ML20056G698
| ML20056G698 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 06/30/1993 |
| From: | NORTHERN STATES POWER CO. |
| To: | |
| Shared Package | |
| ML20056G694 | List: |
| References | |
| NUDOCS 9309070063 | |
| Download: ML20056G698 (17) | |
Text
s l
l TRANSMITTAL MANIFEST NORTHERN STATES POWER COMPANY l
NUCLEAR GENERATION DEPARTMENT l
PRAIRIE ISLAND NUCLEAR GENERATING PLANT i
Effluent and Waste Disposal Semiannual Report i
for Januarv throuah June 1993 MANIFEST DATE: August 27,1993 l
USNRC 4
ANI Library 1
i Regional Admin 111 Fluor Danicis 2
l l
NRR Project Manager T Synder
<DCD_:
Corporate Library Resident inspector Shaw Pittman Potts & Trowbridge 1
l J K Gilchrist 1
J Silberg D A Schuelke 1
Westinghouse Electric 3
(
P H Kamman 1
T J Kitchen i
Pl Site General Mgr.
1 J A Usem ERAD Dept.
K C Hoskins Attn: Records Clerk 1
Safety Audit Committee 10 l
Communications Department R O Anderson
- i Attn: Tom Bushee 1
A B Cutter NRS File (4131116) 1 R L Hannen N S S File 1
F W Hartley MDH 1
W J Hill Attn: Kris Sanda D D Lanning D S Mendele M D Wadley C R Steinhardt Secretary Manifest File l
l 9309070063 930827 N
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1993 Effluent Semiannual Report REV. O s
Fage l of 8 Retention: Lifetime EFFLUENT SEMIANNUAL REPORT 04-JAN-93 THROUGH 04-JUL-93 SUPPLEMENTAL INFORMATION i
Facility:
Prairie Island Nuclear Generating Plant i
Licensee:
Northern States Power Company License Numbers: DPR-42 & DPR-60 i
A.
Regulatory Limits
?
1.
Liquid Effluents:
a.
The dose or dose commitment to an individual from radioactive materials in liquid effluents released from the site shall be limited to:
for the quarter 3.0 mrem to the total body 5
10.0 mrem to any organ for the year 6.0 mrem to the total body
(
20.0 mrem to any organ 2.
Gaceous Effluents:
a.
The dose rate due to radioactive materials released in gaseous effluents from the site shall be limited to:
noble gases 4500 mrem / year total body 43000 mrem / year skin i
i I-131, H-3, LLP 41500 mrem / year to any organ b.
The dose due to radioactive gaseous effluents shall be limited to:
l noble gases 410 mrad / quarter gamma 420 mrad / quarter beta i
420 mrad / year gamma 440 mrad / year beta I-131, H-3, LLP (15 mrom/ quarter to any organ 430 mrem / year to any organ i
I
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i 1993 EFFLUENT SEMIANNUAL REPORT REV. O PAGE 2 s
i J
B.
Maximum Permissible Concentration i
1.
Fission and activation gases in gaseous releases:
10 CFR 20, Appendix B, Table 2, Column 1 l
2.
Iodine and particulates with halflives greater than 8 days in gaseous releases:
l 10 CFR 20, Appendix B, Table 2, Column 1 l
3.
Liquid effluents for radionuclides other than dissolved or entrained gases:
l 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
2 1.
Fission and activation gases Total GeLi 25%
in gaseous releases:
Nuclide GeLi 2.
Iodines in gaseous releases:
Total GeLi 25%
Nuclide GeLi 3.
Particulates in gaseous releases:
rotal GeLi 25%
Nuclide GeLi i
4.
Liquid effluents Total GeLi 25%
Nuclide GeLi 5
i E.
Manual Revisions t
1.
Offsite Dose Calculations Manual latest Revision number:
12
{
Revision date
- 17-JUN-91 i
i t
l 1993 EFFLUENT SEMIANNUAL REPORT
.REV. O PAGE-3 i
1.0 BATCH RELEASES (LIQUID)
QTR: 01 QTR: 02 1.1 NUMBER OF BATCH RELEASES 3.40E+01 2.20E+01 1.2 TOTAL TIME PERIOD (HRS) 5.40E+01 3.44E+01.
