Text

Annual Radioactive Effluent Release Report
	ML14133A156
Person / Time
Site:	McGuire, Mcguire
Issue date:	04/29/2014
From:	Capps S; Duke Energy Carolinas
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	MNS-14-036
	Download: ML14133A156 (145)
	v • d • e

Steven D. Capps Vice President ENERGY McGuire Nuclear Station Duke Energy MGOIVP 1 12700 Hagers Ferry Road Huntersville, NC 28078 o 980.875.4805 t 980.875.4809 April 29, 2014 Steven.Capps@duke-energy.com MNS-14-036 U. S. Nuclear Regulatory Commission Document Control Desk Washington, D.C. 20555

Subject:

Duke Energy Carolinas, LLC McGuire Nuclear Station, Units 1 and 2 Docket Nos. 50-369 and 50-370 Annual Radioactive Effluent Release Report Pursuant to the requirements of Technical Specification 5.6.3 and Section 16.11.17 of the Selected Licensee Commitments (SLC) Manual, attached is the Annual Radioactive Effluent Release Report. Also included in this report is a CD-Rom of the Offsite Dose Calculation Manual (Revision 54) pursuant to the requirements of Technical Specification 5.5.1. The following Attachments form the contents of the report:

Attachment 1 - Summary of Gaseous and Liquid Effluents Report Attachment 2 - Supplemental Information Attachment 3 - Solid Radioactive Waste Disposal Report Attachment 4 - Meteorological Data Attachment 5 - Unplanned Offsite Releases Attachment 6 - Assessment of Radiation Dose from Radioactive Effluents to Members of the Public (Includes Fuel Cycle Dose Calculation Results)

Attachment 7 - Updated Final Safety Analysis Report Radiological Effluent Controls Section 16.11 Attachment 8 - Revisions to the Radioactive Waste Process Control Program Manual Attachment 9 - Information to Support the Nuclear Energy Institute (NEI) Groundwater Protection Initiative Attachment 10 - Non-Functional Monitoring Equipment Attachment 11 - Radioactive Waste Systems Changes Questions concerning this report should be directed to Kay Crane, McGuire Regulatory Affairs at (980) 875-4306.

Steven D. Capps Attachments IV www.duke-energy.com

U. S. Nuclear Regulatory Commission April 29, 2014 Page 2 cc: Mr. Victor McCree, Regional Administrator U. S. Nuclear Regulatory Commission, Region II Marquis One Tower 245 Peachtree Center Ave., NE Suite 1200 Atlanta, Georgia 30303-1257 Without CDs:

Mr. Ed Miller McGuire Project Manager Nuclear Regulatory Commission 11555 Rockville Pike (Mail Stop 0-8 G9A)

Rockville, MD 28052-2738 Mr. John Zeiler Senior Resident Inspector McGuire Nuclear Station Mr. W.L. Cox, III, Section Chief North Carolina Department of Environment and Natural Resources Division of Environmental Health Radiation Protection Section 1645 Mail Service Center Raleigh, NC 27699-1645 INPO Records Center 700 Galleria Place, Suite 100 Atlanta, GA 30339-5957

ATTACHMENT 1 Summary of Gaseous and Liquid Effluents Report This attachment includes a summary of the quantities of radioactive liquid and gaseous effluents as outlined in Regulatory Guide 1.21, Revision 1, Appendix B.

TABLE 1A EFFLUENT AND WASTE DISPOSAL ANNUAL REPORT PERIOD 1/1/13 TO 1/1/14 GASEOUS EFFLUENTS - SUMMATION OF ALL RELEASES McGuire Nuclear Station Units 1 & 2 REPORT FOR 2013 Units QTR 1 QTR 2 QTR 3 QTR 4 YEAR A. Fission and Activation Gases

1. Total Release Ci 5.50E-01 3.77E-01 4.42E-01 5.07E-01 1.88E+00

2. Avg. Release Rate uCi/sec 7.07E-02 4.79E-02 5.56E-02 6.38E-02 5.95E-02 B. Iodine-131

1. Total Release Ci 1.24E-07 0.OOE+00 0.OOE+00 0.OOE+00 1.24E-07

2. Avg. Release Rate uCi/sec 1.60E-08 O.OOE+00 0.OOE+00 0.OOE+00 3.94E-09 C. Particulates Half Life >= 8 days

1. Total Release Ci 1. 79E-06 0.OOE+00 0.OOE+00 1.49E-06 3.28E-06

2. Avg. Release Rate uCi/sec 2.30E-07 0.OOE+00 0.OOE+00 1.88E-07 1.04E-07 D. Tritium

1. Total Release Ci 1.87E+01 1.91E+01 2.09E+01 2.33E+01 8.20E+01

2. Avg. Release Rate uCi/sec 2.40E+00 2.43E+00 2.62E+00 2.93E+00 2.60E+00 E. Carbon-14

1. Total Release Ci 4.72E+00 4.66E+00 5.44E+00 5.44E+00 2.03E+01

2. Avg. Release Rate uCi/sec 6.07E-01 5.92E-01 6.85E-01 6.85E-01 6.42E-01 F. Gross Alpha Radioactivity

1. Total Release Ci 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00

2. Avg. Release Rate uCi/sec 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00

TABLE lB EFFLUENT AND WASTE DISPOSAL ANNUAL REPORT PERIOD 1/1/13 TO 1/1/14 GASEOUS EFFLUENTS - ELEVATED RELEASES - CONTINUOUS MODE McGuire Nuclear Station Units 1 & 2 REPORT FOR 2013 Unit QTR 1 QTR 2 QTR 3 QTR 4 YEAR

1. Fission and Activation Gases

- No Nuclide Activities **

2. Iodines

- No Nuclide Activities **

3. Particulates Half Life >= 8 days

- No Nuclide Activities **

4. Tritium

- No Nuclide Activities **

5. Carbon-14

- No Nuclide Activities **

6. Gross Alpha Radioactivity

- No Nuclide Activities **

TABLE lB EFFLUENT AND WASTE DISPOSAL ANNUAL REPORT PERIOD 1/1/13 TO 1/1/14 GASEOUS EFFLUENTS - ELEVATED RELEASES - BATCH MODE McGuire Nuclear Station Units 1 & 2 REPORT FOR 2013 Unit QTR 1 QTR 2 QTR 3 QTR 4 YEAR

1. Fission and Activation Gases

- No Nuclide Activities **

2. Iodines

- No Nuclide Activities **

3. Particulates Half Life >= 8 days

- No Nuclide Activities **........

4. Tritium

- No Nuclide Activities **

5. Carbon-14

- No Nuclide Activities **

6. Gross Alpha Radioactivity

- No Nuclide Activities ** ........

TABLE IC EFFLUENT AND WASTE DISPOSAL ANNUAL REPORT PERIOD 1/1/13 TO 1/1/14 GASEOUS EFFLUENTS - GROUND RELEASES - CONTINUOUS MODE McGuire Nuclear Station Units 1 & 2 REPORT FOR 2013 Units QTR 1 QTR 2 QTR 3 QTR 4 YEAR

1. Fission and Activation Gases AR-41 Ci 6.15E-05 0.OOE+00 0. OOE+00 0. OOE+00 6. I5E-05 KR-85 Ci 2.43E-05 0.OOE+00 0. OOE+00 0.001E+00 2.43E-05 KR-85M Ci 3.03E-04 0.00E+00 0. OOE+00 0. OOE+00 3.03E-04 KR-87 Ci 1.68E-05 0.OOE+00 0. OOE+00 0. OOE+00 1.68SE-05 KR-88 Ci 2.36E-04 0.00E+00 0 00.E+00 0. OOE+00 2.36E-04 XE-131M Ci 1.54E-05 0.OOE+00 0. OOE+00 0. OOE+00 1.54E-05 XE-133 Ci 1.13E-03 0.OOE+00 0. OOE+00 0. OOE+00 1. 13E-03 XE-133M Ci 1.13E-03 0.OOE+00 0. OOE+00 0. OOE+00 1. 13E-03 XE-135 Ci 1.09E-02 0.OOE+00 0. OOE+00 0.001E+00 1. 09E-02 Totals for Period... Ci 1.39E-02 0.OOE+00 0.001E+00 0.001E+00 1.39E-02

2. Iodines 1-131 Ci 1.24E-07 0.00E+00 0.OOE+00 0.OOE+00 1.24E-07

3. Particulates Half Li .fe >= 8 days CO-58 Ci 1.07E-06 0. OOE+00 0.001E+00 0. OOE+00 1.07E-06 CO-60 Ci 0.OOE+00 0.001E+00 0.001E+00 1. 49E-06 1. 49E-06 MN-54 Ci 7.13E-07 0.001E+00 0.001E+00 0.001E+00 7.13E-07 Totals for Period... Ci 1.79E-06 0. OOE+00 0.001E+00 1. 49E-06 3.27E-06

4. Tritium H-3 Ci 1.76E+01 1.88E+01 2.04E+01 2.26E+01 7.94E+01

5. Carbon-14 C-14 Ci 1.42E+00 1.40E+00 1.63E+00 1.63E+00 6.09E+00

6. Others BE-7 Ci 1.48E-05 0.00E+00 0.OOE+00 0.00E+00 1.48E-05

7. Gross Alpha Radioactivity

- No Nuclide Activities **

TABLE 2A EFFLUENT AND WASTE DISPOSAL ANNUAL REPORT PERIOD 1/1/13 TO 1/1/14 LIQUID EFFLUENTS - SUMMATION OF ALL RELEASES McGuire Nuclear Station Units 1 & 2 REPORT FOR 2013 Unit QTR 1 QTR 2 QTR 3 QTR 4 YEAR A. Fission and Activation Products

1. Total Release Ci 4.19E-03 5.49E-03 8.44E-03 8.43E-03 2.66E-02

2. Average Diluted Concentration

a. Continuous Releases wCi/ml 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.OOE+00

b. Batch Releases iCi/ml 5.35E-12 6.10E-12 8.29E-12 8.84E-12 7.26E-12 B. Tritium

1. Total Release Ci 4.84E+02 1.70E+02 1.18E+02 5.10E+02 1.28E+03

2. Average Diluted Concentration

a. Continuous Releases aCi/ml 6.63E-09 8.56E-09 1.19E-09 3.35E-09 4.92E-09

b. Batch Releases wCi/ml 6.17E-07 1.88E-07 1.16E-07 5.35E-07 3.50E-07 C. Dissolved and Entrained Gases

1. Total Release Ci 0.OOE+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

2. Average Diluted Concentration

a. Continuous Releases pCi/ml 0.OOE+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

b. Batch Releases pCi/ml 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 D. Gross Alpha Radioactivity

1. Total Release Ci 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

2. Average Diluted Concentration

a. Continuous Releases pCi/ml 0.OOE+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

b. Batch Releases pCi/ml 0.OOE+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 E. Volume of Liquid Waste

1. Continuous Releases liters 4.09E+07 8.64E+07 4.82E+07 7.35E+07 2.49E+08

2. Batch Releases liters 1.54E+06 3.23E+06 7.90E+05 1.07E+06 6.64E+06 F. Volume of Dilution Water

1. Continuous Releases liters 7.53E+10 9.77E+10 1.08E+11 5.15E+10 3.33E+11

2. Batch Releases liters 7.84E+11 9.01E+11 1.02E+12 9.54E+11 3.66E+12

TABLE 2B EFFLUENT AND WASTE DISPOSAL ANNUAL REPORT PERIOD 1/1/13 TO 1/1/14 LIQUID EFFLUENTS - CONTINUOUS MODE McGuire Nuclear Station Units 1 & 2 REPORT FOR 2013 Units QTR 1 QTR 2 QTR 3 QTR 4 YEAR

1. Fission and Activation Products

- No Nuclide Activities ** ........ ........ ........ ........

2. Tritium H-3 Ci 5.OOE-01 8.37E-01 1.28E-01 1.73E-01 1. 64E+00

3. Dissolved and Entrained Gases

- No Nuclide Activities **

4. Gross Alpha Radioactivity

- No Nuclide Activities **

TABLE 2B EFFLUENT AND WASTE DISPOSAL ANNUAL REPORT PERIOD 1/1/13 TO 1/1/14 LIQUID EFFLUENTS - BATCH MODE McGuire Nuclear Station Units 1 & 2 REPORT FOR 2013 Units QTR 1 QTR 2 QTR 3 QTR 4 YEAR

1. Fission and Activation Products AG-108M Ci 1.22E-06 1. 92E-06 1. 44E-06 0. 00E+00 4.58E-06 AG-IIOM Ci 0.OOE+00 0. OOE+00 2. 93E-05 0. OOE+00 2. 93E-05 BA-142 Ci 0.OOE+00 0. OOE+00 0 OOE+00 1. 66E-04 1. 66E-04 BE-7 Ci 0.OOE+00 0. OOE+00 1. 43E-04 0. OOE+00 1. 43E-04 BR-84 Ci 0.OOE+00 0. OOE+00 1. 35E-05 0. 00E+00 1.35E-05 CE-141 Ci 9.66E-07 0. OOE+00 0. OOE+00 0. OOE+00 9. 66E-07 CE-143 Ci 3.33E-06 0. OOE+00 0. 00E+00 0. OOE+00 3.33E-06 CO-57 Ci 9.44E-07 0. OOE+00 0. OOE+00 0. OOE+00 9.44E-07 CO-58 Ci 4.40E-04 1. 47E-03 6. 28E-04 2. 22E-04 2.76E-03 CO-60 Ci 1.48E-03 1. 01E-03 1. 55E-03 6. 93E-04 4.73E-03 CR-51 Ci 4.17E-05 1. 81E-04 4. 24E-05 0. OOE+00 2. 65E-04 CS-137 Ci 1.80E-05 1. 18E-05 8. 37E-06 6. 59E-06 4.47E-05 FE-55 Ci 3.51E-04 1. 26E-03 3. 06E-03 4. 26E-03 8. 93E-03 FE-59 Ci 0.OOE+00 0. OOE+00 3. 11E-06 0. OOE+00 3. 11E-06 K-40 Ci 1.45E-05 0. OOE+00 0. OOE+00 0. OOE+00 1. 45E-05 MN-54 Ci 9.79E-05 1. 75E-05 8. 75E-05 1. 15E-05 2.14E-04 NB-95 Ci 7.92E-06 1. 63E-05 6. 62E-05 3. 28E-06 9.37E-05 NB-97 Ci o. OOE+00 0. OOE+00 5. 33E-06 0. OOE+00 5.33E-06 SB-124 Ci 6.26E-05 4.22E-05 8. 08E-05 3. 57E-05 2.21E-04 SB-125 Ci 1.67E-03 1. 47E-03 2. 70E-03 3. 04E-03 8. 87E-03 SN-113 Ci 0.OOE+00 1.27E-06 0. OOE+00 0. OOE+00 1. 27E-06 ZN-65 Ci 0.OOE+00 2.83E-06 0. OOE+00 0. OOE+00 2. 83E-06 ZR-95 Ci 6.47E-06 1. 19E-05 2. 15E-05 0. OOE+00 3. 99E-05 Totals for Period... Ci 4.19E-03 5. 49E-03 8.44E-03 8.43E-03 2. 66E-02

2. Tritium H-3 Ci 4.83E+02 1. 69E+02 1. 18E+02 5.10E+02 1.28E+03

3. Dissolved and Entrained Gases

- No Nuclide Activities **

4. Gross Alpha Radioactivity

- No Nuclide Activities **

ATTACHMENT 2 Supplemental Information to the Gaseous and Liquid Effluents Report

McGuire 2013 ARERR - Carbon-14 Supplemental Information Carbon-14 (C-14), with a half-life of 5730 years, is a naturally occurring isotope of carbon produced by cosmic ray interactions in the atmosphere. Nuclear weapons testing in the 1950s and 1960s significantly increased the amount of C-14 in the atmosphere. C-14 is also produced in commercial nuclear reactors, but the amounts produced are much less than those produced naturally or from weapons testing.

In Regulatory Guide 1.21, Revision 2, "Measuring, Evaluating, and Reporting Radioactive Material in Liquid and Gaseous Effluents and Solid Waste", the NRC recommends U.S. nuclear power plants evaluate whether C-14 is a "principal radionuclide", and if so, report the amount of C-14 released. At McGuire, improvements over the years in effluent management practices and fuel performance have resulted in a decrease in gaseous radionuclide (non-C-14) concentrations, and a change in the distribution of gaseous radionuclides released to the environment. As a result, C-14 has become a "principal radionuclide" for the gaseous effluent pathway at McGuire, as defined in Regulatory Guide 1.21, Rev. 2. McGuire's 2013 Annual Radioactive Effluent Release Report (ARERR) contains estimates of C-14 radioactivity released in 2013, and estimates of public dose resulting from the C-14 effluent.

Because the dose contribution of C-14 from liquid radioactive waste is much less than that contributed by gaseous radioactive waste, evaluation of C-14 in liquid radioactive waste at McGuire is not required (Ref. Reg. Guide 1.21, Rev.

2). The quantity of gaseous C-14 released to the environment can be estimated by use of a C-14 source term scaling factor based on power generation (Ref. Reg. Guide 1.21, Rev. 2). Many documents provide information related to the magnitude of C-14 in typical effluents from commercial nuclear power plants. Those documents suggest that nominal annual releases of C-14 in gaseous effluents are approximately 5 to 7.3 curies from PWRs (Ref. Reg. Guide 1.21, Rev. 2).

A more recent study recommends a higher C-14 gaseous source term scaling factor of approximately 9.0 to 9.8 Ci/GWe-yr for a PWR (Westinghouse) (Ref. EPRI 1021106). For the 2013 McGuire ARERR a source term scaling factor of 9.4 Ci/GWe-yr is assumed. Using a source term scaling factor of 9.4 Ci/GWe-yr and actual electric generation (MWe-hrs) from McGuire in 2013 results in a site total C-14 gaseous release estimate to the environment of -20 Curies. 70% of the C-14 gaseous effluent is assumed to be from batch releases (e.g. WGDTs), and 30% of C-14 gaseous effluent is assumed to be from continuous releases through the unit vents (ref. IAEA Technical Reports Series no. 421, "Management of Waste Containing Tritium and Carbon-14", 2004).

C-14 releases in PWRs occur primarily as a mix of organic carbon and carbon dioxide released from the waste gas system.

Since the PWR operates with a reducing chemistry, most, if not all, of the C-14 species initially produced are organic (e.g., methane). As a general rule, C-14 in the primary coolant is essentially all organic with a large fraction as a gaseous species. Any time the RCS liquid or gas is exposed to an oxidizing environment (e.g. during shutdown or refueling), a slow transformation from an organic to an inorganic chemical form can occur. Various studies documenting measured C-14 releases from PWRs suggest a range of 70% to 95% organic with an average of 80% organic with the remainder being CO, (Ref. EPRI TR-105715). For the McGuire 2013 ARERR a value of 80% organic C-14 is assumed.

Public dose estimates from airborne C-14 are performed using dose models in NUREG-0133 and Regulatory Guide 1.109.

The dose models and assumptions used are documented in the McGuire ODCM. The estimated C-14 dose impact on the maximum organ dose from airborne effluents released from McGuire in 2013 is well below the 1OCFR50, Appendix I, ALARA design objective (i.e., 15 mrem/yr per unit).

McGUIRE NUCLEAR STATION 2013 EFFLUENT AND WASTE DISPOSAL SUPPLEMENTAL INFORMATION I. REGULATORY LIMITS - PER UNIT A. NOBLE GASES - AIR DOSE B. LIQUID EFFLUENTS - DOSE

1. CALENDAR QUARTER - GAMMA DOSE = 5 MRAD 1. CALENDAR QUARTER - TOTAL BODY DOSE = 1.5 MREM

2. CALENDAR QUARTER - BETA DOSE = 10 MRAD 2. CALENDAR QUARTER - ORGAN DOSE - 5 MREM

3. CALENDAR YEAR - GAMMA DOSE = 10 MRAD 3. CALENDAR YEAR - TOTAL BODY DOSE = 3 MREM

4. CALENDAR YEAR - BETA DOSE = 20 MRAD 4. CALENDAR YEAR - ORGAN DOSE = 10 MREM C. GASEOUS EFFLUENTS - IODINE - 131 AND 133, TRITIUM, PARTICULATES W/T 1/2 > 8 DAYS - ORGAN DOSE

1. CALENDAR QUARTER = 7.5 MREM

2. CALENDAR YEAR = 15 MREM II. MAXIMUM PERMISSIBLE EFFLUENT CONCENTRATIONS A. GASEOUS EFFLUENTS - INFORMATION FOUND IN OFFSITE DOSE CALCULATION MANUAL B. LIQUID EFFLUENTS - INFORMATION FOUND IN 1OCFR20, APPENDIX B, TABLE 2, COLUMN 2 III. AVERAGE ENERGY - NOT APPLICABLE IV. MEASUREMENTS AND APPROXIMATIONS OF TOTAL RADIOACTIVITY ANALYSES OF SPECIFIC RADIONUCLIDES IN SELECTED OR COMPOSITED SAMPLES AS DESCRIBED IN THE SELECTED LICENSEE COMMITMENTS ARE USED TO DETERMINE THE RADIONUCLIDE COMPOSITION OF THE EFFLUENT. A

SUMMARY

DESCRIPTION OF THE METHOD USED FOR ESTIMATING OVERALL ERRORS ASSOCIATED WITH RADIOACTIVITY MEASUREMENTS IS PROVIDED AS PART OF THE "SUPPLEMENTAL INFORMATION" ATTACHMENT.

V. BATCH RELEASES A. LIQUID EFFLUENT

1. 2.47E+02 = TOTAL NUMBER OF BATCH RELEASES

2. 2.35E+04 = TOTAL TIME (MIN.) FOR BATCH RELEASES.

3. 4.11E+03 = MAXIMUM TIME (MIN.) FOR A BATCH RELEASE.

4. 9.50E+01 = AVERAGE TIME (MIN.) FOR A BATCH RELEASE.

5. 1.00E+00 = MINIMUM TIME (MIN.) FOR A BATCH RELEASE.

6. 1.84E+06 = AVERAGE DILUTION WATER FLOW DURING RELEASES (GPM).

B. GASEOUS EFFLUENT

1. 3.40E+01 = TOTAL NUMBER OF BATCH RELEASES.

2. 1.05E+06 = TOTAL TIME (MIN.) FOR BATCH RELEASES.

3. 4.48E+04 = MAXIMUM TIME (MIN.) FOR A BATCH RELEASE.

4. 3.10E+04 = AVERAGE TIME (MIN.) FOR A BATCH RELEASE.

5. 9.40E+01 = MINIMUM TIME (MIN.) FOR A BATCH RELEASE.

VI. ABNORMAL RELEASES (SEE "UNPLANNED OFFSITE RELEASES" ATTACHMENT)

McGUIRE NUCLEAR STATION Overall Estimate of Error for Effluent Radioactivity Release Reported The estimated percentage of overall error for both Liquid and Gaseous effluent release data at McGuire Nuclear Station has been determined to be + 30.3%. This value was derived by taking the square root of the sum of the squares of the following discrete individual estimates of error:

(1) Flow Rate Determining Devices = +/-20%

(2) Counting Statistical Error = +/-20%

(3) Calibration Error = +10%

(4) Calibration Source Error = +/-2.5%

(5) Sample Preparation Error = -3%

Attachment 3 Solid Radioactive Waste Disposal Report

REPORT PERIOD McGUIRE NUCLEAR STATION JANUARY - DECEMBER 2013 SOLID RADIOACTIVE WASTE SHIPPED TO DISPOSAL FACILITIES TYPES OF WASTES SHIPPED Number of Number of Container Disposal Volume Waste Total 3

Waste from Liquid Systems Shipments Containers Type ftW m Class Curies (A) dewatered powdex resin (brokered) none (B) dewatered powdex resin none (C) dewatered bead resin (brokered) none (D) dewatered bead resin none (E) dewatered radwaste system resin none (F) dewatered primary bead resin none (G) dewatered mechanical filter media none (H) dewatered mechanical filter media (brokered) none (I) solidified waste none Dry Solid Waste (A) dry active waste (compacted) none dry active waste (non-compacted) none dry active waste (brokered/compacted) none dry active waste (brokered/non-compacted) 18 45 DBP 4098.85775 116.07 A/U 1.941 E+00 (B) sealed sources/smoke detectors none (C) sealed sources none (D) irradiated components none Totals 18 45 4098.85775 116.07 1.941 E+00 2/19/2014

MCGUIRE NUCLEAR SITE

SUMMARY

OF MAJOR RADIONUCLIDE COMPOSITION 2013 I[Type of waste Niucli'de % Abundance Waste ftom liquid systems:

A. Dewatered Powdex Resin (brokered) No shipments in 2013 B. Dewatered Powdex Resin No shipments in 2013 C. Dewatered Bead Resin (brokered) No shipments in 2013

0. Dewatered Bead Resin No shipments in 2013 E. Dewatered Radwaste System Resin (brokered) No shipments in 2013 F. Dewatered Primary Bead Resin (brokered) No Shipments in 2013 G. Dewatered Mechanical Filter Media No shipments in 2013 H. Dewatered Mechanical Filter Media (brokered) No shipments in 2013 I. Solidified Waste No shipments in 2013

.Dry lidW7aste:-7 j A. Dry Active Waste (compacted) Com paction no longer performed on-site.

