L-PI-08-027, Nuclear Management Company, Inc. 2007 Annual Radioactive Effluent Report and Offsite Dose Calculation Manual
| ML081370317 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 05/13/2008 |
| From: | Wadley M Nuclear Management Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| FOIA/PA-2010-0209, L-PI-08-027 | |
| Download: ML081370317 (29) | |
Text
Committed to Nuclear Ex6ellence Prairie Island Nuclear Generating Plant Operated by Nuclear Management Company, LLC L-PI-08-027 MAY 1 3 20o8 TS 5.5.1.c TS 5.6.3 U S Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555 Prairie Island Nuclear Generating Plant Units 1 and 2 Dockets 50-282, 50-306 and 72-10 License Nos. DPR-42, DPR-60 and SNM-2506 2007 Annual Radioactive Effluent Report and Offsite Dose Calculation Manual Pursuant to the applicable Prairie Island Nuclear Generating Plant (PINGP) Technical Specifications (TS), Appendix A to Operating Licenses DPR-42 and DPR-60, and the requirements of the Offsite Dose Calculation Manual (ODCM), Nuclear Management Company, LLC (NMC) submits the 2007 Annual Radioactive Effluent Report which is comprised of the following reports:
Enclosure 1 contains the Off-Site Radiation Dose Assessment for the period January 1, 2007 through December 31, 2007 in accordance with the requirements of the ODCM.
Enclosure 2 contains the Annual Radioactive Effluent Report, Supplemental Information, for the period January 1, 2007 through December 31, 2007 in accordance with the requirements of TS 5.6.3 and the ODCM.
Enclosure 3 contains the Effluent and Waste Disposal Annual Report, Solid Waste and Irradiated Fuel Shipments, for the period January 1, 2007 through December 31, 2007 in accordance with the requirements of TS 5.6.3 and the ODCM.
Enclosure 4 is an attachment to the 2007 Annual Effluent Report which contains a dose assessment, for the first quarter of 2007. In accordance with industry guidelines on groundwater monitoring, the report includes the dose assessment for a secondary steam condensate leakage which occurred March 21, 2007.
1717 Wakonade Drive East e Welch, Minnesota 55089-9642 Telephone: 651.388.1121 Atoc~
Document Control Desk Page 2 contains a complete copy of the entire ODCM, Revision 21, dated 7/25/07.
In accordance with the requirements of TS 5.5.1 .c., the changes are identified by markings in the margin of the affected pages. The manual also contains a Record of Revisions which includes a summary of the revision changes (refer to page 8 of the ODCM).
Summary of Commitments This letter contains no new commitments and no revisions to existing commitments.
Michael D. Wadley Site Vice President, Prairie Island Nuclear Generating Plant Nuclear Management Company, LLC Enclosures (5) cc: Regional Administrator, USNRC, Region III Project Manager, Prairie Island Nuclear Generating Plant, USNRC, NRR NRC Resident Inspector - Prairie Island Nuclear Generating Plant Minnesota Department of Health - Radiation Unit
ENCLOSURE 1 OFF-SITE RADIATION DOSE ASSESSMENT January 01, 2007 - December 31, 2007 6 pages follow
PRAIRIE ISLAND NUCLEAR GENERATING PLANT OFF-SITE RADIATION DOSE ASSESSMENT FOR January through December 2007 An Assessment of the radiation dose due to releases from Prairie Island Nuclear Generating Plant during 2007 was performed in accordance with the Offsite Dose Calculation Manual as required by Technical Specifications. Computed doses were well below the 40 CFR Part 190 Standards and 10 CFR Part 50 Appendix I Guidelines.
Off-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 Annual Radioactive Effluent and Waste Disposal Report prepared for NRC review for the year of 2007.
Off-site Doses from Gaseous Release Computed doses due to gaseous releases are reported in Table 1. Critical receptor location and pathways for organ doses are reported in Table 2. Gaseous release doses are a small percentage of Appendix I Guidelines.
Off-site Doses from Liquid Release Computed doses due to liquid releases are reported in Table 1. Critical receptor information is reported in Table 2. Liquid release doses, both whole body and organ, are a small percentage of Appendix I Guidelines.
Doses to Individuals Due to Activities Inside the Site Boundary Occasionally sportsmen enter the Prairie 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 traffic exists through this area.
For purposes of estimating the dose due to recreational and river water transportation activities within the site boundary, it is assumed that the limiting dose within the site boundary would be received by an individual who spends a total of seven days per year on the river just off-shore from the plant buildings (ESE at 0.2 miles). The gamma dose from noble gas releases and the whole body and organ doses from the inhalation pathway due to Iodine 13 1, Iodine-1 33, tritium and long-lived particulates were calculated for this location and occupancy time. These doses are reported in Table 1.
