18E-06 TOTAL CI 1.71E-04 0.00E+00 4.29E-03 1.84E-03 L_____-_____-_-______--_-____.

... ~ ~,

of1955IEFFLUENT;SEMIANNUALREPORT REV. 1 PAGE 9

' ATTACHMENT #1 i

-ABNORMAL GASEOUS RELEASE - SUMMATION OF ALL ACTIVITY.

CONTINUOUS MODE BATCH MODE-

~'

NUCLIDE UNITS QTR: 01 QTR: 02 QTR: 01 QTR: 02 AR-41

'CI-1.69E+03 KR-85M CI 8.04E+02 KR-87 CI 2.96E+02 KR-88 CI 1.32E+03 XE-133 CI 9.17E+03 XE-135 CI 1.15E+04 XE-135M CI 1.60E+02 TOTAL CI 0.00E+00 0.00E+00 2.94E+04 0.00E+00 l

1 l

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TPINGP 7.53','Rev. 1

f 1

Retention:

Lifetime

-Page 1-of 5 i

-PRAIRIE ISLAND NUCLEAR GENERATING PLANT Period:

1-1-85/7-1-85 NORTHERN STATES POWER.

License No. DPR-42 SOLID RADIOACTIVE WASTE DISPOSAL SEMI-ANNUAL REPORT Table I:- Solid Waste and Ir' radiated Fuel-Shipments -

A.

Solid Vaste Total Volumes and Measured Curie Quantities:

1.

Type'of Waste:

Container

_ Units Total Volumes-A.

Resins Ft3 818 170/3 @ 76 Ci 388.254 1 @ 80 l -',

B.-

Dry Compacted Ft3 2577 7.5 Ci 2.106 C.

Non Compacted Ft3 1862 98 Ci

.1 D.

Wst. Concentrates Ft3 493 7.5 Ci 8.796 L

4 4

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PINGP 753, Rev. 1 Retention:

Lifetime Page 2 of 5 t t, PRAIRIE ISLAND NUCLEAR GENERATING PLANT Period: 1-1-85/7-1-85 NORTHERN STATES POWER ~

License No. DPR-42 SOLID MDIOACTIVE WASTE DISPOSAL SDil-ANNUAL REPORT Table 1:

Solid Waste and Irradiated Fuel Shipments (Continued) 2.

Measured Major Nuclide Composition by Type of Waste:

TYPE (From Page 1)

Nuclide Percent A

Co6o 60 Ni63 26.2 Csl37 5.1 Co38 4.1

,Cs134 3.0 Fe55 23.8 B

Ni63 16.0 Co58 18.1 Co60 13.7 Cs137 12.3 Csl34 9.5 NB95 1.7 Mn54 1.6 Zr95 1.1

• = Inferred - Not Measured on Site

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PINGP 753, Rev. 1 l

Retention:

Lifetime l

[

Page 3 of 5 PRAIRIE ISLAND NUCLEAR GENERATING PLANT Period: 1-1-85/7-1-85' NORTHERN STATES POWER License No. DPR-42 SOLID RADIOACTIVE WASTE DISPOSAL SEMI-ANNUAL REPORT Table 1:

Solid Waste and Irradiated Fuel Shipments (Continued) 2.

Measured Major Nuclid's Composition by Type of Waste (Continuation):

TYPE (From Page 1)

Nuclide Percent C

Fe55 41.2 Ni63 22.9 Co60 11.4 Co58 8.2 Zr95 3.4 NB95 5.4 Mn54 1.7 Agl10 1.2 Cr51 1.2 D

Fe55 25.0 Co60 19.2 Ni63 16.9 Cs137 14.9 Cs134 12.4 CoS8 6.1 Agl10 3.2 Mn54 2.2

• = Inferred - Not Measured on Site

+

PINGP 753', Rev. 1

~

Retention:

Lifetime

..Page 4 of 5 4

PRAIRIE ISLAND NUCLEAR GENERATING PLANT Period: 1-1-85/7-1-85 NORTHERN STATES POWER License No. DPR-42 SOLID RADIOACTIVE WASTE DISPOSAL SEMI-ANNUAL REPORT Table I:

Solid Waste and Irradiated Fuel Shipments (Continued) 3.

