ML20029A801
| ML20029A801 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 12/31/1990 |
| From: | NORTHERN STATES POWER CO. |
| To: | |
| Shared Package | |
| ML20029A797 | List: |
| References | |
| NUDOCS 9103040277 | |
| Download: ML20029A801 (93) | |
Text
. - ~.
3 1990 Effluont Semiannual Report REV. 1 Pag ~e 1 of 9 Retention: Lifetime l
EFFLUENT SEMIANNUAL REPORT 31-DEC-89 TilHOUGII 30-JUN-90 SUPPLEMENTAL ~INFORMATION Facility:
Prairie Island Nuclear Generating Plant Licensee:
-Northern States Power Company l
License Numbers: DPR-42 & DPR-60 o
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 mrom to the total body 10.0 mrom 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 (500 mrem / year total body 43000 mrom/ year skin I-131, H-3, LLP 41500 mrem / year to any organ b.
The dose due to radioacti"e gaseous effluents shall be limited to:
noble gases 410 mrad / quarter gamma 420 mrad / quarter beta j
420 mrad / year gamma L
440 mrad / year beta I-131, 11 - 3, LLP (15 mrem / quarter to any organ 430 mrem / year to any organ 9103040277 910227 PDR ADOCK 03000282 R
PDR J
1990 EFFLUENT SEMIANNUAL REPORT REV. 1 PAGE 2 B.
Maximum Permisrible 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 125%
in gaseous releases:
Nuclide GeLi
-2.
Iodines in gaseous releases:
Total GeLi 125%
Nuclide GeLi 3.
Particulates in gaseous releases:
Total GeLi 25%
Nuclide GeLi 4.
Liquid effluents Total GeLi 125%
Nuclide GeLi
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'1990"ErfLUENTiSEMIANNUAL REPORT REV. 1-PAGE 3 21.0 BATCH-RELEASES (LIQUID)
QTR:.01 QTR: 02 I
.1.1 NUMBER OF BATCH RELEASES 7.70E+01:
3.40E+01 1.2-. TOTAL TIME-PERIOD (HRS) 1.25E+02 6.05E+01 1.3. MAXIMUM' TIME PERIOD---(HRS) 2.75E+00-7.47E+00 1.4 AVERAGE TIME PERIOD (HRS).
1.63E+00 1.78E+00 1
1.5-MINIMUM TIME. PERIOD (HRS) 1.33E+00 1.33E+00
'1. 6 AVERAGE MISSISSIPPI RIVER FLOW (CFS) 9.38E+03 2.23E+04
'2. 0 BATCH RELEASES (GASEOUS)
QTR:. 01 -
- QTR: 02-2.1 ; NUMBER OF BATCH RELEASES 3.6CE+01 1.50E+01 2.21' TOTAL TIME PERIOD (HRS) 2.67E+02 1.43E+02-2.3-' MAXIMUM' TIME PERIOD (HRS)'
2.35E+01-12.40E+01 2.4---AVERAGE. TIME PERIOD (HRS) 7.42E+00
. 9. 51 E+ 0 0-2.5 ~ MINIMUM TIME PERIOD (HRS)'
1.00E-02 7.00E,
i 13.0ii ABNORMAL: RELEASES (LIQUID)
' QTR: 01.
QTR: 02:
l 3.-l'
. NUMBER-OF RELEASEE LO.00E+00:
0.00E+00 E3.2' TOTAL _ ACTIVITY RELEASED-(CI) 0.00E+00 0.00E+00 3 ~. 3 f TOTALLTRITIUM RELEASED (CI) 0.00E+00 0.00E+00
=
-4.0-ABNORMAL RELEASES (GASEOUS)-
QTR: 101-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 i
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l 1990 EFFLUENT SEMIANNUAL REPORT REV. 1 PAGE 4 TABLE 1A GASEOUS EFFLUENTS - SUMMATION OF ALL RELEASES QTR: 01 QTR: 02 5.0 FISSION AND ACTIVATION GASES 5.1 TOTAL RELEASE (CI) 0.02E+01 3.91E-01 5.2 AVERAGE RELEASE RATE (UCI/SEC) 1.02E+01 4.97E-02 5.3 GAMMA DOSE (MRAD) 5.05E-02 3.25E-05 5.4 BETA DOSE (MRAD) 1.56E-01 3.17E-03 5.5 PERCENT OF GAMMA TECH SPEC (%)
5.05E-01 3.25E-04 5.6 PERCENT OF DETA TECH SPEC (%)
7.80E-01 1.59E-02 6.0 IODINES 6.1 TOTAL I-131 (CI) 1.42E-03 1.79E-06 6.2 AVERAGE RELEASE RATE (UCI/SEC) 1.81E-04 2.28E-07 7.0 PARTICULATES 7.1 TOTAL RELEASE (CI) 2.19E-05 3.25E-06 7.2 AVERAGE RELEASE RATE (UCI/SEC) 2.79E-06 4.13E-07 8.0 TRITIUM 8.1 TOTAL RELEASE (CI) 5.91E+01 3.74E+01 8.2 AVERAGE RELEASE RATE (UCI/SEC) 7.52E+00 4.76E+00 9.0 TOTAL IODINE, PARTICULATE AND TRITIUM (UCI/SEC) 7.52E+00 4.76E+00 10.0 DOSE (MREM) 2.01E-01 6.10E-02 11.0 PERCENT OF TECH' SPEC (%)
1.34E+00 4.07E-01 12.0 GROSS ALPHA (CI) 1.56E-07 2.48E-07
19k0EPPudENTSEM8ANNUALREPORT REV. 1 PAGE 5 TABLE 1C GASEOUS EFFLUENTS - GROUND LEVEL RELEASES 13.0 FISSION AND ACTIVATION GASES CONTINUOUS MODE BATCil MODE NUCLIDE UNITS QTR: 01 QTR: 02 QTR: 01 QTR: 02 AR-41 CI 5.01E-02 KR-85 CI 1.47E+00 3.86E-01 KR-85M CI 1.51E-02 3.02E-03 KR-88 CI 1.28E-04 XE-131M CI 4.25E-01 4.64E-01 3.92E-03 XE-133 CI 6.11E+01 1.54E+01 1.46E-03 XE-133M CI 5.62E-01 9.37E-02 XE-135 CI 6.45E-01 6.43E-02 TOTAL CI 6.27E+01 0.00E+00 1.75E+01 3.91E-01 14.0 IODINES CONTINUOUS MODE BATCil MODE NUCLIDE UNITS QTR: 01 QTR: 02 QTR: 01 QTR: 02 I-131 CI 1.42E-03 1.79E-06 3.91E-06 I-133 CI 4.48E-05 4.54E-06 2.30E-06 TOTAL CI 1.46E-03 6.33E-06 6.21E-06 0.00E+00
i o
1990 ErrLUENT SEMIANNUAL REPORT REV. 1 PAGE 6 TABLE 2A GASEOUS EFFLUENTS - SUMMATION OF ALL RELEASES 15.0 PARTICULATES CONTINUOUS MODE BATCil MODE NUCLIDE UNITS OTR: 01 QTR: 02 QTR: 01 QTR: 02 CD-109 CI 2.64E-06 Co-60 CI 3.87E-06 CS-134 CI 2.34E-06 1.18E-06 1.38E-06 CS-137 CI 8.92E-06 1.38E-06 1.57E-06 SB-125 CI 1.19E-06 SR-89 CI 6.92E-07 SR-90 CI TOTAL CI 1.89E-05 3.25E-06 2.95E-06 0.00E+00
o-19,9,0iEFFLUENT SEMIANNUAL ~ REPORT-REV. 1J PAG E - 7..
TABLE-s2A1
- LIQUID-EFFLUENTS.- SUMMATION OF ALL RELEASES.
i
'I QTR: _01 QTR: 02_
l L16' 0 VOLUME OF WASTE PRIOR TO DILUTION 1(LITERS) 5.62E+07 6.47E+07 17.0 _VOLUMELOF DILUTION WATER (LITERS)
.1.09E+11 7.95E+10
+
t 18.0 FISSION AND ACTIVATION PRODUCTS 18.1; TOTAL RELEASELW/O H-3, RADGAS, ALPHA (CI) 8.39E-03 9.12E-03 i
I8.2 ' AVERAGE-DILUTED. CONCENTRATION (UCI/ML) 7-.69E-11' 1.15E r i19.0 TRITIUM
.*** SEE ATTACHED NOTE FOR ADDITIONAL TRITIUM DOSE CALCULATIONS.
E19.1 TOTAL' RELEASEL(CI):
1.69E+02-8.63E+01:
J 19.2 : AVERAGE DILUTED CONCF.NTRATION-(UCI/ML)~
.1~.55E-06 1.08E-06' 20.0 1 DISSOLVED ~AND ENTRAINED GASES 2 0.= 1. _ TOTAL'. RELEASE (CI) 1.17E-02 3.32E-03
-:20.24.
o, AVERAGE DILUTED CONCENTRATION-(UCI/ML) 1.~07E-10 4.18E-11) 121. 0 I - GROSS, ALPHAi( CI-)
0.00E+0'0-4.40E-05 22.0fLTOTALhTRITIUMk-FISSION AND. ACTIVATION i
- PRODUCTS 1(UCI/ML)'
1.55E-06 1.08E.
?23s011 TOTAL, BODY: DOSE-(MREM)-
4.33E-04--
2.64E-04 J24.0: LCRITICAL ORGAN-a q
.24.1 DOSE (MREM) 4.335-04 2.64E-04 i
124;2 "ORGANJ TTL LODY-
-TTL BODY:
Q J25.0. PERCENT-OF TOTAL BODY TECH SPECJLIMIT (%)
1.44E-02
- 8.80E-03
- 26.0 PERCENT OP CRITICAL ORGANETECH SPEC LIMIT-(%)
1.44E-02 8.80E-03
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19Y0 EFFLUENT SEMIANNUAL REPORT REV. 1 PAGE 8 TABLE 2A LIQUID EFFLUENTS - SUMMATION OF ALL RELEASES 27.0 INDIVIDUAL LIQUID EFFLUENT CONTINUOUS MODE BATCH MODE NUCLIDE UNITS QTR: 01 QTR: 02 QTR: 01 QTR: 02 AG-110M CI 1.76E-03 2.15E-03 BE-7 C-1.11E-05 2.00E-04 CO-58 CI 2.00E-03 9.42E-04 CO-60 CI 7.63E-04 1.08E-03 CS-134 CI 3.10E-06 CS-137 CI 8.90E-06 1.83E-05 CR-51 CI 9.73E-05 PE-55 CI 1.99E-03 3.46E-03 FE-59 CI 3.10E-05 I-131 CI 2.75E-04 MN-54 CI 9.64E-06 2.02E-05 NB-97 CI 3.75E-05 9.35E-05 SB-122 CI 4.10E-06 SD-124 CI 2.22E-04 SB-125 CI 1.54E-04 8.19E-04 SC-47 CI 5.34E-05 3.25E-06 SN-113 CI 3.59E-05 1.06E-04 SR-89 CI SR-90 CI SR-92 CI 2.38E-06
-TC-99M CI 2.32E-06 ZR-97 CI 6.07E-05 TOTAL CI 0.00E00 0.00E+00 8.39E-03 9.12E-03 i
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19'90 EFFLUENT SEM1 ANNUAL REPORT REV. 1 PAGE 9 TABLE 2A LIQUID EFFLUENTS - SUMMATION OF ALL RELEASES 38.0 DISSOLVED AND ENTRAINED GASES CONTINUOUS MODE BATCH MODE NUCLIDE UNITS QTR: 01 QTR: 02 QTR: 01 QTR: 02 XE-131M CI 4.50E-04 XE-133 CI 1.12E-02 2.99E-03 XE-133M CI 2.65E-05 1.81E-05 XE-135 CI 3.65E-05 3.16E-04 TOTAL CI 1.17E-02 3.32E-03 L
l
a:
i ATTACHMENT TO THE 1990 SEMIANNUAL EFFLUENT REPORT REV. 1 Liquid Pathway Dose Calculation Quarters 1 and 2, 1990 Summary The total liquid doses to the critical receptor for the first and second quarters of 19ss are 0.0399 and 0.0278 mrem, respectively.
l These doses are reported in this attachment to the semiannual report.
Background-In October, 1989 a well water sample of a residence close to PI showed low levels of tritium.
Initial dose calculation results showed estimated liquid pathway doses a hundred times higher than those reported in past semiannual reports.
These results indicated that a new dose pathway should be considered.
Investigation into the cause of the tritium contamination led to the assumption that the flow of water is from the discharge canal downhill toward the Vermilion-river.
The resident's well draws from the groundwater between the canal and the Vermilion river.
.0DCM Considerations The following' calculation is independent of the ODCM.
At this time no change will be made to the ODCM.
Efforts are currently under way to Edetermine if the tritium levels in the well can be reduced by physical modifications-to the discharge structures and/or adjustments to the
= discharge methods.
Results of these efforts should be available in 1990 and :the problem will be eliminated or appropriate ODCM changes will be made.
Dose Calculation Assumptions
~
For the purpose of dose calculation, the dose-maximizing assumption was.made that.the receptor's concentration of tritium in body-water Land organic molecules is equal to the average concentration in the discharge canal.
The dose conversion factor for this assumption is-taken from page 9-3 of NUREG/CR-3332.
Its value is.102 mrem / year per
,uci/llter of triti'um in the. body.~
When adjusted for the units
. reported;.for-the discharge. canal concentration, the dose. conversion factor is 2.55E4 mrem / quarter per uCi/ml.
h Discussion l
.The elevated tritium levels were found at the residence which is the l
critical receptor (0.6 miles to the SSE of the site) for Prairie E
Island.
