0CAN041504, Annual Radioactive Effluent Release Report for 2014

Annual Radioactive Effluent Release Report for 2014

ML15118A553
Person / Time
Site:	Arkansas Nuclear
Issue date:	04/28/2015
From:	Pyle S Entergy Operations
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
0CAN041504
Download: ML15118A553 (228)

Text

s Entergy Operations, Inc.

1448 S.R. 333 Russellville, AR 72802 Tel 479-858-4704 Stephenie L. Pyle Manager, Regulatory Assurance Arkansas Nuclear One 0CAN041504 April 28, 2015 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555

SUBJECT:

Annual Radioactive Effluent Release Report for 2014 Arkansas Nuclear One, Units 1 and 2 Docket No. 50-313 and 50-368 License No. DPR-51 and NPF-6

Dear Sir or Madam:

Arkansas Nuclear One, Units 1 and 2 (ANO-1 and ANO-2) Technical Specifications (TSs) 5.6.3 and 6.6.3, respectively, require the submittal of an Annual Radioactive Effluent Release Report (ARERR). The information to fulfill this reporting requirement for ANO-1 and ANO-2 for the 2014 calendar year is enclosed.

ANO-1 TS 5.6.3 and ANO-2 TS 6.6.3 require this report to be submitted prior to May 1 of each year in accordance with 10 CFR 50.36a. 10 CFR 50.36a(a)(2) states that the interval between submittals for this report must not exceed 12 months.

Liquid and gaseous release data show that the dose from both ANO-1 and ANO-2 was considerably below the Offsite Dose Calculation Manual (ODCM) limits. The data reveals that radioactive effluents had an overall minimal dose contribution to the surrounding environment.

Pursuant to ANO-1 TS 5.5.1 and TS 6.5.1, the latest revision of the ODCM is submitted as an attachment to the ARERR.

As noted in the ARERR, the Process Control Program, EN-RW-105, was revised in 2014. A copy of the latest revision of this program is also attached, for information only.

No new commitments have been identified in this letter.

0CAN041504 Page 2 of 2 If you have any questions or require additional information, please contact me.

Sincerely, ORIGINAL SIGNED BY STEPHENIE L. PYLE SLP/rwc

Enclosure:

Annual Radioactive Effluent Release Report for 2014 Attachments: 1. Offsite Dose Calculation Manual

2. EN-RW-105, Process Control Program cc: Mr. Marc L. Dapas Regional Administrator U. S. Nuclear Regulatory Commission Region IV 1600 East Lamar Boulevard Arlington, TX 76011-4511 NRC Senior Resident Inspector Arkansas Nuclear One P. O. Box 310 London, AR 72847 U. S. Nuclear Regulatory Commission Attn: Ms. Andrea E. George MS O-8B1 One White Flint North 11555 Rockville Pike Rockville, MD 20852 Mr. Bernard R. Bevill Arkansas Department of Health Radiation Control Section 4815 West Markham Street Slot #30 Little Rock, AR 72205

Enclosure to 0CAN041504 Annual Radioactive Effluent Release Report for 2014

ANO-1 & 2 Radioactive Effluent Release Report for 2014 TABLE OF CONTENTS

1. INTRODUCTION.................................................................................................................. 2

2. REGULATORY LIMITS ........................................................................................................ 3

3.

SUMMARY

OF LIQUID EFFLUENT DATA .......................................................................... 5

4.

SUMMARY

OF GASEOUS EFFLUENT DATA .................................................................. 14

5.

SUMMARY

OF RADIATION DOSES ................................................................................. 23

6.

SUMMARY

OF DOSE TO MEMBERS OF THE PUBLIC .................................................. 25

7. HISTORICAL EFFLUENT DATA ....................................................................................... 26

8. SOLID WASTE

SUMMARY

............................................................................................... 43

9. UNPLANNED RELEASES ................................................................................................. 92

10. RADIATION INSTRUMENTATION .................................................................................... 92

11. CHANGES TO THE PROCESS CONTROL PROGRAM .................................................. 93

12. CHANGES TO THE OFFSITE DOSE CALCULATION MANUAL ...................................... 93

13. LLD LEVELS ...................................................................................................................... 93

14. RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM (REMP)....................... 93

15.

SUMMARY

OF HOURLY METEOROLOGICAL DATA ..................................................... 94

16. DESCRIPTION OF MAJOR CHANGES TO RADIOACTIVE WASTE SYSTEMS ............. 94

17. RADIOACTIVE GROUND WATER MONITORING PROGRAM DATA ............................. 94

18. C-14 REPORTING ............................................................................................................. 96 Page 1 of 106

ANO-1 & 2 Radioactive Effluent Release Report for 2014

1. INTRODUCTION Arkansas Nuclear One (ANO) is a two unit site consisting of a Babcock & Wilcox (Unit 1) and a Combustion Engineering (Unit 2) nuclear steam supply system. Both liquid and gaseous effluents are released in accordance with the Offsite Dose Calculation Manual (ODCM). This report is a summary of the effluent data in accordance with Unit 1 Technical Specification (TS) 5.6.3 and Unit 2 TS 6.6.3. This report provides the following information:

A. Routine radioactive effluent release reports covering the operation of the units during the reporting period.

B. Description of unplanned releases to unrestricted areas.

C. Description of changes to the ODCM.

D. Description of changes to the Process Control Program (PCP).

E. Summary of radiation doses due to radiological effluents during the previous calendar year.

F. Radiation dose to members of the public due to activities inside the site boundary.

G. Description of licensee initiated major changes to the radioactive waste systems during the previous calendar year.

H. Items to be reported in the Annual Radioactive Effluent Release Report (ARERR) per other miscellaneous ODCM requirements.

I. Applicable Radioactive Ground Water Monitoring Program data.

J. ARERR data for 2014 Solid Waste Shipments.

K. Carbon-14 release quantification details are discussed in Section 18.

Page 2 of 106

ANO-1 & 2 Radioactive Effluent Release Report for 2014 This report covers the period from January 1 through December 31, 2014.

2. REGULATORY LIMITS The ODCM contains the limits to which ANO must adhere. Because of the "as low as reasonably achievable" (ALARA) philosophy at ANO, actions are taken to reduce the amount of radiation released to the environment. Liquid and gaseous release data show that the dose from both Unit 1 and Unit 2 is considerably below the ODCM limits. This data reveals that the radioactive effluents have an overall minimal dose contribution to the surrounding environment.

The following are the limits required by the ODCM:

A. Gaseous Effluents

1. Dose rate due to radioactive materials released in gaseous effluent to unrestricted areas shall be limited to the following:

a. Noble gases Less than or equal to 500 mrem/year to the total body Less than or equal to 3000 mrem/year to the skin

b. Iodine-131, tritium, and for all radionuclides in particulate form with half lives greater than 8 days Less than or equal to 1500 mrem/yr to any organ

2. Dose - Noble Gases Quarterly Less than or equal to 5 mrads gamma Less than or equal to 10 mrads beta Yearly Less than or equal to 10 mrads gamma Less than or equal to 20 mrads beta

3. Dose - Iodine-131, Tritium, and Radionuclides in Particulate Form Quarterly Less than or equal to 7.5 mrem to any organ Yearly Less than or equal to 15 mrem to any organ Page 3 of 106

ANO-1 & 2 Radioactive Effluent Release Report for 2014 B. Liquid Effluents

1. Concentration The concentration of radioactive material released to the discharge canal shall be limited to the concentration specified in 10 CFR 20, Appendix B, Table II, Column 2, for radionuclides other than dissolved or entrained noble gases. For dissolved or entrained noble gases, the total concentration released shall be limited to 2E-4 microcuries/ml.

2. Dose Quarterly Less than or equal to 1.5 mrem total body Less than or equal to 5 mrem critical organ Yearly Less than or equal to 3 mrem total body Less than or equal to 10 mrem critical organ Page 4 of 106

ANO-1 & 2 Radioactive Effluent Release Report for 2014

3.

SUMMARY

OF LIQUID EFFLUENT DATA As required by Regulatory Guide 1.21, Revision 1, Radioactivity in Solid Wastes and Releases of Radioactive Materials in Liquid and Gaseous Effluents from Light-Water-Cooled Nuclear Power Plants, a summary of data for liquid releases is provided in the ARERR. The summary of liquid effluents for both Unit 1 and Unit 2 is as follows:

Unit 1 Batch Continuous Number of releases: 120 12 Total time for all releases (minutes): 25,000 110,000 Maximum time for a release (minutes): 1580 10,200 Average time for a release (minutes): 208 9130 Minimum time for a release (minutes): 21 2560 Unit 2 Batch Continuous Number of releases: 47 0 Total time for all releases (minutes): 15900 0 Maximum time for a release (minutes): 628 0 Average time for a release (minutes): 339 0 Minimum time for a release (minutes): 4 0 The Unit 1 liquid releases consisted of:

131 Planned Releases 1 Unplanned Releases The Unit 2 liquid releases consisted of:

47 Planned Releases 0 Unplanned Releases Page 5 of 106

ANO-1 & 2 Radioactive Effluent Release Report for 2014 ANNUAL SUMMATION FOR ALL RELEASES BY QUARTER (ALL LIQUID EFFLUENTS)

January 1 through June 30, 2014 Unit 1 Est. Total Type of Effluent Units Quarter 1 Quarter 2 Error %

A. Fission and Activation Products

1. Total Release (Not Including Curies 4.187E-03 4.683E-03 25 Tritium, Gases, Alpha)

2. Average Diluted Ci/ml 2.739E-12 1.267E-11 Concentration During Period

3. Percent of Applicable Limit  % 9.129E-04 4.223E-03 B. Tritium

1. Total Release Curies 4.673E+01 1.617E+02 25

2. Average Diluted Ci/ml 3.057E-08 4.373E-07 Concentration During Period

3. Percent of Applicable Limit  % 1.019E-03 1.458E-02 C. Dissolved and Entrained Gases

1. Total Release Curies 0.000E+00 3.092E-04 25

2. Average Diluted Ci/ml 0.000E+00 8.364E-13 Concentration During Period

3. Percent of Applicable Limit  % 0.000E+00 4.182E-07 D. Gross Alpha Radioactivity

1. Total Release Curies 0.000E+00 0.000E+00 25 E. Waste Vol Released (Pre-Dilution) Liters 4.785E+06 4.040E+06 25 F. Volume of Dilution Water Used Liters 1.529E+12 3.697E+11 25 Page 6 of 106

ANO-1 & 2 Radioactive Effluent Release Report for 2014 ANNUAL SUMMATION FOR ALL RELEASES BY QUARTER (ALL LIQUID EFFLUENTS)

July 1 through December 31, 2014 Unit 1 Est. Total Type of Effluent Units Quarter 3 Quarter 4 Error %

A. Fission and Activation Products

1. Total Release (Not Including Curies 9.369E-03 1.045E-02 25 Tritium, Gases, Alpha)

2. Average Diluted Ci/ml 2.414E-11 3.125E-11 Concentration During Period

3. Percent of Applicable Limit  % 8.046E-03 1.042E-02 B. Tritium

1. Total Release Curies 1.910E+02 1.823E+02 25

2. Average Diluted Ci/ml 4.922E-07 5.452E-07 Concentration During Period

3. Percent of Applicable Limit  % 1.641E-02 1.817E-02 C. Dissolved and Entrained Gases

1. Total Release Curies 8.209E-04 2.399E-03 25

2. Average Diluted Ci/ml 2.115E-12 7.175E-12 Concentration During Period

3. Percent of Applicable Limit  % 1.057E-06 3.587E-06 D. Gross Alpha Radioactivity

1. Total Release Curies 0.000E+00 0.000E+00 25 E. Waste Vol Released (Pre-Dilution) Liters 9.933E+06 5.752E+06 25 F. Volume of Dilution Water Used Liters 3.881E+11 3.343E+11 25 Page 7 of 106

ANO-1 & 2 Radioactive Effluent Release Report for 2014 UNIT 1 REPORT CATEGORY: ANNUAL LIQUID CONTINUOUS AND BATCH RELEASES TOTALS FOR EACH NUCLIDE RELEASED TYPE OF ACTIVITY: ALL RADIONUCLIDES REPORTING PERIOD: QUARTER # 1 AND QUARTER # 2 YEAR 2014 CONTINUOUS RELEASES BATCH RELESES NUCLIDE UNIT QUARTER 1 QUARTER 2 QUARTER 1 QUARTER 2 AG-110M CURIES 0.000E+00 0.000E+00 2.381E-04 3.986E-04 CO-58 CURIES 0.000E+00 0.000E+00 1.429E-04 8.186E-04 CO-60 CURIES 0.000E+00 0.000E+00 6.487E-04 1.185E-03 CR-51 CURIES 0.000E+00 0.000E+00 0.000E+00 3.578E-04 CS-134 CURIES 0.000E+00 0.000E+00 1.923E-05 1.253E-06 CS-137 CURIES 0.000E+00 0.000E+00 2.038E-03 1.459E-04 FE-55 CURIES 0.000E+00 0.000E+00 4.888E-04 1.202E-03 FE-59 CURIES 0.000E+00 0.000E+00 0.000E+00 1.859E-05 I-131 CURIES 0.000E+00 0.000E+00 0.000E+00 0.000E+00 I-133 CURIES 0.000E+00 0.000E+00 0.000E+00 0.000E+00 MN-54 CURIES 0.000E+00 0.000E+00 1.519E-04 4.900E-05 MO-99 CURIES 0.000E+00 0.000E+00 0.000E+00 0.000E+00 NA-24 CURIES 0.000E+00 0.000E+00 0.000E+00 3.589E-05 NB-95 CURIES 0.000E+00 0.000E+00 2.396E-05 2.576E-04 NB-97 CURIES 0.000E+00 0.000E+00 6.715E-06 1.289E-05 RU-105 CURIES 0.000E+00 0.000E+00 0.000E+00 0.000E+00 RU-106 CURIES 0.000E+00 0.000E+00 0.000E+00 2.617E-05 SB-124 CURIES 0.000E+00 0.000E+00 3.389E-05 7.310E-06 SB-125 CURIES 0.000E+00 0.000E+00 3.861E-04 2.791E-05 SN-113 CURIES 0.000E+00 0.000E+00 0.000E+00 1.561E-05 SR-92 CURIES 0.000E+00 0.000E+00 0.000E+00 6.951E-06 TC-99M CURIES 0.000E+00 0.000E+00 0.000E+00 0.000E+00 ZR-95 CURIES 0.000E+00 0.000E+00 9.568E-06 1.163E-04 H-3 CURIES 4.816E-03 1.137E-02 4.673E+01 1.617E+02 XE-133 CURIES 0.000E+00 0.000E+00 0.000E+00 3.092E-04 XE-135 CURIES 0.000E+00 0.000E+00 0.000E+00 0.000E+00 TOTAL CURIES 4.816E-03 1.137E-02 4.673E+01 1.617E+02 Page 8 of 106

ANO-1 & 2 Radioactive Effluent Release Report for 2014 UNIT 1 REPORT CATEGORY: ANNUAL LIQUID CONTINUOUS AND BATCH RELEASES TOTALS FOR EACH NUCLIDE RELEASED TYPE OF ACTIVITY: ALL RADIONUCLIDES REPORTING PERIOD: QUARTER # 3 AND QUARTER # 4 YEAR 2014 CONTINUOUS RELEASES BATCH RELEASES NUCLIDE UNIT QUARTER 3 QUARTER 4 QUARTER 3 QUARTER 4 AG-110M CURIES 0.000E+00 0.000E+00 3.842E-05 1.920E-04 CO-58 CURIES 0.000E+00 0.000E+00 3.459E-03 1.753E-03 CO-60 CURIES 0.000E+00 0.000E+00 2.725E-03 2.527E-03 CR-51 CURIES 0.000E+00 0.000E+00 3.751E-05 2.367E-04 CS-134 CURIES 0.000E+00 0.000E+00 2.565E-06 1.600E-05 CS-137 CURIES 0.000E+00 0.000E+00 4.050E-05 8.928E-04 FE-55 CURIES 0.000E+00 0.000E+00 2.074E-03 2.717E-03 FE-59 CURIES 0.000E+00 0.000E+00 0.000E+00 0.000E+00 I-131 CURIES 0.000E+00 0.000E+00 0.000E+00 1.977E-05 I-133 CURIES 0.000E+00 0.000E+00 2.483E-05 0.000E+00 MN-54 CURIES 0.000E+00 0.000E+00 1.754E-04 6.709E-05 MO-99 CURIES 0.000E+00 0.000E+00 2.428E-05 4.120E-05 NA-24 CURIES 0.000E+00 1.217E-04 3.086E-04 1.324E-04 NB-95 CURIES 0.000E+00 0.00E+00 2.909E-04 5.965E-04 NB-97 CURIES 0.000E+00 0.000E+00 0.000E+00 0.000E+00 RU-105 CURIES 0.000E+00 0.000E+00 1.210E-05 1.808E-05 RU-106 CURIES 0.000E+00 0.000E+00 0.000E+00 0.000E+00 SB-124 CURIES 0.000E+00 0.000E+00 0.000E+00 1.496E-04 SB-125 CURIES 0.000E+00 0.000E+00 0.000E+00 6.357E-04 SN-113 CURIES 0.000E+00 0.000E+00 0.000E+00 0.000E+00 SR-92 CURIES 0.000E+00 0.000E+00 9.879E-06 0.000E+00 TC-99M CURIES 0.000E+00 0.000E+00 2.472E-05 4.194E-05 ZR-95 CURIES 0.000E+00 0.000E+00 1.210E-04 2.874E-04 H-3 CURIES 1.961E-02 6.411E-03 1.910E+02 1.823E+02 XE-133 CURIES 0.000E+00 0.000E+00 8.175E-04 2.391E-03 XE-135 CURIES 0.000E+00 0.000E+00 3.454E-06 8.009E-06 TOTAL CURIES 1.961E-02 6.411E-03 1.910E+02 1.823E+02 Page 9 of 106

ANO-1 & 2 Radioactive Effluent Release Report for 2014 ANNUAL SUMMATION FOR ALL RELEASES BY QUARTER (ALL LIQUID EFFLUENTS)

January 1 through June 30, 2014 Unit 2 Est. Total Type of Effluent Units Quarter 1 Quarter 2 Error %

A. Fission and Activation Products

1. Total Release (Not Including Curies 1.622E-02 7.038E-02 25 Tritium, Gases, Alpha)

2. Average Diluted Ci/ml 1.061E-11 1.904E-10 Concentration During Period

3. Percent of Applicable Limit  % 3.537E-03 6.345E-02 B. Tritium

1. Total Release Curies 1.465E+02 7.198E+01 25

2. Average Diluted Ci/ml 9.582E-08 1.947E-07 Concentration During Period

3. Percent of Applicable Limit  % 3.194E-03 6.489E-03 C. Dissolved and Entrained Gases

1. Total Release Curies 5.348E-03 4.090E-02 25

2. Average Diluted Ci/ml 3.498E-12 1.106E-10 Concentration During Period

3. Percent of Applicable Limit  % 1.749E-06 5.531E-05 D. Gross Alpha Radioactivity

1. Total Release Curies 0.000E+00 0.000E+00 25 E. Waste Vol Released (Pre-Dilution) Liters 9.910E+05 7.893E+06 25 F. Volume of Dilution Water Used Liters 1.529E+12 3.697E+11 25 Page 10 of 106

ANO-1 & 2 Radioactive Effluent Release Report for 2014 ANNUAL SUMMATION FOR ALL RELEASES BY QUARTER (ALL LIQUID EFFLUENTS)

July 1 through December 31, 2014 Unit 2 Est. Total Type of Effluent Units Quarter 3 Quarter 4 Error %