Critical Receptor location and pathways for organ doses are reported in Table 2.
ABNORMAL RELEASES There were a total of two (2) abnormal releases for 2007. The 2007 abnormal releases are summarized below:
- 1. Leak in Waste Gas System On 10/16/07 during data review, operations noted a negative trend in total volume for the routine Waste Gas System inventory. Further investigation determined that a Waste Gas Decay Tank Release had not occurred for an atypical length of time. Ventilation monitor trend plots and weekly gas grabs were reviewed and indicated no activity. It was determined that a very small leak had been present for as much as 6 months. Engineering determined that approximately 3,000 cubic feet of waste gas was lost.
Cause: Leakage was identified at the Gas Analyzer Panel pump. The Gas Analyzer was isolated until repairs could be performed. Leakage stopped. From this location all release would have exited through Unit One Auxiliary Building Ventilation.
Corrective Action: 129 Waste Gas Decay Tank (WGDT), the inservice tank, was sampled for nuclide mix.
The identified mix was used in the release calculations but, activity levels were determined to be unrepresentatively low, due to the extended time period of release.
Activity levels of the identified mix were extrapolated to the level of the sample taken for a last WGDT release performed prior to the leak:
Nuclide uCi Released Gamma Beta Dose (mrad) Dose (mrad)
Ar-41 1.08E+02 1.08E-06 3.80E-07 Kr-85 9.59E+04 2.OOE-04 1.77E-06 Kr-85M 7.62E+0 1 1.61E-07 1.00E-07 Xe-133 2.44E+04 2.74E-04 9.22E-05 Xe-135 1.32E+03 3.48E-06 2.72E-06 TOTAL 4.79E-04 9.72E-05 H3 1.26E+03 uCi 2.16E-06 mrem Activity was applied to abnormal release file RACO 193, as a Unit One Auxiliary Building Release. Release duration was conservatively set at 1 week and total dose was applied to the month of October, the 4 th Quarter.
Event was captured in the site's Action Request System: CAP-01 115005.
Repairs were accomplished and the Gas Analyzer was returned to service.
Result: The dose from the activity released represented a small percentage of the total dose and was a very small percentage of limits. The dose did not impose upon the health and safety of the public.
The event was reported to the NRC Region 3 Radiation Protection (RP) Inspector, at the time of the event.
- 2. Leaking 1. Steam Generator Relief On 6/17/07, while performing a surveillance procedure on I1 Steam Generator Safety Relief, CV-31084 did not completely reseat. Discharge piping temperatures increased. The isolation was shut and the valve was stroked in an attempt to reset. It was determined that CV-31084 did reseat as evidenced by decreasing downstream temperatures.
Cause: CV-31084 did not fully reseat during performance of surveillance procedure. When the valve was unisolated, following performance of the SP, it leaked.
Corrective Action: CV-31084 was reisolated and stroked. Leakage ended. Work Request #25903 was issued.
Engineering provided a volume released. Based on this volume it was determined that a one second release with the valve full open would conservatively represent the release volume.
The Steam Generators-were sampled and a release file was created to document the release.
The dose consequences were determined to be:
H3 1.53E+01 uCi 2.6E-09 mrem Activity was applied to abnormal release file RAB0060, as an 11 Steam Generator
-Steam Release.
Event was captured in the site's Action Request System: CAP-01097198.
Result: The dose from the activity released represented a small percentage of the total dose and was a very small percentage of limits. The dose did not impose upon the health and safety of the public.
The event was reported to the NRC Region 3 Radiation Protection (RP) Inspector, at the time of the event.
40CFR190 COMPLIANCE The calculated dose from the release of radioactive materials in liquid or gaseous effluents did not exceed twice the limits of I OCFR50, Appendix I, therefore compliance with 40CFRI190 is not required to be assessed, in this report.
SAMPLING. ANALYSIS AND LLD REQUIREMENTS The minimum sampling frequency, minimum analysis frequency and lower limit of detection (LLD) requirements, as specified in ODCM Tables 2.1 and 3.1 were not exceeded in 2007.
MONITORING INSTRUMENTATION There were no occurrences when less than the minimum required radioactive liquid and/or gaseous effluent monitoring instrumentation channels were operable as required by ODCM Tables 2.2 and 3.2.
Doses to Individuals Due to Effluent Releases from the Independent, Spent Fuel Storage Facility (ISFSI)
Two (2) fuel casks were loaded and placed in the storage facility during the 2007 calendar year. The total number of casks in the ISFSI is twenty-four (24). There has been no release of radioactive effluents from the ISFSJ.