Therefore, the dose calculated for waterborne tritium is L
added to the critical receptor's fish pathway dose, This l
overestimates the dose because the tritium dose f rom eating fish is
-accounted for twice.
i 1
i.
Page 3 It should be noted that the total airborne dose' (99% plus of which is due.to tritium) is greater than the total wate borne tritium dose.
- Even though_the. newly identified pathway delivers additional tritium to the critical. receptor.it is a lower concentration than that already in~the body due to airborne exposure.
. Dose calculation Average
- Dose, Diluted Whole Fish Total Quarter Conversion X
Tritium Body
+ Pathway a Liquid Factor Concentration Dose Dose Dose (mrem / quarter (uCi/ml)
(mrem)
(mrem)
(mrem) per uci/ml) 1-2.55E4 1.55E-6 3.95E-2 4.33E-4 3.99E-2 2
2.55E4 1.08E-6 2.75E-2 2.64E-4 2.78E-2
. Dose-Report
. LIQUID EFFLUENTS - SUMMATION OF WATERBORNE TRITIUM AND FISH CATHWAYS QTR:
01 OTR:
02
. TOTAL. BODY DOUE (MREM) 3.99E-02 2.78E-02 CRITICAL ORGAN DOSE (MREM) 3.99E-02
-2.78E-02 ORGAN TTL BODY TTL BODY PERCENT OF TOTAL' BODY TECH SPEC LIMIT (%)
1.33E+00 9.27E-01
' PERCENT OF CRITICAL' ORGAN TECH SPEC LIMIT-(%)
1.33E+00 9.27E-01 I
n-
1990 Effluent Semiannual Report REV. O Page 1 of 9
-Retention: Lifetime EFFLUENT SEMIANNUAL REPORT 01-JUL-90 TilROUGil 29-DEC-90 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 odividual from radioactive materials-in liquid effluents teleased 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 tota 2 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 1-131, H-3, LLP 41500 mrem / year to any organ b.
The doce due to radioactive gaseous effluents shall be limited to:
noble gases 410 mrad / quarter gamma
$20 mrad / quarter beta 420 mrad / year gamma (40 mrad / year beta I-131, H-3, LLP (15 mrem / quarter to any organ 430 mrem / year to any organ l
e 1990 ErrLUENT SEMIANNUAL REPORT nEv. O PAGE 2 B.
Maximum Permissibio_ Concentration 1.
Fission and activation
's in gaseous release.
10 CrR 20, Appendix b 2, Column 1 2.
Iodine and particulates with halflives greater than a days in gaseous telcases:
10 CFR 20, Appendix D, Table 2, Column 1 3
Liquid effluents for radionuclides other than dissolved or entrained gases:
10 CrR 20, Appendix D, Table 2, Column 2 4.
L(quid effluent dissolved and entrained gases:
2.0E-04 uti/ml Total Activity c.
Average Energy Not applicable to Prairie Island regulatory limits.
D.
Measurements and_ar.proximations of total activity 1.
rission and activation gases Total ocLi 125%
in gaseous releases:
Nu;11de GeLi 2.
Iodines in gaseous releases:
Total GeLi 125%
Nuclide GeLi 3.
Particulates in gaseout. releases:
Total GeLi 125%
Nuclide GeLi 4.
Liquid effluents Total GeLi 125%
Nuclide GeLi E.
Manual Revisions 1
1.
Offsite Dose Calculations Manual latest Revision number:
11 Revision date
- 05-OCT-09 l
2.
Process Control Program Manual latest Revision number 3
Revision date
- 31-MAY-90 l
l
.--- -_ _=
i i
e 1990 EFFLUENT SEMIAMMUAL REPORT REV. O PAGE 3 1
1.0 BA'tCH RELEASES (LIQUID)
QTR: 03 QTR: 04
.1 NUMBER OF BATCH RELEASES 4.60E+01 2.50E+01 1.2 TOTAL TIME PERIOD (HRS) 6.83E+01 3.00E+01 1.3 MAXIMUM TIME PERIOD (HRS) 1.92E+00 1.92E+00 1.4 AVERAGE TIME PERIOD (HRS) 1.40E+00 1.52E+00 1.5 MINIMUM TIME PERIOD (HRS) 1.05E+00 1.25E+00 J.6 AVERAGL MISSISSIPPI RIVER FLOW (CFS) 1.52E+04 9.09E+03 2.0 BATCH RELEASES (GASEOUS)
QTR 03 QTR: 04 2.1 NUMBER OF BATCH RELEASES 6.00E+00 1.00E+01 2.2 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 RLLEASES (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.C ' 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 l'
1:
4 e
1990 ErrLUENT SEMIANNUhy REPORT REV. O PAGE 4 TABLE 1A GASEOUS EFFLUENTS - SUMMATION OF ALL RELEASES QTR 03 QTR: 04 5.0 FISSION AND ACTIVATION GASES 5.1 TOTAL RELEASE (CI) 1.35E+00 7.42E-01 5.2 AVERAGE RELEASE RATE (UCI/SEC) 1.72E-01 9.44E-02 5.3 GAMMA DOSE (MRAD) 7.73E-03 3.72E-03 5.4 DETA DOSE (MRAD) 6.0EE-03 4.27E-03 5.S
?ERCENT OF GAMMA TECH SPEC (%)
7.73E-02 3.72E-02 5.6 SERCENT 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 R3 LEASE (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 (%)-
1.0dE-01 2.58E-01 13.0 GROSS ALPHA (CI) 2.38E-07 0.00E+00 l
L
s 1990 EFFLUENT SEMIANNUAL REPORT REV. O PAGE 5 TABLE 1C GASEOUS ErrLUENTS - GROUND LEVEL RELEASES 13.0 FISSION AND ACTIVATION GASES CONTINUOUS MODE BATCil MODE NUCLIDE UNITS OTR: 03 QTR: 04 QTR 00 QTR: 04 AR-41 CI 7.64E-01 3.50E-01 3.55E-03 KR-85 CI 3.70E-01 3.40E-01 KR-85M CI KR-07 CI KR-08 CI 4.84E-02 XE-133 CI 1.95E-01 1.62E-02 XE-133M C1 2.76E-04 XE-135 CI XE-135M CI XE-130 CI TOTAL CI 9.59E-01 3.50E-01 3.94E-01 3.92E-01 14.0 IODINES CONTINUOUS MODE BATCil MODE NUCLIDE UNITS OTR: 03 QTR: 04 QTR: 03 QTR: 04 I-131 CI 4.56E-07 I-132 CI I-133 CI I-134 CI I-135 CI TOTAL CI 0.00E+00 4.56E-07 0.00E+00 0.00E+00 t
l 4
l
- - - ~
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1930 EFFLUENT SEMIANNUAL REPORT REV. O PAGE 6 TABLE 1C GASEOUS EfrLUENTS - GROUND LEVEL RELEASES 15,0 P ARTI CUI.ATES CONTINUOUS MODE BATCil MODE NUCLIDE UNITS QTR 03 QTR: 04 QTR: 03 QTR 04 SR-89 CI SR-90 CI CL-109 CI 3.91E-00 2.96E-05 C0-58 CI 4.83E-07 1.90E-06 1.71E-05 CO-60 CI 5.82E-07 2.53E-07 3.05E-07 0.00E+00 1.74E-05 TOTAL CI 1.10E-06 3.18E-05 l
1990 EFPLUENT SEM8 ANNUAL REPORT REV. O PAGE 7 TABLE 3A LIQUID EFFLUENTS - SUMMATION OF ALL RELEASES QTR 03 OTR: 04 16.0 VOLUME OF WASTE PRIOR TO DILUTION (LITERS) 2.14E+07 3.09E+07
)7.0 VOLUME OF DILUTION WATER (LITERS) 2.45E+11 1.58E+11 18.0 FISSION AND ACTIVATION PRODUCTS 18.1 TOTAL RELEASE W/O H-3, RADGAS, ALPl!A (CI) 2.07E-01 2.92E-02 18.2-AVERAGE DILUTED CONCENTRATION (UCI/ML) 8.45E-10 1<85E-10 19.0 TRITIUM 19.1 TOTAL RELEASE (CI) 7.12E+01 7.10E+01 19.2 AVERAGE DILUTF,D CONCENTRATION (UCI/ML) 2.91E-07 4.49E-07 30.0 DISSOLVED AND ENTRAINED GASES 20.1 1?TAL RELEASE (CI) 4.18E-03 1.29E-03 20.2 AVERAGE DILUTED CONCENTRATION (UCI/ML) 1.71E-11 8.16E-12 21.0 GROSS ALPHA (CI) 1.40E-05 0.00E+00 23.0 TOTAL TRITIUM, FISSION AND ACTIVATION PRODUCTS (UCI/ML) 2.92E-07 4.49E-07 23.0 TOTAL BODY DOSE (MREM) 1.64E-03 3.26E-04 24.0 CRITICAL ORGAN 24.1 DOSE (MREM) 1.64E-03 3.26E-04 24.2 ORGAN TTL BODY TTL BODY 25.0 PERCENT OF TOTAL BODY TECH SFEC LIMIT (%)
5.47E-02 1.09E-02 26.0 PERCENT OF CRITICAL ORGAN TECH SPEC LIMIT (t) 2.20E-02 1.09E-01 y
1990 ErrLUENT SEMIANNUAL REPORT REV. O PAGE 8 TABLE 2A LIQUID ErrLUENTS - SUMMATION Or ALL RELEASES 27.0 INDTVIDUAL LIQUID ErrLUENT CONTINUOUS MODE BATCH MODE NUCLIDE UNITS QTR: 03 QTR 04 QTR 03 QTR: 04 AG-110M CI 1.38E-02 9.71E-03 BE-7 CI 3.45E-04 5.55E-05 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 C1 2.76E-06 1.88E-04 CS-136 CI 2.77E-06 CR-51 CI 1.49E-03 1.35E-03
~
C0-57 CI 2.20E-05 8.85E-05 CE-144 CI 4.65E-05 FE-06 CI 1.03E-02 2.27E-03 FE-59 CI 2.375-03 2.63E-03 LA-140 CI 2.23E-05 1.56E-06 LA-142 CI 3.60E-06 MN-54 CI 2.02E-04 1.45E-04 ND-97 CI 5.90E-06 ND-147 CI 7.12E-06 SB-122 CI 7.21E-04 SB-124 CI 3.90E-03 2.71E-03 SB-125 CI
- 1. 71 E- 0 3 2.435-03 l
SB-126 CI 1.44E-05 SC-47 CI 2.03E-04 2.33E-04 t
SN-113 CI 2.90E-04 8.62E-04 l
CONTINUED l
l u
1990 EFFLUENT SEMIANNUAL REPORT REV. O PAGE 9 TABLE 2A LIQUID EFFLUENTS - SUMMATION OF ALL RELEASES INDIVIDUAL LIQUID EFPLUENT (CONTINUED)
CONTINUOUS MODE DATCH MODE NUCLIDE UNITS OTR: 03 QTR: 04 QTR: 03 QTR: 04 SR-92 CI 5.82E-06 TC-99M CI S.00E-06 2.00E-06 RH-105 CI 4.78E-06 ZN-65 CI 3.34E-05 ZR-97 CI 2.95E-06 SR-89 CI 1.31E-03 SR-90 CI TOTAL CI 1.
10-03 0.00E+00 7.56E-02 2.93E-02 20.0 DISSOLVED AND ENTRAINED GASES CONTINUOUS MODE DATCH MODE NUCLIDE UNITS OTR: 03 QTR: 04 QTR: 03 QTR: 04 AR-41 CI 2.50E-05 l
KR-85 CI l
KR-85M CI KR-87 CI KR-88 CI l
XE-133 CI 4.03E-03 1.27E-03 XE-133M CI 2.88E-05 1.57E-05 XE.35 CI 9.82E-05 4.27E-06 l
l XE-l'"M CI XE-138 CI TOTAL Ci 0.00E+00 0.00E+00 4.18E-03 1.29E-03 l
i o'
PINGP 753, Rev. 2 Page 1 of 5 Retention:
Lifetime O
7-1-90 to 12-31-90 PRAIRIE ISLAND NUCLEAR GENERATING PLANT Period NORTilERN STATES POWER License No. DPR-42 i
SOLID RADIOACTIVE WASTE DISPOSAL SEMI-ANNUAL REPORT Table I:
Solid Waste and Irradiated Fuel Shipments 1.
Solid Waste Total Volumes and Total Curie Quantities:
l' Container l
Disposal Type of Waste Units Totals l
Volumes i-(List) i 3
988.9 133,8 A. Resins-l ft C1 282.690 178.0 1
183.3 3
i 11. Dry-Compacted ft l
Ci l
l l
l l
3
'C.
,Non-Compacted l
ft 1
83.0 l
96 l
l l
l l
l Ci l
0.232 l
l.
l l
l l
l' l
l 3
I D. Filter Media l
ft l
l
_l l
l 1
_Ci l
l I
l l
1:
-l i
I l
l 3
l'S. Spent Fuel l
ft l
l
-l I_
l
__ ll l
l Ci l
l l
l
(
l l
I I
I l'
l l
1
_l l
l IBM
~.
B 1
+
PINGP 753, Rev. 2 Page 2 O
PRAIRIE ISLAND NUCLEAR GENERATING PLANT Period: 7-1-90 to 12-31-90 NORTHERN STATES POWER License No. DMW SOLID RADIOACTIVE VASTE DISPOSAL SEMI-ANNUAL REPORT Table I:
Solid Waste and Irradiated Fuel Shipments (Continued) 2.
