L-PI-07-033, 2006 Annual Radioactive Effluent Report and Offsite Dose Calculation Manual

From kanterella
Jump to navigation Jump to search
2006 Annual Radioactive Effluent Report and Offsite Dose Calculation Manual
ML071370287
Person / Time
Site: Prairie Island  Xcel Energy icon.png
Issue date: 05/14/2007
From: Wadley M
Nuclear Management Co
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
FOIA/PA-2010-0209, L-PI-07-033
Download: ML071370287 (32)
v • d • e


Text

NMC Committed to Nuclear Excellence Prairie Island Nuclear Generating Plant Operated by Nuclear Management Company, LLC MAY 1 4 2007, L-PI-07-033 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 2006 Annual Radioactive Effluent ReDort 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 2006 Annual Radioactive Effluent Report which is comprised of the following reports: contains the Off-Site Radiation Dose Assessment for the period January 1, 2006 through December 31, 2006 in accordance with the requirements of the ODCM. contains the Annual Radioactive Effluent Report, Supplemental Information, for the period January 1, 2006 through December 31, 2006 in accordance with the requirements of TS 5.6.3 and the ODCM. contains the Effluent and Waste Disposal Annual Report, Solid Waste and Irradiated Fuel Shipments, for the period January 1, 2006 through December 31, 2006 in accordance with the requirements of TS 5.6.3 and the ODCM. is an attachment to the 2006 Annual Effluent Report which contains an Amended Liquid Pathway Dose Calculation for the third quarter of 2006. In accordance with industry guidelines on groundwater monitoring, the report includes the dose calculation and dose report for a secondary steam condensate leakage which occurred in August 2006.

1717 Wakonade Drive East

Telephone: 651.388.1121 L Vt8 A-Coli NiL(S-sDf

Document Control Desk Page 2 contains a complete copy of the entire ODCM, Revision 20, dated 10/20/06. 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 (7 Site Vice President, Prairie Isla'rd Nuclear Generating Plant Nuclear Management Company, LLC Enclosures (4) cc: Regional Administrator, USNRC, Region III Project Manager, Prairie Island Nuclear Generating Plant, USNRC, NRR NRC Resident Inspector - Prairie Island Nuclear Generating Plant Tim Donakowski, State of Minnesota

ENCLOSURE 1 OFF-SITE RADIATION DOSE ASSESSMENT January 01, 2006 - December 31, 2006 8 pages follow

PRAIRIE ISLAND NUCLEAR GENERATING PLANT OFF-SITE RADIATION DOSE ASSESSMENT FOR January through December 2006 An Assessment of the radiation dose due to releases from Prairie Island Nuclear Generating Plant during 2006 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 2006..

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 131, Iodine-133, tritium and long-lived particulates were calculated for this location and occupancy time. These doses are reported in Table 1.

ABNORMAL RELEASES There were a total often (10) abnormal releases for 2006. Of these, six (6) abnormal releases were due to containment openings as a result of the Ul Refueling Outage. These abnormal releases were combined into one abnormal release covering the U1 Refueling Outage. The resulting 2006 abnormal releases are summarized below:

1. Loss of Pressure in 129 Waste Gas Decay Tank (WGDT)

Operations noted an unexpected drop in 129 Waste Gas Decay Tank (WGDT) pressure. An evaluation of the waste gas system indicated that 129 WGDT pressure dropped from 79.0 psig to 37.0 psig between 2/8/2006 12:41 and 2/8/2006 18:05. During this time period 122/123 WGDTs were released per C21.3-10.10.

Cause: Leakage was identified at WG-3-13, the outlet of 129 WGDT to exhaust ducts. This valve is verified closed per C21.3-10.10, Waste Gas Decay Tank release procedure.

WG-3-13 is a diaphragm valve.

The incident was due to a ruptured diaphragm or a misadjusted stem travel nut. Valve maintenance and preventive maintenance for diaphragm valves is the cause.

Corrective Action: 129 WGDT was sampled for nuclide mix.

Activity released based on sample taken from 129 GDT post release:

Nuclide uCi/ml uCi Released Gamma Beta Dose (mrad) Dose (mrad)

Kr-85 2.28E-03 8.66E+04 7.81E-08 8.86E-06 Xe-131M 3.85E-04 1.46E+04 1.20E-05 8.52E-05 Xe-133 3.08E-02 1.17E+06 1.79E-03 5.32E-03 Xe-133M 3.49E-04 1.313E+04 2.29E-05 1.03E-04 Xe-135 4.33E-05 1.65E+03 1.67E-05 2.13E-05 TOTAL 1.84E-03 5.54E-03 H3 9.57E-06 uCi/ml 3.64E+02 uCi 2.16E-06 mrem CE 0 1014047-03, was written to "Evaluate the diaphragm valve PM program". This action is assigned to the system engineer for resolution.

