Text

Annual Radioactive Effluent Release Report for January 1, 2013 Through December 31, 2013
	ML14113A447
Person / Time
Site:	Summer
Issue date:	04/17/2014
From:	Lippard G; South Carolina Electric & Gas Co
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	Download: ML14113A447 (203)
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George A. Lippard GeneralManager,Nuclear Plant Operations 803.345.4810 A SCANA COMPANY April 17, 2014 Document Control Desk U. S. Nuclear Regulatory Commission Washington, DC 20555

Dear Sir / Madam:

Subject:

VIRGIL C. SUMMER NUCLEAR STATION UNIT 1 DOCKET NO. 50-395 OPERATING LICENSE NO. NPF-12 ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT Enclosed is the South Carolina Electric & Gas Company Annual Radioactive Effluent Release Report as required by 1OCFR50.36a, Section 6.9.1.8 of the Virgil C. Summer Nuclear Station Technical Specifications and Section 1.6.2 of the Offsite Dose Calculation Manual (ODCM). Also enclosed for your review is a copy of Revision 29 of the ODCM. This submittal covers the period of January 1 through December 31, 2013.

Should there be any questions, please contact Ms. Susan B. Reese at (803) 345-4591.

Very truly yours, George A. Lippard SBR/GAL/bd

Enclosures:

I. Annual Radioactive Effluent Release Report II. Offsite Dose Calculation Manual c: w/o enclosure unless noted K. B. Marsh NRC Resident Insp~etor S. A. Byrne M. Coleman J. B. Archie T. D. Riley N. S. Carns NSRC J. H. Hamilton Susan Jenkins (SCDHEC)

J. W. Williams C. D. Stewart (ANI), (w/enclosures)

W. M. Cherry RTS (LTD 31,3)

V. M. McCree File (818.02-1'., RR 8350)

S. A. Williams PRSF (RC-14-01066), (w/enclosures)

K. M. Sutton nII-)

Of Virgil C.Summer Station - Post Office Box 88 - Jenkinsville, SC .29065

F (803) 941-9776

VIRGIL C. SUMMER JENKINSVILLE, SOUTH CAROLINA ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT VIRGIL C. SUMMER NUCLEAR STATION FOR THE OPERATING PERIOD JANUARY 1, 2013 - DECEMBER 31, 2013 APRIL 2014 A SCANA COMPANY Prepared by:

Michelle Westbury, Specialist Health Physics R wed by: Approved by:

Timothy D. Riley, Supervisor Moses Coleman, Manager Health Physics Health Physics and Safety Services

ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT JANUARY - DECEMBER, 2013 VIRGIL C. SUMMER NUCLEAR STATION SOUTH CAROLINA ELECTRIC & GAS Introduction This report is being submitted as a summary of quantities of radioactive liquid and gaseous effluents and solid waste released from the Virgil C. Summer Nuclear Station. This report is submitted in fulfillment of the requirements in the V. C.

Summer Nuclear Station Operating License Technical Specifications Section 6.9.1.8, Offsite Dose Calculation Manual (ODCM) Sectionl.6.2 and 10CFR50.36(a) and follows the reporting details specified in USNRC Regulatory guide 1.21(1).

Summary information of radioactive gaseous and liquid effluents is presented along with a summary of radioactive waste disposal as well as an evaluation of the radiological impact on man due to operation of the Virgil C. Summer Nuclear Station. Supplemental information including release limits also required by USNRC Regulatory Guide 1.21 is provided as Appendix A.

During the reporting period there were three ODCM reportable incidents. There was one change made to the Offsite Dose Calculation Manual (ODCM) during the 12-month period. Details are presented below.

A. Supplemental Information Regulatory limits for doses, dose rate and effluent concentration limits presented in Supplemental Information are from the Virgil C. Summer Nuclear Station ODCM and 40 CFR 190. Average energy (E-bar) is not applicable to the method for determining release rate limits for fission and activation gaseous effluents; therefore, it has been omitted. A compilation of required supplemental information is provided in Appendix A.

B. Gaseous Effluents Gaseous effluents released from ground level are summarized in Tables 1 and 2. An elevated release pathway does not exist at Virgil C. Summer Nuclear Station. Cumulative doses are discussed in Section E.

The errors for gaseous effluent totals are given as the square root of the sum of squares of counting errors and flow or volume measurement errors. A systematic error estimate of 15% has been 1

ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT JANUARY - DECEMBER, 2013 VIRGIL C. SUMMER NUCLEAR STATION SOUTH CAROLINA ELECTRIC & GAS included in the calculation of total error.

C. Liquid Effluents Liquid effluents are summarized in Tables 3 and 4. Estimated total errors are expressed as in Section B above.

D. Solid Waste Shipments Solid waste shipments are summarized in Table 5. Curie content of radioactive waste packages is determined by dose rates and/or gamma spectroscopy analysis of samples. The total error for each type of curie content determination is conservatively estimated to be the sum of a 15% systematic error and a 20% photon response error for the detector used.

E. Radiological Impact on Man Dose to the maximum exposed individual in the unrestricted area was calculated using measured plant gaseous effluents and meteorological data in accordance with the Offsite Dose Calculation Manual. The source term involved 0.4 days of Waste Gas Decay Tank (WGDT) releases, 2.4 days of 6-inch Reactor Building purge releases, 6.3 days of 36-inch Reactor Building purge releases and a continuous 12-month Main Plant vent release. Doses are summarized in Table 6.

The total activities released are presented in Tables 1 and 2. The highest quarterly air doses at the station boundary resulting from the release of noble gases were 9.80E-04 mrad for gamma during the second quarter and 3.70E-04 mrad for beta also during the second quarter. The maximum quarterly organ dose attributed to the releases, excluding Carbon-14, was 1.68E-04 mrem. Cumulative annual dose was 1.07E-03 mrad, 4.67E-04 mrad and 1.69E-04 mrem for gamma, beta, and organ dose, respectively. Discussion of the impact of Carbon-14 is included in Section K.

Measured plant liquid effluent data was used to calculate estimates of doses to individuals in accordance with the Offsite Dose Calculation Manual. The source term consisted of the isotopic contents of 144 Waste Monitor Tank batch releases, 4 Condensate Backwash Receiver Tank batch releases, 1 NaOH Sump batch release, 17.6 days of Steam Generator Blowdown release and a continuous 2

ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT JANUARY - DECEMBER, 2013 VIRGIL C. SUMMER NUCLEAR STATION SOUTH CAROLINA ELECTRIC & GAS Turbine Building Sump release.

Doses are summarized in Table 6 and total radioactivity released is described in Tables 3 and 4. The highest quarterly total body dose to the maximum exposed individual resulting from the release of radioactive liquid was 1.78E-03 mrem during the first quarter. The highest organ dose was 1.94E-03 mrem to the gastrointestinal-lower large intestine (GI-LLI) also for the first quarter. Cumulative annual doses for the hypothetical maximum exposed individual were 4.01 E-03 mrem for the total body and 4.21 E-3 mrem for the GI-LLI, the maximum annual organ. The GI-LLI was the maximum exposed organ from the first through third quarters and the liver was the maximum exposed organ for the fourth quarter.

Dose rates and concentrations were below station limits as specified in Supplemental Information,Section II A, B, and C during all the effluent releases.

Radiation exposure to members of the public within the site boundary was assessed through calculation of gamma and beta air dose at 0.25 miles of the gaseous effluent release point and direct measurement of exposure using thermoluminescent dosimeters.

Onsite air dose for this reporting period was 3.49E-03 mrad gamma and 1.52E-03 mrad beta, well below levels that can be distinguished above background. Quarterly thermoluminescent dosimetry data from four onsite monitoring locations within 0.2 miles of the Reactor Building and eight locations at the site boundary were analyzed and compared with respective pre-operational background and previous year history. Results showed that the 2013 quarterly dose rates did not differ significantly from the pre-operational or 2012 dose rates. It was concluded that doses to members of the public inside the site boundary were indistinguishable from normal background dose.

Radiation doses from radioactive effluents to workers at the Fairfield Hydro Station for this reporting period were calculated to be 5.06E-05 and 2.21 E-05 mrad for gamma and beta, respectively.

Radiation doses from radioactive effluents to workers at the New Nuclear Site for this reporting period were calculated to be 4.36E-04 and 1.90E-04 mrad for gamma and beta, respectively.

3

ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT JANUARY - DECEMBER, 2013 VIRGIL C. SUMMER NUCLEAR STATION SOUTH CAROLINA ELECTRIC & GAS Radiation doses from nearby uranium fuel cycle sources were not assessed. The ODCM, Sections 1.3.1 and B/1.3 establish a five (5) mile limit beyond which doses from nearby plants are insignificant. There are no uranium fuel cycle plants within a five (5) mile radius of Virgil C. Summer Nuclear Station.

F. Abnormal Releases None G. Meteorology The meteorological data for 2013 was collected and analyzed.

An annual meteorological summary report providing joint frequency distributions of wind direction and speed by atmospheric stability class is maintained in plant records.

The wind direction and wind speed data used were acquired from the 10-meter level of the primary monitoring tower.

Stability was determined by the primary differential temperature (61 to 10 meter).

The combined annual data recovery for wind direction, wind speed and stability was 99.2%. Primary variable recovery rates were as follows: wind direction (10 m) - 99.2%, wind speed (10 m) - 99.2%,

and differential temperature (61 - 10 m) - 99.2%.

H. Offsite Dose Calculation Manual The Virgil C. Summer Nuclear Station ODCM was revised on August 27, 2013. See Enclosure A for a complete copy.

Revision 29 incorporates an alternate setpoint methodology for use with batch releases in which tritium activity is present with very low gamma activity. The alternate setpoint methodology is consistent with the traditional methodology with the exception that the methodology sets a fixed value for tritium and assumes that during the release tritium concentrations cannot increase above that level.

In essence, a fraction of the dilution volume is allocated for dilution of 4

ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT JANUARY - DECEMBER, 2013 VIRGIL C. SUMMER NUCLEAR STATION SOUTH CAROLINA ELECTRIC & GAS tritium with the remaining available for all other radionuclides. The use of a conservative assumption for tritium concentration based on known or historic conditions along with actual waste monitor tank sampling prior to release ensures that instantaneous release rate limits for tritium will not be exceeded. For all other radionuclides, a general safety factor of 0.5 is included for conservatism consistent with the traditional methodology in calculation of the ECL fraction along with an assumption that only 90% of the assumed dilution flow is available.

Other changes to the ODCM included removal of controls for oil incineration which is no longer performed onsite and a change to the Lower Limit of Detection (LLD) formula consistent with NUREG/CR-4007(2) for use with low background conditions. Additionally, one ground water site was removed from the environmental monitoring program after analysis of ground water flow conditions and locations of other well locations.

Offsite Dose Calculation Manual Reportable Incidents The Turbine Building Sump Effluent Line Radiation Monitor RM-L8 required by ODCM Control 1.1.1.1 was out of service from January 13, 2013 at 7:00 until February 28, 2013 at 16:35 due to frequent low flow alarms. Compensatory actions were taken during this time period in accordance with the requirements of the ODCM. The alarms were caused by the sample pump clogging. Initial troubleshooting of the monitor revealed debris inside the flow tube. The pump and monitor were backflushed several times. Large chunks of paint and debris were seen floating in the sump and were eventually filtered out through a screening process. The flushing process returned proper flow indication to the system. A preventative maintenance action has been implemented to periodically clean the sample pump.

The Condensate Backwash Effluent Radiation Monitor RM-L1 1 require by ODCM Control 1.1.1.1 was taken out of service from November 16, 2013 at 2100 until December 27, 2013 at 1415 after observing detector background being less than half the normal indication. Initial testing indicated failure of the preamp 5

ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT JANUARY- DECEMBER, 2013 VIRGIL C. SUMMER NUCLEAR STATION SOUTH CAROLINA ELECTRIC & GAS with additional testing indicating degraded power supply performance. Both replacement of the power supply and preamp was required along with a slight modification to the system to integrate a new preamp. Although releases are allowed for 30 days provided compensatory measures are satisfied, no releases were made during the period in which the monitor was out of service.

The Main Plant Vent flow rate measuring device required by ODCM Control 1.2.1.1 was out of service from July 16, 2013 at 0900 until November 25, 2013 at 1100 during which time all required compensatory actions were met. The flow rate measuring device was taken out of service after observing vent flows approximately 3%

above the normal operating band. During troubleshooting, flow fluctuations were observed inconsistent with normal system operation and decision was made to replace the transmitter. The replacement transmitter was not an exact replacement thus additional time was needed for offsite calibration. While waiting for a suitable replacement transmitter, review of system performance and design identified a success path to increase overall accuracy of the main plant flow indications by use of a transmitter with a longer probe. The decision was made to pursue system improvement through the stations engineering process. After the system was modified, testing at various flow configurations was performed for both trains demonstrating improved system accuracy.

J, Major Changes to Radioactive Waste Treatment Systems During 2013, there were no major changes to the Radioactive Waste Treatment System.

K. Carbon-14 Gaseous Effluents Carbon-14 production and release estimates were calculated using EPRI Report 1021106(3", "Estimation of Carbon-14 in Nuclear Plant Gaseous Effluents". This calculation uses active core coolant mass, average neutron flux by energy and reactor coolant nitrogen concentrations to determine Carbon-14 generation based upon an effective full power year. The 6

ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT JANUARY - DECEMBER, 2013 VIRGIL C. SUMMER NUCLEAR STATION SOUTH CAROLINA ELECTRIC & GAS estimated generation for VC Summer Nuclear station for 2013 was 9.87 curies.

Public dose estimates were performed using Regulatory Guide 1.109 (4) methodology. Carbon dioxide is assumed to make up 20% of the Carbon-14 gaseous emissions from the station based upon available references and on-site testing. Carbon-14 is the highest dose contributor of all radionuclides released in gaseous effluents. Annual dose resulting from Carbon-14 releases in gaseous effluents is estimated to be 1.61 E-01 mrem total body and 8.06E-01 mrem to the maximum organ (bone).

L. Supplemental Groundwater Monitoring Ten temporary monitoring wells were installed down gradient from the Industrial and Sanitary Waste Treatment System ponds on July 18, 2013 as requested by South Carolina Department of Health and Environmental Control (SCDHEC) related to the station's NPDES permit. Samples from three of the wells identified tritium in a range of 2.01 E+03 to 3.21 E+03 pCi/I with no gamma activity above the Minimum Detectable Activity for the Environmental Monitoring Program. The three affected wells are located closest to the Plant Waste Surge Basin which receives the discharge of the Turbine Building Sump and contains tritium at levels which are normally a small fraction of effluent discharge limits but well within environmental measurement detection capability. This plume was reported to the NRC and SCDHEC during our initial response to NEI 07-07 (5), Industry Groundwater Protection Initiative. After receiving permission from SCDHEC, one of the wells was converted to a permanent monitoring well and added to the station's Supplemental Radioactive Monitoring Program. The others were decommissioned on November 12, 2013.

7

ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT JANUARY - DECEMBER, 2013 VIRGIL C. SUMMER NUCLEAR STATION SOUTH CAROLINA ELECTRIC & GAS References

1. Regulatory Guide 1.21, "Measuring,Evaluating, and Reporting Radioactivity in Solid Wastes and Releases of Radioactive Materialsin Liquid and Gaseous Effluents From Light Water Cooled Nuclear Power Plants", Revision 1 ,US Nuclear Regulatory Commission, Washington D.C.,

1974.

2. NUREG/CR-4007, "Lower Limits of Detection:Definitionand Elaborationof a Proposed Position for RadiologicalEffluent and Environmental Measurements",National Technical Information Service, Springfield , Virginia, 1984.

3. EPRI Report 1021106, "Estimationof Carbon-14 in Nuclear PowerPlant Gaseous Effluents", Electric Power Research Institute, Palo Alto, CA, 2010.

4. Regulatory Guide 1.109, "Calculationof Annual Doses to Man from Routine Releases of 'Reactoreffluents for the Purposeof Demonstrating compliance with 10 CFR50, Appendix f', US Nuclear Regulatory Commission, Washington D.C., 1974.

5. NEI Report NEI 07-07, "IndustrialGround Water Protection Initiative-Final Guidance Document Report", Nuclear Energy Institute, Washington, D.C.,

2007.

8

APPENDIX A ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT JANUARY - DECEMBER, 2013 VIRGIL C. SUMMER NUCLEAR STATION SOUTH CAROLINA ELECTRIC & GAS SUPPLEMENTAL INFORMATION Regulatory Dose Limits:

A. Fission and Activation Gases:

The air dose to an individual due to noble gases released in gaseous effluents shall be limited to less than or equal to 5 mrad for gamma radiation and 10 mrad for beta radiation during any calendar quarter and 10 mrad for gamma radiation and 20 mrad for beta radiation during any calendar year (ODCM, Section 1.2.3.1).

B. lodines, Particulates (half-lives > 8 days) and Tritium:

The dose to an individual from radioidines, tritium and radioactive materials in particulate form with half-lives greater than 8 days in gaseous effluents shall be limited to less than or equal to 7.5 mrem to any organ during any calendar quarter and 15 mrem to any organ during any calendar year (ODCM, Section 1.2.4.1).

C. Liquid Effluents:

The dose or dose commitment to an individual from radioactive materials in liquid effluents released shall be limited to less than or equal to 1.5 mrem to the total body and 5 mrem to any organ during any calendar quarter and 3 mrem to the total body and 10 mrem to any organ during any calendar year (ODCM, Section 1.1.3.1).

D. All Sources:

The annual dose equivalent shall not exceed 25 mrem to the whole body, 75 mrem to the thyroid and 25 mrem to any other organ (40 CFR 190).

I1. Dose Rate and Effluent Concentration Limits:

A. Fission and Activation Gases The dose rate in unrestricted areas due to radioactive materials 9

APPENDIX A ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT JANUARY - DECEMBER, 2013 VIRGIL C. SUMMER NUCLEAR STATION SOUTH CAROLINA ELECTRIC & GAS released in gaseous effluents shall be limited to less than or equal to 500 mrem/year to the total body and less than or equal to 3000 mrem/year to the skin (ODCM, Section 1.2.2.1).

B. lodines, Particulates (half-lives > 8 days) and Tritium:

The dose rate in unrestricted areas due to radioactive materials in effluents shall be limited to less than or equal to 1500 mrem/year to any organ (ODCM, Section 1.2.2.1).

C. Liquid Effluents:

The concentration of radioactive materials released from the site shall be limited to 10 times the concentrations specified in 10 CFR 20, Appendix B, Table 2, Column 2 for radionuclides other than dissolved or entrained noble gases. For dissolved or entrained noble gases, the concentration shall be limited to 2E-04 pCi/ml total activity (ODCM, Section 1.1.2.1).

Ill. Average Energy:

Not Applicable IV. Measurements and Approximations of Total Radioactivity:

A. Fission and activation gases: Gamma spectrometry (HPGe)

B. lodines: Gamma spectrometry (HPGe)

C. Particulates: Gamma spectrometry (HPGe), beta proportional counting, alpha proportional counting D. Tritium: Liquid scintillation E. Liquid effluents:Gamma spectrometry (HPGe), liquid scintillation (H-3), beta proportional counting, alpha proportional counting.

10

APPENDIX A ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT JANUARY- DECEMBER, 2013 VIRGIL C. SUMMER NUCLEAR STATION SOUTH CAROLINA ELECTRIC & GAS V. Batch Releases:

A. Gaseous:

1. Number of batch releases: 1

2. Total time period for batch releases: 633 min.

3. Maximum time period for a batch release: 633 min.

4. Average time period for a batch release: 633 min.

5. Minimum time period for a batch release: 633 min.

B. Liquid:

1. Number of batch releases:

55 For first quarter, 2013 44 For second quarter, 2013 24 For third quarter, 2013 26 For fourth quarter, 2013

2. Total time period for batch releases:

4.55E+03 min. for first quarter, 2013 3.45E+03 min. for second quarter, 2013 2.24E+03 min. for third quarter, 2013 2.41 E+03 min. for fourth quarter 2013

3. Maximum time period for a batch release:

9.70E+01 min. for first quarter, 2013 9.60E+01 min. for second quarter, 2013 1.42E+02 min. for third quarter, 2013 1.OOE+02 min. for fourth quarter, 2013 11

APPENDIX A ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT JANUARY - DECEMBER, 2013 VIRGIL C. SUMMER NUCLEAR STATION SOUTH CAROLINA ELECTRIC & GAS

4. Average time period for batch releases:

82.8E+01 min. for first quarter, 2013 7.84E+01 min. for second quarter, 2013 9.33E+01 min. for third quarter, 2013 9.25E+01 min. for fourth quarter, 2013

5. Minimum time period for a batch release:

7.70E+01 min. for first quarter, 2013 1.OOE+00 min. for second quarter, 2013*

8.40E+01 min. for third quarter, 2013 8.OOE+01 min. for fourth quarter, 2013

NaOH sump release

6. Average stream flow during periods of release of effluent into a flowing stream:

4.55E+06 gpm for first quarter, 2013 4.66E+06 gpm for second quarter, 2013 6.28E+06 gpm for third quarter, 2013 4.31 E+06 gpm for fourth quarter, 2013 VI. Abnormal Releases:

A. Gaseous:

1. Number of releases: 0

2. Total activity released: 0 B. Liquid:

1. Number of releases: 0

2. Total activity released: 0 12

ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT January - December 2013 Virgil C. Summer Nuclear Station South CarolinaElectric & Gas Table 1 Gaseous Effluents Summation of All Releases UN In Hil5 I SIEUCU TIRDI~ I-UUR I H zuW1; hST.

QUARTER QUARTER V QUATER QUARTER TOTAL ERROR %/

A. Fission & Activation Gases

1. Total release Ci 2.77E-01 5.84E-01 0.00E+00 0.00E+00 8.60E-01 1.83E+01

2. Average release rate uCi /sec 3.56E-02 7.42E-02 0.OOE+00 0.OOE+00 2.73E-02

3. Percent ODCM Qtr. gamma air dose limit 1.88E-03 1.96E-02 0.OOE+00 0.OOE+00 N/A

4. Percent ODCM annual gamma air dose limit 9.42E-04 1.07E-02

1.07E-02

1.07E-02

1.07E-02

5. Percent ODCM Qtr. beta air dose limit 9.66E-04 3.70E-03 0.OOE+00 0.OOE+00 N/A

6. Percent ODCM annual beta air dose limit 4.83E-04 2.33E-03

2.33E-03

2.33E-03

2.33E-03 B. lodines 1.Total iodine - 131 Ci 0.00E+00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 3.73E+01

2. Average release rate uCi / sec 0.00E+00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 C. Particulates

1. Particulates with half - lifes > 8 days Ci 7.91 E-04 0.OOE+00 0.OOE+00 5.03E-06 7.96E-04 6.31E+01

2. Average release rate uCi / sec 1.02E-04 0.OOE+00 0.OOE+00 6.32E-07 2.52E-05

3. Gross alpha radioactivity Ci 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 D. Tritium

1. Total release Ci 3.16E-01 1.40E-03 0.OOE+00 0.OOE+00 3.18E-01 3.56E+01

2. Average release rate uCi / sec 4.07E-02 1.77E-04 0.OOE+00 0.OOE+00 1.01E-02 E. Carbon 14

1. Total release Ci 2.45E+00 2.45E+00 2.48E+00 2.48E+00 9.87E+00 N/A

2. Average release rate uCi / sec 3.12E-01 3.12E-01 3.12E-01 3.12E-01 3.12E-01 F. Organ Dose (from B,C,and D)

1. Percent ODCM Qtr. organ dose limit 2.24E-03 9.80E-06 0.OOE+00 0.OOE+00 N/A

2. Percent ODCM annual organ dose limit 1.12E-03 1.13E-03

1.13E-03

1.13E-03

1.13E-03

Cumulative

ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT JANUARY - DECEMBER, 2013 TABLE 2 GASEOUS EFFLUENTS -- GROUND-LEVEL RELEASES Continuous Mode Batch Mode Nucides Released Units First Second Third Fourth IAnnual I rs Second Third Fourth Annual Quarter Quarter Quarter Quarter Total uarter Quarter Quarter Quarter Total 1 Fission gases Krypton-85 Ci 0 0 0 0 0 0 2.38E-04 0 0 2.38E-04 Krypton-85m Ci 0 0 0 0 0 0 0 0 0 0 Krypton-88 Ci 0 0 0 0 0 0 0 0 0 0 Xenon-131m Ci 0 0 0 0 0 0 2.35E-05 0 0 2.35E-05 Xenon-133 Ci 1.54E-01 0 0 0 1.54E-01 0 1.84E-03 0 0 1.84E-03 Xenon-133m Ci 0 0 0 0 0 0 7.32E-06 0 0 7.32E-06 Xenon-135 Ci 9.81 E-02 6.77E-02 0 0 1.66E-01 0 0 0 0 0 Xenon-135m Ci 0 0 0 0 0 0 0 0 0 0 Xenon-138 Ci 0 0 0 0 0 0 0 0 0 0 Other: Ar-41 Ci 2.46E-02 5.14E-01 0 0 5.39E-01 0 0 0 0 0 Unidentified: None Ci 0 0 0 0 0 0 0 0 0 0 Total for Period Ci 2.77E-01 5.82E-01 0 0 8.58E-01 0 2.11E-03 0 0 2.11E03 2 lodines and other halogens Iodine-131 Ci 0 0 0 0 0 0 0 0 0 0 Iodine-132 Ci 0 0 0 0 0 0 0 0 0 0 Iodine-133 Ci 0 0 0 0 0 0 0 0 0 0 Arsenic-76 Ci 7.91 E-04 0 0 0 7.91E-04 0 0 0 0 0 Unidentified: None Ci 0 0 0 0 0 0 0 0 0 0 Total for Period Ci 7.91E-04 0 0 0 7.91 E-04 0 0 0 0 0 3 Particulates Cromium-51 Ci 0 0 0 0 0 0 0 0 0 0 Manganese-54 Ci 0 0 0 0 0 0 0 0 0 0 Cobalt-58 Ci 0 0 0 0 0 0 0 0 0 0 Cobalt-60 Ci 1.61 E-07 0 0 0 1.61 E-07 0 0 0 0 0 Zinc-65 Ci 0 0 0 0 0 0 0 0 0 0 Stronium-89 Ci 0 0 0 0 0 0 0 0 0 0 Stronium-90 Ci 0 0 0 0 0 0 0 0 0 0 Niobium-95 Ci 0 0 0 0 0 0 0 0 0 0 Zirconium-95 Ci 0 0 0 0 0 0 0 0 0 0 Silver-110m Ci 0 0 0 0 0 0 0 0 0 0 Tellurium-123m Ci 0 0 0 0 0 0 0 0 0 0 Antimony-124 Ci 0 0 0 0 0 0 0 0 0 0 Antimony-125 Ci 0 0 0 0 0 0 0 0 0 0 Tellurium-125m Ci 0 0 0 0 0 0 0 0 0 0 Antimony-127 Ci 0 0 0 0 0 0 0 0 0 0 Cesium-134 Ci 0 0 0 0 0 0 0 0 0 0 Cesium-137 Ci 0 0 0 0 0 0 0 0 0 0 Other: Be-7 Ci 0 0 0 5.03E-06 5.03E-06 0 0 0 0 0 Unidentified: None Ci 0 0 0 0 0 0 0 0 0 0 Total for Period Ci 1.61E-07 0 0 5.03E-06 5.19E-06 0 0 0 0 0

Tritium and Carbon-14 not included. See Table 1.

ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT JANUARY-DECEMBER, 2013 Virgil C. Summer Nuclear Station South CarolinaElectric & Gas TABLE 3 Liquid Effluents Summation of All Releases I-H* I bI:LUUN U I HIIKU I-UUh I H zgl;j I- I UNITS QUARTER QUARTER QUARTER QUARTER TOTAL ERROR %

A. Fission & Activation Products

1. Total release Ci 2.05E-03 1.85E-03 3.32E-04 7.86E-04 5.02E-03 2.21E+01

2. Average diluted concentration uCilml 6.85E-12 5.94E-12 1.04E-12 2.51 E-12 4.04E-12 B. Tritium

1. Total release Ci 1.40E+02 8.37E+01 1.46E+01 1.12E+02 3.50E+02 1.82E+01

2. Average diluted concentration uCi/ml 4.69E-07 2.68E-07 4.58E-08 3.57E-07 2.82E-07 C. Dissolved and entrained gases

1. Total release Ci 5.80E-04 1.07E-04 O.OOE+00 1.35E-04 8.21E-04 2.16E+01

2. Average diluted concentration uCilml 1.94E-12 3.43E-13 O.OOE+00 4.30E-13 6.61 E-13 3.Percent ODCM limit ( 2.OE-4 uCi/ml) % 9.69E-07 1.71 E-07 O.OOE+00 2.15E-07 3.31 E-07 D. Gross alpha radioactivity

1. Total release Ci O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 N/A E. Volume of waste released (undiluted) liters 1.29E+07 1.03E+07 8.54E+06 9.67E+07 1.28E+08 3.OOE+00 F. Volume of dilution water liters 2.99E+11 3.12E+11 3.18E+11 3.13E+11 1.24E+12 4.30E+00 G. ODCM limits (from A and B1)

1. Percent of ODCM Qtr total body limit % 1.19E-01 6.09E-02 9.53E-03 7.84E-02 N/A

2. Percent of ODCM annual total body limit % 5.93E-02 8.98E-02

9.45E-02

1.34E-01

1.34E-01 *

3. Percent of ODCM Qtr max. organ limit** % 3.88E-02 1.94E-02 2.87E-03 2.38E-02 N/A

4. Percent of ODCM annual max. organ limit** % 1.94E-02 2.91 E-02

3.05E-02

4.03E-02

4.21 E-02 *

Cumulative

- See Section E for max. organ for each quarter and cumulative.

ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT JANUARY - DECEMBER, 2013 Virgil C. Summer Nuclear Station, South CarolinaElectric & Gas TABLE 4 LIQUID EFFLUENTS d- ý:_ All .1 D + 16 X/F A Nuclides Released* Units First Second Third Fourth Annual Totali First Second Third Fourth Annual Total I___I__" _r _ Ouart___________ I rterI Ourer I Ouarter Ouarter IOuarter IOarter Strontium-89 Ci 0 0 0 0 0 0 0 0 0 0 Strontium-90 Ci 0 0 0 0 0 0 0 0 0 0 Cesium-134 Ci 0 0 0 0 0 0 0 0 0 0 Cesium-137 Ci 0 0 0 0 0 5.46E-06 3.40E-09 2.37E-06 5.39E-05 6.17E-05 Iodine-131 Ci 0 0 0 0 0 0 0 0 0 0 Iodine-132 Ci 0 0 1 0 0 0 0 0 0 0 0 Cobalt-57 Ci 0 0 0 0 0 0 0 0 0 0 Cobalt-58 Ci 0 0 0 0 0 1.25E-04 1.18E-04 1.95E-05 5.82E-05 3.21E-04 Cobalt-60 Ci 0 0 0 0 0 7.27E-04 1.14E-03 1.94E-04 4.85E-04 2.55E-03 Iron-59 Ci 0 0 0 0 0 0 0 0 0 0 Zinc-65 Ci 0 0 0 0 0 0 1.43E-05 0 0 1.43E-05 Manganese-54 Ci 0 0 0 0 0 8.57E-05 8.59E-05 1.32E-05 3.69E-05 2.22E-04 Chronium-51 Ci 0 0 0 0 0 1.76E-05 1.35E-04 0 0 1.53E-04 Zirconium-Niobium-95 Ci 0 0 0 0 0 1.46E-04 2.73E-05 0 0 1.73E-04 Zirconium-97 Ci 0 0 0 0 0 0 0 0 0 0 Molybdenurn-99 Ci 0 0 0 0 0 0 0 0 0 0 Technetium-99m Ci 0 0 0 0 0 0 0 0 0 0 Cerium-144 Ci 0 0 0 0 0 0 0 0 0 0 Other: Be-7 Ci 0 0 0 0 0 0 0 0 0 0 Na-24 Ci 0 0 0 0 0 0 0 0 0 0 Fe-55 Ci 0 0 0 0 0 0 0 0 0 0 As-76 Ci 0 0 0 0 0 2.88E-05 1.44E-05 0 0 4.32E-05 Nb-94 Ci 0 0 0 0 0 0 0 0 0 0 AF-110m Ci 0 0 0 0 0 4.23E-07 0 0 0 4.23E-07 Sb-122 Ci 0 0 0 0 0 0 0 1.89E-05 3.69E-06 2.26E-05 Sb-124 Ci 0 0 0 0 0 3.38E-06 3.04E-05 0 0 3.38E-05 Sb-125 Ci 0 0 0 0 0 9.34E-05 8.10E-05 8.37E-05 1.45E-04 4.03E-04 Sb-127 Ci 0 0 0 0 0 0 0 0 1.52E-06 1.52E-06 Te-123m Ci 0 0 0 0 0 2.19E-05 3.65E-06 0 9.85E-07 2.65E-05 Te-125m Ci 0 0 0 0 0 7.96E-04 2.08E-04 0 0 1.00E-03 Te-132 Ci 0 0 0 0 0 0 0 0 0 0 Total for Period (above) Ci 0 0 0 0 0 2.05E-03 1.86E-03 3.32E-04 7.85E-04 5.03E-03 Ar-41 Ci 0 0 0 0 0 0 0 0 0 0 Kr-85 Ci 0 0 0 0 0 0 0 0 1.08E-04 1.08E-04 Xenon-133 Ci 0 0 0 0 0 5.60E-04 1.07E-04 0 2.67E-05 6.94E-04 Xenon-133mn Ci 0 0 0 0 0 6.73E-06 0 0 0 6.73E-06 Xenon-135 Ci 0 0 0 0 0 1.30E-05 0 0 0 1.30E-05 Total Entrained Gases Ci 0 0 0 0 0 5.80E-04 1.07E-04 0 1.35E-04 8.21E-04

Tritium not included. See Table 3 for tritium numbers No Unidentified nuclides round

ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT January - December, 2013 Virgil C. Summer NuclearStation South CarolinaElectric & Gas Table 5 SOLID WASTE SHIPMENTS

1. Solid Waste Shipped Offsite for Burial or Disposal (Not irradiated fuel).

Type of Waste Unit 2013 Est. Total Total Error, %

a. Spent resins, filters, sludge, m3 1.26E+01 +/-2.5E+01 evaporator bottoms, etc. Ci 7.19E-01

b. Dry compressible waste m3 1.87E+02 +/-2.5E+01 contaminated equip., etc. Ci 2.20E-01

c. Irradiated components, m 0 control rods, etc. Ci 0 N/A

d. Other m3 5.15E+00 +/-2.5E+01 Ci 3.81E-03

2. Estimate of major nuclide composition for the year (by type of waste) for concentrations above 1.0%.

a.

Ni-63 40.2% 2.89E-01 Ci Co-60 22.7% 1.64E-01 Ci C-14 9.9% 7.12E-02 Ci Fe-55 8.4% 6.06E-02 Ci Cs-137 5.3% 3.84E-02 Ci Sb-125 5.2% 3.75E-02 Ci Mn-54 2.2% 1.56E-02 Ci Co-58 1.8% 1.26E-02 Ci b.

Co-60 39.0% 8.57E-02 Ci Fe-55 25.6% 5.64E-02 Ci Ni-63 10.5% 2.32E-02 Ci Mn-54 5.7% 1.26E-02 Ci Co-58 4.3% 9.42E-03 Ci Nb-95 3.9% 8.47E-03 Ci Cr-51 2.7% 5.92E-03 Ci Zr-95 2.6% 5.78E-03 Ci Sb-125 2.0% 4.40E-03 Ci Pu-241 1.1% 2.35E-03 Ci C.

I None

ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT January - December, 2013 Virgil C. Summer Nuclear Station South CarolinaElectric & Gas Table 5 SOLID WASTE SHIPMENTS d

C-60 47.6% 1.81 E-03 Ci Fe-55 28.4% 1.08E-03 Ci Ni-63 13.8% 5.24E-04 Ci Mn-54 4.1% 1.57E-04 Ci Sb-125 2.1% 7.98E-05 Ci Pu-241 1.5% 5.75E-05 Ci

3. Solid Waste Disposition Numbers of Shipments Mode of Transportation Destination 2 Southern Pines Alaron Nuclear Services 4 Hittman Trucking Barnwell Processing Facility 1 Hittman-Tenn Energy Solutions/Duratek 4 Hittman-Tenn Energy Solutions/Duratek 1 Tri-State Motor transit Energy Solutions/Duratek 1 Cast Transportation Perma-Fix of Florida Notes:

1. (4) Shipments of DAW/Metal/Carpet were sent to Energy Solutions for processing.

2. (4) Shipments of Gross Dewatered Resin were sent to Barnwell Waste Processing facility for Total Dewatering cycles for disposal at Barnwell Disposal facility.

3. (1) Shipment of Charcoal Plenum and Charcoal Plenum filters were sent to Energy Solutions for processing.

4. (1) Shipment of 18 drums of Waste Oil was sent to Energy Solutions for processing.

5. (2) Shipments of FHB Crane Metal Waste was sent to Alaron for processing.

6. (1) Shipment of Mixed Waste containing 8 drums was sent to Perma Fix of Florida for processing.

ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT JANUARY - DECEMBER 2013 Virgil C. Summer Nuclear Station South CarolinaElectric & Gas Table 6 GASEOUS AND LIQUID DOSES I I I - I ro'w I WU*K I IrK ork.'urU wumr I r I NIU WUMIt I IJ1 I'UUKI IN QUARTIEI- ITUAL Y 4 q P p P q 4 ODCM Percent Percent Percent Percent ANNUAL Section GASEOUS LIMITS DOSE of Limit DOSE of Limit DOSE of Limit DOSE of Limit DOSE I

1.2.3. 1. a,b I 5 mrad gamma / qtr. I I I . _ _

9.42E-05 mrad 1.88E-03 9.80E-04 mrad 1.96E-02 0.00E+00 mrad 0.00E+00 0.OOE+00 mrad O.0OE+00 1.07E-03 10 mrad gamma / yr. 9.42E-04 1.07E-02* 1.07E-02

1.07E-02

1.2.3.1 .a,b 10 mrad beta / qtr. 9.66E-05 mrad 9.66E-04 3.70E-04 mrad 3.70E-03 0.00E+00 mrad 0.00E+00 0.00E+00 mrad 0.00E+00 4.67E-04 20 mrad beta / yr. 4.83E-04 2.33E-03 2.33E-03

2.33E-03

1.2.4.1.a,b 7.5 mrem organ/qtr 2.00E-01 mrem** 2.67E+00 2.OOE-01 mrem** 2.67E+00 2.03E-01 mrem** 2.70E+00 2.03E-01 mrem** 2.70E+00 8.06E-01 15 mrem organ/yr. 1.34E+00 2.67E+00

4.02E+00

5.37E+00

LIQUID LIMITS 1.1.3.la,b 1.5 mrem /qtr. 1.78E-03 mrem 1.19E-01 9.14E-04 mrem 6.09E-02 1.43E-04 mrem 9.53E-03 1.18E-03 mrem 7.84E-02 4.01 E-03 3 mrem / yr. 5.93E-02 8.98E-02

9.45E-02

1.34E-01

1.1.3.1a,b 5 mrem organ/qtr*** 1.94E-03 mrem 3.88E-02 9.72E-04 mrem 1.94E-02 1.44E-04 mrem 2.87E-03 1.19E-03 mrem 2.38E-02 4.21 E-03

t "/t 10 mrem organ/yr. (GI-LLI) 1.94E-02 (GI-LLI) 2.91E-02 (GI-LLI) 3.05E-02 * (LIVER) 4.03E-02 (GI-LLI)

Includes contribution from previous quarters

- Includes dose from all nuclides including Carbon-14 See Section E for max organ for each quarter

ENCLOSURE A ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT JANUARY - DECEMBER, 2013 VIRGIL C. SUMMER NUCLEAR STATION SOUTH CAROLINA ELECTRIC & GAS OFFSITE DOSE CALCULATION MANUAL Revision 29 (August 2013)

OFFSITE DOSE CALCULATION MANUAL FOR SOUTH CAROLINA ELECTRIC AND GAS COMPANY VIRGIL C. SUMMER NUCLEAR STATION SAFETY RELATED Approval Original signed by George Lippard IR/27/1 ,:

General Manager, Date Nuclear Plant Operations PSRC Approval Original signed by Robert Justice 8/27/13 Date Revision 29 (August 2013)

Reviewed by: Original signed by Michael Roberts / 8/26/13 Date Approved by: Original signed by Moses Coleman / 8/27/13 Date NUCLEAR OPERATIONS COPY NO.

