Text

Annual Radioactive Effluent Release Report
	ML12135A595
Person / Time
Site:	Catawba
Issue date:	04/26/2012
From:	Morris J; Duke Energy Carolinas
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	Download: ML12135A595 (162)
	v • d • e

JAMES R. MORRIS

ý§Khk Duke Vice President rO-Energy.

Duke Energy Catawba Nuclear Station 4800 Concord Road York, SC 29745 April 26, 2012 803-701-4251 803-701-3221 fax U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, D.C. 20555

Subject:

Duke Energy Carolinas, LLC Catawba Nuclear Station, Units 1 and 2 Docket Nos. 50-413 and 50-414 2011 Annual Radioactive Effluent Release Report Pursuant to Catawba Nuclear Station Technical Specification (TS) 5.6.3 and Selected Licensee Commitment 16.11-16, please find attached the Annual Radioactive Effluent Release Report for the period of January 1, 2011, through December 31, 2011. In accordance with Catawba TS 5.5.1, the Offsite Dose Calculation Manual (ODCM) is included in this submittal.

Attachment 1 Summary of Gaseous and Liquid Effluents Report Attachment 2 Supplemental Information Attachment 3 Solid 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 Revisions to UFSAR Section 16.11, Radiological Effluent Controls Attachment 8 Revisions to the Radioactive Waste Process Control Program Manual (Compact Disc)* if changes to program Attachment 9 Information to Support the NEI Groundwater Protection Initiative Attachment 10 Inoperable Equipment Attachment 11 Radioactive Waste Systems Changes Enclosure 2011 Offsite Dose Calculation Manual (Compact Disc) www. duke-energy.corn

U.S. Nuclear Regulatory Commission 2011 Annual Radioactive Effluent Release Report April 26, 2012 Page 2 Any questions concerning this report should be directed to Toni Pasour at (803) 701-3566.

Sincerely, James R. Morris Attachments and Enclosures (Process Control Program [PCP] Revision Compact Disc [CD]* if changes to program and Offsite Dose Calculation Manual [ODCM] Compact Disc [CD]* ifrevised from previous submittal)

U.S. Nuclear Regulatory Commission 2011 Annual Radioactive Effluent Release Report April 26, 2012 Page 3 xc (with attachments and enclosures):

V. M. McCree Regional Administrator U.S. Nuclear Regulatory Commission - Region II Marquis One Tower 245 Peachtree Center Ave., NE Suite 1200 Atlanta, GA 30303-1257 J.H. Thompson NRC Project Manager U.S. Nuclear Regulatory Commission Mail Stop 8-G9A 11555 Rockville Pike Rockville, MD 20852-2738 xc (with attachments and PCP CD):

G. A. Hutto, III NRC Senior Resident Inspector Russell Keown, Supervisor Analytical & Radiological Environmental Services Division 2600 Bull Street Columbia, SC 29201 xc (with attachments only):

Sandra Flemming, Director Analytical & Radiological Environmental Services Division 8231 Parklane Road Columbia, SC 29201 Susan E. Jenkins, Manager Radioactive & Infectious Waste Management Division of Waste Management S.C. Department of Health and Environmental Control 2600 Bull Street Columbia, SC 29201 Tom Knight Contamination Mitigation Section Bureau of Land & Waste Management SC Department of Health and Environmental Control 2600 Bull Street Columbia, SC 29201

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, Appendix B. Radioactive liquid and gaseous wastes are sampled and analyzed per the requirements in Selected Licensee Commitment (SLC) Table 16.11 1, "Radioactive Liquid Waste Sampling and Analysis Program," and SLC Table 16.11-6-1, "Radioactive Gaseous Waste Sampling and Analysis Program." Included in the gaseous effluent releases is an estimate of Carbon-14 radioactivity released in 2011 (Ref. "Carbon-14Supplemental Information," contained in the ARERR for further information).

TABLE 1A EFFLUENT AND WASTE DISPOSAL ANNUAL REPORT PERIOD 1/1/11 TO 1/1/12 GASEOUS EFFLUENTS - SUMMATION OF ALL RELEASES Catawba Nuclear Station Units 1 & 2 REPORT FOR 2011 Unit QTR 1 QTR 2 QTR 3 QTR 4 YEAR A. Fission and Activation Gases

1. Total Release Ci 1.01E+00 6.42E-01 7.60E-01 8.43E-01 3.26E+00

2. Avg. Release Rate pCi/sec 1.30E-01 8.17E-02 9.56E-02 1.06E-01 1.03E-01 B. Iodine-131

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

2. Avg. Release Rate pCi/sec O.00E+00 0.OOE+00 0.OOE+00 O.OOE+00 0.OOE+00 C. Particulates Half Life >= 8 days

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

2. Avg. Release Rate pCi/sec 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 D. Tritium

1. Total Release Ci 6.59E+01 5.45E+01 4.83E+01 5.70E+01 2.26E+02

2. Avg. Release Rate pCi/sec 8.47E+00 6.93E+00 6.08E+00 7.17E+00 7.16E+00 E. Carbon-14

1. Total Release Ci 5.44E+00 4.01E+00 5.45E+00 5.32E+00 2.02E+01

2. Avg. Release Rate pCi/sec 6.99E-01 5.10E-01 6.86E-01 6.69E-01 6.41E-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 pCi/sec 0.OOE+00 0.OOE+00 0.OOE+00 O.OOE+00 O.OOE+00

TABLE 1B EFFLUENT AND WASTE DISPOSAL ANNUAL REPORT PERIOD 1/1/11 TO 1/1/12 GASEOUS EFFLUENTS - ELEVATED RELEASES - CONTINUOUS MODE Catawba Nuclear Station Units 1 & 2 REPORT FOR 2011 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 1/1/11 TO 1/1/12 GASEOUS EFFLUENTS - ELEVATED RELEASES - BATCH MODE Catawba Nuclear Station Units 1 & 2 REPORT FOR 2011 Jnit *TR 1 QTR 2 QTR 3 QTR 4 YEAR

1. Fission and Activation Ga ses

- 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 1/1/11 TO 1/1/12 GASEOUS EFFLUENTS - GROUND RELEASES - CONTINUOUS MODE Catawba Nuclear Station Units 1 & 2 REPORT FOR 2011 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 H-3 Ci 6.57E+01 5.43E+01 4.82E+01 5.68E+01 2.25E+02

5. Carbon-14 C-14 Ci 1.63E+00 1.20E+00 1. 64E+00 1. 60E+00 6.07E+00

6. Gross Alpha Radioactivity

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

TABLE IC EFFLUENT AND WASTE DISPOSAL ANNUAL REPORT PERIOD 1/1/11 TO 1/1/12 GASEOUS EFFLUENTS - GROUND RELEASES - BATCH MODE Catawba Nuclear Station Units 1 & 2 REPORT FOR 2011 Unit QTR 1 QTR 2 QTR 3 QTR 4 YEAR

1. Fission and Activation Gases AR-41 Ci 7.87E-01 5.53E-01 6. 96E-01 7.40E-01 2.78E+00 KR-85 Ci 0.OOE+00 0.OOE+00 8.17E-04 0. OOE+00 8.17E-04 XE-133 Ci 2.11E-01 8. 60E-02 6.25E-02 1. OIE-01 4. 60E-01 XE-135 Ci 1.43E-02 3.29E-03 2.58E-04 2.08E-03 1. 99E-02 Totals for Period... Ci 1.01E+00 6.42E-01 7. 60E-01 8.43E-01 3.26E+00

2. Iodines

- No Nuclide Activities **

3. Particulates Half Life >= 8 days

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

4. Tritium H-3 Ci 1.34E-01 209E-0l l.82E-0l l7O~-O1 6Q~E-O1

5. Carbon-14 C-14 Ci 3.81E+00 2.81E+00 3.82E+00 3.72E+00 1.42E+01

6. Gross Alpha Radioactivity

- No Nuclide Activities **

TABLE 2A EFFLUENT AND WASTE DISPOSAL ANNUAL REPORT PERIOD 1/1/11 TO 1/1/12 LIQUID EFFLUENTS - SUMMATION OF ALL RELEASES Catawba Nuclear Station Units 1 & 2 REPORT FOR 2011 Unit QTR 1 QTR 2 QTR 3 QTR 4 YEAR A. Fission and Activation Products

1. Total Release Ci 4.26E-03 3.47E-02 1.31E-02 1.60E-02 6.81E-02

2. Average Diluted Concentration

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

b. Batch Releases pCi/ml 1.63E-10 1.11E-09 3.06E-10 6.36E-10 5.43E-10 B. Tritium

1. Total Release Ci 1.59E+02 2.76E+02 2.OOE+02 8.38E+02 1.47E+03

2. Average Diluted Concentration

a. Continuous Releases pCi/ml 1.33E-07 1.28E-06 6.74E-08 1.38E-07 4.06E-07

b. Batch Releases pCi/ml 6.06E-06 8.72E-06 4.66E-06 3.32E-05 1.17E-05 C. Dissolved and Entrained Gases

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

2. Average Diluted Concentration

a. Continuous Releases pCi/ml O.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00

b. Batch Releases pCi/ml 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 D. Gross Alpha Radioactivity

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

2. Average Diluted Concentration

a. Continuous Releases lICi/ml 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 O.OOE+00

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

1. Continuous Releases liters 7.08E+07 1.66E+08 3.60E+07 3.04E+07 3.03E+08

2. Batch Releases liters 3.83E+05 1.46E+06 1.38E+06 1.49E+06 4.72E+06 F. Volume of Dilution Water

1. Continuous Releases liters 2.61E+09 3.12E+09 4.28E+09 2.52E+09 1.25E+10

2. Batch Releases liters 2.61E+I0 3.12E+I0 4.28E+10 2.52E+10 1.25E+11

TABLE 2B EFFLUENT AND WASTE DISPOSAL ANNUAL REPORT PERIOD 1/1/11 TO 1/1/12 LIQUID EFFLUENTS - CONTINUOUS MODE Catawba Nuclear Station Units 1 & 2 REPORT FOR 2011 Unit QTR 1 QTR 2 QTR 3 QTR 4 YEAR

1. Fission and Activation Products

- No Nuclide Activities **

2. Tritium H-3 Ci 3.58E-01 4.21E+00 2.91E-01 3.53E-01 5.21E+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 1/1/11 TO 1/1/12 LIQUID EFFLUENTS - BATCH MODE Catawba Nuclear Station Units 1 & 2 REPORT FOR 2011 Unit QTR 1 QTR 2 QTR 3 QTR 4 YEAR

1. Fission and Activation Products AG-108M Ci 0.00E+00 0.00E+00 0. 00E+00 7. lIE-06 7.11E-06 AG-IIOM Ci 0. OOE+00 1.10E-03 3. 10E-04 1. 94E-04 1.60E-03 BE-7 Ci 0.00E+00 4.15E-05 0. 00E+00 1. 38E-03 1.42E-03 BI-214 Ci 0. OOE+00 8.98E-06 3. 97E-06 0. 00E+00 1.29E-05 CO-57 Ci 2. OOE-05 1.16E-04 3.81E-05 2.78E-05 2.02E-04 CO-58 Ci 2.76E-03 1. 06E-02 3.24E-03 3. 18E-03 1.98E-02 CO-60 Ci 7.74E-04 1.40E-02 7. 66E-03 7. 68E-03 3.01E-02 CR-51 Ci 0. OOE+00 4.24E-03 1. 99E-04 9. 94E-04 5.43E-03 CS-137 Ci 4. 69E-06 1. 19E-04 4. 96E-06 2.35E-06 1.31E-04 FE-59 Ci 0.00E+00 1. 40E-05 0. 00E+00 0.00E+00 1.40E-05 1-134 Ci 0. OOE+00 0. 00E+00 0.00E+00 2.18E-05 2.18E-05 MN-54 Ci 4.83E-05 8.40E-04 4.32E-04 2. 64E-04 1. 58E-03 NB-95 Ci 2.77E-08 3.24E-04 1. 33E-04 1. 07E-04 5. 64E-04 NB-97 Ci 0. OOE+00 2.79E-05 4. 66E-05 2. 00E-05 9.45E-05 PB-212 Ci 0.00E+00 0.00E+00 0. 00E+00 1. 09E-05 1. 09E-05 PB-214 Ci 0.00E+00 9. 00E-06 7.83E-06 0. 00E+00 1.68E-05 SB-124 Ci 3.09E-05 3.60E-04 6. 63E-05 9.82E-05 5.56E-04 SB-125 Ci 6.26E-04 2.73E-03 9.17E-04 1. 92E-03 6.19E-03 SN-113 Ci 0.OOE+00 2.31E-05 5.03E-06 0. 00E+00 2.82E-05 SR-92 Ci 0.00E+00 7.75E-06 9. 58E-06 3.73E-06 2.11E-05 ZN-65 Ci 0.00E+00 2. 68E-05 0. 00E+00 4. 66E-05 7,34E-05 ZR-95 Ci 0.00E+00 9.47E-05 1.59E-05 6.37E-05 .1. 74E-04 Totals for Period... ci 4.26E-03 3.47E-02 1.31E-02 1.60E-02 6.81E-02

2. Tritium H-3 Ci 1.58E+02 2.72E+02 2.00E+02 8.38E+02 1.47E+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 This attachment includes:

(1) Carbon- 14 Supplemental Information (2) Regulatory Guide 1.21, Revision 1, Supplemental Information (3) Overall Error Estimate for Liquid and Gaseous Effluent Release Data

. Page 1 of 2 Catawba 2011 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 Catawba, 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 Catawba, as defined in Regulatory Guide 1.21, Rev. 2. Catawba's 2011 Annual Radioactive Effluent Release Report (ARERR) contains estimates of C-14 radioactivity released in 2011, 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 Catawba 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 2011 Catawba 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 Catawba in 2011 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 2 (Ref. EPRI TR-105715). For the Catawba 2011 ARERR a value of 80% organic C-14 is assumed.

Page 2 of 2 Public dose estimates from airborne C-i14 are performed using dose models in NUREG-0 133 and Regulatory Guide 1.109. The dose models and assumptions used are documented in the Catawba ODCM. The estimated C-14 dose impact on the maximum organ dose from airborne effluents released from Catawba in 2011 is well below the 10CFR50, Appendix I, ALARA design objective (i.e., 15 mrem/yr per unit).

CATAWBA NUCLEAR STATION 2011 EFFLUENT AND WASTE DISPOSAL SUPPLEMENTAL INFORMATION I. REGULATORY LIMITS - PER UNIT A. NOBLE GASES - AIR DOS] 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 WITH HALF-LIVES > 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 10CFR20, 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 L:ICENSEE 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. 9.80E+01 = TOTAL NUMBER OF BATCH RELEASES

2. 6.38E+03 = TOTAL TIME (MIN.) FOR BATCH RELEASES.

3. 8.60E+01 = MAXIMUM TIME (MIN.) FOR A BATCH RELEASE.

4. 6.51E+01 = AVERAGE TIME (MIN.) FOR A BATCH RELEASE.

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

6. 6.28E+04 = AVERAGE DILUTION WATER FLOW DURING RELEASES (GPM).

B. GASEOUS EFFLUENT

1. 5.70E+01 = TOTAL NUMBER OF BATCH RELEASES.

2. 1.02E+06 = TOTAL TIME (MIN.) FOR BATCH RELEASES.

3. 4.99E+04 = MAXIMUM TIME (MIN.) FOR A BATCH RELEASE.

4. 1.79E+04 = AVERAGE TIME (MIN.) FOR A BATCH RELEASE.

5. 1.13E+02 = MINIMUM TIME (MIN.) FOR A BATCH RELEASE.

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

CATAWBA 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 Catawba 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

CATAWBA NUCLEAR STATION - SOLID RADIOACTIVE WASTE SHIPPED TO A DISP OSAL FACILITY REPORT PERIOD 1/1/2011 TO 12/31/2011 Total Number of Number of Waste Container Burial Volume Activity Type of Waste Shipped Shipments Containers Class Type (ft3) (m3) (Curies)

1. Waste from Liquid Systems (A) Dewatered Secondary Resins 0 0 NA NA 0.0 0.00 0.000 2AS (B) Dewatered Primary Resins 5 5 3B 5 HIC 772.5 21.88 497.790 (C) Evaporator Concentrates 0 0 NA NA 0.0 0.00 0.000 (D) Dewatered Mechanical Filters 2 2 2C 2 HIC 240.6 6.81 A8.436 (E) Dewatered Demineralizers 0 0 NA NA 0.0 0.00 0.000 (F) Solidified (Cement) Acids, Oils, Sludges 0 0 NA NA 0.0 0.00 0.000

2. Dry Solid Waste (A) Dry Active Waste (compacted) 0 0 NA NA 0.0 0.00 0.000 (B) Dry Active Waste (non-compacted) 0 0 NA NA 0.0 0.00 0.000 (C) Dry Active Waste (brokered) NA NA NA NA 17,263.9 488.91 1.513 (D) Irradiated Components 0 0 NA NA 0.0 0.00 0.000

3. All Solid Waste 7 7 NA NA 18,277.0 517.60 537.739

Does not included brokered Dry Active Waste totals.

