3F0419-03, 2018 Annual Radioactive Effluent Release Report

2018 Annual Radioactive Effluent Release Report

ML19120A413
Person / Time
Site:	Crystal River
Issue date:	04/30/2019
From:	Hobbs T Duke Energy Florida
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
3F0419-03
Download: ML19120A413 (97)

Text

ef_~ DUKE Crystal River Nuclear Plant 15760 W. Power Line Street

~ ENERGY Crystal River, FL 34428 Docket 50-302 Docket 72-1035 Operating License No. DPR-72 10 CFR 50.36a(a)(2)

ODCM April 30, 2019 3F0419-03 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001

Subject:

Crystal River Unit 3 - 2018 Annual Radioactive Effluent Release Report

Dear Sir:

Duke Energy Florida, LLC (DEF), hereby provides the 2018 Radioactive Effluent Release Report for Crystal River Unit 3 (CR-3) in accordance with 10 CFR 50.36a(a)(2) and the Offsite Dose Calculation Manual (ODCM). The attached report (Attachment 1) includes a summary of the quantities of radioactive liquid and gaseous effluents, and solid waste released from the CR-3 site during 2018. The data provided in this report is consistent with the objectives outlined in the ODCM and the Process Control Program (PCP), and is in conformance with 10 CFR 50, Appendix I, Section IV.B.1.

A CR-3 administrative procedure requires submittal of licensee initiated changes to the ODCM as part of the Radioactive Effluent Release Report for the period of the report in which any changes were made. The ODCM was revised in 2018; changes are described in this report and a copy of the ODCM is attached (Attachment 2). The PCP was not revised in 2018.

This letter contains no new regulatory commitments.

If you have any questions regarding this submittal, please contact Mr. Mark Van Sicklen, Licensing Lead, Nuclear gulatory Affairs, at (352) 501-3045.

T~x~1 Sincerely, General Manager, Decommissioning - SAFSTOR TDH/mvs Attachment 1: 2018 Annual Radioactive Effluent Release Report Attachment 2: CR3 Offsite Dose Calculation Manual xc: NMSS Project Manager Regional Administrator, Region I

DUKE ENERGY FLORIDA, LLC DOCKET NUMBER 50 - 302 / LICENSE NUMBER DPR - 72 ATTACHMENT 1 2018 ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT

ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 2018 DUKE ENERGY FLORIDA, LLC CRYSTAL RIVER UNIT 3 Facility Operating License No. DPR-72 Docket No. 50-302

CONTENTS Introduction .................................................................................................................... 1 Tabular Data Summaries Gaseous Effluents - Quarters 1 to 4 ............................................................... 2 Liquid Effluents - Quarters 1 to 4 .................................................................... 7 Batch Release Summary .............................................................................................. 10 Abnormal Release Summary ........................................................................................ 12 Dose Evaluation ............................................................................................................ 14 Appendix I Dose Summary ........................................................................................... 16 Radwaste Shipments .................................................................................................... 18 Unplanned Releases ..................................................................................................... 19 Radioactive Waste Treatment Systems ....................................................................... 19 Annual Land Use Census ............................................................................................. 19 Effluent Monitor Instrument Operability ........................................................................ 19 Meteorology Instrumentation Evaluation ...................................................................... 20 ODCM Changes ............................................................................................................ 20 Process Control Program Changes .............................................................................. 23 Carbon-14 Evaluation ................................................................................................... 23 NEI 07-07 Required Data.............................................................................................. 23

INTRODUCTION This report is submitted as required by procedure CP-500, section 4.4.1.2 and in accordance with 10 CFR 50.36a. All 40 CFR 190 limits have been met. There were no NEI 07-07 groundwater protection reportable events in 2018.

The scope of this report includes:

• A summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the plant.

• Quarterly and annual dose summaries.

• A list and description of unplanned releases to unrestricted areas.

• A description of any changes to the:

Process Control Program (PCP), and Offsite Dose Calculation Manual (ODCM).

• Significant changes to any radioactive waste treatment system.

• A list of new dose calculation location changes identified by the annual land-use census.

• Information relating to effluent monitors or required supporting instrumentation being inoperable for 30 or more days.

• Information required to be included in this report per NEI 07-07 Industry Ground Water Protection Initiative-Final Guidance Document issued in August 2007.

Note for reporting purposes, N/D = Not Detected.

1

EFFLUENT and WASTE DISPOSAL REPORT-2018 Table 1A - Regulatory Guide 1.21 Gaseous Effluents - Summation of All Releases Unit: 3 Units 1st Qtr 2nd Qtr 3rd Qtr 4th Qtr Est. Total

% Error A. Fission & Activation Gases

1. Total Release Curies 0.00E+00 0.00E+00 0.00E+00 0.00E+00 3.00E+01

2. Average Release Rate for Period uCi/sec 0.00E+00 0.00E+00 0.00E+00 0.00E+00 B. Iodines

1. Total Iodine-131 Curies 0.00E+00 0.00E+00 0.00E+00 0.00E+00 3.00E+01

2. Average Release Rate for Period uCi/sec 0.00E+00 0.00E+00 0.00E+00 0.00E+00 C. Particulates

1. Particulates with half-lives > 8 days Curies 0.00E+00 2.04E-07 0.00E+00 0.00E+00 3.00E+01

2. Average Release Rate for Period uCi/sec 0.00E+00 2.60E-08 0.00E+00 0.00E+00

3. Gross Alpha Radioactivity Curies 0.00E+00 0.00E+00 1.23E-07 5.12E-09 D. Tritium

1. Total Release Curies 1.52E-01 1.67E-01 3.43E-01 4.52E-03 3.00E+01

2. Average Release Rate for Period uCi/sec 1.96E-02 2.13E-02 4.31E-02 5.69E-04 2

EFFLUENT and WASTE DISPOSAL REPORT-2018 Table 1B - Regulatory Guide 1.21 Gaseous Effluents - Elevated Batch Mode Unit: 3 (This Table Does Not Apply to Crystal River Unit 3)

Units 1st Qtr 2nd Qtr 3rd Qtr 4th Qtr Fission & Activation Gases Total Curies N/D N/D N/D N/D Iodines Total Curies N/D N/D N/D N/D Particulates Total Curies N/D N/D N/D N/D H-3 Curies N/D N/D N/D N/D Gross Alpha Curies N/D N/D N/D N/D 3

EFFLUENT and WASTE DISPOSAL REPORT-2018 Table 1B - (Continued) Regulatory Guide 1.21 Gaseous Effluents - Elevated Continuous Mode Unit: 3 (This Table Does Not Apply to Crystal River Unit 3)

Units 1st Qtr 2nd Qtr 3rd Qtr 4th Qtr Fission & Activation Gases Total Curies N/D N/D N/D N/D Iodines Total Curies N/D N/D N/D N/D Particulates Total Curies N/D N/D N/D N/D H-3 Curies N/D N/D N/D N/D Gross Alpha Curies N/D N/D N/D N/D 4

EFFLUENT and WASTE DISPOSAL REPORT-2018 Table 1C - Regulatory Guide 1.21 Gaseous Effluents - Ground Batch Mode Unit: 3 Units 1st Qtr 2nd Qtr 3rd Qtr 4th Qtr Fission & Activation Gases Total Curies N/D N/D N/D N/D Iodines Total Curies N/D N/D N/D N/D Particulates Total Curies N/D N/D N/D N/D H-3 Curies N/D N/D N/D N/D Gross Alpha Curies N/D N/D N/D N/D 5

EFFLUENT and WASTE DISPOSAL REPORT-2018 Table 1C - (Continued) Regulatory Guide 1.21 Gaseous Effluents - Ground Continuous Mode Unit: 3 Units 1st Qtr 2nd Qtr 3rd Qtr 4th Qtr Fission & Activation Gases Total Curies N/D N/D N/D N/D Iodines Total Curies N/D N/D N/D N/D Particulates Cs-137 Curies N/D 2.04E-07 N/D N/D Total Curies N/D 2.04E-07 N/D N/D H-3 Curies 1.52E-01 1.67E-01 3.43E-01 4.52E-03 Gross Alpha Curies N/D N/D 1.23E-07 5.12E-09 6

EFFLUENT and WASTE DISPOSAL REPORT-2018 Table 2A - Regulatory Guide 1.21 Liquid Effluents - Summation of All Releases Unit: 3 Units 1st Qtr 2nd Qtr 3rd Qtr 4th Qtr Est. Total

% Error A. Fission & Activation Products

1. Total Release (not including tritium, gases, alpha) Curies 3.54E-04 2.867E-05 3.43E-03 4.06E-02 2.50E+01

2. Average diluted concentration during period uCi/ml 1.71E-12 1.24E-13 1.49E-11 2.07E-10

3. Percent of Applicable Limit  % 3.92E-06 1.24E-06 4.17E-05 1.04E-03 B. Tritium

1. Total Release Curies 2.02E-02 4.56E-04 3.32E-02 3.59E+00 3.00E+01

2. Average diluted concentration during period uCi/ml 9.98E-11 1.97E-12 1.44E-10 1.83E-08

3. Percent of Applicable Limit  % 9.98E-07 1.97E-08 1.44E-06 1.83E-04 C. Dissolved and Entrained Gases

1. Total Release Curies 0.00E+00 0.00E+00 0.00E+00 0.00E+00 2.50E+01

2. Average diluted concentration during period uCi/ml 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 D. Gross Alpha Radioactivity

1. Total Release Curies 0.00E+00 0.00E+00 0.00E+00 0.00E+00 3.00E+01 E. Waste Volume Released (Pre-Dilution) Liters 5.07E+04 3.16E+05 6.77E+05 2.04E+06 1.00E+01 F. Dilution Water Volume During Period Liters 2.02E+11 2.32E+11 2.31E+11 1.96E+11 1.00E+01 7

EFFLUENT and WASTE DISPOSAL REPORT-2018 Table 2B - Regulatory Guide 1.21 Liquid Effluents - Batch Mode Unit: 3 Units 1st Qtr 2nd Qtr 3rd Qtr 4th Qtr Fission & Activation Products Fe-55 Curies 2.57E-05 N/D 4.04E-04 1.04E-02 Co-60 Curies 2.11E-04 N/D 2.74E-03 2.47E-03 Ni-63 Curies 1.00E-04 N/D 2.43E-04 8.59E-03 Sr-89 Curies N/D N/D N/D 3.60E-04 Sr-90 Curies N/D N/D N/D 4.74E-04 Cs-134 Curies N/D N/D N/D 1.95E-04 Cs-137 Curies 7.56E-06 2.87E-05 4.37E-05 1.81E-02 Total Curies 3.45E-04 2.87E-05 3.43E-03 4.06E-02 Dissolved and Entrained Gases Total Curies N/D N/D N/D N/D H-3 Curies 2.02E-02 4.56E-04 3.32E-02 3.59E+00 Gross Alpha Curies N/D N/D 3.26E-06 N/D 8

EFFLUENT and WASTE DISPOSAL REPORT-2018 Table 2B - (Continued) Regulatory Guide 1.21 Liquid Effluents - Continuous Mode Unit: 3 Units 1st Qtr 2nd Qtr 3rd Qtr 4th Qtr Fission & Activation Products Total Curies N/D N/D N/D N/D Dissolved and Entrained Gases Total Curies N/D N/D N/D N/D H-3 Curies N/D N/D N/D N/D Gross Alpha Curies N/D N/D N/D N/D 9

EFFLUENT and WASTE DISPOSAL REPORT-2018 Regulatory Guide 1.21 Gaseous Batch Release Summary Unit: 3 Jan - Jun Jul - Dec Number of Batch Releases 0 0 Total Time Period for Batch Releases 0.00E+00 min 0.00E+00 min Maximum Time Period for a Batch Release 0.00E+00 min 0.00E+00 min Average Time Period for a Batch Release 0.00E+00 min 0.00E+00 min Minimum Time Period for a Batch Release 0.00E+00 min 0.00E+00 min 10

EFFLUENT and WASTE DISPOSAL REPORT-2018 Regulatory Guide 1.21 Liquid Batch Release Summary Unit: 3 Jan - Jun Jul - Dec Number of Batch Releases 3 12 Total Time Period for Batch Releases 6.59E+02 min 8.20E+03 min Maximum Time Period for a Batch Release 4.80E+02 min 5.86E+03 min Average Time Period for a Batch Release 2.22E+02 min 6.83E+02 min Minimum Time Period for a Batch Release 8.80E+01 min 8.00E+01 min Average Stream Flow During Release Periods 2.18E+05 gpm 2.84E+05 gpm 11

EFFLUENT and WASTE DISPOSAL REPORT-2018 Regulatory Guide 1.21 Gaseous Abnormal Release Summary Unit: 3 Jan - Jun Jul - Dec Number of Abnormal Releases 0 0 Total Time Period for Abnormal Releases 0.00E+00 min 0.00E+00 min Maximum Time Period for an Abnormal Release 0.00E+00 min 0.00E+00 min Average Time Period for an Abnormal Release 0.00E+00 min 0.00E+00 min Minimum Time Period for an Abnormal Release 0.00E+00 min 0.00E+00 min Total Activity for Abnormal Releases 0.00E+00 Ci 0.00E+00 Ci 12

EFFLUENT and WASTE DISPOSAL REPORT-2018 Regulatory Guide 1.21 Liquid Abnormal Release Summary Unit: 3 Jan - Jun Jul - Dec Number of Abnormal Releases 0 0 Total Time Period for Abnormal Releases 0.00E+00 min 0.00E+00 min Maximum Time Period for an Abnormal Release 0.00E+00 min 0.00E+00 min Average Time Period for an Abnormal Release 0.00E+00 min 0.00E+00 min Minimum Time Period for an Abnormal Release 0.00E+00 min 0.00E+00 min Total Activity for Abnormal Releases 0.00E+00 Ci 0.00E+00 Ci 13

EFFLUENT and WASTE DISPOSAL REPORT-2018 Regulatory Guide 1.21 Gaseous NNG Organ Dose Unit: 3 Receptor Name: Infant Max Ind NW at 1.34 km 1st Quarter 2nd Quarter 3rd Quarter 4th Quarter Calendar Year

% of  % of  % of  % of  % of Organ Dose ODCM Dose ODCM Dose ODCM Dose ODCM Dose ODCM Limit Limit Limit Limit Limit Bone 3.66E-05 4.88E-04 4.78E-05 6.38E-04 8.23E-05 1.10E-03 1.09E-06 1.45E-05 1.68E-04 1.12E-03 Liver 3.66E-05 4.88E-04 4.89E-05 6.52E-04 8.23E-05 1.10E-03 1.09E-06 1.45E-05 1.69E-04 1.13E-03 Total Body 3.66E-05 4.88E-04 4.20E-05 5.60E-04 8.23E-05 1.10E-03 1.09E-06 1.45E-05 1.62E-04 1.08E-03 Thyroid 3.66E-05 4.88E-04 4.15E-05 5.53E-04 8.23E-05 1.10E-03 1.09E-06 1.45E-05 1.61E-04 1.08E-03 Kidney 3.66E-05 4.88E-04 4.35E-05 5.80E-04 8.23E-05 1.10E-03 1.09E-06 1.45E-05 1.63E-04 1.09E-03 Lung 3.66E-05 4.88E-04 4.23E-05 5.64E-04 8.23E-05 1.10E-03 1.09E-06 1.45E-05 1.62E-04 1.08E-03 GI-Lli 3.66E-05 4.88E-04 4.15E-05 5.53E-04 8.23E-05 1.10E-03 1.09E-06 1.45E-05 1.61E-04 1.08E-03 Maximum Organ was LIVER.