LIST OF EFFECTIVE PAGES Page Revision P Revision i 29 1.0-34 29 ii 29 1.0-35 23 iii 24 1.0-36 24 iv 21 1.0-37 23 v 24 1.0-38 24 vi 24 1.0-39 24 vii 21 1.0-40 21 viii 21 1.0-41 21 ix 28 1.0-42 21 x 28 1.0-43 23 xi 28 1.0-44 21 1.0-45 21 1.0-46 29 1.0-1 24 1.0-47 29 1.0-2 24 1.0-48 24 1.0-3 27 1.0-49 29 1.0-4 27 1.0-50 25 1.0-5 24 1.0-51 25 1.0-6 21 1.0-52 25 1.0-7 24 1.0-53 29 1.0-8 25 1.0-54 26 1.0-9 24 1.0-55 26 1.0-10 24 1.0-56 21 1.0-11 29 1.0-57 24 1.0-12 24 1.0-13 25 2.0-1 21 1.0-14 24 2.0-2 21 1.0-15 24 2.0-3 21 1.0-16 24 2.0-4 21 1.0-17 24 2.0-5 21 1.0-18 24 2.0-6 21 2.0-7 21 1.0-19 27 2.0-8 21 1.0-20 24 2.0-9 21 1.0-21 24 2.0-10 21 1.0-22 29 2.0-11 21 1.0-23 29 2.0-12 21 1.0-24 24 2.0-13 21 1.0-25 24 2.0-14 29 2.0-14A 29 1.0-26 29 2.0-15 21 1.0-27 24 2.0-16 21 1.0-28 24 2.0-17 21 1.0-29 25 2.0-18 21 1.0-30 25 2.0-19 28 1.0-31 23 2.0-20 21 1.0-32 23 2.0-21 21 1.0-33 26 2.0-22 21 ODCM, V. C. Summer/SCE&G: Revision 29 (August 2013) i

LIST OF EFFECTIVE PAGES (continued)

Page Revision Page Revision 2.0-23 21 3.0-26 28 2.0-24 21 3.0-27 28 2.0-25 21 3.0-28 28 2.0-26 21 3.0-29 28 2.0-27 21 3.0-30 28 2.0-28 21 3.0-31 28 2.0-29 21 3.0-32 28 2.0-30 21 3.0-33 28 2.0-31 21 3.0-34 28 2.0-32 28 3.0-35 28 2.0-33 21 3.0-36 28 2.0-34 21 3.0-37 16 2.0-35 21 3.0-38 17 2.0-36 22 3.0-39 17 2.0-37 21 3.0-40 28 2.0-38 24 3.0-41 13 2.0-39 24 3.0-42 13 2.0-40 21 3.0-43 28 2.0-41 28 3.0-44 13 2.0-42 21 3.0-45 16 2.0-43 21 3.0-46 16 3.0-47 16 3.0-1 13 3.0-48 13 3.0-2 13 3.0-49 13 3.0-3 17 3.0-50 13 3.0-4 13 3.0-51 13 3.0-5 13 3.0-52 13 3.0-6 28 3.0-7 13 4.0-1 26 3.0-8 18 4.0-2 26 3.0-9 13 4.0-3 26 3.0-10 28 4.0-4 24 3.0-10A 24 4.0-5 29 3.0-11 13 4.0-6 28 3.0-12 15 4.0-7 28 3.0-13 28 4.0-8 23 3.0-14 13 4.0-9 26 3.0-15 16 4.0-10 23 3.0-16 28 4.0-11 23 3.0-17 16 4.0-12 28 3.0-18 13 4.0-13 28 3.0-19 13 3.0-20 13 3.0-21 28 3.0-22 28 3.0-23 28 3.0-24 28 3.0-25 28 ODCM, V. C. Summer/SCE&G: Revision 29 (August 2013) ii

Table of Contents PAGE List of Effective Pages ............................................................................................................. i Table of Contents .............................................................................................................. iii List of Tables ................................................................................................................... v List of Figures ................................................................................................................... vi References ............................................................................................................................... vii Introduction .............................................................................................................................. viii Responsibilities ....................................................................................................................... ix 1.0 SPECIFICATION OF CONTROLS 1.1 Liquid Effluents ................................................................................................ 1.0-1 1.1.1 Radioactive Liquid Effluent Monitoring Instrum entation .................................................................................... 1.0-1 1.1.2 Liquid Effluents: Concentration ........................................................ 1.0-8 1.1.3 Liquid Effl uents: Dose ........................................................................ 1.0-14 1.1.4 Liquid W aste Treatm ent ...................................................................... 1.0-15 1.2 Gaseous Effluents ........................................................................................... 1.0-17 1.2.1 Radioactive Gaseous Effluent Monitoring Instrum entation .................................................................................... 1.0-17 1.2.2 Gaseous Effl uents: Dose Rate ........................................................... 1.0-22 1.2.3 Gaseous Effluents: Dose - Noble Gas ............................................... 1.0-25 1.2.4 Gaseous Effluents: Dose - Radioiodines, Tritium and Radioactive Materials in Particulate Form ................................. 1.0-26 1.2.5 Gaseous Radw aste Treatm ent ........................................................... 1.0-27 1.3 Radioactive Effluents: Total Dose .................................................................. 1.0-29 1.4 Radiological Environm ental Monitoring ........................................................ 1.0-31 1.4.1 Monitoring Program ............................................................................ 1.0-31 1.4.2 Land Use Census ................................................................................ 1.0-41 1.4.3 Interlaboratory Com parison Program ................................................ 1.0-43 1.5 Bases ............................................................................................................... 1.0-44 1.6 Reporting Requirem ents ................................................................................. 1.0-52 1.6.1 Annual Radiological Environmental Operating Report ................................................................................................... 1.0-52 1.6.2 Annual Radioactive Effluent Release Report .................................... 1.0-53 1.6.3 Major Changes to Radioactive Waste Treatment System (Liquid and Gaseous) ............................................................ 1.0-54 1.6.4 V. C. Summer Groundwater Protection Program ........................ 1.0-55 1.7 Definitions ........................................................................................................ 1.0-56 ODCM, V. C. Summer/SCE&G: Revision 25 (January 2007) iii

Table of Contents 2.0 LIQUID EFFLUENT 2.1 Liquid Effluent Monitor Setpoint Calculation ................................................ 2.0-1 2.1.1 Liquid Effluent Monitor Setpoint Calculation Param eters ........................................................................................... 2.0-2 2.1.2 Liquid Radwaste Effluent Line Monitors ............................................ 2.0-6 2.1.3 Liquid Radwaste Discharge Via Industrial and Sanitary Waste System ....................................................................... 2.0-14 2.1.4 Steam Generator Blowdown, Turbine Building Sump, and Condensate Demineralizer Backwash Effluent Lines ....................................................................................... 2.0-15 2.2 Dose Calculation for Liquid Effluents ............................................................ 2.0-32 2.2.1 Liquid Effluent Dose Calculation Parameters ................................... 2.0-32 2.2.2 Methodology ........................................................................................ 2.0-33 2.3 Liquid Effluent Releases through the Neutralization B asin ............................................................................................................... 2.0-35 2.3.1 Rainwater Tank .................................................................................... 2.0-35 2.3.2 NaOH Spray Tank and Stored NaOH .................................................. 2.0-36 3.0 GASEOUS EFFLUENT 3.1 Gaseous Effluent Monitor Setpoint ................................................................ 3.0-1 3.1.1 Gaseous Effluent Monitor Setpoint Calculation Param eters ........................................................................................... 3.0-1 3.1.2 Station Vent Noble Gas Monitors ....................................................... 3.0-5 3.1.3 Waste Gas Decay System Monitor ..................................................... 3.0-7 3.1.4 Alternative Methodology for Establishing Conservative Setpoints ....................................................................... 3.0-8 3.1.5 O il Incineration ..................................................................................... 3.0-10 3.1.6 Meteorological Release Criteria for Batch Releases ............................................................................................... 3.0-10 3.2 Dose Calculation for Gaseous Effluent .......................................................... 3.0-12 3.2.1 Gaseous Effluent Dose Calculation Parameters ............................... 3.0-12 3.2.2 Unrestricted Area Boundary Dose ..................................................... 3.0-14 3.2.3 Unrestricted Area Dose to Individual ................................................. 3.0-15 3.3 Meteorological Model for Dose Calculations ................................................ 3.0-45 3.3.1 Meteorological Model Parameters ...................................................... 3.0-45 3.3.2 Meteorological Model .......................................................................... 3.0-46 4.0 RADIOLOGICAL ENVIRONMENTAL MONITORING .................................................. 4.0-1 ODCM, V. C. Summer/SCE&G: Revision 21 (March 1996) iv

LIST OF TABLES Table No. Page No.

1.1-1 Radioactive Liquid Effluent Monitoring Instrumentation ............................. 1.0-2 1.1-2 Radioactive Liquid Effluent Monitoring Instrumentation Surveillance Requirements ............................................................................. 1.0-5 1.1-3 Frequency Notation ......................................................................................... 1.0-7 1.1-4 Radioactive Liquid Waste Sampling and Analysis Program ........................ 1.0-10 1.2-1 Radioactive Gaseous Effluent Monitoring Instrumentation ......................... 1.0-18 1.2-2 Radioactive Gaseous Effluent Monitoring Instrumentation Surveillance Requirements ............................................................................. 1.0-20 1.2-3 Radioactive Gaseous Waste Sampling and Analysis P rogram ............................................................................................................ 1.0-23 1.4-1 Radiological Environmental Monitoring Program ........................................ 1.0-33 1.4-2 Reporting Levels for Radioactivity Concentrations in Environ-mental Samples Reporting Levels ................................................................. 1.0-38 1.4-3 Maximum Values for the Lower Limits of Detection (LLD) a,c Reporting Levels .............................................................................. 1.0-39 2.2-1 Bioaccumulation Factors ................................................................................ 2.0-37 2.2-2 Adult Ingestion Dose Factors ......................................................................... 2.0-38 2.2-3 Site Related Ingestion Dose Commitment Factor (Ai,) ................................. 2.0-40 3.1-1 Dose Factors for Exposure to a Semi-Infinite Cloud of Noble G ases ..................................................................................................... 3.0-4 3.1-2 Favorable Meteorology ................................................................................... 3.0-10A 3.2-1 Pathway Dose Factors for Section 3.2.2.2. (Pi) ................... 3.0-18 3.2-2 Pathway Dose Factors for Section 3.2.3.2. (Rj) ............................................. 3.0-21 3.2-3 Pathway Dose Factors for Section 3.2.3.3. (R1 ) (Infant) ............. 3.0-24 3.2-4 Pathway Dose Factors for Section 3.2.3.3. (Rj)(Child) ................................. 3.0-27 3.2-5 Pathway Dose Factors for Section 3.2.3.3. (R1 ) (Teenager) ........... 3.0-30 3.2-6 Pathway Dose Factors for Section 3.2.3.3. (Rj)(Adult)................................. 3.0-33 3.2-7 Controlling Receptors, Locations, and Pathways ........................................ 3.0-37 3.2-8 Atmospheric Dispersion Parameters for Controlling Receptor Locations ......................................................................................... 3.0-39 3.2-9 Parameters Used in Dose Factor Calculations ............................................. 3.0-40 4.0-1 Radiological Environmental Monitoring Program ........................................ 4.0-2 ODCM, V. C. Summer/SCE&G: Revision 24 (May 2006)

V

LIST OF FIGURES Figure No. Page No.

2.1-1 Example Liquid Monitor Calibration Curve ................... 2.0-31 2.2-1 Liquid Radwaste Treatment System ................................................. 2.0-42 3.1-1 Example Noble Gas Monitor Calibration Curve ................ 3.0-11 3.2-1 Gaseous Radwaste Treatment System .............................................. 3.0-44 3.3-1 Plume Depletion Effect for Ground Level Releases (5) ........... 3.0-49 3.3-2 Vertical Standard Deviation of Material in a Plume (cr)............... 3.0-50 3.3-3 Relative Deposition for Ground Level Releases (Dg) ............................ 3.0-51 3.3-4 Open Terrain Recirculation Factor ...................................................... 3.0-52 4.0-1 Radiological Environmental Sampling Locations (Remote) ....... 4.0-10 4.0-2 Radiological Environmental Sampling Locations (Local) ................. 4.0-11 4.0-3 Radiological Environmental Sampling Locations (Local) .................. 4.0-12 4.0-4 Radiological Environmental Sampling Locations (Local) .................. 4.0-13 ODCM, V. C. SummerISCE&G: Revision 25 (January 2007) vi

REFERENCES

1. Boegli, T.S., R.R. Bellamy, W.L. Britz, and R.L. Waterfield, "Preparation of Radiological Effluent Technical Specifications for Nuclear Power Plants" NUREG-0133 (October 1978).

2. "Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10CFR 50, Appendix I", U.S. NRC Regulatory Guide 1.109 (March 1976).

3. "Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10CFR 50, Appendix I", U.S. NRC Regulatory Guide 1.109, Rev. 1 (October 1977).

4. "Final Safety Analysis Report", South Carolina Electric and Gas Company, Virgil C.

Summer Nuclear Station.

5. "Operating License Environmental Report", South Carolina Electric and Gas Company, Virgil C. Summer Nuclear Station.

6. Wahlig, B.G., "Estimation of the Radioactivity Release Rate/Equilibrium Concentration Relationship for the Parr Pumped Storage System", Applied Physical Technology, Inc.,

February 1981.

7. "Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light - Water - Cooled Reactors", U.S. NRC Regulatory Guide 1.111 (March 1976).

8. "Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light - Water - Cooled Reactors", U.S. NRC Regulatory Guide 1.111, Rev. 1 (July 1977).

9. Slade, D.H., (editor), "Meteorology and Atomic Energy"; U.S. Atomic Energy Commission, AEC TID-24190, 1968.

10. "Measuring, Evaluating, and Reporting Radioactivity in Solid Wastes and Releases of Radioactive Materials in Liquid and Gaseous Effluents from Light-Water-Cooled Nuclear Power Plants", U.S. NRC Regulatory Guide 1.21, Rev. 1 (June 1974).

11. "Standard Radiological Effluent Technical Specifications for Pressurized Water Reactors",

NUREG-0472, Revision 3 (January 1983).

12. "Quality Assurance for Radiological Monitoring Programs (Normal Operations) - Effluent Streams and the Environment", USNRC Regulatory Guide 4.15, Revision 1 (February 1979).

13. "Age-Specific Radiation Dose Commitment Factors for a One-Year Chronic Intake",

NUREG-0172 (November 1977).

14. Generic Letter 89-01, "Implementation of Programmatic Controls for Radiological Effluent Technical Specifications", January 31, 1989.

15. "Offsite Dose Calculation Manual Guidance: Standard Radiological Effluent Controls for Pressurized Water Reactors", Generic Letter 89-01, Supplement No. 1, NUREG-1301, April 1, 1991.

ODCM, V. C. SummeriSCE&G: Revision 21 (March 1996) vii

INTRODUCTION The OFFSITE DOSE CALCULATION MANUAL (ODCM) is an implementing and C02-- supporting document of the RADIOLOGICAL EFFLUENT TECHNICAL SPECIFICATIONS (RETS). In accordance with USNRC Generic Letter 89-01, entitled "Implementation of Programmatic Controls for Radiological Effluent Technical Specifications in the Administrative Controls Section of the Technical Specifications and the Relocation of Procedural Details of RETS to the Offsite Dose Calculation Manual or to the Process Control Program", the procedural details for implementing the Radiological Controls have been incorporated into the ODCM. The ODCM describes the methodology and parameters to be used in the calculation of offsite doses due to radioactive liquid and gaseous effluents and in the calculation of liquid and gaseous effluent monitoring instrumentation alarm/trip setpoints. The ODCM contains a list and graphical description of the specific sample locations for the radiological environmental monitoring program.

Configurations of the liquid and gaseous radwaste treatment systems are also included.

The ODCM will be maintained at the Station as the reference which details the Radiological Effluent Controls of the V. C. Summer Nuclear Station. Additionally the ODCM will be maintained as the guide for accepted calculational methodologies. Changes in calculation methods or parameters will be incorporated into the ODCM in order to ensure that the ODCM represents the current methodology in all applicable areas. Computer software to perform described calculations will be maintained current with this ODCM.

ODCM, V. C. Summer/SCE&G: Revision 21 (March 1996) viii

RESPONSIBILITIES The ODCM contains the radiological effluent controls, their applicability, remedial actions, surveillance requirements, and their bases. Plant procedures implement responsibilities for compliance with the ODCM that include:

The Operations group is responsible for:

Declaring radioactive liquid and gaseous effluent monitor channels operable or inoperable.

0 Ensuring the minimum number of operable channels for radioactive liquid and gaseous effluent monitors.

a Notifying the responsible group to implement appropriate action if less than the minimum number of radioactive liquid and gaseous effluent monitor channels are operable.

0 Initiating a Condition Report (CR) in accordance with SAP-999 when less than the minimum number of channels operable condition prevails for more than 30 days.

a Restoring to within limits, the concentration of liquid radioactive material exceeding ODCM limits released from the site.

0 Ensuring radioactive liquid and gaseous effluent monitor setpoints are set as prescribed in the effluent release permit.

a Suspending release if radioactive liquid and gaseous effluent monitor setpoints are less conservative than ODCM requirements.

Declaring liquid and gaseous radwaste treatment systems operable or inoperable.

a Ensuring operability of gaseous and liquid radwaste treatment systems and ventilation exhaust treatment system.

Ensuring appropriate portions of the gaseous and liquid radwaste treatment systems are used to reduce the radioactive materials in liquid and gaseous waste prior to their discharge when the projected doses exceed limits specified by the ODCM.

a Initiating a CR in accordance with SAP-999 when liquid or gaseous radwaste system is inoperable for more than 31 days.

0 Performing channel check and source check at the frequencies shown in Tables 1.1-2 and 1.2-2 for each radioactive liquid and gaseous effluent monitoring instrumentation channel.

ODCM, V. C. Summer/SCE&G: Revision 28 (July 2012) ix

Instrumentation and Controls group is responsible for:

Performing channel calibration and analog channel operational test at the frequencies shown in Tables 1.1-2 and 1.2-2 for each radioactive liquid and gaseous effluent monitoring instrumentation channel.

Informing the Operations group of surveillance test results.

The Health Physics group is responsible for:

a Establishing setpoints for radioactive liquid and gaseous effluent monitors, consistent with ODCM methodology, and providing setpoints information to Operations.

0 Implementing remedial actions as requested by Operations. These actions include grab sampling and analysis and providing the results to Operations.

a Performing periodic radioactive effluent monitor checks to determine backgrounds, normal indications and verifying monitor correlation graphs, and providing this information as necessary to Operations.

a Implementing radioactive gaseous and liquid waste sampling and analysis program in accordance with ODCM Tables 1.1-4 and 1.2-3.

Informing Operations when at least one Circulating Water Pump or the Circulating Water Jockey Pump is required to provide dilution to the discharge structure.

Calculating cumulative dose contributions and performing dose projections from liquid and gaseous effluents in accordance with the ODCM and providing the information to Operations.

0 Initiating a CR in accordance with SAP-999 when calculated dose from the discharge of radioactive materials in liquid or gaseous effluents are in excess of the limits specified by ODCM Sections 1.1.3.1 or 1.2.3.1.

Initiating a CR in accordance with SAP-999 when liquid or gaseous waste is discharged without treatment and is in excess of the limits specified by ODCM Sections 1.1.4.1 or 1.2.3.1.

0 Initiating a CR in accordance with SAP-999 when the dose or dose commitment to any member of the public due to releases of radioactivity and radiation is in excess of 25 mrem to the total body or any organ (except the thyroid, which shall be limited to less than or equal to 75 mrem) over 12 consecutive months.

Implementing the Radiological Environmental Monitoring Program as specified in Section 1.4 of the ODCM.

ODCM, V. C. Summer/SCE&G: Revision 28 (July 2012) x

Initiating a CR in accordance with SAP-999 when the Radiological Environmental Monitoring Program limiting conditions for operation are exceeded.

Preparation of the Annual Radioactive Effluent Release Report and the Annual Environmental Operating Report.

ODCM, V. C. Summer/SCE&G: Revision 28 (July 2012) xi

1.0 SPECIFICATION OF CONTROLS 1.1 LIQUID EFFLUENTS 1.1.1 Radioactive Liquid Effluent Monitoring Instrumentation CONTROLS 1.1.1.1 The radioactive liquid effluent monitoring instrumentation channels shown in Table 1.1-1 shall be OPERABLE with their alarm/trip setpoints set to ensure that the limits of ODCM Specification 1.1.2.1 are not exceeded. The alarm/trip setpoints of these channels shall be determined in accordance with ODCM, Section 2.1.

APPLICABLE: At all times.

ACTION:

a. With a radioactive liquid effluent monitoring instrumentation channel alarm/ trip setpoint less conservative than required by the above specification, immediately suspend the release of radioactive liquid effluents monitored by the affected channel or declare the channel inoperable.

b. With less than the minimum number of radioactive liquid effluent monitoring instrumentation channels OPERABLE, take the ACTION shown in Table 1.1-1. Additionally, if this condition prevails for more than 30 days, in the next Annual Radioactive Effluent Release Report explain why this condition was not corrected in a timely manner.

c. The provisions of Technical Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 1.1.1.2 Each radioactive liquid effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION and ANALOG CHANNEL OPERATIONAL TEST operations at the frequencies shown in Table 1.1-2.

ODCM, V. C. Summer, SCE&G: Revision 24 (May 2006) 1.0-1

Table 1.1-1 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION MINIMUM CHANNELS OPERABLE ACTION INSTRUMENT

1. GROSS RADIOACTIVITY MONITORS PROVIDING ALARM AND AUTOMATIC TERMINATION OF RELEASE

a. Liquid Radwaste Effluent Line - RM-L5 or RM-L9 1 1

b. Nuclear (Processed Steam Generator) Blowdown 1 1 Effluent Line RM-L7 or RM-L9

c. Steam Generator Blowdown Effluent Line

1. Unprocessed during Power Operation - 1 2 RM-L10 or RM-L3

2. Unprocessed during Startup - RM-L3 1 2

d. Turbine Building Sump Effluent Line - RM-L8 1 3

e. Condensate Demineralizer Backwash Effluent Line 1 6 RM-L11

2. FLOW RATE MEASUREMENT DEVICES*

a. Liquid Radwaste Effluent Line - Tanks 1 and 2 1/tank 4

b. Penstock Minimum Flow Interlock** 1 4

c. Nuclear Blowdown Effluent Line 1 4

d. Steam Generator (Unprocessed) Blowdown 1 4 Effluent Line

3. TANK LEVEL INDICATING DEVICES

a. Condensate Storage Tank 1 5 In the event that simultaneous releases from both WMT and NBMT are required (which normally will be prevented by procedure) the flow rate for monitor RM-L9 will be determined by adding flow rates for monitors RM-L5 and RM-L7.

Minimum dilution flow is assured by an interlock that terminates liquid waste releases if the minimum dilution flow is not available.

ODCM, V. C. Summer, SCE&G: Revision 24 (May 2006) 1.0-2

Table 1.1-1 (Continued)

TABLE NOTATION ACTION 1 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases may continue provided that prior to initiating a release:

a. At least two independent samples are analyzed in accordance with ODCM Specification 1.1.2.4 and

b. At least two technically qualified members of the Facility Staff independently verify the release rate calculations and discharge line valving.

Otherwise, suspend release of radioactive effluents via this path-way.

ACTION 2 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided grab samples are analyzed by gamma spectroscopy for radioactivity at the LLD specified in Table 1.1-4 or samples are analyzed for gross radioactivity (beta and gamma) at a limit of detection of at least 1 E-7 microcuries/gram.

a. At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the specific activity of the secondary coolant is greater than 0.01 microcuries/gram DOSE EQUIVALENT 1-131, or

b. At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when the specific activity of the secondary coolant is less than or equal to 0.01 micro-curies/gram DOSE EQUIVALENT 1-131.

ACTION 3 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided that, at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , grab samples are collected and analyzed by gamma spectroscopy for radioactivity at the LLD specified in Table 1.1-4 or samples are analyzed for gross radioactivity (beta and gamma) at a limit of detection of at least 1 E-7 microcuries/gram.

ODCM, V. C. Summer, SCE&G: Revision 27 (January 2011) 1.0-3

Table 1.1-1 (Continued)

TABLE NOTATION ACTION 4 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided the flow rate is estimated at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months during actual releases. Pump curves may be used to estimate flow.

ACTION 5 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, liquid additions to this tank may continue for up to 30 days provided the tank liquid level is estimated during all liquid additions to the tank to prevent overflow.

ACTION 6 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases may continue for up to 30 days provided that samples are analyzed in accordance with ODCM Specification 1.1.2.2 and Technical Specifi-cation 4.11.1.5.

ODCM, V. C. Summer, SCE&G: Revision 27 (January 2011) 1.0-4

Table 1.1-2 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS ANALOG CHANNEL CHANNEL CHANNEL SOURCE CALIBRA- OPERATIONAL INSTRUMENT CHECK CHECK TION TEST GROSS RADIOACTIVITY MONITORS PROVIDING ALARM AND AUTOMATIC TERMINATION OF RELEASE

a. Liquid Radwaste Effluent Line - D P R(2) Q(1)

RM-L5 or RM-L9

b. Nuclear Blowdown Effluent Line D P R(2) Q(1)

RM-L7

c. Steam Generator Blowdown D M R(2) Q0(1)

Effluent Line - RM-L3, RM-L1 0

d. Turbine Building Sump Effluent D M R(2) Q(1)

Line - RM-L8

e. Condensate Demineralizer D M R(2) Q(4)

Backwash Effluent Line RM-L1 1

2. FLOW RATE MEASUREMENT DEVICES Q

a. Liquid Radwaste Effluent Line D(3) N.A. R

b. Penstocks Minimum Flow D(3) N.A. R Q Interlock

c. Nuclear Blowdown Effluent Line D(3) N.A. R Q

d. Steam Generator Blowdown Effluent Line D(3) N.A. R Q

3. TANK LEVEL INDICATING DEVICES

a. Condensate Storage Tank D N.A. R Q See Table 1.1-3 for explanation of frequency notation.

ODCM, V. C. Summer, SCE&G: Revision 24 (May 2006) 1.0-5

Table 1.1-2 (Continued)

TABLE NOTATION (1) The ANALOG CHANNEL OPERATIONAL TEST shall also demonstrate that automatic isolation of this pathway and control room alarm annunciation occurs if any of the following conditions exists:

1. Instrument indicates measured levels above the alarm/trip setpoint.

2. Loss of Power (alarm only).

3. Low Flow (alarm only).

4. Instrument indicates a Downscale Failure (alarm only).

5. Normal/Bypass switch set in Bypass (alarm only).

6. Other instrument controls not set in Operate mode.

(2) The initial CHANNEL CALIBRATION shall be performed using one or more of the reference standards certified by the National Institute of Standards and Technology (NIST) or using standards that have been obtained from suppliers that participate in measurement assurance activities with NIST. These standards shall permit calibrating the system over its intended energy and measurement range. For subsequent CHANNEL CALIBRATION, sources that have been related to the initial calibration shall be used.

(3) CHANNEL CHECK shall consist of verifying indication of flow during periods of release. CHANNEL CHECK shall be made at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months on days on which continuous, periodic or batch releases are made.

(4) The ANALOG CHANNEL OPERATIONAL TEST shall also demonstrate that automatic isolation of this pathway and local panel alarm annunciation occurs if any of the following conditions exists:

1. Instrument indicates measured levels above the alarm/trip setpoint.

2. Loss of Power (alarm only).

3. Low Flow (alarm only).

4. Instrument indicates a Downscale Failure (alarm only).

5. Normal/Bypass switch set in Bypass (alarm only).

6. Other instrument controls not set in Operate mode.

ODCM, V. C. Summer, SCE&G: Revision 21 (March 1996) 1.0-6

Table 1.1-3 FREQUENCY NOTATION Notation II Frequency D At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

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 184 days.

R At least once per 18 months.

P Completed prior to each release.

N.A. Not applicable.

Note: Each surveillance requirement shall be performed within the specified surveillance interval with a maximum allowable extension of 25% of the specified surveillance interval.

ODCM, V. C. Summer, SCE&G: Revision 24 (May 2006) 1.0-7

1.1.2 Liquid Effluents: Concentration CONTROLS 1.1.2.1 The concentration of radioactive material released from the site (see Technical Specification Figure 5.1-4) shall be limited to 10 times the concentration values specified in 10 CFR Part 20, Appendix B, Table 2, Column 2 for radionuclides other than dissolved or entrained noble gases. For dissolved or entrained noble gases, the concentration shall be limited to 2E-4 microcuries/ml total activity.

APPLICABLE: At all times.

ACTION:

a. With the concentration of radioactive material released to unrestricted areas exceeding the above limits, immediately restore the concentration to within the above limits.

b. The provisions of Technical Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 1.1.2.2 The radioactivity content of each batch of radioactive liquid waste shall be determined prior to release by sampling and analysis in accordance with Table 1.1-4. The results of pre-release analyses shall be used with the calculational methods in ODCM Section 2.1 to assure that the concentration at the point of release is maintained within the limits of ODCM Specification 1.1.2.1.

1.1.2.3 Post-release analyses of samples composited from batch releases shall be performed in accordance with Table 1.1-4. The results of the previous post-release analyses shall be used with the calculational methods in ODCM Section 2.1 to assure that the concentrations at the point of release were maintained within the limits of ODCM Specification 1.1.2.1.

ODCM, V. C. Summer, SCE&G: Revision 25 (January 2007) 1.0-8

1.1.2.4 The radioactivity concentration of liquids discharged from continuous release points shall be determined by collection and analysis of samples in accordance with Table 1.1-4. The results of the analyses shall be used with the calculational methods in ODCM Section 2.1 to assure that the concentrations at the point of release are maintained within the limits of ODCM Specification 1.1.2.1.

1.1.2.5 At least one Circulating Water Pump or the Circulating Water Jockey Pump shall be determined to be in operation and providing dilution to the discharge structure at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months whenever dilution is required to meet the site radioactive effluent concentration limits of ODCM Specification 1.1.2.1.

ODCM, V. C. Summer, SCE&G: Revision 24 (May 2006) 1.0-9

Table 1.1-4 RADIOACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM Lower Limit Minimum Type of of Detection Liquid Release Sampling Analysis Activity (LLD)

Type Frequency Frequency Analysis (pCilml)a A. Batch Waste Released P P Principal Gamma 5X10.1 Tanks Each Batch Each Batch Emittersf 1-131 1X10o5

1. Waste Monitor Tanks P M Dissolved and 1X10-One Batch/M Entrained Gases (Gamma Emitters)

2. Condensate Demin- P M H-3 lX10-5 eralizer Backwash Each Batch Compositeb Receiving Tank Gross Alpha 1X10.7

3. Nuclear Blowdown P Q Sr-89, Sr-90 5X10-"

Monitor Tank Each Patch Compositeb Fe-55 1X106 B. Continuous Releasee D W Principal Gamma 5X10-'

Grab Sample Compositec Emittersf 1-131 1X10"0

1. Steam Generator M M Dissolved and lXl0 5 Blowdown Grab Sample Entrained Gases (Gamma Emitters)

2. Turbine Building D M H-3 1X10o' Sump Grab Sample Compositec Gross Alpha lX10/

3. Service Water D Q Sr-89, Sr-90 5X10-Grab Sample Compositec Fe-55 1X 10-See Table 1.1-3 for explanation of frequency notation.

ODCM, V. C. Summer, SCE&G: Revision 24 (May 2006) 1.0-10

Table 1.1-4 (Continued)

TABLE NOTATION

a. The Lower Limit of Detection (LLD) is the smallest concentration of radioactive material in a sample that will yield a net count above background that will be detected with a 95% probability. LLD also yields a 5% probability of falsely concluding that a blank observation represents a "real" signal.

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

LLD - 2.71 + 4 .6 6 sb (E) (V) (2.22) (Y) (exp) (-AAt)

Where:

LLD is the "a priori" lower limit of detection as defined above (as pCi per unit mass or volume). Current literature defines the LLD as the detection capability for the instrumentation only and the MDC, the minimum detectable concentration, as the detection capability for a given instrument procedure and type of sample.

4.66 is a factor which corrects for the smallest activity that has a probability, p, of being detected, and a probability, 1-p, of falsely concluding its presence.

2.71 may be used to account for minor deviations of the Poisson distribution from Normality.

4.66 = 2k F1 + (tb /ts) k = a constant whose value depends on the chosen confidence level (NRC recommends a confidence level of 95%)

= 1.6545 at 95% confidence level tb = background time ts = sample time sb is the standard deviation of the background counting rate or the counting rate of blank sample as appropriate (as counts per minute).

E is the counting efficiency (as counts per transformation).

ODCM, V. C. Summer, SCE&G: Revision 29 (August 2013) 1.0-11

Table 1.1-4 (Continued)

TABLE NOTATION V is the sample size (in units of mass or volume).

2.22 is the number of transformations per minute per picoCurie.

Y is the fractional radiochemical yield (when applicable).

k is the radioactive decay constant for the particular radionuclide.

At is the elapsed time between midpoint of sample collection and time of counting (for plant effluents, not environmental samples).

The value of s b used in the calculation of the LLD for a detection system shall be used on the actual observed variance of the back-ground counting rate or of the counting rate of the blank samples (as appropriate) rather than on an unverified theoretically predicted variance. In calculating the LLD for a radionuclide determined by gamma-ray spectrometry the background should include the typical contributions of other radionuclides normally present in the samples.

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

It should be recognized that the LLD is defined as an a Driori (before the fact) limit representing the capability of a measurement system and not as a posteriori (after the fact) limit for particular measurement.*

For a more complete discussion of the LLD, and other detection limits, see the following:

(1) HASL Procedures Manual, HASL-300 (revised annually).

(2) Currie, L. A., "Limits for Qualitative Detection and Quantitative Determination -

Application to Radiochemistry" Anal. Chem. 40, 586-93 (1968).

(3) Hartwell, J. K., "Detection Limits for Radioisotopic Counting Techniques," Atlantic Richfield Handford Company Report ARH-2537 (June 22, 1972).

ODCM, V. C. Summer, SCE&G: Revision 24 (May 2006) 1.0-12

Table 1.1-4 (Continued)

TABLE NOTATION

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. To be representative of the quantities and concentrations of radioactive materials in liquid effluents, samples shall be composited in proportion to the rate of flow of the effluent stream. 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.

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, by a method described in ODCM Section 2.0, to assure representative sampling.

e. A continuous release is the discharge of liquid wastes of a nondiscrete volume; e.g., from a volume of system that has an input flow during the continuous 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, Mo-99, Cs-134, Cs-137, and Ce-141. Ce-144 shall also be measured, but with an LLD of 5 X 10 -6. 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.

ODCM, V. C. Summer, SCE&G: Revision 25 (January 2007) 1.0-13

1.1.3 Liquid Effluents: Dose CONTROLS 1.1.3.1 The dose or dose commitment to an individual from radioactive materials in liquid effluents released from the site (see Technical Specification Figure 5.1-4) shall be limited:

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

APPLICABLE: At all times.

ACTION:

a. With the calculated dose from the release of radioactive materials in liquid effluents exceeding any of the above limits, in lieu of any other report required by ODCM Section 1.6, prepare and submit to the Commission within 30 days, pursuant to Technical Specification 6.9.2, a Special Report which identifies the cause (s) for exceeding the limit (s) and defines the corrective actions to be taken to the releases and the proposed actions to be taken to assure that subsequent releases will be in compliance with ODCM Specification 1.1.3.1.

b. The provisions of Technical Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 1.1.3.2 Dose Calculations Cumulative dose contributions from liquid effluents shall be determined in accordance with ODCM Section 2.2 at least once per 31 days.

ODCM, V. C. Summer, SCE&G: Revision 24 (May 2006) 1.0-14

1.1.4 Liauid Waste Treatment CONTROLS 1.1.4.1 The liquid radwaste treatment system shall be OPERABLE. The appropriate portions of the system shall be used to reduce the radioactive materials in liquid wastes prior to their discharge when the projected doses due to the liquid effluent from the site (See Technical Specification Figure 5.1-4) when averaged on a sliding 31 day calendar basis, would exceed 0.06 mrem to the total body or 0.2 mrem to any organ.

APPLICABLE: At all times.

ACTION:

a. With the liquid radwaste treatment system inoperable for more than 31 days or with radioactive liquid waste being discharged without treatment and in excess of the above limits, in lieu of any other report required by ODCM Section 1.6, prepare and submit to the Commission within 30 days, pursuant to Technical Specification 6.9.2, a Special Report which includes the following information:

1. Identification of the inoperable equipment or subsystems and the reason for inoperability.

2. Action(s) taken to restore the inoperable equipment to OPERABLE status.

3. Summary description of action(s) taken to prevent a recurrence.

b. With radioactive waste being discharged, the requirements to process effluents are:

1. If all streams are unprocessed and projected dose(s) exceed the limits of ODCM Specification 1.1.4.1, process the appropriate streams to the point that the projected dose is within limits.

ODCM, V. C. Summer, SCE&G: Revision 24 (May 2006) 1.0-15

2. With a combination of processed and unprocessed streams producing a dose projection exceeding the limits of specification 1.1.4.1, process the unprocessed streams if they contribute greater than or equal to 10 percent of Specification 1.1.4.1 limits.

c. The provisions of Technical Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 1.1.4.2 Doses due to liquid releases shall be projected at least once per 31 days.

1.1.4.3 The liquid radwaste treatment system shall be demonstrated OPERABLE by operating the liquid radwaste treatment system equipment for at least 30 minutes at least once per 92 days unless the liquid radwaste system has been utilized to process radioactive liquid effluents during the previous 92 days.

ODCM, V. C. Summer, SCE&G: Revision 24 (May 2006) 1.0-16

1.2 GASEOUS EFFLUENTS 1.2.1 Radioactive Gaseous Effluent Monitoring Instrumentation CONTROLS 1.2.1.1 The radioactive gaseous effluent monitoring instrumentation channels shown in Table 1.2.-1 shall be OPERABLE with their alarm/trip setpoints set to ensure that the limits of ODCM Specification 1.2.2.1 are not exceeded. The alarm/trip setpoints of these channels shall be determined in accordance with ODCM Section 3.1.

APPLICABLE: As shown in Table 1.2-1 ACTION:

a. With a radioactive gaseous effluent monitoring instrumentation channel alarm/trip setpoint less conservative than required by the above ODCM Specification, immediately suspend the release of radioactive gaseous effluents monitored by the affected channel or declare the channel inoperable.

b. With less than the minimum number of radioactive gaseous effluent monitoring instrumentation channels OPERABLE, take the ACTION shown in Table 1.2-1. Additionally if this condition prevails for more than 30 days, in the next Annual Radioactive Effluent Release Report, explain why this condition was not corrected in a timely manner.

c. The provisions of Technical Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 1.2.1.2 Each radioactive gaseous effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION and an ANALOG CHANNEL OPERATIONAL TEST operations at the frequencies shown in Table 1.2-2.

ODCM, V. C. Summer, SCE&G: Revision 24 (May 2006) 1.0-17

Table 1.2-1 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION MINIMUM CHANNELS APPLICA-INSTRUMENT OPERABLE BILITY ACTION

1. WASTE GAS HOLDUP SYSTEM

a. Noble Gas Activity Monitor - Providing 1 7 Alarm and Automatic Termination of Release (RM-A10 or RM-A3)

2. MAIN PLANT VENT EXHAUST SYSTEM

a. Noble Gas Activity Monitor - Providing 9 Alarm and Automatic Termination of Release from Waste Gas Holdup System (RM-A3)

b. Iodine Sampler 1 11

c. Particulate Sampler 1 11

d. Flow Rate Measuring Device 1 8

e. Sampler Flow Rate Measuring Device 1 8

3. REACTOR BUILDING PURGE SYSTEM

a. Noble Gas Activity Monitor Providing 10 Alarm & Automatic Termination of 1 Release (RM-A4)

b. Iodine Sampler 11

c. Particulate Sampler 1 11

d. Flow Rate Measuring Device

1. For 36" Purge (IFT09287) 1 8

2. For 6" Purge (IFT08252) 1 8

e. Sampler Flow Rate Measuring Device 1 8 ODCM, V. C. Summer, SCE&G: Revision 24 (May 2006) 1.0-18

Table 1.2-1 (Continued)

TABLE NOTATION

At all times during releases via this pathway.

ACTION 7 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, the contents of the tank(s) may be released to the environment provided that prior to initiating the release:

a. At least two independent samples of the tank's contents are analyzed.

b. At least two technically qualified members of the Facility Staff independently verify the release rate calculations and discharge valve lineup.

Otherwise, suspend release of radioactive effluents via this pathway.

ACTION 8 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided the flow rate is estimated at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

ACTION 9 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided grab samples are taken at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and these samples are analyzed for gross activity within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

ACTION 10 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, immediately suspend PURGING of radioactive effluents via this pathway.

ACTION 11 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via the affected pathway may continue provided samples, as specified in Table 1.2-3, are continuously collected with auxiliary sampling equipment.

a. With the monitor taken out of service, by a condition other than a planned action, the action statement is considered met if a conscious, concerted and continuous effort is being made to initiate the collection of the required sample(s) with auxiliary sampling equipment.

b. A planned removal of the monitor from service requires that the auxiliary sampling equipment be staged in the area to reduce the amount of time for the change over from sampling by the installed monitor to the auxiliary sampling equipment.

ODCM, V. C. Summer, SCE&G: Revision 27 (January 2011) 1.0-19

Table 1.2-2 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS ANALOG MODES INWHICH CHANNEL CHANNEL SURVEILLANCE CHANNEL SOURCE CALIBRATION OPERATIONAL REQUIRED INSTRUMENT CHECK CHECK TEST

1. WASTE GAS HOLDUP SYSTEM

a. Noble Gas Activity P P R(3) Q(1)

Monitor - RM-A10 or RM-A3

2. MAIN PLANT VENT EXHAUST SYSTEM *

a. Noble Gas Activity D M R(3) Q(2)

Monitor - RM-A3

b. Iodine Sampler W N.A. N.A. N.A.

c. Particulate W N.A. N.A. N.A.

Sampler

d. Flow Rate D N.A. R Q Measuring Device

e. Sampler Flow Rate D N.A. R Q Monitor

3. REACTOR BUILDING PURGE SYSTEM

a. Noble Gas Activity D P,M R(3) Q(1)

Monitor - RM-A4

b. Iodine Sampler W N.A. N.A. N.A.

c. Particulate W N.A. N.A. N.A.

Sampler

d. Flow Rate Measuring Device D N.A. R Q

1. For 36" Purge (IFT09287) D N.A. R Q

2. For 6" Purge (IFT08252)

D N.A. R Q

e. Sampler Flow Rate Monitor See Table 1.1-3 for explanation of frequency notation.

ODCM, V. C. Summer, SCE&G: Revision 24 (May 2006) 1.0-20

Table 1.2-2 (Continued)

TABLE NOTATION At all times.

