L-PI-05-046, Annual Radioactive Effluent Report and Offsite Dose Calculation Manual

Annual Radioactive Effluent Report and Offsite Dose Calculation Manual

ML051360426
Person / Time
Site:	Prairie Island
Issue date:	05/16/2005
From:	Solymossy J Nuclear Management Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
FOIA/PA-2010-0209, L-PI-05-046
Download: ML051360426 (28)

Text

U S Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555 Prairie Island Nuclear Generating Plant Units 1 and 2 Dockets 50-282 and 50-306 License Nos. DPR-42, DPR-60 2004 Annual Radioactive Effluent Report and Offsite Dose Calculation Manual Pursuant to the applicable Prairie Island Nuclear Generating Plant Technical Specifications (TS), Appendix A to Operating Licenses DPR-42 and DPR-60, and the requirements of the Offsite Dose Calculation Manual (ODCM), the Nuclear Management Company, LLC (NMC) by this letter submits the 2004 Annual Radioactive Effluent Report which comprises the following reports: contains the Off-Site Radiation Dose Assessment for January through December 2004 in accordance with the requirements of the ODCM; contains the Annual Radioactive Effluent Report, Supplemental Information, Revision 0, for the period January I , 2004 through December 31, 2004 in accordance with the requirements of TS 5.6.3 and the ODCM; and contains the Effluent and Waste Disposal Annual Report, Solid Waste and Irradiated Fuel Shipments, in accordance with the requirements of TS 5.6.3 and the ODCM.

TS 5.5.1 .c requires submittal of a complete copy of the entire ODCM and a summary of changes to the ODCM, when there are changes. The ODCM has not been changed since 2003 and the current revision was submitted last year. Hence, no copy of the ODCM is being submitted.

1717 Wakonade Drive East a Welch, Minnesota 55089-9642 Telephone: 651.388.1 121

Document Control Desk Page 2 Summarv of Commitments This letter contains no new commitments and no revisions to existing commitments.

Generating Plant Units 1 and 2 Enclosures (3) cc: Regional Administrator, USNRC, Region Ill 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 FOR January throunh December 2004 (8 pages follow)

An Assessment of the radiation dose due to releases from Prairie Island Nuclear Generating Plant during 2004 was perfonned 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 fi-om 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 2004.

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.

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.

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.

1. Uncontrolled release of gases from vent header into Auxiliary Building:

From June 21-24, 2,430 cubic feet of vent header gases were released into the Auxiliary Building during maintenance activities. A control valve used for isolation leaked. The release was via the Auxiliary Building normal ventilation system. There was no detectable activity on grab samples collected in the area of the leak and there was no detectable increase on the applicable radiation stack monitors.

Cause: Leaking valve.

Corrective Based on gas decay tank activity, the radioactive gases released totaled:

Action:

Kr 5.65E+04 uCi Xe-133 - 5.83E+02 uCi Based on the conservative assumption that the gas was released evenly, out the unit's stack over the entire period of release, the dose off site as a result of this release was:

Gamma Dose - 1.26E-06 mRad Beta Dose - 1.19E-04 mRad Release files RACO 115 and RACO 116 were created to document this 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-037271 The event was reported to the NRC Region 3 Radiation Protection (RP) Inspector, at the time of the event.

2. Uncontrolled release of gases from Chemical Volume Control System due to relief valve leak:

From June 25-30, 16,350 cubic feet of vent header gases were released while diverting to 121 Chemical and Volume Control System (CVCS) Holdup Tank during a down power evolution. A relief valve on the 121 CVCS Holdup Tank leaked. Release from the plant was via the Auxiliary Building normal ventilation system. There was no detectable activity on grab samples collected in the area of the leak and there was a slight increase (10-15 CPM) above background on the applicable radiation stack monitors.

Cause: Leaking relief valve Corrective Based on gas decay tank analysis, the radioactive gases released totaled:

Action:

Kt- 3.80E+05 uCi Xe- 133 - 3.92E+03 uCi Based on the conservative assumption that all the gases escaped evenly out a single units vent stack over the entire period of the release the dose off site as a result of this release was:

Gamma Dose - 8.47E-06 mRad Beta Dose - 7.09E-04 rnRad Release files RAC0137 and RAC0138 were created to document this 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-037309.

