ML20082M285
| ML20082M285 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 06/30/1991 |
| From: | Fey F NORTHERN STATES POWER CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9109050118 | |
| Download: ML20082M285 (179) | |
Text
Northom States Power Company 114 Nicollet Mall M,nneapohs, Minnesota 55401 1927 Telephone (612) 3M 5500 August 27, 1991 Prairie Island Technical Specification TS6.7A 5&n U.S.
Nuclear Regulatory Commission Document Control Desk washington, D.C.
20555 Prairie Island Nuclear Generatin?, Plant Docket No. 50-282 License No. DPR-42 Docket No. 50-306 License No. DPR-60 Effluent and Waste Disposal Semiannual Report for January 1.
1991 throuch dyne _)0, 199.
In accordance with the Prairie Island Technical Specifications, Appendix A to Operating License DPR-42 and DPR-60, we are rubmitting one copy of the Effluent and Waste Disposal Semiannual saport, covering the last half year of 1991.
analysis for isotopos Sr-89 and Sr-90 for the second quarter were not completed in time to be included with this report.
These cnalyses results are not expected to significantly change the computed off-site doses, and vill be included with the next semiannual report.
Enclosed with this report is an amended Effluent and Waste Disposal Semiannual Report for the second half of 1990 which includes the previously omitted fourth quarter analyses results of Sr-89 and Sr-90.
Cop es of the ODCM and PCP, including a description of the changes, are also a t ched.
,j
,[
AK e'
F>dl. Fey, r.,b Manager Nuclear Radiological Services Attachment
((Ab!
.J 9109050110 710927 I \\
PDR ADOCK 05000232 R
's i
TRANSM1TTAL MANIFEST l
NORTHERN STATES POWER COMPANY i
NUCLEAR GENERATION DEPARTMENT j
PRAIRIE ISLAND NUCLEAR GENERATING PLANT L
Effluent and Waste Disposal Semiannual Report l
for January 1.1991 throc "h June 30. 1991 1
e Manifest Date: August 27, 1991 USNRC 4
ANI Library I
- Regional Admin-Ill Fluor Daniels 3
7
- NRR Project Manager J Gelston i
- DCD T Synder
- Resident Inspector Corporate Library Shaw Pittman Potts &
K M Beadell 1
Trowbridge 1
i D A Schuelke 1
J Silberg P H Kamman 1
Wesi-inghouse Electric 2
Prairie Island Site General Mgr 1
G Goldberg Monticello Plant Manager 1
0 Sommers ERAD Dept.
1 Safety /.udit Committee 10 i
Attn: Records Clerk
- K J Albrecht Communications Department 1
- A B Cutter
- NRS File 1
- F W Hartley NSS File 1
- D M Musolf
- MDH 1
"G H Neils Attn:
'mmissioner of Health
- T H Parker
- MPCA 1
- H B Sellman Attn: J W Ferman
- C Steinhardt
- J A Thie
- Secretary Manifest File u
L I
1991 Effluent Semiannual Report REV. O Page 1 of 8 l
Retention: Lifetime EFFL'JENT SEMIANNUAL REPORT 30-DEC-90 THROUGH 30-JUN-91 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 individual from radioactive materials in liquid effluents released from the site shall be limited to:
for the quarter 3.0 mrem 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 43000 mrem / 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 410 mrad / quartet gamma 420 mrad / quarter beta 420 mrad / year gcmma 440 mrad / year beta I-131, H-3, LLP (15 mrem / quartet to any organ 430 mrem / year to any organ
5 1991 EFFLUENT SEMIANNUAL REPORT REV. O PAGE 2 B.
Maximum Permissible Concentration 1.
Fission and activation gases in gaseous releases:
10 CFR 20, Appendix B, Table 2, Column 1 2.
Iodine and particulates 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.
Fission and activation gases Total GeLi 25%
in gaseous releases:
Nuclide GeLi 2.
Iodines in gaseous releases:
Total GeLi 25%
Nuclide GeLi 3.
Particulates in gaseous releases:
Total GeLi 25%
Nuclide GeLi 4.
Liquid effluents Total GeLi i25%
Nuclide GeLi E.
Manual Revisions 1.
Offsite Dose Calculations Manual latest Revision number:
12 Revision date
- 30-JUN-91 2.
Process Control Program Manual latest Revision number:
4 Revision date
- 23-APR-91
4 1991 EFFLUENT SEMIANNUAL REPORT REV. O PAGE 3 i
1.0 BATCH RELEASES (LIQUID)
QTR: 01 QTR: 02 1.1 NUMBER OF BATCH RELEASES 3.00E+01 6.90E+01 1.2 TOTAL TIME PERIOD (HRS) 4.42E+01 1.19E+02 1.3 MAXIMUM TIME PERIOD (HRS) 2.00E+00 3.58E+00 1.4 AVERAGE TIME PERIOD (HRS) 1.47E+00 1.72E+00 1.5 MINIMUM TIME PERIOD (HRS) 4.20E-01 1.17E+00 1.6 AVERAGE MISSISSIPPI RIVER FLOW (CFS) 9.21E+03 4.14E+04 2.0 BATCH RELEASES (GASEOUS)
QTR: 01 QTR: 02 2.1 NUMBER OF BATCH RELEASES 3.00E+00 1.80E+01 2.2 TOTAL TIME PERIOD (HRS) 1.40E+00 1.37EA02 2.3 MAXIMUM TIME PERIOD (HRS) 1.20E+00 1.99E+01 2.4 AVERAGE TIME PERIOD (HRS) 4.67E-01 7.60E+00 2.5 MINIMUM TIME PERIOD (HRS) 6.00E-02 1.00E-01 3.0 ABNORMAL RELEASES (LIQUID)
QTR: 01 QTR: 02 3.1 NUMBER OF 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 RELEASES 0.00E+00 0.00E+00 4.2 TOTAL ACTIVITY RELEASED (CI) 0.00E+00 0.00E+00
1991 EFFLUENT SEMIANNUAL REPORT REV. O PAGE 4 l
TABLE 1A l
GASEOUS EFFLUENTS - SUMMATION OF ALL RELEASES l
QTR: 01 QTR: 02 5.0 FISSION AND ACTIVATION GASES 5.1 TOTAL RELEASE (CI) 2.13E-01 5.56E+01 5.2 AVERAGE RELEASE RATE (UCI/SEC) 2.71E-08 7.07E-06 5.3 GAMMA DOSE (MRAD) 8.05E-05 2.24E-02 5.4 BETA DOSE (MRAD) 2.39E-04 7.34E-02 5.5 PERCENT OF GAMMA TECH SPEC (t) 8.05E-04 2.24E-01 5.6 PERCENT OF BETA TECH SPEC (%)
1.20E-03 3.67E-01 6.0 IODINES 6.1 TOTAL I-131 (CI) 0.00E+00 1.14E-04 6.2 AVERAGE RELEASE RATE (UCI/SEC) 0.00E+00 1.45E-ll 7.0 PARTICULATES 7.1 TOTAL RELEASE (CI) 0.00E+00 3.68E-04 7.2 AVERAGE RELEASE RATE (UCI/SEC) 0.00E+00 4.68E-ll 8.0 TRITIUM 8.1 TOTAL RELEASE (CI) 1.46E+01 2.40E+01 8.2 AVERAGE RELEASE RATE (UCI/SEC) 1.86E-06 3.05E-06 9.0 TOTAL IODINE, PARTICULATE AND TRITIUM (UCI/SEC) 1.86E-06 3.05E-06 10.0 DOSE (MREM) 2.70E-02 1.25E-01 11.0 PERCENT OF TECH SPEC (%)
1.80E-01 8.33E-01 12.0 GROSS ALPHA (CI) 0.00E+00 4.43E-08
1991 EFFLUENT SEMIANNUAL REPORT REV. O PAGE 5 TABLE 1C GASEOUS EFFLUENTS - GROUND LEVEL RELEASES 13.0 FISSION AND ACTIVATION GASES CONTINUOUS MODE BATCll MODE NUCLIDE UNITS Q1R: 01 QTR: 02 QTR: 01 QTR: 02 AR-41 CI 6.54E-03 KR-85 CI 1.18E+00 8.05E-01 XE-131M CI 8.71E-03 XE-133 CI 2.13E-01 5.16E+01 9.43E-01 XE-133M CI 5.69E-01 XE-135 CI 4.45E-01 3.42E-03 0.00E+00 1.77E+00 TOTAL CI 2.13E-01 5.38E+01
,I
.mJ 14.0 IODINES CONTINUOUS MODE BATCH MODE' NUCLIDE UNITS QTR: 01 QTR: 02 QTR: 01 QTR: 02 lI-131 CI 1.14 E- 0 '.
1-133 CI 1.32E-05 9.07E-09 l
0.00E+00 9.07E-09 TOTAL CI 0.00E+00 1.27E-04 15.0 PARTICULATES CONTINUOUS MODE BATCH MODE NUCLIDE UNITS QTR: 01 QTR: 02 QTR: 01 QTR: 02 CS-137 CI 7.95E-05 CS-134 CI 6.67E-05 CO-58 CI 2.39E-07 2.14E-05 CO-60 CI 1.88E-04 MN-54 CI 1.18E-05 NB-95 CI 1.15E-06 TOTAL CI 0.00E+00 1.39E-06 0.00E+00 3.67E-04
t i
1991 EFFLUENT SEMIANNUAL REPORT REV. O PAGE 6 TABLE 2A LIQUID EFFLUENTS - SUMMATION OF ALL RELEASES QTR: 01 QTR: 02 16.0 VOLUME OF WASTE PRIOR TO DILUTION (LITERS) 2.04E+07 3.66E+07 17.0 VOLUME OF JILUTION WATER (LITERS) 1.46E+11 6.94E+10 18.0 FISSION AND ACTIVATION PRODUCTS 18.1 TOTAL RELEASE W/O H-3, RADGAS, ALPHA (CI) 3.29E-02 9.29E-02 18.2 AVERAGE DILUTED CONCENTRATION (UCI/ML) 2.25E-10 1.34E-09 19.0 TRITIUM 19.1 TOTAL RELEASE (CI) 1.64E+02 7.64E+01 19.2 AVERAGE DILUTED CONCENTRATION (UCI/ML) 1.12E-06 1.10E-06 20.0 DISSOLVED AND ENTRAINED GASES 20.1 TOTAL RELEASE (CI) 5.99E-03 1.07E-03 20.2 AVERAGE DILUTED CONCENTRATION (UCI/ML) 4.10E-11 1.54E-11 21.0 GROSS ALPHA (CI) 0.00E+00 0.00E+00 22.0 TOTAL TRITIUM, FISSION AND ACTIVATION PRODUCTS (UCI/ML) 1.12E-06 1.10E s6 23.0 TOTAL BODY DOSE (MREM) 5.98E-04 1.23E-03 24.0 CRITICAL ORGAN 24.1 DOSE (MREM) 5.98E-04 1.23E-03 24.2 ORGAN TTL BODY TTL BODY 25.0 PERCENT OF TOTAL BODY TECH SPEC LIMIT (%)
1.99E-02 4.10E-02 26.0 PERCENT OF CRITICAL ORGAN TECH SPEC LIMIT (%)
1.99E-02 4.10E-02
1991 EFFLUENT SEMIANNUAL' REPORT ~
REV. 0-PAGE 7 LTABLE 2A LIQUID EFFLUENTS - SUMMATIOi' OF ALL RELEASES 27.0 INDIVIDUAL LIQUID EFFLUENT-CONTINtJOUS-MODE
' BATCH MODE' NUCLIDE UNITS CTR: 01 QT3: 02 QTR: 01 QTR: 02 l
AG-110M aCI 1.06E 3.39E-02 I
5 BE-7 CI 5.38E-05 l
Co-57 CI 7.12E-06 j CO-58 CI 6.02E-06 4.99E-04 1.36E-02 CO-60 CI 4.57E-06 2.15E-03 2.42E 03 l
9 CR-51 CI 5.35E-04 4.64E-03 l CS-134 CI 6.37E-05 1.22E-05 1.27E-05 i
CS-137 CI 1.25E-04 3.90E-05 3.66E-05 FE-55 CI 2.69E-04 1.71E-02 2.89E-02 i
e FE-59 CI 2.06E-04 8.56E-04 l
I I-131 CI 1.50E-04 l
LA-140 CI 3.95E-05
{
MN-54 CI 3.84E-05 2.2iE-04 f
NB-95 CI
-8.40E-05 i
l RH-105 CI 1.21E-05 SB-122 CI 2.64E-04 SB-124 CI 2.16E-04 4.06E-03 i
SB-125 CI 7.40E-04 2.47E-03 SB-126 CI 5.96E-06 l
SC-47 CI 4.41E-06 2.20E-04 SE-75 CI 1.34E-05 SN-113 CI 3.04E-04 5.29E-04 j
I SR-92 CI 2.86E-06 3.22E-04 CONTINUED r
1
[
1991 EFFLUENT SEMIANNUAL REPORT REV. O PAGE P TABLE 2A LIQUID EFFLUENTS - SUMMATION OF ALL RELEASES 27.0 INDIVIDUAL LIQUID EFFLUENT (CONTINUED)
CONTINUOUS MODE BATCH MODE NUCLIDE UNITS QTR: 01 QTR: 02 QTR: 01 QTR: 02 W-187 CI 1.45E-05 ZN-65 CI 6.26E-06 ZR-95 CI 4.10E-05 ZR-97 CI 1.33E-06 TOTAL CI 2.74E-04 1.95E-04 3.26E-02 9.27E-02 28.0 DISSOLVED AND ENTRAINED GASES CONTINUOUS MODE BATCH MODE NUCLIDE UNITS CTR: 01 QTR: 02 QTR: 01 QTR: 02 XE-131M CI 3.18E-05 XE-133 CI 2.64E-05 5.96E-03 7.62E-04 XE-133M CI 2.72E-05 4.61E-06 XE-137 CI 2.40E-04 TOTAL CI 0.00E+00 2.64E-05 5.99E-03 1.04E-03 l
-.- ~ -.
PINGP:753,'Rev. 3 Page 1 of 6-Retention:
Lifetime
~
- PRAIRIE ISLAND NUCLEAR. GENERATING PLANT Period: 1-1-91 to 6-30-9]
NORTHERN STATES POWER License-No. DPR SOLID RADIOACTIVE WASTE DISPOSAL SEMI-ANNUAL REPORT Table I:
Solid Vaste and Irradiated Fuel Shipments 1.
Solid Waste Total Volumes and Total Curie Quantities:
i l
l l
l Container I
l-l l
l Disposal
[
l Type:of Waste l
Units I
Totals l
Volumes l
l (List) l 1
I 3
l A. Resins l
ft l
0 l
l l
1 I
I I
1 1
Ci l
l l
l l
l l
l l-1 I
I I
I l
l J
I 3
+M l
B.
Dry-Compacted I~ft l
0 l
l 9
.I I
l I
I l
l Ci l
l l
.I I
i i
I I
I i
i l
l I
i l
I I
3 l
C.
Non-Compacted l
ft 1
35.0 1
96 l
i
-l l
I I
-l
.]
l Ci l
1.0337 l
1 1
l l
l l
l l
1 l
I i
I I
I I
I 3'
l D.
Filter Media i
ft l
O l
l 1
1 1
I i
~l I
Ci l
I
_I I-1 I
I I
l 1
1 I
I I
I l
l
-d l
S.
Spent Fuel I
ft l
0 l
l 3
l-l l
l l
l l
Ci l
l l
1 I
I I
l l
I I
I I
l l
I I
IBM i
PINGP 753, Rev. 3 Page 2 i
i PRAIRIE ISLAND NUCLEAR GENERATING PLANT Pe riod : 1-1-91 to 6-30-91 NORTHERN STATES POWER License No. DPR-42 SOLID RADIOACTIVE WASTE DISPOSAL SEMI-ANNUAL REPORT Table 1:
Solid Waste and Irradiated Fuel Shipiaents (Continued) 2.
Principal Radionuclide Composition by Type of Waste:
Percent TYPE (From Page 1)
Nuclide Abundance FE-55 46.5 C
CO--6 0 29.5 C-14 11.3
.,. v,-
- = Inferred - Not Measured on Site IBM
P1NGP 753. Rev. -. 3 Page 3 PRAIRIE ISLAND' NUCLEAR GENERATING PLANT Period : 1-1-91 to 6-30 NORTHERN STATES POWER License No. DPR-42 SOLID RADIOACTIVE WASTE DISPOSAL SEMI-ANNUAL REPORT Table I:
Solid Waste and Irradiated Fuel Shipment (Continued) 2.
Principal Radionuclide Composition by Type of_Vaste (Continuation):
Percent TYPE (From Page 1)
Nuclide Abundance sw-
!I.h:C]
7 r
I I
I t
2 -
4 h
i l
i I
- = Inferred - Not Measured on Site f
i IBM I
P'INGP 753, Rev. 3 Page 4 PRAIRIE ISLAND NUCLEAR GENERATING PLANT Period: 1-1-91 to 6-30-91 NORTHERN STATES POWER License No. ~DPR-47 SOLID RADIOACTIVE WASTE DISPOSAL SEMI-ANNUAL REPORT Table 1:
Solid Waste and Irradiated Fuel Shipments (Continued) 3.
Solid Waste Disposition:
Number of Shipments Mode Destination 1
TRIXZ OM3UDGE, 'IN, (QUADREX) 4.
Irradiated Fuel Shipments:
([3,)
Number of Shipments Mode Destin_ation 0
IBM
a
?
P'INCP 753, Rev.
3-Page S.
PRAIRIE ISLAND NUCLEAR GENERATING PLANT Period :1-1-91 to 6-30-91 NORTHERN STATES POWER License No. DPR-4'2 i
SOLID RADIOACTIVE WASTE DISPOSAL SEMI-ANNUAL REPORT' Table I:
Solid Waste and Irradiated Fuel Shipments (Continued)
'5.
Shipping Container and Solidification Method:
Disposal No.
Volume Activity Type of Container Solidif.
(Ft3)
(Ci)
Waste code Code 90-22 35.0 1.0337 C
L N/A r%
- -ici
+
1
'l
'l TOTALS 35.0 1.0337 CONTAINER CODES:
L = LSA (Shipment Type)
A = Type A B-: Type B Q = Highway Route Controlled Quantity SOLIDIFICATION CODES:
C - Cement TYPES OF WASTE:
A = Resins B = Dry Compacted C = Non-Compacted D = Filter Media S = Spent Fuel i
l IBM
1990 Effluent Semiannual Report REV. 1 Page lEof 8 Retention: Lifetime i
EFFLUENT SEMIANNUAL REPORT j
l~
01-JUL-90 THROUGH 29-DEC-90 SUPPLEMENTAL INFORMATION I
Facility:
Prairie Island Nuclear Generating Plant Licensee:
Northern States Power Company License Numbers: DPR-42 &'DPR-60 A.
Regulatory Limits l
1.
Liquid Effluents:
a.
The dose or dose commitment to an individual from radioactive materials in liquid effluents released from the site shall be limited to:
f for the quarter 3.0 mrem to the total body 10.0 mrem to any organ l
for the year 6.0 mrem to.the total body i
l 20.0 mrem to any organ 2.
Gaseous Effluents:
f i
a.
The dose rate due to radioactive materials released in j
gaseous effluents from the site shall be limited to:
noble gases 4500 mrem / year total body 43000 mrem / year skin j
1 l
I-131, H-3, LLP 41500 mrem / year to any organ I
t l
b.
The dose due to radioactive gaseous effluents shall be limited to:
i noble gases 410 mrad / quarte r gamma i
(20 mrad / quarter beta l
420 mrad / year gamma t
440 mrad / year beta i
r I-131, H-3, LLP 415 mrem / quarter to any organ 430 mrem / year to any organ t
9 i
I
1990 EFFLUENT SEMIANNUAL REPORT REV. 1 PAGE 2 i
B.
Maximum Permissible Concentration 1.
Fission and activation gases in gaseous releases:
10 CFR 20, Appendix B, Table 2,' Column 1 2.
Iodine and particulates with half-lives 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.
Fission and activation gases Total GeLi 25%
in gaseous releases:
Nuclide GeLi 2.
Iodines in gaseous releases:
Total GeLi 25%
Nuclide GeLi 3.
Particulates in gaseous releases:
Total GeLi 125%
Nuclide GeLi 4.
Liquid effluents Total GeLi
!25%
Nuclide GeLi E.
Manual Revisions 1.
Offsite Dose Calculations Manual latest Revision number:
11 Revision date
- 05-0CT-89 2.
Process Control Program Manual latest Revision number:
3 Revision date
- 31-MAY-90 l
R
~
~1990 EFFLUENT SEMIANNUAL REPORT REV..1 PAGE-3 100 -BATCH RELEASES (LIQUID)
' QTR: 03 QTR: 04 1.1 NUMBER-OF DATCH RELEASES 4.60E+01 2.50E+01 1-2 TOTAL TIME PERIOD (HRS) 6.83E+01-3.80E+01
.1.3 _ MAXIMUM TIME PERIOD (HRS) 1".92E+00 1.92E+00 1.4 AVERAGE TIME' PERIOD (HRS) 1.48E+00 1.52E+00-1.5 MINIMUM TIME PERIOD (HRS) 1.05E+00 1.25E+00 1.6 AVERAGE MISSISSIPPI RIVER FLOW (CFS) 1.52E+04 9.09E+03 i
2.0 BATCH RELEASES (GASEOUS)
QTR: 03 QTR: 04 2.1 NUMBER OF BATCH RELEASES 6.00E+00 1.80E+01 1.2 TOTAL TIME PERIOD -(HRS) 1.27E+01 4.55E+02 2.3 MAXIMUM TIME PERIOD (HRS) 8.20E+00 6.71E+01 2.4 - AVERAGE TIME' PERIOD (HRS)-
1.37E+00 3.73E+00 2.5 MINIMUM TIME PERIOD (HRS) 2.00E-02 2.00E-02 3.0 ~ ABNORMAL RELEASES (LIQUID)
QTR: 03 QTR: 04 3.1-NUMBER OF 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: 03 QTR: 04 4.1 NUMBER OF RELEASES 0.00E+00 0.00E+00 4.2 TOTAL ACTIVITY RELEASED (CI) 0.00E+00 0.00E+00 r
~
sm.
m
_---,-m s
~ - -
..-m--
19'90 EFFLUENT SEMIANNUAL REPORT REV. 1 PAGE 4 TABLE 1A i
GASEOUS EFFLUENTS - SUMMATION OF ALL RELEASES i
QTR: 03 QTR: 04 5.0 FISSION AND ACTIVATION GASES =
5.1 TOTAL RELEASE (CI) 1.35E+00 7.43E-01 5.2 AVERAGE RELEASE RATE (UCI/SEC) 1.72E-01 9.45E-02 5.3 GAMMA DOSE (MRAD) 7.73E-03 3.72E-03 5.4 BETA DOSE (MRAD) 6.05E-03 4.27E-03 5.5 PERCENT OF GAMMA TECH SPEC (%)
7.73E-02 3.72E-02 5.6 PERCENT OF BETA TECH SPEC (%)
3.03E-02 2.14E-02 6.0 IODINES 6.1 TOTAL I-131 (CI) 0.00E+00 4.56E-07 6.2 AVERAGE RELEASE RATE (UCI/SEC) 0.00E+00 5.80E-08 7.0 PARTICULATES 7.1 TOTAL RELEASE (CI) 1.10E-06 4.89E-05 7.2 AVERAGE RELEASE RATE (UCI/SEC) 1.40E-07 6.22E-06 8.0 TRITIUM
=
8.1 TOTAL RELEASE (CI) 8.48E+00 2.07E+01 8.2 AVERAGE RELEASE RATE (UCI/SEC) 1.08E+00 2.63E+00 9.0 TOTAL IODINE, PARTICULATE AND TRITIUM (UCI/SEC) 1.08E+00 2.63E+00 10.0 DOSE (MREM) 1.53E-02 3.87E-02 11.0 PERCENT OF TECH SPEC (%)
1.02E-01 2.58E-01 12.0 GROSS ALPHA (CI) 2.38r-07 0.00E+00
19'90 EFFLUENT SEMIANNUAL REPORT REV. 1 PAGE 5 TABLE 1C GASEOUS EFFLUENTS - GROUND LEVEL RELEASES 13.0 FISSION AND ACTIVATION GASES CONTINUOUS MODE BATCH MODE i-NUCLIDE UNITS QTR: 03 QTR: 04 QTR: 03 QTR: 04 AR-41 CI 7.64E-01 3.50E-01 3.55E-03 KR-85 CI 3.78E-01 3.40E-01 XE-131M CI 1.08E-03 XE-133 CI 1.95E-01 1.62E-02 4.84E-02 XE-133M CI 2.76E-04 TOTAL CI 9.59E-01 3.50E-01 3.94E-01 3.92E-01 14.0 IODINES CONTINUOUS MODE BATCH MODE NUCLIDE UNITS QTR: 03 QTR: 04 QTR: 03 QTR: 04 I-131 CI 4.56E-07 TOTAL CI 0.00E+00 4.56E-07 0.00E+00 0.00E+00 15.0 PARTICULATES CONTINUOUS MODE BATCH MODE NUCLIDE UNITS QTR: 03 QTR: 04 QTR: 03
-QTR: 04 CD-109 CI 3.91E-08 2.96E-05 CO-58 CI 4.83E-07 1.90E-06 1.71E-05 C0-60 CI 5.82E-07 2.53E-07 3.05E-07 TOTAL CI 1.10E-06 3.18E-05 0.00E+00 1.74E-05 t
4 1990 EFFLUENT SEMIANNUALJREPORT REV. 1 PAGE 6' TABLE 2A f
LIQUID EFFLUENTS - SUMMATION OF.ALL RELEASES r
l t
QTR: 03 QTR: 04 i
16.0 VOLUME OF WASTE PRIOR TO DILUTION (LITERS) 2.14E+07 3.09E+07 f
17.0 VOLUME OF DILUTION WATER (LITERS) 2.45E+11 1.58E+11 i
18.0 FISSION AND ACTIVATION PRODUCTS
[
18.1 TOTAL RELEASE W/O H-3, RADGAS, ALPHA (CI) 7.69E-02 3.71E-02 i
18.2 AVERAGE DILUTED CONCENTRATION (UCI/ML) 3.14E-10 2.35E-10
?