CATAWBA NUCLEAR STATION - SOLID RADIOACTIVE WASTE

SUMMARY

OF PRINCIPAL RADIONUCLIDE COMPOSITION REPORT PERIOD 1/1/2011 TO 12/31/2011 Type of Waste Shipped Radionuclide % Abundance*

1. Waste from Liquid Systems (A) Dewatered Secondary Resins (None shipped to a 10CFR61 facility this period)

(B) Dewatered Primary Resins H-3 0.1%

Cr-51 0.0%

Mn-54 0.8%

Co-57 0.1%

Co-58 0.0%

Fe-59 0.0%

Co-60 6.2%

Zn-65 0.1%

Nb-94 0.0%

Nb-95 0.0%

Zr-95 0.0%

Ag-108m 0.0%

Ag-11Om 0.0%

Sn-113 0.0%

Sb-122 0.0%

Sb-124 0.0%

Sb-125 0.2%

Te-125m 0.0%

1-131 0.0%

Ba-1 33 0.0%

Cs-1 34 0.0%

Cs-1 37 0.1%

W-187 0.0%

Ba/La-140 0.0%

Ce-141 0.0%

Ce-144 0.0%

Pu-238 0.0%

Pu-239 0.0%

C-14 0.1%

Fe-55 12.3%

Ni-59 0.5%

Ni-63 79.5%

Sr-89 0.0%

Sr-90 0.0%

Tc-99 0.0%

1-129 0.0%

Am-241 0.0%

Pu-241 0.0%

Cm-242 0.0%

Cm-243 0.0%

(C) Evaporator Concentrates (None shipped to a 10CFR61 facility this period)

(D) Dewatered Mechanical Filters H-3 0.0%

Cr-51 8.0%

Mn-54 1.8%

Co-57 0.2%

Average percent abundance for all shipments during period. Page 1 of 4

CATAWBA NUCLEAR STATION - SOLID RADIOACTIVE WASTE

SUMMARY

OF PRINCIPAL RADIONUCLIDE COMPOSITION REPORT PERIOD 1/1/2011 TO 12/31/2011 Type of Waste Shipped Radionuclide %Abundance*

Co-58 64.8%

Fe-59 0.4%

Co-60 4.1%

Zn-65 0.1%

Nb-94 0.0%

Nb-95 9.9%

Zr-95 2.5%

Ag-1 08m 0.0%

Ag- 110m 0.0%

Sn-113 0.2%

Sb-1 22 0.0%

Sb-124 0.1%

Sb-125 0.1%

Te-125m 0.0%

1-131 0.0%

Ba-1 33 0.0%

Cs-134 0.0%

Cs-137 0.0%

W-187 0.0%

Ba/La-140 0.0%

Ce-141 0.0%

Ce-144 0.1%

Pu-238 0.0%

Pu-239 0.0%

C-14 0.7%

Fe-55 5.5%

Ni-59 0.0%

Ni-63 1.5%

Sr-89 0.0%

Sr-90 0.0%

Tc-99 0.0%

1-129 0.0%

Am-241 0.0%

Pu-241 0.0%

Cm-242 0.0%

Cm-243 0.0%

(E) Dewatered Demineralizers (None shipped to a 10CFR61 facility this period)

(F) Solidified (Cement) Acids, Oils, Sludges (None shipped to a 10CFR61 facility this period)

2. Dry Solid Waste (A) Dry Active Waste (compacted) (None shipped to a 10CFR61 facility this period)

(B) Dry Active Waste (non-compacted) (None shipped to a 10CFR61 facility this period)

(C) Dry Active Waste (brokered) H-3 0.4%

Cr-51 1.7%

Mn-54 2.1%

Co-57 0.2%

Average percent abundance for all shipments during period. Page 2 of 4

CATAWBA NUCLEAR STATION - SOLID RADIOACTIVE WASTE

SUMMARY

OF PRINCIPAL RADIONUCLIDE COMPOSITION REPORT PERIOD 1/1/2011 TO 12/31/2011 Type of Waste Shipped Radionuclide % Abundance

Co-58 46.5%

Fe-59 0.0%

Co-60 12.5%

Zn-65 0.2%

Nb-94 0.0%

Nb-95 0.5%

Zr-95 0.3%

Ag-108m 0.0%

Ag-110m 0.0%

Sn-113 0.0%

Sb-1 22 0.0%

Sb-124 0.0%

Sb-125 0.7%

Te-125m 0.0%

1-131 0.0%

Ba-133 0.0%

Cs-134 0.0%

Cs-137 1.6%

W-187 0.0%

Ba/La- 140 0.0%

Ce-141 0.0%

Ce-144 0.2%

Pu-238 0.0%

Pu-239 0.0%

C-14 0.2%

Fe-55 21.7%

Ni-59 0.0%

Ni-63 11.1%

Sr-89 0.3%

Sr-90 0.0%

Tc-99 0.0%

1-129 0.0%

Am-241 0.0%

Pu-241 0.0%

Cm-242 0.0%

Cm-243 0.0%

(D) Irradiated Components (None shipped to a 10CFR61 facility this period)

3. All Solid Waste H-3 0.1%

Cr-51 0.6%

Mn-54 0.9%

Co-57 0.1%

Co-58 4.8%

Fe-59 0.0%

Co-60 6.1%

Zn-65 0.1%

Nb-94 0.0%

Nb-95 0.7%

Zr-95 0.2%

Ag-108m 0.0%

Average percent abundance for all shipments during period. Page 3 of 4

CATAWBA NUCLEAR STATION - SOLID RADIOACTIVE WASTE

SUMMARY

OF PRINCIPAL RADIONUCLIDE COMPOSITION REPORT PERIOD 1/112011 TO 12/31/2011 Type of Waste Shipped Radionuclide %Abundance*

Ag-110m 0.0%

Sn-113 0.0%

Sb-122 0.0%

Sb-124 0.0%

Sb-125 0.2%

Te-125m 0.0%

1-131 0.0%

Ba-133 0.0%

Cs-134 0.0%

Cs-137 0.1%

W-187 0.0%

Ba/La-140 0.0%

Ce-141 0.0%

Ce-144 0.0%

Pu-238 0.0%

Pu-239 0.0%

C-14 0.1%

Fe-55 11.8%

Ni-59 0.5%

Ni-63 73.8%

Sr-89 0.0%

Sr-90 0.0%

Tc-99 0.0%

1-129 0.0%

Am-241 0.0%

Pu-241 0.0%

Cm-242 0.0%

Cm-243 0.0%

Average percent abundance for all shipments during period. Page 4 of 4

ATTACHMENT 4 Meteorological Data Meteorological Joint Frequency Distributions of Wind Speed, Wind Direction and Atmospheric Stability using winds at the 10 M Level (Hours of Occurrence)

The SAS System -/i* Qot SECTOR N NNE NE ENE E ESE SE SSE S SSW SW WSW W WNW NW NNW No. No. No. No. No. No. No. No. No. No. No. No. No. No. No. No.

STAB WSCLS (mis)

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 0 0 0 0 0 0 1 0 0 1 0 1 0 1.01-1.25 0 0 0 0 0 1 0 0 1 1 4 1 0 0 1 0 1.26-1.50 3 2 1 0 1 1 8 3 4 3 5 2 4 2 3 1 1.51-2.00 3 3 6 2 6 6 22 44 31 33 46 31 15 8 2 14 2.01-3.00 55 24 6 6 4 5 20 55 60 133 121 53 38 29 10 13 3.01-4.00 59 28 7 5 1 4 5 5 21 69 37 13 13 11 21 17 4.01-5.00 25 29 20 1 0 0 3 2 6 26 13 0 2 8 13 15 5.01-6.00 4 2 3 0 0 0 0 0 2 6 9 2 0 5 4 6 6.01-8.00 0 0 0 0 0 0 0 0 0 1 0 0 0 0 12 8 8.01-10.00 0 0 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 0 0 0 0 2 0 0 0 0 0 0 0 1.01-1.25 0 0 0 0 0 0 0 1 0 0 0 1 1 0 0 0 1.26-1.50 2 0 1 1 0 1 1 3 2 3 1 2 4 1 2 1 1.51-2.00 7 1 1 0 2 0 3 15 6 6 8 4 3 1 0 2 2.01-3.00 16 9 3 4 1 2 3 28 17 23 9 7 8 9 5 8 3.01-4.00 20 9 11 2 0 1 4 1 4 16 4 2 1 6 3 3 4.01-5.00 11 13 6 0 0 0 0 0 1 6 1 2 0 2 2 1 5.01-

6.00 0 6 3 0 0 0 0 0 0 1 4 0 0 0 3 2 6.01-8.00 1 1 0 0 0 0 0 0 0 1 1 0 0 1 4 0 8.01-10.00 0 0 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 1 0 0 1 0 0 0 1.01-1.25 0 0 0 0 0 0 1 1 1 0 0 0 0 1 1 0 1.28-1.50 1 0 1 0 0 0 1 4 2 5 2 2 2 1 1 2 1.51-2.00 8 0 2 0 1 2 5 9 11 11 12 5 5 2 2 1 2.01-3.00 20 10 3 4 0 1 11 11 11 25 10 3 4 2 0 8 3.01-4.00 25 13 6 1 0 1 5 2 8 21 7 1 1 2 6 5 4.01-5.00 12 30 10 1 1 0 2 1 2 6 1 1 0 3 2 0 5.01-6.00 0 4 2 0 0 0 0 0 2 3 1 0 0 1 2 2 6.01-8.00 1 2 0 0 0 0 0 0 0 0 0 0 0 0 5 0 8.01-10.00 0 1 0 0 0 0 0 0 0 0 0 0 0 0 0 0 D 0.46-0.75 0 0 0 0 0 0 0 2 1 0 1 4 2 0 0 0 0.76-1.00 1 0 1 0 0 3 2 5 4 8 7 1 5 5 2 1 1.01-1.25 2 0 0 2 0 3 7 6 6 9 11 13 6 4 4 4 1.26-1.50 3 2 1 4 3 2 9 13 31 25 18 11 11 9 16 9 1.51-2.00 26 8 4 7 8 8 9 22 49 52 28 25 13 16 14 22 2.01-3.00 91 26 15 8 9 6 21 53 113 76 61 15 17 13 19 55 3.01-4.00 138 121 27 13 4 3 18 22 42 42 23 3 8 10 9 33 4.01-5.00 107 140 50 5 4 1 13 11 20 20 6 5 1 7 6 19 5.01-6.00 11 20 19 3 0 0 6 4 5 5 5 0 0 2 4 21 6.01-8.00 4 5 4 0 0 0 0 0 0 0 2 0 2 0 1 6

8.01-10.00 1 1 0 0 0 0 0 0 0 0 0 0 0 0 0 0 E 0.46-0.75 0 1 0 0 0 0 0 2 1 5 2 3 0 1 3 0 0.76- 1 1.00 1 1 0 0 1 1 2 9 18 18 32 18 6 9 1 2 1.01-1.25 1 0 0 2 2 1 6 17 31 35 37 19 21 15 14 4 1.26-1.50 3 1 4 0 2 2 9 24 63 88 34 20 17 21 12 8 1.51-2.00 13 2 2 1 0 1 14 59 128 81 57 24 41 35 50 48 2.01-3.00 64 5 0 1 3 9 42 62 140 74 33 13 17 40 56 109 3.01-4.00 51 5 2 3 2 3 11 25 21 19 3 1 3 13 13 39 4.01-5.00 7 3 11 3 4 4 17 4 16 4 1 0 0 2 5 4 5.01-6.00 2 4 7 0 1 0 2 0 2 4 0 0 0 0 1 0 6.01-8.00 0 0 2 0 1 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 F 0.46-0.75 0 0 0 0 0 0 3 3 5 5 5 6 1 1 1 1 0.76-1.00 0 0 0 1 0 0 0 9 23 14 24 23 11 11 8 0 1.01-1.25 0 0 0 0 0 0 1 13 29 31 30 21 17 11 10 6 1.26-1.50 0 0 0 0 0 1 2 20 50 50 28 22 14 19 21 23 1.51-2.00 5 1 0 0 0 0 0 20 41 33 11 15 29 22 25 53 2.01-3.00 22 0 0 1 0 0 1 7 13 3 0 4 4 12 15 70 3.01-4.00 8 0 0 0 0 0 0 1 0 0 0 0.0 1 0 6 4.01-5.00 0 1 1 0 0 0 2 1 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 G 0.46-

0.75 0 o o o 1 o o 2 3 8 10 4 11 4 9 0.76-1.00 0 o 0 o o o o 4 26 28 34 18 19 11 19 1.01-1.25 0 o o o o o o 5 38 33 34 14 11 11 20 1.26-1.50 1 o 0 o o o o 7 29 23 40 18 8 16 16 1.51-2.00 3 o o o o o o 7 11 10 14 16 11 10 16 2.01-3.00 18 o 0 o o o o 1 0 0 3 1 2 4 4 3.01-4.00 8 o o o o o o 0 0 0 0 0 0 0 0 4.01-5.00 0 1 0 o o o o 0 0 0 0 0 0 0 0 5.01-6.00 0 o o o o o 1 0 0 0 0 0 0 0 0 6.01-8.00 1 o o o o o o 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

ATTACHMENT 5 Unplanned Offsite Releases

Catawba Nuclear Station Summary of Unplanned Radioactive Releases for 2011 Date: July 26, 2011 Event: An estimated 2 psig decrease in the in-service Waste Gas (WG) Decay Tank 'D' occurred following maintenance on a WG pressure transmitter. The leak was from a tubing fitting. Maintenance personnel attempted to tighten the fitting but were not successful.

Maintenance isolated the leaking fitting per procedure to stop the leak.

Consequences: No observable increases above background or alarms were generated on the Auxiliary Building Ventilation Noble Gas Monitor (OEMF41) and either Unit Vent Noble Gas Monitors (1/2EMF36). No observed increase in Monthly Unit Vent (Unknown) Release Activity or Effluent Dose observed for July 2011.

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 ten miles of Catawba 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 1/1/11 TO 1/1/12 GASEOUS ANNUAL DOSE

SUMMARY

REPORT Catawba Nuclear Station Units 1 & 2 1st Quarter 2011 IODINE, H3, AND PARTICULATE DOSE LIMIT ANALYSIS======= Quarter 1 2011 Critical Critical Dose Limit Max % of Period-Limit Group Organ (mrem) (mrem) Limit Q1 - Maximum Organ Dose CHILD BONE 1.28E+00 1.50E+01 8.51E+00 Maximum Organ Dose Receptor Location: 0.5 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 1 2011 Dose Limit % of Period-Limit (mrad) (mrad) Limit Ql - Maximum Gamma Air Dose 8.26E-03 1.00E+01 8.26E-02 Maximum Gamma Air Dose Receptor Location: 0.5 Mile NNE Major Isotopic Contributors (5% or greater to total)

Nuclide Percentage AR-41 9.86E+01 Ql - Maximum Beta Air Dose 3.16E-03 2.00E+01 1.58E-02 Maximum Beta Air Dose Receptor Location: 0.5 Mile NNE Major Isotopic Contributors (5% or greater to total)

Nuclide Percentage AR-41 9.10E+01 XE-133 7.79E+00

EFFLUENT AND WASTE DISPOSAL ANNUAL REPORT PERIOD 1/1/11 TO 1/1/12 GASEOUS ANNUAL DOSE

SUMMARY

REPORT Catawba Nuclear Station Units 1 & 2 2 nd Quarter 2011 IODINE, H3, AND PARTICULATE DOSE LIMIT ANALYSIS======= Quarter 2 2011 Critical Critical Dose Limit Max % of Period-Limit Group Organ (mrem) (mrem) Limit Q2 - Maximum Organ Dose CHILD BONE 9.42E-01 1.50E+01 6.28E+00 Maximum Organ Dose Receptor Location: 0.5 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 2 2011 Dose Limit % of Period-Limit (mrad) (mrad) Limit Q2 - Maximum Gamma Air Dose 5.76E-03 1.OOE+01 5.76E-02 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 Q2 - Maximum Beta Air Dose 2.13E-03 2.OOE+01 1.06E-02 Maximum Beta Air Dose Receptor Location: 0.5 Mile NNE Major Isotopic Contributors (5% or greater to total)

Nuclide Percentage AR-41 9.49E+01

EFFLUENT AND WASTE DISPOSAL ANNUAL REPORT PERIOD 1/1/11 TO 1/1/12 GASEOUS ANNUAL DOSE

SUMMARY

REPORT Catawba Nuclear Station Units 1 & 2 3 rd Quarter 2011 IODINE, H3, AND PARTICULATE DOSE LIMIT ANALYSIS======= Quarter 3 2011 Critical Critical Dose Limit Max % of Period-Limit Group Organ (mrem) (mrem) Limit Q3 - Maximum Organ Dose CHILD BONE 1.28E+00 1.50E+01 8.53E+00 Maximum Organ Dose Receptor Location: 0.5 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 3 2011 Dose Limit % of Period-Limit (mrad) (mrad) Limit Q3 - Maximum Gamma Air Dose 7.23E-03 1.OOE+01 7.23E-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 Q3 - Maximum Beta Air Dose 2.62E-03 2.00E+01 1.31E-02 Maximum Beta Air Dose Receptor Location: 0.5 Mile NNE Major Isotopic Contributors (5% or greater to total)

Nuclide Percentage AR-41 9.71E+01

EFFLUENT AND WASTE DISPOSAL ANNUAL REPORT PERIOD 1/1/11 TO 1/1/12 GASEOUS ANNUAL DOSE

SUMMARY

REPORT Catawba Nuclear Station Units 1 & 2 4 th Quarter 2011 IODINE, H3, AND PARTICULATE DOSE LIMIT ANALYSIS======= Quarter 4 2011 Critical Critical Dose Limit Max % of Period-Limit Group Organ (mrem) (mrem) Limit Q4 - Maximum Organ Dose CHILD BONE 1.25E+00 1.50E+01 8.32E+00 Maximum Organ Dose Receptor Location: 0.5 Mile NE Critical Pathway: Vegetation Major Isotopic Contributors (5% or greater to total)

Nuclide Percentage C-14 1.OOE+02

=== NOBLE GAS DOSE LIMIT AN]*LYSIS= ----------- Quarter 4 2011 Dose Limit % of Period-Limit (mrad) (mrad) Limit Q4 - Maximum Gamma Air Dose 7.70E-03 1.OOE+01 7.70E-02 Maximum Gamma Air Dose Receptor Location: 0.5 Mile NNE Major Isotopic Contributors (5% or greater to total)

Nuclide Percentage AR-41 9.94E+01 Q4 - Maximum Beta Air Dose 2.82E-03 2.OOE+01 1.41E-02 Maximum Beta Air Dose Receptor Location: 0.5 Mile NNE Major Isotopic Contributors (5% or greater to total)

Nuclide Percentage AR-41 9.56E+01

EFFLUENT AND WASTE DISPOSAL ANNUAL REPORT PERIOD 1/1/11 TO 1/1/12 GASEOUS ANNUAL DOSE

SUMMARY

REPORT Catawba Nuclear Station Units 1 & 2 ANNUAL 2011 IODINE, H3, AND PARTICULATE DOSE LIMIT ANALYSIS======= Annual 2011 Critical Critical Dose Limit Max % of Period-Limit Group Organ (mrem) (mrem) Limit Yr - Maximum Organ Dose CHILD BONE 4.75E+00 3.OOE+01 1.58E+01 Maximum Organ Dose Receptor Location: 0.5 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===--- Annual 2011----

Dose Limit % of Period-Limit (mrad) (mrad) Limit Yr - Maximum Gamma Air Dose 2.90E-02 2.00E+01 1.45E-01 Maximum Gamma Air Dose Receptor Location: 0.5 Mile NNE Major Isotopic Contributors (5% or greater to total)

Nuclide Percentage AR-41 9.92E+01 Yr - Maximum Beta Air Dose 1.07E-02 4.OOE+01 2.68E-02 Maximum Beta Air Dose Receptor Location: 0.5 Mile NNE Major Isotopic Contributors (5% or greater to total)

Nuclide Percentage AR-41 9.45E+01 XE-133 5.01E+00

EFFLUENT AND WASTE DISPOSAL ANNUAL REPORT PERIOD 1/1/11 TO 1/1/12 LIQUID ANNUAL DOSE

SUMMARY

REPORT Catawba Nuclear Station Units 1 & 2 1 st Quarter 2011

BATCH LIQUID RELEASES Quarter 1 2011 ===

Critical Critical Dose Limit Max % of Period-Limit Age Organ (mrem) (mrem) Limit Q1 - Maximum Organ Dose CHILD LIVER 1.77E-02 1.00E+01 1.77E-01 Qi - Total Body Dose CHILD I 76r-02 3.OOE+00 5.86E-01 Maximum Organ Critical Pathway: Potable Water Major Isotopic Contributors (5% or greater to total)

Nuclide Percentage H-3 9.81E+01 Total Body Critical Pathway: Potable Water Major Isotopic Contributors (5% or greater to total)

Nuclide Percentage H-3 9.87E+01

=== CONTINUOUS LIQUID RELEASES (WC)--- Quarter 1 2011 ......

Critical Critical Dose Limit Max % of Period-Limit Age Organ (mrem) (mrem) Limit Ql - Maximum Organ Dose CHILD LIVER 3.82E-04 1.00E+01 3.82E-03 Ql - Total Body Dose CHILD 3.82E-04 3.00E+00 1.27E-02 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 1.00OE+02

EFFLUENT AND WASTE DISPOSAL ANNUAL REPORT PERIOD 1/1/11 TO 1/1/12 LIQUID ANNUAL DOSE

SUMMARY

REPORT Catawba Nuclear Station Units 1 & 2 2 nd Quarter 2011

=== BATCH LIQUID RELEASES Quarter 2 2011 =.....