Created release file RABOOO 10 to account for release.

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 captured in the site's Action Request Process, CAP- 01014047.

The event was reported to the NRC Region 3 Radiation Protection (RP) Inspector, at the time of the event.

2. Release from Containment Openings during Unit 1 Refueling Outage Unit One experienced a Fuel Element Defect during cycle 1R24. During refueling outage 1R24 (May 2006) routine air samples, taken at the containment openings (Equipment Hatch, airlocks), detected activity. It was determined that a release had occurred.

Procedural guidance was incomplete for the quantification and reporting of these releases, therefore these releases were determined to be abnormal and were documented in this report as such.

Cause: Typically, air samples taken at containment openings are negative. Due to the fuel element defect, an increased potential for elevated airborne levels in containment and increase probability of release at containment openings existed. Additional assessment should have occurred pre-outage and additional controls implemented.

Insufficient control of the containment openings was identified. An airborne limit, at which containment would be isolated, should have been established, to support effluent release goals, Definitive control of the weather curtain at the Equipment Hatch was lacking.

Additionally, the differential pressure across the containment, created when the Equipment Hatch and the Fuel Receipt Rollup Doors are opened simultaneously, was not sufficiently assessed or controlled. This led to airborne radioactivity escaping the Maintenance Airlock and being entrained in other ventilation pathways, not equipped with charcoal.

Even with sufficient controls the possibility of release exists. Guidance on quantifying and reporting was insufficient.

Corrective Actions: Each week the air samples taken at the various containment openings were assessed.

Conservative assumptions were made with regards to flow and activity. Release files were created to quantify and report the activity released:

Nuclide Activity Released (uCi) Dose (mrem)

Co-58 3.36E+00 4.66E-06 Co-60 5.53E-01 2.01E-05 1-131 3.60E+02 2.33E-02 H3 1.78E+03 3.18-06 Total 2.33E-02 Nuclide Activity Released Gamma Dose Beta Dose (uCi) nRad) (mRad)

Xe-133 2.73E+06 1.03E-03 3.06E-03 Xe-1.35 1.26E+04 2. 60E-05 3.32E-05 Total 1.06E-03 3.09E-03 The activity released was captured and reported in release files number RACO 102, RACO I10, RACO132, RACO133, RACO145 and RAC0146.

Procedural guidance was generated to correct the deficiencies in monitoring and control.

Training was conducted.

The procedure written also includes the conservative assumptions upon which such releases will be based on in the future.

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 captured in the site's Action Request Process, CAP- 01027608.

The event was reported to the NRC Region 3 Radiation Protection (RP) Inspector, at the time of the event.

3. Failed Primary Sample Valve Sprays Primary Coolant.

Various primary sample drag valves were removed, cleaned and reinstalled as part of pre-outage maintenance. When reinstalled, the Pressurizer Steam Space sample path was lined up as part of post maintenance testing. The cooler inlet, 2SM-5-4 sprayed steam. Elevated levels were noted on 2R30 and 2R37, Unit Two Auxiliary Building Stack monitors. The lineup was promptly secured.

Cause: The valve which failed was not one of the valves upon which maintenance was performed, however the Pressurizer Steam Space Sample path is typically only used during outage. This particular valve had not been operated in several months. Spray was determined to be from a packing leak.

Corrective Actions: Increase in radiation monitors readings for Unit Two Auxiliary Building Stack required integration and generation of an abnormal release report.

No samples were taken during this short duration release, therefore the nuclide mix was based on the Gaseous Source Term. Activity, released and dose cosequences:

NUCLIDE uCi Beta (mRad) Gamma (mRad)

Kr-85m 1.10E+03 1.45E-06 2.32E-06 Kr-85 6.90E+02 1.27E-08 1.44E-06 Kr-87 4.14E+02 2.74E-06 4.57E-06 Kr-88 1.79E+03 2.92E-05 5.62E-06 Xe-131m 6.90E+02 1.1 5E-07 8.21E-07 Xe-133m 1.79E+03 6.27E-07 2.84E-06 Xe-133 1.331E+05 4.95E-05 1.47E-04 Xe-135 2.76E+03 .5.68E-06 7.27E-06 Xe-138 4.14E+02 4.09E-06 2.11E-06 TOTAL 9.34E-05 1.74E-04 The activity released was captured and reported as release file RAC0260.

Work Order #100264 was generated to implement repairs.