14

EFFLUENT and WASTE DISPOSAL REPORT-2018 Regulatory Guide 1.21 Liquid Organ & Whole Body Dose Unit: 3 Receptor Name: Adult W at 1.34 km 1st Quarter 2nd Quarter 3rd Quarter 4th Quarter Calendar Year

% of  % of  % of  % of  % of Organ Dose ODCM Dose ODCM Dose ODCM Dose ODCM Dose ODCM Limit Limit Limit Limit Limit Bone 1.16E-07 2.32E-06 7.52E-09 1.50E-07 3.49E-06 6.98E-05 6.38E-04 1.28E-02 6.41E-04 6.41E-03 Liver 2.93E-08 5.85E-07 1.03E-08 2.06E-07 3.89E-06 7.78E-05 3.62E-04 7.25E-03 3.66E-04 3.66E-03 Total Body 1.86E-08 1.24E-06 6.73E-09 4.49E-07 2.60E-06 1.73E-04 1.67E-04 1.11E-02 1.69E-04 5.64E-03 Thyroid 1.02E-10 2.03E-09 3.84E-12 7.68E-11 2.09E-09 4.19E-08 5.31E-07 1.06E-05 5.33E-07 5.33E-06 Kidney 6.08E-10 1.22E-08 3.49E-09 6.98E-08 1.08E-06 2.16E-05 5.34E-05 1.07E-03 5.45E-05 5.45E-04 Lung 9.51E-09 1.90E-07 1.16E-09 2.33E-08 3.43E-07 1.29E-05 1.23E-04 2.47E-03 1.24E-04 1.24E-03 GI-Lli 1.01E-07 2.02E-06 2.03E-10 4.05E-09 3.60E-06 7.20E-05 1.58E-04 3.15E-03 1.61E-04 1.631-03 Liquid Effluent Dose Limits Total Body: 1.5 mrem/quarter, 3 mrem/year Any Organ: 5 mrem/quarter, 10 mrem/year 15

EFFLUENT and WASTE DISPOSAL REPORT-2018 Regulatory Guide 1.21 Liquid App I Dose Assessment Unit: 3 Adult W at 1.34km Units 1st Qtr 2nd Qtr 3rd Qtr 4th Qtr Annual Maximum Organ Dose mrem 1.67E-7 7.52E-9 3.49E-6 6.38E-4 6.41E-4 ODCM Limit mrem 5.00 5.00 5.00 5.00 10.00

% of ODCM Limit  % 2.32E-6 1.50E-7 6.98E-5 1.28E-2 6.41E-3 Maximum Organ was Bone Units 1st Qtr 2nd Qtr 3rd Qtr 4th Qtr Annual Total Body mrem 1.86E-8 6.73E-9 2.60E-6 1.67E-4 1.69E-4 ODCM Limit mrem 1.50 1.50 1.50 1.50 3.00

% of ODCM Limit  % 1.24E-6 1.73E-4 1.73 E-4 1.11E-2 5.64E-3 16

EFFLUENT and WASTE DISPOSAL REPORT-2018 Regulatory Guide 1.21 App I Dose Assessment Unit: 3 Airborne Noble Gas Doses Child Site Boundary NW at 1.34 km Units 1st Qtr 2nd Qtr 3rd Qtr 4th Qtr Annual Beta Air mRad 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 ODCM Limit mRad 10.00 10.00 10.00 10.00 20.00

% of ODCM Limit  % 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Units 1st Qtr 2nd Qtr 3rd Qtr 4th Qtr Annual Gamma Air mRad 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 ODCM Limit mRad 5.00 5.00 5.00 5.00 10.00

% of ODCM Limit  % 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Gaseous Release Dose Summary - There was no measurable noble gases released in 2018 due to the plant shutdown in 2009. All fuel has since been moved to the ISFSI pad.

Gaseous Effluent Dose Limits Gamma Air Dose: 5 mrad/quarter, 10 mrad/year Beta Air Dose: 10 mrad/quarter, 20 mrad/year Any Organ: 7.5 mrem/quarter, 15 mrem/year 17

TABLE 3 EFFLUENT and WASTE DISPOSAL REPORT-2018 SOLID WASTE AND IRRADIATED FUEL SHIPMENTS A. SOLID WASTE SHIPPED OFFSITE FOR PROCESSING OR BURIAL (Non-irradiated fuel)

1. Type of waste Unit 12 month period Est. Total Error %

m3 8.46E+0 25

a. Spent resins, filter sludge, evaporator bottoms, etc.

Ci 2.03E+2 m3 2.85E+1 25

b. Dry compressible waste, contaminated equipment, etc.

Ci 4.78E+0 m3 2.38E-1 25

c. Irradiated components, control rods, etc.

Ci 2.68E+3 m3 0 25

d. Other (describe):

Ci 0

2. Estimate of major nuclide composition (by type of waste in %)*

Fe-55 2.5 NA NA a.

Co-60 3.0 Ni-63 15.0 Cs-134 1.5 Cs-137 76.1 Fe-55 51.3 NA NA

b. Ni-63 41.0 Co-60 6.6 Fe-55 18.7

c. Co-60 69.2 NA NA Ni-63 12.0

d. NA NA NA

• Curie values and principle radionuclides are estimates based on a combination of direct and indirect methods.

3. Solid Waste Disposition Number of Shipments Mode of Transportation Destination (Truck Shipments) 10.00 3 Hittman Transport Services Erwin Resin Solutions, Erwin, TN 12 Hittman Transport Services Energy Solutions - Bear Creek 1 Landstar Inway Energy Solutions - Bear Creek 1 Hittman Transport Services WCS - Andrews, Texas B. IRRADIATED FUEL SHIPMENTS (Disposition)

Number of Shipments Mode of Transportation Destination 0 N/A N/A 18

Unplanned Releases There were no unplanned releases in 2018.

Radioactive Waste Treatment Systems There were no significant changes to the radioactive waste treatment systems in 2018. Due to the shutdown status of Crystal River Unit 3, liquid waste volume and radioactivity will continue to decrease while in SAFSTOR. The spent fuel pool water was released in 2018 as part of the spent fuel pool cleanout. This is why the amount of liquid tritium released in 2018 was greater than in 2017. The liquid waste demineralizers did not have to be recharged in 2018 due to the very small amounts of water processed.

Annual Land Use Census The 2018 land-use census did not identify any new dose calculation locations.

Effluent Monitor Instrument Operability For the year 2018, the main gaseous effluent pathway was the auxiliary building ventilation exhaust system. Radiation monitor RM-A2N is the effluent monitor for this pathway. This monitor was in service for all of 2018. The Reactor Building was also lined-up for continuous venting via RM-A2 in 2018 as a means of maintaining a habitable atmosphere inside containment.

The liquid effluent pathways are the primary plant liquid waste stream. All primary liquid effluent releases in 2018 were monitored by radiation monitor RM-L2. All of the secondary plant liquid waste stream releases were monitored by radiation monitor RM-L7, with the exception of one release performed in December. Both RM-L2 and RM-L7 were removed from service, in support of SAFSTOR, on November 29th. Note: No primary releases were made after this time in 2018.

Assessment of Direct Radiation from ISFSI The Independent Spent Fuel Storage Installation (ISFSI) pad was loaded with spent fuel beginning in June 2017 and completed in January 2018. Calculation N16-0003, performed prior to fuel movement into dry storage, documents a conservative annual dose estimate at 700 meters from the ISFSI pad of about 0.06 mrem. A distance of 700 meters is well within the site boundary controlled area in all directions. This small dose is not distinguishable from normal background fluctuations of several mrem per year as measured by the REMP TLDs. These TLDs are deployed in the controlled area, at locations adjacent to the site boundary, and at offsite locations. REMP TLD results for the second half of 2018 show no detectable changes in dose beyond the expected fluctuations. Based on this, and also including the dose contributions from CR3 effluent releases, the 40 CFR 190 annual dose limit of 25 mrem was not exceeded in 2018.

19

Meteorology Instrumentation Evaluation In 2015 the metrology tower was abandoned. It was concluded that an on-site meteorological data collection system was no longer needed at CR3 to support its effluents program because:

- There is no explicit regulatory requirement or license condition to maintain an on-site meteorological program for a decommissioning unit (or facility), and

- From a technical perspective, a reasonably conservative estimate of dose to a member of the public in the unrestricted area can be performed without periodically assessing changes in atmospheric dispersion and deposition based on our low site source term and the conservative nature of the dispersion factors.

See NTM 229460-80 Met Tower Abandonment White Paper for additional details.

Offsite Dose Calculation Manual (ODCM) Changes The ODCM was revised in 2018 to revision #38 to support various system abandonments as the facility is configured for SAFSTOR. The following changes were incorporated:

ODCM Revision 38 Summary of Changes Page(s) CHANGE Introduction Removed sentence referring to methodology for calculating effluent monitor setpoints.

1-2 1) Removed definitions for: Channel calibration, Channel check, Channel function test, Mode, Ventilation Exhaust Treatment System, Purge - Purging, Operable - Operability and Source Checks. All of these pertain to Effluent monitors which will no longer be required per procedure CP-500, as supported by CR3 calculation N18-0004 and associated 50.59 evaluation. 2)

Removed reference to Improved Tech Specs in the Note box. 3) Removed reference to evaporators in the Liquid radwaste Treatment System definition. 4) Removed reference to Alarm Trip Setpoints in the ODCM definition. 5) Renumbered remaining steps.

2 Definition for Unplanned Release updated to remove reference to the plant ventilation system and to indicate the reduced risk due to plant shutdown.

Removed definitions for Ventilation treatment System and Purge. Re-numbered steps.

3 Removed Sections 2.1, 2.2 and 2.4. These pertain to liquid and gaseous effluent rad monitoring and gaseous treatment. Gas and liquid monitors, and gas treatment, will no longer be required due to the decreased CR3 source term as documented in calculation N18-0004 and the associated 50.59 evaluation for CP-500.

5 Table 2-5: 1) Removed reference to dissolved and entrained gas as CR3 no longer has measurable quantities of radioactive gas. 2) Added Ni-63 to the quarterly composite sample. 3) Removed reference to Continuous Releases as CR3 does not make continuous liquid releases.

7 Table 2-5 Notation: Added guidance in notation e. Removed short lived radionuclides Mo-99 and Ce-141 from notation f.

20

ODCM Revision 38 Summary of Changes Page(s) CHANGE 8 Step 2.6.1: Added clarification that a dose calculation only has to be performed if a release has been made.

9 1) Removed old Step 2.7.a pertaining to noble gas dose rate and renumbered old Step 2.7.b as 2.7.a. Due to radioactive decay and relocation of spent fuel to dry storage, CR3 no longer has measurable dose rates from noble gasses.

2) Removed old step 2.7.1 pertaining to noble gas and renumbered old step 2.7.2 as 2.7.1. 3) Deleted information in Step 2.8 pertaining to noble gas dose as CR3 no longer has a noble gas dose source term.

10 Table 2-6: Removed reference to continuous noble gas monitor. It is not needed as per CP-500, as supported by Calculation N18-0004 and the associated 50.59 evaluation. Added superscript (d) to Aux Building Exhaust in column 1.

12 Table 2-6 Notation: Notation (d): Removed note for which pertained to a continuous air sampler and replaced with note to determine flow rate.

Notation (f) Removed short lived radionuclides Mo-99 and Ce-141 from step f.

Removed row for continuous noble gas monitoring.

13 Added step 2.9.2 14 Step 2.10.a: Removed reference to old Specifications 2.6.a, 2.6.b, 2.8.a, and 2.8.b.

Step 2.10.1: Removed reference to old Specifications 2.6.1 and 2.8.1. Added sentence to account for direct radiation shine and the 25 mrem annual limit of 40 CFR 190 to address Nuclear Oversight NCR 2199373 from audit 2018-CR3-SEC/RP-01.

16 Table 2.7: Removed five mile direct radiation location C78. This TLD location at the end of the north bank of the intake canal road will no longer be readily accessible in the future. ANI and State of Florida DEP agreed with C78 TLD removal.

22 Step 2.14 - Contents of old Special Reports section removed based on results of Calculation N18-0004 and the associated CP-500 50.59 evaluation that CR3 is not likely to exceed any NRC regulatory effluent release limit.

23 Old sections 3.1, 3.2 and 3.4 deleted. These pertained to liquid and effluent monitoring and the ventilation treatment system. As per Calculation N18-0004 and the associated CP-500 50.59 evaluation, these are no longer needed.

24 Old section 3.8 pertaining to Gaseous Effluents Dose Noble Gases Basis removed because there is no longer any noble gas dose at CR3.

ODCM Part II Changes Below 28 Old Section 1.0: Steps pertaining to Radioactive Effluent Monitor Setpoints Specifications was removed as per guidance in Calculation N18-0004 and the associated CP-500 50.59 evaluation. Only steps 1.3-1 and 1.3-2 were kept.

29 Section 1.3.-1: Calculations related to noble gas gamma and beta emissions were deleted due to no longer having a noble gas source term.

21

ODCM Revision 38 Summary of Changes Page(s) CHANGE 30 Section 1.3-2: Removed noble gas source term from the equation and notes and added Ni-63 to account for current CR3 source term. Removed sections 1.4-3, 1.4-4, 1.4-5, 1.4-7 and 1.4-8. These pertained to effluent monitor setpoint calculations.

31 Part II, Section 2.0: Removed old Table II and waste/dose reduction content associated with the liquid and ventilation treatment systems. As per Calculation N18-0004 and the associated CP-500 50.59 evaluation, there is no significant source term to require treatment. Also, changed title of Section 2.0 to indicate dose projection calculations instead of dose reduction specifications.