Critical Critical Dose Limit Max % of Period-Limit Age Organ (mrem) (mrem) Limit Q2 - Maximum Organ Dose ADULT GI-LLI 6.33E-02 1.OOE+01 6.33E-01 Q2 - Total Body Dose CHILD 2.82E-02 3.OOE+00 9.40E-01 Maximum Organ Critical Pathway: Fresh Water Fish Major Isotopic Contributors (5% or greater to total)

Nuclide Percentage NB-95 5.30E+01 H-3 3.31E+01 CO-60 9. 73E+00 Total Body Critical Pathway: Potable Water Major Isotopic Contributors (5% or greater to total)

Nuclide Percentage H-3 8. 95E+01 CO-60 7. 13E+00 CONTINUOUS LIQUID RELEASES (WC) Quarter 2 2011 -- ====

Critical Critical Dose Limit Max % of Period-Limit Age Organ (mrem) (mrem) Limit Q2 - Maximum Organ Dose CHILD LIVER 3.71E-03 1.OOE+01 3. 71E-02 Q2 - Total Body Dose CHILD 3.71E-03 3.OOE+00 1. 24E-01 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 1/1/11 TO 1/1/12 LIQUID ANNUAL DOSE

SUMMARY

REPORT Catawba Nuclear Station Units 1 & 2 3td Quarter 2011

= BATCH LIQUID RELEASES Quarter 3 2011 -

Critical Critical Dose Limit Max % of Period-Limit Age Organ (orem) (mrem) Limit Q3 - Maximum Organ Dose ADULT GI-LLI 2. 46E-02 1.OOE+01 2.46E-01 Q3 - Total Body Dose CHILD 1. 45E-02 3.OOE+00 4.84E-01 Maximum Organ Critical Pathway: Fresh Water Fish Major Isotopic Contributors (5% or greater to total)

Nuclide Percentage H-3 4. 60E+01 NB-95 4.13E+01 CO-60 1.01E+01 Total Body Critical Pathway: Potable Water Major Isotopic Contributors (5% or greater to total)

Nuclide Percentage H-3 9.38E+01 CO-60 5.59E+00 CONTINUOUS LIQUID RELEASES (WC) Quarter 3 2011 =-===.

Critical Critical Dose Limit Max % of Period-Limit Age Organ (mrem) (mrem) Limit Q3 - Maximum Organ Dose CHILD LIVER 1.97E-04 1.OOE+01 1.97E-03 Q3 - Total Body Dose CHILD 1.97E-04 3.OOE+00 6.57E-03 Maximum Organ Critical Pathway: Potable Water Major Isotopic Contributors (5% or greater to total)

Nuclide Percentage m-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 1/1/11 TO 1/1/12 LIQUID ANNUAL DOSE

SUMMARY

REPORT Catawba Nuclear Station Units 1 & 2 4 th Quarter 2011

BATCH LIQUID RELEASES Quarter 4 2011 ===

Critical Critical Dose Limit Max % of Period-Limit Age Organ (mrem) (trem) Limit Q4 - Maximum Organ Dose CHILD GI-LLI 1.03E-01 1.OOE+01 1.03E+00 Q4 - Total Body Dose CHILD 9.88E-02 3.OOE+00 3.29E+00 Maximum Organ Critical Pathway: Potable Water Major Isotopic Contributors (5% or greater to total)

Nuclide Percentage H-3 9.46E+01 Total Body Critical Pathway: Potable Water Major Isotopic Contributors (5% or greater to total)

Nuclide Percentage H-3 9.83E+01

=== CONTINUOUS LIQUID RELEASES (WC) Quarter 4 2011 ---

Critical Critical Dose Limit Max % of Period-Limit Age Organ (mrem) (mrem) Limit

-gan Dose CHILD LIVER 4.04E-04 Q4 - Maximum Or

Dose CHILD 4.04E-04 1.OOE+01 4. 04E-03 Q4 - Total Body 3.OOE+00 1. 35E-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 1/1/11 TO 1/1/12 LIQUID ANNUAL DOSE

SUMMARY

REPORT Catawba Nuclear Station Units 1 & 2 ANNUAL 2011

=== BATCH LIQUID RELEASES Annual 2011-Critical Critical Dose Limit Max % of Period-Limit Age Organ (mrem) (mrem) Limit Yr - Maximum Organ Dose ADULT GI-LLI 1.89E-01 2.OOE+01 9.46E-01 Yr - Total Body Dose CHILD 1.42E-01 6.OOE+00 2.36E+00 Maximum Organ Critical Pathway: Fresh Water Fish Major Isotopic Contributors (5% or greater to total)

Nuclide Percentage H-3 5.96E+01 NB-95 3.08E+01 CO-60 7.OOE+00 Total Body Critical Pathway: Potable Water Major Isotopic Contributors (5% or greater to total)

Nuclide Percentage H-3 9. 60E+01 CONTINUOUS LIQUID RELEASES (WC) Annual 2011 Critical Critical Dose Limit Max % of Period-Limit Age Organ (mrem) (mrem) Limit Yr - Maximum Organ Dose CHILD LIVER 4.71E-03 2.OOE+01 2.36E-02 Yr - Total Body Dose CHILD 4.71E-03 6.OOE+00 7.85E-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

Catawba Nuclear Station 2011 Radioactive Effluent and ISFSI 40CFR190 Uranium Fuel Cycle Dose Calculation Results In accordance with the requirements of 40CFR 190, 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 Catawba Nuclear Station only includes liquid and gaseous effluent dose contributions from Catawba and direct and air-scatter dose from Catawba's onsite Independent Spent Fuel Storage Installation (ISFSI) since no other uranium fuel cycle facility contributes significantly to Catawba's maximum exposed individual. Included in the gaseous effluent dose calculations is an estimate of the dose contributed by Carbon-14 (Ref. "Carbon-14 SupplementalInformation", contained in the ARERR for further information). The combined dose to a maximum exposed individual from Catawba's effluent releases and direct and air-scatter dose from Catawba's ISFSI is below 40CFRI 90 limits as shown by the following summary:

I. 2011 Catawba 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 = 2.18E+00 mrem Maximum Location: 0.5 Mile, Northeast Sector Critical Age: Child Gas non-NG Contribution: 92.4%

Gas NG Contribution: 1.1%

Liquid Contribution: 6.5%

Maximum Organ (other than TB) Dose = 4.75E+00 mrem Maximum Location: 0.5 Mile, Northeast Sector Critical Age: Child Critical Organ: Bone Gas Contribution: 99.8%

Liquid Contribution: 0.2%

II. 2011 Catawba 40CFR190 ISFSI Dose Summary Direct and air-scatter radiation dose contributions from the onsite Independent Spent Fuel Storage Installation (ISFSI) at Catawba have been calculated and documented in the "Catawba Nuclear Station, ISFSI, 10CFR72.212 Evaluation" report. The maximum dose rate to the nearest resident from the Catawba ISFSI is conservatively calculated to be 16.6 mrem/year.

The attached excerpt from the "Catawba Nuclear Station, ISFSI, 10CFR72.212 Evaluation" report is provided to document the method used to calculate the Catawba ISFSI 16.6 mrem/year dose estimate.

The following three pages are excerpted from the "Catawba Nuclear Station, Independent Spent-Fuel Storage Installation, 10CFR72.212 Evaluation, NAC-UMS Universal Storage System" report.

CNS 10CFR72.212 Evaluation NAC-UMS Universal Storage System Page 31 of 56 7.3 10CFR72.212(b)(2)(i)(C) - Requirements of §72.104

"(C) the requirementsof§72.104 have been met. A copy of this record must be retaineduntil spentfuel is no longer stored under the general license issued under §72.210."

The requirements of §72.104 are as follows:

(a) During normal operations and anticipated occurrences, the annual dose equivalent to any real individual who is located beyond the controlled area must not exceed 0.25 mSv (25 mrerm) to the whole body, 0.75 mSv (75 mrem) to the thyroid and 0.25 mSv (25 mrem) to any other critical organ as a result of exposure to:

(1) Planned discharges of radioactive materials, radon and its decay products excepted, to the general environment, (2) Direct radiation from ISFSI or MRS operations, and (3) Any other radiation from uranium fuel cycle operations within the region.

Doses from 24 loaded storage casks located at the ISFSI have been calculated. This represents the placement of a loaded canister at all available locations on the current ISFSI storage pad, completing the projected loading for Phase I.

The methodology and results of the dose calculations are discussed in detail in References 7.3-3 and 7.3-4. A summary of the methodology and results is presented below.

There are two calculations used to estimate the impact of the ISFSI direct radiation doses. The first calculation (Reference 7.3-3) determines a fuel assembly source term to be used in the subsequent shielding model. In order to bound fuel assemblies loaded into canisters in the past and projected to be loaded in the future, the same, bounding fuel assembly is modeled for all 24 spaces in each of the 24 casks. The source term was developed to bound all types of LEU fuel at Catawba (Westinghouse OFA, RFA, and Mk-BW). Axial flux profiles, fuel hardware activation, component activation, and the potential impacts from burnable poisons were modeled. Both gamma and neutron source spectrums were produced. In order to ensure that the gamma flux was conservative, the model includes the impact from activation of components, fuel hardware, and light elements. Thus, each spent fuel location models a bounding fuel assembly with a bounding activated component (thimble plug).

CNS 10CFR72.212 Evaluation NAC-UMS Universal Storage System Page 32 of 56 The source term was modeled using the SAS2H coupled shielding and depletion analysis module of the SCALE code suite. This module utilizes the ORIGEN-S point depletion code to compute the source spectra. An appropriate 44 group library was employed. Use of this code is a standard industry application for source term depletion and decay calculations. It is utilized in a manner consistent with its development.

The results from the source term calculation (Reference 7.3-3) are used as the source term spectra input to the shielding model (Reference 7.3-4).

MCNP, a Monte-Carlo code for neutron and photon transport, was utilized for the shielding computations. This code is an industry standard and is typically applied to problems of this type. The fuel related source term was normalized to the 20 kW administrative decay heat limit and the component source term was normalized to an eight-year decay duration.

The MCNP models were set up using the source terms developed for the four source regions in Reference 7.3-3: fuel (neutron and gamma), fuel hardware, upper plenum, and upper nozzle. These source regions include contributions from both the fuel assembly and the component, as physically appropriate.

The same mesh tally scheme was applied to each source case so the results for each source term could then be summed to produce the final result. A detailed cask model was developed (the work was performed by the cask vendor) and replicated ina 2 by 12 array mimicking the planned arrangement of the loaded canisters on the Catawba ISFSI pad. This represents a full pad of loaded canisters.

Detector locations were laid out on a grid in three dimensions and plots for both near and far field doses were obtained. However, because the coordinate axes align with the cask array orientation, the highest doses are seen along the axes. Thus, for a given distance the highest dose will be found along the x axis, as the long part of the array defines the y axis. The results are as expected for the near and far field doses. Conservatively, the coordinate system was eschewed in the evaluation of the results. for 72.104 purposes in favor or the straight line distance to the limiting receptor location (nearest real individual). The nearest real individual is over 450 meters from the ISFSI, but a conservative evaluation distance of 405 meters is adopted. This distance from the ISFSI is within the site boundary. No real individual can live within the (site controlled) boundary, so this distance (location) bounds any real individual (living off-site). As shown in Table 6.7-5 of Reference 7.3-4, the annual dose to the nearest real individual from.a full 2 by. 12 array of loaded canisters with limiting 20 kW fuel sources and inserts decayed for eight years is 16.6 mrem/yr. The maximum dose at this distance is found, as expected, along

CNS I OCFR72.212 Evaluation NAC-UMS Universal Storage System Page 33 of 56 the x axis. This is a conservative application of the shielding analysis results.

The shielding analysis contains many receptor locations, and the results from these cases could be used with a plot of the location of the nearest individual on the shielding model XY coordinates system. This would produce a more precise and lower result.

The computed direct shine dose from the ISFSI to the nearest individual will be added to the plant generated dose to show compliance with 72.104.

General Office Radiation Protection has responsibility for this function.

This information is submitted to the NRC as part of the Annual Radioactive Effluent Release Report for Catawba.

(b) Operational restrictions must be established to meet as low as is reasonably achievable objectives for radioactive materials in effluents and direct radiation levels associated with ISFSI or MRS operations.

(c) Operational limits must be established for radioactive materials in effluents and direct radiation levels associated with ISFSI or MRS operations to meet the limits given in paragraph (a) of this section.

The requirements are met through implementation of the CNS Radiation Protection Program (References 7.3-1 and 7.3-2).

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

There were revisions to the Catawba Nuclear Station Updated Final Safety Analysis Report, Section 16.11, Radiological Controls, in 2011:

Section 16.11-7, "Radioactive Gaseous Effluent Monitoring Instrumentation," was changed on 10/24/11.

Section 16.11-16, "Annual Radiological Environmental Operating Report," was changed on 10/24/11.

As per TS 5.5.5.b. Licensee initiated changes to the Radiological Effluent Controls of the UFSAR, Catawba is attaching the entire Section 16.11 of the UFSAR and the List of Effective Sections which will demonstrate when each section was revised.

LIST OF EFFECTIVE SECTIONS SECTION REVISION NUMBER REVISION DATE TABLE OF CONTENTS 13 03/05/12 16.1 1 08/27/08 16.2 2 08/21/09 16.3 1 08/21/09 16.5-1 2 02/20/12 16.5-2 Deleted 16.5-3 1 02/20/04 16.5-4 0 10/09/02 16.5-5 1 01/28/10 16.5-6 1 08/21/09 16.5-7 0 10/09/02 16.5-8 2 12/22/08 16.5-9 1 02/20/12 16.5-10 Deleted 16.6-1 0 10/09/02 16.6-2 Deleted 16.6-3 1 08/21/09 16.6-4 1 08/21/09 16.6-5 1 08/21/09 16.7-1 1 08/21/09 16.7-2 4 02/03/11 16.7-3 2 02/03/11 16.7-4 2 08/21/09 16.7-5 2 08/21/09 Catawba Units 1 and 2 Page 1 Revision 50

LIST OF EFFECTIVE SECTIONS SECTION REVISION NUMBER REVISION DATE 16.7-6 2 08/21/09 16.7-7 1 08/21/09 16.7-8 2 08/21/09 16.7-9 7 04/11/12 16.7-10 4 10/24/11 16.7-11 1 08/21/09 16.7-12 1 08/21/09 16.7-13 2 08/21/09 16.7-14 1 08/21/09 16.7-15 1 08/21/09 16.7-16 0 06/08/09 16.8-1 3 08/21/09 16.8-2 2 02/20/12 16.8-3 1 10/24/06 16.8-4 2 11/05/07 16.8-5 3 08/21/09 16.9-1 7 10/24/11 16.9-2 5 10/24/11 16.9-3 3 02/03/11 16.9-4 3 08/21/09 16.9-5 6 06/23/10 16.9-6 7 08/21/09 16.9-7 4 08/21/09 16.9-8 5 08/21/09 Catawba Units 1 and 2 Page 2 Revision 50

LIST OF EFFECTIVE SECTIONS SECTION REVISION NUMBER REVISION DATE 16.9-9 3 08/21/09 16.9-10 5 08/21/09 16.9-11 3 08/21/09 16.9-12 2 08/21/09 16.9-13 3 08/21/09 16.9-14 1 09/25/06 16.9-15 2 08/21/09 16.9-16 2 08/21/09 16.9-17 0 10/09/02 16.9-18 0 10/09/02 16.9-19 3 02/20/12 16.9-20 0 10/09/02 16.9-21 0 10/09/02 16.9-22 1 08/21/09 16.9-23 4 10/24/11 16.9-24 2 10/24/06 16.9-25 2 08/21/09 16.9-26 0 03/05/12 16.10-1 1 08/21/09 16.10-2 1 10/24/06 16.10-3 1 08/21/09 16.11-1 0 10/09/02 16.11-2 3 06/23/10 16.11-3 0 10/09/02 Catawba Units 1 and 2 Page 3 Revision 50

LIST OF EFFECTIVE SECTIONS SECTION REVISION NUMBER REVISION DATE 16.11-4 1 08/21/09 16.11-5 0 10/09/02 16.11-6 1 06/08/09 16.11-7 6 10/24/11 16.11-8 0 10/09/02 16.11-9 0 10/09/02 16.11-10 1 08/21/09 16.11-11 1 03/20/03 16.11-12 0 10/09/02 16.11-13 0 10/09/02 16.11-14 0 10/09/02 16.11-15 0 10/09/02 16.11-16 1 10/24/11 16.11-17 0 10/09/02 16.11-18 1 08/21/09 16.11-19 0 10/09/02 16.11-20 1 08/21/09 16.11-21 0 10/09/02 16.12-1 0 10/09/02 16.13-1 0 10/09/02 16.13-2 Deleted 16.13-3 Deleted 16.13-4 0 10/09/02 Catawba Units 1 and 2 Page 4 Revision 50

Liquid Effluents 16.11-1 16.11 RADIOLOGICAL EFFLUENTS CONTROLS 16.11-1 Liquid Effluents COMMITMENT: The concentration of radioactive material released in liquid effluents to UNRESTRICTED AREAS (see Figure 16.11-16-1 in SLC 16.11-16) shall be limited to:

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 Table to Table 16.11-1-1. 16.11-1-1 Catawba Units 1 and 2 16.11-1-1 Revision 0

Liquid Effluents 16.11-1 Table 16.11-1-1 Radioactive Liquid Waste Sampling and Analysis Program (page 1 of 3)

LIQUID SAMPLING MINIMUM TYPE OF LOWER RELEASE TYPE FREQUENCY ANALYSIS ACTIVITY LIMIT OF FREQUENCY ANALYSIS DETECTION (LLD)(1 )

(ICi/ml)

1. Batch Waste Prior to each Prior to each Principal Gamma 5x10-'

Release release release Emitters(3)

Tanks(2) Each Batch Each Batch 1-131 1x10-6 Any tank Prior to each 31 days Dissolved and lx10-5 which release Entrained Gases discharges One Batch/31 (Gamma Emitters) liquid wastes days by either liquid effluent monitor, EMF-49 or EMF-57 Prior to each 31 days H-3 lx10. 5 release Composite(4 )

Each Batch Gross Alpha 1x10.1 Prior to each 92 days Sr-89, Sr-90 5x10 8 release Composite(4 )

Each Batch

2. Continuous Continuous(6) 7 days Principal Gamma 5x10-1 Releases(5) Composite(6) Emitters(3)

Conventional Waste Water Treatment Line 1-131 lx10x-31 days 31 days Dissolved and 1x10-*

Grab Sample Entrained Gases (Gamma Emitters)

Continuous"6 ' 31 days H-3 lxi 0-5 Composite(6 )

Gross Alpha lxi 0-1 Continuous"6 ' 92 days Sr-89, Sr-90 5 xl0-O Composite 6 I Catawba Units 1 and 2 16.11-1-2 Revision 0

Liquid Effluents 16.11-1 Table 16.11-1-1 Radioactive Liquid Waste Sampling and Analysis Program (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 = 4.66s b E. V .2.22 x 10 6

Y" exp (-AAt)

Where:

LLD = the "a priori" lower limit of detection (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 (counts per minute),

E = the counting efficiency (counts per disintegration),

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

2.22 x 106 = the number of disintegrations per minute per microCurie, Y = the fractional radiochemical yield, when applicable, X = the radioactive decay constant for the particular radionuclide (sec 1 ), and At = the elapsed time between midpoint of sample collection and time of counting (sec).

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

(3) The principal gamma emitters for which the LLD specification applies include the following radionuclides:

Catawba Units 1 and 2 16.11-1-3 Revision 0

Liquid Effluents 16.11,1 Table 16.11-1-1 Radioactive Liquid Waste Sampling and Analysis Program (page 3 of 3)

Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, Cs-134, Cs-137, and Ce-141. The LLD for Ce-144 is 5x10-6 gtCi/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 Radioactive Effluent Release Report pursuant to Technical Specification 5.6.3 in the format outlined in Regulatory Guide 1.21, Appendix B, Revision 1, June 1974.

(4) 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 that is representative of the liquids released.