Solid Waste Disposition:

Number of Shipments Mode Destination 9

Truck Richland 3

Truck Barnwell B.

Irradiated Fuel Shipments:

Number of Shipments Mode Destination None 1.

)

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.4 PINGP 753, Rev. 1 g

Retention:

Lifetime

' Page 5 of 5 l

.g 1

i l'

PRAIRIE ISLAND NUCLEAR GENERATING PLANT Period: 1-1-85/7-1-85 NORTHERN' STATES-POWER License No. DPR-42 I

SOLID RADIOACTIVE WASTE DISPOSAL SEMI-ANNUAL REPORT Table I:

Solid Waste and Irradiated Fuel Shipments (Continued)

C.

Shipping Container and Solidification Method:

No.

Volume Activity Type of Container Solidify (Ft3)

(Ci)

Waste Code Code 85-01 1078 0.069 C

L 85-12 630 0.448 B

L 85-12 98 0.002 C

L 85-14 76 126.563 A

B 85-13 76 128.545 A

B 85-15 420 0.701 B

L 85-15 210 0.701 D

L C

85-15 98 0.001 C

L 85-17 270 0.098 B

L 85-17 196 0.020 C

L 85-17 178 0.905 D

L C

85-17 80 0.315 A

L 85-20 170 0.793 A

A 85-21 196 0.006 C

L 85-21 893 0.632 B

L 85-22 105 7.190 D

L C

85-23 170 9.750 A

A 85-24 170 0.188 A

L 85-24 196 0.002 C

L 85-24 364 0.227 B

L 85-19 76 122.100 A

B CONTAINER CODES:

L = LSA A = Type A B = Type B Q = Large Quantity SOLIDIFICATION CODES:

C = Cement TYPES OF WASTE:

A = Resins B = Dry Compacted C = Non-Compacted D = Wst. Conc.

S = Spent Fuel

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Page Revisions to Effluent and Waste Disposal Semi-Annual Report for for January 1.1985 throuch June 30, 1985 The Semi-Annual Effluent and Waste Report for January 1,1985 through June 30, 1985 has been revised to incorporate Fe-55 analysis results that arrived from our contractor too late to be incorporated in the report submitted on August 30, 1985.

Please replace pages 6, 7, 8, and Attachment 1 of the subject report with the attached revised pages.

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rimUt' 133, nev. O Revised 10-1-85 P:g2 6'cf 8 1

I TABLE 2A

• 3,,.

EFFLUENT SEMIANNUAL REPORT LIQUID EFFLUENTS - SUMMATION OF ALL RELEASES I

I I

I I

TOTAL l

QTR__1, QU 2-WIT ERROR".l 16.0 VOLUME OF WASTE (Prior to Dilution) l Liters 1 6.43ro7 Ic - torn 7 l9 mnen1 l

17.0 VOLUME OF DILUTION Water (liters) l, Liters i 1.03E11 l 8.14E10 12.50E01 l

18.0 FISSION AND ACTIVATION PRODUCTS

,e 18.1 Total Release W/0 H-3, (Ci) l l

l l

!.*[y Rad Gas, Alpha l 6.20E-03 l 4. oRv-0 t 12.50E01 l

18.2 Average Diluted (pCi/ml) l l

l Concentration l 6.02E-11 Is.?6v-11 l

19.0 TRITIUM 19 1 Total Release (Ci) l 1.48E02 -

1 1.79E02 12.50E01 l

19.2 Average Diluted (pci/ml)

I i

l Concentration l 1.44E-06

2.20E-06 l

20.0 DISSOLVED AND ENTPAINED GASES 20.1 Total Release (Ci) l 4.46E-03 1 1.8cE-03 12.50E01 l

20.2 Average Diluted (pCi/ml) l J

l Concentration l 4.33E-11 l 2.26E-11 l

21.0 GROSS ALPHA 1

21.1 Total Release (Ci) 1 0.00E00 1 0.00E00 12.50E01 l

?.2.0 TOTAL TRITIL'M. FISSION /ED ACTIVATION PRODUCTS (pCi/ml) l:.--E-06 i 2.20E-06 f2.50E01 l

I I

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__ _ _ _ ____ A

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3 EINCP 150, R:7. 6 Revissd 10-1-85 Pega 7 of 8 i