32 Part II, Section 3.0: Re-wrote sampling guidance to account for RM-A2 removal and reduced tritium source term after the spent fuel pool is drained.

Old Table III removed.

33 Part II, Section 4.0: Removed cumulative dose calculation specifications in old sections 4.1-1 to 4.3-1 as they were redundant to Part I or no longer applicable due to lack of a noble gas source term per the guidance in calculation N18-0004 and the associated CP-500 50.59. Old Table IV removed.

37 Table 4.4.2: Removed short lived radionuclides which are no longer present at CR3 in measurable quantities.

39 Table 4.4.3: Removed short lived radionuclides which are no longer present at CR3 in measurable quantities.

40 Old Tables 4.4.4 and 4.4.5: Removed Inhalation Dose factor tables for Teen and Adult as Infant dose factors are used for dose calculations and child dose factors are used for dose rate calculations. Teen and Adult are not used.

Table 4.4.6: Removed short lived radionuclides which are no longer present at CR3 is measurable quantities.

42 Old Tables 4.4.7 to 4.4.15: Removed Ingestion Dose Factor tables for all age groups and pathways except for infant and grass-cow-milk pathway. Dose to the other age groups and via other pathways is not assessed as infant grass cow milk is considered limiting.

Table 4.4-16: Removed short lived radionuclides which are no longer present at CR3 in measurable quantities.

44 Table 4.4-17: Removed short lived radionuclides which are no longer present at CR3 is measurable quantities.

45 Rewrote Section 4.6 to account for decreased gaseous release dose pathway and removed discussion on iodine and Kr-85 source term in a gaseous release. Removed old note before skin dose rate section that pertained to monitor setpoints. Removed old section pertaining to iodines under the skin dose rate calculation.

22

ODCM Revision 38 Summary of Changes Page(s) CHANGE 46 Removed reference to Permanently Defueled Emergency Plan and used a value of 1.5 mrem in the calculation instead of the old emergency plan Alert limit of 10 mrem. The new IOEP does not have an Alert dose threshold limit.

49 Table 5.1-1: Updated distance for Stations C13 and C14G. These were moved slightly closer to CR3 to make them more accessible.

52 Table 5.1-4: Removed TLD location C78. This TLD location at the end of the north bank of the intake canal road will no longer be readily accessible in the future.

57 Section 6.0: Administrative Controls - Removed this section as administrative controls are already incorporated in plant procedures. Removing this section also addresses Nuclear Oversight IMPR recommendation 2199386 from audit 2018-CR3-SEC/RP-01 about releases to Settling Ponds.

Process Control Program (PCP) Changes The PCP was not revised in 2018.

Carbon-14 Evaluation The plant has not operated since 2009 so there is no source term generation for carbon-14 production. Since the decision to retire the facility has been announced, there will be no C-14 source term generated ever again at CR-3.

Nuclear Electric Institute (NEI) Required Information The following environmental data is being included in this report per objective 2.4.b.i and 2.4.b.ii of NEI 07-07 Industry Ground Water Protection Initiative, as this groundwater well data is used to assist in evaluation of groundwater at the site, but is not officially included in the Radiological Environmental Monitoring Program (REMP) or the Offsite Dose Calculation Manual (ODCM).

These 2 graphs are of tritium measurements in units of pCi/l, taken from groundwater monitoring wells located west of CR-3 on either side (north and south) of the site settling percolation ponds.

There are many other groundwater monitoring wells included in the REMP that are used for evaluating the groundwater in the vicinity of the CR-3 site. These two wells are providing supplemental information. The LLD for tritium measurement of these environmental well samples is ~180 pCi/l. Measurements over the past several years have not showed tritium above LLD.

23

Tritium Measurements GW Well # MWC-IF2 600 500 400 pCi/l 300 200 100 0

10/10/06 02/22/08 07/06/09 11/18/10 04/01/12 08/14/13 12/27/14 05/10/16 09/22/17 02/04/19 06/18/20 Tritium Measurements GW Well # MWC-27 1400 1200 1000 800 pCi/l 600 400 200 0

10/10/06 02/22/08 07/06/09 11/18/10 04/01/12 08/14/13 12/27/14 05/10/16 09/22/17 02/04/19 06/18/20 24

Additional Information On February 5, 2013, Duke Energy announced that a decision has been made to permanently retire Crystal River Unit 3. The decision was made due to the high cost of repair and risk associated with repairing the containment buildings delaminated concrete wall. The company is working to develop a comprehensive decommissioning plan and intends to place the facility in SAFSTOR for the immediate future and eventual dismantling. The plant staff (called SAFSTOR 1 organization) is working to shut down and abandon as many systems as possible, by removing energy sources, lubrications, greases, electrical, and system fluids to prepare the unit for SAFSTOR and eventual dismantlement. All spent fuel has been relocated from the spent fuel pool to the ISFSI facility as of January 2018.

25

DUKE ENERGY FLORIDA, LLC DOCKET NUMBER 50 - 302 / LICENSE NUMBER DPR - 72 ATTACHMENT 2 CR3 OFFSITE DOSE CALCULATION MANUAL

CRYSTAL RIVER UNIT 3 OFF-SITE DOSE CALCULATION MANUAL PNSC & PM approval documented in PNSC Meeting 2018-02 on 10/9/18 APPROVED BY: Bryant Akins on file Radiation Protection & Chemistry Manager DATE: 11/1/18 REVISION: 38 APPROVED BY: Interpretation Contact Chuck Burtoff on file Lead Scientist

INTRODUCTION The Off-Site Dose Calculation Manual (ODCM) is provided to support implementation of the Crystal River Unit 3 radiological effluent controls. The ODCM is divided into two parts.

Part I contains the specifications for liquid and gaseous radiological effluents and the radiological environmental monitoring program which were relocated from the Technical Specifications in accordance with the provisions of Generic Letter 89-01 issued by the NRC in January 1989. Part II of the ODCM contains the calculation methods used in determining the dose to members of the public resulting from routine radioactive effluents released from Crystal River Unit 3.

The ODCM shall become effective after acceptance by the Plant Nuclear Safety Committee and approval by the Plant Manager in accordance with plant procedures. Changes to the ODCM shall be documented and records of reviews performed shall be retained. This documentation shall contain sufficient information to support the change (including analyses or evaluations), and a determination that the change will maintain the level of radioactive effluent control required by the regulations listed in Technical Specification and not adversely impact the accuracy or reliability of effluent, dose, or setpoint calculations.

Changes shall be submitted to the NRC in the form of a complete and legible copy of the entire ODCM as part of, or concurrent with, the Radioactive Effluent Release Report for the period of the report in which any change to the ODCM was made. Each change shall be identified by markings in the margin of the affected pages, clearly indicating the area of the page that was changed, and shall indicate the date (e.g. month/year) the change was implemented.

TABLE OF CONTENTS PAGE PART I - SPECIFICATIONS 1.0 DEFINITIONS ..................................................................................................................1 Frequency .............................................................................................................1 Liquid Radwaste Treatment System .....................................................................1 Member Of The Public ..........................................................................................1 Offsite Dose Calculation Manual (ODCM) .............................................................1 Site Boundary ........................................................................................................1 Unplanned Release ...............................................................................................2 Unrestricted Area ..................................................................................................2 2.0 SPECIFICATIONS ...........................................................................................................3 Radioactive Liquid Effluent Monitoring Instrumentation.........................................3 Radioactive Gaseous Effluent Monitoring Instrumentation ....................................3 Liquid Radwaste Treatment System .....................................................................3 Ventillation Exhaust Treatment System.................................................................3 Liquid Effluents Concentration ..............................................................................4 Liquid Effluents - Dose ..........................................................................................8 Gaseous Effluents - Dose Rate ............................................................................9 Dose - Noble Gases ..............................................................................................9 Dose - Tritium And Radioactive Particulates ......................................................13 Total Dose ...........................................................................................................14 Radiological Environmental Monitoring ...............................................................15 Land Use Census ................................................................................................21 Interlaboratory Comparison Program ..................................................................22 Special Reports ...................................................................................................22 3.0 SPECIFICATION BASES...............................................................................................23 Not Used .............................................................................................................23 Not Used .............................................................................................................23 Liquid Radwaste Treatment System Basis ..........................................................23 Not Used .............................................................................................................23 Liquid Effluents Concentration Basis ...................................................................23 Liquid Effluents Dose Basis.................................................................................24 Gas Effluents Dose Rate Basis ...........................................................................24 Not Used .............................................................................................................24 Gaseous Effluents Dose Tritium And Radioactive Particulate Basis ...................25 Total Dose Basis .................................................................................................26 Radiological Environmental Monitoring Program Basis .......................................26 OFF-SITE DOSE CALCULATION MANUAL Rev. 38 Page i

TABLE OF CONTENTS PAGE Radiological Environmental Monitoring Program Land Use Census Basis ...................................................................................................................26 Radiological Environmental Monitoring Interlaboratory Comparison Program Basis .....................................................................................................27 Not Used .............................................................................................................27 Not Used .............................................................................................................27 Not Used .............................................................................................................27 Not Used .............................................................................................................27 OFF-SITE DOSE CALCULATION MANUAL Rev. 38 Page ii

TABLE OF CONTENTS PAGE PART II - METHODOLOGIES .............................................................................................................. 28 1.0 Pre-Release Calculations ......................................................................................................... 29 2.0 Radioactive Effluents Dose Projection Calculations .................................................................. 31 3.0 Radioactive Effluents Sampling Specifications ......................................................................... 32 4.0 Radioactive Effluents Dose Calculation Specifications ............................................................. 33 5.0 Environmental Monitoring ......................................................................................................... 48 OFF-SITE DOSE CALCULATION MANUAL Rev. 38 Page iii

PART I LIST OF TABLES Table Page 2-5 Radioactive Liquid Waste Sampling and Analysis Program ..................................................... 5 2-6 Radioactive Gaseous Waste Sampling and Analysis Program .............................................. 10 2-7 Operational Radiological Environmental Monitoring Program ................................................ 16 2-8 Reporting Levels for Radioactivity Concentrations in Environmental Samples ....................... 18 2-9 Maximum Values For The Lower Limits Of Detection (LLD) a, d............................................ 19 PART II LIST OF TABLES Table 4.4-2 Inhalation Dose Factors (Ri) - Infant ...................................................................................... 37 4.4-3 Inhalation Dose Factors (Ri) - Child ...................................................................................... 39 c

4.4-6 Ingestion Dose Factors ( R i ) Grass-Cow-Milk Pathway (Infant) .............................................. 40 4.4-16 Dose Factors Ground Plane Pathway .................................................................................... 42 4.4-17 Liquid Effluent - Adult Ingestion Dose Factors ....................................................................... 44 5.1-1 Environmental Radiological Monitoring Station's Locations ................................................... 49 5.1-2 Environmental Radiological Monitoring Station's Locations -

Groundwater Monitoring Wells............................................................................................... 50 5.1-3 Ring TLDs (Inner Ring) .......................................................................................................... 51 5.1-4 Ring TLDs (5 Mile Ring) ........................................................................................................ 52 Note: Numbered tables not listed are not used.

LIST OF FIGURES Figure 5.1 Environmental Monitoring Sample Station Locations ............................................................. 53 5.2 Environmental Monitoring TLD Locations .............................................................................. 54 5.3 Environmental Monitoring TLD Environmental Monitoring TLD Locations (5 mile) ................. 55 5.4 CR3 Groundwater Monitoring Well Locations ........................................................................ 56 Change Summary ................................................................................................................................ 58 OFF-SITE DOSE CALCULATION MANUAL Rev. 38 Page iv

PART I SPECIFICATIONS OFF-SITE DOSE CALCULATION MANUAL Rev. 38 Page v

1.0 DEFINITIONS FREQUENCY NOTATION FREQUENCY S At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

D At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

W At least once per 7 days.

M At least once per 31 days.

Q At least once per 92 days.

SA At least once per 6 months.

Y At least once per 12 months.

R At least once per 18 months.

P Completed prior to each release.

N.A. Not applicable.

NOTE: Surveillance frequencies are met if the surveillance is performed within 1.25 times the interval specified, as measured from the previous performance or as measured from the time a specified condition of the frequency is met.

LIQUID RADWASTE TREATMENT SYSTEM The Liquid Radwaste Treatment System shall be any available equipment (e.g., filters) capable of reducing the quantity of radioactive material, in liquid effluents, prior to discharge.

MEMBER OF THE PUBLIC MEMBER OF THE PUBLIC means an individual in a controlled or unrestricted area.

However, an individual is not a member of the public during any period in which the individual receives an occupational dose.

OFFSITE DOSE CALCULATION MANUAL (ODCM)

The ODCM contains the methodology and parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluents, and in the conduct of the Radiological Environmental Monitoring Program (REMP). The ODCM also contains information that should be included in the Annual Radiological Environmental Operating and Radioactive Effluent Release Reports.

SITE BOUNDARY The Site Boundary shall be that line beyond which the land is not owned, leased, or otherwise controlled by the licensee.

OFF-SITE DOSE CALCULATION MANUAL Rev. 38 Page 1

UNPLANNED RELEASE An Unplanned Release is an unintended discharge of liquid or airborne radioactivity to the environment. Due to plant shutdown, the chance of having an unplanned release is significantly reduced. Minor equipment failures which cause an increase in plant releases are not unplanned as it is expected that minor failures will occur from time-to-time. Human error which results in a release of radioactivity to the environment is considered unplanned.

EXAMPLES: 1. Releasing the wrong waste tank.

2. Plant leakage which exceeds reporting limits such as those of 10 CFR 50.72 and 10 CFR 50.73.

UNRESTRICTED AREA An Unrestricted Area shall be any area at or beyond the site boundary, access to which is not controlled by the licensee for purposes of protection of individuals from exposure to radiation and radioactive materials, or any area within the site boundary used for residential quarters or industrial, commercial, institutional, and/or recreational purposes.

OFF-SITE DOSE CALCULATION MANUAL Rev. 38 Page 2

2.0 SPECIFICATIONS RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION Not Used RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION Not Used LIQUID RADWASTE TREATMENT SYSTEM The Liquid Radwaste Treatment System shall be used, as required, to reduce radioactive materials in liquid wastes prior to their discharge, when projected monthly doses due to liquid effluents discharged to Unrestricted Areas would exceed the following values:

a. 0.06 mrem whole body;

b. 0.2 mrem to any organ APPLICABILITY: At all times.