(1) The ANALOG CHANNEL OPERATIONAL TEST shall also demonstrate that automatic isolation of this pathway and control room alarm annunciation occurs if any of the following conditions exists:

1. Instrument indicates measured levels above the alarm/trip setpoint.

2. Loss of Power (alarm only).

3. Low Flow (alarm only).

4. Instrument indicates a Downscale Failure (alarm only).

5. Normal/Bypass switch set in Bypass (alarm only).

6. Other instrument controls not set in Operate mode.

(2) The ANALOG CHANNEL OPERATIONAL TEST shall also demonstrate that control room alarm annunciation occurs if any of the following conditions exists:

1. Instrument indicates measured levels above the alarm setpoint.

2. Loss of Power.

3. Low Flow.

4. Instrument indicates a Downscale Failure.

5. Instrument controls not set in Operate mode.

(3) The initial CHANNEL CALIBRATION shall be performed using one or more of the reference standards certified by the National Institute of Standards and Technology (NIST) or using standards that have been obtained from suppliers that participate in measurement assurance activities with NIST. These standards shall permit calibrating the system over its intended range of energy and measurement.

For subsequent CHANNEL CALIBRATION, sources that have been related to the initial calibration shall be used.

ODCM, V. C. Summer, SCE&G: Revision 24 (May 2006) 1.0-21

1.2.2 Gaseous Effluents: Dose Rate CONTROLS 1.2.2.1 The dose rate in unrestricted areas due to radioactive materials released in gaseous effluents from the site ( see Technical Specification Figure 5.1-3) shall be limited to the following:

a. For noble gases: Less than or equal to 500 mrem/yr to the total body and less than or equal to 3000 mrem/yr to the skin.

b. For Iodine-1 31, Iodine-1 33 and for all radioactive materials in particulate form and tritium with half lives greater than 8 days: Less than or equal to 1500 mrem/yr to any organ.

APPLICABLE: At all times.

ACTION:

With the dose rate(s) exceeding the above limits, immediately decrease the release rate to within the above limit(s).

SURVEILLANCE REQUIREMENTS 1.2.2.2 The dose rate due to noble gases in gaseous effluents shall be determined to be within the above limits in accordance with the methods and procedures of the ODCM.

1.2.2.3 The dose rate due to radioiodines, tritium and radioactive materials in particulate form with half lives greater than 8 days in gaseous effluents shall be determined to be within the above limits in accordance with the methods and procedures of ODCM Section 3.2.2 by obtaining representative samples and performing analyses in accordance with the sampling and analysis program specified in Table 1.2-3.

ODCM, V. C. Summer, SCE&G: Revision 29 (August 2013) 1.0-22

Table 1.2-3 RADIOACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM Lower Limit of Minimum Type of Detection Gaseous Release Sampling Analysis Activity (LLD)

Type Frequency Frequency Analysis (ILCi/mI)a A. Waste Gas Storage P P Principal Gamma 1X10A Tank Each Tank Each Tank Emitters9 Grab Sample B1 Reactor Building P P Principal Gamma 1X1O0-9

- 36" Purge Line Each Purge Each Purge Emitters H-3 1X1 0-

- 6" Purge Line B2 Reactor Building Mb Mb Principal Gamma 1X10-4

- 6" Purge Line Grab Sample Emittersg (if continuous) H-3 1X10-C Main Plant Vent Mbe Mb Principal Gamma 1X10-4 Grab Sample Emitters 9 H-3 1X10,5 D1. Reactor Building Continuous Wd 1-131 1X10 1102 Purge Samplerý Charcoal Sample 1-133 1X10 11

2. Main Plant Vent Continuous WO Principal Gamma 1X10 Samplerf Particulate Emitters9 Sample 1-131, others Continuous M Samplerý Composite Parti- Gross Alpha 1X10-11 culate Sample Continuous Q 11 Sampler1 Composite Parti- Sr-89, Sr-90 1X10 culate Sample Continuous Noble Gas Noble Gases 1X106 Monitor Monitor Gross Beta See Table 1.1-3 for explanation of frequency notation.

ODCM, V. C. Summer, SCE&G: Revision 29 (August 2013) 1.0-23

Table 1.2-3 (Continued)

TABLE NOTATION

a. See Table 1.1-4 notation (a) for definition of LLD.

b. Analyses shall be also be performed within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following shutdown, startup, or a THERMAL POWER change exceeding 15 percent of the RATED THERMAL POWER within a one hour period.

c. Tritium grab samples shall be taken at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when the refueling canal is flooded.

d. Samples shall be changed at least once per 7 days and analyses shall be completed within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months after changing (or after removal from sampler).

Sampling shall also be performed at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for at least 7 days following each shutdown, startup or THERMAL POWER change exceeding 15 percent of RATED THERMAL POWER in one hour and analyses shall be completed within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months of changing. When samples collected for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months are analyzed, the corresponding LLD's may be increased by a factor of 10.

e. Tritium grab samples shall be taken at least once per 7 days from the ventilation exhaust from the spent fuel pool area, whenever spent fuel is in the spent fuel pool.

f. 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 ODCM Specifications 1.2.2.1, 1.2.3.1 and 1.2.4.1.

g. The principal gamma emitters for which the LLD specification applies exclusively are the following radionuclides: Kr-87, Kr-88, Xe-1 33, Xe-1 33m, Xe-1 35 and Xe-138 for gaseous emissions and Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, Cs-134, Cs-1 37, Ce-141 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.

h. Prior to sampling for analysis, each batch of oil shall be isolated and representative samples obtained by methods described in ASTM D 4057-81, Volume 05.03, "Standard Practice for Manual Sampling of Petroleum and Petroleum Products".

This LLD refers to the liquid sample.

ODCM, V. C. Summer, SCE&G: Revision 24 (May 2006) 1.0-24

1.2.3 Gaseous Effluents: Dose - Noble Gas CONTROLS 1.2.3.1 The air dose due to noble gases released in gaseous effluents from the site (see Technical Specification Figure 5.1-3) shall be limited to the following:

a. During any calendar quarter: Less than or equal to 5 mrad for gamma radiation and less than or equal to 10 mrad for beta radiation.

b. During any calendar year: Less than or equal to 10 mrad for gamma radiation and less than or equal to 20 mrad for beta radiation.

APPLICABLE: At all times.

ACTION:

a. With the calculated air dose from radioactive noble gases in gaseous effluents exceeding any of the above limits, in lieu of any other report required by ODCM Section 1.6, prepare and submit to the Commission within 30 days, pursuant to Technical Specification 6.9.2, a Special Report which identifies the cause(s) for exceeding the limit(s) and defines the corrective actions to be taken to releases and the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with ODCM Specification 1.2.3.1.

b. The provisions of Technical Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 1.2.3.2 Dose Calculations Cumulative dose contributions for the current calendar quarter and current calendar year shall be determined in accordance with ODCM Section 3.2.3 at least once per 31 days.

ODCM, V. C. Summer, SCE&G: Revision 24 (May 2006) 1.0-25

1.2.4 Gaseous Effluents: Dose - Radioiodines, Tritium, and Radioactive Materials in Particulate Form.

CONTROLS 1.2.4.1 The dose to an individual from radioiodines, tritium, and radioactive materials in particulate form, and radionuclides (other than noble gases) with half-lives greater than 8 days in gaseous effluents (see Technical Specification Figure 5.1-3) shall be limited to the following:

a. During any calendar quarter: Less than or equal to 7.5 mrem to any organ.

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

APPLICABLE: At all times.

ACTION:

a. With the calculated dose from the release of tritium, radioiodines, and radioactive materials in particulate form with half lives greater than 8 days in gaseous effluents exceeding any of the above limits, in lieu of any other report required by ODCM Section 1.6, prepare and submit to the Commission within 30 days, pursuant to Technical Specification 6.9.2, a Special Report which identifies the cause(s) for exceeding the limit and defines the corrective actions to be taken to releases and the proposed actions to be taken to assure that subsequent release will be in compliance with ODCM Specification 1.2.4.1.

b. The provisions of Technical Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 1.2.4.2 Dose Calculations Cumulative dose contributions for the current calendar quarter and current calendar year shall be determined in accordance with ODCM Section 3.2.3 at least once per 31 days.

ODCM, V. C. Summer, SCE&G: Revision 29 (August 2013) 1.0-26

1.2.5 Gaseous Effluents: Gaseous Radwaste Treatment CONTROLS 1.2.5.1 The GASEOUS RADWASTE TREATMENT SYSTEM and the VENTILATION EXHAUST TREATMENT SYSTEM shall be OPERABLE. The appropriate portions of the GASEOUS RADWASTE TREATMENT SYSTEM shall be used to reduce radioactive materials in gaseous waste prior to their discharge when the projected gaseous effluent air doses due to gaseous effluent releases from the site (See Technical Specification Figure 5.1-3), when averaged over 31 days, would exceed 0.2 mrad for gamma radiation and 0.4 mrad for beta radiation.

The appropriate portions of the VENTILATION EXHAUST TREATMENT SYSTEM shall be used to reduce radioactive materials in gaseous waste prior to their discharge when the projected doses due to gaseous effluent releases from the site when averaged over 31 days would exceed 0.3 mrem to any organ.

APPLICABLE: At all times*.

ACTION:

a. With the GASEOUS RADWASTE TREATMENT SYSTEM and/or the VENTILATION EXHAUST TREATMENT SYSTEM inoperable for more than 31 days or with gaseous waste being discharged without treatment and in excess of the above limits, in lieu of any other report required by ODCM Section 1.6, prepare and submit to the Commission within 30 days, pursuant to Technical Specification 6.9.2, a Special Report which includes the following information:

1. Identification of the inoperable equipment or subsystems and the reason for inoperability.

2. Action(s) taken to restore the inoperable equipment to OPERABLE status.

The Waste Gas System may be secured during refueling and defueled operations since there is no gas in the system to be removed and processed. The system is considered "inoperable" during these conditions due to the instrumentation being out of calibration when flow is stopped through the recombiner. This "inoperable" state is the normal system condition during refueling and defueled modes.

ODCM, V. C. Summer, SCE&G: Revision 24 (May 2006) 1.0-27

3. Summary description of action(s) taken to prevent a recurrence.

b. The provisions of Technical Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 1.2.5.2 Doses due to gaseous releases from the reactor shall be projected at least once per 31 days.

1.2.5.3 The GASEOUS RADWASTE TREATMENT SYSTEM and VENTILATION EXHAUST TREATMENT SYSTEM shall be demonstrated OPERABLE by opera-ting the GASEOUS RADWASTE TREATMENT SYSTEM equipment and VENTILATION EXHAUST TREATMENT SYSTEM equipment for at least 30 minutes, at least once per 92 days unless the appropriate system has been utilized to process radioactive gaseous effluents during the previous 92 days.

ODCM, V. C. Summer, SCE&G: Revision 24 (May 2006) 1.0-28

1.3 RADIOACTIVE EFFLUENTS: TOTAL DOSE CONTROLS 1.3.1 The 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) over 12 consecutive months.

APPLICABLE: 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 ODCM Specification 1.1.3.1.a, 1.1.3.1.b, 1.2.3.1.a, 1.2.3.1.b, 1.2.4.1.a or 1.2.4.1.b, in lieu of any other report required and ODCM Section 1.6, prepare and submit to the Commission, within 30 days, pursuant to Technical Specification 6.9.2, a Special Report which defines the corrective action to be taken to reduce subsequent releases to prevent recurrence of exceeding the limits of ODCM Specification 1.3.1. This Special Report, defined in 10 CFR 20.2203(a)(4), shall include an analysis which estimates the radiation exposure (dose) to a member of the public from uranium fuel cycle sources (including all effluent pathways and direct radiation) for a 12 consecutive month period that includes the release(s) covered by this report. The report 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 limits of ODCM Specification 1.3.1, and if the release condition resulting in violation of 40 CFR 190 has not already been corrected, the Special Report shall include a request for a variance in accordance with the provisions of 40 CFR 190 and including information of § 190.11 (b). Submittal of the report is considered a timely request, and a variance is granted until staff action on the request is complete. The variance only relates to the limits of 40 CFR 190, and does not apply in any way to the requirements for dose limitation of 10 CFR Part 20, as addressed in ODCM Specifications 1.1.2 and 1.2.2.

b. The provisions of Technical Specifications 3.0.3 and 3.0.4 are not applicable.

ODCM, V. C. Summer, SCE&G: Revision 25 (January 2007) 1.0-29

SURVEILLANCE REQUIREMENTS 1.3.2 Dose Calculations Cumulative dose contributions from liquid and gaseous effluents shall be determined in accordance with ODCM Specifications 1.1.3.2, 1.2.3.2 and 1.2.4.2.

ODCM, V. C. Summer, SCE&G: Revision 25 (January 2007) 1.0-30

1.4 RADIOLOGICAL ENVIRONMENTAL MONITORING 1.4.1 Monitoring Pro-gram CONTROLS 1.4.1.1 The radiological environmental monitoring program shall be conducted as specified in Table 1.4-1.

APPLICABILITY: At all times.

ACTION:

a. With the radiological environmental monitoring program not being conducted as specified in Table 1.4-1 in lieu of any other report required by ODCM Section 1.6, 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 in an environmental sampling medium exceeding the reporting levels of Table 1.4-2 when averaged over any calendar quarter, in lieu of any other report required by ODCM Section 1.6, prepare and submit to the Commission within 30 days from the end of the affected calendar quarter a Special Report. When more than one of the radionuclides in Table 1.4-2 are detected in the sampling medium, this report shall be submitted if:

Concentration(1) Concentration (2) .. 1.0 Limit Level (1) Limit Level (2)

When radionuclides other than those in Table 1.4-2 are detected and are the result of plant effluents, this report shall be submitted if the potential annual dose to an individual is equal to or greater than the calendar year limits of ODCM Specifications 1.1.3.1, 1.2.3.1 and 1.2.4.1. 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.

ODCM, V. C. Summer, SCE&G: Revision 23 (September 1999) 1.0-31

c. With milk or fresh leafy vegetable samples permanently unavailable from one or more of the sample locations required by Table 1.4-1, in lieu of any other report required by ODCM Section 1.6 prepare and submit to the Commission within 30 days, pursuant to Technical Specification 6.9.2, a Special Report which identifies the cause of the unavailability of samples and identifies locations for obtaining replacement samples. The locations from which samples were unavailable may then be deleted from those required by Table 1.4-1, provided the locations from which the replacement samples were obtained are added to the environmental monitoring program as replacement locations.

d. The provisions of Technical Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 1.4.1.2 The radiological environmental monitoring samples shall be collected pursuant to Table 1.4-1 and shall be analyzed pursuant to the requirements of Tables 1.4-1 and 1.4-3.

ODCM, V. C. Summer, SCE&G: Revision 23 (September1999) 1.0-32

Table 1.4-1 Radiological Environmental Monitoring Program Virgil C. Summer Nuclear Station Exposure Path-way and/or Minimum Number of Sample Locations and Sampling and Type & Frequency Sample Criteria for Selection Collection Frequency of Analysis AIRBORNE:

I. Particulates A) 3 Indicator samples from locations close to the Continuous sampler operation Gross beta following filter site boundary, in different sectors, of the with weekly collection, change; quarterly 6 highest calculated annual average ground level composite (by location) for D/Q or dose. gamma isotopic.

B) 1 Indicator sample to be taken close to the site Continuous sampler operation Gross beta following filter boundary in the sector corresponding to the with weekly collection. change; quarterly 6 residence having the highest anticipated offsite composite (by location) for ground level concentration or dose. gamma isotopic.

C) 1 Indicator sample to be taken at the location of Continuous sampler operation Gross beta following filter one of the dairies2 being sampled meeting the with weekly collection, change; quarterly 6 criteria of VII(A). composite (by location) for gamma isotopic.

D) 1 Control sample to be taken at a location at Continuous sampler operation Gross beta following filter least 10 air miles from the site and not in the with weekly collection. change; quarterly 6 most prevalent wind directions. composite (by location) for gamma isotopic.

I1. Radioiodine A) 3 Indicator samples to be taken at two Continuous sampler operation Gamma isotopic for 1-131 locations as given in I(A) above, with weekly canister collection, weekly.

B) 1 Indicator sample to be taken at the location Continuous sampler operation Gamma isotopic for 1-131 as given in I(B) above, with weekly canister collection. weekly.

C) 1 Indicator sample to be taken at the location Continuous sampler operation Gamma isotopic for 1-131 as given in I(C) above, with weekly canister collection, weekly.

D) 1 Control sample to be taken at a location as Continuous sampler operation Gamma isotopic for 1-131 given in I(D) above, with weekly canister collection. weekly.

Ill. Direct A) 13 Indicator stations with two or more Monthly 5 or quarterly 6. Gamma dose monthly 5 or dosimeters to form an inner dng of stations in quarterdy 6 the 13 accessible sectors within I to 2 miles of the plant.

B) 16 Indicator stations with two or more Monthly 5 or quarterly 6. Gamma dose monthly 5 or dosimeters to form an outer ring of stations in quarterly 6 the 16 accessible sectors within 3 to 5 miles of the plant.

C) 11 Stations with two or more dosimeters to be Monthly 5 or quarterly . Gamma dose monthly 5 or placed in special interest areas such as quarterly 6 population centers, nearby residences, schools and in 4 or 5 areas to serve as control stations.

ODCM, V. C. Summer, SCE&G: Revision 26 (September 2007) 1.0-33

Table 1.4-1 Radiological Environmental Monitoring Program Virgil C. Summer Nuclear Station Exposure Path-way and/or Minimum Number of Sample Locations and Sampling and Type & Frequency Sample Criteria for Selection Collection Frequency of Analysis WATERBORNE: A) 1 Indicator sample downstream to be taken at a Time composite samples with Gamma isotopic monthly 5 IV. Surface Water location which allows for mixing and dilution in collection every month, 5 with quarterly 6 composite the ultimate receiving river. (by location) or monthly 5 sample to be analyzed for tritium.

B) 1 Control sample to be taken at a location on Time composite samples with Gamma isotopic monthly 5 the receiving river sufficiently far upstream such collection every month. 5 with quarterly 6 composite that no effects of pumped storage operation are (by location) or monthly 5 anticipated, sample to be analyzed for tritium.

C) 1 Indicator sample to be taken in the upper Time composite samples with Gamma isotopic monthly 5 reservoir of the pumped storage facility in the collection every month. 5 with quarterly 6 composite plant discharge canal. (by location) or monthly 5 sample to be analyzed for tritium.

V. Ground Water A) 12 Indicator samples to be taken within the Quarterly 6 grab sampling. Gamma isotopic and exclusion boundary and in the direction of tritium analyses quarterly 6 potentially affected ground water supplies.

B) 1 Control sample from unaffected location. Quarterly 6 grab sampling. Gamma isotopic and tritium analyses quarterly 6 VI. Drinking Water A) 1 Indicator sample from a nearby public ground Monthly 5 grab sampling. Monthly 5gamma isotopic water supply source. and gross beta analyses and quarterly 6 composite for tritium analyses.

B) 1 Indicator (finished water) sample from the Monthly 5 composite sampling. Monthly 5 gamma isotopic nearest downstream water supply. and gross beta analyses and quarterly 6 composite for tritium analyses.

C) 1 Control (finish water) sample from the nearest Monthly 5 composite sampling. Monthly 5 gamma isotopic unaffected public water supply. and gross beta analyses and quarterly 6 composite for tritium analyses.

ODCM, V. C. Summer, SCE&G: Revision 29 (August 2013) 1.0-34

Table 1.4-1 Radiological Environmental Monitoring Program Virgil C. Summer Nuclear Station Exposure Path-way and/or Minimum Number of Sample Locations and Sampling and Type & Frequency Sample Criteria for Selection Collection Frequency of Analysis INGESTION: A) Samples from milking animals in 3 locations with- Semimonthly 4 when animals Gamma isotopic and VII. Milk 2 in 5 km distance having the highest dose are on pasture, monthly 5 other 1-131 analysis potential. Ifthere are none then 1 sample from times. semimonthly 4 when milking animals in each of 3 areas between 5 to 8 animals are on pasture; km distance where doses are calculated to be monthly 5 at other times.

greater than 1 mrem per year 1.

B) 1 Control sample to be taken at the location of a Semimonthly 4 when animals Gamma isotopic and dairy greater than 20 miles distance and not in are on pasture, monthly 5 other 1-131 analysis the most prevalent wind direction, times . semimonthly 4 when animals are on pasture; monthly 5 at other times.

C) 1 Indicator grass (forage) sample to be taken at Monthly 5 when available. Gamma isotopic.

the location of one of the dairies being sampled meeting the criteria of VII(A), above, when animals are on pasture.

when available .8 5

D) 1 Control grass (forage) sample to be taken at the Monthly Gamma isotopic.

location of VII(B) above.

VIII. Food Products A) 2 samples of broadleaf vegetation grown in the 2 Monthly 5 when available. Gamma Isotopic on nearest offsite locations of highest calculated edible portion.

annual average ground level D/Q if milk sampling is not performed within 3 km or if milk sampling is not performed at a location within 5 to 8 km where the doses are calculated to be greater than 1 mrem/yr 1.

Gamma Isotopic on B) 1 Control sample for the same foods taken at a Monthly when available, edible portion.

location at least 10 miles distance and not in the most prevalent wind direction if milk sampling is not performed within 3 km or if milk sampling is not at a location within 5 to 8 km where doses are calculated to be greater than 1 mrem/yr 1.

IX. Fish A) 1 Indicator sample to be taken at a location in the Semiannual 7 collection of the Gamma isotopic on upper reservoir, following specie types if edible portions available: bass; bream, semiannually 7 crappie; catfish, carp.

B) 1 Indicator sample to be taken at a location in the Semiannual 7 collection of the Gamma isotopic on lower reservoir, following specie types if edible portions available: bass; bream, semiannually 7.

crappie; catfish, carp.

C) 1 Control sample to be taken at a location on the Semiannual 7 collection of the Gamma isotopic on receiving river sufficiently far upstream such that following specie types if edible portions no effects of pumped storage operation are available: bass; bream, semiannually 7.

anticipated, crappie; catfish, carp.

ODCM, V. C. Summer, SCE&G: Revision 23 (September 1999) 1.0-35

Table 1.4-1 Radiological Environmental Monitoring Program Virgil C. Summer Nuclear Station Exposure Path-way and/or Sample Minimum Number of Sample Locations Sampling and Type & Frequency and Criteria for Selection Collection Frequency of Analysis AQUATIC: A) I Indicator sample to be taken at a location in Semiannual 7 grab sample. Gamma isotopic.

X. Sediment the upper reservoir.

B) 1 Indicator sample to be taken on or near the Semiannual 7 grab sample. Gamma isotopic.

shoreline of the lower reservoir.

C) 1 Control sample to be taken at a location on Semiannual 7 grab sample. Gamma isotopic.

the receiving river sufficiently far upstream such that no effects of pumped storage operation are anticipated.

ODCM, V. C. Summer, SCE&G: Revision 24 (May 2006) 1.0-36

Table 1.4-1 (Continued)

TABLE NOTATION

1. The dose shall be calculated for the maximum organ and age group, using the guidance/methodology contained in Regulatory Guide 1.109, Revision 1 and the parameters particular to the Site. The locations are selected based on potential for highest exposure.

2. Milking animal and garden survey results will be analyzed annually. Should the survey indicate new dairying activity, the owners shall be contacted with regard to a contract for supplying sufficient samples. If contractual arrangements can be made, Site(s) will be added for additional milk sampling up to a total of 3 Indicator Locations.

3. Time composite samples are samples which are collected with equipment capable of collecting an aliquot at time intervals which are short (e.g., hourly) relative to the compositing period.

4. At least once per 18 days.

5. Not to exceed 35 days.

6. At least once per 100 days

7. At least once per 200 days.

8. Milk and grass (forage) sampling at the control location is only required when locations meeting the criteria of VII(A) are being sampled.

NOTE: Deviations from this sampling schedule may occasionally be necessary if sample media are unobtainable due to hazardous conditions, seasonal unavailability, insufficient sample size, malfunctions of automatic sampling or analysis equipment and other legitimate reasons. If specimens are unobtainable due to sampling equipment malfunction, every effort shall be made to complete corrective action prior to the end of the next sampling period. Deviations from sampling analysis schedules will be described in the annual report.

ODCM, V. C. Summer, SCE&G: Revision 23 (September 1999) 1.0-37

Table 1.4-2 Reporting Levels for Radioactivity Concentrations in Environmental Samples Airborne Fish Food Water Particulate (pCi/Kg, Milk Products Analysis (pCi/I) or Gases wet) (pCill) (pCi/Kg, wet)

(pCi/m 3 )

H-3 20,000(a) N.A. N.A. N.A. N.A.

Mn-54 1,000 N.A. 30,000 N.A. N.A.

Fe-59 400 N.A. 10,000 N.A. N.A.

Co-58 1,000 N.A. 30,000 N.A. N.A.

Co-60 300 N.A. 10,000 N.A. N.A.

Zn-65 300 N.A. 10,000 N.A. N.A.

Zr-95 400 N.A. 20,000 N.A. N.A.

Nb-95 400 N.A. 20,000 N.A. N.A.

1-131 2 0.9 N.A. 3 100 Cs-134 30 10 1,000 60 1,000 Cs-1 37 50 20 2,000 70 2,000 Ba-140 200 N.A. N.A. 300 N.A.

La-140 200 N.A. N.A. 300 N.A.

(a) For drinking water samples. This is the 40 CFR Part 141 value.

ODCM, V. C. Summer, SCE&G: Revision 24 (May 2006) 1.0-38

Table 1.4-3 Maximum Values for the Lower Limits of Detection (LLD)a'c Airborne Food Particulate Fish Products Sediment Water or Gases (pCi/Kg, Milk (pCi/Kg, (pCi/Kg, dry)

Analysis (pCi/I) (pCi/m3 ) wet) (pCi/l) wet)

Gross 4 1 X 1102 N.A. N.A. N.A. N.A.

Beta H-3 2000(b) N.A. N.A. N.A. N.A. N.A.

Mn-54 15 N.A. 130 N.A. N.A. N.A.

Fe-59 30 N.A. 260 N.A. N.A. N.A.

Co-58 15 N.A. 130 N.A. N.A. N.A.

Co-60 15 N.A. 130 N.A. N.A. N.A.

Zn-65 30 N.A. 260 N.A. N.A. N.A.

Zr-95 30 N.A. N.A. N.A. N.A. N.A.

Nb-95 15 N.A. N.A. N.A. N.A. N.A.

1-131 1b 7 X 10-2 N.A. 1 60 N.A Cs-134 15 5 X 10-2 130 15 60 150 Cs-137 18 6 X 10-2 150 18 80 180 Ba-140 60 N.A. N.A. 60 N.A. N.A.

La-140 15 N.A. N.A. 15 N.A. N.A.

ODCM, V. C. Summer, SCE&G: Revision 24 (May 2006) 1.0-39

Table 1.4-3 (Continued)

TABLE NOTATION

a. Table 1.4-3 lists detection capabilities for radioactive materials in environmental samples. These detection capabilities are tabulated in terms of the lower limits of detection (LLDs). See Table 1.1-4 notation (a) for definition of LLD.

b. LLD for drinking water samples.

c. Other peaks potentially due to reactor operations (fission and activation products) which are measurable and identifiable, together with the radionuclides in Table 1.4-3, shall be identified and reported.

ODCM, V. C. Summer, SCE&G: Revision 21 (March 1996) 1.0-40

1.4.2 Land Use Census CONTROLS 1.4.2.1 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 16 meteorological sectors within a distance of five miles.

APPLICABILITY: At all times.

ACTION:

a. With a land use census identifying a location(s) which yields a calculated dose or dose commitment greater than the values currently being calculated in ODCM Specification 1.2.4.2, in lieu of any other report required by ODCM Section 1.6, prepare and submit to the Commission within 30 days, pursuant to Technical Specification 6.9.2, a Special Report which identifies the new location(s).

b. With a land use census identifying a location(s) which yields a calculated dose or dose commitment (via the same exposure pathway) 20 percent greater than at a location from which samples are currently being obtained in accordance with ODCM Specification 1.4.1.1, in lieu of any other report required by ODCM Section 1.6, prepare and submit to the Commission within 30 days, pursuant to Technical Specification 6.9.2, a Special Report which identifies the new location. The new location shall be added to the radiological environmental monitoring program within 30 days. The sampling location, excluding the control station location, having the lowest calculated dose or dose commitment (via the same exposure path-way) may be deleted from this monitoring program after October 31 of the year in which this land use census was conducted.

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.

ODCM, V. C. Summer, SCE&G: Revision 21 (March 1996) 1.0-41

c. The provisions of Technical Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 1.4.2.2 The land use census shall be conducted at least once per 12 months between the dates of June 1 and October 1 using that information which will provide the best results, such as by a door-to-door survey, aerial survey, or by consulting local agriculture authorities.

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1.4.3 Interlaboratory Comparison Program CONTROLS 1.4.3.1 Analyses shall be performed on radioactive materials supplied by a National Institute of Standards and Technology traceable Laboratory as part of an Interlaboratory Comparison Program.

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.

b. The provisions of Technical Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 1.4.3.2 A summary of the results obtained as part of the above required Interlaboratory Comparison Program shall be included in the Annual Radiological Environmental Operating Report.

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1.5 BASES B/1. 1 LIQUID EFFLUENTS B/1 .1.1 Radioactive Liquid Effluent Monitoring Instrumentation The radioactive liquid effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in liquid effluents during actual or potential releases of liquid effluents. The alarm/trip setpoints for these instruments shall be calculated in accordance with the procedures in the ODCM to ensure that the alarm/trip will occur prior to exceeding 10 times the concentration levels specified in 10 CFR 20, Appendix B, Table 2, Column 2. The OPERABILITY and use of this instrumentation is consistent with the requirements of General Design Criteria 60, 63 and 64 of Appendix A to 10 CFR Part 50.

B/I.1.2 Concentration This control is provided to ensure that concentration of radioactive materials released in liquid waste effluents from the site (see Technical Specification Figure 5.1-4) will be less than 10 times the concentration levels specified in 10 CFR Part 20, Appendix B, Table 2, Column 2. It provides operational flexibility for releasing liquid effluents in concentrations to follow the Section II.A design objectives of Appendix I to 10 CFR 50. This limitation provides additional assurance that the levels of radioactive materials in bodies of water outside the site will result in exposures within:

(1) the Section II.A design objectives of Appendix 1, 10 CFR 50, to an individual and (2) restrictions authorized by 10 CFR 20.1301 (e).

The concentration limit for dissolved or entrained noble gases is based upon the assumption that Xe-135 is the controlling radionuclide and its Effluent concentration in air (submersion) was converted to an equivalent concentration in water. This specification does not affect the requirement to comply with the annual limitations of 10 CFR 20.1301 (a).

B/1.1.3 Dose This control is provided to implement the requirements of Sections II.A, III.A and IV.A of Appendix 1, 10 CFR Part 50. The CONTROLS ODCM, V. C. Summer, SCE&G: Revision 21 (March 1996) 1.0-44

implement the guides set forth in Section II.A. 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 material in liquid effluents will be kept "as low as is reasonably achievable". Also, for fresh water sites with drinking water supplies which can be potentially affected by plant operations, there is reasonable assurance that the operation of the facility will not result in radionuclide concentrations in the finished drinking water that are in excess of the requirements of 40 CFR 141. The dose calculations in the ODCM implement the requirements in Section III.A of Appendix I that conformance with guides of Appendix I be shown by calculational procedures based on models and data, such that the actual exposure of an individual 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 NUREG-0133, "Preparation of Radiological Effluent Technical Specifications for Nuclear Power Plants", section 4.3. NUREG-0133 implements Regulatory Guide 1.109, Revision 1, October 1977 (Section C.1 and Appendix A) and Regulatory Guide 1.113, April 1977.

Regulatory Guide 1.109, October 1977, is titled "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". Regulatory Guide 1.113, April 1977, is titled "Estimating Aquatic Dispersion of Effluents from Accidental and Routine Reactor Releases for the Purpose of Implementing Appendix I".

B/1.1.4 Liquid Waste Treatment The OPERABILITY of the liquid radwaste treatment system ensures that this system will be available for use whenever liquid effluents require treatment prior to release to the environment. The requirement that the appropriate portions of this system be used when specified provides assurance that the releases of radioactive materials in liquid effluents will be kept "as low as is reasonably achievable". This control 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 ODCM, V. C. Summer, SCE&G: Revision 21 (March 1996) 1.0-45

fraction of the dose design objectives set forth in Section II.A of Appendix 1, 10 CFR Part 50, for liquid effluents.

B/1.2 GASEOUS EFFLUENTS B/1.2.1 Radioactive Gaseous Effluent Monitoring Instrumentation The radioactive gaseous effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in gaseous effluents during actual or potential releases of gaseous effluents. The alarm/trip setpoints for these instruments shall be calculated in accordance with the procedures in the ODCM to ensure that the alarm/trip will occur prior to exceeding the limits of 10 CFR Part 20. The OPERABILITY and use of this instrumentation is consistent with the requirements of General Design Criteria 60, 63 and 64 of Appendix A to 10 CFR Part 50.

B/1.2.2 Dose Rate This control along with controls 1.2.3 and 1.2.4 provide reasonable assurance that radioactive material discharged in gaseous effluents will not result in the exposure of a member of the public in an unrestricted area, either at or beyond the site boundary in excess of the design objectives of Appendix I to 10 CFR 50. This control is provided to ensure that gaseous effluent from all units on the site will be appropriately controlled yet provides operational flexibility for releasing gaseous effluents to satisfy the section II.B and II.C design objectives of Appendix I to 10 CFR 50.

The restrictions of Control 1.2.3 along with limited occupancy times for a member of the public within the site boundary are sufficient to control exposure to gaseous effluent within 10 CFR 20, Appendix B, Table 2, Column 1 effluent concentrations.

The specified release rate limits restrict, at all times, the corresponding gamma and beta dose rates above background to an individual 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 3000 mrem/year to the skin. These release rate limits also restrict, at all times, the corresponding thyroid dose rate above ODCM, V. C. Summer, SCE&G: Revision 29 (August 2013) 1.0-46

background to a child via the inhalation pathway to less than or equal to 1500 mrem/year.

This control does not affect the requirement to comply with the annual limitations of 10 CFR 20.1301 (a).

B/1.2.3 Dose - Noble Gases This control is provided to implement the requirements of Sections ll.B, III.A and IV.A of Appendix 1, 10 CFR Part 50. The CONTROLS implement the guides set forth in Section II.B 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 release of radioactive material in gaseous effluents will be kept "as low as is reasonably achievable". 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 an individual through appropriate pathways is unlikely to be substantially underestimated. The dose calculations established in the ODCM for calculating the doses due to the actual release rates of radioactive noble gases in gaseous effluents are consistent with the methodology provided in NUREG-0133, "Preparation of Radiological Effluent Technical Specifications for Nuclear Power Plants", section 5.3. NUREG-0133 implements Regulatory Guide 1.109, Revision 1, October 1977 and Regulatory Guide 1.111, Revision 1, July 1977. Regulatory Guide 1.109 is entitled "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 is entitled "Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water Cooled Reactors",

Revision 1, July 1977. The ODCM equations provided for determining the air doses at the site boundary are based upon the historical average atmospheric conditions.

This control applies to the release of gaseous effluents from all reactors at the site.

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B/1.2.4 Dose-Radioiodines, Tritium and Radioactive Materials in Particulate Form This control is provided to implement the requirements of Sections II.C, III.A and IV.A of Appendix 1, 10 CFR Part 50. The Limiting Conditions for Operation are the guides set forth in Section I1.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 Appendix I to assure that the releases of radioactive materials in gaseous effluents will be kept "as low as is reasonably achievable". The ODCM 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 an individual through appropriate pathways in unlikely to be substantially underestimated. The ODCM calculational methods for calculating the doses due to the actual release rates of the subject materials are consistent with the methodology provided in NUREG-0133, "Preparation of Radiological Effluent Technical Specifications for Nuclear Power Plants", section 5.3. NUREG-0133 implements Regulatory Guide 1.109, Revision 1, October 1977 and Regulatory Guide 1.111, Revision 1, July 1977. Regulatory Guide 1.109 is entitled "Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix "

Revision 1, October 1977 and Regulatory Guide 1.111 is entitled "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 controls for radioiodines, tritium, and radioactive materials in particulate form are dependent on the existing radionuclide pathways to man, in the unrestricted area. The pathways which were examined in the development of these calculations were: 1) individual inhalation of airborne radionuclides, 2) deposition of radionuclides onto green leafy 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.

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This specification applies to the release of gaseous effluents from all reactors at the site.

B/1.2.5 Gaseous Radwaste Treatment The OPERABILITY of the GASEOUS RADWASTE TREATMENT SYSTEM and the VENTILATION EXHAUST TREATMENT SYSTEM ensures that the systems will be available for use whenever gaseous effluents require treatment prior to release to the environment. The requirement that the appropriate portions of these systems be used, when specified, provides reasonable assurance that the releases of radioactive materials in gaseous effluents will be kept "as low as is reasonably achievable". 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 objectives given in Section IL.D of Appendix I to 10 CFR Part 50.

The specified limits governing the use of appropriate portions of the systems were specified as a suitable fraction of the dose design objectives set forth in Sections II.B and 11.C of Appendix 1, 10 CFR Part 50, for gaseous effluents.

B/1.3 RADIOACTIVE EFFLUENTS: TOTAL DOSE The control is provided to meet the dose limitations of 40 CFR 190 which have been incorporated into 10 CFR 20.1301(d). The specification requires the preparation and submittal of a Special 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 190 if the individual reactors remain within twice the 10 CFR 50 Appendix I dose design objectives and if direct radiation doses from the units (including outside storage tanks, etc.) are kept small.

The Special Report will describe a course of action which should result in the limitation of dose to a member of the public for 12 consecutive months to within the 40 CFR 190 limits. For the purposes of the Special 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 5 miles must be considered. If the dose to any member of ODCM, V. C. Summer, SCE&G: Revision 29 (August 2013) 1.0-49

the public is estimated to exceed the requirements of 40 CFR 190, the Special Report with a request for a variance (provided the release conditions resulting in violation of 40 CFR 190 have not already been corrected), in accordance with the provisions of 40 CFR 190.11, is considered to be a timely request and fulfills the requirements of 40 CFR 190 until NRC staff action is completed. The variance only relates to the limits of 40 CFR 190 and does not apply in any way to other dose requirements for dose limitation of 10 CFR 20, as addressed in ODCM Controls 1.1.2.1 and 1.2.2.1. An individual is not considered a member of the public during any period in which he/she is engaged in carrying out any operation which is part of the nuclear fuel cycle.

Demonstration of compliance with the limits of 40 CFR 190 or with the design objectives of Appendix I to 10 CFR 50 will be considered to demonstrate compliance with the 0.1 rem limit of 10 CFR 20.1301.

B/1.4.1 Monitoring Program The radiological monitoring program required by this control provides measurements of radiation of radioactive materials in those exposure pathways and for those radionuclides, which lead to the highest potential radiation exposures of individuals 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. The initially specified monitoring program will be effective for at least the first three years of commercial operation.

Following this period, program changes may be initiated based on operational experience.

The detection capabilities required by Table 1.4-3 are state-of-the-art for routine environmental measurements in industrial laboratories. It should be recognized that the LLD is defined as an a Drio (before the fact) limit representing the capability of a measurement system and not as a posteriori (after the fact) limit for a particular measurement. Analyses shall be performed in such a manner that the stated LLDs will be achieved under routine conditions. Occasionally background fluctuations, unavoidably small sample sizes, the presence of interfering nuclides, or other uncontrollable circumstances may render these LLDs unachievable. In such cases, the contributing factors will be identified and described in the Annual Radiological Environmental Operating Report.

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B/1.4.2 Land Use Census This control is provided to ensure that changes in the use of unrestricted areas are identified and that modifications to the monitoring program are made if required by the results of this census. The best survey information from the door-to-door, aerial or consulting 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 assumptions 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.

B/1.4.3 Interlaboratory Comparison Program 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.

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1.6 REPORTING REQUIREMENTS 1.6.1 Annual Radiological Environmental Operating Report 1.6.1.1 Routine radiological environmental operating reports covering the operation of the unit during the previous calendar year shall be submitted prior to May 1 of each year. The initial report shall be submitted prior to May 1 of the year following initial criticality.

1.6.1.2 The Annual Radiological Environmental Operating Report shall include summaries, interpretations, and an analysis of trends of the results of the radiological environmental surveillance activities for the report period, including a comparison with preoperational studies, operational controls (as appropriate), and previous environmental surveillance reports and an assessment of the observed impacts of the plant operation on the environment. The reports shall also include the results of land use censuses required by ODCM Specification 1.4.2.1. If harmful effects or evidence of irreversible damage are detected by the monitoring, the report shall provide an analysis of the problem and a planned course of action to alleviate the problem.

The Annual Radiological Environmental Operating Report shall include summarized and tabulated results in the format of Regulatory Guide 4.8, December 1975 of all radiological environmental samples taken during the report period. In the event that some results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for missing results. The missing data shall be submitted as soon as possible in a supplementary report.

The report shall also include the following: a summary description of the radiological environmental monitoring program; a map of all sampling locations keyed to a table giving distances and directions from one reactor; and the results of licensee participation in the Interlaboratory Comparison Program, required by ODCM Specification 1.4.3.1.

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1.6.2 Annual Radioactive Effluent Release Report 1.6.2.1 A radioactive effluent release report covering the operation of the unit during the previous year of operation shall be submitted prior to May 1 of each year. The period of the first report shall begin with the date of initial criticality.

1.6.2.2 The Radioactive Effluent Release Report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit as outlined in Regulatory Guide 1.21, "Measuring, Evaluating, and Reporting Radioactivity in Solid Wastes and Releases of Radioactive Materials in Liquid and Gaseous Effluents from Light-Water-Cooled Nuclear Power Plants",

Revision 1, June 1974, with data summarized on a quarterly basis following the format of Appendix B thereof.

The Radioactive Effluent Release Report shall include an annual summary of hourly meteorological data collected over the previous year. This annual summary may be either in the form of an hour-by-hour listing of wind speed, wind direction, and atmospheric stability, and precipitation (if measured) on magnetic tape, or in the form of joint frequency distributions of wind speed, wind direction, and atmospheric stability. This same report shall include an assessment of the radiation doses due to the radioactive liquid and gaseous effluents released from the unit during the previous calendar year. This same report shall also include an assessment of the radiation doses from radioactive liquid and gaseous effluents to members of the public due to their activities inside the site boundary (Figures 5.1-3 and 5.1-4 of the VCSNS Technical Specifications) during the year. All assumptions used in making these assessments (i.e., specific activity, exposure time and location) shall be included in these reports. Historical annual average meteorology or meteorological conditions concurrent with the time of release of radioactive materials in gaseous effluents (as determined by sampling frequency and measurement) shall be used for determining the gaseous pathway doses.