(5) A continuous release is the discharge of liquid wastes of a non-discrete volume, e.g., from a volume of a system that has an input flow during the continuous release.

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

Catawba Units 1 and 2 16.11-1-4 Revision 0

Liquid Effluents 16.11-1 BASES The basic requirements for SLCs concerning effluents from nuclear power reactors are stated in 10 CFR 50.36a. These requirements indicate that compliance with effluent SLCs will keep average annual releases of radioactive material in effluents to small percentages of the limits specified in the old 10 CFR 20.106 (new 10 CFR 20.1302). 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 10 CFR 20.106 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 10 CFR 50.36a 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 10 CFR 50, Appendix I.

As stated in the Introduction to Appendix B of the new 10 CFR 20, the liquid 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 10 CFR 50, 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 has demonstrated that the use of the concentration values associated with the old 10 CFR 20.106 as SLC limits has resulted in calculated maximum individual doses to a MEMBER OF THE PUBLIC that are small percentages of the limits of 10 CFR 50, Appendix I.

Therefore, the use of concentration values which correspond to an annual dose of 500 mrem (ten times the concentration values stated in the new 10 CFR 20, Appendix B, Table 2, Column 2) should not have a negative impact on the ability to continue to operate within the limits of 10 CFR 50, Appendix I and 40 CFR 190.

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 10 CFR 20, 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 10 CFR 20, 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 10 CFR 20, Appendix B, Table II, Column 2, are based, is a factor of 10 higher than annual dose of 50 mrem, upon which the concentrations in the new 10 CFR Catawba Units 1 and 2 16.11-1-5 Revision 0

Liquid Effluents 16.11-1 BASES (continued) 20, Appendix B, Table 2, Column 2, are based. Compliance with the limits of the new 10 CFR 20.1301 will be demonstrated by operating within the limits of 10 CFR 50, Appendix I and 40 CFR 190. The concentration limit for dissolved or entrained noble gases is based upon the assumption that Xe-135 is the controlling radioisotope and its MPC 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 radioactive materials in liquid effluents from all units at the site.

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," Annal. 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. Catawba Offsite Dose Calculation Manual.

2. 10 CFR Part 20, Appendix B.

Catawba Units 1 and 2 16.11-1-6 Revision 0

Radioactive Liquid Effluent Monitoring Instrumentation 16.11-2 16.11 RADIOLOGICAL EFFLUENTS 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: At all times, except when the effluent pathway is isolated and a release to the environment is not possible.

REMEDIAL ACTIONS

NOT E-------------------------------------------

Separate Condition entry is allowed for each Function.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more A.1 Suspend the release of Immediately Radioactive Liquid radioactive liquid effluents Effluent Monitoring monitored by the affected Instrumentation channel(s).

channel(s) Alarm/Trip Setpoint less OR conservative than required. A.2 Declare the channel(s) Immediately non-functional.

B. One or more B.1 Enter the applicable Immediately Radioactive Liquid Conditions and Required Effluent Monitoring Actions specified in Table Instrumentation 16.11-2-1 for the channel(s) non- channel(s).

functional.

(continued)

Catawba Units 1 and 2 16.11-2-1 Revision 3

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 Testing release Requirement 16.11-1-1.

AND C.1.2 Perform independent Prior to initiating a verification of the release discharge line valving.

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

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

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

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

(continued)

Catawba Units 1 and 2 16.11-2-2 Revision 3

Radioactive Liquid Effluent Monitoring Instrumentation 16.11-2 REMEDIAL ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME D. One flow rate D. 1 --------- NOTE-------

measurement device Pump performance curves channel 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 D.2 Restore channel to 30 days FUNCTIONAL status.

E. One channel non- E.1 Perform an analysis of Once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months functional, grab samples for during releases radioactivity at a lower limit when secondary of detection of 10-7 specific activity is >

microCurie/ml. 0.01 microCurie/gm DOSE EQUIVALENT 1-131 AND Once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months during releases when secondary specific activity is <

0.01 microCurie/gm DOSE EQUIVALENT 1-131 AND E.2 Restore channel to 30 days FUNCTIONAL status.

(continued)

Catawba Units 1 and 2 16.11-2-3 Revision 3

Radioactive Liquid Effluent Monitoring Instrumentation 16.11-2 REMEDIAL ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME F. One channel non- F. 1 Collect and analyze grab Once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months functional. samples for principal gamma emitters (listed in Table 16.11-1-1, NOTE 3) at a lower limit of detection 7

of no more than 5x1 0-microCurie/ml.

AND F.2 Restore non-functional 30 days channel to FUNCTIONAL status.

G. Required Action and G.1 Explain why the non- In the next associated Completion functionality was not scheduled Time of Condition C, D, corrected within the Radioactive Effluent E, or F not met. specified Completion Time. Release Report pursuant to Technical Specification 5.6.3 Catawba Units 1 and 2 16.11-2-4 Revision 3

Radioactive Liquid Effluent Monitoring Instrumentation 16.11-2 TESTING REQUIREMENTS

,lIl vr I

------------- I--I-------------------------------------

Refer to Table 16.11-2-1 to determine which TRs apply for each Radioactive Liquid Effluent Monitoring Instrumentation 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. 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 Perform COT. 92 days TR 16.11-2-6 ------------------- NOTE ------------------

For Instrument 1, the COT shall also demonstrate that automatic isolation of this pathway and control room alarm annunciation (for EMF-57, alarm annunciation is in the Monitor Tank Building control room and on the Monitor Tank Building control panel remote annunciator panel) occur if any of the following conditions exist:

a. Instrument indicates measured levels above the Alarm/Trip Setpoint, or

b. Circuit failure/instrument downscale failure (alarm only)

Perform COT. 9 months (continued)

Catawba Units 1 and 2 16.11-2-5 Revision 3

Radioactive Liquid Effluent Monitoring Instrumentation 16.11-2 TESTING REQUIREMENTS (continued)

TEST FREQUENCY TR 16.11-2-7 ------------------- NOTE ------------------

For Instrument 1, the initial CHANNEL CALIBRATION shall be performed using one or more of the reference standards certified by the National Bureau of Standards (NBS) or using standards that have been obtained from suppliers that participate in measurement assurance activities with NBS. 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 CHANNEL CALIBRATION. 18 months Catawba Units 1 and 2 16.11-2-6 Revision 3

Radioactive Liquid Effluent Monitoring Instrumentation 16.11-2 Table 16.11-2-1 Radioactive Liquid Effluent Monitoring Instrumentation INSTRUMENT REQUIRED CONDITIONS TESTING CHANNELS REQUIREMENTS

1. Radioactivity Monitors Providing Alarm and Automatic Termination of Release 1.a Waste Liquid Discharge Monitor 1 per station A, C, G TR 16.11-2-1 (EMF Low Range) TR 16.11-2-3 TR 16.11-2-6 TR 16.11-2-7 1.b Turbine Building Sump Monitor 1 A, E, G TR 16.11-2-1 (EMF-31) TR 16.11-2-4 TR 16.11-2-6 TR 16.11-2-7 1.c Monitor Tank Building Liquid Discharge Monitor 1 per station A, C, G TR 16.11-2-1 (EMF Low Range) TR 16.11-2-3 TR 16.11-2-6 TR 16.11-2-7

2. Continuous Composite Samplers and Sampler Flow Monitor 2.a Conventional Waste Water Treatment Line 1 per station E, G TR 16.11-2-2 (no alarm/trip function) TR 16.11-2-7

3. Flow Rate Measurement Devices 3.a Waste Liquid Effluent Line 1 per station D, G TR 16.11-2-2 (no alarm/trip function) TR 16.11-2-7 3.b Conventional Waste Water Treatment Line 1 per station D, G TR 16.11-2-2 (no alarm/trip function) TR 16.11-2-7 3.c Low Pressure Service Water Minimum Flow Interlock 1 per station D, G TR 16.11-2-2 TR 16.11-2-5 TR 16.11-2-7 3.d Monitor Tank Building Waste Liquid Effluent Line 1 per station D, G TR 16.11-2-2 (no alarm/trip function) TR 16.11-2-7 4; Radioactivity Monitors Providing Alarm 4.a Service Water Monitor on Containment Spray Heat 1 per heat A, F, G TR 16.11-2-1 Exchanger exchanger TR 16.11-2-4 (EMF-45 A & B - Low Range) TR 16.11-2-6 TR 16.11-2-7 Catawba Units 1 and 2 16.11-2-7 Revision 3

Radioactive Liquid Effluent Monitoring Instrumentation 16.11-2 BASES The Radioactive Liquid Effluent Monitoring Instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in liquid effluents during actual or potential releases of liquid effluents. The Alarm/Trip Setpoints for these instruments shall be calculated 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 of 10 CFR Part 20. 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. Catawba Offsite Dose Calculation Manual.

2. 10 CFR Part 20.

3. 10 CFR Part 50, Appendix A.

Catawba Units 1 and 2 16.11-2-8 Revision 3

Dose 16.11-3 16.11 RADIOLOGICAL EFFLUENTS CONTROLS 16.11-3 Dose 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 Figure 16.11-16-1 in SLC 16.11-16) shall be limited:

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

b. During any calendar year to < 3 mrem to the whole body and to < 10 mrem to any organ.

APPLICABILITY: At all times.

REMEDIAL ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Calculated dose from A.1 --------- NOTE-------

release of radioactive If drinking water supply is materials in liquid taken from receiving water effluents exceeding body within 3 miles above limits, downstream of plant discharge, the Special Report shall also include the results of radiological analyses of the drinking water source and the radiological impact on finished drinking water supplies with regard to 40 CFR 141, Safe Drinking Water Act.

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.

Catawba Units 1 and 2 16.11-3-1 Revision 0

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

This SLC applies to the release of radioactive materials in liquid effluents from each unit at the site. When shared radwaste treatment systems are used by more than one unit on a site, the wastes from all units are mixed for shared treatment; by such mixing, the effluent releases cannot accurately be ascribed to a specific unit. An estimate should be made of the contributions from each unit based on input conditions, e.g., flow rates and radioactivity concentrations, or, if not practicable, the treated effluent releases may be allocated equally to each of the radioactive waste producing units sharing the radwaste treatment system. For determining conformance to COMMITMENTS, these allocations from shared radwaste treatment systems are to be added to the releases specifically attributed to each unit to obtain the total releases per unit.

Catawba Units 1 and 2 16.11-3-2 Revision 0

Dose 16.11-3 REFERENCES 1. Catawba Offsite Dose Calculation Manual.

2. 40 CFR Part 141.

3. 10 CFR Part 50, Appendix I.

Catawba Units 1 and 2 16.11-3-3 Revision 0

Liquid Radwaste Treatment System 16.11-4 16.11 RADIOLOGICAL EFFLUENTS 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 Figure 16.11-16-1 in SLC 16.11-16) would exceed 0.06 mrem to the whole 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 Special Report to the NRC without treatment and in which identifies the excess of above limits, reasons 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.

Catawba Units 1 and 2 16.11-4-1 Revision 1

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 and SLC 16.11-3.

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 the Liquid Radwaste Treatment System is not being fully utilized.

BASES The FUNCTIONALITY of the Liquid Radwaste Treatment System ensures that this system will be available for use whenever liquid effluents require treatment prior to release to the environment. The requirement that the appropriate portions of this system be used when specified provides assurance that the releases of radioactive materials in liquid effluents will be kept "as low as is reasonably achievable". This COMMITMENT implements the requirements of 10 CFR Part 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50, and the design objective given in Section II.D of Appendix I to 10 CFR Part 50. The specified limits governing the use of appropriate portions of the Liquid Radwaste Treatment System were specified as a suitable fraction of the dose design objectives set forth in Section II.A of Appendix 1,10 CFR Part 50, for liquid effluents.

This SLC applies to the release of radioactive materials in liquid effluents from each unit at the site. When shared radwaste treatment systems are used by more than one unit on a site, the wastes from all units are mixed for shared treatment; by such mixing, the effluent releases cannot accurately be ascribed to a specific unit. An estimate should be made of the contributions from each unit based on input conditions, e.g., flow rates and radioactivity concentrations, or, if not practicable, the treated effluent releases may be allocated equally to each of the radioactive waste producing units sharing the radwaste treatment system. For determining conformance to COMMITMENTS, these allocations from shared radwaste treatment systems are to be added to the releases specifically attributed to each unit to obtain the total releases per unit.

REFERENCES 1. Catawba Offsite Dose Calculation Manual.

2. 10 CFR Part 50, Appendix A.

Catawba Units 1 and 2 16.11-4-2 Revision 1

Liquid Radwaste Treatment System 16.11-4 REFERENCES (continued)

3. 10 CFR Part 50, Appendix I.

Catawba Units 1 and 2 16.11-4-3 Revision 1

Chemical Treatment Ponds 16.11-5 16.11 RADIOLOGICAL EFFLUENTS CONTROLS 16.11-5 Chemical Treatment Ponds COMMITMENT The quantity of radioactive material contained in each Chemical Treatment Pond (CTP) shall be limited by the following expression:

264 ___. A.' <1.0 V j (Cj x10) excluding tritium and dissolved or entrained noble gases, where:

A = CTP inventory limit for single radionuclide "j", in Curies; C = 10 CFR 20, Appendix B, Table 2, Column 2, concentration for single radionuclide "j", microCuries/milliliter; V = design volume of liquid and slurry in the CTP, 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 CTP radioactive material to the exceeding above limit. CTP.

AND A.2 Initiate corrective action to Immediately reduce the CTP contents to within limits.

Catawba Units 1 and 2 16.11-5-1 Revision 0

Chemical Treatment Ponds 16.11-5 TESTING REQUIREMENTS TEST FREQUENCY TR 16.11-5-1 Verify that the quantity of radioactive material contained Prior to each in each batch of resin/water slurry to be transferred to transfer the CTPs is within limits by analyzing a representative sample of the batch to be transferred. Each batch to be transferred to the CTPs shall be limited by:

x cj S< 0.006, j (C 1 x 10) where:

Cj= radioactive resin/water slurry concentration for radionuclide )j"entering the UNRESTRICTED AREA CTPs, in microCuries/milliliter; and C= 10 CFR 20, Appendix B, Table 2, Column 2, concentration for single radionuclide "j", in microCuries/milliliter.

BASES The inventory limits of the CTPs are based on limiting the consequences of an uncontrolled release of the pond inventory. The expression in this SLC 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 1 to Appendix B of 10 CFR Part 20 applies.

The batch limits of resin/water slurry transferred to the CTP assure that radioactive material transferred to the CTP are "as low as is reasonably achievable" in accordance with 10 CFR 50.36a. The expression in SLC 16.11-5 assures no batch 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:

ci 3mrem/yr -0.006, (Cj xlO) 500 mrem/yr where:

Catawba Units 1 and 2 16.11-5-2 Revision 0

Chemical Treatment Ponds 16.11-5 BASES (continued) c = radioactive resin/water slurry concentration for radionuclide "j" entering the UNRESTRICTED AREA CTP, in microCuries/milliliter; and, C= 10 CFR Part 20, Appendix B, Table 2, Column 2, concentration for single radionuclide "j", in microCuries/milliliter.

The filter/demineralizers using powdered resin and the blowdown demineralizer are backwashed or sluiced to a holding tank. The tank will be agitated to obtain a representative sample of the resin inventory in the tank. A known weight of the wet, drained resin (moisture content approximately 55 to 60%, bulk density of about 58 pounds per cubic foot) will then be counted. The concentration of the resin slurry to be pumped to the CTPs will then be determined by the formula:

QjWR VT where:

Q = concentration of radioactive materials in wet, drained 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 microCuries/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);

W R = total weight of resin in the storage tank in grams (determined from chemistry logs procedures); and, V T = total volume of resin water mixture in storage tank to be transferred to the CTPs in milliliters.

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

REFERENCES 1. Catawba Offsite Dose Calculation Manual.

2. 10 CFR Part 20, Appendix B.

3. 10 CFR Part 50, Appendix I.

Catawba Units 1 and 2 16.11-5-3 Revision 0

Gaseous Effluents 16.11-6 16.11 RADIOLOGICAL EFFLUENTS CONTROLS 16.11-6 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 Figure 16.11-16-1 in SLC 16.11-16) 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-1 31, for Iodine-1 33, for tritium, and for all radionuclides in particulate form with half-lives > 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 that the dose rate due to noble gases in gaseous In accordance with effluents is within limits in accordance with the the methodology methodology and parameters in the ODCM. and parameters in the ODCM TR 16.11-6-2 Verify that the dose rate due to Iodine-1 31, Iodine-1 33, According to Table tritium, and all radionuclides in particulate form with half- 16.11-6-1 lives > 8 days in gaseous effluents is within limits in accordance with the methodology and parameters in the ODCM by obtaining representative samples and performing analyses according to Table 16.11-6-1.

Catawba Units 1 and 2 16.11-6-1 Revision 1

Gaseous Effluents 16.11-6 Table 16.11-6-1 Radioactive Gaseous Waste Sampling and Analysis Program (page 1 of 4)

GASEOUS RELEASE TYPE SAMPLING MINIMUM TYPE OF ACTIVITY ANALYSIS LOWER LIMIT OF FREQUENCY ANALYSIS DETECTION FREQUENCY (LLD)"1 ' (pCi/ml)

1. Waste Gas Storage Tank Prior to each Prior to each Principal Gamma Emitters(2) lx1i0.

release release Each Tank Each Tank Grab Sample

2. Containment Purge Prior to each Prior to each Principal Gamma Emitters(2 ) 1x104 release release Each PURGE(3) Each PURGE(31 Grab Sample 31 days H-3 (oxide) 1x105

3. Unit Vent 7 days(3)(4) 7 days"' Principal Gamma Emitters(Z) 1x10-4 Grab Sample H-3 (oxide) lxi 0-6

4. Containment Air Release and 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months (3)(5 ) 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ("(" Principal Gamma Emitters(2) lx10-Addition System Grab Sample 31 days H-3 (oxide) lx1 0-

5. All Release Types as Listed in 3. Continuous"') 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ("' 1-131 lx10-11 Above Charcoal Sample 1-133 1xl0.9 Continuous"6 ' 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months "'7 Principal Gamma Emitters") l xi01 Particulate Sample Continuous"') 31 days Gross Alpha(8) lx10-11 Composite Particulate Sample Continuous"') 92 days Sr-89, Sr-90 1x10-11 Composite Particulate Sample _

(continued)

Catawba Units 1 and 2 16.11-6-2 Revision 1

Gaseous Effluents 16.11-6 Table 16.11-6-1 Radioactive Gaseous Waste Sampling and Analysis Program (page 2 of 4)

GASEOUS RELEASE TYPE SAMPLING MINIMUM TYPE OF ACTIVITY ANALYSIS LOWER LIMIT OF FREQUENCY ANALYSIS DETECTION FREQUENCY (LLD)(1) (pCi/ml)

6. Waste Monitor Tank Building 7 days 7 days Principal Gamma Emitters(2 ) 1x10 4 Ventilation Exhaust Grab Sample H-3 (oxide) 1x10.6 Continuous"') 7 days(9) 1-131 lxl 012 Charcoal Sample 1-133 1x1010 2 )

Continuous"6 ' 7 days9 Principal Gamma Emitters~ lX 1011 Particulate Sample Continuous"') 31 days Gross Alpha lxi 011 Composite Particulate Sample Continuous"') 92 days Sr-89, Sr-90 1x10-1 Composite Particulate Sample Catawba Units 1 and 2 16.11-6-3 Revision 1

Gaseous Effluents 16,11-6 Table 16.11-6-1 Radioactive Gaseous Waste Sampling and Analysis Program (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 = 4.66 sb E-V .2.22 x 106

Y. exp (-2At)

Where:

LLD = the "a priori" lower limit of detection (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 (counts per minute);

E = the counting efficiency (counts per disintegration);

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

2.22 x 106= the number of disintegrations per minute per microCurie; Y = the fractional radiochemical yield, when applicable;

?, the radioactive decay constant for the particular radionuclide (sec-1); and At = the elapsed time between midpoint of sample collection and time of counting (sec).