ACTION:

a. When radioactive liquid waste, in excess of the above limits, is discharged without prior treatment, prepare and submit to the Commission within 30 days, a Special Report which includes the following information:

1. Identification of inoperable equipment and the reasons for inoperability.

2. Actions taken to restore the inoperable equipment to operable status.

3. Actions taken to prevent recurrence.

SURVEILLANCE REQUIREMENTS 2.3.1 Doses due to liquid releases shall be projected at least once per 31 days if a liquid release was made.

VENTILLATION EXHAUST TREATMENT SYSTEM Not Used OFF-SITE DOSE CALCULATION MANUAL Rev. 38 Page 3

LIQUID EFFLUENTS CONCENTRATION The concentration of radioactive material released to Unrestricted Areas shall be less than or equal to 10 times the concentrations specified in 10 CFR Part 20, Appendix B, Table II, Column 2 for radionuclides other than dissolved or entrained noble gases.

APPLICABILITY: At all times.

ACTION:

a. With the concentration of radioactive materials released to Unrestricted Areas exceeding the above limits, without delay restore the concentration of radioactive materials being released to Unrestricted Areas to within the above limits.

SURVEILLANCE REQUIREMENTS 2.5.1 Radioactive liquid wastes shall be sampled and analyzed in accordance with the sampling and analysis program of Table 2-5.

2.5.2 The results of the radioactivity analyses shall be used to assure the concentrations of radioactive material released from the site are maintained within the limits of Specification 2.6.

OFF-SITE DOSE CALCULATION MANUAL Rev. 38 Page 4

TABLE 2-5 RADIOACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM Minimum Lower Limit of Sampling Type of Activity Detection (LLD)

Analysis Liquid Release Type Frequency Analysis Frequency (µCi/ml)a A. Batch Waste P P Principal Gamma 5x10-7 Release Tanksd,e Each Batch Each Batch Emittersf

1. Evaporator Condensate Storage Tanks (2)

2. Laundry & Shower Sump Tanks (2)

P M

3. Secondary Drain H-3 1x10-5 Each Batch Compositeb,c, Tank Gross Alpha 1x10-7 P Q Sr-89, Sr-90 5x10-8 Each Batch Compositeb,c Fe-55, Ni-63 1x10-6 Note: The BWST may be used as a batch waste tank for final disposal of spent fuel pool water. The ODCM criteria relating to ECST batch releases (e.g. representative sampling, etc.) will be followed.

OFF-SITE DOSE CALCULATION MANUAL Rev. 38 Page 5

TABLE 2-5 (Continued)

TABLE NOTATION

a. The LLD* is the smallest concentration of radioactive material in a sample that will be detected with 95% probability with 5% probability of falsely concluding that a blank observation represents a "real" signal.

For a particular measurement system (which may include radiochemical separation):

LLD = 4.66sb / (2.22x106 E V Y e-t)

Where:

LLD is the lower limit of detection as defined above (as microcurie per unit mass or volume),

sb is the standard deviation of the background counting rate or of the counting rate of a blank sample as appropriate (as counts per minute),

E is the counting efficiency (as counts per disintegration),

V is the sample size (in units of mass or volume),

2.22x106 is the number of disintegrations per minute per microcurie, Y is the fractional radiochemical yield (when applicable),

is the radioactive decay constant for the particular radionuclide, and t is the elapsed time between midpoint of sample collection and time of counting (for plant effluents, not environmental samples).

Typical values of E, V, Y, and t shall be used in the calculation.

• The LLD is defined as an a priori (before the fact) limit representing the capability of a measurement system and not as an a posteriori (after the fact) limit for a particular measurement.

b. A composite sample is one in which the quantity of liquid sampled is proportional to the quantity of liquid waste discharged and in which the method of sampling employed results in a specimen which is representative of the liquids released.

c. Prior to analyses, all samples taken for the composite shall be thoroughly mixed in order for the composite sample to be representative of the effluent release.

OFF-SITE DOSE CALCULATION MANUAL Rev. 38 Page 6

TABLE 2-5 (Continued)

TABLE NOTATION

d. A batch release is the discharge of liquid wastes of a discrete volume. Prior to sampling for analyses, each batch shall be isolated, and then thoroughly mixed to assure representative sampling.

e. Means shall exist to determine waste volume, waste discharge flow rate, and dilution flow rate for each batch release.

f. The principal gamma emitters for which the LLD specification applies exclusively are the following radionuclides: Mn-54, Fe-59, Co-58, Co-60, Zn-65, Cs-134, Cs-137, and Ce-144. This list does not mean that only these nuclides are to be detected and reported. Other peaks, which are measurable and identifiable, together with the above nuclides, shall also be identified and reported. Nuclides which are below the LLD for the analyses shall be reported as "less than" the nuclide's LLD, and shall not be reported as being present at the LLD level for that nuclide. The "less than" values shall not be used in the required dose calculations.

OFF-SITE DOSE CALCULATION MANUAL Rev. 38 Page 7

LIQUID EFFLUENTS - DOSE The dose or dose commitment to a Member Of The Public from radioactive materials in liquid effluents released to Unrestricted Areas shall be limited as follows:

a. During any calendar quarter to less than or equal to 1.5 mrem to the total body and less than or equal to 5 mrem to any organ.

b. During any calendar year to less than or equal to 3 mrem to the total body and to less than or equal to 10 mrem to any organ.

APPLICABILITY: At all times.

ACTION:

a. With the calculated dose from the release of radioactive materials in liquid effluents exceeding any of the above limits, prepare and submit to the Commission, within 30 days, which includes:

1. Identification of the cause for exceeding the limit(s);

2. Corrective action taken to reduce the release of radioactive materials in liquid effluents during the remainder of the current calendar quarter an during the remainder of the current calendar year so that the dose or dose commitment to a Member Of The Public from this source is less than or equal to 3 mrem total body and less than or equal to 10 mrem to any organ during the calendar year.

SURVEILLANCE REQUIREMENTS 2.6.1 DOSE CALCULATIONS. Cumulative dose contributions from liquid effluents shall be determined at least once per 31 days if a release has been made.

OFF-SITE DOSE CALCULATION MANUAL Rev. 38 Page 8

Gaseous Effluents - Dose Rate The dose rate at or beyond the Site Boundary, due to radioactive materials released in gaseous effluents, shall be limited as follows:

a. Tritium and radioactive particulates with half-lives of greater than 8 days: less than or equal to 1500 mrem/year to any organ.

APPLICABILITY: At all times ACTION:

a. With dose rate (s) exceeding the above limits, without delay decrease the dose rate to within the above limit(s).

SURVEILLANCE REQUIREMENTS 2.7.1 The dose rate due to radioactive materials specified above, other than noble gases, in gaseous effluents shall be determined to be within the above limits by obtaining representative samples and performing analyses in accordance with Table 2-6.

Dose - Noble Gases Not Used OFF-SITE DOSE CALCULATION MANUAL Rev. 38 Page 9

TABLE 2-6 RADIOACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM Lower Limit of Sampling Minimum Analysis Detection (LLD)

Gaseous Release Type Type of Activity Analysis Frequency Frequency

(µCi/ml)a A. Not Used B. Not Used C. Auxiliary Building and Fuel Handling Area M Principal Gamma Emitters f 1x10-4 Exhaust d Grab Sample H-3 1x10-6 M

Principal Gamma Emitters f Grab Sample e Particulate (Others) 1x10-11 Sample M

Composite Grab Sample e Particulate Gross Alpha 1x10-11 Sample Q

Composite Grab Sample e Particulate Sr-89, Sr-90 1x10-11 Sample OFF-SITE DOSE CALCULATION MANUAL Rev. 38 Page 10

TABLE 2-6 (Continued)

TABLE NOTATION

a. The LLD* is the smallest concentration of radioactive material in a sample that will be detected with 95% probability with 5% probability of falsely concluding that a blank observation represents a "real" signal.

For a particular measurement system (which may include radiochemical separation):

LLD = 4.66sb/(2.22x106 EVY e-t)

Where:

LLD is the lower limit of detection as defined above (as microcurie per unit mass or volume),

sb is the standard deviation of the background counting rate or of the counting rate of a blank sample as appropriate (as counts per minute),

E is the counting efficiency (as counts per disintegration),

V is the sample size (in units of mass or volume),

2.22x106 is the number of disintegrations per minute per microcurie, Y is the fractional radiochemical yield (when applicable),

is the radioactive decay constant for the particular radionuclide, and t is the elapsed time between midpoint of sample collection and time of counting (for plant effluents, not environmental samples).

Typical values of E, V, Y, and t shall be used in the calculation.

• The LLD is defined as an a priori (before the fact) limit representing the capability of a measurement system and not as an a posteriori (after the fact) limit for a particular measurement.

b. Not Used OFF-SITE DOSE CALCULATION MANUAL Rev. 38 Page 11

TABLE 2-6 (Continued)

TABLE NOTATION

c. Not Used

d. Means shall exist to determine ventilation exhaust flow rate. Samples shall be obtained using calibrated samplers

e. The ratio of the sample flow rate to the sampled stream flow rate shall be known for the time period covered by each dose or dose rate calculation made in accordance with the Specifications 2.7 and 2.9.

f. The principal gamma emitters for which the LLD specification applies exclusively are the following radionuclides: Mn-54, Fe-59, Co-58, Co-60, Zn-65, Cs-134, Cs-137 and Ce-144 for particulate emissions. This list does not mean that only these nuclides are to be detected and reported.

Other peaks, which are measurable and identifiable, together with the above nuclides, shall also be identified and reported. Nuclides which are below the LLD for the analyses shall be reported as "less than" the nuclide's LLD and shall not be reported as being present at the LLD level for that nuclide. The "less than" values shall not be used in the required dose calculations.

OFF-SITE DOSE CALCULATION MANUAL Rev. 38 Page 12

DOSE - TRITIUM AND RADIOACTIVE PARTICULATES The dose to a Member Of The Public from Tritium and radioactive particulates with half-lives greater than 8 days in gaseous effluents released from the site to areas at or beyond the Site Boundary shall be limited as follows:

a. During any calendar quarter: less than or equal to 7.5 mrem to any organ, and

b. During any calendar year: less than or equal to 15 mrem to any organ.

APPLICABILITY: At all times.

ACTION:

a. With the calculated dose from the release of Tritium and radioactive particulates with greater than 8 day half-lives, in gaseous effluents, exceeding any of the above limits, prepare and submit to the Commission, within 30 days, a report which includes:

1) Identification of the cause for exceeding the limits(s);

2) Corrective action to reduce those releases during the remainder of the current calendar quarter and the remainder of the current calendar year so that the average dose to any organ is less than or equal to 15 mrem.

b. If the projected monthly dose exceeds 0.3 mrem to any organ, then action shall be taken to reduce the quantity of radioactive material in the effluents:

SURVEILLANCE REQUIREMENTS 2.9.1 DOSE CALCULATIONS: Cumulative dose calculations for the current calendar quarter and current calendar year shall be determined at least once per 31 days.

2.9.2 DOSE PROJECTIONS: Dose shall be projected at least once per 31 days OFF-SITE DOSE CALCULATION MANUAL Rev. 38 Page 13

TOTAL DOSE The calendar year dose or dose commitment to any Member Of The Public, due to releases of radioactivity and radiation, from uranium fuel cycle sources shall be limited to less than or equal to 25 mrem to the total body or any organ (except the thyroid, which shall be limited to less than or equal to 75 mrem).

APPLICABILITY: At all times.

ACTION:

a. With the calculated doses from the release of radioactive materials in liquid or gaseous effluents exceeding twice the limits of Specification 2.9.a or 2.9.b, calculations should be made, which include direct radiation contributions from the reactor, to determine whether the above limits of Specification 2.10 have been exceeded. If such is the case, prepare and submit to the Commission within 30 days, a report that defines the corrective action to be taken to reduce subsequent releases to prevent recurrence of exceeding the above limits and includes the schedule for achieving conformance with the above limits. This report, as defined in 10 CFR Part 20.2203, shall include an analysis that estimates the radiation exposure (dose) to a Member Of The Public from uranium fuel cycle sources, including all effluent pathways and direct radiation, for the calendar year that includes the release(s) covered by this report. It shall also describe levels of radiation and concentrations of radioactive material involved, and the cause of the exposure levels or concentrations. If the estimated dose(s) exceeds the above limits, and if the release condition resulting in violation of 40 CFR Part 190 has not already been corrected, the report shall include a request for a variance in accordance with the provisions of 40 CFR Part 190. Submittal of the report is considered a timely request, and a variance is granted until staff action on the request is complete.

SURVEILLANCE REQUIREMENTS 2.10.1 DOSE CALCULATIONS - Cumulative dose contributions from liquid and gaseous effluents shall be determined in accordance with Specifications 2.9.1. Total dose must also consider shine from the ISFSI pad and must be less than 25 mrem in a year as per 40 CFR 190. Total dose must also consider shine from the ISFSI pad and must be less than 25 mrem in a year per 40 CFR 190.

Note: Procedure ISFS-212, section 5.2.1, equates the ISFSI controlled area boundary to CR3s owner controlled area boundary as described in the DSAR. Real individuals beyond the controlled area are where the ISFSI dose limits of 72.104 apply. These are the same dose limits imposed by 40 CFR 190, which is the basis for the limits of ODCM Specification 2.10 above. ISFSI calculation 11182-0502 demonstrates that the highest annual dose at the controlled area boundary is ~ 0.1 mrem.

OFF-SITE DOSE CALCULATION MANUAL Rev. 38 Page 14

RADIOLOGICAL ENVIRONMENTAL MONITORING The radiological environmental monitoring program shall be conducted as specified in Table 2-7.

APPLICABILITY: At all times.

ACTION:

a. With the radiological environmental monitoring program not being conducted as specified in Table 2-7, prepare and submit to the Commission, in the Annual Radiological Environmental Operating Report, a description of the reasons for not conducting the program as required and the plans for preventing a recurrence.

b. With the level of radioactivity, resulting from plant effluents, in an environmental sampling medium exceeding the reporting levels of Table 2-8 when averaged over any calendar quarter, prepare and submit to the Commission, within 30 days of obtaining analytical results from the affected sampling period, a report, which identifies the cause(s) for exceeding the limit(s) and defines corrective actions to be taken to reduce radioactive effluents so that the potential annual dose to a Member Of The Public is less than the calendar year limits of Specifications 2.7, 2.8, and 2.9. When more than one of the radionuclides in Table 2-8 are detected in the sampling medium, this report shall be submitted if:

concentration (1) concentration (2) limit level (1) + limit level (2) + .. 1.0 When radionuclides other than those in Table 2-8 are detected and are the result of plant effluents, this report shall be submitted if the potential annual dose to a MEMBER OF THE PUBLIC is greater than or equal to the calendar year limits of Specifications 2.7, 2.8, and 2.9. This report is not required if the measured level of radioactivity was not the result of plant effluents; however, in such an event, the condition shall be reported and described in the Annual Radiological Environmental Operating Report.

c. With milk or fresh leafy vegetation samples unavailable from one or more of the sample locations required by Table 2-7, identify the cause of the unavailability of samples and identify locations for obtaining replacement samples in the next Annual Radiological Environmental Operating Report. The locations from which samples were unavailable may then be deleted from those required by Table 2-7, provided the locations from which the replacement samples were obtained are added to the environmental monitoring program as replacement locations.