Result: The dose from the activity released represented a small percentage of the total dose and were a very small percentage of limits. The dose did not impose upon the health and safety of the public.

The event was captured in the site's Action Request Process, CAP-1055506.

The event was reported to the NRC Region 3 RP Inspector.

4. Unit Two Steam Generator Manway Removal Trips Containment Ventilation:

Unit Two experienced a Fuel Element Defect during cycle 2R24. During Unit Two refueling outage 2R24, in an effort to control containment airborne levels, HEPA hoses from Steam Generator Manways were directed to Containment Ventilation. Upon removal of the Steam Generator Inserts, Inservice Purge promptly tripped, due to. increased radiation monitor count rate. Following the initial purge of the Steam Generator Bowls activity quickly decreased and purging of the Steam Generators directly to containment ventilation was recommenced without incident.

Cause: Unanticipated activity levels when initially venting Steam Generator Bowls.

Corrective Action: The duration of the release was determined to be the time that inserts were removed to the time that ventilation tripped. This was determined to be 5 minutes.

Radiation monitors were integrated to determine release activity. Total activity was ratioed to the nuclide mix previously identified:

Nuclide Activity (uCi) Gamma Dose (mRad) Beta Dose (mRad)

Xe-131m 3.09E+04 3.68E-05 5.16E-06 Xe-,133 1.78E+06 2.OOE-03 6.73E-04 Xe-133m 1.61E+04 2.56E-05 5.64E-06 Total 2.06E-03 6.84E-04 Release file RAB0290 was created to document this release.

It was determined that additional constraints would be implemented to preclude this event, including continuous monitoring and throttling of Steam Generator Bowl purging until such time as the activity levels decreased to less than 50% of monitor set points.

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 captured in the site's Action Request Process, CAP- 01062770.

The event was reported to the NRC Region 3 Radiation Protection (RP) Inspector, at the time of the event.

5. Leaking Steam Relief On December 14,,2006, Following the Unit Two Refueling outage 2R24 it was noted that the down stream temperatures on CV-3 1104, STM GEN 2A STM DUMP TO ATM, were elevated indicating leak by and an on-going release.

Cause: CV-31104, STM GEN 2A STM DUMP TO ATM leaked by.

Corrective Action: Samples taken quantified no activity.

Release file RABOO88 was created to document this release.

Repairs we implemented to CV-31104.

Result: No activity was released and no there was no dose consequence to the general public.

This release did not impose upon the health and safety of the public.

The event was captured in the site's Action Request Process, CAP-01068657.

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 10CFR50, Appendix I, therefore compliance with 40CFR1 90 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 2006.

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 2006 calendar year. The total number of casks in the ISFSI is twenty-two (22). There has been no release of radioactive effluents from the ISFSI.

CURRENT ODCM REVISION The Offsite Dose Calculation Manual was revised in 2006. The current revision is 20. The revision date is October 20, 2006. A copy is submitted with this year's report.

PROCESS CONTROL PROGRAM There were no changes made to the Process Control Program in 2006. Current manual is revision 8, August 25, 1999.

Table 1 OFF-SITE RADIATION DOSE ASSESSMENT - PRAIRIE ISLAND PERIOD: JANUARY through DECEMBER 2006 10 CFR Part 50 Appendix I Guidelines for a 2-unit site per year Gaseous Releases Maximum Site Boundry Gamma Air Dose (mrad) 5. 39E-02 20 Maximum Site Boundry Beta Air Dose (mrad) 1. 62E-01 40 Maximum Off-site Dose to any organ (mrem)* 4. 55E-01 30 Offshore Location Gamma Dose (mrad) 3.95E-03 Total Body (mrem)* 9. 66E-04 Organ (mrad)* 9.54E-03 30 Liquid Releases Maximum Off-site Dose Total Body (mrem) 3. 16E-03 6 Maximum Off-site Dose Organ - GI TRACT (mrem) 3. 96E-03 20 Limiting Organ Dose Organ - TOTAL BODY (mrem) 3. 16E-03 6

  • Long-Lived Particulate, 1-131, 1-133 and Tritium

Table 2 OFF-SITE RADIATION DOSE ASSESSMENT - PRAIRIE ISLAND SUPPLEMENTAL INFORMATION PERIOD: JANUARY throuiih DECEMBER 2006 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, 2006 - December 31, 2006 9 pages follow

2006 Annual Radioactive Effluent Report REV. 0 Page 1 of 9 Retention: Lifetime ANNUAL RADIOACTIVE EFFLUENT REPORT 01-JAN-06 THROUGH 31-DEC-06 SUPPLEMENTAL INFORMATION Facility: Prairie Island Nuclear Generating Plant Licensee: Northern States Power Company License Numbers: DPR-42 & DPR-60 A. Regulatory Limits