The assessment of radiation doses shall be performed in accordance with the OFFSITE DOSE CALCULATION MANUAL (ODCM).

The Radioactive Effluent Release Report shall also include an assessment of radiation doses to the likely most exposed member of the public from reactor releases and other nearby uranium fuel cycle sources (including doses from primary effluent pathways and direct radiation) for the previous 12 consecutive months to show conformance with 40 CFR 190, Environmental Radiation Protection Standards for Nuclear Power Operation. Acceptable methods for calculating the dose contribution from liquid and gaseous effluents are given in Regulatory Guide 1.109, Rev. 1.

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The Radioactive Effluent Release Report shall include unplanned releases from site to unrestricted areas of radioactive materials in gaseous and liquid effluents on a quarterly basis.

The Radioactive Effluent Release Report shall also include the following: an explanation as to why the inoperability of liquid or gaseous effluent monitoring instrumentation was not corrected within the time specified in ODCM Specifications 1.1.1.1 and 1.2.1.1, respectively.

1.6.3 Major Changes To Radioactive Waste Treatment Systems (Liquid and Gaseous) 1.6.3.1 Licensee initiated major changes to the radioactive waste systems (liquid and gaseous):

1. Shall be reported to the Commission in the Annual Radioactive Effluent Release Report for the period in which the evaluation was reviewed by the Plant Safety Review Committee. The discussion of each change shall contain:

a. A summary of the evaluation that led to the determination that the change could be made in accordance with 10 CFR 50.59.

b. Sufficient detailed information to totally support the reason for the change without benefit of additional or supplemental information.

c. A detailed description of the equipment, components and processes involved and the interfaces with other plant systems.

d. An evaluation of the change which shows the predicted releases or radioactive materials in liquid and gaseous effluents that differs from those previously predicted in the license application and amendments thereto.

e. An evaluation of the change which shows the expected maximum exposures to individual in the unrestricted area and to the general population that differ from those previously estimated in the license application and amendments thereto.

f. A comparison of the predicted releases of radioactive materials, in liquid and gaseous effluents, to the actual releases for the period prior to when the changes are to be made.

g. An estimate of the exposure to plant operating personnel as a result of the change.

h. Documentation of the fact that the change was reviewed and found acceptable by the PSRC.

2. Shall become effective upon review and acceptance as set forth in Technical Specification 6.5.

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1.6.4 V. C. Summer Groundwater Protection Program 1.6.4.1 The NEI Industry Groundwater Protection Initiative was established to address operating experience with groundwater contamination at several nuclear power stations.

1.6.4.2 The following criteria have been established to require notification of the NRC and South Carolina Department of Health and Environmental Control (SCDHEC).

1. A radioactive leak or spill that exceeds 100 gallons or of an unknown volume likely to exceed 100 gallons.

2. Any leak or spill, regardless of the volume or activity, deemed by station management to warrant voluntary communication.

3. A water sample from offsite groundwater or surface water which exceeds the reporting criterion of Table 1.4-2.

4. A water sample from an onsite groundwater monitoring well or surface water that is hydrologically connected to groundwater that exceeds the reporting criterion of Table 1.4-2.

1.6.4.3 If any of the above criteria are met, a telephone notification of the NRC and SCDHEC shall be made in accordance with plant procedures.

1.6.4.4 Prepare and submit a written report to the Commission within 30 days.

1.6.4.5 A description of any spill or leak will be included in the Annual Radioactive Effluent Release Report 1.6.4.6 A description of any sample exceeding the reporting criteria of Table 1.4-2 will be included in the Annual Radiological Environmental Operating Report.

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1.7 Definitions ACTION 1.7.1 ACTION shall be that part of a specification which prescribes measures required under designated conditions.

ANALOG CHANNEL OPERATIONAL TEST 1.7.2 An ANALOG CHANNEL OPERATIONAL TEST shall be the injection of a simulated signal into the channel as close to the sensor as practicable to verify OPERABILITY of alarm, interlock and/or trip functions. The ANALOG CHANNEL OPERATIONAL TEST shall include adjustments, as necessary, of the alarm, interlock and/or trip setpoints such that the setpoints are within the required range and accuracy.

CHANNEL CALIBRATION 1.7.3 A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel such that it responds within the required range and accuracy to known values of input. The CHANNEL CALIBRATION shall encompass the entire channel including the sensors and alarm, interlock and/or trip functions, and may be performed by any series of sequential, overlapping or total channel steps such that the entire channel is calibrated.

CHANNEL CHECK 1.7.4 A CHANNEL CHECKS shall be the qualitative assessment of channel behavior during operation by observation. This determination shall include, where possible, comparison of the channel indication and/or status with other indications and/or status derived from independent instrument channels measuring the same parameter.

GASEOUS RADWASTE TREATMENT SYSTEM 1.7.5 A GASEOUS RADWASTE TREATMENT SYSTEM is any system designed and installed to reduce radioactive gaseous effluents by collecting primary coolant system off gases from the primary system and providing for delay or holdup for the purpose of reducing the total radioactivity prior to release to the environment.

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OPERABLE - OPERABILITY 1.7.6 A system, subsystem, train, component or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified function(s), and when all necessary attendant instrumentation, controls, electrical power, cooling or seal water, lubrication or other auxiliary equipment that are required for the system, subsystem, train, component or device to perform its function(s) are also capable of performing their related support function(s).

SOURCE CHECK 1.7.7 A SOURCE CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to a radioactive source.

VENTILATION EXHAUST TREATMENT SYSTEM 1.7.8 A VENTILATION EXHAUST TREATMENT SYSTEM is any system designed and installed to reduce gaseous radioiodine or radioactive material in particulate form in effluents by passing ventilation or vent exhaust gases through charcoal absorbers and/or HEPA filters for the purpose of removing iodines or particulates from the gaseous exhaust stream prior to the release to the environment (such a system is not considered to have any effect on noble gas effluents). Engineered Safety Feature (ESF) atmospheric cleanup systems are not considered to be VENTILATION EXHAUST TREATMENT SYSTEM components.

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2.0 LIQUID EFFLUENT 2.1 Liquid Effluent Monitor Setpoint Calculation The Virgil C. Summer Nuclear Station is located on the Monticello Reservoir which provides supply and discharge for the plant circulating water. This reservoir also provides supply and discharge capacity for the Fairfield Pumped Storage Facility. The Parr Reservoir located below the pumped storage facility is formed by the Parr Dam.

There are two analyzed release pathways and sources of dilution for liquid effluents: the circulating water discharge canal and the liquid effluent line to the penstocks of the pumped storage facility. All liquid effluent pathways discharge to one of these release points. Generally speaking, very low concentrations of radioactive waste are discharged to the circulating water discharge while higher concentrations of radioactive waste are released to the penstocks of the pumped storage facility during the generation cycle.

The calculated setpoint values will be regarded as upper bounds for the actual setpoint adjustments. That is, setpoint adjustments are not required to be performed if the existing setpoint level corresponds to a lower count rate than the calculated value.

Setpoints may be established at values lower than the calculated values, ifdesired.

Calculated monitor setpoints may be added to the ambient background count rate.

GENERAL NOTE: If no discharge is planned for a specific pathway or if the sum of the effluent concentrations of gamma emitting nuclides equals zero, the monitor setpoint should be established as close to background as practical to prevent spurious alarms and yet alarm should an inadvertent release occur.

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2.1.1 Liquid Effluent Monitor Setpoint Calculation Parameters Term Definition* Section of Initial Use A Penstock discharge adjustment factor which will allow the set point 2.1.2 to be established in a convenient manner and to prevent spurious alarms.

= ft/f*x B = Steam Generator Blowdown adjustment factor which will allow the 2.1.4.1 set point to be established in a convenient manner and to prevent spurious alarms.

= fd/fds CECL = the effluent concentration limit (ODCM Control 1.1.2.1) 2.1.2 implementing 10 CFR 20 for the site, in pCi/ml.

Ca = the effluent concentration of alpha emitting nuclides observed by 2.1.2 gross alpha analysis of the monthly composite sample, in p[Ci/mI.

Cf = the measured concentration of Fe-55 in liquid waste as determined 2.1.2 by analysis of the most recent available quarterly composite sample, in CiCVml.

Cg = the effluent concentration of a gamma emitting nuclide, g, observed 2.1.2 by gamma-ray spectroscopy of the waste sample, in [.Ci/ml.

Ci = the concentration of nuclide i, in pCi/ml, as determined by the 2.1.2 analysis of the waste sample.

Cir = the concentration of radionuclide i, in ýiCi/ml, in the Monticello 2.1.2 Reservoir. Inclusion of this term will correct for possible long-term buildup of radioactivity due to recirculation and for the presence of activity recently released to the Monticello Reservoir by plant activities.

CS = the concentration of Sr-89 or Sr-90 in liquid wastes as determined 2.1.2 by analysis of the quarterly composite sample, in pCi/ml.

C' = the measured concentration of H-3 in liquid waste as determined 2.1.2 by analysis of the monthly composite, in iiCi/ml.

c = the setpoint, in gCi/ml, of the radioactivity monitor measuring the 2.1.2 radioactivity concentration in the effluent line prior to dilution and subsequent release. This setpoint which is proportional to the volumetric flow to the effluent line and inversely proportional to the volumetric flow of the dilution stream plus the effluent stream, represents a value which, if exceeded, would result in concentrations exceeding the limits of 10 CFR 20 in the unrestricted area.

All concentrations are in units of ptCi/ml unless otherwise noted.

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Term Definition Section of Initial Use CB = the monitor setpoint concentration for RM-L7, the Nuclear Blowdown 2.1.2.2 Monitor Tank discharge line monitor, in pCi/ml.

CC = the monitor setpoint concentration for RM-L9, the combined Liquid Waste 2.1.2.3 Processing System and Nuclear Blowdown System effluent discharge line monitor, in pCi/ml.

CD the monitor setpoint concentration for RM-L1 1, the Condensate 2.1.4.2.2 Demineralizer Backwash discharge line monitor, in gCi/ml.

CM = the monitor setpoint concentration for RM-L5, the Waste Monitor Tank 2.1.2.1 discharge line monitor, in [tCi/ml.

Csa the monitor setpoint concentration for RM-L3, the initial Steam Generator 2.1.4.1.1 Blowdown Effluent line monitor, in pCi/ml.

CSb = the monitor setpoint concentration for RM-L10, the final Steam Generator 2.1.4.1.1 Blowdown Effluent line monitor, in pCi/ml.

Cr = the monitor setpoint concentration for RM-L8, the Turbine Building Sump 2.1.4.2.1 Effluent line monitor, in pCi/ml.

CFD = the Condensate Demineralize Backwash Effluent Concentration Factor. 2.1.4.2 CFs = the Steam Generator Blowdown Effluent Concentration Factor. 2.1.4.3 CFT = the Turbine Building Sump Effluent Concentration Factor. 2.1.4.2 DF = the dilution factor, which is the ratio of the total dilution flow rate to the 2.1.2 effluent stream flow rate(s).

F = the dilution water flow setpoint as determined prior to the release, in 2.1.2 volume per unit time.

Fd = the flow rate of the Circulating Water System during the time of release of 2.1.4.1 the Turbine Building Sump and/or the Steam Generator Blowdown, in volume per unit time.

Fdc = the dilution flow rate of the Circulating Water System used for effluent 2.1.4.1 monitor setpoint calculations, based on 90 percent of expected Circulating Water System flow rate during the time of release and corrected for recirculated Monticello Reservoir activity, in volume per unit time.

Fdp = the dilution flow rate through the penstock(s) receiving the radioactive 2.1.2 liquid release upon which the effluent monitor setpoint is based, as corrected for any recirculated radioactivity, in volume per unit time.

Fk = The near field dilution factor for Ci during release from Turbine Building 2.1.4.4.1 sump.

Ft the flow rate of water through the Fairfield Pumped Storage Station 2.1.2 penstock(s) to which radioactive liquids are being discharged during the period of effluent release. This flow rate is dependent upon operational status of Fairfield Pumped Storage Station, in volume per unit time.

f = the effluent line flow setpoint as determined for the radiation monitor 2.1.2 location, in volume per unit time.

fd the maximum permissible discharge flow rate for releases to the 2.1.4.1 ODCM, V. C. Summer, SCE&G: Revision 21 (March 1996) 2.0-3

Term Definition Section of Initial Use Circulating Water, in volume per unit time.

fdb* = the flow rate of the Nuclear Blowdown Monitor Tank discharge, in volume 2.1.2 per unit time.

fdm* = the flow rate of a Waste Monitor Tank discharge, in volume per unit time. 2.1.2 fds* = the flow rate of the Steam Generator Blowdown discharge, in volume per 2.1.4.1 unit time.

fdx = the flow rate of the tank discharge, either fdm or fdb, in volume per unit 2.1.2 time.

fr = the recirculation flow rate used to mix the contents of a tank, in volume 2.1.2 per unit time.

ft = the maximum permissible discharge flow rate for batch releases to the 2.1.2 penstocks, in volume per unit time.

ECLi = ECLg, ECLa, ECLS, ECLf, and ECLt = the limiting concentrations of the 2.1.2 appropriate gamma emitting, alpha emitting, and strontium radionuclides, Fe-55, and tritium, respectively, from 10 CFR, Part 20, Appendix B, Table 2, Column 2.

SF = the safety factor, a conservative factor used to compensate for 2.1.2 engineering and measurement uncertainties. SF = 0.5, corresponding to a 100 percent variation.

[C]LLD = the Lower Limit of Detection (LLD) for radionuclide i in liquid waste in the 2.1.3 Waste Monitor Tank, as determined by the analysis required in ODCM Table 1.1-4, in [iCi/ml.

[Cj]M = the concentration of radionuclide i in the waste contained within the 2.1.3 Waste Monitor Tank serving as the holding facility for sampling and analysis prior to discharge, in [tCi/ml.

--Cg = the sum of the concentrations Cg of each measured gamma emitting 2.1.2 g nuclide observed by gamma-ray spectroscopy of the waste sample, in LX Cg

=

plCi/ml.

the gamma isotopic concentrations of the Nuclear Blowdown Monitor Tank as obtained from the sum of the measured concentrations 2.1.2 JB determined by the analysis required in ODCM Table 1.1 -4, in [tCi/ml.

(Conservatively this value will be either zero, if no release is to be conducted from this system, or the maximum measured capacity of the discharge pump if a release is to be conducted.)

ODCM, V. C. Summer, SCE&G: Revision 21 (March 1996) 2.0-4

TIerm Definition Section of Initial Use

= the gamma isotopic concentrations of the Condensate Demineralizer 2.1.4.2.2

.Cg Backwash effluent (including solids) as obtained from the sum of the 911 D measured concentrations determined by the analysis required in ODCM Table 1.1-4, in liCi/ml.

-C = the gamma isotopic concentrations of the Waste Monitor Tank as 2.1.2 gj obtained from the sum of the measured concentrations determined by 911 M the analysis required in ODCM Table 1.1-4, in [iCi/ml.

Cg]

1 = the gamma isotopic concentrations of the Steam Generator Blowdown as obtained from the sum of the measured concentrations 2.1.4.1.1 911 S determined by the analysis required in ODCM Table 1.1-4, in pCi/ml.

C = the gamma isotopic concentrations of the Turbine Building Sump as 2.1.4.2.1 Cg] obtained from the sum of the measured concentrations determined by 911 T the analysis required in ODCM Table 1.1-4, in OCi/ml.

tr = the minimum time for recirculating the contents of a tank prior to 2.1.2 sampling, in minutes.

V = the volume of liquid in a tank to be sampled, in gallons. 2.1.2 V. = release volume for Turbine Building sump release permit j, in gallons. 2.1.4.4.1

-j ODCM, V. C. Summer, SCE&G: Revision 21 (March 1996) 2.0-5

2.1.2 Liauid Radwaste Effluent Line Monitors (RM-L5, RM-L7, RM-L9)

Liquid Radwaste Effluent Line Monitors provide alarm and automatic termination of release functions prior to exceeding 10 times the concentration limits specified in 10 CFR 20, Appendix B, Table 2, Column 2 at the release point to the unrestricted area. To meet this specification, the alarm/trip setpoints for liquid effluent monitors and flow measurement devices are set to assure that the following equation is satisfied:

I0CECL > cf (1)

-F + f where:

CECL the effluent concentration limit specified in 10 CFR 20 Appendix B, Table 2, Column 2. Note that Control 1.1.2.1 limits release concentrations to 10 times the Appendix B, Table 2, Column 2 values.

c =the setpoint, in gCi/ml, of the radioactivity monitor measuring the radioactivity concentration in the effluent line prior to dilution and subsequent release; the setpoint, which is inversely proportional to the volumetric flow of the effluent line and proportional to the volumetric flow of the dilution stream plus the effluent stream, represents a value which, if exceeded, would result in concentrations exceeding 10 times the effluent concentrations of 10 CFR 20 in the unrestricted area.

F = the dilution water flow setpoint as determined prior to the release point, in volume per unit time.

f the effluent line flow setpoint as determined at the radiation monitor location, in volume per unit time.

At the Virgil C. Summer Nuclear Station the Liquid Waste Processing System (LWPS) and the Nuclear Blowdown System (NBS) both discharge to the penstocks of the Fairfield Pumped Storage (FPS) Facility through a ODCM, V. C. Summer, SCE&G: Revision 21 (March 1996) 2.0-6

common line. The available dilution water flow (Fdp) is assumed to be 90 percent of the flow through the FPS penstock(s) to which liquid effluent is being discharged and is dependent upon operational status of the FPS Facility. The waste tank flow rates (fdm and fdb) and the monitor setpoints (CM, CB and cc) are set to meet the condition of equation (1) for a given effluent concentration, C. The three monitor setpoints are determined in accordance with the monitor system configuration for this discharge pathway. The LWPS discharges through RM-L5, which has setpoint CM for alarm/control functions over releases from either Waste Monitor Tanks 1 or

2. The Nuclear Blowdown discharges through RM-L7, which has setpoint cB for alarm/ control functions over releases from the Nuclear Blowdown Monitor Tank. These two release pathways merge into a common line monitored by RM-L9, which has setpoint cc for control functions over the common effluent line. Although the piping is arranged so that simultaneous batch releases from the two systems could be practiced, operational releases shall be from only one of the two batch systems at any given time.

The method by which their setpoints are determined is as follows:

1) The isotopic concentration for a waste tank to be released is obtained from the sum of the measured concentrations as determined by the analysis required in Table 1.1-4:

Ci- 1: Cg +C'a +-Cs++'Ct'+Cf (2) g where:

C, the concentration of nuclide i, in jiCi/ml, as determined by the analysis of the waste sample.*

-- Cg the sum of the concentrations C 9 of each measured gamma g emitting nuclide observed by gamma-ray spectroscopy of the waste sample, in jiCi/ml.

Values for C., Cs, Ct and Cf will be based on most recent available composite sample analyses as required by Table 1.14.

ODCM, V. C. Summer, SCE&G: Revision 21 (March 1996) 2.0-7

Ca* = the effluent concentration of alpha emitting nuclides observed by gross alpha analysis of the monthly composite sample, in pCi/ml.

CS* the concentration of Sr-89 and Sr-90 in liquid waste as determined by analysis of the quarterly composite sample, in pCi/ml.

Ct* = the measured concentration of H-3 in liquid waste as determined by analysis of the monthly composite sample, in liCi/ml.

Cf* the measured concentration of Fe55 in liquid waste as determined by analysis of the quarterly composite sample, in

_LCi/ml.

The Cg term will be included in the analysis of each batch; terms for alpha, strontium, Fe-55, and tritium shall be included as appropriate*. Isotopic concentrations for both the Waste Monitor Tanks (WMT) and the Nuclear Blowdown Monitor Tank (NBMT) may be calculated using equation (2).

Prior to being sampled for analysis, the contents of a tank shall be isolated and recirculated. The minimum recirculation time shall be:

tr = 2V/fr (3) tr = the minimum time for recirculating the contents of a tank prior to sampling.

V = the volume of liquid in the tank to be sampled.

fr the recirculation flow rate used to mix the contents of a tank.

This is done to ensure that a representative sample will be obtained.

Mechanical mixers shall ensure a similar minimum turnover.

Based on most recent available composite sample analysis as required by Table 1.1-4.

ODCM, V. C. Summer, SCE&G: Revision 21 (March 1996) 2.0-8

2) Once isotopic concentrations for either Waste Monitor Tank or the Nuclear Blowdown Monitor Tank have been determined, these values are used to calculate a Dilution Factor, DF, which is the ratio of dilution flow rate to tank flow rate(s) required to assure that 10 times the limiting concentration of 10 CFR 20, Appendix B, Table 2, Column 2 are met at the point of discharge for whichever tank is having its contents discharged.

C - + SF (4) 10 (E CL)i x (5)

. + + L+ + ' xSF D 10(ECL)g [10(ECL)a 10(ECL), I0(ECL)f I0(ECL),

where:

Ci S10(ECL),

1 the sum of the ratios of the measured concentration of nuclide i to 10 times its limiting ECL value for the tank whose contents are being considered for release. For a WMT, X = M. For the NBMT, X = B.

ECLi ECLg, ECLa, ECLS, ECLf, and ECLt, = effluent concentration limits of the appropriate gamma emitting, alpha emitting, and strontium radionuclides, Fe-55, and tritium, respectively, given in 10 CFR, Part 20, Appendix B, Table 2, Column 2.

SF = the safety factor; a conservative factor used to compensate for engineering and measurement uncertainties.

0.5, Corresponding to a 100 percent variation.

ODCM, V. C. Summer, SCE&G: Revision 21 (March 1996) 2.0-9

3) The maximum permissible discharge flow rate, ft, may be calculated for the release of either the WMT or NBMT. First the appropriate Dilution Factor is calculated by applying equation (4), using the appropriate concentration ratio term (i.e. M or B).

Then, Fdp +fdx _ Fdp f, -F D -F for Fdp >> fdx (6) where:

Fdp dilution flow rate to be used in effluent monitor setpoint calculations, based on 90 percent FPS Station expected flow rate, as corrected for any recirculated radioactivity:

Fdp = (0.9) F,(1- C (7) 10 (E CL)(

where:

Ft the flow rate through the Fairfield Pumped Storage Station penstock(s) to which radioactive liquids are being discharged. Ft should normally fall between 2500 and 44800 cfs.

Cir the concentration of radionuclide i, in pCi/ml, in the intake of Fairfield Pumped Storage Station (that is, in the Monticello Reservoir). Inclusion of this term will correct for possible long-term buildup of radioactivity due to recirculation and for the presence of activity recently released to the Monticello Reservoir by plant activities.

For expected discharges of liquid wastes, the summation will be much less than 1.0 and can be ignored (Reference 6).

fdx = the flow rate of the tank discharge, either fdm or fdb.

fdb = flow rate of Nuclear Blowdown Monitor Tank discharge.

(Conservatively this value will be either zero, if no release is ODCM, V. C. Summer, SCE&G: Revision 21 (March 1996) 2.0-10

to be conducted from this system, or the maximum measured capacity of the discharge pump if a release is to be conducted.)

fdm = flow rate of Waste Monitor Tank discharge. (Conservatively this value will either be zero, if no release is to be conducted from this system, or the maximum measured capacity of the discharge pump if a release is to be conducted.)

DF = the Dilution Factor from Step 2.

If ft >- fdx, the release may be made as planned and the flow rate monitor setpoints should be established as in Step 4 (below). Because FdP is normally very large compared to the maximum discharge pump capacities for the Waste Monitor Tank and the Nuclear Blowdown Monitor Tank, it is extremely unlikely that ft < fdx. However, if a situation should arise such that ft < fdx, steps must be taken to assure that equation (1) is satisfied prior to making the release. These steps may include decreasing fdx by decreasing the flow rate of fdm or fdb, and/or increasing Fdp.

When new candidate flow rates are chosen, the calculations above should be repeated to verify that they combine to form an acceptable release. If they do, the establishment of flow rate monitor setpoints may proceed as follows in Step 4. If they do not, the choice of candidate flow rates must be repeated until an acceptable set is identified.

Note that if DF < 1, the waste tank concentration for which the calculation is being performed includes safety factors in Step 2 and meets the instantaneous release rate limits without further dilution.

Even though no dilution would be required, there will be no discharge if minimum dilution flow is not available, since the penstock minimum flow interlock will prevent discharge.

ODCM, V. C. Summer, SCE&G: Revision 21 (March 1996) 2.0-11

4) The dilution flow rate setpoint*, F, is established at 90 percent of the expected available dilution flow rate:

F = (0.9) Ft (8)

The flow rate monitor setpoint* for the effluent stream shall be set at the selected discharge pump rate (normally the maximum discharge pump rate or zero) fdm or fdb chosen in Step 3 above.

5) The radiation monitor setpoints may now be determined based on the values of Y- Ci, F, and f which were specified to ensure releases are limited to 10 times the values of 10 CFR 20, Appendix B, Table 2, Column 2. The monitor response is primarily to gamma radiation, therefore, the actual setpoint is based on YCg.

The setpoint concentration, c, is determined as follows:

C:! E CgXA (9) g A= Adjustment factor which will allow the setpoint to be established in a practical manner for convenience and to prevent spurious alarms.

A = ft / fax (10)

If A _>1, Calculate c and determine the maximum value for the actual monitor setpoint (cpm) from the monitor calibration graph.

Setpoints for flow rates are administrative limits.

ODCM, V. C. Summer, SCE&G: Revision 21 (March 1996) 2.0-12

If A < 1, No release may be made. Reevaluate the alternatives presented in Step 3.

NOTE: If calculated setpoint values are near actual concentrations planned for release, it may be impractical to set the monitor alarm at this value. In this case a new setpoint may be calculated following the remedial methodology presented in Step 3 for the case of ft < fdx.

Within the limits of the conditions stated above, the specific monitor setpoint concentrations for the three liquid radiation monitors RM-L5, RM-L7, and RM-L9 are determined as follows:

2.1.2.1 RM-L5, Waste Monitor Tank Discharge Line Monitor:

CM is in ý.Ci/ml

See GENERAL NOTE under 2.1.

2.1.2.2 RM-L7, Nuclear Blowdown Monitor Tank Discharge Line Monitor:

CB*[< Cý'B(A) (12)

CB is in eiý/ml NOTE: In no case should discharge be made directly from the Nuclear Blowdown Holdup Tank to the penstocks.

See GENERAL NOTE under 2.1.

2.1.2.3 RM-L9, Combined Liquid Waste Processing System and Nuclear Blowdown Waste Effluent Discharge Line Monitor The monitor setpoint concentration on the common line, cc, should be the same as the setpoint ODCM, V. C. Summer, SCE&G: Revision 21 (March 1996) 2.0-13

concentration for the monitor on the active individual discharge line (i.e., cM, or CB as determined above):

Cc*< MAX (CM, CB) (13)

See GENERAL NOTE under 2.1.

NOTE: In all cases, CM, Cs, and c, are the setpoint concentration values in ItCi/ml. The actual monitor setpoints (cpm) for RM-L5, RM-L7, and RM-L9 are determined from the calibration graph for the particular monitor. Initially, the calibration curves were determined conservatively from families of response curves supplied by the monitor manufacturers. A sample is shown in Figure 2.1-1. As releases occur, a historical correlation will be prepared and placed in service when sufficient data are accumulated.

2.1.2.4 RML-5 / RML-9 Alternate Setpoint Methodology For low gamma activity releases with relatively high tritium concentrations, an alternate setpoint methodology can be used as follows which assumes that the tritium concentrations will not increase greater than the value used in the calculation. The tritium concentration used for the calculation is established based on known plant conditions.

Setpoint concentration, c , is determined as follows:

0.9Ft c*C!gICx XCfH. 3 g [Zi *H-3 (1 (ECLD)]

SF ODCM, V. C. Summer, SCE&G: Revision 29 (August 2013) 2.0-14

Where, CfH- 3 = tritium correction factor 10ECLH-O.9Ft fdx CH-3 --- Concentration of tritium in pCi/mL ECLH- 3 = tritium ECL 2.1.3 Liquid Radwaste Discharge Via Industrial and Sanitary Waste System (RM-L5)

In the Virgil C. Summer Nuclear Station liquid waste effluent system design, there exists a mechanism for discharging liquid wastes via the Industrial Sanitary Waste System. The sample point prior to discharge is one of the Waste Monitor Tanks. The analysis requirements are the requirements listed in Table 1.14.

ODCM, V. C. Summer, SCE&G: Revision 29 (August 2013) 2.0-14A

This effluent pathway shall only be used when the following condition is met for all radionuclides, i:

[ci] ,, _[ci] LLD (14)

[Ci] M = the concentration of radionuclide i in the waste contained within the Waste Monitor Tank serving as the holding facility for sampling and analysis prior to discharge, in pCi/ml.

[Ci] LLD = the Lower Limit of Detection, (LLD) for radionuclide i in the liquid waste in the Waste Monitor Tank as determined by the analysis required in Table 1.1-4, in [tCi/ml.

When the conditions of equation (14) are met, liquid waste may be released via the Industrial and Sanitary Waste System pathway. The RM-L5 setpoint should be established as close to background as practical to prevent spurious alarms and yet alarm should an inadvertent high concentration release occur.

2.1.4 Steam Generator Blowdown, Turbine Building Sump, and Condensate Demineralizer Backwash Effluent Lines (RM-L3, RM-L10, RM-L8, RM-L11)

Concentrations of radionuclides in the liquid effluent discharges made via the Turbine Building Sump, Steam Generator Blowdown, and Condensate Demineralizer Backwash are expected to be very low or nondetectable. The first two releases are expected to be continuous in nature and the last a batch release. All will be sampled in an appropriate manner as specified in Table 1.1-4 of the ODCM. The Steam Generator Blowdown Monitors, the Turbine Building Sump Monitor, and the Condensate Demineralizer Backwash Monitor provide alarm and automatic termination.

ODCM, V. C. Summer, SCE&G: Revision 21 (March 1996) 2.0-15

In reality, all of these effluent pathways utilize the circulating water as dilution to the effluent stream, with the circulating water discharge canal being the point of release into an unrestricted area. Steam Generator Blowdown Effluent may be released to the Circulating Water either directly in the Condenser outflow (the normal flow path) or in the first hours following startup via the Industrial and Sanitary Waste System (ISWS) for chemical reasons. The Turbine Building Sump and Condensate Demineralizer Backwash Effluents enter Circulating Water via the sumps and ponds of the Industrial and Sanitary Waste System.

CO3ý To ensure compliance with ODCM specification 1.1.2.1, normally no dilution is assumed for discharges to the Industrial and Sanitary Waste System. Additionally, releases are normally limited to 1 ECL to ensure that the conditions of 10 CFR 20.1301 are met. These administrative controls provide assurance that ODCM specification 1.1.2.1 would not be compromised in the event circulating water dilution is lost. To add operational flexibility for abnormal conditions (radionuclide concentration in Turbine Building sump > 1 ECL), discharges from the Turbine Building sump and concentrations in the discharging ponds of the ISWS may exceed the operational objective, 1 ECL, provided circulating water dilution is sufficient to ensure compliance with ODCM specification 1.1.2.1 and liquid effluents are being discharged in compliance with ODCM specification 1.1.4.1.

Two separate setpoint calculations are given for Turbine Building sump discharges (RM-L8). Section 2.1.4.2.1 describes the setpoint calculation normally used, limiting discharges to 1 ECL. Section 2.1.4.4 provides an alternate setpoint methodology which may be used during abnormal conditions. RM-L8 setpoints are considered in compliance with ODCM specification 1.1.1.1 provided the setpoints C031 are adequate to prevent releases in excess of ODCM specification 1.1.2.1.

Two mutually exclusive setpoint calculation processes are outlined below for steam generator blowdown. Section 2.1.4.1 is to be used whenever Steam Generator Blowdown is being released directly to the Circulating Water in the Condenser outflow, which is the normal mode. Section 2.1.4.2 is to be used whenever Steam Generator Blowdown own is being released to the Industrial and Sanitary Waste System, or diverted to the Nuclear Blowdown Processing System, both of which are alternate modes.

Normally, water collected by the Nuclear Blowdown Processing System has very low specific activity. This water may be processed to the Turbine Building sump.

NOTE: When Circulating Water is unavailable for effluent dilution and water is being directed to a releasing ISWS pond, releases containing activity above LLD (excluding tritium) should be discouraged via pathways which lead to it. Steam Generator Blowdown should be diverted to the Nuclear Blowdown ODCM, V. C. Summer, SCE&G: Revision 21 (March 1996) 2.0-16

Processing System. Condensate Demineralizer Backwash may be diverted to the Turbine Building sump or not released. Turbine Building sump effluent should be processed through temporary demineralizers or diverted to the Excess Liquid Waste Processing System. (These steps are to keep the calculated dose to individuals as low as reasonably achievable.)

An option for directing water from the TBS with specific activity > LLD to a non releasing pond is provided in Section 2.1.4.5.

2.1.4.1 Steam Generator Blowdown Effluent Direct to Circulating Water (Normal Mode)

Equation (1) is again used to assure that effluents are in compliance with the aforementioned specification:

cf lO CECL >

(F+f)

The available dilution water flow (FdC) is dependent upon the mode of operation of the Circulating Water System. Any change in this value will be accounted for in a recalculation of equation (1). The Steam Generator Blowdown flow rate (fds) and the Steam Generator Blowdown monitor setpoints (Csa and csb) are set to meet the condition of equation (1).

RM-L3, the first monitor in the Steam Generator Blowdown discharge pathway, alarms and terminates release of the stream. The discharge is then automatically diverted to the Nuclear Blowdown Processing System. RM-L10, the last monitor in the Steam Generator Blowdown discharge pathway, alarms and terminates the release. Thus, RM-L10 is redundant to RM-L3 and the setpoint (Csb) will be determined in the same manner as RM-L3(csa).

The method by which the monitor setpoints are determined is as follows:

1) The isotopic concentrations for any release source to be or being released are obtained from the sum of the measured ODCM, V. C. Summer, SCE&G: Revision 21 (March 1996) 2.0-17

concentrations as determined in Table 1.1-4. Equation (2) is again employed for this calculation:

I C,=E Cg+Ca,+ C,+q+Cf S g where:

__C, the sum of the measured concentrations as determined by the analysis of the waste sample, in 1tCi/ml.

,_Cg the sum of the concentrations Cg of each measured gamma g emitting nuclide observed by gamma-ray spectroscopy of the waste sample, in ýiCi/ml.

Ca the measured concentration Ca of alpha emitting composite sample, in iiCi/ml.

C = the measured concentrations of Sr-89 and Sr-90 in liquid waste as determined by analysis of the most recent available quarterly composite sample, in lpCi/ml.

Ct the measured concentration of H-3 in liquid waste determined by analysis of the monthly composite sample, in giCi/ml.

Cf the measured concentration of Fe-55 in liquid waste as determined by analysis of the most recent available quarterly composite sample, in ptCi/ml.

Isotopic concentrations for the Steam Generator Blowdown System effluent, the Turbine Building Sump Effluent, and the Condensate Demineralizer Backwash effluent may be calculated using equation (2).

2) Once isotopic concentrations for the Steam Generator Blowdown have been determined, these values are used to calculate a Dilution Factor (DF) which is the ratio of the total dilution flow rate to effluent stream flow rate ODCM, V. C. Summer, SCE&G: Revision 21 (March 1996) 2.0-18

required to limit the effluent concentration at the point of discharge to less than 10 times the values in 10 CFR 20, Appendix B, Table 2, Column 2.

(15) i 10 (E CL)i + SF (16)

DF = 10(ECL)g +(ECL). + C(ECL)f 10(ECL) I0(ECL), s where:

Cj Cg, Ca, Cs, Cf, and Ct; measured concentrations as defined in Step 1. Terms Ca, Cs, Cf, and Ct will be included in the calculation as appropriate.

the sum of the ratios of the measured concentration of Io(ECL), S nuclide i to its limiting value ECLj for the Steam Generator Blowdown effluent.

ECLi ECLg, ECLa, ECLS, ECLf, and ECLt are limiting concentrations of the appropriate radionuclide from 10 CFR, Part 20, Appendix B, Table 2, Column 2 limits.

SF the same generic term as used in Section 2.1.2, Step 2.

0.5

3) The maximum permissible effluent discharge flow rate, fd, may now be calculated for a release from the Steam Generator Blowdown.

ODCM, V. C. Summer, SCE&G: Revision 28 (July 2012) 2.0-19

Fdc + FdC Ifd for FdC >> fd (17)

DF DF where:

Fdc = Dilution flow rate for use in effluent monitor setpoint calculations, based on 90 percent of the expected flow rate of the Circulating Water System during the time of release and corrected for any recirculated activity:

Fdc =(0.9)FdIl--*1 (ECL),C (18) where:

Fd = the flow rate of the Circulating Water System during the time of the 5

release. Fd should normally fall between 1.78 X 10 and 5.34 X 105 gpm when the plant is operating and should be 5000 gpm when the plant is shutdown and the Circulating Water Jockey pump is operating.

Cir = the concentration of radionuclide i, in itCi/ml, in the Circulating Water System intake, (that is, in the Monticello Reservoir). Inclusion of this term will correct for possible long-term buildup of radioactivity due to recirculation and for the presence of activity recently released to the Monticello Reservoir by plant activities. For expected discharges of liquid wastes, the summation will be much less than 1.0 and can be ignored (Reference 6).

fds = Flow rate of Steam Generator Blowdown discharge. (This value normally will be either zero, if no release is to be conducted, or the maximum rated capacity of the discharge pump (250 gpm), if a release is to be conducted.)

ODCM, V. C. Summer, SCE&G: Revision 21 (March 1996) 2.0-20

Note that the equation is valid only for DF > 1; for DF _<1, the effluent concentration meets the release criteria without dilution as well as being in compliance with the conservatism imposed by the Safety Factor in Step 2.

If fd -- fds, releases may be made as planned. Because Fdc is normally very large compared to the maximum discharge pump capacity of the Steam Generator Blowdown System, it is extremely unlikely that fd < fds. However, if a situation should arise such that fd < fds, steps must be taken to assure that equation (1) is satisfied prior to making the release. These steps may include diverting Steam Generator Blowdown to the Nuclear Blowdown Processing System or decreasing the effluent flow rate.

When new candidate flow rates are chosen, the calculations above should be repeated to verify that they combine to form an acceptable release. If they do, the establishment of flow rate monitor setpoints should proceed as follows in Step 4. If they do not provide an acceptable release, the choice of candidate flow rates must be repeated until an acceptable set is identified.

4) The dilution flow rate setpoint for minimum flow rate, F, is established at 90 percent of the expected available dilution flow rate:

F = (0.9) (Fd) (19)

Flow rate monitor setpoints for the Steam Generator Blowdown effluent stream shall be set at the selected discharge pump rate (normally the maximum discharge pump rate) fds chosen in Step 3 above.

5) The Steam Generator Monitor setpoints may be specified based on the values of I Ci, F, and f which were specified to ODCM, V. C. Summer, SCE&G: Revision 21 (March 1996) 2.0-21

limit discharge within 10 times the limits of 1OCFR 20, Appendix B, Table 2, Column 2. Since the monitor responds primarily to gamma radiation, therefore, the actual setpoint is based on Cg. The monitor setpoint in cpm which corresponds to the calculated value c is taken from the monitor calibration graph.

(See NOTE, page 2.0-14.) The setpoint concentration, c, is determined as follows:

c-* Cc XB (20) g B = fd / fdr (21)

If B >_1, Calculate c and determine the maximum value for the actual monitor setpoint (cpm) from the monitor calibration graph.

If B < 1, No release may be made. Reevaluate the alternatives presented in step 3.

NOTE: If the calculated setpoint value is near actual concentrations being released or planned for release, it may be impractical to set the monitor alarm at this value. In this case a new setpoint may be calculated following the remedial methodology presented in steps 3 and 4 for the case fd < fds.

Within the conditions stated above, the specific monitor setpoint concentrations for the two steam generator blowdown monitors RM-L3 and RM-L10 are calculated as shown on the following page.

2.1.4.1.1 For RM-L3, Steam Generator Blowdown Discharge initial monitor, and for RM-L10, Steam Generator Blowdown Discharge final monitor:

ODCM, V. C. Summer, SCE&G: Revision 21 (March 1996) 2.0-22

cSaor cSb !ýEC ()(22)

= the isotopic concentration of the Steam Generator Blowdown Cs effluent as obtained from the sum of the measured concentrations determined by the analysis required in ODCM Table 1.1-4, in jiCi/ml.

See GENERAL NOTE under 2.1.

2.1.4.2 Turbine Building Sump and Condensate Demineralizer Backwash (Normal Mode)

For conservatism, the Turbine Building Sump and Condensate Demineralizer Backwash monitor setpoints (CT and cD) will claim no dilution from the Circulating Water, and will be set at the applicable concentration limit. That is:

CECL > C (23)

The Turbine Building sump monitor, RM-L8, alarms and terminates release upon exceeding the monitor setpoint (CT). The discharge can then be manually diverted to the Excess Waste Processing System.

RM-L1 1, the Condensate Demineralizer Backwash monitor, alarms and terminates release upon exceeding the monitor setpoint (Co). The discharge may then be manually diverted to the Turbine Building sump or simply delayed.

The Turbine Building Sump and Condensate Demineralizer Backwash monitor setpoints are to be established independently of each other and without crediting dilution. They are to be based on the measured radionuclide concentrations of the effluent stream and are to ensure that discharge concentrations do not exceed the ECLs specified in 10 CFR 20, Appendix B, Table 2, Column 2 prior to discharge.

For each effluent stream, a concentration factor (CF) is calculated by summing the ratios of detected radionuclides in the ODCM, V. C. Summer, SCE&G: Revision 21 (March 1996) 2.0-23

effluent stream to the applicable ECLs, the calculated values normalize the effluent mixture to terms of ODCM Control 1.1.2.1 release criteria and includes a safety factor for engineering uncertainty.