SURVEILLANCE REQUIREMENTS 2.11.1 The radiological environmental monitoring samples shall be collected pursuant to Table 2-7 from the locations given in the table and Figures 5.1, 5.2, and 5.3 and shall be analyzed pursuant to the requirements of Tables 2-7 and 2-9.

OFF-SITE DOSE CALCULATION MANUAL Rev. 38 Page 15

TABLE 2-7 OPERATIONAL RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM Exposure Pathway Number of Samples Sampling/

Type/Frequency of Analysis and/or Sample and Locations Collection Frequency

1. AIRBORNE One sample each: Continuous sampler/

Particulate sampler:

particulates C07, C18, C40, C41, Weekly collection C46 and Control a) Gross at 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />s/

Location C47 following weekly filter change.

b) Composite gamma special analysis (by location)/

quarterly. (Gamma Spectral Analysis shall also be performed on individual samples if gross beta activity of any sample is greater than 1.0 pCi/m3 and which is also greater than ten times the control sample activity.

2. DIRECT 1) Site Boundary: Continuous Gamma exposure rate/quarterly RADIATION C60, C61, C62, placement/Quarterly C63, C64, C65, collection C66, C67, C68, C69, C41, C70, C27, C71, C72, C73

2) Five Miles:

C18, C03, C04, C74, C75, C76, C08, C77, C09, C14G, C01, C79

3) Control Location:

C47 OFF-SITE DOSE CALCULATION MANUAL Rev. 38 Page 16

TABLE 2-7 (Continued)

OPERATIONAL RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM Exposure Pathway Number of Samples Sampling/

Type/Frequency of Analysis and/or Sample and Locations Collection Frequency

3. WATERBORNE One sample each: Grab sample/Monthly Gamma spectral analysis/monthly Seawater C14H, C14G Control Location C13 Tritium analysis on each sample or on a quarterly composite of monthly samples Ground water One sample: Grab Gamma spectral and Tritium C40 (Control Location) sample/semiannual analysis/each sample Site Ground Water One sample each: Grab sample/quarterly Gamma spectral and Tritium CR3-1S, CR3-1D, analysis/each sample CR3-2, CR3-3S, CR3-3D, CR3-4, CR3-5, CR3-6S, CR3-6D, CR3-7 CR3-8, CR3-9, CR3-10 Drinking water One sample each: Grab sample/quarterly Gamma spectral and Tritium C07, C10, C18 (All analysis/each sample Control Locations)

Shoreline Sediment One sample each: Semiannual sample Gamma spectral analysis/each C14H, C14M, C14G sample Control Location C09

4. INGESTION One sample each: Quarterly: Gamma spectral analysis on Fish & C29, Control Location Oysters and carnivorous edible portions/each sample Invertebrates C30 fish Food Products One sample each: Monthly (when Gamma spectral analysis/each C48a*, C48b*, Control available): Sample sample Location C47 comprised of three (3) types of broad leaf vegetation from each location One sample: C19 Annual during harvest: Gamma spectral analysis/each Citrus sample One sample: C04 Annual during harvest: Gamma spectral analysis/each Watermelon sample

• Stations C48a and C48b are located near the site boundary for gaseous effluents in the two sectors which yield the highest historical annual average D/Q values.

OFF-SITE DOSE CALCULATION MANUAL Rev. 38 Page 17

TABLE 2-8 REPORTING LEVELS FOR RADIOACTIVITY CONCENTRATIONS IN ENVIRONMENTAL SAMPLES Water Airborne Particulate or Fish Milk Food Products Analysis (pCi/l) Gases (pCi/m3) (pCi/Kg, wet) (pCi/l) (pCi/Kg, wet)

H-3 20,000(a)

Mn-54 1,000 30,000 Fe-59 400 10,000 Co-58 1,000 30,000 Co-60 300 10,000 Zn-65 300 20,000 Zr-Nb-95(b) 400 Cs-134 30 10 1,000 60 1,000 Cs-137 50 20 2,000 70 2,000 Ba-La-140(b) 200 300 (a) For drinking water samples. This is 40 CFR Part 141 value. If no drinking water pathway exists, a value of 30,000 pCi/l may be used. At the Crystal River site, there is no drinking water pathway due to the direction of groundwater flow being west-southwest towards the Gulf of Mexico and the fact that the groundwater at the site is too saline for human consumption.

(b) An equilibrium mixture of the parent and daughter isotope which contains the reporting value of the parent isotope.

OFF-SITE DOSE CALCULATION MANUAL Rev. 38 Page 18

TABLE 2-9 MAXIMUM VALUES FOR THE LOWER LIMITS OF DETECTION (LLD) a, d Water Airborne Particulate or Fish Milk Food Products Sediment Analysis (pCi/l) Gases (pCi/m )

3 (pCi/Kg, wet) (pCi/l) (pCi/Kg, wet) (pCi/Kg, dry) gross beta 0.01 3H 2000b 54Mn 15 130 59Fe 30 260 58Co 15 130 60Co 15 130 65Zn 30 260 95Zr-Nb 15c 134Cs 15 0.05e 130 15 60 150 137Cs 18 0.06e 150 18 80 180 140Ba-La 15c 15c OFF-SITE DOSE CALCULATION MANUAL Rev. 38 Page 19

TABLE 2-9 (Continued)

TABLE NOTATION

a. The LLD* is the smallest concentration of radioactive material in a sample that will be detected with 95% probability with 5% probability of falsely concluding that a blank observation represents a "real" signal.

For a particular measurement system (which may include radiochemical separation):

LLD = 4.66sb / (2.22 E V Y e-t)

Where:

LLD is the lower limit of detection as defined above (as picocurie per unit mass or volume),

sb is the standard deviation of the background counting rate or of the counting rate of a blank sample as appropriate (as counts per minute),

E is the counting efficiency (as counts per disintegration),

V is the sample size (in units of mass or volume),

2.22 is the number of disintegrations per minute per picocurie, Y is the fractional radiochemical yield (when applicable),

is the radioactive decay constant for the particular radionuclide, and t is the elapsed time between environmental collection, or end of the sample collection period, and time of counting.

Typical values of E, V, Y, and t shall be used in the calculation.

• The LLD is defined as an a priori (before the fact) limit representing the capability of the measurement system and not as an a posteriori (after the fact) limit for a particular measurement.

Analyses shall be performed in such a manner that the stated LLD's will be achieved under routine conditions. Occasionally, background fluctuations, unavoidable small sample sizes, the presence of interfering nuclides, or other uncontrollable circumstances may render these LLD's unachievable. In such cases, the contributing factors shall be identified and described in the Annual Radiological Environmental Operating Report.

b. LLD for drinking water. If no drinking water pathway exists, a value of 3000 pCi/l may be used.

See drinking water pathway discussion in Table 2-8(a).

c. The specified LLD is for an equilibrium mixture of parent and daughter nuclides which contain 15 pCi/l of the parent nuclide.

d. Other peaks which are measurable and identifiable, together with the radionuclides in Table 2.9, shall be identified and reported.

e. Cs-134, and Cs-137 LLD's apply only to the quarterly composite gamma spectral analysis, not to analyses of single particulate filters.

f. Not Used OFF-SITE DOSE CALCULATION MANUAL Rev. 38 Page 20

LAND USE CENSUS A land use census shall be conducted and shall identify the location of the nearest milk animal, the nearest residence and the nearest garden* of greater than 500 square feet producing fresh leafy vegetables in each of the land based meteorological sectors within a distance of five miles.

APPLICABILITY: At all times.

ACTION:

a. With a land use census identifying a location(s) that yields a calculated dose or dose commitment greater than the values currently being calculated by Specification 2.9.1, identify the new location in the next Annual Radiological Environmental Operating Report.

b. With a land use census identifying a location(s) which yields a calculated dose or dose commitment (via the same exposure pathway) which is at least 20% greater than at a location from which samples are currently being obtained in accordance with Specification 2.11, this location shall be added to the radiological environmental monitoring program within 30 days. The new sampling location shall replace the present sampling location, which has the lower calculated dose or dose commitment (via the same exposure pathway), after June 30 following this land use census. Identification of the new location and revisions of the appropriate figures shall be submitted with the next Radioactive Effluent Release Report.

• Broad leaf vegetation sampling may be performed at the site boundary in the direction sector with the highest D/Q in lieu of the garden census.

SURVEILLANCE REQUIREMENTS 2.12.1 The land use census shall be conducted at least once per 12 months during the growing season by a door-to-door survey, aerial survey, or by consulting local agriculture authorities, using that information which will provide adequate results.

OFF-SITE DOSE CALCULATION MANUAL Rev. 38 Page 21

INTERLABORATORY COMPARISON PROGRAM Analyses shall be performed on radioactive materials supplied as part of an Interlaboratory Comparison Program which has been approved by the Commission. A summary of the results obtained from this program shall be included in the Annual Radiological Environmental Operating Report.

APPLICABILITY: At all times.

ACTION:

a. With analyses not being performed as required above, report the corrective actions taken to prevent a recurrence to the Commission in the Annual Radiological Environmental Operating Report.

SURVEILLANCE REQUIREMENTS 2.13.1 No surveillance requirements other than those required by the Interlaboratory Comparison Program.

SPECIAL REPORTS Not Used OFF-SITE DOSE CALCULATION MANUAL Rev. 38 Page 22

3.0 SPECIFICATION BASES Not Used Not Used LIQUID RADWASTE TREATMENT SYSTEM BASIS The requirement that these systems be used when specified provides assurance that the releases of radioactive materials in liquid effluents will be kept "as low as reasonably achievable" (ALARA). This specification implements the requirements of 10 CFR Part 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50 and the design objective given in Section II.D of Appendix I to 10 CFR Part 50. The specified limits governing the use of appropriate portions of the liquid radwaste treatment system were specified as a suitable fraction of the dose design objectives set forth in Section II.A of Appendix I, 10 CFR Part 50, for liquid effluents.

Not Used LIQUID EFFLUENTS CONCENTRATION BASIS This specification is provided to ensure that the concentration of radioactive materials released in liquid waste effluents to Unrestricted Areas will be less than 10 times the effluent concentration limits (ECLs) specified in 10 CFR Part 20. This limitation provides additional assurance that the levels of radioactive materials in bodies of water in Unrestricted Areas will result in exposures within the Section II.A design objectives of Appendix I, 10 CFR 50, to a Member Of The Pubic.

There are no Noble gases seen in the typical waste water due to lengthy radioactive decay.

OFF-SITE DOSE CALCULATION MANUAL Rev. 38 Page 23

3.0 SPECIFICATION BASES (Cont'd)

LIQUID EFFLUENTS DOSE BASIS This specification is provided to implement the requirements of Sections II.A. Ill-A and IV.A of Appendix I, 10 CFR Part 50. The Limiting Condition for Operation implements the guides set forth in Section II.A of Appendix I. The ACTION statement provides the required operating flexibility and at that same time implements the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive material in liquid effluents will be kept "as low as is reasonably achievable" (ALARA). The dose calculations in the ODCM implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data, such that the actual exposure of a Member Of The Public through appropriate pathways is unlikely to be substantially underestimated. The equations specified in the ODCM for calculating the doses due to the actual release rates of radioactive materials in liquid effluents are consistent with the methodology provided in Regulatory Guide 1.109, "Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I," Revision 1, October 1977 and Regulatory Guide 1.113, "Estimating Aquatic Dispersion of Effluents from Accidental and Routine Reactor Releases for the Purpose of Implementing Appendix I," April 1977.

GAS EFFLUENTS DOSE RATE BASIS These limits provide reasonable assurance that radioactive material discharged in gaseous effluents will not result in the exposure of a member of the public, either within or outside the site boundary to annual average concentrations exceeding the limits specified in Appendix B, Table 2 of 10 CFR 20. For a member of the public who may at time be within the site boundary, the occupancy of the member of the public will be sufficiently low to compensate for any increase in the atmospheric diffusion factor above that for the site boundary. The specified release rate limits restrict, at all times, the corresponding gamma and beta dose rates above to a member of the public at or beyond the site boundary to less than or equal to 500 mrem/year to the total body or to less than or equal to 3000 mrem/year to the skin. These release rate limits also restrict, at all times, the corresponding thyroid dose rate above background to a child via the inhalation pathway to less than or equal to 1500 mrem/year.

Not Used OFF-SITE DOSE CALCULATION MANUAL Rev. 38 Page 24

3.0 SPECIFICATION BASES (Cont'd)

GASEOUS EFFLUENTS DOSE TRITIUM AND RADIOACTIVE PARTICULATE BASIS This specification is provided to implement the requirements of Sections II.C, III.A, and IV.A of Appendix I, 10 CFR Part 50. The Limiting Conditions for Operation are the guides set forth in Section II.C of Appendix I. The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive materials in gaseous effluent will be kept "as low as is reasonably achievable" (ALARA). The calculational methods specified in the Surveillance Requirements implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data, such that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated. The methods for calculating the dose due to the actual release rates of the subject materials are consistent with the methodology provided in Regulatory Guide 1.109, "Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I," Revision 1, October 1977 and Regulatory Guide 1.111, "Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors,"

Revision 1, July 1977. These equations also provide for determining the actual doses based upon the historical average atmospheric conditions. The release rate specifications for Tritium, and radioactive particulates with half-life greater than eight days are dependent on the existing radionuclide pathways to man, in areas at and beyond the SITE BOUNDARY. The pathways which were examined in the development of these calculations were:

1) Individual inhalation of airborne radionuclides,

2) deposition of radionuclides onto green leaf vegetation with subsequent consumption by man,

3) deposition onto grassy areas where milk animals and meat producing animals graze with consumption of the milk and meat by man, and

4) deposition on the ground with subsequent exposure of man.