CURRENT ODCM REVISION The Offsite Dose Calculation Manual was revised in 2007. The current revision is 21. The revision date is July 25, 2007. A copy is submitted with this year's report.
PROCESS CONTROL PROGRAM There were no changes made to the Process Control Program in 2007. Current manual is revision 8, August 25, 1999.
Table 1 OFF-SITE RADIATION DOSE ASSESSMENT - PRAIRIE ISLAND PERIOD: JANUARY through DECEMBER 2007 10 CFR Part 50 Appendix I Guidelines for a 2-unit site per year GaseQus-Releases Maximum Site Boundry Gamma Air Dose (mrad) 6. 57E-06 20 Maximum Site Boundry Beta Air Dose (mrad) 2. 05E-04 40 Maximum Off-site Dose to any organ (mrem)* 3.26E-02 30 Offshore Location Gamma Dose (mrad) 4.86E-07 Total Body (mrem)* 1. 28E-03 Organ (mrad)* 1.28E-03 30 Liquid Releases Maximum Off-site Dose Total Body (mrem) 1. 72E-03 6 Maximum Off-site Dose Organ - GI TRACT (mrem) 2. 50E-03 20 Limiting Organ Dose Organ - TOTAL BODY (mrem) 1. 72E-03 6
- Long-Lived Particulate, 1-131, 1-133 and Tritium
Table 2 OFF-SITE RADIATION DOSE ASSESSMENT - PRAIRIE ISLAND SUPPLEMENTAL INFORMATION PERIOD: JANUARY throunh DECEMBER 2007 Gaseous Releases Maximum Site Boundary Dose Location (From Building Vents)
Sector WNW Distance (miles) 0.4 Offshore Location Within Site Boundary Sector ESE Distance (miles) 0.2 Pathway Inhalation Maximum Off-site Sector SSE Distance (miles) 0.6 Pathways Plume, Ground, Inhalation, Vegetables Age Group Child Liquid Releases Maximum Off-site Dose Location Downstream Pathway Fish
ENCLOSURE 2 ANNUAL RADIOACTIVE EFFLUENT REPORT SUPPLEMENTAL INFORMATION January 01, 2007 - December 31, 2007 9 pages follow
2007 Annual Radioactive Effluent Report REV. 0
ý,Page 1 'of 9
'Retention: Lifetime ANNUAL RADIOACTIVE EFFLUENT REPORT 01-JAN-07 THROUGH 31--DEC-07ý'.,
SUPPLEMENTAL INFORMATION Prairie Island Nuclear Generating Plant Licensede: Northern States Power Company License Numbers: DPR-42 & DPR-60 A. Regulatory Limits 1 Liquid Effluent s:
a.'The & dose '0o.- dos'e commitment to an;individual fromdradioactive rnai: i*mater als in liquid effluents released from the site shall be limited to:
for t he-qua r t Or 3.0 mrem to theototal body 10.0 mrem to any organ for the year 6.0 mrem to the total body 20.0 mrem to any organ-
- 2. Gaseous Effluents:
- a. The dose rate due to radioactive materials released in gaseous effluents from the site shall be limited to:
noble gases < 500 mrem/year total body
<3000 mrem/year skin 1-131, 1-133, H-3, LLP <i500 mrem/year to any organ
- b. The dose due to radioactive gaseous effluents released from the site shall be limited to:
noble gases <10 mrad/quarter gamma
- 20 mrad/quarter beta
- 20 mrad/year gamma
- 40 mrad/year beta 1-131,1-133, H-3, !LLP *15 mrem/quarter, to any organ
<30 mrem/year to any organ
2007 Annual Radioactive Effluent Report Rev. 0 PAGE 2 B. Water Effluent Concentration
- 1. Fission and activation gases in gaseous releases:
10 CFR 20, Appendix B, Table 2, Column 1
- 2. Iodine and particulates with half lives greater than 8 days in gaseous releases:
J-0 CFR 20,, Appendix B,,. Table-.. 2-,,Column i-.
- 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.OE-04 uCi/mi Total Activity C. verage Energy Not applicable to Prairie Island regulatory limits'.