A. Fission and Activation Products

1. Total Release (Not Including Curies 2.317E-03 1.944E-03 25 Tritium, Gases, Alpha)

2. Average Diluted Ci/ml 5.971E-12 5.814E-12 Concentration During Period

3. Percent of Applicable Limit  % 1.990E-03 1.938E-03 B. Tritium

1. Total Release Curies 1.405E+01 1.464E+02 25

2. Average Diluted Ci/ml 3.619E-08 4.378E-07 Concentration During Period

3. Percent of Applicable Limit  % 1.206E-03 1.459E-02 C. Dissolved and Entrained Gases

1. Total Release Curies 1.590E-04 3.493E-03 25

2. Average Diluted Ci/ml 4.097E-13 1.045E-11 Concentration During Period

3. Percent of Applicable Limit  % 2.048E-07 5.225E-06 D. Gross Alpha Radioactivity

1. Total Release Curies 0.000E+00 0.000E+00 25 E. Waste Vol Released (Pre-Dilution) Liters 2.712E+05 3.719E+05 25 F. Volume of Dilution Water Used Liters 3.881E+11 3.343E+11 25 Page 11 of 106

ANO-1 & 2 Radioactive Effluent Release Report for 2014 UNIT 2 REPORT CATEGORY: ANNUAL LIQUID CONTINUOUS AND BATCH RELEASES TOTALS FOR EACH NUCLIDE RELEASED TYPE OF ACTIVITY: ALL RADIONUCLIDES REPORTING PERIOD: QUARTER # 1 AND QUARTER # 2 YEAR 2014 CONTINUOUS RELEASES BATCH RELEASES NUCLIDE UNIT QUARTER 1 QUARTER 2 QUARTER 1 QUARTER 2 AG-110M CURIES 0.000E+00 0.000E+00 5.478E-05 2.291E-05 BE-7 CURIES 0.000E+00 0.000E+00 0.000E+00 0.000E+00 CO-57 CURIES 0.000E+00 0.000E+00 0.000E+00 6.995E-05 CO-58 CURIES 0.000E+00 0.000E+00 1.229E-03 2.209E-02 CO-60 CURIES 0.000E+00 0.000E+00 4.065E-04 6.864E-03 CR-51 CURIES 0.000E+00 0.000E+00 8.395E-04 2.538E-03 CS-134 CURIES 0.000E+00 0.000E+00 5.789E-05 1.331E-05 CS-137 CURIES 0.000E+00 0.000E+00 3.351E-04 1.663E-05 CS-138 CURIES 0.000E+00 0.000E+00 0.000E+00 0.000E+00 FE-55 CURIES 0.000E+00 0.000E+00 9.672E-03 2.605E-02 FE-59 CURIES 0.000E+00 0.000E+00 2.744E-04 5.403E-05 I-131 CURIES 0.000E+00 0.000E+00 0.000E+00 1.006E-05 I-133 CURIES 0.000E+00 0.000E+00 0.000E+00 0.000E+00 LA-140 CURIES 0.000E+00 0.000E+00 7.805E-06 1.183E-05 MN-54 CURIES 0.000E+00 0.000E+00 7.763E-05 4.960E-03 MO-99 CURIES 0.000E+00 0.000E+00 0.000E+00 0.000E+00 NB-95 CURIES 0.000E+00 0.000E+00 3.117E-04 4.234E-04 NB-97 CURIES 0.000E+00 0.000E+00 1.656E-05 1.494E-05 RU-105 CURIES 0.000E+00 0.000E+00 0.000E+00 2.341E-05 SB-124 CURIES 0.000E+00 0.000E+00 2.960E-04 2.796E-04 SB-125 CURIES 0.000E+00 0.000E+00 2.460E-03 9.395E-04 SE-75 CURIES 0.000E+00 0.000E+00 0.000E+00 0.000E+00 SN-113 CURIES 0.000E+00 0.000E+00 0.000E+00 1.763E-05 SR-89 CURIES 0.000E+00 0.000E+00 0.000E+00 5.706E-03 TC-99M CURIES 0.000E+00 0.000E+00 0.000E+00 0.000E+00 ZR-95 CURIES 0.000E+00 0.000E+00 1.840E-04 2.821E-04 H-3 CURIES 0.000E+00 0.000E+00 1.465E+02 7.198E+01 KR-88 CURIES 0.000E+00 0.000E+00 0.000E+00 2.802E-05 XE-133 CURIES 0.000E+00 0.000E+00 5.348E-03 4.015E-02 XE-133M CURIES 0.000E+00 0.000E+00 0.000E+00 3.605E-04 XE-135 CURIES 0.000E+00 0.000E+00 0.000E+00 3.647E-04 TOTAL CURIES 0.000E+00 0.000E+00 1.465E+02 7.198E+01 Page 12 of 106

ANO-1 & 2 Radioactive Effluent Release Report for 2014 UNIT 2 REPORT CATEGORY: ANNUAL LIQUID CONTINUOUS AND BATCH RELEASES TOTALS FOR EACH NUCLIDE RELEASED TYPE OF ACTIVITY: ALL RADIONUCLIDES REPORTING PERIOD: QUARTER # 3 AND QUARTER # 4 YEAR 2014 CONTINUOUS RELEASES BATCH RELEASES NUCLIDE UNIT QUARTER 3 QUARTER 4 QUARTER 3 QUARTER 4 AG-110M CURIES 0.000E+00 0.000E+00 2.551E-05 3.086E-05 BE-7 CURIES 0.000E+00 0.000E+00 0.000E+00 3.855E-04 CO-57 CURIES 0.000E+00 0.000E+00 0.000E+00 0.000E+00 CO-58 CURIES 0.000E+00 0.000E+00 7.326E-05 1.193E-04 CO-60 CURIES 0.000E+00 0.000E+00 3.495E-05 6.347E-05 CR-51 CURIES 0.000E+00 0.000E+00 6.563E-05 1.810E-04 CS-134 CURIES 0.000E+00 0.000E+00 7.783E-06 0.000E+00 CS-137 CURIES 0.000E+00 0.000E+00 0.000E+00 1.047E-05 CS-138 CURIES 0.000E+00 0.000E+00 3.722E-05 0.000E+00 FE-55 CURIES 0.000E+00 0.000E+00 1.880E-03 3.254E-04 FE-59 CURIES 0.000E+00 0.000E+00 0.000E+00 2.646E-05 I-131 CURIES 0.000E+00 0.000E+00 3.214E-05 8.502E-05 I-133 CURIES 0.000E+00 0.000E+00 0.000E+00 6.613E-06 LA-140 CURIES 0.000E+00 0.000E+00 0.000E+00 0.000E+00 MN-54 CURIES 0.000E+00 0.000E+00 0.000E+00 1.174E-05 MO-99 CURIES 0.000E+00 0.000E+00 0.000E+00 3.605E-05 NB-95 CURIES 0.000E+00 0.000E+00 1.797E-05 4.220E-05 NB-97 CURIES 0.000E+00 0.000E+00 0.000E+00 0.000E+00 RU-105 CURIES 0.000E+00 0.000E+00 0.000E+00 0.000E+00 SB-124 CURIES 0.000E+00 0.000E+00 1.626E-05 3.349E-05 SB-125 CURIES 0.000E+00 0.000E+00 9.187E-05 4.815E-04 SE-75 CURIES 0.000E+00 0.000E+00 1.879E-05 3.541E-05 SN-113 CURIES 0.000E+00 0.000E+00 0.000E+00 0.000E+00 SR-89 CURIES 0.000E+00 0.000E+00 0.000E+00 0.000E+00 TC-99M CURIES 0.000E+00 0.000E+00 0.000E+00 3.670E-05 ZR-95 CURIES 0.000E+00 0.000E+00 1.643E-05 3.255E-05 H-3 CURIES 0.000E+00 0.000E+00 1.405E+01 1.464E+02 KR-88 CURIES 0.000E+00 0.000E+00 0.000E+00 0.000E+00 XE-133 CURIES 0.000E+00 0.000E+00 1.590E-04 3.400E-03 XE-133M CURIES 0.000E+00 0.000E+00 0.000E+00 3.684E-05 XE-135 CURIES 0.000E+00 0.000E+00 0.000E+00 5.617E-05 TOTAL CURIES 0.000E+00 0.000E+00 1.405E+01 1.464E+02 Page 13 of 106

ANO-1 & 2 Radioactive Effluent Release Report for 2014

4.

SUMMARY

OF GASEOUS EFFLUENT DATA As required by Regulatory Guide 1.21, Revision 1, a summary of data for gaseous releases is provided in the ARERR. The summary of gaseous effluents for both Unit 1 and Unit 2 is as follows:

Unit 1 Unit 2 Number of releases: 96 120 Total time for all releases (minutes): 876,000 880,000 Maximum time for a release (minutes): 10,600 10,500 Average time for a release (minutes): 9130 7340 Minimum time for a release (minutes): 1240 1 The Unit 1 gaseous releases consisted of:

96 Planned vent and tank releases 0 Unplanned releases The Unit 2 gaseous releases consisted of:

120 Planned vent and tank releases 0 Unplanned releases Page 14 of 106

ANO-1 & 2 Radioactive Effluent Release Report for 2014 ANNUAL SUMMATION FOR ALL RELEASES BY QUARTER (ALL AIRBORNE EFFLUENTS)

January 1 through June 30, 2014 Unit 1 Est. Total Type of Effluent Units Quarter 1 Quarter 2 Error %

A. Fission and Activation Products

1. Total Release Curies 0.000E+00 0.000E+00 25

2. Average Release Rate for Ci/Sec 0.000E+00 0.000E+00 Period

3. Percent of Applicable Limit  % 0.000E+00 0.000E+00 B. Radioiodines

1. Total Iodine-131 Curies 0.000E+00 0.000E+00 25

2. Average Release Rate for Ci/Sec 0.000E+00 0.000E+00 Period

3. Percent of Applicable Limit  % 0.000E+00 0.000E+00 C. Particulates

1. Particulates (Half-Lives > 8 Days) Curies 0.000E+00 0.000E+00 25

2. Average Release Rate for Ci/Sec 0.000E+00 0.000E+00 Period

3. Percent of Applicable Limit  % 0.000E+00 0.000E+00

4. Gross Alpha Radioactivity Curies 9.996E-08 1.640E-07 D. Tritium

1. Total Release Curies 4.580E+00 3.437E+00 25

2. Average Release Rate for Ci/Sec 5.809E-01 4.360E-01 Period

3. Percent of Applicable Limit  % 8.136E-04 6.106E-04 Page 15 of 106

ANO-1 & 2 Radioactive Effluent Release Report for 2014 ANNUAL SUMMATION FOR ALL RELEASES BY QUARTER (ALL AIRBORNE EFFLUENTS)

July 1 through December 31, 2014 Unit 1 Est. Total Type of Effluent Units Quarter 3 Quarter 4 Error %

A. Fission and Activation Products

1. Total Release Curies 0.000E+00 0.000E+00 25

2. Average Release Rate for Ci/Sec 0.000E+00 0.000E+00 Period

3. Percent of Applicable Limit  % 0.000E+00 0.000E+00 B. Radioiodines

1. Total Iodine-131 Curies 0.000E+00 0.000E+00 25

2. Average Release Rate for Ci/Sec 0.000E+00 0.000E+00 Period

3. Percent of Applicable Limit  % 0.000E+00 0.000E+00 C. Particulates

1. Particulates (half-lives > 8 days) Curies 0.000E+00 0.000E+00 25

2. Average Release Rate for Ci/Sec 0.000E+00 0.000E+00 Period

3. Percent of Applicable Limit  % 0.000E+00 0.000E+00

4. Gross Alpha Radioactivity Curies 1.268E-07 1.196E-07 D. Tritium

1. Total Release Curies 1.638E+00 2.710E+00 25

2. Average Release Rate for Ci/Sec 2.077E-01 3.438E-01 Period

3. Percent of Applicable Limit  % 2.909E-04 4.815E-04 Page 16 of 106

ANO-1 & 2 Radioactive Effluent Release Report for 2014 UNIT 1 REPORT CATEGORY: ANNUAL AIRBORNE GROUND LEVEL CONTINUOUS AND BATCH RELEASES TOTALS FOR EACH NUCLIDE RELEASED TYPE OF ACTIVITY: FISSION GASES, IODINES, AND PARTICULATES REPORTING PERIOD: QUARTER # 1 AND QUARTER # 2 YEAR 2014 CONTINUOUS RELEASES BATCH RELEASES NUCLIDE UNIT QUARTER 1 QUARTER 2 QUARTER 1 QUARTER 2 Fission Gases NONE CURIES 0.000E+00 0.000E+00 0.000E+00 0.000E+00 Total for CURIES 0.000E+00 0.000E+00 0.000E+00 0.000E+00 Period Iodines NONE CURIES 0.000E+00 0.000E+00 0.000E+00 0.000E+00 Total for CURIES 0.000E+00 0.000E+00 0.000E+00 0.000E+00 Period Particulates NONE CURIES 0.000E+00 0.000E+00 0.000E+00 0.000E+00 Total for CURIES 0.000E+00 0.000E+00 0.000E+00 0.000E+00 Period Other H-3 CURIES 0.000E+00 0.000E+00 4.580E+00 3.437E+00 G-ALPHA CURIES 0.000E+00 0.000E+00 9.996E-08 1.640E-07 Total for CURIES 0.000E+00 0.000E+00 4.580E+00 3.437E+00 Period Page 17 of 106

ANO-1 & 2 Radioactive Effluent Release Report for 2014 UNIT 1 REPORT CATEGORY: ANNUAL AIRBORNE GROUND LEVEL CONTINUOUS AND BATCH RELEASES TOTALS FOR EACH NUCLIDE RELEASED TYPE OF ACTIVITY: FISSION GASES, IODINES, AND PARTICULATES REPORTING PERIOD: QUARTER # 3 AND QUARTER # 4 YEAR 2014 CONTINUOUS RELEASES BATCH RELEASES NUCLIDE UNIT QUARTER 3 QUARTER 4 QUARTER 3 QUARTER 4 Fission Gases NONE CURIES 0.000E+00 0.000E+00 0.000E+00 0.000E+00 Total for CURIES 0.000E+00 0.000E+00 0.000E+00 0.000E+00 Period Iodines NONE CURIES 0.000E+00 0.000E+00 0.000E+00 0.000E+00 Total for CURIES 0.000E+00 0.000E+00 0.000E+00 0.000E+00 Period Particulates NONE CURIES 0.000E+00 0.000E+00 0.000E+00 0.000E+00 Total for CURIES 0.000E+00 0.000E+00 0.000E+00 0.000E+00 Period Other H-3 CURIES 0.000E+00 0.000E+00 1.638E+00 2.710E+00 G-ALPHA CURIES 0.000E+00 0.000E+00 1.268E-07 1.196E-07 Total for CURIES 0.000E+00 0.000E+00 1.638E+00 2.710E+00 Period Page 18 of 106

ANO-1 & 2 Radioactive Effluent Release Report for 2014 ANNUAL SUMMATION FOR ALL RELEASES BY QUARTER (ALL AIRBORNE EFFLUENTS)

January 1 through June 30, 2014 Unit 2 Est. Total Type of Effluent Units Quarter 1 Quarter 2 Error %

A. Fission and Activation Products

1. Total Release Curies 0.000E+00 2.990E+00 25

2. Average Release Rate for Ci/Sec 0.000E+00 3.792E-01 Period

3. Percent of Applicable Limit  % 0.000E+00 5.311E-03 B. Radioiodines

1. Total Iodine-131 Curies 0.000E+00 0.000E+00 25

2. Average Release Rate for Ci/Sec 0.000E+00 0.000E+00 Period

3. Percent of Applicable Limit  % 0.000E+00 0.000E+00 C. Particulates

1. Particulates (half-lives > 8 days) Curies 0.000E+00 0.00E+00 25

2. Average Release Rate for Ci/Sec 0.000E+00 0.00E+00 Period

3. Percent of Applicable Limit  % 0.000E+00 0.00E+00

4. Gross Alpha Radioactivity Curies 6.281E-08 9.660E-08 D. Tritium

1. Total Release Curies 1.805E+00 5.321E+00 25

2. Average Release Rate for Ci/Sec 2.290E-01 6.749E-01 Period

3. Percent of Applicable Limit  % 3.207E-04 9.452E-04 Page 19 of 106

ANO-1 & 2 Radioactive Effluent Release Report for 2014 ANNUAL SUMMATION FOR ALL RELEASES BY QUARTER (ALL AIRBORNE EFFLUENTS)

July 1 through December 31, 2014 Unit 2 Est. Total Type of Effluent Units Quarter 3 Quarter 4 Error %

A. Fission and Activation Products

1. Total Release Curies 1.925E+00 2.474E+00 25

2. Average Release Rate for Ci/Sec 2.442E-01 3.138E-01 Period

3. Percent of Applicable Limit  % 3.420E-03 4.395E-03 B. Radioiodines

1. Total Iodine-131 Curies 0.000E+00 0.000E+00 25

2. Average Release Rate for Ci/Sec 0.000E+00 0.000E+00 Period

3. Percent of Applicable Limit  % 0.000E+00 0.000E+00 C. Particulates

1. Particulates (half-lives > 8 days) Curies 0.000E+00 0.000E+00 25

2. Average Release Rate for Ci/Sec 0.000E+00 0.000E+00 Period

3. Percent of Applicable Limit  % 0.000E+00 0.000E+00

4. Gross Alpha Radioactivity Curies 5.419E-07 1.698E-07 D. Tritium

1. Total Release Curies 6.522E+00 6.396E+00 25

2. Average Release Rate for Ci/Sec 8.272E-01 8.112E-01 Period

3. Percent of Applicable Limit  % 1.159E-03 1.136E-03 Page 20 of 106

ANO-1 & 2 Radioactive Effluent Release Report for 2014 UNIT 2 REPORT CATEGORY: ANNUAL AIRBORNE GROUND LEVEL CONTINUOUS AND BATCH RELEASES TOTALS FOR EACH NUCLIDE RELEASED TYPE OF ACTIVITY: FISSION GASES, IODINES, AND PARTICULATES REPORTING PERIOD: QUARTER # 1 AND QUARTER # 2 YEAR 2014 CONTINUOUS RELEASES BATCH RELEASES NUCLIDE UNIT QUARTER 1 QUARTER 2 QUARTER 1 QUARTER 2 Fission Gases AR-41 CURIES 0.000E+00 0.000E+00 0.000E+00 2.943E+00 XE-133 CURIES 0.000E+00 0.000E+00 0.000E+00 4.709E-02 Total for CURIES 0.000E+00 0.000E+00 0.000E+00 2.990E+00 Period Iodines NONE CURIES 0.000E+00 0.000E+00 0.000E+00 0.000E+00 Total for CURIES 0.000E+00 0.000E+00 0.000E+00 0.000E+00 Period Particulates NONE CURIES 0.000E+00 0.000E+00 0.000E+00 0.000E+00 Total for CURIES 0.000E+00 0.000E+00 0.000E+00 0.000E+00 Period Other G-ALPHA CURIES 0.000E+00 0.000E+00 6.281E-08 9.660E-08 H-3 CURIES 0.000E+00 0.000E+00 1.805E+00 5.321E+00 Total for CURIES 0.000E+00 0.000E+00 1.805E+00 5.321E+00 Period Page 21 of 106

ANO-1 & 2 Radioactive Effluent Release Report for 2014 UNIT 2 REPORT CATEGORY: ANNUAL AIRBORNE GROUND LEVEL CONTINUOUS AND BATCH RELEASES TOTALS FOR EACH NUCLIDE RELEASED TYPE OF ACTIVITY: FISSION GASES, IODINES, AND PARTICULATES REPORTING PERIOD: QUARTER # 3 AND QUARTER # 4 YEAR 2014 CONTINUOUS RELEASES BATCH RELEASES NUCLIDE UNIT QUARTER 3 QUARTER 4 QUARTER 3 QUARTER 4 Fission Gases AR-41 CURIES 0.000E+00 0.000E+00 1.925E+00 2.474E+00 XE-133 CURIES 0.000E+00 0.000E+00 0.000E+00 0.000E+00 Total for CURIES 0.000E+00 0.000E+00 1.925E+00 2.474E+00 Period Iodines NONE CURIES 0.000E+00 0.000E+00 0.000E+00 0.000E+00 Total for CURIES 0.000E+00 0.000E+00 0.000E+00 0.000E+00 Period Particulates NONE CURIES 0.000E+00 0.000E+00 0.000E+00 0.000E+00 Total for CURIES 0.000E+00 0.000E+00 0.000E+00 0.000E+00 Period Other G-ALPHA CURIES 0.000E+00 0.000E+00 5.419E-07 1.698E-07 H-3 CURIES 0.000E+00 0.000E+00 6.522E+00 6.396E+00 Total for CURIES 0.000E+00 0.000E+00 6.522E+00 6.396E+00 Period Page 22 of 106

ANO-1 & 2 Radioactive Effluent Release Report for 2014

5.