OFF-SITE DOSE CALCULATION MANUAL Rev. 38 Page 25

3.0 SPECIFICATION BASES (Cont'd)

TOTAL DOSE BASIS This specification is provided to meet the dose limitations of 40 CFR Part 190 that have now been incorporated into 10 CFR Part 20 by 46 FR 18525. This specification applies to Members Of The Public beyond the site boundary (i.e. in the unrestricted area). The specification requires the preparation and submittal of a report whenever the calculated doses from plant radioactive effluents exceed twice the design objective doses of Appendix I. For sites containing up to 4 reactors, it is highly unlikely that the resultant dose to a Member Of The Public will exceed the dose limits of 40 CFR Part 190 if the individual reactors remain within the reporting requirement level. The report will describe a course of action that should result in the limitation of the annual dose to a Member Of The Public to within the 40 CFR Part 190 limits. For the purposes of the report, it may be assumed that the dose commitment to the Member Of The Public from other uranium fuel cycle sources is negligible, with the exception that dose contributions from other nuclear fuel cycle facilities at the same site or within a radius of 8 km must be considered. If the dose to any Member Of The Public is estimated to exceed the requirements of 40 CFR Part 190, the report with a request for a variance (provided the release conditions resulting in violation of 40 CFR Part 190 have not already been corrected), in accordance with the provisions of 40 CFR part 190.11 and 10 CFR Part 20.405c, is considered to be a timely request and fulfills the requirements of 40 CFR Part 190 until NRC staff action is completed. The variance only relates to the limits of 40 CFR Part 190 and does not apply in any way to the other requirements for dose limitation of 10 CFR Part 20, as addressed in Specifications 2.5 thru 2.9. An individual is not considered a Member Of The Public during any period in which he/she is engaged in carrying out any operation that is part of the nuclear fuel cycle.

RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM BASIS The radiological monitoring program required by this specification provides measurements of radiation and of radioactive materials in those exposure pathways and for those radionuclides which lead to the highest potential radiation exposures of Member Of The Public resulting from the station operation. This monitoring program thereby supplements the radiological effluent monitoring program by verifying that the measurable concentrations of radioactive materials and levels of radiation are not higher than expected on the basis of the effluent measurements and modeling of the environmental exposure pathways. Program changes may be initiated based on operational experience.

The LLD's required by Table 2-9 are considered optimum for routine environmental measurements in industrial laboratories. The LLD's for drinking water meet the requirements of 40 CFR 141.

RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM LAND USE CENSUS BASIS This specification is provided to ensure that changes in the use of areas at or beyond the Site Boundary are identified and that modifications to the monitoring program are made if required by the results of this census. Adequate information gained from door-to-door or aerial surveys or through consultation with local agricultural authorities shall be used. This census satisfies the requirements of Section IV.B.3 of Appendix I to 10 CFR Part 50. Restricting the census to gardens of greater than 500 square feet provides assurance that significant exposure pathways via leafy vegetables will be identified and monitored since a garden of this size is the minimum required to produce the quantity (26 kg/year) of leafy vegetables assumed in Regulatory Guide 1.109 for consumption by a child. To determine this minimum garden size, the following assumption were used:

1) that 20% of the garden was used for growing broad leaf vegetation (i.e., similar to lettuce and cabbage), and

2) a vegetation yield of 2 kg/square meter.

OFF-SITE DOSE CALCULATION MANUAL Rev. 38 Page 26

3.0 SPECIFICATION BASES (Cont'd)

RADIOLOGICAL ENVIRONMENTAL MONITORING INTERLABORATORY COMPARISON PROGRAM BASIS The requirement for participation in an Interlaboratory Comparison Program is provided to ensure that independent checks on the precision and accuracy of the measurements of radioactive material in environmental sample matrices are performed as part of the quality assurance program for environmental monitoring in order to demonstrate that the results are reasonably valid.

Not Used Not Used Not Used Not Used OFF-SITE DOSE CALCULATION MANUAL Rev. 38 Page 27

PART II METHODOLOGIES OFF-SITE DOSE CALCULATION MANUAL Rev. 38 Page 28

SECTION 1.0 PRE-RELEASE CALCULATION PRE-RELEASE CALCULATION 1.3-1 GASEOUS RADWASTE RELEASE I. INTRODUCTION Prior to initiating a release of gaseous radwaste, it must be determined that the concentration of radionuclides to be released, and the flow rates at which they are released will not cause the dose rate limitation of 1500 mrem/yr (any organ via the inhalation pathway) to be exceeded from tritium and radioactive particulates with greater than 8 day half-lives at the Site Boundary.

II. INFORMATION REQUIRED Results of appropriate Nuclide Analysis III. CALCULATIONS Tritium, Radioactive Particulates Dose Rate (T,P) = (X/Q)PiQi mrem/yr.

where:

Pi = The dose parameter for radionuclides other than noble gases for the inhalation pathway, in mrem/yr per µCi/m3. (See Table 4.4-3).

Qi = The release rate of radionuclides, i, in gaseous effluent from individual release sources, in µCi/sec (per unit, unless otherwise specified). Qi = Effluent stream nuclide concentration x flow rate.

Flow Rates (Variable - based on setpoint needs, nominal or maximum values listed below.)

1) Auxiliary Building and Fuel Handling Area Exhaust Duct = 156,000 cfm = 7.4 x 107 cc/sec (X/Q) = 2.5 x 10-6 sec/m3. For all vent releases. The highest calculated annual average relative concentration for any area at or beyond the unrestricted area boundary.

OFF-SITE DOSE CALCULATION MANUAL Rev. 38 Page 29

SECTION 1.0 PRE-RELEASE CALCULATION (Contd)

PRE-RELEASE CALCULATION 1.3-2 LIQUID RADWASTE RELEASE I. INTRODUCTION Prior to initiating a release of liquid radwaste, it must be determined that the concentration of radionuclides to be released and the flow rates at which they will be released will not lead to a release concentration greater than the limits of 10 times the effluent concentrations specified by 10 CFR 20 at the point of discharge.

II. INFORMATION REQUIRED Results of appropriate Nuclide Analysis III. CALCULATIONS

+

Discharge Concentration = 0.1 + 63 + + + + 55 ÷ 63 55 where:

Ci = The concentration of isotope i, in the gamma spectrum excluding dissolved or entrained noble gases.

CT = Tritium Concentration from most recent analysis.

Ca = Gross alpha concentration from most recent analysis.

CS = Sr-89, 90 concentration from most recent analysis.

CNi = Ni-63 concentration from most recent analysis.

CFe = Fe-55 concentration from most recent analysis.

E = Effluent Stream Flow Rate D = Dilution Stream Raw Water Flow Rate ECL = 10 CFR 20 Appendix B, effluent concentration limit.

If Discharge Concentration is less than or equal to 1, the discharge may be initiated. If Discharge Concentration is greater than 1, then release parameters must be changed to assure that Discharge Concentration is not greater than 1. Changes include reducing tank concentration by decay or dilution, reducing the waste stream release rate, or increasing dilution water flow rate.

1.4-3 Not Used 1.4-4 Not Used 1.4-5 Not Used 1.4.7 Not Used 1.4-8 Not Used OFF-SITE DOSE CALCULATION MANUAL Rev. 38 Page 30

SECTION 2.0 RADIOACTIVE EFFLUENTS DOSE PROJECTION CALCULATIONS 2.1 Dose Projection Methodology Gaseous Radwaste Calculation Dp = 31Dc/NDQ where:

Dp = Projected Dose (monthly).

Dc = Current quarter cumulative dose, including projection for release under evaluation.

NDQ = Number of days into quarter, where the quarterly periods are:

January 1 through March 31, April 1 through June 30, July 1 through September 30, October 1 through December 2.2 Dose Projection Methodology - Liquid Radwaste I. Introduction Crystal River Unit 3 operating practices require liquid radwaste (except for Laundry and Shower Sump waste and Secondary Drain Tank waste) to be processed prior to releasing them to the environment.

As long as these practices are maintained the radwaste reduction requirements of Section 2.3 of Part I of the ODCM are met, and there is no need to project doses prior to the release of liquid radwaste.

II. Calculation Dose projection calculations will be necessary if there is a malfunction of liquid radwaste treatment system equipment and liquid radwaste must be released without prior treatment.

Dp = 31Dc/NDQ where:

Dp = Projected Dose (monthly).

Dc = Current quarter cumulative dose, including projection for release under evaluation.

NDQ = Number of days into quarter, where the quarterly periods are:

January 1 through March 31, April 1 through June 30, July 1 through September 30, October 1 through December 31.

OFF-SITE DOSE CALCULATION MANUAL Rev. 38 Page 31

SECTION 3.0 RADIOACTIVE EFFLUENTS SAMPLING SPECIFICATIONS Representative Sampling Method (Liquid)

(Evaporator Condensate Storage Tanks, Laundry & Shower Sump Tanks, Borated Water Storage Tanks, Secondary Drain Tank)

To obtain representative samples from these tanks, the contents of the tank to be sampled will be recirculated through two contained volumes and a grab sample will be collected upon completion. No additions of liquid waste will be made to the tank until completion of the release.

Representative Sampling Method (Gas Particulates and Tritium)

(Auxiliary Building & Fuel Handling Building)

Representative particulate samples will be taken on a monthly basis after RM-A2 operation is abandoned. The volume of air released will also be made on a monthly basis to determine the total release. Sample time and volumes should be established to assure the LLD limits of 1E-11 µCi/ml for principal gamma emitters, gross alpha and Sr-89/90 are met. Tritium samples should be taken semi-annually and counted to assure an LLD of 1E-5 µCi/ml is met. Note: This sample may be obtained using RM-A2 while it is still in service.

OFF-SITE DOSE CALCULATION MANUAL Rev. 38 Page 32

SECTION 4.0 RADIOACTIVE EFFLUENTS DOSE CALCULATIONAL SPECIFICATIONS OFF-SITE DOSE CALCULATION MANUAL Rev. 38 Page 33

DOSE CALCULATION 4.3-2 (TRITIUM & PARTICULATES)

The dose to an individual at or beyond the Site Boundary due to Tritium and radioactive particulates with half lives of greater than 8 days is calculated as follows:

D = 3.17 x 10-8 WRiQi mrem where:

D = The radiation dose to an individual at or beyond the Unrestricted Area BOUNDARY, in mrem.

Ri = The dose factor for each identified radionuclide, i, in m2(mrem/year) per uCi/sec or mrem/year per uCi/m3.

W = X/Q for inhalation pathway, 2.5 x 10-6 sec/m3 at the site boundary and 7.5 x 10-7 sec/m3 at the critical receptor.

W = D/Q for food and ground plane pathway, 1.9 x 10-8m-2 at the site boundary and 5.7 x 10-9 m-2 at the critical receptor.

Qi = Total µCi of isotope i released during the calendar quarter or calendar year, as appropriate 3.17 x 10-8 = The number of years in one second

Reference:

NUREG 0133, Section 5.3.1 OFF-SITE DOSE CALCULATION MANUAL Rev. 38 Page 34

DOSE CALCULATION 4.3-3 (LIQUID EFFLUENTS)

The dose or dose commitment to a Member Of The Public from radioactive materials in liquid effluents released to Unrestricted Areas is calculated as follows:

D = Ai tkCikFk i k where:

D = The cumulative dose commitment to the total body or any organ, T, from the liquid effluents for the total time period tk in mrem.

tk = The length of the kth time period over which C is averaged for all liquid releases, ik in hours.

Cik = The average concentration of radionuclide, i, in undiluted liquid effluent during time period tk from any liquid release, in µCi/ml.

Ai = The site related ingestion dose commitment factor to the total body or any organ for each identified principal gamma and beta emitter as shown in Table 4.4-17 of this manual, in mrem-ml per hour-µCi.

Fk = Waste flow rate / (Waste flow rate + Dilution flow rate)*

Dilution flow rate is the sum of available circulating water and Raw Water. Units 1 and 2 circulating water flow may be included.

References:

1) NUREG 0133, Section 4.3.

2) *Telecon/Meeting Summary with C. Willis (USNRC) dated 01/16/85 regarding Fk OFF-SITE DOSE CALCULATION MANUAL Rev. 38 Page 35

CALCULATION OF INHALATION PATHWAY DOSE FACTOR (Ri)

Ri = K ( BR ) DFAi mrem / year per uCi / m 3 where:

K = A constant unit of conversion - 106 pCi/uCi BR = The Breathing Rate of the represented age group:

1400 m3/yr - infant 3700 m3/yr - child 8000 m3/yr - teen 8000 m3/yr - adult DFAi = The maximum organ inhalation dose factor for the represented age group for the ith radionuclide, in mrem/pCi. From Reference 2 below. The dose factor for tritium in bone is not provided in Reference 2 below (i.e. no data), therefore the dose factor used for all other organs is applied. This is conservative and consistent with more current radiation protection guidance (e.g. Federal Guidance Report 11 and ICRP-56).

References:

1) NUREG-0133, Section 5.3.1.1

2) Regulatory Guide 1.109, Table E-5, and Tables E-7 through E-10 OFF-SITE DOSE CALCULATION MANUAL Rev. 38 Page 36

TABLE 4.4-2 Inhalation Dose Factors (Ri) - Infant Nuclide Bone Liver T. Body Thyroid Kidney Lung GI-LLI H-3 6.47E2 6.47E2 6.47E2 6.47E2 6.47E2 6.47E2 6.47E2 Mn-54 ND 2.53E4 4.98E3 4.98E3 4.98E3 9.95E5 7.06E3 Fe-55 1.97E4 1.17E4 3.33E3 ND ND 8.69E4 1.09E3 Fe-59 1.36E4 2.35E4 9.48E3 ND ND 1.02E6 2.48E4 Co-60 ND 8.02E3 1.18E4 ND ND 4.51E6 3.19E4 Ni-63 3.39E5 2.04E4 1.16E4 ND ND 2.09E5 2.42E3 Zn-65 1.93E4 6.26E4 3.11E4 ND 3.25E4 6.47E5 5.14E4 Sr-89 3.98E5 ND 1.14E4 ND ND 2.03E6 6.40E4 Sr-90 4.09E7 ND 2.59E6 ND ND 1.12E7 1.31E5 Ag-110m 9.98E3 7.22E3 5.00E3 ND 1.09E4 3.67E6 3.30E4 Cs-134 3.96E5 7.03E5 7.45E4 ND 1.90E5 7.97E4 1.33E3 Cs-137 5.49E5 6.12E5 4.55E4 ND 1.72E5 7.13E4 1.33E3 Ce-144 3.19E6 1.21E6 1.76E5 ND 5.38E5 9.84E6 1.48E5 OFF-SITE DOSE CALCULATION MANUAL Rev. 38 Page 37

Calculation of Ingestion Dose Factor Grass-Cow-Milk Pathway QF(Uap ) fpfs (1 - fpfs )e -ith -itf R ic [D/Q] = K' Fm (r )(DFLi )a + e i + w Y p Y s where: Unit = m2mrem/yr per µCi/sec Reference Table R.G. 1.109 K = A constant of unit conversion, 106 pCi/Ci.

