ML20217Q825

Off-Site Radiation Dose Assessment for NSP,Jan-Dec 1997

ML20217Q825
Person / Time
Site:	Prairie Island
Issue date:	12/31/1997
From:	NORTHERN STATES POWER CO.
To:
Shared Package
ML20217Q814	List: ML20217Q811 ML20217Q825 ML20217Q832 ML20217Q839
References
NUDOCS 9805110133
Download: ML20217Q825 (7)

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Attachment 1 1

1 PRAIRIE ISLAND NUCLEAR GENERATING PLANT l NORIITERN STATES POWER COMPANY OFF-SITE RADIATION DOSE ASSESSMENT FOR NORHTERN STATES POWER COMPANY l l

January Through December 1997 9805110133 980504 PDR ADOCK 05000202:

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. i NORTHERN STATES POWER COMPANY PRAIRIE ISLAND NUCLEAR GENERATING PLANT OFF-SITE RADIATION DOSE ASSESSMENT FOR January throuah December 1997 An Assessment of the radiation dose due to the release from Prairie Island Nuclear Generating Plant during 1997 was performed in accordance with the Technical Specifications. Computed doses were well below the 40 CFR Part 190 Standards and 10 CFR Part 50 Appendix I Guidelines.

Of f-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 1997.

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. Doses are a small percentage of Appendix I Guidelines.

Off-site Doses from Liould Release Computed doses due to Liquid releases are reported in Table 1.

Receptor information is reported in Table 2. 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 j transportation activities within the site boundary, it is assumed that  ;

the limiting dose within the site boundary would be received by en j 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 l from noble gas-releases and the whole body and organ doses from the l

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

Doses to Individuals Due to Effluent Releases from the ISFSI Two additional loaded fuel casks were placed in the storage facility during the 1997 calendar year bringing the total number of casks to 7.

There has been no release of radioactive effluents from the ISFSI.

Radiation Effluent Monitorina Samplina Deviations Two liquid waste tank composite samples were accidentally discarded prior to preparing the composite sample for analysis in accordance with the Offsite Dose Calculation Manual Table 2.1 and were therefore not analyzed for Sr-89, Sr-90, Fe-55 or alpha.

Cause: The release month for the samples was incorrectly entered into the computer database as March rather than April. The database then provided a report to assist the user in preparing the composite volumes at the end of the quarter. Since the two tanks were incorrectly i entered, they were accidently discarded with the quarter

1. 1 samples even though they had not been included in the composite preparation.

Corrective 1) The computer program used to enter the release Actions: dates has been upgraded to prevent users from entering dates that are not logical.

2) The procedure to prepare the composites has been revised to require the user to review and inventory all available samples prior to composite preparation and before discarding any samples.

Result: A review of previous results showed that the composite values from waste tank samples are normally less than LLD for Sr-89, Sr-90 and alpha and that the Fe-55 values are consistent over time. No unusual plant evolutions were occurring at the time of the release. The results of the quarter one and quarter two waste tank composites were evaluated and the activity values were assigned to the tanks in question in a conservative manner.

LIOUID ABNORMAL RELEASE On January 30, 1997 a volume of liquid radioactive waste was released without sampling and analysis prior to the release in accordance with Table 2.1 of the Offsite Dose Calculations Manual.

Cause: A leaking valve diaphragm allowed liquid rad waste to I mix with processed waste water during a planned release. l The liquid waste monitor (R-18) tripped within seconds  :

limiting the unplanned release volume to a calculated 100 gallons of unprocessed water.

Corrective A number of valves in the liquid rad waste system were Actions: rebuilt or repaired, and piping identified as unnecessary was cut and blank flanged to prevent i reoccurrence.

Result: Post release samples were collected from the waste system piping and the source of the inleakage.

Activities and subsequent doses were conservatively determined and compared to release rate and dose limitations. The calculated Total Body Dose due to the release was 1.90E-05 mrem. The dose from the activity released was insignificant and compared to only a fraction of a normal processed waste tank release, and did not impose upon the health and safety of the public.  ;

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. AIRBORNE ABNORMAL RELEASE During the fourth quarter an unplanned release resulting from equipment inadequacy resulted in 9456 cubic feet of a radioactive waste gas being released.

Cause: A small diaphragm leak in a waste gas compressor outlet pressure control valve caused a 9456 cubic feet of waste gas to be released into the auxiliary building over an 18 day period. The leak was initially identified by an evaluation of gas system inventory records.

Correctl'e The leaking valve diaphragm was repaired and additional Action: valves were rebuilt.

Result: The waste gases leaked into the auxiliary building and were released via the sampled and monitored auxiliary building waste path. The leaking gas radioactivity concentration was so low as to not be detectable on any grab samples or detectable on rad effluent monitors.

Concentrations of gases released were calculated from waste gas decay t. : samples taken for routine i surveillance.

The calculated gamma and beta dose off site resulting from the release of the airborne noble gases was:

l Gamma Dose: 7.28E-06 mrad Beta Dose: 5.85E-04 mrad The dose from the gases released was insignificant and posed no health or safety concerns with personnel on or off site.

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Table 1 l l

OFF-SITE RADIATION DOSE ASSESSMENT - PRAIRIE ISLAND PERIOD: JANUARY through DECEMBER 1997 l 10 CFR Part 50 Appendix I Guidelines per 2-units site per year Gaseous Releases Maximum Site Boundry Gamma Air Dose (mrad) 3.83E-05 20 l

Maximum Site Boundry Bota Air Dose (mrad) 3.98E-03 40 -

Maximum Off-site Dose I l

to any organ (mrem)* 2.83E-01 30 l Offshore Location Gamma Dose (mrad) 2.18E-06 Total Body (mrem)* 3.69E-03 Organ (mrad)* 3.84E-03 30 Liquid Releases Maximum Off-site Dose Total Body (mrem) 1.50E-02 6 Maximum Off-sita Dose Organ - LIVER (mrem) 2.12E-02 20 Limiting Organ Dose Organ - TOTAL BODY (mrem) 1.50E-02 06

• Long-Lived Particulate, I-131, I-133 and Tritium l

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Table 2 I

OFF-SITE RADIATION DOSE ASSESSMENT - PRAIRIE ISLAND )

SUPPLEMENTAL INFORMATION )

PERIOD: JAPDARY throuoh DECEMBER 1997 i

Gaseous Releases I

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 Liould Releases Maximum Off-site Dose Location Downstream Pathway Fish

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