ML25132A133
| ML25132A133 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 05/10/2025 |
| From: | Currier B Northern States Power Company, Minnesota, Xcel Energy |
| To: | Office of Nuclear Reactor Regulation, Document Control Desk |
| References | |
| L-Pl-25-015 | |
| Download: ML25132A133 (1) | |
Text
fl Xcel Energy May 10, 2025 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 Prairie Island Nuclear Generating Plant, Units 1 and 2 Docket Nos. 50-282 and 50-306 Renewed Facility Operating License Nos. DPR-42 and DPR-60 2024 Annual Radioactive Effluent Report 1717 Wakonade Drive Welch, MN 55089 L-Pl-25-015 10 CFR 50.36a Tech Spec 5.6.3 Pursuant to 10 CFR 50.36a, "Technical specifications on effluents from nuclear power reactors," paragraph (a)(2), and in accordance with Prairie Island Nuclear Generating Plant (PINGP) Technical Specification (TS) 5.6.3 "Radioactive Effluent Report," the Northern States Power Company (NSPM), a Minnesota corporation, d/b/a Xcel Energy, is submitting the following enclosures: - Radioactive Effluent Report. - Radioactive Effluent Report, Supplemental Information. - D59 PROCESS CONTROL PROGRAM FOR PROCESSING /
DEWATERING OF RADIOACTIVE WASTE FROM LIQUID SYSTEMS Summary of Commitments This le ter
- e,::
commitments and no revisions to existing commitments.
Bryan C
- rier Plant Manager, Prairie Island Nuclear Generating Plant Northern States Power Company - Minnesota Enclosures (3)
Document Control Desk L-Pl-25-015 Page 2 cc:
Administrator, Region 111, USNRC Project Manager, Prairie Island, USNRC Resident Inspector, Prairie Island, USNRC Department of Health, State of Minnesota Pl Dakota Community Environmental Coordinator
ENCLOSURE 1 RADIOACTIVE EFFLUENT REPORT JANUARY 1, 2024 - DECEMBER 31, 2024 8 pages to follow
Page 1 of 8 PRAIRIE ISLAND NUCLEAR GENERATING PLANT OFF-SITE RADIATION DOSE ASSESSMENT FOR January 1, 2024 - December 31, 2024 An Assessment of the 2024 radiation dose, due to operation of The Prairie Island Nuclear Generating Plant, was performed in accordance with the Offsite Dose Calculation Manual, and 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 historical meteorological data were used in making this assessment. Source terms were obtained from the Annual Radioactive Effluent and Waste Disposal Report and prepared for NRC review, for the year of 2024.
OFFSITE 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.
OFFSITE 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 Boundary 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 and beta doses from noble gas releases and the maximum organ doses from the inhalation pathway due to Iodine 131, lodine-133, tritium, long-lived particulates and Carbon-14 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.
Page 2 of 8 40 CFR 190 COMPLIANCE:
REMP environmental TLD results for 2024 were reviewed per ANSI/HPS N13.37-2014 methodology for determining any plant effect above ambient gamma radiation measurements. All measurements are within the range of variations in natural background radiation.
Neutron sky shine dose from the ISFSI was evaluated. The maximum neutron sky shine dose was determined to be 1.13 mrem, to the nearest resident, at 724 meters from the ISFSI. Neutron sky shine dose is greater than the effluent dose to the Critical Receptor, therefore, 40 CFR190 compliance was evaluated to the location of the maximum neutron sky shine dose.
The 40 CFR 190 evaluation location was determined to be 0.7 miles west of the plant.
Dose due to gaseous effluents was calculated to the 40 CFR 190 evaluation location.
Gamma Direct Radiation Dose:
Neutron Sky Shine Dose:
Noble Gas Gamma Dose:
Noble Gas Beta Dose:
Iodine, particulate, H-3 and C-14 Dose:
MREM 0.00E+00 1.13E+00 3.76E-06 5.90E-06 5.34E-03*
- Calculated values were identical for Whole Body, Thyroid and Maximum "Other" Organs SUMMATION OF 40 CFR 190 DOSE:
40 CFR 190 LIMIT (MREM) 40 CFR 190 DOSE (MREM)
WHOLE BODY 25 1.14E+00 THYROID 75 1.14E+00 OTHER ORGANS (TEEN - WHOLE BODY) 25 1.14E+00
Page 3 of 8 ABNORMAL RELEASES:
There were zero (0) abnormal releases in 2024.
SAMPLING, ANALYSIS AND LLD REQUIREMENTS:
The lower limit of detection (LLD) requirements, as specified in ODCM Tables 2.1 and 3.1, were met for 2024. The minimum sampling frequency requirements, as specified in ODCM Tables 2.1 and 3.1, were met for 2024.
MONITORING INSTRUMENTATION:
For 2024, there were four (4) occurrences when less than the minimum required radioactive liquid and/or gaseous effluent monitoring instrumentation channels were functional, as required by ODCM Tables 2.2 and 3.2.
1R30, Unit 1 Auxiliary Building Stack Monitor Flow Integrator 1R30, Unit 1 Auxiliary Building Stack Monitor Flow Integrator was out of service from 10/30/2023 06:00 to 4/23/2024 07:50. Total time out of service was 176.1 days.
The ODCM Table 3.2 contingency action is that effluent releases via this pathway may continue provided that the flow rate is estimated at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
Contingency actions were performed as prescribed.
1R30, Unit 1 Auxiliary Building Stack Monitor Flow Integrator was not restored within a timely manner due to resource loading.
2R12, Unit Two Shield Building Stack Monitor 2R12, Unit Two Shield Building Stack Monitor was out of service from 3/6/2024 13:15 to 5/20/2024 11:04. Total time out of service was 74.9 days.
2R22, Unit Two Shield Building Stack Monitor, is a redundant monitor to 2R12. 2R22 remained in service during the period that 2R12 was not functional. No contingency actions were required.