l 19.0 TRITIUM i
19.1 TOTAL RELEASE (CI) 7.12E+01 7.10E+01 l
l 19.2 AVERAGE DILUTED CONCENTRATION (UCI/ML) 2.91E-07 4.49E-07 i
20.0 DISSOLVED AND ENTRAINED GASES l
f 20.1 TOTAL RELEASE (CI) 4.18E-03 1.29E-03 20.2 AVERAGE DILUTED CONCENTRATION (UCI/ML) 1.71E-11 8.16E-12 I
21.0 GROSS ALPHA (CI) 1.40E-05 0.00E+00 l
122.0 TOTAL TRITIUM, FISSION AND ACTIVATION
[
PRODUCTS (UCI/ML) 2.92E-07 4.49E-07 I
i I
23.0 TOTAL BODY DOSE (MREM)
-1.64E-03 3.34E-04 I
24.0- CRITICAL ORGAN i
24.1 DOSE (MREM) 1.64E-03 3.34E-04 r
i 24.2 ORGAN TTL BODY TTL BODY-
{
I 25.0 PERCENT OF TOTAL BODY TECH SPEC LIMIT (%)
5.47E-02 1.11E-02
[
i 26.0 PERCENT OF CRITICAL ORGAN TECH SPEC LIMIT (%)
5.47E-02 1.11E-02
}
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- I
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m 1990 EFFLUENT SEMIANNUAL _ REPORT
_REV..1 PAGE 7 TABLE 2A LIQUID EFFLUENTS --SUMMATION OF ALL RELEASES 27.0 INDIVIDUAL LIQUID EFFLUENT CONTINUOUS MODE BATCH MODE I
NUCLIDE UNITS QTR: 03 QTR: 04 QTR: 03 QTR: 04 AG-110M CI 1.30E-02 9.71E-03 BE-7 CI 3.45E-04 5.55E-05 i
CO-58 CI 2.81E-02 2.61E-03
)
CO-60 CI 3.73E-03 4.09E-03
{
CS-134 CI 1.26E-04 1.81E-05
{
CS-137 CI 2.76E-06 1.88E-04
^
CS-136 CI 2.77E-06 CR-51 CI 1.49E-03 1.35E-03 I
C0-57 CI 2.20E-05 8.85E-07
(
i CE-144 CIL 4.65E-05 i
FE-55 CI 1.83E-02 1.01E-02 FE-59 CI 2.37E-03 2.63E-03 LA-140 CI 2.23E-05 1.56E-06 i
t i
LA-142 CI 3.60E-06 l
MN-54 CI 2.02E-04 1.45E-04 NB-97 CI 5.90E-06 t
.i ND-147 CI-7.12E-06 l
I
.l SB-122 CI 7.21E-04 SB-124 CI 3.90E-03 2.71E-03 i
SB-125 CI 1.71E-03 2.43E-03 i
SB-126 CI 1.44E-05 I
SC-47 CI 2.03E-04 2.33E-04 1
1 SN-113 CI 2.90E-04 8.62E-04 l
p I
CONTINUED l
i
)
I i
l -
PAGE 8 i9'90 EFFLUENT SEMIANNUAL REPORT REV. 1 TABLE 2A LIQUID EFFLUENTS - SUMMATION OF ALL RELEASES INDIVIDUAL LIQUID EFFLUENT (CONTINUED)
CONTINUOUS MODE BATCH MODE NUCLIDE UNITS QTR: 03 QTR: 04 QTR: 03 QTR: 04 5.82E-06 SR-92 CI 5.00E-06 2.00E-06 TC-99M CI 4.78E-06 RH-105 CI 3.34E-05 ZN-65 CI 2.958-06 ZR-97 CI SR-89 CI 1.31E-03 TOTAL CI 1.31E-03 0.00E+00 7.56E-02 3.71E-02 28.0 DISSOLVED AND ENTRAINED GASES CONTINUOUS MODE BATCH MODE NUCLIDE UNITS QTR: 03 QTR: 04 QTR: 03 QTR: 04 2.50E-05 AR-41 CI 4.03E-03 1.27E-03 XE-133 CI 2.88E-05 1.57E-05 XE-133M CI 9.82E-05 4.27E-06 XE-135 CI TOTAL CI 0.00E+00 0.00E+00 4.lSE-03 1.29E-03
-Rev. 12 i
Page 1 of 127 l
l i
.)
r
. PRAIRIE ISLAND NUCLEAR GENERATING PLANT l
i OFFSITE DOSE CALCULATION MANUAL 1
.d (ODCM) 1
)
l DOCKET NO. 50-282 and 50-306 l
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t NORTHERN STATES POWER COMPANY MINNEAPOLIS, MINNESOTA f}7f
$?y J
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o*******,***********we************************************************
l-l'
^ OC-REVIEW DATE:
d?- 7 O.7/
o
' / /$ /
REVIEWED BY:
M
- ) ///<ntt<
DATE:
ff A -f q d / /.e d -
DATt:
7- /7-9/
AgPaOvED uY:
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H4 Rev. 10 TABLE OF CONTENTS Section-
. Title Page
-Table of Contents.-..........
..................... 2 List'of Figures....................................
3 List of Tab 1es................................
.... 4 Record-of Revisions................................
5 o
- 1. 0-INTRODUCTION.................................................
6 2.0 LIQUID EFFLUENTS.............................................
6 2.1 --Monitor Alarm Setpoint Determination...............
.... 6 2.2 Liquid Ef fluent Concentration - Compliance with 10 CFR 20.........................................
15-
~
2.3 Liquid Effluent Doses - Compliance with 10 CFR 50,.............................................
16-Refer.nces............................................. 20 stW QM
~
3.0 GASEOUS EFFLUENTS...........................................
23 3.1 Monitor Alarm Setpoint De t e rmina ti on................... 2 3 3.2 Gaseous Effluent Dose Rate - Compliance with 10 CFR 20..............................................
29 3.3 Gaseous Effluent Doses - Compliance with 10 CFR 50.....................,........................
33 References..........
............. 38 4.0 tINFORMATION RELATED TO 40 CFR 190 and 40 CFR 141............
59
-5.0: RADIATION ENVIRONMENTAL MONITORING PROGRAM................. 61
-APPENDIX A - Meteorological Dispersion Parameters...........
69
' APPENDIX B - Joint Frequency Distribution Tables........... 108 APPENDIX C - Dose Parameters for Radiciodines, Particulates and Tritium......................
116 IBM i
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Rev. 12 Page 3 m
-LIST OF FIGURES
-i
,;4i Figure Pm
-5.1-1
- S i t e Bour.d a ry TLD Lo c a tion s....................... 6 6 -
-5.1-2 Radiological ~ Environmental. Sample Points l
Within 10-Mile Radius...............................,67 5.1-3 Radiological Environmental Sample Points outside 10-Mil-Radius........................... 68 I-L i
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'*Ek 3
l Y5f?.$ -
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,._2 2._
,,m
..m._2 m,
a-_..ma_.m in..
._o w.
...sw=
.e + m.
m _41. 2
-1
-H4 Rev.12 Page 4
- LIST OF TABLES Tables Page.
' 2.1-1L Liqui d Sourc e Te rms for PINGP.................... 14
-2.3-1 A
Va lu e s f o r t h e - P I' 'c P... -........................ 21 g
3.1-1 Monitor Alarm Setpoln. Determination for PINGP.... 27 3.1-2 Gaseous Source Terms for PINGP....................
28 3.1-3 Dose Factors and Constants........................
29
. 3.2-1 P. Values for the PINGP...........................
32 1
=3.3-1 Dose Factors for Noble Gases and Daughters........ 39 3.3 R Values for the PINGP -' Ground, All Ages.........40 7
3.3-3 R
Values for the PINCP - Vegetable, Adult........
41 3
L 3.3-4 R
Values for the PINGP - Vegetable Teen.........
42 f
j 3.3-5 R
values for the PINGP - Vegetable, Child........
43 l
1 3.3-6 R. Values for the PINGP - Meat, Adult.............
44 l
1
- 3.3 R. Values for the PINGP - Meat, Teen..............
45 i
r 1
3.3-8 R. Values for the PINGP - Mest, Child.............
46 l
1 y
' 3.3-9 R. Values for the PINGP - Cow Milk, Adult..........
47 i
. ma 1
' Ch?
3.3-10 R. Values for the PINGP - Cow Milk, Teen..........
48 f
1
- 3.3-11 R. Values for the PINGP - Cow Milk, Child.........
49-
'l 1
3.3-12 R; Values for-the PINGP - Cow Milk, Infant........
50 l
~ 3.3-13 R. Values for the PINGP - Goat Milk, Adult........
51 l
1 3.3-14 R -Values for the PINGP - Goat Milk, Teen.........
52-j 1
- 3.3-15
-R.
Values for the PINGP - Goat Milk, Child........
53 f
1
- 3.3-16 R. Values for the PINGP - Goat Milk, Infant.......
54 8
1 1
- 3.3-17 R. Values for the PINGP - Inhalation, Adult....... 55 I
p 3.'3-18 R
Values for the PINGP - Inhalation, Teen........
56 i
3.3-19 R. Values for the PINGP - Inhalation, Child.......
57 1
~ 3.3-20 R. Values for the PINGP - Inhalation, In f an t...... 58 1
1 5.1-1 PINGP-- Radiation Environmental Monitoring Program Sampling-Locations....................... 62 j
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+
IBM
. _.....~ -.. - - - - -
s H4
'i Rev. 12 i
Page 5 RECORD OF REVISICNS Revision No.
Date-Reason for Revision e
original June 7, 1979 j
1 April 15, 1980 Incorporation of !EC Staff coments and crrrection of miscellaneous errors i
=2 August 6, 1982 Incorporation of NRC Staff coments f
3 February 21, 1983 Change in trJ.lk semling location i
4 Nover er 14, 1903 Change in milk sanpling location I
and change in cooling tower blowdown 5.
March 27, 1984 Change Table 3.2-1 l
6 Februat'f 14, 1996 Change it location to collect cultivated crops (leafy green
[
veg.) and removal of meat animals from the land use census
-I 7
July 31, 1986 Retype and format COCM.
l-No chang', in content
{
8 January 8,1987 Addition of Discharge Canal
[
monitor Setpoint calculation 9-June 29, 1987 Change innalation dose factor
[
a>
to child and address change in N.
land use survey 10 April 27, 1989 Change in method for calculating i
liquid effluent neniter setpoints.
j rix of various typing errors.
Change in locotton of two REMP samoling locations. Deletion of one REMP sampling location.
j 11 October 5, 1989 Change in Tables 3.3-6 thru
{
3.3-16.
Appendix C equationa j
corrected. Section 5 figures j
replaced. Sanple point definitions corrected.
}
12 June 17, 1991 Changes in REMP sampling L
locations Tables 5.1-1.
Added text to address the~
increased volune of the
-[
new discharge pipng.
j i
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[
H4 Rev. 12 1.0 INTRODUC110N This offsite Dose Calculation Manual (ODCM) provides the information and methodologies to be used by the Prairie Island Nuclear Generating Plant (PNGP) to assure compliance with certain portions of PNGP's Operating Technical Specification.
These portions are those related to liquid and gaseous radiological effluents.
They are intended to show compliance with-10 CFR 20, 10 CFR 50.36A, 10 CPR 50, Appendix A (GDC 60 &'64) and Appendix 1, and 40 CrR 190.
This ODCM is based on "RadioloEical Effluent Technical Specifications fsr PVR's.(NUREG-0472, October, 1978),"
" Preparation of Radiological Effluent Technical Specifica-tions for Naole. c Power Plants (NUREG-0133, October 1978),"
and other inputs from the Nuclear Regulatory Commission (USNRC).
Specific plant procedures for implementation of this manual are provided elsewhere.
These-procedures will be utilized by the operating staff of the PNGP to assure compli-ance-with the Technical-Specifications.
Also included in this manual is information related to the Radiological Environmental Monitoring Program (REMP) in the form of rigures 5.1-1,_5-.1-2, 5.1-3 and Table 5.1-1.
These figures.and table designate specific sample types and locations evrrently used to satisfy the technical specification require-43$
ments for the REMP.
They are subject to change based on the 98E" results of the periodic land use census.
Although the CDCM has been prepared as generically as possible in order to minimize the need for future revisions, some changes to the ODCM may be required in the future.
Any such changes will be properly reviewed and approved as required by the PNGP Technical Specifications.
2.0 LIQUID EFFLUCNTS 2.1' Moni cor Alarm Setpoint Det ermination This procedure determines the monitor alarm setpoint that indicates if the concentration ofiradionuclides in the liquid effluent released at the site bounda-'
exceeds the concentrations specified in 10 C Appendix B Table II, Column 2 for radionuci
..ie r than. dissolved or entrained noble gases or exceeds a concentration of 2 x 10-4 pCi/ml for dissolved or i
[
entrained noble gases.
Since Fe-55, St-89, Sr-90, and alpha concentrations are determined from composite samples, the liquid moni tor setpoint determinations should be completed using the most recent available composite sample results.
IBM L
H4 Rev. 12 Page 7 Monitor high alarm or isolation setpoints will be established in one of these ways:
a.
Monthly calculation of setpoints using the methodology of Sections 2.1.1 and 2.1.3.
b.
Calculation of alarm setpoint based on analysis prior to dischcrge using methodology of Section 2.1.2.
c.
Alarm setpoint determined using. methodology of Sections 2.1.1 and 2.1.3 assuming all radionuclides have an MPC of IE-7 pC1/ml.
No recalculation of setpoints is necessary unless an increase in alarm setpoint i desired.
PVR GALE Code source :erms (Table 2.
.)
may be used if there were no detectable isotopes in tne previous morth or in the analysit prior to release.
If the newly calculated setpoint is less than the-existing monitor setpoint.-the setpoint will be reduced to the new value.
If the calculated setpoint is greater than the existing setpoint, the setpoint may remain at the lower
+
value or increased to the new value.
2.1.1 Liquid Effluent Monitor Setpoints
. vr The following method applica when determining C'd the isolation setpoints for the Waste-Effluent Liquid Monitor (R-18), Steam Generator Blowdown Liquid Monitor - Unit 1 (IR-19), and Steam Generator Blowdown Liquid Monitor - Unit 2 (2R-19) during all operational conditions when the radwaste discharge flow rate is maintaine'.
constant at the maximum design flow rate.
2.1.1.1 Determine the " mix" (radionuclides and composition) of the liquid effluent, a.
Determine the liquid source terms that are representative of the
" mix" of the liquid ef fluent.
Liquid source terms are the total curies of each isotope released during the previous month.
Table 2.1-1 source terms may be used if there have been no liquid releases, b.
Determine the activity concentrations (ACi) of all non-gamma emitters including H-3, Sr-89, Sr-90, Fe-55, and alpha activity.
IBM
H4 Rev. 12 Page 8 c.
Determine NGF (the total fraction of the MPC in the liquid effluent) for all non-gamma emitting nuclides.
NGF={}
ACf MPCi (2.1-1) i where:
ACi = activity concentration of suclide 'i' in the liquid i
effluent (pci/ml).
MPCf = The liquid effluent radioactivity concentration limit for radionuclide 'i' (pci/ml) from Table 2.1-1 or Reference 3.
d.
Determine 51 (the fraction of thre gamma emitting radioactivity in the liquid effluent comprised by radionuclide
'i') for each i
individual radionuclide in the liquid effluent.
Si =
Ai
}{i Ai (2.1-2) m eu
'ig*
where:
Ai =
the radioactivity of gamma emitting radionuclide 'i' in the liquid effluent, e.
Determine WGP (the sum of fractional
~
activities weighted by the MPC) for the gamma emitting nuclides in the l
liouid ef fluent.
1 VGF = []
Si MPCi (2.1-3) o i
where:
MPCi :
The liquid effluent radioactivity concentration limit for radionuclide 'i' (pCi/ml) from Tabit 2.1-1 or Reference 3.
j l
IBM i
i
H4 Rev. 12 Page 9 3
2.1.1.2 Determine C (the maximum acceptable total radio 5ctivity concentration of gamma emitting nuclides in the liquid effluent prior to dilution (pCi/ml).
C
=
1 x
F (2.1-4) t
_ ggy WGF f
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) 60 gpm from the ADT tank pump (2)
=
100 gpm from the CVCS tank pump (2}
=
I 60 gpm from the SGBD tank pump ' )
=
2.1.1.3 Determine C.R.
(the calculated monitor count rate above background attributed
- e.
- to the radionuclides (nepm)).
t91 C.R.
is obtaired by using the applicable Effluent Monitor Efficiency Curve located in the Radiation Monitor Calibration file.
C.R.
is the count rate that corresponds.to the " adjusted"'
total radioactivity concentration'(C ).
t 2.1.1.4 Determine HSP (the monitor high alarm setpoint above background (nepm)).
HSP = T,C.R.
(2.1-5)
T* = Fraction of the radioactivity from the site that may be released via each release point to ensure that the site boundary limit is not exceeded due to simultaneous releases from several release points.
0.90 for the Waste Effluent Liquid
=
Monitor (R-18) 0.05 for the Steam Generator Blowdown
=
Liquid Monitor - Unit 1 (IR-19) 0.05 for the Steam Generator Blowdown
=
Liquid Monitor - Unit 2 (2R-19)
IBM
f I
i l
H4 Rev. 12 i
i Page 10 T
values may be revised from the i
L VIlues given above.
The summation L
of all the T values for active release points shall not be greater than unity.
2.1.1.5 The monitor high alarm setpoint above f
background (nepm), shall be set at or below the HSP value.
i 2.1.2 Setpoint Based on Analysis of Liquid Prior to l
Discharge (Op tion a D This method may be used in lieu of the method ta I
Section 2.1.1 to determine the setpoints for the f
maximum acceptable discharge flow rate prior to dilution and to determine the associated high alarm setpoint based on this flow rate for the Waste Effluent Liquid Monitor (R 18), Steam Generator Blowdown Liquid Monitor - Unit 1 1
(lR-39), and Steam Generator Blowdown Liquid--
Monitor - Unit 2 (2R-19), during all-operational conditions.
2.1.2.1 Determine f (the maximum acceptable i
discharge flow rate prior to dilution (gpm)).
i
, ax L 9 '5 f
=
0.8 F T I
~
m c
(2.1-6) f MPC i
g i
F: Dilution water flow rate (gpm) i
= 67,300 gpm from cooling tower blowdown i
L Cg: Concentration of radMnuclide "i" in i
the. liquid effluent prior to dilution (pCi/ml) from analysis of i
the liquid effluent to be released.
j MPCg: The liquid effluent radicactivity
)
concentration limit for r:dlonuclide i
"f" (pCi/ml) from Table 2.1 ; or if not listed in Table 2.1-1 fro *a J
Reference 3.
I T* = Fraction of the radioactivity from the site that may be released via i
each release point to ens.tre that j
l the site boundary limit is not exceeded due to simultaneous releases from several release points.
Refer to 7
Section 2.1.1.4.
}i l
-IBM 1
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~
Rev, 12-l Page 11 l
2.1.2.2 Determine the monitor setpoint based on the radionuclide mix of the liquid 1
effluent.
l a.
Determine C.R.
(the calculated I
monitor count rate above background
[
attributed to the radionuclides (nepm)).
j r
C.R.
is obtained by using the l
applicable Effluent Monitor Ef ficiency Curve located in the Radiation Honitor Calibration file.
C.R.
is th: count rate point that corresponds to the " adjusted" total radioactivity concentration (C ).
t t
C' = The total radioactivity concen-I tration of the radionuclides l
(minus tritium and other radionuclides that are only i
beta emitters) in the liquid discharge prior to dilution i
(pC1/ml) as determined using l
Equation 2.1-4.
i fCi b.
Determine HSP (the monitor high MEjl alarm setpoint above background (nepm)).
j HSP = C.R.
I (2.)-7) j 0.8 u
i 0.8 : A correction factor to increase the monitor setpoint to prevent spurious alarms caused by deviations in the j
mixture of radionuclides that l
affects monitor response.
c.
The monitor high alarn setpoint above background shall be set at or below this HSP value when this optional method is selected.
The i
maximum discharge flow shall not i
exceed the value of f as determined r
in Section 2.1.2.1 when this optional method is selected.
l r
2.1.3 Discharge Canal Monitor l
The following method determines the high alarm setpoint for the Discharge Canal Monitor (R-21) i during all operational conditions.
IBM 1
j
I H4 Rev. 12 Page 12
.v 2.1. 3.1 De termine the " mix" ( radionuclides anc composition) of the liquid effluent.
a.
Determine the liquid source terms that are representative of the
" mix" of all liquids released _into the discharge canal.
Liquid source terms are the total curies of each isotope released during the previous month.
Table 2.1-1 source terms may be used if there have been no liquid releases.
b.
Determine the activity concentrations (ACi) of all non-gamma emitters including H-3, Sr-89, Sr-90, Fe-55, and alpha activity.
c.
Determine NGr (the total fraction of the MPC in the liquid released to the discharge canal) for all non-gamma emitting nuclides.
The volume used to calculate the non-gamma emitting activity concentrations is the volume released via cooling tower blowdown during a one month 40%
period at the minimum flow rate of
'TEJ 67,300 gpm.
AC_1 NGF MPCi i
Activity concentration of where:
ACi nuclide 'i' released to the discharge canal (pCi/ml) l The liquid effluent MPCi radioactivity concentration limit for radionuclide 'i' (pci/ml) from Table 2.1-1 or Reference 3.
Detshmmaemittingradioactivity ine Si (the fraction of d.
the in the liquid released to the discharge canal comprised by radionuclide 'i') for each individual radionuclide released to the discharge canal.
IBM
114 Rev. 12 Page 13 l
Si Ai
)]i
^
where:
Ai The radioactivity of gamma emitting radionuclide 'i' released to the discharge canal.
e.
Determine VGF (the sum of fractional activities weighted by the MPC) for the gamma emitting nuclides released to the discharge canal.
VGF
=
Si MPCi i
where:
MPCi The liquid effluent radioactivity concentration limit for radionuclide 'i' (pCi/ml) from Table 2.1-1 or Reference 3.
.e-2.1.3.2 Determine C (the maximum acceptable total radiohetivity concentration of
- - O ganna emitting nuclides released to the discharge canal (pCi/ml)).
Ct
=
3-NGF WGF 2.1.3.3 Determine C.R. (the calculated monitor count rate above background attributed to the radionuclides (nepm)).
C.R.
is obtained by using the applicable Effluent Monitor Efficiency Curve located in the Radiation Monitor Calibration file.
C.R.
is the count rate that corresponds to the " adjusted" total radioactivity concentration (Ct).
2.1.3.4 The monitor high alarm se point above background (nepm) shall be set at or below the C.R. value.
IBM
l H4 I
Rev. 12
(
Page 14 j
TABLE 2.1-1 LIQUID SOURCE TERMS WASTE EFFLUENT SGBD RADIONUCLIDE MFC (pCi/ml)
(A,)(Ci/Yr)
(A,)(C1/Yr) l
-Mo-99 4E-5 6.42E-3 1.415E-2 i
1-131 3E-7 3.061E-2 4.11E-2 i
Te 132 2E-5 2.12E-3 3.61E-3 I-132 8E-6 2.83E-3 1.88E !
I-133 IE-6 2.365E-2 4.856E !
Cs-134 9E-6 1.464E-1 4.047E-2 6
1-135 4E-6 4.84E-3 1.792E-2 Cs-136 6E-5 5.743E-2 1.862E-2 Cs-137 2E-5 8.214E-2 2.69E-2 All Others 1E-7 0
2E-5 H-3 3E-3 1.89E2 1.41E2 Noble gases-2E-4 Total 1.894E2 1.412E2
.~.
t Whhn h
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. -. _, _, - -..=
.. ~ -.. -. -
. -.. ~..-
.. -. - - - ~.
Rev. 12 i
Page 15 2.2 ' Comp _11ance With_10 CFR-20 3
In order to comply.with 10 CFR 20, the concentrations of-radionuclides-in liquid effluents will not exceed the maximum permissible concentrations (MPC) as defined in j
Appendix B. Table 11 of 10 CFR 20.
For continuous releases, the alarm trip setpoints discussed in Section 2.1 will assure that these concentrations-are not exceeded.
For bate; aleases, concentrations of radioactivity in effluents l
priot
.o dilution will be determined, providing protection in addition to the alarm trip setpoint discussed in l
Section 2.1 Concentration in diluted effluents will be calculated using these results.
t 2.2.1 Continuous Releases i
~
Continuous liquid releases can occur from PNGP
(
through steam generator blowdown.
The alarm i
2 trip setpoints discussed in Section 2.1 will 7
assure that releases from this pathway will not exceed the limits of 10 CFR 20.
]
Other minor releases of a continuous nature have i
occurred at PNGP through the turbine building i
sump system.
These releases were minor and are not expected to occur in the future.
However, a 44%
continuous composite sampler will be maintained 63 9 at the discharge from the-turbine building sump with sar.ples being taken and analyzed weekly.
l If these samples indicate detectable levels of radionuclides, the methodologies given in i
Section 2.2.2 will be applied to the turbine sump weekly releases and the limit in Equation i
2.-2-2_will be lowered to account for this source term.
i 2.2.2 Batch Releases
}
To further show compliance with 10 CFR 20, the radioactivity content of each batch release will be determined prior to release.
The concentration of the various radionuclides in 1
the batch release prior to dilution, is divided by the minimum dilution flow to obtain the i
concentration at the site boundary.-
This j
calculation is shown in the following equation:
1 i
C R
r I
(2.2-1) i Conc.
=
1 MDF I
where l
Conc
= concentration of radionuclide i at the
{
site boundary, pCi/ml; l
h IBM I
e
~- ----a-----
H4 Rev. 12 Page 16 concentration of radionuclide i in the C.
=
2 potential batch release, pCi/ml; release rate of the batch R
=
MDF
= minimum dilution flov (=67,300 gpm)
The projected concentration at the site boundary is compared to the MPCs in Appendix B. Table 11 of 10 CFR 20 which are given in Table 2.1-1.
Before a release may occur, Equation 2.2-2 must be met for all isotopes.