Dry Active Waste (non-compacted) No shipments in 2013 Dry Active Waste (brokered/compacted) No shipments in 2013 Dry Active Waste (brokered/non-compacted)

I

2013-0001 Nuclide %Abundance Cr-51 .94 Mn-54 2.74 Co-57 .08 Co-58 1.61 Co-60 43.36 Cs-137 .11 Fe-55 37.24 Ni-63 5.95 H-3 1.02 C-14 .05 Nb-95 2.44 Ce-144 .26 Sb-1 25 2.30 Zr-95 1.69 Sn-113 .20 2013-0002 Nuclide %Abundance Cr-51 1.19 Mn-54 2.77 Co-57 .08 Co-58 1.75 Co-60 42.74 Cs-137 .11 Fe-55 36.91 Ni-63 5.86 H-3 1.00 C-14 .05 Nb-95 2.91 Ce-144 .27 Sb-125 2.28 Zr-95 1.85 Sn-I 13 .21 2013-0003 Nuclide %Abundance Cr-51 .77 Mn-54 2.73 Co-57 .08 Co-58 1.48 Co-60 43.77 Cs-137 .12 Fe-55 37.56 Ni-63 6.07 H-3 1.03 C-14 .05 Nb-95 2.06 Ce-144 .26 Sb-125 2.31 Zr-95 1.54 Sn-113 .19 2

2013- 005 Nuclide %Abundance Cr-51 1.39 Mn-54 2.78 Co-57 .08 Co-58 1.84 Co-60 42.32 Cs-1 37 .11 Fe-55 36.63 Ni-63 5.81 H-3 .99 C-14 .05 Nb-95 3.28 Ce-144 .27 Sb-1 25 2.26 Zr-95 1.97 Sn-113 .21 2013- 006 Nuclide %Abundance Cr-51 30.13 Mn-54 1.53 Co-57 .06 Co-58 14.88 Co-60 10.98 Cs-137 .08 Fe-55 30.80 Fe-59 1.83 Ni-63 1.36 Nb-95 4.72 Ce-144 .29 Sb-1 24 .24 Zr-95 2.32 Sn-I 13 .21 Zn-65 .58 2013-008 Nuclide %Abundance Cr-51 29.02 Mn-54 1.56 Co-57 .06 Co-58 14.96 Co-60 11.27 Cs-137 .08 Fe-55 31.59 Fe-59 1.81 Ni-63 1.38 Nb-95 4.61 Ce-144 .29 Sb-124 .24 Zr-95 2.32 Sn-113 .21 Zn-65 .59 3

2013-009 Nuclide %Abundance Cr-51 28.28 Mn-54 1.59 Co-57 .06 Co-58 14.96 Co-60 11.46 Cs-1 37 .09 Fe-55 32.15 Fe-59 1.80 Ni-63 1.42 Nb-95 4.53 Ce-144 .30 Sb-124 .24 Zr-95 2.32 Sn-113 .21 Zn-65 .60 2013-010 Nuclide %Abundance Cr-51 28.24 Mn-54 1.59 Co-57 .06 Co-58 14.96 Co-60 11.45 Cs-1 37 .09 Fe-55 32.20 Fe-59 1.79 Ni-63 1.42 Nb-95 4.53 Ce-144 .30 Sb-124 .24 Zr-95 2.32 Sn-113 .21 Zn-65 .60 2013-013 Nuclide %Abundance Cr-51 27.35 Mn-54 1.62 Co-57 .06 Co-58 14.99 Co-60 11.72 Cs-1 37 .09 Fe-55 32.84 Fe-59 1.77 Ni-63 1.45 Nb-95 4.43 Ce-144 .30 Sb-124 .24 Zr-95 2.32 Sn-113 .21 Zn-65 .61 4

2013-017 Nuclide %Abundance Cr-51 26.88 Mn-54 1.63 Co-57 .06 Co-58 14.98 Co-60 11.81 Cs-I 37 .09 Fe-55 33.26 Fe-59 1.76 Ni-63 1.46 Nb-95 4.38 Ce-144 .30 Sb-124 .24 Zr-95 2.31 Sn-113 .21 Zn-65 .62 2013- 018 Nuclide %Abundance Cr-51 25.56 Mn-54 1.68 Co-57 .06 Co-58 15.03 Co-60 12.17 Cs-137 .09 Fe-55 34.22 Fe-59 1.73 Ni-63 1.51 Nb-95 4.24 Ce-144 .31 Sb-124 .24 Zr-95 2.30 Sn-113 .21 Zn-65 .63 2013-019 Nuclide %Abundance Cr-51 23.35 Mn-54 1.74 Co-57 .06 Co-58 14.94 Co-60 12.83 Cs-137 .10 Fe-55 35.98 Fe-59 1.67 Ni-63 1.60 Nb-95 4.00 Ce-144 .33 Sb-124 .23 Zr-95 2.29 Sn-113 .22 Zn-65 .65 5

2013-021 Nuclide %Abundance Cr-51 25.58 Mn-54 1.68 Co-57 .06 Co-58 14.98 Co-60 12.19 Cs-1 37 .09 Fe-55 34.22 Fe-59 1.73 Ni-63 1.51 Nb-95 4.25 Ce-144 .31 Sb-124 .24 Zr-95 2.31 Sn-113 .22 Zn-65 .63 2013- 024 Nuclide %Abundance Cr-51 18.00 Mn-54 1.92 Co-57 .07 Co-58 14.53 Co-60 14.57 Cs-I 37 .11 Fe-55 40.47 Fe-59 1.48 Ni-63 1.82 Nb-95 3.33 Ce-144 .36 Sb-124 .22 Zr-95 2.19 Sn-113 .23 Zn-65 .71 2013- 025 Nuclide %Abundance Mn-54 2.33 Co-57 .06 Co-58 .04 Co-60 61.77 Cs-1 37 .91 Fe-55 24.21 Ni-63 6.11 Ce-144 .46 Sb-125 3.81 Zr-95 .01 Sn-113 .02 Zn-65 .27 6

2013-027 Nuclide %Abundance Cr-51 .72 Mn-54 .13 Co-58 .85 Co-60 13.34 Cs-I 37 .82 Fe-55 17.79 Fe-59 .09 Ni-63 59.44 H-3 6.44 Nb-95 .15 Ce-144 .02 Sb-124 .01 Zr-95 .13 Sn-113 .01 Zn-65 .05 2013-028 Nuclide %Abundance Cr-51 16.62 Mn-54 1.90 Co-57 .07 Co-58 14.35 Co-60 14.43 Cs-137 .11 Fe-55 40.21 Fe-59 1.48 Ni-63 1.80 Nb-95 3.36 Ce-144 .35 Sb-124 .22 Zr-95 2.16 Sn-113 .22 Zn-65 .71 2013-029 Nuclide %Abundance Cr-51 17.67 Mn-54 1.93 Co-57 .07 Co-58 14.33 Co-60 14.75 Cs-1 37 .11 Fe-55 40.92 Fe-59 1.45 Ni-63 1.84 Nb-95 3.26 Ce-144 .36 Sb-124 .22 Zr-95 2.15 Sn-113 .22 Zn-65 .71 7

B. Sealed Sources No shipments in 2013 C. Sealed Sources/Smoke Detectors No shipments in 2013 D. Irradiated Components No shipments in 2013 8

Attachment 4 Meteorological Data

- Meteorological Data MNS 2013 JFD SECTOR N NNE NE EN E ESE SE SSE S SSW SW WSW W WNW NW NNW E

No. No. No. No. No. No. No. No. No. No. No. No. No. No. No. No.

A 0.46-0.75 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0.76-1.00 0 0 0 1 0 0 0 0 0 0 0 0 0 0 0 0 1.01-1.25 0 1 0 0 0 0 0 0 0 0 0 0 0 0 0 0 1.26-1.50 0 0 0 0 0 0 0 0 0 0 0 0 0 1 2 0 1.51-2.00 0 0 1 1 0 0 0 0 0 0 0 0 0 0 0 0 2.01-3.00 0 0 0 0 0 0 1 0 1 0 0 0 0 0 0 0 3.01-4.00 1 0 0 0 0 0 0 0 0 0 2 1 0 0 0 0 4.01-5.00 2 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 5.01-6.00 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 6.01-8.00 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 1 8.01-10.00 1 0 0 0 0 0 0 0 0 0 0 0 0 0 0 1 10.01-Max 4 2 0 0 0 0 0 0 0 0 0 0 0 0 0 0 B 0.46-0.75 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0.76-1.00 0 0 0 0 0o 0o 0 0 0 0 0 0 0 0 0 0 1.01-1.25 0 1 0 0 0 0 0 0 0 0 0 0 0 0 1 0 1.26-1.50 0 0 1 0 0 0 0 0 0 0 0 1 0 0 0 0 1.51-2.00 1 1 0 1 0 0 0 0 0 1 0 0 1 0 0 0 2.01-3.00 0 0 1 2 1 0 0 0 0 1 2 31 0 0 0 3.01-4.00 0 0 0 0 2 1 0 0 0 3 2 01 1 0 0

- Meteorological Data 4.01-5.00 0 0 0 0 0 0 0 0 1 1 3 1 0 0 0 0 5.01-6.00 1 1 0 0 0 0 0 0 0 0 2 0 0 0 1 1 6.01-8.00 0 1 0 0 0 0 0 0 0 0 0 0 0 0 1 1 8.01-10.00 3 1 0 0 0 0 0 0 0 0 0 0 0 0 0 2 10.01-Max 0 1 0 0 0 0 0 0 0 0 0 0 0 0 0 0 C 0.46-0.75 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0.76-1.00 0 0 0 0 0 0 0 0 0 0 01 0 0 0 0 0 1.01-1.25 0 1 0 0 0 0 0 0 0 0 0 0 0 0 0 0 1.26-1.50 0 1 0 0 0 0 0 0 0 0 0 0 0 1 0 1 1.51-2.00 4 0 3 3 1 1 0 0 00 1 0 0 1 1 1 0 2.01-3.00 2 4 4 9 3 2 1 1 4 14 5 11 2 0 1 1 3.01-4.00 1 3 2 8 3 041 4 15 17 10 2 3 2 0 4.01-5.00 3 4 4 9 0 1 0 1 1 8 21 13 2 0 0 3 5.01-6.00 2 0 1 1 0 0 0 0 0 3 7 2 0 0 2 3 6.01-8.00 9 2 0 0 0 0 0 0 0 0 0 1 0 1 1 10 8.01-10.00 5 2 0 0 0 0 0 0 0 0 0 0 0 0 3 4 10.01-Max 2 0 0 0 0 0 0 0 0 0 0 0 0 0 0 1 D 0.46-0.75 1 2 0 1 0 1 0 0 0 0 0 2 1 1 0 4 0.76-1.00 4 5 3 0 0 0 1 3 2 1 1 2 0 2 4 3 1.01-1.25 10 11 4 8 2 3 3 6 8 3 3 6 5 4 12 6 1.26- I 1.50 24 21 10 10 6 8 12 13 11 6 22 15 19 13 13 14 1.51-2.00 46 44 38 29 30 21 25 38 58 26 45 48 38 22 24 30 2.01-3.00 85 106 189 119 122 94 102 52 154 151 229 150 67 45

- Meteorological Data 3.01-4.00 53 86 308 147 125 47 10 9 50 116 265 105 53 40 55 35 4.01-5.00 69 101 201 63 35 8 4 6 13 52 114 36 25 22 50 62 5.01-6.00 63 70 88 14 9 3 0 0 3 16 43 9 9 19 36 42 6.01-8.00 59 47 18 1 0 3 0 0 2 9 18 13 6 16 32 61 8.01-10.00 10 5 3 0 00 0 0 0 0 0 0 0 3 9 10 13 10.01-Max 0 0 0 0 0 0 0 0 0 0 0 0 0 2 2 3 T0.46-0.75 0 0 0 0 1 0 1 3 2 1 3 2 0 1 1 0 0.76-1.00 0 3 3 0 1 3 3 3 5 13 8 6 7 2 2 1 1.01-1.25 0 2 6 3 5 1 6 10 11 9 14 12 14 5 4 3 1.26-1.50 1 2 1 2 2 5 12 20 15; 18 31 24 25 12 6 8 1.51-2.00 8 8 7 6 7 11 23 29 81 59 77 56 27 15 12 8 2.01-3.00 11 6 9 11 11 11 45 28 98 145 160 50 20 17 24 11 3.01-4.00 4 5 8 4 7 3 2 7 13 26 51 6 71 23 12 12 4.01-5.00 3 4 12 1 2 0 0 1 2 6 7 6 21 6 13 2 5.01-6.00 5 0 1 1 1 0 0 0 3 1 5 4 0 0 0 3 6.01-8.00 2 0 3 0 0 0 0 01 1 1 2 2 1 1 2 3 8.01-10.00 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 10.01-Max 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 F 0.46-0.75 0 0 0 1 1 0 1 1 2 8 6 4 3 1 0 0 0.76-1.00 1 1 0 1 0 0 3 5 8 18 15 12 7 1 0 0 1.01-1.25 1 0 2 0 0 0 2 0 10 16 11 14 8 2 1 0 1.26-1.50 0 1 0 0 0 0 0 10 14 17 14 10 6 3 0 0 1.51-2.00 2 1 1 1 0 0 4 9 38 25 14 17 14 4 2 0

- Meteorological Data 2.01-3.00 1 0 0 0 0 0 1 3 13 12 18 12 4 3 1 0 3.01-4.00 0 0 0 0 0 0 0 1 0 0 3 0 1 0 0 0 4.01-5.00 0 0 0 0 0 0 0 0 0 0 0 0 0 0 1 0 5.01-6.00 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 6.01-8.00 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 8.01-10.00 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 10.01-Max 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 G 0.46-0.75 0 0 0 0 0 0 0 0 0 2 6 0 0 0 0 0 0.76-1.00 1 0 0 0 0 0 0 1 4 19 22 10 7 1 0 0 1.01-1.25 0 0 0 0 0 0 0 1 3 19 28 6 3 1 0 0 1.26-1.50 0 0 0 0 0 0 0 1 4 19 7 3 3 0 0 0 1.51-2.00 0 00 0 0 0 0 0 0 7 10 1 2 0 0 0 0 2.01-3.00 0 0 0 0 0 0 0 0 1 2 6 0 0 0 0 0 3.01-4.00 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 4.01-5.00 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 5.01-6.00 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 6.01-8.00 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 8.01-10.00 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 10.01-Max 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0

Attachment 5 Unplanned Offsite Releases

Januazy 13,2013 Memorandum To: Annual Radioactive Effluent Release Report CC: Steve Mooneyhan, H. J. Sloan, Chris Whitener, C.D. Ingram, Kay Crane, Duncan Brewer From: William C. Spencer RP Staff Radiation Protection McGuire Nuclear Station Re: Unplanned release to the Unit 1 Vent Reference PIP M-13-0314 Event Summary:

See referenced PIP for details.

On 1/12/13 night shift Radwaste Chemistry perforning waste gas (WG) system monitoring of WG tank levels noted a volume loss of - ten psig over the previous three days from (WGDT-C). No indication of waste gas leakage was identified by Unit Vent gas monitors during this period. It was noted that a very slight increase was identified on sampling point # four on 0 EMF 41 (Aux building gas monitor). Gas sample grabs were obtained in the area of sample point # four and no radioactive noble gas was identified. The following day shift obtained gas sample from WGDT-C and secured NV VCT purges to the affected tank.

Operations Shift Manager (OSM) and Radiation Protection (RP) were notified and trending of Unit Vent gas monitor (IEMF 36) as well as the Auxiliary bid noble gas monitor (OEMF 41) were put into place on a more frequent basis. No radioactivity was identified during monitoring. Follow up leak investigation identified I WG 237 (B WG Compressor discharge isolation valve) as the source of the leak.

The total Noble gas activity released during the unplanned event (1.37E-2 Curies) was conservatively reported on Gaseous Waste Release (GWR) # 2013004. Two additional WGDTs E and B were involved with the leak investigation and small volume loss was identified. This was not considered unplanned.

Noble gas curies released from WGDT E and B was accounted for on GWRs 2013005 and 2013006.

The unplanned release of Noble gas radioactivity was evaluated against off site dose limits using current ODCM methodology on the attached spreadsheet. Procedure SRPMP 8-2 "Investigation of Unusual Radiological Occurrences" was completed to document the event.

Safety Significance:

The health and safety of the public were not compromised by this event. The total Noble gas activity released during the unplanned event was 0.0137 curies. Calculated dose and doserate to the Total Body, Skin, Gamma Air, and Beta Air were all orders of magnitude below the limits specified by Selected Licensee Commitments and Code of Federal Regulations.

W.C. Spencer Harry J Sloan RP Staff Support General Supervisor Radiation Protection Radiation Protection McGuire Nuclear Station McGuire Nuclear Station

Unplanned Release on 1113113 PIP M-13-0314 WGDT 'C' Release Summary SLC Limits Release % of Limit Total Body Doserate (mrem/yr) 500 6.87E-03 1.37E-03 Skin + Gamma Air Doserate (mrem/yr) 3000 1 .44E-02 4.80E-04 Gamma Air Dose (mrad) 40 1 6.36E-05 1.59E-04 Beta Air Dose (mrad) 20 7.55E-05 3.78E-041 Release Duration (sec) 2.75E+05 Release Volume (cu. ft.) 644.9 Release Duration (min) 4590 Average Release Rate (cu.ft. per min) 0.14 Average Release Rate (cu meters per sec) 6.63E-05 Average X/Q (sec per cu meter) 7.611E-05 Total Body ý Skin + 1.1*Gamn ia Air Gamma Air - Beta Air WGDT 'E' Sample ucilcc i Curies Cil/sec ICi/mA3 @ SB mrem/yr mrem/yr mrad mrad Kr-85 0.OOE+00 O.OOE+00 O0.OE+00 O.OOE+00 0.OOE+00 O.OOE+00 0.OOE+00 0.OOE+00 Kr-87 9.22E-07 1.68E-05 6.11E-111 4.65E-15 2.75E-05 _ 7.68E-05 2.51E-07 4.18E-07 Xe-133 5.66E-05 1.03E-03 3.76E-09 2.86E-13 8.40E-05 1.98E-04 8.81 E-07 2.62E-06 Xe-133 M 6.01 E-05 1.1OE-03 3.98E-091 3.03E-13 7.61E-05, 4.1OE-04 8.65E-07 3.92E-06 Xe-1 35 5.99E-04 1.09E-02 3.97E-081 3.02E-121 5.48E-03 1.20E-02 5.07E-05 6.50E-05 Kr-85M 1.66E-05 3.03E-04 1.10E-091 8.38E-14 9.81E-05 2.36E-04 9.OOE-07 1.44E-06 Kr-88 1.29E-05 2.36E-04 8.57E-10 6.53E-14 9.59E-04 1.25E-03 8.66E-06 1.67E-06 Ar-41 3.37E-06 6.15E-05 2.23E-101 1.70E-14 1.50E-04 2.20E-04 1.38E-06 4.87E-07 t

1.37E-02 1.44E-02 6.36E-05 7.55E-05 TotalI 6.87E-031I ODCM Values from Appendix A 1 Ki Li Mi Ni Kr-85 16.1 1340 17.2 1950 SKr-87 _ 5920 9730 6170 10300 - _

Xe-133 _ 294 306 353 1050 F Xe-133m 251 994 327 1480 Xe-135 - _ 1810 1860 1920 2460 Ki= Total Body Dose Factor Kr-85M 1170 1460 1230 1970 _ Li= Skin Dose Factor Kr-88 14700 2370 1 15200 2930 Mi= Gamma Air Dose Factor Ar-41 - 7 8840 2690 9300 3280 Ni= Beta Air Dose Factor

_ __ __ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __ _ _ _ _ _ _ _ __ _I_ _ _-_ _

Febmuiy20,2013 Memorandum To: Annual Radioactive Effluent Release Report CC: Steve Mooneyhan, H. J. Sloan, Chris Whitener, C.D. Ingram, Kay Crane, Duncan Brewer From: William C. Spencer RP Staff Radiation Protection McGuire Nuclear Station Re: Unplanned release to the Unit 1 Vent Reference PIP M-13-1568 Event Summary:

See referenced PIP for details.

On 2/18/13Radwaste Chemistry performing waste gas (WG) system monitoring of WG tank levels noted a volume loss of- four psig over the previous four days from (WGDT-E). No indication of waste gas leakage was identified by Unit Vent gas monitors or on 0 EMF 41 (Aux building gas monitor) during the period. The NV VCT purges to the affected tank were secured and WGDT-B was placed in service with the B WGDT compressor to allow for leak search on the B compressor components. During this evolution 1.3 psig volume was lost from the B WGDT. The leak was identified on components associated with the B WGDT compressor.

To properly shutdown the waste gas system the E tank was place back in service. During this period an additional volume loss of 1.9 psig occurred.

The total Noble gas activity released during the unplanned event from WGDT-E was conservatively estimated at 2.2E-2 curies of noble gas. The release was reported on Gaseous Waste Release (GWR) #

2013012. The volume loss from WGDT-B was involved with the leak investigation and small volume loss was identified. This was not considered unplanned. Noble gas curies released from WGDT-B was 2.88E-4 curies and accounted for on GWR# 2013013.

The unplanned release of Noble gas radioactivity was evaluated against off site dose limits using current ODCM methodology on the attached spreadsheet. Procedure SRPMP 8-2 "Investigation of Unusual Radiological Occurrences" was completed to document the event.

Safety Significance:

The health and safety of the public were not compromised by this event. Calculated dose and doserate to the Total Body, Skin, Gamma Air, and Beta Air were all orders of magnitude below the limits specified by Selected Licensee Commitments and Code of Federal Regulations.