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.

Catawba Units 1 and 2 16.11-6-4 Revision 1

Gaseous Effluents 16.11-6 Table 16.11-6-1 Radioactive Gaseous Waste Sampling and Analysis Program (page 4 of 4)

(2) The principal gamma emitters for which the LLD specification applies include the following radionuclides: Kr-87, Kr-88, Xe-133, Xe-133m, Xe-135, and Xe-138 in noble gas releases based on grab samples 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 based on continuous samples. The LLD for Ce-144 is 5x10-9 gCi/ml and is based on continuous samples. 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 Radioactive Effluent Release Report, pursuant to Technical Specification 5.6.3 in the format outlined in Regulatory Guide 1.21, Appendix B, Revision 1, June 1974.

(3) Sampling and analysis shall also be performed following shutdown, startup, or a THERMAL POWER stabilization (power level constant at desired power level) after a THERMAL POWER change exceeding 15% of RATED THERMAL POWER within a 1-hour period, for at least one of the three gaseous release types with this notation.

.(4) 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.

(5) Required sampling and analysis frequency during effluent release via this pathway.

(6) 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.

(7) Samples shall be changed at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and analyses shall be completed within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months after changing, or after removal from sampler.

(8) The composite filter(s) will be analyzed for alpha activity by analyzing one filter per week to ensure that at least four filters are analyzed per collection period.

(9) Samples shall be changed at least once per 7 days and analyses shall be completed within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months to meet LLDs after changing, or after removal from sampler. If the particulate and charcoal sample frequency is changed to a 24-hour frequency, the corresponding LLDs may be increased by a factor of 10 (e.g., LLD for 1-131 from lx10 1 2 to 1x10-11 pCi/ml).

Catawba Units 1 and 2 16.11-6-5 Revision 1

Gaseous Effluents 16.11-6 BASES The basic requirements for SLCs concerning effluents from nuclear power reactors are stated in 10 CFR 50.36a. These requirements indicate that compliance with effluent SLCs will keep average annual releases of radioactive material in effluents to small percentages of the limits specified in the old 10 CFR 20.106 (new 10 CFR 20.1301). 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 10 CFR 20.106 which references Appendix B, Table II concentrations (MPCs). These referenced concentrations are specific values which relate to an annual dose of 500 mrems. It is further indicated in 10 CFR 50.36a 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 10 CFR 50, Appendix I.

As stated in the Introduction to Appendix B of the new 10 CFR 20, the gaseous effluent concentration (EC) limits given in Appendix B, Table 2, Column 1, are based on an annual dose of 50 mrems for isotopes for which inhalation or ingestion is limiting or 100 mrems for isotopes for which submersion (noble gases) is limiting. Since release concentrations corresponding to limiting dose rates less than or equal to 500 mrems/year to thewhole body, 3000 mrems/year to the skin from noble gases, and 1500 mrems/year to any organ from Iodine-131, lodine-133, tritium, and all radionuclides in particulate form with half-lives greater than 8 days at the site boundary has been acceptable as a SLC limit for gaseous effluents to assure that the limits of 10 CFR 50, Appendix I and 40 CFR 190 are not likely to be exceeded, it should not be necessary to restrict the operational flexibility by incorporating the dose rate associated with the EC value for isotopes based on inhalation/ingestion (50 mrems/year) or the dose rate associated with the EC value for isotopes based on submersion (100 mrems/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 10 CFR 20, Appendix B, Table 2, Column 1, relate to a dose of 50 or 100 mrems in a year. When applied on an instantaneous basis, this corresponds to a dose rate of 50 or 100 mrems/year.

These low values are impractical upon which to base effluent monitor setpoint calculations for many gaseous 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 gaseous release rate SLCs will be maintained at the current instantaneous dose rate limit for noble gases of 500 mrems/year to the whole body and 3000 mrems/year to the skin; and for Iodine-1 31, for Iodine-1 33, for tritium, and for all radionuclides in particulate Catawba Units 1 and 2 16.11-6-6 Revision 1

Gaseous Effluents 16.11-6 BASES (continued) form with half-lives greater than 8 days, an instantaneous dose rate limit of 1500 mrems/year to any organ.

Compliance with the limits of the new 10 CFR 20.1301 will be demonstrated by operating within the limits of 10 CFR 50, Appendix I and 40 CFR 190.

Operational history at Catawba has demonstrated that the use of the dose rate values listed above (i.e., 500 mrems/year, 3000 mrems/year, and 1500 mrems/year) as SLC limits has resulted in calculated maximum individual doses to MEMBERS OF THE PUBLIC that are small percentages of the limits of 10 CFR 50, Appendix I and 40 CFR 190.

The specified 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 to less than or equal to 3000 mrem/year to the skin from noble gases, and to less than or equal to 1500 mrem/year to any organ from Iodine-1 31, lodine-133, tritium, and all radionuclides in particulate form with half-lives greater than eight days.

This commitment applies to the release of radioactive materials in gaseous effluents from all units at the site.

The required detection capabilities for radioactive material in gaseous waste samples are tabulated in terms of the lower limits of detection (LLDs). Based on NUREG-1301 and Regulatory Guide 1.21, the LLD value of lx10- VCi/ml for grab samples is only applicable to noble gases grab samples and the LLD values for particulate and iodine radionuclides are applicable to continuous charcoal and particulate samples. The Table 16.11-6-1 Gaseous Release Type Number 6 (Waste Monitor Tank Building Ventilation Exhaust) LLDs are based on weekly samples per NUREG-1301. The Table 16.11-6-1 Gaseous Release Type Number 5 (All Release Types as Listed in 3. Above) LLDs, for the 24-hour charcoal and particulate samples, are based on daily (once per 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ) samples per NUREG-1301. There are two isotopes with associated LLDs that do not agree directly with NUREG-1 301: Ce-144, LLD of 5x10 9 VtCi/ml, which has historically been applied and achieved for analytical results, and 1-133, LLD of lx10Q10 pCi/ml, which again has been historically listed, as 1x10-9 pVCi/ml, for Radioactive Gaseous Waste Sampling but changed to be in agreement with 1-131 for weekly (7-day) samples and is not specified in NUREG-1301. 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).

Catawba Units 1 and 2 16.11-6-7 Revision 1

Gaseous Effluents 16.11-6 REFERENCES 1. Catawba Offsite Dose Calculation Manual.

2. 10 CFR Part 20, Appendix B.

3. 10 CFR Part 20.

4. 10 CFR Part 50.

5. 40 CFR Part 190.

6. NUREG-1301.

7. Regulatory Guide 1.21.

Catawba Units 1 and 2 16.11-6-8 Revision 1

Radioactive Gaseous Effluent Monitoring Instrumentation 16.11-7 16.11 RADIOLOGICAL EFFLUENTS 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 their Alarm/Trip Setpoints set to ensure that the limits of SLC 16.11-6 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: As shown in Table 16.11-7-1.

REMEDIAL ACTIONS

NOT r---------------------------------------------------------

Separate Condition entry is allowed for each FL CONDITION REQUIRED ACTION COMPLETION TIME A. One or more A.1 Suspend the release of Immediately Radioactive Gaseous radioactive gaseous Effluent Monitoring effluents monitored by the Instrumentation affected channel(s).

channel(s) Alarm/Trip Setpoint less OR conservative than required. A.2 Declare the channel(s) Immediately non-functional.

B. One or more B.1 Enter the applicable Immediately Radioactive Gaseous Conditions and Required Effluent Monitoring Actions specified in Table Instrumentation 16.11-7-1 for the channel(s) non- channel(s).

functional.

(continued)

Catawba Units 1 and 2 16.11-7-1 Revision 6

Radioactive Gaseous Effluent Monitoring Instrumentation 16.11-7 REMEDIAL ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C. One channel non- C. 1 Verify that EMF-36 (Low Prior to initiating a functional. Range) is FUNCTIONAL. release OR C.2.1 Analyze two independent Prior to initiating a samples of the tank's release contents.

AND C.2.2 Perform independent Prior to initiating a verification of the release discharge line valving.

AND C.2.3.1 Perform independent Prior to initiating a verification of manual release portion of the computer input for release rate calculations performed by computer.

OR C.2.3.2Perform independent Prior to initiating a verification of entire release calculations for release rate calculations performed manually.

AND C.2.4 Restore channel to 14 days FUNCTIONAL status.

OR C.3 Suspend release of Immediately radioactive effluents via this pathway.

(continued)

Catawba Units 1 and 2 16.11-7-2 Revision 6

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 device the release. during releases channel(s) non-functional. AND D.2 Restore 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 effluent pathway. during releases channel(s) non-functional. AND E.2 Perform an analysis of Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of grab samples for obtaining the sample radioactivity.

AND E.3 Restore channel to 30 days FUNCTIONAL status.

(continued)

Catawba Units 1 and 2 16.11-7-3 Revision 6

Radioactive Gaseous Effluent Monitoring Instrumentation 16.11-7 REMEDIAL ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME F. Noble Gas Activity F. 1 --------- NOTE --------------

Monitor (EMF Low In order to utilize Required Range) providing Action F.1, the following automatic termination of conditions must be release via the satisfied:

Containment Purge 1. The affected unit is in Exhaust System (CPES) MODES 5 or 6.

non-functional. 2. EMF-36 is FUNCTIONAL and in service for the affected unit.

3. The Reactor Coolant System for the affected unit has been vented.

4. Either the reactor vessel head is in place (bolts are not required),

or if it is not in place, the lifting of heavy loads over the reactor vessel and the movement of irradiated fuel assemblies within containment have been suspended.

Restore the non-functional 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months channel to FUNCTIONAL status.

G. Required Action and G.1 Suspend PURGING of Immediately associated Completion radioactive effluents via Time of Condition F not this pathway.

met.

OR Required Action F.1 not utilized.

(continued)

Catawba Units 1 and 2 16.11-7-4 Revision 6

Radioactive Gaseous Effluent Monitoring Instrumentation 16.11-7 REMEDIAL ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME H. One or more sampler H.1 Perform sampling with Continuously channel(s) non- auxiliary sampling functional. equipment as required by Table 16.11-6-1.

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

One Condenser 1.1 ---------- NOTE-------

Evacuation System Applicable to effluent Noble Gas Activity releases via the Condenser Monitor (EMF-33) Steam Air Ejector (ZJ) channel non-functional. System.

Obtain grab samples from Once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months effluent pathway. during releases AND 1.2 ---------- NOTE-------

Applicable to effluent releases via the Condenser Steam Air Ejector (ZJ)

System.

Perform an analysis of Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of grab samples for obtaining the sample radioactivity.

AND (continued)

Catawba Units 1 and 2 16.11-7-5 Revision 6

Radioactive Gaseous Effluent Monitoring Instrumentation 16.11-7 REMEDIAL ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME (continued) 1.3 ---------- NOTE -------------

Applicable to effluent releases via the Steam Generator Blowdown (BB)

System atmospheric vent valve (BB-27) in the off-normal mode.

Perform an analysis of Once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months grab samples for during releases radioactivity at a lower limit when secondary of detection of 10-7 specific activity is >

microCurie/ml. 0.01 microCurie/gm DOSE EQUIVALENT 1-131 AND Once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months during releases when secondary specific activity is <

0.01 microCurie/gm DOSE EQUIVALENT 1-131 AND 1.4 Restore channel to 30 days FUNCTIONAL status.

J. Noble Gas Activity J. 1 Verify that EMF-36 is Prior to initiating a Monitor (EMF Low FUNCTIONAL. release Range) providing automatic termination of OR release via the Containment Air J.2.1 Analyze two independent Prior to initiating a Release and Addition samples of the release System non-functional. containment atmosphere.

AND (continued)

Catawba Units 1 and 2 16.11-7-6 Revision 6

Radioactive Gaseous Effluent Monitoring Instrumentation 16.11-7 REMEDIAL ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME J. (continued) J.2.2 Perform independent Prior to initiating a verification of the release discharge line valving.

AND J.2.3.1 Perform independent Prior to initiating a verification of manual release portion of the computer input for release rate calculations performed by computer.

OR J.2.3.2 Perform independent Prior to initiating a verification of entire release calculations for release rate calculations performed manually.

AND J.2.4 --------- NOTE -------

If channel remains or is anticipated to remain non-functional for > 90 days, re-evaluate the configuration of the affected unit in accordance with the applicable portions of 10 CFR 50.59 and 10 CFR 50.65(a)(4) prior to expiration of the 90-day period.

Restore channel to 30 days FUNCTIONAL status.

(continued)

Catawba Units 1 and 2 16.11-7-7 Revision 6

Radioactive Gaseous Effluent Monitoring Instrumentation 16.11-7 REMEDIAL ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME K. Required Action and K.1 Explain why the non- In the next associated Completion functionality was not scheduled Time of Condition C, D, corrected within the Radioactive Effluent E, F, H, I, or J not met. specified Completion Time. Release Report pursuant to Technical Specification 5.6.3 TESTING REQUIREMENTS

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

Refer to Table 16.11-7-1 to determine which TRs apply for each Radioactive Gaseous Effluent Monitoring Instrumentation channel.

TEST FREQUENCY TR 16.11-7-1 Perform CHANNEL CHECK. Prior to each release TR 16.11-7-2 -------------------

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

For Instruments la, 4, and 5, a SOURCE CHECK for these channels shall be the qualitative assessment of channel response when the channel sensor is exposed to a light-emitting diode.

Perform SOURCE CHECK. Prior to each release TR 16.11-7-3 Perform CHANNEL CHECK. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months TR 16.11-7-4 Perform CHANNEL CHECK. 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months TR 16.11-7-5 Perform CHANNEL CHECK. 7 days (continued)

Catawba Units 1 and 2 16.11-7-8 Revision 6

Radioactive Gaseous Effluent Monitoring Instrumentation 16.11-7 TESTING REQUIREMENTS (continued)

TEST FREQUENCY TR 16.11-7-6 ------------------- NOTE ------------------

For Instruments 2 and 3a, a SOURCE CHECK for these channels shall be the qualitative assessment of channel response when the channel sensor is exposed to a light-emitting diode.

Perform SOURCE CHECK. 31 days TR 16.11-7-7 ------------------- NOTE -----------------

For Instruments Ia, 2, 3a, 3c, 5, and 6a, the COT shall also demonstrate, as applicable, that automatic isolation of this pathway and control room alarm annunciation (for EMF-58, alarm annunciation is in the Monitor Tank Building control room and on the Monitor Tank Building control panel remote annunciator panel) occur if any of the following conditions exist:

a. Instrument indicates measured levels above the Alarm/Trip Setpoint, or

b. Circuit failure/instrument downscale failure (alarm only)

Perform COT. 9 months TR 16.11-7-8 ------------------- NOTE -----------------

For Instrument 4, the COT shall also demonstrate that automatic isolation of this pathway and control room alarm annunciation occur if any of the following conditions exist:

a. Instrument indicates measured levels above the Alarm/Trip Setpoint, or

b. Circuit failure/instrument downscale failure (alarm only)

Perform COT. 18 months (continued)

Catawba Units 1 and 2 16.11-7-9 Revision 6

Radioactive Gaseous Effluent Monitoring Instrumentation 16.11-7 TESTING REQUIREMENTS (continued)

TEST FREQUENCY TR 16.11-7-9 -NOTE ---------------------------------

For Instruments la, 2, 3a, 3c, 4, 5, and 6a, the initial CHANNEL CALIBRATION shall be performed using one or more of the reference standards certified by the National Bureau of Standards (NBS) or using standards that have been obtained from suppliers that participate in measurement assurance activities with NBS. 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 CHANNEL CALIBRATION. 18 months Catawba Units 1 and 2 16.11-7-10 Revision 6

Radioactive Gaseous Effluent Monitoring Instrumentation 16.11-7 Table 16.11-7-1 Radioactive Gaseous Effluent Monitoring Instrumentation (page 1 of 2)

INSTRUMENT REQUIRED CONDITIONS APPLICABLE TESTING CHANNELS MODES REQUIREMENTS

1. Waste Gas Holdup System l.a Noble Gas Activity Monitor - Providing 1 per station A, C, K At all times except TR 16.11-7-1 Alarm and Automatic Termination of when the isolation TR 16.11-7-2 Release valve is closed and TR 16.11-7-7 (EMF Low Range) locked TR 16.11-7-9 1.b Effluent System Flow Rate Measuring 1 per station D, K At all times except TR 16.11-7-1 Device when the isolation TR 16.11-7-9 valve is closed and locked

2. Condenser Evacuation System Noble 1 A, I, K When air ejectors TR 16.11-7-3 Gas Activity Monitor are in operation TR 16.11-7-6 (EMF-33) (BB-27 is only isolation (Apply Required TR 16.11-7-7 function required) (Note 1) Action 1.3 when air TR 16.11-7-9 ejectors are not in operation)

3. Vent System 3.a Noble Gas Activity Monitor 1 A, E, K At all times TR 16.11-7-4 (EMF Low Range) TR 16.11-7-6 TR 16.11-7-7 TR 16.11-7-9 3.b Deleted.

3.c Particulate Sampler 1 A, H, K At all times TR 16.11-7-4 (EMF-35) (Note 2) TR 16.11-7-6 TR 16.11-7-7 TR 16.11-7-9 3.d Unit Vent Stack Flow Rate Meter 1 D, K At all times TR 16.11-7-4 (no alarm/trip function) (Note 2) TR 16.11-7-9 3.e Unit Vent Radiation Monitor Flow Meter 1 E, K At all times TR 16.11-7-4 (Note 2) TR 16.11-7-9

4. Containment Purge System Noble Gas 1 A, F, G, K 5,6 TR 16.11-7-2 Activity Monitor - Providing Alarm and TR 16.11-7-3 Automatic Termination of Release TR 16.11-7-8 (EMF Low Range) TR 16.11-7-9 (continued)

Catawba Units 1 and 2 16.11-7-11 Revision 6

Radioactive Gaseous Effluent Monitoring Instrumentation 16.11-7 Table 16.11-7-1 Radioactive Gaseous Effluent Monitoring Instrumentation (page 2 of 2)

INSTRUMENT REQUIRED CONDITIONS APPLICABLE TESTING CHANNELS MODES REQUIREMENTS

5. Containment Air Release and Addition I A, J, K 1, 2, 3,4,5,6 TR 16.11-7-2 System Noble Gas Activity Monitor- TR 16.11-7-3 Providing Alarm and Automatic TR 16.11-7-7 Termination of Release TR 16.11-7-9 (EMF Low Range)

6. Monitor Tank Building HVAC 6.a Noble Gas Activity Monitor - Providing 1 per station A, E, K At all times TR 16.11-7-4 Alarm (Note 2) TR 16.11-7-6 (EMF Low Range) TR 16.11-7-7 TR 16.11-7-9 6.b Effluent Flow Rate Measuring Device 1 per station D, K At all times TR 16.11-7-4 (Note 2) TR 16.11-7-9 Note 1: The setpoint is as required by the primary to secondary leak rate monitoring program.

Note 2: Except when the effluent pathway is isolated and a release to the environment is not possible.

Catawba Units 1 and 2 16.11-7-12 Revision 6

Radioactive Gaseous Effluent Monitoring Instrumentation 16.11-7 BASES The Radioactive Gaseous Effluent Monitoring Instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in gaseous effluents during actual or potential releases of gaseous effluents. The Alarm/Trip Setpoints for these instruments shall be calculated 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 of 10 CFR Part 20. 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 sensitivity of any noble gas activity monitor used to show compliance with the gaseous effluent release requirements of SLC 16.11-8 shall be such that concentrations as low as 1 x 10-6 pCi/cc are measurable.