SUMMARY

OF RADIATION DOSES The following is a summary of the annual radiation doses due to radiological effluents during 2014 calculated in accordance with the ODCM.

UNIT 1 Liquid Radwaste Effluents Dose Limits (mRem): Total Body = 1.5/Qtr 3/Yr, Other Organs = 5/Qtr 10/Yr Organ Qtr 1  % Qtr 2  % Qtr 3  % Qtr 4  % Year  %

TBody 1.118E-3 7.454E-2 3.150E-4 2.100E-2 2.843E-4 1.895E-2 7.156E-4 4.771E-2 2.433E-3 8.110E-2 Bone 1.149E-3 2.298E-2 8.576E-5 1.715E-3 2.769E-5 5.538E-4 4.781E-4 9.562E-3 1.740E-3 1.740E-2 Liver 1.662E-3 3.324E-2 3.539E-4 7.077E-3 2.940E-4 5.879E-3 9.364E-4 1.873E-2 3.246E-3 3.246E-2 Thyroid 7.795E-5 1.559E-3 2.356E-4 4.712E-3 2.532E-4 5.064E-3 2.867E-4 5.735E-3 8.535E-4 8.535E-3 Kidney 6.142E-4 1.228E-2 2.735E-4 5.470E-3 2.618E-4 5.237E-3 4.964E-4 9.927E-3 1.646E-3 1.646E-2 Lung 2.566E-4 5.133E-3 2.501E-4 5.003E-3 2.575E-4 5.149E-3 3.547E-4 7.095E-3 1.119E-3 1.119E-2 GI-LL 1.313E-4 2.626E-3 2.865E-4 5.730E-3 3.377E-4 6.755E-3 3.918E-4 7.835E-3 1.147E-3 1.147E-2 Skin 0.000E+0 0.000E+0 0.000E+0 0.000E+0 0.000E+0 0.000E+0 0.000E+0 0.000E+0 0.000E+0 0.000E+0 Gaseous Radwaste Effluents Iodine, H-3, and Particulate (ITP) - Dose Limits (mRem) = 7.5/Qtr 15/Yr Organ Qtr 1  % Qtr 2  % Qtr 3  % Qtr 4  % Year  %

TBody 2.014E-2 2.686E-1 1.512E-2 2.016E-1 7.203E-3 9.604E-2 1.192E-2 1.590E-1 5.439E-2 3.626E-1 Bone 0.000E+0 0.000E+0 0.000E+0 0.000E+0 0.000E+0 0.000E+0 0.000E+0 0.000E+0 0.000E+0 0.000E+0 Liver 2.014E-2 2.686E-1 1.512E-2 2.016E-1 7.203E-3 9.604E-2 1.192E-2 1.590E-1 5.439E-2 3.626E-1 Thyroid 2.014E-2 2.686E-1 1.512E-2 2.016E-1 7.203E-3 9.604E-2 1.192E-2 1.590E-1 5.439E-2 3.626E-1 Kidney 2.014E-2 2.686E-1 1.512E-2 2.016E-1 7.203E-3 9.604E-2 1.192E-2 1.590E-1 5.439E-2 3.626E-1 Lung 2.014E-2 2.686E-1 1.512E-2 2.016E-1 7.203E-3 9.604E-2 1.192E-2 1.590E-1 5.439E-2 3.626E-1 GI-LLI 2.014E-2 2.686E-1 1.512E-2 2.016E-1 7.203E-3 9.604E-2 1.192E-2 1.590E-1 5.439E-2 3.626E-1 Skin 0.000E+0 0.000E+0 0.000E+0 0.000E+0 0.000E+0 0.000E+0 0.000E+0 0.000E+0 0.000E+0 0.000E+0 Noble Gas Air Dose Limits (mRad) = Gamma 5/Qtr 10/Yr, Beta 10/Qtr 20/Yr Type Qtr 1  % Qtr 2  % Qtr 3  % Qtr 4  % Year  %

Gamma 0.000E+0 0.000E+0 0.000E+0 0.000E+0 0.000E+0 0.000E+0 0.000E+0 0.000E+0 0.000E+0 0.000E+0 Beta 0.000E+0 0.000E+0 0.000E+0 0.000E+0 0.000E+0 0.000E+0 0.000E+0 0.000E+0 0.000E+0 0.000E+0 Page 23 of 106

ANO-1 & 2 Radioactive Effluent Release Report for 2014 UNIT 2 Liquid Radwaste Effluents Dose Limits (mRem): Total Body = 1.5/Qtr 3/Yr, Other Organs = 5/Qtr 10/Yr Organ Qtr 1  % Qtr 2  % Qtr 3  % Qtr 4  % Year  %

TBody 4.488E-4 2.992E-2 2.553E-4 1.702E-2 2.470E-5 1.646E-3 2.247E-4 1.498E-2 9.534E-4 3.178E-2 Bone 2.394E-4 4.788E-3 9.589E-4 1.918E-2 9.768E-6 1.954E-4 6.948E-6 1.390E-4 1.215E-3 1.215E-2 Liver 5.655E-4 1.131E-2 3.827E-4 7.654E-3 2.952E-5 5.904E-4 2.283E-4 4.566E-3 1.206E-3 1.206E-2 Thyroid 2.345E-4 4.690E-3 1.187E-4 2.373E-3 3.109E-5 6.217E-4 2.595E-4 5.191E-3 6.438E-4 6.438E-3 Kidney 3.336E-4 6.673E-3 1.641E-4 3.283E-3 2.031E-5 4.063E-4 2.218E-4 4.437E-3 7.399E-4 7.399E-3 Lung 2.849E-4 5.698E-3 1.605E-4 3.209E-3 2.161E-5 4.323E-4 2.205E-4 4.410E-3 6.875E-4 6.875E-3 GI-LL 3.372E-4 6.743E-3 1.067E-3 2.135E-2 2.368E-5 4.737E-4 2.287E-4 4.574E-3 1.657E-3 1.657E-2 Skin 0.000E+0 0.000E+0 0.000E+0 0.000E+0 0.000E+0 0.000E+0 0.000E+0 0.000E+0 0.000E+0 0.000E+0 Gaseous Radwaste Effluents Iodine, H-3, and Particulate (ITP) - Dose Limits (mRem) = 7.5/Qtr 15/Yr Organ Qtr 1  % Qtr 2  % Qtr 3  % Qtr 4  % Year  %

TBody 7.941E-3 1.059E-1 2.340E-2 3.120E-1 2.869E-2 3.825E-1 2.813E-2 3.751E-1 8.817E-2 5.878E-1 Bone 0.000E+0 0.000E+0 0.000E+0 0.000E+0 0.000E+0 0.000E+0 0.000E+0 0.000E+0 0.000E+0 0.000E+0 Liver 7.941E-3 1.059E-1 2.340E-2 3.120E-1 2.869E-2 3.825E-1 2.813E-2 3.751E-1 8.817E-2 5.878E-1 Thyroid 7.941E-3 1.059E-1 2.340E-2 3.120E-1 2.869E-2 3.825E-1 2.813E-2 3.751E-1 8.817E-2 5.878E-1 Kidney 7.941E-3 1.059E-1 2.340E-2 3.120E-1 2.869E-2 3.825E-1 2.813E-2 3.751E-1 8.817E-2 5.878E-1 Lung 7.941E-3 1.059E-1 2.340E-2 3.120E-1 2.869E-2 3.825E-1 2.813E-2 3.751E-1 8.817E-2 5.878E-1 GI-LLI 7.941E-3 1.059E-1 2.340E-2 3.120E-1 2.869E-2 3.825E-1 2.813E-2 3.751E-1 8.817E-2 5.878E-1 Skin 0.000E+0 0.000E+0 0.000E+0 0.000E+0 0.000E+0 0.000E+0 0.000E+0 0.000E+0 0.000E+0 0.000E+0 Noble Gas Air Dose Limits (mRad) = Gamma 5/Qtr 10/Yr, Beta 10/Qtr 20/Yr Type Qtr 1  % Qtr 2  % Qtr 3  % Qtr 4  % Year  %

Gamma 0.000E+0 0.000E+0 1.737E-2 3.473E-1 1.136E-2 2.271E-1 1.459E-2 2.918E-1 4.331E-2 4.331E-1 Beta 0.000E+0 0.000E+0 6.152E-3 6.152E-2 4.005E-3 4.005E-2 5.416E-3 5.146E-2 1.530E-2 7.652E-2 Page 24 of 106

ANO-1 & 2 Radioactive Effluent Release Report for 2014

6.

SUMMARY

OF DOSE TO MEMBERS OF THE PUBLIC The following is a summary of the annual radiation dose to members of the public (in mrem) due to activities inside the site boundary.

UNIT 1 BONE LIVER TBODY THYROID KIDNEY GI-LLI LUNG SKIN Gaseous Effluent Iodine/Tritium 0.000E+00 5.439E-02 5.439E-02 5.439E-02 5.439E-02 5.439E-02 5.439E-02 0.000E+00 Particulate Noble Gas 0.000E+00 0.000E+00 Liquid Effluent Fish 1.740E-03 3.246E-03 2.433E-03 8.535E-04 1.646E-03 1.147E-03 1.119E-03 0.000E+00 Sediment 2.074E-04 2.436E-04 Unit 1 Total 1.740E-03 5.764E-02 5.703E-02 5.524E-02 5.604E-02 5.554E-02 5.551E-02 2.436E-04 UNIT 2 Gaseous Effluent Iodine/Tritium 0.000E+00 8.817E-02 8.817E-02 8.817E-02 8.817E-02 8.817E-02 8.817E-02 0.000E+00 Particulate Noble Gas 2.882E-02 4.619E-02 Liquid Effluent Fish 1.215E-03 1.206E-03 9.534E-04 6.438E-04 7.399E-04 1.657E-03 6.875E-04 0.000E+00 Sediment 1.954E-04 2.297E-04 Unit 2 Total 1.215E-03 8.938E-02 1.181E-01 8.881E-02 8.891E-02 8.983E-02 8.886E-02 4.642E-02 Site Total 2.955E-03 1.470E-01 1.751E-01 1.441E-01 1.450E-01 1.454E-01 1.444E-01 4.666E-02 Limit (40CFR190) 25 25 75 25 25 25 25 25

% Limit 1.182E-02 5.881E-01 7.005E-01 5.762E-01 5.798E-01 5.815E-01 5.775E-01 1.867E-01 Page 25 of 106

ANO-1 & 2 Radioactive Effluent Release Report for 2014

7. HISTORICAL EFFLUENT DATA The following graphs show the historical release data for both units on a yearly basis. These graphs compare data from 2004 through 2014.

Page 26 of 106

ANO-1 & 2 Radioactive Effluent Release Report for 2014 UNIT 1 LIQUID EFFLUENTS FISSION AND ACTIVATION PRODUCTS 1.00E+00 CURIES 1.00E-01 1.00E-02 2004 2005 2006 2007 2008 2009 2010 2011 2012 2013 2014 YEAR UNIT 1 LIQUID EFFLUENTS TRITIUM 1.00E+03 CURIES 1.00E+02 2004 2005 2006 2007 2008 2009 2010 2011 2012 2013 2014 YEAR Page 27 of 106

ANO-1 & 2 Radioactive Effluent Release Report for 2014 UNIT 1 LIQUID EFFLUENTS DISSOLVED AND ENTRAINED GASES 1.00E+02 1.00E+01 1.00E+00 1.00E-01 CURIES 1.00E-02 1.00E-03 1.00E-04 1.00E-05 1.00E-06 2004 2005 2006 2007 2008 2009 2010 2011 2012 2013 2014 YEAR UNIT 1 LIQUID EFFLUENTS TOTAL VOLUME RELEASED 1.00E+08 GALLONS 1.00E+07 1.00E+06 2004 2005 2006 2007 2008 2009 2010 2011 2012 2013 2014 YEAR Page 28 of 106

ANO-1 & 2 Radioactive Effluent Release Report for 2014 UNIT 1 LIQUID EFFLUENTS CRITICAL ORGAN DOSE 1.00E+00 1.00E-01 MREM 1.00E-02 1.00E-03 2004 2005 2006 2007 2008 2009 2010 2011 2012 2013 2014 YEAR UNIT 1 LIQUID EFFLUENTS TOTAL BODY DOSE 1.00E+00 1.00E-01 MREM 1.00E-02 1.00E-03 2004 2005 2006 2007 2008 2009 2010 2011 2012 2013 2014 YEAR Page 29 of 106

ANO-1 & 2 Radioactive Effluent Release Report for 2014 UNIT 1 LIQUID EFFLUENTS COLLECTIVE DOSES 0.140 0.120 0.100 0.080

% LIMIT 0.060 0.040 0.020 0.000 2004 2005 2006 2007 2008 2009 2010 2011 2012 2013 2014 YEAR BONE LIVER GI-LLI THYROID KIDNEY LUNG UNIT 1 GASEOUS EFFLUENTS FISSION AND ACTIVATION PRODUCTS 1.00E+04 1.00E+03 1.00E+02 1.00E+01 CURIES 1.00E+00 1.00E-01 1.00E-02 1.00E-03 1.00E-04 2004 2005 2006 2007 2008 2009 2010 2011 2012 2013 2014 YEAR Page 30 of 106

ANO-1 & 2 Radioactive Effluent Release Report for 2014 UNIT 1 GASEOUS EFFLUENTS RADIOIODINES 1.00E+00 1.00E-01 1.00E-02 1.00E-03 CUIRES 1.00E-04 1.00E-05 1.00E-06 1.00E-07 1.00E-08 2004 2005 2006 2007 2008 2009 2010 2011 2012 2013 2014 YEAR UNIT 1 GASEOUS EFFLUENTS GROSS GAMMA 1.00E-02 1.00E-03 CURIES 1.00E-04 1.00E-05 2004 2005 2006 2007 2008 2009 2010 2011 2012 2013 2014 YEAR Page 31 of 106

ANO-1 & 2 Radioactive Effluent Release Report for 2014 UNIT 1 GASEOUS EFFLUENTS PARTICULTES 1.00E-01 1.00E-02 1.00E-03 1.00E-04 CURIES 1.00E-05 1.00E-06 1.00E-07 1.00E-08 1.00E-09 2004 2005 2006 2007 2008 2009 2010 2011 2012 2013 2014 YEAR UNIT 1 GASEOUS EFFLUENTS TRITIUM 1.00E+02 CURIES 1.00E+01 1.00E+00 2004 2005 2006 2007 2008 2009 2010 2011 2012 2013 2014 YEAR Page 32 of 106

ANO-1 & 2 Radioactive Effluent Release Report for 2014 UNIT 1 GASEOUS EFFLUENTS BETA DOSE 1.00E+00 1.00E-01 MRAD 1.00E-02 1.00E-03 1.00E-04 2004 2005 2006 2007 2008 2009 2010 2011 2012 2013 2014 YEAR UNIT 1 GASEOUS EFFLUENTS TOTAL BODY DOSE 1.00E+00 1.00E-01 MREM 1.00E-02 1.00E-03 2004 2005 2006 2007 2008 2009 2010 2011 2012 2013 2014 YEAR Page 33 of 106

ANO-1 & 2 Radioactive Effluent Release Report for 2014 UNIT 1 GASEOUS EFFLUENTS CRITICAL ORGAN DOSE 1.00E+00 1.00E-01 MREM 1.00E-02 1.00E-03 2004 2005 2006 2007 2008 2009 2010 2011 2012 2013 2014 YEAR UNIT 1 GASEOUS EFFLUENTS COLLECTIVE DOSES 0.700 0.600 0.500 0.400

% LIMIT 0.300 0.200 0.100 0.000 2004 2005 2006 2007 2008 2009 2010 2011 2012 2013 2014 YEAR BONE LIVER THYROID KIDNEY LUNG GI-LLI Page 34 of 106

ANO-1 & 2 Radioactive Effluent Release Report for 2014 UNIT 2 LIQUID EFFLUENTS FISSION AND ACTIVATION PRODUCTS 1.00E+00 CURIES 1.00E-01 1.00E-02 2004 2005 2006 2007 2008 2009 2010 2011 2012 2013 2014 YEAR UNIT 2 LIQUID EFFLUENTS TRITIUM 1.00E+04 1.00E+03 CURIES 1.00E+02 1.00E+01 2004 2005 2006 2007 2008 2009 2010 2011 2012 2013 2014 YEAR Page 35 of 106

ANO-1 & 2 Radioactive Effluent Release Report for 2014 UNIT 2 LIQUID EFFLUENTS DISSOLVED AND ENTRAINED GASES 1.00E+03 1.00E+02 1.00E+01 CURIES 1.00E+00 1.00E-01 1.00E-02 1.00E-03 2004 2005 2006 2007 2008 2009 2010 2011 2012 2013 2014 YEAR UNIT 2 LIQUID EFFLUENTS TOTAL VOLUME RELEASED 1.00E+08 GALLONS 1.00E+07 1.00E+06 2004 2005 2006 2007 2008 2009 2010 2011 2012 2013 2014 YEAR Page 36 of 106

ANO-1 & 2 Radioactive Effluent Release Report for 2014 UNIT 2 LIQUID EFFLUENTS TOTAL BODY DOSE 1.00E+00 1.00E-01 MREM 1.00E-02 1.00E-03 1.00E-04 2004 2005 2006 2007 2008 2009 2010 2011 2012 2013 2014 YEAR UNIT 2 LIQUID EFFLUENTS CRITICAL ORGAN DOSE 1.00E+00 1.00E-01 MREM 1.00E-02 1.00E-03 1.00E-04 2004 2005 2006 2007 2008 2009 2010 2011 2012 2013 2014 YEAR Page 37 of 106

ANO-1 & 2 Radioactive Effluent Release Report for 2014 UNIT 2 LIQUID EFFLUENTS COLLECTIVE DOSES 0.100 0.090 0.080 0.070 0.060

% LIMIT 0.050 0.040 0.030 0.020 0.010 0.000 2004 2005 2006 2007 2008 2009 2010 2011 2012 2013 2014 YEAR BONE LIVER GI-LLI THYROID KIDNEY LUNG UNIT 2 GASEOUS EFFLUENTS FISSION AND ACTIVATION PRODUCTS 1.00E+04 1.00E+03 1.00E+02 CURIES 1.00E+01 1.00E+00 1.00E-01 1.00E-02 2004 2005 2006 2007 2008 2009 2010 2011 2012 2013 2014 YEAR Page 38 of 106

ANO-1 & 2 Radioactive Effluent Release Report for 2014 UNIT 2 GASEOUS EFFLUENTS TRITIUM 1.00E+02 CURIES 1.00E+01 1.00E+00 2004 2005 2006 2007 2008 2009 2010 2011 2012 2013 2014 YEAR UNIT 2 GASEOUS EFFLUENTS RADIOIODINES 1.00E-01 1.00E-02 1.00E-03 CURIES 1.00E-04 1.00E-05 1.00E-06 1.00E-07 2004 2005 2006 2007 2008 2009 2010 2011 2012 2013 2014 YEAR Page 39 of 106

ANO-1 & 2 Radioactive Effluent Release Report for 2014 UNIT 2 GASEOUS EFFLUENTS PARTICULATES 1.00E-01 1.00E-02 1.00E-03 CURIES 1.00E-04 1.00E-05 1.00E-06 1.00E-07 2004 2005 2006 2007 2008 2009 2010 2011 2012 2013 2014 YEAR UNIT 2 GASEOUS EFFLUENTS GAMMA DOSE 1.00E+00 1.00E-01 1.00E-02 CURIES 1.00E-03 1.00E-04 1.00E-05 2004 2005 2006 2007 2008 2009 2010 2011 2012 2013 2014 YEAR Page 40 of 106

ANO-1 & 2 Radioactive Effluent Release Report for 2014 UNIT 2 GASEOUS EFFLUENTS BETA DOSE 1

0.1 CURIES 0.01 0.001 0.0001 2004 2005 2006 2007 2008 2009 2010 2011 2012 2013 2014 YEAR UNIT 2 GASEOUS EFFLUENTS TOTAL BODY DOSE 1.00E+00 1.00E-01 MREM 1.00E-02 1.00E-03 2004 2005 2006 2007 2008 2009 2010 2011 2012 2013 2014 YEAR Page 41 of 106

ANO-1 & 2 Radioactive Effluent Release Report for 2014 UNIT 2 GASEOUS EFFLUENTS CRITICAL ORGAN DOSE 1.00E+00 1.00E-01 MREM 1.00E-02 1.00E-03 2004 2005 2006 2007 2008 2009 2010 2011 2012 2013 2014 YEAR UNIT 2 GASEOUS EFFLUENTS COLLECTIVE DOSES 2.500 2.000 1.500

% LIMIT 1.000 0.500 0.000 2004 2005 2006 2007 2008 2009 2010 2011 2012 2013 2014 YEAR BONE LIVER THYROID KIDNEY LUNG GI-LLI Page 42 of 106

ANO-1 & 2 Radioactive Effluent Release Report for 2014

8. SOLID WASTE

SUMMARY

As required by Regulatory Guide 1.21, Revision 1, a summary of data for solid wastes shipped offsite is provided in the Annual Radioactive Effluent Release Report (ARRERR).