D. Measurements and approximations of total activity
- 1. Fission and activation gases Total Gem +/-25%
in gaseous releases: Nuclide Gem 2.. Iodines in gaseous releases: Total Gem +/-25%
Nuclide Gem
- 3. Particulates in gaseous releases: Total Gem +/-25%
. Nuclide Gem
- 4. Liquid effluents Total Gem +/-25%
Nucuclide Gem E. Manual Revisions
- 1. Offsite Dose Calculations Manual latest Revision number: 4-Revision date: 7
- 57
2007 ANNUAL RADIOACTIVE EFFLUENT REPORT REV. 0 PAGE 3 1.0 BATCH RELEASES (LIQUID)
QTR: 01ý QTR: 02 QTR: 03 QTR: 04 1.1 NUMBER OF BATCH RELEASES 4.10E+01 3..08+01 2.00E+01 1S.20E+01 1.2 TOTAL'ITIME PERIOD (HRS) 7.36B+01 6.63B+01 3.66E+01 9.18Eo-01 1.3 MAXIMUM TIME PERIOD (HRS) 2 . 92E+00 2.30'+00 2.25E+00 2 2.22E+00 1.4 AVERAGE TIME PERIOD (HRS) 1.80E+00 :.74E+00 1.83E+00 1.77E+00 1.5 MINIMUM TIME PERIOD (HRS) 1.48E+00 9.83E-01 1.52E+00 1.48E+00 1.6 AVERAGE MISSISSIPPI RIVER FLOW (CFS) 1.213+04 2.90E+04 7.24E+03 14.86E¶t04 2.0 BATCH RELEASES (AIRBORNE)
QTR: 01 QTR: 02 QTR: 03 QTR:.64 1 2.1 NUMBER OF BATCH RELEASES 8.OOE+00 t1.103+01 0.00E+00 0.OOE+00 2.2 TOTAL TIME PERIOD (HRS) 3.18E+01 1.42E+02 0.003+00 10.OOE+00 2.3 MAXIMUM TIME PERIOD (HRS) 1.14E+01 2.48E+01 0.00E+00 0.00E+01 2.4 AVERAGE TIME PERIOD (HRS) 3.98E+00- 1.29E+01 0.OOE+00 0.00E*001 2.5 MINIMUM TIME PERIOD (HRS). .1.43E-01' 3.33E-04 0.0-0E.00. 0. 003E00 3.0 ABNORMALREL BASES (LIQUID)
QTR: 01 QTR: 02 QTR: 03 QTR: 04 3.1 NUMBER ýOF BATCH RELEASES 0.00E+00 0.00E+00 0.OOE+00 0.OOE+00 3.2 TOTAL ACTIVITY RELEASED (CI) 0.00E+00 6.00E+00 0.00E+00 0.00E3,00 3.3 TOTAL -TRITIUM RELEASED (CI) o.003+00. O.OOE+00 0.OOE+00 0.00E-00 4.0 ABNORMAL RELEASES (AIRBORNE)
QTR: 01 QTR: 02 QTR: 03 4.1 NUMBER OF BATCH RELEASES 0.003+00 1.00E+00 0.00E+00 1. 00EI0oo 4.2 TOTAL ACTIVITY RELEASED (CI) 1.53E-05 j 0.00E+00 1.00E- 1 0.00E+00
2007 ANNUAL RADIOACTIVE EFFLUENT REPORT REV. 0 PAGE 4 TABLE IA GASEOUS EFFLUENTS - SUMMATION OF ALL RELEASES QTR: 01 QTR:02 QTR:'03 QTR: 4 5.0 FISSION AND ACTIVATION GASES 5.1. TOTAL RELEASE (CI) 0.OOE+0od 0.003+00 0.OOE+00 9.693E!02 5.2 AVERAGE RELEASE RATE (UCI/SEC)} 0.00E+00 0.003+00 0.OOE+00 1.263E-02 5.3 GAMMA DOSE (MRAD) 0.00E+00* *0.00E+00 O.OOE+00 6.573'106I 5.4 BETA DOSE (MRAD) 0.00E+00, 1 0.00-E+00 2.05E-104 0.00E+00 E0.003+00 0.00E+00 6.57E'05 5.5 PERCENT OF GAMMA TECH SPEC (%)
5.6 PERCENT OF BETA-TECH SPEC (%) 0.00E+00j 0.00E+00 0.OOE+00 1.03E303 L__ _ _J! _ _ _ I !___