1.3 MAXIMUM TIME PERIOD (HRS) 2.12E+00-2.55E+00 1.4 AVERAGE TIME PERIOD (HRS) 1.59E+00 1.56E+00 1.5 MINIMUM. TIME PERIOD (HRS) 1.28E+00 1.33E+00 1.6 AVERAGE MISSISSIPPI RIVER FLOW (CFS) 1.04E+04 5.68E+04
-l 2.0 BATCH RELEASES (GASEOUS)
QTR: 01 QTR: 02 2.1 NUMBER OF BATCH RELEASES 1.60E+01 2.00E+00' 2.2 TOTAL TIME PERIOD (HRS) 7.84E+01 5.30E+00 i
2.3 MAXIMUM TIME PERIOD (HRS) 4.88E+01 3.10E+00 l
2.4 AVERAGE TIME PERIOD (HRS) 3.05E+00 2.65E+00-2.5 MINIMUM TIME PERIOD (HRS) 1.20E-01 2.18E+001 3.0 ABNORMAL RELEASES (LIQUID)
QTR: 01 QTR: 02 3.1 NUMBER OF RELEASES 0.00E+00 0.00E+00 3.2 TOTAL ACTIVITY RELEASED (CI) 0.00E+00 0.00E+00 3.3 TOTAL TRITIUM RELEASED (CI) 0.00E+00 0.00E+00 4.0 ABNORMAL RELEASES (GASEOUS)
QTR: 01 QTR: 02 4.1 NUMBER OF RELEASES 0.00E+00 0.00E+00-4.2 TOTAL ACTIVITY RELEASED (CI) 0.00E+00 0.00E+00
1993 EFFLUENT SEMIANNUAL REPORT REV.
0~
PAGE 4 TABLE 1A
. GASEOUS EFFLUENTS - SUMMATION OF ALL RELEASES OTR: 01 QTR: 02 5.0 FISSION AND ACTIVATION GASES 5.1 TOTAL RELEASE (CI) 3.51E-01 3.02E-01 5.2 AVERAGE RELEASE RATE (UCI/SEC) 4.47E-02 3.84E 5.3 GAMMA DOSE (MRAD) 2.55E-05 2.47E-05 5.4 BETA DOSE (MRAD) 2.87E-03 2.46E-03 5.5 PERCENT OF GAMMA TECH SPEC (%)
2.55E-04 2.47E-04 5.6 PERCENT OF BETA TECH SPEC (%)
1.44E-02 1.23E-02 6.0 IODINES 6.1 TOTAL I-131 (CI) 0.00E+00 0.00E+00 6.2 AVERAGE RELEASE RATE (UCI/SEC) 0.00E+00 0.00E+00 7.0 PARTICULATES 7.1 TOTAL RELEASE (CI) 4.61E-05 1.33E-06 7.2 AVERAGE RELEASE RATE (UCI/SEC) 5.87E-06 1.69E-07 8.0 TRITIUM 8.1 TOTAL RELEASE (CI) 2.21E+01 1.21E+01 8.2 AVERAGE RELEASE RATE (UCI/SEC) 2.81E+00 1.54E+00 H9. 0 TOTAL IODINE, PARTICULATE AND TRITIUM (UCI/SEC) 2.81E+00 1.54E+00 10.0 DOSE (MREM) 5.46E-02 2.23E-02 11.0 PERCENT OF TECH SPEC (%)
3.64E-01 1.49E 12.0 GROSS ALPHA (CI) 9.67E-09 5.24E-08
~..
. =.