The summary for solid waste shipments for Unit 1 is as follows:

Solid Waste Shipped Offsite for Disposal and Estimates of Major Nuclides by Waste Class and Stream During Period From 01/01/2014 to 06/30/2014 Waste Stream: Resins, Filters and Evaporator Bottom Volume Waste Class Ft³ M³ Curies Shipped  % Error (Ci)

A 7.20E+02 2.04E+01 1.47E-02 +/-25%

B 0.00E+00 0.00E+00 0.00E+00 +/-25%

C 0.00E+00 0.00E+00 0.00E+00 +/-25%

All 7.20E+02 2.04E+01 1.47E-02 +/-25%

Waste Stream: Dry Active Waste Volume Waste Class Ft³ M³ Curies Shipped  % Error (Ci)

A 1.62E+03 4.59E+01 5.17E-03 +/-25%

B 0.00E+00 0.00E+00 0.00E+00 +/-25%

C 0.00E+00 0.00E+00 0.00E+00 +/-25%

All 1.62E+03 4.59E+01 5.17E-03 +/-25%

Waste Stream: Irradiated Components Volume Waste Class Ft³ M³ Curies Shipped  % Error (Ci)

A 0.00E+00 0.00E+00 0.00E+00 +/-25%

B 0.00E+00 0.00E+00 0.00E+00 +/-25%

C 0.00E+00 0.00E+00 0.00E+00 +/-25%

All 0.00E+00 0.00E+00 0.00E+00 +/-25%

Waste Stream: Other Wastes Volume Page 43 of 106

ANO-1 & 2 Radioactive Effluent Release Report for 2014 Waste Class Ft³ M³ Curies Shipped % Error (Ci)

A 0.00E+00 0.00E+00 0.00E+00 +/-25%

B 0.00E+00 0.00E+00 0.00E+00 +/-25%

C 0.00E+00 0.00E+00 0.00E+00 +/-25%

All 0.00E+00 0.00E+00 0.00E+00 +/-25%

Waste Stream: Sum of All 4 Categories Volume Waste Class Ft³ M³ Curies Shipped % Error (Ci)

A 2.34E+03 6.63E+01 1.99E-02 +/-25%

B 0.00E+00 0.00E+00 0.00E+00 +/-25%

C 0.00E+00 0.00E+00 0.00E+00 +/-25%

All 2.34E+03 6.63E+01 1.99E-02 +/-25%

Number of Shipments Mode of Transportation Destination 1 Hittman Transport Bear Creek Operations 1 Landstar Ranger Bear Creek Operations Resins, Filters and Evaporator Bottoms Waste Class A Nuclide Name Percent Abundance Curies Co-60 1.43% 2.11E-04 Cs-137 98.57% 1.45E-02 Resins, Filters and Evaporator Bottoms Waste Class ALL Nuclide Name Percent Abundance Curies Co-60 1.43% 2.11E-04 Cs-137 98.57% 1.45E-02 Page 44 of 106

ANO-1 & 2 Radioactive Effluent Release Report for 2014 Dry Active Waste Waste Class A Nuclide Name Percent Abundance Curies C-14 1.11% 5.76E-05 Cr-51 0.28% 1.47E-05 Mn-54 1.14% 5.90E-05 Fe-55 32.13% 1.66E-03 Co-57 0.31% 1.61E-05 Co-58 0.49% 2.51E-05 Co-60 24.00% 1.24E-03 Ni-63 29.03% 1.50E-03 Zr-95 0.51% 2.61E-05 Nb-95 0.90% 4.64E-05 Sb-125 0.28% 1.44E-05 Cs-134 0.09% 4.70E-06 Cs-137 9.73% 5.03E-04 Dry Active Waste Waste Class All Nuclide Name Percent Abundance Curies C-14 1.11% 5.76E-05 Cr-51 0.28% 1.47E-05 Mn-54 1.14% 5.90E-05 Fe-55 32.13% 1.66E-03 Co-57 0.31% 1.61E-05 Co-58 0.49% 2.51E-05 Co-60 24.00% 1.24E-03 Ni-63 29.03% 1.50E-03 Zr-95 0.51% 2.61E-05 Nb-95 0.90% 4.64E-05 Sb-125 0.28% 1.44E-05 Cs-134 0.09% 4.70E-06 Cs-137 9.73% 5.03E-04 Page 45 of 106

ANO-1 & 2 Radioactive Effluent Release Report for 2014 Sum of All 4 Categories Waste Class A Nuclide Name Percent Abundance Curies C-14 0.29% 5.76E-05 Cr-51 0.07% 1.47E-05 Mn-54 0.30% 5.90E-05 Fe-55 8.35% 1.66E-03 Co-57 0.08% 1.61E-05 Co-58 0.13% 2.51E-05 Co-60 7.30% 1.45E-03 Ni-63 7.55% 1.50E-03 Zr-95 0.13% 2.61E-05 Nb-95 0.23% 4.64E-05 Sb-125 0.07% 1.44E-05 Cs-134 0.02% 4.70E-06 Cs-137 75.48% 1.50E-02 Sum of All 4 Categories Waste Class All Nuclide Name Percent Abundance Curies C-14 0.29% 5.76E-05 Cr-51 0.07% 1.47E-05 Mn-54 0.30% 5.90E-05 Fe-55 8.35% 1.66E-03 Co-57 0.08% 1.61E-05 Co-58 0.13% 2.51E-05 Co-60 7.30% 1.45E-03 Ni-63 7.55% 1.50E-03 Zr-95 0.13% 2.61E-05 Nb-95 0.23% 4.64E-05 Sb-125 0.07% 1.44E-05 Cs-134 0.02% 4.70E-06 Cs-137 75.48% 1.50E-02 Waste Volume Burial Volume Manifest Number Date Shipped Used Used RSR 2014-016 3/11/2014 X RSR 2014-045 4/24/2014 X Page 46 of 106

ANO-1 & 2 Radioactive Effluent Release Report for 2014 Solid Waste Shipped Offsite for Disposal and Estimates of Major Nuclides by Waste Class and Stream for Unit 1 During Period From 07/01/2014 to 12/31/2014 Waste Stream: Resins, Filters and Evaporator Bottom Volume Waste Curies Class Ft³ M³ Shipped  % Error (Ci)

A 8.12E+02 2.30E+01 4.75E+01 +/-25%

B 5.52E+02 1.58E+01 1.03E+02 +/-25%

C 0.00E+00 0.00E+00 0.00E+00 +/-25%

All 1.36E+03 3.88E+01 1.51E+02 +/-25%

Waste Stream: Dry Active Waste Volume Waste Curies Class Ft³ M³ Shipped  % Error (Ci)

A 0.00E+00 0.00E+00 0.00E+00 +/-25%

B 0.00E+00 0.00E+00 0.00E+00 +/-25%

C 0.00E+00 0.00E+00 0.00E+00 +/-25%

All 0.00E+00 0.00E+00 0.00E+00 +/-25%

Waste Stream: Irradiated Components Volume Waste Curies Class Ft³ M³ Shipped  % Error (Ci)

A 0.00E+00 0.00E+00 0.00E+00 +/-25%

B 0.00E+00 0.00E+00 0.00E+00 +/-25%

C 0.00E+00 0.00E+00 0.00E+00 +/-25%

All 0.00E+00 0.00E+00 0.00E+00 +/-25%

Page 47 of 106

ANO-1 & 2 Radioactive Effluent Release Report for 2014 Waste Stream: Other Wastes Volume Waste Curies Class Ft³ M³ Shipped  % Error (Ci)

A 0.00E+00 0.00E+00 0.00E+00 +/-25%

B 0.00E+00 0.00E+00 0.00E+00 +/-25%

C 0.00E+00 0.00E+00 0.00E+00 +/-25%

All 0.00E+00 0.00E+00 0.00E+00 +/-25%

Waste Stream: Sum of All 4 Categories Volume Waste Curies Class Ft³ M³ Shipped  % Error (Ci)

A 8.12E+02 2.30E+01 4.75E+01 +/-25%

B 5.52E+02 1.58E+01 1.03E+02 +/-25%

C 0.00E+00 0.00E+00 0.00E+00 +/-25%

All 1.36E+3 3.88E+01 1.51E+02 +/-25%

Number of Shipments Mode of Transportation Destination 5 Hittman Transport Bear Creek Operations 2 Hittman Transport Erwin Resin Solution Gallaher Road 1 Hittman Transport Operations Page 48 of 106

ANO-1 & 2 Radioactive Effluent Release Report for 2014 Resins, Filters and Evaporator Bottoms Waste Class A Nuclide Name Percent Abundance Curies H-3 0.27% 1.30E-01 C-14 4.36% 2.07E+00 Mn-54 3.68% 1.75E+00 Fe-55 13.58% 6.45E+00 Co-57 0.12% 5.93E-02 Co-58 0.47% 2.23E-01 Co-60 67.79% 3.22E+01 Ni-59 0.02% 1.04E-02 Ni-63 6.08% 2.89E+00 Zn-65 0.50% 2.38E-01 Sr-90 0.01% 4.94E-03 Tc-99 0.00% 1.34E-03 Cs-134 0.54% 2.55E-01 Cs-137 2.56% 1.22E+00 Resins, Filters and Evaporator Bottoms Waste Class B Nuclide Name Percent Abundance Curies H-3 0.28% 2.86E-01 C-14 4.01% 4.14E+00 Mn-54 1.23% 1.27E+00 Fe-55 20.23% 2.09E+01 Co-57 0.03% 3.27E-02 Co-58 0.07% 7.61E-02 Co-60 29.04% 3.00E+01 Ni-59 0.21% 2.20E-01 Ni-63 27.83% 2.87E+01 Zn-65 0.18% 1.89E-01 Sr-90 0.08% 8.47E-02 Tc-99 0.01% 5.48E-03 Sb-125 0.83% 8.55E-01 Cs-134 1.84% 2.00E+00 Cs-137 13.91% 1.44E+01 Ce-144 0.11% 1.12E-01 Pu-238 0.00% 1.34E-04 Pu-241 0.00% 4.28E-03 Am-241 0.00% 5.64E-05 Page 49 of 106

ANO-1 & 2 Radioactive Effluent Release Report for 2014 Resins, Filters and Evaporator Bottoms Waste Class All Nuclide Name Percent Abundance Curies H-3 0.28% 4.16E-01 C-14 4.12% 6.21E+00 Mn-54 2.01% 3.02E+00 Fe-55 18.13% 2.73E+01 Co-57 0.06% 9.20E-02 Co-58 0.20% 2.99E-01 Co-60 41.26% 6.22E+01 Ni-59 0.15% 2.31E-01 Ni-63 20.97% 3.16E+01 Zn-65 0.28% 4.27E-01 Sr-90 0.08% 8.96E-02 Tc-99 0.01% 6.82E-03 Sb-125 0.57% 8.55E-01 Cs-134 1.50% 2.26E+00 Cs-137 10.33% 1.56E+01 Ce-144 0.07% 1.12E-01 Pu-238 0.00% 1.34E-04 Pu-241 0.00% 4.28E-03 Am-241 0.00% 5.64E-05 Sum of All 4 Categories Waste Class A Nuclide Name Percent Abundance Curies H-3 0.27% 1.30E-01 C-14 4.36% 2.07E+00 Mn-54 3.68% 1.75E+00 Fe-55 13.58% 6.45E+00 Co-57 0.12% 5.93E-02 Co-58 0.47% 2.23E-01 Co-60 67.79% 3.22E+01 Ni-59 0.02% 1.04E-02 Ni-63 6.08% 2.89E+00 Zn-65 0.50% 2.38E-01 Sr-90 0.01% 4.94E-03 Tc-99 0.00% 1.34E-03 Cs-134 0.54% 2.55E-01 Cs-137 2.56% 1.22E+00 Page 50 of 106

ANO-1 & 2 Radioactive Effluent Release Report for 2014 Sum of All 4 Categories Waste Class B Nuclide Name Percent Abundance Curies H-3 0.28% 2.86E-01 C-14 4.01% 4.14E+00 Mn-54 1.23% 1.27E+00 Fe-55 20.23% 2.09E+01 Co-57 0.03% 3.27E-02 Co-58 0.07% 7.61E-02 Co-60 29.04% 3.00E+01 Ni-59 0.21% 2.20E-01 Ni-63 27.83% 2.87E+01 Zn-65 0.18% 1.89E-01 Sr-90 0.08% 8.47E-02 Tc-99 0.01% 5.48E-03 Sb-125 0.83% 8.55E-01 Cs-134 1.84% 2.00E+00 Cs-137 13.91% 1.44E+01 Ce-144 0.11% 1.12E-01 Pu-238 0.00% 1.34E-04 Pu-241 0.00% 4.28E-03 Am-241 0.00% 5.64E-05 Page 51 of 106

ANO-1 & 2 Radioactive Effluent Release Report for 2014 Sum of All 4 Categories Waste Class All Nuclide Name Percent Abundance Curies H-3 0.28% 4.16E-01 C-14 4.12% 6.21E+00 Mn-54 2.01% 3.02E+00 Fe-55 18.13% 2.73E+01 Co-57 0.06% 9.20E-02 Co-58 0.20% 2.99E-01 Co-60 41.26% 6.22E+01 Ni-59 0.15% 2.31E-01 Ni-63 20.97% 3.16E+01 Zn-65 0.28% 4.27E-01 Sr-90 0.08% 8.96E-02 Tc-99 0.01% 6.82E-03 Sb-125 0.57% 8.55E-01 Cs-134 1.50% 2.26E+00 Cs-137 10.33% 1.56E+01 Ce-144 0.07% 1.12E-01 Pu-238 0.00% 1.34E-04 Pu-241 0.00% 4.28E-03 Am-241 0.00% 5.64E-05 Waste Burial Volume Manifest Number Date Shipped Volume Used Used RSR 2014-087 8/14/2014 X RSR 2014-090 8/19/2014 X RSR 2014-091 8/21/2014 X RSR 2014-092 9/3/2014 X RSR 2014-099 9/11/2014 X RSR 2014-117 11/4/2014 X RSR 2014-124 12/2/2014 X RSR 2014-126 12/18/2014 X Page 52 of 106

ANO-1 & 2 Radioactive Effluent Release Report for 2014 Solid Waste Shipped Offsite for Disposal and Estimates of Major Nuclides by Waste Class and Stream for Unit 1 During Period From 01/01/2014 to 12/31/2014 Waste Stream: Resins, Filters and Evaporator Bottom Volume Waste Curies Class Ft³ M³ Shipped  % Error (Ci)

A 1.53E+03 4.34E+01 4.75E+01 +/-25%

B 5.52E+02 1.57E+01 1.03E+02 +/-25%

C 0.00E+00 0.00E+00 0.00E+00 +/-25%

All 2.08E+03 5.91E+01 1.51E+02 +/-25%

Waste Stream: Dry Active Waste Volume Waste Curies Class Ft³ M³ Shipped  % Error (Ci)

A 1.62E+03 4.59E+01 5.17E-03 +/-25%

B 0.00E+00 0.00E+00 0.00E+00 +/-25%

C 0.00E+00 0.00E+00 0.00E+00 +/-25%

All 1.62E+03 4.59E+01 5.17E-03 +/-25%

Waste Stream: Irradiated Components Volume Waste Curies Class Ft³ M³ Shipped  % Error (Ci)

A 0.00E+00 0.00E+00 0.00E+00 +/-25%

B 0.00E+00 0.00E+00 0.00E+00 +/-25%

C 0.00E+00 0.00E+00 0.00E+00 +/-25%

All 0.00E+00 0.00E+00 0.00E+00 +/-25%

Page 53 of 106

ANO-1 & 2 Radioactive Effluent Release Report for 2014 Waste Stream: Other Wastes Volume Waste Curies Class Ft³ M³ Shipped  % Error (Ci)

A 0.00E+00 0.00E+00 0.00E+00 +/-25%

B 0.00E+00 0.00E+00 0.00E+00 +/-25%

C 0.00E+00 0.00E+00 0.00E+00 +/-25%

All 0.00E+00 0.00E+00 0.00E+00 +/-25%

Waste Stream: Sum of All 4 Categories Volume Waste Curies Class Ft³ M³ Shipped  % Error (Ci)

A 3.15E+03 8.93E+01 4.75E+01 +/-25%

B 5.52E+02 1.58E+01 1.03E+02 +/-25%

C 0.00E+00 0.00E+00 0.00E+00 +/-25%

All 3.70E+03 1.05E+02 1.51E+02 +/-25%

Number of Shipments Mode of Transportation Destination 6 Hittman Transport Bear Creek Operations 1 Landstar Ranger Bear Creek Operations 2 Hittman Transport Erwin Resin Solution Gallaher Road 1 Hittman Transport Operations Page 54 of 106

ANO-1 & 2 Radioactive Effluent Release Report for 2014 Resins, Filters and Evaporator Bottoms Waste Class A Nuclide Name Percent Abundance Curies H-3 0.27% 1.30E-01 C-14 4.36% 2.07E+00 Mn-54 3.68% 1.75E+00 Fe-55 13.58% 6.45E+00 Co-57 0.12% 5.93E-02 Co-58 0.47% 2.23E-01 Co-60 67.77% 3.22E+01 Ni-59 0.02% 1.04E-02 Ni-63 6.08% 2.89E+00 Zn-65 0.50% 2.38E-01 Sr-90 0.01% 4.94E-03 Tc-99 0.00% 1.34E-03 Cs-134 0.54% 2.55E-01 Cs-137 2.59% 1.23E+00 Resins, Filters and Evaporator Bottoms Waste Class B Nuclide Name Percent Abundance Curies H-3 0.28% 2.86E-01 C-14 4.01% 4.14E+00 Mn-54 1.23% 1.27E+00 Fe-55 20.23% 2.09E+01 Co-57 0.03% 3.27E-02 Co-58 0.07% 7.61E-02 Co-60 29.04% 3.00E+01 Ni-59 0.21% 2.20E-01 Ni-63 27.83% 2.87E+01 Zn-65 0.18% 1.89E-01 Sr-90 0.08% 8.47E-02 Tc-99 0.01% 5.48E-03 Sb-125 0.83% 8.55E-01 Cs-134 1.94% 2.00E+00 Cs-137 13.91% 1.44E+01 Ce-144 0.11% 1.12E-01 Pu-238 0.00% 1.34E-04 Pu-241 0.00% 4.28E-03 Am-241 0.00% 5.64E-05 Page 55 of 106