6.0 IODINES 6.1 TOTAL 1-131 (CI)' o.ooE+00 0.002+00 0.00E+OO 0.00E-a0O
.6.2 AVERAGE RELEASE RATE (UCI/SEC) 0.00E+00 0.00E+00 0.OOE+00 0.00E+100 7.0 PARTICULATES 7.1 TOTAL RELEASE (CI) 4.178-06 0.o0o+00 o .OOE+00 [ 0.00Eoo 7.2 AVERAGE RELEASE RATE (UCI/SEC) 5.31E-07 0.00E+00 0.00H+00 0.00E+,00 I __ __ _ __ _ _ _L_ _ _ _ I _ I___
8.0 TRITIUM-8.1 TOTAL RELEASE (CI) 3.33E+0 3.35E+00. 2.73E+00 2.38E+00 8.2 AVERAGE RELEASE RATE (UCI/SEC) 4.23E-01 4.26E-01 3.47E-01 3.02'E-01 9.0 TOTAL IODINE, PARTICULATE AND TRITIUM (UCI/SEC) 4.23E-01 4 4,26E-01 3.47E-01 3.02E-,01 10.0 DOSE FROM IODINE, LLP, AND TRITIUM (MREM) 1.423-0.e.*r 9.22E-03 4.88E-03 4.253-03 11.0 PERCENT OF TECH SPEC (%) 9.50E-02 6.15E-02 3.253-02 2.83E-12.0 GROSS ALPHA (CI) o.OOE0+00 0.00E+00 0.003+00 0.00E+÷0 I I0
2007 ANNUAL RADIOACTIVE EFFLUENT REPORT REV. 0 PAGE 5 TABLE IC GASEOUS EFFLUENTS - GROUND LEVEL RELEASES (CI) 13.0 FISSION AND ACTIVAT ION GASES CONTINUOUS MODE BATCH MODE NUCLIDE UNITS] QTR: 01 QTR: .02 QTR: 03 QTR: 04 QTR: .01 QTR: 02 QTRi 03 QTR: 04 AR-41 CI [1_1.08E-04 K I ,I KR-85 CI 9.50E-02 II KR-85M CI _7.62E-05 XE-133 CI 2.44E-03 XE-135 CI 1.32E-03 0.00E÷00 0.ooE+00 0.OOE+00 TOTALS CI 0.OOE+00 0.0OE+00 0.OOE+00 9.89E-02 14.0 IODINES CONTINUOUS MODE BATCH MODE NUCLIDE UNITS QTR: 01 QTR: 02 1 QTR: 03 QTR: 04 QTR: 01 QTR: 02 QTR: 03 QTR: 04 TOTALS CI 0.OOE4-00 f0.009+00 0 OOE+00 0.OOEI-00 i.OOE+00 o.ooE+ O 0.OOE+00 0.OOE+00
2007 ANNUAL RADIOACTIVE EFFLUENT REPORT REV. 0 PAGE 6 TABLE iC GASEOUS EFFLUENTS - GROUND LEVEL RELEASES (CONTINUED) 15.0 PARTICULATES CONTINUOUS MODE BATCH NMODE NUCLIDE UNITS QTR: 01 QTR: 02 QTR: 03 QTR: 04 QTR: 01 QTR: 02 QTR:, 03 QTR: 04 CS-137 TOTALS CI CI 0.OOE+00 0.00E-400 0.OOE+00 0.00OE+00 4.'17-06 4.173-06 1
o.ooE+oo0 O.OOE0EO i 0.00E+00 E
2007 ANNUAL RADIOACTIVE EFFLUENT REPORT : REV. 0 PAGE ,7 TABLE IA LIQUID EFFLUENTS - SUMMATION OF ALL RELEASES QTR: 01 QTR:->02 QTRzO3 'QTR:: 04 I_
IQ iý 0:2 QTR 16.0 VOLUME OF WASTE PRIOR TO DILUTION .(.LITERS) 1. 2.95E+07 2.17E+07 2 32E"'07 17.0 VOLUME OF DILUTION WATER (LITERS;) 1. 84E+11 I.0E+11 2.57E+11-. 2.2-2Et11 18.0 FISSION AND ACTIVATION PRODUCTS
-18.1 TOTAL'RELEASES W/O H-3, ýRADGAS,, ALPHA (CI) r4.84E-02 1.2"4E-62 4.E03 4.90E..072 18.2 AVERAGE DILUTION CONCENTRATTON :(UCI/ML) 2.63E-10 1.18E-4o0 1.S9E-11 }2.20EI10 19.0 TRITIUM 19.1 TOTAL RELEASE (CI) .