2R12, Unit Two Shield Building Stack Monitor was not restored within a timely manner due to resource loading.
Page 4 of 8 1R37, Unit One Auxiliary Building Stack Monitor 1R37, Unit One Auxiliary Building Stack Monitor was out of service from 10/19/23 18:56 to 10/14/2024 14:30. Total time out of service was 360.8 days.
1R30, Unit 1 Auxiliary Building Stack Radiation Monitor, is a redundant monitor to 1R37, Unit One Auxiliary Building Stack Radiation Monitor. 1R30 remained in service during the period that 1R37 was not functional. No contingency actions were required.
1R37, Unit One Auxiliary Building Stack Monitor, parts availability delayed restoring the monitor in a timely manner.
22 Steam Generator Blowdown Flow 22 Steam Generator Blowdown Flow was out of service from 10/23/2024 15:44 to 12/15/2024 03:20. Total time out of service was 52.5 days. Maintenance schedule did not satisfy ODCM requirements for calibration frequency. Maintenance schedule was corrected.
The failure to perform the required calibration was not noted until late in the out of service period. Therefore, contingency actions were not performed. The ODCM Table 2.2 contingency action is to perform a flow estimate at least every four hours, during periods of release.
During the out of service period, Unit Two Steam Generator Blowdown flow was directed to the river twice for a total of 352 hours0.00407 days <br />0.0978 hours <br />5.820106e-4 weeks <br />1.33936e-4 months <br />.
22 Steam Generator Blowdown Flow was not restored in a timely manner, because the site was not aware of the scheduling error.
DOSES TO INDIVIDUALS DUE TO EFFLUENT RELEASES FROM THE INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI):
Two (2) fuel casks were loaded and placed in the ISFSI during the 2024 calendar year. The total number of casks in the ISFSI, as of 12/31/2024, was fifty-two (52). There were zero (0) releases of radioactive effluents from the ISFSI.
CURRENT OFFSITE DOSE CALCULATIONS MANUAL (ODCM) REVISION:
The Offsite Dose Calculation Manual (ODCM) WAS NOT revised in 2024. The Current revision is Revision 33, issued on 3/24/2023.
The Offsite Dose Calculation Manual (ODCM) Supporting Data WAS NOT revised in 2024. The current revision is Revision 8, issued on 3/24/2023.
Page 5 of 8 PROCESS CONTROL PROGRAM:
D59, The Process Control Program for Solidification/Dewatering of Radioactive Waste from Liquid Systems, WAS revised in 2024. Current revision is Revision 13, issued 3/8/2024.
INDUSTRY INITIATIVE ON GROUND WATER PROTECTION:
For 2024, there was zero (0) events for inclusion in the Annual Effluent Report, as part of the NEI Ground Water Initiative.
CRITICAL RECEPTOR:
Based on the Annual Land Use Census, the critical receptor did not change. The critical receptor is defined as The Suter Residence, at 0.6 miles, in the SSE sector.
Page 6 of 8 LOW LEVEL WASTE DISPOSAL ANNUAL REPORT SOLID WASTE AND IRRADIATED COMPONENTS SHIPMENTS PERIOD: 1/1/24 TO 12/31/24 LICENSE NUMBER: DPR-42/60 SOLID WASTE SHIPPED OFFSITE FOR BURIAL OR DISPOSAL (NOT IRRADIATED FUEL):
Resins, Filters and Evaporator Bottoms Volume Curies Shipped Waste Class ft3 m3 Curies A
0.00E+00 0.00E+00 0.00E+00 B
0.00E+00 0.00E+00 0.00E+00 C
0.00E+00 0.00E+00 0.00E+00 ALL 0.00E+00 0.00E+00 0.00E+00 Major Nuclides NA Dry Active Waste Volume Curies Shipped Waste Class ft3 m3 Curies A
2.56E+04 7.25E+02 5.18E-01 B
0.00E+00 0.00E+00 0.00E+00 C
0.00E+00 0.00E+00 0.00E+00 ALL 2.56E+04 7.25E+02 5.18E-01 Major Nuclides H-3, C-14, Cr-51, Mn-54, Fe-55, Fe-59, Co-58, Co-60, Ni-63, Sr-90, Zr-95, Nb-94, Nb-95, Tc-99, Ag-110m, Sn-117m, Sb-125, I-129, Cs-137, Ce-144, Hf-181, Pu-238, Pu-239, Pu-240, Pu-241, Am-241, Cm-242, Cm-243, Cm-244 Irradiated Components Volume Curies Shipped Waste Class ft3 m3 Curies A
0.00E+00 0.00E+00 0.00E+00 B
0.00E+00 0.00E+00 0.00E+00 C
0.00E+00 0.00E+00 0.00E+00 ALL 0.00E+00 0.00E+00 0.00E+00 Major Nuclides N/A Other Waste Volume Curies Shipped Waste Class ft3 m3 Curies A
1.28E+03 3.62E+01 8.25E-02 B
0.00E+00 0.00E+00 0.00E+00 C
0.00E+00 0.00E+00 0.00E+00 ALL 1.28E+03 3.62E+01 8.25E-02 Major Nuclides H-3, C-14, Cr-51, Mn-54, Fe-55, Co-58, Co-60, Ni-63, Sr-90, Zr-95, Nb-94, Nb-95, Tc-99, Sb-125, I-129, Cs-137, Ce-144, Pu-238, Pu-239, Pu-240, Pu-241, Am-241, Cm-242, Cm-243, Cm-244 Sum of All Low Level Waste Shipped from Site Volume Curies Shipped Waste Class ft3 m3 Curies A
2.69E+04 7.61E+02 6.00E-01 B
0.00E+00 0.00E+00 0.00E+00 C
0.00E+00 0.00E+00 0.00E+00 ALL 2.69E+04 7.61E+02 6.00-01 Major Nuclides H-3, C-14, Mn-54, Fe-55, Fe-59, Co-58, Co-60, Ni-63, Sr-90, Zr-95, Nb-94, Nb-95, Tc-99, Ag-110m, Sn-117m, Sb-125, I-129, Cs-137, Ce-144, Hf-181, Pu-238, Pu-239, Pu-240, Pu-241, Am-241, Cm-242, Cm-243, Cm-244 Total curie quantity and principal radionuclides identification are calculated estimates determined for packaged waste using gross gamma radiation measurements, direct sample data or swipe data within WMGs Radman Suite Software.