ANO-1 & 2 Radioactive Effluent Release Report for 2014 Resins, Filters and Evaporator Bottoms Waste Class All Nuclide Name Percent Abundance Curies H-3 0..28% 4.16E-01 C-14 4.12% 6.21E+00 Mn-54 2.00% 3.02E+00 Fe-55 18.12% 2.73E+01 Co-57 0.06% 9.20E-02 Co-58 0.20% 2.99E-01 Co-60 41.26% 6.22E+01 Ni-59 0.15% 2.31E-01 Ni-63 20.97% 3.16E+01 Zn-65 0.28% 4.27E-01 Sr-90 0.06% 8.96E-02 Tc-99 0.00% 6.82E-03 Sb-125 0.57% 8.55E-01 Cs-134 1.50% 2.26E+00 Cs-137 10.34% 3.97E+00 Ce-144 0.07% 1.12E-01 Pu-238 0.00% 1.34E-04 Pu-241 0.00% 4.28E-03 Am-241 0.00% 5.64E-05 Dry Active Waste Waste Class A Nuclide Name Percent Abundance Curies C-14 1.11% 5.76E-05 Cr-51 0.28% 1.47E-05 Mn-54 1.14% 5.90E-05 Fe-55 32.13% 1.66E-03 Co-57 0.31% 1.61E-05 Co-58 0.49% 2.51E-05 Co-60 24.00% 1.24E-03 Ni-63 29.03% 1.50E-03 Zr-95 0.51% 2.61E-05 Nb-95 0.90% 4.64E-05 Sb-125 0.28% 1.44E-05 Cs-134 0.09% 4.70E-06 Cs-137 9.73% 5.03E-04 Dry Active Waste Page 56 of 106

ANO-1 & 2 Radioactive Effluent Release Report for 2014 Waste Class All Nuclide Name Percent Abundance Curies C-14 1.11% 5.76E-05 Cr-51 0.28% 1.47E-05 Mn-54 1.14% 5.90E-05 Fe-55 32.13% 1.66E-03 Co-57 0.31% 1.61E-05 Co-58 0.49% 2.51E-05 Co-60 24.00% 1.24E-03 Ni-63 29.03% 1.50E-03 Zr-95 0.51% 2.61E-05 Nb-95 0.90% 4.64E-05 Sb-125 0.28% 1.44E-05 Cs-134 0.09% 4.70E-06 Cs-137 9.73% 5.03E-04 Sum of All 4 Categories Waste Class A Nuclide Name Percent Abundance Curies H-3 0.27% 1.30E-01 C-14 4.36% 2.07E+00 Cr-51 0.00% 1.47E-05 Mn-54 3.68% 1.75E+00 Fe-55 13.58% 6.45E+00 Co-57 0.12% 5.93E-02 Co-58 0.47% 2.23E-01 Co-60 67.77% 3.22E+01 Ni-59 0.02% 1.04E-02 Ni-63 6.08% 2.89E+00 Zn-65 0.50% 2.38E-01 Sr-90 0.01% 4.94E-03 Zr-95 0.00% 2.61E-05 Nb-95 0.00% 4.64E-05 Tc-99 0.00% 1.34E-03 Sb-125 0.00% 1.44E-05 Cs-134 0.54% 2.55E-01 Cs-137 2.59% 1.23E+00 Page 57 of 106

ANO-1 & 2 Radioactive Effluent Release Report for 2014 Sum of All 4 Categories Waste Class B Nuclide Name Percent Abundance Curies H-3 0.28% 2.86E-01 C-14 4.01% 4.14E+00 Mn-54 1.23% 1.27E+00 Fe-55 20.23% 2.09E+01 Co-57 0.03% 3.27E-02 Co-58 0.07% 7.61E-02 Co-60 29.04% 3.00E+01 Ni-59 0.21% 2.20E-01 Ni-63 27.83% 2.87E+01 Zn-65 0.18% 1.89E-01 Sr-90 0.08% 8.47E-02 Tc-99 0.01% 5.48E-03 Sb-125 0.83% 8.55E-01 Cs-134 1.94% 2.00E+00 Cs-137 13.91% 1.44E+01 Ce-144 0.11% 1.12E-01 Pu-238 0.00% 1.34E-04 Pu-241 0.00% 4.28E-03 Am-241 0.00% 5.64E-05 Page 58 of 106

ANO-1 & 2 Radioactive Effluent Release Report for 2014 Sum of All 4 Categories Waste Class All Nuclide Name Percent Abundance Curies H-3 0.28% 4.16E-01 C-14 4.12% 6.21E+00 Cr-51 0.00% 1.47E-05 Mn-54 2.01% 3.02E+00 Fe-55 18.13% 2.73E+01 Co-57 0.06% 9.20E-02 Co-58 0.20% 2.99E-01 Co-60 41.25% 6.22E+01 Ni-59 0.15% 2.31E-01 Ni-63 20.97% 3.16E+01 Zn-65 0.28% 4.27E-01 Sr-90 0.06% 8.96E-02 Zr-95 0.00% 2.61E-05 Tc-99 0.00% 6.82E-03 Sb-125 0.57% 8.55E-01 Cs-134 1.50% 2.26E+00 Cs-137 10.34% 3.97E+00 Ce-144 0.07% 1.12E-01 Pu-238 0.00% 1.34E-04 Pu-241 0.00% 4.28E-03 Am-241 0.00% 5.64E-05 Waste Burial Volume Manifest Number Date Shipped Volume Used Used RSR 2014-016 3/11/2014 X RSR 2014-045 4/24/2014 X RSR 2014-087 8/14/2014 X RSR 2014-090 8/19/2014 X RSR 2014-091 8/21/2014 X RSR 2014-092 9/3/2014 X RSR 2014-099 9/11/2014 X RSR 2014-117 11/4/2014 X RSR 2014-124 12/2/2014 X RSR 2014-126 12/18/2014 X Page 59 of 106

ANO-1 & 2 Radioactive Effluent Release Report for 2014 The summary for solid waste shipments for Unit 2 is as follows:

Solid Waste Shipped Offsite for Disposal and Estimates of Major Nuclides by Waste Class and Stream During Period From 01/01/2014 to 06/30/2014 Waste Stream: Resins, Filters and Evaporator Bottom Volume Waste Curies Class Ft³ M³ Shipped  % Error (Ci)

A 1.44E+03 4.08E+01 1.77E-02 +/-25%

B 0.00E+00 0.00E+00 0.00E+00 +/-25%

C 0.00E+00 0.00E+00 0.00E+00 +/-25%

All 1.44E+03 4.08E+01 1.77E-02 +/-25%

Waste Stream: Dry Active Waste Volume Waste Curies Class Ft³ M³ Shipped  % Error (Ci)

A 0.00E+00 0.00E+00 0.00E+00 +/-25%

B 0.00E+00 0.00E+00 0.00E+00 +/-25%

C 0.00E+00 0.00E+00 0.00E+00 +/-25%

All 0.00E+00 0.00E+00 0.00E+00 +/-25%

Waste Stream: Irradiated Components Volume Waste Curies Class Ft³ M³ Shipped  % Error (Ci)

A 0.00E+00 0.00E+00 0.00E+00 +/-25%

B 0.00E+00 0.00E+00 0.00E+00 +/-25%

C 0.00E+00 0.00E+00 0.00E+00 +/-25%

All 0.00E+00 0.00E+00 0.00E+00 +/-25%

Page 60 of 106

ANO-1 & 2 Radioactive Effluent Release Report for 2014 Waste Stream: Other Wastes Volume Waste Curies Class Ft³ M³ Shipped  % Error (Ci)

A 0.00E+00 0.00E+00 0.00E+00 +/-25%

B 0.00E+00 0.00E+00 0.00E+00 +/-25%

C 0.00E+00 0.00E+00 0.00E+00 +/-25%

All 0.00E+00 0.00E+00 0.00E+00 +/-25%

Waste Stream: Sum of All 4 Categories Volume Waste Curies Class Ft³ M³ Shipped  % Error (Ci)

A 1.44E+03 4.08E+01 1.77E-02 +/-25%

B 0.00E+00 0.00E+00 0.00E+00 +/-25%

C 0.00E+00 0.00E+00 0.00E+00 +/-25%

All 1.44E+03 4.08E+01 1.77E-02 +/-25%

Number of Shipments Mode of Transportation Destination 2 Hittman Transport Bear Creek Operations Resins, Filters and Evaporator Bottoms Waste Class A Nuclide Name Percent Abundance Curies Co-60 23.21% 4.11E-03 Cs-137 76.79% 1.36E-02 Page 61 of 106

ANO-1 & 2 Radioactive Effluent Release Report for 2014 Resins, Filters and Evaporator Bottoms Waste Class ALL Nuclide Name Percent Abundance Curies Co-60 23.21% 4.11E-03 Cs-137 76.79% 1.36E-02 Sum of All 4 Categories Waste Class A Nuclide Name Percent Abundance Curies Co-60 23.21% 4.11E-03 Cs-137 76.79% 1.36E-02 Sum of All 4 Categories Waste Class All Nuclide Name Percent Abundance Curies Co-60 23.21% 4.11E-03 Cs-137 76.79% 1.36E-02 Waste Burial Manifest Number Date Shipped Volume Used Volume Used RSR 2014-041 4/10/2014 X RSR 2014-043 4/16/2014 X Page 62 of 106

ANO-1 & 2 Radioactive Effluent Release Report for 2014 Solid Waste Shipped Offsite for Disposal and Estimates of Major Nuclides by Waste Class and Stream for Unit 2 During Period From 07/01/2014 to 12/31/2014 Waste Stream: Resins, Filters and Evaporator Bottom Volume Waste Curies Class Ft³ M³ Shipped  % Error (Ci)

A 2.26E+03 6.41E+01 2.03E+01 +/-25%

B 1.89E+02 5.35E+00 7.29E+01 +/-25%

C 0.00E+00 0.00E+00 0.00E+00 +/-25%

All 2.45E+03 6.95E+01 9.32E+01 +/-25%

Waste Stream: Dry Active Waste Volume Waste Curies Class Ft³ M³ Shipped  % Error (Ci)

A 0.00E+00 0.00E+00 0.00E+00 +/-25%

B 0.00E+00 0.00E+00 0.00E+00 +/-25%

C 0.00E+00 0.00E+00 0.00E+00 +/-25%

All 0.00E+00 0.00E+00 0.00E+00 +/-25%

Waste Stream: Irradiated Components Volume Waste Curies Class Ft³ M³ Shipped  % Error (Ci)

A 0.00E+00 0.00E+00 0.00E+00 +/-25%

B 0.00E+00 0.00E+00 0.00E+00 +/-25%

C 0.00E+00 0.00E+00 0.00E+00 +/-25%

All 0.00E+00 0.00E+00 0.00E+00 +/-25%

Page 63 of 106

ANO-1 & 2 Radioactive Effluent Release Report for 2014 Waste Stream: Other Wastes Volume Waste Curies Class Ft³ M³ Shipped  % Error (Ci)

A 0.00E+00 0.00E+00 0.00E+00 +/-25%

B 0.00E+00 0.00E+00 0.00E+00 +/-25%

C 0.00E+00 0.00E+00 0.00E+00 +/-25%

All 0.00E+00 0.00E+00 0.00E+00 +/-25%

Waste Stream: Sum of All 4 Categories Volume Waste Curies Class Ft³ M³ Shipped  % Error (Ci)

A 2.26E+03 6.41E+01 2.03E+01 +/-25%

B 1.89E+02 5.35E+00 7.29E+01 +/-25%

C 0.00E+00 0.00E+00 0.00E+00 +/-25%

All 2.45E+03 6.95E+01 9.32E+01 +/-25%

Number of Shipments Mode of Transportation Destination 4 Hittman Transport Bear Creek Operations 1 Southern Pines Bear Creek Operations Gallaher Road 1 Hittman Transport Operations Page 64 of 106

ANO-1 & 2 Radioactive Effluent Release Report for 2014 Resins, Filters and Evaporator Bottoms Waste Class A Nuclide Name Percent Abundance Curies C-14 2.39% 4.65E-01 Mn-54 2.56% 5.22E-01 Fe-55 34.01% 6.93E+00 Co-57 0.06% 1.30E-02 Co-58 0.01% 1.28E-03 Co-60 17.28% 3.52E+00 Ni-59 0.27% 5.60E-02 Ni-63 40.19% 8.19E+00 Zn-65 0.28% 5.70E-02 Sr-90 0.05% 9.36E-03 Tc-99 0.01% 1.55E-03 Ag-110m 0.01% 2.05E-03 Sn-113 0.00% 4.12E-04 Sb-125 1.69% 3.45E-01 Cs-134 0.16% 3.27E-02 Cs-137 1.01% 2.05E-01 Ce-144 0.02% 4.52E-03 Page 65 of 106

ANO-1 & 2 Radioactive Effluent Release Report for 2014 Resins, Filters and Evaporator Bottoms Waste Class B Nuclide Name Percent Abundance Curies C-14 1.22% 8.92E-01 Mn-54 1.99% 1.45E+00 Fe-55 17.70% 1.29E+01 Co-57 0.07% 5.12E-02 Co-58 0.12% 9.04E-02 Co-60 14.82% 1.08E+01 Ni-59 0.34% 2.50E-01 Ni-63 47.20% 3.44E+01 Zn-65 0.33% 2.44E-01 Sr-90 0.05% 3.50E-02 Sb-125 2.73% 1.99E+00 Cs-134 1.96% 1.43E+00 Cs-137 11.31% 8.24E+00 Ce-144 0.14% 1.01E-01 Pu-238 0.00% 1.55E-04 Am-241 0.00% 8.41E-05 Cm-242 0.00% 1.70E-05 Cm-243 0.00% 4.88E-05 Cm-244 0.00% 4.83E-05 Page 66 of 106

ANO-1 & 2 Radioactive Effluent Release Report for 2014 Resins, Filters and Evaporator Bottoms Waste Class All Nuclide Name Percent Abundance Curies C-14 1.48% 1.36E+00 Mn-54 2.11% 1.97E+00 Fe-55 21.27% 1.98E+01 Co-57 0.07% 6.42E-02 Co-58 0.10% 9.17E-02 Co-60 15.36% 1.43E+01 Ni-59 0.33% 3.06E-01 Ni-63 45.67% 4.26E+01 Zn-65 0.32% 3.01E-01 Sr-90 0.05% 4.44E-02 Tc-99 0.00% 1.55E-03 Ag-110m 0.00% 2.05E-03 Sn-113 0.00% 4.12E-04 Sb-125 2.50% 2.34E+00 Cs-134 1.57% 1.46E+00 Cs-137 9.06% 8.45E+00 Ce-144 0.11% 1.06E-01 Pu-238 0.00% 1.55E-04 Am-241 0.00% 8.41E-05 Cm-242 0.00% 1.70E-05 Cm-243 0.00% 4.88E-05 Cm-244 0.00% 4.83E-05 Page 67 of 106

ANO-1 & 2 Radioactive Effluent Release Report for 2014 Sum of All 4 Categories Waste Class A Nuclide Name Percent Abundance Curies C-14 2.28% 4.65E-01 Mn-54 2.56% 5.22E-01 Fe-55 34.05% 6.93E+00 Co-57 0.06% 1.30E-02 Co-58 0.01% 1.28E-03 Co-60 17.29% 3.52E+00 Ni-59 0.28% 5.60E-02 Ni-63 40.24% 8.19E+00 Zn-65 0.28% 5.70E-02 Sr-90 0.05% 9.36E-03 Tc-99 0.01% 1.55E-03 Ag-110m 0.01% 2.05E-03 Sn-113 0.00% 4.12E-04 Sb-125 1.69% 3.45E-01 Cs-134 0.16% 3.27E-02 Cs-137 1.01% 2.05E-01 Ce-144 0.02% 4.52E-03 Page 68 of 106

ANO-1 & 2 Radioactive Effluent Release Report for 2014 Sum of All 4 Categories Waste Class B Nuclide Name Percent Abundance Curies H-3 0.04% 3.13E-02 C-14 1.22% 8.92E-01 Mn-54 1.99% 1.45E+00 Fe-55 17.69% 1.29E+01 Co-57 0.07% 5.12E-02 Co-58 0.12% 9.04E-02 Co-60 14.81% 1.08E+01 Ni-59 0.34% 2.50E-01 Ni-63 47.18% 3.44E+01 Zn-65 0.33% 2.44E-01 Sr-90 0.05% 3.50E-02 Tc-99 0.01% 4.99E-03 Sb-125 2.73% 1.99E+00 I-129 0.00% 1.93E-03 Cs-134 1.96% 1.43E+00 Cs-137 11.30% 8.24E+00 Ce-144 0.14% 1.01E-01 Pu-238 0.00% 1.55E-04 Am-241 0.00% 8.41E-05 Cm-242 0.00% 1.70E-05 Cm-243 0.00% 4.88E-05 Cm-244 0.00% 4.83E-05 Page 69 of 106

ANO-1 & 2 Radioactive Effluent Release Report for 2014 Sum of All 4 Categories Waste Class All Nuclide Name Percent Abundance Curies H-3 0.03% 3.13E-02 C-14 1.45% 1.36E+00 Mn-54 2.11% 1.97E+00 Fe-55 21.26% 1.98E+01 Co-57 0.07% 6.42E-02 Co-58 0.10% 9.17E-02 Co-60 15.35% 1.43E+01 Ni-59 0.33% 3.06E-01 Ni-63 45.66% 4.26E+01 Zn-65 0.32% 3.01E-01 Sr-90 0.05% 4.44E-02 Tc-99 0.01% 6.54E-03 Ag-110m 0.00% 2.05E-03 Sn-113 0.00% 4.12E-04 Sb-125 2.50% 2.34E+00 I-129 0.00% 1.93E-03 Cs-134 1.57% 1.46E+00 Cs-137 9.05% 8.45E+00 Ce-144 0.11% 1.06E-01 Pu-238 0.00% 1.55E-04 Am-241 0.00% 8.41E-05 Cm-242 0.00% 1.70E-05 Cm-243 0.00% 4.88E-05 Cm-244 0.00% 4.83E-05 Waste Burial Volume Manifest Number Date Shipped Volume Used Used RSR 2014-082 7/29/2014 X RSR 2014-083 7/31/2014 X RSR 2014-086 8/12/2014 X RSR 2014-094 9/9/2014 X RSR 2014-119 11/6/2014 X RSR 2014-125 12/16/2014 X Page 70 of 106

ANO-1 & 2 Radioactive Effluent Release Report for 2014 Solid Waste Shipped Offsite for Disposal and Estimates of Major Nuclides by Waste Class and Stream for Unit 2 During Period From 01/01/2014 to 12/31/2014 Waste Stream: Resins, Filters and Evaporator Bottom Volume Waste Curies Class Ft³ M³ Shipped  % Error (Ci)

A 3.70E+03 1.05E+02 2.04E+01 +/-25%

B 1.89E+02 5.35E+00 7.29E+01 +/-25%

C 0.00E+00 0.00E+00 0.00E+00 +/-25%

All 3.89E+03 1.10E+02 9.33E+01 +/-25%

Waste Stream: Dry Active Waste Volume Waste Curies Class Ft³ M³ Shipped  % Error (Ci)

A 0.00E+00 0.00E+00 0.00E+00 +/-25%

B 0.00E+00 0.00E+00 0.00E+00 +/-25%

C 0.00E+00 0.00E+00 0.00E+00 +/-25%

All 0.00E+00 0.00E+00 0.00E+00 +/-25%

Waste Stream: Irradiated Components Volume Waste Curies Class Ft³ M³ Shipped  % Error (Ci)