y 6
1993 EFFLUENT SEMIANNUAL REPORT REV. O PAGE 5 r
. TABLE 1C GASEOUS EFFLUENTS - GROUND LEVEL RELEASES I
l t
13.0 FISSION AND ACTIVATION GASES i
CONTINUOUS MODE BATCH MODE
{
NUCLIDE UNITS QTR: 01 QTR: 02 QTR: 01 QTR: 02 i
i KR-85 CI 3.52E-01 3.00E-01 XE-133 CI 1.31E-04 1.98E-03 i
XE-131M CI 3.20E-06 XE-135 CI 2.85E-06 l
TOTAL CI 0.00E+00 0.00E+00 3.52E-01 3.02E-01 i
l 14.0 IODINES CONTINUOUS MODE BATCH MODE l
I NUCLIDE UNITS QTR: 01 QTR: 02 QTR: 01 QTR: 02 TOTAL CI 0.00E+00 0.00E+00 0.00E+00 0.00E+00 i
i j
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15.0 PARTICULATES CONTINUOUS MODE BATCH MODE
{
L 9
NUCLIDE UNITS QTR: 01 QTR: 02 QTR: 01 QTR: 02 i
BR-82 CI 1.02E-06 4
CO-58 CI 3.98E-06 CO-60 CI 9.24E-06
]
CS-134 CI 1.16E-05 CS-137 CI 1.63E-05 i
MN-54 CI 4.70E-06 NB-95 CI 3.03E-07 SR-90 CI 3.09E-07 TOTAL CI 0.00E+00 1.33E-06 4.61E-05 0.00E+00
~.l
.1993 EFFLUENT SEMIANNUAL REPORT REV. O PAGE 6 j
TABLE 2A l
LIQUID EFFLUENTS - SUMMATION OF ALL RELEASES j
QTR: 01 QTR: 02 r
f 16.0 VOLUME OF WASTE PRIOR TO DILUTION (LITERS) 3.12E+10 2.52E+10 17.0 VOLUME OF DILUTION WATER (LITERS) 1.36E+11 9.20E+10 18.0. FISSION AND ACTIVATION PRODUCTS 18.1 TOTAL RELEASE W/O H-3, RADGAS, ALPHA (CI) 5.74E-02 3.06E-02 i
18.2 AVERAGE DILUTED CONCENTRATION (UCI/ML) 3.44E-10 2.62E-10 i
19.0 TRITIUM l
19.1 TOTAL RELEASE (CI) 5.16E+01 4.79E+01 l
19.2 AVERAGE DILUTED CONCENTRATION (UCI/ML) 3.09E-07 4.09E-07 l
20.0 DISSOLVED AND ENTRAINED GASES i
20.1 TOTAL RELEASE (CI) 6.13E-06 2.56E-06 20.2 AVERAGE DILUTED CONCENTRATION (UCI/ML) 3.67E-14 2.19E-14 l
21.0 GROSS ALPHA (CI) 0.00E+00 0.00E+00
.f 22.0 TOTAL TRITIUM, FISSION AND ACTIVATION l
PRODUCTS (UCI/ML) 3.09E-07 4.09E-07 23.0 TOTAL BODY DOSE (MREM) 1.44E-02 5.34E-03 i
24.0 CRITICAL ORGAN 24.1 DOSE (MREM) 1.44E-02 5.34E-03 24.2 ORGAN TTL BODY TTL BOBY 25.0 PERCENT OF TOTAL BODY TECH SPEC LIMIT (%)
4. 8 0 E 1.78E-01 26.0 PERCENT OF CRITICAL ORGAN TECH SPEC LIMIT (%)
4.80E-01 1.78E-01
,r
4 1993 EFFLUENT SEMIANNUAL REPORT REV. O PAGE 7 TABLE 2A j
LIQUID EFFLUENTS - SUMMATION OF ALL RELEASES l
27.0 INDIVIDUAL LIQUID EFFLUENT l
CONTINUOUS MODE BATCH MODE NUCLIDE UNITS OTR: 01 QTR: 02 QTR: 01 QTR: 02 5
AG-110M CI 1.90E-02 1.53E-03 CO-58 CI 1.37E-02 2.09E-03 I
I CO-60 CI 5.17E-06 7.81E-03 2.01E-03 CR-51 CI 2.47E-03 5.71E-05 l
FE 59 CI 7.48E-04 2.69E-05 3
MN-54 CI 4.51E-04 5.08E-05 I
NB-95 CI 9.40E-04 9.40E-05 l
l SB-124 CI 2.64E-03 5.41E-05 SB-125 CI 4.50E-03 4.96E-04 i
SN-113 CI 7.68E-04 1.21E-04 ZR-95 CI 6.85E-04 5.75E-05 l
i CS-134 CI 1.48E-03 5.61E-04 CO-57 CI 1.90E-05 6.27E-06
[
CS-137 CI 1.77E-04 1.54E-03 5.99E-04 f
AG-108M CI 5.43E-05 BE-7 CI 3.38E-04 5.05E-04 ZN-65 CI 9.62E-06 7.11E-06 NB-97 CI 1.48E-05 2.82E-06 2.31E-06 l
SC-47 CI 1.11E-05 d-I NA-24 CI 5.56E-06 I
l FE-55 CI 1.65E-05 6.62E-06 2.23E-02
]
TOTAL CI 1.97E-04 1.65E-05 5.72E-02 3.06E-02 J
1993 EFFLUENT SEMIANNUAL REPORT REv. O PAGE 8 TABLE 2A LIQUID EFFLUENTS - SUMMI. TION OF ALL RELEASES 28.0 DISSOLVED AND ENTRAINED GASES CONTINUOUS MODE BATCH MODE NUCLIDE UNITE QTR: 01 QTR: 02 QTR: 01 QTR: 02 KR-85M CI 2.56E-06 XE-133 CI 6.13E-06 i
TOTAL CI 0.00E+00 0.00E+00 6.13E-06 2.56E-06 m
1
4 PINGP 753. Rev. 4 Page 1 of 6 Retention: Life i
i PRAIRIE ISLAND NUCLEAR GENERATING PLANT Period: 1-1-93 to7-1-93 NORTHERN STATES POWER License No. DPR-42 i
SOLID RADIOACTIVE WASTE DISPOSAL SEMI-ANNUAL REPORT i
TABLE 1 SOLID WASTE AND IRRADIATED FUEL SHIPMENTS 1.