1. Liquid Effluents:
a. The dose or dose commitment to an individual from radioactive materials in liquid effluents released from the site shall be limited to:

for the quarter 3.0 mrem to the total body 10.0 mrem to any organ for the year 6.0 mrem to the total body 20.0 mrem to any organ

2. Gaseous Effluents:
a. The dose rate due to radioactive materials released in gaseous effluents from the site shall be limited to:

noble gases

  • 500 mrem/year total body

<3000 mrem/year skin 1-131, 1-133, H-3, LLP <1500 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

2006 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:

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.OE-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 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%

Nuclide Gem E. Manual Revisions

1. Offsite Dose Calculations Manual latest Revision number: 20 Revision date  : _10/20/06_

2006 ANNUAL RADIOACTIVE EFFLUENT REPORT REV. 0 PAGE 3 1.0 BATCH RELEASES (LIQUID)

QTR: 01 QTR: 02 QTR: 03 QTR: 04 A 2.60E+01 8.50E+01 3.50E+01 7.40E+01i 1.1 NUMBER OF BATCH RELEASES 1.2 TOTAL TIME PERIOD (HRS) 4.92E+01 1.59E+02 6.51E+01 1.30E+02 1.3 MAXIMUM TIME PERIOD (HRS) 2.40E+00 3.27E+00 2.32E+00 2.68E+00 1.4 AVERAGE TIME PERIOD (HRS) 1.89E+00 1.87E+00 1.86E+00 1.75E+00 1.5 MINIMUM TIME PERIOD (HRS) 1.60E+00 6.17E-01 1.43E+00 1.23E+00 1.6 AVERAGE MISSISSIPPI RIVER FLOW (CFS) 1.89E+04 3.79E+04 7.37E+03 7.06E+03 2.0 BATCH RELEASES (AIRBORNE)

QTR: 01 QTR: 02 QTR: 03 TQTR: 04-1 2.1 NUMBER OF BATCH RELEASES 1.00E+01 2.40E+01 0.OOE+00 1.70E+01 2.2 TOTAL TIME PERIOD (HRS) 1.20E+02 3.56E+02 0.OOE+00 1.92E+021 2.3 MAXIMUM TIME PERIOD (HRS) 2.40E+01 3.25E+01 0.OOE+00 2.40E+011 2.4 AVERAGE TIME PERIOD (HRS) 1.20E+01 1.49E+01 0.OOE+00 1.13E+011 2.5 MINIMUM TIME PERIOD (HRS)

L 2.83E-03

__ _ I 7.17E-03 I 0.OOE+00

__ I_ _i__

I 1.67E-021 3.0 ABNORMAL RELEASES (LIQUID)

QTR: 01 QTR: 02 QTR: 03 QTR: 04 3.1 NUMBER OF BATCH RELEASES 0.OOE+00 0.OOE+00 0.00E+00 0.OOE+00 3.2 TOTAL ACTIVITY RELEASED (CI) 0.OOE+00 0.OOE+00 0.00E+00 ] 0.00E+00 3.3 TOTAL TRITIUM RELEASED (CI) 0.OOE+00 0.OOE+00 0.OOE+00 J 0.OOE+00 4.0 ABNORMAL RELEASES (AIRBORNE) fQTR: 01 QTR: 02 QTR: 03 QTR: 04 4.1 NUMBER OF BATCH RELEASES 1.OOE+00 6.OOE+00 0.OOE+00 3.OOE+00 4.2 TOTAL ACTIVITY RELEASED (CI) 1.29E+00 2.74E+00 ] 0.OOE+00 J1.97E+001

2006 ANNUAL RADIOACTIVE EFFLUENT REPORT REV. 0 PAGE 4 TABLE IA GASEOUS EFFLUENTS - SUMMATION OF ALL RELEASES QTR: 01 QTR: 02 QTR: 03 QTR: 04 SI I I I 5.0 FISSION AND ACTIVATION GASES 5.1 TOTAL RELEASE (CI) 4.44E+00 T 3.83E+01 0.OOE+00 9.56E+01 1 5.2 AVERAGE RELEASE RATE (UCI/SEC) 5.65E-01 4.87E-o00 I o.OOE+00 1 1.-22E+01 5.3 GAMMA DOSE (MRAD) I.oE-o03 t 1.48E-02 I o.OOE+00 3.62E-012 5.4 BETA DOSE (MRAD) 1.10E-02 t4.34E-02 t .OOE+00 t1. 08E-011]

5.5 PERCENT OF GAMMA TECH SPEC (%) 2.99E-02 I1.48E-01 I 0.IE+ 3.62E-011 5.6 PERCENT OF BETA TECH SPEC (%) 5.50E-02

__ __ I _

2.17E-01

_ _ I 0.OOE+00 15.