CF = (C SF (24)

CFT (F_,CL)1 SF (25)

CFD = C D S (26) where:

the sum of the ratios of the measured concentration of nuclide i

[x(E CL)] T to its limiting value ECLi for the Turbine Building sump effluent.

Ci .] D (ECL) the sum of the measured concentration of nuclide i (in liquid only) to its limiting value ECLi for the Condensate Demineralizer Backwash effluent.

CFT = the concentration factor for the Turbine Building Sump Effluent.

CFD = the concentration factor for the Condensate Demineralizer Backwash Effluent.

SF the generic engineering safety factor used in Section 2.1.2, Step 2.

= 0.5 If CF < 1, calculate c and determine the actual monitor setpoint (cpm) from the calibration curve.

If CF> 1, no release may be made via this path. The release must either be delayed, diverted, or processed.

ODCM, V. C. Summer, SCE&G: Revision 21 (March 1996) 2.0-24

If the concentration factor cannot be reduced to less than or equal to 1, proceed to section 2.1.4.4 or 2.1.4.5 for additional guidance for processing Turbine Building Sump releases.

Within the limits of the conditions stated above, the specific monitor setpoint concentrations for RM-L8 and RM-L1 1 may now be calculated. Because they are primarily sensitive to gamma radiation, their setpoints will be based on the concentrations of gamma emitting radionuclides as follows:

2.1.4.2.1 For RM-L8, Turbine Building Sump Discharge Monitor:

CT !cg Y ý] T+CFT (27)

Where:

= The gamma isotopic concentration of the Turbine Building sump 9[X~cg]T effluent as obtained from the sum of the measured concentrations determined by the analysis required in ODCM Table 1.1-4, in IaCi/ml.

OFT = The Turbine Building sump Effluent Concentration Factor from equation (25).

See GENERAL NOTE under 2.1.

ODCM, V. C. Summer, SCE&G: Revision 21 (March 1996) 2.0-25

2.1.4.2.2 For RM-L1 1, Condensate Demineralizer Backwash Discharge Monitor:

CD ~ce]

I CFD (28) where:

[Zc]D = The gamma isotopic concentration of the Condensate Demineralizer

, D Backwash effluent (including solids) as obtained from the sum of the measured concentrations determined by the analysis required ODCM Table 1.1-4, in piCi/ml.

CFo = The Condensate Demineralizer Backwash Effluent Concentration Factor from equation (26).

See GENERAL NOTE under 2.1.

2.1.4.3 Steam Generator Blowdown Effluent Not Directly to Circulating Water (Alternate Mode)

Equation (23) is again used to assure that effluents are in compliance with the aforementioned specification before dilution in the receiving water:

CECL Ž! C Because dilution is not considered in the setpoint calculation, it is not necessary to calculate maximum permissible discharge flow rates or anticipated available dilution flow rate.

The functions of the two monitors whose setpoints are to be established are described in Section 2.1.4.1 above. The method for the determination is as follows:

1) If a release is found to be permissible, flow rate monitors for the active effluent streams (Steam Generator Blowdown - fds, Turbine Building sump - fdt, and Condensate Demineralizer - fdd) may have their setpoints established at any operationally convenient value.

ODCM, V. C. Summer, SCE&G: Revision 21 (March 1996) 2.0-26

2) The Concentration Factor of equation (24) is again used to ensure the permissibility of the release.

CF = (ECL)i + SF

= (ECL) + SF (29)

All terms are defined in subsection 1.1.3.1 and subscript S refers to the Steam Generator Blowdown Effluent.

If CF < 1, calculate c and determine the actual monitor setpoint (cpm) from the calibration curve.

If CF > 1, no release may be made via this path. The release must either be delayed or diverted for additional processing.

Within the above limitation, setpoint concentrations may now be calculated for the two effluent monitors. Because they are primarily sensitive to gamma radiation, their setpoints will be based on the concentrations of gamma emitting radionuclides as follows:

2.1.4.3.1 For RM-L3, Steam Generator Blowdown Discharge initial monitor, and RML-10, Steam Generator Blowdown Discharge final monitor:

CSa or C'b ~ZgSC~ss (30)

ODCM, V. C. Summer, SCE&G: Revision 21 (March 1996) 2.0-27

Where:

l,*g] The isotopic concentration of the Steam Generator Blowdown C S effluent as obtained from the sum of the measured concentrations determined by the analysis required in ODCM Table 1.1-4, in

[tCi/ml.

CFs = The Steam Generator Blowdown Effluent Concentration Factor from equation (29).

See GENERAL NOTE under 2.1.

C03-> 2.1.4.4 Turbine Building Sump (Abnormal Conditions)

Provided circulating water is available, 1 to 3 circulating water pumps, effluent exceeding 1 ECL may be released from the Turbine Building sump to the industrial and sanitary waste system, using the setpoint in this section, provided the following conditions are met:

1) Instantaneous release rate limits of ODCM Specification 1.1.2.1 are not exceeded in the circulating water discharge canal.

2) Annual average concentrations of radioactivity in ISWS ponds will not exceed 1 ECL.

3) The limits of ODCM specification 1.1.4.1 will not be exceeded with actual liquid effluent releases over a 31 day period.

4) Average discharge flow does not exceed values used in setpoint determination.

In addition, the source of radioactivity should be identified and isolated.

Radionuclide concentration in Turbine Building sump effluent should be restored to <1 ECL as soon as possible and normal setpoint reestablished. Radionuclide concentration in Pond 6B should be restored to < LLD (excluding tritium) using dilution as necessary (normal flow from the TBS would normally be adequate). Turbine Building sump samples should be obtained and analyzed every twelve hours while the alternate setpoint is being used to ensure that the setpoint remains conservative with respect to the isotopic mixture and to ensure offsite doses are within ODCM limits.

ODCM, V. C. Summer, SCE&G: Revision 21 (March 1996) 2.0-28

Alternate setpoint methodology for Turbine Building sump (RM-L8) is available to ensure operational flexibility in the event radioactivity is detected in the Turbine Building sump > 1 ECL and the release would result in minimal offsite dose. The alternate setpoint methodology is not intended to be used continuously. To remove restrictions on operation of circulating water, pond concentrations should be restored to < LLD as soon as possible. The setpoint methodology follows:

2.1.4.4.1 For RM-L8, Turbine Building Sump (alternate methodoloqy)

Cg 1 CT<9 x- (57)

CFT Fk

where, Fk The near field dilution factor for C, during release from Turbine Building sump.

(average undiluted waste flow)

(average flow from discharge structure)

For purpose of implementing section 2.1.4.4 release condition 2, the following must be satisfied.

j=

L(C, /ECLI)ITJJ/

~1 t zl < 1.0 (58) j=1 where, [E(C1 /ECLj)]Tj = the sum of the ratios of the measured concentration of nuclide i to its limiting value ECLj for the Turbine Building sump effluent for release permit j, including proposed permit, Vj = Release volume for Turbine Building sump release permit j (gal), and j = index for batch release permits during the calendar year.

ODCM, V. C. Summer, SCE&G: Revision 21 (March 1996) 2.0-29

2.1.4.5 Turbine Building Sump - Special Considerations During Station Shutdowns During periods in which circulating water (CW) is not available for diluting Turbine Building Sump (TBS) discharges, effluent from the TBS may be directed to a non-releasing pond and offsite dose calculations required by Specification 1.1.3.1 deferred until CW is restored. RM-L8 setpoint requirement specified by Specification 1.1.1.1 is not applicable when directing water from the TBS to a non-releasing ISWS pond provided the following conditions are met.

1) Sufficient freeboard is available in the non-releasing pond to ensure that pond contents will not be released to the CW discharge canal prior to reestablishing CW flow.

2) Release of ISWS contents will be in compliance with Specifications 1.1.2.1, 1.1.3.1 and 1.1.4.1 once CW flow has been reestablished.

3) ISWS pond radioactivity will not exceed 1 ECL.

4) TBS samples are obtained and analyzed every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months while water is being directed to a non-releasing pond.

5) Sample non-releasing pond within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months of adding water > 1 ECL and every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter while TBS discharge concentrations exceed 1 ECL.

Once samples have been obtained and release acceptability determined, RM-L8 setpoint may be increased to 2 times indication to allow release of sump contents to a non-releasing pond.

Demonstrating compliance with item 3 can be performed by calculations using TBS samples and discharge volumes or by sampling ISWS ponds.

ODCM, V. C. Summer, SCE&G: Revision 21 (March 1996) 2.0-30

Figure 2.1-1 Example Liquid Effluent Monitor Calibration Curve E

C) z 0

C-z 0

z 0

0

+02 1E+03 IE+04 COUNT RATE (cpm)

ODCM, V. C. Summer, SCE&G: Revision 21 (March 1996) 2.0-31

2.2 Dose Calculation for Liquid Effluents The method of this section is to be used in all cases for calculating doses to individuals from routine liquid effluents. Five notes at the end of the section confirm the values which certain parameters are to be assigned in some special cases.

2.2.1 Liquid Effluent Dose Calculation Parameters Section of Term Definition Initial Use Aj, = the site related ingestion dose commitment factor 2.2.2 to the total body or any organ r, for each identified principal gamma and beta emitter listed in Table 2.2-3 in mrem-ml per hr-pCi.

BFj = Bioaccumulation Factor for nuclide i, in fish, 2.2.2 pCi/Kg per pCi/I, from Table 2.2-1.

Cik the average concentration of radionuclide, i, in 2.2.2 undiluted liquid effluent during time period Atk from any liquid released, in pCi/ml.

DFT = a dose conversion factor for nuclide, i, for adults 2.2.2 in preselected organ, T, in mrem/pCi found in Table 2.2-2.

D= the cumulative dose commitment to the total 2.2.2 body or any organ, T, from the liquid effluents for the total time period, ZAtk in mrem (Ref. 1).

D= Dilution Factor from the near field area within 2.2.2 one-quarter mile of the release points to the potable water intake for adult water consumption; for V. C. Summer, Dw = 1.

Fk = the near field average dilution factor for Cik 2.2.2 during any liquid effluent release.

Ko = 1.14 x 10 5, units conversion factor = 2.2.2 (106 pCi/ptCi) (103 mill) - 8760 hr/yr Atk the length (in hours) of a time period over which 2.2.2 concentrations and flow rates are averaged for dose calculations.

UF = 21 kg/yr, fish consumption (adult) (Reference 3). 2.2.2 U11 = 730 kg/yr, water consumption (adult) (Reference 2.2.2 3).

Z = applicable near-field dilution factor when no 2.2.2 additional dilution is to be considered; Z = 1.

ODCM, V. C. Summer, SCE&G: Revision 28 (July 2012) 2.0-32

2.2.2 Methodology The dose contribution from all radionuclides identified in liquid effluents released to unrestricted areas is calculated using the following expression:

D, A, >3't A.C& J (31)

A,, =K. ( (U,,,/ D.,) + UFBF,) Dl,, (32)

F1, (average undiluted liquid waste flow)

(average flow from the discharge structure) (Z)

NOTE 1: If radioactivity in the Monticello Reservoir (Cir) becomes > the LLD specified in ODCM, Table 1.1-4, that concentration must be included in the Dose determination. For this part of the dose calculation, Fk = 1 and Atk = the entire time period for which the dose is being calculated.

NOTE 2: Prior to termination of Circulating Water Pumps, an assessment of the dose resulting from pond radioactivity concentrations and discharge flow rates from the Industrial And Sanitary Waste System (ISWS) will be performed as follows. Sampling of the liquid in the ISWS wilt be initiated, and the measured concentrations of radionuclides will be used in the dose calculations with Fk = I and Atk = the entire time period for which the dose is being calculated.

NOTE 3: For releases through the ISWS pathway when circulating water is not available, dose projections for assessment of release acceptability should be based on the most representative samples obtained from in plant sumps. Normally sump samples are also used to assess actual release. However, due to the ultraconservative assumptions when circulating water is not available, i.e. dose calculations are based on radioactive material concentration in the discharge stream regardless of release volume, representative samples from the ISWS may be used to evaluate impact of releases.

ODCM, V. C. Summer, SCE&G: Revision 21 (March 1996) 2.0-33

NOTE 4: During periods when the Circulating Water Pumps are in operation, any releases to the ISWS are to be credited with dilution in Circulating Water for dose calculation purposes, even though such dilution is normally not claimed in the setpoint calculation. When taken in union with the note above, this procedure results in some overestimation of dose to the population because discharges made to the ISWS just before loss of Circulating Water will be counted twice in the dose calculation process.

NOTE 5: If radioactivity in the Service Water becomes > LLD as determined by the analysis required by ODCM, Table 1.1-4, that concentration must be included in the Dose determination. For this part of the dose calculation, Fk =1 and A tk = the entire time since the last Service Water sample was taken.

ODCM, V. C. Summer, SCE&G: Revision 21 (March 1996) 2.0-34

2.3 Liquid Effluent Releases through the Neutralization Basin Releases of slightly contaminated liquids from pathways feeding the Neutralization Basin (Pond 007) through Circulating Water (CW) may be made under strictly controlled conditions.

Releases from these pathways (e.g., NaOH sump, RWST sump) will be allowed if the following conditions are adhered to in controlling the radioactive materials released.

2.3.1 Rainwater Tank Rainwater collected in the RWST sump is pumped to the RWST Pit Drain Tank (Rainwater Tank) for analysis and subsequent release. Normally the rainwater is sampled, found to contain no detectable radioactivity, and is released to the environment via the storm drain system. If measurable amounts of radioactive materials are found in Rainwater Tank samples, the tank may be pumped to a Waste Monitor Tank and released without processing. In order to allow for operational flexibility, the Rainwater Tank containing radioactive materials may be drained to the NaOH sump and discharged to the circulating water (CW) system via the Neutralization basin (007). The following constraints are to be applied for releases through this pathway:

(1) At least one CW pump must be used for dilution to release through this pathway.

(2) Chemistry Services must be notified to verify that conditions in the Neutralization Basin are such that additions to the basin can be made.

(3) Using the Rainwater Tank analysis and available circulating water, a release calculation must be performed that shows that releases will be less than 6.OE-4 mrem (whole body) and 2.OE-3 mrem (any organ). These limits represent 1% of unprocessed effluent 31-day dose limits (ODCM Section 1.1.4.1).

ODCM, V. C. Summer, SCE&G: Revision 21 (March 1996) 2.0-35

(4) If these limits are met, the Rainwater Tank may be drained through the NaOH sump to the Neutralization Basin. Chemistry will then release the Neutralization Basin contents through circulating water as soon as possible once their procedural and NPDES release criteria are met.

2.3.2 NaOH Spray Tank and Stored NaOH (1) The same limits and conditions as 2.3.1(1-4) apply for releases from the NaOH sump.

(2) Samples should be obtained and analyzed during performance of NaOH Spray Tank activities that require the draining of NaOH from the tank or sight glass.

(3) If the sample show concentrations of radionuclides that would exceed the dose limits above and the tank must have liquid removed from it, the contaminated NaOH may be drained to appropriate holding containers for temporary storage. Once the conditions for release become favorable (e.g.

return of CW), the containers used for temporary storage may be sampled and analyzed for release. If the dose limitations in 2.3.1 (3) are met and Chemistry approval is obtained, a release permit is generated and the containers can be drained through the NaOH sump or emptied directly to the Neutralization Basin for release through CW.

ODCM, V. C. Summer, SCE&G: Revision 22 (August 1996) 2.0-36

TABLE 2.2-1 BIOACCUMULATION FACTORS*

(pCi/kg per pCi/liter)

ELEMENT FRESHWATER FISH H 9.OE- 01 C 4.6E 03 F 1.OE 01 Na 1.0E 02 P 1.0E 05 Cr 2.OE 02 Mn 4.OE 02 Fe 1.0E 02 Co 5.0E 01 Ni 1.0E 02 Cu 5.OE 01 Zn 2.OE 03 Br 4.2E 02 Rb 2.OE 03 Sr 3.OE 01 Y 2.5E 01 Zr 3.3E 00 Nb 3.OE 04 Mo 1.OE 01 Tc 1.5E 01 Ru 1.OE 01 Rh 1.OE 01 Sb 1.OE 00 Te 4.0E 02 I 1.5E 01 Cs 2.OE 03 Ba 4.OE 00 La 2.5E 01 Ce 1.OE 00 Pr 2.5E 01 Nd 2.5E 01 W 1.2E 03 Np 1.OE 01

Values in Table 2.2-1 are taken from Reference 3, Table A-I.

ODCM, V. C. Summer, SCE&G: Revision 25 (January 2007) 2.0-37

TABLE 2.2-2 Page 1 of 2 ADULT INGESTION DOSE FACTORS*

(mrem/pCi ingested)

NUCLIDE BONE LIVER T-BODY THYROID KIDNEY LUNG GI-LLI H-3 NO DATA 1.05E-07 1.05E-07 1.05E-07 1.05E-07 1.05E-07 1.05E-07 C-14 2.84E-06 5.68E-07 5.68E-07 5.68E-07 5.68E-07 5.68E-07 5.68E-07 tF-18 6.24E-07 NO DATA 6.92E-08 NO DATA NO DATA NO DATA 1.85E-08 NA-24 1.70E-06 1.70E-06 1.70E-06 1.70E-06 1.70E-06 1.07E-06 1.70E-06 P-32 1.93E-04 1.20E-05 7.46E-06 NO DATA NO DATA NO DATA 2.17E-05 CR-51 NO DATA NO DATA 2.66E-09 1.59E-09 5.86E-10 3.53E-09 6.69E-07 MN-54 NO DATA 4.57E-06 8-72E-07 NO DATA 1.36E-06 NO DATA 1.40E-05 MN-56 NO DATA 1.15E-07 2.04E-08 NO DATA 1.46E-07 NO DATA 3.67E-06 FE-55 2.75E-06 1.90E-06 4.43E-07 NO DATA NO DATA 1.06E-06 1.09E-06 FE-59 4.34E-06 1.02E-05 3.91 E-06 NO DATA NO DATA 2.85E-06 3.40E-05 tCO-57 NO DATA 1.75E-07 2.91 E-07 NO DATA NO DATA NO DATA 4.44E-06 CO-58 NO DATA 7.45E-07 1.67E-06 NO DATA NO DATA NO DATA 1.51 E-05 CO-60 NO DATA 2.14E-06 4.72E-06 NO DATA NO DATA NO DATA 4.02E-05 NI-63 1.30E-04 9.01 E-06 4.36E-06 NO DATA NO DATA NO DATA 1.88E-06 NI-65 5.28E-07 6.86E-08 3.13E-08 NO DATA NO DATA NO DATA 1.74E-06 CU-64 NO DATA 8.33E-08 3.91 E-08 NO DATA 2.10E-07 NO DATA 7.1OE-06 ZN-65 4.84E-06 1.54E-05 6.96E-06 NO DATA 1.03E-05 NO DATA 9.70E-06 ZN-69 1.03E-08 1.97E-08 1.37E-09 NO DATA 1.28E-08 NO DATA 2.96E-09 tZn-69mj 1.70E-07 4.08E-07 3.73E-08 NO DATA 2.47E-07 NO DATA 2.49E-05 tBR-82 NO DATA NO DATA 2.26E-06 NO DATA NO DATA NO DATA 2.59E-06 BR-831 NO DATA NO DATA 4.02E-08 NO DATA NO DATA NO DATA 5.79E-08 BR-84 NO DATA NO DATA 5.21 E-08 NO DATA NO DATA NO DATA 4.09E 13 BR-85 NO DATA NO DATA 2.14E-09 NO DATA NO DATA NO DATA LT E-24**

RB-86 NO DATA 2.11E-05 9.83E-06 NO DATA NO DATA NO DATA 4.16E-06 RB-88 NO DATA 6.05E-8 3.21 E-08 NO DATA NO DATA NO DATA 8.36E-19 RB-89t NO DATA 4.01 E-8 2.82E-08 NO DATA NO DATA NO DATA 2.33E-21 SR-891 3.08E-04 NO DATA 8.84E-06 NO DATA NO DATA NO DATA 4.94E-05 SR-90t 7.58E-03 NO DATA 1.86E-03 NO DATA NO DATA NO DATA 2.19E-04 SR-911 5.67E-06 NO DATA 2.29E-07 NO DATA NO DATA NO DATA 2.70E-05 SR-921 2.15E-06 NO DATA 9.30E-08 NO DATA NO DATA NO DATA 4.26E-05 Y-90 9.62E-09 NO DATA 2.58E-10 NO DATA NO DATA NO DATA 1.02E-04 Y-91Mt 9.09E-11 NO DATA 3.52E-12 NO DATA NO DATA NO DATA 2.67E-10 Y-91 1.41 E-07 NO DATA 3.77E-09 NO DATA NO DATA NO DATA 7.76E-05 Y-92 8.45E-1 0 NO DATA 2.47E-1 1 NO DATA NO DATA NO DATA 1.48E-05 Y-93 2.68E-09 NO DATA 7.40E-1 1 NO DATA NO DATA NO DATA 8.50E-05 ZR-951 3.04E-08 9.75E-09 6.60E-09 NO DATA 1.53E-08 NO DATA 3.09E-05 ZR-97t 1.68E-09 3.39E-1 0 1.55E.10 NO DATA 5.12E-10 NO DATA 1.05E-04 NB-95 6.22E-09 3.46E-09 1.86E-09 NO DATA 3.42E-09 NO DATA 2.10E-05 tNB-97 5.22E-1 1 1.32E- 11 4.82E-1 2 NO DATA 1.54E-1 1 NO DATA 4.87E-08 MO-99; NO DATA 4.31 E-06 8.20E-07 NO DATA 9.76E-06 NO DATA 9.99E-06 Daughter contributions are included (see Reference 13).

t Values taken from Reference 13, Table 4.

Values other than those footnoted in Table 2.2-2 are taken from Reference 3, Table E-1 1.

- Less than E-24.

ODCM, V. C. Summer, SCE&G: Revision 24 (May 2006) 2.0-38

TABLE 2.2-2 (continued)

Page 2 of 2 NUCLIDE BONE LIVER T-BODY THYROID KIDNEY LUNG GI-LLI TC-99M 2.47E-10 6.98E-09 8.89E-09 NO DATA 1.06E-08 3.42E-10 4.13E-07 TC-101 2.54E-10 3.66E-10 3.59E-09 NO DATA 6.59E-09 1.87E-10 1.10E-21 RU-1 03t 1.85E-07 NO DATA 7.97E-08 NO DATA 7.06E-07 NO DATA 2.16E-05 RU-105t 1.54E-08 NO DATA 6.08E-07 NO DATA 1.99E-07 NO DATA 9.42E-06 RU-106: 2.75E-06 NO DATA 3.48E-07 NO DATA 5.31 E-06 NO DATA 1.78E-04 AG-110M4 1.60E-07 1.48E-07 8.79E-08 NO DATA 2.91 E-07 NO DATA 6.04E-05 tSB-124 2.80E-06 5.29E-08 1.11E-06 6.79E-09 NO DATA 2.18E-06 7.95E-05 tST-125 1.79E-06 2.OOE-08 4.26E-07 1.82E-09 NO DATA 1.38E-06 1.97E-05 tSB-126 1.15E-06 2.34E-08 4.15E-07 7.04E-09 NO DATA 7.05E-07 9.40E.05 tSB-127 2.58E-07 5.65E-09 9.90E-08 3.1OE-09 NO DATA 1.53E-07 5.90E-05 TE-125M 2.68E-06 9.71 E-07 3.59E-07 8.06E-07 1.09E-05 NO DATA 1.07E-05 TE-127Mt 6.77E-06 2.42E-06 8.25E-07 1.73E-06 2.75E-05 NO DATA 2.27E-05 TE-127 1.10E-07 3.95E-08 2.38E-08 8.15E-08 4.48E-07 NO DATA 8.68E-06 TE-129Mt 1.15E-05 4.29E-06 1.82E-06 3.95E-06 4.80E-05 NO DATA 5.79E-05 TE-129 3.14E-08 1.18E-08 7.65E-09 2.41 E-08 1.32E-07 NO DATA 2.37E-08 TE-131Mt 1.73E-06 8.46E-07 7.05E-07 1.34E-06 8.57E-06 NO DATA 8.40E-05 TE-131t 1.97E-08 8.23E-09 6.22E-09 1.62E-08 8.63E-08 NO DATA 2.79E-09 TE-132t 2.52E-06 1.63E-06 1.53E-06 1.80E-06 1.57E-05 NO DATA 7.71E-05 1-130 7.56E-07 2.23E-06 8.80E-07 1.89E-04 3.48E-06 NO DATA 1.92E-06 1-131:t 4.16E-06 5.95E-06 3.41 E-06 1.95E-03 1. 02E-05 NO DATA 1.57E-06 1-132 2.03E-07 5.43E-07 1.90E-07 1.90E-05 8.65E-07 NO DATA 1.02E-07 1-1 33t 1.42E-06 2.47E-06 7.53E-07 3.63E-04 4.31 E-06 NO DATA 2.22E-06 1-134 1.06E-07 2.88E-07 1.03E-07 4.99E-06 4.58E-07 NO DATA 2.51E-10 I-1 35t 4.43E-07 1.16E-06 4.28E-07 7.65E-05 1.86E-06 NO DATA 1.31 E-06 CS-134 6.22E-05 1.48E-04 1.21 E-04 NO DATA 4.79E-05 1.59E-05 2.59E-06 CS-136 6.51E-06 2.57E-05 1.85E-05 NO DATA 1.43E-05 1.96E-06 2.92E-06 CS-1 37t 7.97E-05 1.09E-04 7.14E-05 NO DATA 3.70E-05 1.23E-05 2.11E-06 CS-1 38 5.52E-08 1.09E-07 5.40E-08 NO DATA 8.01 E-08 7.91 E-09 4.65E-13 BA-1 39 9.70E-08 6.91 E-11 2.84E-09 NO DATA 6.46E-11 3.92E-11 1.72E-07 BA-140: 2.03E-05 2.55E-08 1.33E-06 NO DATA 8.67E-09 1.46E-08 4.18E-05 BA-141t 4.71 E-08 3.56E-11 1.59E-09 NO DATA 3.31 E-11 2.02E-11 2.22E-17 BA-142t 2.13E-08 2.19E-11 1.34E-09 NO DATA 1.85E-11 1.24E-11 3.OOE-26 LA-140 2.50E-09 1.26E-09 3.33E-10 NO DATA NO DATA NO DATA 9.25E-05 LA-142 1.28E-10 5.82E-11 1.45E-11 NO DATA NO DATA NO DATA 4.25E-07 CE-141 9.36E-09 6.33E-09 7.18E-10 NO DATA 2.94E-09 NO DATA 2.42E-05 CE-143: 1.65E-09 1.22E-06 1.35E-10 NO DATA 5.37E-10 NO DATA 4.56E-05 CE-144t 4.88E-07 2.04E-07 2.62E-08 NO DATA 1.21 E-07 NO DATA 1.65E-04 PR-143 9.20E-09 3.69E-09 4.56E-10 NO DATA 2.13E-09 NO DATA 4.03E-05 PR-144 3.01 E-11 1.25E-11 1.53E-12 NO DATA 7.05E-12 NO DATA 4.33E-18 ND-147t 6.29E-09 7.27E-09 4.35E-10 NO DATA 4.25E-09 NO DATA 3.49E.05 W-187 1.03E-07 8.61 E-08 3.01 E-08 NO DATA NO DATA NO DATA 2.82E-05 NP-239 1.19E-09 1.17E-10 6.45E-11I NO DATA 3.65E-10 NO DATA 2.40E-05 ODCM, V. C. Summer, SCE&G: Revision 24 (May 2006) 2.0-39

TABLE 2.2-3 SITE RELATED INGESTION DOSE COMMITMENT FACTOR, AiR*

(mrem/hr per LCi/ml)

Page 1 of 2 NUCLIDE BONE LIVER T-BODY THYROID KIDNEY LUNG GI-LLI H-3 NO DATA 8.96E + 00 8.96E + 00 8.96E + 00 8.96E + 00 8.96E + 00 8.96E + 00 C-14 3.15E + 04 6.30E + 03 6.30 + 03 6.30E + 03 6.30E + 03 6.30E + 03 6.30E + 03 F-18 6.69E + 01 NO DATA 7.42E + 00 NO DATA NO DATA NO DATA 1.98E + 00 NA-24 5.48E + 02 5.48E + 02 5.48E + 02 5.48E + 02 5.48E + 02 5.48E + 02 5.48E + 02 P-32 4.62E + 07 2.87E + 06 1.79E + 06 NO DATA NO DATA NO DATA 5.20E + 06 CR-51 NO DATA NO DATA 1.49E + 00 8.94E-01 3.29E-01 1.98E + 00 3.76E + 02 MN-54 NO DATA 4.76E + 03 9.08E + 02 NO DATA 1.42E + 03 NO DATA 1.46E + 04 MN-56 NO DATA 1.20E + 02 2.12E + 01 NO DATA 1.52E + 02 NO DATA 3.82E + 03 FE-55 8.87E + 02 6.13E + 02 1.43E + 02 NO DATA NO DATA 3.42E + 02 3.52E + 02 FE-59 1.40E + 03 3.29E + 03 1.26E + 03 NO DATA NO DATA 9.19E + 02 1.10E + 04 CO-57 NO DATA 3.55E + 01 5.91E + 01 NO DATA NO DATA NO DATA 9.01E + 02 CO-58 NO DATA 1.51E + 02 3.39E + 02 NO DATA NO DATA NO DATA 3.06E + 03 CO-60 NO DATA 4.34E + 02 9.58E + 02 NO DATA NO DATA NO DATA 8.16E + 03 NI-63 4.19E + 04 2.91E + 03 1.41E + 03 NO DATA NO DATA NO DATA 6.07E + 02 NI-65 1.70E + 02 2.21E + 01 1.01 E + 01 NO DATA NO DATA NO DATA 5.61E + 02 CU-64 NO DATA 1.69E + 01 7.93E + 00 NO DATA 4.26E + 01 NO DATA 1.44E + 03 ZN-65 2.36E + 04 7.50E + 04 3.39E + 04 NO DATA 5.02E + 04 NO DATA 4.73E + 04 ZN-69 5.02E + 01 9.60E + 01 6.67E + 00 NO DATA 6.24E + 01 NO DATA 1.44E + 01 Zn-69mn 8.28E + 02 1.99E + 03 1.82E + 02 NO DATA 1.20E + 03 NO DATA 1.21 E + 05 BR-82 NO DATA NO DATA 2.46E + 03 NO DATA NO DATA NO DATA 2.82E + 03 BR-83t NO DATA NO DATA 4.38E + 01 NO DATA NO DATA NO DATA 6.30E + 01 BR-84 NO DATA NO DATA 5.67E + 01 NO DATA NO DATA NO DATA 4.45E - 04 RB-85 NO DATA NO DATA 2.33E + 00 NO DATA NO DATA NO DATA 1.09E - 15 RB-86 NO DATA 1.03E + 05 4.79E + 04 NO DATA NO DATA NO DATA 2.03E + 04 RB-88 NO DATA 2.95E + 02 1.56E + 02 NO DATA NO DATA NO DATA 4.07E - 09 RB-891 NO DATA 1.95E + 02 1.37E + 02 NO DATA NO DATA NO DATA 1.13E - 11 SR-891 4.78E + 04 NO DATA 1.37E + 03 NO DATA NO DATA NO DATA 7.66E + 03 SR-90t 1.18E + 06 NO DATA 2.88E + 05 NO DATA NO DATA NO DATA 3.48E + 04 SR-91$ 8.79E + 02 NO DATA 3.55E + 01 NO DATA NO DATA NO DATA 4.19E + 03 SR-92t 3.33E + 02 NO DATA 1.44E + 01 NO DATA NO DATA NO DATA 6.60E + 03 Y-90 1.38E + 00 NO DATA 3.69E - 02 NO DATA NO DATA NO DATA 1.46E + 04 Y-91Mt 1.30E - 02 NO DATA 5.04E - 04 NO DATA NO DATA NO DATA 3.82E - 02 Y-91 2.02E + 01 NO DATA 5.39E - 01 NO DATA NO DATA NO DATA 1.11E + 04 Y-92 1.21 E - 01 NO DATA 3.53E - 03 NO DATA NO DATA NO DATA 2.12E + 03 Y-93 3.83E - 01 NO DATA 1.06E - 02 NO DATA NO DATA NO DATA 1.22E + 04 ZR-951 2.77E + 00 8.88E - 01 6.01E - 01 NO DATA 1.39E + 00 NO DATA 2.82E + 03 ZR-971 1.53E - 01 3.09E - 02 1.41 E - 02 NO DATA 4.67E - 02 NO DATA 9.57E + 03 NB-95 4.47E + 02 2.49E + 02 1.34E + 02 NO DATA 2.46E + 02 NO DATA 1.51E + 06 NB-97 3.75E + 00 9.49E - 01 3.47E - 01 NO DATA 1.11E +00 NO DATA 3.50E + 03 I Daughter contributions are included (see Reference 13).

Calculated using equation (32) and Tables 2.2-1 and 2.2-2 ODCM, V. C. Summer, SCE&G: Revision 21 (March 1996) 2.040

TABLE 2.2-3 SITE RELATED INGESTION DOSE COMMITMENT FACTOR, A,,*

(mrem/hr per jiCi /ml)

Page 2 of 2 NUCLIDE BONE LIVER T-BODY THYROID KIDNEY LUNG GI-LLI MO-99t NO DATA 4.62E + 02 8.79E + 01 NO DATA 1.05E + 03 NO DATA 1.07E + 03 TC-99M 2.94E-02 8.32E - 02 1.06E - 00 NO DATA 1.26E + 00 4.07E - 02 4.92E + 01 TC-101 3.03E - 02 4.36E - 02 4.28E - 01 NO DATA 7.85E - 01 2.23E - 02 1.31E - 13 RU-103t 1.98E + 01 NO DATA 8.54E - 01 NO DATA 7.57E + 01 NO DATA 2.31E + 03 RU-105t 1.65E + 00 NO DATA 6.52E - 01 NO DATA 2.13E + 01 NO DATA 1.01E + 03 RU-106: 2.95E + 02 NO DATA 3.73E + 01 NO DATA 5.69E + 02 NO DATA 1.91E + 04 AG-110MWt 1.42E + 01 1.31E + 01 7.80E + 00 NO DATA 2.58E + 01 NO DATA 5.36E + 03 SB-124 2.40E + 02 4.53E + 00 9.50E + 01 5.81E-01 NO DATA 1.87E + 02 6.81E + 03 SB-125t 1.53E + 02 1.71E + 00 3.65E + 01 1.56E-01 NO DATA 1.18E + 02 1.69E + 03 SB-1 26 9.85E + 01 2.OOE + 00 3.55E + 01 6.03E-01 NO DATA 6.04E + 01 8.05E + 03 SB-1 27 2.21E + 01 4.84E-01 8.47E + 00 2.65E-01 NO DATA 1.31E + 01 5.05E + 03 TE-125M 2.79E + 03 1.01E + 03 3.74E + 02 8.39E + 02 1.13E + 04 NO DATA 1.11E + 04 TE-127M: 7.05E + 03 2.52E + 03 8.59E + 02 1.80E + 03 2.86E + 04 NO DATA 2.36E + 04 TE-127 1.14E + 02 4.11E + 01 2.48E + 01 8.48E + 01 4.66E + 02 NO DATA 9.03E + 03 TE-129Mt 1.20E + 04 4.47E + 03 1.89E + 03 4.11E + 03 5.OOE + 04 NO DATA 6.03E + 04 TE-129 3.27E + 01 1.23E + 01 7.96E + 00 2.51E + 01 1.37E + 02 NO DATA 2.47E + 01 TE-131Mt 1.88E + 03 8.81E + 02 7.34E + 02 1.39E + 01 8.92E + 03 NO DATA 8.74E + 04 TE-131: 2.05E + 01 8.57E + 00 6.47E + 00 1.69E + 01 8.98E + 01 NO DATA 2.90E + 00 TE-132t 2.62E + 03 1.70E + 03 1.59E + 03 1.87E + 03 1.63E + 04 NO DATA 8.02E + 04 1-130 9.01E + 01 2.66E + 02 1.05E + 02 2.25E + 04 4.15E + 02 NO DATA 2.29E + 02 1-131 t 4.96E + 02 7.09E + 02 4.06E + 02 2.32E + 05 1.22E + 02 NO DATA 1.87E + 02 1-132 2.42E + 01 6.47E + 01 2.26E + 01 2.26E + 03 1.03E + 02 NO DATA 1.22E + 011 1-1 33t 1.69E + 02 2.94E + 02 8.97E + 01 4.32E + 04 5.13E + 02 NO DATA 2.64E + 02 1-134 1.26E + 01 3.43E + 01 1.23E + 01 5.94E + 02 5.46E + 01 NO DATA 2.99E - 02 1-135t 5.28E + 01 1.38E + 02 5.10E + 01 9.11E + 03 2.22E + 02 NO DATA 1.56E + 02 CS-134 3.03E + 05 7.21E + 05 5.89E + 05 NO DATA 2.33E + 05 7.75E + 04 1.26E + 04 CS-136 3.17E + 04 1.25E + 05 9.01E + 04 NO DATA 6.97E + 04 9.55E + 03 1.42E + 04 CS-1 37T 3.88E + 05 5.31E + 05 3.48E + 05 NO DATA 1.88E + 05 5.99E + 04 1.03E + 04 CS-1 38 2.69E + 02 5.31E + 02 2.63E + 02 NO DATA 3.90E + 02 3.85E + 01 2.27E - 03 BA-1 39 9.OOE + 00 6.41 E - 03 2.64E - 01 NO DATA 5.99E - 03 3.64E - 03 1.60E + 01 BA-140t 1.88E + 03 2.37E + 00 1.23E + 02 NO DATA 8.05E - 01 1.35E + 00 3.88E + 03 BA-141t 4.27E + 00 3.30E - 03 1.48E - 01 NO DATA 3.07E - 03 1.87E - 03 2.06E - 09 BA-1421 1.98E + 00 2.03E - 03 1.24E - 01 NO DATA 1.72E - 03 1.15E - 03 2.78E - 18 LA-140 3.58E - 01 1.80E - 01 4.76E - 02 NO DATA NO DATA NO DATA 1.32E + 04 LA-142 1.83E - 02 8.33E - 03 2.07E - 03 NO DATA NO DATA NO DATA 6.08E + 01 CE-141 8.01E - 01 5.42E - 01 6.15E - 02 NO DATA 2.52E - 01 NO DATA 2.07E + 03 CE-143t 1.41 E - 01 1.04E + 02 1.16E - 02 NO DATA 4.60E - 02 NO DATA 3.90E + 03 CE-144: 4.18E + 01 1.77E + 01 2.24E + 00 NO DATA 1.04E + 01 NO DATA 1.41E + 04 PR-143 1.32E + 00 5.28E - 011 6.52E - 02 NO DATA 3.05E - 01 NO DATA 5.77E + 03 PR-144 4.31E - 03 1.79E - 03 2.19E - 04 NO DATA 1.01E - 03 NO DATA 6.19E - 10 ND-147t 9.OOE - 01 1.04E + 00 6.22E - 02 NO DATA 6.08E - 01 NO DATA 4.99E + 03 W-187 3.04E + 02 2.55E + 02 8.90E + 01 NO DATA NO DATA NO DATA 8.34E + 04 NP-239 1.28E - 01 1.25E - 02 6.91 E - 03 NO DATA 3.91 E - 02 NO DATA 2.57E + 03 ODCM, V. C. Summer, SCE&G: Revision 28 (July 2012 2.0-41

LIQUID RADWASTE TREATMENT SYSTEM FIGURE 2.2-1 ON em U

I~-j I-','

I~5I I~I ODCM, V. C. Summer, SCE&G: Revision 21 (March 1996) 2.0-42

FIGURE 2.2-1 NOTES: 1. Turbine Building Sump contents may be processed to the main condenser cleaning sump through a portable demineralizer. This is an optional treatment pathway which provides processing flexibility in the event processing through excess liquid waste is not desirable. Since a temporary demineralizer is used for this optional treatment pathway, operability tests specified in ODCM specification 1.1.4.1 are not required. To ensure adequacy of the RM-L8 setpoint while using the alternate process pathway, samples must be obtained from the discharge side of the demineralizers or condenser cleaning sump and analyzed every twelve hours.

ODCM, V. C. Summer, SCE&G: Revision 21 (March 1996) 2.0-43

3.0 GASEOUS EFFLUENT 3.1 Gaseous Effluent Monitor Setpoints The calculated setpoint values will be regarded as upper bounds for the actual setpoint adjustments. That is, setpoint adjustments are not required to be performed if the existing setpoint level corresponds to a lower count rate than the calculated value. Setpoints may be established at values lower than the calculated values, if desired.

Calculated monitor setpoints may be added to the ambient background count rate.

3.1.1 Gaseous Effluent Monitor Setpoint Calculation Parameters Term Definition Section of Initial Use Cv = count rate of a station vent monitor (3.1.2) corresponding to grab sample radionuclide concentrations, X1 v, as determined from the monitor's calibration curve, in cpm.

Cv = the count rate of the monitor on vent v (3.1.4) corresponding to Xv' uCi/cc of Xe-133, in cpm.

c = count rate of the gas decay system monitor for (3.1.3) measured radionuclide concentrations corrected to discharge pressure, in cpm.

c = the count rate of the waste gas decay system (3.1.4) monitor corresponding to the total noble gas concentration in cpm.

Dss = limiting dose rate to the skin (3.1.2)

(3000 mrem/year).

DTB = limiting dose rate to the total body (3.1.2)

(500 mrem/year).