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

3. Temporary filter installed in Spent Fuel Pool exhaust without tritium monitor:

A high efficiency particulate air filter (HEPA) unit installed by RP in the Spent Fuel Pool (SFP) area discharged outside the SFP and did not measure or account for the tritium released. The Chemistry department was not informed of the ventilation change and there were no samples collected for the airborne tritium removed from the SFP area IAW the ODCM, Table 3.1. The HEPA was in place from Aug 13,2004 at 07:OO through Sep 08,2004 at 17:08.

Cause: RP was unaware of the requirement and did not contact the Chemistry department, who is responsible for the effluent requirements and would have ensured the required samples were collected.

Corrective Action: The period of release was conservatively determined to be the entire period of Aug 13, 04 07:OO to Sep 08,04 17:08.

Tritium activity was based on SFP Normal Ventilation routine measurements.

Hours were adjusted in the release files, to compensate for the difference in flow of the HEPA and SFP Normal Ventilation.

Release files RAC0167, RAC0168, RAC0169, RAC0170 and RAC0175 were created to document this release.

RP was counseled on this issue.

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-038133.

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

4. Inadequate controls of airflow from Unit 1 Containment while equipment hatched removed:

Between September 18,2004 and September 22,2004, there was a period of 27.6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> when the Unit 1 Equipment Hatch was removed, with no Inservice Purge or Large Containment Purge fans in service.

One train of Shield Building Ventilation and Shield Building Recirculation Fans were providing ventilation to the Shield Building, exhausting to the Shield Building Exhaust Stack.

During periods when the West Fuel Receipt Rollup Door was opened and certain weather conditions existed, it is believed that movement of air may have been out of containment.

Cause: Control of the roll-up door was not considered for its potential impact on containment air flow.

Corrective All air samples collected in the vicinity of the Equipment Hatch were reviewed. Only Actions: one particulate filter had activity. The air sampler was an AMS-4, located at the hatch, just inside the hatch. Activity was 1.5E-12 uCi/cc Co-58 and 3.7E-14 uCi/cc Nb-95.

This activity was assigned to the release. Tritium level, for the release, was assigned based on silica gel tritium sampling for Unit 1 Shield Building Stack.

A release file (RAC0236) was created for Effluent Surveillance Week 37, to document the release. The conservative estimates of a release flow rate of 1000 CFM, a release duration of 27.6 were used in calculating release volume. Release volume was 4.69E+10 cc. Conservative maximum activity and maximum off site dose are calculated as follows:

Co-58 7.03E-02 uCi 9.78E-08 mRem Nb-95 1.74E-03 uCi 1.35E-09 rnRem H-3 5.77E+02 uCi 1.03E-06 mRem Total 1.13E-06 mRem Additional controls were placed on the roll up door, during outage.

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-038784.

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

The calculated dose from the release of radioactive materials in liquid or gaseous effluents d U m t exceed twice the limits of IOCFRSO, Appendix I, therefore compliance with 40CFR190 i s .required to be assessed, in this report.

The minimum sampling frequency, minimum analysis frequency and lower limit of detection (LLD) requirements, as specified in ODCM Tables 2.1 and 3.1 H Yexceeded.

There yverew occurrences when less than the minimum required radioactive liquid andlor gaseous effluent monitoring instrumentation channels were operable as required by ODCM Tables 2.2 and 3.2.

No fuel casks were loaded and placed in the storage facility during the 2004 calendar year. The total number of casks in the ISFSI is seventeen. There has been no release of radioactive effluents from the ISFSI.

The Offsite Dose Calculation Manual was revised in 2004. The current revision is 18. The revision date is June 26,2003. A copy was submitted with last year's report.

There were changes made to the Process Control Program in 2004. Current manual is revision 8, August 8, 1999.

Table 1 OFF-SITE RADIATION DOSE ASSESSPENT - PRAIRIE ISLAND PERIOD: JANUARY through DECEMBER 2004 10 CFR Part 50 Appendix I Guidelines for a 2-unit site per year Gaseous Releases Maximum Site Boundry Gamma Air Dose (mrad)

Maximum Site Boundry Beta Air Dose (mrad)

Maximum Off-site Dose to any organ (mrem)*

Offshore Location Gamma Dose (mrad)

Total Body (mrem)*

Organ (mrad)*

Liquid Releases Maximum Off-site Dose Total Body (mrem) 1.873-03 Maximum Off-site Dose Organ - GI TRACT (mrem) 4.813-03 Limiting Organ Dose Organ - TOTAL BODY (mrem) 1.873-03