I _

39E-01 1

6.0 IODINES 6.1 TOTAL 1-131 (CI) 3.57E-06 6.27E-03 i 0.OOE+00 i 3.68E-04 6.2 AVERAGE RELEASE RATE (UCI/SEC) 4.55E-07 7.98E-04 I 0.OOE+00 4.68E-05 I 7.0 PARTICULATES 7.1 TOTAL RELEASE (CI) 3.74E-0 2.93E-05 0.OOE+00 1.36E-0*5 7.2 AVERAGE RELEASE RATE (UCI/SEC) 4.76E-07 I3.73E-06 I 0.OOE+00 I 1.73E-016 I __ _ _ _ _ _ I _ I__ _ I__

8.0 TRITIUM 8.1 TOTAL RELEASE (CI) 3.03E+00 2.54E+00 2.11E+00 2.37E4-00 8.2 AVERAGE RELEASE RATE (UCI/SEC) 3.86E-01 3.23E-01 1 2.68E-01 j 3.02E-01 J 9.0 TOTAL IODINE, PARTICULATE AND TRITIUM (UCI/SEC) 3.86E-01 3.24E-01 2.68E-01 3.02E-01 10.0 DOSE FROM IODINE, LLP, AND TRITIUM (MREM) 9.21E-03 4.11E-01 3.77E-03 3.11E-02 11.0 PERCENT OF TECH SPEC (%) 6.14E-02 2.74E+00 2.51E-02 2.07E-01 12.0 GROSS ALPHA (CI) o.OOE+0O 0.OOE+00 0.OOE+00 1.OOE+00

2006 ANNUAL RADIOACTIVE EFFLUENT REPORT REV. 0 PAGE 5 TABLE IC GASEOUS EFFLUENTS - GROUND LEVEL RELEASES (CI) 13.0 FISSION AND ACTIVATION GASES CONTINUIOUS MODE BATCH MODE 1I NUCLIDE UNITS QTR: 01 QTR: 02 QTR: 03 QTR: 04 QTR:

QT: 01 01 1QTR: 02 I QTR: 03 I QTR: 04 I II II AR-41 CI [ I S2. OOE-04 KR-85 CI [ S6.90E-04 2.65E-01 KR-85M CI 4 1.10E-03 1.41E-04 4 KR-87 CI 4 4.14E-04 KR-88 CI 4 1.79E-03 XE-131M CI 4 2.38E-01 XE-133 CI 2.93E+00 4 3.81E+01 9.52E+01 f1.22E+0044{

XE-133M CI 2.04E-01 1.33E-02 4 XE-135 CI [ 1.87E-01 7.45E-03 4.11E-03 4 XE-138 CI _ _ 4.14E-04 TOTALS CI I 2.93E+00 j 3.83E+01 0.00E+00 9.56E+01 JS1.51E+00 0.OOE+00 0.OOE+00 000E÷00 14.0 IODINES CONTINUqOUS MODE BATCH MODE 0TR: 01 QTR: 02 QTR: 03 QTR: 04 NUCLIDE i

UNITS II jQTR: 01 QTR: 02 QTR: 03 QTR: 04 QTR 01 I 1-131 CI 3.57E-06 6.27E-03 3.6 8E-04 2.09E-07 6.02E-10 1-133 CI 1.55E-04 3.21E-05 TOTALS CI 3.57E-06 j6.43E-03 0.OOE+00 4.OOE-04 O.OOE00 2.09E-07 0.OOE+00 6.02E-10

2006 ANNUAL RADIOACTIVE EFFLUENT REPORT REV. 0 PAGE 6 TABLE IC GASEOUS EFFLUENTS - GROUND LEVEL RELEASES (CONTINUED) 15.0 PARTICULATES CONTINUOUS MODE BATCH MODE NUCLIDE T UNITS i1 QTR: 01 1QTR: 02 1 QTR: 03 QTR: 04 QTR: 01 QTR: 02 QTR: 03 QTR: 04 IIIIIII I BE-7 CI 4.11E-08 I _

CO-58 CI 3.67E-06 [ 2.91E-07 ] 1.86E-07 CO-60 CI f5.53E-07 4I fCS-137 CI3.56E-08 ______I]{3.74E-06 ______II1.31E-05 R-105 2.50 I2.50-° I I-0o I TOTALS CI 0.OOE+00 j 2.93E-05 0.00+E00 2.91E-07 J{3.74E-06 0.00+E00 0.00+E00 1.33E-05