Characterization of radioactive waste is performed in accordance with 10 CFR 20, 10 CFR 61, and NRCs Branch Technical Positions.
Page 7 of 8 Table 1 OFF-SITE RADIATION DOSE ASSESSMENT JANUARY 2024 THROUGH DECEMBER 2024 DOSE LIMIT*
Gaseous Releases Maximum Site Boundary Gamma Air Dose (mrad) 1.08E-05 20 Maximum Site Boundary Beta Air Dose (mrad) 7.33E-06 40 Maximum Off-site Dose to any Organ (mrem)**
6.05E-02 30 Organ Child - bone Offshore Location Maximum Site Boundary Gamma Air Dose (mrad) 2.17E-07 20 Maximum Site Boundary Beta Air Dose (mrad) 1.47E-07 40 Maximum Off-site Dose to any Organ (mrem)**
9.51E-04 30 Organ Teen - Total Body Liquid Releases Maximum Off-site Dose Total Body (mrem) 4.09E-03 6
Maximum Off-site Dose to any Organ (mrem) 4.93E-03 20 Organ Adult - Gi-LLi
- 10 CFR part 50, Appendix I Guidelines (2-unit site per year)
Page 8 of 8 Table 2 OFF-SITE RADIATION DOSE ASSESSMENT-PRAIRIE ISLAND SUPPLEMENTAL INFORMATION January 1, 2024 - December 31, 2024 Gaseous Releases Maximum Site Boundary Dose Location (From Building Vents)
Sector W
Distance (miles) 0.36 Offshore Location Within Site Boundary Sector ESE Distance (miles) 0.2 Pathway Inhalation Critical Receptor Location Sector SSE Distance (miles) 0.60 Pathways Ground Inhalation Vegetable Age Group Child Liquid Releases Maximum Off-Site Dose Location Sector SSE Distance (miles) 0.43 Pathway Fish
ENCLOSURE 2 RADIOACTIVE EFFLUENT REPORT SUPPLEMENTAL INFORMATION JANUARY 1, 2024 - DECEMBER 31, 2024 8 pages to follow
Page 1 of 8 ANNUAL RADIOACTIVE EFFLUENT REPORT SUPPLEMENTAL INFORMATION 01-JAN-24 THROUGH 31-DEC-24 Facility:
Licensee:
Prairie Island Nuclear Generating Plant Northern States Power Company License Numbers: DPR-42 & DPR-60 A. Regulatory Limits
- 1.
Liquid Effluents:
- a.
The dose or dose commitment to an individual from radioactive materials in liquid effluents released from the site shall be limited to:
for the quarter 3.0 mrem to the total body 10.0 mrem to any organ for the year 6.0 mrem to the total body 20.0 mrem to any organ
- 2.
Gaseous Effluents:
- a.
The dose rate due to radioactive materials released in gaseous effluents from the site shall be limited to:
noble gases 500 mrem/year total body 3000 mrem/year skin I-131, I-133, H-3, LLP, C-14 l500 mrem/year to any organ
- b.
The dose due to radioactive gaseous effluents released from the site shall be limited to:
noble gases l0 mrad/quarter gamma 20 mrad/quarter beta 20 mrad/year gamma 40 mrad/year beta I-131, I-133, H-3, LLP, 15 mrem/quarter to any organ C-14 30 mrem/year to any organ
Page 2 of 8 B.
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:
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:
Offsite Dose Calculation Manual C.
Average Energy Not applicable to Prairie Island regulatory limits.
D.
Measurements and approximations of total activity
- 1.
Fission and activation gases Total HPGe in gaseous releases:
Nuclide
- 2.
Iodines in gaseous releases:
Total HPGe Nuclide
- 3.
Particulates in gaseous releases:
Total HPGe Nuclide
- 4.
Liquid effluents Total HPGe Nuclide
+/-25%
+/-25%
+/-25%
+/-25%
E.