Fv the flow rate in vent v (cc/sec) (3.1.2)

(1 cc/sec = 0.002119 cfm).

f = the maximum permissible waste gas discharge (3.1.3) rate, based on the actual radionuclide mix and skin dose rate (cc/sec).

the maximum permissible waste gas discharge (3.1.3) rate, based on the actual radionuclide mix and ODCM, V. C. Summer, SCE&G: Revision 17 (April 1993) 3.0-1

Term Definition Section of Initial Use total body dose rate (cc/sec).

f = the maximum permissible waste gas discharge (3.1.3) rate, the lesser of fs and ft (cc/sec).

f = the conservative maximum permissible waste (3.1.4) gas discharge rate based on Kr-89 skin dose rate (cc/sec).

ft = the conservative maximum permissible waste (3.1.4) gas discharge rate based on Kr-89 total body dose rate (cc/sec).

K = total body dose factor due to gamma emissions (3.1.2) from isotope i (mrem/year per uCi/m 3) from Table 3.1-1.

Kkr-89 total body dose factor for Kr-89, the most (3.1.3) restrictive isotope from Table 3.1-1 (mrem/yr per uCi/m 3).

L = Skin dose factor due to beta emissions from (3.1.2) isotope i (mrem/yr per uCi/m 3) from Table 3.-1-1.

LKr-89 = Skin dose factor for Kr-89, the most restrictive (3.1.3) isotope, from Table 3.1-1 (mrem/yr per uCi/m 3).

Mi= air dose factor due to gamma emissions from (3.1.2) isotope i (mrad/yr per uCi/m 3) from Table 3.1-1.

MKr-89 = air dose factor for Kr-89, the most restrictive (3.1.3) isotope, from Table 3.1-1 (mrad/yr per uCi/m 3).

Rs = count rate per mrem/yr to the skin. (3.1.2)

Rt = count rate per mrem/yr to the total body. (3.1.2)

Rs'= conservative count rate per mrem to the skin (Xe- (3.1.4) 133 detection, Kr-89 dose).

Rt'= conservative count rate per mrem to the total (3.1.4) body (Xe-1 33 detection, Kr-89 dose).

Term Definition Section of Initial Use Sd = count rate of the waste gas decay system noble (3.1.3)

ODCM, V. C. Summer, SCE&G: Revision 17 (April 1993) 3.0-2

Term Definition Section of Initial Use gas monitor at the alarm setpoint, in cpm.

S, count rate of a station vent noble gas monitor at (3.1.2) the alarm setpoint, in cpm.

Svc count rate of the containment purge noble gas (3.1.2) monitor at the alarm setpoint, in cpm.

SVP count rate of the plant vent noble gas monitor at (3.1.2) the alarm setpoint, in cpm.

Xid = the concentration of noble gas radionuclide i in a (3.1.3) waste gas decay tank, as corrected to the pressure of the discharge stream at the point of its flow measurement in uCi/cc.

Xiv the measured concentration of noble gas (3.1.2) radionuclide i in the last grab sample analyzed for vent v in uCi/cc.

Xd' = the total noble gas concentration in a waste gas (3.1.4) decay tank, as corrected to the pressure of the discharge stream at the point of its flow measurement in uCi/cc.

X = a concentration of Xe-133 chosen to be in the (3.1.4) operating range of the monitor on vent v in uCi/cc.

X = the highest annual average relative concentration (3.1.2) in any sector, at the site boundary in sec/m 3.

1.1 = mrem skin dose per mrad air dose (3.1.2) 0.25 the safety factor applied to each of the two vent (3.1.2) noble gas monitors (plant vent and containment purge) to assure that the sum of the releases has a combined safety factor of 0.5 which allows a 100 percent margin for cumulative uncertainties of measurements.

ODCM, V. C. Summer, SCE&G: Revision 17 (April 1993) 3.0-3

TABLE 3.1-1 DOSE FACTORS FOR EXPOSURE TO A SEMI-INFINITE CLOUD OF NOBLE GASES.*

Nuclide y-Body***(Ki) B-Skin***(Li. Y-Air- .il 1.3Air*+,0 Kr-85m 1.17E + 03**** 1.46E + 03 1.23E + 03 1.97E + 03 Kr-85 1.61 E + 01 1.34E + 03 1.72E + 01 1.95E + 03 Kr-87 5.92E + 03 9.73E + 03 6.17E + 03 1.03E + 04 Kr-88 1.47E + 04 2.37E + 03 1.52E + 04 2.93E + 03 Kr-89 1.66E + 04 1.01 E + 04 1.73E + 04 1.06E + 04 Kr-90 1.56E + 04 7.29E + 03 1.63E + 04 7.83E + 03 Xe-131m 9.15E + 01 4.76E + 02 1.56E + 02 1.11E + 03 Xe-1 33m 2.51E + 02 9.94E + 02 3.27E + 02 1.48E + 03 Xe-1 33 2.94E + 02 3.06E + 02 3.53E + 02 1.05E + 03 Xe-1 35m 3.12E + 03 7.11E + 02 3.36E + 03 7.39E + 02 Xe-1 35 1.81 E + 03 1.86E + 03 1.92E + 03 2.46E + 03 Xe-1 37 1.42E + 03 1.22E + 04 1.51 E + 03 1.27E + 04 Xe-138 8.83E + 03 4.13E + 03 9.21E + 03 4.75E + 03 Ar-41 8.84E + 03 2.69E + 03 9.30E + 03 3.28E + 03

Values taken from Reference 3, Table B-1

- mrad -M3 pCi - yr

mrem - m3 pCi - yr

- - 1.17E + 03 = 1.17 x 10 3 ODCM, V. C. Summer, SCE&G: Revision 17 (April 1993) 3.0-4

CO1-> 3.1.2 Station Vent Noble Gas Monitors (RM-A3 and RM-A4 For the purpose of implementation of section 1.2.1 of the ODCM, the alarm setpoint level for the station vent noble gas monitors will be calculated as follows:

S = count rate of the plant vent noble gas monitor (= Sv, for RM-A3) or the containment purge noble gas monitor (= Sv, for RM-A4) at the alarm setpoint level.

0.25 x Rt x DTB (34)

< the lesser of or 0.25 x R. x Dss (35) 0.25 the safety factor applied to each of the two vent noble gas monitors (plant vent and containment purge) to assure that the sum of the releases has a combined safety factor of 0.5 which allows a 100 percent margin for cumulative uncertainties of measurements.

DTB = Dose rate limit to the total body of an individual

= 500 mrem/yr R, = count rate per mrem/yr to the total body

= C,,/((X/Q) x F,, x Y- K, Xi,,) (36)

Dss = Dose rate limit to the skin of the body of an individual in an unrestricted area.

= 3000 mrem/year.

Rs count rate per mrem/yr to the skin.

C, + [X/Qx F,, x Y (L, + 1.1 M,) X,] (37)

Xiv = the measured concentration of noble gas radionuclide i in the last grab sample analyzed for vent v, pCi/ml. (For the plant vent, grab samples are taken at least ODCM, V. C. Summer, SCE&G: Revision 28 (July 2012) 3.0-5

monthly. For the 6" and 36" containment purge lines, the sample is taken just prior to the release and also monthly, ifthe release is continuous.)

F = the flow rate in vent v, cc/sec. (1 cc/sec = 0.002119 cfm)

Cv = count rate, (cpm) of the monitor on station vent v corresponding to grab sample noble gas concentrations, X1v, as determined from the monitor's calibration curve; i.e.,

product of the monitor response curve slope (cPm/uCi/ml) and the sum of the noble gas concentrations in the grab sample (uCi/ml). (Initial calibration curves of the type shown in Figure 2.1-1 have been determined conservatively from families of response curves supplied by the monitor manufacturers. As releases occur, a historical correlation will be prepared and placed in service when sufficient data are accumulated.)

X/Q = the highest annual average relative concentration in any sector, at the site boundary (seven year average).

6.3E-6 sec/m 3 in the ENE sector.

K = total body dose factor due to gamma emissions from isotope i (mrem/yr per 1iCi/m 3) from Table 3.1-1.

L = skin dose factor due to beta emissions from isotope i (mrem/yr per ýiCi/m 3) from Table 3.1-1.

1.1 = mrem skin dose per mrad air dose.

M = air dose factor due to gamma emissions from isotope i (mrad/yr per pCi/i 3) from Table 3.1-1.

ODCM, V. C. Summer, SCE&G: Revision 28 (July 2012) 3.0-6

NOTE: At plant startups when no grab sample analysis is available for the continuous releases, the Alternate Methodology of Section 3.1.4 must be used.

3.1.3 Waste Gas Decay System Monitor (RM-A10)

The permissible conditions for discharge through the waste gas decay system monitor (RM-A10) will be calculated in a manner similar to that for the plant vent noble gas monitor. In the case of the waste gas system, however, the discharge flow rate is continuously controllable by valve HCV-014 and permissible release conditions are therefore defined in terms of both flow rate and concentration.

Therefore, RM-A10 is used only to insure that a representative sample was obtained.

For operational convenience, (to prevent spurious alarms due to fluctuations in background) the setpoint level will be established at 1.5 times the measured waste concentration.

The maximum permissible flow rate will be set on the same basis but include the engineering safety factor of 0.5. The RM-A10 setpoint level Sd is defined as:

Sd _*1.5c (38) where:

c = count rate in CPM of the waste gas decay system monitor corresponding to the measured concentration (taken from the monitor calibration curves).

The maximum permissible waste gas flow rate f, (cc/sec) is calculated from the maximum permissible dose rates at the site boundary according to:

fw < the lesser of f, or fs (39)

ODCM, V. C. Summer, SCE&G: Revision 13 (June 1990) 3.0-7

= the maximum permissible discharge rate based on total body dose rate.

0. 2 5xDTB / [X/Q x 1.5 Z Xd K,] (40) f, the maximum permissible discharge rate based on skin dose rate.

= 0.25xDss/[X/Qxl15 X,d(L,+1.1M,)] (41)

Xid = the concentration of noble gas radionuclide i in the waste gas decay tank whose contents are to be discharged, as corrected to the pressure of the discharge stream at the point of the flow rate measurement. The maximum discharge pressure as governed by the diaphragm valve, 7896, is 30 psia.

NOTE: The factor of 1.5 in the denominators of equations (40) and (41) places fw on the same basis as Sdo When a discharge is to be conducted, valve HCV-014 is to be opened until (a) the waste gas discharge flow rate reaches 0.9 x f, or (b) the count rate of the plant vent noble gas monitor RM-A3 approaches its setpoint, whichever of the above conditions is reached first.

When no discharges are being made from the Waste Gas Decay System, the RM-A10 setpoint should be established as near background as practical to prevent spurious alarms and yet alarm in the event of an inadvertent release.

3.1.4 Alternative Methodology for Establishing Conservative Setpoints As an alternate to the methodology of section 3.1.2, to minimize necessity for frequent adjustment of setpoint, a conservative setpoint may be calculated as follows:

For a plant vent:

Rt'= conservative count rate per mrem/yr to the total body (Xe-1 33 detection, Kr-89 dose).

ODCM, V. C. Summer, SCE&G: Revision 18 (September 1994) 3.0-8

= C' + [X/Q x Kkr-89 x X,' x F],

(42) where:

X,' = a concentration of Xe-133 chosen to be in the operating range of the monitor on vent v, pCi/cc.

C' = the count rate in CPM of the monitor on vent v corresponding to Xv,'

[tCi/cc of XE-133.

KKr-89 = total body dose factor for Kr-89, the most restrictive isotope from Table 3.1-1.

R, = count rate per mrem/yr to the skin.

= Or'÷[X/Q x(LKr89 + 1.1Mkr8 9 ) X Xv' X Fv] (43) where:

LKr-89 = skin dose factor for Kr-89, the most restrictive isotope from Table 3.1-1.

MKr89 = air dose factor for Kr-89, the most restrictive isotope, from Table 3.1-1.

For the waste gas decay system:

t' = the conservative maximum permissible discharge rate based on Kr-89 total body dose rate.

= 0.25XDTB+[X/Q xl.5xXd' Kkr89] (44) fs = the conservative maximum permissible discharge rate based on Kr-89 skin dose rate.

0.25 x Dss[ X / Q x 1.5 X Xd' x (LKr8 9 + 1.1MKr-89)] (45)

ODCM, V. C. Summer, SCE&G: Revision 13 (June 1990) 3.0-9

Xd' = the total concentration of noble gas radionuclides in the waste gas decay tank whose contents are to be discharged, as corrected to the pressure of the discharge stream at the point of the flow measurement.

c' = count rate in cpm of the waste gas decay system monitor corresponding to Xd' gCi/cc of Kr-85.

3.1.5 Oil Incineration 3.1.5.1 Oil incinerator was removed from service in 2010 and the structure decommissioned under Engineering Change Request 71498.

3.1.6 MeteoroloQical Release Criteria for Batch Releases Planned gaseous batch releases (WGDT) and will be performed during favorable meteorology. Limiting releases to favorable meteorology provides assurance that release conditions will be conservative with respect to annual average dispersion values, (X/Q, X/Q') . Favorable meteorology is defined in Table 3.1-2.

ODCM, V. C. Summer, SCE&G: Revision 28 (July 2012) 3.0-10

Table 3.1-2 Favorable Meteorology Differential Temperature (AT) 1 Stability Wind Speed2 (mph)

OF Class 61m- 10m 40m- 10m lo1n 61m AT:< -1.74 AT* -1.03 A * *

-1.74 < AT* -1.56 -1.03 < AT <-0.92 B * *

-1.56 < AT < -1.38 -0.92 < AT <-0.81 C 1.3 1.6

-1.38 < AT <-0.46 -0.81 < AT <-0.27 D 3.1 4.1

-0.46 < AT <1.38 -0.27 < AT:< 0.81 E 3.5 6.6 1.38 < AT< 3.67 0.81 < AT< 2.16 F 5.2 14.0 3.67 < AT 2.16 < AT G 7.0 18.9 Notes:

1 The AT values for 61m 10m are considered as primary indicators for determination of stability class. The 40m - 10M AT values are used only when 61mn - 1Om values are not available. All AT values are listed in OF and are based on values in USNRC Regulatory Guide 1.23.

2 The 1Om wind speed is considered the primary indication for windspeed. The 61mn wind speed indication should only be used if 10m is not available.

No wind is required for planned releases.

ODCM, V. C. Summer, SCE&G: Revision 24 (May 2006) 3.0-1 OA

Figure 3.1-1 Example Noble Gas Monitor Calibration Curve 10- 1 C

0 n

c e 10-2 n

t r

a 10-3 t

i 0

n io-4 (uCi/ml) 10"$

1o-6 10 10, 104 105 ODCM, V. C. Summer, SCE&G: Revision 13 (June 1990) 3.0-11

3.2 Dose Calculation for Gaseous Effluent 3.2.1 Gaseous Effluent Dose Calculation Parameters Term Definition Section of Initial Use Do average organ dose rate in the current year (3.2.2.2)

(mrem/yr).

Dp dose to an individual from radioiodine and (3.2.3.2) radionuclides in particulate form and radionuclides (other than noble gases), with half-lives greater than eight days (mrem).

Ds average skin dose rate in current year (3.2.2.1)

(mrem/year).

Dt current total body dose rate (mrem/yr) (3.2.2.1)

D = air dose due to beta emissions from noble gas (3.2.3.1) radionuclides (mrad).

D, air dose due to gamma emissions from noble gas (3.2.3.1) radionuclides (mrad).

Ki total body dose factor due to gamma emissions (3.2.2.1) from isotope i (mrem/year per uCim 3) from Table 3.1-1.

Li skin dose factor due to beta emissions from (3.2.2.1) noble gas radionuclide i (mrad/yr per jiCi/m3) from Table 3.1-1.

M = air dose factor due to gamma emissions from (3.2.2.1) noble gas radionuclide i (mrad/yr per [iCi/m 3) from Table 3.1-1.

Ni air dose factor due to beta emissions from noble (3.2.3.1) gas radionuclide i (mrad per iiCi/m3) from Table 3.1-1.

Pi dose parameter for radionuclide i, (mrem/yr per (3.2.2.2) p.Ci/m 3) for inhalation, from Table 3.2-1.

, the release rate of noble gas radionuclide i as (3.2.2.1) determined from the concentrations measured in the analysis of the appropriate sample required by Table 1.2-3 (iiCi/sec).

ODCM, V. C. Summer, SCE&G: Revision 15 (February 1991) 3.0-12

Section of Term Definition Initial Use

= the release of non-noble gas radionuclide i as (3.2.2.2) determined from the concentrations measured in the analysis of the appropriate sample required by Table 1.2-3 (lICi/sec).

= cumulative release of noble gas radionuclide i (3.2.3.1) over the period of interest (1iCi).

Q' = cumulative release of non-noble gas radionuclide (3.2.3.2) i (required by ODCM Specification 1.2.4.1) over the period of interest (lCi).

R = dose factor for radionuclide i and pathway j, (3.2.3.2)

(mrem/yr per uCi/m 3) or (m 2-mrem/yr per pCi/sec) from Tables 3.2-2 through 3.2-6.

= relative dispersion parameter for the maximum (3.2.3.2) exposed individual, as appropriate for his exposure pathway j and radionuclide i.

X / Q'for inhalation and all tritium / carbon-14 pathways D / Q' for other pathways and non-tritium radionuclides X/ Q = the highest annual average relative concentration (3.2.2.1) in any sector, at the site boundary in sec/m 3.

3.17 x 10i= the fraction of one year per one second (3.2.3.1)

X /Q Annual average relative concentration for the (3.2.3.2) location of the maximum exposed individual for the site (sec/m3).

D/Q Annual average relative deposition for the (3.2.3.2) location of the maximum exposed individual for the site (m-2).

ODCM, V. C. Summer, SCE&G: Revision 28 (July 2012) 3.0-13

3.2.2 Unrestricted Area Boundary Dose 3.2.2.1 For the purpose of implementation of section 1.2.2.1a, (ý 500 mrem/year - total body, < 3000 mrem/year - skin) the dose at the unrestricted area boundary due to noble gases shall be calculated as follows:

Dt= current total body dose rate (mrem/yr)

= X/Q K, Q, (46)

Ds= current skin dose rate (mrem/yr)

- X/Q (L + 1.1M,) Qi (47) where:

Q = the release rate of noble gas radionuclide i as determined from the concentration measured in the analysis of the appropriate sample required by Table 1.2-3 (pCi/sec.).

X/Q = the highest annual average relative concentration in any sector, at the site boundary (for value, see Section 3.1.2).

Ki, Li, and Mi will be selected for the appropriate radionuclide from Table 3.1-1.

3.2.2.2 For the purpose of implementation of section 1.2.2.1.b (ý 1500 mrem/yr - any organ) organ doses due to radioiodines and all radioactive materials in particulate form and radionuclides (other than noble gases) with half-lives greater than eight days, will be calculated as follows:

Do= current organ dose rate (mrem/yr)

= XX/Q P,Q(' (48) where:

ODCM, V. C. Summer, SCE&G: Revision 13 (June 1990) 3.0-14

X/Q = the highest annual average relative concentration in any sector, at the site boundary (for value, see Section 3.1.2)

Pi = dose parameter for radionuclide i, (mrem/yr per itCi/m3 ) for inhalation, from Table 3.2-1.

Q ' = the release rate of non-noble gas radionuclide i as determined from the concentrations measured in the analysis of the appropriate sample required by Table 1.2-3 (jiCi/sec.).

3.2.3 Unrestricted Area Dose (Air Dose and Dose to Individual) 3.2.3.1 For the purpose of implementation of section 1.2.3.1 (Calendar quarter: <5 mrad - y and < 10 mrad - 0, Calendar year: < 10 mrad - y and < 20 mrad - P) and section 1.2.5.1 (air dose averaged over 31 days: < 0.2 mrad - y and < 0.4 mrad - 3), the air dose in unrestricted areas shall be determined as follows:

Dy = air dose due to gamma emissions from noble gas radionuclide i (mrad)

= 3.17x10- 1 M, X/QQ0 (49) where:

3.17 x 10-8 = the fraction of one year per one second Q, = cumulative release of noble gas radionuclide i over the period of interest ([.Ci).

ODCM, V. C. Summer, SCE&G: Revision 16 (September 1991) 3.0-15

D = air dose due to beta emissions from noble gas radionuclide i (mrad).

= 3.17 x 10-8 Y N, X/ Q (50) where, Ni = air dose factor due to beta emission from noble gas radionuclide i (mrad/yr per uCi/m 3) from Table 3.1-1.

3.2.3.2 For all gaseous effluents, dose to an individual from radioiodines and radioactive materials in particulate form and radionuclides (other than noble gases), with half-lives greater than eight (8) days (Calendar quarter: < 7.5 mrem any organ, Calendar year:

< 15 mrem any organ) will be calculated for the purpose of implementation of section 1.2.4.1 as follows:

DP= dose to an individual from radioiodines and radionuclides in particulate form, with half-lives greater than eight days (mrem)

F - 1

= 3.17 x 10-8 E Rj WuJQ (51) where:

Ii relative concentration or relative deposition for the maximum exposed individual, as appropriate for exposure pathway j and radionuclide i.

X / Q' for inhalation and all tritium / carbon-14 pathways 3

= 3.5 x 10-6sec/m DIQ' for other pathways excluding tritium and carbon-14 8

= 1.1x10- m-2 (See the notes to Table 3.2-7 and 3.2-8 for the origin of these factors.)

ODCM, V. C. Summer, SCE&G: Revision 28 (July 2012) 3.0-16

Aj = dose factor for radionuclide i and pathway j, (mrem/yr per pCi/m 3 ) or (M 2 - mrem/yr per liCi/sec) from Table 3.2-2.

=Cumulative release of non-noble gas radionuclide i (required by ODCM Specification 1.2.4.1) over the period of interest (jiCi).

3.2.4 For the purpose of initial assessments of the impact of unplanned gaseous releases, dose calculations for the critical receptor in each affected sector may be performed using section 3.2.3.1 and section 3.2.3.2 equations as follows:

(1) For each affected sector, X/Q and D/Q will be calculated for one mile and critical receptor locations using actual meteorological conditions occurring during the unplanned release. Actual X/Q and D/Q values will be compared to annual average dispersion coefficients (X/Q, X/Q', and X/Q'). The more limiting dispersion coefficients will be used along with methodology in sections 3.2.3.1 and 3.2.3.2 for the initial assessment.

(2) The location of the critical receptors and the pathways j which should be analyzed are shown in Table 3.2-7. (For very rough calculations, the annual average dispersion coefficients (X / Q and D / Q) for each receptor are shown in Table 3.2-8.)

(3) The Rij for the appropriate exposure pathways and age groups will be selected from Tables 3.2-3 through 3.2-6.

ODCM, V. C. Summer, SCE&G: Revision 16 (September 1991) 3.0-17

TABLE 3.2-1 PATHWAY DOSE FACTORS FOR SECTION 3.2.2.2 (PI)*

Page 1 of 3 AGE GROUP (CHILD)

ISOTOPE INHALATION H-3 1.125E + 03 C-14 3.589E + 04 NA-24 1.610E + 04 P-32 2.605E + 06 CR-51 1.698E + 04 MN-54 1.576E + 06 MN-56 1.232E + 05 FE-55 1.110E + 05 FE-59 1.269E + 06 CO-58 1.106E + 06 CO-60 7.067E + 06 NI-63 8.214E + 05 NI-65 8.399E + 04 CU-64 3.670E + 04 ZN-65 9.953E + 05 ZN-69 1.018E + 04 BR-83 4.736E + 02 BR-84 5.476E + 02 BR-85 2.531E + 01 RB-86 1.983E + 05 RB-88 5.624E + 02 RB-89 3.452E + 02 SR-89 2.157E + 06 SR-90 1.010E + 08 SR-91 1.739E + 05

See note, page 3.0-20 3

Units - mrem/yr per iiCi/m ODCM, V. C. Summer, SCE&G: Revision 13 (June 1990) 3.0-18

TABLE 3.2-1 PATHWAY DOSE FACTORS FOR SECTION 3.2.2.2 (Pi)

Page 2 of 3 AGE GROUP (CHILD)

ISOTOPE INHALATION SR-92 2.424E + 05 Y-90 2.679E + 05 Y-91M 2.812E + 03 Y-91 2.627E + 06 Y-92 2.390E + 05 Y-93 3.885E + 05 ZR-95 2.231E + 06 ZR-97 3.511E + 05 NB-95 6.142E + 05 MO-99 1.354E + 05 TC-99M 4.810E + 03 TC-101 5.846E + 02 RU-103 6.623E + 05 RU-105 9.953E + 04 RU-106 1.476E + 07 AG-110M 5.476E + 06 TE-125M 4.773E + 05 TE-127M 1.480E + 06 TE-127 5.624E + 04 TE-129M 1.761E + 06 TE-129 2.549E + 04 TE-131M 3.078E + 05 TE-131 2.054E + 03 TE-132 3.774E + 05 1-130 1.846E + 06

See note, page 3.0-20 3

Units - mrem/yr per ViCi/m ODCM, V. C. Summer, SCE&G: Revision 13 (June 1990) 3.0-19

TABLE 3.2-1 PATHWAY DOSE FACTORS FOR SECTION 3.2.2.2 (Pi)

Page 3 of 3 AGE GROUP (CHILD)

ISOTOPE INHALATION 1-131 1.624E + 07 1-132 1.935E + 05 1-133 3.848E + 06 1-134 5.069E + 04 1-135 7.918E + 05 CS-134 1.014E + 06 CS-136 1.709E + 05 CS-137 9.065E + 05 CS-1 38 8.399E + 02 BA-139 5.772E + 04 BA-140 1.743E + 06 BA-141 2.919E + 03 BA-142 1.643E + 03 LA-140 2.257E + 05 LA-142 7.585E + 04 CE-141 5.439E + 05 CE-143 1.273E + 05 CE-144 1.195E + 07 PR-143 4.329E + 05 PR-144 1.565E + 03 ND-147 3.282E + 05 W-187 9.102E + 04 NP-239 6.401E + 04 NOTE: The Pi values of Table 3.2-1 were calculated according to the methods of Reference 1, Section 5.2.1, for children. The values used for the various parameters and the origins of those values are given in Table 3.2-9 and its notes.

3 Units - mrem/yr per jiCi/m ODCM, V. C. Summer, SCE&G: Revision 13 (June 1990) 3.0-20

TABLE 3.2-2 PATHWAY DOSE FACTORS FOR SECTION 3.2.3.2 (R1 )*

Page 1 of 3 AGE GROUP (CHILDI IN.A.1 (CHILDI ISOTOPE INHALATION GROUND PLANE VEGETATION H-3 1.125E+03 (Total Body) 0.OOOE+00 (Skin) 3.627E+03 (Total Body)

C-14 3.589E+04 (Bone) 0.OOOE+00 (Skin) 7.OOOE+05 (Bone)

NA-24 1.610E+04(Total Body) 3.33E+08 (Skin) 3.729E+05(Total Body)

P-32 2.605E+06 (Bone) O.00OE+00 (Skin) 3.366E+09 (Bone)

CR-51 1.698E+04 (Lung) 5.506E+06 (Skin) 6.213E+06 (GI-LLI)

MN-54 1.576E+06 (Lung) 1.625E+09 (Skin) 6.648E+08 (Liver)

MN-56 1.232E+05 (GI-LLI) 1.068E+06 (Skin) 2.723E+03 (GI-LLI)

FE-55 1.11OE+05 (Lung) 0.OOOE+00 (Skin) 8.012E+08 (Bone)

FE-59 1.269E+06 (Lung) 3.204E+08 (Skin) 6.693E+08 (GI-LLI)

CO-58 1.106E+06 (Lung) 4.464E+08 (Skin) 3.771E+08 (GI-LLI)

CO-60 7.067E+06 (Lung) 2.532E+10 (Skin) 2.095E*09 (GI-LLI)

NI-63 8.214E+05 (Bone) 0.OOOE+00 (Skin) 3.949E+10 (Bone)

NI-65 8.399E+04 (GI-LLI) 3.451E+05 (Skin) 1.211E+03 (GI-LLI)

CU-64 3.670E+04 (GI-LLI) 6.876E+05 (Skin) 5.159E+05 (GI-LLI)

ZN-65 9.953E+05 (Lung) 8.583E+08 (Skin) 2.164E+09 (Liver)

ZN-69 1.018E+04 (GI-LLI) 0.OOE+00 (Skin) 9.893E-04 (GI-LLI)

BR-83 4.736E+02(Total Body) 7.079E+03 (Skin) 5.369E+00(Total Body)

BR-84 5.476E+02(Total Body) 2.363E+05 (Skin) 3.822E - 11 (Total Body)

BR-85 2.531E+01 (Total Body) 0.OOOE+00 (Skin) 0.000E+00jTotal Body)

RB-86 1.983E+05 (Liver) 1.035E+07 (Skin) 4.584E+08 (Liver)

RB-88 5.624E+02 (Liver) 3.779E+04 (Skin) 4.374E - 22 (Liver)

RB-89 3.452E+02 (Liver) 1.452E+05 (Skin) 1.642E - 26 (Liver)

SR-89 2.157E+06 (Lung) 2.509E+04 (Skin) 3.593E+10 (Bone)

SR-90 1.010E+08 (Bone) 0.OOOE+00 (Skin) 1.243E+12 (Bone)

SR-91 1.739E+ 05 (GI-LLI) 2.511E+06 (Skin) 1.157E+06 (GI-LLI)

See note, page 3.0-36

- Reference 1, section 5.3.1, page 30, paragraph 1 explains the logic used in selecting these specific pathways.

Critical organs for each pathway by nuclide in parentheses.

Units -

Inhalation and all tritium / Carbon 14 - mrem/yr per ipCi/m 3 Other pathways for all other radionuclides - M2 e mrem/yr per ýtCi/sec ODCM, V. C. Summer, SCE&G: Revision 28 (July 2012) 3.0-21

TABLE 3.2-2 (continued)

PATHWAY DOSE FACTORS FOR SECTION 3.2.3.2 (R1 )

Page 2 of 3 AGE GROUP (CHILD) (N.A.) (CHILD)

ISOTOPE INHALATION GROUND PLANE VEGETATION SR - 92 2.424E+05 (GI-LLI) 8.631 E+05 (Skin) 1.378E+04 (GI-LLI)

Y - 90 2.679E+05 (GI-LLI) 5.308E+03 (Skin) 6.569E+07 (GI-LLI)

Y - 91M 2.812E+03 (Lung) 1.161E+05 (Skin) 1.737E - 05 (GI-LLI)

Y - 91 2.627E+06 (Lung) 1.207E+06 (Skin) 2.484E+09 (GI-LLI)

Y - 92 2.390E+05 (GI-LLI) 2.142E+05 (Skin) 4.576E+04 (GI-LLI)

Y - 93 3.885E+05 (GI-LLI) 2.534E+05 (Skin) 4.482E+06 (GI-LLI)

ZR- 95 2.231E+06 (Lung) 2.837E+08 (Skin) 8.843E+08 (GI-LLI)

ZR - 97 3.511E+05 (GI-LLI) 3.445E+06 (Skin) 1.248E+07 (GI-LLI)

NB- 95 6.142E+05 (Lung) 1.605E+08 (Skin) 2.949E+08 (GI-LLI)

MO - 99 1.354E+05 (Lung) 4.626E+06 (Skin) 1.647E+07 (Kidney)

TC - 99M 4.810E+03 (GI-LLI) 2.109E+05 (Skin) 5.255E+03 (GI-LLI)

TC - 101 5.846E+02 (Lung) 2.277E+04 (Skin) 4.123E- 29 (Kidney)

RU - 103 6.623E+05 (Lung) 1.265E+08 (Skin) 3.971 E+08 (GI-LLI)

RU - 105 9.953E+04 (GI-LLI) 7.212E+05 (Skin) 5.981 E+04 (GI-LLI)

RU - 106 1.476E+07 (Lung) 5.049E+08 (Skin) 1.159E+10 (GI-LLI)

AG - 110M 5.476E+06 (Lung) 4.019E+09 (Skin) 2.581 E+09 (GI-LLI)

TE - 125M 4.773E+05 (Lung) 2.128E+06 (Skin) 3.506E+08 (Bone)

TE - 127M 1.480E+06 (Lung) 1.083E+05 (Skin) 3.769E+09 (Kidney)

TE - 127 5.624E+04 (GI-LLI) 3.293E+03 (Skin) 3.903E+05 (GI-LLI)

TE - 129M 1.761 E+06 (Lung) 2.312E+07 (Skin) 2.430E+09 (GI-LLI)

TE - 129 2.549E+04 (GI-LLI) 3.076E+04 (Skin) 7.200E - 02 (GI-LLI)

TE - 131M 3.078E+05 (GI-LLI) 9.459E+06 (Skin) 2.163E+07 (GI-LLI)

TE- 131 2.054E+03 (Lung) 3.450E+07 (Skin) 1.349E - 14 (GI-LLI)

TE - 132 3.774E+05 (Lung) 4.968E+06 (Skin) 3.111 E+07 (GI-LLI)

I - 130 1.846E+06 (Thyroid) 6.692E+06 (Skin) 1.371 E+08 (Thyroid)

Units - 3 Inhalation and all tritium / Carbon 14 - mrem/yr per gCi/m Other pathways for all other radionuclides -mr a mrem/yr per [iCi/sec ODCM, V. C. Summer, SCE&G: Revision 28 (July 2012) 3.0-22

TABLE 3.2-2 (continued)

PATHWAY DOSE FACTORS FOR SECTION 3.2.3.2 (R1 )

Page 3 of 3 AGE GROUP (CHILD) (N.A.) (CHILD)

ISOTOPE INHALATION GROUND PLANE VEGETATION 1-131 1.624E+07 (Thyroid) 2.089E+07 (Skin) 4.754E+10 (l-hyroid) 1-132 1.935E+05 (Thyroid) 1.452E+06 (Skin) 7.314E+03 (Thyroid) 1-133 3.848E+06 (hyroid) 2.981 E+06 (Skin) 8.113E+08 (Thyroid) 1-134 5.069E+04 (Thyroid) 5.305E+05 (Skin) 6.622E - 03 (Thyroid) 1-135 7.918E+05 (Thyroid) 2.947E+06 (Skin) 9.973E+06 (Thyroid)

CS-134 1.014E+06 (Liver) 8.007E+09 (Skin) 2.631E+10 (Liver)

CS-1 36 1.709E+05 (Liver) 1.710E+08 (Skin) 2.247E+08 (Liver)

CS-137 9.065E+05 (Bone) 1.201 E+10 (Skin) 2.392E+ 10 (Bone)

CS-138 8.399E+02 (Liver) 4.102E+05 (Skin) 9.133E - 11 (Liver)

BA-139 5.772E+04 (GI-LLI) 1.194E+05 (Skin) 2.950E+00 (GI-LLI)

BA-140 1.743E+06 (Lung) 2.346E+07 (Skin) 2.767E+08 (Bone)

BA-141 2.919E+03 (Lung) 4.734E+04 (Skin) 1.605E - 21 (Bone)

BA-142 1.643E+03 (Lung) 5.064E+04 (Skin) 4.105E - 39 (Bone)

LA-140 2.257E+05 (GI-LLI) 2.180E+07 (Skin) 3.166E+07 (GI-LLI)

LA-142 7.585E+04 (Lung) 9.117E+05 (Skin) 2.141E+01 (GI-LLI)

CE-141 5.439E+05 (Lung) 1.540E+07 (Skin) 4.082E+08 (GI-LLI)

CE-143 1.273E+05 (GI-LLI) 2.627E+06 (Skin) 1.364E+07 (GI-LLI)

CE-144 1.195E+07 (Lung) 8.042E+07 (Skin) 1.039E+10 (GI-LLI)

PR-143 4.329E+05 (Lung) 0.OOOE+00 (Skin) 1.575E+08 (GI-LLI)

PR-144 1.565E+03 (Lung) 2.112E+03 (Skin) 3.829E- 23 (GI-LLI)

ND-147 3.282E+05 (Lung) 1.009E+07 (Skin) 9.197E+07 (GI-LLI)

W-187 9.102E+04 (GI-LLI) 2.740E+06 (Skin) 5.380E+06 (GI-LLI)

NP-239 6.401E+04 (GI-LLI) 1.976E+06 (Skin) 1.357E+07 (GI-LLI)

Units - 3 Inhalation and all tritium / Carbon 14 - mrem/yr per lCi/m Other pathways for all other radionuclides -m e mrem/yr per [tCi/sec ODCM, V. C. Summer, SCE&G: Revision 28 (July 2012) 3.0-23

TABLE 3.2-3 PATHWAY DOSE FACTORS FOR SECTION 3.2.4 (R1 )*

Page 1 of 3 AGE GROUP (INFANT) (N.A.) (INFANT) (INFANT) (INFANT) (INFANT) (INFANT) (INFANT)

ISOTOPE INHALATION GOUND GRS/COWI GRS/COW/ GRS/COWI GRS/GOT/ GRSIGOT/ VEGETATION PLANE MILK MEAT MILK MEAT MILK H-3 6.468E+02 0.000E+00 2.157E+03 0.OOOE+00 2.157E+03 0.OOOE+00 4.398E+03 0.OOOE+00 C-14 2.646E+04 0.OOOE+00 6.445E+05 0.000E+00 6.445E+05 0.OOOE+00 6.445E+05 0.OOOE+00 NA-24 1.056E+04 1.385E+07 1.542E+07 0.OOOE+00 2.300E-37 0.OOOE+00 1.851E+06 O.OOOE+00 P-32 2.030E+06 0.OOOE+00 1.602E+11 0.000E+00 7.088E+08 O.OOOE+00 1.924E+11 O.OOOE+00 CR-51 1.284E+04 5.506E+06 4.700E+06 0.000E+00 1.729E+05 O.OOOE+00 5.641E+05 O.OOOE+00 MN-54 9.996E+05 1.625E+09 3.900E+07 O.OOOE+00 1.118E+07 O.OOOE+00 4.680E+06 0.OOOE+00 MN-56 7.168E+04 1.068E+06 2.862E+00 O.OOOE+00 0.OOOE+00 0.OOOE+00 3.436E - 01 O.OOOE+00 FE-55 8.694E+04 O.OOOE+00 1.351E+08 O.OOOE+00 4.439E+07 O.OOOE+00 1.757E+06 O.OOOE+00 FE-59 1.015E+06 3.204E+08 3.919E+08 O.OOOE+00 3.384E+07 O.OOOE+00 5.096E+06 O.000E+00 CO-58 7.770E+05 4.464E+08 6.055E+07 O.OOOE+00 8.824E+06 O.OOOE+00 7.251 E+06 O.OOOE+00 CO-60 4.508E+06 2.532E+10 2.098E+08 O.OOOE+00 7.107E+07 O.OOOE+00 2.517E+07 O.OOOE+00 NI-63 3.388E+05 O.OOOE+00 3.493E+10 O.OOOE+00 1.221E+10 O.OOOE+00 4.192E+09 0.000E+00 NI-65 5.012E+04 3.451E+05 3.020E+01 O.OOOE+00 O.OOOE+00 0.OOOE+00 3.635E+00 0.000E+00 CU-64 1.498E+04 6.876E+05 3.807E+06 O.OOOE+00 7.934E-46 0.OOOE+00 4.246E+05 O.000E+00 ZN-65 6.468E+05 8.583E+08 1.904E+10 O.OOOE+00 5.160E+09 0.OOOE+00 2.285E+09 0.OOOE+00 ZN-69 1.322E+04 O.OOOE+00 3.855E-09 O.OOOE+00 O.OOOE+00 O.OOOE+00 3.581E - 10 0.OOOE+00 BR-83 3.808E+02 7.079E+03 9.339E-01 O.OOOE+00 O.000EO+00 0.000E+00 1.124E - 01 O.OOOE+00 BR-84 4.004E+02 2.363E+05 1.256E-22 O.OOOE+00 0.000E+00 0.OOOE+00 1.527E - 23 0.OOOE+00 BR-85 2.044E+01 0.000E+00 0.00E+00 0.OOOE+00 0.OOOE+00 O.OOOE+00 0.OOOEOO O.OOOE+00 RB-86 1.904E+05 1.035E+07 2.234E+10 0.OOOE+00 2.827E+08 0.OOOE+00 2.671E+09 0.OOOE+00 RB-88 5.572E+02 3.779E+04 1.874E-44 O.OOOE+00 0.OOOE+00 O.OOOE+00 2.304E - 45 0.OOOE+00 RB-89 3.206E+02 1.452E+05 3.414E-52 0.000E+00 0.OOOE+00 0.OOOE+00 4.056E - 53 O.OOOE+00 SR-89 2.030E+06 2.509E+04 1.258E+10 0.OOOE+00 1.280E+09 0.OOOE+00 2.643E+10 0.OOOE+00 SR-90 4.088E+07 0.000E+00 1.216E+11 0.OOOE+00 4.230E+10 0.OOOE+00 2.553E+11 0.OOOE+00 SR-91 7.336E+04 2.511E+06 3.215E+05 0.OOOE+00 0.OOOE+00 O.OOOE+00 6.758E+05 0.000E-+00 (PASTURE) I (PASTURE) (FEED) (PASTURE)l (PASTURE)

See note, page 3.0-36 Units - 3 Inhalation and all tritium / Carbon 14 - mrem/Yr per .Ci/m Other pathways for all other radionuclides -m

mrem/yr per jiCi/sec ODCM, V. C. Summer, SCE&G: Revision 28 (July 2012) 3.0-24

TABLE 3.2-3 (continued)