NBLE 2A 6;

EFFLUENT SEMIANNUAL REPORT LIQUID EFFLUENTS - SUMMATION OF ALL RELEASES UNIT l

QTR 1 l

QTR 2 l

1 I

I I

I 23.0 TOTAL BODY DOSE (mrem)

I 1 91r n1 i a_ur_na l

24.0 CRITICAL ORGAN DOSE (mrem) l 1.62E-03 4.59E-04 l

(organ) l T $ver t4ver l

25.0 % TOTAL BODY T.S. LIMIT (%)

l 4.03r-02 1 1.457-02 l

P.

$'fj6/

26.0 % OF CRITICAL ORGAN

(%)

l l

l 1 4.'9E-03 l

T.S. Limit l 1.62E-02 5

4 I

PINGP 158, Rev. 6 Page 8 of 8

' h2 TABLE 2A EFFLUENT SEM1 ANNUAL RPEORT LIQUID EFFLUENTS - SUMMATION OF ALL RELEASES CONTINUOUS MODE BATCH MODE I

i i

l i

l l

l NUCLIDE j UNIT l

QTR 1 l

QTR 9 l

QTR 3 l

QTR_2-I J

l l

l 1

I l

27.0 INDIVIDUAL LIOUID FIFILINT 4

1 Sr89 L Ci i

i i

I l

l Sr90 Ci I

i l

l l Cs134 Ci 1

! 2.g4r-Os ll 6.10r-07 l

I 2.50E-05 I

l l Cs137 i Ci 1.90E-04 l

J l

l 1131 l Ci l

l~CoS8 i ci 8.62E-04 i

1_cor-04 1

1_79r n, 1 Anr na l Cc60 I ci I 4.48E-04 1

4 3.51E-04 3. 2 9E-Ot.

l 4

( Fe59 i Ci J

l i

l l

i Ci i

l i

I l

tin 54 Ci l

i I 8.34E-05 4 6.76E-06

,l

$llZn65 l Cr51 Ci l

i 1

l l

l Zr-Nb95 I Ci i

i

! 4.07E-05 I

l

~

l Mo99 i Ci i

i l

i l

l Ba-La140 l Ci i

l i

i l

5r i C1 1 3.75E-04 1

0.00E00 1 2.07E-03 I 3.62F-03 l

1 l

I l

l j.

l Total Ci i 1.8SE-03 I

1.59E-04 I 4.32E-03 1 4.1?E-03 l

l

• Individual Liquid Effluen: Activity from Attachmen: #1 CONTINUOUS MODE BATCH MODE l

i I

l l

I i

l NUCLIDE I UNIT l

QTR 1 l

QTR 2 l

QTR 1 l

QTR 2 1

f

]

I I

l I

I I

28.0 DISSOLVED AND ENTRAINE's GASES l Xe133 i C,i l 1.71E-04 i

! 3.30E-03 1 1.84E-03 l

I i

i l

l Xe133m 1 Ci i

i

__ l l Xe131m t Ci i

i i

I l Xe135 I ci i

i i 2.99E-05 1 3.1EE-06 l

i l

l Kr85m i Ci f

1 i

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l KrSS I Ci i

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8 i

l KrSS I C:

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- f Revised 10-1-85 RTTACHMENT 1 ADDITIONAL NUCLIDES o