W.C. Spencer Hary J Sloan RP Staff Support General Supervisor Radiation Protection Radiation Protection McGuire Nuclear Station McGuire Nuclear Station

Planned Release on 2118-2/19 PIP M-13-01568_

WGDT 'B' Release Summary SSLC Limits Release of Limit Total Body Doserate (mrem/yr) 500 1.12E-04 2.23E-05 Skin + Gamma Air Doserate (mrem/yr) - 3000 2.68E-04 8.94E-06 Gamma Air Dose (mrad) 40 3.49E-07 8.73E-07 -

Beta Air Dose (mrad) 20 8.51 E-07 4.25E-061 _

Release Duration (sec) 8.64E+04 _

Release Volume (cu. ft.) 53.1 Release Duration (min) 1440 ______ _

Average Release Rate (cu.ft. per min) 0.04 ____

Average Release Rate (cu meters per sec) 1.74E-05 ____________

Average X/Q (sec per cu meter) 7.611 E-05 Total Body Skin + 1.1*Gamma Air Gamma Air Beta Air WGDT 'E' Sample uci/cc Curies Cl/sec CCi/mA3 @ SB [mrem/yr inmrem/yr imrad mrad Kr-85 4.63E-06 6.96E-061 8.05E-1i __6.13E-15 9.87E-08 8.33E-06 2.89E-10 3.27E-08 Kr-87 0.OOE+00 0.OOE+00! 0.OOE+00 0.OOE+001 0.OOE+00 0.00E+00 0.OOE+00 0.OOE+00 Xe-133 1.63E-04 2.46E-04 2.84E-09 2.16E-13 6.36E-05 1.50E-04 2.09E-07 6.22E-07 Xe-133 M 4.20E-06 6.32E-06 7.31E-11 . 5.57E-15 1.40E-06 7.53E-06 4.98E-09 2.26E-08 Xe-135 1.92E-05 2.89E-05 3.34E-101 2.54E-14 4.61E-05 1.01E-04 1.34E-07 1.71E-07 Kr-85M 2.30E-07 3.46E-07 4.01E-12 3.05E-161 3.57E-07 8.59E-07 1.03E-09 I1.65E-09 Kr-88 0.OOE+00 .OOE+00E O0.OE+00 O.OOE+00I O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 Xe-131M O.OOE+00 0.OOE+00 O.OOE+O0 O.OOE+00 0.OOE+00 O.OOE+00 .OOE+00 O.OOE+00

____ 1____________ ________________

Total I

2.88E-041 i 1.12E-04 2.68E-04 I 3.49E-07 i

8.51E-07 ODCM Values from Appendix A Ki Li I Mi Ni Kr-85 _ 16.1- 1340 17.2 _ 1950 I _

Kr-87 5920 9730 6170 10300 Xe-133 _ 294 306 353 1050 _ _

Xe-133m 251 994 327 1480 __ _

Xe-i 35 1810 1860 1920 2460 Ki= Total Body Dose Factor Kr-85M 1170 1460 1230 1970 Li= Skin Dose Factor -

Kr-88 14700 2370 15200 2930 Mi= Gamma Air Dose Factor Xe-131M 91.5 476 156 1110 Ni= Beta Air Dose Factor

Unplanned Rel on 2113113-2119113 _

PIP M-13-01568 _

WGDT 'E' Release Summary 3LC Limits tRelease 01 of Limit Total Body Doserate (mrem/yr) 500 1.74E-03 3.47E-04 Skin + Gamma Air Doserate (mrem/yr) 3000 4.01 E-03 1.34E-04 Gamma Air Dose (mrad) 40 2.83E-05 7.08E-05 Beta Air Dose (mrad) 20 6.46E-05 3.23E-04 Release Duration (sec) 4.52E+05 ___

Release Volume (cu. ft.) 249 __,

Release Duration (min) 7528 _____

Average Release Rate (cu.ft. per min) 0.03 Average Release Rate (cu meters per sec) 1.56E-05 Average X/Q (sec per cu meter) 7.611E-05 Tota IlBody Skin + 1.1*Gamma Air Gamma Air Beta Air WGDT 'E' Sample ucilcc Curies Cil/sec Ci/MA3§SB mn em/yr mrem/yr mrad mrad Kr-85 O.OOE+00 O.OOE+O- O.OOE+00 O.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 Kr-87 0.OOE+00 0.00E+001 0.OOE+00 0.00E+00 0. 00E+00 0.00E+0 0.00E+00 0.00E+00 Xe- 133 2.70E-03 1.90E-02 4.21E-08 3.20E-12 9 .42E-04 2.23E-03 1.62E-05 4.82E-05 Xe-133 M 6.76E-05 4.77E-04 1.06E-09 8.04E-14 2 .02E-05 1.09E-04 3.76E-07 1.70E-06 Xe- 135 3.46E-04 2.44E-03 5.40E-09 4.11E-13 7 A44E-04 1.63E-03 1.13E-05 1.45E-05 Kr-85M 4.37E-06 3.08E-05 6.82E-11 5.19E-15 6 .08E-06' 1.46E-05 9.15E-08 1.47E-07 Kr-88 1.26E-06 8.90E-06, 1.97E-1 1 1.50E-115 2 .20E-05 2.86E-05 / 3.26E-07 6.29E-08 Ar-41 1.53E-07 1.08E-061 2.38E-1 2 1.81E-16 1 .60E-06 2.34E-06 2.42E-08 L 8.52E-09 Total II 1.74E-03 4.01 E-03 2.20E-02i ODCM Values from Appendix A Ki Li Mi Ni Kr-85 16.1 1340 17.2 1950 Kr-87 5920 9730 Xe- 133 294 306 Xe-1 33m Xe-1i35 Ki= Total Body Dose Factor Kr-85M Li= Skin Dose Factor Kr-88 Mi= Gamma Air Dose Factor Ar-41 Ni= Beta Air Dose Factor

August 15,2013 Memorandum To: Annual Radioactive Effluent Release Report CC: Steve Mooneyhan, Chris Whitener, C.D. Ingram, Kay Crane, Jeff Robertson, Jeff Nolin From: William C. Spencer RP Staff Radiation Protection McGu.iL Nuclear Station Re: Unplanned release to the Unit 1 Vent Reference PIP M-13-7959 Event Summary:

See referenced PIP for details.

On 8/15/13 Radwaste Chemistry was performing a planned waste gas decay tank E (WGDT-E) release. In conjunction with the release a correlation check of OEMF 50L was also planned which requires a gas sample of the release discharge stream.

The prior to release gas sample identified one nuclide, Kr-85 @ 4.45E-6 uci/ml. The release was initiated at 11: 14 hrs and OEMF 50L reached it's expected counts per minute (cpm) of -124 cpm. All release parameters were normal. A gas sample was collected from the discharge stream at OEMF 50L (monitoring instrument) at 11:32 hrs and analyzed. The analysis indicated two additional nuclides Xe-133 @2.48E-7uci/ml and Xe-135

@5.73E-8 uci/ml. The release was terminated due to the unexpected identification of two shorter lived nuclides in the discharge stream at 12:48 hours. A maximum of 132 counts per minute above background (ccpm) was seen on OEMF-50L during the release. The OEMF 50L expected was 124 ccpm. OEMF 50L trip 2 alarm setpoint was established at 350 cpin and was not challenged. Approximately 25 cpm above bkg was seen on the unit vent gas monitor which is typical for this level of activity and is most likely a response to Carbon-14.

Carbon-14 is a beta emitter constituent of the WGDT storage system. No SLC or T/S limits were challenged.

The total Noble gas activity released during the event from WGDT-E was determined to be 1.68E-4 Curies.

The release was reported on Gaseous Waste Release (GWR) # 2013063. The quantity of Kr-85 released (1.57E-4 Curies) was not considered unplanned but included in the evaluation for conservancy.

The total release of Noble gas radioactivity was evaluated against off site dose limits using current ODCM methodology on the attached spreadsheets. Procedure SRPMP 8-2 "Investigation of Unusual Radiological Occurrences" was completed to document the event.

Safety Significance:

The health and safety of the public were not compromised by this event. Calculated dose and doserate to the Total Body, Skin, Gamma Air, and Beta Air were all orders of magnitude below th limits specified by Selected Licensee C~mmitments and Code of Federal Regulations.

W.C. Spencer Willard Osburn RP Staff Support General Supervisor Radiation Protection Radiation Protection McGuire Nuclear Station McGuire Nuclear Station

Unplanned Rel on 8115113 PIP M-13-07959 WGDT 'E' Release Summary SLC Umits Release %ofUmit Total Body Doserate (mrem/yr)_ 500 1.18E-04 2.361E-05 Skin + Gamma Air Doserate (mremlyr) 3.07E-03 I .02E-04i Gamma Air Dose (mrad) 40 2.33E-08 5.83E-08,-

Beta Air Dose (mrad) 20 7.72E-07 3.86E-06, Release Duration (sec) 5.64E403 Release Volume (cu. ft.) 1245 Release Duration (min) 94 Average Release Rate (cu.ft. per min) 13.24 Average Release Rate (cu meters per sec) 6.25E-03 Average X/Q (sec per cu meter) 7.611 E-05 Total Body Skin + I.IGamma Air Gamma Air Beta Air WGDT 'E' Sample ucl/cc Curl.e ClIsec CIm3@SB mremlyr jmremlyr mrad mrad Kr-85 4.45E-06 1.57E-04 2.78E-08 2.12E-12; 3.41 E-05 2.88E-03 6.51 E-09 -7.38E-07 Kr-87 0.OOE+00 0.OOE+O0 O.OOE+O0 o.OOE+00 0.OOE+00 0.00E+00 o.OOE+00 0.OOE.00 Xe-133 2.48E-07 8.74E-06 1.55E-09 1.18E-13 3.47E-05 8.19E-05 7.45E-09 2.22E-08 Xe-133 M 0.OOE+00 0.00E+00 0.OOE+00 O.OOE+00: 0.OOE+00 O.OOE+00 O.OOE+00 0.OOE+OO Xe-135 5.73E-08 2.02E-06 3.58E-1 0 2,73E-14 4.93E-05 1.08E-04 9.36E-09 1.20E-08 Kr-85M o0OOE+00 0.00E+0O O.OOE+00 0.OOE+O00 0.00E+00 0.OOE+00 0.OOE+00 0.OOE+00 Kr-88 0.OOE+00 O.OOE+00i O.OOE+00 O.OOE+O0 0.OOE+00 O.OOE+00 0.OOE+00 0.OOE+00 Ar-41 0.OOE+00 O.OOE+00 0ý.OOE+00 O.00E+00 0.OOE+00 0 OOE+00 0.OOE+00 O.OOE+00 Total 1.68E-04 1.I1E-04 3.07E-03 2.33E-08 7.72E-07 ODCM Values from Appendix A Ki U Mi Ni Kr-85 16.1 1340 17.2 1950 Kr-87 5920 9730 6170 10300 Xe-133 294 306 353 1050 Xe-1 33m 251 994 327 1480 Xe-135 1810 1860 1920 2460 Ki= Total Body Dose Factor Kr-85M 1170 1460 1230 1970 L= Skin Dose Factor Kr-_. 14700 2370 15200 2930 Mi= Gamma Air Dose Factor Ar-41 8840 2690 9300 3280 Ni= Beta Air Dose Factor

ATTACHMENT 6 Assessment of Radiation Dose from Radioactive Effluents to Members of the Public (includes fuel cycle dose calculation results)

This attachment includes an assessment of radiation doses to the maximum exposed member of the public due to radioactive liquid and gaseous effluents released from the site for each calendar quarter for the calendar year of the report as well as the total dose for the calendar year.

This attachment also includes an assessment of radiation doses to the maximum exposed member of the public from all uranium fuel cycle sources within 8 km of the site for the calendar year of this report to show conformance with 40 CFR 190.

Methods for calculating the dose contribution from liquid and gaseous effluents are given in the Offsite Dose Calculation Manual (ODCM).

EFFLUENT AND WASTE DISPOSAL ANNUAL REPORT PERIOD 1/1/13 TO 1/1/14 GASEOUS ANNUAL DOSE

SUMMARY

REPORT McGuire Nuclear Station Units 1 & 2 1it Quarter 2013

- IODINE, H3, AND PARTICULATE DOSE LIMIT ANALYSIS -- Quarter 1 2013 Critical Critical Dose Limit Max % of Period-Limit Group Organ (mrem) (mrem) Limit Q1 - Maximum Organ Dose CHILD BONE 2.14E-01 1.50E+01 1.43E+00 Maximum Organ Dose Receptor Location: 1.5 Mile NE Critical Pathway: Vegetation Major Isotopic Contributors (5% or greater to total)

Nuclide Percentage C-14 1.00E+02 NOBLE GAS DOSE LIMIT ANALYSIS Quarter 1 2013 =

Dose Limit % of Period-Limit (mrad) (mrad) Limit Q1 - Maximum Gamma Air Dose 1.10E-02 1.OOE+01 1.10E-01 Maximum Gamma Air Dose Receptor Location: 0.5 Mile NNE Major Isotopic Contributors (5% or greater to total)

Nuclide Percentage AR-41 9.88E+01 Ql - Maximum Beta Air Dose 4.07E-03 2.OOE+01 2.03E-02 Maximum Beta Air Dose Receptor Location: 0.5 Mile NNE Major Isotopic Contributors (5% or greater to total)

Nuclide Percentage AR-41 9.44E+01

EFFLUENT AND WASTE DISPOSAL ANNUAL REPORT PERIOD 1/1/13 TO 1/1/14 GASEOUS ANNUAL DOSE

SUMMARY

REPORT McGuire Nuclear Station Units 1 & 2 22d Quarter 2013 IODINE, H3, AND PARTICULATE DOSE LIMIT ANALYSIS Quarter 2 2013 Critical Critical Dose Limit Max % of Period-Limit Group Organ (mrem) (mrem) Limit Q2 - Maximum Organ Dose CHILD BONE 2.11E-01 1.50E+01 1.41E+00 Maximum Organ Dose Receptor Location: 1.5 Mile NE Critical Pathway: Vegetation Major Isotopic Contributors (5% or greater to total)

Nuclide Percentage C-14 1.OOE+02 NOBLE GAS DOSE LIMIT ANALYSIS Quarter 2 2013 Dose Limit % of Period-Limit (mrad) (mrad) Limit Q2 - Maximum Gamma Air Dose 8.01E-03 1.00E+01 8.01E-02 Maximum Gamma Air Dose Receptor Location: 0.5 Mile NNE Major Isotopic Contributors (5% or greater to total)

Nuclide Percentage AR-41 9.97E+01 Q2 - Maximum Beta Air Dose 2.87E-03 2.00E+01 1.44E-02 Maximum Beta Air Dose Receptor Location: 0.5 Mile NNE Major Isotopic Contributors (5% or greater to total)

Nuclide Percentage AR-41 9.80E+01

EFFLUENT AND WASTE DISPOSAL ANNUAL REPORT PERIOD 1/1/13 TO 1/1/14 GASEOUS ANNUAL DOSE

SUMMARY

REPORT McGuire Nuclear Station Units 1 & 2 3rd Quarter 2013

- IODINE, H3, AND PARTICULATE DOSE LIMIT ANALYSIS Quarter 3 2013 Critical Critical Dose Limit Max % of Period-Limit Group Organ (mrem) (mrem) Limit Q3 - Maximum Organ Dose CHILD BONE 2.47E-01 1.50E+01 1.65E+00 Maximum Organ Dose Receptor Location: 1.5 Mile NE Critical Pathway: Vegetation Major Isotopic Contributors (5% or greater to total)

Nuclide Percentage C-14 1.OOE+02

- NOBLE GAS DOSE LIMIT ANALYSIS Quarter 3 2013 Dose Limit % of Period-Limit (mrad) (mrad) Limit Q3 - Maximum Gamma Air Dose 9.35E-03 1.OOE+01 9.35E-02 Maximum Gamma Air Dose Receptor Location: 0.5 Mile NNE Major Isotopic Contributors (5% or greater to total)

Nuclide Percentage AR-41 9.96E+01 Q3 - Maximum Beta Air Dose 3.37E-03 2.00E+01 1.68E-02 Maximum Beta Air Dose Receptor Location: 0.5 Mile NNE Major Isotopic Contributors (5% or greater to total)

Nuclide Percentage AR-41 9.75E+01

EFFLUENT AND WASTE DISPOSAL ANNUAL REPORT PERIOD 1/1/13 TO 1/1/14 GASEOUS ANNUAL DOSE

SUMMARY

REPORT McGuire Nuclear Station Units 1 & 2 4S Quarter 2013 IODINE, H3, AND PARTICULATE DOSE LIMIT ANALYSIS Quarter 4 2013 Critical Critical Dose Limit Max % of Period-Limit Group Organ (mrem) (mrem) Limit Q4 - Maximum Organ Dose CHILD BONE 2.47E-01 1.50E+01 1.65E+00 Maximum Organ Dose Receptor Location: 1.5 Mile NE Critical Pathway: Vegetation Major Isotopic Contributors (5% or greater to total)

Nuclide Percentage C-14 1.00E+02

- NOBLE GAS DOSE LIMIT ANALYSIS Quarter 4 2013 Dose Limit % of Period-Limit (mrad) (mrad) Limit Q4 - Maximum Gamma Air Dose 1.03E-02 1.00E+01 1.03E-01 Maximum Gamma Air Dose Receptor Location: 0.5 Mile NNE Major Isotopic Contributors (5% or greater to total)

Nuclide Percentage AR-41 9.93E+01 Q4 - Maximum Beta Air Dose 3.67E-03 2.00E+01 1.88E-02 Maximum Beta Air Dose Receptor Location: 0.5 Mile NNE Major Isotopic Contributors (5% or greater to total)

Nuclide Percentage AR-41 9.58E+01

EFFLUENT AND WASTE DISPOSAL ANNUAL REPORT PERIOD 1/1/13 TO 1/1/14 GASEOUS ANNUAL DOSE

SUMMARY

REPORT McGuire Nuclear Station Units 1 & 2 ANNUAL 2013 IODINE, H3, AND PARTICULATE DOSE LIMIT ANALYSIS Annual 2013 Critical Critical Dose Limit Max % of Period-Limit Group Organ (mrem) (mrem) Limit Yr - Maximum Organ Dose CHILD BONE 9.20E-01 3.OOE+01 3.07E+00 Maximum Organ Dose Receptor Location: 1.5 Mile NE Critical Pathway: Vegetation Major Isotopic Contributors (5% or greater to total)

Nuclide Percentage C-14 1.OOE+02 NOBLE GAS DOSE LIMIT ANALYSIS Annual 2013 Dose Limit % of Period-Limit (mrad) (mrad) Limit Yr - Maximum Gamma Air Dose 3.87E-02 2.OOE+01 1.93E-01 Maximum Gamma Air Dose Receptor Location: 0.5 Mile NNE Major Isotopic Contributors (5% or greater to total)

Nuclide Percentage AR-41 9.93E+01 Yr - Maximum Beta Air Dose 1.41E-02 4.OOE+01 3.52E-02 Maximum Beta Air Dose Receptor Location: 0.5 Mile NNE Major Isotopic Contributors (5% or greater to total)

Nuclide Percentage AR-41 9.63E+01

EFFLUENT AND WASTE DISPOSAL ANNUAL REPORT PERIOD 1/1/13 TO 1/1/14 LIQUID ANNUAL DOSE

SUMMARY

REPORT McGuire Nuclear Station Units 1 & 2 12-t Quarter 2013

- BATCH LIQUID RELEASES Quarter 1 2013 Critical Critical Dose Limit Max % of Period-Limit Age Organ (mrem) (mrem) Limit Qi - Maximum Organ Dose CHILD LIVER 6. 38E-02 1.OOE+01 6.38E-01 Qi - Total Body Dose CHILD 6. 31E-02 3.OOE+00 212E+00 Maximum Organ Critical Pathway: Potable Water Major Isotopic Contributors (5% or greater to total)

Nuclide Percentage H-3 9.98E+01 Total Body Critical Pathway: Potable Water Major Isotopic Contributors (5% or greater to total)

Nuclide Percentage H-3 9.99E+01

- CONTINUOUS LIQUID RELEASES (WC) Quarter 1 2013 Critical Critical Dose Limit Max % of Period-Limit Age Organ (mrem) (mrem) Limit Ql - Maximum Organ Dose CHILD LIVER 6.84E-04 1.00E+01 6.84E-03 Q1 - Total Body Dose CHILD 6.84E-04 3.OOE+00 2.28E-02 Maximum Organ Critical Pathway: Potable Water Major Isotopic Contributors (5% or greater to total)

Nuclide Percentage H-3 1.OOE+02 Total Body Critical Pathway: Potable Water Major Isotopic Contributors (5% or greater to total)

Nuclide Percentage H-3 1.00E+02

EFFLUENT AND WASTE DISPOSAL ANNUAL REPORT PERIOD 1/1/13 TO 1/1/14 LIQUID ANNUAL DOSE

SUMMARY

REPORT McGuire Nuclear Station Units 1 & 2 22-d Quarter 2013

- BATCH LIQUID RELEASES Quarter 2 2013 -

Critica 1 Critical Dose Limit Max % of Period-Limit Age Organ (mrem) (mrem) Limit Q2 - Maximum Organ Dose CHILD GI-LLI 1.97E-02 1.00E+01 1. 97E-01 Q2 - Total Body Dose CHILD 196E-02 3_00E+00 6. SSE-01 Maximum Organ Critical Pathway: Potable Water Major Isotopic Contributors (5% or greater to total)

Nuclide Percentage H-3 9.94E+01 Total Body Critical Pathway: Potable Water Major Isotopic Contributors (5% or greater to total)

Nuclide Percentage H-3 9.98E+01 CONTINUOUS LIQUID RELEASES (WC) Quarter 2 2013 Critical Critical Dose Limit Max % of Period-Limit Age Organ (mrem) (mrem) Limit Q2 - Maximum Organ Dose CHILD LIVER 8.94E-04 1.00E+01 8. 94E-03 Q2 - Total Body Dose CHILD 8.94E-04 3.00E+00 2. 98E-02 Maximum Organ Critical Pathway: Potable Water Major Isotopic Contributors (5% or greater to total)

Nuclide Percentage H-3 1.OOE+02 Total Body Critical Pathway: Potable Water Major Isotopic Contributors (5% or greater to total)

Nuclide Percentage H-3 1.00OE+02

EFFLUENT AND WASTE DISPOSAL ANNUAL REPORT PERIOD 1/1/13 TO 1/1/14 LIQUID ANNUAL DOSE

SUMMARY

REPORT McGuire Nuclear Station Units 1 & 2 3Ld Quarter 2013 BATCH LIQUID RELEASES Quarter 3 2013 Critical Critical Dose Limit Max % of Period-Limit Age Organ (mrem) (mrem) Limit Q3 - Maximum Organ Dose CHILD GI-LLI 1.25E-02 1.00E+01 1.25E-01 Q3 - Total Body Dose CHILD 1.23E-02 3.OOE+00 4.09E-01 Maximum Organ Critical Pathway: Potable Water Major Isotopic Contributors (5% or greater to total)

Nuclide Percentage H-3 9.77E+01 Total Body Critical Pathway: Potable Water Major Isotopic Contributors (5% or greater to total)

Nuclide Percentage H-3 9.95E+01 CONTINUOUS LIQUID RELEASES (WC) Quarter 3 2013 Critical Critical Dose Limit Max % of Period-Limit Age Organ (mrem) (mrem) Limit Q3 - Maximum Organ Dose CHILD LIVER 1.25E-04 1.OOE+01 1.25E-03 Q3 - Total Body Dose CHILD 1.25E-04 3.OOE+00 4.17E-03 Maximum Organ Critical Pathway: Potable Water Major Isotopic Contributors (5% or greater to total)

Nuclide Percentage H-3 1.00E+02 Total Body Critical Pathway: Potable Water Major Isotopic Contributors (5% or greater to total)

Nuclide Percentage H-3 1.OOE+02

EFFLUENT AND WASTE DISPOSAL ANNUAL REPORT PERIOD 1/1/13 TO 1/1/14 LIQUID ANNUAL DOSE

SUMMARY

REPORT McGuire Nuclear Station Units 1 & 2 4ý- Quarter 2013

- BATCH LIQUID RELEASES Quarter 4 2013 -

Critical Critical Dose Limit Max % of Period-Limit Age Organ (mrem) (mrem) Limit Q4 - Maximum Organ Dose CHILD LIVER 5.65E-02 1.OOE+01 5.65E-01 Q4 - Total Body Dose CHILD 5. 64E-02 3.OOE+00 1. SE+00 Maximum Organ Critical Pathway: Potable Water Major Isotopic Contributors (5% or greater to total)

Nuclide Percentage H-3 9.99E+01 Total Body Critical Pathway: Potable Water Major Isotopic Contributors (5% or greater to total)

Nuclide Percentage H-3 9.99E+01 CONTINUOUS LIQUID RELEASES (WC) Quarter 4 2013 -

Critical Critical Dose Limit Max % of Period-Limit Age Organ (mrem) (mrem) Limit Q4 - Maximum Organ Dose CHILD LIVER 3.54E-04 1.00E+01 3.54E-03 Q4 - Total Body Dose CHILD 3.54E-04 3.OOE+00 1. 18E-02 Maximum Organ Critical Pathway: Potable Water Major Isotopic Contributors (5% or greater to total)

Nuclide Percentage H-3 1. OOE+02 Total Body Critical Pathway: Potable Water Major Isotopic Contributors (5% or greater to total)

Nuclide Percentage H-3 1.OOE+02

EFFLUENT AND WASTE DISPOSAL ANNUAL REPORT PERIOD 1/1/13 TO 1/1/14 LIQUID ANNUAL DOSE

SUMMARY

REPORT McGuire Nuclear Station Units 1 & 2 ANNUAL 2013

- BATCH LIQUID RELEASES Annual 20:13 Critical Critical Dose Limit Max % of Period-Limit Age Organ (mrem) (mrem) Limit Yr - Maximum Organ Dose CHILD GI-LLI 1.47E-01 2.OOE+01 7.36E-01 Yr - Total Body Dose CHILD 1.47E-01 6.OOE+00 2.45E+00 Maximum Organ Critical Pathway: Potable Water Major Isotopic Contributors (5% or greater to total)

Nuclide Percentage H-3 9.96E+01 Total Body Critical Pathway: Potable Water Major Isotopic Contributors (5% or greater to total)

Nuclide Percentage H-3 9.99E+01 CONTINUOUS LIQUID RELEASES (WC) Annual 2013 Critical Critical Dose Limit Max % of Period-Limit Age Organ (mrem) (mrem) Limit Yr - Maximum Organ Dose CHILD LIVER 2.06E-03 2.00E+01 1.03E-02 Yr - Total Body Dose CHILD 2.06E-03 6.OOE+00 3.43E-02 Maximum Organ Critical Pathway: Potable Water Major Isotopic Contributors (5% or greater to total)

Nuclide Percentage H-3 1.OOE+02 Total Body Critical Pathway: Potable Water Major Isotopic Contributors (5% or greater to total)

Nuclide Percentage H-3 1.OOE+02

McGuire Nuclear Station 2013 Radioactive Effluent and ISFSI 40CFR190 Uranium Fuel Cycle Dose Calculation Results In accordance with the requirements of 40CFR190, the annual dose commitment to any member of the general public shall be calculated to assure that doses are limited to 25 millirems to the total body or any organ with the exception of the thyroid which is limited to 75 millirems. The fuel cycle dose assessment for McGuire Nuclear Station only includes liquid and gaseous effluent dose contributions from McGuire and direct and air-scatter dose from McGuire's onsite Independent Spent Fuel Storage Installation (ISFSI) since no other uranium fuel cycle facility contributes significantly to McGuire's maximum exposed individual. Included in the gaseous effluent dose calculations is an estimate of the dose contributed by Carbon-14 (Ref. "Carbon-14Supplemental Information", contained in the ARERR for further information). The combined dose to a maximum exposed individual from McGuire's effluent releases and direct and air-scatter dose from McGuire's ISFSI is below 40CFR190 limits as shown by the following summary:

I. 2013 McGuire 40CFR190 Effluent Dose Summary The 40CFR190 effluent dose analysis to the maximum exposed individual from liquid and gas releases includes the dose from noble gases (i.e., total body and skin).