{'
Conc 1
0.9 (2.2-2) i MPC g Maximum permissible concentration MFC
=
1 of radionuclide i from Appendix B, Table 11 of 10 CFR 20. pCi/ml The surration has been reduced frcra 1.0 to 0.9 to account for simultaneous continuous releases from steam generator blevdown as given in Section 2.1.1.5.
As noted earlier, this traction nay be adjusted based on experience. The su: ration
,'M cf all source terms shall not be greater than 1.0 cf the 10 CG 20 limit.
Since the volume of the discharge pipe vill contain the volume of 2 to 3 waste batch tatnks, to ensure ccepliance with 10 CG 20 when the traximum acceptable discharge flow rate, as calculated in section 2.1.2, in less than the maximum possible release rate from all release sources, the discharge pipe rhall be flushed with a volume of at least the volume of the f.iecharge pipe. The flush rate shall not exceed the naximum discharge flow rate and may be accortplished with water from other release paths.
If mere than one vaste batch tank requiring flushing are to be released, the discharge pipe may be flushed following the final tank release.
15,500 gal.
Velume of discharge pipe
=
2.3 Llauid Effluent Doses - Comoliance with 10 cm 50 Doses resulting fron. liquid effluents vill be calculated mont.hly to show cocpliance with 10 cm 50. A curulative su:mation of total body and organ doses for each calendar quarter and calender year vill be r:aintained as well as projected doses for the next month.
Since Fe-55, Sr-89, Sr-90, and alpha concentrations are determined from composite samples, the monthly liquid effluent dose calculations and comparisons to quarterly and annual limits should be completed using the most recent available composite sample results. The quarterly and annual dose calculations shall be ccepleted using the actual composite sample results.
-H4 Rev. 12 Page 17 The limits of 10 cm 50 are on a per reactor unit basis. 'Ihe liquid ra haste system at PINGP is shared by both reactor units naking it impossible to separate the releases of the two units. The releases that can be separated by unit, steam generator blevdown and turbine building sump releases, contribute a very s::all portion of the total liquid releases from PINGP. Therefore, for compliance with 10 CPR 50 the releases from both units vill be summed and the limits of Appendix I will be doubled.
~
2.3.1 Determination of Liquid Ef fluent Dilution To determine doses from liquid effluents the near field average dilution factor for the period of release must be calculated.
This dilution factor must be calculated for each batch release and each continuous release mode.
The dilution factor is determined by:
Rk F
(2.3-1) k X ADF p where Y
R
=
release rate of the batch or k
continuous releases during the period, k, gpm ADF
=
k average dilution flow durinF the time period of release k, gpm.
The value of X is the site specific factor for the mixing effect of the PNGP discharge structure.
This value is 10 for PNGP while (1) operating in the closed cycle cooling mode.
The product o is limited to 1000 (4.5x10(XandADFTh!!re f ore, since blowdown cfs gpm).
flow in closed cycle is 150 cfs, the d'" "i"*' #
5 of Equation 2.3-1 is always 4.5 x 10 in closed cycle.
In once through or helper made, the value of X is reduced to 1.0.
l
1 H4 I
Rev. 12 Page 18 cs n-2.3.2 Dose Calculations The dose contribution from the release of liquid effluents will be calculated monthly.
The dose I
contribution-will be calculated using the-following equation:
where:
j Dt
=
IIA t
C F
(2.3-2)
{
it k
ik k
ki t
where:
Dt
=
the dose commitment to the total body or any' organ t, from the liquid ef fluents for the period of release, mrem, C
=
ik the average concentration of
-radionuclide, 1 in undiluted liquid effluent for liquid release k, pCi/ml; fi t
A, the site related ingestion dora
{
=
g commitment-factor to the tota' (jg.y, body or any organ t for each g
identified principal. gamma and beta emitter, mrem /hr per pCi/ml; F
=
k the near field average dilution j
factor for C during liquid effluent reibbsek, j
g the duration of release k, hours.
t
=
h The dose factor A was calculated for an adult for each isotope bling the following equation:
1.14 x 105 [21BF DF J
(2.3-3)
A
=
g7 f
17 where:
8 5
6 3
1.14 x 30
=
10 pCi x 10 ml x 1 yr pCi
~T 8760 hr 21
=
adult fish consumption, Kg/yr; BF
=
bioaccumulation factor for i
radionuclide i in fish from Table A-1 of Regulatory Guide 1 109 Rev. 1 (5) pC1/Kg per.pCi/1; dose conversion factor for DF..
=
ll radionuclide i for adults for a particular organ I from Table E-ll of Regulatory Guide 1.109 Rev. 1, (5) mrem /pC1.
t l
1 1
H4 Rev. 12 Page 19 i
f f
~
A table of A values for an adult at the PNGP are presentebl in Table 2.3-1.
Mississippi River water is not used as a potable water supply within 300 miles downstream of the PNGP.
Wells i
are used for irrigation downstream of the plant.
]
2.3.3 cumulation of Doses i
Doses calculated monthly will be sunned fer comparison with quarterly and annual limits.
The monthly results should be'added to the dosts cumulated f rom the other months in the quarter i
of interest and in the year of interest for the combined releases of both reactor units.
The following relat anships should hold:
i 1
For the quarter, j
7 3.0 mrem total body (2.3-4)
D D
10 mrem any organ 1(2.3-5) a.3 r
, ( M4 For-the calendar year.
t l
D 6.0 mrem total body (2.3-6) t c.
20 mrem any organ (2.3-7)
{
D, i
The quarterly limits given above represent one half i
of the annual design objective.
If these quarterly I
or annual limits are exceeded, a special report
[
should be submitted to the USNRC identifying the i
cause and correct'ive action to be taken.
If twice the quarterly or annual limits are exceeded, a special report shall be submitted showing i
j compliance with 40 CFR 190.
i l
i-
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?
L I
t
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-pir+ys-rP=e-wea-*
--emf ^
w mesr ae.* paved
- -=M--
-adde--w-eetete m v m.
ww w ee r-ar - -+e wn M w s '-'
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9--
P-wMw 'w W-'
!!4 Rev. 12 Page 20 2.3.4 Projection of Doses Anticipated doses resulting from the release of liquid effluents will be projected monthly.
If the projected doses for the month exceed 2 percent of Equation 2.3 6 or 2.3-7, additional components of the liquid radwaste treatment system will be used to process waste.
The projected doses will be calculated using Equation 2.3-2.
The dilution factor, T "ill be calculated ty replacing the term ADFk'in Equation 2.3-1withthetermMDFfromChuation 2.2-1.
The total source term utilized for the most recent dose calculation should be used for the projections unlest information exist s indicating that actual releases could differ significantly in the next month.
In this case, the source term would be adjusted to reflect this information and the justification for the adjustment noted.
This adjustment should account for any radwaste equipment which was operated during the previous month that could be out of service in the esming month, m1%
REFERENCES 1.
" Prairie Island Final Environmentai Statement," USAEC, May,
- 1973, p.
\\'-26.
Prairie Island Nuclear Generating Plant. Appendix 1 2.
"NSP Analysis Supplement No. 1 - Docket No. 50-282 and 50-306,"
Table 2.1-1.
3.
"10 CFR 20," Appendix B, Table ll, Column 2.
4.
"NSP - Prairie Island Nuclear Generating Plant, Appendix I Analysis Supplement No. 1 docket 50-282 and 50-306," July 21, 1976, Table 2.1-2.
5.
U.
S.
Nuclear Regulatory Commission, " Regulatory Guide 1.109
- Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Compliante with 10 CFR 50, Appendix 1," Rev.
1, 1977 i
I i
H4 Rev. 12 Page 21 TABLE 2.3-1 a
A VALUES FOR THE PRAIRIE ISLAND NUCLEAR GENERATING PLANT tt (MREM /HR PER pC1/ML)
KUCLIDE BONE-LISTR T. BODY THYROID KIDNEY LUNO GI-LL1 2H i
0.00E-01 2.26E-01 2.26E-01 2.26E-01 2.26E-01 2.26E-01 2.26E-01 6C 14 3.13E 04 6.26E 03 6.26E 03 6.26E 03 6.26E 03 6.26E 03 6.26E 03 11NA 24 4.07E 02 4.07E 02 4.07E 02 4.07E 02 4.07E 02 4.07E 02 4.07E 02 24CR 51 0.00E-01 0.00E-01 1.27E 00 7.61E-01 2.61E-01 1.69E-00 3.20E 02 25MN 54 0.00E-01 4.38E 03 8.35E 02 0.00E-01 1.30E 03 0.00E-01 1.34E 04 25MN 56 0.00E-01 1.10E 02 1.95E 01 0.00E-01 1.40E 02 0.00E-01 3.51E 03 26FE 55 6.58E 02 4.15E 02 1.06E 02 0.00E-01 0.00E-01 2.54E 02 2.61E 02 26FE 59 1.04E 03 2.44E 03 9.36E 02 0.00E-01 0.00E-01 6.E2E 02 8.14E 03 27CO $8 0.00E-01 8.92E 01 2.00E 02 0.00E-01 0.00E-01 0.00E-01 1.81E 03 27C0 60 0.00E-01 2.56E 02-5.65E 02 0.00E-01 0.00E-01 0.00E-01 4.81E 03 28NI 63 3.11E 04 2.16E 03 1.04E 03 0.00E-01 0.00E-01 0.00E-01 4.50E 02 28NI 65 1.26E 02 1.64E 01 7.49E 00 0.00E-01 0.00E-01 0.00E-01 4.17E 02 29CU 64 0.00E-01 9.97E 00 4.68E 00 0.00E-01 2.51E 01 0.00E-01 8.50E 02 30ZN 65 2.32E 04 7.37E 04 3.33E 04 0.00E-01 4.93E 04 0.00E-01 4.64E 04
-30ZN 4.93E 01 9.43E 01 6.56E 00 0.00E-01 6.13E 01
__0.00E-01 1.42E 01 35BR 83 0.00E-01 0.00E-01 4.04E 01 0.00E-01 0.00E-01 0.00E-01 5.82E 01 35BR 84 0.00E-01 0.00E-01 5.24E 01 0.00E-01 0.00E-01 0.00E-01 4.11E-04 35BR 85 0.00E-01 0.00E-01 2.15E 00 0.00E-01 0.00E-01 0.00E-01 1.01E-15 37RB 86 0.00E-01 1.01E 05 4.71E 04 0.00E-01 0.00E-01 0.00E-Oi 1.99E 04 37RB 88 0.00E-01 2.90E 02 1.54E 02 0.00E-01 0.00E-01 0.00E-01 4.00E-09 n;:_ s 37RB 89 0.00E-01 1.92E 02 1.35E 02 0.00E-01 0.00E-01 0.00E-01 1.12E-11
\\U ?
385R 89 2.21E 04 0.00E-01 6.35E 02 0.00E-01 0.00E-01 0.00E-01 3.55E 03 38SR 90 5.44E 05 0.00E-01 1.34E 05 0.00E-01 0.00E-01 0.00E-01 1.57E 04 38SR 91 4.07E 02 0.00E-01 1.64E 01 0.00E-01 0.00E-01 0.00E-01 1.94E 03 385R 92 1.54E 02 0.00E-01 6.68E 00 0.00E-01 0.00E-01 0.00E-01 3.06E 03 l
39Y 90 5.76E-01 0.00E-01 1.54E-02 0.00E-01 0.00E-01 0.00E-01 6.10E 03 l
39Y 91M 5.44E-03 0.00E-01 2.11E-04 0.00E-01 0.00E-01 0.00E-01 1.60E-02 39Y 91 8.4I.E 90 6.00E-01 2.26E-01 0.00E-01 0.00E-01 0.00E-01 4.64E 03 39Y 92 5.06E-02 0.00E-01 1.48E-03 0.00E-01 0.00E-01 0.00E-01 8.86E 02 39Y 93 1.60E-01 0.00E-01 4.43E-03 0.00E-01 0.00E-01 0.00E-01 5.09E 03 40ZR 95 2.40E-01 7.70E-02 5.21E-02 0.00E-01 1.21E-01 0.00E-01 2.44E 02 40ZR 97 1.33E-02 2.68E-03 1.22E-03 0.00E-01 4.04E-03 0.00E-01 8.30E 02 41NB 95 4.47E 02 2.48E 02 1.34E 02 0.00E-01 2.46E 02 0.00E-01 1.51E 06 42MO 99 0.00E-01 1.03E 02 1.96E 01 0,00E-01 2.34E 02 0.00L-01 2.39E 02 437C 99M 8.87E-03 2.51E-02 3.19E-01 0.00E-01 3.81E-01 1.23E-02 1.48E 01 43TC101 9.12E-03 1.31E-02 1.29E-01 0.00E-01 2.37E-01 6.72E-03 3.95E-14 44RU103 4.43E 00 0.00E-01 1.91E 00 0.00E-01 1.69E 01 0.00E-01 5.17E 02 44RU105 3.69E-01 0.00E-01 1.46E-01 0.00E-01 4.76E 00 0.00E-01 2.26E 02 44RU106 6.58E 01 0.00E-01 8.33E 00 0.00E-01 1.27E 02 0.00E-01 4.26E 03 47AG110M 8.81E-01 8.15E-01 4.84E-01 0.00E-01 1.60E 00 0.00E-01 3.33E 02 l
52TE125M 2.57E 03 9.30E 02 3.44E 02 7.72E 02 1.04E 04 0.00E-01 1.02E 04 52TE127M 6.48E 03 2.32E 03 7.90E 02 1.66E 03 2.63E 04 0.00E-01 2.17E 04 52TE127 1.05E 02 3.78E 01 2.28E 01 7.80E 01 4.29E 02 0.00E-01 8.31E 03 i
52TE129M 1.10E 04 4.11E 03 1.74E 03 3.78E 03 4.60E 04 0.00E-01 5.54E 04 l
52TE129 3.01E 01 1.13E 01 7.33E 00 2.31E 01 1.26E 02 0.00E-01 2.27E 01 l
52TE131M 1.66E 03 8.10E 02 6.75E 02 1.28E 03 8.21E 03 0.00E-01 8.04E 04 l'
52TE131 1.89E 01 7.88E 00 5.96E 00 1.55E 01 8.26E 01 0.00E-01 2.67E 00 52TE132 2.41E 03 1.56E 03 1.47E 03 1.72E 03 1.50E 04 0.00E-01 7.38E 04 53I 130 2.71E 01 8.01E 01 3.16E 01 6.79E 03 1.25E 02 0.00E-01 6.89E 01 53I 131 1.49E 02 2.14E 02 1.22E 02 7.00E 04 3.66E 02 0.00E-01 5.64E 01 L
IBM l
~..
___.__-___-_.m.__
Rev. 12 i
Page 22 i
TABLE 2.3-1 (CONT.)-
l A
VALUES EOR THE PRAIRIE ISLAND NUCLEAR GENERATING PLANT j
g (MRIM/}D1 PER pCI/ML) i NUCLIDE BONE LIVER T. BODY THYROID KIDNEY LUFG GI-LLI l
531 132 7.29E 00 1.95E 01 6.82E 00 6.82E 02 3.11E 01 0.00E-01 3.66E 00
}
$31 133 5.10E 01 8.87E 01 2.70E 01 1.30E 04 1.55E 02 0.00E-01 7.97E 01 l
53I 134 3.61E 00 1.03E 01 3.70E 00 1.79E 02 1.64E 01 0.00E-01 9.01E-03 53I 135 1.59E 01 4.17E 01 1.54E 01 2.75E 03 6.68E 01 0.00E-01 4.70E 01
$5CS134 2.98E 05 7.09E 05 5.79E 05 0.00E-01 2.29E 05 7.61E 04 1.24E 04 55CS136 3.12E 04 1.23E 05 8.86E 04 0.00E-01 6.85E 04 9.38E 03 1.40E 04
$5C5137 3 82E 05 5.22E 05 3.42E 05 0.00E-01 1.77E 05 5.89E 04 1.01E 04 l
55CS138 2.64E 00 5.22E 02 2.59E 02 0.00E-01 3.84E 02 3.79E 01 2.23E-03 56BA139 9.29E-01 6.62E-04 2.72E-02 0.00E-01 6.19E-04 3.75E-04 1.65E 00 l
56BA140 1.94E 02 2.44E-01 1.27E 01 0.00E-01 E.30E-02 1.40E-01 4.00E 02
?
56BA141 4.51E-01 3.41E-04 1.52E-02 0.00E-01 3.17E-04 1.93E-04 2.13E-10
)
56BA140 2.04E-01 2.10E-0*
1.2BE-02 0.00E-01 1.77E-04 1.19E-04 2.37E-19 j
57LA140 1.50E-01 7.54E-02 1.99E-02 0.00E-01 0.00E-01 0.00E-01 5.54E OS 571A162 7.66E-03 3.46E-03 8.68E-04 0.00E-01 0.00E-01 0.00E-01 2.54E 01
[
58CE141 2.24E 1.52E-02 1.72E-03 0.00E-01 7.04E-03 0.00E-01 5.79E 01 j
SSCE143 3.0$E-03 2.92E 00 3.23E-04 0.00E-01 1.29E-03 0.00E-01 1.09E 02 58CE144 1.17E 00 4.88E-01 6.27E-02 0.00E-01 2.90E-01 0.00E-01 3.95E 02-i I
59PR143 5.51E-01 2.21E-01 2.73E-02 0.00E-01 1.27E-01 0.00E-01 2.41E 03 59PR144 1.80E-03 7.48E-04 9.16E-05 0.00E-01 4.22E-04 0.00E-01 2.59E-10 60ND147 3.76E-01 4.35E-01 2.60E-02 0.00E-01 2.54E-01 0.00E-01 2.09E 03 yW.3 74W 187 2.96E 02 2.47E 02 8.65E 01 0.00E-01 0.00E-01 0.00E-01 8.10E 04 4.V 93NP239 2.85E-02 2.80E-03 1.54E-03 0.00E-01 8.74E-03 0.00E-01 5.75E 02 l
+
l i
l
[
I t
i
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i 6
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f I
t 1
1
{
IBM l
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,, _ ~... _. _ _ _. _.. _., - _, ~,.. - _.. _ _,.... _ _... - _ _ _ -..,. _., _. _ -...,,.,... _ - - - - -,., -,.. -.. _.., _.,. - _........
114 Rev. 12 Page 23 3.0 GASEOUS EFFLUENTS 3.1 Moni tor Ala rm Setpoint Determination This procedure determines the monitor alarm setpoint that indicates if the dose rate at or beyond the site boundary due to noble gas radionuclides in the gaseous effluent released from the site exceeds 500 mrem / year to the whole body or exceeds 3000 mrem / year to the skin.
Monitor high alarm or isolation setpoints will be established in one of the following ways:
a.
Monthly calculation of setpoint using the methodology of Section 3.1.1 for continuous releases using previous month releases as cource term, b.
Prior to each containment purge, recalculation of the setpoint using the methodology of Section 3.1.1 based on the sample taken prior to purging.
c.
In lieu of (a) and (b) above, alarm setpoints may be established using the methodology of section 3.1.1 using conservative assumptions (e.g.,
100%
Kr-89).
No recalculation of setpoints is necessary unless an increase is desired.
/9 PWR GALE Code source terms (Table 3.1-2) may be used if there were no detectable isotopes in the previous month or in the analysis prior to purging.
If the newly calculated setpoint is less than the existing monitor setpoint, the setpoint will be reduced to the new value.
If the calculated setpoint is greater than the existing setpoint, the setpoint may remain at the lower value or increased to the new value.
3.1.1 Effluent Monitors The following method applies when determining the isolation or high alarm setpoint for the monitors listed in Table 3.1-1.
3.1.1.1 Determine the " mix" (noble gas radio-nuclides and composition) of the gaseous er' fluent.
a.
Determine the gaseous source terms that are representative of the gaseous effluent.
Gaseous source te rms are the total curies of each noble gas released during the previous month or a repre-sentative analysis of the gaseous effluent.
Table 3.1-2 source terms may be used if the releases for the previous month were below the lower limits of detection (LLD).
IBM
_ -. _ ~...
I H4 Rev. 12 Page 24 b.
Determine S (the fraction of the total radiobctivity in the gaseous
{
effluent compriced by noble gas i
radionuclide "i") for each individual nobic gas radionuclide in the gaseous effluent.
l S*.
=
A (3.1-3) 1A 1 i
1 The radioactivity of noble A
gas radionuclide "i" in the gaseous effluent from Table 3.1-2.
3.1.1.2 Determine Q. (the maximum acceptable total releabe rate of all noble gas i
radionuclides in the gaseous effluent (pci/sec)) based upon the whole body.
exposure limit.
=
500 (3.1-2)
Q t (x/Q) 1 KSgg I
i (x/Q)
The highest calculated annual
=
t average relative concentration
~
of ef fluents released via the plant vents for any area at or i
beyond the site boundary for 3
all sectors (sec/m ) from the
[
"x/Q" column in Table 3.1-1.
K
=
The total whole body dose factor due g
to gamma emissions from noble gas, radionuclide "i" (mrem / year /pCi/m#)
from Table 3.1-3.
i 3.1.1. 3 Dete rmine Q based upon the skin exposure i
t limit.
=
3000 (3.1-3)
Q (x/Q) 1 (Lg+l1M)
S 1
i 3
1 i
t The total skin dose factor Lt + 1.1 Mg
=
due to emissions from noble gas radionuclide "i" (mrem /
i year /pci/m ) from Table 3.1-3.
J 8
f 1BM.
t
- l,
H4 Rev. 12 Page 25 3-.1.1.4 Dete rmine C (the maximum acceptable total radiohetivity concentration of all noble gas radionuclides in the gaseous effluent (pCi/cc)).
C
=
2.12 E-3 9 t (3.1-4) t 7
NOTE:
Use the lower of the Q'1.1.2 values obtained in Sections 3.
and 3.1.1.3.
F = The maximum effluent flow rate at the point of release (cfm) from the." Effluent Flow Rate" column in Table 3.1-1.
2.12 E-3 = Unit conversion constant to convert pCi/sec/cfm to pCi/cc.
-3.1.1.5 Determine C.R. (the calculated monitor count rate above background attributed to the noble gas radionuclides (ncpm)).
C.R.
is obtained by using the applicable Ef fluent Monitor Efficiency Curve located
~ng in the Radiation Monitor Calibration file.
WW
~
C.R.
is the count rate point that corresponds to the total radioactivity concentration (C ).
g 3.1.1.6 Determine HSP (the monitor high alarm setpoint above background (nepm)).
HSP = T, C.R.
(3.1-5)
T*
= Fraction of the total radioactivity from the site that may be released via each release point to ensure that the site boundary limit is not exceeded due to simultaneous releases from several release l
points from the " Release Fraction" column in Table 3.1-1.
3.1.1.7 The isolation or high alarm setpoints above background (nepm) for the monitors should be set at or below the HSP values.
3.1.2 Air Ejector Monitors l
Radiation monitors 1R-15 and 2R-15 provide an L
indication of gross noble gas activity at the l-main condenser air ejector of Unit I and Unit 2, l
respectively.
These monitors are provided to l
l IBM i
t-
H4 Rev. 12 Page 26 give rapid indication of steam generator tube leakage.
They are not effluent monitors since the air ejectors are vented to the auxiliary building vents during normal plant operation and releases are monitored by the auxiliary building vent monitoring system, a
IBM
n W.h g,/ :
' lI6 Rev. 12 Page 27 TABLE 3.1-1
?!ONITOR ALARt1 SETPOINT DETER 111 NATION FOR PNCP l
l l
SOURCE l
SOURCE
-l l
EFFLUENT 1
REl. EASE I
l l
RELEASE I
0F l
TERMS (A.) l l
FLOW RATE l FRACTION I
1 ){
X/Q(sec/m )
l (F) (cfm) l (Tm)
.I l
MONITOR I
POINT l
RELEASE l
(TAnl.E 3.1 2
l 1
l l
I I
i l
l l
l l
l
[
1 1R-30 1
Aux. Bldg.
l Aux. B l.lg.
l and i Vent - Unit. 1 I Unit 1 Exhaust l Aux. Bldg.
l 3.38E-5 1
2.9E+4 l
0.7 l
l 1R-37 l
l Air Ejector l
Air Eje tur i
3.3ME-5 1
2.9E*4 I
l l
l Unit I l
l l
{
_t f
I I
I l
1 i
i I
I 2R-30 i
Aux. Bldg.
1 Aux.' Bldg. -
l l
l 1
l l
and l Vent - Unit 2 l Unit 2 Exhausti Aux. Bldg.
l 3.38E-5
_I-4.1E+4 1
0.3 I
[
l 2R-37 l
Itlas Uecay Tanksi Xe-133 (1007.)l 1.32E-4 l
4.IE+4 l
l t
'l i
I I
i I
l l
l Air Ejector i Air Ejector i
3.38E-5 1
4.lE+4 l
1 I
I I Unit 2 l
l 1
i
_1 i
i l
l l
l l
l l
1 i
l 1R-12 l Shield Bldg. 1 Cont. - Units l Shield Bldg. I 1.32E-4 1
3.2E+4 1
0.3 l
l l
and l Vent - Unit 11 IF.2 Purge, I
I I
(Note 2)
I j
l l
1R-22 l
l Unit. 1 Inserv-l l
l l
l i
l 1
l l ice Purge i
l l
l 1
l 1
1 I
l 1
1 l
j l
2R-12 l Shiel.1 Bldg. ! Cont.-Unit 2 l Shield Bldg. I 1.32E-4 l
4.6E+3 1
0.3 l
l l
and i
Vent Unit 2l Inservice l
l l
1 l
l l
2R-22 l
l Purge _
l 1,
I l
l
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i i
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i l
i i
l l
l R-35 l Radwaste l Radwaste l
Aux. Bl.lg.
I 3.38E-5 1
6.1E+3 1
0.1 l
[
{
l l Bldg. Vent l Bldg. Exhaust l l
l l
l 1
l l
1 1
I I
l 1
i l
l R-25 l Spent Fuel l Spent Fuel l
Aux. B l.l g.
l 3.38E-5 l
1.8E*4 1
0.1 l
l l
and l Pool l Pool Air l
l l
l l
l l
l R-31 i Air Vent l Exhaust l
l l
1 l
[
l NOTES:
{
l i
1.
Values listed for T are norninal values only.
They may be adjasted as necessary to allow a reasonable i
margin to the monitEr setpoint. Duplicate values of T are assigned to both Shield Building vents since f
only one containment will be purged at any one time.