2006 ANNUAL RADIOACTIVE EFFLUENT REPORT REV. 0 PAGE 7 TABLE LA LIQUID EFFLUENTS - SUMMATION OF ALL RELEASES QTR: 01 QTR: 02 QTR: 03 QTR: 04-1 II I II 16.0 VOLUME OF WASTE PRIOR TO DILUTION (LITERS) 3.85E+07 4.46E+07 1.88E+07 2.29E°07 17.0 VOLUME OF DILUTION WATER (LITERS) 1.62E+*1 1.04E+l1 2.66E+11 1.84E+1-1 18.0 FISSION AND ACTIVATION PRODUCTS 18.1 TOTAL RELEASES W/O H-3, RADGAS, ALPHA (CI) 1.13E-02 1 2.85E-02 1.13E-02 3.23E-02 18.2 AVERAGE DILUTION CONCENTRATION (UCI/ML) 7.OOE-11 2.73E-10 4.26E-11 I 1.76E-10 19.0 TRITIUM 19.1 TOTAL RELEASE (CI) 3.40E+02 1.63E+02 1.29E+02 1.75E+02 19.2 AVERAGE DILUTION CONCENTRATION (UCI/ML) 2.11E-06 1.57E-06 4.87E-07 9.54E-07 20.0 DISSOLVED AND ENTRAINED GASES 20.1 TOTAL RELEASE (CI) 1.35E-03 11.61E-02 1.75E-03 1.32E-02 20.2 AVERAGE DILUTION CONCENTRATION (UCI/ML) 8.34E-12 1.55E-10 6.59E-12 7.18E-11 21.0 GROSS ALPHA (CI) O.OOE+00 0.OOE+00 0.OOE+00 0.0OE+00 22.0 TOTAL TRITIUM, FISSION & ACTIVATION PRODUCTS (UCI/ML) 2.11E-06 1.57E-06 4.87E-07 9.55E-07 23.0 TOTAL BODY DOSE (MREM) 1.19E-03 7.16E-04 3.40E-04 9.10E-04 24.0 CRITICAL ORGAN 24.1 DOSE (MREM) 1.19E-03 7.16E-04 3.40E-04 9.10E-04 24.2 ORGAN TOT BODY TOT BODY TOT BODY TOT BODY 25.0 PERCENT OF TECHNICAL SPECIFICATIONS LIMIT (%) 3.98E-02 2.39E-02 1.13E-02 3.03E-02 26.0 PERCENT OF CRITICAL ORGAN TECH SPEC LIMIT (%) 3.98E-02 2.39E-02 1.13E-02 3.03E-02

2006 ANNUAL RADIOACTIVE EFFLUENT REPORT REV. 0 PAGE 8 TABLE 2A LIQUID EFFLUENTS - SUMIATION OF ALL RELEASES (CI) 27.0 INDIVIDUAL LIQUID EFFLUENT CONTINUOUS MODE BATCH MODE NUCLIDE UNITS FQTR: QTR: 0 2 QTR: 03 QTR: 04 QTR: 01 QTR: 02 QTR: 03 QTR: 04 I I I t I *T 6.45E-06 I 2.OOE-05 9.67E-07 AG-108M CI

[AG-IlOM CI 1.18E-03 2.90E-03 1.46E-03 5.81E-03 4 4 +

CO-57 CI 5.86E-05 6.58E-05 1.07E-05 2.77E-05 CO-58 CI 2.55E-03 5.81E-03 2.73E-03 8.75E-03

. 2 -

CO-60 CI 1.40E-03 8.07E-03 1.72E-03 4.39E-03 CR-51 CI 4.49E-04 2.07B-03 4.83E-04 1.95E-03 CS-134 CI 3.60E-05 iI 2.46E-05 7.22E-06 1.19E-06 CS-137 CI 6. 28E-05 ______1__ ___ 9.49E-05 6.55E-05 7.46E-06 3.24E-06 FE-55 CI 4.09E-03 2.90E-03 5.78E-03 FE-59 CI 2.13E-04 4.36E-04 1.315-04 2.95E-04 1.0 E-0 1-131 CI 5.04E-04 7.53E-05 LA-140 CI 5.89E-05 MN-54 CI 7.47E-05 4.10E-04 6.49E-05 1.78E-04 NB-95 CI 1.63E-04 2.43E-04 1.08E-04 1.72E-04 NB-97 CI 2.05E-06 7.88E-06 2.59E-06 8.28E-04 SB-124 CI 7.75E-05 1.33E-04 3.22E-06 2.63E-04 SB-125 CI 7.82E-04 2.66E-03 [1.11E-03 2.70E-03 SN-113 CI 5.19E-05 1.01E-04 4.70E-05 1.01E-04 SSi-85 CI 8.30E-07 1.24E-06 1.36E-06 SR-92 CI 1.37E-05 3.15E-05 1.62E-05 4.71E-05 TE-123M TE-125M CI CI