PATHWAY DOSE FACTORS FOR SECTION 3.2.4 (R1 )*

Page 2 of 3 AGE GROUP I IINFANTI (INFAN'n (INFANT) 1INFANT~ I 1INFANT~ (INFANfl (INFANT*

ISOTOPE INHALATION GROUND GRS/COW/ GRS/COWI GRS/COW/ GRS/GOT/ GRSIGOTI VEGETATION PLANE MILK MEAT MILK MEAT MILK SR - 92 1.400E+05 8.631 E+05 5.005E+01 O.OOOE+00 0.OOOE+00 O.OOOE+00 1.054E+02 O.OOOE+00 Y - 90 2.688E+05 5.308E+03 9.406E+05 O.OOOE+00 2.335E-05 0.OOOE+00 1.129E+05 0.OOOE+00 Y - 91M 2.786E+03 1.161E+05 1.876E-15 O.000E+00 O.OOOE+00 0.OOOE+00 2.290E - 16 0.OOOE+00 Y - 91 2.450E+06 1.207E+06 5.251 E+06 O.OOOE+00 6.324E+05 0.OOOE+00 6.302E+05 O.OOOE+00 Y -92 1.266E+05 2.142E+05 1.026E+01 0.OOOE+00 0.OOOE+00 O.000E+00 1.234E+00 0.OOOE+00 Y - 93 1.666E+05 2.534E+05 1.776E+04 0.OOOE+00 2.386E-61 0.OOOE+00 2.046E+03 0.OOOE+00 ZR -95 1.750E+06 2.837E+08 8.257E+05 O.000E+00 1.090E+05 0.OOOE+00 9.910E+04 0.OOOE+00 ZR - 97 1.400E+05 3.445E+06 4.446E+04 O.OOOE+00 4.980E-35 O.OOOE+00 5.339E+03 O.OOOE+00 NB -95 4.788E+05 1.605E+08 2.062E+08 O.OOOE+00 1.213E+07 O.OOOE+00 2.475E+07 O.000E+00 MO - 99 1.348E+05 4.626E+06 3.108E+08 0.OOOE+00 1.523E-02 O.OOOE+00 3.731E+07 0.OOOE+00 TC - 99M 2.030E+03 2.109E+05 1.646E+04 0.OOOE+00 O.OOOE+00 O.000E+00 1.978E+03 0.OOOE+00 TC - 101 8.442E+02 2.277E+04 1.423E-56 O.000E+00 0.OOOE+00 O.OOOE+00 6.530E - 58 0.OOOE+00 RU - 103 5.516E+05 1.265E+08 1.055E+05 O.OOOE+00 7.573E+03 0.OOOE+00 1.265E+04 O.OOOE+00 RU - 105 4.844E+04 7.212E+05 3.204E+00 0.OOOE+00 O.OOOE+00 O.OOOE+00 3.851E - 01 O.OOOE+00 RU - 106 1.156E+07 5.049E+08 1.445E+06 O.OOOE+00 4.266E+05 0.OOOE+00 1.734E+05 O.OOOE+00 AG - 110M 3.668E+06 4.019E+09 1.461E+10 O.OOOE+00 3.984E+09 O.OOOE+00 1.752E+09 O.OOOE+00 TE - 125M 4.466E+05 2.128E+06 1.508E+08 O.OOOE+00 1.799E+07 0.OOOE+00 1.809E+07 0.000E+00 TE - 127M 1.312E+06 1.083E+05 1.037E+09 O.OOOE+00 2.046E+08 0.OOOE+00 1.244E+08 O.OOOE+00 TE - 127 2.436E+04 3.293E+03 1.359E+05 0.OOOE+00 1.269E-65 O.000E+00 1.594E+04 0.OOOE+00 TE - 129M 1.680E+06 2.312E+07 1.392E+09 0.OOOE+00 7.559E+07 0.OOOE+00 1.672E+08 O.OOOE+00 TE - 129 2.632E+04 3.076E+04 2.187E-07 O.OOOE+00 0.000E+00 O.OOOE+00 2.624E - 08 O.OOOE+00 TE - 131M 1.988E+05 9.459E+06 2.288E+07 0.OOOE+00 1.653E-15 0.OOOE+00 2.747E+06 0.OOOE+00 TE - 131 8.218E+03 3.450E+07 1.384E-30 0.OOOE+00 O.OOOE+00 0.OOOE+00 1.688E - 31 0.000E+00 TE - 132 3.402E+05 4.968E+06 6.513E+07 0.OOOE+00 1.041E-01 0.OOOE+00 7.842E+06 O.OOOE+00 I -130 1.596E+06 6.692E+06 8.754E+08 0.000E+00 7.115E-45 0.OOOE+00 1.051E+09 O.OOOE+00 (PASTURE) I (PASTURE) (FEED) (PASTURE) I (PASTURE)

Units - 3 Inhalation and all tritium / Carbon 14 - mrem/Yr per [iCi/m Other pathways for all other radionuclides -m - mrem/yr per pCi/sec ODCM, V. C. Summer, SCE&G: Revision 28 (July 2012) 3.0-25

TABLE 3.2-3 (continued)

PATHWAY DOSE FACTORS FOR SECTION 3.2.4 (R=)*

Page 3 of 3 AGE GROUP I (INFANT) (N.A.) (INFANT) (INFANT) (INFANTn (INFANT) I (INFANT) (INFANT)

ISOTOPE INHALATION GROUND GRS/COW/ GRS/COWI GRS/COWI GRS/GOTI GRS/GOTI VEGETATION PLANE MILK MEAT MILK MEAT MILK

- 131 1.484E+07 2.089E+07 1.053E+12 O.OOOE+00 1.567E+08 0.000E+00 1.264E+12 O.OOOE+00 1-132 1.694E+05 1.452E+06 1.188E+02 O.OOOE+00 0.OOOE+00 0.000E+00 1.638E+02 0.000E+00 I - 133 3.556E+06 2.981E+06 9.601E+09 O.000E+00 1.776E-22 0.OOOE+00 1.153E+10 0.OOOE+00 I - 134 4.452E+04 5.305E+05 8.402E-10 O.OOOE+00 0.OOOE+00 0.000E+00 1.017E - 09 0.OOOE+00 I - 135 6.958E+05 2.947E+06 2.002E+07 0.000E+00 O.000E+00 O.000E+00 2.406E+07 O.000E+00 CS - 134 7.028E+05 8.007E+09 6.801E+10 0.000E+00 2.191E+10 O.000E+00 2.040E+11 O.000E+00 CS - 136 1.345E+05 1.710E+08 5.795E+09 0.000E+00 1.729E+07 O.000E+00 1.744E+10 0.000E+00 CS - 137 6.118E+05 1.201E+10 6.024E+10 O.000E+00 2.096E+10 0.OOOE+00 1.087E+12 0.000E+00 CS - 138 8.764E+02 4.102E+05 2.180E-22 0.000E+00 O.000E+00 0.OOOE+00 6.628E - 22 O.OOOE+00 BA - 139 5.096E+04 1.194E+05 2.874E-05 0.000E+00 O.000E+00 0.000E+00 3.265E - 06 0.000E+00 BA - 140 1.596E+06 2.346E+07 2.410E+08 0.000E+00 6.409E+05 O.000E+00 2.893E+07 0.000E+00 BA - 141 4.746E+03 4.734E+04 4.916E-44 0.000E+00 0.OOOE+00 0.OOOE+00 5.899E - 45 O.OOOE+00 BA - 142 1.554E+03 5.064E+04 1.049E-78 0.000E+00 0.000E+00 0.000E+00 1.259E - 79 0.000E+00 LA- 140 1.680E+05 2.180E+07 1.880E+05 O.000E+00 4.563E-12 0.000E+00 2.253E+04 0.000E+00 LA - 142 5.950E+04 9.117E+05 1.078E-05 O.OOOE+00 0.OOOE+00 0.000E+00 1.278E - 06 0.000E+00 CE - 141 5.166E+05 1.540E+07 1.366E+07 O.OOOE+00 7.008E+05 0.OOOE+00 1.640E+06 0.000E+00 CE - 143 1.162E+05 2.627E+06 1.536E+06 O.OOOE+00 1.039E-14 0.000E+00 1.844E+05 0.000E+00 CE - 144 9.842E+06 8.042E+07 1.334E+08 0.000E+00 3.749E+07 0.OOOE+00 1.601E+07 0.OOOE+00 PR - 143 4.326E+05 O.OOOE+00 7.845E+05 0.OOOE+00 2.771E+03 0.OOOE+00 9.407E+04 0.OOOE+00 PR - 144 4.284E+03 2.112E+03 1.171E-48 0.000E+00 0.OOOE+00 O.OOOE+00 1.259E - 49 0.000E+00 ND - 147 3.220E+05 1.009E+07 5.743E+05 0.OOOE+00 6.902E+02 0.OOOE+00 6.885E+04 0.OOOE+00 W -187 3.962E+04 2.740E+06 2.501E+06 0.000E+00 5.275E-22 0.OOOE+00 2.983E+05 0.000E+00 NP - 239 5.950E+04 1.976E+06 9.400E+04 0.OOOE+00 1.025E-07 0.OOOE+00 1.132E+04 0.000E+00 (PASTURE) (PASTURE) (FEED) (PASTURE) I (PASTURE)

Units - 3 Inhalation and all tritium / Carbon 14 - mrem/Yr per gCi/m Other pathways for all other radionuclides -m 9 mrem/yr per JICi/sec ODCM, V. C. Summer, SCE&G: Revision 28 (July 2012) 3.0-26

TABLE 3.2-4 PATHWAY DOSE FACTORS FOR SECTION 3.2.4 (Ri)*

Page 1 of 3 AGE GROUP ICHILD) (N.A.i VCHILD) VCHILD) (CHILD) ICHILDb ICHILD1 ICHILDb GRS/GOT/ VEGETATION

_ _

NH OP ISO

__ _

I LA IONI RO ND PLANE RSCO MLK M

I JMEAT GRS/COW I GRSICOW I MILK GRS/GOTI 4MEAT MILK4 H-3 1.125E+03 4

O.OOOE+00

+

1.421E+03 I 4

2.118E+02 4

1.421 E+03

+

2.543E+01 I 4

2.899E+03 4 I 3.627E+03 C-14 3.589E+04 0.OOOE+00 3.290E+05 I 1.056E+05 3.290E+05 1.267E+04 I 3.290E+O5 7.000E+05 I NA-24 1.610E+04 1.385E+07 8.853E+06 1.725E - 03 1.321 E - 37 2.070E - 04 1.063E+06 3.729E+05 P-32 2.605E+06 0.OOOE+00 7.775E+10 7.411E+09 3.440E+08 8.893E+08 9.335E+10 3.366E+09 CF-51 1.698E+04 5.506E+06 5.398E+06 4.661E+05 1.985E+05 5.593E+04 6.478E+05 6.213E+06 MN-54 1.576E+06 1.625E+09 2.097E+07 8.011E+06 6.012E+06 9.613E+05 2.517E+06 6.648E+08 MN-56 1.232E+05 1.068E+06 1.865E÷00 2.437E - 51 0.OOOE+00 2.924E - 52 2.238E - 01 2.723E+03 FE-55 1.110E+05 0.OOOE+00 1.118E+08 4.571E+08 3.673E+07 5.486E+07 1.453E+06 8.012E+08 FE-59 1.269E+06 3.204E+08 2.025E+08 6.338E+08 1.749E+07 7.605E+07 2.633E+06 6.693E+08 CO-58 1.106E+06 4.464E+08 7.080E+07 9.596E+07 1.032E+07 1.152E+07 8.487E+06 3.771E+08 CO-60 7.067E+06 2.532E+10 2.391E+08 3.838E+08 8.103E+07 4.605E+07 2.870E+07 2.095E+09 NI-63 8.214E+05 0.OOOE+00 2.964E+10 2.912E+10 1.036E+10 3.495E+09 3.557E+09 3.949E+10 NI-65 8.399E+04 3.451E+05 1.909E+01 4.061E - 51 0.OOOE+00 4.873E - 52 2.298E+00 1.211E+03 CU-64 3.670E+04 6.876E+05 3.502E+06 1.393E - 05 7.299E - 46 1.672E - 06 3.907E+05 5.159E+05 ZN-65 9.953E+05 8.583E+08 1.101E+10 1.OOOE +09 2.985E+09 1.200E+08 1.322E+09 2.164E+09 ZN-69 1.018E+04 0.OOOE+00 1.123E- 09 0.OOOE+00 0.OOOE+00 0.OOOE+00 1.043E - 10 9.893E-04 BR-83 4.736E+02 7.079E+03 4.399E - 01 9.519E - 57 0.OOOE+00 1.142E - 57 5.190E - 02 5.369E+00 BR-84 5.476E+02 2.363E+05 6.508E- 23 0.000E+00 0.000E+00 0.OOOE+00 7.758E - 24 3.822E - 11 BR-85 2.531 E+01 0.OOOE+00 O.OOOE+00 0.OOOE+00 O.OOOE+00 0.OOOE+00 O.OOOE+00 O.OOOE+00 RB-86 1.983E+05 1.035E+07 8.804E+09 5.816E+08 1.114E+08 6.979E+07 1.053E+09 4.584E+08 RB-88 5.624E+02 3.779E+04 7.150E - 45 0.OOOE+00 0.000E+00 0.OOOE+00 8.789E - 46 4.374E - 22 RB-89 3.452E+02 1.452E+05 1.397E - 52 0.OOOE+00 0.OOOE+00 0.OOOE+00 1.659E - 53 1.642E - 26 SR-89 2.157E+06 2.509E+04 6.618E+09 4.815E+08 6.730E+08 5.778E+07 1.390E+10 3.593E+10 SR-90 1.010E+08 0.000E+00 1.117E+11 1.040E+10 3.887E+10 1.248E+09 2.346E+11 1.243E+12 SR-91 1.739E+ 05 2.511E+06 2.878E+05 55.292E-10 0.000E+00 6.351E - 11 6.050E+05 1.157E+06 (PASTURE) I(PASTURE) (FEED) (PASTURE) I (PASTURE)

See note, page 3.0-36 Units - 3 Inhalation and all tritium / Carbon 14 - mrem/*r per ViCi/m Other pathways for all other radionuclides -m = mrem/yr per ItCi/sec ODCM, V. C. Summer, SCE&G: Revision 28 (July 2012) 3.0-27

TABLE 3.2-4 (continued)

PATHWAY DOSE FACTORS FOR SECTION 3.2.4 (Ri)*

Page 2 of 3 AGE GROUP I (CHILD) (N.A.) (CHILD) (CHILD) (CHILD) (CHILD) (CHILD) (CHILD)

ISOTOPE INHALATION GROUND GRS/COW, GRS/COW/ GRS/COW/ GRSGOT/ GRS/GOT/ VEGETATION PLANE MILK MEAT MILK MEAT MILK SR -92 2.424E+05 8.631E+05 4.134E+01 3.492E -48 0.OOOE+00 4.191E - 49 8.706E+01 1.378E+4 Y - 90 2.679E+05 5.308E+03 9.171E+05 4.879E+05 2.277E - 05 5.855E+04 1.101E+05 6.569E+7 Y - 91M 2.812E+03 1.161E+05 5.622E - 16 0.OOOE+00 0.OOOE+00 0.OOOE+00 6.344E - 17 1.737E-5 Y - 91 2.627E+06 1.207E+06 5.199E+06 2.400E+08 6.261E+05 2.880E+07 6.240E+05 2.484E+9 Y - 92 2.390E+05 2.142E+05 7.310E+00 6.959E - 35 0.OOOE+00 8.350E - 36 8.791E - 01 4.576E+4 Y - 93 3.885E+05 2.534E+05 1.573E+04 1.547E- 07 9.134E - 61 1.857E - 08 1.888E+03 4.482E+6 ZR -95 2.231 E+06 2.837E+08 8.786E+05 6.106E+08 1.160E+05 7.328E+07 1.054E+05 8.843E+8 ZR - 97 3.511E+05 3.445E+06 4.199E+04 7.015E - 01 4.703E - 35 8.418E - 02 5.042E+03 1.248E+7 NB - 95 6.142E+05 1.605E+08 2.287E+08 2.288E+09 1.346E+07 2.673E+08 2.747E+07 2.949E+8 MO - 99 1.354E+05 4.626E+06 1.738E+08 2.456E+05 8.512E - 03 2.947E+04 2.086E+07 1.647E+7 TC - 99M 4.810E+03 2.109E+05 1.474E+04 6.915E - 18 0.OOOE+00 8.298E - 19 1.771E+03 5.255E+3 TC - 101 5.846E+02 2.277E+04 5.593E - 58 0.OOOE+00 0.OOOE+00 0.OOOE+00 2.566E - 59 4.123E-29 RU - 103 6.623E+05 1.265E+08 1.108E+05 4.009E+09 7.952E+03 4.811E+08 1.329E+04 3.971E+8 RU - 105 9.953E+04 7.212E+05 2.493E+00 5.885E - 25 0.OOOE+00 7.061E - 26 2.997E - 01 5.981E+4 RU - 106 1.476E+07 5.049E+08 1.437E+06 6.902E+10 4.243E+05 8.282E+09 1.725E+05 1.159E+10 AG - 110M 5.476E+06 4.019E+09 1.678E+10 6.742E+08 4.576E+09 8.090E+07 2.013E+09 2.581E+9 TE - 125M 4.773E+05 2.128E+06 7.377E+07 5.690E+08 8.802E+06 6.828E+07 8.853E+06 3.506E+8 TE - 127M 1.480E+06 1.083E+05 5.932E+08 5.060E+09 1.171E+08 6.072E+08 7.118E+07 3.769E+9 TE - 127 5.624E+04 3.293E+03 1.191E+05 1.607E-08 0.OOOE+00 1.929 -09 1.396E+04 3.903E+5 TE - 129M 1.761E+06 2.312E+07 7.961 E+08 5.245E+09 4.324E+07 6.294E+08 9.563E+07 2.46E+9 TE - 129 2.549E+04 3.076E+04 7.96E-08 0.OOOE+00 0.OOOE+00 0.OOOE+00 9.641E - 09 7.204E-2 TE - 131M 3.078E+05 9.459E+06 2.244E+07 9.815E+03 1.621E - 15 1.178E+03 2.094E+06 2.163E+7 TE - 131 2.054E+03 3.450E+07 8.489E - 32 0.OOOE+00 0.OOOE+00 0.OOOE+00 1.036E - 32 1.349E-14 TE - 132 3.774E+05 4.968E+06 4.551E+07 9.325E+06 7.272E - 02 1.119E+06 5.480E+06 3.111E+7 I -130 1.846E+06 6.692E+06 3.845E+08 6.758E - 04 3.125E - 45 8.109E - 05 4.617E+08 1.371E+8 (PASTURE) (PASTURE) (FEED) (PASTURE) (PASTURE)

Units 3 Inhalation and all tritium / Carbon 14 - mrem/yr per [iCi/m Other pathways for all other radionuclides -m

mrem/yr per JiCi/sec ODCM, V. C. Summer, SCE&G: Revision 28 (July 2012) 3.0-28

TABLE 3.2-4 (continue)

PATHWAY DOSE FACTORS FOR SECTION 3.2.4 (R=)*

Page 3 of 3 AGE GROUP I (CHILD) (N.A.1 ICHILDI (CHILD) (CHILD) (CHILDI (CHILD) (CHILDi ISOTOPE INHALATION GROUND GRSICOW/ GRSICOWI GRS/COWI GRSIGOTI GRSIGOTI VEGETATION PLANE MILK MEAT MILK MEAT MILK 1-131 1.624E+07 2.089E+07 4.333E+11 5.503E+09 6.448E+07 6.604E+08 5.201E+11 4.754E+10 1-132 1.935E+05 1.452E+06 5.129E+01 2.429E - 57 0.OOOE+00 2.915E - 58 7.072E+01 7.314E+03 1-133 3.848E+06 2.981E+06 3.945E+09 1.304E+02 7.299E - 23 1.564E+01 4.737E+09 8.113E+08 1-134 5.069E+04 5.305E+05 3.624E - 10 0.000E+00 0.OOOE+00 0.000E+00 4.386E - 10 6.622E - 03 1-135 7.918E+05 2.947E+06 8.607E+06 1.039E - 14 0.000E+00 1.247E - 15 1.034E+07 9.973E+06 CS-134 1.014E+06 8.007E+09 3.715E+10 1.513E+09 1.197E+10 1.816E+08 1.115E+11 2.631E+10 CS-136 1.709E+05 1.710E+08 2.773E+09 4.426E+07 8.276E+06 5.311E+06 8.344E+09 2.247E+08 CS-137 9.065E+05 1.201E+10 3.224E+10 1.334E+09 1.122E+10 1.600E+08 9.672E+10 2.392E+10 CS-138 8.399E+02 4.102E+05 5.528E -23 0.000E+00 0.OOOE+00 0.OOOE+00 1.681E - 22 9.133E - 11 BA-139 5.772E+04 1.194E+05 1.231E --05 O.000E+00 0.000E+00 0.OOOE+00 1.398E - 06 2.950E+00 BA-140 1.743E+06 2.346E+07 1.171E+08 4.384E+07 3.114E+05 5.261E+06 1.406E+07 2.767E+08 BA-141 2.919+03 4.734E+04 1.894E - 45 0.000E+00 O.000E+00 0.000E+00 2.273E - 46 1.605E-21 BA-142 1.643E+03 5.064E+04 1.208E - 79 0.OOOE+00 0.000E+00 0.OOOE+00 1.450E - 80 4.105E - 39 LA-140 2.257E+05 2.180E+07 1.894E+05 5.492E+02 4.596E -12 6.590E+01 2.269E+04 3.166E+07 LA-142 7.585E+04 9.117E+05 5.203E - 06 0.000E+00 0.OOOE+00 0.OOOE+00 6.166E - 07 2.141E+01 CE-141 5.439E+05 1.540E+07 1.361E+07 1.382E+07 6.980+05 1.658E+06 1.633E+06 4.082E+08 CE-143 1.273E+05 2.627E+06 1.488E+06 2.516E+02 1.006E - 14 3.020E+01 1.787E+05 1.364E+07 CE-14 1.195E+07 8.042E+07 1.326E+08 1.893E+08 3.727E+07 2.271E+07 1.592E+07 1.039E+10 PR-143 4.329E+05 0.OOOE+00 7.754E+05 3.609E+07 2.738E+03 4.331E+06 9.297E+04 1.575E+08 PR-144 1.565E+03 2.112E+03 2.040E - 50 0.OOOE+00 0.000E+00 0.OOOE+00 2.353E - 51 3.829E - 23 ND-147 3.282E+05 1.009E+07 5.712E+05 1.505E+07 6.864E+02 1.805E+06 6.846E+04 9.197E+07 W-187 9.102E+04 2.740E+06 2.420E+06 2.790E+00 5.103E - 22 3.348E - 01 2.886E+05 5.380E+06 NP-239 6.401E+04 1.976E+06 9.138E+04 2.232E+03 9.336E -08 2.679E+02 1.100E+04 1.357E+07 (PASTURE) (PASTURE) (FEED) (PASTURE) (PASTURE)

Units 3 Inhalation and all tritium / Carbon 14 - mrem/yr per giCi/m Other pathways for all other radionuclides -m a mrem/yr per ltCi/sec ODCM, V. C. Summer, SCE&G: Revision 28 (July 2012) 3.0-29

TABLE 3.2-5 PATHWAY DOSE FACTORS FOR SECTION 3.2.4 (R1 )*

Page 1 of 3 Ar.I: r'DnIID 1T==kAf'9D~ IN AI IT==k1A('-=P I T=MkArf'D% I TC:CAf'-D% 1T==hIAt'-ED% I TCEEMA1(-C0 ITIE EM AjDI:

ISOTOPE INHALATION GROUND GRS/COW/ GRS/COW/ GRS/COW/ GRS/GOTI GRS/GOT/ VEGETATION PLANE MILK MEAT MILK MEAT MILK H-3 1.272E+03 O.OOOE+00 8.993E+02 1.754E+02 8.993E+02 2.104E+01 1.835E+03 2.342E+03 C-14 2.600E+04 0.OOOE+08 1.338E+05 5.618E+04 1.338E+05 6.741E+03 1.338E+05 2.904E+05 NA-24 1.376E+04 1.385E+07 4.255E+06 1.084E - 03 6.347E - 38 1.301E - 04 5.110E+05 2.389E+05 P-32 1.888E+06 0.OOOE+00 3.153E+10 3.931E+09 1.395E+08 4.717E+08 3.785E+10 1.608E+09 CR-51 2.096E+04 5.506E+06 8.387E+06 9.471E+05 3.085E+05 1.137E+05 1.006E+06 1.037E+07 MN-54 1.984E+06 1.625E+09 2.875E+07 1.436E+07 8.240E+06 1.723E+06 3.450E+06 9.320E+08 MN-56 5.744E+04 1.068E+06 4.856E - 01 8.302E - 52 O.OOOE+00 9.962E - 53 5.829E - 02 9.451EE+02 FE-55 1.240E+05 0.OOOE+00 4.454E+07 2.382E+08 1.463E+07 2.859E+07 5.790E+05 3.259E+08 FE-59 1.528E+06 3.204E+08 2.861EE+08 1.171E+09 2.470E+07 1.405E+08 3.720E+06 9.895E+08 CO-58 1.344E+06 4.464E+08 1.095E+08 1.942E+08 1.596E+07 2.330E+07 1.313E+07 6.034E+08 CO-60 8.720E+06 2.532E+10 3.621E+08 7.600E+08 1.227E+08 9.120E+07 4.345E+07 3.238E+09 NI-63 5.800E+05 0.OOOE+00 1.182E+10 1.519E+10 4.130E+09 1.823E+09 1.419E+09 1.606E+10 NI-65 3.672E+04 3.451E+05 4.692E+00 1.305E - 51 0.OOOE+00 1.566E - 52 5.647E - 01 3.966E+02 CU-64 6.144E+04 6.876E+05 3.293E+06 1.713E - 05 6.863E - 46 2.072E - 06 3.673E+05 6.465E+05 ZN-65 1.240E+06 8.583E+08 7.315E+09 8.688E+08 1.983E+09 1.043E+08 8.779E+08 1.471E+09 ZN-69 1.584E+03 0.OOOE+00 1.760E - 11 0.OOOE+00 0.OOOE+00 0.OOOE+00 1.635E - 12 2.067E - 05 BR-83 3.440E+02 7.079E+03 1.790E - 01 5.066E - 57 0.000E+00 6.079E -58 2.112E - 02 2.911E+00 BR-84 4.328E+02 2.363E+05 2.877E - 23 0.OOOE+00 0.OOOE+00 0.OOOE+00 3.429E - 24 2.251E - 11 BR-85 1.832E+01 0.OOOE+00 0.OOOE+00 0.OOOE+00 0.OOOE+00 0.000+00 0.OOOE+00 0.OOOE+00 RB-86 1.904E+05 1.035E+07 4.746E+09 4.101E+08 6.006E+07 4.921E+07 5.675E+08 2.772E+08 RB-88 5.456E+02 3.779E+04 3.886E - 45 0.OOOE+00 0.OOOE+00 0.OOOE+00 4.777E - 46 3.168E - 22 RB-89 3.520E+02 1.452E+05 7.957E - 53 0.OOOE+00 0.OOOE+00 0.OOOE+00 9.454E - 54 1.247E - 26 SR-89 2.416E+06 2.509E+04 2.674E+09 2.545E+08 2.719E+08 3.054E+07 5.617E+09 1.513E+10 SR-90 1.080E+08 0.OOOE+00 6.612E+10 8.049E+09 2.301E+10 9.659E+08 1.389E+11 7.507E+11 SR-91 2.592E+05 2.511E+06 2.409E+05 5.794E - 10 0.OOOE+00 6.953E - 11 5.064E+05 1.291E+06 (PASTURE) (PASTURE) (FEED) (PASTURE) (PASTURE)

See note, page 3.0-36 Units - 3 Inhalation and all tritium / Carbon 14 - mrem/Yr per [tCi/m Other pathways for all other radionuclides -m e mrem/yr per [tCi/sec ODCM, V. C. Summer, SCE&G: Revision 28 (July 2012) 3.0-30

TABLE 3.2-5 (continued)

PATHWAY DOSE FACTORS FOR SECTION 3.2.4 (R1 )*

Page 2 of 3 AGE GROUP I (TEENAGER) (N.A.) (TEENAGER) I (TEENAGER) I (TEENAGER) I (TEENAGER) I (TEENAGER) I (TEENAGER)

ISOTOPE INHALATION GROUND GRS/COW/ GRS/COW/ GRS/COW/ GRSIGOT/ GRS/GOT/ VEGETATION PLANE MILK MEAT MILK MEAT MILK SR - 92 1.192E+05 8.631E+05 2.277E+01 2.516E - 48 0.OOOE+00 3.019E - 49 4.795E+01 1.012E+04 Y-90 5.592E+05 5.308E+03 1.074E+06 7.470E+05 2.666E - 05 8.965E+04 1.289E+05 1.025E+08 Y - 91M 3.200E+03 1.161E+05 5.129E-18 O.OOOE+00 0.OOOE+00 0.OOOE+00 6.260E - 19 2.285E - 07 Y - 91 2.936E+06 1.207E+06 6.147E+06 3.910E+08 7.797E+05 4.691E+07 7.780E+05 3.212E+09 Y - 92 1.648E+05 2.142E+05 2.828E+00 3.522E - 35 0.OOOE+00 4.226E - 36 3.402E - 01 2.360E+04 Y - 93 5.792E+05 2.534E+05 1.312E+04 1.688E - 07 7.620E - 61 2.026E - 08 1.511E+03 4.983E+06 ZR -95 2.688E+06 2.837E+08 1.201E+06 1.092E+09 1.585E+05 1.310E+08 1.441E+05 1.253E+09 ZR - 97 6.304E+05 3.44E+06 4.225E+04 9.231E - 01 4.732E - 35 1.108E - 01 5.073E+03 1.673E+07 NB - 95 7.512E+05 1.605E+08 3.338E+08 4.251E+09 1.963E+07 5.101E+08 4.008E+07 4.551E+08 MO - 99 2.688E+05 4.626E+06 1.023E+08 1.892E+05 5.013E - 03 2.270E+04 1.228E+07 1.293E+07 TC - 99M 6.128E+03 2.109E+05 1.055E+04 6.471E - 18 0.OOOE+00 7.766E - 19 1.267E+03 5.011E+03 TC - 101 6.672E+02 2.277E+04 1.343E - 58 0.OOOE+00 0.OOOE+00 0.OOOE+00 1.508E - 59 3.229E - 29 RU - 103 7.832E+05 1.265E+08 1.513E+05 7.162E+09 1.086E+04 8.595E+08 1.815E+04 5.706E+08 RU - 105 9.040E+04 7.212E+05 1.263E+00 3.900E - 25 0.000E+00 4.680E - 26 1.519E - 01 4.039E+04 RU - 106 1.608E+07 5.049E+08 1.799E+06 1.130E+11 5.312E+05 1.356E+10 2.159E+05 1.484E+10 AG - 110M 6.752E+06 4.019E+09 2.559E+10 1.345E+09 6.982E+09 1.614E+08 3.071E+09 4.031E+09 TE - 125M 5.360+05 2.128E+06 8.863E+07 8.941E+08 1.058E+07 1.073E+08 1.064E+07 4.375E+08 TE - 127M 1.656E+06 1.083E+05 3.420E+08 3.816E+09 6.753E+07 4.580E+08 4.105E+07 2.236E+09 TE - 127 8.080E+04 3.293E+03 9.572E+04 1.689E - 08 0.OOOE+00 2.027E - 09 1.122E+04 4.180E+05 TE - 129M 1.976E+06 2.312E+07 4.602E+08 3.966E+09 2.500E+07 4.759E+08 5.528E+07 1.514E+09 TE - 129 3.296E+03 3.076E+04 2.834E - 09 0.OOOE+00 0.OOOE+00 0.OOOE+00 3.433E - 10 3.916E-03 TE - 131M 6.208E+05 9.459E+06 2.529E+07 1.447E+04 1.827E - 15 1.736E+03 3.036E+06 3.248E+07 TE - 131 2.336E+03 3.450E+07 2.879E - 32 0.OOOE+00 0.OOOE+00 0.OOOE+00 3.515E - 33 6.099E - 15 TE - 132 4.632E+05 4.968E+06 8.581E+07 2.300E+07 1.371E - 01 2.760E+06 1.033E+07 7.818E+07 I -130 1.488E+06 6.692E+06 1.742E+08 4.005E - 04 1.416E - 45 4.806E - 05 2.092E+08 8.276E+07 (PASTURE) (PASTURE) (FEED) (PASTURE) I (PASTURE)

Units 3 Inhalation and all tritium / Carbon 14 - mrem/yr per jiCi/m Other pathways for all other radionuclides -M 2

mrem/yr per igCi/sec ODCM, V. C. Summer, SCE&G: Revision 28 (July 2012) 3.0-31

TABLE 3.2-5 (continued)

PATHWAY DOSE FACTORS FOR SECTION 3.2.4 (R1 )*

Page 3 of 3 AGE GROUP I(TEENAGER) (N.A.) I TEENAGER) I(TEENAGER) I(TEENAGER) I(TEENAGER) I TEENAGER) I(TEENAGER)

ISOTOPE AG INHALATION (TENAER

........ GROUND (N.. GRSICOWI (TENGR GRSICOWI (TEAGR GRSICOW/

(TEAGR GRSIGOTI (TEAGR GRSIGOTI (TEAGR VEGETATION (TEENAGER)

PLANE MILK MEAT MILK MEAT MILK I -131 1.464E+07 2.089E+07 2.195E+11 3.645E+09 3.266E+07 4.375E+08 2.634E+11 3.140E+10 I -132 1.512E+05 1.452E+06 2.242E+01 1.389E - 57 0.000E+00 1.667E - 58 3.092E+01 4.262E+03 I -133 2.920E+06 2.981E+06 1.674E+09 7.234E+01 3.096E - 23 8.680E+00 2.009E+09 4.587E+08 1-134 3.952E+04 5.305E+05 1.583E - 10 0.OOOE+00 0.OOOE+00 0.OOOE+00 1.915E - 10 3.854E - 03 I - 135 6.208E+05 2.947E+06 3.777E+06 5.963E - 15 O.OOOE+00 7.156E - 16 4.538E+06 5.832E + 06 CS - 134 1.128E+06 8.007E+09 2.310E+10 1.231E+09 7.443E+09 1.477E+08 6.931E+10 1.671E+10 CS - 136 1.936E+05 1.710E+08 1.759E+09 3.671E+07 5.249E+06 4.405E+06 5.292E+09 1.708E+08 CS - 137 8.480E+05 1.201E+10 1.781E+10 9.634E+08 6.197E+09 1.156E+08 5.342E+10 1.348E+10 CS - 138 8.560+02 4.102E+05 3.149E - 23 0.OOOE+00 0.000E+00 0.OOOE+10 9.576E - 23 6.935E - 11 BA - 139 6.464E+03 1.194E+05 7.741 E - 07 0.OOOE+00 0.000E+00 0.OOOE+00 8.794E - 08 2.403E - 01 BA - 140 2.032E+06 2.346E+07 7.483E+07 3.663E+07 1.990E+05 4.396E+06 8.981E+06 2.130E+08 BA - 141 3.288E+03 4.734E+04 7.703E - 46 0.OOOE+00 0.OOOE+00 0.OOOE+00 9.244E - 47 8.699E - 22 BA - 142 1.912E+03 5.064E+04 5.010E - 80 0.OOOE+00 0.OOOE+00 0.OOOE+00 6.012E - 81 5.613E - 39 LA - 140 4.872E+05 2.180E+07 2.291E+05 8.689E+02 5.560E - 12 1.043E+02 2.745E+04 5.104E+07 LA - 142 1.200E+04 9.117E+05 4.611E - 07 0.OOOE+00 0.OOOE+00 0.OOOE+00 5.465E - 08 2.529E+00 CE - 141 6.136E+05 1.540E+07 1.696E+07 2.252E+07 8.700E+05 2.703E+06 2.036E+06 5.404E+08 CE - 143 2.552+05 2.627E+06 1.671E+06 3.695E+02 1.130E - 14 4.434E+01 2.006E+05 2.040E+07 CE - 144 1.336E+07 8.042E+07 1.655E+08 3.089E+08 4.650E+07 3.706E+07 1.986E+07 1.326E+10 PR - 143 4.832E+05 0.OOOE+00 9.553E+05 5.817E+07 3.374E+03 6.980E+06 1.146E+05 2.310E+08 PR - 144 1.752E+03 2.112E+03 1.238E - 53 0.OOOE+00 0.OOOE+00 0.OOOE+00 1.331E - 54 3.097E - 26 ND - 147 3.720E+05 1.009E+07 7.116E+05 2.453E+07 8.552E+02 2.942E+06 8.530E+04 1.424E+08 W-187 1.768E+05 2.740E+06 2.646E+06 3.989E+00 5.579E - 22 4.787E - 01 3.155E+05 7.839E+06 NP - 239 1.320E+05 1.976E+06 1.060E+05 3.387E+03 1.083E - 07 4.064E+02 1.276E+04 2.097E+07 (PASTURE) (PASTURE) (FEED) (PASTURE) (PASTURE)

Units -

Inhalation and all tritium / Carbon 14 - mrem/Yr per jiCi/m 3 Other pathways for all other radionuclides -m e mrem/yr per JiCi/sec ODCM, V. C. Summer, SCE&G: Revision 28 (July 2012) 3.0-32

TABLE 3.2-6 PATHWAY DOSE FACTORS FOR SECTION 3.2.4 (Ri)*

Page 1 of 3 AGE GROUP I (ADULT) (N.A.) (ADULT) (ADULT (ADULT) (ADULT) (ADULT) (ADULT)

ISOTOPE INHALATION GROUND GRS/COWI GRSICOWI GRS/COWI GRSIGOTI GRSIGOTI VEGETATION PLANE MILK MEAT MILK MEAT MILK H-3 1.264E+03 0.000E+00 6.904E+02 2.940E+02 6.904E+02 I 3.528E+01 1.408E+03 2.845E+03 C-14 1.816E+04 0.000E+00 7.255E+04 I 6.650E+04 I 7.255E+04 7.980E+03 I 7.255E+04 1.791E+05 NA-24 1.024E+04 1.385E+07 2.438E+06 1.356E - 03 3.636E - 38 1.628E - 04 2.926E+05 2.690E+05 P-32 1.320E+06 0.OOOE+00 1.709E+10 4.651E+09 7.559E+07 5.582E+08 2.052E+10 1.403E+09 CR-51 1.440E+04 5.506E+06 7.187E+06 1.772E+06 2.644E+05 2.127E+05 8.624E+05 1.168E+07 MN-54 1.400E+06 1.625E+09 2.578E+07 2.812E+07 7.389E+06 3.375E+06 3.091E+06 9.585E+08 MN-56 2.024E+04 1.068E+06 1.328E - 01 4.958E - 52 0.000E+00 5.949E - 53 1.594E - 02 5.082E+02 FE-55 7.208E+04 0.000E+00 2.511E+07 2.933E+08 8.250E+06 3.519E+07 3.265E+05 2.096E+08 FE-59 1.016E+06 3.204E+08 2.327E+08 2.080E+09 2.009E+07 2.495E+08 3.024E+06 9.875E+08 CO-58 9.280E+05 4.464E+08 9.565E+07 3.703E+08 1.394E+07 4.443E+07 1.147E+07 6.252E+08 CO-60 5.968E+06 2.532E+10 3.082E+08 1.413E+09 1.044E+08 1.695E+08 3.7E+06 3.139E+09 NI-63 4.320E+05 0.OOOE+00 6.729E+09 1.888E+10 2.351E+09 2.266E+09 8.075E+08 1.040E+10 NI-65 1.232E+04 3.451 E+05 1.219E+00 7.405E - 52 0.OOOE+00 8.886E - 53 1.464E - 01 2.026E+02 CU-64 4.896E+04 6.876E+05 2.031 E+06 2.307E - 05 4.233E - 46 2.769E - 06 2.415E+05 7.841 E+05 ZN-65 8.640E+05 8.583E+08 3.798E+09 1.132E+09 1.183E+09 1.358E+08 4.588E+08 1.009E+09 ZN-69 9.200E+02 0.OOOE+00 4.031E - 12 0.OOOE+00 0.OOOE+00 0.OOOE+00 4.837E - 13 1.202E - 05 BR-83 2.408E+02 7.079E+03 1.399E - 01 8.648E - 57 0.OOOE+00 1.038E - 57 1.698E - 02 4.475E+00 BR-84 3.128E+02 2.363E+05 1.69E - 23 0.OOOE+00 0.OOOE+00 0.OOOE+00 2.029E - 24 2.475E - 11 BR-85 1.280E+01 0.000E+00 0.000E+00 0.OOOE+00 0.OOOE+00 0.OOOE+00 0.OOOE+00 0.OOOE+00 RB-86 1.352E+05 1.027E+07 2.595E+09 4.870E+00 3.201E+07 5.845E+07 3.113E+08 2.194E+08 RB-88 3.872E+02 3.779E+04 2.139E - 45 0.OOOE+00 0.OOOE+00 0.OOOE+00 2.573E - 46 3.428E - 22 RB-89 2.560E+02 1.476E+05 4.496E - 53 0.OOOE+00 0.OOOE+00 0.OOOE+00 5.396E - 54 3.961 E - 26 SR-89 1.400E+06 2.509E+04 1.451E+09 3.014E+08 1.475E+08 3.617E+07 3.046E+09 9.961E+09 SR-90 9.920E+07 0.OOOE+00 4.680E+10 1.244E+10 1.628E+10 1.493E+09 9.828E+10 6.846E+11 SR-91 1.912E+05 2.511E+06 1.377E+05 7.233E-10 0.OOOE+00 8.680E - 11 2.872E+05 1.451E+06 (PASTURE) I (PASTURE) (FEED) (PASTURE) I (PASTURE)

See note, page 3.0-36 Units - 3 Inhalation and all tritium / Carbon 14 - mrem/yr per jiCi/m Other pathways for all other radionuclides -m 9 mrem/yr per jiCi/sec ODCM, V. C. Summer, SCE&G: Revision 28 (July 2012) 3.0-33

TABLE 3.2-6 (continued)