The assigned T values of all active release points i
shall not be greater than unity.
l t
=
l 2.
When purging the Unit I containment via the inservice purge system, the monitor setpoints s>ay be bas-d on 4.6E+3 cfm for th7 duration of the release.
l
,i l
..-- -.,,_, - ~,,....-.-..,,_,,,...- -,,.-......_ - -,
.~. -._ __.-.,-..~._
_.-.m ll4 Rev. 12 i
II)
TABLE 3.1-2 GASEOUS SOURCE TER}1s AUX. BLDG SHIELD BLDG.
AIR EJECTOR RAD _I_ONUC L I DE (A ).(Ci/Yr)
(A )-(C1/Yr)
(A ) (Ci/Yr) f g
g Kr-85m 3E0 2EO-Kr-85 2E0 2.2E1 Kr-87 1EO Kr-88 SEO 1EO 3E0 Xe-131m 2E0 2.1El 1EO Xe-133m SEO 2E1 3E0 Xe-133 3.7E2 2.7E3 2.3E2 Xe-135 8E0 6EO SEO t
Xe-138 1EO j
Total 3.97E2 2.77E3 2.44E2 i
h indicates that the release is less than 1 Ci/yr.
t 1
l t
i Ar E rf,3
(
f t
i I
t I
3 f
I f
i I
t I
I t
ii i
IEM
'l I'
-.....-. ---...a
i t
H4.
Rev. 12 l
Page 29 i
TABLE 3.1-3
-l DOSE FACTORS AND CONSTANTS i
SKIN TOTAL VHOLE BODY DOSE FACTOR DOSE FACTOR (L
+ 1.1M )
RADIONUCLIDE (K ) (mrem /yr/pCi/m )
(mrkm/yr/pCk/m) g I
Kr-83m 7.56E-2 2.12E1 Kr-85m 1.17E3 2.81E3 l
Kr-85 1.61El 1.36E3 i
Kr-87 5.92E3 1.65E4 l
Kr-88 1.47E4 1.91E4 1
Kr-89 1.66E4 2.91E4 l
Kr-90 1.56E4 2.52E4 i
Xe-131m 9.15El 6.48E2 l
Xe-133m 2.51E2 1.35E3 Xe-133 2.94E2 6.94E2 Xc-135m 3.12E3 4.41E3 Xe-135 1.81E3 3.97E3 Xe-137 1.42E3 1.39E4 Xe-138 8.83E3 1.43E4 l
Ar-41 8.84E3 1.29E4 r
i 46 3.2 Gaseous Effluent
- _;;w Dose Rate - Compliance with 10 CFR 20 r
Dose rates resulting'from the release of noble gases, I
and radioiodines and particulates must be calculated to show compliance with 10 CFR 20.
The limits of 10 CFR 20 i
must be met on an instantaneous basis at the hypothetical worst case location, and apply on a per site basis.
l Since Sr-89 and Sr-90 concentrations are determined from composite samples, the pre-release, weekly and monthly
}
airborne dose calculations and comparisons to quarterly and annual limits should be completed using the most recent available composite sample results.
The quarterly dose values and critical receptors reported semi-annually to the USNRC shall be calculated using the actual composite results.
i f
f i
i I
-IBM r
H4 Rev. In Page 36 3.2.1 Noble Gases The dose rate at the site boundary resulting from noble gas effluents is limited by 10 CFR 20 to 500 mrem /yr to the total body and 3000 mrem /yr to the skin.
The setpoint determinations discussed in the previous section are based on the dose calculational method presented in NUREG-0133.(2)
They represent a backward solution to the limiting dose equations in NUREG-0133.
Setting alarm set trip points-in t his n.anner will assure that the limits of 10 CFh 20 are met for noble gas releason.
Therefore, no routine dose calculations for noble gases will be needed to show compliance with this part.
Routine calculations will be made for doses from noble gas releases to show compliance with 10 CFR 50, Appendix 1 as discussed in l
. Section 3.3.1.
3.2.2-Radiciodine, Radioactive Particulates, and Other Radionuclides I
The dose rate at the site boundary resulting from the release of radiciodines and particulates with half lives greater than 8 days is limited by 10 q$2; CFR 20 to 1500 mrem /yr to any organ.
Calculations Cf showing compliance with this dose rate limit will be performed for batch releases prior to the release and weekly for all releases.
The calculations l
will be based on the results of sample analyses l
obtained pursuant to the PNGP Technical Specifications.
To show compliance, Equations 3.2-1 will be evaluated for 1-131, tritium, and radioactive particulates with half-lives greater than eight l
days.
(x/Q )
h
< 1500 mrem /yr (3.2-1)
P;I I
y fy where P
child critical organ dose parameter for 1
=
7 radionuclide i for the inhalation pathway, 3 (Table 3.2-1);
mrem /yr per pCi/m annual average relative concentration for (x/Q )
lor.g-te rm release at the critical location, y
4 sec/m (Appendix A, Table A-3);
3 l
1BM 4
,, _.. ~, -
,,..___c.
-. -,,. ~ -.... ~. _,. _. - _,.., _,, _.
ti H4 Rev. 12 Page 31 the total release rate of radionuclide i Q
=
IV
'from all vents from both units for the batch or week of interest, pCi/sec; Radiofodines, tritium, and radioactive particulates will be released from up to six individual vents all within 300 feet of each other.
For showing com-pliance with 10 CFR 20, calculations based on Equa-tion 3.2-1 will be made once per week.
The source terms (QIV) will be determined from the results of analysis of vent particulate filters and charcoal canisters and vent flow rate.
These source terms include all gaseous releases from PNGP.
Significant short-term batch releases of long-lived radioactive particulates and tritium will result from containment purges.
Calculations will be made for these releases separately-to further assure compliance with 10 CFR Part.20 prior to release.
These calculations will be used only to determine whether or not the purge release will be allowed to occur.
Sou:.te te rms will be de termined f rom the -
results of isotopic analyses of samples from containment prior to release.
Equation 3.2.1 will be used in conjunction with the'following relationship to demonstrate that the batch Aug release does not exceed the dose rate limit:
M&P
~
BL u 1500 - (D
-D)
(3.2-2) y p
where BL =
limiting dose rate for the batch, mrem /yr; previous week's dose rate from all continuous D
=
y and batch releases mrem /yr; previous week's dose rate from all purge D
=
P releases mrem /yr.
3.2.3 Crftical Receptor Identification The atmospheric dispersion parameters given in Appendix A will be used to identify the critical receptor.
Compliance with Part 20 will always be shown at the site boundary location with the highest x/Q.
As discussed previously, weekly and batch dose calculations will be performed at this location.
The critical site boundary location, based upon long term ground level releases x/Q i
(Table A-3), is 0.36 miles in the WNW sector.
IBM l
L
114 Rev. 12 Page 32 t
TABLE 3.2-1 Values for Child p
l I
mremfyr I
3 I
ISOTOPE I
P pCi/m l
g7 l
l l
l H-3 l
1.12 E 3 l
l Cr-51 1
1.70 E 4 l
l Mn-54 l
1.38 E 6 l
l Fe-59 l
1.27 E 6 l
l Co-58 l
1.11 E 6 l
l Co-60 1
7.07 E 6 l
l Zn-65 I
9.95 E 5 l
l Rb-86 1
1.98 E 5 l
l Sr-89 l
2.16 E 6 i
i Sr-90 l
1.01 E 8 l
l Y-91 l
2.63 E 6 l
i.$)
I Zr
<.5 l
2.23 E 6 l
l Nb-95 l
6.14 E 5 l
l Ru-103 l
6.62 E 5 1
)
Ru-106 l
1.43 E 7 l
l Ag-110m l
5.48 E 6 I
I Te-127m i
1.48 E 6 l
l Te-129m i
1.76 E 6 l
1 Cs-134 1
1.01 E 6 l
l Cs-136 l
1.71 E 5 l
l Cs-137 l
9.07 E 5 l
l Ba-140 1
1.74 E 6 l
1 Ce-141 1
5.44 E 5 l
l Ce-144 l
- 1. 20 E 7 l
l I-131 l
1.62 E 7 1
1 1
1 IBM
t i
Rev 12 Page 33 Compliance with 10 CFR 50_
3.3 Gaseous Effluents Doses resulting from the release of noble gases, and radioiodines and particulates must be calculated to show e
compliance with Appendix 1 of 10 CFR 50.
The calculations will be performed monthly for all gaseous effluents.
The limits of 10 CFR 50 are on a per reactor unit basis.
The
[
gaseous radwaste treatment system and the auxiliary building at Ph'GP is shared by both reactor units making it impossible to separate the releases of the two units.
The releases that can be separated by unit contribute a very small portion of the total gaseous releases from PNGP.
Therclore, i
for compliance with 10 CFR 50 the releases from both units i
will be summed and the limits of Appendix 1 will be doubled.
Since Sr-89 and Sr-30 concentrations are determined from composite samples, the pre-release, weekly and monthly airborne dose calculations and comparisons to quarterly and annual limits should be completed using the most recent available composite sample results.
The quarterly dose values and critical receptors reported semi-annually to the USNRC shall be calculated using the actual composite results.
3.3.3 Noble Gas
[
<:x n 0 ;2 3.3.1.1 Dose Equations The air dose at the critical receptor due to noble gases released in gaseous ef fluents is determined by Equations l
3.3-1 and 3.3-2.
The critical receptor i
will be identified as described in Section 3.3.4 i
For gamma radiation:
3.17 x 10-8 y g{
(X79),
q (Xjq) q i
< 10 mrad for any calendar quarter l
< 20 mrad for any calendar year (3.3-t)
For beta radiation:
3 17 x 10'b i N
(x/Q)y Q
+ (x/q)yqiy g
3y
^
-l
< 20 mrad for any calendar quarter
< 40 mrad for any calendar year (3.3-2)
IBM l
t
H4 Rev. 12 Page 34 where:
.c air dose factor due to gamma emission P..
=
1 tor each identified noble gas radionuclide 1, mrad /yr per pCi/m3; (Table 3.3-1) the air dcsc factor due to beta emissions N
=
i for each identified noble gas radionuclide i, mrad /yr per pCi/m3; (Table 3.3-1)
(x/Q)',
the annual average relative concentration
=
for areas at or beyond the restricted area boundary for long-term vent releases 3
(greater than 500 hr/ year), sec/m (Appendix A Table A-4);
the relative concentration for areas at or (x/q)v
=
beyond the restricted area boundary for short-term vent releases (equal to or less than 500 hrs / year), sec/m3 (Appendix A.
Table A-7);
=
the total release of noble gas radionucli' q.iv i in gaseous ef fluents for short-term ve t releases from both units (equal to or less than 500 hrs / year), pCi; a
=
the total release of noble gas radionuclide SeFA Q.iv i in gaseous effluents for long-term vent releases from both units (greater than 500 hrs /yr), pCi;
-8 the inverse of the number of seconds in 3.17 x 10
=
a year.
Noble gases will be relcased from PNGP from up to six vents.
Long-term x/Q's were given in Appendix A.
Short-term x/q's were calculated using the USNRC computer code "XOQD0Q" assuming 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> per year short term releases and are given in Appendix A (Table A-7).
Values of M and N were calculated using the methodology presented in NUREG-0133 and are given in Table 3.3-1.
3.3.1.2 Cumulation of Doses Doses calculated monthly will be summed for comnarison with quarterly and annual limits.
The monthly results will be added to the doses cumulated from the other months in the quarter of interest and the year of interest and compared to IBM
H4 7
Rev.12 Page 35 the limits given in Equations 3.3-1 and-3.3-2.
If these limits are exceeded, a special report will be submitted to the USNRC in accordance with the PNGP Technical--Specifications.
If twice the limits are exceeded, a special report showing compliance with 40 CFR 190 will be submitted.
3.3.2 Radioiodine, Particulacea, and Other Radionuclides 3.3.2.1 Dose Equations The worst case dose to an indiv! dual from j
1-131, tritium, and radioactive particulates j
with half-lives greater than eigh* days i
in gaseous effluents released to unrestricted areas is determined by the following expressions:
During any calendar quarter or year-3.17 x 10'O IIR "v 1v * "v iv 9
9 ijak J i i
< 15 mrem (per_ quarter)
< 30 mrem _(per calendar year) ca (3.3-3) u v;r.n
~
where:
release of radionuclide i for long-term Q.
=
iv vent releases from both units (greater than 500 hrs /yr), pCi;
=
release of radionuclide i for short te rm q*."
purge.eleases from both units (equal to or less than 500 hrs /yr); pCi; the dispersion parameter for estimating W
=
y the dose to an individual at the controlling I
location for L ig-term vent releases (greater tb.a f00 hrs /yr);
the dispersion parameter for estimating the w
=
y dose to art individual at the controlling location for short-term vent releases (equal to or less than 500 hrs /yr);
-8 the inverse of the number of seconds in 3,17 x 10
=
a year, R
ijak =
the dose factor for each identified radionuclide i, pathway j, age group a, and organ k, m mrem /yr per pCi/sec or mrem /yr 2
per pCi/m3 IBM
H4 Rev. 12 i
I Page 36 The above equation will be applied to each combination of age group and organ.
Values of R
, have been calculated using the methodology gkdNi in NUREG-0133 and are given in Tables 3.3-2 through 3.3-20.
Dose factors for isotopes not listed will be determined in accordance with the methodology in Appendix C.
Equation 3.3-3 will be applied to a controlling location which will have one or more of the following:
residence, vegetable garden and milk animal.
The selection of the actual receptor is discussed in Section 3.3.4.
The source terms and dispersion parameters in Equation 3.3-3 are obtained in the same manner as in Section 3.2.
The V 3
values are in terms of x/Q(sec/m ) for the inhalation pathways and for tritium (Tables 2
A-4 and A-7) and in terms of D/Q(1/m ) for all other pathways (Tables A-5 and A-8).
3.3.2.2 Cumulation of Doses Doses calculated monthly will be summed for comparison with quarterly and annual limits.
The monthly results should be added to the doses cumulated from the other months in the quarter of interest
'A and in the year of interest and compared
' ^>
with the limits in Equation 3.3-3.
If these limits are exceeded, a special report will be submitted in accordance with the PNGP Technical Specifications.
If twice the limits are exceeded, a special report showing compliance with 40 CFR 190 will be submitted.
3.3.3 Projection of Doses Doses resulting from the release of gaseous effluents will be projected monthly.
The doses calculated for the present month will be used as the projected doses unless information exists indicating that actual releases could differ significantly in the next month.
In this case the source terms will be adjusted to reflect this information and the justification for the adjustment noted.
If the projected release of noble gases for the month exceeds 2 percent of the calendar year limits of equation 3.3-1 or 3.3-2, additional waste gas treatment will be provided.
If the projected release of I-132, tritium, and radioactive particulates with half-lives greater than 8 days exceeds 2 percent of the calendar year limit of equation 3.3-3, operation of the ventilation exhaust treatment equipment is required if not currently in use.
IBM
H4 Rev. 12 Page 37 3.3.4 critical Receptor-Identification s
The critical receptors for compliance with 10 CFR 50, Appendix 1 will be identified.
For the noble-gas specification-the critical location will-be based on the beta and gamma air doses only.
This location will be the offsite-location with the higt.est long term vent x/Q and will be selected using the x/Q values given in Appendix A, Table A-3.
This location will remain the same unless 1
meteorological data is reevaluated or the site boundary changes.
The critical location for the 1-131, tritium, and long-lived particulate pathway will be selected once each year.
The selection will follow the annual land use census performed within 5 miles of the PNGP.
Each of the following locations will be evaluated as potential critical receptors prior to implementation of the Radioactive Effluent Technical Specifications:
- 1. Residence in each sector
- 2. Vegetable garden producing leafy green vegetables
- 3. All ioentified milk animal locations i
40%
OF#
Following the annual survey, doses will. be calculated using Equation 3.3-3 for all new i
identified receptors and those receptors whose characteristics have changed significantly. The calculation'will include appropriate information about each new location.
The dispersion parar:eters given in this manual should be employed.
The total releases reported for the previous calendar year should be used as the source terms.
IBM
. _ -.. _ _ _ _.. =. _. _. -
t H4 s
Rev..12 Page 38.
}
REFERENCES t
(1) - NSP-Prairie Island Nuclear Generating Plant, Appendix I
'f Analysis - Supplement-No. 1 - Docket No. 50-282 and 50-306",
l Table 2.1-4.
t i
e r
ii i
i I
i t
i f
i,
-1 i
f 3
i I
r t
i i
- S.
- me i
t l
t j
i l
l-I t
i t
t L
l-t, k
l-6 t-g l-
?
F h
I i
t I
L.t i
IBM i
I i
1
?
I i
l-t
Q;
'114 4
Rev. 12 Page 39 TABLE 3.3-1 DOSE FACTORS FOR NOBLE GASES AND DAUC11TERS*
l l
Total Body Gamma Air Beta Air Dose Factor Skin Dose Fector Dose Factor Dose Factor j
K_
L.
3 N.
3 N
3 3
_ mrem /yr*per_pci/m.)
(mrem /y[perlCi/m );
'(m rad /yr
- pe rj aci /m )
(mrad /y r_3per pCi/m_)_.
M.onuclide
(
l Kr-83m 7.56E-02d 1.93E+0!
2.88E+02' Kr-85m 1.17E403 1.46E+03 1.23E+03 1.97E+03 Kr-85 1.61E+01 1 34E+03 1.72E+01 1 95E+03 Kr-87 5.92E+03 9.73E+03
' 6.17E+03 1.03E+04' Kr-88 1.47E+04 2.37E+03 1.52E+04 2.93E+03 i
Kr-89 1.66E+04 1.01E+04 1.73E+04
- 1.06E+04.
Kr-90 1.56E+04 7.29E+03 1.63E+04 7.83E+03.
Xe-131m 9.15E+01 4.76E+02 1.56E+02 1.11E+03 Xe-133m 2.51E+02 9.94E+02 3.27E+02 1.48E+03' Xe-133 2.94E+02 3.06E+02 3.53E+02 1.05E+03-Xe-135m 3.12E+03 7.11E+02 3.36E+03 7.39E+02 Xe-135 1.81E+03 1.86E+03 1.92E+03 2.46E+03-Xe-137 1.42E+03 1.22E+04 1.51E+03 1.27E+04 Xe-138 8.83E*03 4.13E+03 9.21E+03 4.75E403
[
Ar-41 8.84E+03 2.69E+03 9.30E+03' 3.28E+03 t
t i
i
)
- The listed dose factors are for radionuclides that may be detected in gaseous effluents. All others are D.
- 7.56E-02 = 7.56 x 10_2 b
'wWe"-m.*we'*v*-e-e e *
'n ' v* -f w e v e
=e'i==(r w*~w**we-*pr-w=N+-.=w-ww= www vv" 4mwme'
'e ve ww--'e'-*"N't'-+="*We~
me g-y.e p. g. -
y.,ym,,,,-,+.9.,.p-..,
..g...
w..e.v p e "4
- ws' Mw w -w v f e w-w e = w o*- y e *'"
Mv*im-'e*
l;[
J
,LlI!
f
!lii}
l;!fl f :i ii!;
- ib d,
II i i iI 1II II IIlIl1IIlIIiI 8
8 0
4' 7'
9' 0
}
. 0's 0
0 1
0 0
0 1
~
~
N E
E E
E E
E E
E' I
7 3'
4 2
8 9
6' 0
K 5
2 4
- 5. '
5 0
9 2
v S
^
2 2
7 1
r
- 1 3
4 2
~
R
~
E I lI I 1i I 1 1I1II1I Illl1 I Ill P
9 8
8 0
4 7
9 0
R
. 0 0
0 1
0 0
0 1
Y
/
G E
E E
E E
E E
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t 4
5 9
5 3
2 2
3 E
N 3
7 7
1 2
8
- 0. ~
R A
1 2
. 3' 2
2
. 7
. 6 1
T i
M N
L
~
1 L
P M
l1 Ii1I I1 lIlI IlIllIilI I ll G
F N
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9 8
8 0'
4 7
9 0
T D
0 0
0 '
. 0 0'
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A w
2 0 R
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1 4 E
R 4
5 9
5 3
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N Y
3 7
7 2
7 8
- 0. ~
U 1
E
.e I
1 2
3 2
2 1
6 1
N C
T vg I
4ea R
D RP A
1 IlI 1lI III I 1i IlIIlIIiI 1 1 I 1
E N
L A
C 9
8 8
0 4
7 9
0 U
Y 0
- 0 0
1 0
0 0
1 N
E M
N E
E E
E E
E E
E U
D D
4 5
9 5
3 2
2 3
I
~
3 7
7 2
7 8
0 T
N I
1 A
K I
L 1
2 3
2 2
1 6
1 R
S T
I D
I1 I i1I Ill I1II 1IIli IIl l1I E
N A
I R
9 8
8 0
4 7
9 0
0 0
0 1
0 0
0 1
N I
A R
O R
E E
E E
E E
E E
E I
'I'.
V 4
5 9
5
. 3 2
2 3
T 3
2 7
8 0
A 7
7 1
I L
I L
E 1
2 3
2 2
1 6
1 A
IT 1
1N R
IlI I 1 II1 l I1IilIIlIIIi i1I I
O T
R 9
8 8
0 4
7 9
0 O
S 0
0 0
1 0
0 0
1 F
r E
E U
N E
E E
E E
E E
E L
0 4
5 9
5 3
2 2
3 M
3 7
7 2
7 8
0 1
/
A 1
1 V
I 1
2 3
2 2
1 6
1 C
R p
R 1 liIlI I1II1I I1IIliIIi I 1 I 2
E P
3 T
9 8
8 0
. 4 7
9 0
m C
0 0
0 1
0 0
0 1
R 3
A Y
R
. E E
E E
E E
E E
/
m E
T 4
5 9
5 3
2 2
3 M
3 7
7 2
7 8
0 E
L 1
R l
I lA G
1 2
3 2
2 1
6 1
MS r
T R
FE I
OI I ll IlI I1II1I I1lIlIIII I 1I T
SO 0
T Y
9 8
8 0
4 7
9
. 1 I L D
0 0
0 1
0 0
0 D 0 NL E
E E
E E
E E
E UA l
N I
U 4
5 9
5 3
2 2
3 3
7 2
7 8
0 NR O
7 1
I O
+
R T C
1 2
3 2
2 1
6 1
F S
EC
=
l1i I1!
I 1I I11 ill1lI I11 11I UE Y E LS
. A D A/
4 9
8 0
9 1
4 7
VI N
I I
L 5
5 5
6 8
3 3
3 C
T C 1
1 1
R p r
A U N
E O
O R
S S
P N M
F C
C S
I C
C v
+
1 s
- i
!;l
l m
O C
C O
O O
O C
O C
1 1
1 8
4 0
4 t
6 Z
W W
W W
W W
W W
W W
.=
cc C
C O
O O
O C
C C
O.
O.
C.
O O.
O.
O.
O.
C.
W 44 g
N O
C C
C C
C C
C C
2 y
b m
N ce
- =
m Cd g
C C
C O
O O
C C
O C
>=
4 e
4 e
i e
N C
W W
W W
W W
W W
W W
E cc O
N O
C o
C o
c N
W 3
O.
C:
N.
O.
C.
C.
C.
o.
C W
=
h W
N O
CC C
C C
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~ H4 -
Rev. 12 4.0 INFORMATION RELATED-TO 40 CFR 190 and 40 CFR 141 The PNGP Technical Specifications indicate that if twice the per unit limits of 10 CFR 50, Appendix-I are exceeded, a
-special report is required.
Therefore,.if twice the annual limits of Equations 2.3-6, 2.3-7, 3.3-1, 3.3-2 or 3.3-3 are exceeded, the licensee shall prepare and submit a Special Report to the Commission and limit subsequent releases such that the dose or dose commitment to a real individual from all uranium fuel cycle sources is limited to < 25 mrem to the total body or any organ (except the thyroid, which is limited to < 75 mrem) over 12 consecutive months.
This special Report is to include an analysis which demonstrates that radiation exposures to all real individuals from all uranium fuel cycle sources (including all liquid and gaseous effluent pathways and direct radiation) are less than the standards in 40 CFR Part 190, Environmental Radiation Protection Standards for Nuclear Power Operations.
If analysis indicates that releases resulting in doses that exceed the 40 CFR 190 Standard may have occurred, then a variance from the Commission to permit such releases will be obtained.
The Standard Technical Specifications 3.11.1.2, 3.11.2.2 and 3.11.2.3 consider doses to a real individual and apply-to each reactor but do not include any other portion of the uranium fuel cycle or direct shine from the reactor.
The " Uranium fuel cycle" is defined in 40 CFR Part 190.02(b) as:
" Uranium fuel cycle means the operations of milling of uranium ore, chemical conversion of uranium, isotopic enrichment of uranium, fabrication of uranium fuel, generation of electricity by a light-water-cooled nuclear power plant using uranium fuel, and reprocessing of spent uranium fuel, to the extent that these directly support the production of electrical power fvt public use utilizing nuclear energy, but excludes mining operations, operations at waste disposal sites, transportation of any radioactive material in support of these operations, and the' reuse of recovered non-uranium special nuclear and by-product materials from the cycle."
The Special Report will contain:
1)
A determination of which uranium fuel cycle facilities or operations, in addition to the nuclear power reactor units at the site, contribute to the annual dose to the maximum exposed member of the public.
Only those other fuel cycle facilities within-5 miles of PNGP need be considered.
No such facilities exist or are planned at present.
IBM
H4 Rev. 12 Page 60 2)
A determination of the maximum exposed member of the public as given for determining compliance with 30 CFR 50, Appendix 1.
3)
A determination of the total annual dose to this person from all existing pathways and sources of radioactive effluents using the methodologies described for i
compliance with 10 CFR 50, Appendix 1 in this ODCM.
Where additional information on pathways and nuclides is needed, the best available information will be used and documented.
4)
A determination of the dose resulting from direct radiation from the plant and storage facilities.
The various organ doses resulting from liquid effluents from the PNGP will be summed with the organ doses resulting from releases of noble gases, radiciodines and particulates.
These doses will be based upon releases from the PNGP during the past 3 quarters and from the quarter in which twice the specification was exceeded.
The doses from the PNGP will be summed with the doses to the maximum exposed individual contributed from other operations of the uranium fuel cycle.