A 0.00E+00 0.00E+00 0.00E+00 +/-25%

B 0.00E+00 0.00E+00 0.00E+00 +/-25%

C 0.00E+00 0.00E+00 0.00E+00 +/-25%

All 0.00E+00 0.00E+00 0.00E+00 +/-25%

Page 71 of 106

ANO-1 & 2 Radioactive Effluent Release Report for 2014 Waste Stream: Other Wastes Volume Waste Curies Class Ft³ M³ Shipped  % Error (Ci)

A 0.00E+00 0.00E+00 0.00E+00 +/-25%

B 0.00E+00 0.00E+00 0.00E+00 +/-25%

C 0.00E+00 0.00E+00 0.00E+00 +/-25%

All 0.00E+00 0.00E+00 0.00E+00 +/-25%

Waste Stream: Sum of All 4 Categories Volume Waste Curies Class Ft³ M³ Shipped  % Error (Ci)

A 3.70E+03 1.05E+02 2.04E+01 +/-25%

B 1.89E+02 5.35E+00 7.29E+01 +/-25%

C 0.00E+00 0.00E+00 0.00E+00 +/-25%

All 3.89E+03 1.10E+02 9.33E+01 +/-25%

Number of Shipments Mode of Transportation Destination 6 Hittman Transport Bear Creek Operations 1 Sothern Pines Bear Creek Operations Gallaher Road 1 Hittman Transport Operations Page 72 of 106

ANO-1 & 2 Radioactive Effluent Release Report for 2014 Resins, Filters and Evaporator Bottoms Waste Class A Nuclide Name Percent Abundance Curies C-14 2.38% 4.86E-01 Mn-54 2.56% 5.22E-01 Fe-55 33.98% 6.93E+00 Co-57 0.06% 1.30E-02 Co-58 0.01% 1.28E-03 Co-60 17.28% 3.52E+00 Ni-59 0.27% 5.60E-02 Ni-63 40.16% 8.19E+00 Zn-65 0.28% 5.70E-02 Sr-90 0.05% 9.36E-03 Tc-99 0.01% 1.55E-03 Ag-110m 0.01% 2.05E-03 Sn-113 0.00% 4.12E-04 Sb-125 1.69% 3.45E-01 Cs-134 0.16% 3.27E-02 Cs-137 1.07% 2.19E-01 Ce-144 0.02% 4.52E-03 Page 73 of 106

ANO-1 & 2 Radioactive Effluent Release Report for 2014 Resins, Filters and Evaporator Bottoms Waste Class B Nuclide Name Percent Abundance Curies C-14 1.22% 8.92E-01 Mn-54 1.99% 1.45E+00 Fe-55 17.70% 1.29E+01 Co-57 0.07% 5.12E-02 Co-58 0.12% 9.04E-02 Co-60 14.82% 1.08E+01 Ni-59 0.34% 2.50E-01 Ni-63 47.20% 3.44E+01 Zn-65 0.33% 2.44E-01 Sr-90 0.05% 3.50E-02 Sb-125 2.73% 1.99E+00 Cs-134 1.96% 1.43E+00 Cs-137 11.31% 8.24E+00 Ce-144 0.14% 1.01E-01 Pu-238 0.00% 1.55E-04 Am-241 0.00% 8.41E-05 Cm-242 0.00% 1.70E-05 Cm-243 0.00% 4.88E-05 Cm-244 0.00% 4.83E-05 Page 74 of 106

ANO-1 & 2 Radioactive Effluent Release Report for 2014 Resins, Filters and Evaporator Bottoms Waste Class All Nuclide Name Percent Abundance Curies C-14 1.48% 1.38E+00 Mn-54 2.11% 1.97E+00 Fe-55 21.26% 1.98E+01 Co-57 0.07% 6.42E-02 Co-58 0.10% 9.17E-02 Co-60 15.36% 1.43E+01 Ni-59 0.33% 3.06E-01 Ni-63 45.66% 4.26E+01 Zn-65 0.32% 3.01E-01 Sr-90 0.05% 4.44E-02 Tc-99 0.00% 1.55E-03 Ag-110m 0.00% 2.05E-03 Sn-113 0.00% 4.12E-04 Sb-125 2.50% 2.34E+00 Cs-134 1.57% 1.46E+00 Cs-137 9.07% 8.46E+00 Ce-144 0.11% 1.06E-01 Pu-238 0.00% 1.55E-04 Am-241 0.00% 8.41E-05 Cm-242 0.00% 1.70E-05 Cm-243 0.00% 4.88E-05 Cm-244 0.00% 4.83E-05 Page 75 of 106

ANO-1 & 2 Radioactive Effluent Release Report for 2014 Sum of All 4 Categories Waste Class A Nuclide Name Percent Abundance Curies C-14 2.38% 4.86E-01 Mn-54 2.56% 5.22E-01 Fe-55 33.98% 6.93E+00 Co-57 0.06% 1.30E-02 Co-58 0.01% 1.28E-03 Co-60 17.28% 3.52E+00 Ni-59 0.27% 5.60E-02 Ni-63 40.16% 8.19E+00 Zn-65 0.28% 5.70E-02 Sr-90 0.05% 9.36E-03 Tc-99 0.01% 1.55E-03 Ag-110m 0.01% 2.05E-03 Sn-113 0.00% 4.12E-04 Sb-125 1.69% 3.45E-01 Cs-134 0.16% 3.27E-02 Cs-137 1.07% 2.19E-01 Ce-144 0.02% 4.52E-03 Page 76 of 106

ANO-1 & 2 Radioactive Effluent Release Report for 2014 Sum of All 4 Categories Waste Class B Nuclide Name Percent Abundance Curies C-14 1.22% 8.92E-01 Mn-54 1.99% 1.45E+00 Fe-55 17.70% 1.29E+01 Co-57 0.07% 5.12E-02 Co-58 0.12% 9.04E-02 Co-60 14.82% 1.08E+01 Ni-59 0.34% 2.50E-01 Ni-63 47.20% 3.44E+01 Zn-65 0.33% 2.44E-01 Sr-90 0.05% 3.50E-02 Sb-125 2.73% 1.99E+00 Cs-134 1.96% 1.43E+00 Cs-137 11.31% 8.24E+00 Ce-144 0.14% 1.01E-01 Pu-238 0.00% 1.55E-04 Am-241 0.00% 8.41E-05 Cm-242 0.00% 1.70E-05 Cm-243 0.00% 4.88E-05 Cm-244 0.00% 4.83E-05 Page 77 of 106

ANO-1 & 2 Radioactive Effluent Release Report for 2014 Sum of All 4 Categories Waste Class All Nuclide Name Percent Abundance Curies C-14 1.48% 1.38E+00 Mn-54 2.11% 1.97E+00 Fe-55 21.26% 1.98E+01 Co-57 0.07% 6.42E-02 Co-58 0.10% 9.17E-02 Co-60 15.36% 1.43E+01 Ni-59 0.33% 3.06E-01 Ni-63 45.66% 4.26E+01 Zn-65 0.32% 3.01E-01 Sr-90 0.05% 4.44E-02 Tc-99 0.00% 1.55E-03 Ag-110m 0.00% 2.05E-03 Sn-113 0.00% 4.12E-04 Sb-125 2.50% 2.34E+00 Cs-134 1.57% 1.46E+00 Cs-137 9.07% 8.46E+00 Ce-144 0.11% 1.06E-01 Pu-238 0.00% 1.55E-04 Am-241 0.00% 8.41E-05 Cm-242 0.00% 4.88E-05 Cm-243 0.00% 4.88E-05 Cm-244 0.00% 4.83E-05 Waste Burial Volume Manifest Number Date Shipped Volume Used Used RSR 2014-041 4/10/2014 X RSR 2014-043 4/16/2014 X RSR 2014-082 7/29/2014 X RSR 2014-083 7/31/2014 X RSR 2014-086 8/12/2014 X RSR 2014-094 9/9/2014 X RSR 2014-119 11/6/2014 X RSR 2014-125 12/16/2014 X Page 78 of 106

ANO-1 & 2 Radioactive Effluent Release Report for 2014 The summary for solid waste shipments for Unit 1 and 2 Consolidated Shipments is as follows:

Solid Waste Shipped Offsite for Disposal and Estimates of Major Nuclides by Waste Class and Stream During Period From 01/01/2014 to 06/30/2014 Waste Stream: Resins, Filters and Evaporator Bottom Volume Waste Curies Class Ft³ M³ Shipped  % Error (Ci)

A 0.00E+00 0.00E+00 0.00E+00 +/-25%

B 0.00E+00 0.00E+00 0.00E+00 +/-25%

C 0.00E+00 0.00E+00 0.00E+00 +/-25%

All 0.00E+00 0.00E+00 0.00E+00 +/-25%

Waste Stream: Dry Active Waste Volume Waste Curies Class Ft³ M³ Shipped  % Error (Ci)

A 1.10E+04 3.11E+02 4.99E-01 +/-25%

B 0.00E+00 0.00E+00 0.00E+00 +/-25%

C 0.00E+00 0.00E+00 0.00E+00 +/-25%

All 1.10E+04 3.11E+02 4.99E-01 +/-25%

Page 79 of 106

ANO-1 & 2 Radioactive Effluent Release Report for 2014 Waste Stream: Irradiated Components Volume Waste Curies Class Ft³ M³ Shipped  % Error (Ci)

A 0.00E+00 0.00E+00 0.00E+00 +/-25%

B 0.00E+00 0.00E+00 0.00E+00 +/-25%

C 0.00E+00 0.00E+00 0.00E+00 +/-25%

All 0.00E+00 0.00E+00 0.00E+00 +/-25%

Waste Stream: Other Wastes Volume Waste Curies Class Ft³ M³ Shipped  % Error (Ci)

A 0.00E+00 0.00E+00 0.00E+00 +/-25%

B 0.00E+00 0.00E+00 0.00E+00 +/-25%

C 0.00E+00 0.00E+00 0.00E+00 +/-25%

All 0.00E+00 0.00E+00 0.00E+00 +/-25%

Waste Stream: Sum of All 4 Categories Volume Waste Curies Class Ft³ M³ Shipped  % Error (Ci)

A 1.10E+04 3.11E+02 4.99E-01 +/-25%

B 0.00E+00 0.00E+00 0.00E+00 +/-25%

C 0.00E+00 0.00E+00 0.00E+00 +/-25%

All 1.10E+04 3.11E+02 4.99E-01 +/-25%

Number of Shipments Mode of Transportation Destination 6 Hittman Transport Bear Creek Operations 1 Landstar Ranger Bear Creek Operations Page 80 of 106

ANO-1 & 2 Radioactive Effluent Release Report for 2014 Dry Active Waste Waste Class A Nuclide Name Percent Abundance Curies C-14 1.12% 5.59E-03 Cr-51 0.25% 1.23E-03 Mn-54 1.13% 5.66E-03 Fe-55 32.20% 1.61E-01 Co-57 0.31% 1.54E-03 Co-58 0.46% 2.30E-03 Co-60 23.99% 1.20E-01 Ni-63 29.12% 1.45E-01 Zr-95 0.48% 2.38E-03 Nb-95 0.80% 3.99E-03 Sb-125 0.28% 1.39E-03 Cs-134 0.09% 4.54E-04 Cs-137 9.78% 4.89E-02 Dry Active Waste Waste Class ALL Nuclide Name Percent Abundance Curies C-14 1.12% 5.59E-03 Cr-51 0.25% 1.23E-03 Mn-54 1.13% 5.66E-03 Fe-55 32.20% 1.61E-01 Co-57 0.31% 1.54E-03 Co-58 0.46% 2.30E-03 Co-60 23.99% 1.20E-01 Ni-63 29.12% 1.45E-01 Zr-95 0.48% 2.38E-03 Nb-95 0.80% 3.99E-03 Sb-125 0.28% 1.39E-03 Cs-134 0.09% 4.54E-04 Cs-137 9.78% 4.89E-02 Page 81 of 106

ANO-1 & 2 Radioactive Effluent Release Report for 2014 Sum of All 4 Categories Waste Class A Nuclide Name Percent Abundance Curies C-14 1.12% 5.59E-03 Cr-51 0.25% 1.23E-03 Mn-54 1.13% 5.66E-03 Fe-55 32.20% 1.61E-01 Co-57 0.31% 1.54E-03 Co-58 0.46% 2.30E-03 Co-60 23.99% 1.20E-01 Ni-63 29.12% 1.45E-01 Zr-95 0.48% 2.38E-03 Nb-95 0.80% 3.99E-03 Sb-125 0.28% 1.39E-03 Cs-134 0.09% 4.54E-04 Cs-137 9.78% 4.89E-02 Sum of All 4 Categories Waste Class All Nuclide Name Percent Abundance Curies C-14 1.12% 5.59E-03 Cr-51 0.25% 1.23E-03 Mn-54 1.13% 5.66E-03 Fe-55 32.20% 1.61E-01 Co-57 0.31% 1.54E-03 Co-58 0.46% 2.30E-03 Co-60 23.99% 1.20E-01 Ni-63 29.12% 1.45E-01 Zr-95 0.48% 2.38E-03 Nb-95 0.80% 3.99E-03 Sb-125 0.28% 1.39E-03 Cs-134 0.09% 4.54E-04 Cs-137 9.78% 4.89E-02 Page 82 of 106

ANO-1 & 2 Radioactive Effluent Release Report for 2014 Waste Volume Burial Volume Manifest Number Date Shipped Used Used RSR 2014-025 3/25/2014 X RSR 2014-044 4/17/2014 X RSR 2014-047 5/1/2014 X RSR 2014-055 5/21/2014 X RSR 2014-056 5/29/2014 X RSR 2014-063 6/3/2014 X RSR 2014-070 6/23/2014 X Page 83 of 106

ANO-1 & 2 Radioactive Effluent Release Report for 2014 The summary for solid waste shipments for Unit 1 and 2 Consolidated Shipments is as follows:

Solid Waste Shipped Offsite for Disposal and Estimates of Major Nuclides by Waste Class and Stream During Period From 07/01/2014 to12/31/2014 Waste Stream: Resins, Filters and Evaporator Bottom Volume Waste Curies Class Ft³ M³ Shipped  % Error (Ci)

A 0.00E+00 0.00E+00 0.00E+00 +/-25%

B 0.00E+00 0.00E+00 0.00E+00 +/-25%

C 0.00E+00 0.00E+00 0.00E+00 +/-25%

All 0.00E+00 0.00E+00 0.00E+00 +/-25%

Waste Stream: Dry Active Waste Volume Waste Curies Class Ft³ M³ Shipped  % Error (Ci)

A 5.24E+03 1.48E+02 1.34E-01 +/-25%

B 0.00E+00 0.00E+00 0.00E+00 +/-25%

C 0.00E+00 0.00E+00 0.00E+00 +/-25%

All 5.24E+03 1.48E+02 1.34E-01 +/-25%

Waste Stream: Irradiated Components Volume Waste Curies Class Ft³ M³ Shipped  % Error (Ci)

A 0.00E+00 0.00E+00 0.00E+00 +/-25%

B 0.00E+00 0.00E+00 0.00E+00 +/-25%

C 0.00E+00 0.00E+00 0.00E+00 +/-25%

All 0.00E+00 0.00E+00 0.00E+00 +/-25%

Page 84 of 106

ANO-1 & 2 Radioactive Effluent Release Report for 2014 Waste Stream: Other Wastes Volume Waste Curies Class Ft³ M³ Shipped  % Error (Ci)

A 0.00E+00 0.00E+00 0.00E+00 +/-25%

B 0.00E+00 0.00E+00 0.00E+00 +/-25%

C 0.00E+00 0.00E+00 0.00E+00 +/-25%

All 0.00E+00 0.00E+00 0.00E+00 +/-25%

Waste Stream: Sum of All 4 Categories Volume Waste Curies Class Ft³ M³ Shipped  % Error (Ci)

A 5.24E+03 1.48E+02 1.34E-01 +/-25%

B 0.00E+00 0.00E+00 0.00E+00 +/-25%

C 0.00E+00 0.00E+00 0.00E+00 +/-25%

All 5.24E+03 1.48E+02 1.34E-01 +/-25%

Number of Shipments Mode of Transportation Destination 2 Hittman Transport Bear Creek Operations 1 Landstar System Inc Bear Creek Operations Dry Active Waste Waste Class A Nuclide Name Percent Abundance Curies Co-58 5.00% 6.71E-03 Co-60 22.42% 3.01E-02 Sb-125 20.48% 2.75E-02 Cs-134 3.40% 4.57E-03 Cs-137 48.70% 6.54E-02 Page 85 of 106

ANO-1 & 2 Radioactive Effluent Release Report for 2014 Dry Active Waste Waste Class All Nuclide Name Percent Abundance Curies Co-58 5.00% 6.71E-03 Co-60 22.42% 3.01E-02 Sb-125 20.48% 2.75E-02 Cs-134 3.40% 4.57E-03 Cs-137 48.70% 6.54E-02 Sum of All 4 Categories Waste Class A Nuclide Name Percent Abundance Curies Co-58 5.00% 6.71E-03 Co-60 22.42% 3.01E-02 Sb-125 20.48% 2.75E-02 Cs-134 3.40% 4.57E-03 Cs-137 48.70% 6.54E-02 Sum of All 4 Categories Waste Class All Nuclide Name Percent Abundance Curies Co-58 5.00% 6.71E-03 Co-60 22.42% 3.01E-02 Sb-125 20.48% 2.75E-02 Cs-134 3.40% 4.57E-03 Cs-137 48.70% 6.54E-02 Waste Volume Burial Volume Manifest Number Date Shipped Used Used RSR 2014-110 10/14/2014 X RSR 2014-111 10/16/2014 X RSR 2014-122 11/12/2014 X Page 86 of 106

ANO-1 & 2 Radioactive Effluent Release Report for 2014 The summary for solid waste shipments for Unit 1 and 2 Consolidated Shipments is as follows:

Solid Waste Shipped Offsite for Disposal and Estimates of Major Nuclides by Waste Class and Stream During Period From 01/01/2014 to12/31/2014 Waste Stream: Resins, Filters and Evaporator Bottom Volume Waste Curies Class Ft³ M³ Shipped  % Error (Ci)

A 0.00E+00 0.00E+00 0.00E+00 +/-25%

B 0.00E+00 0.00E+00 0.00E+00 +/-25%

C 0.00E+00 0.00E+00 0.00E+00 +/-25%

All 0.00E+00 0.00E+00 0.00E+00 +/-25%

Waste Stream: Dry Active Waste Volume Waste Curies Class Ft³ M³ Shipped  % Error (Ci)

A 1.62E+04 4.59E+02 6.34E-01 +/-25%

B 0.00E+00 0.00E+00 0.00E+00 +/-25%

C 0.00E+00 0.00E+00 0.00E+00 +/-25%

All 1.62E+04 4.59E+02 6.34E-01 +/-25%

Waste Stream: Irradiated Components Volume Waste Curies Class Ft³ M³ Shipped  % Error (Ci)

A 0.00E+00 0.00E+00 0.00E+00 +/-25%

B 0.00E+00 0.00E+00 0.00E+00 +/-25%

C 0.00E+00 0.00E+00 0.00E+00 +/-25%

All 0.00E+00 0.00E+00 0.00E+00 +/-25%

Page 87 of 106

ANO-1 & 2 Radioactive Effluent Release Report for 2014 Waste Stream: Other Wastes Volume Waste Curies Class Ft³ M³ Shipped  % Error (Ci)

A 0.00E+00 0.00E+00 0.00E+00 +/-25%

B 0.00E+00 0.00E+00 0.00E+00 +/-25%

C 0.00E+00 0.00E+00 0.00E+00 +/-25%

All 0.00E+00 0.00E+00 0.00E+00 +/-25%

Waste Stream: Sum of All 4 Categories Volume Waste Curies Class Ft³ M³ Shipped  % Error (Ci)