Principal Radionuclide Composition by Type of Waste:
Percent TYPE (From Page 1)
Nuclide Abundance A
CO-60 38.0
- N1-63 21.9
~
CO-38 13.8
c FE-55 46.5 Co-60 29.5 C-i-11.3 N1-63' 11.1 H-3 1.6 O
- = Inferred - Not Measured on Site IBM
PINGP 753, Rev. 2 Page 3 O
PRAIRIE ISLAND NUCLEAR GENERATING PLANT Period: 7-1-90 to 12-31-90 NORTHERN STATES POWER License No. DPR-42 SOLID RADIOACTIVE WASTE DISPOSAL SEMI-ANNUAL REPORT Table 1:
Solid Waste and Irradiated Fuel Shipments (Continued) 2.
Principal Radionuclide Composition by Type of Waste (Continuation):
Percent TYPE (From Page 1)
Nuclide Abundance O
O
- = Inferred - Not Measured on Site IBM
PINGP 753, Rev. 2 Page 4 O
PRAIRIE ISLAND NUCLEAR GENERATING PLANT Period: 7-1-90 t 12-31-90 NORTHERN STATES POWER License No. DPR-42 SOLID RADIOACTIVE VASTE DISPOSAL SEMI-ANNUAL REPORT Table 1:
Solid Waste and Irradiated Fuel Shipments (Continued) 3.
Solid Waste Disposition:
Number of Shipments Mode Destination 2
TRUCK IIARNWEl.L 2
TRUCK OAK RIDGl:
4 TRUCK RICllLAND 4.
Irradiated Fuel Shipments:
Number of Shipments Mode Destination 0
O IBM
a o
r
.PINGP 753, Rev. 2 Page 5 O
V PRAIRIE ISLAND NUCLEAR GENERATING PLANT Period: 7-1-90 tp 12-31-90 NORTHERN STATES POWER License No. DPR-42 SOLID RADIOACTIVE WASTE DISPOSAL SEMI-ANNUAL REPORT Table I:
Solid Waste and Irradiated Fuel Shipments (Continued) 5.
Shipping Container and SolidiTication Method:
Disposal No.
Volume Activity Type of Container Solidif.
(Pt3)
(C1)._
. Waste Code Code 90-22 44.8 0.108 C
L N/A
~~
90-25 135.3 196.777 A
A N/A 87-39 38.2 0.123 C-L N/A 90-33.
135.8 77.044 A
A N/A
.. 90-3 4._
.178.0 1.100 A
L N/A
.90-35
,_123 d_
2.626 A
L N/A 90-36 llfLH_
1.347 A
L N/A
_90-37
_183.3 3.796 A
L N/A TOTALS 8 1071.9 282.921 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
' O B = Dry Compacted l
C = Non-Compacted D = Filter Media S = Spent Fuel IBM
o NORTHERN DTATES POWER COMPANY ERh1RILJMhED_EEREAR_9EEURATIRO PLhMT QFF-BITE RAplAT19.H_D9E_hEERQAMI;KT_IQ.B l
J.allMAry_1-December 3 b_1990 An assessment of radiation dose due to release from the Prairic Island Nuclear Generating Plant during 1990 was performed in accordance with the Technical Specifications. Computed doses were well below the 40 CFR Part 190 Standards and 10 CFR Part 50 Appendix I Guidelines.
Of f-site dose calculation formulas and meteorological data from the Off-site Dose Calculation Manual were used in making this assessment. Source terms were obtained from the two Effluent and Waste Disposal Semiannual Reports prepared for NRC review during the year of 1990.
Off-site _. Doses from Gaseous Relings Computed doses due to gaseous releases are reported in Table 1.
Critical receptor location and pathways for organ doses are reported in Tobic 2.
Doses, both whole body and organ, are a small percentage of Appendix I Guidelines.
Off-s1.te Dosea fr_qm_LLquid Religgg Computed doses due to liquid releases are reported in Table 1.
Receptor information is reported in Table 2.
Both whole body dose and organ dose are a small percentage of Appendix I Guidelines.
Doses to Jndividuals Due to Activities Inside the site,_ Boundary occasionally sportsmen enter the Prairic Island site for recreational activities. These individuals are not expected to spend more than a few hours per year within the site boundary.
Commercial and recreational river traf fic exists through this area.
For purposes of estimating the dose due tc recreational and river water transportation activities within tha site boundary, it is assumed. hat the limiting dose within the site boundary would be roccived by an individual who spends a total of neven days per year on the river just off shore from the main plant buildings (ESE at 0.2 miles). Whole cody and inhalation organ doses were calculated i
for this location and occupancy time. These doses were reported in Table 1.
t l
l l
l l
- - - =.. - --_ - -.. _ - - -. - -. _ - -.. -. -. -
p l
p_gses to Host ExDosed MeRL1;tstr of - the. 0eagraLL_hlblig_ f rom._ Rup.ts_t lut)Sases aDi_Other UratnigmJgelCyc1e Ro_11rg_n Thoro are no other uranium fuel facilities in the vicinity of the Prairie Island site. The only other artificial source of exposure to the general public in addition to the plant ef fluent releases is from direct radiation of the reactors. This direct radiation from pressurized water reactors has boon shown to be negligible. An array of TLD monitoring stations around the perimeter of the cite boundary has consistently indicated that plant operation in the past years has no effect on ambient gamma radiation.
Therefore, the most exposed member of the general public will not receive an annual radiation dose from reactor effluent release, and all other fuel cycle activities in excess of the sum of the liquid and gaseous whole body and organ doses reported in Tablo 1 for the site boundary and critical receptor, respectively. These dosos are well below 40 CFR Part 190 standards of 25 mrem to the whole body, 75 mrom to the thyroid, and 25 mrom to any other orgari.
l
~ -.
_ _.. =. _ _ _.
3 p
6 i
TABLE 1 OFF-DITE RADIATION DOSE ASSE08 MENT - PRAIRIE ISLAND PERIODI J1diRRY_1.. thr.pygh DECMp3R 31,__1990 10 CFR Part 50 Appendix I Guidelines por 2-units sito per year QAmeous Releases Maximum Site Boundary Gamma Air Doso (mrad) 0.0434 20 Maximum Site Boundary Beta Air Doso (mrad) 0.10 40 Maximum Off-site Doso to Any Organ (mrom)*
j Total 0.351 30 Offshore Location (mrem)
Whole Body 3.4 E-04 10 Organ 4.34E-04 30 Liggi_d Releases Maximum Off-site Dose Whole Body (mrom) 0.0886 6
Maximum Off-site Dose Organ, Total 0.0886 20
I I
e TABLE 2 OFF-SITE RADIATION DODE ASSESDHENT - PRAIRIE ISLAND BUPPLEMENTAL INFORMATION J.Al{ VAR'l 1 tihtoygh.lF&ERER,lL_,1E1Q PERIODt o
gaseous Relaci s Maximum Sito,,.:dary Doso Location (from Building vente)
Sector WNW Distanco (miles) 0.36 Offshcro Location Within Site Doundary Sector ESE Distanco (milos) 0.2 Maximum Off-site Dono Location Sector SSE Distance (miles) 0.6 Pathways
- Plume, Ground, Inhalation, Vogotablos Age Group child Organ Thyroid LigMjd RolqAppa Maximum Off-site Dose Location Downstream Pathways Fish, well water Ago Group child Organ Whole Body
ATTACHMENT To THE 1990 SEMIANNUAL ErrLUENT REPORT Liquid Pathway Dose Calculation Quarters 3 and 4, 1990 Summary The total liquid doses to the critical receptor for the third and fourth quarters of 1990 are 0.0091 and 0.0118 mrem, respectively.
These doses are reported in this attachment to the semiannual report.
Background
In October, 1989 a well water sample of a residence close to PI showed low levels of tritium.
Initial dose calculation results showed estimated liquid pathway doses a hundred times higher than those reported in past semiannual reports.
These results indicated that a new dose p
...way should be considered.
Investigation into the cause of the tritium contamination led to the assumption that the flow of water is from the discharge canal downhill toward the Vermilion river.
The resident's well draws from the groundwater between the canal and the Vermilion river.
ODCM Considerations The following calculation is independent of the ODCM.
At this time no change will be made to the ODCM, Efforts are currently under way to reduce the tritium levels in the well by extending the current radioactive dischargo point to the point where the discharge canal empties into the river.
Results of these efforts should be available in 1991 and the problem will be eliminated or appropriate ODCM changes will be made.
Dose Calculation Assumptions For the purpose of dose calculation, the dose-maximizing assumption was made that the receptor's concentration of tritium in body water and organic molecules is equal to the average concentration in the discharge canal.
The dose conversion factor for this assumption is taken from page 9-3 of NUREG/CR-3332.
Its value is 102 mrem / year per uCi/ liter of tritium in the body.
When adjusted for the units reported for the discharge canal concentration, the dose conversion factor is 2.55E4 mrem / quarter per uCi/ml.
Discussion The elevated tritium levels were found at the residence which is the critical receptor (0.6 miles to the SSE of the site) for Prairie Island.
Therefore, the dose calculated for waterborne tritium is added to the critical receptor's fish pathway dose.
This overestimates the dose because the tritium dose from eating fish is accounted for twice.
q e
Page 2 It should be noted that the total airborne dose (991 plus of which is due to tritium) is greater than the total waterborne tritium dose.
Even though the newly identified pathway delivers additional tritium to the critical receptor it is a lower concentration than that already in the body due to airborne exposure.
Dose calculation Average Dose Diluted whole Fish Total Quarter Conversion X
Tritium Body
+ Pathway -
Liquid
=
Factor Concentration Dose Dose Dose (mrem / quarter (uCi/ml)
(mrem)
(mrem)
(mrem) per uci/ml) 3 2.55E4 2.91E-7 7.42E-3 1.64E-3 9.06C-3 4
2.55C4 4.49C-7 1.14C-2 3.26E-4 1.18E-2 Dose Report LIQUID CrrLUENTS - SUMMATION OF WATERBORNE TRITIUM AND FISH PATilWAYS OTR 03 OTR:
04 TOTAL BODY DOSE (MREM) 9.06E-03 1.10E-02 CRITICAL ORGAN DOSE (MREM) 9.06E-03 1.18E-02 ORGAN TTL BODY TTL BODY PERCENT OF TOTAL BODY TECil SPEC LIMIT (%)
3.02E-01 3.93E-01 PERCENT OF CRITICAL ORGAN TECH SPEC LIMIT (t) 3.02E-01 3.93E-01 1
' PRAIRIE l$LAt4D fiUCLEAR OENERATtHQ PLANT UPDATED P.l. OPNS
. NORTHERN STATES POWER COIAPANY DAAINTENAf4CE PROCEDURES i
TITLE:
NUlADER:
M PROCESS CONTROL PROGRAM D59 w
FOR SOLIDIFICATION / DEWATERING REV:
3 OF RADIOACTIVE WASTE section FROM LIQUID SYSTEMS Page 1 of 24 0.C. REVIEW DATE:
REVIEWED DY:
D AT E:
)
O APPROVED DY:
DATE:
i UPDATED P.l. OPN>AAlHTEMANCE PROCEDURES
' PRAIRIE ISLAt4D MucLEAR oEl4ERATING PLAldT
.idoRTHERid STATES POWER CotAPAt4Y 71TLE:
HulADER:
M PROCESS CONTROL PROGRAM DS9 W
FOR SOLIDIFICATION / DEWATERING REV:
3 section OF RADIOACTIVE WASTE FROM LIQUID SYSTEMS Pago 2 of 24 TABLE OF CONTENTS Section Title Pago
- 1. 0 G E N E R AL....................................
4 1.1 Purpose................................
4 1.2 Scope..............
4 1.3 Defi niti on s...............................
4 1.4 Applicable Tech. Spec.,
5 2.0 SOLIDIFICATION OF LIOUID WASTE CONCEt'TRATES..........
5 2.1 Purpose................................
5 2.2 Applic abili ty..............................
5 2.3 R ef e r en c e s..............................
5 2.4 System Des cription..........................
5 2.5 Sequence of Operation....
8 2.6 Sample Solidification of Liquid Waste Concentrates.......
9 3.0 MANUAL SOLIDIFICATION OF WASTE LIOUIDS,
12 3.1 P u r po s e......................,.........
12 3.2 Applic abili ty..............................
12 3.3 Sequence of Operation.....,,.................
12 3.4 C u r e Ti m e...............................
12 3.5 Verification of Solidification,
12 4.0 PROuESSING OF CERTAIN WASTE LIOUIDS THRU SPENT BEAD R E SIN.............
13 4.1 Pu r p o s e................................
13 4.2 Applic abili ty..............................
13 4.3 Sequence of Operation........................
14 4.4 Dewatering Procedure........................
14 5.0 MANUAL SOLIDIFICATION OF WET TRASH BY SUBMERSION.....
14 5.1 P u r po s e.............................,
14 5.2 Applic abili ty..............................
14 5.3 Sequence of Operation.............
14 5.4 Cure Time.......
15 5.5 Verification of Solidification...............
15 5.6 Disocsition.......