The direct dose components will be determined by either calculation or actual measurement.
The N-16 component of m.-4 direct radiation may be calculated using SKYSHINF A Computer CJ1 Procedure Evaluating Ef fects of Structure Design on N-16 Gamma-Rav Dose Rates.
Radiation Research Associates, Inc.
Report RRA-T7209, November 1972.
(Available in the NSP-NSS Library).
The calculation or actual measurement will be documented in this Special Report.
If the quarterly or annual doses due to liquid releases exceed the values listed in 3ection 24 3.3, a special report shall be submitted to the USNRC and shall include information related to 40 CFR 141 such as analysis of Mississippi River water and an analysis of possible impacts through the drinking water pathway.
IBM
H41 Rev. 12 5.0 RADI ATION ' ENVIRONMENTAL MONITORING PROGRAM
~'
y 5.1 Sampling Table 5.1-1 and Figures 5.1-1, 5.1-2, and 5.1-3 specify the current sampling locations based on the latest land use census.
If it is learned from an annual census that milk animals or gardens are present at a location which yields a calculated thyroid dose greater than those previously sampled, the-new milk animal or garden
-locations resulting in higher calculated doses shall be added to the surveillance program as soon as practicable.
Sample locations (except the control) having lower calculated doses may be dropped from the program at the end of the grazing or growing season (October 31) to keep the total number of sample locations constant.
An annual land use survey will be conducted to determine whether corn fields are irrigated with water taken from the Mississippi. River between the plant and a point 5 miles downstream.
If corn fields are being irrigated with Mississippi River water, appropriate samples will be collected and analyzed per Technical Specifications, vse Table 4.10-1.
MF 5.2 Interlaboratory Comparison Program Analyses shall be performed on radioactive samples supplied by the EFA crosscheck program.
This program involves the analyses of samples provided by a control laboratory and comparison of results with those of the control laboratory as well as with other laboratories which receive portions of the same samples.
Media used in this program (air, milk, water, etc.) nhall be limited to those found in the radiation environmental monitoring program.
The results of analyses performed as a part of the crosscheck program shall be included in the Annual Radiation Environmental Monitoring Report.
IBM
H4 Rev. 12 TABLE 5.1-1
. 41RIE ISLAND NUCLEAR GENERATING PLANT KADIATION ENVIRONMENTAL HONITORING PROGRAM SAMPLING LOCATIONS Type of 4
sample Codri Collection Site Location C
River Water P-5 Upstream of Plant 2.3 mi 0 348*/NNW River Water P-6 Lock & Dam #3 1.6 mi @ 129*/SE Drinking Water-P-11 Red Wing Service Center 3.3 mi @ 158*/SSE Well Vater--
P-25 Rohl Farm 12.9 mi @ 352*/N l-C 14111 Water P-6 Lock & Dam #3 1.6 mi @ 129*/SE Well Water P-8 Community Center 1.0 mi @ 321*/VNW Well Water P-9 Plant Well #2 0.3 mi @ 306*/NW C
Sediment-River P-20 Upstream of-Plant 0.9 mi @ 45*/NE Sediment-River F-6 Lock & Dam #3 1.6 mi @ 129*/SE Sediment-Shoreline P-12 Downstream of Plant 3.0 mi @ 116*/ESE C
7g, Pariphyton or P-5 Upstream of Plant 2.3 mi @ 348*/NNW 4Y#
Invertebrates P '2 Downstrema of Plant 3.0 mi @ 116*/ESE c
Fish P-19 Upstream of Plant 1.3 mi @ 0*/N Fish P-13 Downstream of Plant 3.5 mi @ 113*/ESE c
Milk P-25 Rohl Farm 12.9 mi @ 352*/N' Milk P-14 Gustafson Farm 2.3 mi 0 173*/SSE Milk P-16 Johnson Farm 2.6 mi @ 60*/ENE Milk P-17 Dosdall Farm 3.9 mi @
9*/N Milk P-18 Christensen Farm 3.8 mi @ 88*/E IBM
H4 Rev. 12 Page 63 TABLE 5.1-1 (Continued)
PRAIRIE ISLAND NUCLEAR GENERATING PLANT I
RADIATION ENVIRONMENTAL HONITORING PROGRAM t
t SAMPLING LOCATIONS i
t Type of Sample Code Collection Site Location i-Cultivated Crops C
l (Leafy Green Veg)
P-25 Rohl Farm
- 12. 9 mi @ 3 52*/N 1
P-24 Suter Residence 0.6 mi @ 158*/SSE C-
.Particulates and P-1 Air Station P-1 11.8 mi e 316*/NW Radiciodine (afr)
Particulates and P-2 Air Station P-2 0.5 mi 0 294*/VNV i
Radiciodine (air) r, Particulates and P-3 Air Station P-3 0.8 mi @ 313*/NW
{
.s Radioiodine (air) 4
+ - u+
\\J'gf Particulates and P-4 Air Station P-4 0.4 mi @ 359*/N f
Radioiodine (air) i Particulates and P-6 Air Station P-6 1.6 mi @ 129*/SE l
Radioiodine (air)
}
t Direct Radiation f
-(TLD)
POlA Property Line 0.4 mi @ 359*/N
[
Direct Radiation
-(TLD)
P02A Property Line 0.3 mi 0 10*/N Direct Radiation (TLD)
PO3A Property Line 0.5 mi 0 183 /S i
J.I Direct Radiation (TLD)
PO4A Property Line 0.4 mi 0 204*/SSW Direct Radiation (TLD)-
POSA Property Line 0.4 mi @ 225*/tW i
L IBM i
i i
I
114 Rev. 12 TABLE 5.1-1 (Continued)
-PRAIRIE ISLAND NUCLEAR GENERATING PLANT I
RADIATION ENVIRONMENTAL MONIT0 RING PROGRAM SAMPLING LOCATIONS y
i
]
Type of Sample Code Collection Site Location
'4 i
Direct Radiation i
(TLD)
P06A Property Line 0.4 mi 0 249'/WSW l
Direct Radiation (TLD)
-P07A Property Line 0.4 mi @ 268'/W Dire-Radiation (TLb>
PO8A Property Line 0.4 mi 0 291*/VNW Direct Radiation
[
(TLD)
P09A Property Line 0.7 mi @ 317'/NW t
Direct Radiation (TLD)
P10A Property Line 0.5 mi @ 333*/NNW Direct Radiation
,jyj (TLD)
PolB Tom Killian Res.
4.7 mi @ 355'/N I
Direct Radiation (TLD)
P02B Roy Kinneman Farm 4.8 mi @ 17 /NNE 1
Direct Radiation (TLD)
P03B Wayne Anderson Farm 4.9 mi @ 46 /NE Direct Radiation (TLD)
PO4b Nelson Drive (Road) 4.2 mi 0 61'/ENE Direct Radiation (TLD)
PO5B County Rd E & Coulee 4.1 mi @ 102*/ESE I
-Direct Radiation (TLD)
P06B Villiam Houschildt Res.
4.4 mi @ II2*/ESE l
Direct Radiation
_(TLD)
PO7B Red Wing Public Works 4.7 mi @ 140 /SE Direct Radiation 4
-(TLD)-
PO8B David Wnuk Res.
4.1 mi 0 165'/SSE 2
Direct Radiation (TLD)
P09B Hwy 19 South 4.2 mi @ 187o/S i
Direct Radiation l
(TLD)
P10B Cannondale Farm 4.9 mi @ 200o/SSW IBM
{
11 4 i
Rev. la Page 65 l
TABLE 5.1-1 (Continued) i PRAIRIE ISLAND NUCLEAR GENERATING PLAN 1 RADIATION ENVIRONMENTAL MONITORING PROGRAM SAMPLING LOCATIONS
[
Type of Sample Code Collection Site Location Direct Radiation (TLD)
P11B Wallace Weberg Farm 4.5 mi @ 221*/SW Direct Radiation
-(TLD)
P12B Roy Gergen Farm 4.5 mi @ 247'/VSW Direct Radiation (TLD)
P13B Thomas O'Rourke Farm 4.4 mi 0 270'/W Direct Radiation (TLD)
-P14B David Anderson Farm 4.9 mi @ 306'/NW-Direct Radiation (TLD)
P15B Holst Farms 4.2 mi @ 347'/NNW Direct Radiation (TLD)
POIS Federal Lock & Dam #3 1.6 mi @ 129'/SE Direct Radiation (TLD)
P02S Charles Suter Res.
0.5 mi @ 155*/SSE Direct Radiation (TLD)
P03S Carl Gustafson Farm 2.2 mi @ 173 /S Direct Radiation (TLD)
P04S Richard Burt Res.
2.0 mi 0 202*/SSW Direct Radiation (TLD)
POSS Kinney Store 2.0 mi @ 270'/V Direct Radiation (TLD)
P06S Earl Flynn Farm 2.5 mi @ 299'/VNW Direct Radiation (TLD)
P25 Robert Kinneman Farm 11.1 mi @ 331'/NNW
- "c" denotes control location.
All other locations are indicators.
The letters after numbered TLD's are as follows:
"A" denotes locations in the general area of the site boundary.
"B" denotes locations about 4 to 5 miles distance from the plant.
l "S" denotes special interest locations.
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11 4 Rev. 12 Page 66 FIGURE 5.11 SITE BOUNDARY TLD LOCATIONS
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t, H4 Rev. 12 Page 69 APPEND 1X A
- METEOROLOGICAL ANALYSES l
Table A-1 Release Conditions i
. Table A 2 Unrestricted Area Boundaries l
Table A-3 Long Term - Ground Level - Unrestricted Area Boundary - x/Q and D/Q.
[
Table A-4 Long Term - Ground Level - Standard Distances - x/Q j
Table A-5 Long Term - Ground Level - Standard Distances - D/Q Table A-6 Short Term - Ground Level - Unrestricted Area
[
Boundary - x/q and D/q Table A-7 Short Term - Ground Level - Standard Distances - x/q j
Table A-8 Short Term - Ground Level - Standard. Distances - D/q.
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Page 70 l
APPENDIX - A l
I hmmary of Dispersion Calculational Procedures j
Undepleted, undecayed dispersion parameters were comput.ed using the computer program XOQD0Q (Sagendorf and Coll, 1977).
Specifically, sector average x/Q and D/Q values were obtained for a sector width of 22.5 degrees.
Building wake corrections were used to adjust calculations for ground-level releases.
Standard i
open terrain recirculation correction factors were also applied I
as available as default values in XOQD0Q.
Dispersion calculations were based on ground level releases for the. shield buildings, turbine building, and auxiliary building (hereafter referred to as the plant complex).
A summary of j
release conditions used as input to XOQD0Q is presented in Table i
A-1 and controlling unrestricted boundary distances are defined in Table A-2.
Computed y/Q and D/Q values for unrestricted area boundary. locations (relative to release points) and for standard distances (to five miles from the source in 0.1 ~ile increments) are presented in Tables A-3 through A-8.
Onsite meteorological data for the period April 1, 1977 through March 31, 1978 (as presented in Appendix B) were used as input to XOQDOQ.
Data were collected and AT stability classes were j
eQ defined in conformance with NRC Regulatory Guide 1.23.
Dispersion j
"* T' calculations for the plant complex were based on AT 42.7-12.2m and 12.2 meter wind data (joint data recovery of 96 percent).
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i 1.
Sagendorf, J.F.
and Goll, J.T., XOQDOQ Prograrn for the Evaluation of Routine Ef fluent Relear:es at Nuclear Power Stations, NUREG-0374,
[T.S. ' Nuclear R. egulatory Comruission7 Sept. ember 1977.
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Pane 72 l-TABI.E A-1 i
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- Applied to short-term calculations only i
114 llev. 12 Page 73 TABLE A 2 Distances (Miles) To controlling Unrestricted Area Boundary Locations As Measured _from Edge of Plant Complex Sector _
D_istance N
0.28 NNE 0.26 NE 0.84*
ENE 0.62*
E 0.59*
ESE 0.61*
0.43 2.k.
0.36 VNV 0.36 NV 0.43 NNW 0.48
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114 Rev. 12 1
Page 79 i
TABLE A-4 (Cont.)
1 3
]
. Prairie Island Dispersion Parameters'(X/9), sec/m j-for Long Term Ground Level Reicases > 500 Hrs /Yr or > 150 Ilts/QTR for Standard Distances (As Measuted from Efge c,f Plant Compicx) i rtiles i
i Sector 2.9 3.0 3.1 3.2 3.1 3.4 3.5 4
l l
N 2.01E-07 1.88E-07 1.77E-07 J.67E-07 1.58E-07 1.49E-07 1.42E-07 NNE 1.36E-07 1.28E-07 1.20E-07 1.14E-07
'1.07E-07 1.02E-07 9.65E-n8 NE 1.29E-07 1.21E-07 1.14E-07 1.08E-07 1.02E-07 9.69E-08 9.20E ENE 1.29E-07 1.21E-07 1.14E-07 1.07E-07
- 1. DIE-07 9.60E-08 9.11E-03 i
E 3.84E-07 3.61E-07 3.40E-07 3.21E-07 3.04E-07 2.8BE-07 2.74E-07 l'
l ESE 7.23E-07 6.80E-07 6.41E-07 6.05E-07 5.73E-07 5.43E-07 5.16E-07 SE 7.67E-07 7.21E-07 6.80E-07 6.42E-07 6.08E-07 5.77E-07 5.48E-07 l
SSE 4.10E-07 3.85E-07 3.63E-07 3.43E-07 3.24E-07 3.08E-07 2.92E-07 S
1.75E-07 1.64E-07 1.55E-07' 1.46E-07 1.38E-07 1.31E-07 1.24E-07 SSW 1.38E-07 1.29E-07 1.22E-07 1.15E-07 1.09E-07 1.03E-07 9.77E-08 SW 1.49E-07 1.40E-07 1.32E-07 1.24E-07 1.17E-07 1.IIE-07.
1.05E-07 i
WSW 1.85E 1.74E-07 1.64E-07 1.54E-07 1.46E-07 1.39E-07 1.32E-07 l
l W
4.17E-07 3.92E-07 3.70E-07 3.49E-07 3.30E-07 3.13E-07 2.98E-07 WNW 5.60E-07 5.26E-07 4.95E-07 4.67E-07 4.42E-07 4.19E-07 3.98E-07 NW 4.79E-07 4.50E-07 4.24E-07 4.00E-07 3.79E-07 3.59E-07 3.41E-07 NNW 2.95E-07 2.77E-07 2.61E-07 2.46E-07 2.32E-07 2.20E-07 2.09E-07 Period of Record: '4/1/77 - 3/31/78 i
1 IBM L.-.,_,._..__,-,.__..--..____,-,._.._.--,.-.____...._,.
Pik N34 114 l
Rev. 12 l
'Page 80
. i TABLE A-4 (Cont.)-
i 3
1 Prairie Island Dispersion Parameters (X/Q), sec/m,
for Long Term Ground Level Releases > -500 lirs/Yr or > 150 Ifrs/QTR f
for Standard Distances (As fic sured from Edge of Plant Compicx) i j-11iles 1
\\
. I l
t Sector' 3.6 3.7
- 3.3 3.9 4.0 4.1 4.2 4
l N
1.35E-07 1.28E-07
. 1.22E-07 1.16E-07 1.11E-07 1.07E-07 1.02E-07 j
l NNE 9.18E-08 8.74E-08
' 8.34E-08 7.97E-08 7.62E-08 7.30E-08 7.00E-08 I
NE 8.75E-08 8.34E-08 7.96E-08 7.61E-08 7.28E-08
'6.98E-08 6.70E-08 ENE 8.66E-08 8.24E-08 7.86E-08 7.50E 08 7.17E-08 6.86E-08 6.58E-08 E
2.60E-07 2.48E-07 2.37E-07 2.26E-07 2.17E-07
- 2. ORE-07 1.99E-07 ESE 4.91E-07 4.68E-07 4.47E-C7 4.27E-07
- .09E-07 3.92E-07 3.76E-07 i
SE 5.22E-07 4.97E-07 4.75E-07 4.54E-07 4.35E-07 4.17E-07 4.00E-07 i
l SSE 2.78E-07 2.65E-07 2.53E-07 2.42E-07 2.32E-07' 2.22E-07 2.13E-07 I,
S 1.18E-07 1.12E-07 1.07E-07 1.02E-07 9.79E-08 9.38E-08 8.99E-DE l
SSV 9.29E-08 8.85E-08 8.44E-08 8.07E-08 7.72E-08 7.39E-08 7.09E-08 i
SW' 1.00E-07 9.54E-08 9.10E-08 8.69E-08 8.31E-08 7.95E-08 7.6?E-08 l
i WSW 1.25E-07 1.19E-07 1.14E-07 1.09E-07 1.04E-07 9.08E-08 9.57E-08 t
I W
2.83E-07 2.70E-07 2.58E-07 2.47E-07 2.36E-07 2.26E-07 2.11F-07 l
WNW 3.78E-07 3.60E-07 3.44E-07 3.28E-07
. 3.14E-07 1.01E-07 2.89E-07 i
NV 3.24E-07 3.09E-07 2.95E-07 2.82C-07 2.70E-07 2.58E-07 2.48E-07 a
l NNW I.99E-07 1.89E-07 1.80E-07 1.72E-07 1.65E-07 1.58E-07 1.51E-07 4
i Period of Record: 4/1/77 - 3/31/78 i
i j
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i s-114
[
F Rev. 12
[
j' Page 81 ll TABLE A-4 (Cont.)
3 i
'Prairic Islanet Dispersson Parameters (X/Q), sec/m'
.for Long Team Ground Level Releases '> 500 lirs/Yr or > 150 lirs/QTR l
for Standar.1 Distances (As Me.isured (tow Edge of Plant Complex)
[
i Miles i
h i
Sector 4.3 4.4 4.5 4.6-4.7 4.8 4.9 I
i E
N 9.79E 9.41E-OS 9.04E-08.
8.71E-08 8.39E-08 8.09E-08 7.81E-08 l
NNE 6.72E-08 6.46E-08 6.22E 5.99E-08 5.77E-08 5.57E-08 5.38E-08 t
j NE 6.43E-08 6.18E-08 5.95E-08 5.74E-08 5.53E-08 5.34C-08 5.16E-08 j
t ENE 6.31E-08 6.06E-08 5.83E-08 5.62E-08 5.41E-08 5.22E-08 5.04E-08 l
f E
1.91E-07 1.84E-07 1.77E-07 1.71E-07 1.65E-07 1.59E-07 1.54E-07 ESE 3.62E-07 3.48E-07 3.35E-07' 3.23E-07 3.11E-07 3.01E-07 2.91E-07 SE 3.85E-07 3.70E-07 3.56E-07 3.44E-07 3.31E-07 0.20E-07 3.09E-07 I
l SSE 2.05E-07 1.97E-07
't.90E-07 1.33E-07 1.76E-07 1.70E-07 1.65E-07 i
i S
8.63E-08 8.30E-08 7.98E-08 7.69E-08 7.41E-08 7.15E-08 6.91E-03 i'
SSW 6.31E-08 6.55E-08 6.30E-08' 6.07E-08 5.85E-08 5.65E-08 5.46E-08 t
SW 7.32E-08 7.03E-08 6.76E-08 6.51E-08 6.28E-08 6.06E-08 5.35E-08 I
I WSW 9.19E-08 8.84E-08 8.51E-08 8.20E-08 7.91E-08 7.64E-08 7.38E-08 i
i j
W 2.09E-07 2.01E-07 1.93E-07 1.86E-07 1.80E-07 1.73E-07 1.68E-07 l
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NNV 1.45E-07 1.39E-07
't.34E-07 1.29E-07 1.24E-07 1.20E-07 1.16E-07 4
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w-114
' i Rev. 12 i
Page 91' TAllLE A-6 Prairie Island Dispersion Parameters
'for Short Term Ground
<cl Releases 5 500 IIrs/Yr or 5 150 Hrs /Qtr
[
i fcr Unrest ricted Area.soundary locations (identified in. Table A-2) i t
i Unrest.ricted Arca-Roundary Sector 2
1 Sector X/fL(sec/m' )
D/<1 (1/m )
i N
7.09E-05 4.60E-07 t
' 4.11E-07 NNE 7.32E-05 r
NE 1.60E-05 6.71E-08 r
ENE I.97E-05 1.11E-07 E
4.92E-05 1.99E-07 ESE 6.40E-05 2.52E-07 SE 5.98E-05 2.43E-07 1
SSE 8.79E-05 3.08E-07 S
5.18E-05 2.04E-07 I
t SSV 5.26E-05 1.89E-07 I
SW 5.25E-05 1.90E-07 WSW 1.83E-05 2.44E -
I W
1.32E-04 3.78E-07 i
WNW 1.10E-04 4.61E-07 NW 7.67E-05 3.25E-07 NNW 4.79E-05 2.34E-07 i
I i
i Period of Record:
4/1/77 - 3/31/78 i
F k r,.,. m + -
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' '! l 114 R(nr. 12 Pafs? 95 TABI.E A-7 (Cont.)
Prairie Island Dispersion Paramet ers (x/q), sec/m,
for Short Term Ground Level Releases 5 500 llrs/Yr or < 150 !!rs/Qt r for Standard Distances (As Measured from Edge of Plant Complex) tiiles Sector 2.2 2.3 2.4 2.5 2.6 2.7 2._8 N
2.87E-06 2.67E-06 2.50E-06 2.32i;-06 2.27E-06 1.94F-06 1.84E-06 NNE 2.56E-06 2.39E-06 2.24E-06 2.09E-06 2.03E-06 1.76E-06 1.66E-06 NE 3.19E-06 2.98E-06 2.79E-06 2.63E-06 2.57E-06 2.24E-06 2.11E-06 ENE 2.51E-06 2.34E-06 2.19E-06 2.05E-06 1.97E-06 1.74E-06 1.64E-06 E
5.82E-06 5.44E-06 5.10E-06 4.89E-06 4.75E-06 4.14E-06 3.92E-06 ESE 8.41E-06 7.87E-06 7.38E-06 7.16E-06 7.09E-06 6.11E-06 5.78E-06 SE 8.93E-06 8.35E-06 7.82E-06 7.70E-06 7.60E-06 6.62E-06 6.26E-06 SSE 6.56E-06 6.14E-06 5.76E-06 5.55E-06 5.36E-06 4.73E-06 4.47E-06 S
3.76E-06 3.51E-06 3.29E-06 3.18E-06 3.06E-06 2.73E-06 2.59E-06 SSW 3.39E-06 3.16E-06 2.96E-06 2.89E-06 2.75E-06 2.42E-06 2.28E-06 SW 3.37E-06 3.15E-06 2.95E-06 2.84E-06 2.70E-06 2.44E-06 2.30E-06 WSW 4.47E-06 4.18E-06 3.92E-06 3.85E-06 3.67E-06 3.28E-06 3.10E-06 W
7.21E-06 6.74E-06 6.32E-06 6.45E-06 6.03E-06 5.30E-06 5.02E-06 WNW 6.42E-06 6.00E-06 5.65E-06 5.32E-06 5.23E-06 4.55E-06 4.30E-06 NW 6.16E-06 5.76E-06 5.40E-06 5.16E-06 5.03E-06 4.37E-06 4.13E-06 NNW 4.31E-06 4.02E-06 3.76E-06 3.57E-06 3.46E-06 3.00E-06 2.83E-06 Period of Record:
4/1/77 - 3/31/78 IBM
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3 2 6 4 7 5 4 3 7 8 3 0 3 9 5 1
1 8
ns u l
4 3 3 3 5 7 8 5 3 2 2 3 5 6 5 4 oas i
i ea.
M A
sl e E
r ei t
L eR B
p s
A sl A T
i e(
D v 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 es 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 dL e d.
E E E E E E E E E E E r.
E E E 1
n c
0 9 1
3 9 0 6 8 4 1
8 4 1
4 3 9 ad n.
l na 3
5 7 3 3 6 9 6 5 9 9 5 3 6 2 7 1
sut I os 4 3 3 3 5 8 8 5 3 2 2 3 5 6 6 4 ri eGD i
rmd.
i r i 8
aea 7
rTd
.P n
9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9
/
t a 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 1
3 7
rt E E E E E E E E E E E E E E E E
/
oS S r.
3 5 0 9 1
0 4 6 9 3 6 4 3 9 0 6 5 3
h 0 4 5 0 5 4 9 0
7 5 0 5 0 7
1 7
o rf 4 4 3 3 6 8 9 5 3 3 3 3 5 7 6 5 7
o 7
f
/
1
/
4 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 d
6 E E E E E E E E E E E E E E E E r
3 1
2 8 1
3 2 9 3 4 3 1
4 9 9 3 3 oc 0 2 6 0 9 3 9 2 3 9 8 3 9 3 7
3 e
5 4 3 3 6 9 9 6 3 3 3 3 5 7 6 5 R
r o
do i
o N E E E E E E E S W W W W W W V r
M r
t N N N S S S S S S N N N e
B c
N E
E S
S W
W N
P I
e S
N M
m w
w P
9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 M
1 E E E E E E E E E E E E E E E E 9
9 6 0 1
2 7 2 8 8 5 9 0 5 7 4 7
1 4
7 3 0 0 4 9 3 4 7 8 2 0 7 9
.w 1
1 e
t 2 2 2 2 3 4 5 3 2 1
1 2 3 4 3 2
+
w T
W 9
e 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 v
0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 e
8 E E E E E E E E E E E E E E E E r
4 7
9 9 0 2 3 2 3 4 6 7 4 0 1
2 3 3 3 0 1
6 6 6 2 8 8 2 4 2 9 0 W
1 t)
Qx m
/ e 2 2 2 2 3 5 5 3 2 1
1 2 3 4 3 3 sl r p f
n
.4 I
ro M
6
- 0C e
20 2
5 11 m1 t
/
n 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 e
t 1 5a 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 w
.e vg
,rP 7
E E E E E E E E E E E E E E E E t
l t
9 9 8 9 7 4 6 8 4 4 5 3 4 R 8 6 4 ea
) o.
4 9 4 7 3 8 7 3 9 9 3 5 3 0 M
S 1
HP q
f 1
1 1
1
/ ro e
DY 2 2 2 2 3 5 5 3 2 1
1 2 3 4 4 3
( / e P
sg srd h
U rfiE e
)
t 0 m e
W e0 o t
m5 r 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 W
n r <_f a
e o
C a
d 6
E E E E E E E E F.