______ I ______ I ______ I______ 6.02E-05 4.65E-03 2.37E-05 4.32E-04 2.18E-04 3.23E-04 4 4

[ TE-132 CI 3.22E-07 W-187 CI 6.20E-06 4.23E-05 5.52E-05 4 4 I-

[ ZN-65 CI 1.56E-05 4.16E-05 (CONTINUED)

2006 ANNUAL RADIOACTIVE EF FLUENT REPORT REV. 0 PAGE 9 TABLE 2A LIQUID EFFLUENTS - SUNKATII ON OF ALL RELEASES (CONTINUED) 27.0 INDIVIDUAL LIQUID N]FFLUENT CONTINU( )US MODE BATCH MODE NUCLIDE UNITS QTR: 01 iTR: 02 QTR: 03 QTR: 04 QTR: 01 QTR: 02 QTR: 03 QTR: 04 II I I I I I I I I II ZR-95 CI 7.85E-05 1.25E-04 5.95E-05 1.76E-04 ZR-97 CI 3.56E-06 2.90E-06 1.91E-06 TOTALS CI 9.89E-05 0.00E+00 0.00E+00 1.21E-04 1.12E-02 2.85E-02 1.13E-02 3.22E-02 28.0 DISSOLVED AND ENTRA: INED GASES CONTINUOUS MODE BATCH MODE NUCLIDE UNITS QTR: 01 QTR: 02 QTR: 03 1QTR: 04 QTR:

II 01 QTR: 02 j I

QTR: 03 I

QTR: 04 I

[ I R-85

~I iCI 1.91E-04 2.85E-04 3.14E-04 XE-131M CI 5.24E-05 5.47E-05 1.14E-04 XE-133 CI 4.23E-05 2.52E-05 1.81E-05 1.06E-03 1.57E-02 1.75E-03 1.26E-02 XE-133M CI 8.96E-06 7.61E-05 XE-135 CI 5.OOE-05 TOTALS CI 4 . 23E-05 [ 2.52E-05 0.OOE+00 1.81E-05 1.30E-03 1.61E-02 1.75E-03 1.32E-02

ENCLOSURE3 EFFLUENT AND WASTE DISPOSAL ANNUAL REPORT SOLID WASTE AND IRRADIATED FUEL SHIPMENTS January 01, 2006 - December 31, 2006 7 pages follow

PINGP 753, Rev. 7 Page 1 of 7 Retention: Life PRAIRIE ISLAND NUCLEAR GENERATING PLANT Period: 01/01/06-12/31/06 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:

A. Iesins m-3 1 .bzt+U1 I 1U.4 f

5.38E+02 Ci 1.35E+02 2.50E+01 3

B. Dry-Compacted mf Ci m3 C. Non-Compacted 4.91 E+02 1280 ft3 1.73E+04 260 Ci3 5.44E-01 2.50E+01 D. Filter Media m3 ft Ci 3

S. Other (furnish description) m 5.54E+01 1958 U1 Rx Head ft3 1.96E+03 Ci 1.39E+01 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.

Document4

PINGP 753, Rev. 7 Page 2 of 7 PRAIRIE ISLAND NUCLEAR GENERATING PLANT Period: 01-01-06/12-31-06 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 6.43E+01 Co-58 8.20E+00 Co-60 8.20E+00

  • Ni-63 1.21E+01 Zr-95 1.84E+00 Nb-95 1.40E+00
  • C-14 1.OOE+O0 1% cutoff S *Fe-55 3.50E+01 Co-58 3.76E+01 Co-60 5.56E+00 Nb-95 2.39E+00 Zr-95 1.35E+00 Cr-51 2.67E+00
  • Ni-63 1.09E+01
  • C-14 3.07E+00 1% cutoff
  • = Inferred - Not Measured on Site

PINGP 753, Rev. 7 Page 3 of 7 PRAIRIE ISLAND NUCLEAR GENERATING PLANT Period: 01-01-06/12-31-06 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)

  • H-3 1.52E+00 A
  • Fe-55 1.29E+01
  • Ni-63 4.61 E+01 Co-60 1.25E+01 Cs-1 34 5.08E+00 Cs-1 37 2.OOE+01 1% cutoff
  • = Inferred - Not Measured on Site