15
' PRAIRIE isLAt4D IJUcLEAR o[l4ERATlf40 PLAT 4T UPDATED P.l. OPNS
' lioRTHERid STATES POWER CotAPAt4Y
_h All4TEMAlicE PROCEDURES TITLE 14UtA DER:
PROCESS CONTROL PROGRAM DS9 D
FOR SOLIDIFICATION / DEWATERING REV :
3 sobtion OF RADIOACTIVE WASTE FROM LIQUID SYSTEMS Page 3 of 24 Section Title Pago 6.0 PROCESSING OF WET TRASH BY COMPACTION / CEMENTATION...
15 6.1 Pu r p o s o................................
16 6.2 Applicability..............................
15 6.3 Sequence nf Ope ration........................
16 6.4 Curo Time......
16 6.5 Verification of Absence of Fr 9 Water...............
16 6.6 Di s po sition...............................
16 7.0 DEWATERING OF BEAD RESIN....
16 7.1 P u r po s e................................
16 7.2 Applic ability..............,...............
17 7.3 Dewaterir.g Procedure...,
17 7.4 Verification of Dewatering.
17 8.0 DEWATERING OF POWDERED RESIN....................
17 8.1 Purpose......................
17 8.2 Applicability.....
17 8.3 System Description..........................
18 8.4 Di s po s al................................
18 9.0 DEWATERING OF SPENT FILTER ELEMENTS...............
19 9.1 Purpose................................
19 9.2 Applic ability..........................
19 9.3 Description of Filling Process....................
19 9.4 Dewatering...........
20 95 Verification of Dowatering......................
20 Appendix A Process Control Program for in-Container Solidification of B ead R e s i n.............................
21 FIGURE 1: Solid Radwaste Flow Diagram...........
22 ATTACHMENT 1 A: Sample Verification Form...............
23 ATTACHMENT 18: Sample Verification Form...............
24
UPDATED P.I. OPN{
e PRAIRIE isLANo HuCLEAR GENERATN40 PLANT-
. NORTHERN STATES POWER COMPANY MAINTENANCE PROCEDURES j
TITLE:
hum DER:
4 PROCESS CONTROL PROGRAM D59 i-FOR SOLIDIFICATION / DEWATERING
' i sestion -
OF RADIOACTIVE WASTE REV:
3 FROM LIQUID SYSTEMS Page 4 of 24 1.0 DEEERAL E
1.1 Purpose l
The purpose of this Process Control Jagram (PCP)is to detail the means by which the dewatering and/or solidification of radioactive waste from liquid systems can be assured:
1.2 Scoog Tnis PCP includes the following processes:
-1.2.1 Solidification of liquid waste concentrates.
1.2.2 Manual solidification of waste liquids.
1.2.3 Manual solidificailon of wet trash by' submersion.
1.2.4 Processing of wet trash by compaction / cementation.
1.2.5 Dewatering of bead resin.
1.2.6 Dewatering of powered resin, 1.2.7-Dewatering of spent filter elements.
1.2.8 Appendix A - PCP for in-container Solidification of Bead Resin.
1.3 pofinitions 1.3.1 Batch 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, 1.3.2 Solidification The conversion of wet radioactive wastes into a form that meets shipping and burial ground requirements.
OPRAIRIE lsLAND NUCLEAR oENERATINo PLANT
' UPDATED P.I. Umd
. NSRTHERN STATES POWER COMPANY MAINTENANCE PROCEDURES
.o y TliLEt NuMDER:
9M PROCESS CONTROL PROGRAM D59 S
W FOR SOLIDIFICATION / DEWATERING REV:
3 s$ction OF RADIOACTIVE WASTE i
FROM LIQUID SYSTEMS Page 5 of 24 4
1.3.3 powalerinn The process of removing water from a substance to meet specific limits.
1.4 Ap31icable Toch, Sp3_c2 T.S.3.9.C 2.0 SALlDIFlQMION OF L,_ QUID. WASTE CONCENTRATES 2.1 Purpose To establish the process parameters which provide reasonable assurance of complete solidification of liquid waste concentrates.
2.2 Apyllcability This section of the PCP is applicable to solidification of hquid waste concentrates using the Atcor Solidification System and related equipment.
2.3 Referenq.pf.
2.3.1 C21.2.1 Solid Radioactive Waste Operating Procedure 2.3.2 C21.2.2 Trash Compactor Operation Operating Procedure 2.4 SyJJ.em Descripj!gn -
2.4.1 General Description The solidification system for liquid waste concentrates includes 121 Waste Concentrates Tank (WCT), the Atcor Solidification System and related pumps, piping and equipment. Concentrates are accumulated from the 5 GPM ADT evaporator or 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 i
are held in the Atcor drum storage aisles until solidifier; ion can be confirmed. The drums are then capped, deconned, and surveyed prior to storage for subsequent shipment and disposal. A flow diagram is shown on Figure 1.
,' PRAlRIE ISLAND NUCLEAR GENERATING PLANT UPDATED P.I. OPNS NORTHERN STATES POWER COMPANY MAINTENANCE PROCEDuREG TITLE:
HuMBER:
M PROCESS CONTROL PROGRAM D59 Em#g FOR SOLIDIFICATION / DEWATERING se'etiM OF RADIOACTIVE WASTE REv:
3 1 ~,
FROM LIQUID SYSTEMS Page 6 of 24 2.4.2 Detailed Descripilon A.121 Waste Concentrates Tank 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 protection and is equipped with a high level alarm to prevent over-filling. A direct reading float-type level gauge provides level indication from outside of the shielded vault.
121 WCT pump and discharge piping are arranged for recirculation and mixing of the tank contents or for pumping the contents to lhe Atcor System for solidification. A sample valve is provided near the pump discharge.
B. Atcor Solidification System The Atcor Solidification System is designed to mix waste liquid concentrates with cement, to convey the blended mass into 55 gal. drums, and to store the filled drums in a shielded area for curing. The system consists of the following principle components:
- 1. Waste Meterina Tank The waste metering tank is' a tank of approximately 700 gal, process capacity. The tank is equipped with heaters to maintain contents in solution and is equipped with an agitator to ensure homogeneity of liquid. The tank is equipped with a positive displacement discharge pump having discharge rate variable up to 10 GPM. The pump L
discharges directly into the mixer eeder.
I
- _. _ _. = - _ _ _ _ _ _. _. - -. _. _ _ _ = _ _ _ _ _.. _ _ _ - _ _ = _ _ _.. _ _ _ _ _...
UPDATED P.I. OPUS O PRAIRIE ISLAND NUCLEAR GENERATING PLANT
, NORTHERN STATES POWER COMPANY MAINTFNANCE PHoCEDuREs j
TITLE:
HuMBER:
4 M.
' Em# '
FOR SOLIDIFICATION / DEWATERING REV:
3
~
Jsecad OF RADIOACTIVE WASTE FROM LIQUID SYSTEMS Page 7 of 24
- 2. Cement Bin The cement bin is a bin of approx.100 cu. ft. process capacity. The bin is equipped with a vibrating lower cone to preclude bridging of coment and to ensure uniform flow of material having a consistent bulk density. The coment bin 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.
- 3. Mixer /Feedet 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 discharge chute directs the blended mass into the shipping container by gravity flow.
- 4. Controls Controls for the solidification system are contained or, o panel shielded from the waste materials.
Gauges indicating feed rate of cemen't and waste liquid are located on the control panel.
Rates of cement feed and liquid feed are adjustable from the control panel during processing.
A closed circuit TV camera and monitor are provided for viewing drum movements from the control panel.
- 5. Cement Type Cement normally used is type "M" masonry cement having 50% lime and 50% portland cement, conforming to ASTM-C-91-64 and ASTM-C-270-61T.
l
UPDATED P.I. OPNS PRAIRIE ISLAND NUCLEAR GENERATm0 PLANT l
' NORTHERN STATES POWER CoMPINY MAINTENANCE PROCEDURES TITLE:
ItuMDER:
PROCESS CONTROL PROGRAM DS9 D
FOR SOLIDIFICATION / DEWATERING REV:
3 section OF RADIOACTIVE WASTE l
FROM LIQUID SYSTEMS Page 8 of 24 2,5 SAnience of Operati.on 2.5.1 Recirculption of 121WCT Before beginning the soUdification process, the contents of 121 WCT should be recirculated for at least three volume changes to assura complete mixing and homogeneity.
2.5.2 After recirculation, a sample of the 121 WCT is to be drawn and analyz9d for isotopte content, pH and % boric acid.
If pH is greater than 5.0, no adjustments need be made. If the pH is less than 5.0, it must be increased to twtween 5.5 and 7.0 w th the addition of time. Adjustmonts to pH, ir required, should be mado to the liquid in 121 WCT As an alternato, pH adjustments may be made in the Atcor Metering Tank.
2.5.3 After sampling and pH adjustment, if requitod, the waste liquid is transferred to the Atcor Waste Metering Tank for solidification.
Filled drums are stored in the Atcut Drum Storage Aisles until solidification can be verified.
2.5.4 flowraten Normal flowrates with operating tolerance together with the discharge volume are as follows for typical evaporator bottoms:
Waste Liquid Flow 5.01 - 2% gpm Masonry Cement 0.81 - 10% cfm Prvduct Discharge 1.0 cfm Other flowrates may be used if demonstrated to result in solidification.
2.5.5 Curo 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.
I im m=da
mm _
0
- PRAIRIE lsLANo NUCLEAR oENERATINo PLANT UPDATED P.I. OPNS e
e NORTHERN STATES POWER CotAPANY M AINTEN ANci' PROCEDURES TITLE:
HuMDER:
PROCESS CONTROL PROGRAM D59 D
FOR SOLIDIFICATION / DEWATERING
<a REV :
3
' isectiisn-OF RADIOACTIVE WASTE P
FROM LlOUID SYSTEMS Page 9 of 24 2.5.6 Verification oLSolidification Representative barrels of each batch are to be 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 determined and remedies are defined.
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 Alsles, and after they are capped and deconned, the drum number is recorded together with the batch nurnber, contents and radiation level. The drums are then placed in storage to awalt shipping and burial.
2.6 Sample Solidification of LiquicLWpste__Concentratos 2.0,1 Samplinci Requirements If it is not feasible to verify sohdification and the abstnce of free water in the full-scale product, sample solidification EHALL be conducted for at least every tenth batch of liquid waste concentrates.
2.0.2 Proroqul@cn o
Before drawing a specimen from 121 WCT for sample solidification, the cont 6..t3 must be adequately mixed to achieve a representative l
mixture.
2.G.3 Sample Preparation A. Obtain a specimen from 121 WCT in the required volume. The volume required will be approximately 200 mi for each sample mixed plus 10 mi for a boric acid analysis.
' PRAIRIE lsLAND NUCLEAR GENERATING PLANT UPDATED P.I. OPNS c
. NORTHERN STATES POWER COMPANY taAINTENANcE PROCEDURES TITLE:
HuMDER:
PROCESS CONTROL PROGRAM DS9 D
FOR SOLIDIFICATION / DEWATERING REV:
3
> section >
OF RADIOACTIVE WASTE FROM LIQUID SYSTEMS Page 10 of 24 b
B. Remove approximately 10 ml for boric acid ans., sis. Record
% boric acid on Attachment 1 A.
C, Plate 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 Attachment 1 A.
b D. Check the mixture pH and record this value on Athchment 1 A.
If the pH is less than 5.0, slow'y add lime to the liquid while continuo' sly stirring until a pH value of 5.5 to 7.0 is achieved.
Record the final pH and the weight of lime added on A.
E. Because of the relatively long sure time required, three samples should be mixed from the initial test specimen using different liquid / cement ratios. One sample will be mixed at the recommended full scale operating mix ratio. The other two samples should have more and less liquid than recommended for full scale mixing.
Additional se p!es may be mixed frorn the initial test specimen at the discretion of the Rad Waste System Engincor using additional mix ratios or using different pH vt The following table defines the mN ratios which should be v VOLUME OF VOLUME WT OF LIQUID / CEMENT WASTE LIQUID CEMENT CEMENT RATIO (ml)
(ml)
(gm)
(Volume) 176 (NOTE #1) 218
.88 200 166 200 218 0.83 (NOTE -#2) 156 200 218 0.78 NOTE n 1: Cement volume is theoretical and is listed for reference only. For accurate sample preparation, cement must be measured by weight.
NOTE #2: Liquid /coment ratlo (volume) recommended by Atcor for full scale mixing.
F.
Place the required amount of cement in a beaker. Measure out the correct amount of waste liquid for the sample. Thoroughly i
mix the liquid and cement together to ensure homogeneity.
. ~
~ -
' PRAIRIE ISLAND NUCLEAR GENERATlHG PLANT UPDATED P.I. OPNS'
. NORTHERN STATES POWER COMPANY MAINTENANCE PRocEDur.fE_
- .;4 TITLE
NuMDER:
- he PROCESS CONTROL-PROGRAM DS9 e
Ear m FOR SOLIDIFICATION / DEWATERING lsectik,
OF RADIOACTIVE WASTE REV:
3 FROM LIQUID SYSTEMS Page 11 of 24 G. Cover the sample and store in a shielded erea.
H. Observe the sample immediately after mixing nnd intermittently thereafter as appropriate till solidification is complete. Record the results in the space provided on Attachmont 18.
WON',:
s me water may app ar n the surface and be reabsorbed during solidification,
- l. Set the sample aside for future d;sposal J
Complete Attachment 1 A before proceeding with full scale solidification.
2.0.4 Samolo Acceptance Critoria A. Visual inspection after mixing will confirm tnat the sample is homogeneous.
B. Visual inspection of the sample affei curing will confirm that no free water exists on the surfac.' of the sample.
t C. Physical inspection of the samp'a after curing will confirm that the end product is a uniform, liquid free, froe standing solid that I-resists penetration when probed with a pencil-sized probe.
l-D. If test samples from the initial specimen fall to produce a mixture which will solidify, additional specimens SHALL be drawn and mixed to determine the proper solidification parameters before full scsle solidification can commence.