E E E E E E E C
2 0 7
9 3 7
2 3 3 2 3 5 9 7 5 9
(
Pse s
6 2 2 9 5 9 4 0 0 4 6 5 2 2 1
er e
4 1
'P M
ns u t
.3 2 2 2 3 5 6 3 2 2 2 2 3 4 4 3
~
O oas i
l t ea f
A f
sl e 1
E r eM w
L eR w
B p
s W
A sl A T
e(
T D v 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9
=
4
. es 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 dL e 5
E E E E E E E E E E E E E E E E T'
n c
6 1
7 9 0 1
8 0 4 1
2 5 4 6 4 3 ad n W
l na 4
2 7 3 3 8 3 5
5 8 7 4 4 1
1 1
1
+
N P
sut I os 3 2 2 2 4 5 6 4 2 2 2 2 3 4 4 3 M
ri eGD i
9 rmd i r r 8
M n
a ea 7
rTd P
n 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9
/
t a 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 1
T 3
1u 4
rt E E E E E E E E E E E E E E E E
/
T oS h
4 0 4 8 9 8 7 8 8 5 0 1
7 1
8 6 0 3
9 2 6 0 9 6 6 SW 4 8 4 4 2 0 6 2 6 2
S r g
o f
rf 3 2 2 2 4 6 6 4 2 2 2 2 4 4 4 3 t
7 w
o 7
v f
/
M m
1 e
e
/
t 4
r v
9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 9 s
g 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 w
d
- e 3
E E E E E E E E E E E E E E E E r
4' 4
5 7 0 1
8 5 9 8 8 0 1
9 9 0 8 6 o
'd 1
7 c
5 9 6 6 4 3 9 4 7 3 3 7 2 8 y
e v
3 2 2 2 4 6 6 4 2 2 2 2 4 5 4 3 R
e t
f e
o u
d w
o
=
i o
N E E E E E E E 5 W W W W V W W r
M t
r
+
t N N N S S S S S S N N N e
B w
w c
N E
E S
S W
W N
P I
w
.-e e
S 3
t' e
w t
'g
+
m
~
i' f
ii l
l i'
i
mi.
- ii:td ns 11 4 Rev. 12 Page 107 TA!!!.E A-8 (Cont.)
2 Prairie Island Dispersion Parameters (D/q), 1/m f or Short Term Ground I.evel Releases < 500 lirs/Yr or < 150 lirs/QTH for Standard 11istances ( As fleasured ' f rom E.1; c of Plant Cocplex) flites Sector 5;.0 N
2.65E-09 NNE 2.21E-09 NE 1.93E-09 ENE 1.94E-09 E
3.34E-09 ESE 4.73E-09 SE 5.17E-09 SSE 3.35E-09 S
2.07E-09 i
SSW 1.73E-09 SW 1 73E-09 WSW 2.07E-09 W
3.15E-09 VNV 3.89E-09 NV 3.63E-09 NNW 2.80E-09 Period of Record: 4/1/77 - 3/31/78 I Htt
_. _. _... _.... ~.. _.....
H4
=Rev. 12-t Page 108
.l i'r i
t i
-.s APPENDIX - B i
e Stability Prairie Island 12.2m ind and AT 42,7-12.2m
?
Joint' Frequency Distributions (4/1/77 - 3/31/78) f I.t
)
t I
I 1
Ii
)
v i
b r
~
f i
4
't t
IBM
[
_. - _ =
l H4 i
Rev, 12 i'
Page 109 NORTHERN STATES POWER COMPANY j
PRAIRIE ISLAND NUCLEAR GENERATING PLANT r
SITE METEOROLOGY - FREQUENCY DISTRIBUTION TABLES HOURS AT EACH VIND SPEED AND DIRECTION PERIOD:
4/1/77 THROUGH 3/31/78 i
i STABILITY CLASS A ELEVATION 40 FT-i WIND SPEED (MPli) AT 40 FT LEVEL f
DIRECTION 1 TO 3 4 TO 7 8 TO 12 13 TO 18 19 TO 24 ABOVE 24 TOTAL-i N
7 22 29 11 0
0 69
[
Nh1 13
.19 20 4
0 0
56 j
hT 11 35 16 1
0 0
63 l
t EhT 11 33 20 0
0 0
64 i
i E
14 37 24 0
0 0
75 l
JM ESE 4
45 49 7
2 0
107 i
17;M
?
SE 4
10 22 13 1
0 50 l
SSE I
7 19 12 2
0 41 S
2 23 45 27 0
0 97 i
SSW 3
22 39 14 0
0 76 SW 2
17 30 3
0 0
52 WSW 0
21 25 11 0
0 57 j
i W-1 29 46 18 2
0 96 WNW 6
34 64 56 20 1
181 f
NV.
12 42 72 53 20 0
199 NNV 11 43 49 20 2
0 125 VAR 0
0 0
0 0
0 0
t i
TOTAL IIOURS TIIIS CLASS 1408 lI0URS OF CALM Tl!IS CLASS 0
PERCENT Of ALL DATA THIS CLASS 16.81 i
IBM i
i t
H4 Rev. 12 Page 110 NORTHERN STATES POWER COMPANY-PRAIRIE ISLAND NUCLEAR GENERATING PLANT SITE METEOROLOGY - FREQUENCY DISTRIBUTION TABLES HOURS AT EACH WIND SPEED AhT DIRECTION PERIOD:
4/1/77 THROUGH-3/31/78.
STABILITY CLASS B' ELEVATION 40 FT WIND SPEED (MPH) AT 40 FT IIVEL DIRECTION 1 TO 3 4 100 7 8 TO 12 13 TO 18 19 TO 24 ABOVE 24 TOTAL N
O 3-5 1
o 0
9 NNE 1
2 1
1 0
0 5
NE O
2 0
0 0
0 2
ENE O
3 2
0 0
0 5
E O
1 1
0 0
0 2
ENk
.ESE 1
5 10 6
1 0
23 9.iiy SE 2
2 3
4 0
0 16 SSE O
3 4
3 0
0 10 E
1 0
7 9
0 0
17 SSW 0
1 7
0 0
0 8
SW 0
4 1
0 0
0 5
VSV 1
2-5 1
0 0
9 W
0 8
7 3
0 0
18 WNV 1
5 8
6 3
0 23 NW 2
4 11 10 1
0 28 t
- NNV 1
5 3
1 0
1 11 VAR 0
0 0
0 0
0 0
TOTAL HOURS THIS CLASS 191 i
HOURS OF CALM THIS CLASS 0
PERCENT OF ALL DATA THIS CLASS 2.28 IBM
..-.a
H4 Rev, 12 l
Page 111 NCRTHERN STATES POWER COMPAh7 PRAIRTE ISLAND hTCLEAR GENERATING PLANT SITE METEOROLOGY - FREQUENCY DISTRIBUTION TABLES HOURS AT EACH WIND SPEED AND DIRECTION PERIOD:
4/1/77 THROUGH 3/31/78 STABILITY CLASS C ELEVATION 40 TT WIND SPEED (MPH) AT 40 TT LEVEL DIRECTION 1 TO 3 4 TO 7 8 TO 12 13 TO 18 19 TO 24 AB0VE 24 TOTAL N
2 4
4 1
0 0
11 NKE 2
3 1
0 0
0 6
hT 1
5 1
0 0
0 7
ENE 0
3 1
0 0
0 4
E 1
8 3
0 0
0 12 ESE O
7 11 2
0 0
'O SE O
2 5
6 0
0 13 SSE O
2 6
7 1
0 16 5
0 2
10 4
0 0
16 SSW I
6 4
0 0
0 11 Sk 2
2 3
2 0
0 9
WSW 1
6 5
1 2
0 15 W
0-2 11 4
1 0
18 WNW 1
3 6
7 1
0 18 NW 2
7 11 16 6
1 43 NNW 3
5 3
3 0
21 VAR 0
0 0
0 0
0 0
TOTAL HOURS THIS CLASS 240 HOURS OF CALM THIS CLASS 0
PERCENT OF ALL DATA THIS CLASS 2.87 IBM
~... ~. = _ - -.. -...
i H4 Rev. 12
-Page 11' NORTIERN STATES POWER COMPAhT j
PRAIRIE ISLAND NUCLEAR GENERATING PLANT
-SIIE METEOROLOGY - FREQUENCY DISTRIBUTION. TABLES l
~
l Il0URS AT EACH WIND SPEED AND DIRECTION j
PERIOD:- 4/1/77 THROUGH 3/31/78 STABILITY CLASS D ELEVATION-40 FT i
VIND SPEED (MPil).AT 40 FT LEVEL i
DIRECTION 1 TO 3 4 TO 7 8 TO 12 13 TO 18 19 TO 24 ABOVE 24 TOTAL l,
t f
N
- 9 24 11 8
0 0
52 NNE 6
22 9
0 0
0 37 I
NE 16 26 4
0 0
0 46-
+
EhT 11 41 4
0 0
0 56-E.
11-95 27 0
0 0
133
'i X'
ESE 8-57 154 19 0
0 236
- c.g":
I
-SE.
10 30 90 38 5
0 173 SSE 8
40 59 51 10 0
168 5
1 51 72 17 4
0 145 SSV 5
29 30 12 0
0 76 SV 4
15 17
- 4 0
0 40 WSW
-5 23 31 21 3
4 87 j
V.
6 61 28 6
1 155 WNW 14
- 57 76 75 21 0
243
[
l
.NW 14 44 72 110 41 0
281 NNW 22 41 25 13-0 117 j
VAR 0
0' O
O O
O O
l l
TOTAL HOURS THis CLASS 2051 I=
HOURS OF CALM THIS CLASS 4
i PERCENT OF ALL DATA THIS CLASS 24.49 j
l i
IBM l
5 I
J
l H4 Rev.- 12 Page 113 NORTHERN'3TATES POWER COMPANY PRAIRIE ISLAND NUCLEAR GENERATING PIANT i
SITE METEOROLOGY - FREQUENCY DISTRIBUTION TABLES j
i HOURS AT EACH WIND SPEED AND DIRECTION i
PERIOD:
4/1/77 THROUGH 3/31/78 STABILITY CLASS E ELEVATION 40 FT l
~
WIND SPEED (MPH) Af~40 IT LEVEL DIRECTION 1 TO 3 4 TO 7 8 TO 12 13 TO 18 19 TO 24 ABOVE 24 TOTAL N
22 30 9
11 1
0 73
[
i NNE la 29 7
0 0
0 54 t
NE 22 26 7
1 0
0 56
[
t ENE 19 30 5
1 0
0 55
[
E 25 96 10 0
0 0
131 ESE 28 144 140 27 0
0 349
-l yfy SE 24 107 125 41 2
0 299
{
SSE 21 67 74 23 0
0 185 l
S 11 56 73 29 1
0 170 SSW-3 26 29 40 1
0-99 l
SW 14 22 17 12 0
0 65 VSW 14 24 24 11 1
0 74 I
W 26 73 48 18 1
0 166 t
WNW 46-136 127
-44 4
0 357 j
l
.NW 46 98 101 62 8
0 315 l
NNV 43 53 48 10 3
0 157 VAR-0 0
0 0
0 0
0 t
i I
TOTAL HOURS THIS CLASS 2612 HOURS OF CALM THIS CLASS 7
[
PERCENT OF ALL DATA THIS CLASS 31.18
[
1 IBM i
h
_ _ _ ~. _
l H4 Rev 12 Page 114 NORTHERN STATES POWER COMPANY-PRAIRIE ISLAND NUCLEAR GENERATING PLANT SITE METEOROLOGY - FREQUENCY DISTRIBUTION TABLES HOURS AT EACH WIND SPEED AND DIRECTION PERIOD:
4/1/77 THROUGH 3/31/78 STABILITY CLASS F ELEVATION 40 FT WIND SPEED (MPH) AT 40 FT LEVEL DITICTION 1 TO 3 4 TO 7 8 TO 12
-13 TO 18 19 T0 24 ABO \\T 24 TOTAL N
18 8
3 0
0 0
29 NNE 11 6
1 0
0 0
18 hT 11 5
2 0
0 0
18 ENE 13 7
0 0
0 3
20
-E 29-33 2
0 0
0 64 c
ESE 39 61 9
1 0
0 110
-.5 SE 38 69 36 3
0 0
146 1
SSE 27 32 17 2
1 0
79 S
12 16 21 7
0 0
56 SSW 6
11 17 0
0 0
40 SV 5
3 9
4 0
0 21 WSW 8
8 8
0 0
0 24 W
25 39-
'12 2
0 0
78 WNW 56 63 12 0
0 0
131 hN 66 71 16 3
0 0
156 Nhv 29 19 6
2 0
0 56 VAR 0
0 0
0 0
0 0
TOTAL-HOURS THIS CLASS 1053 HOURS OF CALM THIS CLASS 7
PERCENT OF ALL DATA THIS CLASS 12.57 IBM.
~
11 4 Rev. 12 Page 115 NORTl[ERN STATES POWER COMlitiY PRAIRIE ISLAND NUCLEAR GENERATING PLANT SITE METEOROLOGY - FREQUENCY DISTRIBUTION TABLES HOURS AT EACH WIND SPEIO 4D DIRECTION PERIOD:
4/1/77 THROUGH 3/31/78 STABILITY CLASS G ELEVATION 40 FT WIND SPEED (MPH) er 40 FT LEVEL DIRECTION 1 TO 3 4 TG 7 8 TO 12 13 10 16 19 TO 24 ABOVE 24 TOTAL N
14 5
0 0
0 0
19 NKE 13 2
1 0
0 0
16 NE 12 2
1 0
0 0
15 s
ENE 22 1
2 0
0 0
25 E
52 8
2 0
0 0
62 ESE
$0 17 1
0 0
0 68 N)
SE 37 23 8
1 0
0 69 SSE 18 8
7 2
0 0
35 S
11 4
4 0
0 0
19 SSV 13 2
2 0
0 0
17 SW 15 5
1 0
0 0
21 WSW 10 1
O O
O 12 W
41 19 1
0 0
0 61 VNW 75 50 0
0 0
0 125 gW 80 66 3
0 0
0 149 NNV 47 19 5
0 0
0 71 VAR 0
0 0
0 0
0 0
TOTAL HOURS THIS CLASS 821 HOURS OF CALM THIS CLASS 37 PERCENT OF ALL DATA THIS CLASS 9.80 IBM t --
H4 Rev. 12 Page 116 APPENDIX C DOSE PARAMETERS FOR RADIOIODINES, PARTICULATES AND TRITIUM This appendix contains the methodology whicu was used to calculate-the doce parameters for radioiodines, perticulates, and tritium to show compliance with 10 CFR 20 and Appendix 1 of 10 CFR 50 for gaseous effluents.
These dose parameters.
. and R, were g
calculated using the methodology outlined in NUh2G-0133 along with Regulatory Guide 1.109 Revision 1.
The following sections provide the specific methodology which was utilized in calculating the P and R values for the various exposure g
1 pathways.
C.1 Calculation of P; The parameter, P contained in the radioiodine and 4,
particulater, portion of Section 3.2, includes pathway transport parameters of the ith radionuclide, the receptor's gn;)
usage of the pathway media and the dosimetry of the Gn exposure.
Pathway usage rates and the internal dosimetry are functions of the receptor's age: however, the child age group, will always cr.ceive the maximum dose under the exposure conditions assumed.
C.l.1 Inh _alation Pathway K'(BR) DFA (C.1-1)
P g
4^I where:
dose parameter for radionuclide i for the P.
=
1 inhalation pathway, mrem /yr per pCi/m3; 1
K'
=
a constant of unit conversion:
6 10 pCi/pCi;
=
the breathing rate of the child age group, BR
=
m /yr; the maximum organ inhalation dose factor DFA.
=
for the child age group for radionuclide i, 1
mrem /pci.
IBM
m.
H4 Rev. 12 Page 117 Theagegroupconsideredisthecgildgroup.
The child's breathing rate is taken as 3700 m /yr from Table E-5 of
~
Regulatory Guide 1.109 Revision 1.
The inhalation dose factors for the child DFA1, are presented in Table E-9 of-Regulatory Guide 1.109 1n units of mrem /pC1.
The total body is considered as an organ in the selection of DFA Theincorporationofbreathingrateofthechildandtke.
unit conversion factor results in the following:
9 3.7 x 10 DFA.
(C.1-2)
P.
=
1 1
7 C.2 Calculation of R.
The radiciodine and particulate Technical Specification is applicable-to the location in the unrestricted area where the combination of existing pathway.' and receptor age groups
. indicates-the maximum potential exposure occurs.
Tne inhalation and ground plane exposure pathways shall be considered to exist at all locations.
The grass-goat-milk, the* grass-cow-milk, grass-cow-meat, and vegetation pathways are considered based on their existence at the various locations.
R. values have been calculated for the adult, teen, child, 3nd infant age groups for the ground plane, cow l milk, goat milk, vegetable and. beef ingestion pathways.
The
. methodology which was utilized to calculate these values is
.^"
presented below, i
i C.2.1 Inhalation Pathway 1
1 g
K'(BR)a (DFAg),
(C.2-1)
[
R
=
I i
where:
l R,
=
dose factor for each identified 1
l 1
radionuclide i of the organ of interest, L
mrem /yr per pCi/m3; r
a constant of unit conversion:
K'
=
6
=
10 pCi/pCi; 1
breathing ratg of the receptor of l
l (3R)a
=
l age group a,m fyr:
i organ inhalation dose factor for (DFA ),
=
f l
radionuclide i for the receptor of l
age group a, mrem /pCi.
{
~The breathing rates (BR) for the various age groups are tabulated below, as gives in Table E-5 of the Regulatory Guide 1.109 Revision 1.
i IBM l
r
i H4 Rev. 12 Page 118 3
Ace Group (a)
Breathinn Rate (0 /yr)
Infant 1400 Child 3700 Teen 8000 Adult 8000 Inhalation dose factors (DFA )
for the various age groups g
r.tven in Tables E-7 throug8 E-10 of Regulatory Guide are 1.109 Revision 1.
C.2.2 Ground Plane Pathway
-l t I K'K"(SF)D7G (1-e i)A (C.2-2)
R
=
g g
3 g
G where:
R'.
dose factor for the ground plane pathway
=
c for each identified radianuclide i for the organ of interest, m' mrem /yr per pC1/sec per;
- c4 K'
a constant of unit conversion; nn 6
=
10 pCi/pci; 1:"
=
a constant of unit conversion; 8750 hr/ year;
=
A
=
the radiological decgy constant for g
radionuclide i, sec the exposure time, sec; t
=
8 4.73 x 10 sec (15 years);
=
the ground plane dose conversion factgr DFG
=
g for radionuclide i; mrem /hr per pCi/m";
the shielding factor (dimensionless);
SF
=
I
=
factor to account for fractional f
deposition of radionuclide i.
For radionuclides other than iodine, the factor I is to one.
For radiciodines, the value of I kay equal However, a value of 1.0 was used in calcblating vary.
the R values in Table 3.3-2.
IBM 7
11 4 Rev. 12 Page 119 A shielding factor of 0.7 is suggested in Table E-15 of Regulatory Guide 1.109 Revision 1.
A tabulation of DFG. values is presented in Table E-6 of Regulafory Guide 1.109 Revision 1.
C.2.3 Grann-Cow or Goat-Milk Pathway
~
-At E
O fI~'
g 3)
.x t
i*f r(1-c i 'P) +
iv
+
~
R I
f p 9,(Ufb )a '
1
" 1 E,Q V i
ps YA PA 3
3 Pf i
i-At"
-A t
~
r (1 -
i "_) +
B, (1-c 1 b)
,y
(),,g g )
g
~
where:
E dose factor for the cow milk or goat milk g
M pathway, for each identified tadionuclide i for the organ,gf interest, mrem /yr per pCi/sec per m K'
=
a constant of unit conversion; f
10 ' pC1/pCi;
=
g,9
=
the cow's or goat's feed consumption rate, Q 7 kg/ day (wet weight);
the receptor's milk consumption rate for age U
=
"P group a, liters /yr; Y
the agricultural productivigy by unit area E
of pasture feed grass, kg/m ;
Y
=
the agricultural procjuctivity oy unit area y
of stored feed, kg/m';
the stable element transfer moefficients, F*
=
pCi/ liter per pCi/ day; fraction of deposited activity retained on r
=
cow's feed grass; the organ ingestion dose factor for (DFL )"
g radionuclide i for the recept.or in age group a, mrem /pci; 1BM
11 4 Rev.12 Page 120 A
=
A g+Ay; g
g the radiological decfy constant for A
=
radionuclide i, sec the decay constant for removal of activity A"
=
on lyaf and plant surfaces by weathering, sec 5.73 x 10*7
~1 (corresponding to 2
=
see 14 day half-life);
the transport time from feed to cow or goat t
=
g to milk, to receptor, see; h
the transport time from harvest, to cow or t
=
goat, to consumption, sect period of time that activity builds up in t
=
b soil sec; concentration factor for uptake of B
=
iv radionuclide i from the soil by the edible parts of crops, pC1/Kg (wet weight) per pCi/Kg (dry soil);
effectlye surface density for soil, (dry P
=
soil)/m';
f
=
fraction of the year that the cow or goat p
is on pasture; f
=
fraction of the cow feed that is pasture 8
grass while the cow is on pasture; t,p period of pasture grass exposure during I
=
the growing season, see; period of crop exposure during the growing t
=
e, season, sec; j'
1
=
factor to account for fractional deposition 1
of radionuclide 1.
For radionuclides other than iodine, the factor 1 is equal to one.
For radioiodines, the value of I day vary, liowever, a value of 1.0 was used in calcuIating the R va'ues Tables 3.3-9 through 3.3-16.
Milk cattle and goats are considered to be fed from two potential sources, pasture grass and stored feeds.
Following the development in Regulatory Guide 1.109 Revision 1, the value of f was considered unity in lieu of site-specific infoFmation.
The value of f p was 0.5 based upon an 6-month grazing period, IBM
H4 Rev. 12 Page 121 i
Table C-1 contains the appropriate parameter values and j
their source in Regulatory Guide 1.109 Revision 1.
l
~
The concentration of tritium in milk is based on the airborne concentration rather than the deposition.
Therefore, the R is based on X/Q:
l g
T m F apIDII'i)a 0.75(0.5/H)
(C.2-4)
{
E'E"'I 0 U R
H i
where:
[
d se factor for the cow or goat milk pathway l
R
=
T fortritiumfortgeorganofinterest, l
n mrem /yr per pCi/m ;
j a constant of unit conversion; i
K"'
=
3
=
10 Em/kg; 3
absolute humidity of the atmosphere, gm/m ;
H
=
e the fraction of total feed that is water; 0.75
=
t the ratio of the specific activity of the 0.5
=
feed grass water to the atmospheric water.
f
. a._"
and other parameters 3nd values are given above.
A value of H of 8 grams / meter, was used in lieu of site-specific i
information.
{
C.2.4 ' Grass-Cow-Meat Pathway The integrated concentration in meat follows in a similar manner to the development for the milk pathway, f
therefore:
j
~
-A t
~
ga r(1-e i ep) +Biv(1-e
)
-A t E
+
1 K'Q UF ap f(DTL ),.e gg R
F
=
3 1
7
_3 g r(1-e i e s) +
Biv(1-e I D)
-l t e
ih (C.2-5) ff)
Ps Y,AE 1
l IA l
I l
l where:
[
t R
=
dose factor for the meat ingestion pathway i
g f r radionuclide i for any organ of interest, j
9 a
m' mrem /yr per pCi/sec ;
[
f i
L IBM i
H4 Rev. 12 Page 122 r.
=
the stable element transfer coefficients.
I pCi/Kg per pCi/ day; the receptor's meat consumption rate for age U
=
- P group a, kg/yr; the transport time from slaughter to
=
t s consumption, sect the transport time from harvest to animal t
=
h consumpt. ion, sect period of pasture grass exposure during t
"P the growing season, sec; period of crop exposure during the growing t
es season, sec; factor to account for fractional deposition I
=
f of radionuclide 1.
mayvaby.sequal i
the factor 1 For radionuclides other than iodine, For radiciodines, the value of I However,avalueof1.0wasusedincalcu$atingtheH to one.
values in Tables 3.3-6 through 3.3-8.
All other terms remain the same as defined in Equation
!*4,
'3h Table C-2 contains the values which were used C.2-3.
for the meat pathway.
in calculating Rg is based on its The concentration of tritium in meat airborne concentration rather than the depocition.
Therefore, the R is based on X/Q.
g 0.75(0.5/H)
(C.2-6)
K'K"'F Q U,p(DFL ),
f7 1
R
=
T B
where:
R
=
dose factor for the meat ingestion pathway for tritium for agy organ of interest, T
B mrem /yr per pCi/m All other terms are defined in Equation C.2-4 and C.2-5, above.
C.2.5 Vegetation pathway The integrated concentration in vegetation consumed by man follows the expression developed in the derivation Man is considered to consume two of the milk factor.
that differ only types of vegetation (fresh and stored) time period between ha in the therefore:
IBM
_. - - - -.. ~..
i 114 Rev. 12 Page 123
~A t I
-A t
b II ~' i h) f L
iL r(1-e i ")
+
iv
~
y g, g )a pg g
1 i
aL, YA PA 1,
E g
i a
-At
-A t
I E
Bgy(1-e i b)
S e r(1-e i *)
(C.2-7) l gg 4
ag YA N
{
yg g
i I
where:
l t< l dose factor for vegetable pathway for v
radionuclidg,1 for the organ of interest, m' mram/yr per pCi/sec h' '
a constant of unit conversion:
i 6
=
10 pCi/pci; b
I U*
the consuw> tion rate of fresh leafy vegetation by the receptor in age group a, kg/yr; in 3
the consumption rate of stored U*
=
vegetation by the receptor in age group a, kg/yr; the fraction of the annual intake of l
f g
fresh leafy vegetation grown locally; f
=
the fraction of the annual intake of 8
stored vegetation grown locally; the average time between harvest of t
=
g leafy vegetation and its consumption, r
sec; the average time between harvest of t
=
h stored vegetation and its consumption, sec; the vegetation areal density, kg/m2; Y
=
y pe'iod of leafy vegetable exposure t
=
during growing season, see; fact $r to account for fractional I.