GASEOUS FITLUENTS CONTINUOUS MODE BATCH HODE NUCLIDE UNIT OTR 1 QTR 2 QTR 1 QTR 2 i_________l____g____________;____________i____________i____________;

l Ga-72 l Ci l l

l 6.06E-07 1

1 g_________l____l____________;____________l___7_________;____________l l Mn-54 1 Ci l l

l

.61E-07 I

I i_________l____l____________i____________!____________[____________l l Se 75 i Ci 1 5.00E-07 1

l 1

I l _ _2_ _5_3_ _ _ _ I _ C_f_ _ I _ _ _ _ _ _ _ _ _ _ _ _ I _ _ _ _ _ _ _ _ _ _ _ _ I _ _ _ _ _ _

i __F_-

____i_

_i____________ ____________l____________i____________i i Subtotall Ci l 5.00E-07 l

0.00E00 1

1.57E-06 l

0.00E00 l

l_________l____i____________[____________i____________i____________i LIQUID r:rLUENTS CONTINUOUS MODE BATCH MODE NUCLIUE UNIT OTR 1 QTR 2 QTR 1 QTR 2 e

_________i qfj-________i____i____________l____________l____________[___8.62E-04 l Eg-110m l Ci l l

1 9.03E-05 1

1

~

i_________i____i____________l____________i____________i____________;

1 Cs-136 i Ci i i

i 1

5.17E-05 1

,;_________[____i____________j____________l____________i____________i l Ga-72 l Ci i i

i B.83E-05 l

l i_________l____i____________i____________i____________i____________;

l Nb-97 l Ci l l

l 7.10E-07 l

l j.________l____l____________;____________l____________;____________i i sb-124 1 Ci I l

l 1.43E-03 1

1.20E-04 l

_________j-___i____________g____________;____________i____________i i sb-125 i Ci l 3.75E-04 l

l 4.61E-04 l

1.90E-04 I

l_________[____i____________l____________l____________i____________;

i Sc-47 I Ci i l

l 1.50E-06 I

i

_________i____;____________j____________i____________i____________;

I sr-92 l Ci l l

1 l

2.66E-04 l

i_________i____;____________i____________[____________i____________i I Ir-97 i Ci i l

l 1

3.66E-05 l

l _ _J_e _5_5_ _ _ _ ;i _ _C_1_ i _ _ _ _ _ _ _ _ _ _ _ _ ; _ _ _ _ _ _ _ _ _ _ _ _ i _ _ _ _ _

____l____________i____________i____________i____________

i 2

g_

l Eubtotall Ci [

3.75E-04 1

0.00E00 1

2.07E-03 l

i_________i____i____________;____________[____________l___3_.6_2_E__0_3___i 1

i i

i l

L _._

____________________w

p *. ;

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DESCRIPTION OF REVISION 10 ODCM MANUAL CHANGES 1)

A change in method for calculating liquid effluent monitor setpoints.

I See attached pages for further explanation and justification.

2)

Equation 2.1-6 variable definitions contained a typographical error.

The Dilution Water Flow Rate "F" was corrected to 67,300 gpm.

3)

Revision 9 of the ODCM changed the critical organ dose parameter from the infants inhalation pathway to the childs inhalation pathway.

The variable definition for equation 3.2-1 still incorrectly referenced the infants inhalation pathway and was corrected.

4)

In Table 5.1-1, Radiation Environmental Monitoring Program Sampling Locations, sample point P-8 for Well Water was changed from Kinney's Store to the Community Center since Kinney's Store no longer exists.

5)

In Table 5.1-1, Radiation Environmental Monitoring Program Sampling Locations, sample point P07B for TLD location was changed from the Red Wing Service Center to the Red Wing Public Works since it is a more accessable locations for sample changing.

The new TLD location represents the same sector and distance from the plant site.

6)

In Table 5.1-1, Radiation Environmental Monitoring Program Sampling Locations, the Corn sample location was deleted from the table since there are no cornfields irrigated with water from the Mississippi River within 5 miles downstream of the plant.

Section 5.1 of the ODCM states that we will conduct an annual land use survey to determine if any irrigation is taking place between the plant and a distance of 5 miles downstream.

l; CHANGE TG THE ODCM LIQUID EFFLUENT l

MON 2 TORS SETPOINT CALCULATION 1

A review of section 2.1 of the ODCM (Revision 9) which deals with liquid effluents revealed a discrepancy in the formulas used to calculate monitor setpoints.