i1.47E+02 j2.S3E+02 j8.42,E101 2. 36E+j02]
19.2 AVERAGE DILUTION CONCENTRATION ,(UCI/ML) 8.01E-07 J 2.41E-06 3.28E-07 1.06E106 20.0 DISSOLVED AND ENTRAINED GASES
- 20.1 TOTAL RELEASE (CI) 3.51E-06 4.OOE-05 j .OOE+00 8 8.73E5-04]
20.2 AVERAGE DILUTION CONCENTRATION (UCI/ML) 1.91E-14ý SI _
3.80OE-13
_ _ _ _I I _
.OOE+00
_ _ __I I3.*93E 12]
21.0 -GROSS ALPHA (CI) o.oo,.+oo I 0.0O0+0 0.00E+00 j 0.00E+ý0, 22.0 TOTAL TRITIUM, FISSION :& ACTIVATION PRODUCTS (UCI/ML) 8.OE-07 2.41E-06 3.28E-07 1.06E-,06 23.0 TOTAL BODY DOSE (MREM) 4.16E-*04 5.77E,,94 1.89E-04 5.35E-b-4 24.0 CRITICAL ORGAN 24.1 DOSE (EREM) 4-.16E-04 I5.77F,-0'4 1.89E-04 5.35E-, 4
.24.2 'ORGAN TOT BODY" TOT BODY TOT BODY TOT BODY 25.0 PERCENT OF TECHNICAL SPECIFICATIONS LIMIT (%) 1.39E-02 1.92E-02 6.31E-03 1.78E-162 26.0 PERCENT OF CRITICAL ORGAN TECH SPEC LIMIT (%) 1.39E-02 j 1.92E-02 I 6.31E-03 I 1.78E-b2
2007 ANNUAL RADIOACTIVE EFFLUENT REPORT REV. 0 PAGE 8 TABLE 2A LIQUID EFFLUNTB -- SUIOIATION OF ALL RELEASBS (CI) 27.0 INDIVIDUAL LIQUID EFFLUENT CONTINUOU S MODE BATCH MODE NUCLIDE UNITS. QTR: 01 QTR: 02. QTR:03 1QTR:
I ~ 04~ IIQTR:
II -: "..
- 01. . QTR: 02* QTR: 03 TR 0 AG-108M CI 4.1E-0 AG-11OM CI [
BA-139: CI 4 6.01E-03 T.58E-06 4.16E-03 1.50E-03.Ii 1.490 1.88E-03 CE-139 CI CO-57 { CI : 1.53E-06j 1.30-O~_________
. . I 1.84E-06
[
4{_37-3 1.55E-03 1.21E-04 2.32E___04.
CO-60 { CI . " '.. ' -
CR-51 CI " -
J.8E0 f.1.27Z-02 7 .98E-05 -1~.28Eý0 1.89E-04 CS-137 CI I 1.34E-05 FE-59 CI 144iO03 4.75E3 FE'5 - " CI :.-. 3. 03E-04 2.40E _ _ __ __ _ _05_ _
LA-140 [ CI 8,30E-06J MN-54 [CI . -- [678E-05 .4.62E-06 18E0 NA-24 CI _. I 9.29E0 S NB--95 NB-97 CI CI -
[7
'.14E-05 _____.
[ .56E-06 j7.60E-06 1.2-5 .3E0 SB-124 CI ".-_ . 2 . 2E-3ý.88-06 3.13E-05 97E0 SB-125 CI 1.31E02 5.7E-04 7.72E-04 4.0E 2 SN-113 " CI 7.7 -5 5.91E-06 ____ __I SR-85 CI [ 3 .23E-06 SR-92 CI " * [ 9.674.57 05 2.32 -05 2.15E-05 TE-123M CI-
.26
-04 6.53E-05j TE-1.25M CI- J " j.06Eý3- 1. 52E--03:
TE-132
- CI *2.1E0 TL-201- CI J_'_'L i0 5E-05 ]5*
(CONTINUED)
2007 ANNUAL RADIOACTIVE EF FLUENT REPORT :-REV. 0 PAGE 9 TABLE 2A LIQUID EFFLUENTS - SUMKATI ON OF ALL RELEASES (CONT INUED) 27.0 INDIVIDUAL LIQUID 9 FFLUEET CONTINU OUS MODE BATCH.-MODER NUCLIDE UNITS 0QTR: 01 QTR: 02 QTR: 03 1 QTR: 044 0R` -,1 QTR- 02 QTR.:, 03. QTR: 04 II I ZR-95 CI - '.1.74E-04 ZR-97 CI . ; 3.64-*06 5.96E-06 2.20E-06 4.60E-06 TOTALS CI 1.34E-05 0.0.0E+00]
o 0.005+00 0.00+E00 4.84E-02. 1.24E-02 4.08E-03 4.90E-02 28.0 DISSOLVED AND ENTRA INED GASES BATCH MODE CONTINUOUS MODE NUCLIDE UNITS QTR: 01 QTR:04 QTR:-02 II QTR, 03 I QTR*: 04 t I* QiR: 01. I QTR 0 QT-R ý jKR-85 {CI -t -I -I XE-133 CI 3.51E-06 3.71E-05 1.27E--04 XE-135 CI 2.88E-06 S3.S.E-06 ,-!4"00E-°0 O.OOE+00. 8.73E-04 TOTALS CI 0.00E+00 0. OOE+00. d.OOE+00 0.OOE+00
ENCLOSURE3 EFFLUENT AND WASTE DISPOSAL ANNUAL REPORT SOLID WASTE AND IRRADIATED FUEL SHIPMENTS January 01, 2007 - December 31, 2007 7 pages follow