PATHWAY DOSE FACTORS FOR SECTION 3.2.4 (RI)-

Page 2 of 3 AGE GROUP I (ADULT (N.A.) (ADULT) (ADULn (ADULT) (ADULT) (ADULT) (ADULn ISOTOPE INHALATION GROUND GRS/COW/ GRS/COWI GRSICOWI GRS/GOTI GRS/GOTI VEGETATION PLANE MILK MEAT MILK MEAT MILK SR - 92 4.304E+04 8.631 E+05 9.675E+00 2.334E - 48 0.OOOE+00 2.801 E - 49 2.05E+01 8.452E+03 Y - 90 5.056E+05 5.308E+03 7.511E+05 1.141E+06 1.865E - 05 1.369E+05 9.028E+04 1.410E+08 Y - 91M 1.920E+03 1.161E+05 1.883E - 19 0.OOOE+00 0.000E+00 0.OOOE+00 2.262E - 20 1.527E - 08 Y - 91 1.704E+06 1.207E+06 4.726E+06 6.231 E+08 5.691 E+05 7.477E+07 5.672E+05 2.814E+09 Y - 92 7.352E+04 2.142E+05 9.772E - 01 2.657E - 35 0.OOOE+00 3.188E - 36 1.17E - 01 1.603E+04 Y - 93 4.216E+05 2.534E+05 7.091E+03 2.075E - 07 4.290E - 61 2.490E - 08 8.43E+02 5.517E+06 ZR -95 1.768E+06 2.837E+08 9.587E+05 1.903E+09 1.265E+05 2.284E+08 1.151E+05 1.194E+09 ZR - 97 5.232E+05 3.445E+06 2.707E+04 1.292E+00 3.032E - 35 1.550E - 01 3.24E+03 2.108E+07 NB -95 5.048E+05 1.605E+08 2.787E+08 7.748E+09 1.639E+07 9.297+08 3.344E+07 4.798+08 MO - 99 2.480E+05 4.626E+06 5.741E+07 2.318E+05 2.813E - 03 2.781 E+04 6.878E+06 1.426E+07 TC - 99M 4.160E+03 2.109E+05 5.553E+03 7.439E - 18 0.OOOE+00 8.927E - 19 6.641E+02 5.187E+03 TC - 101 3.992E+02 2.277E+04 7.406E - 59 0.OOOE+00 0.OOOE+00 0.OOOE+00 8.888E - 60 3.502E - 29 RU - 103 5.048E+05 1.265E+08 1.189E+05 1.229E+10 8.537E+03 1.475E+09 1.426E+04 5.577E+08 RU - 105 4.816E+04 7.212E+05 5.240E - 01 3.533E - 25 0.OOOE+00 4.239E - 26 6.245E - 02 3.294E+04 RU - 106 9.360E+06 5.049E+08 1.320E+06 1.811E+11 3.898E+05 2.173E+10 1.584E+05 1.247E+10 AG - 110M 4.632E+06 4.019E+09 2.198E+10 2.523E+09 5.996E+09 3.028E+08 2.638E+09 3.979E+09 TE - 125M 3.136E+05 2.128E+06 6.626E+07 1.460E+09 7.906E+06 1.751E+08 7.955E+06 3.927E+08 TE - 127M 9.600E+05 1.083E+05 1.860E+08 4.531E+09 3.671E+07 5.437E+08 2.223E+07 1.418E+09 TE - 127 5.736E+04 3.293E+03 5.278E+04 2.034E - 08 0.OOOE+00 2.441E - 09 6.172E+03 4.532E+09 TE - 129M 1.160E+06 2.312E+07 3.028E+08 5.698E+09 1.645E+07 6.838E+08 3.636E+07 1.261E+09 TE - 129 1.936E+03 3.076E+04 1.183E - 09 0.OOOE+00 0.OOOE+00 0.OOOE+00 1.42E - 10 2.80E-03 TE - 131M 5.560E+05 9.459E+06 1.753E+07 2.190E+04 1.266E - 15 2.628E+03 2.102E+06 4.428E+07 TE - 131 1.392E+03 3.450E+07 1.578E - 32 0.OOOE+00 0.OOOE+00 0.OOOE+00 1.927E - 33 6.575E - 15 TE - 132 5.096E+05 4.968E+06 7.356E+07 4.287E+07 1.170E - 01 5.144E+06 8.827E+06 1.312E+08 I - 130 1.136E+06 6.692E+06 1.050E+08 5.272E - 04 8.535E - 46 6.326E - 05 1.254E+08 9.809+07 (PASTURE) (PASTURE) (FEED) (PASTURE) (PASTURE)

Units - Inhalation and all tritium / Carbon 14 - mrem/yr per gCi/m 3 2

Other pathways for all other radionuclides -M a mrem/yr per [iCi/sec ODCM, V. C. Summer, SCE&G: Revision 28 (July 2012) 3.0-34

TABLE 3.2-6 (continued)

PATHWAY DOSE FACTORS FOR SECTION 3.2.4 (R,)*

Page 3 of 3 AGE GROUP I (ADULT) (N.A.1 (ADULT) (ADULT) (ADULT) (ADULT) (ADULT) (ADULT)

ISOTOPE INHALATION GROUND GRSICOWI GRSICOWI GRS/COW/ GRSIGOTI GRSIGOTI VEGETATION PLANE MILK MEAT MILK MEAT MILK I -131 1.192E+07 2.089E+07 1.388E+11 5.034E+09 2.065E+07 6.040E+08 1.665E+11 3.785E+10 I -132 1.144E+05 1.452E+06 1.541E+01 1.816E - 57 0.000E+00 2.179E - 58 1.849E+01 5.016E+03 I -133 2.152E+06 2.981E+06 9.891E+08 9.336E+01 1.830E - 23 1.120E+01 1.189E+09 5.331E+08 I -134 2.984E+04 5.305E+05 8.886E - 11 0.000E+00 0.000E+00 0.000E+00 1.066E-10 4.563E-03 I - 135 4.480E+05 2.947E+06 2.217E+06 7.644E - 15 0.000E+00 9.172E - 16 2.676E+06 6.731E+06 CS - 134 8.480+05 8.007E+09 1.345E+10 1.565E+09 4.333E+09 1.878E+08 4.035E+10 1.110E+10 CS - 136 1.464E+05 1.710E+08 1.039E+09 4.724E+07 3.093E+06 5.669E+06 3.117E+09 1.675E+08 CS - 137 6.208E+05 1.201E+10 1.010E+10 1.193E+09 3.513E+09 1.431E+08 3.03E+10 8.696E+09 CS - 138 6.208E+02 4.102E+05 1.786E - 23 0.000E+00 0.000E+00 0.OOOE+00 5.146E-23 7.730E - 11 BA - 139 3.760E+03 1.194E+05 7.863E - 08 0.000E+00 0.000E+00 0.000E+00 9.435E-09 5.225E - 02 BA - 140 1.272E+06 2.346E+07 5.535E+07 5.917E+07 1.472E+05 7.100E+06 6.643E+06 2.646E+08 BA - 141 1.936E+03 4.734E+04 4.327E - 46 O.000E+00 0.OOOE+00 0.000E+00 5.193E-47 9.463E - 22 BA - 142 1.192E+03 5.064E+04 2.509E-80 0.000E+00 0.OOOE+00 0.OOOE+00 3.011E-81 2.463E - 39 LA - 140 4.584E+05 2.180E+07 1.672E+05 1.385E+03 4.059E - 12 1.662E+02 2.006E+04 7.319E+07 LA - 142 6.328E+03 9.117E+05 6.273E - 08 0.000E+00 0.000E+00 0.OOOE+00 7.531E-09 6.768E - 01 CE - 141 3.616E+05 1.540E+07 1.25E+07 3.632E+07 6.424E+05 4.358E+06 1.503E+06 5.097E+08 CE - 143 2.264E+05 2.627E+06 1.15E+06 5.547E+02 7.768E - 15 6.656E+01 1.38E+05 2.758E+07 CE - 144 7.776E+06 8.042E+07 1.21E+08 4.928E+08 3.398E+07 5.914E+07 1.451E+07 1.112E+10 PR - 143 2.808E+05 0.OOOE+00 6.918E+05 9.204E+07 2.445E+03 1.104E+07 8.297E+04 2.748E+08 PR - 144 1.016E+03 2.112E+03 6.716E - 54 0.OOOE+00 0.OOOE+00 0.OOOE+00 7.745E-55 3.303E - 26 ND - 147 2.208E+05 1.009E+07 5.231E+05 3.935E+07 6.286E+02 4.722E+06 6.273E+04 1.853E+08 W - 187 1.552E+05 2.740E+06 1.796E+06 5.912E+00 3.787E - 22 7.094E - 01 2.14E+05 1.046E+07 NP - 239 1.192E+05 1.976E+06 7.409E+04 5.152E+03 7.545E - 08 6.182E+02 8.876E+03 2.872E+07 (PASTURE) (PASTURE) (FEED) I (PASTURE) I (PASTURE)

Units - 3 Inhalation and all tritium / Carbon 14 - mrem/*r per jtCi/m Other pathways for all other radionuclides -m e mrem/yr per [iCi/sec ODCM, V. C. Summer, SCE&G: Revision 28 (July 2012) 3.0-35

NOTE: The Ri values of Table 3.2-2 through 3.2-6, with the exception of C-14 were calculated in accordance with the methods of Section 5.3.1 of Reference 1.

C-14 values were calculated using Reference 2. Columns in those tables marked "Pasture" are for freely-grazing animals (fp = fs = 1). Columns marked "Feed" are for animals fed solely locally-grown stored feed (fp = f%= 0). The values used for each parameter and the origins of the values are given in Table 3.2-9 and its notes.

ODCM, V. C. Summer, SCE&G: Revision 28 (July 2012) 3.0-36

Table 3.2-7 CONTROLLING RECEPTORS, LOCATIONS, AND PATHWAYS*

DISTANCE AGE ORIGIN SECTOR (METERS) PATHWAY GROUP (FOR INFORMATION ONLY N**

6,100 Vegetation Child -Vegetable Garden NNE** 5,300 Vegetation Child -Vegetable Garden NE 4,500 Vegetation Child -Vegetable Garden 4,500 Grass/Cow/Meat Child Grazing Beef Cattle ENE 2,600 Vegetation Child -Vegetable Garden 2,600 Grass/Cow/Meat Grazing Beef Cattle E 1,800 Vegetation Child -Vegetable Garden ESE 1,800 Vegetation Child -Vegetable Garden SE 2,400 Vegetation Child -Vegetable Garden SSE 4,300 Vegetation Child -Vegetable Garden S**

6,300 Vegetation Child -Vegetable Garden SSW** 5,500 Vegetation Child -Vegetable Garden SW** 5,300 Vegetation Child -Vegetable Garden WSW 3,100 Grass/Cow/Meat Child -Grazing Beef Cattle W 4,300 Vegetation Child -Vegetable Garden 3,500 Grass/Cow/Meat Child Grazing Beef Cattle WNW** 7,700 Vegetation Child -Vegetable Garden NW** 6,600 Vegetation Child -Vegetable Garden 6,600 Grass/Cow/Meat Child Grazing Beef Cattle NNW 4,800 Vegetation Child -Vegetable Garden 4,800 Grass/Cow/Meat Child Grazing Beef Cattle

See note on the following page for the method used to identify these controlling receptors.

- If a cow were located at 5.0 miles (8,000 meters) in this sector, an infant consuming only its milk would receive a greater total radiation dose than would the real receptor listed. However, such an infant would not be the Maximum Exposed Individual for the site.

ODCM, V. C. Summer, SCE&G: Revision 16 (September 1991) 3.0-37

NOTE: The controlling receptor in each sector was identified in the following way. Receptor locations and associated pathways were obtained from the August 1991 field survey.

A child was assumed at each location, except that where a milk cow was listed, an infant was assumed. X / Q' and D / Q' for each candidate receptor was calculated using five year averaged meteorological data. XOQDOQ8 2 software was used to analyze the meteorological data. Expected annual releases of each nuclide were taken from Table 5.2-2 of Reference 5. The specific dispersion values for each candidate are used with the methodology of ODCM section 3.2.3.2 to calculate a hypothetical dose. The controlling receptor for each sector was then chosen as the candidate receptor with the highest total annual dose of any candidate receptor in the given sector. All listed pathways are in addition to inhalation and ground plane exposure.

ODCM, V. C. Summer, SCE&G: Revision 17 (April 1993) 3.0-38

Table 3.2-8 ATMOSPHERIC DISPERSION PARAMETERS FOR CONTROLLING RECEPTOR LOCATIONS*

DISTANCE SECTOR X/Q' D/Q' (MILES/METERS)

N 2.3 E-7 6.3 E-10 3.8/6,100 NNE 2.9 E-7 8.5 E-10 3.3 /5,300 NE 5.4 E-7 1.5 E-9 2.8 /4,500 ENE 1.8 E-6 5.4 E-9 1.6/2,600 E 3.5 E-6 1.1 E-8 1.1/ 1,800 ESE 2.1 E-6 6.8 E-9 1.1/1,800 SE 6.5 E-7 2.4 E-9 1.5/2,400 SSE 1.2 E-7 5.3 E-10 2.7/4,300 S 7.6 E-8 3.5 E-10 3.9 / 6,300 SSW 1.2 E-7 7.0 E-10 3.4/5,500 SW 1.3 E-7 9.6 E-10 3.3/5,300 WSW 3.6 E-7 2.5 E-9 1.9/3,100 W 1.8 E-7 7.7 E-10 2.7/4,300 W 2.8 E-7 1.3 E-9 2.2 / 3,500 WNW 3.8 E-8 1.3 E-10 4.8/7,700 NW 9.8 E-8 2.8 E-10 4.1 /6,600 NNW 3.3 E-7 9.0 E-10 3.0 / 4,800 Annual average relative dispersion and deposition values for the receptor locations in Table 3.2-7. Values were calculated from 5 year averaged meteorological data using the XOQDOQ-82 software. Dispersion values were calculated assuming ground-level release, open terrain recirculation, dry depletion, and using decay with a half-life of 8.0 days. As a result of the analysis described in the note to Table 3.2-7, the location of the maximum exposed individual for the site is assumed to be the vegetable garden at 1.1 miles in the E sector. Therefore, the site X / Q' and D / Q' (Section 3.2.3.2 and following) are those from this table for that location.

ODCM, V. C. Summer, SCE&G: Revision 17 (April 1993) 3.0-39

Table 3.2-9 Page 1 of 4 PARAMETERS USED IN DOSE FACTOR CALCULATIONS Ori-gin of Value Parameter Value Section of Table in NUREG- Site-R.G. 1.109 0133 Specific

'For P**

DFAj Each radionuclide E-9 Note 2 3

BR 3700 m /yr E-5

'For Ri (Vegetation)***

r Each element type E-1 2

Yv 2.0 kg/m E-15

?,w 5.73 E-7 sec-1 5.3.1.3 DFL1 Each age group and radio- E-1 1 thru Note 2 nuclide E-14 UaL Each age group E-5 fL 1.0 5.3.1.5 tL 8.6 E + 4 seconds E-15 Uas Each age group E-5 f9 0.76 5.3.1.5 th 5.18 E + 6 seconds E-15 H 8.84 gm/m 3 Note 1

- - For Ri (inhalation)***

BR Each age group E-5 DFAj Each age group and E-7 thru Note 2 nuclide E-10 ODCM, V. C. Summer, SCE&G: Revision 28 (July 2012) 3.0-40

Table 3.2-9 Page 2 of 4 PARAMETERS USED IN DOSE FACTOR CALCULATIONS Origin of Value Parameter Value Section of Table in NUREG- Site-R.G. 1.109 0133 Specific

- - For R, (Ground Plane)***

SF 0.7 E-15 DFGi Each radionuclide E-6 t 4.73 E + 8 sec 5.3.1.2

'For Ri (Grass/AnimallMeat)***

QF (Cow) 50 kg/day E-3 QF (Goat) 6 kg/day E-3 Uap Each age group E-5

?w 5.73 E-7 sec 1 5.3.1.3 Ff (Both) Each element E-1 r Each element type E-15 DFLI Each age group and nuclide E-1 1 thru Note 2 E-14 fp 1.0 Note 3 fs 1.0 Note 3 3

Yp 0.7 kg/M E-15 th 7.78 E + 6 sec E-15 YS 2.0 kg/M 2 E-15 tf 1.73 E + 6 sec E-15 H 8.84 gm/m 3 Note 1 ODCM, V. C. Summer, SCE&G: Revision 13 (June 1990) 3.0-41

Table 3.2-9 Page 3 of 4 PARAMETERS USED IN DOSE FACTOR CALCULATIONS Origin of Value Parameter Value Section of Table in NUREG- Site-R.G. 1.109 0133 Specific

- - For R, Note 4 (Grass/Animal/Milk)***

QF (Cow) 50 kg/day E-3 QF (Goat) 6 kg/day E-3 Uap Each age group E-5 xw 5.73 E-7 sec-l 5.3.1.3 Fm Each element E-1 & E-2 r Each element type E-15 DFLj Each age group and nuclide E-1 1 thru Note 2 E-14 Yp 0.7 kg/M 2 E-15 th 7.78 E + 6 sec E-15 YS 2.0 kg/m 2 E-15 tf 1.73 E + 5 sec E-15 fp 1.0 Note 5 fs 1.0 Note 5 fp 0.0 Note 5 f 0.0 Note 5 3

H 8.84 gm/m Note 1 ODCM, V. C. Summer, SCE&G: Revision 13 (June 1990) 3.0-42

Table 3.2-9 (Continued)

Page 4 of 4 NOTES

1. Site-specific annual average absolute humidity. For each month, an average absolute humidity was calculated from the 7 years of monthly average temperatures in Table 2.3-49 of Reference 4 and 5 years of monthly average dew points in Table 2.3-64 of Reference 4. The 12 monthly values were averaged to obtain the annual average of 8.84 gm/m 3 . (Section 5.2.1.3 of Reference 1 gives a default value of 8 gm/m 3.)

2. Inhalation and ingestion dose factors were taken from the indicated source. For each age group, for each nuclide, the organ dose factor used was the highest dose factor for that nuclide and age group in the referenced table.

3. Typically beef cattle are raised all year on pasture. Annual land surveys have indicated that the small number of goats raised within 5 miles typically are used for grass control and not food or milk. Nevertheless, the goats were treated as full meat and milk sources where present, despite the fact that their numbers cannot sustain the meat consumption rates of Table E-5 of Reference 3.

4. According to the August 1990 land use census, dairy cattle possibly graze at 4.9 miles in the West sector. If dairy cattle graze at this location, the dose to an infant consuming milk from these animals would be less than the dose received by the critical receptor identified for the sector. No other milking activity within five miles of the plant was identified. These values are included for reference only.

5. Two columns of Re's were calculated - one for cows kept exclusively on local pasture (fp = f, = 1), and one for cows kept exclusively on locally grown stored feed (fp = f, = 0).

See the note on page 3.0-36.

ODCM, V. C. Summer, SCE&G: Revision 28 (July 2012) 3.0-43

GASEOUS RADWASTE TREATMENT SYSTEM Figure 3.2-1 ODCM, V. C. Summer, SCE&G: Revision 16 (September 1991) 3.0-44

3.3 Meteorological Model for Dose Calculations 3.3.1 Meteorological Model Parameters Term Definition Section of Initial Use b = height of the containment building. (3.3.2.1)

Dg = deposition rate for ground-level releases (3.3.2.2) relative to the distance from the containment building (from Figure 3.3-3).

D/Q the sector averaged relative deposition2 (3.3.2.2) for any distance in a given sector (m ).

= wind speed class. The wind speed (3.3.2.1) classes are given in Table 4A of Reference 10 as 1-3, 4-7, 8-12, 13-18, 19-24, and > 24 miles per hour.

N = total hours of valid meteorological data. (3.3.2.1) nij = number of hours meteorological (3.3.3.1) conditions are observed to be in a given wind direction, wind speed class i, and atmospheric stability class j.

n = number of hours wind is in given (3.3.2.1) direction.

r = distance from the containment building to (3.3.2.1) the location of interest for dispersion calculations (m).

AT/AZ = temperature differential with vertical (3.3.2.1) separation (OK/100m).

T = terrain recirculation factor, (3.3.2.1)

Figure 3.3-4.

u = wind speed (midpoint of wind speed (3.3.2.1) class i) at ground level (m/sec).

X/Q = the sector average relative concentration (3.3.2.1) at any distance in a given sector.

(sec/m 3).

6 = plume depletion factor at distance r from (3.3.3.1)

Figure 3.3-1.

ODCM, V. C. Summer, SCE&G: Revision 16 (September 1991) 3.0-45

Term Definition Section of Initial Use

-3, vertical standard deviation of the plume (in (3.3.2.1) meters), at distance r for ground level releases under the stability category indicated by AT/AZ, from Figure 3.3-2.

2.032 = (2/I1)1 2 divided by the width in radians of a (3.3.2.1) 22.50 sector (0.3927 radians).

2.55 = the inverse of the number of radians in a 22.50 (3.3.2.2) sector 1

(22.50) (0.0175 Radians/°)

3.3.2 Meteorological Model 3.3.2.1 Atmospheric dispersion for routine venting or other routine gaseous effluent releases is calculated using a ground-level, wake-corrected form of the straight line flow model.

X/Q the sector-averaged relative concentration at any distance in the given sector (sec/m3 )

2.032,5 T Z ~.Nru,nj*" (52) where:

2.032 (2/7t) 112 divided by the width in radians of a 22.50 sector (0.3927 radians).

6 = plume depletion factor at distance r for the appropriate stability class from Figure 3.3-1.

= wind speed class. The wind speed classes are given in Table 4A of Reference 10 as 1-3, 4-7, 8-12, 13-18, 19-24, and > 24 miles per hour.

nij = number of hours meteorological conditions are observed to be in a given wind direction, wind speed class i, and atmospheric stability class j.

ODCM, V. C. Summer, SCE&G: Revision 16 (September 1991) 3.0-46

N = total hours of valid meteorological data.

r = distance from the containment building to location of interest (m)

Ui = wind speed (midpoint of wind speed class i) at ground level (m/sec).

-z the lesser of (oz +b /2,)2 or (53) where:

Uz = vertical standard deviation of the plume (in meters) at distance r for ground level releases under the stability category indicated by AT/ AZ, from Figure 3.3-2.

T = terrain recirculation factor, from Figure 3.3-4 t= 3.1416 b = height of the containment building (50.9m)

AT/AZ = temperature differential with vertical separation (OK/1 00m).

Note: For calculation of X/Q using actual meteorological data for a particular release, ui

= the average wind speed for hour i and nij = number of hours with wind speed and stability class j.

3.3.2.2 Relative deposition per unit area for all releases is calculated for a ground-level release.

D/Q = the sector-averaged relative deposition at any distance in a given sector (m-2).

255 Dg" (54) rN where, D9 deposition rate for ground-level releases relative to distance (r) from the containment building (from Figure 3.3-3).

ODCM, V. C. Summer, SCE&G: Revision 16 (September 1991) 3.0-47

2.55 = the inverse of the number of radians in a 22.50 sector I

(225') (0.0175 Radians!0 )

n number of hours wind is in given direction (sector).

N = total hours of valid meteorological data.

ODCM, V. C. Summer, SCE&G: Revision 13 (June 1990) 3.0-48

FIGURE 3.3-1 Plume Depletion Effect for Ground Level Releases (8)

(All Atmospheric Stability Classes)

Graph taken from Reference 8, Figure 2 D.

p In w

9-mu a

-J mu 4

mu

-J

a. 0.

o q r~ ** t* ~ r~ ~1 -

a AR C T O R N aPU FRACTION REMAINING IN PLUME ODCM, V. C. Summer, SCE&G: Revision 13 (June 1990) 3.0-49

FIGURE 3.3-2 Vertical Standard Deviation of Material in a Plume (,5)

(Letters denote Pasquill Stability Clases)

Graph taken from Reference 8, Figure 1 E

I-PLUME TRAVEL DISTANCE (KILOMETERS)

Temperature change Pasquill Stability with Height A T/A Z (*K/lOOm) Categories Classification

<-1.9 A Extremely Unstable

-1.9 to -1.7 B Moderately Unstable

-1.7 to 1.5 C Slightly Unstaable

-1.5 to -0.5 D Neutral

-0.5 to 1.5 E Slightly Stable 1.5 to 4.0 F Moderately Stable

>4.0 G Extremely Stable ODCM, V. C. Summer, SCE&G: Revision 13 (June 1990) 3.0-50

FIGURE 3.3-3 Relative Deposition for Ground Level Releases (D9)

(All Atmospheric Stability Classes)

Graph taken from Reference 8, Figure 6 1o-4 I-I-=

z<

0 10-6 0=o I t i ll 10-6 10-7 0.1 1.0 ¶0.0 100.0 200.0 PLUME TRAVEL DISTANCE (KILOMETERS)

ODCM, V. C. Summer, SCE&G: Revision 13 (June 1990) 3.0-51

FIGURE 3.3-4 Open Terrain Recirculation Factor Graph taken from Reference 7, Figure 2

-I-. - -. -- - _____

a U;

cc a

______ d Q R CORRECTION FACTOR ODCM, V. C. Summer, SCE&G: Revision 13 (June 1990) 3.0-52

4.0 RADIOLOGICAL ENVIRONMENTAL MONITORING Sampling locations as required in section 1.4.1 of the ODCM Specifi-cations are described in Table 4.0-1 and shown on Figures 4.0-1 through 4.0-4. As indicated by the ditto (") marks in the table, entries in the sampling frequency and analysis frequency columns apply to all samples below the entry until a new entry appears.

ODCM, V. C. SummerISCE&G: Revision 26 (September 2007) 4.0-1

RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM VIRGIL C. SUMMER NUCLEAR STATION TABLE 4.0-1 Exposure Criteria for Selection Sampling and Collection Sample" Locations Type & Frequency of Pathway of Sample Number & Location Frequency Location Mi/Dir Analysis and/or Sample AIRBORNE: A) 3 Indicator samples from locations close to the site Continuous sampler operation 2 1.2 SW Gross beta following filter

1. Particulate boundary, in different sectors, of the highest with weekly collection. 7 1.0 E hange; Quarterly calculated annual average ground level D/Q or 30 0.5 SSW Composite (by location) for dose. 9 gamma isotopic.

B) 1 Indicator sample to be taken close to the site Continuous sampler operation 6 1.0 ESE Gross hange;beta following Quarterly 6 filter boundary in the sector corresponding to the with weekly collection.

residence having the highest anticipated offsite Composite (by location) for ground level concentration or dose. 9 gamma isotopic.

C) 1 Indicator sample to be taken at the location of Continuous sampler operation N/A N/A Gross hange;beta following Quarterly 6 filter one of the dairies being sampled meeting the with weekly collection.

criteria of VII(A). 2,9 Composite (by location) for gamma isotopic.

D) 1 Control samples to be taken at a location at Continuous sampler operation 17 25.0 SE Sross beta following filter least 10 air miles from the site and not in the most with weekly collection. change; Quarterly 6 prevalent wind direction. 9 Composite (by location) for gamma isotopic.

ODCM, V. C. Summer/SCE&G: Revision 26 (September 2007) 4.0-2

RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM VIRGIL C. SUMMER NUCLEAR STATION TABLE 4.0-1 Exposure Criteria for Selection Sampling and Collection Sample "' Locations Type & Frequency of Pathway of Sample Number & Location Frequency Location Mi/Dir Analysis and/or Sample II. Radioiodine A) 3 Indicator samples to be taken at two locations Continuous sampler operation 2 1.2 SW Gamma Isotopic for as given in I(A) above. with weekly canister collection. 7 1.0 E 1-131 weekly.

30 0.5 SSW B) 1 Indicator to be taken at the location as given in Continuous sampler operation 6 1.0 ESE Gamma Isotopic for I(B) above with weekly canister collection. 1-131 weekly.

C) 1 Indicator sample to be taken at the location as Continuous sampler operation N/A N/A Gamma Isotopic for given in I(C) above with weekly canister collection. 1-131 weekly.

D) 1 Control sample to be taken at a location similar Continuous sampler operation 17 25.0 SE Gamma Isotopic for in nature to I(D) above. with weekly canister collection. 1-131 weekly.

I1l. Direct A) 13 Indicator stations to form an inner ring of Monthly 5 or quarterly 6; 1,2 1.2 S, 1.2 SW Gamma dose monthly 5 stations in the 13 accessible sectors within 1 to 2 exchange' two or more 3,4 1.2W, 1.2WNW or quarterly. 6 miles of the plant. dosimeters at each location. 5,6 0.9 SE, 1.0 7,8 ESE 9,10 1.0 E, 1.5 ENE 29 2.3 NE, 2.5NNE 30 1.0 WSW, 47 1.0 SSW 1.0 NW ODCM, V. C. Summer/SCE&G: Revision 26 (September 2007) 4.0-3

RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM VIRGIL C. SUMMER NUCLEAR STATION TABLE 4.0-1 Exposure Criteria for Selection Sampling and Collection Sample" Locations Type & Frequency of Pathway of Sample Number & Location Frequency Location Mi/Dir Analysis and/or Sample I1l. Direct B) 16 Indicator stations to form an outer ring of Monthly 5 or quarterly 6, 12,13 4.2 N, 2.9 NNW (Continued) stations in the 16 accessible sectors within 3 to exchange two or more 32,33 4.6 NNE, 4.2 ENE Gamma dose monthly 5 5 miles of the plant. dosimeters at each location. 34,35 4.9 ESE, 4.6 SE or quarterly. 6 36,37 3.1 SSE, 4.9 NW 41,42 3.8 S, 3.8 SSW 43,44 5.2 SW, 2.8 WSW 46,60 3.7 WNW, 3.5W 53,55 3.0 NE, 2.8 E 5

C) 11 Stations to be placed in special interest areas Monthly 5 or quarterly 6 16, 17 28.0 W, 25.0 SE Gamma dose monthly such as population centers, nearby residences, exchange two or more 18,19 16.5 S, 21.0 SSW or quarterly. 6 schools and in 4 or 5 areas to serve as controls. dosimeters at each location. 20,31 22.0 NW, 6.6 NNE 45,52 5.8 WSW, 3.8 NNE 54,56 1.7 ENE, 2.0 SE 58 2.5 SSE WATERBORNE:

IV. Surface A) 1 Indicator sample downstream to be taken at a Time composite samples with 21 3,11 2.7 SSW Gamma isotopic Water location which allows for mixing and dilution in collection every month. 5 monthly 5 with the ultimate receiving river. quarterly 6 composite (by location) to be nalyzed for tritium.

B) 1 Control sample to be taken at a location on the Time composite samples with 22 1 26.0 NNW Gamma isotopic receiving river, sufficiently far upstream such collection every month. 5 nonthly 5 with that no effects of pumped storage operation are quarterly 6 composite anticipated. (by location) to be rnalyzed for tritium.

ODCM, V. C. Summer/SCE&G: Revision 24 (May 2006) 4.0-4

RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM VIRGIL C. SUMMER NUCLEAR STATION TABLE 4.0-1 Exposure Criteria for Selection Sampling and Collection Sample" Locations Type & Frequency of Pathway of Sample Number & Location Frequency Location Mi/Dir Analysis and/or Sample IV. Surface Gamma isotopic monthly Water C) 1 Indicator sample to be taken in the upper Time composite samples with 23 11 0.5 ESE with quarterly 6 (Continued) reservoir of the pumped storage facility at the collection every month.,5 composite (by location) plant discharge canal. to be analyzed for tritium.

V. Ground A) 12 Indicators samples to be taken within the Quarterly 6 grab sampling. 6 1.0 ESE Gamma isotopic and Water exclusion boundary and in the direction of 26, 27 Onsite ritium analyses 6

potentially affected ground water supplies. 101-103 Onsite uarterly.

106 Onsite 108,110 Onsite 112 0.36 SSE 113 0.33 SSE 114 0.39 SE Gamma isotopic and

1) 1 Control sample from unaffected location. Quarterly 6 grab sampling. 59 2.6 SSE tritium analyses Iluarterly.6 VI. Drinking A) 1 Indicator sample from a nearby public Monthly 5 grab sampling. 28 2.6 SSE Monthly 5 gamma Water ground water supply source. isotopic and gross beta nalyses and quarterly 6 composite for tritium analyses.

1 Indicator (finished water) sample from the Monthly 5 composite sampling. 17 25.0 SE Monthly 5 gamma nearest downstream water supply. isotopic and gross beta nalyses and quarterly 6 composite for tritium analyses.

C) 1 Control (finished water) sample from an Monthly 5 composite sampling. 39 14.0 SSE Monthly 5 gamma unaffected water supply. Isotopic and gross beta analyses and quarterly 6 composite for tritium

-analyses.

ODCM, V. C. Summer/SCE&G: Revision 29 (August 2013) 4.0-5

RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM VIRGIL C. SUMMER NUCLEAR STATION TABLE 4.0-1 Exposure Criteria for Selection Sampling and Collection Sample " Locations Type & Frequency Pathway of Sample Number & Location Frequency Location Mi/Dir of Analysis and/or Sample INGESTION:

VII. Milk 2 A) Samples from milking animals in 3 locations within Semimonthly 4 when animals To be supplied Gamma isotopic and 5 km having the highest dose potential. If there are on pasture, monthly 5 when milk 1-131 analysis are none then 1 sample from milking animals in other times. animals are semimonthly 4 when each of 3 areas between 5 to 8 km distance where found in animals are on doses are calculated to be greater than 1 mrem accordance with pasture monthly 5 per year.1 criteria VII(A). other times.

B) 1 Control sample to be taken at the location of a Semimonthly 4 when animals 16 20.0 W Gamma isotopic and dairy > 20 miles distance and not in the most are on pasture, monthly 5 1-131 analysis 2 84 prevalent wind direction. other times. semimonthly 4 when animals are on pasture monthly other times.

1 Indicator grass (forage) sample to be taken at Monthly 5 when available. To be supplied Gamma isotopic.

the location of one of the dairies being sampled when milk meeting the criteria of VII(A), above, when animals animals are are on pasture. found in accordance with criteria VII(A).

D) 1 Control grass (forage) sample to be taken at the Monthly 5 when available. 8 16 20.0 W Gamma isotopic.

location of VII(B) above.

ODCM, V. C. Summer/SCE&G: Revision 28 (July 2012) 4.0-6

RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM VIRGIL C. SUMMER NUCLEAR STATION TABLE 4.0-1 Exposure Criteria for Selection Sampling and Collection Sample "U Locations Type & Frequency Pathway of Sample Number & Location Frequency Location Mi/Dir of Analysis and/or Sample VIII. Food 4) 2 samples of broadleaf vegetation grown in the 2 Monthly 5 when available. 6 1.0 ESE Gamma isotopic on Products nearest offsite locations of highest calculated annual 7 1.0 E edible portion.

average ground level D/Q if milk sampling is not performed within 3 km or if milk sampling is not performed at a location within 5 to 8 km where the doses are calculated to be greater than 1 mrem/yr.i B) 1 Control sample for the same foods taken at a Monthly 5 when available. 40 11.9 SSE Gamma isotopic on location at least 10 miles distance and not in the dible portion.

most prevalent wind direction if milk sampling is not performed within 3 km or if milk sampling is not performed at a location within 5 to 8 km where the doses are calculated to be greater than 1 mrem/yr. 1 IX. Fish A) 1 Indicator sample to be taken at a location in the Seminannual 7 collection of 23 1 0.3-5 Gamma isotopic on upper reservoir, the following specie types if edible portions 7 available: bass; bream, seminannually.

crappie; catfish, carp.

B) 1 Indicator sample to be taken at a location in the Seminannual 7 collection of 2111 1-3 Gamma isotopic on lower reservoir. the following specie types if edible portions 7 available: bass; bream, seminannually.

crappie; catfish, carp.

ODCM, V. C. Summer/SCE&G: Revision 28 (July 2012) 4.0-7

RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM VIRGIL C. SUMMER NUCLEAR STATION TABLE 4.0-1 Exposure Criteria for Selection Sampling and Collection Sample 1" Locations Type & Frequency Pathway of Sample Number & Location Frequency Location Mi/Dir of Analysis and/or Sample IX. Fish C) 1 Control sample to be taken at a location on the Semiannual 7 collection of the 22 26.0 NNW Gamma Isotopic on (Continued) receiving river sufficiently far upstream such that following specie types if edible portions available: bass; bream, semiannually. 7 no effects of pumped storage operation are anticipated. crappie; catfish, carp.

AQUATIC:

X. Sediment A) 1 Indicator sample to be taken at a location in the Semiannual 7 grab sample. 23 0.5 ESE Gamma isotopic.

upper reservoir.

B) 1 Indicator sample to be taken on or near the Semiannual 7 grab sample. 2111 2.7 SSW Gamma isotopic.

shoreline of the lower reservoir.

C3) 1 Control sample to be taken at a location on the Semiannual 7 grab sample. 22 26.0 NNW Gamma isotopic.

receiving river sufficiently far upstream such that no effects of pumped storage operation are

_ _ _ anticipated.

ODCM, V. C. Summer/SCE&G: Revision 23 (September 1999) 4.0-8

RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM VIRGIL C. SUMMER NUCLEAR STATION TABLE 4.0-1 TABLE NOTATION

1. The dose shall be calculated for the maximum organ and age group, using the guidance/methodology contained in Regulatory Guide 1.109, Revision 1 and the parameters particular to the Site. The locations are selected based on the potential for the highest exposure.

2. Milking animal and garden survey results will be analyzed annually. Should the survey indicate new dairying activity the owners shall be contacted with regard to a contract for supplying sufficient samples. If contractual arrangements can be made, Site(s) will be added for additional milk sampling up to a total of 3 Indicator Locations.

3. Time composite samples are samples which are collected with equipment capable of collecting an aliquot at time intervals which are short (e.g. hourly) relative to the compositing period.

4. At least once per 18 days.

5. Not to exceed 35 days.

6. At least once per 100 days.

7. At least once per 200 days

8. Milk and forage sampling at the control location is only required when locations meeting the criteria of VII(A) are being sampled.

9. Sample site locations are based on 5 year average meteorological analysis.

10. Location numbers refer to Figures 4.0-1 through 4.0-4.

11. Though generalized areas are noted for simplicity of sample site enumeration, airborne, water and sediment sampling is done at the same location whereas biological sampling sites are generalized areas in order to reasonably assure availability of samples.

ODCM, V. C. Summer, SCE&G: Revision 26 (September 2007) 4.0-9

FIGURE 4.0-1 Radiological Environmental Sampling Locations (Remote) d i-li a 1~

0.

o :

-

0

-~

En U]

z U) w U] w Ian

in zW U- I-.

0. o *~.

-j 0 S cc -a- in ar aiw

-L U] I-2w 4-

~ii 30 0 A ODCM, V. C. Summer, SCE&G: Revision 23 (September 1999) 4.0-10

FIGURE 4.0-2 Radiological Environmental Sampling Locations (Local)

ODCM, V. C. Summer, SCE&G: Revision 23 (September 1999) 4.0-11

FIGURE 4.0-3 Radiological Environmental Sampling Locations (Local)

ODCM, V. C. Summer, SCE&G: Revision 28 (July 2012) 4.0-12

FIGURE 4.0-4 Radiological Environmental Sampling Locations (Local)

T V

9 Mi U Z4 1 0

7 7'

', 77~ K-

, ,

77'

- 7 7

- A A 7 ~7 7 7 w

-'-3

_ i~

I-J.

ODCM, V. C. Summer, SCE&G: Revision 28 (July 2012) 4.0-13

SAP-01 39 ATTACHMENT II PAGE 1 OF 2 REVISION 34 DOCUMENT REVIEW FORM Page 1 of 2 Doumn Idetiicaio Originators Name: Mike. Roberts Ext: 89908 Mail Code: P-40 Date: 3/06/13 Document No.: ODCM Revision No.: . 29 Change Letter:

Title: Off Site Dose Calculation Manual [0 SR E]QR F-1 NNS Development Process:

Permanent: (check one) N Normal Rev/Chg or [] Editorial Correction [] Temporary Approval

==

Description:==

This revision updates the Lower Limit of Detection (LLD) calculation, removes a groundwater sampling location which is no longer required by the latest hydrogeological study and to remove/the oil incinerator following decommissioning.

Reason/Basis for Change: NUREG/CR-4007; CR 13-00163 and ECR 71498 Is the SCOPE of the procedure affected by this change? NO [a YES [F Temporary Approval Final approval required by:

(30 days)

OR DC&R (Person Notified) SS Date Dcomments Reiwr CommentseC Comments Comments Position Type/Print Name Position Type/Print Name Yes/No Yes/No V1[:] -

")

QR Riley MCHS Harmon L1N

0 I.. MOPS Justice rqL1 .. -.

MOM Johnson E1l C) 0=*

n) MMS Shue FII[. 0:) MPS Ray E]J MPSE Weir EI1] MNL Thompson E1]

MHPSS Coleman GMNPO Lippard O-K QR Qualification Verified? M Yes Comment Due Date 3128/2013 olurrence Y n Spervisor Standard review period is 21 days Sup'*rvisor/Date or enter CR # (per 6.4.8.C) GM concurrence __ for expedited review period Pr- imlmntto Actions All Comments Resolved El None Received Yes ri a Commitments Addressed per SAP-0630 1[ NA EL Yes P/CAP # __ LI MLSA Initial/Date 50.59 Applicability/Review Completed (SAP-0107) EL NA [RYes, Attached Security Compliance Review Completed (SAP-0163) J NA [L_Yes (Security review required)

Pre-implementation Training Completed [ NA LI Yes Training required after implementation If NA LI Yes, CR #

PSRC Review Completed El NA M Yes, Mtg. No. a /1 -1/3 NSRC Review Completed & NA EL Yes, Mtg. No.

CMMS Update Required [NA [L Yes Planner Notified Initial/Date Supervisor/Date A k Approval Autho/ty/Date i-;pr Failure by the "Additional Reviewers" to provide comments within 5 working days following the comment due date may be considered as "No Comment".

SAP-01 39 ATTACHMENT II PAGE 2 OF 2 REVISION 34 DOCUMENT REVIEW FORM Page 2 of 2 Document No.: ODCM Rev. No. 29 Chg. Ltr.

DESCRIPTION CONTINUED: N/A REASON/BASIS FOR CHANGE CONTINUED: N/A DOCUMENT REVIEWERS CONTINUED:

Comments Position Comments Position Comments Ty- rnt Nar e Type/Print Name Yes/No Yes/No QA LID NL

") ;=.C L_.

MEP Williamson El 0!=* EDl

°*

LDD DLI] ]DL DLI LDE

Failure by the "Additional Reviewers" to provide comments within 5 working days following the comment due date may be considered as "No Comment".

ML14113A447

Text

Dear Sir / Madam:

Subject:

Enclosures:

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