A 1.62E+04 4.59E+02 6.34E-01 +/-25%

B 0.00E+00 0.00E+00 0.00E+00 +/-25%

C 0.00E+00 0.00E+00 0.00E+00 +/-25%

All 1.62E+04 4.59E+02 6.34E-01 +/-25%

Number of Shipments Mode of Transportation Destination 8 Hittman Transport Bear Creek Operations 1 Landstar Ranger Bear Creek Operations 1 Landstar System Inc Bear Creek Operations Page 88 of 106

ANO-1 & 2 Radioactive Effluent Release Report for 2014 Dry Active Waste Waste Class A Nuclide Name Percent Abundance Curies C-14 0.88% 5.59E-03 Cr-51 0.19% 1.23E-03 Mn-54 0.89% 5.66E-03 Fe-55 25.37% 1.61E-01 Co-57 0.24% 1.54E-03 Co-58 1.42% 9.01E-03 Co-60 23.66% 1.50E-01 Ni-63 22.95% 1.45E-01 Zr-95 0.38% 2.38E-03 Nb-95 0.63% 3.99E-03 Sb-125 4.56% 2.89E-02 Cs-134 0.79% 5.02E-03 Cs-137 18.03% 1.14E-01 Dry Active Waste Waste Class All Nuclide Name Percent Abundance Curies C-14 0.88% 5.59E-03 Cr-51 0.19% 1.23E-03 Mn-54 0.89% 5.66E-03 Fe-55 25.37% 1.61E-01 Co-57 0.24% 1.54E-03 Co-58 1.42% 9.01E-03 Co-60 23.66% 1.50E-01 Ni-63 22.95% 1.45E-01 Zr-95 0.38% 2.38E-03 Nb-95 0.63% 3.99E-03 Sb-125 4.56% 2.89E-02 Cs-134 0.79% 5.02E-03 Cs-137 18.03% 1.14E-01 Page 89 of 106

ANO-1 & 2 Radioactive Effluent Release Report for 2014 Sum of All 4 Categories Waste Class A Nuclide Name Percent Abundance Curies C-14 0.88% 5.59E-03 Cr-51 0.19% 1.23E-03 Mn-54 0.89% 5.66E-03 Fe-55 25.37% 1.61E-01 Co-57 0.24% 1.54E-03 Co-58 1.42% 9.01E-03 Co-60 23.66% 1.50E-01 Ni-63 22.95% 1.45E-01 Zr-95 0.38% 2.38E-03 Nb-95 0.63% 3.99E-03 Sb-125 4.56% 2.89E-02 Cs-134 0.79% 5.02E-03 Cs-137 18.03% 1.14E-01 Sum of All 4 Categories Waste Class All Nuclide Name Percent Abundance Curies C-14 0.88% 5.59E-03 Cr-51 0.19% 1.23E-03 Mn-54 0.89% 5.66E-03 Fe-55 25.37% 1.61E-01 Co-57 0.24% 1.54E-03 Co-58 1.42% 9.01E-03 Co-60 23.66% 1.50E-01 Ni-63 22.95% 1.45E-01 Zr-95 0.38% 2.38E-03 Nb-95 0.63% 3.99E-03 Sb-125 4.56% 2.89E-02 Cs-134 0.79% 5.02E-03 Cs-137 18.03% 1.14E-01 Page 90 of 106

ANO-1 & 2 Radioactive Effluent Release Report for 2014 Waste Volume Burial Volume Manifest Number Date Shipped Used Used RSR 2014-025 3/25/2014 X RSR 2014-044 4/17/2014 X RSR 2014-047 5/1/2014 X RSR 2014-055 5/21/2014 X RSR 2014-056 5/29/2014 X RSR 2014-063 6/3/2014 X RSR 2014-070 6/23/2014 X RSR 2014-110 10/14/2014 X RSR 2014-111 10/16/2014 X RSR 2014-122 11/12/2014 X Page 91 of 106

ANO-1 & 2 Radioactive Effluent Release Report for 2014

9. UNPLANNED RELEASES An unplanned release is the unintended discharge of a volume of liquid or airborne radioactivity to unrestricted areas.

During 2014, there was one unplanned releases (liquid or gaseous) to an unrestricted area on ANO-1.

On 12/23/2014 Unit One Operations commenced Treated Waste Monitor Tank T-16A liquid release 1LR-2014-0135. At that time, Treated Waste Monitor Tank T-16B was on recirculation for Chemistry sampling. When the T-16A release automatically terminated on low tank level, it was observed release flow indication was approximately 55 gallons per minute (gpm) rather than the expected 0 gpm. After a brief investigation of this anomaly, Operations personnel closed the Liquid Radwaste Isolation Valve CV-4642. Upon closing CV-4642, flow was observed to drop to 0 gpm. During this time T-16B level lowered from 82% to 78.7%. This equates to approximately 400 gallons of water released from T-16B.

Upon a review of the Piping and Instrumentation Diagram of the system, it was determined that one possible leakage path to explain this condition was the Treated Waste Discharge Valve CZ-55B to the Header from T-16B. CZ-55B is a reach-rod operated manual valve on the 326' elevation of the Unit 1 Auxiliary Building. The pin on the position indicator was observed not bottomed out, and the handwheel would not turn any further in the closed direction. Operations personnel went behind the wall and found CZ-55B to be partially open. Due to interference from the reach rod operator, CZ-55B could not be fully closed.

To prevent re-occurrence, CZ-55B Discharge Valve was repaired as noted in the immediate actions of CR-ANO-1-2014-02082. The condition above was documented in release 1LR2014-0141 as an unplanned T-16B release. Volume released was 5% of the tank volume to be conservative. (82% - 78.7% = 3.3%)

During 2014, there were zero unplanned releases (liquid or gaseous) to an unrestricted area on ANO-2.

10. RADIATION INSTRUMENTATION As required by ODCM Appendix 1, any radioactive effluent instrumentation inoperable for more than 30 days shall be reported in the ARERR.

On 4/11/14 condition report CR-ANO-1-2014-0652 documents the Liquid Radwaste Process Monitor (RE-4642) remaining out of service for greater than 30 days. ODCM L2.1.1 Condition E.1 of the specification states "Initiate a condition report to document and track the condition for inclusion in the Radioactive Effluent Release Report pursuant to TS 5.6.3 (ANO-1)

/ TS 6.6.3 (ANO-2)." Although the radmonitor was non-functional the release pathway was not active for the full 30 days, but for conservatism it is still being documented in the ARERR.

ODCM L2.1.1 states the applicability is when liquid releases are in progress, RE-4642 is required to be operational. The radmonitor has since been returned to service and remains functional.

Page 92 of 106

ANO-1 & 2 Radioactive Effluent Release Report for 2014

11. CHANGES TO THE PROCESS CONTROL PROGRAM As required by ODCM Section 5.0, a description of changes made to the Process Control Program (EN-RW-105) shall be included in the ARERR for the period in which the change was made effective.

Changes include the following:

Editorial revision to address the issue identified in CR-HQN-2014-00858, CA-02 (Develop a draft procedure that includes instructions for vendors processing waste still owned by Entergy to comply with the PCP program.)

Reworded Step 5.1[1](b) to improve clarity: inserted text processed on-site OR off-site by vendors.

12. CHANGES TO THE OFFSITE DOSE CALCULATION MANUAL In accordance with Unit 1 and Unit 2 TS, changes to the ODCM shall be included in the ARERR for the period in which the change(s) was made effective.

Changes include the following:

The ODCM was revised two times during 2014 to include the following:

Rev 024:

Limitation (L) 2.2.1, Radioactive Gaseous Effluent Monitoring Instrumentation, is modified to provide guidance for appropriate actions when a Super Particulate Iodine and Noble Gas (SPING) monitor is non-functional on a vent path that is in the auto - standby condition (i.e.,

Emergency Penetration Room Ventilation, SPINGs 4 and 8). The associated Bases were revised accordingly.

Rev 025:

Basis (B) 2.5.1, Radiological Environmental Monitoring, background data is updated to include explanation for sample location choice associated with air station #6.

13. LLD LEVELS In accordance with ODCM Appendix 1, lower limits of detection (LLDs) higher than required shall be documented in the ARERR.

During 2014, there were no LLDs higher than required.

14. RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM (REMP)

In accordance with ODCM Appendix 1 Limitations L2.5.1.B and L2.5.2.A, unavailability of milk or fresh, leafy vegetable samples, or an increase in an environmental sample location's calculated dose commitment must be identified in the ARERR.

Page 93 of 106

ANO-1 & 2 Radioactive Effluent Release Report for 2014 A. Changes in Sample Locations During 2014, there were no changes to milk or fresh leafy vegetable sample locations or instances where fresh leafy vegetable samples were unavailable.

B. Increase in Calculated Dose Commitment There were no environmental sampling locations identified during 2014 that would yield a calculated dose commitment greater than the values previously calculated.

15.

SUMMARY

OF HOURLY METEOROLOGICAL DATA In accordance with ODCM Appendix 1 Section 5.2, in lieu of including a summary of the meteorological data in this report, the 2014 data is retained at ANO. This data is available for NRC review.

16. DESCRIPTION OF MAJOR CHANGES TO RADIOACTIVE WASTE SYSTEMS There were no major changes made to the Unit 1 liquid and gaseous radwaste systems or the solid radwaste system during 2014.

There was one major addition to Unit 2 liquid radwaste system to improve processing of liquid radwaste. Engineering Change EC-47007 was implemented for use of a Radwaste Reverse Osmosis (RWRO) to aid in processing liquid radwaste. It further removes radioactive particulates in solution to allow faster processing rates and maintain ALARA. There were no major changes to the Unit 2 gaseous radwaste system.

17. RADIOACTIVE GROUND WATER MONITORING PROGRAM DATA NEI 07-07, Industry Ground Water Protection Initiative - Final Guidance Document, Objective 2.4, Annual Reporting, requires documentation of all on-site ground water sample results and a description of any significant on-site leaks/spills into ground water for each calendar year in the ARERR as contained in the appropriate reporting procedure.

A. NEI 07-07 Objective 2.4, Annual Reporting, Acceptance Criteria b.i requires that ground water sample results that are taken in support of the Ground Water Protection Initiative (GPI) but are not part of the REMP program (e.g. samples obtained during the investigatory phase of the action plan) are reported in the ARERR. Additionally, Entergys procedure EN-CY-111, Radiological Ground Water Monitoring Program, Step 5.15 [3] requires that a listing of non-REMP wells and a summary of pertinent sample results from the RGWMP are reported in the ARERR and an estimate of the doses to a member of the public associated with off-site releases of licensed radioactive material via ground water is included in the ARERR.

In 2014, there were no non-REMP designated ground water wells installed at ANO.

There were no new REMP designated ground water wells installed in 2014. There were four previously installed (prior to 2010) REMP designated ground water wells. The results of the samples collected from the REMP designated ground water wells are included in the 2014 Annual Radiological Environmental Operating Report (AREOR) as required by NEI 07-07. The AREOR for the calendar year 2014 has not been submitted as of the date of this transmittal. The AREOR will be submitted to the NRC in accordance with the ANO-1 and ANO-2 TSs.

ANO did not show any positive results during storm water sampling activities in 2014.

Page 94 of 106

ANO-1 & 2 Radioactive Effluent Release Report for 2014 B. NEI 07-07 Objective 2.4, Acceptance Criteria c.ii requires that a description of all spills or leaks that were communicated per NEI 07-07 Objective 2.2, Voluntary Communication be included in the ARERR. Additionally, Entergys procedure EN-RP-113, Response to Contaminated Spills/Leaks, requires that the following be included in the ARERR:

1. Spills/leaks documented on Attachment 9.1 that were released to the environment or outside the spent fuel pool enclosure, SHALL be documented in the next ARERR.

2. The documentation in the ARERR report will contain:

(a) Description of event (b) Impact of event (c) Remediation of event (d) Radioactive contamination content and levels of event (e) Discussion of impact on groundwater, if any In 2014, there were no spills/leaks that required communication per NEI 07-07 Objective 2.2 or inclusion in the ARERR per EN-RP-113.

A PCRS search was conducted using radioactive spill(s) as the search criteria. Zero items were found.

Entergys procedure EN-CY-108, Monitoring of Non-Radioactive Systems, requires that verified positive results associated with the sampling of designated nonradioactive or cross-contaminated systems are to be included in the sites ARERR, unless already reported under an existing monitored ODCM release point.

In 2014, there was one sample point, Ground Water Monitoring Well-17 (MW-17), where a positive detection of tritium was found. During normal monthly sampling of this well a detectable amount of tritium was confirmed by offsite vendor analysis. The increase in tritium in MW-17 is suspected to be atmospheric recapture; however, this condition is still under investigation. There are no indications of underground piping leaks. See results below.

Date Result Units 02/19/2014 12:00 3.49E+02 pCi/L 03/24/2014 12:00 < 2.93E+02 pCi/L 04/28/2014 12:00 2.39E+003 pCi/L 05/21/2014 12:00 7.66E+002 pCi/L 06/06/2014 12:00 7.53E+002 pCi/L 07/21/2014 12:00 3.89E+02 pCi/L 08/26/2014 12:00 6.58E+002 pCi/L 09/22/2014 11:09 6.28E+02 pCi/L 10/28/2014 11:14 4.38E+02 pCi/L 10/28/2014 12:00 < 3.17E+02 pCi/L 11/19/2014 12:00 4.08E-02 pCi/L 12/15/2014 13:20 3.69E+02 pCi/L Page 95 of 106

ANO-1 & 2 Radioactive Effluent Release Report for 2014

18. C-14 Reporting The purpose of this section is to provide the required input for Carbon-14 (C-14) effluent source term calculations. In Revision 2 of RG 1.21, the NRC has recommended that U.S. nuclear power plants evaluate whether C-14 is a principle gaseous effluent, and if so, report the amount of C-14 released.

Carbon-14 (C-14) is a naturally occurring isotope of carbon. C-14 is produced in commercial nuclear reactors, but the amounts produced are much less than those produced naturally.

Radioactive effluents from commercial nuclear power plants have decreased to the point that C-14 can become a principle radionuclide in gaseous effluents, as defined in RG 1.21.

Therefore, concentrations and offsite dose from C-14 have been estimated and included in this report for ANO.

In 2010, ANO and other facilities participated in an EPRI task force to build a model to accurately estimate C-14 releases, given some key site-specific plant parameters (e.g., mass of the primary coolant, average thermal neutron cross section, rated thermal power). For purposes of industry standardization, the output from the EPRI model is presented in this report in the following spreadsheets.

Page 96 of 106

ANO-1 & 2 Radioactive Effluent Release Report for 2014 Facility Name: Arkansas Nuclear One - Unit 1 Reactor Type WPWR WPWR, CEPWR, or BWR Reactor Power Rating 2568 MegaWatts(thermal)

Equivalent Full Power Operation 365 Days Critical Receptor Undepleted X/Q 5.00E-07 sec/m3 Fractional Equilibrium Ratio, p 1.00 RG-1.109 Equation C-8; assume 1.0 for continuous release Milk Ingestion Pathway Y Enter "Y"' to assume milk ingestion pathway, "N" to ignore pathway Meat Ingestion Pathway Y Enter "Y"' to assume meat ingestion pathway, "N" to ignore pathway Leafy Vegetable Ingestion Pathway Y Enter "Y"' to assume leafy vegetable ingestion pathway, "N" to ignore pathway Enter "Y"' to assume garden produce (non-leafy vegetable+fruit+grain) ingestion pathway, N" to Garden Produce Ingestion Pathway Y ignore pathway Custom C-14 Production Rate 8.92 Ci Produced; Enter 'D' to assume default values Custom Gaseous Release Fraction D  % C-14 production released as a gaseous effluent; Enter 'D' to assume default values Custom Gaseous Release Rate D Ci Released; Enter 'D' to assume default values Custom Carbon Dioxide Fraction D  % C-14 released as Carbon Dioxide; Enter 'D' to assume default values Normalized values listed in this section are taken from EPRI Technical Report 1021106, "Estimation of Carbon-14 in Nuclear Power Plant Gaseous Effluents", 2010, pages 4-28 and 3-26 Parameter WPWR CEPWR BWR Site Values Normalized Production - Ci/GWt-yr 3.4 3.9 5.1 8.92 Ci Produced Gaseous Release Fraction 98% 98% 99% 98%  % Released Normalized Release - Ci/GWt-yr 3.33 3.82 5.05 8.74 Ci Released C-14 Carbon Dioxide Fraction 30% 30% 95% 30% %CO2 Total Dioxide C-14 Release - Ci 8.74E+00 2.62E+00 MAXIMUM DOSE VALUES Organ Age mrem/yr RG-1.109 Bone Child 2.41E-01 RG-1.109 T.Body/Other Child 4.81E-02 ICRP-72 CEDE Teen 7.29E-02 Page 97 of 106

ANO-1 & 2 Radioactive Effluent Release Report for 2014 Facility Name: Arkansas Nuclear One - Unit 2 Reactor Type CEPWR WPWR, CEPWR, or BWR Reactor Power Rating 3026 MegaWatts(thermal)

Equivalent Full Power Operation 365 Days Critical Receptor Undepleted X/Q 5.00E-07 sec/m3 Fractional Equilibrium Ratio, p 1.00 RG-1.109 Equation C-8; assume 1.0 for continuous release Milk Ingestion Pathway Y Enter "Y"' to assume milk ingestion pathway, "N" to ignore pathway Meat Ingestion Pathway Y Enter "Y"' to assume meat ingestion pathway, "N" to ignore pathway Leafy Vegetable Ingestion Pathway Y Enter "Y"' to assume leafy vegetable ingestion pathway, "N" to ignore pathway Enter "Y"' to assume garden produce (non-leafy vegetable+fruit+grain) ingestion pathway, "N" to Garden Produce Ingestion Pathway Y ignore pathway Custom C-14 Production Rate 11.166612 Ci Produced; Enter 'D' to assume default values Custom Gaseous Release Fraction D  % C-14 production released as a gaseous effluent; Enter 'D' to assume default values Custom Gaseous Release Rate D Ci Released; Enter 'D' to assume default values Custom Carbon Dioxide Fraction D  % C-14 released as Carbon Dioxide; Enter 'D' to assume default values Normalized values listed in this section are taken from EPRI Technical Report 1021106, "Estimation of Carbon-14 in Nuclear Power Plant Gaseous Effluents", 2010, pages 4-28 and 3-26 Parameter WPWR CEPWR BWR Site Values Normalized Production - Ci/GWt-yr 3.4 3.9 5.1 11.17 Ci Produced Gaseous Release Fraction 98% 98% 99% 98%  % Released Normalized Release - Ci/GWt-yr 3.33 3.82 5.05 10.94 Ci Released C-14 Carbon Dioxide Fraction 30% 30% 95% 30% %CO2 Total Dioxide C-14 Release - Ci 1.09E+01 3.28E+00 MAXIMUM DOSE VALUES Organ Age mrem/yr RG-1.109 Bone Child 3.02E-01 RG-1.109 T.Body/Other Child 6.03E-02 ICRP-72 CEDE Teen 9.13E-02 Page 98 of 106

Attachment 1 to 0CAN041504 Offsite Dose Calculation Manual

ARKANSAS NUCLEAR ONE OFFSITE DOSE CALCULATION MANUAL REVISION 25 Changes are indicated by beginning the affected information with a revision bar on the right side of the page which stops at the end of the change. Deletions of entire paragraphs or sections have a revision bar to the right of the page where text was deleted. The amendment number is indicated at the bottom of the affected page near the left margin and indicates the latest revision to the information contained on that page. Absence of a revision bar on a replacement page means the page was reprinted for word processing purposes only. However, general formatting changes may have been made to all pages.