Initiation of the Containment Purge Exhaust System (CPES) with EMF-39 non-functional is not permissible. The basis for Required Action F.1 is to allow the continued operation of the CPES with EMF-39 initially FUNCTIONAL. Continued operation of the CPES is contingent upon the ability of the affected unit to meet the requirements as noted in Required Action F.1.

TR 16.11-7-7 requires the performance of a COT on the applicable Radioactive Gaseous Effluent Radiation Monitors. The test ensures that a signal from the control room module can generate the appropriate alarm and actuations. The required actuations/isolations for a High Radiation condition (i.e., radiation level above its Trip 2 setpoint) are listed below for each monitor.

OEMF Waste Gas Discharge Monitor 1WG160 closes when EMF-50 detects radiation level above its setpoint.

1/2EMF Condensate Steam Air Ejector Exhaust Monitor The following actuations occur when EMF-33 detects radiation level above its setpoint:

1. Closure of BB27 is required in order to isolate the Blowdown Tank from the environment. Because of plant limitations/restrictions:

a. Opening the valve (in order to verify it goes closed on a High Radiation signal) is only possible during outages due to the negative effects on the Blowdown System with the unit at power.

b. Testing during innages will be by verification of relay contacts opening in the valve circuit.

2. Closure of BB24, BB65, BB69, and BB73 is required to minimize the amount of potentially contaminated material being delivered to the Blowdown Tank.

3. Closure of NM269, NM270, NM271, and NM272 is required to minimize the amount of potentially contaminated material being delivered to the Conventional Sampling System.

4. Closure of NM267 is required to minimize the amount of potentially contaminated material being delivered to the Condensate Storage Tank by isolating flow through EMF-34.

5. Closure of BB48 is required to minimize the amount of potentially contaminated material being delivered from the Blowdown System discharge to the Turbine Building sump.

Catawba Units 1 and 2 16.11-7-13 Revision 6

Radioactive Gaseous Effluent Monitoring Instrumentation 16.11-7 BASES (continued) 1/2EMF Unit Vent Noble Gas Monitor The following actuations occur when EMF-36 detects radiation level above its setpoint:

1. Containment Air Release and Addition System fans discharge to unit vent valve VQ10 closes.

2. Auxiliary Building unfiltered ventilation exhaust fans A and B stop.

3. Fuel Handling Ventilation Exhaust System (FHVES) exhaust trains align to the filter units.

4. (For 1EMF-36 only) 1WG160 closes.

1/2EMF Unit Vent Particulate Monitor (Sampler)

The following actuations occur when EMF-35 detects radiation level above its setpoint:

1. Containment Air Release and Addition System fans discharge to unit vent valve VQ10 closes.

2. Auxiliary Building unfiltered ventilation exhaust fans A and B stop.

3. Fuel Handling Ventilation Exhaust System (FHVES) exhaust trains align to the filter units.

4. ((For 1EMF-35 only) 1WG160 closes.

1/2EMF Containment Noble Gas Monitor The following actuations occur when EMF-39 detects radiation level above its setpoint:

1. Signals are provided to both trains of the Solid State Protection System (SSPS) to initiate a CPES isolation. This is verified by observing that Relays K615 in the SSPS A output cabinet and the SSPS B output cabinet are latched.

2. EMF-39 isolates the CPES without going through the SSPS by stopping CPES supply fans.A and B, CPES exhaust fans A and B, and by closing the appropriate valves and dampers.

3. Containment Evacuation Alarm, unless the source range trip is blocked.

OEMF-58 This monitor provides no control function.

TR 16.11-7-8 requires the performance of a COT on the Containment Noble Gas Monitor, 1/2EMF-39. The test ensures that a signal from the control room module can generate the appropriate alarm and actuations. The required actuations/isolations for a High Radiation condition (i.e., radiation level above its Trip 2 setpoint) are listed below.

1/2EMF Containment Noble Gas Monitor The following actuations occur when EMF-39 detects radiation level above its setpoint:

1. Containment Air Release and Addition System fans discharge to unit vent valve VQ10 closes.

2. Containment Evacuation Alarm, unless the source range trip is blocked.

Catawba Units 1 and 2 16.11-7-14 Revision 6

Radioactive Gaseous Effluent Monitoring Instrumentation 16.11-7 REFERENCES 1. Catawba Offsite Dose Calculation Manual.

2. 10 CFR Part 20.

Catawba Units 1 and 2 16.11-7-15 Revision 6

Dose - Noble Gases 16.11-8 16.11 RADIOLOGICAL EFFLUENTS CONTROLS 16.11-8 Dose - Noble Gases COMMITMENT The air dose due to noble gases released in gaseous effluents, from each unit, to areas at and beyond the SITE BOUNDARY (see Figure 16.11-16-1 in SLC 16.11-16) shall be limited to the following:

a. During any calendar quarter: < 5 mrad for gamma radiation and < 10 mrad for beta radiation, and

b. During any calendar year: < 10 mrad for gamma radiation and

< 20 mrad for beta radiation.

APPLICABILITY: At all times.

REMEDIAL ACTIONS 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.

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.

Catawba Units 1 and 2 16.11-8-1 Revision 0

Dose - Noble Gases 16.11-8 BASES This SLC is provided to implement the requirements of Sections II.B, III.A and IV.A of Appendix 1,10 CFR Part 50. The COMMITMENT implements the guides set forth in Section II.B of Appendix I. The REMEDIAL ACTION statement provides 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 III.A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data such that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated. The 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 to the release of radioactive materials in gaseous effluents from each unit at the site. When shared radwaste treatment systems are used by more than one unit on a site, the wastes from all units are mixed for shared treatment; by such mixing, the effluent releases cannot accurately be ascribed to a specific unit. An estimate should be made of the contributions from each unit based on input conditions, e.g., flow rates and radioactivity concentrations, or, if not practicable, the treated effluent releases may be allocated equally to each of the radioactives waste producing units sharing the radwaste treatment system. For determining conformance to COMMITMENTS, these allocations from shared radwaste treatment systems are to be added to the releases specifically attributed to each unit to obtain the total releases per unit.

REFERENCES 1. Catawba Offsite Dose Calculation Manual.

2. 10 CFR Part 50, Appendix I.

Catawba Units 1 and 2 16.11-8-2 Revision 0

Dose - Iodine-1 31, Iodine-1 33, Tritium, and Radioactive Material in Particulate Form 16.11-9 16.11 RADIOLOGICAL EFFLUENTS CONTROLS 16.11-9 Dose - Iodine-131, Iodine-133, Tritium, and Radioactive Material in Particulate Form COMMITMENT The dose to a MEMBER OF THE PUBLIC from Iodine-131, Iodine-133, tritium, and all radionuclides in particulate form with half-lives > 8 days in gaseous effluents released, from each unit, to areas at and beyond the SITE BOUNDARY (see Figure 16.11-16-1 in SLC 16.11-

16) shall be limited to the following:

a. During any calendar quarter: < 7.5 mrem to any organ, and

b. During any calendar year: < 15 mrem to any organ.

APPLICABILITY: At all times.

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

TESTING REQUIREMENTS TEST FREQUENCY TR 16.11-9-1 Determine cumulative dose contributions from Iodine- 31 days 131, Iodine-133, tritium, and radioactive material in particulate form with half-lives > 8 days in gaseous effluents for current calendar quarter and current calendar year in accordance with the methodology and parameters in the ODCM.

Catawba Units 1 and 2 16.11-9-1 Revision 0

Dose - Iodine-1 31, Iodine-1 33, Tritium, and Radioactive Material in Particulate Form 16.11-9.

BASES This SLC is provided to implement the requirements of Sections II.C, III.A and IV.A of Appendix I, 10 CFR Part 50, and are the guides set forth in Section II.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 III.A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data, such that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated. The 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 COMMITMENTS for Iodine-1 31, Iodine-1 33, 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 the 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.

This commitment applies to the release of radioactive materials in gaseous effluents from each unit at the site. When shared radwaste treatment systems are used by more than one unit on a site, the wastes from all units are mixed for shared treatment; by such mixing, the effluent releases cannot accurately be ascribed to a specific unit. An estimate should be made of the contributions from each unit based on input conditions, e.g., flow rates and radioactivity concentrations, or, if not practicable, the treated effluent releases may be allocated equally to each of the radioactive waste producing units sharing the radwaste treatment system. For determining conformance to COMMITMENTS, these allocations from shared radwaste treatment systems are to. be added to the releases specifically attributed to each unit to obtain the total releases per unit.

REFERENCES 1. Catawba Offsite Dose Calculation Manual.

2. 10 CFR Part 50, Appendix I.

Catawba Units 1 and 2 16.11-9-2 Revision 0

Gaseous Radwaste Treatment System 16.11-10 16.11 RADIOLOGICAL EFFLUENTS CONTROLS 16.11-10 Gaseous Radwaste Treatment System COMMITMENT The VENTILATION EXHAUST TREATMENT SYSTEM and the WASTE GAS HOLDUP SYSTEM 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 16.11-16-1 in SLC 16.11-16) would exceed either:

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 gaseous A.1 Prepare and submit a 30 days waste being discharged Special Report to the NRC without treatment and in which identifies non-excess 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.

Catawba Units 1 and 2 16.11-10-1 Revision 1

Gaseous Radwaste Treatment System 16.11-10 TESTING REQUIREMENTS

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

The installed VENTILATION EXHAUST TREATMENT SYSTEM and WASTE GAS HOLDUP SYSTEM shall be demonstrated FUNCTIONAL by meeting SLC 16.11-6, SLC 16.11-8, and SLC 16.11-9.

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 Radwaste Treatment Systems are not being fully utilized.

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 10 CFR 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 Sections lI.B and II.C of Appendix I, 10 CFR Part 50, for gaseous effluents.

This SLC applies to the release of radioactive materials in gaseous effluents from each unit at the site. When shared radwaste treatment systems are used by more than one unit on a site, the wastes from all units are mixed for shared treatment; by such mixing, the effluent releases cannot accurately be ascribed to a specific unit. An estimate should be made of the contributions from each unit based on input conditions, e.g., flow rates and radioactivity concentrations, or, if not practicable, the treated effluent releases may be allocated equally to each of the radioactive waste producing units sharing the radwaste treatment system. For determining conformance to COMMITMENTS, these allocations from shared radwaste treatment systems are to be added to the releases specifically attributed to each unit to obtain the total releases per unit.

Catawba Units 1 and 2 16.11-10-2 Revision 1

Gaseous Radwaste Treatment System 16.11-10 REFERENCES 1. Catawba Offsite Dose Calculation Manual.

2. 10 CFR Part 50, Appendix I.

Catawba Units 1 and 2 16.11-10-3 Revision 1

Solid Radioactive Wastes 16.11-11 16.11 RADIOLOGICAL EFFLUENTS CONTROLS 16.11-11 Solid Radioactive Wastes COMMITMENT Radioactive wastes shall be processed and packaged to ensure compliance with the applicable requirements of 10 CFR Part 20, 10 CFR 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 the PROCESS CONTROL PROGRAM for the solidification of liquid or wet radioactive wastes or the dewatering of wet radioactive wastes to be shipped for direct disposal at a 10 CFR Part 61 licensed disposal site. Wastes shipped for offsite processing in accordance with the processor's specifications and transportation requirements are not required to be solidified or dewatered to meet disposal requirements.

APPLICABILITY: At all times.

REMEDIAL ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Applicable regulatory A.1 Suspend shipment of Immediately requirements for inadequately processed solidified or dewatered waste.

wastes not satisfied.

AND A.2 Take action to correct the Prior to next PROCESS CONTROL shipment for disposa PROGRAM, procedures, of solidified or or solid waste equipment dewatered wastes as necessary to prevent recurrence.

(continued)

Catawba Units 1 and 2 16.11-11-1 Revision 1

Solid Radioactive Wastes 16.11-11 REMEDIAL ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B. Solidification test as B.1 Suspend solidification of Immediately described in the the batch under test and PROCESS CONTROL follow PROCESS PROGRAM fails to CONTROL PROGRAM verify solidification, guidance for test failures.

AND B.2 --------- NOTE-------------

Once a subsequent test verifies solidification, solidification of the batch may be resumed as directed by the PROCESS CONTROL PROGRAM.

Modify the PROCESS Prior to next CONTROL PROGRAM as solidification for required to assure shipment of waste solidification of subsequent for disposal at a 10 batches of waste. CFR Part 61 disposal site C. Solidification or C.1 Reprocess the waste in Prior to shipment for dewatering for disposal accordance with disposal of the not performed in PROCESS CONTROL inadequately accordance with the PROGRAM requirements. processed waste tha PROCESS CONTROL requires solidificatior PROGRAM. OR or dewatering C.2 Follow PROCESS Prior to shipment for CONTROL PROGRAM or disposal of the procedure guidance for inadequately alternative free-standing processed waste tha liquid verification to ensure requires solidificatior the waste in each or dewatering container meets disposal requirements and take appropriate administrative action to prevent recurrence.

(continued)

Catawba Units 1 and 2 16.11-11-2 Revision 1

Solid Radioactive Wastes 16.11-11 REMEDIAL ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME D. Solid waste equipment D.1 Restore the equipment to a In a time frame incapable of supporting status capable of supporting COMMITMENT. supporting COMMITMENT. COMMITMENT OR D.2 Provide for alternative In a time frame capability to process supporting wastes as necessary to COMMITMENT satisfy all applicable transportation and disposal requirements.

TESTING REQUIREMENTS TEST FREQUENCY TR 16.11-11-1 Verify, using the PROCESS CONTROL PROGRAM, the Every tenth batch solidification of at least one representative test specimen of each type of from at least every tenth batch of each type of radioactive waste radioactive waste to be solidified for disposal at a 10 to be solidified CFR Part 61 disposal site.

BASES This SLC implements the requirements of 10 CFR Part 50.36a and General Design Criterion 60 of Appendix A to 10 CFR Part 50 and requirements to use a PROCESS CONTROL PROGRAM to meet applicable 10 CFR Part 61 waste form criteria for solidified and dewatered radioactive wastes.

The PROCESS CONTROL PROGRAM describes administrative and operational controls used for the solidification of liquid or wet solid radioactive wastes in order to meet applicable 10 CFR Part 61 waste form requirements.

The PROCESS CONTROL PROGRAM describes the administrative and operational controls used for the dewatering of wet radioactive wastes to meet 10 CFR Part 61 free-standing water requirements.

The process parameters used in establishing the PROCESS CONTROL PROGRAM 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.

Catawba Units 1 and 2 16.11-11-3 Revision I

Solid Radioactive Wastes 16.11-11 REFERENCES 1. 10 CFR Part 50, "Domestic Licensing of Production and Utilization Facilities."

2. 10 CFR Part 50, Appendix A.

3. 10 CFR Part 20, "Standards for Protection Against Radiation."

4. 10 CFR Part 61, "Licensing Requirements for Land Disposal of Radioactive Waste."

5. 10 CFR Part 71, "Packaging and Transportation of Radioactive Materials."

6. PROCESS CONTROL PROGRAM Manual.

7. Generic Letter 84-12, "Compliance with 10 CFR Part 61 and Implementation of the Radiological Effluent Technical Specifications (RETS) and Attendant Process Control Program (PCP)."

8. Generic Letter 89-01, "Implementation of Programmatic Controls for Radiological Effluent Technical Specifications in the Administrative Controls Section of the Technical Specifications and the Relocation of Procedural Details of RETS to the Offsite Dose Calculation Manual or to the Process Control Program."

Catawba Units 1 and 2 16.11-11-4 Revision 1

Total Dose 16.11-12 16.11 RADIOLOGICAL EFFLUENTS 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 to 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, that Immediately releases exceeding the cumulative dose from twice the specified limits direct radiation of SLC 16.11-3, SLC contributions and outside 16.11-8, or SLC 16.11- storage tanks and

9. 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.

Catawba Units 1 and 2 16.11-12-1 Revision 0

Total Dose 16.11-12 TESTING REQUIREMENTS

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

Cumulative dose contributions from liquid and gaseous effluents shall be determined in accordance with SLC 16.11-3, SLC 16.11-8, and SLC 16.11-9, 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 and from radwaste storage tanks doses from in accordance with the methodology and parameters effluent releases specified in the ODCM. exceed twice the limits of SLC 16.11-3, SLC 16.11-8, or SLC 16.11-9 BASES This SLC 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 SLC requires the preparation and submittal of a Special Report whenever the calculated doses due to releases of radioactivity and to radiation from uranium fuel cycle sources 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 Appendix I, and if direct radiation doses from the units and from outside storage tanks are kept small.

This Special Report, as defined in 10 CFR 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.

Catawba Units 1 and 2 16.11-12-2 Revision 0

Total Dose 16.11-12 BASES (continued)

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 accordance with the provisions of 40 CFR 190.11 and 10 CFR 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 SLC 16.11-1 and SLC 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. Catawba Offsite Dose Calculation Manual.

2. 10 CFR Part 20.

3. 40 CFR Part 190.

Catawba Units 1 and 2 16.11-12-3 Revision 0

Monitoring Program 16.11-13 16.11 RADIOLOGICAL EFFLUENTS CONTROLS 16.11-13 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 In 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 16.11- the Annual Radiological Operating Report 13-1. Environmental Operating pursuant to Report. Technical Specification 5.6.2 B. Radioactivity level B.1 Prepare and submit a 30 days resulting from plant Special Report that effluents of identifies the cause(s) for environmental sampling exceeding the limits and medium at a specified defines the corrective location in excess of actions to be taken to reporting limits of Table reduce radioactive 16.11-13-2 when effluents so that the averaged over any potential annual dose to a calendar quarter. MEMBER OF THE PUBLIC is less than the calendar year limits of SLC 16.11-3, SLC 16.11-8, and SLC 16.11-9.

(continued)

Catawba Units 1 and 2 16.11-13-1 Revision 0

Monitoring Program 16.11-13 REMEDIAL ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C. Milk or fresh leafy C. 1 --------- NOTE-------

vegetation samples Specific location(s) from unavailable from one or which samples were more sample location(s) unavailable may be deleted required by Table 16.11- from the program.

13-1.

Revise the Radiological 30 days Environmental Monitoring Program to identify location(s) for obtaining replacement samples.

AND C.2 Identify the cause of the In the next unavailability of samples scheduled Annual and identify and justify new Radioactive Effluent location(s) for obtaining Release Report replacement samples in pursuant to the Annual Radioactive Technical Effluent Release Report Specification 5.5.1 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 ------------------ NOTE ------------------

The maximum values for the lower limits~of detection shall be as specified in Table 16.11-13-3.

Collect and analyze radiological environmental In accordance with monitoring samples pursuant to Table 16.11-13-1 from Table 16.11-13-1 the specific locations given in the table and figure(s) in the ODCM.

Catawba Units 1 and 2 16.11-13-2 Revision 0

Monitoring Program 16.11-13 Table 16.11-13-1 Radiological Environmental Monitoring Program (page 1 of 7)

EXPOSURE PATHWAY AND/OR NUMBER OF REPRESENTATIVE SAMPLING AND COLLECTION TYPE AND FREQUENCY OF SAMPLE SAMPLES AND SAMPLE FREQUENCY ANALYSIS LOCATIONS"1 '

1. Direct Radiation(2) Forty routine monitoring stations Quarterly Gamma dose quarterly either with 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 to be placed in special interest areas such as population centers, nearby residences, schools, and in one or two areas to serve as control stations.