=
1 deposition of radionuclide 1.
l k
IBM
i H4 Rev. 12 Page 124 i
i For radionuclides other than iodine, the factor I, is i
For radiofodines, the value of I
&ay i
equal to one.
However,avalueof1.0wasusedincaleblating vary.
j the R values in cables 3.3-9 through 3.3-12.
All other factors were defined above.
Table C-3 presents the appropriate parameteb values and I
their source in Regulatory Guide 1.109 Revision 1.
In lieu of site specific data default values for f and i
g f, 1.0 and 0.76, respectively were used in the l
cElculation of R These values were obtained from TableE-15ofReku.latory Guide 1.109 Revision 1.
i The concentration of tritium in vegetation is based on the airborne concentration rather than the deposition.
Therefore, the R is based on X/Q g
j R
K'K"'
Uf
+
Uf (DTt ),
0.75(0.5/H)
(c.2-8)
T.
=
g g
g i
i where:
.. c, dose factor for the vegetable pathway for MV!
R
=
T tritium fgr any organ of interest, mrem /yr y
per pCi/m i
All other terms remain the same as those in Equations C.2-4=and C.2-7.
t t
h I
i I
i h
-i.
i l
l IBM f
[
H4 Rev. 12 TABLE C-1 Parameters for Cow and Goat Milk Pathways Parameter Yelue Reference Q7 (kg/ day)
(Reg, Guide 1.109 Rev.
50 (cow)
_1
{
6 (goat) 7able E-3 Table E-3 Yp (kg/m )
0.7 Table E-15 tg (seconds) 1.73 x 105 (2 days) l Table E-15 r
1.0 (radiciodines) 0.2 (particulates)
Table E-15 Table E-15 (DFL ),
(mrem /pci)
Each radionuclide Tables E-11 to E-14 F, (pci/ day per pCi/ liter)
Each stable element Table E-1 (cow)
W as Table E-2 (goat) y/r tb (see nds) 4.73 x 103 (15 yr)
Table E-15 Y, (hg/m )
2.0 Table E-15 Y (kg/nS) 0.7 Table E-15 th (seconds) 7.78 x 106 (90 days)
Table E-15 V8P (liters /yr) 330 infant 330 child Table E-5 400 teen Table E-5 310 adult Table E-5 Table E-5 t
(seconds) 2.59 x 10 (30 days)
Table E-15 t
(seconds) 5.18 x 106 (60 days)
Table E-15 B.
fdr(pC1/Kg (dry soil))pCi/Kg (wet weight)
Each stable element Table E-1 P (Kg (dry soil /m )
IBM 240 Table E-15
k H4 Rev. 12 Page 126 TABLE C-2 Parameters For The Heat Pathway j
Parameter Value Reference (Reg. Guide 1.109 Rev.-1 i
I r
1.0 (radiciodines)
Table E-15 i
0.2 (particulates)
Table E-15 l
Ff (pCi/Kg per pCi/ day)
Each stable element Table t-1 l
8P (Kg/yr) 0 infant Table E-5 U
41 child Table E-5
{
65 teen Table E-5 i
110 adult Table E-5 (DFL ), (mrem /pci)
Each radionuclide Tables E-11 to E-14 f
g f
2 Yp (kg/m )
0.7 Table E-15 cu'A W
Y, (kg/m')
2.0 Table E-i$
l b (seconds) 4.73 x 108 (15 yr)
Table E-15 t
0 (20 days)
Table E-15 t, (seconds) 1.73 x 10 f
h (see nds) 7.78 x 100 (90 days)
Table E-15 f
t 2.59 x 100 (30 days)
Table E-15 t,p (seconds) t (seconds) 5.18 x 10 (60 days)
Table E-15 f
Qg (kg/ day) 50 Table E-3
?
Y Bg (pCi/K3 (wet weight)
Each stable element Table E-1 j
per pCi/Kg (dry soil))
i A
i P (Kg (dry soil)/m )
240 Table E-15 1
IliM L
E
_. _ _ _ _ _ _ _ _ _ _ _ _.. _ j
11 4 Rev. 12 l
Page 127 l
TABLE C-3 l
Parameters For The Vegetable Pathway j
i l
f Parameter Value Refference (Reg. Guide 1.109 Rev. 1 e
r (dimensionless) 1.0 (radiofodines)
Table E-1 l
0.2 (particulates)
Table E-1 l
(DFL ), (mrem /pci)
Each radionuclide Tables E-11 to E-14 g
U (kg/yr) - Infant 0
Table E-5
- Child 26 Table E-5
- Teen 42 Table E-5
- Adult 64 Table E-5 i
=U (kg/yr)
Infant 0
Table E-5
- Child 520 Table E-5
- Teen 630 Table E-5
- Adult 520 Table E-5
.[^j g (seconds) 8.6 x 10 (1 day)
Table E-15 t
h (see ads) 5.18 x 106 (60 days)
Table E-15 t
Y (kg/m )
2.0 Table E-15 y
0 (60 days)
Table E-15 t, (seconds) 5.18 x 10 g (seconds) 4.73 x 100 (15 yr)
Table E-15 P(Kg(dry soil)/m )
240 Table E-15 l-i Bgy(pci/Kg(wet weight)
Each stable element Table E-1 per pCi/kg (dry soil))
IBM
o PINGP 753, Rsv. 3 Page 6 of 6 r
PRAIRIE ISLAND NUCLF).R GENERATING PLANT Period: 1-1-91 to 6-30-91 i
IORTHERN STATES POWER License No. DPR-42 l
SOLID RADIOACTIVE WASTE DISPOSAL SEMI-A!NUAL REPORT Tablo II: Process Control Program Changes
Title:
Process Control Program for Solidification /Devatering of Radioactive Waste From Liquid Systems.
Current Revision Numbor 4
Effective Date: _ 5-23-91 tmE If the effective date of the PCP is within the period covered by this
[
retx>rt, then a description and justification of the changes to the PCP is required (T.S. 6.7.A.4).
Attach the sidelined pages to i
this report.
j i
Changes / Justification:
i
~
n 1.0 Section 1.1 Putposeg Changer Colon to period.
Justifications Correct punctuation.
l l
2.0 Section 1.2 Scope; Change Added section 1.2.9, Reporting'Ruquirementu.
Renamed section 1.2.0.
Justification: N/A.
3.0 Section 2.5.4 Flowrateng Changes Removed extra minus signs.
Justification Incorrect as found.
i 4.0 Section 9.0 Processing /Dewatoring of Spent Filter Elements; Change Rewrote sections 9.1, 9.2, and 9.3.
l Justification: Sections rewritten to provido a generic methodology f or filter disposal.
5.0 Section 10.0 In-Container Solidification of Dead Resing l
Changes: This section in its entirety is new.
i Justifications This section previously existed as an Appendix to this PCP.
In-Container solidi fication al-though not normally used at Prairie Island is a viable I
method f or complying with 10 CFR 61 stability require-ments.
As such it is included in this PCP.
i i
6.0 Section 11.0 Reporting Requirements; Changes: This section in its entirety is new.
l Justifications This section is a direct response to NRC l
.m m-.
,,. _ ~ _. -. _. -, -.. -... _,
DESCRIPTION OF REVISION 12 ODCM MANUAL CllANGES Effective Date:
20-JUti-1991 The following is a list of changes made to the Offsite Dose Calculation Manual during the first reporting period of 1991.
Some changes in operation were discussed via phone conversation with the NRC at the time of-the. changes.
This is the formal notification of our ODCM change in accordance with T.S.
6.5.E.1.
Item 1:
In section 2,2 text was added to describe the increased operational controls of liquid waste batch releases to ensure compliance with 10 CFR 20 following the release of batch tanks with maximum acceptable release-rates less than that of the maximum possible release rate from all release source paths.
Plushing-the discharge pipe at or below the calculated rate with lower activity water prior to additional releases prevents the higher activity release from being pushed out of the pipe at a possible higher release rate than originally calculated.
These controls-are necessary since our discharge pipe has been lengthened and its capacity will now hold the volume of 2 to 3 tanks.
Item 2:
In section 3.2.2, equation 3.2-2 the definitions for DV and Dp were edited to reflected the descriptions agreed upon with the NRC office via phone conversations.
The definition changes do not change the equation, they clarify its use.
Item 3:
Table 5.1-1 was revised to reflect changes in the Radiation Environmental Monitoring Progr m Sampling Locations and/or description of sample locations.
Changes were made-to the location / description of the following sample points.
Well Water Milk cultivated crops (leafy green vegetables)
--.. -..- =...
"*l l'
PRAIRIE ISLAMD tJUCLEAR GEtJERAllMO PLAlli tJORVHEnti STATES POWER COMPAtJV IAAltJTElJAllCE PROCEDURES TITLE:
IJUM DER:
M PROCESS CONTROL PROGRAM DS9 l
L.F FOR SOLIDIFICATION / DEWATERING REV :
4 OF RADIOACTIVE WASTE section FROM LIQUID SYSTEMS Page 1 of 27 O.C. REVIEW DATE:
REVIEWED DY:
i D AT E: -
g rmm - - /2 / M e a t o n n
L-n w *.e. vt s n uI PRAlRIE ISLAND NUCLEAR GENERATING PLANT j
NORTHERN STATES POWER COMPANY MAINTENANCE PROCEDURES TITLE:
NuuBER:
M PROCESS CONTROL PROGRAM D59 t
Ear FOR SOLIDIFICATION / DEWATERING REV:
4 ssetion.
OF RADIOACTIVE WASTE FROM LlOUlD SYSTEMS page 2 of 27 TABLE OF CONTENTS Section Title Pago 1.0 GENERAL...................................
5 i
1 1.1 Purpose...........
5 f
1.2 Scope..................................
5 1.3 D e fi n i ti en s...............................
5 l
1.4 Applicable Tech. Spec.........................
6 2.0 SOLIDIFICATION OF LIQUID WASTE CONCENTRATES.........
6 i
2.1 Purpose................................
6 l
2.2 A pplic ability..............................
6 l
2.3 Referencet 6
2.4 System Description............
6 2.5 Sequence of Operation........................
9
(
2.6 Sample Solidification of Liquid Waste Concentrates.......
10 l
3.0 MANUAL SOLIDIFICATION OF WASTE LIQUIDS.............
13 i
t I
3.1 Purpose...........
13 3.2 A ppli c a bili ty....................,.....,,..
13
[
3.3 Sequence of Operation................,.......
13 l
3.4 Cure Time 13 3.5 Verification of Solidification 13 t
4.0 PROCESSING OF CERTAIN WASTE LIQUIDS THRU SPENT BEAD l
RESIN.....................................
14 f
E 4.1 Purpos e.............
14 1
4.2 Applicability................,.........
14 4.3 Sequence of Operation..
15 i
4.4 Dewatering Procedure..
15 t
5.0 MANUAL SOLIDIFICATION OF WET TRASH BY SUBMERSION....
15 5.'
Purpose,,..,.....
15 5.2 Applicability...........
15 l
5.3 Sequence of Operation.....
15 5.4 Cure Time....,.......
16 l
5.5 Verification of Solidification 16 l
5.6 Disposition...
16 l
f 1
. - -,.~,,, -,,, n
. -. -. -.,. -,...,., - - -,, - -, - - - -. - - -,. -..,, +, - -, -, - -,...
PRAIRIE ISLAND NUCLEAR GENERATING PLANT
'IM f
' NORTHERN STATES POWER COMPANY MAINTENANCE PROCEDURES r TITLE:
NUMDER:
l M'
PROCESS CONTROL PROGRAM D59 I
- W FOR SOLIDIFICATION / DEWATERING 4
section -
OF RADIOACTIVE WASTE FROM LIQUID SYSTEMS pago 4 of 27 Soction Titio Pago 11.0 REPORTING REQUIREMENTS [ continued]
J l
11.2 Applicability..............................
23 l
11.3 References..............................
23 l
11.4 PCP Revisions............................
23 11.5 Reports of Mishaps..........................
24 11.6 PCP Specimon Summary Reports.................
24 12.0 ATTAC H M E NTS...............................
24 Attachment A - Sample Verification Form....................
26 Attachment B - Sample Verification Form....................
27 j
List of F10uros i
l 1
SOLID RADWASTE FLOW DIAGRAM 25 t
i i
r i
I I
I
PRAIRIE ISLAND NUCLEAR oENERATING PLAN 7
'*" ' '" * ' M NORTHERN STATES POWER COMPANY MAINTENANCE PROCEDURES TITLE:
NUM DER:
o, v.
h7 PROCESS CONTROL PROGRAM D59 Ear FOR SOL!DIFICATION/ DEWATERING 4
i sec'tlon -
OF RADIOACTIVE WASTE FROM LIQUID SYSTEMS page a of n Section Title Pa00 6.0 PROCESSING OF WET TRASH BY COMPACTION / CEMENTATION..
16 i
i 6.1 Purpo s e................................
16 6.2 Applic ability..............................
16 l
6.3 Sequence of Operation.........,..............
17 i
6.4 C u r e Ti m e...............................
17 6.5 Verification of Absence of Free Water...............
17 6.6 Dis po s ition.........
17 l
7.0 DEWATERING OF BEAD RESlN......................
17 l
)
{
7.1 P u r po s e................................
17 7.2 Applicability..............................
18 j
7.3 Dewatering Drocedure........................
18
[
7.4 Verification of Dewatering..............,,......
18 l
t 8.0 DEWATERING OF POWDERED RESIN..................
18 l
i 8.1 Purpose................................
18 i
8.2 Applicability........................
18 8.3 System Des cription..........................
19 8.4 Di s po s a l................................
19 i
9.0 PROCESSING / DEWATERING OF SPENT FILTER ELEMENTS.,,...
20 l
1 l
9.1 Purpose..........
20 l
9.2 A ppli c abili ty..............................
20 9.3 Description of Filling Process..........
20 9.4
. Dewat e rin g...............................
20 j
9.5 Ve%cntion of Dewatering............
21 i
10.0 IN-CONTAINER SOLIDIFICATION OF BEAD RESIN...........
21 10.1 Purpose...............
21 10.2 Applic ability...................,
22 10.3 R efe ren c e s..............................
22 10.4 System Description.....
22 10.5 Sam pling............................
23 11.0- REPORTING REQUIREMENTS 23 11.1 Purpose....
23 l-l i
i i
i i
l h
l
-.. - - - =.
PRAIRIE ISLAND Nuct. EAR GENERATING PLANT P
f
' NORTHERN STATES POWER COMPANY MAINTENANCE PROCEDURES j
TITLE:
NuMDER:
j
'h PROCESS CONTROL PROGRAM D59 j
Ear FOR SOLIDIFICATION / DEWATERING
~
l 4
~ section.
OF RADIOACTIVE WASTE
)
FROM LIQUID SYSTEMS page s of n l
~
1.0 - GENERAL 1.1 Purpos_q The purpose of this Process Control Program (PCP) is to detail the means l
by which the dewatering and/or solidification of radioactive waste from liquid systems can be assured.
l
-l 1.2 hqp_q This PCP includes the following processes:
l 1.2.1 Solidification of liquid waste concentrates.
l 1.2.2 Manual solidification of ~ waste liquids.
1.2.3 Manual solidificatlon of wet trash by submersion.
l 1.2.4 Processing of wet trash by compaction / cementation.
j i
1.2.5 Dewatering of bead resin.
1.2.6 Dewatering of powered resin.
l 1
1.2.7 Dewatering of spent filter elements, 1
1.2.8 In-container Solidification of Bead Rosin.
i 1.2.9 Reporting Requirements.
1.3 Definitions
[
1.3.1 Datch A quantity of liquid waste concentrates (for example, the contents of 121 Waste Concentrates Tank) to be solidified. A batch of waste
[
concentrates can normally be drummed in not more than two days.
l l
1.3.2 Solidification h
i
?
s l
The conversion of wet radioactive wastes into a form that meets shipping and bunal ground requirements.
i l
I I
L a
E
PRAthlE isLAt4D tiUCLEAR GEf4ERAlltJG PLAN 7 NORTHERN STATES POWER CoMPAtJY tAAll4TENANCE PROCEDURES TITLE:
NUMDER:
PROCESS CONTROL PROGRAM DS9 D
FOR SOLIDIFICATION / DEWATERING "EV 4
section OF RADIOACTIVE WASTE FROM LIQUID SYSTEMS Page o of 27 1.3.3 Dewatering The process of removing water from a substance to meet specific limits.
1.4 Appil. cable Tech. Sp_ep T.S.3.9.C 2.0 SOLIDIFICATIOf1 OF LIQUID WASTE COtJCENTRATES 2.1 P_u ritgs e To establish the process parameters which provide reasonable assurance of complete solidification of liquid waste concentrates.
2.2 Applicability This section of the PCP is applicable to solidification of liquid waste concentrates using the Atcor Solidification System and related equipment.
2.3 R e f e r.e n_.c e_s.
2.3.1 C21.2.1 Solid Radioactive Waste Operating Procedure 2.3.2 C21.2.2 Trash Compactor Operation Operating Procedure 2.4 Sy_s t e_m_D es c ription 2.4.1 General Des _c_ription The solidification system for liquid waste concentrates includes 121 Waste Concentrates Tank ONCT), the Atcor Solidification System and related pumps. piping and equipment. Concentrates are accumulated from the 5 GPM ADT evaporator of the 2 GPM waste evaporator and stored in 121 WCT. When a sufficient quantity exists in 121 WCT. the contents are transferred to the Atcor system for solidification in 55 gallon drums. The filled drums are held in the Atcor drum storage aisles until solidification can be confirmed. The drums are then capped. decenned, and surveyed prior to storage for subsequent shipment and disposal. A flow diagram is shown on Figure 1.
t l
i PRAIPIE ISLAND NUCLEAR GENERATING PLANT N
[
NORTHERN STATES POWER COMPANY MAINTENANCE PROdETURES i
7 W
i TITLE:
NUMBER:
[
' M PROCESS CONTROL PROGRAM D59 Bar P
FOR SOLIDIF10ATION/ DEWATERING REW 4
(
- ssction _
OF RADIOACTIVE WASTE FROM LIQUID SYSTEMS Page 7 of 27 l
I 2.4.2 Detailed Description i
1 A.121 Waste Concentrates Tank t
121 WCT is an upright cylindrical, vented tank of approximately 1700 gal, capacity. The tank is electrically heated to keep the contents in solution.
121 WCT receives concentrates from either the 5 GPM ADT Evaporator, the 2 GPM Waste Evaporator, or the coagulation tank. The tank is located in a shielded vault for radiation i
protection and is equipped with a high level alarm to prevent I
over-filling. A direct reading float-type level gauge provides level indication from outside of the shielded vault.
i 121 WCT pump and discharge piping are arranged for j
recirculation and mixing of the tank contents or for pumping the contents to the Atcor System for solidification. A sample valve is provided near the pump discharge.
l
- 8. Atcor Solidification System I
I L
1 i
The Atcor Solidification System is designed to mix waste liquid l
concentrates with cement, to convey the blended mass into 55 i
gal, drums, and to store the filled drums in a chielded area for i
curing. The system consists of the following principle i
components:
l 1.
Waste Meterino Tank The waste metering tank is a tank of approximately 700 gal process capacity. The tank is equipped with heaters to l
maintain contents in solution and is equipped with an l
agitator to ensure homogeneity of liquid. The tank is equipped with a positive displacement discharge pump l
having discharge rate variable up to 10 GPM. The pump discharges directly into the mixer feeder.
[
i l
I I
PRAIRIE ISLAND NUCLEAR GENERATlHG PLANT lUl' DATED Pj. OPNSl NoMTHERN STATES POWER COMPANY MAINTENANCE PROCEDURES
'M M
TITLE:
NuMDEFl M
^
"1
. Em#
FOR SOLIDIFICATION / DEWATERING 4
4.etion :
OF RADIOACTIVE WASTE FROM LIQUID SYSTEMS page a of 27
- 2. Cement Bin i
The cement bin is a bin of approx.100 cu. ft. process capacity. The bin is equipped with a vibrating lowsr cone to preclude bridging of cement and to ensure uniform flow of material having a consistent bulk density. The cement bin 3
is fitted with a discharge auger having a discharge rate variable from 0.3 to 3.3 CFM. The auger discharges directly into the mixer / feeder.
i
- 3. Mixer / Feeder l
The mixer / feeder is a double enveloping screw type mixer which simultaneously blends the liquid waste and cement while conveying the mass to the discharge chute. The dist.;arge rNw (rects the blended mass into the shipping container by yavity flow.
4.
'pnts,,Is rmtroit for the solidification system are contained on a j
panel shielded from the waste materials.
Gauges indicating feed rate of cement and waste liquid are j
located on the control panel.
i Rates of cement feed and liquid feed are adjustable from j
the control panel during processing.
l A closed circuit TV camera and monitor are provided for viewing drum movements frorn the control panel.
l l
S.
Gement Tvoe
{
i Cement normally used is type "M" masonry cement havino I
50% lime and 50% portland cement, conforming to ASTM-C-91-64 and ASTM-C-270-61T.
I t
t 5
j s.
i i
, ~.. _ _ _....
"*"'#'N' PRAIRIE ISLAND NUCLEAR GENERATlHQ PLANT l
NORTHERN STATES POWER COMPANY MAINTENANCE PROCEDURES TITLE:
NUMDER:
PROCESS CONTROL PROGRAM D59 DY FOR SOLIDIFICATION / DEWATERING
+
REV:
4 isection 9 OF RADIOACTIVE WASTE h
FROM LIQUID SYSTEMS Page 9 of 27 l
t l
2.5 Secuence of Operation
- 2. ? -
Gygenslaition of 121 WCT f*Wu ' winning the solidification process, the contents of 121 N d thd be recirculated for at.least three volume changes to t Lwe qymplete mixing and homogenelty.
2.5.2 After recirculation, a sample of the 121 WCT is to be drawn and analyzed for isotepic content, pH and % boric acid.
i If pH is greater than 5.0, no adjustments need be inade. If the pH l
Is less than 5.0, it must be increased to between 5.5 and 7.0 with the addition of lime. Adjustments to pH,11 required, should be l
made to the liquid in 121 WCT As an alternate, pH adjustments may be made in the Atcor Metering Tank.
(
2.5.3 After sampling and pH adjustment, if required, the waste liquid is l
transferred to the Atcor Waste Metering Tank for solidification.
[
Filled drums are stored in the Atcor Drum Storage Alsles until i
solidification can be verified.
2.5.4 Flowrates Normal flowrates with operating tolerance together with the j.
discharge volume are as fo!!ows for typical evaporator bottoms:
Waste Liquid Flow 5.012% gpm f
Masonry Cement 0.8 10% cfm t
Product Discharge 1.0 cfm I
Other flowrates may be used if demonstrated to result in i
solidification.
L l
2.5.5 Cure Time Cure time is variable and depends upon waste pH, Boron concentration, and mix ratios. Normally, a two to three week cure time can be expected for complete solidification.
l
-l I
- u. a. - -
PRAIRIE ISLAND NUCLEAR GENERATlHO PLANT gumalcur'.l.OPNSj j
NoRTNERN STATES POWER COMPANY MAINTENANCE PROCEDURES l s
TITLE:
NUMBER:
l
.m
~*W PROCESS CONTROL PROGRAM D59 C_
57 FOR SOLIDIFICATION / DEWATERING e
4 0
isetion d OF RADIOACTIVE WASTE FROM LIQUID SYSTEMS Page io of 27 lj 2.5.0 Verification of Solidi _fication Representative barrels of each batch are to be inspected to verify I
solidification and the absence of free water. A drum may be l
considered solid when the cemented mass offers significant j
resistance to penetration by a hammer, or similar object. Absence of free water may be determined visually.
If solidification fails to take place, the process SHALL be suspended until the cause is determined and remedies are defined.
3 2.5.7 When solidification and absence of free water has been verified, the drums may be capped, deconned and removed from the Atcor Drum Storage Alsles. As an alternate to this sequence and in the interest of minimizing personnel exposure, the drums may be removed individually for capping and deconning 2.5.8 When the drums are removed from the Atcor Drum Storage Aisles, I
and after they are capped and deconned, the drum number is i
recorded together with the oatch number, contents and radiation j
level. The drums are then placed in storage to await shipping and j
- burial, i
I 2.6 Sample Solidification of Linuid Waste Concentrates
[
2.6.1 S.amplina Requirernents
}
i If it is not feasible to verify solidification and the absence of free water in the full-scale product, sample solidification SHALL be conducted for at least every tenth batch of liquid waste concentrates.
2.6.2 Prereauisites Before drawing a specimen from 121 WCT for sample solidificati.on, j
i the contents must be adequately mixed to achieve a representative mixture.
2.6.3 Sample Preparation j
A. Obtain a specimen from 121 WCT in the required volume. The f
volume required will be approximately 200 ml for each sample j
mixed plus 10 mi for a boric acid analysis.
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PRAlRIE ISLAND NUCLEAR GENERATING PLAN?
[
~NONTHERN STATES PCWER COMPANY MAINTENANCE PROCEDURES i
TITLE:
NUMi1ER:
I hi' PROCESS CONTROL PROGRAM D59 B#
FOR SOLIDIFICATION / DEWATERING I
REV :
4
'ssetton.
OF RADIOACTIVE WASTE FROM LIQUID SYSTEMS Page 11 of 27 l
B. Remove approximately 10 mi for boric acid analysis. Record l
% boric acid on Attachment 1 A.
T l
C. Place the remaining waste liquid in a beaker Maintain the temperature of the liquid to prevent precipitation of boron.
Record the volume of waste in the beaker on Attachnmnt 1 A.
D. Chock the mixturo pH and record this value on Attachment 1 A.
i If the pH is less than 5.0i slowly add lime to the liquid while cont!nuously stirring until a pH value of 5.5 to 7.0 is achieved.
Record the final pH and the weight of lime added on i A.
l f
E. Because of the relatively long cure time required, three samples should be mixed from the initial test specimen using different liquid / cement ratios. One sample will be mixed at the l
recommended full scale operating mix ratio. The o!her two samples should have more and less liquid than recommended i
for full scale mixing.