PINGP 753, Rev. 7 Page 4 of 7 PRAIRIE ISLAND NUCLEAR GENERATING PLANT Period: 01-01-06/12-31-06 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 6 Studsvik Logistics RACE, LLC 3 Hittman EnergySolutions, LLC (Bulk) 1 Perkins EnergySolutions, LLC (Bulk) 3 Studsvik Logistics Studsvik Processing Facility, LLC

PINGP 753, Rev. 7 Page 5 of 7 PRAIRIE ISLAND NUCLEAR GENERATING PLANT Period: 01-01-06/12-31-06 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 ) (mCi) Waste Code Code 06-018 260/7.36 1.30E-01 C L N/A 06-022 568.5/16.1 1.70E+01 C L N/A 06-023 1958/55.44 1.39E+04 S L N/A 06-021 1137/32.2 2.07E+01 C L N/A 06-035 2560/72.5 5.01 E+01 C L N/A 06-037 2560/72.5 5.76E+01 C L N/A 06-039 2560/72.5 3.36E+01 C L N/A 06-040 2560/72.5, 1.24E+01 C L N/A 06-041 2560/72.5 1.64E+02 C L N/A 06-042 2560/72.5 1.89E+02 C L N/A 06-053 179.4/5.08 8.18E+03 A N/A 06-059 179.4/5.08 1.03E+05 A A N/A L

06-060 179.4/5.08 2.43E+04. A N/A TOTAL 13 198581561.4 1.50E+05 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: 01-01-06/12-31-06 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: 01-01-06/12-31-06 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.5.5.4). Attach the sidelined pages to this report.

Changes/Justification: N/A

ENCLOSUREF4 ATTACHMENT TO THE 2006 ANNUAL EFFLUENT REPORT AMENDED LIQUID PATHWAY DOSE CALCULATION Quarter 3, 2006 2 pages follow

ATTACHMENT TO THE 2006 ANNUAL EFFLUENT REPORT Amended Liquid Pathway Dose Calculation Quarter 3, 2006 Summary The total liquid dose to the critical receptor for the third quarter of 2006 is 0.0198 mrem.

This dose is reported in this attachment to the annual report.

Backqround On August 4 and 5, 2006, approximately 168 gallons of secondary steam condensate leaked to the ground outside the northeast side of the turbine building. This water had a tritium concentration of 19,100 pCi/L. On August 5, 2006, the leak was contained and no further leakage to the environment occurred. It is assumed that the discharged water could potentially enter the groundwater and be incorporated in drinking water at the nearest resident who is also the critical receptor.

ODCM Considerations The following calculation is independent of the ODCM. A revision to the ODCM to incorporate the Industry Initiative on Groundwater Protection is planned for 2007.

Corrective actions have been taken to prevent a similar spill in the future.

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 concentration of the released water diluted by a factor of 1000 (a dilution factor of approximately 1000 was calculated when tritium was discharged into the discharge canal versus sample results from a well 700 feet from the canal). (In this case, the critical receptor is 0.6 miles from the release point.) The tritium dose conversion factor is taken from page 9-3 of NUREG/CR-3332. Its value is 1.02E-4 mrem/year per pCi/liter of tritium in the body.

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.

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. It should be noted that the total airborne dose (99% plus of which is due to tritium) is greater than the total waterborne tritium dose. Even though the leaked tritium pathway delivers additional dose to the critical receptor, it is a lower concentration than that already in the body due to airborne exposure.

Dose Calculation Dose Diluted Whole Fish Total Quarter Conversion X Tritium = Body + Pathway = Liquid Factor Concentration Dose Dose Dose (mrem/per (pCi/L) (mrem) (mrem) (mrem) pCi/L) 3 1. 04E-4 1. 91E+2 1.95E-2 3.40E-4 1.98E-2 Dose Report LIQUID EFFLUENTS - SUMMATION OF WATERBORNE TRITIUM AND FISH PATHWAYS QTR: 03 TOTAL BODY DOSE (MREM) 1. 98E-02 CRITICAL ORGAN DOSE (MREM) 1. 98E-02 ORGAN TTL BODY PERCENT OF TOTAL BODY TECH SPEC LIMIT (%) 6.50E-01 PERCENT OF CRITICAL ORGAN TECH SPEC LIMIT (%) 6.50E-01

Retrieved from "https://kanterella.com/w/index.php?title=L-PI-07-033,_2006_Annual_Radioactive_Effluent_Report_and_Offsite_Dose_Calculation_Manual&oldid=2660585"