Additionally, if test samples from the initial specimen fail to L
produce a mixture which will solidify, sample solidification of specimens from successive batches SHALL be conducted until l
at least three samples from consecutive batches demonstrate L
solidification.
f PRAIRIE lsLAND NUCLEAR GENERATING PLANT UPDATED P.I. OPNS NORTHERN STATES POWER COMPANY MAINTENANc ycEDuREs TITLE:
NuMDEH:
m
'M PROCESS CONTROL PROGRAM DS9 5#
FOR SOLIDIFICATION / DEWATERING REV :
3 fssetionf OF RADIOACTIVE WASTE FROM LIQUID SYSTEMS Pago 12 of 24 3.0 MANUAL _ SOLIDIFICATION OF WASTE LIQUIDS 3.1 Purpose To establish parameters which provide reasonable msurance of complete solidification of waste liquids when mixed manually.
3.2 Ap_plicability This section of the PCP is applicable to manual solidification of waste liquids with masonry cement. Manual solidification may include the use of a portable, power-operated mixer.
Waste liquids which are normally solidified manually include:
I Laundry sludg(
- 2 Decon solutio. 8, etc. not suitable for evaporation 3.3 Secuence of Operation 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 tc thicken. Continue to mix until all of the cement is incorporMed and the mixture is l
smooth.
Remove the mixer. (if applicable).
l 3.4 Cure Time Solidification can normally be expected within two to three days.
3.5 Verification of Solidification 3.5.1 Nh drum of manually sollaified waste liquid SHALL be inspected
<erify solidification and the absence of free water. A drum may be considered solid when the cemented mass offers significaat resistance to penetration by a hammer or similar object. Absence of fcee water may be determined visuaUy.
- PRAlRIE ISLAND NUCLEAR GENERATING PLANT UPDATED P.I. OPNS
- HORVHERN STATES POWER COMPANY MAINTENANCE PROCEDURES
- 43 TITLE:
NUMDER:
sM., M PROCESS CONTROL PROGRAM D59 E#
FOR SOLIDIFICATION / DEWATERING
' 5ectient OF RADIOACTIVE WASTE REV:
3 FROM LIQUID SYSTEMS Pago 13 of 24 If solidification fails to take place, the process SHALL be suspended until the cause is determined and remedies are defined.
3.5.2 When solidification and absence of free wa;ar 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 PROCE_SSING OF CERTAIN WASTE LIQUIDS THRU SPENT BEAD RESIN 4.1 Purp_olo To establish an alternate method of. processing certa:n 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 Ap_plicability.
The following waste liquids may be processed using this procedure:
4.2.1 Laundry sludge.
4.2.2 Decon solutions, etc. not suitable for evaporation.
4.2.3 Filter sludge.
4.2.4 Mop bucket slurry.
l 4.2.5 Tank bottoms, l
l 4.2.6 Sump bottoms.
L TNOTE l Evaporator Concentrates may not be processed using this proceduro.
The above list is not to be considered complete. Items may be added or deleted upon evaluation of the Rad Waste System Engineer.
l l
PRAIRIE ISLAND NUCLEAR GENERATING PLANT UPDATED P.I. OPNS NORTHERN STATES POWER COMPANY MAINTENANCE PROCEDURES TITLE:
NuMDER:
>M PROCESS CONTROL PROGRAM DS9 5.#
FOR SOLIDIFICATION / DEWATERING REV:
3 isocilon :
OF RADIOACTIVE WASTE FROM LIQUID SYSTEMS Page 14 of 24 4.3 Sequence of Operation 4.3.1 Ensure there is a layer of bead resin in the liner to act as a filter (the type of liner is determined by the actit ity of the material to be disposed of).
4.3.2 Ensure adequate volume for the quantity of material to be processed.
4.3.3 Pump / pour liquid slurry into liner.
4.3.4 Flush drum and/or container, pump and hoses to liner.
4.4 Dewaterinn Procedure Dewater as per Section 7.0 " Dewatering of Bead Resin" to easure there is no free standing water in either the resin or the s...-je.
5.0 MANUAL SOLIDIFICATION OF WET TRASH BY SUBMERSION 5.1 P_ ump _s_e.
To establish parameters which provide reasonable assurance of comple+o solidification of liquid contained in wet trash.
5.2 App _ligability This section of the PCP is applicable to solidificatiori of wet trash with masonry cement.
Wet trash includes contaminated material such as mopheads, wet rags, paper towels, etc.
5.3 Sp_que.nce of Op_eration 5.3.1 Place desired amount of liquid in 55 gal drum (normally 1/2 to 2/3 full).
NOTE; ContaV%tod Hquids may be used for this purpose.
.____ _ _ __~
~ _ ___.______ _._
.._____m._
__a.______m__-.___m_.-_-___-_____-______m.-___m
_a.__
_. _ _. _ _.___.__.u_m___._._m._m______
_.________________m________
__.____.-_m
-_.__m_____m-_ _ - _ _ _ - - -
' PRAIRIE ISLAND NUCLEAR GENEnATING PLANT bPDATED P.I. OPNS
. NORTHERN STATES POWER COMPANY MA!NTENANCE PROCEDURES TITLE:
NuMDER:
PROCESS CONTROL PROGRAM DS9 D
FOR SOLIDIFICATION / DEWATERING PdV:
3 OF RADIOACTIVE WASTE
- section7 FROM LIQUID SYSTEMS Page 15 of 24 5.3.2 Commence mixing.
5.3.3 Add cement while continuing to mix ' '" " ate of 1 cu. ft. (one bag) per 6.25 gal of liquid or until the mixture oegnis 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 cemented mass using a stick or similar device. Attempt to put as many items of trash as possible into the barrel within the limits of ALARA.
5.4 Cure Time Solidification can normally be expected within two to three days.
5.5 Verification of Solidification Each drum SHALL be inspecwd 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 SMALL be suspended until the cause is identified and ren' dies are datermined.
5.6 Disposition 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 leve'. The drum is then placed in storage to await shipment and burial.
6.0 PROCESSING OF WET' TRASH BY COMPACTION / CEMENTATION 6.1 Purpose 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 vf wet trash using the trash compactor while cancurrently absorbing any free water with masonry cement.
.. _ ~
~
PRA!RIE ISLAND NUCLEAR GENERATING PLANT UPDATED P.I. OPNb
. NORTHERN STATES POWER COMPANY MAINTENANCE PROCEDURES W
TITLE:
NUMBER:
_ h, PROCESS CONTROL PROGRAM DS9 Eur-FOR SOLIDIFICATION /DEWATEPMG ls,ection[
OF RADIOACTIVE WASTE REV:
3
~
FROM LIQUID SYSTEMS Page 16 of 24 Wet trash includes conta.e Tied material such as rnop heads, wet rags, paper towels, etc.
6.3 Seouence of Operatiom 6.3.1 Place approximately 2" of masonry cement in the bottom of a 55 gal. drum.
6.3.2 Place a layer of wet trash items into the drum while integrating cement into each 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 required to incorporate any free water thus produced.
6.3.4 Repeat the preceding two steps until the drum is filled.
6.4 C.ure Timo Absence of free water can normally be determined visually immediately following the final compaction cycle.
6.5 Verification of Absence of Free Water Each drum of processed wet trash SHALL be inspected to verify the absence of free water, if free water is detected, additional cement SHALL be added to solidify the free water.
6.6 Disposition
- When the absence of free water has been verified, trie drum SHALL 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.
7.0 DEWATERING OF BEAD RESIN 7.1 Purpose To' describe the process used to provide reasonable assurance that' bead resin is dewatered to meet appli %Ie burial site criteria.
I
PRAIRIE ISLAND NUCLEAR GENERATING PLANT UPDATED P.I. OPNS
' NORTHERN STATES POWER.;oMPANY -
MAINTENANCE PROCEDURES
/9,+
TITLE:
HuMDER:
m,@
PROCESS CONTROL PROGRAM D59 mer FOR SOLIDIFICATION / DEWATERING REV:
3 d'?seationo,
OF RADIOACTIVE WASTE i
E.
FROM LIQUID SYSTEMS Pa0017 of 24 7.2 Applicability This section of the PCP is applicable to all radioactively contaminated bead resin which is intended to be shipped dewatered (not solidified) for disposal.
7.3 Dewaterinn Procedure The dewatering procedure varies with the supplier of the resin liner, with the type of liner, whether a steel liner or a high integrity container (HIC),
L and ~/th the dewatering requirement of the burial site. Individual shipping l
procedures unique to the particular container and burial site refer to the appropriate dewatering procedure.
In general, however, the dewatering process normally consists of the following steps after the liner has been filled:
7.3.1 -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 process is complete.
l-If water is found, pump until vacuum is lost. Rer, eat the pump / wait cycle as required.- When'no more water can be removed, the dewatering l
process is complete.
l 8.0 ' DEWATERING OF POWDERED RESIN 8.1 Purpose
.To describe the. cess used to provide reasonable assurance ihat powdered res..._ Jewatered to meet applicable bu isl site criteria.
8.2 A_pplicability This section of the PCP is applicable to all radioactively contaminated powdered resin which is intended to be shipped for bsrial,
.__m
-- j
' PRAIRIE ISLAND NUCLEAR GENERATING PLANT
' NORTHERN STATES POWER COMPANY MAINTENANCE PROCEDURES TITLE:
NUMBER:
id PROCESS CONTROL PROGRAM D59
<~
,n
?:,
FOR SOLIDIFICATION / DEWATERING M\\setlog OF RADIOACTIVE WASTE REV:
3 FROM LIQUID SYSTEMS Page 18 of 24 l
8.3 System Description
Contaminated powdered resin originates in the Ov'3nsate Polishing System Filter Demineralizers of both units.
Spent resin is purged from the Filter Cemineralizers to the Backwash Waste Receiving Tank where it awaits the dewatering / drying proc ^ss.
The dewatering / drying process takes place in the Clamshell Backwash Waste Filter (" Clamshell").
There are two Clamshells to serve tN 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 pumped from the Backwash Waste Receiving Tank, thw the Clamshells. When a cake of resin develc~ 'n the Clarshell ta a prer/aarmined thickness, the filtering process au.c
. aally swiu
- to a purge phase followed by a forced air drying phase. Tne duratica of the air drving phase can be adjusted.
Experience, hvwever, 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 contaminatior ".azard.
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 chute to a container below, if the resin is insufficiently dried it will not flow freely down the chute.
8.4 Disposal Powdered recin which has_ been processed thru the Clamshell system does not normally receive further dewatering treatment. Powdered resin may, therefore, be shippad in a container not fitted with dewatering equipment such as a steel drum or box. Because processed powdered esin is sufficiently dry to flow freely, and because powdered resin is rmally 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.
0
' PRAIRIE ISLAt4D NUCLEAR GEt4ERATit1G PLAT 4T UPDATED P.I. OPNS
. NORTHERN STATES POWER COMPANY MAINTENANCE PROCEDURES TITLE:
HUMBER:
M
- M PROCESS CONTROL PROGRAM DS9 4
FOR SOLIDIFICATION / DEWATERING W
REV:
3
/ section)
OF RADIOACTIVE WASTE 4
FROM LIQUID SYSTEMS Page 19 of 24 9.0 DEWATERING OF SPENT FIL'LER ELEMENTS 9.1 fLurpose To describe 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 all radioactively contaminated filter elements intended for shipment for burial in the dewatered state (not solidified). Normally a High Integrity Container (HIC) is used for this purpose. Procedures specific to the appropriate type of container SHALL i
be employed.
9.3 Description of Fillinct Process 9.3.1 Verify that the container to be used is approved by the manufacturer for disposal of filter elements.
9 3.2 Install dewatering element with attached hose in the container. The dewatering elements must be compatible with the dewatering pump (normally a vacuum pump). Conduct hose to.outside of container for later attachment to dewatering pump.
9.3.3 Allow filter elements to drain of excess water prior to placing in container.
9.3.4 Place a layer of processed powdered resin or similar material on the bottom of the container if required.
9.3.5 Place a layer of filter elements into the container while attempting to avoid bridging of filters and observing the principles of ALARA.
9.3.6 Fill voids with processed powdered resin or similar material if required.
' NOTE:.
Powdered resin may be used to fill volds between filter elements while observing principles of ALARA even if not required to be used an packing material by the container rnanuf acturer.
i l
' PRAIRIE lsLAtlD flucLEAR GEt4ERATitlG PLAtlT UPDATED P.I. OPNS I40RTHERt1 STATES POWER COMPAt4Y IAAlt4TEllAticE PROCEDURES 4
TITLE:
flutADER:
M PROCESS CONTROL PROGRAM DS9 w
FOR SOLIDIFICATION / DEWATERING REV:
3 secuon-OF RADIOACTIVE WASTE l'
FROM LIQUID SYSTEMS Page 20 of 24 9.3.7 Repeat the preceding two steps until the container is full.
9.4 DewaterLna The dewatering process may vary with type and manufacture of container and with requirsinents of the burial site. Typically, however, the dewatering process consists of the following steps:
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.
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.
9.5 Verification of Dewatering Preceding shipment, connect and operate the dewatering pump as before.
'f no water is present, the dowatering process is complete.
i 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.