Maximum Total Body Dose = 4.07E-01 mrem Maximum Location: 1.5 Mile, Northeast Sector Critical Age: Child Gas non-NG Contribution: 63.32%

Gas NG Contribution: 0.67%

Liquid Contribution: 36.01%

Maximum Organ (other than TB) Dose = 9.20E-01 mrem Maximum Location: 1.5 Mile, Northeast Sector Critical Age: Child Critical Organ: Bone Gas Contribution: 99.95%

Liquid Contribution: 0.05%

II. 2013 McGuire 40CFR190 ISFSI Dose Summary Direct and air-scatter radiation dose contributions from the onsite Independent Spent Fuel Storage Installation (ISFSI) at McGuire have been calculated and documented in the "McGuire Nuclear Site 1OCFR72.212 Evaluation Report". The maximum dose rate to the nearest real individual from the McGuire ISFSI is conservatively calculated to be less than 4 mrem/yr.

The attached excerpt from the "McGuire Nuclear Site 10CFR72.212 Evaluation Report" is provided to document the method used to calculate the McGuire ISFSI less than 4 mrem/year dose estimate to the nearest real individual.

6.0 10 CFR 72.212(b)(5)(iii) -Radioactive Materials in Effluents and Direct Radiation 6.1 Purpose 10 CFR 72.212(b)(5)(iii) requires the general licensee to perform written evaluations, before use and before applying the changes authorized by an amended CoC to a cask loaded under the initial CoC or an earlier amended CoC, that establish that the requirements of 10 CFR 72.104 have been met. A copy of this record shall be retained until spent fuel is no longer stored under the general license issued under 10 CFR 72.210.

10 CFR 72.104 provides the regulatory criteria for radioactive materials in effluents and direct radiation from an independent spent fuel storage installation (ISFSI) during normal operation and anticipated occurrences.

Specifically, 10 CFR 72.104(a) limits the annual dose equivalent to any real individual who is located beyond the controlled area to 25 mrem to the whole body, 75 mrem to the thyroid, and 25 mrem to any other critical organ. This dose equivalent must include contributions from (1) planned discharges of radioactive materials (radon and its decay products excepted) to the general environment, (2) direct radiation from ISFSI operations, and (3) any other radiation from uranium fuel cycle operations within the region. In addition, 10 CFR 72.104(b) requires that operational restrictions be established to meet as low as is reasonably achievable (ALARA) objectives for radioactive materials in effluents and direct radiation levels associated with ISFSI operations. Also, 10 CFR 72.104(c) requires that operational limits be established for radioactive materials in effluents and direct radiation levels associated with ISFSI operations to meet the above-mentioned dose limits.

This section provides the written evaluation required by 10 CFR 72.212(b)(5)(iii), demonstrating Duke Energy's compliance with the requirements of 10 CFR 72.104 for the MNS ISFSI.

6.2 Evaluation This evaluation addresses the radiological dose rate from a composite population of all MNS ISFSI cask types.

6.2.1 §72.104(a)-Dose Limits Duke Energy Engineering Instruction MCEI-0400-241 determined that the distance from the nearest residence to the ISFSI is 0.65 miles (1046 meters). Hence, it is conservative to assume that the closest real individual is at least 700 meters from the ISFSI.

Enercon determined the annual total dose (gamma plus neutron) at a distance of 700 meters from all currently loaded casks (10 TN-32A casks and 28 NAC-UMScasks) to be approximately 1.62 mrem. The evaluation was based on actual cask average bum-up (as loaded) and considering cooling time on the storage pads as of September 1, 2010. The distance at which this dose is

calculated (700 meters) is conservative compared to the distance to the closest real individual.

NAC International determined the annual total dose (gamma plus neutron) at a distance of 700 meters from a (future) 2x6 array of MAGNASTOR casks to be approximately 1.01 mrem (2.02 mrem for two arrays). The evaluation was conservatively based on full cask loads of 37 fuel assemblies at the maximum allowable heat load of 35.5 kW. The distance at which this dose is calculated (700 meters) is conservative compared to the distance to the closest real individual.

The total calculated annual public dose from liquid and gaseous effluent pathways averaged over a ten-year period is less than 1 mrem. No other uranium fuel cycle facility contributes significantly to the dose received by the closest real individual.

Based on the above, the calculated annual dose to the closest real individual due to the ISFSI, which is comprised of the currently existing ten TN-32A casks and 28 NAC-UMScasks, and up to two 2x6 arrays of MAGNASTOR casks (see Note below), is determined to be less than 4 mrem, and the estimated annual dose due to McGuire power generation is less than 1 mrem. Hence, the total annual dose to the closest real individual (less than 5 mrem) is within the 10 CFR 72.104(a) limit.

Note: As stated above, up to two 2x6 arrays of MAGNASTOR casks are assumed in this evaluation. The first eight MAGNASTOR casks are planned to be placed on a concrete pad currently containing four NAC-UMS casks. This will conservatively count as one 2x6 array. Additional MAGNASTOR casks will be placed on their own concrete pad (the second 2x6 array). Hence, this §72.104(a) evaluation bounds up to 20 MAGNASTOR casks, arranged as described.

Attachment 7 Revisions to the Updated Final Safety Analysis Report Radiological Effluent Controls Section 16.11

Liquid Effluents - Concentration 16.11.1 16.11 RADIOLOGICAL EFFLUENT CONTROLS 16.11.1 Liquid Effluents - Concentration COMMITMENT The concentration of radioactive material released in liquid effluents to UNRESTRICTED AREAS (see Figure 16.11.1-1) shall be limited:

a. For radionuclides other than dissolved or entrained noble gases, 10 times the effluent concentrations specified in 10 CFR Part 20, Appendix B, Table 2, Column 2, and

b. For dissolved or entrained noble gases, the concentration shall be limited to 2 x 10-4 microCurie/ml total activity.

APPLICABILITY At all times.

REMEDIAL ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Concentration of A.1 Restore the concentration Immediately radioactive material to within limits.

released in liquid effluents to UNRESTRICTED AREAS not within limits.

TESTING REQUIREMENTS TEST FREQUENCY TR 16.11.1.1 ------------------- NOTE-----------------

The results of the radioactivity analyses shall be used in accordance with the methodology and parameters in the ODCM to assure that the concentrations at the point of release are maintained within the limits.

Sample and analyze radioactive liquid wastes according According to to Table 16.11.1-1. Table 16.11.1-1 McGuire Units 1 and 2 16.11.1-1 Revision 137

Liquid Effluents - Concentration 16.11.1 TABLE 16.11.1-1 (Page 1 of 3)

RADIOACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM LIQUID RELEASE SAMPLING MINIMUM TYPE OF LOWER LIMIT TYPE FREQUENCY ANALYSIS ACTIVITY OF DETECTION FREQUENCY ANALYSIS (LLD) microCi/ml (1)

1. Batch Waste P P Principal Release Tanks Each Batch Each Batch Gamma 5x10-7 (WMT and Emitters(6) 4 RMT)( )

1-131 1x10-6 P M Dissolved and One Batch/M Entrained 1X1 0.5 Gases (Gamma emitters)(7)

P M H-3 1x100-Each Batch Composite (2)

Gross Alpha lx10-7 P Q Sr-89, Sr-90 5xl 0-8 Each Batch Composite (2)

2. Continuous Continuous(3) W Principal Releases Composite(3 ) Gamma 5xl0-(VUCDT Emitters (6) discharge, CWWTS outlet and Turbine Building Sump to RC)(5) 1-131 1x10-6 M M Dissolved and Grab Sample Entrained lx10-i Gases (Gamma 7

emitters)( )

Continuous(3) M H-3 1xl 0s Composite(3)

Gross Alpha 1x10.7 Continuous(3) Q Sr-89, Sr-90 5x10-8 Composite(3)

McGuire Units 1 and 2 16.11.1-2 Revision 137

Liquid Effluents - Concentration 16.11.1 TABLE 16.11.1-1 (Page 2 of 3)

NOTES:

(1) The LLD is defined, for purposes of these commitments, as the smallest concentration of radioactive material in a sample that will yield a net count, above system background, that will be detected with 95% probability with only 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 / T) + 4 .6 5 Sb E-V-2.22 x10 6 - Y. exp (-2At)

Where:

LLD is the "a priori" lower limit of detection as defined above (as microCurie per unit mass or volume),

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

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

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

2.22 x 106 is the number of disintegrations per minute per microCurie, Y is the fractional radiochemical yield (when applicable),

X 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), and T is the background and sample counting time in minutes.

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

(2) 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.

McGuire Units 1 and 2 16.11.1-3 Revision 137

Liquid Effluents - Concentration 16.11.1 TABLE 16.11.1-1 (Page 3 of 3)

(3) To be representative of the quantities and concentrations of radioactive materials in liquid effluents, samples shall be collected continuously or intermittently 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.

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

(5) 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.

(6) 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-1 34, Cs-1 37, and Ce-1 41. The LLD for Ce-1 44 is 5x1 06 microCi/ml. 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 be identified and reported in the Annual Radioactive Effluent Release Report.

(7) The principal gas gamma emitters for which the LLD specification applies are Xe-1 33 and Xe-135. These are the reference nuclides in Regulatory Guide 1.21.

McGuire Units 1 and 2 16.11.1-4 Revision 137

Liquid Effluents - Concentration 16.11.1 FIGURE 16.11.1-1 SITE BOUNDARY/EXCLUSION AREA BOUNDARY McGuire Units 1 and 2 16.11.1-5 Revision 137

Liquid Effluents - Concentration 16.11.1 BASES This commitment is provided to ensure that the concentration of radioactive materials released in liquid waste effluents to UNRESTRICTED AREAS will be less than 10 times the effluent concentration levels specified in 10 CFR Part 20, Appendix B, Table 2, Column 2.

This limitation provides additional assurance that the levels of radioactive materials in bodies of water in UNRESTRICTED AREAS will result in exposures within: (1) the Section II.A design objectives of Appendix 1,10 CFR Part 50, to a MEMBER OF THE PUBLIC, and (2) the limits of 10 CFR Part 20.1301 to the population. The concentration limit for dissolved or entrained noble gases is based upon the assumption that Xe-1 35 is the controlling radioisotope and its EC in air (submersion) was converted to an equivalent concentration in water using the methods described in International Commission on Radiological Protection (ICRP) Publication 2. This commitment applies to the release of liquid effluents from all reactors at the site.

The basic requirements for the Selected Licensee Commitments concerning effluents from nuclear power reactors are stated in 10CFR50.36a. These requirements indicate that compliance with effluent Selected Licensee Commitments will keep average annual releases of radioactive material in effluents to small percentages of the limits specified in the old 1OCFR20.106 (new 1OCFR20.1301). These requirements further indicate that operational flexibility is allowed, compatible with considerations of health and safety, which may temporarily result in releases higher than such small percentages, but still within the limits specified in the old 10CFR20.106 which references Appendix B, Table II concentrations (MPCs). These referenced concentrations are specific values which relate to an annual dose of 500 mrem. It is further indicated in 10CFR50.36a that when using operational flexibility, best efforts shall be exerted to keep levels of radioactive materials in effluents as low as is reasonably achievable (ALARA) as set forth in 1OCFR50, Appendix I.

As stated in the Introduction to Appendix B of the new 1OCFR20, the effluent concentration (EC) limits given in Appendix B, Table 2, Column 2, are based on an annual dose of 50 mrem. Since a release concentration corresponding to a limiting dose rate of 500 mrem/year has been acceptable as a SLC limit for liquid effluents, which applies at all times as an assurance that the limits of 1OCFR50, Appendix I are not likely to be exceeded, it should not be necessary to reduce this limit by a factor of 10.

Operational history at Catawba/McGuire/Oconee has demonstrated that the use of the concentration values associated with the old 10CFR20.106 as SLC limits has resulted in calculated maximum individual doses to members of the public that are small percentages of the limits of 10CFR50, Appendix I. Therefore, the use of concentration values which correspond to an annual dose of 500 mrem should not have a negative impact on the ability to continue to operate within the limits of 1OCFR50 Appendix I and 40CFR1 90.

Having sufficient operational flexibility is especially important in establishing a basis for effluent monitor setpoint calculations. As discussed above, the concentrations stated in the new 10CFR20, Appendix B, Table 2, Column 2, relate to a dose of 50 mrem in a year.

When applied on an instantaneous basis, this corresponds to a dose rate of 50 mrem/year.

This low value is impractical upon which to base effluent monitor setpoint calculations for many liquid effluent release situations when monitor background, monitor sensitivity, and monitor performance must be taken into account.

BASES (continued)

McGuire Units 1 and 2 16.11.1-6 Revision 137

Liquid Effluents - Concentration 16.11.1 Therefore, to accommodate operational flexibility needed for effluent releases, the limits associated with SLC 16.11.1 are based on ten times the concentrations stated in the new 1OCFR20, Appendix B, Table 2, Column 2 to apply at all times. The multiplier of ten is proposed because the annual dose of 500 mrem, upon which the concentrations in the old 1OCFR20, Appendix B, Table II, Column 2 are based, is a factor of ten higher than the annual dose of 50 mrem, upon which the concentrations in the new 1OCFR20, Appendix B, Table 2, Column 2, are based. Compliance with the limits of the new 10CFR20.1301 will be demonstrated by operating within the limits of 1OCFR50, Appendix I and 40CFR1 90.

The required detection capabilities for radioactive materials in liquid waste samples are tabulated in terms of the lower limits of detection (LLDs). Detailed discussion of the LLD, and other detection limits can be found in HASL Procedures Manual, HASL-300 (revised annually), Currie, L. A., "Limits for Qualitative Detection and Quantitative Determination -

Application to Radiochemistry," Anal. Chem. 40, 586-93 (1968), and Hartwell, J. K.,

"Detection Limits for Radioanalytical Counting Techniques," Atlantic Richfield Hanford Company Report ARH-SA-215 (June 1975).

REFERENCES

1. McGuire Nuclear Station Offsite Dose Calculation Manual (ODCM)

2. International Commission on Radiological Protection (ICRP) Publication 2 McGuire Units 1 and 2 16.11.1-7 Revision 137

Radioactive Liquid Effluent Monitoring Instrumentation 16.11.2 16.11 RADIOLOGICAL EFFLUENT CONTROLS 16.11.2 Radioactive Liquid Effluent Monitoring Instrumentation COMMITMENT The radioactive liquid effluent monitoring instrumentation channels.

shown in Table 16.11.2-1 shall be FUNCTIONAL with their Alarm/Trip Setpoints set to ensure that the limits of SLC 16.11.1 are not exceeded.

AND The Alarm/Trip Setpoints of these channels shall be determined and adjusted in accordance with the methodology and parameters in the OFFSITE DOSE CALCULATION MANUAL (ODCM).

APPLICABILITY i ,s shown in Table 16.11.2-1.

REMEDIAL ACTIONS

NOTIr__

Separate Condition entry is allowed for each Funci CONDITION REQUIRED ACTION COMPLETION TIME A. One or more radioactive A.1 Suspend the release of Immediately liquid effluent monitoring radioactive liquid effluents channels Alarm/Trip monitored by the affected setpoint less channel.

conservative than required. OR A.2 Declare the channel non- Immediately functional.

OR A.3 Adjust setpoint to within Immediately limit.

B. One or more radioactive B.1 Enter the Remedial Action Immediately liquid effluent monitoring specified in Table 16.11.2-instrument channels 1 for the channel(s).

non-functional.

(continued)

McGuire Units 1 and 2 16.11.2-1 Revision 134

Radioactive Liquid Effluent Monitoring Instrumentation 16.11.2 REMEDIAL ACTIONS (continued~

CONDITION REQUIRED ACTION COMPLETION TIME C. One channel non- C.1.1 Analyze two independent Prior to initiating a functional. samples per TR 16.11.1.1. release AND C.1.2 Perform independent Prior to initiating a verification of the discharge release line valving.

AND C.1.3.1 Perform independent Prior to initiating a verification of manual release portion of the computer input for the release rate calculations performed by computer.

OR C.1.3.2 Perform independent Prior to initiating a verification of entire release release rate calculations for calculations performed manually.

AND 14 days C.1.4 Restore channel to FUNCTIONAL status.

OR C.2 Suspend the release of Immediately radioactive effluents via this pathway.

(continued)

McGuire Units 1 and 2 16.11.2-2 Revision 134

Radioactive Liquid Effluent Monitoring Instrumentation 16.11.2 REMEDIAL ACTIONS (continued CONDITION REQUIRED ACTION COMPLETION TIME D. One or more channels D.1 Obtain grab samples from Once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months non-functional. the effluent pathway. during releases.

AND D.2 Perform an analysis of To meet LLD grab samples for requirements per radioactivity. Table 16.11.1-1.

AND D.3 Restore the channel to 30 days FUNCTIONAL status.

E. One or more flow rate E.1 NOTE--

measurement channels Pump performance curves non-functional. generated in place may be used to estimate flow.

Estimate the flow rate of Once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months the release. during releases AND E.2 Restore the channel to 30 days FUNCTIONAL status.

F. RC minimum flow F.1 Verify that the number of Once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months interlock non-functional. pumps providing dilution is during releases greater than or equal to the number of pumps required.

AND F.2 Restore the channel to 30 days FUNCTIONAL status.

(continued)

McGuire Units 1 and 2 16.11.2-3 Revision 134

Radioactive Liquid Effluent Monitoring Instrumentation 16.11.2 REMEDIAL ACTIONS (continued' CONDITION REQUIRED ACTION COMPLETION TIME G. Required Action and G.1 Explain why the non- In the next associated Completion functionality was not scheduled Annual Time of Condition C, D, corrected within the Radioactive Effluent E or F not met. specified Completion Time Release Report in the Annual Radioactive Effluent Release Report.

McGuire Units 1 and 2 16.11.2-4 Revision 134

Radioactive Liquid Effluent Monitoring Instrumentation 16.11.2 TESTING REQUIREMENTS

NOTE------------------

Refer to Table 16.11.2-1 to determine which TRs apply for each Radioactive Liquid Effluent Monitoring channel.

TEST FREQUENCY TR 16.11.2.1 Perform CHANNEL CHECK. 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months TR 16.11.2.2 NOTE The CHANNEL CHECK shall consist of verifying indication of flow.

Perform CHANNEL CHECK. Every 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months during periods of release TR 16.11.2.3 Perform SOURCE CHECK. Prior to each release TR 16.11.2.4 Perform SOURCE CHECK. 31 days TR 16.11.2.5 NOTES-

1. For Instrument 1, the COT shall also demonstrate that automatic isolation of the pathway occurs if the instrument indicates measured levels above the Alarm/Trip Setpoint.

2. For Instruments 1 and 2, the COT shall also demonstrate that control room alarm annunciation occurs if the instrument indicates measured levels above the Alarm/Trip Setpoint; circuit failure and, a downscale failure.

Perform CHANNEL OPERATIONAL TEST. 92 days TR 16.11.2.6 Perform a CHANNEL CALIBRATION. 18 months (continued)

McGuire Units 1 and 2 16.11.2-5 Revision 134

Radioactive Liquid Effluent Monitoring Instrumentation 16.11.2 TESTING REQUIREMENTS (continued)

TEST FREQUENCY 4

TR 16.11.2.7 -------- NOTE.

The initial CHANNEL CALIBRATION shall be performed using standards certified by the National Institute of Standards and Technology (NIST) or using standards 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 range. For subsequent CHANNEL CALIBRATION, sources that have been related to the initial calibration shall be used.

Perform a CHANNEL CALIBRATION. 24 months McGuire Units 1 and 2 16.11.2-6 Revision 134

Radioactive Liquid Effluent Monitoring Instrumentation 16.11.2 TABLE 16.11.2-1 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION INSTRUMENT MINIMUM REMEDIAL APPLICABILITY TESTING CHANNELS ACTION REQUIREMENTS FUNCTIONAL

1. Radioactivity Monitors Providing Alarm And Automatic Termination of Release

a. Waste Liquid Effluent Line (EMF-49) 1 per station A, C, G During liquid TR 16.11.2.1 effluent releases TR 16.11.2.3 TR 16.11.2.5 TR 16.11.2.7

b. EMF-49 Minimum Flow Device 1 per station C, G During liquid TR 16.11.2.5 (2) effluent releases TR 16.11.2.7

c. Containment Ventilation Unit Condensate 1 A, D, G At all times TR 16.11.2.1 Line (EMF-44) TR 16.11.2.4 TR 16.11.2.5 TR 16.11.2.7

d. EMF-44 Minimum Flow Device 1 D. G At all times TR 16.11.2.5 (2) TR 16.11.2.7

2. Radioactivity Monitors Providing Alarm But Not Automatic Termination of Release

a. Conventional Waste Water Treatment 1 A. D, G At all times TR 16.11.2.1 Line or Turbine Building Sump to RC TR 16.11.2.4 (EMF- 31) TR 16.11.2.5 TR 16.11.2.7 1 D, G At all times TR 16.11.2.5

b. EMF-31 Minimum Flow Device (2) TR 16.11.2.7

3. Continuous Composite Samplers

a. Containment Ventilation Unit Condensate 1 D. G At all times TR 16.11.2.2 Line TR 16.11.2.5 TR 16.11.2.6

b. Conventional Waste Water Treatment Line 1 per station D. G At all times TR 16.11.2.2 TR 16.11.2.5 TR 16.11.2.6

c. Turbine Building Sump to RC 1 D, G At all times TR 16.11.2.2 TR 16.11.2.6 (Continued)

McGuire Units 1 and 2 16.11.2-7 Revision 134

Radioactive Liquid Effluent Monitoring Instrumentation 16.11.2

4. Flow Rate Measurement Devices

a. Waste Liquid Effluent Line I per station E. G During liquid TR 16.11.2.2 effluent releases TR 16.11.2.5 TR 16.11.2.6

b. Containment Ventilation Unit Condensate E, G At all times TR 16.11.2.2 Line TR 16.11.2.5 TR 16.11.2.6

c. Conventional Waste Water Treatment Une I per station E, G At all times TR 16.11.2.2 TR 16.11.2.5 TR 16.11.2.6

d. Turbine Building Sump to RC 1 E, G At all times TR 16.11.2.2 TR 16.11.2.6

5. RC Minimum Flow Interlock (1) 1 per station F, G At all times TR 16.11.2.5 NOTES:

1. Minimum flow dilution is assured by an interlock which terminates waste liquid release ifthe number of RC pumps running falls below the number of pumps required for dilution. The required number of RC pumps for dilution Is determined per station procedures.

2. Radioactivity Monitor (EMF) shall not be declared functional unless both the EMF and the associated EMF's Minimum Flow Device are rendered functional.

BASES 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 minimum flow devices for EMFs listed in Table 16.11.2-1 are required to provide assurance of representative sampling during actual or potential releases of liquid effluents. An interlock between the EMF's minimum flow device and its associated flow rate measurement device disables the remove alarm during non-release timeframes for the purpose of the control room black board annunciator criteria that disable expected alarms. An EMF flow rate measurement device measures total flow of the effluent while the EMF minimum flow device measures the sample flow rate through the EMF. The Alarm/Trip Setpoints of these instruments shall be calculated and adjusted in accordance with the methodology and parameters in the ODCM to ensure that the Alarm/Trip will occur prior to exceeding the limits stated in SLC 16.11.1. The FUNCTIONALITY 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. The Turbine Building Sump to RC Discharge Flow Measurement and Sampler Devices are for monitoring only and do not alarm or have any controls that require a COT.