This analysis reviews the old setpoint methodology, points out the discrepancy in this methodology, explains the new setpoint methods (which have been put into practice) and compares the setpoints obtainod using the old method versus the revised method.

As soon es the discrepancy in the ODCM was noted, the liquid effluents were temporarily halted until it was resolved.

i Use of the new method was instituted as soon as it became available and had been approved.

Revision 10 of the ODCM was issued soon thereafter.

Old Methodology (Revision 9 of ODCM)

Section 2.1 of the ODCM contains three sets of setpoint calculations, namely:

1.

2.1.1 " Liquid Effluent Monitor Setpoints",

2.

2.1.2 "Setpoint Based on Analysis of Liquid Prior to Discharge".and 3.

2.1.3 " Discharge Canal Monitor".

No problems were noted with 2.1.2 so this section is ignored.

Sections 2.1.1 and 2.1.3 have identical calculations with the exception of the dilution factor.

This review will address section 2.1.1 but the results will also be applied to section 2.1.3.

An explanation of each st ;p in section 2.1.1 is given below:

2.1.1.1 This step involves the determination of the radionuclides mix for the waste stream and the fractional activity of each nuclide according to the following formula:

Ai Si

= -------

jf Ai a

where: Ai the radioactivity of

=

radionuclides 'i' in the liquid effluent

p

~ ~~~ ~~

uy

.Page 3 2.1.1.2.This step presents a formula for calculating the maximum. acceptable total radioactivity concentration of all radionuclides in the liquid efflue'nt prior to dilution (Ct) according to the following-formula:

s.'/f.

Ct = -------------

jf___Si f

MPCi where: F = dilution water flow rate (gpm)

= 67,300 gpm from cooling tower blowdown f = the maximum attainable discharge flow rate prior to dilution (gpm)

This formula is an adaptation of the equation in the addendum to NUREG-0133.

Weighting of the MPCi's is achieved by the ratio Si/MPCi.

Summing of these individual weights produces the inverse of a " composite MPC" which is the maximum allowable activity in the diluted sample.

Maximum allowable activity in the undiluted sample is calculated by multiplying this quantity by the ratio of dilution to sample flows.

Note that this equation includes both the gamma and non-gamma emitting activity.

2.1.1.3'This step adjusts the maximum allowable activity calculated in the last step for the contribution of non-gamma emitters according to the formula:

Cm = Ct - (Ct x Sh) where: Sh - The fraction of the total radioactivity in the liquid effluent contributed by tritium and other radionuclides that do not emit gamma or X-ray radiation.

2.1.1.4, 2.1.1.5 and 2.1.1.6 These steps convert the maximum allowable activity concentration to the monitor setpoint value using the applicable monitor l

efficiency curve and fraction of total activity released via a particular release point.

No problems were found with these steps and no changes were made.

l

I Page 3 DISCREPANCY A discrepancy was found in step 2.1.1.3 of Section 2.1.

The intent.of the formula in this step is to reduce the value of the maximum allowable activity by the amount of non-gamma.

emitting activity.

Since the effluent monitor is only sensitive to gamma radiation this appe;ars to be appropriate but in fact, this step overcompensates for the' contribution of the non-gamma emitters to the maximum allowable activity concentration.

Since the variable "Sh" is the fraction of total activity due to the non-gamma emitters and the variable "Ct" is the maximum' allowable activity bared on a summation of weighted MPC's for individual nuclides, the equation is reducing the maximum allowable activity based on MPC-weighted activities by a fraction which has not been weighted by the MPC's.

In.this case the non-weighted reduction overcompensates for the effect of the non-gnmma emitters.

NOTE:

The overcompensation effect of the non-gamma emitters occurs for the mix of radionuclides normally seen at Prairie Island.

It also occurs for all anticipated radionuclides mixtures.

Since the exicting formula overcompensates for the non-gamma emitters and therefore produces a setpoint which tends to be lower than the setpoint which would correspond to the MPC it is tempting to retain it.