PINGP 753, Rev. 7 Page 1 of 7 Retention: Life PRAIRIE ISLAND NUCLEAR GENERATING PLANT Period: O1/01/07-12/31/07 NORTHERN STATES POWER License No. DPR-42/60 EFFLUENT AND WASTE DISPOSAL ANNUAL REPORT SOLID WASTE AND IRRADIATED FUEL SHIPMENTS A. SOLID WASTE SHIPPED OFFSITE FOR BURIAL OR DISPOSAL (NOT IRRADIATED FUEL)
- 1. Solid Waste Total Volumes and Total Curie Quantities:
m- 1.UzL-+U 1 -I 1.4 ft 3 3.59E+02 Ci 5.59E-01 2.50E+01 3
B. Dry-Compacted ft Ci3 ft 3
C. Non-Compacted m3 2.90E+02 1280 ft 1.02E+04 Ci 8.08E-01 2.50E+01 D. Filter Media ft 3
Ci m3 S. Other (furnish description) 7.25E+01 1280 Combined DAW/Charcoal/Grit 2.56E+03 Ci 6.35E-02 2.50E+01 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 disposal.
Document2
PINGP 753, Rev. 7 Page 2 of 7 PRAIRIE ISLAND NUCLEAR GENERATING PLANT Period: 01-01-07/12-31-07 NORTHERN STATES POWER License No. DPR-42/60 EFFLUENT AND WASTE DISPOSAL ANNUAL REPORT SOLID WASTE AND IRRADIATED FUEL SHIPMENTS A. SOLID WASTE SHIPPED OFFSITE FOR BURIAL OR DISPOSAL (NOT IRRADIATED FUEL) [continued]
- 2. Principal Radionuclide Composition by Type of Waste:
(Bold letter designation from Page 1)
TYPE Percent %
Abundance Nuclide (0.00E0)
C *Fe-55 6.35E+01 Co-58 8.85E+00 Co-60 8.07E+00
- Ni-63 1.19E+01 Zr-95 2.01 E+00 Nb-95 1.59E+00 1% cutoff S *Fe-55 2.77E+01 Co-58 4.40E+00 Co-60 4.96E+00 Nb-95 1.16E+00 Zr-95 1.03E+00
- C-14 7.85E+00
- Ni-63 1.58E+01
- H-3 3.50E+01 1% cutoff
- = Inferred - Not Measured on Site
PINGP 753, Rev. 7 Page 3 of 7 PRAIRIE ISLAND NUCLEAR GENERATING PLANT Period: 01-01-07/12-31-07 NORTHERN STATES POWER License No. DPR-42/60 EFFLUENT AND WASTE DISPOSAL ANNUAL REPORT SOLID WASTE AND IRRADIATED FUEL SHIPMENTS A. SOLID WASTE SHIPPED OFFSITE FOR BURIAL OR DISPOSAL (NOT IRRADIATED FUEL) [continued]
- 2. Principal Radionuclide Composition by Type of Waste (Continuation):
(Bold letter designation from Page 1)
TYPE Percent %
Abundance Nuclide (0.00E0)
A *H-3 1.18E+00
- Fe-55 3.48E+01
- Ni-63 3.53E+01 Co-60 1.33E+01 Co-58 1.14E+00 Cs-1 37 3.83E+00 C-14 3.51E+00 Sb-125 5.61 E+00 1% cutoff
- = Inferred - Not Measured on Site
PINGP 753, Rev. 7 Page 4 of 7 PRAIRIE ISLAND NUCLEAR GENERATING PLANT Period: 01-01-07/12-31-07 NORTHERN STATES POWER License No. DPR-42/60 EFFLUENT AND WASTE DISPOSAL ANNUAL REPORT SOLID WASTE AND IRRADIATED FUEL SHIPMENTS A. SOLID WASTE SHIPPED OFFSITE FOR BURIAL OR DISPOSAL (NOT IRRADIATED FUEL) [continued]
- 3. Solid Waste Disposition:
Number of Shipments Mode Destination 5 StudsvikLogistics StudsvikRACE, LLC 2 Hittman Transport Studsvik Processing Facility, Services LLC
PINGP 753, Rev. 7 Page 5 of 7 PRAIRIE ISLAND NUCLEAR GENERATING PLANT Period: 01-01-07/12-31-07 NORTHERN STATES POWER License No. DPR-42/60 EFFLUENT AND WASTE DISPOSAL ANNUAL REPORT SOLID WASTE AND IRRADIATED FUEL SHIPMENTS A. SOLID WASTE SHIPPED OFFSITE FOR BURIAL OR DISPOSAL (NOT IRRADIATED FUEL) [continued]
- 4. Shipping Container and Solidification Method:
Disposal No. Volume Activity Type of Container Solidif.