' PRAIRIE lsLAND NUCL EAR oENERATmo PLANT UPDATED P.I. OPNS
. NORTHERN STATES POWER COMPANY MAINTENANCE PROCEDURES TITLE:
NuMDER:
PROCESS CONTROL PROGRAM DS9 D
FOR SOLIDIFICATION / DEWATERING REV:
3 section=
OF RADIOACTIVE WASTE FROM LIQUID SYSTEMS Page 21 of 24 Appendix A Process Control Pronram for in-Container Solidification of Bead Rosin 1.0 GENERAL Bead resin is normally shipped in the bulk dewatered form. However, high activity resin may be solidified if desired.
Following a brief system description, the Hittman Nuclear and Development Corp. PCP for in-Container Solidification of Bead Resi, is appended.
This document is proprietary and is reproduced in its entirety as an appendix to the Prairie Island Process Control Program for Solidification / Dewatering of Waste From Liquid Systems.
Certain plant specific exceptions to the Hittman document are noted in the system description.
2.0 SYSTEM DESCRIPTION The resin disposal system for the purposes of this PCP censists of 121 Spent Resin Tank,122 Spent Resin Pump, a portable dewatering pump and related piping, hoses, and valves. In addition are included those items furnished by the resin disposal contractor including a shipping cask, shipping liner, solidification equipment and related controls and appurtenances.
Resin is pumped in a water slurry from 121 Spent Resin Tank to the shipping liner in the proper amount. The water is then pumped out to the 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 nonradioactive resin is normal. References in the PCP to sampling the spent resin tank, therefore, do not apply.
NOTE.
Because of its proprietary nature, the Hittrnan
{
Nuclear and Development Corporation Process p
Control Program for in-Container Sohdification of i
Bead Rosin #STD-P-05-004 is retained in the Rad Protection Files for reference.
PRAIRIE ISLAND NUCLEAR GENERATING PLANT UPDATED P.l. OPNS
+.
NORTHERN STATES PCWER COMPANY MAINTEllANCE PROCEDURES TITLE:
HUMDER:
fyl PROCESS CONTROL PROGRAM DS9
~
w FOR SOLIDIFICATION / DEWATERING REV:
3-
^ isestionj '
OF RADIOACTIVE WASTE FROM LIQUID SYSTEMS Page 22 of 24 FIGURE 1 SOLID RADWASTE FLOW DIAGRAM 2 GPM or 5 GPM EVAPORATOR I
l
- 121 WASTE RECIRC LINE CONCENTRATES TANK
-CCOR CEMENT BIN
,.TCOR WASTE METERING TANK 8
0
\\
i WASTE LIQUID CEMENT FEEDER l
FEEDER V C l
l MIXER / FEEDER L
i BARREL l
L y
4 y --
vi-"'
_ _ _ ~
? PRAIRIE ISLAND NUCLEAR GENERATING P!. ANT UPDATED P.l. OPNS
.- NORTHERN STATES POWER COMPANY MAINTENANCE PROCEDURES p-. '.
TITLE:
NUMBER:
- M PROCESS CONTROL PROGRAM DS9 i
-W FOR SOLIDIFICATIOb8/ DEWATERING REV:
3 4section1 OF RADIOACT VE WASTE FROM-LlOl %) SYSTEMS Page 23 of 24 ATT ACHMENT 1 A SAMPLE VERIFICATION 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 Lime Ratio -
P3 X 8.34 =.
Ibs P2 gal SAMPLE PROPORTIONS L
Sample No.
S1 Sample Waste Liquid Voi ml S2 Sample Csmant wt gm l
Liquid / Cement Ratio (vol) = - S1 X 1.089* =
S2
- Density correction factor.
I l
,' PRAIRIE ISLAND 'lOCLEAR GENERATING PLANT l
NORTHERN STATES POWER COMPANY MAINTENANCE PROCEDURES (N-1
,7j! ;ijf[::
TITL E:
NUMBER:
s, [ M %' ' '
'W FOR SOLIDIFICATION / DEWATERING b f ' :(sktienl7[,
OF RADIOACTIVE WASTE 3
REV:
- 1.. -
FROM LIQUID SYSTEMS Page 24 of 24 '
A1TACHMENT 1B SAMPLE VERIFICATION FORM Sample No._
Describe sample appearance, water amount, hardness, etc.
3REkTIME :
CONDITIONS NOTED'L 4% '-
- /
0' HRS l
r L
I l
Samp;e is:i: not gog;g Date RPS S.ignature
-- -.. - + -,
n--
p 3 i
t.--
DESCRIPTION OF REVISION 11 ODCM MANUAL CHANGES E2 factive Date:
05-OCT-1989 The following is a list of changes made to the offsite Dose Calculation Manual in response to the EG&G Technical Evaluation Report of what was ODCM revision 9 in 1987 and submitted to Northern-States Power 5/18/89.
NSP's response was discussed and submitted to the NRC prior to the revision of the ODCM Rev. 11 in October.of 1969.
This is the formal notification of our ODCM change in accordance with T.S.
6.5.E.1.
Item 1:
In section 2.3.1, the dilution flow ADFk, was supposed to be the average flow during the reporting period (e.g.,
one. month, one quarter, or one year) and not the " actual dilution flow during the period of releace" as previously defined.
The definition was-correctly defi. led in revision 11.
Though incorrectly defined, this varioble has been used correctly in calculations.
stem 2:
In section 3.2.2, equation 3.2-1 and the del a t ton for X/Qv were corrected for typographical errors.
-Item 3':
Table 3.2.1 was corrected fm a typographical error.
Item 4:
Section 3.3.1 was corrected to ceference six plant vents points instead o.f nine vents, which was incorrect.
Item 5:
Section 3.3.1.2 was corrected for a. typographical error.
-Item 6:
Section 3.3 2.1 was corrected to reference-the appropriate tables dispersion tables.
The tables contained the correct values and were being used corcectly, but the ODCM reference were incorrect.
Item 7:-
Table 5.1-1 was revised to reflect changes in the Radiation Environmental Monitoring Program Eampling Locations and/or description of sample locations.
Changes were nade to the location / description of the following cample points.
-River Water Drinking Water Sediment-River Sediment-Shoreline Periphyton or Invertebrates Fish-Milk Particulate and Radiciodine Air Sampling
' Direct Radiation TLDs Item 8:
Figures 3.1-1, 5.1-2 and 5.1-3 were replaced with clearer and more visible maps.
Item 9:
Section C.1 of Appendix C were corrected to calculate the value of P1 using the " child" breathing rate (3700 m*3/yr) instead of the " infant" breathing rate (1400 m'3/yr).
-The actual calculation of Pi was changed to use the child i
breathing rate instead of the infants in revision 9 to the
.ODCM but the Appendix C description was not corrected.
l
o r
' Item 110 r In Tables 3.3-6 through 3.3-18, there were some inconsistt-ncies between the ODCM methodology and the manner in which the calculations of the R-value (dose factor) was actually performed.
The R-values according to the ODCM formulas exceed the ODCM table values by 18. percent for Cs-l'37 and up to 34 percent for I-131 for the cow milk, goat milk, meat and inhalation pathways.
No inconsistencies were found for the inhalation pathway R-values in Tables 3.3-17 and 3.3-18. It was noted that the ODCM table
-values could be rcproduced if a value of 2.59E+06 seconds was used for "te" and a value of 0.5 was used for "fp".
Explanations of the use of these variables and their definitions are given below:
The "te" variable This variable is defined in the ODCM as the " period of pasture grass and crop exposure during the growing season". ODCM tables C-1 and C-2 give two values for "te",
one for pasture and one for stored feed. ODCM formulas C.2-3 and C.2-5 each contain two occurrences of "te" with no distinction made for use of the assigned values. The first occurrence is in the part of the equation devoted to pasture and the second occurrence to stored feed. It should be obvious that the pasture value should be used in the first occurrence and stored feed value in the second occurrence. However, the ODCM table values were generated by substituting the pasture value in both variable occurrences.
Use of the pasture value in place of the stored feed value gives a non-conservative result, i.e.,
the calculated dose factor produces a smaller dose than would be generated using the stored feed value.
The "fp" variable This variable is defined in the ODCM as the " fraction of the year that the cow or acat is on pasture". A value of 0.667 was assigned to this varitble on page 119 of the ODCM, Revision 9.
However, a value of 0.5 is used in the NRC GASPAR code annual critical receptor determihation; and in the PINGP USAR. The L
value of 0.5 is also the value used to-generate the ODCM table values.
Oce of the 0.5 value for "fp" instead of 0.667 gives a non-conservative result, i.e.,
the calculated dose factor
!=
produces a smaller dose than would be generated using the l--
ODCM-assigned value.
L The ODCM was modified in-the following manner to address these n
j inconsistencies:
- a. The "te" variable was split into the variables "tep" and L
"tes" with the former used in the first part of the milk l
and meat equations related to animal intake of ra'ionuclides from pasture and the latter will be used in the sscond part of the equation related to intake from stored feed.
l The assigned values are 2.59E+06 nd 5.18E+06 seconds, respectively. This change will ensure the correct factor for exposure of pasture and stored feeds is used in the dose calculation.
i a
e e
b.
The "fp" varinole was assigned a value of 0.5 based upon a more realistic 6-month grazing ceriod.
Changing the values of these two variables produced R-values which are a maximum of 6 percent greater than the factors listed in revision 10 of the ODCt4.
Past dose calculations for the critical receptor have included neither the milk nor the meat pathway since these pathways do not apply at the critical receptor location. Therefore, the revised R-values have no effect on previously calculated doses.
The above described change effected changes to the equations and descriptions listed in Appendix C sections C.2.2, C.2.3, C.2.4 and C.2.5.
Changes were also made to Tables C-1, C-2 C-3 to list the values of the variables described above, i
e H4 Rev. 11 Page 5 RECORD OF REVISIONS Revision No.
Date Reason for Revision Original June 7, 1979 1
April 15, 1980 Incorporation of NRC Staff comments and correction of miscellaneous errors 2
August 6, 1982 Incorporation of NRC Staff comments 3
February 21, 1983 Change in milk sampling location 4
November 14, 1983 Change in milk sampling location and change in cooling tower blowdown 5
March 27, 1984 Change Table 3.2-1 6
February 14, 1986 Change in location to collect cultivated crops (leafy
/
green veg.) and removal of meat animals from the land use census.
7 July 31, 1986 Retype and format ODCM.
No change in content.
8 January 8, 1987 Addition of Discharge Canal Monitor Setpoint calculation.
9 June 29, 1987 Change inhilation dose factor to child and address change in land 4
use survey, 10 April 27, 1989 Change in method for calculating liquid effluent monitor setpoints.
Fix of various typing errors.
Change in location of two REMP sampling locations.
Deletion of one REMP sampling location.
11 October 5, 1989 Change in Tables 3.3-6 thru 3.3-16.
Appendix C equations corrected.
Section 5 figures replaced.
Sample point definitions I
corrected.
IBM
s s
H4 Rev. 13 Page 17 2.3.1 Determination of Liquid Effluent 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 (2.3-1)
F k
X ADFk where release rate of the batch or R
=
k continuous releases during the period, k, gpm average dilution flow during the ADF
=
k time period of release k, gpm.
The value of X is the site specific factor for the mixing effect of the PNGP discharge This value is 10 for PNGP while (1) structure.
operating in the closed cycle cooling mode.
The product of X and ADFu is limited to 1000 cfs (4.5 x 10 gpm).
Th8refore, since blowdown flow in closed cycle is 150 cfs, the denominator 5
of Equation 2.3-1 is always 4.5 x 10 in closed cycle.
In once through or helper mode, the value of X is reduced to 1.0.
2.3.2 Dose Calculations The dose contribution from the release of liquid effluents will be calculated monthly.
The dose
. contribution will.be calculated using the following equation:
where:
Dt
=
IIA t
C F
(2.3-2) fz k
gy k
k i.
where:
the dose commitment to the total DI
=
body or any organ t, from the liquid effluents for the period of release, mrem; 1
1BM
e H6 Rev. 11 Page 30 3.2.1 Noble Gases The dose rate at the site boundary resulting from noble gas effh.ents is limited by 10 CFR 20 to 500 mrem /yr t<
the total body and 3000 mrem /yr to the skin.
The setpoint determinations dis cussed in the previous section are based on the doca, 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 this manner will assure that the limits of 10 CFR 20 are met for noble gas releases.
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 Section 3.3.1.
3.2.2 Radiciodine, Radioactive Particulates, and other Radionuclides The dose rate'at the site boundary resulting from the release of radioiodines and particulates with half lives greater than 8 days is limited by 16 CFR 20 to 1500 mrem /yr to any organ.
Calculations showing compliance with this dose rate limit will be performed for batch releases prior to the release and weekly for all releases.
The calculations will be based on the results of sample analyses 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 days.