(continued)

Catawba Units 1 and 2 16.11-13-3 Revision 0

Monitoring Program 16.11-13 Table 16.11-13-1 Radiological Environmental Monitoring Program (page 2 of 7)

EXPOSURE PATHWAY AND/OR NUMBER OF REPRESENTATIVE SAMPLING AND COLLECTION TYPE AND FREQUENCY OF SAMPLE SAMPLES AND SAMPLE FREQUENCY ANALYSIS LOCATIONS(1"

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

Particulates sample collection weekly, or more 1-131 analysis weekly.

frequently if required by dust Three samples from close to the loading.

three SITE BOUNDARY locations, Particulate Sampler:

in different sectors, of the highest Gross beta radioactivity calculated annual average ground- analysis following filter level D/Q; change; (3) and gamma isotopic analysis (4) of One sample from the vicinity of a composite (by location) community having the highest quarterly.

calculated annual average ground-level D/Q; and One sample from a control location, as for example 15 to 30 km distant and in the least prevalent wind direction.

(continued)

Catawba Units 1 and 2 16.11-13-4 Revision 0

Monitoring Program 16.11-13 Table 16.11-13-1 Radiological Environmental Monitoring Program (page 3 of 7)

EXPOSURE PATHWAY AND/OR NUMBER OF REPRESENTATIVE SAMPLING AND COLLECTION TYPE AND FREQUENCY OF SAMPLE SAMPLES AND SAMPLE FREQUENCY ANALYSIS LOCATIONS"1 )

3. Waterborne

a. Surface(5) One sample upstream. Composite sample over 1-month Gamma isotopic analysis (4)

One sample downstream. period (6) monthly. Composite for tritium analysis quarterly.

b. Ground Samples from one or two sources Quarterly Gamma isotopic (4) and tritium only if likely to be affected(7 ). analysis quarterly.

c. Drinking One sample of each of one to Composite sample over 2-week 1-131 analysis on each composite three of the nearest water supplies period (6)when 1-131 analysis is when the dose calculated for the that could be affected by its performed; monthly composite consumption of the water is discharge. otherwise. greater than 1 mrem per year(8 ).

Composite for gross beta and One sample from a control gamma isotopic analyses (4) location. monthly. Composite for tritium analysis quarterly.

d. Sediment from Shoreline One sample from downstream Semiannually Gamma isotopic analysis (4) area with existing or potential semiannually.

recreational value.

(continued)

Catawba Units 1 and 2 16.11-13-5 Revision 0

Monitoring Program 16.11-13 Table 16.11-13-1 Radiological Environmental Monitoring Program (page 4 of 7)

EXPOSURE PATHWAY AND/OR NUMBER OF REPRESENTATIVE SAMPLING AND COLLECTION TYPE AND FREQUENCY OF SAMPLE SAMPLES AND SAMPLE FREQUENCY ANALYSIS LOCATIONS"1 .

4. Ingestion

a. Milk Samples from milking animals in Semimonthly when animals are on Gamma isotopic (4) and 1-131 three locations within 5-km pasture; monthly at other times. analysis semi-monthly when distance having the highest dose animals are on pasture; monthly at potential. If there are none, then other times.

one sample from milking animals in each of three areas between 5 to 8 km distant where doses are calculated to be greater than 1 mrem per year(B). One sample from milking animals at a control location 15 to 30 km distant and in the least prevalent wind direction.

b. Fish and Invertebrates One sample each of a predatory Sample in season, or Gamma isotopic analysis (4) on species, a bottom feeder and a semiannually if they are not edible portions.

forage species in vicinity of plant seasonal.

discharge area.

One sample each of a predatory species, a bottom feeder and a forage species in areas not influenced by plant discharge.

(continued)

Catawba Units 1 and 2 16.11-13-6 Revision 0

Monitoring Program 16.11-13 Table 16.11-13-1 Radiological Environmental Monitoring Program (page 5 of 7)

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

4. Ingestion (Continued)

c. Food Products One sample of each principal At time of harvest (9). Gamma isotopic analyses (4) on class of food products from any edible portion.

area that is irrigated by water in which liquid plant wastes have been discharged.

Samples of three different kinds of Monthly, when available. Gamma isotopic (4) and 1-131 broad leaf vegetation grown analysis.

nearest each of two 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 Monthly, when available. Gamma isotopic (4) and 1-131 broad leaf vegetation grown 15 to analysis.

30 km distant in the least prevalent wind direction if milk sampling is not performed.

Catawba Units 1 and 2 16.11-13-7 Revision 0

Monitoring Program 16.11-13 Table 16.11-13-1 Radiological Environmental Monitoring Program (page 6 of 7)

NOTES:

(1) Specific parameters of distance and direction sector from the centerline of the station, 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 circumstances such as hazardous conditions, seasonal unavailability, and malfunction of automatic sampling equipment. If specimens are unobtainable due to sampling equipment malfunction, 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 pursuant to Technical Specification 5.6.2. It is recognized that, at times, it may not be possible or practicable 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 any Licensee Event Report required by 10 CFR 50.73 and pursuant to Technical Specification 5.6.3, identify the cause of the unavailability of samples for that pathway and identify the new location(s) for obtaining replacement samples in the next 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 40 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 sectors 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 within minimal fading.)

(3) 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 10 times the yearly mean of control samples, gamma isotopic analysis shall be performed on the individual samples.

Catawba Units 1 and 2 16.11-13-8 Revision 0

Monitoring Program 16.11-13 Table 16.11-13-1 Radiological Environmental Monitoring Program (page 7 of 7)

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

(5) 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.

(6) A composite sample is one in which the rate at which the liquid sampled is uniform and in which the method of sampling employed results in a specimen that is representative of the time-averaged concentration at the location being sampled. 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.

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

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

(9) 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 tuberous and root food products.

Catawba Units 1 and 2 16.11-13-9 Revision 0

Monitoring Program 16.11-13 Table 16.11-13-2 Reporting Levels for Radioactivity Concentrations in Environmental Samples ANALYSIS WATER AIRBORNE FISH MILK FOOD PRODUCTS (pCi/I) PARTICULATE OR (pCi/kg, wet) (pCi/I) (pCi/kg, wet)

GASES (pCi/m 3)

H-3 20,000(l)

Mn-54 1,000 30,000 Fe-59 400 10,000 Co-58 1,000 30,000 Co-60 300 10,000 Zn-65 300 20,000 Zr-Nb-95 400 1-131 2 0.9 3 100 Cs-134 30 10 1,000 60 1,000 Cs-137 50 20 2,000 70 2,000 Ba-La-140 200 300 (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.

Catawba Units 1 and 2 16.11-13-10 Revision 0

Monitoring Program 16.11-13 Table 16.11-13-3 Lower Limit of Detection (LLD)(3) (page 1 of 3)

ANALYSIS WATER AIRBORNE FISH MILK FOOD SEDIMENT (pCi/I) PARTICULATE (pCi/kg, wet) (pCi/I) PRODUCTS (pCi/kg, dry)

OR GASES (pCi/kg, wet)

(pCi/m 3)

Gross Beta 4 0.01 H-3 2000"51 Mn-54 15 130 Fe-59 30 260 Co-58, 60 15 130 Zn-65 30 260 Zr-Nb-95 15 1-131 1(4) 0.07 1 60 Cs-134 15 0.05 130 15 60 150 Cs-137 18 0.06 150 18 80 180 Ba-La-140 15 15 Catawba Units 1 and 2 16.11-13-11 Revision 0

Monitoring Program 16.11-13 Table 16.11-13-3 Lower Limit of Detection (LLD)(3) (page 2 of 3)

NOTES:

(1) 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 pursuant to Technical Specification 5.6.2.

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

(3) The LLD is defined, for purposes of these commitments, as the smallest concentrations 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 4.66 s b E V .2.22 Y exp (-2 t)

Where:

LLD = the "a priori" lower limit of detection (picoCuries 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 (counts per minute);

E = the counting efficiency (counts per disintegration);

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

2.22 = the number of disintegrations per minute per picoCurie; Y = the fractional radiochemical yield, when applicable;

= the radioactive decay constant for the particular radionuclide (sec-1 ); and At = the elapsed time between environmental collection, or end of the sample collection period, and time of counting (sec).

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

Catawba Units 1 and 2 16.11-13-12 Revision 0

Monitoring Program 16.11-13 Table 16.11-13-3 Lower Limit of Detection (LLD)(3) (page 3 of 3)

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 pursuant to Technical Specification 5.6.2.

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

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

Catawba Units 1 and 2 16.11-13-13 Revision 0

Monitoring Program 16.11-13 BASES The Radiological Environmental Monitoring Program required by this SLC 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 plant 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 riori (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 16.11-13-2 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 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 SLC 16.11-3, SLC 16.11-8, and SLC 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 reporting level (1) reporting 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 from all radionuclides is equal to or greater than the calendar year limits of SLC 16.11-3, SLC 16.11-8, and SLC 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 required by Technical Specification 5.6.2. The methodology and parameters used to estimate the potential annual dose to a MEMBER OF THE PUBLIC shall be indicated in the 30-day Special Report.

Catawba Units 1 and 2 16.11-13-14 Revision 0

Monitoring Program 16.11-13 BASES (continued)

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. Catawba Offsite Dose Calculation Manual.

2. 10 CFR Part 50, Appendix I.

Catawba Units 1 and 2 16.11-13-15 Revision 0

Land Use Census 16.11-14 16.11 RADIOLOGICAL EFFLUENTS CONTROLS 16.11-14 Land Use Census COMMITMENT -----------------------

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

Broad leaf vegetation sampling of at least 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. Requirements for broad leaf vegetation sampling in Table 16.11-13-1 (Item 4c) shall be followed, including analysis of control samples.

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 the nearest milk animal, the nearest residence, and the nearest garden of > 50 m2 (500 ft 2) producing broad leaf vegetation.

APPLICABILITY: At all times.

REMEDIAL ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Location(s) identified A. 1 Identify the new location(s) In the next which yield a calculated in the Annual Radioactive scheduled Annual dose or dose Effluent Release Report. Radioactive Effluent commitment greater Release Report than values currently pursuant to calculated in SLC 16.11- Technical

9. Specification 5.6.3 B. Location(s) identified B.1 Add the new location(s) to 30 days which yield a calculated the Radiological dose or dose Environmental Monitoring commitment (via same Program.

exposure pathway) 20%

greater than at a AND location from which samples are currently being obtained in accordance with SLC 16.11-13.

(continued)

Catawba Units 1 and 2 16.11-14-1 Revision 0

Land Use Census 16.11-14 REMEDIAL ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B. (continued) B.2 Identify the new In the next location(s), revised scheduled Annual figure(s) and table(s) for Radioactive Effluent the ODCM, and Release Report information supporting the pursuant to change in sampling Technical location(s) in the Annual Specification 5.5.1 Radioactive Effluent Release Report.

TESTING REQUIREMENTS TEST FREQUENCY TR 16.11-14-1 ------------------ NOTE -----------------------

The results of the Land Use Census shall be included in the Annual Radiological Environmental Operating Report pursuant to Technical Specification 5.6.2.

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.

BASES This SLC 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 given in the ODCM 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 Appendix I to 10 CFR Part 50. Restricting the census to gardens of greater than 50 m2 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 quantify (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 kg/Im2.

Catawba Units 1 and 2 16.11-14-2 Revision 0

Land Use Census 16.11-14 BASES (continued)

With a Land Use Census identifying a location(s) which yield 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(s) within 30 days to the Radiological Environmental Monitoring Program given in the ODCM. 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. Catawba Offsite Dose Calculation Manual.

2. 10 CFR Part 50, Appendix I.

Catawba Units 1 and 2 16.11-14-3 Revision 0

Interlaboratory Comparison Program 16.11-15 16.11 RADIOLOGICAL EFFLUENTS CONTROLS 16.11-15 Interlaboratory Comparison Program COMMITMENT Analyses shall be performed on all radioactive materials, supplied as part of an Interlaboratory Comparison Program, 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 the next performed as required. taken to prevent scheduled Annual recurrence in the Annual Radiological Radiological Environmental Environmental Operating Report. Operating Report pursuant to Technical Specification 5.6.2 TESTING REQUIREMENTS TEST FREQUENCY TR 16.11-15-1 Report a summary of the results of the Interlaboratory In the Annual Comparison Program in the Annual Radiological Radiological Environmental Operating Report. Environmental Operating Report pursuant to Technical Specification 5.6.2 BASES The requirement for participation in an Interlaboratory Comparison Program is provided to ensure that independent checks on the precision and accuracy of the measurements of radioactive material in environmental sample matrices are performed as part of the quality assurance program for environmental monitoring in order to demonstrate that the results are valid for the purposes of Section IV.B.2 of Appendix I to 10 CFR Part 50.

Catawba Units 1 and 2 16.11-15-1 Revision 0

Interlaboratory Comparison Program 16.11-15 BASES (continued)

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

REFERENCES 1. 10 CFR Part 50, Appendix I.

Catawba Units 1 and 2 16.11-15-2 Revision 0

Annual Radiological Environmental Operating Report And Radioactive Effluent Release Report 16.11-16 16.11 RADIOLOGICAL EFFLUENTS CONTROLS 16.11-16 Annual Radiological Environmental Operating Report and Radioactive Effluent Release Report COMMITMENT Annual Radiological Environmental Operating Report Routine Annual Radiological Environmental Operating Reports covering the operation of the unit during the previous calendar year shall be submitted prior to 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 preoperational 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 the Land Use Census.

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; at least two legible maps (one map shall cover stations near the SITE BOUNDARY, and a second map shall include the more distant stations) covering all sampling locations keyed to a table giving distances and directions from the centerline of one reactor; the results of licensee participation in the Interlaboratory Comparison Program, required by SLC 16.11-15; discussion of all deviations from the sampling schedule of Table 16.11-13-1; and 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 the station.

(continued)

Catawba Units 1 and 2 16.11-16-1 Revision 1

Annual Radiological Environmental Operating Report And Radioactive Effluent Release Report 16.11-16 COMMITMENT (continued)

Radioactive Effluent Release Report The Radioactive Effluent Release Report covering the operation of the unit during the previous calendar year shall be submitted before May I 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.

The Radioactive Effluent Release Report shall include an annual summary of hourly meteorological data collected over the previous year. This annual summary may be either in the form of an hour-by-hour listing 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. (In lieu of submission with the Radioactive Effluent Release Report, the licensee has the option of retaining this summary of required meteorological data on site in a file that shall be provided to the NRC upon request.) 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 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.

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

(continued)

Catawba Units 1 and 2 16.11-16-2 Revision 1

Annual Radiological Environmental Operating Report And Radioactive Effluent Release Report 16.11-16 COMMITMENT (continued)

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 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.

A single submittal may be made for the station. The submittal should combine those sections that are common to both units.

APPLICABILITY: At all times.

REMEDIAL ACTIONS None TESTING REQUIREMENTS None BASES None REFERENCES None Catawba Units 1 and 2 16.11-16-3 Revision 1

Annual Radiological Environmental Operating Report And Radioactive Effluent Release Report 16.11-16

-V..

Figure 16.11-16-1 UNRESTRICTED AREA and SITE BOUNDARY for Radioactive Effluents Catawba Units 1 and 2 16.11-16-4 Revision 1

Liquid Holdup Tanks 16.11-17 16.11 RADIOLOGICAL EFFLUENTS CONTROLS 16.11-17 Liquid Holdup Tanks COMMITMENT The quantity of radioactive material contained in each temporary unprotected outdoor 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(s) radioactive material to the exceeding limit, tank(s).

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

AND A.3 Describe the events In the next leading to this condition in scheduled the Radioactive Effluent Radioactive Effluent Release Report. Release Report pursuant to Technical Specification 5.6.3 TESTING REQUIREMENTS TEST FREQUENCY TR 16.11-17-1 Verify that the quantity of radioactive material contained 7 days in each tank is within limits by analyzing a representative sample of the tank(s) contents when radioactive materials are being added to the tank(s).

Catawba Units 1 and 2 16.11-17-1 Revision 0

Liquid Holdup Tanks 16.11-17 BASES The tanks included in this SLC are 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 tank's 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 1. Letter from NRC to Gary R. Peterson, Duke, Issuance of Improved Technical Specifications Amendments for Catawba, September 30, 1998.

2. Technical Specification 5.5.12, Explosive Gas and Storage Tank Radioactivity Monitoring Program.

Catawba Units 1 and 2 16.11-17-2 Revision 0

Explosive Gas Mixture 16.11-18 16.11 RADIOLOGICAL EFFLUENTS CONTROLS 16.11-18 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 is > 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 and hydrogen concentration > 4% by volume.

B. Concentration of oxygen B.1 Suspend all additions of Immediately in the WASTE GAS waste gases to the system.

HOLDUP SYSTEM >

4% by volume and AND hydrogen 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.

Catawba Units 1 and 2 16.11-18-1 Revision 1

Explosive Gas Mixture 16.11-18 TESTING REQUIREMENTS TEST FREQUENCY TR 16.11-18-1 Verify that the concentrations of hydrogen and oxygen in During WASTE the WASTE GAS HOLDUP SYSTEM are within limits by GAS HOLDUP continuously monitoring the waste gases in the WASTE SYSTEM GAS HOLDUP SYSTEM with the hydrogen and oxygen operation monitors required FUNCTIONAL by SLC 16.11-20.

BASES This SLC 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 1. Letter from NRC to Gary R. Peterson, Duke, Issuance of Improved Technical Specifications Amendments for Catawba, September 30, 1998.

2. Technical Specification 5.5.12, Explosive Gas and Storage Tank Radioactivity Monitoring Program.

Catawba Units 1 and 2 16.11-18-2 Revision 1

Gas Storage Tanks 16.11-19 16.11 RADIOLOGICAL EFFLUENTS CONTROLS 16.11-19 Gas Storage Tanks COMMITMENT The quantity of radioactivity contained in each gas storage tank shall be limited to < 97,000 Curies of noble gases (considered as Xe-133 equivalent).

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(s) radioactive material to the exceeding limit, tank(s).

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

AND A.3 Describe the events In the next leading to this condition in scheduled the Radioactive Effluent Radioactive Effluent Release Report. Release Report pursuant to Technical Specification 5.6.3 TESTING REQUIREMENTS TEST FREQUENCY TR 16.11-19-1 Verify that the quantity of radioactive material contained 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months in each tank is within limits when radioactive materials are being added to the tank(s).

Catawba Units 1 and 2 16.11-19-1 Revision 0

Gas Storage Tanks 16.11-19 BASES The tanks included in this SLC are those tanks for which the quantity of radioactivity contained is not limited directly or indirectly by another SLC.

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 whole body exposure to a MEMBER OF THE PUBLIC at the nearest SITE 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 1. Letter from NRC to Gary R. Peterson, Duke, Issuance of Improved Technical Specifications Amendments for Catawba, September 30, 1998.

2. Technical Specification 5.5.12, Explosive Gas and Storage Tank Radioactivity Monitoring Program.

Catawba Units 1 and 2 16.11-19-2 Revision 0

Explosive Gas Monitoring Instrumentation 16.11-20 16.11 RADIOLOGICAL EFFLUENTS CONTROLS 16.11-20 Explosive Gas Monitoring Instrumentation COMMITMENT The Explosive Gas Monitoring Instrumentation channels shown in Table 16.11-20-1 shall be FUNCTIONAL with their Alarm/Trip Setpoints set to ensure that the limits of SLC 16.11-18 are not exceeded.

APPLICABILITY: During WASTE GAS HOLDUP SYSTEM operation.

REMEDIAL ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required A.1 Declare the channel(s) Immediately Explosive Gas non-functional.