I Additional samples may be mixed from the initial test specimen at the discretion of the Rad Waste System Engineer using j
additional mix ratios or using different pH values. The following table defines the mix ratios which should be used:
h VOLUME OF VOLUME WT OF LIQUID / CEMENT WASTE LIQUID CEMENT CEMENT RATIO
{
(ml)
(ml)
(gm)
(Volume)
}
176 (NOTE #1) 218 0.88 200 l
166 200 218 0.83 (NOTE #2)
[
156 200 218 0.78 i
NOTE #1: Cement volume is theoretical and is listed for reference only. For accurate sample preparation, i
l cement must be measured by weight.
.{
f NOTE #2: 1.lquid/coment ratio (volume) recommended by Atcor L
t f or full scale mixing.
l l
i F. Place the required amount of cement in a beaker, Measure out l
the correct amount of waste liquid for the sample. Thoroughly l
mix the liquid and cement together to ensure homogeneity.
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J PRAIRIE ISLAND NUCLEAR GENERATING PLANT NORTHERN STATES POWER COMPAlW MAINTENANCE PROCEDURES
,p, TITLE:
NUMUER:
M'-
PROCESS CONTROL PROGRAM D59 v,
Em#
FOR SOLIDIFICATION / DEWATERING 4
secnon-OF RADIOACTIVE WASTE
~~
FROM LIQUID SYSTEMS Page 12 of 27 G. Cover the sample and store in a shielded area.
H. Observe the sample immediately after mixing and intermittently thereafter as appropriate till solidification is complete. Record the results in the space provided on Attachment 18.
gg*.
sorne water rnay appt er on the surface and be teabsorbed during solid 6fic.ation.
I.
Set the sample aside for future disposal.
J. Complete Attachment 1 A before proceeding with full scale solidification.
2.0.4 Sam,plo Acceptanco Criterl0 A. Visual inspection after mixing will confirm that the sample is homogeneous.
B. Visual inspection of the sample after curing will confirm that no free water exists on the surface of the sample.
C. Physical inspection of the sample after curing will confirm that the end product is a uniform, liquid free, free standing solid that resists penetration when probed with a pencil-sized probe.
D. If test samples from the initial specimen fail to produce a mixture which will solidify, additional specimens SHALL be drawn and mixed to determine the proper solidification parameters before full scale solidification can commence.
Additionally, if test samples from the initial specimen fail to produce a mixture which will solidify, sample solidification of specimens from successive batches SHALL be conducted until at least three samples from consecutive batches demonstrate solidification.
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PRAIRIE ISLAND IJUCLEAll OENERATINO PLANT NORTHERN STATES POWER COMPANY MAlHTENAhCE PROCEDURES TITLE:
NUMi1E R:
PROCESS CONTROL PROGRAM D39 D
FOR SOLIDIFICATION / DEWATERING l'
REV:
4 section OF RADIOACTIVE WASTE FROM LIQUID SYSTEMS Pago 13 of 27 3.0 h1ANUAL SOLIDIFICATION OF WASTE LIOUlDS 3.1
_P_u tpo s o i
o To establish parameters which provide reasonable assurance of complete
.,olidification of waste liquids when mixed manually.
3.2 Ap.plicability This section of the PCP is applicable to manual solidification of waste liciuids with masonry cement. Manual solidification may include the use of a portable, power-operated mixer.
Waste liquids which are normally solidified manually irclude:
I Laundry sludge f
2 Decon solutions, etc. not suitable for evaporation 3.3 Seguence of Operatior!
3.3.1 Place desired amount of liquid in 55 gal, drum (normally 1/2 to 2/3 full).
3.3.2 Commence mixing.
3.3.3 Add cement while continuing to mix at the rate of 1 ft3 (1 bag) per 6.25 gal. of liquid or until mixture begins to thicken. Continue to mix until all of the cement is incorporated and the mixture is smooth.
c Remove the mixer. (if applicable).
3.4 Cure Time Solidification can normally be expected within two to three days.
3.5 Verification of Solidification 3.5.1 Each drum of manually solidified waste liquid SHALL be inspected to venfy solidification and the absence of free water. A drum may be considered sohd when the cemented mass offers s jnificant resistance to penetration by a hammer or similar object. Absence of free water may be determined visually.
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, PRAlHIE ISLAND NuCl. EAR GENERATlHG PLANT NORTHERN STATES POWER COMPANY MAINTENANCE pro'UbDuRES TITLE:
NUMBER:
'M w
PROCESS CONTROL PROGRAM D59 5.# T S FOR SOLIDIFICATION / DEWATERING REV:
4 sesdon<
OF RADIOAOTIVE WASTE
^
FROM LIQUID SYSTEMS Page 14 of 27 If solidification fails to take place, the process SHALL be suspended until the cause is determined and remedies are ocfined.
3.5.2 When solidification and absence of free water has been verified, the drum may be capped and deconned. The drum number is recorded together with the batch number, contents, and radiation level. The drum is then placed in storage to await shipment and burial.
4.0 P_R_QCESSING OF CERTAIN WASTE LIQUIDS THRU SPENT BEAD RESjj 4.1 Purpose To establish an alternate method of processing certain waste liquids in lieu of solidification. This method utilizes spent bead resin to filter out suspended particulates allowing normal processing of the resultant liquid.
Disposal volumes and personnel exposures are thus reduced.
4.2 Applicability The following waste liquids may be processed using this procedure:
l l
4.2.1 Laundry sludge.
4.2.2 Decon solutions, etc. not suitable for evaporation.
)
l 4.2.3 Filter sludge.
f i
4.2.4 Mop bucket slurry.
l 4.2.5 Tank bottoms.
t 4.2.6 Sump bottoms.
l i
-NOTE:
EvaporW Concentrates may not be processed using l
this procidure.
[
I The above list is not to be considered complete. Items may be i
added or deleted upon evalustion of the Rad Waste System Engineer.
i
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g p. - n e s.... v e..,q PRAJRIE ISLAND NUCLEAR GENERATING PLANT f
NORTHERN CTATES POWER COMPANY MAINTENANCE PROCEDURES TITLE:
NUMBER:
e*
PROCESS CONTROL PROGRAM b59 j
m p
Hi"^
FOR SOLID 1FICATION/ DEWATERING 5#
4 f
sketion -
OF RADIOACTIVE WASTE FROM LIQUID SYSTEMS Page is of 27 i
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4.3 Sequence of Ooorqtion j
4.3.1 Ensure there is a layer of bead resin in the liner to act as a filter f
(the type of liner is determined by the activity of the material to be l
disposed of).
j 4.3.2 Ensure adequate volume for the quantity of material to be processed.
i 4.3.3 Pump / pour liquid slurry into liner.
I I
4.3.4 Flush drum and/or container, pump and hoses to liner.
!t i
4.4 Dewate.rina Procedure i
Dewater as per Section 7.0 " Dewatering of Bead Resin" to ensure there is i
I no free standing water in either the resin or the sludge.
5.0 MANUAL SOLIDIFICATION OF WET TRASH BY SUBMERSION I
5.1 Purpose To establish parameters which provide reasonable assurance of complete f
solidification of liquid contained in wet trash.
[
[
5.2 ep_plcability i
This section of the PCP is applicable to solidi'ication of wet trash with masonry cement.
[
i Wet trash includes contaminated material such as mopheads, w9t rags, paper towels, etc.-
I 5.3 Seauence of Operation 5.3.1 Place desired amount of liquid in 55 gal drum (normally 1/2 to 2/3 full).
j i
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~~~~
NOTE:
Contaminated liquids may be used for this purpose.
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PRAIRIE ISLAND UUCLEAR GENERATING PLANT HoRTl4 (N STATI 5 M.WER Coh PANU MAINTENANCE PROCEDURES
- l 'N TLE:
NUPA DER:
M PHOCESS CONTROL PROGRAM DS9 L#
JOF, SOLIDIFICATION / DEWATERING REV:
4 i
OF RADIOACTIVE WASTE sectiotr FROM LIQUID SYSTEMS Pago to of 27 5.3.2 Commence mixing.
5.3.3 Add cement while continuing to mix at the rate of 1 cu. ft. (one bag) per 6.25 gal of liquid or until the mixture begins to thicken.
Continue to mix until all of the cement is incorporated and the mixture is smooth. Remove the mixer (if applicable).
5.3.4 Immerse items of wet trash into the remented mass using a stick or similar device. Attempt to put as many items of trash as possible into the barrel within tne limits of ALARA.
5.4 C_ure _TimJt Solidification can normally be expected within two to three days.
5.5 Verification of Solidification Each drum SHALL t'e inspected to verify solidification and the absence of free water. A drum may be considered solid when the cemented mass offers significant resistance to penetration by a hammer or similar object.
Absence of free water may be determined visually.
If solidification fails to take place, the process SHALL be suspended until the cause is identified and remedies are determined.
5.6 Dis. position When solidification and the absence of free water has been verified, the drum may be capped and deconned. Record the drum number together with the batch number, contents. and radiation level. The drum is then placed in storage to await shipment and burial.
6.0 PRO _ CESSING OF WET TRASH BY COMPACTION /CFfMENTATION.
G.1 Pigpo_jte To establish parameters which provide reasonable assurance that wet radioactive trash is packaged safely and with an absence of free water.
6.2 Applicability This section of the PCP is applicable to the compaction of wet trash using the trash compactor wnile concurrently absorbing any free water with masonry cement.
._~
- PRAIRI'E ISLAND NUCLEAR GENERATINO PLANT' C-QKl. OPNs/
doRTNERN STATES POWER COMPANY MAINTENANCE PROCwunES EhR..
W/
TITLE:
NUMBER:
n e
PROCESS CONTROL PROGRAM DS9 4 L Bar ;
FOR SOLIDIFICATION / DEWATERING
~$e. tion; OF RADIOACTIVE WASTE-4
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FROM LIQUID SYSTEMS pago 18 of 27
^
7.2 Applicability This section of the PCP is applicablo to all radioactively contaminated bead resin which is intended to be shipped dewatered (not solidified) for disposal.
7.3 Dewaterina Procedurg The dewatering procedure / aries with the supplier of the resin liner, with tho' type of liner, whether, steel liner or a high integrity container (HIC),
a;.d with the dewatering requirement of the burial site. Individual shipping a cedures unique to the particular container and burial site refer to the propriate dewatering procedure.
t general, however, the dewatering process normally consists of the following steps after the liner has been filled:
7.3cl Initial pumpdown with the diaphragm pump until suction is lost.
7.3.2 A waiting period (twenty hours, for example).
7.3.3 Final' dewatering consisting of one or more pumpdowns using a diaphragm pump or a vacuum pumping system.
7.4 Verification of Dewaterina Preceding shipment, connect and operate the dewatering pump as before.
If no water is present, the dewatering procecs is complete.
If water is found, pump until vacuum _is lost. Repeat the pump / wait cycle as required. When no more water can be removed, the dewatering
=
process is complete.
8.0 DEW'TERING OF POWDERED RESIN 8.1 Purpose To describe the process used to provide reasonable assurance that powdered resin is dewatered to rneet applicable burial site criteria.
- 8.2 A_pylicability This section'of the PCP is applicable to all radioactively contaminated powdered resin which is intended to be shipped for burial.
PRAIRIE ISLAND NUCLEAR GENERATING PLANT l
NORTHERN STATES Po?/ER COMPANY MAINTENANCE PPoCEDuRES j
- 5 TITLE:
HuMDER:
j hpf '
PROCESS CONTROL PROGRAM D59 t
9 p --.
FOR SOLIDIFICATION / DEWATERING
- ~ w
(( %
OF RADIOACTIVE WASTE REV:
4-(
FROM LIQUID SYSTEMS Page 17 or 27 Wet trash inclodes contaminued material such as mop heads, wet rags, i
paper towels, etc, i
6.3 Seauence of Operation I
t i
6.3.1 Place approximately 2" of masonry cement in the bottom of a 55 i
gal. drum.
l 6.3.2 Place a layer of wet trash items into the drum while integrating
}
cement into eact. item of trash. Add a small amount of cement to fill voids between items.
6.3.3 Compress the wet trash using the compactor. Add cement as r
required to incorporate any free water thus poduced.
6.3.4 Repeat the preceding two steps until the drum is filled.
G.4 Cure Time i
Absence of free water can normally be determined visually immediately following the final compaction cycle.
6.5 Verification of Absence of Free Water Eacn drum of processed wet trash' SHALL be inspected to verify the absence of free water. -if free water is detected, additional cement SHALL r
be added to solidify the free water.
t 6.6 Disposition
[
When the absence of free wat6. h; een verified, the drum SHALL be capped and deconned. The drum nuinber is recorded together with the batch number, contents, and radiation level. The drum is then placed in
}
storage to await shipment and burial.
i l
7.0 DEWATERING OF BEAD RESIN 7.1 Purpose i
To describe the process used to provide reasonable assurance that bead j
resin is dewatered to meet applicable burial site criteria.
l l
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t.
i PRAIRIE ISLAND NUCLEAR GENERATING PLANT HORTHERN STATES POWER COMPANY MAINTENANCE PROCEDURES VITLE:
NUMDER:
i M
PROCESS CONTROL PROGRAM DS9 W
FOR SOLIDIFICATION / DEWATERING RW:
4 OF RADIOACTIVE WASTE section.
FROM LIQUID SYSTEMS Pago 19 of 27 8,3 System Description Contaminated powdered resin originates in the Condensate Polishing System Filter Demineralizers of both units.
Spent resin is purged from the Filter Demineralizers to the Backwash Waste Receiving Tank where it awaits the dewatering / drying process.
The dewatering /dr/ ng process takes place in the Clamshell Backwash i
Waste Filter (" Clamshell").
There are two Clamshells to serve the needs of both units, each capable of being aligned to either unit. It is the function of the clamshells to filter the powdered resin out of the water-resin slurry that is pumpeo from the Backwash Waste Receiving Tank, thru the Clamshells. When a cake of resin develops in the Clamshell to a predetermined thickness, the filtering process automatically switches to a purge phase followed by a forced air drying phase. The duration of the air drying phase can be adjusted.
Experience, however, has demonstrated that a drying cycle of approximately 12 minutes produces a product sufficiently dry to meet burial site requirements yet not so dry as to create an air-borne contamination hazard.
When the air-dry cycle is completed, the resin is dumped from the Clamshell into a hopper from which it is conducted down an enclosed chu'e to a container below. If the resin is insufficiently dried it will not flow freely down the chute.
Disp _osal i
8.4 Powdered resin which has been processed..
a a Clamshell system does not normally receive further dewatering treatrnent. Powdered resin may, therefore, be shipped in a container not fitted with dewatering equipment such as a steel drum or box. Because processed powdered resin is sufficiently dry to flow freeiy, and because powdered resin is normally very low in specific activity, it may be used to fill interstitial space in shipments of non-compatible trash or to fill voids in other shipping containers where they occur.
1 PRAIRIE ISLAND NUCLEAR GENERATING PLANT NORTHERN STATES POWER COMPANY MAINTENANCE PROCEDURES I NUMDER:
TITLE:
PROCESS CONTROL PROGRAM DS9 D
FOR SOLIDIFICATION / DEWATERING REV :
4 OF RADIOACTIVE WASTE section FROM LIQUID SYSTEMS Page 20 of 27 9.0 PROCESSING / DEWATERING OF SPENT FILTER ELEMENTS 9.1 Purpose This section describes the method for processing spent filter elements and the process used to provide reasonable assurance that spent filter shipments are dewatered to meet applicable burial site criteria.
9.2 Applicability This section of the PCP is applicable to at! radioactively contaminated filter elements intended for shipment for buria' in the dewatered state (not solidified). Procedures specific to the appropriate type of container SHALL be employed.
9.3 Description of Fillinq Process 9.3.1 Verify that the container to be used is approved by the manufacturer for disposal of filter elements.
9.3.2 Ensure a dewatering element with an attached hose is installed in the container. The dewatering elements must be compatible with the dewatering pump.
9.3.3 Filter elements should be drained of excess water prior to placing in the container.
9.3.4 Ensure a layer of processed resin or similar material is on the bottom of the container. The resin acts as a sludge filter and prevents the dewatering element from clogging.
9.3.5 Place filter elements into the container while attempting to avoid bridging of filters and observing the pnnciples of ALARA.
9.3.6 Voids may be filled with prcessed resin or similar material.
9.4 Dewatering The dewatering process may vary with type and manufacture of container and with requirements of the burial site. Typically, however, the dewatering process consists of the following steps:
PRAIRIE ISLAND NUCLEAR GENERATING PLANT
"' * '[jj NORTHERN STATES POWER COMPANY MAINTENANCE PROCEDURES ay>e3 Vxa TITLE:
NuMEER:
D6* '
PROCESS CONTROL PROGRAM DS9 E#
FOR SOLIDIFICATION / DEWATERING REV:
4 4
is dflaid '
OF RADIOACTIVE WASTE Fl~ e ~ %
FROM LIQUID SYSTEMS Page 21 of 27 9.4.1 Allow wait period (typically 20 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />) for water if present to migrate to the bottom of the container.
9.4.2 Connect the dewatering pump to the dewatering element hose.
Conduct the pump discharge hose to a container to enable monitoring of discharge volume.
9.4.3 Start the dewatering pump, if no water is found, the container may be considered to be dewatered.
l If water is found, pump until vacuum is lost, stop the pump and begin another wait period.
Repeat the pump / wait cycle until no more water can be removed.
I 9.5 Verification of Dewaterinn Preceding shipment, connect and 'perate the dewatering pump as before.
process is complete.
If no water is present, the dewate if water is found, pump until vacuum is lost. Repeat the pump / wait cycle f
as required. When no more water can be removed, the dewatering process is complete.
10.0 IN-CONTAINER SOLIDIFICATION OF BEAD RESIN T
10.1 - Purpose High level bead resin is normally shipped in the bulk dewatered form in high integrity containers. However, high activity resin may be solidified to comply with the stability requirements of 10CFR Part 61.
The Westinghouse Hittman Nuclear Incorporated PCP for incontainer Solidification of Bead Resin is incorporated as part of the Prairie Island PCP. Topical reports submitted by SEG/Hittman provide test data that demonstrates waste form stability.
This document is proprietary and may not be reproduced, but may be considered an appendix to the Prairie Island Process Control Program for Solidification / Dewatering of Waste From Uquid Systems.
i Certain plant specific exceptions to the SEG/Hittman document are noted in the system description.
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PRAIRIE ISLAND NUCLEAR GENERATING PLANT NORTHERN STATES POWER COMPANY MAINTENANCE PROCEDURES
,s-Q TITLE:
HUMBER:
0 *; M"
. PROCESS CONTROL PROGRAM D59 E#
FOR SOLIDIFICATION / DEWATERING REV:
4 Nssch6'nt -
OF RADIOACTIVE WASTE
_ZM*-
FROM LIQUID SYSTEMS Page 22 of 27 10.2 Apolicability This section of the PCP is applicable to the in-container solidification of bead resin using the SEG/Hittman PCP and field procedures.
i 10.3 References 10.3.1 STD-R-05-007 Topical Report Cement Solidified Wastes to Meet the Stability Requirements of 10 CFR 61. Westinghouse Radiological Services Incorporated.
10.3.2 STD-P-05-004 Process Control Program for incontainer Solidification of Bead Resin. Westinghouse Hittman Nuclear incorporated.
10.3.3 Waste Form Technical Position, Revision 1. United States Nuclear Regulatory Commission.
10.4 System Description The resin disposal system for the purposes of this PCP consists of 121 Spent Resin Tank,122 Resin Pump, a portable dewatering pump and relat 3d piping, hoses, and valves, in addition are included those items furrthed by the resin disposal contractor including a shipping cask, snipping liner, solidification equipment and related controls and appurtenances.
i l
Resin is pumped in a water slurry from the 121 Spent Resin Tank to the i
shipping liner in the proper amount. The water is then pumped out to trie drains system, after which the solidification process will begin in accordance with the contractor's procedures.
Because of the high activity of the resin requiring solidification, sample solidification using non-radioactive resin is normal. References in the PCP to sampling the Spent Resin Tank for solidification test purposes, therefore, do not apply.
- NOTE::'
Because of the proprietary nature, the referenced procedures in section 10.3 con not be reproduced and are retained in the Radiation Protection Files for i
reference.
1
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UPUAIED P.l. OPNS PRAIRIE ISLAND NUCLEAR GENERATING PLANT MA TENANCE~PRoCE NoRTHERI. STATES POWER COMPANY TITLE:
NuM ER:
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- s
'N PROCESS CONTROL-PROGRAM DS9
~
Barc-FOR SOLIDIFICATION / DEWATERING REV:
4
'secuoris.
OF RADIOACTIVE WASTE
~ ~ ~
FROM LIQUID SYSTEMS Page 23 of 27 10.5 Samplina in the event solidification of waste is required to ensure Part 61 STABILITY requirements, verification and surveillance test specimens SHALL be taken as per SEG/Hittman PCP for incontainer Solidification of Bead Resin, STD-P-05-004. In addition, the preparation, examination and testing of the samples SHALL comply with Waste Form Technical Position, Revision 1.
ii.0 REPORTING REQUIREMENTS 11.1 Purpose This section of the PCP sets forth the reporting requirements are as they apply to this PCP to ensure that the reports are completed accurately and in a timely manner.
11.2 _A_py!icability This section of the PCP, in whole or part, applies to all sections of the PCP.
11.3 References 11.3.1 T.S. 6.7.A.4 Prairie Island Nuclear Plant Technical Specifications.
.11.3.2 Waste Form Technical Position, Revision 1.
United States Nuclear Regulatory Commission.
11.3.3 STD-P-05-004 Process Control Program for incontainer Solidification of Bead Resin.
11.4 PCP Revisions Whenever the PCP is revised or changed, a description of the changes SHALL be included in the Semi-Annual Radioactive Effluent Release Report (T.S. 6.7.A.4).
PRAIRIE ISLAND NUCLEAR GENERATING PLANT D
NORTHERN STATES POWER COMPANY MAINTENANCE PROCEDURES TITLE:
NUMBER:
PROCESS CONTROL PROGRAM D59 D
FOR SOLIDIFICATION / DEWATERING REV:
4
- section OF RADIOACTIVE WASTE FROM LIQUID SYSTEMS Page 24 of 27 11.5 Reports of Mishaps Waste form mishaps SHALL be reported to the NRC (Director of the Division of Low-Level Waste Management and Decommissioning) AND the designated State disposal site regulatory authority within 30 days of knowledge of the incident. Mishaps are defined as failure of misuse of stabilized waste forms or containers that provide stability (HIC's). Such mishaps include, but are not necessarily limited to, the following:
.s#
11.5.1 The failure of high integrity containers used to ensure structural stability.
11.5.2 The misuse of high integrity containers, as evidenced by excesive free liquid, or excessive void space within.the container.
11.5.3 Production of a solidified Class B or Class C waste form that exhibits any of the characteristics listed in the Waste Form Technical Position, Revision 1.
11.6 P_CP Specimen Summary Reports Whenever cement stabilization (as defined by 10 CFR 61) of low-level waste is necessary, PCP test specimens are required for verification and surveillance. Verification specimens are intended to provide assurance that the formulations used in the qualification testing program correspond to those actually used in the field. Surveillance specimens are intended to provide verification that the waste forms remain stable with time. A summary report SHA'.L be prepared annually and submitted to the NRC (Director, Division of Low-Level Waste Management and Decommis'sioning) documenting the results of tests performed on the cement-stabilized waste form surveillance specimens during the calendar year.
The annual report should be submitted within 90 days of the end of each calendar year.
12.0 ATTACHMENTS 12.1 Figure 1 on page 25 - Solid Radwaste Flow Diagram.
12.2 Attachment 1 A on page 26 - Sample Verification Form.
12.3 Attachment 18 on page 27 - Sample Verification Form.
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PRAIRIE ISLAND NUCLEAR GEt4ERATitiG PLAf4T PAAINTENANCE PROCEDURES NORTHERN STATES POWER COMPANY NUMDER:
TITLE:
DS9 PROCESS CONTROL PROGRAM D
FOR SOLIDIFICATION / DEWATERING REV:
4 OF RADIOACTIVE WASTE page 25 or 2T
^
- section FROM LIQUID SYSTEMS j
SOLID RADWASTE FLOW DIAGRAM Figure 1 2 CPM or 5 GPM EVAPORATOR f
J I
RECIRC LINE
- 121 WASTE CONCONTRATES TANK ATCOR CEMENT BIN l
-H f
ATCOR WASTE METERING TANK
\\
y CEMENT FEEDER WASTE LIQUID C "
FEEDER
]
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MIXER / FEEDER L
BARREL
f PRAIRIE ISLAND NUCLEAR GENERATING PLANT
' ~ ' '" W'W Dl NORTHERN STATES POWER COMPANY MAINTENANCE PROCEDURES
/- Q ; $ ;
TITLE:
NUMDER:
PROCESS CONTROL PROGRAM DS9 h * *M Ear FOR SOLIDIFICATION / DEWATERING 4
e :isset'onL
~
OF RADIOACTIVE WASTE i
FROM LIQUID SYSTEMS Page so of 27
. A - Sample Vorification Form RPS Date Time Waste Type PRETREATMENT P1 initial pH Initial Temp F
% Boric Acid P2 Specimen Volume ml
{
P3 Lime Added gm P4 Final pH j
k Lime Ratio -
P3 X 8.34 -
Ibs P2 gal I
SAMPLE PROPORTIONS l
Sample No.
S1 Sample Waste Liquid Voi ml f
S2 Sample Cement wt gm f
?
Liquid /Cem ant Ratio (voi)
_S1 X 1.089* -
I S2
- Density correction factor.
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l cut'.l OFNSl PRAIRIE ISLAND NUCLEAR GENERATING PLANT NORTHERN STATES POWER COMPANY MAINTENANCE PROCEDURES. l
. ; gig. 3 TITLE:
NUM BER:
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y PROCESS CONTROL PROGRAM D59 E#-
FOR SOLIDIFICATION / DEWATERING REV:
4 7 sect 16n; [
OF RADIOACTIVE WASTE i
.M FROM LIQUID SYSTEMS page 27 of 27 i 8 - Sample Verification Form 1
Sample No, j
Describe sample appearance, water amount, hardness, etc.
l
- CURE TIME"
. CONDITIONS NOTED' i
0 HRS I
i I
i l
l t
9 I
l Sample is/is not Solid Date RPS Signature i
c f
w-y
-