REFERENCES

1. McGuire Nuclear Station Offsite Dose Calculation Manual (ODCM)

2. 10 CFR Part 50, Appendix A McGuire Units i and 2 16.11.2-8 Revision 134

Dose - Liquid Effluents 16.11.3 16.11 RADIOLOGICAL EFFLUENT CONTROLS 16.11.3 Dose - Liquid Effluents COMMITMENT The dose or dose commitment to a MEMBER OF THE PUBLIC from radioactive materials in liquid effluents released from each unit to UNRESTRICTED AREAS (see Figure16.11.1-1) shall be limited:

a. During any calendar quarter, to < 1.5 mrem to the total body and to < 5 mrem to any organ, and

b. During any calendar year, to < 3 mrem to the total body and to

_410 mrem to any organ.

APPLICABILITY At all times.

REMEDIAL ACTIONS NOTES-Enter applicable Conditions and Required Actions of SLC 16.11.12, "Total Dose," when the limits of this SLC are exceeded by twice the specified limit.

CONDITION REQUIRED ACTION COMPLETION TIME A. Calculated dose from NOTE release of radioactive The Special Report shall include materials in liquid the results of radiological analyses effluents exceeding of the drinking water source, and above limits, the radiological impact on finished drinking water supplies with regard to the requirements of 40 CFR 141, Safe Drinking Water Act, as applicable.

A.1 Prepare and submit a 30 days Special Report to the NRC which identifies the causes for exceeding the limits, corrective actions taken to reduce releases, and actions taken to ensure that subsequent releases are within limits.

McGuire Units I and 2 16.11.3-1 Revision 0

Dose - Liquid Effluents 16.11.3 TESTING REQUIREMENTS TEST FREQUENCY TR 16.11.3.1 Determine cumulative dose contributions from liquid 31 days effluents for current calendar quarter and current calendar year in accordance with the methodology and parameters in the ODCM.

BASES This commitment is provided to implement the requirements of Sections II.A, III.A and IV.A of Appendix 1,10 CFR Part 50. The commitment implements the guides set forth in Section IH.A of Appendix I. The REMEDIAL 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 to UNRESTRICTED AREAS will be kept "as low as is reasonably achievable." Also, for fresh water sites with drinking water supplies that 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 Part 141. These requirements are applicable only if the drinking water supply is taken from the river 3 miles downstream of the plant discharge.

The dose calculation methodology and parameters in the ODCM implement the requirements in Section IIL.A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data, such that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated. The equations specified in the ODCM for calculating the doses due to the actual release rates of radioactive materials in liquid effluents are consistent with the methodology provided in Regulatory Guide 1.109, "Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I," Revision 1, October 1977 and Regulatory Guide 1.113, "Estimating Aquatic Dispersion of Effluents from Accidental and Routine Reactor Releases for the Purpose of Implementing Appendix I," April 1977.

This commitment applies to the release of liquid effluents from each unit at the site. For units with shared Radwaste Treatment Systems, the liquid effluents from the shared system are to be proportioned among the units sharing that system in accordance with the guidance given in NUREG-01 33, Chapter 3.1.

McGuire Units i and 2 16.11.3-2 Revision 0

Dose - Liquid Effluents 16.

11.3 REFERENCES

1. McGuire Nuclear Station, Off site Dose Calculation Manual

2. 40 CFR Part 141, Safe Drinking Water Act

3. 10 CFR Part 50, Appendix I

4. Regulatory Guide 1.109, "Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I," Revision 1, October 1977.

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

McGuire Units 1 and 2 16.11.3-3 Revision 0

Liquid Radwaste Treatment System 16.11.4 16.11 RADIOLOGICAL EFFLUENT CONTROLS 16.11.4 Liquid Radwaste Treatment System COMMITMENT The Liquid Radwaste Treatment System shall be FUNCTIONAL and appropriate portions of the system shall be used to reduce releases of radioactivity when the projected doses due to the liquid effluent from each unit to UNRESTRICTED AREAS (see Figure16.11.1-1) would exceed 0.06 mrem to the total body or 0.2 mrem to any organ in a 31 day period.

APPLICABILITY At all times.

REMEDIAL ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Radioactive liquid waste A.1 Prepare and submit a 30 days being discharged without Special Report to the NRC treatment and in excess which identifies the reasons of above limits, liquid radwaste was discharged without AND treatment, identification of non-functional equipment Any portion of Liquid and reasons for non-Radwaste Treatment functionality, corrective System not in operation. -actions taken to restore the equipment to FUNCTIONAL status, and actions taken to prevent recurrence.

McGuire Units I and 2 16.11.4-1 Revision 134

Liquid Radwaste Treatment System 16.11.4 TESTING REQUIREMENTS

NOTE----- --------------------

The Liquid Radwaste Treatment System shall be demonstrated FUNCTIONAL by meeting SLC 16.11.1 and16.11.3.

TEST FREQUENCY TR 16.11.4.1 Project liquid release doses from each unit to 31 days UNRESTRICTED AREAS, in accordance with the methodology and parameters in the ODCM, when water systems are being released without being processed by its radwaste treatment system.

BASES 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 specification implements the requirements of 10 CFR Part 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50 and the design objective given in Section 11.D of Appendix I to 10 CFR Part 50. The specified limits governing the use of appropriate portions of the Liquid Radwaste Treatment System were specified as a suitable fraction of the dose design objectives set forth in Section II.A of Appendix I, 10 CFR Part 50, for liquid effluents.

This commitment applies to the release of liquid effluents from each reactor at the site. For units with shared Radwaste Treatment Systems, the liquid effluents from the shared system are to be proportioned among the units sharing that system in accordance with the guidance given in NUREG-0133, Chapter 3.1.

REFERENCES

1. McGuire Nuclear Station, Off site Dose Calculation Manual

2. 10 CFR Part 50

3. 10 CFR Part 50, Appendix I McGuire Units I and 2 16-11.4-2 Revision 134

Chemical Treatment Ponds 16.11.5 16.11 RADIOLOGICAL EFFLUENT CONTROLS 16.11.5 Chemical Treatment Ponds COMMITMENT The quantity of radioactive material contained in each chemical treatment pond shall be limited by the following expression (excluding tritium and dissolved or entrained noble gases):

264 . A, < 1.0 V j(Cx 10)

Where:

A I = pond inventory limit for single radionuclide "j",in Curies Ci = 10 CFR 20, Appendix B, Table 2, Column 2, concentration for single radionuclide "j*, microCuries/ml; V = design volume of liquid and slurry in the pond, in gallons; and 264 = conversion unit, microCuries/Curie per milliliter/gallon.

APPLICABILITY At all times.

REMEDIAL ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Quantity of radioactive A. 1 Suspend all additions of Immediately material in any of the radioactive material to the chemical treatment pond.

ponds exceeding above limit. AND A.2 Initiate corrective action to Immediately reduce the pond contents to within limits.

McGuire Units I and 2 16.11.5-1 Revision 0

Chemical Treatment Ponds 16.11.5 TESTING REQUIREMENTS TEST FREQUENCY TR 16.11.5.1 Verify quantity of radioactive material in each batch of Prior to each slurry (powdex resin) to be transferred to chemical transfer treatment ponds is within limits by analyzing a representative sample of the slurry. Each batch to be transferred to the chemical treatment ponds is limited by:

F- QJ < 6.0 x 10' PC"lgmn j (Cj x10) PCi/ m BASES The inventory limits of the chemical treatment ponds (CTP) are based on limiting the consequences of an uncontrolled release of the pond inventory. The expression in SLC 16.11.5 assumes the pond inventory is uniformly mixed, that the pond is located in an uncontrolled area as defined in 10 CFR Part 20, and that the concentration limit in Note 4 to Appendix B of 10 CFR Part 20 applies.

The batch limits of slurry to the chemical treatment ponds assure that radioactive material in the slurry transferred to the CTP are "as low as is reasonably achievable" in accordance with 10 CFR Part 50.36a. The expression in SLC 16.11.5 assures no batch of slurry will be transferred to the CTP unless the sum-of the ratios of the activity of the radionuclides to their respective concentration limitation is less than the ratio of the 10 CFR Part 50, Appendix I, Section II.A, total body dose level to the instantaneous whole body dose rate limitation, or that:

I X c j 3 mrem/ yr I (C,,xlO) <(-500 mrem/yr =0.006 Where:

ci = Radioactive slurry concentration for radionuclide "j"entering the UNRESTRICTED AREA chemical treatment ponds, in microCuries/milliliter; and C j= 10 CFR 20, Appendix B, Table 2, Column 2, concentration for single radionuclide "j", in microCuries/milliliter.

McGuire Units I and 2 16.11.5-2 Revision 0

Chemical Treatment Ponds 16.11.5 BASES (continued)

For the design of filterldemineralizers using powder resin, the slurry wash volume and the weight of resin used per batch is fixed by the cell surface area, and the slurry volume to resin weight ratio is constant at 100 ml/gram of wet, drained resin with a moisture content of approximately 55 to 60% (bulk density of about 58 pounds per cubic feet). Therefore, Xi =7 _ _ _ _ _ _< 0.006, and j (C, x 10) j (C i x 10) (102 mi/gm) (106 pCi/,uCi)

SQj < 6.0 x 10' PC"/gm j(C i x lO) POCi/mi Where:

Qj concentration of radioactive materials in wet, drained slurry (powdex resin) for radionuclide "j", excluding tritium, dissolved or entrained noble gases, and radionuclides with less than an 8-day half-life. The analysis shall include at least Ce-144, Cs-134, Cs-137, Co-58 and Co-60, in picoCuries/gram. Estimates of the Sr-89 and Sr-90 batch concentration shall be included based on the most recent monthly composite analysis (within 3 months); and C = 10 CFR 20, Appendix B, Table 2, Column 2, concentration for single radionuclide Uj", in microCunes/milliliter.

The batch limits provide assurance that activity input to the chemical treatment ponds will be minimized, and a means of identifying radioactive material in the inventory limitation of SLC 16.11.5.

The basic requirements for the Selected Licensee Commitments concerning effluents from nuclear power reactors are stated in 10CFR50.36a. These requirements indicate that compliance with effluent Selected Licensee Commitments will keep average annual releases of radioactive material in effluents to small percentages of the limits specified in the old 10CFR20.106 (new 10CFR20.1301). These requirements further indicate that operational flexibility is allowed, compatible with considerations of health and safety, which may temporarily result in releases higher than such small percentages, but still within the limits specified in the old 1OCFR20.106 which references Appendix B, Table II concentrations- (MPCs). These referenced concentrations are specific values which relate to an annual dose of 500 mrem. It is further indicated in IOCFR50.36a that when using operational flexibility, best efforts shall be exerted to keep levels of radioactive materials in effluents as low as is reasonably achievable (ALARA) as set forth in IOCFR50, Appendix I.

McGuire Units 1 and 2 16.11.5-3 Revision 0

Chemical Treatment Ponds 16.11.5 BASES (continued)

As stated in the Introduction to Appendix B of the new 10CFR20, the effluent concentration (EC) limits given in Appendix B, Table 2, Column 2, are based on an annual dose of 50 mrem. Since a release concentration corresponding to a limiting dose rate of 500 mremlyear has been acceptable as a SLC limit for liquid effluents, which applies at all times as an assurance that the limits of 10CFR50, Appendix I are not likely to be exceeded, it should not be necessary to reduce this limit by a factor of 10.

Operational history at CatawbalMcGuirelOconee has demonstrated that the use of the concentration values associated with the old 10CFR20.106 as SLC limits has resulted in calculated maximum individual doses to members of the public that are small percentages of the limits of 10CFR50, Appendix I. Therefore, the use of concentration values which correspond to an annual dose of 500 mrem should not have a negative impact on the ability to continue to operate within the limits of IOCFR50, Appendix I and 40CFR190.

Having sufficient operational flexibility is especially important in establishing a basis for effluent monitor setpoint calculations. As discussed above, the concentrations stated in the new 1OCFR20, Appendix B, Table 2, Column 2, relate to a dose of 50 mrem in a year.

When applied on an instantaneous basis, this corresponds to a dose rate of 50 mrem/year.

This low value is impractical upon which to base effluent monitor setpoint calculations for many liquid effluent release situations when monitor background, monitor sensitivity, and monitor performance must be taken into account.

Therefore, to accommodate operational flexibility needed for effluent releases, the limits associated with SLC 16.11.1 are based on ten times the concentrations stated in the new 10CFR20, Appendix B, Table 2, Column 2 to apply at all times. The multiplier of ten is proposed because the annual dose of 500 mrem, upon which the concentrations in the old 1OCFR20, Appendix B, Table II, Column 2 are based, is a factor of ten higher than the annual dose of 50 mrem, upon which the concentrations in the new IOCFR20, Appendix B, Table 2, Column 2, are based. Compliance with the limits of the new 1OCFR20.1301 will be demonstrated by operating within the limits of 10CFR50, Appendix I and 40CFR190.

REFERENCES

1. McGuire Nuclear Station, Off site Dose Calculation Manual

2. 10 CFR 20, Appendix B

3. 10 CFR 50, Appendix I, Section IL.A

4. 10 CFR 20

5. 10 CFR 50.36a McGuire Units I and 2 16.11.5-4 Revision 0

Dose Rate - Gaseous Effluents 16.11.6 16.11 RADIOLOGICAL EFFLUENT CONTROL 16.11.6 Dose Rate - Gaseous Effluents COMMITMENT The dose rate due to radioactive materials released in gaseous effluents from the site to areas at and beyond the SITE BOUNDARY (see Figurel6.11.1-1) shall be limited to the following:

a. For noble gases: < 500 mrem/yr to the whole body and < 3000 mrem/yr to the skin, and

b. For Iodine - 131 and 133, for tritium, and for all radioactive materials in particulate form with half-lives greater than 8 days:

< 1500 mrem/yr to any organ.

APPLICABILITY At all times.

REMEDIAL ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Dose rate not within A.1 Restore the release rate to Immediately limit. within limits.

TESTING REQUIREMENTS TEST FREQUENCY TR 16.11.6.1 Verify dose rates due to noble gases in gaseous effluents In accordance with are within limits in accordance with the methodology and the ODCM parameters in the ODCM.

TR 16.11.6.2 Verify dose rates due to radioactive materials, other than In accordance with noble gases, in gaseous effluents are within limits in Table 16.11.6-1 accordance with the methodology and parameters in the ODCM by obtaining representative samples and performing analyses in accordance with Table16.11.6-1.

McGuire Units 1 and 2 16.11.6-1 Revision 137

Dose Rate - Gaseous Effluents 16.11.6 TABLE 16.11.6-1 (Page 1 of 4)

RADIOACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM Gaseous Release Type Sampling Frequency Minimum Analysis Type of Activity Analysis Lower Limit of Frequency Detection (LLD)01 )

AiCi/ml)

1. Waste Gas Storage Tanks P P Each Tank Each Tank Principal Gas Gamma Emitters(') lxi10 Grab Sample

2. Containment Purge P P Each PURGE Each PURGE Principal Gas Gamma Emitters(6 ) lx10" Grab Sample M H-3 1x1 0-6 6

3. Unit Vent W( 2) W Principal Gas Gamma Emitters(' lx104 Grab Sample H-3 1x1 0-.

4.a. Radwaste Facility Vent W W Principal Gas Gamma Emitters(') lx10-1

b. Waste Handling Building Grab Sample

c. Equipment Staging Building H-3 lx1 0-

5. Unit Vents Continuous(5) W(8) 1-131 lx1 0-12 Charcoal Sample 1-133 1x1 0-10 Continuous( 5) W(8) Principal Gamma Emitters( 6 ) 1x10"11 Particulate (1-131, Others)

Sample Continuous(-) M Gross Alpha(7 ) lx10"11 Composite Particulate Sample Continuous(5) Q Sr-89, Sr-90 lx1011 Composite Particulate Sample McGuire Units 1 and 2 16.11.6-2 Revision 137

Dose Rate - Gaseous Effluents 16.11.6 TABLE 16.11.6-1 (Page 2 of 4)

RADIOACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM Gaseous Release Type Sampling Frequency Minimum Analysis Type of Activity Analysis Lower Limit of Frequency Detection (LLD)01 )

pCi/ml)

6. All Release Types as listed in 4 above. Continuous(5) W(8) 1-131 lx10-12 Charcoal Sample 1-133 1x10-10 Continuous(5) W(8) Principal Gamma Emitters(6) lx10-11 Particulate (1-131, Others)

Sample Continuous(') M Gross Alpha(7) lx1i0-11 Composite Particulate Sample Continuous(5) Q Sr-89, Sr-90 lx10"11 Composite Particulate Sample McGuire Units 1 and 2 16.11.6-3 Revision 137

Dose Rate - Gaseous Effluents 16.11.6 TABLE 16.11.6-1 (Page 3 of 4)

NOTES:

1. The LLD is defined, for purposes of these commitments, as the smallest concentration of radioactive material in a sample that will yield a net count, above system background, that will be detected with 95% probability with only 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 / T) + 4.65Sb E- V- 2.22 x 106

Y. exp (-AAt)

Where:

LLD = the "a priori" lower limit of detection as defined above (as microCurie per unit mass or volume);

Sb = the standard deviation of the background counting rate or of the counting rate of a blank sample as appropriate (as counts per minute);

E = the counting efficiency (as counts per disintegration);

V = the sample size (in units of mass or volume);

2.22 x106 = the number of disintegrations per minute per microCurie; Y = the fractional radiochemical yield (when applicable);

k = the radioactive decay constant for the particular radionuclide; At = the elapsed time between midpoint of sample collection and time of counting (for plant effluents, not environmental samples); and T = The background and sample counting time in minutes.

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 priori" (before the fact) limit representing the capability of a measurement system and not as an "a posteriori" (after the fact) limit for a particular measurement.

McGuire Units 1 and 2 16.11.6-4 Revision 137

Dose Rate - Gaseous Effluents 16.11.6 TABLE 16.11.6-1 (Page 4 of 4)

NOTES:

2. 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.

3. Not used.

4. Not used.

5. The ratio of the sample flow volume to the sampled stream flow volume shall be known for the time period covered by each dose or dose rate calculation made in accordance with SLCs 16.11.6, 16.11.8 and 16.11.9.

6. The principal gamma emitters for which the LLD specification applies include the following radionuclides: Kr-87, Kr-88, Xe-1 33, Xe-1 33m, Xel 35, and Xe-1 38 in noble gas releases and Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, 1-131, Cs-134, Cs-137, and Ce-141 in iodine and particulate releases. The LLD for Ce-144 is 5x10-9 microCi/ml.

This list does not mean that only these nuclides are to be considered. Other gamma peaks that are identifiable, together with those of the above nuclides, shall also be analyzed and reported in the Annual Radioactive Effluent Release Report.

7. The composite filter(s) will be analyzed for alpha activity by analyzing the filter media used during the collection period.

8. Samples shall be changed at least once per 7 days and analyses shall be completed to meet LLD after changing, or after removal from sampler. If the particulate and charcoal sample frequency is changed to a 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months frequency the corresponding LLDs may be increased by a factor of 10 (i.e., LLD for 1-131 from 1 x 10-12 to 1 x 10-11 microCi/ml).

McGuire Units 1 and 2 16.11.6-5 Revision 137

Dose Rate - Gaseous Effluents 16.11.6 BASES Specific release rate limits restrict, at all times, the corresponding gamma and beta dose rates above background to a MEMBER OF THE PUBLIC at or beyond the SITE BOUNDARY to less than or equal to 500 mrem/year to the whole body, and 3000 mrem/year to the skin from noble gases, and 1500 mrem/year to any organ from Iodine 131, Iodine 133, tritium, and all radionuclides in particulate form with half-lives greater than eight days. This commitment applies to the release of gaseous effluents from all reactors at the site. The Exclusion Area Boundary (Site Boundary) is set as the boundary for gaseous effluent release limits. The Exclusion Area Boundary (EAB) is formed by a 2500 ft radius centered on the Reactor Buildings' centerlines as shown on Figure 16.11.1-1.

The basic requirements for the Selected Licensee Commitments concerning effluents from nuclear power reactors are stated in 1OCFR50.36a. These requirements indicate that compliance with effluent Selected Licensee Commitments will keep average annual releases of radioactive material in effluents to small percentages of the limits specified in the old 10CFR20.106 (new 10CFR20.1301). These requirements further indicate that operational flexibility is allowed, compatible with considerations of health and safety, which may temporarily result in releases higher than such small percentages, but still within the limits specified in the old 1OCFR20.106 which references Appendix B, Table II concentrations (MPCs). These referenced concentrations are specific values which relate to an annual dose of 500 mrem. It is further indicated in 1OCFR50.36a that when using operational flexibility, best efforts shall be exerted to keep levels of radioactive materials in effluents as low as is reasonably achievable (ALARA) as set forth in 10CFR50, Appendix I.

As stated in the Introduction to Appendix B of the new 1OCFR20, the effluent concentration (EC) limits given in Appendix B, Table 2, Column 1, are based on an annual dose of 50 mrem for isotopes for which inhalation or ingestion is limiting or 100 mrem for isotopes for which submersion (noble gases) is limiting. Since release concentrations corresponding to limiting dose rates of less than or equal to 500 mrem/year to the whole body, 3000 mrem/year to the skin from noble gases, and 1500 mrem/year to any organ from Iodine 131, Iodine 133, tritium and for all radionuclides in particulate form with half-lives greater than eight days at the site boundary has been acceptable as a SLC limit for gaseous effluents to assure that the limits of 1OCFR50, Appendix I and 40CFR190 are not likely to be exceeded, it should not be necessary to restrict the operational flexibility by incorporating the EC value for isotopes based on ingestion/inhalation (50 mrem/year) or for isotopes with the EC based on submersion (100 mrem/year).

Having sufficient operational flexibility is especially important in establishing a basis for effluent monitor setpoint calculations. As discussed above, the concentrations stated in the new 10CFR20, Appendix B, Table 2, Column 1, relate to a dose of 50 or 100 mrem in a year. When applied on an instantaneous basis, this corresponds to a dose rate of either 50 or 100 mrem/year. These low values are impractical upon which to base effluent monitor setpoint calculations for many effluent release situations when monitor background, monitor sensitivity, and monitor performance must be taken into account. Therefore, to accommodate operational flexibility needed for effluent releases, the limits associated with SLC 16.11.6 will be maintained at the current dose rate limit for noble gases of 500 mrem/year to the whole body and 3000 mrem/year to the skin, for Iodine 131, Iodine 133, tritium and all radionuclides in particulate form with half-lives greater than eight days an instantaneous dose rate limit of 1500 mrem/year to any organ.

McGuire Units 1 and 2 16.11.6-6 Revision 137

Dose Rate - Gaseous Effluents 16.11.6 BASES (continued)

Compliance with the limits of the new 10CFR20.1301 will be demonstrated by operating within the limits of 10CFR50, Appendix I and 40CFR190. Operational history at Catawba/McGuire/Oconee has demonstrated that the use of the dose rate values listed above (i.e. 500 mrem/year, 3000 mrem/year and 1500 mrem/year) as SLC limits has resulted in calculated maximum individual doses to members of the public that are small percentages of the limits of 10CFR50, Appendix I and 40CFR190.

The required detection capabilities for radioactive materials in gaseous waste samples are tabulated in terms of the lower limits of detection (LLDs). Detailed discussion of the LLD, and other detection limits can be found in HASL Procedures Manual, HASL-300 (revised annually), Currie, L.A., "Limits for Qualitative Detection and Quantitative Determination -

Application to Radiochemistry," Anal. Chem. 40, 586 (1968), and Hartwell, J. K.

"Detection Limits for Radioanalytical Counting Techniques," Atlantic Richfield Hanford Company Report ARH-SA-215 (June 1975).

REFERENCES

1. McGuire Nuclear Station, Off site Dose Calculation Manual

2. 10 CFR Part 20, Appendix B

3. 10CFRPart20

4. 10 CFR Part 50 McGuire Units 1 and 2 16.11.6-7 Revision 137

Radioactive Gaseous Effluent Monitoring Instrumentation 16.11.7 16.11 RADIOLOGICAL EFFLUENT CONTROLS 16.11.7 Radioactive Gaseous Effluent Monitoring Instrumentation COMMITMENT The radioactive gaseous effluent monitoring instrumentation channels shown in Table 16.11.7-1 shall be FUNCTIONAL with Alarm/Trip Setpoints set to ensure that the limits of SLC 16.11.6 are not exceeded.

AND The Alarm/Trip setpoints shall be determined and adjusted in accordance with the methodology and parameters in the ODCM.