QF = The cow's consumption rate, 50 kg/day (wet weight) E-3 Uap = The receptor's milk consumption rate for age (a), in liters/yr E-5 Infant & Child - 330, Teen - 400, Adult - 310 Yp = The agricultural productivity by unit area of pasture feed grass E-15 0.7 kg/m2 Ys = The agricultural productivity of unit area of stored feed E-15 2.0 kg/m2 Fm = The stable element transfer coefficients, in days/kg. E-1 r = Fraction of deposited activity retained on cow's feed grass E-15 0.2 particulates tf = Transport time from pasture to receptor, in sec. E-15 1.73x105 sec (2 days) th = Transport time from crop field to receptor, in sec. E-15 7.78x106 sec. (90 days)

( DFLi) a = The maximum organ ingestion dose factor for the ith E-11 to E-14 radionuclide for the receptor in age group (a), in mrem/pCi

-1 i = The decay constant for the ith radionuclide, in sec w = The decay constant for removal of activity on leaf and E-15 plant surfaces by weathering 5.73 x 10-7 sec-1 (corresponding to a 14 day half-life).

fp = Fraction of the year that the cow is on pasture ----

(dimensionless) = 1*.

fs = Fraction of the cow feed that is pasture grass ----

while the cow is on pasture (dimensionless) = 1*.

• Milk cattle are considered to be fed from two potential sources, pasture grass and stored feeds.

Note: The above equation does not apply to the concentration of tritium in meat. A separate equation is provided in NUREG 0133, Section 5.3.1.4 to determine Tritium value.

Reference:

The equation for Rci (D/Q) was taken from NUREG-0133, Section 5.3.1.3 OFF-SITE DOSE CALCULATION MANUAL Rev. 38 Page 38

TABLE 4.4-3 Inhalation Dose Factors (Ri) - Child Nuclide Bone Liver T. Body Thyroid Kidney Lung GI-LLI H-3 1.12E3 1.12E3 1.12E3 1.12E3 1.12E3 1.12E3 1.12E3 Cr-51 ND ND 1.54E2 8.55E1 2.43E1 1.70E4 1.08E3 Mn-54 ND 4.29E4 9.51E3 ND 1.00E4 1.58E6 2.29E4 Fe-55 4.74E4 2.52E4 7.77E3 ND ND 1.11E5 2.87E3 Fe-59 2.07E4 3.34E4 1.67E4 ND ND 1.27E6 7.07E4 Co-60 ND 1.31E4 2.26E4 ND ND 7.07E6 9.62E4 Ni-63 8.21E5 4.63E4 2.80E4 ND ND 2.75E5 6.33E3 Zn-65 4.26E4 1.13E5 7.03E4 ND 7.14E4 9.95E5 1.63E4 Sr-89 5.99E5 ND 1.72E4 ND ND 2.16E6 1.67E5 Sr-90 1.01E8 ND 6.44E6 ND ND 1.48E7 3.43E5 Ag-110m 1.69E4 1.14E4 9.14E3 ND 2.12E4 5.48E6 1.00E5 Cs-134 6.51E5 1.01E6 2.25E5 ND 3.30E5 1.21E5 3.85E3 Cs-137 9.07E5 8.25E5 1.28E5 ND 2.82E5 1.04E5 3.62E3 Ce-144 6.77E6 2.12E6 3.61E5 ND 1.17E6 1.20E7 3.89E5 OFF-SITE DOSE CALCULATION MANUAL Rev. 38 Page 39

TABLE 4.4-6 c

Ingestion Dose Factors ( R i )

Grass-Cow-Milk Pathway (Infant)

Nuclide Bone Liver T. Body Thyroid Kidney Lung GI-LLI H-3 2.38E3 2.38E3 2.38E3 2.38E3 2.38E3 2.38E3 2.38E3 Mn-54 ND 3.89E7 8.83E6 ND 8.63E6 ND 1.43E7 Fe-55 1.35E8 8.72E7 2.33E7 ND ND 4.26E7 1.11E7 Fe-59 2.26E8 3.94E8 1.55E8 ND ND 1.17E8 1.88E8 Co-60 ND 8.81E7 2.08E8 ND ND ND 2.10E8 Ni-63 3.49E10 2.16E9 1.21E9 ND ND ND 1.07E8 Zn-65 5.55E9 1.90E10 8.78E9 ND 9.24E9 ND 1.61E10 Sr-89 1.26E10 ND 3.61E8 ND ND ND 2.59E8 Sr-90 1.22E11 ND 3.10E10 ND ND ND 1.52E9 Ag-110m 3.86E8 2.82E8 1.87E8 ND 4.03E8 ND 1.46E10 Cs-134 3.65E10 6.80E10 6.87E9 ND 1.75E10 7.18E9 1.85E8 Cs-137 5.15E10 6.02E10 4.27E9 ND 1.62E10 6.55E9 1.88E8 Ce-144 2.33E6 9.52E5 1.30E5 ND 3.85E5 ND 1.33E8 OFF-SITE DOSE CALCULATION MANUAL Rev. 38 Page 40

Calculation of Dose Factors G

in the Ground Plane Pathway ( R i [D/Q])

(

i [ D / Q] = K K ( SF)( DFGi ) 1- e RG

- it

) / i where: units = m2 mrem/yr per uCi/sec Reference Table, R.G. 1.109 K = A constant unit of conversion, 106 pCi/µCi.

K = A constant unit of conversion, 8760 hr/yr SF = The shielding factor, 0.7(dimensionless) E-15 i = The decay constant for the ith radionuclide, sec-1 t = The exposure period, 4.73 x 108 sec (15 years)

DFGi = The ground plane dose conversion factor for the E-6 ith radionuclide (mrem/hr per pCi/m2)

G

Reference:

The equation deriving R i [D/Q] was taken from NUREG 0133, Section 5.3.1.2.

OFF-SITE DOSE CALCULATION MANUAL Rev. 38 Page 41

Table 4.4-16 G

Dose Factors Ground Plane Pathway ( R i [D/Q])

Nuclide T. Body Skin Mn-54 1.39E9 1.63E9 Fe-55 0 0 Fe-59 2.73E8 3.21E8 Co-60 2.15E10 2.53E10 Ni-63 0 0 Zn-65 7.47E8 8.57E8 Sr-89 2.17E4 2.52E4 Ag-110m 3.44E9 4.02E9 Cs-134 6.85E9 8.00E9 Cs-137 1.03E10 1.20E10 Ce-144 6.95E7 8.05E7 Units are m2mrem/yr per µCi/sec OFF-SITE DOSE CALCULATION MANUAL Rev. 38 Page 42

CALCULATION OF LIQUID EFFLUENT ADULT INGESTION DOSE FACTORS Ai = 1.14E5 ( 21BFi + 5BIi) DFi Ai = Composite dose parameter for the total body or critical organ of an adult for nuclide i, for all appropriate pathways, mrem/hr per µCi/ml 1.14E5 = units conversion factor, 106pci/µci x 103 ml/kg ÷ 8760 hr/yr BFi = Bioaccumulation factor for nuclide i, in fish, pCi/kg per pCi/L, from Table A-1 of Regulatory Guide 1.109 (Rev. 1) or Table A-8 of Regulatory Guide 1.109 (original draft).

BIi = Bioaccumulation factor for nuclide i, in invertebrates, pCi/kg per pCi/L, from Table A-1 of Regulatory Guide 1.109 (Rev. 1) or Table A-8 of Regulatory Guide 1.109 (original draft).

DFi = Dose conversion factor for nuclide i, for adults in pre-selected organ , in mrem/pCi, from Table E-11 or Regulatory Guide 1.109 (Rev. 1) or Table A-3 of Regulatory Guide 1.109 (original draft).

Reference:

The equation for Saltwater sites from NUREG 0133, Section 4.3.1, where Uw/Dw = 0 since no drinking water pathway exists.

OFF-SITE DOSE CALCULATION MANUAL Rev. 38 Page 43

Table 4.4-17 Liquid Effluent - Adult Ingestion Dose Factors (Ai)

Nuclide Bone Liver T. Body Thyroid Kidney Lung GI-LLI H-3 2.82E-1 2.82E-1 2.82E-1 2.82E-1 2.82E-1 2.82E-1 2.82E-1 Mn-54 ND 7.06E3 1.35E3 ND 2.10E3 ND 2.16E4 Fe-55 5.11E4 3.53E4 8.23E3 ND ND 1.97E4 2.03E4 Fe-59 8.06E4 1.90E5 7.27E4 ND ND 5.30E4 6.32E5 Co-60 ND 1.73E3 3.82E3 ND ND ND 3.25E4 Ni-63 4.96E4 3.44E3 1.67E3 ND ND ND 7.18E2 Zn-65 1.61E5 5.13E5 2.32E5 ND 3.43E5 ND 3.23E5 Sr-89 4.99E3 ND 1.43E2 ND ND ND 8.00E2 Sr-90 1.23E5 ND 3.01E4 ND ND ND 3.55E3 Ag-110m 1.57E3 1.45E3 1.33E1 ND 2.85E3 ND 5.91E5 Sb-124 2.77E2 5.23E0 1.09E2 6.71E-1 ND 2.15E2 7.83E3 Sb-125 1.77E2 1.97E0 4.20E1 1.80E-1 ND 1.36E2 1.94E3 Cs-134 6.84E3 1.63E4 1.33E4 ND 5.27E3 1.75E3 2.85E2 Cs-137 8.78E3 1.20E4 7.85E3 ND 4.07E3 1.35E3 2.32E2 Ce-144 1.79E2 7.47E1 9.59E0 ND 4.43E1 ND 6.04E4 OFF-SITE DOSE CALCULATION MANUAL Rev. 38 Page 44

4.6 SOURCE TERM Gaseous Releases CR-3 has been permanently shut down since September, 2009 and will remain permanently defueled. All fuel is now on the ISFSI pad. As such, there is no significant releasable noble gas or iodine source term remaining at CR3. With the spent fuel pool drained, there will also be a significantly decreased tritium source term in gaseous releases. The particulate component of any gaseous release should also be minimal. However, because of planned changes in ventilation, particulates, along with tritium, will continue to be sampled and assessed for dose.

Total Body Dose Rate:

For a single nuclide:

Dose rate = (X/Q)(K)(Q)

Q is in µCi/sec and is the product of the vent concentration (VC) and vent flow rate (VFR), therefore:

Dose rate = (X/Q)(K)(VC)(VFR)

Substituting values:

(2.5E-6 s/m3)(1.61E1 mrem-m3/µCi-y)(6E-3 µCi/cc)(7.4E7 cc/s)] = 17.9 mrem/y Scaling this result to 250 mrem/y:

(250 mrem/y) ÷ (17.9 mrem/y) x (6E-3 µCi/cc) = 0.084 µCi/cc (250 mrem/y) ÷ (17.9 mrem/y)/(1 fuel assembly) = 14 fuel assemblies Skin Dose Rate:

For a single nuclide:

Dose rate = X/Q(L + 1.1M)Q = X/Q(L + 1.1M)(VC)(VFR)

Substituting values:

(2.5E-6 s/m3)(1.34E3 mrem-m3/µCi-y + 1.1 x 1.72E1 mrad-m3/µCi-y )(6E-3 µCi/cc)(7.4E7 cc/s)] = 1.50E3 mrem/y OFF-SITE DOSE CALCULATION MANUAL Rev. 38 Page 45

4.6 SOURCE TERM (Cont'd)

Liquids This section will characterize normal releases and put this characterization in context by deriving a waste concentration or discharge concentration required to challenge the limits of the ODCM - which is 1.5 mrem per quarter total body.

All releases are made in a batch mode, meaning that waste tanks are first filled, isolated, sampled, and then release permits are prepared specifying the conditions under which the release is to be made.

The highest activity releases are made from WDT-10A and WDT-10B, otherwise known as the evaporator condensate storage tanks (ECSTs).

A review of effluent data from 2005 through 2013 (ref. Radioactive Effluent Release Reports) indicates that the highest annual quantity of activity (excluding noble gases, H-3, and alpha emitters) was 0.1 Ci released in 2005.

For 2011 through 2013 the isotopic (beta and gamma) make-up of liquid effluents was comprised of H-3, Mn-54, Fe-55, Co-58, Co-60, Ni-63, Ag-110m, Sb-125, Cs-134, and Cs-137. Some of these isotopes are expected to be less prominent over the near term due to decay and the removal of their source (high activity spent resins) from the plant and tritium is trending down at a significant rate due to evaporation. Of these, cesium has the most limiting effluent concentration limits (ECL) (ref 10 CFR 20) which are 9E-7 uCi/ml for Cs-134 and 1E-6 uCi/ml for Cs-137.

A typical release from WDT-10A or WDT-10B has a gamma emitter concentration < 1E-6 µCi/ml (ref. 2014 release data from SP-736L), which is < 3E-5 Ci total for the tank.

The gamma activity of spent fuel pool water (ref. 2014 NuclearIQ - Chemistry Data Management System) with the spent fuel demineralizer in service is < 4E-5 µCi/ml with most of this being Cs-137.

Total Cs-137 in the spent fuel pool water is approximately 0.05 Curies.

Calculations:

The ODCM limit for total body dose due to liquid releases is 1.5 mrem per quarter.

Assumptions:

Release Volume: 10,000 gallons (maximum for 1 ECST)

Release Rate: 100 gpm (maximum release rate from ECST)

Release Duration: 100 minutes (1.667 hours0.00772 days <br />0.185 hours <br />0.0011 weeks <br />2.537935e-4 months <br />)

Dilution Flow Rate: 60,000 gpm (Approx. equal to 1 CW pump at units 1 & 2)

Isotopic mix: Cs-137 (due to its high dose factor and dominance in the spent fuel pool water and spent fuel)

OFF-SITE DOSE CALCULATION MANUAL Rev. 38 Page 46

4.6 SOURCE TERM (Cont'd)

Liquids (Cont'd)

Using ODCM dose calculation 4.4-3 and solving for the concentration required to reach 1.5 mrem total body:

Concentration in uCi/ml = Dose limit / [Dose Factor)(Duration)(Dilution)]

0.07 uCi/ml = 1.5 mrem /(7850 mrem/hr per uCi/ml)(1.667 hours0.00772 days <br />0.185 hours <br />0.0011 weeks <br />2.537935e-4 months <br />)(100 gpm/60000 gpm)

(0.07 uCi/ml represents a total tank activity of 2.6 Curies of Cs-137)

Note: Current Cs-137 concentrations in spent fuel pool are on the order of 5E-6 uCi/ml or less.