Manual Revisions Offsite Dose Calculations Manual (ODCM):
Latest Revision number:
33 Revision date:
March 24, 2023 Offsite Dose Calculations Manual (ODCM) Supporting Data:
Latest Revision number: 8 Revision date:
March 24, 2023
Page 3 of 8 Prairie Island Nuclear Generating Station PI 2024 Annual Release Summary Batch Release Summary Liquid Releases Qtr 1 Qtr 2 Qtr 3 Qtr 4 Year Number of Releases:
39 40 40 58 177 Total Time for All Releases (Minutes):
2818.0 2927.0 3087.0 4836.0 13668.0 Maximum Time for All Releases (Minutes):
102.0 102.0 120.0 230.0 230.0 Average Time for All Releases (Minutes):
72.3 73.2 77.2 83.4 77.2 Minimum Time for All Releases (Minutes):
4.0 57.0 60.0 27.0 4.0 Gaseous Releases Qtr 1 Qtr 2 Qtr 3 Qtr 4 Year Number of Releases:
61 1
7 67 136 Total Time for All Releases (Minutes):
213231.0 131040.0 4384.0 220156.0 568811.0 Maximum Time for All Releases (Minutes):
131040 131040.0 1076.0 132480.0 132480.0 Average Time for All Releases (Minutes):
3495.6 131040.0 626.3 3285.9 4182.4 Minimum Time for All Releases (Minutes):
135.0 131040.0 348.0 455.0 135.0 Abnormal Release Summary Liquid Releases Number of Abnormal Releases:
0 Total Activity for Abnormal Releases:
0.00E+00 Curies Gaseous Releases Number of Abnormal Releases:
0 Total Activity for Abnormal Releases:
0.00E+00 Curies
Page 4 of 8 Prairie Island Nuclear Generating Station PI 2024 Annual Release Summary Gaseous Effluents-Summation of All Releases Type of Effluent Units Qtr 1 Qtr 2 Qtr 3 Qtr 4 Est. Total Error, %
A. Fission & Activation Gases
- 1. Total Release
- 2. Average Release Rate for Period
- 3. Percent of Applicable Limit Curies Ci/sec 1.66E-03 2.49E-03 1.63E-03 1.28E-03 2.50E+01 2.12E-04 3.17E-04 2.06E-04 1.60E-04 6.03E-06 9.08E-06 9.41E-05 4.61E-06 B. Iodines
- 1. Total Iodine-131 Curies 0.00E+00 0.00E+00 0.00E+00 0.00E+00 2.50E+01
- 2. Average Release Rate for Period Ci/sec 0.00E+00 0.00E+00 0.00E+00 0.00E+00
- 3. Percent of Applicable Limit 0.00E+00 0.00E+00 0.00E+00 0.00E+00 C. Particulates
- 1. Total Particulates (Half-lives > 8 days)
Curies 3.37E-10 0.00E+00 8.46E-05 4.05E-06 2.50E+01
- 2. Average Release Rate for Period Ci/sec 4.29E-11 0.00E+00 1.06E-05 5.10E-07
- 3. Percent of Applicable Limit 6.38E-09 0.00E+00 9.02E-03 2.77E-05
- 4. Gross Alpha Activity Curies 0.00E+00 0.00E+00 0.00E+00 0.00E+00 2.50E+01 D. Tritium
- 1. Total Release Curies 1.25E+01 9.78E+00 1.54E+01 2.42E+01 2.50E+01
- 2. Average Release Rate for Period Ci/sec 1.60E+00 1.24E+00 1.94E+00 3.04E+00
- 3. Percent of Applicable Limit 2.54E-02 1.80E-02 3.07E-02 4.46E-02 E. Carbon-14
- 1. Total Release Curies 1.22E+00 2.69E+00 2.67E+00 1.23E+00 2.50E+01
Page 5 of 8 Prairie Island Nuclear Generating Station PI 2024 Annual Release Summary Gaseous Effluents - Ground Level Releases Continuous Mode Batch Mode Nuclides Released Units Qtr 1 Qtr 2 Qtr 3 Qtr 4 Qtr 1 Qtr 2 Qtr 3 Qtr 4
- 1.
Fission and Activation Gases Ar-41 Curies 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 1.51E-03 0.00E+00 Xe-133 Curies 0.00E+00 0.00E+00 0.00E+00 0.00E+00 1.62E-03 2.41E-03 1.19E-04 1.24E-03 Xe-135 Curies 0.00E+00 0.00E+00 0.00E+00 0.00E+00 4.67E-05 7.96E-05 0.00E+00 3.20E-05 Total for Period Curies 0.00E+00 0.00E+00 0.00E+00 0.00+00 1.66E-03 2.49E-03 1.63E-03 1.28E-03
- 2. Iodines I-131 Curies 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Total For Period Curies 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00
- 3. Particulates Cd-109 Curies 3.04E-10 0.00E+00 2.29E-06 8.00E-07 0.00E+00 0.00E+00 8.23E-05 0.00E+00 Co-58 Curies 3.35E-11 0.00E+00 0.00E+00 2.77E-06 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Co-60 Curies 0.00E+00 0.00E+00 5.34E-09 2.32E-07 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Mn-54 Curies 0.00E+00 0.00E+00 0.00E+00 2.53E-07 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Nd-147 Curies 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 1.42E-10 0.00E+00 Total for Period Curies 3.37E-10 0.00E+00 2.30E-06 4.05E-06 0.00E+00 0.00E+00 8.23E-05 0.00E+00
- 4.
Tritium H-3 Curies 1.22E+01 9.77E+00 1.51E+01 2.41E+01 3.38E-01 6.32E-03 3.35E-01 4.31E-02
Page 6 of 8 Prairie Island Nuclear Generating Station PI 2024 Annual Release Summary Liquid Effluents - Summation of All Releases Type of Effluent Units Qtr 1 Qtr 2 Qtr 3 Qtr 4 Est. Total Error, %
A. Fission & Activation Products
- 1. Total Release (not including Tritium, Curies 2.04E-03 2.41E-04 5.41E-03 1.02E-02 2.50E+01 Gases, and Alpha)
- 2. Average Diluted Concentration During Ci/ml 2.31E-11 2.87E-12 6.27E-11 1.19E-10 Period
- 3. Percent of Applicable Limit 4.08E-02 4.83E-03 1.08E-01 2.04E-01 B. Tritium
- 1. Total Release
- 2. Average Diluted Concentration During Period Curies Ci/ml 2.29E+02 1.38E+02 1.78E+02 2.24E+02 2.50E+01 2.59E-06 1.65E-06 2.06E-06 2.63E-06
- 3. Percent of Applicable Limit 2.59E-01 1.65E-01 2.06E-01 2.63E-01 C. Dissolved and Entrained Gases
- 1. Total Release Curies 2.62E-05 2.94E-05 6.58E-05 1.58E-05 2.50E+01
- 2. Average Diluted Concentration During Period
- 3. Percent of Applicable Limit Ci/ml 2.96E-13 3.49E-13 7.62E-13 1.85E-13 1.48E-07 1.75E-07 3.81E-07 9.27E-08 D. Gross Alpha Radioactivity