,l*,l[J r.--- - - - -

Brief periods of routine sampling (not to exceed 15 minutes) do not make the instrumentation non-functional.

APPLICABILITY As shown in Table 16.11.7-1.

REMEDIAL ACTIONS Separate Condition entry is allowed for each Function.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more radioactive A.1 Suspend the release of Immediately gaseous effluent radioactive gaseous monitoring channels effluents monitored by the Alarm/Trip setpoint less affected channel.

conservative than required. OR A.2 Declare the channel non- Immediately functional.

OR A.3 Adjust setpoint to within Immediately limit.

(continued)

McGuire Units 1 and 2 16.11.7-1 Revision 134

Radioactive Gaseous Effluent Monitoring Instrumentation 16.11.7 REMEDIAL ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B. One or more radioactive B.1 Enter the Remedial Action Immediately gaseous effluent specified in Table 16.11.7-1 monitoring instrument for the channel(s).

channels non-functional.

C. One channel non-functional.

C. 1.1 Analyze two independent samples of the tank contents.

Prior to initiating a release I AND C.1.2 Perform independent Prior to initiating a verification of the discharge release valve lineup.

AND C. 1.3.1 Perform independent Prior to initiating a verification of manual release portion of the computer input for the release rate calculations performed by computer.

OR C.1.3.2Perform independent Prior to initiating a verification of entire release release rate calculations for calculations performed manually.

AND C.1.4 Restore channel to 14 days FUNCTIONAL status.

OR C.2 Suspend the release of Immediately radioactive effluents via this pathway.

(continued)

McGuire Units I and 2 16.11.7-2 Revision 134

Radioactive Gaseous Effluent Monitoring Instrumentation 16.11.7 REMEDIAL ACTIONS (continued CONDITION REQUIRED ACTION COMPLETION TIME D. One or more flow rate D.1 Estimate the flow rate of Once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months measurement channels the release. during releases non-functional.

AND D.2 Restore the channel to 30 days FUNCTIONAL status.

E. One or more noble gas E.1 Obtain grab samples from Once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months activity monitor channels the effluent pathway. during releases non-functional.

AND E.2 Perform an analysis of grab To meet LLD samples for radioactivity, requirements per Table 16.11.6-1 AND E.3 Restore the channel to 30 days FUNCTIONAL status.

F. Noble gas activity F.1 Suspend PURGING or Immediately monitor providing VENTING of radioactive automatic termination of effluents via this pathway.

release non-functional.

G. One or more sampler G.1 Perform sampling with Continuously channels non-functional. auxiliary sampling equipment as required by Table 16.11.6-1.

AND G.2 Restore the channel to 30 days FUNCTIONAL status.

(continued)

McGuire Units 1 and 2 16.11.7-3 Revision 134

Radioactive Gaseous Effluent Monitoring Instrumentation 16.11.7 REMEDIAL ACTIONS (continued CONDITION REQUIRED ACTION COMPLETION TIME H. One or more Sampler H.1 Verify flow through the Once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Minimum Flow Device sampling apparatus. during releases Channels non-functional.

AND H.2 Restore the channel to 30 days FUNCTIONAL status.

I. Required Action and 1.1 Explain why the non- In the next associated Completion functionality was not scheduled Annual Time of Condition C, D, correctedwithin the Radioactive Effluent E, F, G, or H not met. specified Completion Time Release Report in the Annual Radioactive Effluent Release Report.

TESTING REQUIREMENTS Refer to Table 16.11.7-1 to determine which TRs apply for each Radioactive Gaseous Effluent Monitoring channel.

TEST FREQUENCY TR 16.11.7.1 Perform CHANNEL CHECK. Prior to each release TR 16.11.7.2 - NOTE- Pdor to each The SOURCE CHECK for these channels shall be the release qualitative assessment of channel response when the channel sensor is exposed to a source of increased radioactivity or a simulated source of radioactivity such as a light emitting diode.

Perform SOURCE CHECK.

TR 16.11.7.3 Perform CHANNEL CHECK 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months TR 16.11.7.4 Perform CHANNEL CHECK. 7 days (continued)

McGuire Units I and 2 16.11.7-4 Revision 134

Radioactive Gaseous Effluent Monitoring Instrumentation 16.11.7 TESTING REQUIREMENTS (continued)

TEST FREQUENCY TR 16.11.7.5 NOTE The SOURCE CHECK for these channels shall be the qualitative assessment of channel response when the channel sensor is exposed to a source of increased radioactivity or a simulated source of radioactivity such as a light emitting diode.

Perform SOURCE CHECK. 31 days TR 16.11.7.6 -NOTES-

1. For noble gas activity monitors providing automatic termination of release, the COT shall also demonstrate that automatic isolation of the pathway occurs if the instrument indicates measured levels above the Alarm/Trip Setpoint.

2. For all noble gas activity monitors, the COT shall also demonstrate that control room alarm annunciation occurs if the instrument indicates measured levels above the Alarm/Trp Setpoint; circuit failure and, a downscale failure.

Perform CHANNEL OPERATIONAL TEST. 92 days TR 16.11.7.7 -------- NOTE------- --------

For all noble gas activity monitors, the initial CHANNEL CALIBRATION shall be performed using standards certified by the National Institute of Standards and Technology (NIST) or using standards 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 range. For subsequent CHANNEL CALIBRATION, sources that have been related to the initial calibration shall be used.

18 months Perform a CHANNEL CALIBRATION.

McGuire Units 1 and 2 16.11.7-5 Revision 134

Radioactive Gaseous Effluent Monitoring Instrumentation 16.11.7 TABLE 16.11.7-1 (Page 1 of 3)

RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION INSTRUMENTS MINIMUM REMEDIAL APPLICABILITY TESTING CHANNELS ACTION REQUIREMENTS FUNCTIONAL

1. WASTE GAS HOLDUP SYSTEM

a. Noble Gas Activity Monitor - 1 per station A, C, I During gas effluent TR 16.11.7.1 Providing Alarm and Automatic releases. TR 16.11.7.2 Termination of Release (Low TR 16.11.7.6 Range- EMF-50 or 1 EMF-36, low- TR 16.11.7.7 range)

b. Effluent System Flow Rate 1 per station D, I At all times except TR 16.11.7.3 Measuring Device when isolation TR 16.11.7.6 valve is closed & TR 16.11.7.7 locked.

2. Condenser Evacuation System - Noble 1 A, E, I When air ejectors TR 16.11.7.3 Gas Activity Monitor (EMF-33) are operable. TR 16.11.7.5 TR 16.11.7.6 TR 16.11.7.7

3. Vent System

a. Noble Gas Activity Monitor (Low A, E, I At all times. TR 16.11.7.3 Range - EMF-36) TR 16.11.7.5 TR 16.11.7.6 TR 16.11.7.7

b. Iodine Sampler G, I At all times, except TR 16.11.7.4 during routine sampling.

c. Particulate Sampler (EMF-35) G. I At all times, except TR 16.11.7.4 during routine sampling.

d. Unit Vent Flow Rate Monitor 1 D, I At all times. TR 16.11.7.3 (Totalizer) TR 16.11.7.6 TR 16.11.7.7

e. Iodine Sampler Minimum Flow 1 H,I At all times, except TR 16.11.7.3 Device during routine TR 16.11.7.6 sampling. TR 16.11.7.7 L Particulate Sampler Minimum Flow I G, At all times, except TR 16.11.7.3 Device (1) during routine TR 16.11.7.6 sampling. TR 16.11.7.7

4. Containment Purge System - Noble 1 A, F, I Modes 1 through 6, TR 16.11.7.2 Gas Activity Monitor - Providing Alarm except when TR 16.11.7.3 and Automatic Termination of Release isolation valve is TR 16.11.7.6 (Low Range - EMF-39) closed & locked. TR 16.11.7.7 (continued)

McGuire Units 1 and 2 16.11.7-6 Revision 134

Radioactive Gaseous Effluent Monitoring Instrumentation 16.11.7 TABLE 16.11.7-1 (Page 2 of 3)

RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION INSTRUMENTS MINIMUM REMEDIAL APPLICABILITY TESTING CHANNELS ACTION REQUIREMENTS FUNCTIONAL _

5. Auxiliary Building Ventilation System - I A, E. I At all times. TR 16.11.7.3 Noble Gas Activity Monitor (EMF-41 or TR 16.11.7.5 EMF-36) TR 16.11.7.6 TR 16.11.7.7

6. Fuel Storage Area Ventilation System - 1 A, E, I At all times. TR 16.11.7.3 Noble Gas Activity Monitor (EMF-42 or TR 16.11.7.5 EMF-36) TR 16.11.7.6 TR 16.11.7.7

7. Contaminated Parts Warehouse Ventilation System

a. Noble Gas Activity Monitor 1 per station A, E, I During gaseous TR 16.11.7.3 (EMF-53) effluent releases. TR 16.11.7.5 TR 16.11.7.6 TR 16.11.7.7

b. Flow Rate Monitor 1 per station D, I During gaseous TR 16.11.7.3 effluent releases. TR 16.11.7.6 TR 16.11.7.7

c. EMF-53 Sampler Minimum Flow 1 per station H.1 During gaseous TR 16.11.7.3 Device (1) effluent releases. TR 16.11.7.6 TR 16.11.7.7

8. Radwaste Facility Ventilation System

a. Noble Gas Activity Monitor 1 per station A, E, I During gaseous TR 16.11.7.3 (EMF-52) effluent releases. TR 16.11.7.5 TR 16.11.7.6 TR 16.11.7.7

b. Flow Rate Monitor 1 per station D, I During gaseous TR 16.11.7.3 effluent releases. TR 16.11.7.6 TR 16.11.7.7

c. EMF-52 Sampler Minimum Flow 1 per station H, I During gaseous TR 16.11.7.3 Device (1) effluent releases. TR 16.11.7.6 TR 16.11.7.7 (continued)

McGuire Units 1 and 2 16.11.7-7 Revision 134

Radioactive Gaseous Effluent Monitoring Instrumentation 16.11.7 TABLE 16.11.7-1 (Page 3 of 3)

RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION INSTRUMENTS MINIMUM REMEDIAL APPLICABILITY TESTING CHANNELS ACTION REQUIREMENTS FUNCTIONAL

9. Equipment Staging Building Ventilation System

a. Noble Gas Activity Monitor I per station A, E, I During gaseous TR 16.11.7.3 (EMF-59) effluent releases. TR 16.11.7.5 TR 16.11.7.6 TR 16.11.7.7

b. Flow Rate Monitor 1 per station D, I During gaseous TR 16.11.7.3 effluent releases. TR 16.11.7.6 TR 16.11.7.7

c. EMF-59 Sampler Minimum Flow i per station H, I During gaseous TR 16.11.7.3 Device (1) effluent releases. TR 16.11.7.6 TR 16.11.7.7

10. Containment Air Release and Addition I A, E, I At all times except TR 16.11.7.3 System - Noble Gas Activity Monitor when isolation TR 16.11.7.5 (EMF-39L or EMF-36L) valve is closed & TR 16.11.7.6 locked. TR 16.11.7.7 NOTES:

1. Radioactivity monitor (EMF) shall not be declared FUNCTIONAL unless both the EMF and the associated EMF's Minimum Flow Device are rendered FUNCTIONAL. I McGuire Units 1 and 2 16.11.7-8 Revision 134

Radioactive Gaseous Effluent Monitoring Instrumentation 16.11.7 BASES 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 instrumentation consists of monitoring and sampling instrumentation. Monitors provide continuous display of process parameters with appropriate alarms and trip setpoints established. Samplers collect a portion of the desired process for subsequent laboratory analysis, and do not have alarm/trip capability. Samplers and the analysis program provide a method to assure that long term effluent release quantities do not exceed the requirements of SLC 16.11.6.

Monitors provide assurance that instantaneous effluent releases do not exceed the requirements of SLC 16.11.6. The minimum flow devices for EMFs listed in Table 16.11.7-1 are required to provide assurance of representative sampling during actual or potential releases of gaseous effluents. The flow rate monitor quantifies the total gaseous effluent (both non-radioactive and radioactive) released to the environment.

During routine sampling, instrumentation may be turned off for short periods of time (not to exceed 15 minutes) in order to meet analysis requirements of SLC 16.11.6. This is considered to be a normal function of the equipment. The Alarm/Trip Setpoints for these instruments shall be calculated and adjusted in accordance with the methodology and parameters in the ODCM to ensure that the Alarm/Trip will occur prior to exceeding the limits stated in SLC 16.11.6. The FUNCTIONALITY 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.

REFERENCES

1. McGuire Nuclear Station, Offsite Dose Calculation Manual

2. 10 CFR Part 50, Appendix A McGuire Units 1 and 2 16.11.7-9 Revision 134

Noble Gases 16.11.8 16.11 RADIOLOGICAL EFFLUENT CONTROLS 16.11.8 Noble Gases COMMITMENT Air dose due to noble gases released in gaseous effluents, from each unit, to areas at and beyond the SITE BOUNDARY (see Figurel 6.11.1-1) 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, and

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.

APPLICABILITY At all times.

REMEDIAL ACTIONS I

Enter applicable Conditions and Required Actions of SLC 16.11.12, "Total Dose," when the limits of this SLC are exceeded by twice the specified limit.

CONDITION REQUIRED ACTION COMPLETION TIME A. Calculated air dose from A.1 Prepare and submit a 30 days radioactive noble gases Special Report to the NRC in gaseous effluents which identifies the causes exceeding any of above for exceeding the limits, limits, corrective actions taken to reduce releases, and actions taken to ensure that subsequent releases are within limits.

McGuire Units 1 and 2 16.11.8-1 Revision 0

Noble Gases 16.11.8 TESTING REQUIREMENTS TEST FREQUENCY TR 16.11.8.1 Determine cumulative dose contributions from noble 31 days gases in gaseous effluents for current calendar quarter and current calendar year in accordance with the methodology and parameters in the ODCM.

BASES This commitment is provided to implement the requirements of Sections II.B, III.A and IV.A of Appendix I, 10 CFR Part 50. The Limiting Condition for Operation implements the guides set forth in Section II.B of Appendix I.

The REMEDIAL 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 gaseous effluents to UNRESTRICTED AREAS will be kept "as low as is reasonably achievable."

The TESTING REQUIREMENTS implement the requirements in Section IILA of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data such that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially under-estimated.

The dose calculation methodology and parameters 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 Regulatory Guide 1.109, "Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I," Revision 1, October 1977 and Regulatory Guide 1.111, "Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water Cooled Reactors," Revision 1, July 1977.

The ODCM equations provided for determining the air doses at and beyond the SITE BOUNDARY are based upon the historical average atmospheric conditions.

This commitment applies at all times to the release of gaseous effluents from each reactor at the site. For units with shared Radwaste Treatment Systems, the gaseous effluents from the shared system are to be proportioned among the units sharing that system in accordance with the guidance given in NUREG-0133, Chapter 3.1.

McGuire Units I and 2 16.11.8-2 Revision 0

Noble Gases 16.

11.8 REFERENCES

Manuai Station, Off site Dose Calculation

1. McGuire Nuclear 50, Appendix I

2. 10 CFR Part Revision 0 16.11.8-3 and 2 McGuire Units I

Dose - Iodine-131 and 133, Tritium and Radioactive Materials in Particulate Form 16.11.9 16.11 RADIOLOGICAL EFFLUENT CONTROLS 16.11.9 Dose - Iodine-1 31 and 133, Tritium and Radioactive Materials in Particulate Form COMMITMENT The dose to a MEMBER OF THE PUBLIC from Iodine-1 31 and 133, tritium, and all radioactive materials in particulate form with half-lives greater than 8 days in gaseous effluents released from each unit to areas at and beyond the SITE BOUNDARY (see Figure 16.11.1-1) shall be limited to the following:

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

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

APPLICABILITY At all times.

REMEDIAL ACTIONS

l #'%"lff'* *"*

JIi Enter applicable Conditions and Required Actions of SLC 16.11.12, "Total Dose," when the limits of this SLC are exceeded by twice the specified limit.

CONDITION REQUIRED ACTION COMPLETION TIME A. Calculated dose from A.1 Prepare and submit a 30 days the release of Iodine 131 Special Report to the NRC and 133, tritium, and which identifies the causes radioactive materials in for exceeding the limits, particulate form with corrective actions taken to half-lives greater than 8 reduce releases, and days in gaseous actions taken to ensure effluents exceeding any that subsequent releases of the above limits, are within limits.

McGuire Units I and 2 16.11.9-1 Revision 0

Dose - Iodine-131 and 133, Tritium and Radioactive Materials in Particulate Form 16.11.9 TESTING REQUIREMENTS TEST FREQUENCY TR 16.11.9.1 Determine cumulative dose contributions for Iodine 131 31 days and 133, tritium, and radioactive material in particulate form with half lives greater than 8 days in gaseous effluents for current calendar quarter and current calendar year in accordance with the methodology and parameters in the ODCM.

BASES This commitment is provided to implement the requirements-of Sections- II.C, llI.A and IV.A of Appendix 1,10 CFR Part 50. The Limiting Conditions for Operation are the guides set forth in Section I .C of Appendix I.

The REMEDIAL ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive materials in gaseous effluents to UNRESTRICTED AREAS will be kept "as low as is reasonably achievable.

The ODCM calculational methods specified in the TESTING REQUIREMENTS implement the requirements in Section IlI.A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data, such that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated.

The ODCM calculational methodology and parameters for calculating the doses due to the actual release rates of the subject materials are consistent with the methodology provided in Regulatory Guide 1.109, "Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I," Revision 1, October 1977 and Regulatory Guide 1.111, "Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water Cooled Reactors, Revision 1, July 1977. These equations also provide for determining the actual doses based upon the historical average atmospheric conditions.

The release rate specifications for Iodine-131 and 133, tritium, and radionuclides in particulate form with half-lives greater than 8 days are dependent upon the existing radionuclide pathways to man, in the areas at and beyond the SITE BOUNDARY. The pathways that 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.

McGuire Units 1 and 2 16.11.9-2 Revision 0

Dose - Iodine-131 and 133, Tritium and Radioactive Materials in Particulate Form 16.11.9 BASES (continued)

This commitment applies at all times to the release of gaseous effluents from each reactor at the site. For units with shared Radwaste Treatment Systems, the gaseous effluents from the shared system are to be proportioned among the units sharing that system in accordance with the guidance given in NUREG 0133, Chapter 3.1.

REFERENCES

1. McGuire Nuclear Station, Off site Dose Calculation Manual

2. 10 CFR Part 50, Appendix I McGuire Units 1 and 2 16-11.9-3 Revision 0

Gaseous Radwaste Treatment System 16.11.10 16.11 RADIOLOGICAL EFFLUENT CONTROLS 16.11.10 Gaseous Radwaste Treatment System COMMITMENT The VENTILATION EXHAUST TREATMENT and WASTE GAS HOLDUP SYSTEMS shall be FUNCTIONAL and appropriate portions of these systems shall be used to reduce releases of radioactivity when the projected doses in 31 days due to gaseous effluent releases, from each unit, to areas at and beyond the SITE BOUNDARY (see Figure 16.11.1-1) would exceed:

a. 0.2 mrad to air from gamma radiation, or

b. 0.4 mrad to air from beta radiation, or

c. 0.3 mrem to any organ of a MEMBER OF THE PUBLIC.

APPLICABILITY At all times.

REMEDIAL ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Radioactive gases being A.1 Prepare and submit a 30 days discharged without Special Report to the NRC treatment and in excess which identifies non-of above limits, functional equipment and reasons for non-functionality, actions taken to restore the equipment to FUNCTIONAL status, and actions taken to prevent recurrence.

McGuire Units 1 and 2 16.11.10-1 Revision 134

Gaseous Radwaste Treatment System 16.11.10 TESTING REQUIREMENTS

.-.-.-------.----.-.-------.-- T-------

The installed Gaseous Radwaste Treatment System shall be demonstrated FUNCTIONAL by meeting SLC 16.11.6, 16.11.8 and16.11.9.

TEST FREQUENCY TR 16.11.10.1 Project gaseous release doses from each unit to areas 31 days at and beyond the SITE BOUNDARY, in accordance with the methodology and parameters in the ODCM, when gaseous systems are being released without being processed by its radwaste treatment system.

BASES The FUNCTIONALITY of the WASTE GAS HOLDUP 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 commitment implements the requirements of 19 CFR Part 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50 and the design objectives given in Section II.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 Section lI.B and II.C of Appendix 1, 10 CFR Part 50, for gaseous effluents.

This commitment applies at all times to the release of radioactive materials in gaseous effluents from each unit at the site. For units with shared Radwaste Treatment Systems, the gaseous effluents from the shared system are to be proportioned among the units sharing that system in accordance with NUREG-0133, Chapter 3.1.

REFERENCES

1. McGuire Nuclear Station, Off site Dose Calculation Manual

2. 10 CFR Part 50, Appendix I

3. 10CFRPart50 McGuire Units 1 and 2 16.11.10-2 Revision 134

Solid Radioactive Waste 16.11.11 16.11 RADIOLOGICAL EFFLUENT CONTROLS 16.11.11 Solid Radioactive Waste COMMITMENT Radioactive wastes shall be processed and packaged to ensure compliance with the applicable requirements of 10 CFR Part 20, 1OCFR Part 61, 10 CFR Part 71, and State regulations governing the transportation and disposal of radioactive wastes.

The Solid Radwaste System or an approved alternative process shall be used in accordance with a PROCESS CONTROL PROGRAM (PCP) for the solidification of liquid or wet radioactive wastes or the dewatering of wet radioactive wastes to be shipped for direct disposal at a 10CFR61 licensed disposal site. Wastes shipped for off site processing in accordance with the processor's specifications and transportation requirements are not required to be solidified or dewatered to meet disposal requirements.

The PCP describes administrative and operational controls used for the solidification of liquid or wet solid radioactive wastes in order to meet applicable 10CFR61 waste form requirements.

" The PCP describes the administrative and operational controls used for the dewatering of wet radioactive wastes to meet 10CFR61 free standing water requirements.

The process parameters used in establishing the PCP shall be based on demonstrated processing of actual or simulated liquid or wet solid wastes and must adequately verify that the final product of solidification or dewatering meets all applicable Federal, State and disposal site requirements.

APPLICABILITY At all times.

McGuire Units 1 and 2 16.11.11-1 Revision 134

Solid Radioactive Waste 16.11.11 REMEDIAL ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Applicable regulatory A.1 Suspend shipments of Immediately requirements for defectively packaged solid solidified or dewatered radioactive wastes from the wastes are not site.

satisified.

AND A.2 Initiate action to correct the Prior to next PROCESS CONTROL shipment for disposal PROGRAM, procedures, or of solidified or solid waste equipment as dewatered wastes.

necessary to prevent recurrence.

B. A solidification test as B.1 Suspend solidification of the Immediately described in the PCP batch under test and follow fails to verify PCP guidance for test Solidification. failures.

B.2 Once a subsequent test Prior to next verifies Solidification, solidification for solidification of the batch shipment of waste may then be resumed as for disposal at a directed by the PCP. The 10CFR61 disposal PCP shall be modified as site.

required to assure Solidification of subsequent batches of waste (continued)

McGuire Units 1 and 2 16.11.11-2 Revision 134

Solid Radioactive Waste 16.11.11 REMEDIAL ACTIONS (continued)

C. With solidification or C.1 Reprocess the waste in Prior to shipment for dewatering for accordance with PCP disposal of the disposal not requirements. inadequately processed performed in waste that requires accordance with the ,OR solidification of dewatering PROCESS CONTROL C.2 Follow PCP or procedure PROGRAM. quidance for alternative free standing liquid verification to ensure the waste in each container meets disposal requirements and take appropriate administrative action to prevent recurrence.

D. With the solid waste D.1 Restore the equipment In a time frame that equipment incapable to FUNCTIONAL status supports the I of meeting SLC or provide for alternative COMMITMENT section of 16.11.11 ornot in capability to process SLC 16.11.11 service wastes as necessary to satisfy all applicable disposal requirements TESTING REQUIREMENTS TEST FREQUENCY TR 16.11.11.1 The Process Control Program shall be used to verify the Every tenth batch Solidification of at least one representative test of each type of specimens from at least every tenth batch of each type radioactive waste of radioactive waste to be solidified for disposal at a to be solidified.

1 OCFR61 disposal site per the COMMITMENT of this SLC.

McGuire Units I and 2 16.11.11-3 Revision 134

Solid Radioactive Waste 16.11.11 BASES:

This commitment implements the requirements of 10 CFR Part 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50 and requirements to use a Process Control Program to meet applicable IOCFR61 waste form criteria for solidified and dewatered radioactive wastes.