Another option for approximating the dose due to liquid releases is to evaluate the concentration of the individual radionuclides against their ECLs keeping in mind that ECLs are established at a level which would give a 50 mrem TEDE for 1 year of exposure (ref. 10 CFR 20, Appendix B).

ODCM pre-release calculation 1.3-2 solves for discharge concentration (DC) to assure that the release parameters are established to prevent exceeding an instantaneous dose rate of 500 mrem/year.

A release with a discharge concentration of 1 and a duration of 1 year would yield a dose of 500 mrem TEDE.

From this, it is possible to derive the discharge concentration which equates to 1.5 mrem TEDE provided an assumption is made about the release duration. In this case the release duration is assumed to be 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

Discharge Concentration = (1.5 mrem /500 mrem) x (8760 hours0.101 days <br />2.433 hours <br />0.0145 weeks <br />0.00333 months <br /> per year/1 hour) = 26.3 This means that a release with a calculated discharge concentration of 26.3 uCi/ml and which lasts for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> would impart a dose of 1.5 mrem TEDE.

Note: This method does not account for additional dilution in the discharge canal.

OFF-SITE DOSE CALCULATION MANUAL Rev. 38 Page 47

SECTION 5.0 ENVIRONMENTAL MONITORING OFF-SITE DOSE CALCULATION MANUAL Rev. 38 Page 48

Table 5.1-1 Environmental Radiological Monitoring Stations Locations DIRECTION APPROX. DISTANCE STATION LOCATION FROM PLANT FROM PLANT (mi)

C04 State Park Old Dam on River near road intersection ENE 10.6 C07 Crystal River Public Water Plant ESE 7.4 C09 Fort Island Gulf Beach S 3.2 C10 Indian Waters Public Water Supply ESE 6.0 C13 Mouth of Intake Canal WSW 3.4 C14H Head of Discharge Canal N 0.1 C14M Midpoint of Discharge Canal W 1.2 C14G Discharge Area W 2.0 C18 Yankeetown City Well N 5.3 C19 NW Corner State Roads 488 & 495 ENE 9.6 C29 Discharge Area W 2.0 C30 Intake Area WSW 3.4 Near E. Site Boundary & well pump -

C40 E 3.6 station CR-South #5 C41 Onsite abandoned meteorological tower SW 0.4 C46 North Pump Station N 0.4 C47 Office of Radiation Control, Orlando ESE 78 C48A1 Near C46 North Pump Station N 0.4 C48B1 Onsite NNE of CR 4 & 5 NNE 0.9 NOTE: Distances are approximate. More than one type of sample media,(e.g. air and water) are obtained at some stations. For multi-media stations there may be minor difference in distance for each type of sample.

1 If vegetation is not available, then select another suitable nearby location. The ENE sector is also an acceptable sector based on D/Q values.

OFF-SITE DOSE CALCULATION MANUAL Rev. 38 Page 49

Table 5.1-2 Environmental Radiological Monitoring Stations Locations-Groundwater Monitoring Wells DIRECTION APPROX. DISTANCE STATION LOCATION FROM PLANT FROM PLANT (mi)

• CR3-1S CR-3 Site Perimeter, Just Outside of Protected Area Fence ENE 0.2

• CR3-1D CR-3 Site Perimeter, Just Outside of Protected Area Fence ENE 0.2 CR3-2 CR-3 Site Perimeter, Just Outside of Protected Area Fence E 0.1

• CR3-3S CR-3 Site Perimeter, Just Outside of Protected Area Fence ESE 0.1

• CR3-3D CR-3 Site Perimeter, Just Outside of Protected Area Fence ESE 0.1 CR3-4 CR-3 Site Perimeter, Just Outside of Protected Area Fence SSE 0.086 CR3-5 CR-3 Site Perimeter, Just Outside of Protected Area Fence SSW 0.051 CR3-6S CR-3 Site Perimeter, Just Outside of Protected Area Fence W 0.038 CR3-6D CR-3 Site Perimeter, Just Outside of Protected Area Fence W 0.038 CR3-7 CR-3 Site Perimeter, Just Outside of Protected Area Fence WNW 0.060 CR3-8 CR-3 Site Perimeter, Just Outside of Protected Area Fence WNW 0.073 CR3-9 CR-3 Site Perimeter, Just Outside of Protected Area Fence NW 0.1 CR3-10 CR-3 Site Perimeter, Just Outside of Protected Area Fence NNE 0.1 The above listed wells have been included in the REMP as a result of information provided in the groundwater flow study completed January 22, 2007 by EnHydro, LLC.

• These wells added to REMP as a result of recommendations made from groundwater flow study completed 11/01/2012 by Gaydos Hydro Services, LLC.

OFF-SITE DOSE CALCULATION MANUAL Rev. 38 Page 50

TABLE 5.1-3 RING TLDs (INNER RING)

LOCATION DIRECTION APPROX. DISTANCE (Mi.)

C27 W 0.4 C60 N 0.9 C61 NNE 0.9 C62 NE 1.2 C63 ENE 0.9 C64 E 0.8 C65 ESE 0.3 C66 SE 0.4 C67 SSE 0.3 C68 S 0.3 C69 SSW 0.3 C41 SW 0.4 C70 WSW 0.7 C71 WNW 0.6 C72 NW 0.3 C73 NNW 0.7 OFF-SITE DOSE CALCULATION MANUAL Rev. 38 Page 51

TABLE 5.1-4 RING TLDs (5 MILE RING)

LOCATION DIRECTION APPROX. DISTANCE (Mi.)

C18 N 5.3 C03 NNE 4.9 C04 NE 6.0 C74 ENE 5.1 C75 E 4.0 C76 ESE 5.6 C08 SE 5.7 C77 SSE 3.4 C09 S 3.2 C14G W 2.0 C01 NW 4.8 C79 NNW 5.0 OFF-SITE DOSE CALCULATION MANUAL Rev. 38 Page 52

FIGURE 5.1 Environmental Monitoring Sample Station Locations OFF-SITE DOSE CALCULATION MANUAL Rev. 38 Page 53

FIGURE 5.2 Environmental Monitoring TLD Locations OFF-SITE DOSE CALCULATION MANUAL Rev. 38 Page 54

FIGURE 5.3 Environmental Monitoring TLD Locations (5 mile)

OFF-SITE DOSE CALCULATION MANUAL Rev. 38 Page 55

FIGURE 5.4 CR3 Groundwater Monitoring Well Locations Deep Wells Are Also Installed at #s 1, 3, 6 10 9

1 8

7 2

6 5 3 4

OFF-SITE DOSE CALCULATION MANUAL Rev. 38 Page 56

SECTION 6.0 ADMINISTRATIVE CONTROLS Not Used OFF-SITE DOSE CALCULATION MANUAL Rev. 38 Page 57

Summary of Changes DRRs 2036411, 2167278, 2205109 and 2205242 Page(s) CHANGE Introduction Removed sentence referring to methodology for calculating effluent monitor setpoints.

1-2 1) Removed definitions for: Channel calibration, Channel check, Channel function test, Mode, Ventilation Exhaust Treatment System, Purge - Purging, Operable -

Operability and Source Checks. All of these pertain to Effluent monitors which will no longer be required per procedure CP-500, as supported by CR3 calculation N18-0004 and associated 50.59 evaluation. 2) Removed reference to Improved Tech Specs in the Note box. 3) Removed reference to evaporators in the Liquid radwaste Treatment System definition. 4) Removed reference to Alarm Trip Setpoints in the ODCM definition. 5) Renumbered remaining steps.

2 Definition for Unplanned Release updated to remove reference to the plant ventilation system and to indicate the reduced risk due to plant shutdown.

Removed definitions for Ventilation treatment System and Purge. Re-numbered steps.

3 Removed Sections 2.1, 2.2 and 2.4. These pertain to liquid and gaseous effluent rad monitoring and gaseous treatment. Gas and liquid monitors, and gas treatment, will no longer be required due to the decreased CR3 source term as documented in calculation N18-0004 and the associated 50.59 evaluation for CP-500.

5 Table 2-5: 1) Removed reference to dissolved and entrained gas as CR3 no longer has measurable quantities of radioactive gas. 2) Added Ni-63 to the quarterly composite sample. 3) Removed reference to Continuous Releases as CR3 does not make continuous liquid releases.

7 Table 2-5 Notation: Added guidance in notation e. Removed short lived radionuclides Mo-99 and Ce-141 from notation f.

8 Step 2.6.1: Added clarification that a dose calculation only has to be performed if a release has been made.

9 1) Removed old Step 2.7.a pertaining to noble gas dose rate and renumbered old Step 2.7.b as 2.7.a. Due to radioactive decay and relocation of spent fuel to dry storage, CR3 no longer has measurable dose rates from noble gasses. 2)

Removed old step 2.7.1 pertaining to noble gas and renumbered old step 2.7.2 as 2.7.1. 3) Deleted information in Step 2.8 pertaining to noble gas dose as CR3 no longer has a noble gas dose source term.

10 Table 2-6: Removed reference to continuous noble gas monitor. It is not needed as per CP-500, as supported by Calculation N18-0004 and the associated 50.59 evaluation. Added superscript (d) to Aux Building Exhaust in column 1.

12 Table 2-6 Notation: Notation (d): Removed note for which pertained to a continuous air sampler and replaced with note to determine flow rate. Notation (f)

Removed short lived radionuclides Mo-99 and Ce-141 from step f. Removed row for continuous noble gas monitoring.

OFF-SITE DOSE CALCULATION MANUAL Rev. 38 Page 58

Page(s) CHANGE 13 Added step 2.9.2 14 Step 2.10.a: Removed reference to old Specifications 2.6.a, 2.6.b, 2.8.a, and 2.8.b.

Step 2.10.1: Removed reference to old Specifications 2.6.1 and 2.8.1. Added sentence to account for direct radiation shine and the 25 mrem annual limit of 40 CFR 190 to address Nuclear Oversight NCR 2199373 from audit 2018-CR3-SEC/RP-01.

16 Table 2.7: Removed five mile direct radiation location C78. This TLD location at the end of the north bank of the intake canal road will no longer be readily accessible in the future. ANI and State of Florida DEP agreed with C78 TLD removal.

22 Step 2.14 - Contents of old Special Reports section removed based on results of Calculation N18-0004 and the associated CP-500 50.59 evaluation that CR3 is not likely to exceed any NRC regulatory effluent release limit.

23 Old sections 3.1, 3.2 and 3.4 deleted. These pertained to liquid and effluent monitoring and the ventilation treatment system. As per Calculation N18-0004 and the associated CP-500 50.59 evaluation, these are no longer needed.

24 Old section 3.8 pertaining to Gaseous Effluents Dose Noble Gases Basis removed because there is no longer any noble gas dose at CR3.

ODCM Part II Changes Below 28 Old Section 1.0: Steps pertaining to Radioactive Effluent Monitor Setpoints Specifications was removed as per guidance in Calculation N18-0004 and the associated CP-500 50.59 evaluation. Only steps 1.3-1 and 1.3-2 were kept.

29 Section 1.3.-1: Calculations related to noble gas gamma and beta emissions were deleted due to no longer having a noble gas source term.

30 Section 1.3-2: Removed noble gas source term from the equation and notes and added Ni-63 to account for current CR3 source term. Removed sections 1.4-3, 1.4-4, 1.4-5, 1.4-7 and 1.4-8. These pertained to effluent monitor setpoint calculations.

31 Part II, Section 2.0: Removed old Table II and waste/dose reduction content associated with the liquid and ventilation treatment systems. As per Calculation N18-0004 and the associated CP-500 50.59 evaluation, there is no significant source term to require treatment. Also, changed title of Section 2.0 to indicate dose projection calculations instead of dose reduction specifications.

32 Part II, Section 3.0: Re-wrote sampling guidance to account for RM-A2 removal and reduced tritium source term after the spent fuel pool is drained. Old Table III removed.

OFF-SITE DOSE CALCULATION MANUAL Rev. 38 Page 59

Page(s) CHANGE 33 Part II, Section 4.0: Removed cumulative dose calculation specifications in old sections 4.1-1 to 4.3-1 as they were redundant to Part I or no longer applicable due to lack of a noble gas source term per the guidance in calculation N18-0004 and the associated CP-500 50.59. Old Table IV removed.

37 Table 4.4.2: Removed short lived radionuclides which are no longer present at CR3 in measurable quantities.

39 Table 4.4.3: Removed short lived radionuclides which are no longer present at CR3 in measurable quantities.

40 Old Tables 4.4.4 and 4.4.5: Removed Inhalation Dose factor tables for Teen and Adult as Infant dose factors are used for dose calculations and child dose factors are used for dose rate calculations. Teen and Adult are not used.

Table 4.4.6: Removed short lived radionuclides which are no longer present at CR3 is measurable quantities.

42 Old Tables 4.4.7 to 4.4.15: Removed Ingestion Dose Factor tables for all age groups and pathways except for infant and grass-cow-milk pathway. Dose to the other age groups and via other pathways is no assessed as infant grass cow milk is considered limiting.

Table 4.4-16: Removed short lived radionuclides which are no longer present at CR3 is measurable quantities.

44 Table 4.4-17: Removed short lived radionuclides which are no longer present at CR3 is measurable quantities.

45 Rewrote Section 4.6 to account for decreased gaseous release dose pathway and removed discussion on iodine and Kr-85 source term in a gaseous release.

Removed old note before skin dose rate section that pertained to monitor setpoints. Removed old section pertaining to iodines under the skin dose rate calculation.

46 Removed reference to Permanently Defueled Emergency Plan and used a value of 1.5 mrem in the calculation instead of the old emergency plan Alert limit of 10 mrem. The new IOEP does not have an Alert dose threshold limit.

49 Table 5.1-1: Updated distance for Stations C13 and C14G. These were moved slightly closer to CR3 to make them more accessible.

52 Table 5.1-4: Removed TLD location C78. This TLD location at the end of the north bank of the intake canal road will no longer be readily accessible in the future.

57 Section 6.0: Administrative Controls - Removed this section as administrative controls are already incorporated in plant procedures. Removing this section also addresses Nuclear Oversight IMPR recommendation 2199386 from audit 2018-CR3-SEC/RP-01 about releases to Settling Ponds.

OFF-SITE DOSE CALCULATION MANUAL Rev. 38 Page 60