- 1. Total Release Curies 0.00E+00 0.00E+00 0.00E+00 0.00E+00 2.50E+01 E. Waste Volume Released (Pre-Dilution)
Liters 5.93E+07 5.04E+07 5.93E+07 3.79E+07 2.50E+01 F. Volume of Dilution Water Used Liters 8.83E+10 8.40E+10 8.62E+10 8.52E+10 2.50E+01
Page 7 of 8 Prairie Island Nuclear Generating Station PI 2024 Annual Release Summary Liquid Effluents Continuous Mode Batch Mode Nuclides R Eleased Units Qtr 1 Qtr 2 Qtr 3 Qtr 4 Qtr 1 Qtr 2 Qtr 3 Qtr 4 Ag-110m Curies 0.00E+00 0.00E+00 0.00E+00 0.00E+00 6.07E-06 0.00E+00 1.30E-04 7.37E-04 Co-57 Curies 0.00E+00 0.00E+00 0.00E+00 0.00E+00 3.24E-06 2.68E-06 3.05E-05 4.13E-05 Co-58 Curies 0.00E+00 0.00E+00 0.00E+00 0.00E+00 9.32E-04 1.59E-04 1.26E-04 1.28E-03 Co-60 Curies 0.00E+00 0.00E+00 0.00E+00 0.00E+00 1.48E-04 4.81E-05 3.14E-03 3.04E-03 Cr-51 Curies 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 5.32E-05 4.77E-04 Fe-55 Curies 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 1.76E-03 3.87E-03 Fe-59 Curies 0.00E+00 0.00E+00 0.00E+00 0.00E+00 4.72E-06 0.00E+00 0.00E+00 1.73E-05 H-3 Curies 3.26E-03 1.75E-02 1.26E-01 8.18E-02 2.29E+02 1.38E+02 1.78E+02 2.24E+02 La-140 Curies 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 1.28E-06 2.83E-06 Mn-54 Curies 0.00E+00 0.00E+00 0.00E+00 0.00E+00 9.59E-07 0.00E+00 4.37E-05 9.15E-05 Nb-95 Curies 0.00E+00 0.00E+00 0.00E+00 0.00E+00 5.65E-07 0.00E+00 4.70E-06 1.63E-04 Nb-97 Curies 0.00E+00 0.00E+00 0.00E+00 0.00E+00 4.75E-06 3.69E-06 2.48E-05 1.63E-05 Sb-124 Curies 0.00E+00 0.00E+00 0.00E+00 0.00E+00 5.24E-06 0.00E+00 0.00E+00 0.00E+00 Sb-125 Curies 0.00E+00 0.00E+00 0.00E+00 0.00E+00 8.67E-04 1.15E-05 5.93E-05 2.01E-04 Sn-113 Curies 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 8.35E-06 4.12E-05 Sr-92 Curies 0.00E+00 0.00E+00 0.00E+00 0.00E+00 6.25E-07 3.91E-07 5.54E-06 1.81E-06 Te-123M Curies 0.00E+00 0.00E+00 0.00E+00 0.00E+00 6.88E-05 1.62E-05 4.23E-06 8.46E-05 Xe-133 Curies 0.00E+00 0.00E+00 0.00E+00 0.00E+00 2.62E-05 2.50E-05 6.58E-05 1.58E-05 Xe-135 Curies 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 4.41E-06 0.00E+00 0.00E+00 Zn-65 Curies 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 9.42E-06 Zr-95 Curies 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 1.02E-05 1.02E-04 2024 TOTAL Curies 3.26E-03 1.75E-02 1.26E-01 8.18E-02 2.29E+02 1.38E+02 1.78E+02 2.24E+02
Page 8 of 8 Prairie Island Nuclear Generating Station 2024 Annual Dose Summary Gaseous Effluents Parameter Location Dose Dose Limit
% of Limit Qtr 1 Gamma Air Dose (mrad) 0.58 km W 4.40E-07 1.00E+01 0.00 Beta Air Dose (mrad) 0.58 km W 1.21E-06 2.00E+01 0.00 Total Body Dose (mrem) 0.58 km W 3.73E-07 5.00E+00 0.00 Skin Dose (mrem) 0.58 km W 8.71E-07 1.50E+01 0.00 Max Organ Dose (mrem) 0.97 km SSE 3.81E-03 1.50E+01 0.03 Child - Kidney Qtr 2 Gamma Air Dose (mrad) 0.58 km W 6.68E-07 1.00E+01 0.00 Beta Air Dose (mrad) 0.58 km W 1.82E-06 2.00E+01 0.00 Total Body Dose (mrem) 0.58 km W 5.68E-07 5.00E+00 0.00 Skin Dose (mrem) 0.58 km W 1.32E-06 1.50E+01 0.00 Max Organ Dose (mrem) 0.97 km SSE 2.41E-02 1.50E+01 0.16 Child - Bone Qtr 3 Gamma Air Dose (mrad) 0.58 km W 9.41E-06 1.00E+01 0.00 Beta Air Dose (mrad) 0.58 km W 3.39E-06 2.00E+01 0.00 Total Body Dose (mrem) 0.58 km W 8.94E-06 5.00E+00 0.00 Skin Dose (mrem) 0.58 km W 1.31E-05 1.50E+01 0.00 Max Organ Dose (mrem) 0.97 km SSE 3.64E-02 1.50E+01 0.24 Child - Bone Qtr 4 Gamma Air Dose (mrad) 0.58 km W 3.33E-07 1.00E+01 0.00 Beta Air Dose (mrad) 0.58 km W 9.21E-07 2.00E+01 0.00 Total Body Dose (mrem) 0.58 km W 2.82E-07 5.00E+00 0.00 Skin Dose (mrem) 0.58 km W 6.59E-07 1.50E+01 0.00 Max Organ Dose (mrem) 0.97 km SSE 6.70E-03 1.50E+01 0.04 Child - Kidney Year Gamma Air Dose (mrad) 0.58 km W 1.08E-05 2.00E+01 0.00 Beta Air Dose (mrad) 0.58 km W 7.33E-06 4.00E+01 0.00 Total Body Dose (mrem) 0.58 km W 1.02E-05 1.00E+01 0.00 Skin Dose (mrem) 0.58 km W 1.59E-05 3.00E+01 0.00 Max Organ Dose (mrem) 0.97 km SSE 6.02E-02 3.00E+01 0.20 Child - Bone Liquid Effluents Parameter Max Receptor Dose Dose Limit
% of Limit Qtr 1 Max Organ Dose (mrem)
Adult - Gi-LLi 1.96E-03 1.00E+01 0.02 Total Body Dose (mrem)
Adult - Total Body 1.91E-03 3.00E+00 0.06 Qtr 2 Max Organ Dose (mrem)
Adult - Gi-LLi 6.94E-04 1.00E+01 0.01 Total Body Dose (mrem)
Adult - Total Body 6.80E-04 3.00E+00 0.02 Qtr 3 Max Organ Dose (mrem)
Adult - Gi-LLi 7.43E-04 1.00E+01 0.01 Total Body Dose (mrem)
Adult - Total Body 4.72E-04 3.00E+00 0.02 Qtr 4 Max Organ Dose (mrem)
Adult - Gi-LLi 1.53E-03 1.00E+01 0.02 Total Body Dose (mrem)
Adult - Total Body 1.03E-03 3.00E+00 0.03 Year Max Organ Dose (mrem)
Adult - Gi-LLi 4.93E-03 2.00E+01 0.02 Total Body Dose (mrem)
Adult - Total Body 4.09E-03 6.00E+00 0.07
ENCLOSURE 3 D59 PROCESS CONTROL PROGRAM FOR PROCESSING/DEWATERING OF RADIOACTIVE WASTE FROM LIQUID SYSTEMS REVISION 13, MARCH 08, 2024 12 pages to follow
PRAIRIE ISLAND NUCLEAR GENERATING PLANT RADIATION PROTECTION PROCEDURE D
PROCESS CONTROL PROGRAM FOR PROCESSING/DEWATERING OF RADIOACTIVE WASTE FROM LIQUID SYSTEMS D59 REVISION:
13 Page 1 of 12 Working Copy Verified Initial Date Initial Date MO:
See SAP RESULTS/COMMENTS:
Maintenance Notification Initiated: YES NO Number:
PERFORMERS NAMES AND INITIALS Print Name:
Initials: Date:
Print Name:
Initials: Date:
Print Name:
Initials: Date:
Print Name:
Initials: Date:
Print Name:
Initials: Date:
Print Name:
Initials: Date:
REFERENCE USE
- Procedure should be at the work location.
- Procedure segments may be performed from memory.
- Use the procedure to verify segments have been completed.
- When required, sign or initial appropriate blocks to certify that all segments are complete.
Approval:
602000032674
PRAIRIE ISLAND NUCLEAR GENERATING PLANT RADIATION PROTECTION PROCEDURE D
PROCESS CONTROL PROGRAM FOR PROCESSING/DEWATERING OF RADIOACTIVE WASTE FROM LIQUID SYSTEMS D59 REVISION:
13 Page 2 of 12 TABLE OF CONTENTS Section Title Page 1.0 GENERAL............................................................................................................ 4 1.1 Purpose..................................................................................................... 4 1.2 Scope......................................................................................................... 4 1.3 Definitions.................................................................................................. 4 2.0 PROCESSING OF CERTAIN WASTE LIQUIDS THRU SPENT BEAD RESIN 5
2.1 Purpose..................................................................................................... 5 2.2 Applicability................................................................................................ 5 2.3 Sequence of Operation.............................................................................. 5 2.4 Dewatering Procedure............................................................................... 5 3.0 DEWATERING OF BEAD RESIN......................................................................... 6 3.1 Purpose..................................................................................................... 6 3.2 Applicability................................................................................................ 6 3.3 Dewatering Procedure............................................................................... 6 3.4 Verification of Dewatering.......................................................................... 6 4.0 DEWATERING OF POWDERED RESIN............................................................. 7 4.1 Purpose..................................................................................................... 7 4.2 Applicability................................................................................................ 7 4.3 System Description.................................................................................... 7 4.4 Disposal..................................................................................................... 7 5.0 PROCESSING/DEWATERING OF SPENT FILTER ELEMENTS........................ 8 5.1 Purpose..................................................................................................... 8 5.2 Applicability................................................................................................ 8 5.3 Description of Filling Process..................................................................... 8 5.4 Dewatering................................................................................................. 9
PRAIRIE ISLAND NUCLEAR GENERATING PLANT RADIATION PROTECTION PROCEDURE D
PROCESS CONTROL PROGRAM FOR PROCESSING/DEWATERING OF RADIOACTIVE WASTE FROM LIQUID SYSTEMS D59 REVISION:
13 Page 3 of 12 5.5 Verification of Dewatering.......................................................................... 9 5.6 Spent Filter Storage in Drums.................................................................... 9 6.0 REPORTING REQUIREMENTS........................................................................ 11 6.1 Purpose................................................................................................... 11 6.2 Applicability.............................................................................................. 11 6.3 References.............................................................................................. 11 6.4 PCP Revisions......................................................................................... 11 6.5 Reports of Mishaps.................................................................................. 11 6.6 PCP Specimen Summary Reports........................................................... 12 7.0 ATTACHMENTS................................................................................................. 12
PRAIRIE ISLAND NUCLEAR GENERATING PLANT RADIATION PROTECTION PROCEDURE D
PROCESS CONTROL PROGRAM FOR PROCESSING/DEWATERING OF RADIOACTIVE WASTE FROM LIQUID SYSTEMS D59 REVISION:
13 Page 4 of 12 1.0 GENERAL 1.1 Purpose The purpose of this Process Control Program (PCP) is to detail the means by which the dewatering of radioactive waste from liquid systems can be assured, in accordance with applicable federal regulations and other requirements governing the disposal or processing solid radioactive waste.