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

Table 2 OFF-SITE RADIATION DOSE ASSESSMENT - PRAIRIE ISLAND SUPPLEMENTAL INFORMATION PERIOD: JANIIARY t h r o u g h -

Maximum Site Boundary Dose Location (From Building Vents)

Sector Distance (miles)

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 Maximum Off-site Dose Location Downstream Pathway Fish

ENCLOSURE 2 ANNUAL RADIOACTIVE EFFLUENT REPORT 01JAN-04 THROUGH 31-DEC-04 SUPPLEMENTAL INFORMATION (8 pages follow)

2004 Annual Radioactive Effluent Report REV. 0 Page 1 of 8 Retention: Lifetime ANNUAL RADIOACTIVE EFFLUENT REPORT 01-JAN-04 THROUGH 31-DEC-04 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 5 500 mrem/year total body 53000 mrem/year skin 1-131, 1-133, H-3, LLP 51500 mrem/year to any organ

b. The dose due to radioactive gaseous effluents released from the site shall be limited to:

noble gases 510 mrad/quarter gamma (20 mrad/quarter beta S20 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

2004 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 I Table 2, Column 1

2. Iodine and particulates with half lives greater than 8 days in gaseous releases:

10 CFR 20, Appendix B I 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%

Nucl ide Gem

4. Liquid effluents Total Gem +25%

Nuclide Gem E. Manual Revisions

1. Offsite Dose Calculations Manual latest Revision number: 18 Revision date  : 6/26/03

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2004 ANNUAL RADIOACTIVE EFFLUENT REPORT REV. 0 PAGE 6 TABLE 1A L I Q U I D EPFLUEHTS - SUHHATIOU OF ALL RELEASES I QTR: 01 1 QTR: 02 1 QTR: 03 1 QTR: 04 16.0 VOLUME OF WASTE PRIOR TO DILUTION (LITERS) 17.0 VOLUME OF DILUTION WATER (LITERS) 18.0 P I S S I O U AUD ACTIVATIOU PRODUCTS 18.1 TOTAL RELEASES W/O H-3, RADGAS, ALPHA (CI) 18.2 AVERAGE DILUTION CONCENTRATION (UCI/ML) 19.0 TRITIUW 19.1 TOTAL RELEASE (CI) 19.2 AVERAGE DILUTION CONCENTRATION (UCI/ML) 20.0 D I S S O L V E D AUD EUTRAIUBD GASES 20.1 TOTAL RELEASE (CI) 20.2 AVERAGE DILUTION CONCENTRATION (UCI/ML) 21.0 GROSS ALPHA (CI) 22.0 TOTAL TRITIUM, FISSION 6 ACTIVATION PRODUCTS (UCI/ML) 23.0 TOTAL BODY DOSE (MREM) 24.0 C R I T I C A L ORGAN 24.1 DOSE (MREM) 1.26E-03 4.95E-04 5.733-04 2.383-03 24 - 2 ORGAN G I TRACT TOT BODY TOT BODY GI TRACT 25.0 PERCENT OF TECHNICAL SPECIFICATIONS LIMIT ( % )

26.0 PERCENT OF CRITICAL ORGAN TECH SPEC LIMIT ( % )

2004 ANNUAL RADIOACTIVE EFFLUENT REPORT REV. 0 PAGE 7 TABLE 21 L I Q U I D EFPLUERTS - SUIIIUTIOII OP ALL RELEASES (CI) 27.0 IHDIVIDUAL L I Q U I D EPPLUBUT CONTINUOUS MODE BATCH MODE I I QTR: 01 /

I QTR: 02 II QTR: 03 1 I

QTR: 04 1 I

QTR: 01 /

I QTR: 0 2 /

I QTR: 03 1 I

QTR: 04 I I

AG-108M CI 6.39E-06 AG-11OM CI 1.353-03 2.553-04 1.09E-04 1.31E-03 BE-7 CO-5 7 C1 CI

c. 2.14E-06 6.803-07 3.093-06 7.463-06 1.24E-04 CO-58 CI 1.58E-05 1.03E-03 1.973-04 4.06E-03 3.053-02 CO-6 0 CI 3.363-04 8.43E-05 1.78E-04 6-943-04 c

I I I I 1 CR-51 CI 1.41E-04 7.31.-06 9.278-04 1-1OE-03 CS-137 CI 1.6 0E-06 FE-55 CI 1.293-05 4.35E-03 3.263-03 3.29E-03 2.693-02 FE-5 9 CI 7.773-05 6.453-06 1.15E-05 7.193-04 MN-54 CI 4.60E-06 3.443-06 5.35E-05 NA-24 NB- 9 5