REFERENCES:

1. 10CFR Part 50, "Domistic Licensing of Production and Utilization Facilities"

2. 10 CFR Part 50, Appendix A

3. 10CFR20, "Standards for Protection Against Radiation"

4. 10CFR61, "Licensing Requirements for Land Disposal of Radioactive Waste

5. 10CFR71, "Packaging and Transportation of Radioactive Materials"

6. DPCo Process Control Program Manual

7. NRC Generic Letter 84-12, "Compliance With 10 CFR Part 61 And Implementation Of the Radiological Effulent Technical Specifications (Rets) and Attendant Process Control Program (PCP)"

8. NRC Generic Letter 89-01, "Implementation of Programmatic Controls for Radiological Effulent 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" McGuire Units I and 2 16.11.11-4 Revision 134

Total Dose 16.11.12 16.11 RADIOLOGICAL EFFLUENT CONTROLS 16.11.12 Total Dose COMMITMENT The annual (calendar year) dose or dose commitment to any MEMBER OF THE PUBLIC, due to releases of radioactivity and radiation, from uranium fuel cycle sources shall be limited to < 25 mrem to the whole body or any organ, except the thyroid, which shall be limited to < 75 mrem.

APPLICABILITY At all times.

REMEDIAL ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Calculated doses from A.1 Verify, by calculation, the Immediately releases exceeding cumulative dose from direct twice the specified limits radiation contributions, the of SLC 16.11.3,16.11.8 ISFSI, outside storage or 16.11.9. tanks, and radioactivity releases are within the total dose limit.

AND A.2 NOTE Only required to be performed if the total dose limit is exceeded.

Prepare and submit a 30 days Special Report to the NRC which identifies corrective actions to be taken to reduce subsequent releases to prevent recurrence and schedule for achieving conformance with specified limits.

McGuire Units 1 and 2 16.11.12-1 Revision 67

Total Dose 16.11.12 TESTING REQUIREMENTS

- ---. --------

- -.----------- - TE-- .. .

Cumulative dose contributions from liquid and gaseous effluents shall be determined in accordance with SLC 16.11.3, 16.11.8 andl6.11.9, and in accordance with the methodology and parameters specified in the ODCM.

TEST FREQUENCY TR 16.11.12.1 Determine cumulative dose contributions from direct When calculated radiation from the units, the ISFSI, and from radwaste doses from storage tanks in accordance with the methodology and effluent releases parameters specified in the ODCM. exceeds twice the limits of SLCs 16.11.3, 16.11.8 or 16.11.9 BASES This commitment is provided to meet the dose limitations of 40 CFR Part 190 that have been incorporated into 10 CFR Part 20 by 46 FR 18525. The specification requires the preparation and submittal of a Special Report whenever the calculated doses from plant generated radioactive effluents and direct radiation exceed 25 mrem to the whole body or any organ, except the thyroid, which shall be limited to less than or equal to 75 mrem. For sites containing up to four reactors, it is highly unlikely that the resultant dose to a MEMBER OF THE PUBLIC will exceed the dose limits of 40 CFR Part 190 if the individual reactors remain within twice the dose design objectives of 10 CFR Part 50, Appendix I, and if direct radiation doses from the units and outside storage tanks are kept small.

This Special Report, as defined in 10 CFR Part 20.2203(a)(4), shall include an analysis that estimates the radiation exposure (dose) to a MEMBER OF THE PUBLIC from uranium fuel cycle sources, including all effluent pathways and direct radiation, for the calendar year that includes the release(s) covered by this report. It shall also describe levels of radiation and concentrations of radioactive material involved, and the cause of the exposure levels or concentrations. The Special Report will describe a course of action that should result in the limitation of the annual dose to a MEMBER OF THE PUBLIC to within the 40 CFR Part 190 limits. For the purposes of the 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 8 km must be considered.

If the dose to any MEMBER OF THE PUBLIC is estimated to exceed the requirements of 40 CFR Part 190, the Special Report with a request for a variance (provided the release conditions resulting in violation of 40 CFR Part 190 have not already been corrected), in McGuire Units 1 and 2 16.11.12-2 Revision 67

Total Dose 16.11.12 BASES (continued) accordance with the provisions of 40 CFR Part 190.11 and 10 CFR Part 20.2203(a)(4), is considered to be a timely request and fulfills the requirements of 40 CFR Part 190 and a variance is granted until NRC staff action is completed. The variance only relates to the limits of 40 CFR Part 190, and does not apply in any way to the other requirements for dose limitation of 10 CFR Part 20, as addressed in SLCs 16.11.1 and 16.11.6.

An individual is not considered a MEMBER OF THE PUBLIC during any period in which he/she is engaged in carrying out any operation that is part of the nuclear fuel cycle.

REFERENCES

1. McGuire Nuclear Station, Offsite Dose Calculation Manual

2. 10 CFR Part 20

3. 40 CFR Part 190

4. 10 CFR Part 50, Appendix I McGuire Units 1 and 2 16.11.12-3 Revision 67

Radiological Environmental Monitoring Program 16.11.13 16.11 RADIOLOGICAL EFFLUENT MONITORING 16.11.13 Radiological Environmental Monitoring Program COMMITMENT The Radiological Environmental Monitoring Program shall be conducted as specified in Table 16.11.13-1.

APPLICABILITY At all times.

REMEDIAL ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Radiological A.1 Identify the reasons for not Within the next Environmental conducting the program as scheduled Annual Monitoring Program not required and the plans for Radiological being conducted as preventing a recurrence in Environmental specified in Table the Annual Radiological Operating Report 16.11.13-1. Environmental Operating Report.

B. Radioactivity level of B.1 Prepare and submit a 30 days environmental sampling Special Report that defines medium at a specified the corrective actions to be location in excess of taken to reduce radioactive reporting limits of Table effluents so that the 16.11.13-2. potential annual dose to a MEMBER OF THE PUBLIC is less than the calendar year limits of SLC 16.11.3, 16.11.8, and 16.11.9.

(continued)

McGuire Units 1 and 2 16.11.13-1 Revision 137

Radiological Environmental Monitoring Program 16.11.13 REMEDIAL ACTIONS (continued' CONDITION REQUIRED ACTION COMPLETION TIME C. Milk or fresh leafy C.1 ---------- NOTE-------

vegetable samples Specific locations from unavailable from one or which samples were more required sample unavailable may be deleted locations, from the program.

Revise the Radiological 30 days Environmental Monitoring Program to identify locations for obtaining replacement samples.

AND C.2 Identify the cause of the Within the next unavailability of samples scheduled Annual and identify new location(s) Radioactive Effluent for obtaining replacement Release Report samples in the next Annual Radioactive Effluent Release Report and also include in the report a revised figure(s) and table for the ODCM reflecting the new location(s).

TESTING REQUIREMENTS TEST FREQUENCY TR 16.11.13.1 ------------------ NOTES-----------------

The maximum values for the lower limits of detection shall be as specified in Table16.11.13-3.

The radiological environmental monitoring samples shall In accordance with be collected from the locations given in the table and Table 16.11.13-1 figure in the ODCM and shall be analyzed pursuant to the requirements of Tables16.11.13-1.

McGuire Units 1 and 2 16.11.13-2 Revision 137

Radiological Environmental Monitoring Program 16.11.13 TABLE 16.11.13-1 (Page 1 of 6)

RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM EXPOSURE NUMBER OF REPRESENTATIVE SAMPLES SAMPLING AND TYPE AND FREQUENCY PATHWAY AND/OR AND SAMPLE LOCATIONS(1 ) COLLECTION OF ANALYSIS SAMPLE FREQUENCY

1. Direct Radiation(2) Forty routine monitoring stations either with Quarterly Gamma dose quarterly.

two or more dosimeters or with one instrument for measuring and recording dose rate continuously, placed as follows:

An inner ring of stations, one in each meteorological sector in the general area of the SITE BOUNDARY; An outer ring of stations, one in each meteorological sector in the 6- to 8-km range from the site; and The balance of the stations placed in special interest areas such as population centers, nearby residences, schools, and in one or two areas to serve as control stations.

(continued)

McGuire Units 1 and 2 16.11.13-3 Revision 137

Radiological Environmental Monitoring Program 16.11.13 TABLE 16.11.13-1 (Page 2 of 6)

RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM EXPOSURE NUMBER OF REPRESENTATIVE SAMPLES SAMPLING AND TYPE AND FREQUENCY

1) COLLECTION OF ANALYSIS PATHWAY AND/OR AND SAMPLE LOCATIONSO SAMPLE FREQUENCY

2. Airborne Samples from five locations: Continuous sampler Radioiodine Canister:

Radioiodine and operation with sample 1-131 analysis weekly.

Particulates Three samples from close to the three SITE collection weekly, or BOUNDARY locations, in different sectors, of more frequently if Particulate Sampler:

the highest calculated annual average required by dust Gross beta radioactivity ground level D/Q. loading, analysis following filter change('); Gamma isotopic One sample from the vicinity of a community analysis(5) of composite (by having the highest calculated annual average location quarterly).

ground level D/Q.

One sample from a control location, as for example 15-30 km distant and in the least prevalent wind direction(3 ).

3. Waterborne One sample upstream. Composite sample Gamma isotope analysis( 5 )

a. Surface(6 ) One sample downstream. over 1-month period(7 ). monthly. Composite for tritium analysis quarterly.

b. Ground Samples from one or two sources only if Quarterly Gamma isotopic( 5 ) and likely to be affected(8) tritium analysis quarterly.

(continued)

McGuire Units 1 and 2 16.11.13-4 Revision 137

Radiological Environmental Monitoring Program 16.11.13 TABLE 16.11.13-1 (Page 3 of 6)

RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM EXPOSURE NUMBER OF REPRESENTATIVE SAMPLES SAMPLING AND TYPE AND FREQUENCY PATHWAY AND/OR AND SAMPLE LOCATIONSO 1) COLLECTION OF ANALYSIS SAMPLE FREQUENCY

c. Drinking One sample of each of one to three of the Composite sample 1-131 analysis on each nearest water supplies that could be affected over 2-week period(7) composite when the dose by its discharge. when 1-131 analysis is calculated for the performed; monthly consumption of the water is One sample from a control location, composite otherwise, greater than 1 mrem per yearO9 ). Composite for gross beta and gamma isotopic analyses(5 ) monthly.

Composite for tritium analysis quarterly.

d. Sediment from One sample from downstream area with Semiannually Gamma isotopic analysis(5 )

the shoreline existing or potential recreational value, semiannually.

4. Ingestion Samples from milking animals in three Semimonthly when Gamma isotopic(') and 1-131

a. Milk locations within 5-km distance having the animals are on analysis semimonthly when highest dose potential. If there are none, pasture; monthly at animals are on pasture; then one sample from milking animals in other times. monthly at other times.

each of three areas between 5 to 8 km distant where doses are calculated to be greater than 1 mrem per year(9 ).

One sample from milking animals at a control location 15 to 30 km distant and in the least prevalent wind direction.

(continued)

McGuire Units 1 and 2 16.11.13-5 Revision 137

Radiological Environmental Monitoring Program 16.11.13 TABLE 16.11.13-1 (Page 4 of 6)

RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM EXPOSURE NUMBER OF REPRESENTATIVE SAMPLES SAMPLING AND TYPE AND FREQUENCY PATHWAY AND/OR AND SAMPLE LOCATIONS(1 ) COLLECTION OF ANALYSIS SAMPLE FREQUENCY

b. Fish and One sample each commercially and Sample in season, or Gamma isotopic analysis(5)

Invertebrates recreationally important species in vicinity of semiannually if they on edible portions plant discharge area. are not seasonal One sample of same species in areas not influenced by plant discharge.

c. Food Products One sample of each principal class of food At time of harvest(10 ) Gamma isotopic analyses(5) products from any area that is irrigated by on edible portion.

water in which liquid plant wastes have been discharged.

Samples of three different kinds of broad leaf Monthly, when Gamma isotopic( 5 ) and 1-131 vegetation grown nearest each of two available, analysis.

different offsite locations of highest predicted annual average ground level D/Q if milk sampling is not performed.

One sample of each of the similar broad leaf Monthly, when Gamma isotopic(5) and 1-131 vegetation grown 15 to 30 km distant in the available, analysis.

least prevalent wind direction if milk sampling is not performed.

McGuire Units 1 and 2 16.11.13-6 Revision 137

Radiological Environmental Monitoring Program 16.11.13 TABLE 16.11.13-1 (Page 5 of 6)

RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM NOTES:

1. Specific parameters of distance and direction sector from the centerline of one reactor, and additional description where pertinent, shall be provided for each and every sample location in Table 16.11.13-1 in a table and figure(s) in the ODCM. Refer to NUREG-0133, "Preparation of Radiological Effluent Technical Specifications for Nuclear Power Plants," October 1978, and to Radiological Assessment Branch Technical Position, Revision 1, November 1979. Deviations are permitted from the required sampling schedule if specimens are unobtainable due to hazardous conditions, seasonal unavailability, malfunction of automatic sampling 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. All deviations from the sampling schedule shall be documented in the Annual Radiological Environmental Operating Report. It is recognized that, at times, it may not be possible or practical to continue to obtain samples of the media of choice at the most desired location or time. In these instances suitable alternative media and locations may be chosen for the particular pathway in question and appropriate substitutions made within 30 days in the Radiological Environmental Monitoring Program. In lieu of an Licensee Event Report, identify the cause of the unavailability of samples for that pathway and identify the new locations(s) for obtaining replacement samples in the next Annual Radioactive Effluent Release Report and also include in the report a revised figure(s) and table for the ODCM reflecting the new location(s).

2. One or more instruments, such as a pressurized ion chamber, for measuring and recording dose rate continuously may be used in place of, or in addition to, integrating dosimeters. For the purposes of this table, a thermoluminescent dosimeter (TLD) is considered to be one phosphor; two or more phosphors in a packet are considered as two or more dosimeters. Film badges shall not be used as dosimeters for measuring direct radiation. The forty stations is not an absolute number. The number of direct radiation monitoring stations may be reduced according to geographical limitations; e.g.,

at an ocean site, some sections will be over water so that the number of dosimeters may be reduced accordingly. The frequency of analysis or readout for TLD systems will depend upon the characteristics of the specific system used and should be selected to obtain optimum dose information with minimal fading.

3. The purpose of this sample is to obtain background information. If it is not practical to establish control locations in accordance with the distance and wind direction criteria, other sites that provide valid background data may be substituted.

4. Airborne particulate sample filters shall be analyzed for gross beta radioactivity 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or more after sampling to allow for radon and thoron daughter decay. If gross beta activity in air particulate samples is greater than ten times the yearly mean of control samples, gamma isotopic analysis shall be performed on the individual samples.

McGuire Units 1 and 2 16.11.13-7 Revision 137

Radiological Environmental Monitoring Program 16.11.13 TABLE 16.11.13-1 (Page 6 of 6)

RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM NOTES (continued):

5. Gamma isotopic analysis means the identification and quantification of gamma-emitting radionuclides that may be attributable to the effluents from the facility.

6. The "upstream sample" shall be taken at a distance beyond significant influence of the discharge. The "downstream" sample shall be taken in an area beyond but near the mixing zone. "Upstream" samples in an estuary must be taken far enough upstream to be beyond the plant influence. Salt water shall be sampled only when the receiving water is utilized for recreational activities.

7. A composite sample is one in which the quantity (aliquot) of liquid sampled is proportional to the quantity of flowing liquid and in which the method of sampling employed results in a specimen that is representative of the liquid flow. In this program composite sample aliquots shall be collected at time intervals that are very short (e.g.,

hourly) relative to the compositing period (e.g., monthly) in order to assure obtaining a representative sample.

8. Groundwater samples shall be taken when this source is tapped for drinking or irrigation purposes in areas where the hydraulic gradient or recharge properties are suitable for contamination.

9. The dose shall be calculated for the maximum organ and age group, using the methodology and parameters in the ODCM.

10. If harvest occurs more than once a year, sampling shall be performed during each discrete harvest. If harvest occurs continuously, sampling shall be monthly. Attention shall be paid to including samples of tuborous and root food products.

McGuire Units 1 and 2 16.11.13-8 Revision 137

Radiological Environmental Monitoring Program 16.11.13 TABLE 16.11.13-2 (Page 1 of 1)

REPORTING LEVELS FOR RADIOACTIVITY CONCENTRATIONS IN ENVIRONMENTAL SAMPLES REPORTING LEVELS ANALYSIS WATER WATER AIRBOURNE FISH MILK MILK BROAD LEAF (pCi/I) PARTICULATE OR (pCi/kg, wet) (pCi/I) VEGETATION GASES (pCi/m 3) (pCi/kg, wet)

H-3 20,000(1) 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 20,000 N/A N/A Zr-Nb-95 400 N/A N/A N/A N/A 1-131 2 0.9 N/A 3 100 Cs-134 30 10 1,000 60 1,000 Cs-137 50 20 2,000 70 2,000 Ba-La-1 40 200 N/A N/A 300 N/A NOTES:

1. For drinking water samples. This is 40 CFR Part 141 value. If no drinking water pathway exists, a value of 30,000 pCi/I may be used.

McGuire Units 1 and 2 16.11.13-9 Revision 137

Radiological Environmental Monitoring Program 16.11.13 TABLE 16.11.13-3 (Page 1 of 3)

MAXIMUM VALUES FOR THE LOWER LIMITS OF DETECTION (LLD) (1)(2)(3)

ANALYSIS WATER AIRBORNE FISH MILK BROAD LEAF SEDIMENT (pCi/I) PARTICULATE (pCi/kg, wet) (pCi/I) VEGETATION (pCi/kg, dry)

OR GASES (pCi/kg, wet)

(pCi/m 3)

Gross Beta 4 0.01 N/A N/A N/A N/A H-3 2000* 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, 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 15 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 1(4) 0.07 N/A 1 60 N/A Cs-1 34 15 0.05 130 15 60 150 Cs-1 37 18 0.06 150 18 80 180 Ba-140 15 N/A N/A 15 N/A N/A La-1 40 15 N/A N/A 15 N/A N/A

If no drinking water pathway exists, a value of 3000 pCi/I may be used.

McGuire Units 1 and 2 16.11.13-10 Revision 137

Radiological Environmental Monitoring Program 16.11.13 TABLE 16.11.13-3 (Page 2 of 3)

MAXIMUM VALUES FOR THE LOWER LIMITS OF DETECTION (LLD)

NOTES:

1. The LLD is defined, for purposes of these commitments, as the smallest concentration of radioactive material in a sample that will yield a net count (above system background) that will be detected with 95% probability with 5% probability of falsely concluding that a blank observation represents a "real" signal.

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

(2.71 / T) + 4.65Sb LLD -

E V -2.22 *Y.exp (-AAt)

Where:

LLD is the "a priori" lower limit of detection as defined above (as picoCurie per unit mass or volume),

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

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

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

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

X is the radioactive decay constant for the particular radionuclide, At is the elapsed time between sample collection (or end of the sample collection period) and time of counting (for environmental samples, not plant effluent samples), and T is the background and sample counting time in minutes.

It should be recognized that the LLD is defined as an a priori (before the fact) limit representing the capability of a measurement system and not as an a posteriori (after the fact) limit for a particular measurement. Analyses shall be performed in such a manner that the stated LLDs will be achieved under routine conditions.

Occasionally background fluctuations, unavoidable small sample sizes, the presence of interfering nuclides, or other uncontrollable circumstances may render these LLDs unachievable. In such cases, the contributing factors shall be identified and described in the Annual Radiological Environmental Operating Report.

McGuire Units 1 and 2 16.11.13-11 Revision 137

Radiological Environmental Monitoring Program 16.11.13 TABLE 16.11.13-3 (Page 3 of 3)

MAXIMUM VALUES FOR THE LOWER LIMITS OF DETECTION (LLD)

NOTES (continued):

2. This list does not mean that only these nuclides are to be considered. Other peaks that are identifiable, together with those of the above nuclides, shall also be analyzed and reported in the Annual Radiological Environmental Operating Report.

3. Required detection capabilities for thermoluminescent dosimeters used for environmental measurements shall be in accordance with the recommendations of Regulatory Guide 4.13.

4. LLD for drinking water samples. If no drinking water pathway exists, the LLD of gamma isotopic analysis may be used.

McGuire Units 1 and 2 16.11.13-12 Revision 137

Radiological Environmental Monitoring Program 16.11.13 BASES The Radiological Environmental Monitoring Program is established to monitor the radiation and radionuclides in the environs of the plant. The program provides representative measurements of radioactivity in the highest potential exposure pathways, and verification of the accuracy of the effluent monitoring program and modeling of environmental exposure pathways. The program is contained in SLC 16.11.13 - 16.11.16 and conforms to the guidance of Appendix I to 10 CFR Part 50.

The program includes the following:

1. Monitoring, sampling, analysis, and reporting of radiation and radionuclides in the environment in accordance with the methodology and parameters in the ODCM,

2. A Land Use Census to ensure that changes in the use of areas at and beyond the SITE BOUNDARY are identified and that modifications to the monitoring program are made if required by the results of this census, and

3. Participation in an Interlaboratory Comparison Program to ensure that independent checks on the precision and accuracy of the measurements of radioactive materials in environmental sample matrices are performed as part of the quality assurance program for environmental monitoring.

The portion of the Radiological Environmental Monitoring Program required by this commitment provides representative measurements of radiation and of radioactive materials in those exposure pathways and for those radionuclides that lead to the highest potential radiation exposures of MEMBERS OF THE PUBLIC resulting from the station operation. This monitoring program implementsSection IV.B.2 of Appendix I to 10 CFR Part 50 and 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 the modeling of the environmental exposure pathways. Guidance for this monitoring program is provided by the Radiological Assessment Branch Technical Position on Environmental Monitoring. The initially specified monitoring program will be effective for at least the first 3 years of commercial operation. Following this period, program changes may be initiated based on operational experience.

The required detection capabilities for environmental sample analyses are tabulated in terms of the lower limits of detection (LLDs). The LLDs required by Table 16.11.13-3 are considered optimum for routine environmental measurements in industrial laboratories. It should be recognized that the LLD is defined as an a priori (before the fact) limit representing the capability of a measurement system and not as an a posteriori (after the fact) limit for a particular measurement.

With the level of radioactivity in an environmental sampling medium at a specified location exceeding the reporting levels of Table 16.11.13-3 when averaged over any calendar quarter, in lieu of a Licensee Event Report, prepare and submit to the Commission within 30 days a Special Report that defines the corrective actions to be McGuire Units 1 and 2 16.11.13-13 Revision 137

Radiological Environmental Monitoring Program 16.11.13 BASES (continued) taken to reduce radioactive effluents so that the potential annual dose to a MEMBER OF THE PUBLIC is less than the calendar year limits of SLCs 16.11.6, 16.11.8, and 16.11.9. When more than one of the radionuclides in Table 16.11.13-2 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 16.11.13-2 are detected and are the result of plant effluents, this report shall be submitted if the potential annual dose to a MEMBER OF THE PUBLIC is equal to or greater than the calendar year limits of SLCs 16.11.6, 16.11.8 and 16.11.9. This report is not required if the measured level of radioactivity was not the result of plant effluents; however, in such an event, the condition shall be reported and described in the Annual Radiological Environmental Operating Report. The methodology and parameters used to estimate the potential annual dose to a MEMBER OF THE PUBLIC shall be indicated in this report.

Detailed discussion of the LLD, and other detection limits, can be found in HASL Procedures Manual, HASL-300 (revised annually), Currie, L. A., "Limits for Qualitative Detection and Quantitative Determination - Application to Radiochemistry," Anal. Chem.

40, 586-93 (1968), and Hartwell, J. K., "Detection Limits for Radioanalytical Counting Techniques," Atlantic Richfield Hanford Company Report ARH-SA-215 (June 1975).

REFERENCES

1. McGuire Nuclear Station, Off site Dose Calculation Manual

2. 10 CFR Part 50, Appendix I McGuire Units 1 and 2 16.11.13-14 Revision 137

Land Use Census 16.11.14 16.11 RADIOLOGICAL EFFLUENT CONTROLS 16.11.14 Land Use Census COMMITMENT A land use census shall be conducted and shall identify within a distance of 8 km (5 miles) the location in each of the 16 meteorological sectors of:

a. the nearest milk animal,

b. the nearest residence, and

c. the nearest garden of greater than 50 m 2 (500 ft 2) producing broad leaf vegetation.

For elevated releases as defined in Regulatory Guide 1.111, Revision 1, July 1977, the land use census shall identify within a distance of 5 km (3 miles) the location in each of the 16 meteorological sectors of:

a. all milk animals, and

b. all gardens of greater than 50 m2 producing broad leaf vegetation.

n*dlL

Broad leaf vegetation sampling of three different kinds of vegetation may be performed at the SITE BOUNDARY in each of two different direction sectors with the highest predicted D/Qs in lieu of the garden census. Specifications for broad leaf vegetation sampling in Table 16.11.13-1 4c shall be followed, including analysis of control samples.