I P
(x/Q )
Q
< 1500 mrem /yr (3.2-1) g y
1y where J
P child critical organ dose parameter for 1
=
7 radionuclide i for t ae inhalation pathway, mrem /yr per pCi/m
. Table 3.2-1);
8 nnnual average re?ative concentration for (X/Q )
=
long-term release. at the critical location, y
(Appendis A, Table A-3);
3 sec/m IBM
e H4 Rev. 11 Page 32 TABLE 3.2-1 Values for Child p
l l
. mrem /yr l
3 l
ISOTOPE l
P pCi/m g7 l
I l
l H-3 l
1.12 E 3 l
l Cr-51 l
1.70 E 4 l
l Mn-54 l
1.58 E 6 l
l Fe-59 l
1.27 E 6 l
l Co-58 l
1.11 E 6 l
l Co-60 l
7.07 E 6 l
l Zn-65 l
9.95 E 5 l
l Rb-86 l
1.98 E 5 l
l Sr-89 l
.?.16 E 6 l
l Sr-90 l
1.01 E 8 l
l Y-91 1
2.63 E 6 l
l Zr-95 l
2.23 E 6 l
l Nb-95 l
6.14 E 5 l
l Ru-103 1
6.62 E 5 l
l Ru-106 l
1.43 E 7 l
l Ag-110m l
5.48 E 6 l
l Te-127r l
1.48 E 6 l
l Te-129m i
1.76 E 6 l
l Cs-134 l
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 l
1.74 E 6 l
l Cc-141 1
5.44 E 5 l
l Ce-144 l
1.20 E 7 l
l I-131 l
1.62 E 7 l
1 I
I J
IBM
e Q.
'e H4 Rev. 11-Page 34 where:
.j The air dose factor due to gamma emission M
=
g for each identified noble gas radionuclide i, mrad /yr per pCi/m ; (Table 3.3-1) s f
N
=
the air dose factor due to beta emissions j
g for each identified noble gas radionuclide s
1, mrad /yr per-pC1/m ; (Table 3.3-1) 1 i
(X/Q)v the annual average relative concentration
=
for areas at or beyond the restricted area boundary for long-term vent releases (greater than 500 hr/ year), sec/m8 (Appendix A, Table A-4);
9 (x/q)\\,
the relative concentration for areas-at or
=
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 radionuclide-
=
qgV i in gaseous effluents for short-term vent releases from both units (equal to: or less than 500 hrs / year), pCi; Qgv the total release of noble gas.radionuclide-
=
i in gaseous effluents _for long-term vent i
releases from both units (greater than 500 hrs /yr;, pCi;
,3.17 x 10-8
=
the inverse of the number-of seconds in a year.
Noble gas.es will be released _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 j
USNRC computer code'"XOQDOQ"' assuming 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> L
_per year short term releases and are given in-E Appendix-A-(Table A-7).
Values of' M Land N were I
calculated-using the methodology presented in NUREG-0133 and are giventin Table 3.3-1.
3.3.1.2 Cumulation'of Doses-Doses calculated monthly will be summed for comparison with' quarterly and annual-limits.
The monthly results vill be H
added to the doses cumulated trom the:
nther months in che quarter of interest p
and the year of interest and cor. pared to l
IBML e
~
e c
z.
H4 Rev. 11 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 gl48NinNUREG-0133andaregiveninTables3.3-2 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 c ontrolling location which will have one or more of the followin'i 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 W 8
values are in terms of x/Q(sec/m ) for the inhalation pathwcys 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 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-y CFR 190 will be-submitted.
3.3.3-Projection of Doses Doses-resulting from the release of gaseous.
l effluents will be projected monthly.
The doses H
calculated for:the present' month.will be used as the projected doses unless information exists indicating'that actual releases-7could differ significantly in the next month.
In'this caseL l
L 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 cf equation 3.3-1 or 3.3-2, additional waste gee tieatment will be provided.
If the E
projected release of 1-132', tritium, and radioactive l
particulates with half-lives greater than 8 Mays
- F exceeds 2 percent of the calendar year limit of equation-3.3-3, operat2on of
- the ventilation exhaust treatment equipment is required if-not currently in use.
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- C 0 U
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~~~~~--~ ~~~-----~ ~~~ ---
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O I-1
.; 3 H4 Rev. 11 Page 62 TABLE 5.1-1 PRAIRIE ISLAND NUCLEAR GENERATING PLANT RADIATION ENVIRONMENTAL MONITORING PROGRAM SAMPLING 'RCATIONS Type of Sample Code Collection Site Location River Water P-5 Upstream of Plant 2.3 mi @ 341*/NNW C
River Water P-6 Lock & Dam #3 1.6 mi @ 129 /SE
-Drinking Vater P-11 Red Wing Service Center 3.3 mi 0158'/SSE [
Vell Water P-25 Kinneman Farm 11.1-mi @ 331*/NMi C
Well Water' P-6 Lock & Dam #3 1.6 mi 0 129'/SE Vell Water P-8 Community Center 1.0 mi @ 321*/WNU Well Water P-9 Plant Well #2 0.3 mi 0 306*/NW Sediment-River P-20 Upstream of Plant 0.9 mi @ 45'/NE e
Sediment-River P-6 Lock & Dam #3 1.6 mi 0 129*/SE Sediment-Shoreline P-12 Downstream of Plant 3.0 mi-@ 116 /ESE Periphyton or P-5 Upstream of Plant 2.3 mi 0 348*/NNW c
Invertebrates P-12 Downstream of Plant 3.0 mi 0 116'/ESE Fish P-19 Upstream of Plant 1.3 mi @ O'/N e
Fish P-13 Downstream of Plant 3.5 mi @ ll3'/ESr.
Milk P-25 Kinneman Farm 11.1 mi 0 331 /N:a Milk P-14 Gustafson Farm 2.2 mi @ 173*/SSr Milk P-16
_ Johnson Farm 2.6 mi @-60*/ENE Milk P-17 Place Farm 3.5.mi @ 25 /NNE LMilk
'P-18 Christensen Farm
-3.7 mi 0 88*/E S
1BM
kN s
H4 Rev. 11 Page 63 TABLE 5.1-1 (Continued)
PRAIRIE ISLAND NUCLEAR GENERATING PLANT RADIti ION ENVIRONMENTAL MONITORING PROGRAM SAMPLING LOCATIONS i
Type of Location Sample Code Collection Site 11.1 mi @ 331 /l:
Cultivated Crops c
Kinneman Farm (Leafy Green Veg)
P-25 0.6 mi 0 158*/S:
P-24 Suter Residence Particulates and P-1 Air Station P-1 11.8 mi @ 316 /:,
c Radiciodine (air)
Particulates and P-2 Air Station P-2
- 0. 5 mi @ 294 /W:-
Radioiodine (air)
Particulates and P-3 Air Station P-3 0.8 mi @ 313 /E Radioiodine (air)
Particulates and P-4 Air Station P-4 0.4 mi @ 359 /N Radioiodine (air)
Particulates and P-6 Air Station P-6 1.6 mi @ 129 /F Radiciodine (air)
Direct Radiation 0.4 mi 0 359 /t (TLD)
PO1A Property Line Direct Radiation 0.3 mi @ 10 /N (TLD)
PO2A Property Line Direct Radiation 0.5 mi @ 183 /.
(TLD)
PO3A Property Line Direct Radiation 0.4 mi 0 204 /-
(TLD)
PO4A Property Line Direct Radiation 0.4 mi 0 225 /
(TLD)
POSA Property Line IBM
ib s
H4 Rev. 11 Page 64 TABLE 5.1-1 (Continued)
PRAIRIE ISLAND NUCLEAR GENERATING PLANT RADIATION ENVIRONMENTAL MONITORING PROGRAM SAMPLING LOCATIONS L
Type of Sample Code Collection Site Location Direct Radiation (TLD)
PO6A Property Line 0.4 mi 0 249'/WSW Direct Radiation (TLD)
PO7A Property Line 0.4 mi 0 268'/W i
Direct Radiation (TLD)
PO8A Property Line 0.4 mi @ 291*/VNV Direct Radiation (TLD)
PO9A Property Line 0.7 mi @ 317 /NW j
Direct Radiation (TLD)
P10A Property Line 0.5 mi 0 333 /NNV 1
l 1
l Direct Radiation (TLD)
PO1B Tom Killian Res.
4.7 mi @ 355'/N
)
)
Direct Radiation (TLD)
PO2B Roy Kinneman Farm 4.8 mi @ 17 /NNE 1
TLD PO3B Wayne Anderson Farm 4.9 mi @ 46 /NE l
Direct Radiation (TLD)
PO4B Nelson Drive (Road) 4.2 mi @ 61*/ENE Direct Radiation (TLD)
PO5B County Rd E & Coulee 4.1 mi 0 102'/ES!
Direct Radiation (TLD)
PO6B William Houschildt Res.
4.4 mi @ 112 /ES' l
Direct Radiation (TLD)
PO7B Red Wing Public Works 4.7 mi @ 140 /SL Direct Radiation 4;l mi @ 165 /Ss (TLD)
POBB David Wnuk Res.
I L
Direct Radiation 4.2 mi @ 187 /S (TLD)
F09B Hwy 19 South Direct Radiation 4.9 mi 0 200 /S:
(TLD)
P10B Cannondale Farm IBM l
Sk s
H4 Rev. 11 Page 65 TABLE 5.1-1 (Continued)
PRAIRIE ISLAND NUCLEAR GENERATING PLANT RADIATION ENVIRONMENTAL MONITORING PROGRAM SAMPLING LOCATIONS Type of Sample Code Collection Site Location Direct Radiation (TLD)
PilB Wallace Weberg Farm 4.5 mi @ 221*/SW Direct Radiation (TLD)
P12B Roy Gergen Farm 4.5 mi @ 247 /WSU Direct Radiation (TLD)
P13B Thomas O'Rourke Farm 4.4 mi @ 270*/W Direct Radiation (TLD)
P14B David Anderson Farm 4.9 mi @ 306*/NW Direct Radiation (TLD)
P15B Holst Farms 4.2 mi @ 347 /NN1i Direct Radiation (TLD)
POIS Federal Lock & Dam #3 1.6 mi @ 129*/SE Direct Radiation
(,"LD )
P02S Charles Suter Res.
0.5 mi @ 155 /SS Direct Radiation (TLD)
P03S Carl Gustafson Farm 2.2 mi 0 173 /S Direct Radiation (TLD)
P04S Richard Burt Res, 2.0 mi @ 202*/SSU j
Direct Radiation (TLD)
P0ss Kinney Store 2.0 mi 0 270 /W Direct Radiation (TLD)
P06S Earl Flynn Farm 2.5 mi G 299 /WNU Direct Radiation (TLD)
P25 Robert Kinneman Farm 11.1 mi @ 331 /Nt,,
- "c" denotes control location.
All other locations are indicators.
l 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.
"S" denotes special interest locations.
IBM
),
11 4 i
Rev. 11 l
Page 66 i
l; FIGURE 5.1-1 h
H SITE BOUNDARY TLD LOCATIONS ii l
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(NO SCALE) 5 NL-99P!739-1 [
C
.g FIGURE 5.1-2 j
g RADIOLOGICAL ENVIRONMENTAL SAMPLE PDINTS WITHIN Rev. 11
{
Page 67 10 - MILE RADIUS j
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N.S.P. TLD POINTS
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'rth Rev. 11 FIGURE ' 5.1-3 Page 68 RADIOLOGICAL ENVIRONMENTAL SAMPLE POINTS OUTSIDE 10 - MILE RADIUS pgA11 A4 es
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114 Rev. 11 Page 116 APPENDIX C DOSE PARAMETERS FOR RADIOIODINES, PARTICULATES AND TRITIUM This appendix contains the methodology which was used to calculate the dose parameters for radioiodines, particulates, and tritium to show compliance with 10 CFR 20 and Appendix 1 of 10 CFR 50 for gc.scous effluents.
These dose parameters, P. and R, were g
calculated using the methodology outlined in NUREG-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
g pathways.
l C.1 Calculation of P, I
The parameter, P, contained in the radiciodine and particulatesporkionofSection3.2,includespathway transport parameters of the ith radionuclide, the receptor's usage of the pathway media and the dosimetry of the exposure.
Pathway usage rates and the internal dosimetry are functions of the receptor's age: however, the child age group, will always receive the maximum dose under the exposure conditions assumed.
C.1.1 Inhalation Pathway l
P
=
-K'(BR) DFA (C.1-1) 1 g
l I
where:
dose parameter for radionuclide i for3;he P
=
t f
inhalation pathway, mrem /yr per pCi/m l
I K'
=
a constant of unit conversion:
6
=
10 pC1/pci; l
BR
=
tgebreathingrateofthechildagegroup, l
m /yr; J
DFA
=
the maximum organ inhalation dose factor g
for the child age group for radionuclide i, mrem /pci.
IBM
.k l
H4 Rev. 11 Page 117 The age group considered is the child group.
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 l
factors for the child DFAI, are presented in Table E-9 The total of Regulatory Guide 1.109 in units of mrem /pci.
body is considered as an organ in the selection of DFA Theincorporationofbreathingrateofthechildandtbe.
l unit conversion factor results in the following:
9 P
=
3.7 x 10 DFA (C.1-2) 3 C.2 Calculation of R f The radioiodine and particulate Technical Specification is applicable to the location in the unrestricted area where the combination of existing pathways and receptor age groups indicates che maximum potential exposure occurs.
The inhalation and ground plane exposure pathways shall be considered to exist at all locations.
The grass-goat-milk, the grass-cou-milk, grass-cow-meat, and vegetation pathways are considered based on their existence at the various
?ocations.
R values have been calculated for the adult, teen, child, knd infant age groups for the ground plane, cow milk, goat milk, vegetable and beef ingestion pathways.
The methodology which was utilized to calculate these values is presented below.
j C.2.1 Inhalation Pathway O (BR)a (DFA )a (C.2-1)
R
=
j 1
wher':
l R
=
dose factor for each identified 1
l 1
radionuclide i of the organ of interest, 3
i mrem /yr per pCi/m ;
a constant of unit conversion:
K'
=
6 10 pCi/pci;
=
of the receptor of (BR)a breathing ratg/yr:
=
age group a,m organ inhalation dose factor for (DFA ),
=
f radionuclide i for the receptor of age group a, mrem /pCi.