However, the reason the discrepancy was found was that the formula does not provide reasonable results when the non-gamma emitting fraction (Sh) gets very large.

In this case the maximum. allowable concentration of gamma emitters approaches zero.

From sampling results it is known that this situation is not realistic since the non-gamma emitters have never exceeded more than half of the diluted MPC.

When the non-gamma emitting fraction becomes very large this formula causes the setpoint to be set so low that background fluctuations could cause alarms.

Therefore a correction to the formula is needed to avoid this situation.

RESOLUTION Calculation of the proper maximum allowable activity concentration in the liquid effluent requires that the mix of gamma and non-gamma emitters be assessed and the contribution of the non-gamma emitters be removed.

The present method reduces the maximum allowable activity concentration for the non-gamma emitters using a non-weighted activity fraction.

Instead of calculating a weighted activity fraction for use with the new method, a more straight-forward approach is taken.

In this method L

non-gamma and gamma emitter effects on the maximum acceptable concentration are dealt with separately.

A step-by-step description of the approach follows:

_________O

Page 4 1.

Calculate the number of.MPC's of non-gamma emitting activity in the undiluted sample by summing the quotient of the activity concentration hnd the MPCi for non-gamma emitting nuclides.

Y

'2.

Reduce the dilution factor (F/f) by the number of t

non-gamma emitting MPC's.

The reduced dilution factor is then used to calculate the maximum allowable gamma activity.

For example:

If the number of non-gamma emitting MPC's in a sample is 143, then it must be diluted at least 143 times to achieve an MPC at the outlet.

If the total dilution factor is 673 (67,300 gpm blowdown flow / 100 gpm sample flow), the remaining dilution factor which is applied to the gamma emitting activity is 673 - 143 - 530.

3.

Proceed with the calculation of fractional MPC's (Si/MPCi) for gamma emitting activity and apply the reduced dilution factor.

These changes were incorporated into Revision 10 of the CDCM.

Each of the changes is shown (in abbreviated form) and described for the corresponding step number:

2'.l.1.1 b.

Determine the activity concentrations (Aci) of all non-gamma emitters including H-3, Sr-89, Sr-90, Fe-55, and alpha activity.

This is the first step in separating the non-gamma emitting activity from the gamma activity.

c.

Determine NGF (the total fraction of the MPC in the liquid effluent) for all non-gamma emitting nuclides.

)[-ACi NGF =

i HPCi where: ACi = Activity concentration of non-gamma emitting nuclide

'i' in the liquid effluent (uci/ml)

MPCi = The liquid effluent radioactivity concentration limit for radionuclides 'i' (uci/ml)

l' Page 5 This step calculates the number of MPC's of non-gamma emitting activity present in the undiluted liquid.

d.

Determine si (the fraction of the gamma emitting radioactivity in the liquid effluer' comprised by radionuclides

'i')

for each individual radionuclides in the liquid effluent.

Ai Si

= -------

L F Ai a

L where: Ai = the radioactivity of gamma emitting radionuclides 'i' in the liquid effluent i

This equation is identical to the old. formula (2.1-1) except that non-gamma emitters are not included.

Deter'ine WGF (the sum of fractional m

e.

activities weighted by the MPC) for the gamma emitting nuclides.

ji-Si wGr =

i MPCi where: MPCi - The liquid effluent L

radioactivity concentration limit for radionuclides 'i' l

(uCi/ml)

This intermediate step was separated out from the old formulas to make it easier to understand.

2.1.1.2 Determine Ct (the maximum acceptable total radioactivity concentration of gamma emitting r,uclides in the liquid effluent prior to dilution (uCi/ml)).

1 i

Ct = ----- x

--- - NGF WGF f

where: r = dilution water flow rate (gpm) l f = the maxinun attainable discharge flow rate prior to dilution (gpm) l

Page 6 This step completes the revised method oy calculating the.

product of the inverse of the. weighted 7.PC's for gamma emitting activity by the difference between the dilution factor and.the number of MPC's of non-gamma emitting j

activity.

I t

l l

1 1

1 l

I i

I l

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