APPLICABILITY At all times.

REMEDIAL ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Location(s) identified A.1 Identify the new location in In next scheduled which yields a calculated the Annual Radioactive Annual Radioactive dose/dose commitment Effluent Release Report. Effluent Release greater than values Report currently calculated in SLC 16.11.9.

(continued)

McGuire Units 1 and 2 16.11.14-1 Revision 21

Land Use Census 16.11.14 REMEDIAL ACTIONS (continued)

B. Location(s) identified B.1 Add the new location to the 30 days which yields a Radiological Environmental calculated dose or dose Monitoring Program.

commitment (via same exposure pathway) 20% AND greater than at a location from which samples are 8.2 NOTES----

currently being obtained If samples cannot be in accordance with SLC obtained, an explanation of 16.11.13. why samples are not obtainable (substitute representative locations if possible) shall be included.

Identify the new In the next location(s), revised figures scheduled Annual and tables for the ODCM, Radiological Release in the next Annual Report Radiological Release Report.

TESTING REQUIREMENTS TEST FREQUENCY TR 16.11.14.1 -- N--

The results of the land use census shall be included in the Annual Radiological Environmental Operating Report.

Conduct a land use census during the growing season 12 months using the information which will provide the best results such as a door-to-door survey, aerial survey, or consultation with local agricultural authorities.

McGuire Units 1 and 2 16.11.14-2 Revision 21

Land Use Census 16.11.14 BASES This commitment is provided to ensure that changes in the use of areas at and beyond the SITE BOUNDARY are identified and that modifications to the Radiological Environmental Monitoring Program are made if required by the results of this census. The best information from the door-to-door survey, from aerial survey, or from consulting with local agricultural authorities shall be used. This census satisfies the requirements of Section IV.B.3 of m2 Appendix I to 10 CFR Part 50. Restricting the census to gardens of greater than 50 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 made: (1) 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 kgl/M2.

With a land use census identifying a location(s) which yields a calculated dose or dose commitment (via the same exposure pathway) 20% greater than at a location from which samples are currently being obtained in accordance with SLC 16.11.13, add the new location to the Radiological Environmental Monitoring Program. The sampling location(s),

excluding the control station location, having the lowest calculated dose or dose commitment (via the same exposure pathway) may be deleted from this monitoring program after October 31 of the year in which this land use census was conducted.

REFERENCES

1. McGuire Nuclear Station, Off site Dose Calculation Manual

2. 10 CFR Part 50, Appendix I McGuire Units 1 and 2 16.11.14-3 Revision 21

Interlaboratory Comparison Program 16.11.15 16.11 RADIOLOGICAL EFFLUENT CONTROLS 16.11.15 Interlaboratory Comparison Program COMMITMENT Analyses shall be performed on radioactive materials, supplied as part of an Interlaboratory Comparison Program (ICP), that correspond to samples required by SLC 16.11.13.

APPLICABILITY At all times.

REMEDIAL ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Analyses not being A.1 Report corrective actions In next scheduled performed as required. taken to prevent recurrence Annual Radiological in the Annual Radiological Environmental Environmental Operating Operating Report Report.

TESTING REQUIREMENTS TEST FREQUENCY TR 16.11.15.1 Report a summary of the results of the Interlaboratory 12 months Comparison Program in the Annual Radiological Environmental Operating Report.

McGuire Units I and 2 16.11.15-1 Revision 21

Interlaboratory Comparison Program 16.11.15 BASES This 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 valid for the purposes of Section IV.B.2 of Appendix I to 10 CFR Part 50.

The Interlaboratory Comparison Program (ICP) shall be described in the Annual Radiological Environmental Operating Report.

REFERENCES

1. 10 CFR Part 50, Appendix I McGuire Units 1 and 2 16,11.15-2 Revision 21

Annual Radiological Environmental Operating Report 16.11.16 16.11 RADIOLOGICAL EFFLUENT CONTROLS 16.11.16 Annual Radiological Environmental Operating Report COMMITMENT Routine Annual Radiological Environmental Operating Reports covering the operation of the unit during the previous calendar year shall be submitted by May 15 of each year.

The Annual Radiological Environmental Operating Reports 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 pre-operational studies, with operational controls as appropriate, and with 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 SLC 16.11.14.

The Annual Radiological Environmental Operating Reports shall include the results of analysis of all radiological environmental samples and of all environmental radiation measurements taken during the period pursuant to the locations specified in the Table and Figures in the ODCM, as well as summarized and tabulated results of these analyses and measurements in the format of the table in the Radiological Assessment Branch Technical Position, Revision 1, November 1979. In the event that some individual results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results. The missing data shall be submitted as soon as possible in a supplementary report.

The reports shall also include the following:

a summary description of the Radiological Environmental Monitoring Program; 0 at least two legible maps covering all sampling locations keyed to a table giving distances and directions from the centerline of one reactor (one map shall cover stations near the site boundary; a second shall include the more distant stations);

the results of licensee participation in the Interlaboratory Comparison Program, required by SLC 16.11.15;

a discussion of all deviations from the sampling schedule of Table 16.11.13-1; and McGuire Units I and 2 16.11.16-1 Revision 134

Annual Radiological Environmental Operating Report 16.11.16 COMMITMENT (continued) a discussion of all analyses in which the LLD required by Table 16.11.13-3 was not achievable.

A single submittal may be made for a multiple unit station..

APPLICABILITY At all times.

REMEDIAL ACTIONS None TESTING REQUIREMENTS None BASES None REFERENCES

1. Technical Specification 5.6.2 McGuire Units 1 and 2 16.11.16-2 Revision 134

Radioactive Effluent Release Reports 16.11.17 16.11 RADIOLOGICAL EFFLUENT CONTROLS 16.11.17 Radioactive Effluent Release Reports COMMITMENT Routine Radioactive Effluent Release Reports covering the operation of the unit durng the previous calendar year of operation shall be submitted before May 1 of each year.

The Radioactive Effluent Release Reports 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 calendar year. This annual summary may be either in the form of an hour-by-hour listing on magnetic tape of wind speed, wind direction, atmospheric stability, and precipitation (if measured), 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 or station during the previous calendar year. A five year average of representative onsite meteorological data shall be used in the gaseous effluent dose pathway calculations. Dispersion factors (X/Qs) and deposition factors (D/Qs) shall be generated using the computer code XOQDOQ (NUREG/CR-2919) which implements NRC Regulatory Guide 1.111.

The meteorological conditions concurrent with the time of release shall be reviewed annually to determine if the five-year average values should be revised. The assessment of radiation doses shall be performed in accordance with the methodology and parameters in 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 calendar year to show conformance with 40 CFR Part 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, October 1977.

McGuire Units I and 2 16.11.17-1 Revision 118

Radioactive Effluent Release Reports 16.11.17 COMMITMENT (continued)

The Radioactive Effluent Release Reports shall include the following information for each type of solid waste shipped offsite or disposed of in the site landfill during the report period:

a. Total container volume, in cubic meters,

b. Total Curie quantity (determined by measurement or estimate),

c. Principal radionuclides (determined by measurement or estimate),

d. Type of waste (e.g., dewatered spent resin, compacted dry waste, evaporator bottoms),

e. Number of shipments, and

f. Solidification agent or absorbent (e.g., cement, or other approved agents (media)).

The Radioactive Effluent Release Reports shall include a list and description of unplanned releases from the site to UNRESTRICTED AREAS of radioactive materials in gaseous and liquid effluents made during the reporting period.

The Radioactive Effluent Release Reports shall include any changes made during the reporting period to the PROCESS CONTROL PROGRAM (PCP) and to the OFFSITE DOSE CALCULATION MANUAL (ODCM), as well as a listing of new locations for dose calculations and/or environmental monitoring identified by the land use census pursuant to SLC 16.11.14.

The Radioactive Effluent Release Reports shall also identify any licensee initiated major changes to the Radioactive Waste Systems (liquid, gaseous, and solid). Otherwise, this information may be included in the annual UFSAR update. 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 Part 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 of radioactive materials in liquid and gaseous effluents and/or quantity of solid waste that differ from those previously predicted in the License application and amendments thereto; McGuire Units I and 2 16.11.17-2 Revision 118

Radioactive Effluent Release Reports 16.11.17 COMMITMENT (continued)

e. An evaluation of the change, which shows 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 and in solid waste, 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; and

h. Documentation of the fact that the change was reviewed and found acceptable by the Station Manager or the Chemistry Manager.

A single submittal may be made for a multiple unit station. The submittal should combine those sections that are common to all units at the station; however, for units with separate Radwaste Systems, the submittal shall specify the releases of radioactive material from each unit.

APPLICABILITY At all times REMEDIAL ACTIONS None TESTING REQUIREMENTS None BASES None REFERENCES

1. Technical Specification 5.6.3 McGuire Units I and 2 16.11.17-3 Revision 118

Liquid Holdup Tanks 16.11.18 16.11 RADIOLOGICAL EFFLUENT CONTROLS 16.11.18 Liquid Holdup Tanks COMMITMENT The quantity of radioactive material contained in each unprotected outdoor radwaste tank shall be limited to < 10 Curies, excluding tritium and dissolved or entrained noble gases.

APPLICABILITY At all times.

REMEDIAL ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Quantity of radioactive A. 1 Suspend all additions of Immediately material in tank not radioactive material to the within limit, tank.

AND A.2 Reduce the tank contents 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months to within limit.

AND A.3 Describe the events Within the next leading to this condition in scheduled Annual the next Annual Radioactive Effluent Radioactive Effluent Release Report Release Report.

TESTING REQUIREMENTS TEST FREQUENCY TR 16.11.18.1 Verify the quantity of radioactive material contained in 7 days unprotected outdoor radwaste tanks is within limits by analyzing a representative sample of the tank's contents when radioactive materials are being added to the tank.

McGuire Units 1 and 2 16.11.18-1 Revision 0

Liquid Holdup Tanks 16.11.18 BASES The tanks applicable to this SLC include all those outdoor radwaste tanks that are not surrounded by liners, dikes, or walls capable of holding the tank contents and that do not have tank overflows and surrounding area drains connected to the Liquid Radwaste Treatment System.

Restricting the quantity of radioactive material contained in the specified tanks provides assurance that in the event of an uncontrolled release of the tanks' contents, the resulting concentrations would be less than the limits of 10 CFR Part 20, Appendix B, Table II, Column 2, at the nearest potable water supply and the nearest surface water supply in an UNRESTRICTED AREA.

REFERENCES None McGuire Units 1 and 2 16.111.18B-2 Revision 0

Explosive Gas Mixture 16.11.19 16.11 RADIOLOGICAL EFFLUENT CONTROLS 16.11.19 Explosive Gas Mixture COMMITMENT The concentration of oxygen in the WASTE GAS HOLDUP SYSTEM shall be limited to < 2% by volume whenever the hydrogen concentration exceeds 4% by volume.

APPLICABILITY At all times.

REMEDIAL ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Concentration of oxygen A.1 Reduce oxygen 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months in the WASTE GAS concentration to within HOLDUP SYSTEM limits.

> 2% but < 4% by volume.

B. Concentration of oxygen 13.1 Suspend all additions of Immediately in the WASTE GAS waste gases to the system.

HOLDUP SYSTEM

> 4% and hydrogen AND concentration > 4% by volume. B.2 Reduce the concentration Immediately of oxygen to < 4% by volume.

AND B.3 Reduce oxygen 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months concentration to within limits.

McGuire Units 1 and 2 16.11.19-1 Revision 0

Explosive Gas Mixture 16.11.19 TESTING REQUIREMENTS TEST FREQUENCY TR 16.11.19.1 Verify the concentrations of hydrogen and oxygen in the During WASTE WASTE GAS HOLDUP SYSTEM is within limits by GAS HOLDUP monitoring waste gases in the WASTE GAS HOLDUP SYSTEM SYSTEM with the hydrogen and oxygen monitors operation required by SLC 16.7.8.

BASES This specification is provided to ensure that the concentration of potentially explosive gas mixtures contained in the WASTE GAS HOLDUP SYSTEM is maintained below the flammability limits of hydrogen and oxygen. Automatic control features are included in the system to prevent the hydrogen and oxygen concentrations from reaching these flammability limits. These automatic control features include isolation of the source of hydrogen and/or oxygen, automatic diversion to recombiners, or injection of dilutants to reduce the concentration below the flammability limits. Maintaining the concentration of hydrogen and oxygen below their flammability limits provides assurance that the releases of radioactive materials will be controlled in conformance with the requirements of General Design Criterion 60 of Appendix A to 10 CFR Part 50.

REFERENCES None McGuire Units 1 and 2 16.11.19-2 Revision 0

Gas Storage Tanks 16.11.20 16.11 RADIOLOGICAL EFFLUENT CONTROLS 16.11.20 Gas Storage Tanks COMMITMENT The quantity of radioactivity contained in each gas storage tank shall be limited < 49,000 Curies noble gases (considered as Xe-133).

APPLICABILITY At all times.

REMEDIAL ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Quantity of radioactive A.1 Suspend all additions of Immediately material in tank not radioactive material to the within limit, tank.

AND A.2 Reduce the tank contents 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months to within limit.

TESTING REQUIREMENTS TEST FREQUENCY TR 16.11.20.1 Verify the quantity of radioactive material contained in 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months each gas storage tank is within limit when radioactive materials are being added to the tank.

McGuire Units 1 and 2 16.11.20-1 Revision 0

Gas Storage Tanks 16.11.20 BASES This SLC considers postulated radioactive releases due to a waste gas system leak or failure, and limits the quantity of radioactivity in each pressurized gas storage tank in the WASTE GAS HOLDUP SYSTEM to assure that a release would be substantially below the dose guideline values of 10 CFR Part 100 for a postulated event.

Restricting the quantity of radioactivity contained in each gas storage tank provides assurance that in the event of an uncontrolled release of the tank's contents, the resulting total body exposure to a MEMBER OF THE PUBLIC at the nearest exclusion area boundary will not exceed 0.5 rem. This is consistent with Standard Review Plan 11.3, Branch Technical Position ETSB 11-5, "Postulated Radioactive Releases Due to a Waste Gas System Leak or Failure," in NUREG-0800, July 1981.

REFERENCES None McGuire Units 1 and 2 16.11.20-2 Revision 0

Attachment 8 Revisions to the Radioactive Waste Process Control Program Manual

March 17, 2014 J.N. Robertson McGuire Nuclear Station Nuclear Regulatory Affairs Attention: K.L. Crane

Subject:

McGuire Nuclear Station 2013 Annual Radioactive Effluent Release Report Process Control Program Changes File: GS-764.25, MC-215.06 This memo documents that there have been no revisions to the Radioactive Waste Process Control Program Manual since the 2012 ARERR transmittal in 2013. This statement is to be included in the NRC distribution of the Annual Radioactive Effluent Release Report for McGuire Nuclear Station for the period of January L, 2013 through December 31, 2013.

If you have any questions, please call David Vaught @ 980-373-5302.

James A. Mockridge Supervising Scientist Nuclear Chemistry by: David L. Vaught Lead Engineer Nuclear Chemistry - Radwaste

Attachment 9 Information to Support the Nuclear Energy Institute (NEI) Groundwater Protection Initiative

2013 McGuire ARERR Groundwater Well Data Section Duke Energy implemented a Groundwater Protection Program in 2007. This program was developed to ensure timely and effective management of situations involving inadvertent releases of licensed material to ground water. As part of this program, McGuire Nuclear Station monitored sixty ground water wells during 2013.

Wells are sampled quarterly, semi-annually or annually. Ground water samples are regularly analyzed for tritium and gamma emitters, with selected wells being analyzed for difficult to detect radionuclides. No gamma or difficult to detect radionuclides (other than naturally occurring radionuclides) were identified in well samples during 2013. Results from sampling during 2013 confirmed existing knowledge of tritium concentrations in site ground water (shown in the table below). Lining of the Conventional Waste (WC) Ponds was performed in 2011 and 2012 along with WC piping replacement in 2013. Tritium concentrations in wells near these ponds have continued to decrease.

Results from sampling during 2013 are shown in the table below.

Welll Location / Description Tritium Concentration (pCi/I) # of Name 1st Qtr 2nd Qtr 3rd Qtr 4th Qtr Samples M-100R MNS GWPI / M-100R / SE of WC 3.37E+02 2.20E+02 <MDA 1.97E+02 4 M-101 MNS GWPI/ M-101 /SE of WC 2.69E+02 2.58E+02 <MDA 1.69E+02 4 M-102 MNS GWPI/ M-102/SW of WC 5.66E+03 4.73E+03 4.84E+03 4.39E+03 4 M-103 MNSGWPI/ M-103/SofWC 2.04E+03 1.59E+03 1.61E+03 1.54E+03 4 M-103R MNSGWPI/ M-103R/SofWC 2.17E+03 1.88E+03 1.62E+03 1.46E+03 4 M-104DR MNS GWPI/ M-104DR/WofWC 2.64E+03 2.44E+03 2.13E+03 1.94E+03 4 M-104R MNS GWPI / M-104R / W of WC 1.90E+03 1.60E+03 1.50E+03 1.19E+03 4 M-105 MNS GWPI/ M-105 / Landfarm NS <MDA NS <MDA 2 M-20 MNS GWPI I M-20 / S of Hwy. 73 NS 6.81E+02 NS 6.19E+02 2 M-20R MNS GWPI / M-20R / S of Hwy. 73 NS 4.70E+02 NS 5.51E+02 2 M-21 MNS GWPI/ M-21/ S of Hwy. 73 NS <MDA NS <MDA 2 M-22 MNS GWPI/ M-22/ S of Hwy. 73 NS <MDA NS <MDA 2 M-22R MNS GWPI/ M-22R / S of Hwy. 73 NS <MDA NS <MDA 2 M-23 MNS GWPI/ M-23/ S of Acs. Rd. NS <MDA NS <MDA 2 M-30 MNS GWPI/ M-30 /WWCB NS <MDA NS <MDA 2 M-30R MNS GWPI / M-30R / WWCB NS 2.08E+02 NS <MDA 2 M-31 MNS GWPI/ M-31 /Access road NS <MDA NS <MDA 2 M-32 MNS GWPI M-32 / Main entrance NS <MDA NS <MDA 2 M-34DR MNS GWPI/ M-34DR / Access road <MDA <MDA <MDA <MDA 4 M-34R MNS GWPI/ M-34R / Access road <MDA <MDA <MDA <MDA 4 M-35 MNS GWPI/ M-35 /Access road <MDA <MDA <MDA <MDA 4 M-42 MNS GWPI/ M-42 /U-2 Rx. Bldg. 2.10E+03 2.31E+03 1.80E+03 1.63E+03 4 M-48DR MNS GWPI/ M-48DR / U-2 SFP <MDA 2.14E+02 2.43E+02 <MDA 4 M-48R MNS GWPI/ M-48R / U-2 SFP 7.06E+02 7.35E+02 7.99E+02 5.91E+02 4 M-53 IVINS GWPI / M-53 / N of plant 8.45E+02 I 7.89E+02 8.30E+02 7.7711+02 4

2013 McGuire ARERR Groundwater Well Data Section M-55 MNS GWPI/ M-55/NAB <MDA <MDA 1.76E+02 <MDA 4 M-59 MNS GWPI M-59 U-2 Doghouse 2.11E+03 2.11E+03 1.83E+03 1.74E+03 4 M-60 MNS GWPI/ M-60 /MOC Parking NS <MDA NS <MDA 2 M-62 MNS GWPI/ M-62/ S of RWF <MDA 2.02E+02 <MDA <MDA 4 M-64 MNS GWPI/ M-64 /Rdwst. Bldg. 5.61E+02 6.22E+02 6.54E+02 4.72E+02 4 M-66 MNS GWPI/ M-66/S of SSF 7.25E+02 8.84E+02 8.47E+02 8.15E+02 4 M-66R MNS GWPI/ M-66R / S of SSF <MDA <MDA <MDA <MDA 4 M-68 MNS GWPI / M-68 / U-1 RMWST 4.87E+02 5.55E+02 5.82E+02 NS 3 M-70 MNS GWPI M-70 U-1 SFP 4.43E+02 4.24E+02 4.62E+02 4.58E+02 4 M-70DR MNS GWPI / M-70DR / U-1 SFP <MDA <MDA <MDA <MDA 4 M-70R MNS GWPI/ M-70R / U-1 SFP 1.97E+02 2.08E+02 1.87E+02 1.53E+02 4 M-72 MNS GWPI M-72 / Rdwst. Trench 8.24E+02 8.80E+02 8.49E+02 7.06E+02 4 M-76 MNS GWPI/ M-76/ W of U-1 SFP NS NS NS 3.80E+02 1 M-82 MNS GWPI/ M-82 /River 1.57E+03 1.68E+03 1.37E+03 1.17E+03 4 M-84 MNS GWPI/ M-84 /River 4.13E+03 2.97E+03 3.56E+03 4.18E+03 4 M-84R MNS GWPI/ M-84R / River 6.42E+03 6.24E+03 5.76E+03 5.15E+03 4 M-85 MNS GWPI / M-85 River 1.28E+03 1.46E+03 1.15E+03 1.06E+03 4 M-87 MNS GWPI/ M-87 /Landfarm 4.62E+02 7.21E+02 4.44E+02 4.81E+02 4 M-89 MNS GWPI/ M-89 /Landfarm 4.21E+02 6.OOE+02 5.80E+02 6.34E+02 4 M-90 MNS GWPI / M-90 / Landfarm NS 2.OOE+02 NS 5.21E+02 2 M-91 MNS GWPI M-91 E of WC 4.08E+02 2.99E+02 4.77E+02 4.45E+02 4 M-91R MNS GWPI/ M-91R/E of WC 3.05E+02 2.31E+02 3.64E+02 3.59E+02 4 M-92 MNS GWPI/ M-92 / N of WC Ponds NS 1.83E+02 NS 4.22E+02 2 M-92R MNS GWPI/ M-92R / N of WC Ponds NS <MDA NS <MDA 2 M-93 MNS GWPI / M-93 / N of IHUP NS 5.38E+02 NS 6.39E+02 2 M-93R MNS GWPI/ M-93R / N of IHUP NS <MDA NS <MDA 2 M-94 MNS GWPI/ M-94 /SE of IHUP NS <MDA NS <MDA 2 M-95 MNS GWPI / M-95 / Lower Parking NS <MDA NS <MDA 2 M-95R MNS GWPI / M-95R / Lower Parking NS <MDA NS <MDA 2 M-96 MNS GWPI / M-96 / West Parking NS <MDA NS <MDA 2 M-96R MNS GWPI/ M-96R / West Parking NS <MDA NS <MDA 2 M-97 MNS GWPI / M-97 / East Parking NS 2.09E+02 NS <MDA 2 M-98 MNS GWPI/ M-98/ S of Amin. Bldg. NS <MDA NS <MDA 2 M-98R MNS GWPI/ M-98R / S of Amin. Bldg. NS <MDA NS <MDA 2 MS-1 MNS GWPI MS-1 /Surface Water <MDA <MDA <MDA <MDA 4 MS-2 MNS GWPI / MS-2 / Surface Water 3.28E+02 4.04E+02 3.70E+02 4.60E+02 4 MS-3 MNS GWPI/ MS-3 Surface Water 4.97E+02 3.73E+02 5.59E+02 5.23E+02 4 MS-4 MNS GWPI/ MS-4lSurface Water 4.59E+02 6.23E+02 4.70E+02 6.34E+02 4

2013 McGuire ARERR Groundwater Well Data Section

NS - Not sample due to insufficient volume in well to sample.

pCi/l - pico curies per liter

<MDA - less than minimum detectable activity, typically 250 pCi/liter 20,000 pCi/l - the Environmental Protection Agency drinking water standard for tritium. This standard applies only to water that is used for drinking.

1,000,000 pCi/l - the 1OCFR20, Appendix B, Table 2, Column 2, Effluent Concentration limit for tritium.

Attachment 10 Inoperable Monitoring Equipment (January 1, 2013 through December 31, 2013)

There were no SLC related effluent monitoring instruments out of service greater than the SLC limits for functionality.

Attachment 11 Radioactive Waste Systems Changes This attachment documents the changes made to the radioactive waste systems at the McGuire Nuclear Station during the period of January 1, 2013 to December 31, 2013.

There were no significant changes to the radioactive waste systems during 2013 at the McGuire Nuclear Station.

MNS-14-036, Annual Radioactive Effluent Release Report

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MNS-14-036, Annual Radioactive Effluent Release Report

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