Solid Waste Total Volumes and Total Curie Quantities:
i 6. MONTH L
'. EST.J TOTAL:
CONTAINER TTYPE'OF WASTE.
- UNITS:
iPERIOD
_ ERRORi%y (D.lSPOSAL';
u TOTALS?
- -(0.00- E00);
~
.(0.L.) E00) ;
- VOLUMES $
- (LIST)3 A.
Resins ft3 2.72E+2 135.8 Ci 2.83E+2 2.5E+1 l
l B.
Dry-Compacted ft3 Ci C.
Non-Compacted ft3 Ci D.
Filter Media ft3 Ci S.
Spent Fuel ft3 Ci L NOTE::
The solid waste information provided in this report is the volume and activity of the low-level waste leaving the Prairie Island site. No allowance is made for off-site volume reduction prior to burial.
l
PINGP 753, Rev. 4 Page 2 of 6 PRA!RIE ISLAND NUCLEAR GENERATING PLANT Period: 1-1-93 to 7-1-9; NORTHERN STATES POWER License No. DPR-42 SOLID RADIOACTIVE WASTE DISPOSAL SEMI-ANNUAL REPORT '
r TABLE 1 SOLID WASTE AND IRRADIATED FUEL SHIPMENTS [ continued]
1 2.
Principal Radionuclide Composition oy Type of Waste:
Percent TYPE (From Page 1)
Nuclide Abundance A
Co-60 36.5 Ni-63 35.1 Cs-137 9.5 Fe-55 9.0 l
Cs-134 3.9 Co-58 2.9
{
Mn-54 2.5 i
- " Interred - Not Measured on Site
PINGP 753, Rev. 4
{
Page 3 of 6 PRAIR!E ISLAND NUCLEAR GENERAT!NG PLANT Period: 1-1-93 to7-1-93 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):
i Percent TYPE (From Page 1)
Nuclide Abundance
. t J
i i
i
)
' ~ Inferred - Not Measured on Site
PINGP 753, Rev. 4 Page 4 of 6 PRAIRIE ISLAND NUCLEAR GENERATING PLANT Period: 1-1-93 to 7-1-9:
NORTHERN STATES POWER License No. DPR-42 SOLID RADIOACTIVE WASTE DISPOSAL SEMI-ANNUAL REPORT j
TABLE 1 SOLID WASTE AND IRRADIATED FUEL SHIPMENTS [ continued) i 3.
Solid Waste Disposition:
Number of Shipments Mode Destination 2
TRUCK BARNWELL 4.
Irradiated Fuel Shipments:
Number of Shipments Mode Destination 0
1 4
1 9
r r
PINGP 753, Rev. 4 Page 5 of 6 l
PRAIRIE ISLAND NUCLEAR GENERATING PLANT Period: 1-1-93 to 7-1-92 '
NORTHERN STATES POWER Ucense No. DPR-42 j
SOLID RADIOACTIVE WASTE DISPOSAL SEMI-ANNUAL REPORT TABLE 1 SOLID WASTE AND IRRADIATED FUEL SHIPMENTS [ continued]
l i
5.
Shipping Container and Solidification Method:
i Disposal i
No.
Volume Activity Type of Container Solidif.
(Ft3)
(Ci)
Waste Code Code 93-03 135.8 160 A
L N/A 93-06 135.8 173 A
L N/A l
I 4
i TOTALS 2
271.6 283 i
)
CONTAINER CODES:
L=
LSA (Shipment Type)
A = Type A B = Type 8 j
Q=
Highway Route Controlled Quantity SOLIDIFjCATJON CODES: C = Cement TYPES OF WASTE: A=
Resins B=
Dry Compacted C= Non-Compacted j
D=
r'ilter Media S=
Spent Fuel i
PINGP 753, Rev. 4 Page 6 of 6 r)RAIRIE ISLAND NUCLEAR GENERATING PLANT Period: 1-1-93 to 7-1-92i NORTHERN STATES POWER Ucense No. DPR-42 SOLID RADIOACTIVE WASTE DISPOSAL SEMI-ANNUAL REPORT TABLE 2 PROCESS CONTROL PROGRAM CHANGES TITLE:
Process Control Program for Solidification / Dewatering of Radioactive Waste from Liquid Systems Current Revision Number:
4 Effective Date:
s-23-91 l
TNOTE~.