1.2 Scope This PCP includes the following processes:
1.2.1 Dewatering of bead resin.
1.2.2 Dewatering of powdered resin.
1.2.3 Dewatering of spent filter elements.
1.2.4 Reporting Requirements.
1.3 Definitions 1.3.1 Dewatering The process of removing water from a substance to meet specific limits.
PRAIRIE ISLAND NUCLEAR GENERATING PLANT RADIATION PROTECTION PROCEDURE D
PROCESS CONTROL PROGRAM FOR PROCESSING/DEWATERING OF RADIOACTIVE WASTE FROM LIQUID SYSTEMS D59 REVISION:
13 Page 5 of 12 2.0 PROCESSING OF CERTAIN WASTE LIQUIDS THRU SPENT BEAD RESIN 2.1 Purpose To establish an alternate method of processing certain waste liquids in lieu of solidification. This method utilizes spent bead resin to filter out suspended particulates allowing normal processing of the resultant liquid. Disposal volumes and personnel exposures are thus reduced.
2.2 Applicability The following waste liquids may be processed using this procedure:
2.2.1 Laundry sludge 2.2.2 Decon solutions 2.2.3 Filter sludge 2.2.4 Mop bucket slurry 2.2.5 Tank bottoms 2.2.6 Sump bottoms NOTE:
Evaporator Concentrates may not be processed using this procedure.
The above list is not to be considered complete. Items may be added or deleted upon evaluation of the Rad Materials Shipping Coordinator.
2.3 Sequence of Operation 2.3.1 Ensure there is a layer of bead resin in the liner to act as a filter (the type of liner is determined by the activity of the material to be disposed of).
2.3.2 Ensure adequate volume for the quantity of material to be processed.
2.3.3 Pump/pour liquid slurry into liner.
2.3.4 Flush drum and/or container, pump and hoses to liner.
2.4 Dewatering Procedure Dewater as per Section 3.0 Dewatering of Bead Resin to ensure there is no free standing water in either the resin or the sludge.
PRAIRIE ISLAND NUCLEAR GENERATING PLANT RADIATION PROTECTION PROCEDURE D
PROCESS CONTROL PROGRAM FOR PROCESSING/DEWATERING OF RADIOACTIVE WASTE FROM LIQUID SYSTEMS D59 REVISION:
13 Page 6 of 12 3.0 DEWATERING OF BEAD RESIN 3.1 Purpose To describe the process used to provide reasonable assurance that bead resin is dewatered to meet applicable disposal criteria.
3.2 Applicability This section of the PCP is applicable disposal site or processors criteria to all radioactively contaminated bead resin which is intended to be shipped dewatered (not solidified) for disposal.
3.3 Dewatering 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), and with the dewatering requirement of the disposal site. Individual shipping procedures unique to the particular container and disposal site or processor refer to the appropriate dewatering procedure.
In general, however, the dewatering process normally consists of the following steps after the liner has been filled:
3.3.1 Initial pumpdown with the diaphragm pump until suction is lost.
3.3.2 A waiting period (twenty hours, for example).
3.3.3 Final dewatering consisting of one or more pumpdowns using a diaphragm pump or a vacuum pumping system.
3.4 Verification of Dewatering Preceding shipment, connect and operate the dewatering pump as before.
IF no water is present, THEN the dewatering process is complete.
IF water is found, THEN pump until vacuum is lost. Repeat the pump/wait cycle as required.
WHEN no more water can be removed, THEN the dewatering process is complete.
PRAIRIE ISLAND NUCLEAR GENERATING PLANT RADIATION PROTECTION PROCEDURE D
PROCESS CONTROL PROGRAM FOR PROCESSING/DEWATERING OF RADIOACTIVE WASTE FROM LIQUID SYSTEMS D59 REVISION:
13 Page 7 of 12 4.0 DEWATERING OF POWDERED RESIN 4.1 Purpose To describe the process used to provide reasonable assurance that powdered resin is dewatered to meet applicable disposal site criteria.
4.2 Applicability This section of the PCP is applicable to all radioactively contaminated powdered resin which is intended to be shipped for burial or processing.
4.3
System Description
Contaminated powdered resin originates in the Condensate Polishing System Filter Demineralizers of both units.
Spent resin is purged from the Filter Demineralizers to the Backwash Waste Receiving Tank where it awaits the dewatering/drying process.
The dewatering/drying process takes place in the Clamshell Backwash Waste Filter (Clamshell).
There are two Clamshells to serve the needs of both units, each capable of being aligned to either unit. It is the function of the clamshells to filter the powdered resin out of the water-resin slurry that is pumped from the Backwash Waste Receiving Tank, thru the Clamshells. When a cake of resin develops in the Clamshell to a predetermined thickness, the filtering process automatically switches to a purge phase followed by a forced air drying phase. The duration of the air drying phase can be adjusted. Experience, however, has demonstrated that a drying cycle of approximately 12 minutes produces a product sufficiently dry to meet disposal site requirements yet not so dry as to create an airborne contamination hazard.
When the air-dry cycle is completed, the resin is dumped from the Clamshell into a hopper from which it is conducted down an enclosed chute to a container below. If the resin is insufficiently dried it will not flow freely down the chute.
4.4 Disposal
PRAIRIE ISLAND NUCLEAR GENERATING PLANT RADIATION PROTECTION PROCEDURE D
PROCESS CONTROL PROGRAM FOR PROCESSING/DEWATERING OF RADIOACTIVE WASTE FROM LIQUID SYSTEMS D59 REVISION:
13 Page 8 of 12 Powdered resin which has been processed thru the Clamshell system does not normally receive further dewatering treatment. Powdered resin may, therefore, be shipped in a container not fitted with dewatering equipment such as a steel drum or box. Because processed powdered resin is sufficiently dry to flow freely, and because powdered resin is normally very low in specific activity, if approved, 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.
5.0 PROCESSING/DEWATERING OF SPENT FILTER ELEMENTS 5.1 Purpose This section describes the method for processing spent filter elements and the process used to provide reasonable assurance that spent filter shipments are dewatered to meet applicable disposal site or processors criteria.