Monitoring "

Instrumentation channel(s) Alarm/Trip Setpoint less conservative than required.

B. One required hydrogen B.1 Suspend oxygen supply to Immediately monitor channel non- the recombiner.

functional.

AND B.2 Restore channel to 30 days FUNCTIONAL status.

C. One required oxygen C.1 Obtain and analyze grab 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months monitor channel non- samples.

functional.

AND C.2 Restore channel to 30 days FUNCTIONAL status.

(continued)

Catawba Units 1 and 2 16.11-20-1 Revision 1

Explosive Gas Monitoring Instrumentation 16.11-20 REMEDIAL ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME D. Two required oxygen D.1 Obtain and analyze grab Once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months monitor channels non- samples. during degassing functional. operations AND Once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months during other operations AND D.2 Restore channels to 30 days FUNCTIONAL status.

E. Required Action and E.I1: Prepare and submit a 30 days associated Completion Special Report to the NRC Time of Condition B, C, to explain why the non-or D not met. functionality was not corrected within the time specified.

TESTING REQUIREMENTS

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

Refer to Table 16.11-20-1 to determine which TRs apply for each Explosive Gas Monitoring Instrumentation channel.

TEST FREQUENCY TR 16.11-20-1 Perform CHANNEL CHECK. 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months TR 16.11-20-2Perform COT. 31 days (continued)

Catawba Units 1 and 2 16.11-20-2 Revision 1

Explosive Gas Monitoring Instrumentation 16.11-20 TESTING REQUIREMENTS (continued)

TEST FREQUENCY TR 16.11-20-3 ------------------ NOTE ------------------

The CHANNEL CALIBRATION shall include the use of standard gas samples in accordance with the manufacturer's recommendations. In addition, a standard gas sample of nominal four volume percent hydrogen (for the hydrogen monitors) and four volume percent oxygen (for the oxygen monitors), with the balance nitrogen, shall be used in the calibration to check linearity of the analyzer.

Perform CHANNEL CALIBRATION. 92 days Catawba Units 1 and 2 16.11-20-3 Revision 1

Explosive Gas Monitoring Instrumentation 16.11-20 Table 16.11-20-1 Explosive Gas Monitoring Instrumentation INSTRUMENT REQUIRED TESTING CHANNELS REQUIREMENTS WASTE GAS HOLDUP SYSTEM Explosive Gas Monitoring Instrumentation

1. Hydrogen Monitors 1/inservice train TR 16.11-20-1 per station TR 16.11-20-2 TR 16.11-20-3

2. Oxygen Monitors 2/inservice train TR 16.11-20-1 per station TR 16.11-20-2 TR 16.11-20-3 Catawba Units 1 and 2 16.11-20-4 Revision 1

Explosive Gas Monitoring Instrumentation 16.11-20 BASES The Explosive Gas Monitoring Instrumentation is provided for monitoring and controlling the concentrations of potentially explosive gas mixtures in the WASTE GAS HOLDUP SYSTEM.

REFERENCES 1. Letter from NRC to Gary R. Peterson, Duke, Issuance of Improved Technical Specifications Amendments for Catawba, September 30, 1998.

Catawba Units 1 and 2 16.11-20-5 Revision 1

Major Changes to Liquid, Gaseous, and Solid Radwaste Treatment Systems 16.11-21 16.11 RADIOLOGICAL EFFLUENTS CONTROLS 16.11-21 Major Changes to Liquid, Gaseous, and Solid Radwaste Treatment Systems COMMITMENT Licensee-initiated major changes to the Radwaste Treatment Systems (liquid, gaseous, and solid):

1. Shall be reported to the NRC in the Radioactive Effluent Release Report for the period in which the evaluation was reviewed by the Station Manager. Licensees may choose to submit the information called for in this SLC as part of the periodic Updated Final Safety Analysis Report 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 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;

e. An evaluation of the change, which shows the expected maximum exposures to a MEMBER OF THE PUBLIC 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 (continued)

Catawba Units 1 and 2 16.11-21-1 Revision 0

Major Changes to Liquid, Gaseous, and Solid Radwaste Treatment Systems 16.11-21 COMMITMENT (continued)

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

2. Shall become effective upon review and acceptance by a qualified individual/organization.

APPLICABILITY: At all times.

REMEDIAL ACTIONS None TESTING REQUIREMENTS None BASES None REFERENCES 1. Letter from NRC to Gary R. Peterson, Duke, Issuance of Improved Technical Specifications Amendments for Catawba, September 30, 1998.

Catawba Units 1 and 2 16.11-21-2 Revision 0

ATTACHMENT 8 Revisions to the Radioactive Waste Process Control Program Manual

The following letter dated April 7, 2012, from David L.Vaught, Senior Engineer, Nuclear Chemistry, summarizes how the Process Control Program (PCP) manual has been revised. The updated version of the manual contains all the changes implemented during 2011 and is designated as the "2011 Report Year" on the enclosed Compact Disc.

Duke Energy Radioactive Waste Process Control Program Manual Summary of 2011 Changes April 7, 2012 RD Hart Regulatory Compliance Manager Catawba Regulatory Compliance ATTENTION: MJ Sawicki

SUBJECT:

Catawba Nuclear Station 2011 Annual Radioactive Effluent Release Report Process Control Program Changes File: GS-764.25, CN-215.06 Enclosed are CD copies of the PDF file of the Radioactive Waste Process Control Program Manual to be included in the NRC distribution of the Annual Radioactive Effluent Release Report for Catawba Nuclear Station for the period of January 1, 2011 through December 31, 2011. This version of the Manual contains all the changes implemented during 2011 and is designated on the CD cover as the "2011 Report Year".

The PCP Manual is revised using the review and approval process in APPENDIX F of the PCP Manual, "Administration of the PCP and Support Documents" prior to publication on the NEDL Portal.

The attachment summarizes the scope of the changes during 2011 .A more detailed summary of changes and basis are in the Appendix H of the PCP Manual.

The PDF file "DPCo-2012-PCP-Manual.pdf' on the CDs was reviewed and verified against the control copies of the PCP Manual published on the NEDL Portal. Three CD copies are for internal distribution and DHEC and four CDs are for the NRC as follows:

Duke

1. ELL- CD

2. Master File - CD DHEC

3. DHEC primary contact Russell Keown - CD NRC

4. NRC Document Control Desk

5. Catawba NRC Project Manager

6. Catawba Senior Resident Inspector

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

Larry A Wilson Supervising Scientist Nuclear Chemistry by: David L Vaught Senior Engineer Nuclear Chemistry - Radwaste ATTACHMENT

ATTACHMENT Duke Energy Radioactive Waste Process Control Program Manual Summary of 2011 Changes A brief summary of the 2011 changes to the Duke Energy Radioactive Waste PCP Manual is found below. These are described in more detail in APPENDIX H "Revision Summary -

Licensee Initiated Changes" Appendix B "MNS PCP" Rev 18 (Issued 1/30/08) Minor Change (6/15/11)

Minor Change: Added procedure OPI/0B162001083 "WM Process Skid Demineralizers Sluice, Load, and Maintenance" as a PCP implementing procedure. This is a minor change per the definition in Appendix F:

ADMINISTRATION OF THE PCP AND SUPPORT DOCUMENTS Section 5.1.2.2 for the following reasons:

The change does not alter the scope, results, requirements, or methods by which the dewatering or solidification process as described in the Corporate PCP is performed.

The change does not alter the responsibilities of site personnel in meeting the PCP requirements.

" The change does not alter a PCP QA approved provider dewatering or solidification process,

responsibilities for fulfilling PCP requirements or the vendor interface with the station personnel or work processes.

The procedure OP/O/B/6200/083 is not a new procedure, its scope and use in the PCP dewatering process has not changed.

CHANGE BASIS: MNS PIP M-10-06651 INOS Audit 10-15(INOS)(REC)(MNS) 2010 Radiological Effluent Control Program Audit

The procedure OP/0/B/6200/083 is used in combination with Procedure OP10OB162001064 "Transfer and Dewatering of Media" to perform PCP related activities. Procedure P/0/B/6200/064 contains the final dewatering for disposal activities; however, Procedure OP/0/B/6200/083 is identified as a PCP implementing procedure in the Purpose statement of the procedure because it includes instructions (noted below) that important to the dewatering process.

o the requirement for using the two bottom laterals (L&P # 2.26 "Dewatering Legs #1 and

ý#2 shall be used to dewater mixed media.")

o the requirement for the strainer in the dewatering pump discharge: (L&P # 2.29 "Y-strainer or similar device shall be installed in discharge path of dewatering pump to catch media in case internal filter fails.")

ATTACHMENT 9 Information to Support the Nuclear Energy Institute (NEI)

Groundwater Protection Initiative

2011 ARERR Groundwater Well Data Section Duke Energy implemented a Ground Water Protection program in 2007. This initiative was developed to ensure timely and effective management of situations involving inadvertent releases of licensed material to ground water. As part of this program, Catawba Nuclear Station monitored forty-six wells in 2011.

Wells are currently sampled quarterly or semi-annually. Ground water samples are regularly analyzed for tritium and gamma emitters, select 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 2011.

Results from sampling during 2011 confirmed existing knowledge of tritium concentrations in site ground water (shown in the table below).

Results from sampling during 2011 are shown in the table below.

Avg. Tritium Conc. # of Well Name Well Location Conc.(pCi/1) Range Samples ClOOR U-1 SFP * * *0 C100DR U-1 SFP < < 4 C1OIR U-1 SFP 843 744-1,090 4 C1O1DR U-1 SFP 521 464-608 4 C102 E of U1 SFP O/S 689 624-747 4 protected area C103 E of U1 SFP @ Cooling 769 666 - 836 4 Towers C104 U-1 RMWST 709 685-752 4 C105 Engr. Bldg. 518 255-636 4 C105R Engr. Bldg. 736 632 - 799 4 C106 W Parking Lot < < 4 C106R W Parking Lot < < - 218 4 C107 MET Tower Hill 689 583 - 820 4 C200R U-2 SFP 859 800-973 3 C200DR U-2 SFP 476 434 - 505 3 C201R U-2 SFP 2,940 2,230 - 3,960 4 C201DR U-2 SFP 649 408 - 926 4 C202 S of RMC Tent 755 684-826 2 C203 East of RMC tent @ 581 492 - 640 4 Cooling Towers C204 S of RMC Tent 165 <- 165 2 C205 Adm. Parking < < 4 C205R Adm Parking 188 <- 188 4 C206 W Parking Lot < < 4 C207 Mon. Tank B 697 623-835 4 C207R Mon. Tank B 276 < - 316 4 C208 N of MTB 270 < - 342 4

2011 ARERR Groundwater Well Data Section C209 MTUville S of light pole 189 <- 189 4 23A C210 N of U2 Mech Equip Bldg 236 < - 271 1 C211 West of RL intake O/S 851 617-1,270 4 protected area C212 Behind Aquatic Center 237 < - 237 4 C213 Mon. Tank B 20,250 16,900 - 23,400 4 C213R Mon. Tank B 263 < - 263 4 C214 N of U2 TB 1,104 919-1,440 3 C215 N of U2 TB 2,304 647-5,580 4 C217 N of U2 TB 768 698-875 4 C218 N of U2 TB 2,148 621 -3,860 4 C220 N of U2 TB 8,080 6,920 - 9,260 4 C221 N of U2 TB 489 436-585 4 WCMW-2 WC Ponds 2,623 2,510 - 2,850 4 WCMW-3 WC Ponds 788 616- 898 4 WCMW-4 WC Ponds 502 410- 700 4 WCMW-5 WC Ponds 260 < - 311 4 LMW 2A Landfill < < 2 LMW 3A Landfill < < 2 LMW 4 Landfill < < 2 LMW 5S Landfill < < 2 LMW 5D Landfill < < 2

.No sample available due to insufficient volume.

pCi/1 - pico curies per liter

< - 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 10CFR20, Appendix B, Table 2, Column 2, Effluent Concentration limit for tritium.

ATTACHMENT 10 Inoperable Equipment

Inoperable Monitoring Equipment Report from 1/1/2011 to 3/24/12 per SLC 16.11-2 and 16.11-7 SLC # from Title Completion Determination and Data Reviewed Table 16.11- Time 2-1 1.c EMF 45 A & B 30 Days For 1EMF45B, maximum out of service time is 1 113.87 days (TSAIL C1-11-01403).

1EMF45B was entered into TSAIL 5/12/2011 for tracking only. 1EMF45B was experiencing nuisance 'Loss of Sample Flow' alarms when there was no RN flow through the Containment Spray Heat Exchangers. The circuit is designed to block this alarm if either 1RN-225B or 1RN-229B is closed with power on the valve in order to prevent nuisance alarms. 1EMF45B was left as a TSAIL entry as tracking, because the EMF channel check cannot be met in the Mode Surveillance when in this alignment (1RN-229B closed/power on + 1RN225B NOT closed/power on). This deficiency did not in any way prevent the alarm from coming in if an open flow path through the Containment Spray Heat Exchangers is aligned and the EMF loses sample flow. 1EMF45B was removed from TSAIL when the following Work Orders (WO) were closed -

01983259 - 1EMF 45B; TROUBLE ALARM RECEIVED WITH 1EMF-45B TROUBLE ALARM

01983272 - 1EMF-45B; LOSS OF FLOW ALARM ACTIVE W/RN FLOW PATH ISOLATED

01034573 - 1EMF45 LOSS OF FLOW ALARM ACTIVE WITH RN FLOW PATH ISOLATED.

SLC # from Title Completion Determination and Data Reviewed Table 16.11- Time 7-1 1.a EMF 50 (L) 14 Days OEMF50(L), out of service times are 433.95 days (1/18/10 to 3/28/11 TSAIL CO-10-00146) and 352.26 days (4/7/11 to 3/24/12 TSAIL CO-11-01041).

OEMF50 was non-functional from 1/18/2010 to 3/28/2010 due to issues associated with C-14 interference in the channel count rate. OEM F50 was removed from TSAIL when the following were closed -

" PIP C-09-2848 - Carbon-14 not properly being discriminated

R&R 09-1943 - CE101834 Replace 1WG-309 with DSV-100

R&R 11-108 - Perform CC on EMF50L & EMF50H

R&R 11-728 - OEMF - PERFORM CC ON EMF50L & EMF50H.

WO 01893175 - OEMF - ACOT-EMF 50

WO 01909564 - OEMF50: PRIMARY CALIBRATION IMPLEMENTATION

WO 01933547 - OEMF - ACOT-EMF 50

WO 01953635 - OEMF - PERFORM CC ON EMF50L & EMF50H.

OEMF50 was non-functional from 4/7/11 to present date due to issues associated correlation of OEMF5O actual reading to expected reading based on sample activity. OEMF50 will be removed from TSAIL when the following are all closed -

PIP C-11-6849 - OEMF50L (WG Discharge Monitor) did not respond as expected

Procedure OP/O/A/6500/080 - Enclosure 4.4 - Removing Inoperable EMF RP86A From Alarm Circuitry

WO 01913784 - OEMFRT50(H):REPLACE RP-86 DISPLAY

WO 02002300 - OEMF50: TEST EMF50 FOR DIFFERENT HV AND DISC VALUES

WO 01054681 - OEMF RT SOL: CORRECT REASON FOR CABINET TROUBLE ALARMS Page 1 of 2

Inoperable Monitoring Equipment Report from 1/1/2011 to 3/24/12 per SLC 16.11-2 and 16.11-7 SLC # from Title Completion Determinationand DataReviewed Table 16.11- Time 7-1 3.a and 3.e EMF 36 and Unit Vent 30 Days For 1EMF36, maximum out of service time is Radiation Monitor Flow 41.87 days (TSAIL C1-11-00561).

Meter For 2EMF36, maximum out of service time is 34.03 days (TSAIL C2-11-00562).

1EMF36 and associated Radiation Monitor Flow Meter were listed in TSAIL from 2/22/2011 to 4/5/2011 due to issues associated with testing not being performed within the specified frequency. The Channel Operability Test (COT) was not fully checking the control actions of the EMF. A risk assessment documented in PIP C-11-1401 concluded that 1EMF 36 remained fully functional. 1EMF36 was removed from TSAIL when the following were closed -

PIP C-11-1401 - 1,2 EMF 36 SLC Testing Req. Not Satisfied

" WO 01972665 - 1EMF - ACOT-EMF 36 2EMF36 and associated Radiation Monitor Flow Meter were listed in TSAIL from 2/22/2011 to 3/28/2011 due to issues associated with testing not being performed within the specified frequency. The Channel Operability Test (COT) was not fully checking the control actions of the EMF. A risk assessment documented in PIP C-11-1401 concluded that 2EMF 36 remained fully functional. 2EMF36 was removed from TSAIL when the following were closed -

PIP C-11-1401 - 1,2 EMF 36 SLC Testing Req. Not Satisfied

WO 01973132 - PERFORM COT ON 2EMF36L Page 2 of 2

Addendum to Inoperable Monitoring Equipment Report from 1/1/2011 to 3/24/12 per SLC 16.11-2 and 16.11-7 SLC # from Title Completion Determination and Data Reviewed Table 16.11- Time 2-1 1.c EMF 45 A & B 30 Days For 1EMF45A(L), maximum out of service time is 32.2 days (TSAIL C1-10-03227).

For 1EMF45B(L), maximum out of service time is 113.87 days (TSAIL C1-11-01403).

Report information for 1EMF45B submitted with previous submittal approved by Technical manager on 4/1/12.

For 2EMF45A(L), maximum out of service time is 589.67 days (TSAIL C2-12-00093).

1EMF45A was entered into TSAIL 12/6/2010 due to loss of flow alarm issues. Work to repair and replace the flow transmitter along with channel and flow calibration work resulted in 1EMF45A being out of service for 32.2 days.

1EMF45A was returned to service on 1/7/2011 when the following Work Orders (WO) were closed -

" WO 01926068 - 1EMF TI ATC15/FB: I/R NO LOSS OF FLOW TIME DELAY 1EMF45A

WO 01936878 - 1EMF45A - PERFORM CHANNEL AND FLOW CALIBRATION

WO 01961966 - EC> 105239: REPLACE 1RNFT4980

WR 01024148 - 1RNFT4980:I/R 1EMF45A FLOW TRANSMITTER.

2EMF45A was entered into TSAIL 1/10/2012 for channel and flow calibration work. During this work, it was discovered that EMF detector was not fully seated in the detector well. 2EMF45A was determined to have been non-functional from 6/2/2010, when work was last performed on the EMF, and resulted in 2EMF45A being out of service for 589.67 days. As the 2010 and 2011 non-functionality of 2EMF45A was not identified until 1/11/12, 2EMF45A was not reported with the 2010 Inoperable Monitoring Equipment Report. 2EMF45A was returned to service on 1/12/2012 when Work Orders (WO) 01985050 was completed.

NOTE: Report information for 1EMF45B, EMF50(L), 1EMF36 and 2EMF36 submitted with previous submittal approved by Technical manager on 4/1/12.

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ATTACHMENT 11 Radioactive Waste Systems Changes

ATTACHMENT 11 Radioactive Waste Systems Changes There were no modifications to the Liquid Waste system performed during the calendar year 2011.

The Gaseous Waste System had no significant modifications performed during the calendar year 2011.

The only modifications performed were associated with parts obsolescence issues.

The Solid Waste Disposal system was changed to allow for more controlled venting of Spent Resin Storage Tanks. 1WS-11 and 1WS-32 were replaced with a valve and manual loader assembly. This is to allow for slowly opening the vent path without allow the depressurization to migrate contamination to other areas of the Auxiliary Building. The original plug valves were replaced with a globe type valves that are more capable of throttling flow. The overall intent of operation remains unchanged.

ML12135A595

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ML12135A595

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