If the effective date of the PCP is within the period covered by this report, then a description and justification of the changes to the PCP is required (T.S.6.7.A.4).
Attach the sidelined pages to this report.
i Changes / Justification:
I 6
i 1
m U
4 l
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t
Subject:
Composite results compared to T.S.
LLDs It has been discovered that some of the strontium 89 analysis results i
for the samples collected in 1989 through 1992 were reported as "Less Than" values but greater than the LLD values specified by the Prairie Island Technical Specifications Table TS.4.17-3 and TS.4.17-4.
The Strontium 89 LLDs specified are 5.0E-08 uti/ml for liquids and 1.0E-ll uCi/cc for l
airborne samples.
Per discussions with NRC and per Technical Specifications this attachment is describing the reasons the LLDS were greater that T.S.
The analysis of airborne and liquid composite samples for Strontium 89 is conducted by an offsite laboratory.
During the period involved, two separate offsite laboratories were involved in the composite analysis.
Analysis of sampling procedures and discussions with the laboratory we are currently utilizing revealed that the discrepancies occurred due to l
the following reasons:
1)
We ship the offsite laboratory half of the volume collected and composited during the previous quarter.
The offsite lab then analyzes less than half of the amount they receive in order to maintain a sufficient volume to allow them to reanalyze if necessary.
In the past we have procedurally collected a one liter sample to save for composite analysis with each release.
If a release occurs from a i
specific release point only once during the quarter, the amount of sample analyzed was less than 250 mls.
2)
We have specified to the offsite laboratory that all analysis results be decay corrected back to a specific date part way through the previous quarter.
Strontium 89 has a 50 day half life, so any delay in the shipping, analysis or counting causes the calculated LLD value to be higher.
On one occasion, due to temporary restrictions of shipping radioactive materials over Indian reservation property, the entire quarters composite samples were held on site until the shipping restrictions were lifted.
This caused the entire quarters liquid Sr-89 values
)
I to be higher than LLD due to decay correction.
l 3)
Some airborne batch releases are Containment Post LOCA vents and the J
only particulate filter " ailable for analysis is from a prerelease sample taken from conta...ent prior to the release.
Do to the small sample volume and possible long decay time, depending upon when the vent release occured during the quarter, it can be difficult to meet the airborne T.S.
LLD of 1.0E-11 for Sr-89.
The following steps have been taken to ensure that all Strontium 89 composite results will be analyzed in a manner that can accurately report LLD values less than those specified in our technical specifications:
1)
Plant procedures have been modified to ensure that a minimum of 2 liters of sample be collected and available to ship to the offsite laboratory for analysis from every release point that was used during the quarter.
During winter months, at least 1800 mis will be shipped to allow space in the bottle to prevent freezing from breaking the bottle.
2)
The laboratory indicates that the specified LLD can be obtained if a minimum volume of 600 ml is used for the analysis.
We are requesting that a sample volume of 800 to 1000 mls be used for the analysis.
3}
The analysis request letter that accompanies the composite samples to the offsite laboratory will specifiy the required LLDs for the various samples types.
4)
All containment vent and purge particulate samples will be counted for gross beta gamma.
If the analysis indicates beta gamma activity the particulate filter will be shipped to the offsite laboratory before the decay correction would cause the LLD calculation to be greater than T.S.
5)
The plant review and acceptance of laboratory results will be improved.
The computer program that is used to input the composite results inte our database will be revised to automatically check to insure that analysis results are less than the specified LLD values.
Included are attachments to the Semi-annual Effluent Release Reports that specify the release points and their respective "Less Than" activity values as reported by the offsite laboratory for the periods of Quarter 1 of 1989 through Quarter 4 of 1992 Only the release points with "Less Than" values that exceeded the LLD are included in the attachments.
The Semi-annual Effluent Release Reports results remain unchanged since all original calculations were maae with these release concentrations less than detectable.
Based on plant experience and other analysis results of the involved release paths, we believe the activity was well less than the Tech. Spec. LLD values.
t l
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