(Ft 31m3 ) (Ci) Waste Code Code 07-004 2560/72.5 0.390 C L N/A 07-005 2560/72.5 0.141 C L N/A 07-006 2560/72.5 0.226 C L N/A 07-011 179.4/5.1 0.296 A L N/A 07-012 179.4/5.1 0.263 A L N/A 07-018 2560/72.5 0.065 C L N/A 07-019 2560/72.5 0.049 C L N/A TOTAL 7 13200/373 1.43 S
CONTAINER CODES: L = LSA (Shipment type) A = Type A B = Type B Q = Highway Route Controlled Quantity SOLIDIFICATION CODES: C = Cement TYPES OF WASTES:, A = Resins B = Dry Compacted C = Non-Compacted D = Filter Media S = Other
PINGP 753, Rev. 7 Page 6 of 7 PRAIRIE ISLAND NUCLEAR GENERATING PLANT Period: 01-01-07/12-31-07 NORTHERN STATES POWER License No. DPR-42/60 EFFLUENT AND WASTE DISPOSAL ANNUAL REPORT SOLID WASTE AND IRRADIATED FUEL SHIPMENTS B. IRRADIATED FUEL SHIPMENTS (DISPOSITION)
Number of Shipments Mode Destination 0
PINGP 753, Rev. 7 Page 7 of 7 PRAIRIE ISLAND NUCLEAR GENERATING PLANT Period: 01-01-07/12-31-07 NORTHERN STATES POWER License No. DPR-42/60 EFFLUENT AND WASTE DISPOSAL ANNUAL REPORT SOLID WASTE AND IRRADIATED FUEL SHIPMENTS C. PROCESS CONTROL PROGRAM CHANGES TITLE: Process Control for Solidification/Dewatering of Radioactive Waste from Liquid Systems Current Revision Number: 8 Effective Date: 8/25/1999 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.6.5.D) (IT.S.5A4). Attach the sidelined pages to this report.
Changes/Justification: N/A
ENCLOSURE 4 ATTATCHMENT TO THE 2007 EFFLUENT REPORT Description and Dose Assessment of Quarter 1 of 2007 Leak Communicated per ODCM (H4, Section 8.4) Industry Initiative on Groundwater Protection 1 page follows
ATTACHMENT TO THE 2007 ANNUAL EFFLUENT REPORT 8.4) Industry Initiative on Groundwater Protection Quarter 1, 2007 Summary This dose assessment for the leak described in this attachment did not change the total liquid dose to the critical receptor for the first quarter of 2007.
Backgiround On March 21, 2007, approximately 150 gallons of secondary steam condensate leaked to the ground outside the northeast side of the turbine building during transfer of turbine building sump water to the landlocked canal. This water had a tritium concentration of 5,150 pCi/L. The transfer of water was secured and no further leakage to the environment occurred. This release occurred from a monitored release path.
ODCM Considerations Since the release occurred via a monitored release path, the dose calculations were performed per the ODCM. The released effluent did not flow to the normal landlocked location but was absorbed into the ground closer to the turbine building. At this location, the water has farther to travel to the receptor and is subject to greater dispersion than the normal release location. The ODCM dose calculations over-estimate the dose from this release. Corrective actions have been taken to prevent a similar spill in the future.
Dose Calculation Assumptions The dose calculation was performed per the ODCM for the annual effluent report. This dose over-estimates the dose under the conditions of this release. No revision to the ODCM dose calculation is warranted.
Discussion The critical receptor is located 0.6 miles to the SSE of the Prairie Island site. The leaked water would have to travel in the groundwater under the recycle canal and discharge canal to reach the critical receptor. This assumed water flow maximizes the dose because the normal groundwater flow is towards the Vermillion River which would not carry the tritium toward the critical receptor.