5.2 Applicability This section of the PCP is applicable to all radioactively contaminated filter elements intended for shipment for disposal or processing in the dewatered state (not solidified). Procedures specific to the appropriate type of container SHALL be employed.
5.3 Description of Filling Process 5.3.1 IF spent filters are to be bagged and placed into a drum, THEN proceed to section 5.6, Spent Filter Storage in Drums.
5.3.2 Verify that the container to be used is approved by the manufacturer for disposal of filter elements.
5.3.3 Ensure a dewatering element with an attached hose is installed in the container. The dewatering elements must be compatible with the dewatering pump.
5.3.4 Filter elements should be drained of excess water prior to placing in the container.
5.3.5 Place filter elements into the container while attempting to avoid bridging of filters and observing the principles of ALARA.
PRAIRIE ISLAND NUCLEAR GENERATING PLANT RADIATION PROTECTION PROCEDURE D
PROCESS CONTROL PROGRAM FOR PROCESSING/DEWATERING OF RADIOACTIVE WASTE FROM LIQUID SYSTEMS D59 REVISION:
13 Page 9 of 12 5.4 Dewatering The dewatering process may vary with type and manufacture of container and with requirements of the disposal facility. Typically, however, the dewatering process consists of the following steps:
5.4.1 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.
5.4.2 Connect the dewatering pump to the dewatering element hose. Conduct the pump discharge hose to a container to enable monitoring of discharge volume.
5.4.3 Start the dewatering pump.
IF no water is found, THEN the container may be considered to be dewatered.
IF water is found, THEN pump until vacuum is lost, THEN stop the pump and begin another wait period.
Repeat the pump/wait cycle until no more water can be removed.
5.5 Verification of Dewatering Preceding shipment, connect and operate the dewatering pump as before.
IF no water is present, THEN the dewatering process is complete.
IF water is found, THEN pump until vacuum is lost. Repeat the pump/wait cycle as required.
WHEN no more water can be removed, THEN the dewatering process is complete.
5.6 Spent Filter Storage in Drums 5.6.1 Ensure a drum provided meets the following criteria prior to use:
A.
Drum has been inspected per FP-RP-NISP-07, Control of Radioactive Material.
B.
Drum has been marked per FP-RP-NISP-04, Radiological Posting and Labeling.
C.
Drum has approved absorbents placed in the bottom.
PRAIRIE ISLAND NUCLEAR GENERATING PLANT RADIATION PROTECTION PROCEDURE D
PROCESS CONTROL PROGRAM FOR PROCESSING/DEWATERING OF RADIOACTIVE WASTE FROM LIQUID SYSTEMS D59 REVISION:
13 Page 10 of 12 5.6.2 Ensure approved bags are used to contain loose radioactive materials once the filter is placed inside.
5.6.3 Ensure approved absorbents are placed into the receiving bag prior to pulling the filter.
5.6.4 Place the filter into the prepared bag.
5.6.5 Ensure the bag is secured tightly to prevent leakage of any radioactive contaminants while in storage in the drum.
5.6.6 Once the drum is filled per work plan, ensure a radioactive materials tag is placed on the drum documenting pertinent radiological information and contents.
5.6.7 Ensure lock ring is installed per manufacturers direction.
5.6.8 Store the drum per RMSC direction.
PRAIRIE ISLAND NUCLEAR GENERATING PLANT RADIATION PROTECTION PROCEDURE D
PROCESS CONTROL PROGRAM FOR PROCESSING/DEWATERING OF RADIOACTIVE WASTE FROM LIQUID SYSTEMS D59 REVISION:
13 Page 11 of 12 6.0 REPORTING REQUIREMENTS 6.1 Purpose This section of the PCP sets forth the reporting requirements are as they apply to this PCP to ensure that the reports are completed accurately and in a timely manner.
6.2 Applicability This section of the PCP, in whole or part, applies to all sections of the PCP.
6.3 References Waste Form Technical Position, Revision 1. United States Nuclear Regulatory Commission.
6.4 PCP Revisions Whenever the PCP is revised or changed, a description of the changes AND justifications SHALL be included in the Annual Radioactive Effluent Release Report.
6.5 Reports of Mishaps Waste form mishaps SHALL be reported to the NRC (Director of the Division of Low-Level Waste Management and Decommissioning) AND the designated State disposal site regulatory authority within 30 days of knowledge of the incident.
Mishaps are defined as failure or misuse of stabilized waste forms or containers that provide stability (HICs). Such mishaps include, but are not necessarily limited to, the following:
6.5.1 The failure of high integrity containers used to ensure structural stability.
6.5.2 The misuse of high integrity containers, as evidenced by excessive free liquid, or excessive void space within the container.
6.5.3 Production of a solidified Class B or Class C waste form that exhibits any of the characteristics listed in the Waste Form Technical Position, Revision 1.
PRAIRIE ISLAND NUCLEAR GENERATING PLANT RADIATION PROTECTION PROCEDURE D
PROCESS CONTROL PROGRAM FOR PROCESSING/DEWATERING OF RADIOACTIVE WASTE FROM LIQUID SYSTEMS D59 REVISION:
13 Page 12 of 12 6.6 PCP Specimen Summary Reports WHENEVER cement stabilization (as defined by 10 CFR 61) of low-level waste is necessary, THEN PCP test specimens are required for verification and surveillance.
Verification specimens are intended to provide assurance that the formulations used in the qualification testing program correspond to those actually used in the field.
Surveillance specimens are intended to provide verification that the waste forms remain stable with time. A summary report SHALL be prepared annually and submitted to the NRC (Director, Division of Low-Level Waste Management and Decommissioning) documenting the results of tests performed on the cement-stabilized waste form surveillance specimens during the calendar year.
The annual report should be submitted within 90 days of the end of each calendar year.
7.0 ATTACHMENTS NONE