=I CI c

I I I I I 4.703-06 4.44E-06 3.11E-04 NB-97 SB-122 C1 CI

/ 6*54E-07 1.12E-06 1.623-05 3.773-06 2.083-06 SB-124 CI 3.393-04 1.34E-04 6.78E-06 1.663-03

- I SB-125 CI 3.123-03 1.30E-03 8.79E-04 4.943-03 1

SN-113 3.003-05 SR-9 2 2.79E-05 3.29E-06 2.03E-06 2.123-05 TE-123M CI 3.34E-04 2.163-05 3.65E-05 3.56E-04 TE-125M CI 8.90E-03 6.543-05 1.31E-02 I I TE-132 CI 4.66E-06 ZR-95 CI 4 -708-06 1.71E-04 TOTALS CI 2.00E-02 5-273-03 9.613-03 8.203-02

2004 ANNUAL RADIOACTIVE EFFLUENT REPORT REV. 0 PAGE 8 TABLE PA L I Q U I D EFFLUEHTS - SUllMATIOIW OF ALL RELEASES (CI) (COIWTIIOUED) 28.0 DISSOLVED AlsD EBTRAIIOBD GASES CONTINUOUS MODE BATCH MODE NUCLIDE UNITS QTR: 01 QTR: 02 QTR: 03 QTR: 04 QTR: 01 QTR: 02 QTR: 03 QTR: 04 I I I I I I I I I I I I XE-133 CI 1.223-04 4.183-04 5.303-05 XE-133M XE-135 TOTALS CI CI CI c.

I 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 I

3.283-06 1.26E-04 4.563-05 4.643-04 3.423-06 5.643-05

ENCLOSURE 3 EFFLUENT AND WASTE DISPOSAL ANNUAL REPORT SOLID WASTE AND IRRADIATED FUEL SHIPMENTS (7 pages follow)

PlNGP 753, Rev. 7 Page 1 of 7 Retention: Life PRAIRIE ISLAND NUCLEAR GENERATING PLANT Period: OI/OI/O~-12/31/04 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:

C. Non-Compacted m' 4.35E+02 1280 ft3 1.54E+04 Ci 3.41E-01 2.50E+01 D. Filter Media m3 ft3 Ci S. Other (furnish description) m3 4.79E+02 4071 Old Steam Generators ft3 1.69E+04 4384 Ci 1.25E+03 2.50E+01 and activity of the low-level waste leaving the Prairie Island site.

ance is made for off-site volume reduction prior to H:\word~documents\Documents\2004PlNGP 753.dot

PlNGP 753, Rev. 7 Page 2 of 7 PRAIRIE ISLAND NUCLEAR GENERATING PLANT Period: 01-01-04112-31-04 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

• Fe-55

• = Inferred - Not Measured on Site

PlNGP 753, Rev. 7 Page 3 of 7 PRAIRIE ISLAND NUCLEAR GENERATING PLANT Period: 01-01-04112-31-04 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 (O.OOE0)

• = Inferred - Not Measured on Site

PlNGP 753, Rev. 7 Page 4 of 7 PRAIRIE ISLAND NUCLEAR GENERATING PLANT Period: 01-01-04112-31-04 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 RACE Logistics RACE, LLC Canadian Pacific Envirocare of Utah Railroad Perkins Envirocare of Utah

PlNGP 753, Rev. 7 Page 5 of 7 PRAIRIE ISLAND NUCLEAR GENERATING PLANT Period: 01-01-04112-31-04 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.

(~t~lrn~) (Ci) Waste Code Code 43841124.14 8.38E+02 S DOT-E 13401 NIA 43841124.14 4.14E+02 DOT-E 13401 NIA 40711115.28 NIA 407111 15.28 NIA NIA NIA NIA TOTAL 10 3230019 14 1250.34 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

PlNGP 753, Rev. 7 Page 6 of 7 PRAIRIE ISLAND NUCLEAR GENERATING PLANT Period: 01-01-04112-31-04 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

PlNGP 753, Rev. 7 Page 7 of 7 PRAIRIE ISLAND NUCLEAR GENERATING PLANT Period: 01-01-04112-31-04 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 SolidificationIDewateringof Radioactive Waste from Liquid Systems Current Revision Number: 8 Effective Date: 812511999 ChangesIJustification: NIA