The breathing rates (BR) for the various age groups are l
tabulat'ed below, as givefl in Table E-5 of the Regulatory Guide 1.109 Revision 1.
l IBM
i}
-H4 Rev. 11 Page 118 3
l Age Group (a)_
Breathing Rate (m /yr) 1400 l_ #
Infant 3700 Child 8000 Teen 8000 Adult for the various age groups
. Inhalation dose factors (DFA )8 E-10 of Regulatory Guide are given in Tables E-7 throbg 1.109 Revision 1.
C.2.2 Ground Plane Pathway
-i t
-~
I K'K"(SF)DFG (1-e f )/A (C.2-2) 1 R
=
g g
1 G-where:
dose factor-for the ground plane pathway R
=
I for each identified radignuclide i;for t
o
~
the organ of interest, m mrem /yr per pCi/sec per; a constant of unit conversion';
K'
=
6
=
10 pC1/pci; a constant of unit conversion; K"
=
= ~
8760_hr/ year; for Ag the-radiological decgy. constant
=
radionuclide 1, sec the exposure time, sec; t
=-
8
-4.73 xl10.see-(15 years);
=
-theLground plane dose conversion-factgr DFG{
for L radionuclide i; mrem /hr-.-perJpci/m ;
=
-the1 shielding factor (dimensionless);
SF
-=
factor to account.for fractional I
=
f
, deposition of'radionuclide i.-
is For radionuclides other.than iodine, the; factor 1 4 For radioiodines, the value of If may-equal to one.
- However, a value of 1.0 was used in calc 01ating vary.
the R values in Table 3.3-2.
IBM
a
.h H4 Rev. 11 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 Regulatory Guide 1.109 Revision 1.
C.2.3 Grass-Cow or Goat-Hilk Pathway i'b) l
~
-1 t
E B1y(1-e R
1f r(1-c 1 g),
~
4 F ap m(
ia*
pa YA PA I
"I M
i pE i
g
~
~
~
~
~
~
-X E *es) +
~ b) l 1
B (1-e
-A t r(1-c (g_f g )
gy gh (C.2-3) ps YA gE PA i
i
~
where:
dose factor for the cow milk or goat milk R
=
g M
pathway, for each identified radionuclide i for the organ,gf interest, mrem /yr per pCi/sec per m a constant of unit conversion; K'
=
6 10 pCi/pCi;
=
the cow's or goat's feed consumption rate, Q
=
F kg/ day (wet weight);
the receptor's milk consumption rate for age U
=
- P group a, liters /yr; the agricultural productivi3y by unit area Y
=
p of pasture feed grass, kg/m ;
l
.theagriculturalproguctivitybyunit area Y
=
g of stored feed, kg/m ;
the stable element transfer coefficients,
' F*.
=
pCi/ liter per pC1/ day; fraction of deposited activity retained on r
=
cow's feed grass; the organ ingestion dose factor for (DFL )a
=
g radionuclide i for the receptor in age group a, mrem /pCi; 1
IBM
db H4 Rev. 11 Page 120 A
A
+A I
E i
W g
the radiological decyy constant for
-A
=
radionuclide i, sec A"
the decay constant-for removal of activity
=
on lyaf and plant surfaces by weathering, sec 5.73 x 10-7 sec 'l (corresponding to a
=
14 day half-life);
the. transport time from feed to cow or goat t
=
g to-milk, to receptor, sec; the transport time from harvest, to cow or
.t
=
h goat, to consumption, sec; period of time that activity builds up in t
=
b soil see; B
concentration factor'for uptake of iv radionuclide i from the soil by the edible-parts of crops, pCi/Kg (wet weight) per pCi/Kg-(dry soil);
effectlye surface density for soil, (dry P-
=
soil)/m ;
fraction of the year that the cow or goat f
=
p is on pasture;
-fraction of the cow feed that is-pasture
'f1
=
s grass while the cow.is on pasture; period of pasture grass l exposure during it
=-
ep~
the-growing season, sec; t,,
period'of crop exposurejduring the growing
=
season, sec; I-
=
factor to account for fractional deposition g
of-radionuclide 1.
For radionuclides other than todine, the factor 1 is 4
equal to one.
For radiciodines, the value of I may k
vary'. However, a value-of.l.0 was^used in calcu ating-
- the R. values Tables 3.3-9 through 3.3-16.
Milk cattle and~ goats _are considered to be fed from two fpotential 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 fp-was10.5 based upon an 6-month grazing period.
j
-IBM I
-, ~
so H4 Rev. 11 Page 121 Table C-1 contains the appropriate parameter values and their source in Regulatory Guide 1.109 Revision 1.
The concentration of tritium in milk is based on the airborne concentration rather than the deposition.
Therefore, the R is based on X/Q:
g K'K"'F S U,p(DFL ),
0.75(0.5/H)
(C.2-4)
R
=
mF g
TM where:
dose factor for the cow or goat milk pathway R
=
T for tritium for tge organ of interest, M
mrem /yr per pCi/m ;
a constant of unit conversion; K"'
=
3
=
10 Em/kg; 3
absolute humidity of the atmosphere, gm/m ;
H
=
0.75
=
the fraction of total feed that is water; the ratio of the specific activity of the 0.5
=
feed grass water to the atmospheric water, and other parameters 3nd values are given above.
A value of H of 8 grams / meter was used in lieu of site-specific in fo rma tion.
C.2.4 Grass-Cow-Meat Pathway The integrated concentration in meat follows in a similar manner to the development for the milk pathway, therefore:
-At[)
_x f
E,ep)
,Bh-( l ~'
~
ia r(1-e g
+
F ap f( M ), e ff
,9 F
g p"
7z 1.B i
PE i
i r(1-e g e s),
Bgy(1-e
)
-At E
fh (C.2-5) g_f f )
ps YA IA aE i
I where:
R
=
ce tor for the meat ingestion pathway f
B fJ 11onuclide i for any organ of interest, m'
miem/yr per pCi/sec; IBM
d)
H4 Rev. 11 Page 122 the stable element transfer. coefficients, F
=
f pC1/Kg per pC1/ day; the receptor's meat consumption rate for age
.U
=
ap group a, kg/yr; t,
the transport time from slaughter to
=
consumption, sec; the transport time from harvest to animal t
=
h consumption, sec; period of pasture grass exposure during-t
=
ep the growing-season, see; period of crop exposure during the growing t
es season, see; factor to account for fractional deposition I
=
f of radionuclide 1.
-For radionuclides other than iodinc,-the factor 1 is equal 4
to one.
For radiciodines, the value of I may vary.
However,avalueof1.0wasused.incalculatingtheR
. values in Tables 3.3-6 through 3.3-8.
All other_ terms remain the same as defined in Equation C.2-3; Table C-2 contains the values which were used in calculating R for the meat pathway.
g
- The concentration of tritium in meat is based on its airborne concentrationLrather than'the deposition.
cTherefore,-the R is based on X/Q.
g K'K"'F Qp,p(DFL ),
0.75(0.5/H)
(C.2-6)
U R
=
g g
TB-where:
R -
l dose. factor-for the' meat ingestion pathway
=
T for tritium-forlagy organ of interest, q
B mrem /yr per pCi/m All-other'tcrms areLdefined in Equation C.2-4 and C.2-5,
[
above.
C.2.5 : Vegetation-Pathval The integrated" concentration in vegetation consumed by man follows the expression developed in the derivation of-the milk factor.
Han is considered to consume two types of' vegetation (fresh and stored) that differ only in the time period between harvest and consumption, therefore:
IBM m
2 H4 Rev. 11 Page 123
~
~
.x t
-A't E' e B
(1-e i b)
~"
+
I K'(DFL )a Uf e
+
R
=
i i
ag YA iy yE i
i
-A t [ (1-e
~
-A ib g
E Biv(1-c y
r i e)
S c (C.2-7) pg ag YA PA E
g 1
where:
(
g dose factor for vegetable pathway for R
=
V radionuclidg,i for the organ of l
interest, m mrem /yr per pCi/sec K'
=
a constant of unit conversion:
6
=
10 pCi/pci; b
U, the consumption rat. of fresh leafy
=
vegetation by the receptor.in age group a, kg/yr; b
U, the consumption rate of stored
=
vegetation by the receptor in age group a. kg/yr; f'
the fraction of the annual intake of
=
l fresh leafy vegetation grown locally; i
f
=
the fraction of the annual intake of E
stored vegetation grown locally; the average time between harvest of t
=
g leafy vegetation and its consumption,-
sec; yI the average time between harvest of t
=
stored vegetation and its consumotion, see; 2
the vegetation areal density, kg/m ;
Y
=
y period of leafy vegetable exposure t
=
g during growing season, sec; 4
I.
=
factor to account for fractional 1
depositior of radionuclide 1.
IBM
uD
=
H4 Rev, 11 Page 124 For radionuclides other than iodine, the factor la is For radiciodines, the value of I 6ay equal to one.
However,avalueof1.0wasusedincaleblating vary.
the R values in cables 3.3-9 through 3.3-12.
All other factors were defined above.
Table C-3 presents the appropriate parameter values and their source in Regulatory Guide 1.109 Revision 1.
In lieu of site-specific data default values for f and g
f, 1.0 and 0.76, respectively were used in the cElcalation of R These values were obtained from TableE-15ofReku.latory Guide 1.109 Revision 1.
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 R
K'K"'
Uf
+
Uf (DFL ),
0.75(0.5/H)
(C.2-8)
T
=
g y
t g
where:
dose factor for the vegetable pathway for R
=
T tritium f r any organ of interest, mrem /yr 9
V per pCi/m~
All other terms remain the same as those in Equations C.2-4 and C.2-7.
l l
IBM t
0 Q
H4 Rev. 11 Page 125 T/.BLE C-1 Paramettra For Cow and Goat Milk Pathways Value Reference (Reg. Guide 1.109 Rev. 1 Parameter 50 (cow)
Table E-3 QF (kg/ day) 6 (goat)
Table E-3 Table E-15 2
0.7 Yp (kg/m )
1.73 x 105 ('
days)
Table E-15 g (seconds) t 1.0 (radiciodines)
Table E-15 r
0.2 (particulates)
Table E-15 l
Each radionuclide Tables E-11 to E-14 (LFL,), (mrem /pCi)
Each stable element Table E-1 (cow)
" (pci/ day per pCi/ liter)
Table E-2 (goat)
F 4.73 x 100 (15 yr)
Table E-15 b (seconds) t Table E-15 2.0 Y, (kg/m )
2 Table E-15 Y (kg/m )
0.7 7.78 x 100 (90 days)
Table E-15 h (see nds) t Table E 330 infant ap (liters /yr) 330 child U
Table E-5 Table E-5 400 teen Table E-5 310 adult t
(seconds) 2.59 x 106 (30 days)
Table E-15 t,, (seconds) 5.18 x 100'(60 days)
Table E '15 I
B (pC1/Kg (wet weight)
Each stable element Table E-1 gyper pCi/Kg (dry soil))
2 Table E-15 P (Kg (dry soil /m )
240 IBM
6 O
H4 Rev. 11 Page 126 TABLE C-2 Parameters For The Meat Pathway Pa4ameter Value
,. Reference (Reg. Ouide 1.109 Rev. 1 1.0 (radiciodines)
Table E-15 r
0.2 (particulates)
Teble E-15 Each stable element Table E-1 Ff (pC1/Kg per pCi/ day)
I O infant Table E-5
- P (Kg/yr',
U 41 child Table E-5 65 teen Table E-5 110 adult Table E-5 (DFL ),_(mrem /pCi)
Each radionuclide Tables E-11 to E-14 g
0.7 Table E-15 p (kg/m )
Y 2.0 Table E-15 Y, (kg/m )
4.73 x 108 (15 yr)
Table E-15 b (seconds) t t, (seconds) 1.73 x 106 (20 days)
Table E-15 7.78 x 106 (90 days)
Table E-15 h (seconds) t 2.59 x 106 (30 days)
Table E-15 t,p (seconds) t,, (seconds) 5.18 x 106 (60 days)
Table E-15 50 Table E-3 Qg (kg/ day)
Bg (pC1/Kg (wet weight)
Each stable element Table E-1 per p"i/Kg (dry soil))
P (Kg (dry soil)/m )
240 Table E-15 IBM i
O.
.b e
e H4 Rev. 11 Page 127 TAB 1.E C-3
)
Parameters Tor The Veget ale Patbvay Parameter yalue Reference (Reg._ Guide 1.109 Rev. I r (dimensionless) 1.0 (radiciodines)
Table E-1 0.2 (particulates)
Table E-1 (DFL ), (mrem /pCi)
Each radionuclide Tables E-11 to E-14 g
Uf(l.giyr)
- Infant 0
Table E-5
- Child 26 Table E-5
- Teen 42 Table E-5
- Adult 64 Table E 5 Uf(kg/yr)
- Infant 0
Table E-5
- Child 520 Table E-5
- Teen 630 Table E-5
- f.dult 520 Table E-5 g (seconds) 8.6 x 100 (1 day)
Table E %
t h (see tids) 5.18 x 100 (60 days)
Tabl.i E-15 t
2 Y (kg/m )
2.0 Table E-15 y
t,(seconds) 5.18 x 10' (60 days)
Table E-15 b (see ads) 4.73 x 108 (15 yr)
Table E 15 t
2 P(Kg(dry soil)/m )
240 Table E-15 Bgy(pCi/Kg(wet weight)
Each stable element Table E-1 per pCi/kg (dry soil))
IBM
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