ARKANSAS NUCLEAR ONE ODCM TABLE OF CONTENTS Section Title Page

1.0 INTRODUCTION

..............................................................................................................5 2.0 LIQUID EFFLUENTS .......................................................................................................5 2.1 Radioactive Liquid Effluent Monitor Setpoint ........................................................5 2.2 Liquid Dose Calculation ........................................................................................7 2.2.1 Dose Calculations for Aquatic Foods.....................................................7 2.2.2 Dose Calculations for Potable Water .....................................................9 2.3 Liquid Projected Dose Calculation .....................................................................10 3.0 GASEOUS EFFLUENTS................................................................................................10 3.1 Gaseous Monitor Setpoints ................................................................................10 3.1.1 Batch Release Setpoint Calculations...................................................10 3.1.2 Eberline SPING (Final Effluent) Monitor Setpoint Calculations ...........11 3.2 Airborne Release Dose Rate Effects ..................................................................13 3.2.1 Noble Gas Release Rate .....................................................................13 3.2.2 I-131, Tritium and Particulate Release Dose Rate Effects ..................15 3.3 Dose Due to Noble Gases ..................................................................................15 3.3.1 Beta and Gamma Air Doses from Noble Gas Releases ......................15 3.4 Dose Due to I-131, Tritium and Particulates in Gaseous Effluents ....................16 3.4.1 Total Dose from Atmospherically Released Radionuclide ...................17 3.5 Gaseous Effluent Projected Dose Calculation ...................................................24 3.6 Dose to the Public Inside the Site Boundary ......................................................24 3.6.1 Liquid Releases ...................................................................................24 3.6.2 Airborne Release .................................................................................25 4.0 ENVIRONMENTAL SAMPLING STATIONS - RADIOLOGICAL ...................................26 5.0 REPORTING REQUIREMENTS ....................................................................................27 5.1 Annual Radiological Environmental Operating Report .......................................27 5.2 Radioactive Effluent Release Report .................................................................28 Revision 25 2

ARKANSAS NUCLEAR ONE ODCM TABLE OF CONTENTS (continued)

Figure Title Page FIGURE 4-1 Radiological Sample Stations (Far Field) ....................................................30 FIGURE 4-1A Radiological Sample Stations (Near Field) .................................................31 FIGURE 4-1B Radiological Sample Stations (Site Map) ....................................................32 FIGURE 4-2 Maximum Area Boundary for Radioactive Release Calculation (Exclusion Areas) ........................................................................................33 Table Title Page TABLE 4-1 Environmental Sampling Stations - Radiological ........................................34 APPENDIX 1 RADIOLOGICAL EFFLUENT CONTROLS Section Title Page 1.0 DEFINITIONS ................................................................................................................40 2.0 LIMITATION AND SURVEILLANCE APPLICABILITY ...................................................43 2.1 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION ........45 2.1.1 Radioactive Liquid Effluent Monitoring Instrumentation..........................45 2.2 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION...48 2.2.1 Radioactive Gaseous Effluent Monitoring Instrumentation .....................48 2.3 RADIOACTIVE LIQUID EFFLUENTS ..................................................................53 2.3.1 Liquid Radioactive Material Release ......................................................53 2.4 RADIOACTIVE GASEOUS EFFLUENTS ............................................................57 2.4.1 Gaseous Radioactive Material Release..................................................57 2.5 RADIOLOGICAL ENVIRONMENTAL MONITORING ..........................................62 2.5.1 Sample Locations ...................................................................................62 2.5.2 Land Use Census ...................................................................................70 2.5.2 Interlaboratory Comparison Program .....................................................72 Revision 25 3

ARKANSAS NUCLEAR ONE ODCM APPENDIX 1 RADIOLOGICAL EFFLUENT CONTROLS BASES Section Title Page B 2.0 LIMITATION AND SURVEILLANCE APPLICABILITY..................................................73 B 2.1 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION ....76 B 2.1.1 Radioactive Liquid Effluent Monitoring Instrumentation .................76 B 2.2 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION ..............................................................81 B 2.2.1 Radioactive Gaseous Effluent Monitoring Instrumentation ............81 B 2.3 RADIOACTIVE LIQUID EFFLUENTS .............................................................87 B 2.3.1 Liquid Radioactive Material Release .............................................87 B 2.4 RADIOACTIVE GASEOUS EFFLUENTS........................................................93 B 2.4.1 Gaseous Radioactive Material Release .........................................93 B 2.5 RADIOLOGICAL ENVIRONMENTAL MONITORING .....................................98 B 2.5.1 Sample Locations ..........................................................................98 B 2.5.2 Land Use Census ........................................................................101 B 2.5.3 Interlaboratory Comparison Program...........................................103 Revision 25 4

ARKANSAS NUCLEAR ONE ODCM

1.0 INTRODUCTION

The Offsite Dose Calculation Manual (ODCM) provides guidance for making release rate and dose calculations for radioactive liquid and gaseous effluents from Arkansas Nuclear One -

Units 1 and 2 (ANO-1 and ANO-2). The methodology is drawn from NUREG-0133, Rev. 0.

Parameters contained within this manual were taken from NUREG-0133 and Regulatory Guide (RG) 1.109 except as noted for site specific values. These numbers and the calculational method may be changed as provided for in the Technical Specifications (TSs).

The following references are utilized in conjunction with the limitations included in this manual concerning the indicated subjects:

Subject ANO-1 ANO-2 Process Control Program (PCP) EN RW-105 EN RW-105 Radioactive Effluent Controls Program TS 5.5.4 TS 6.5.4 Annual Radiological Environmental Monitoring Report TS 5.6.2 TS 6.6.2 Radioactive Effluent Release Report TS 5.6.3 TS 6.6.3 ODCM TS 5.5.1 TS 6.5.1 2.0 LIQUID EFFLUENTS 2.1 Radioactive Liquid Effluent Monitor Setpoint ODCM Limitation L 2.1.1, Radioactive Liquid Effluent Instrumentation, requires that the radioactive liquid effluents be monitored with the alarm/trip setpoints adjusted to ensure that the limits of the radioactive liquid effluent concentration limitations are not exceeded. These concentrations are for the site. The alarm/trip setpoint on the liquid effluent monitor is dependent upon the dilution water flow rate, radwaste tank flow rate, isotopic composition of the radioactive liquid to be discharged, a gross gamma count of the liquid to be discharged, background count rate of the monitor, and the efficiency of the monitor. Due to the fact that these are variables, an adjustable setpoint is used. The setpoint must be calculated and the monitor setpoint set prior to the release of each batch of radioactive liquid effluents. The following methodology is used for the setpoint determination for the following monitors.

ANO-1: RE-4642 - Liquid Radwaste Monitor ANO-2: 2RE-2330 - Liquid Radwaste Monitor 2RE-4423 - Liquid Radwaste Monitor

1) A sample from each tank (batch) to be discharged is obtained and counted for gross gamma (Cs-137 equivalent) and a gamma isotopic analysis is performed.

2) A dilution factor (DF) for the tank is calculated based upon the results of the gamma isotopic analysis and the Maximum Permissible Concentration (MPC) of each detected radionuclide.

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ARKANSAS NUCLEAR ONE ODCM DF is calculated as follows:

DF = i(Ci/MPCi) + CTNG/MPCTNG where:

DF = dilution factor; Ci = concentration of isotope i, (µCi/ml);

MPCi = maximum permissible concentration of isotope i, (from 10 CFR 20, Appendix B, Table II, Column 2 in µCi/ml);

CTNG = total concentration of noble gases (µCi/ml); and MPCTNG = 2 x 10-4 (µCi/ml) per Limitation L 2.3.1.a

3) The dilution water flowrate is normally the number of ANO-1 circulating water pumps in operation at the time of release. Each circulating water pump has an approximate flowrate of 191,500 gallons per minute (gpm) (this flowrate may be reduced due to throttling of circulating water pump flow and/or circulating water bay configuration).

However, under specific conditions and under strict controls, lower dilution water flowrates utilizing service water and cooling tower blowdown flowrates may be used.

4) The theoretical release rate, Fm, of the tank (batch) to be released is expressed in terms of the dilution water flowrate, such that for each volume of dilution water released, a given volume of liquid radwaste may be combined. This may be expressed as follows:

Fm = DV/DF where:

Fm = theoretical release rate (gpm);

DV = Dilution volume (gpm). When ANO-1 circulating water pumps are running, DV is the number of ANO-1 circulating water pumps in operation multiplied by the approximate flowrate of an ANO-1 circulating water pump (normally 191,500 gpm) or an indicated flow rate. The minimum total flow rate shall be greater than or equal to 100,000 gpm. Otherwise DV is dilution volume provided by service water and cooling tower blowdown flowrate; and DF = dilution factor as calculated in Step 2 above.

Note: In the above equation, the theoretical release rate (Fm) approaches zero as the dilution factor increases. The actual flowrate (FA) will normally be equal to the theoretical release rate for high activity releases. For low activity releases, the theoretical release rate becomes large and may exceed the capacity of the pump discharging the tank. In these cases, the actual release rate may be set to the maximum flowrate of the discharge pump.

5) The monitor setpoint is calculated by incorporating the monitor reading prior to starting the release (i.e., background countrate), and a factor which is the amount of increase in the release concentration that would be needed to exceed the radioactive liquid concentration limitation. The monitor setpoint is expressed as follows:

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ARKANSAS NUCLEAR ONE ODCM ML = A*(K*FM/FA) + B where:

ML = monitor setpoint (counts per minute or cpm);

A = allocation fraction for the specific unit. (Typically, these values are set at 0.45, but may be adjusted up or down as needed. However, the total site allocation can not exceed 1.0.)

K = monitor countrate (cpm) expected based on the gross activity of the release (this value is obtained from a graph of activity (µCi/ml) versus output countrate for the monitor (cpm));

FM/FA = number of times the activity would have to increase to exceed the radioactive liquid effluent-concentration limitation; and B = background countrate (cpm) prior to the release.

To permit the computer to calculate the setpoint, an equation for the expected countrate (K) is expressed as follows:

K = Offset

• SASlope where:

Log of the detector response in cpm Slope =

Log of activity concentration in Ci/ml SA = Gross gamma (Cs-137 equivalent) activity for the tank (Ci/ml); and Offset = detector response (cpm) for the minimum detectable sample activity calculated from the calibration data.

Note: I&C personnel use varying concentrations of Cs-137 to determine the response curve; therefore, a Cs-137 equivalent activity must be used to accurately predict the countrate.

Combining terms, the equation for determining the monitor setpoint may be expressed as follows:

ML = A[(Offset

• SASlope)FM/FA] + B 2.2 Liquid Dose Calculation The dose or dose commitment to an individual in the unrestricted area shall be less than or equal to the limits specified in Radioactive Liquid Effluents - Dose Limitations. The dose limits are on a per reactor basis. This value is calculated using the adult as the maximum exposed individual via the aquatic foods (Sport Freshwater Fish) and the potable water pathways.

2.2.1 Dose Calculations for Aquatic Foods The concentrations of radionuclides in aquatic foods are assumed to be directly related to the concentrations in water. The equilibrium ratios between the two concentrations are called bioaccumulation factors.

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ARKANSAS NUCLEAR ONE ODCM Two different pathways are calculated for aquatic foods: sport and commercial freshwater fish.

The internal dose d from the consumption of aquatic foods in pathway p to organ j of individuals of age group a from all nuclides i is computed as follows (see Chapter 4 of NUREG-0133 and RG 1.109-12, equation A-3):

dp (r,,a,j) = i[{(1100)(e -itp )(Bi)}(M)(Ua)(F) (Qi)(Daij)]

-1 The total dose from both aquatic food pathways is then:

D(r,,a,j) = dp (r,,a,j)

P where:

r = user-selected distance from the release point to the receptor location, in kilometers.

It may be different from the controlling distance specified for the potable water pathway (0.4 km);

= user-selected sector (one of sixteen 22.5° sectors surrounding the reactor site, designated N, NNE, NE, etc.). This sector may be different from the controlling sector specified for the potable water pathway (S);

A = user-selected age group: infant, child, teen, adult. It is the same controlling age group used in the potable water pathway (adult);

J = user-selected organ: bone, liver, total body, thyroid, kidney, lung, GI-LLI. It is the same controlling organ used in the potable water pathway (liver);

{} = represents the concentration factor stored in the database; Note: Only one concentration factor is needed to represent the two pathways since sport and commercial use the same bioaccumulation factor for a given pathway.

1100 = factor to convert from (Ci/yr)/(ft3/sec) to Ci/liter; i = decay constant of nuclide i in hr-1; tp = environmental transit time, release to receptor; Note: This value should be set to 0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> (i.e., no decay correction) for the above equation in order to be consistent with the equation presented in Chapter 4 of NUREG-0133. For maximum individual dose calculations, this value is set to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, which is the minimum transit time recommended by RG 1.109, Appendix A, 2.b.

Bi = bioaccumulation factor for nuclide i, in Ci/kg per Ci/liter. Cesium has a site specific number based on carnivorous and bottom feeder sport fish of 400 Ci/kg per Ci/liter (0CAN048408, dated April 13, 1984); Niobium has a site specific number based upon freshwater fish of 300 Ci/kg per Ci/liter.

M = dimensionless mixing ratio (reciprocal of the dilution factor) at the point of exposure; Revision 25 8

ARKANSAS NUCLEAR ONE ODCM Ua = annual usage factor that specifies the intake rate for an individual of age group a, in kilograms/year. The program selects this usage factor in accordance with the controlling age group a as specified previously by the user; F = average flow rate in ft3/sec. This value is based on total dilution volume for the quarter divided by time into the quarter; Qi = number of curies of nuclide i released; and Daij = ingestion dose factor for age group a, nuclide i, and organ j, in mrem per Ci ingested. The program selects the ingestion dose factor according to the user-specified controlling age group a and controlling organ j.

2.2.2 Dose Calculations for Potable Water The dose D from ingestion of water to organ j of individuals of age group a due to all nuclides i is calculated as follows (See Chapter 4 of NUREG-0133 and NRC RG 1.109-12, equation A-2):

Note: The potable water pathway is used only during the time that the Russellville Water System is using the Arkansas River as a water source. The Russellville Water Works will notify ANO when they are using the Arkansas River as a water source.

D (r,,a,j) = i [{(1100)(e -itp )}(M)(Ua)(F )(Qi)(Daij)]

-1 where:

r = user-selected distance (0.4 km) from the release point to the receptor location, in kilometers. It may be different from the controlling distance selected for the aquatic food pathway;

= user-selected sector; (one of the sixteen 22.5° sectors surrounding the reactor site, designated N, NNE, NE, etc.). It may be different from the controlling sector for the aquatic food pathway; a = user-selected age group (infant, child, teen, adult). The same controlling age group is used for all liquid pathways (adult);

j = user-selected organ (bone, liver, total body, thyroid, kidney, lung, GI-LLI). The same controlling organ is used for all liquid pathways (liver).

{} = the expression in brackets represents the concentration factor stored in the database; 1100 = factor to convert from (Ci/yr)/(ft3/sec) to Ci/liter; M = dimensionless mixing ratio (reciprocal of the dilution factor) at the point of exposure; i = decay constant of nuclide i in hr-1; and tp = environmental transit time, release to receptor.

Note: This value is set to 0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> (i.e., no decay correction) for the above equation to be consistent with the equation presented in Chapter 4 of NUREG-0133.

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ARKANSAS NUCLEAR ONE ODCM Ua = annual usage factor that specifies the intake rate for an individual of age group a, in liters/year. The program selects this usage factor according to the user-specified controlling age group a; F = average flow rate in ft3/sec; this value is based on total dilution volume for one quarter divided by time into the quarter; Qi = number of curies of nuclide i in the release; and Daij = ingestion dose factor, for age group a, nuclide i, and organ j, in mrem per Ci ingested. The program selects the ingestion dose factor according to the user-specified controlling age group a and controlling organ j.

2.3 Liquid Projected Dose Calculation The quarterly projected dose is based upon the methodology of Section 2.2 and is expressed as follows:

DQP = 92(DQC + DRP)/T where:

DQP = quarterly projected dose (mrem);

92 = number of days per quarter; DQC = cumulative dose for the quarter (mrem);

DRP = dose for current release (mrem); and T = current days into quarter; 3.0 GASEOUS EFFLUENTS 3.1 Gaseous Monitor Setpoints Note: Sections 3.1.1 and 3.1.2 below detail two methods of calculating setpoints at ANO.

These methods cover two different sets of monitors of which only one will be in-service at any one time.

3.1.1 Batch Release Setpoint Calculations 3.1.1.a This section applies to the following gaseous radiation monitors (these releases are also monitored by the SPING monitors in Section 3.1.2):

ANO-1: RE-4830 - Waste Gas Holdup System Monitor*

RX-9820 - Reactor Building Purge and Ventilation SPING ANO-2: 2RE-8233 - Containment Building Purge Monitor*

2RE-2429 - Waste Gas Holdup System Monitor*

2RX-9820 - Containment Building Purge and Ventilation SPING

• These monitors provide automatic isolation.

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ARKANSAS NUCLEAR ONE ODCM The setpoints to be used during a batch type of release (i.e., Reactor Building

[Containment] Purge, release from the Waste Gas Holdup System or any other non-routine release) will be calculated for each release before it occurs.

3.1.1.b The basic methodology for determining a monitor setpoint is based upon the expected concentration at the monitor (CM). This is in turn based upon the fraction of an MPC assigned to this release point. Batch releases are maintained below the assigned MPC fraction by controlling the release rate. The calculated value of S may not exceed the equivalent of 1 MPC at site boundary. If value of S for RX (2RX) -9820 is less than SPING Channel 5 high alarm setpoint, then the high alarm setpoint may be used as a default value. If the value of S for RE-4830 and 2RE-2429 is less than 50,000 cpm, then 50,000 cpm may be used as a minimum setpoint. If the value of S for 2RE-8233 is less than 1,000 cpm, then 1,000 cpm may be used as a minimum setpoint.

S = 1.2(CM)(K) + (2.0)(B) where:

S = monitor setpoint (cpm);

CM = Xe-133 equivalent concentration at the monitor (Ci/ml);

K = conversion factor determined from response curve of monitor (cpm per Ci/ml). This value is 1.0 when calculating S for RX (2RX) -9820.

2.0 = factor to accommodate random count rate fluctuations; B = background count rate at the monitor (cpm).

1.2 = Safety Factor to correct for instrument uncertainties.

3.1.2 Eberline SPING (Final Effluent) Monitor Setpoint Calculations 3.1.2.a This section applies to the following gaseous radiation monitors:

ANO-1: RX-9820 - Reactor Building Purge and Ventilation SPING RX-9825 - Auxiliary Building Ventilation SPING RX-9830 - Spent Fuel Pool Area Ventilation SPING RX-9835 - Emergency Penetration Room Ventilation SPING ANO-2: 2RX-9820 - Containment Building Purge and Ventilation SPING 2RX-9825 - Auxiliary Building Ventilation SPING 2RX-9830 - Spent Fuel Pool Area Ventilation SPING 2RX-9835 - Emergency Penetration Room Ventilation SPING 2RX-9845 - Auxiliary Building Extension Ventilation SPING 2RX-9850 - Radwaste Storage Building Ventilation SPING The determination of setpoints for the above monitors is based on an assigned fraction of the MPC of noble gas activity at the site boundary (Xe-133 equivalent) released from the above release points. The total of these fractions is always less than 1.00. The assigned fractions are based on the vent flow rates, atmospheric dilution rate, and the ventilation system(s) in operation.

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ARKANSAS NUCLEAR ONE ODCM Note: The fact that an effluent monitor is in alarm does not necessarily mean that radioactive gases are being released at such a rate that the MPC limit is being exceeded. The alarm would indicate that radioactive gases are being released at a rate that is exceeding the fractional allocation of an MPC allotted to that particular release point. Consideration must be given to the release rate of radioactive gases via all of the release pathways.

The initial fractions of an MPC allocated to the release points are given below. The allocations may be changed as needed, to allow for operational transients, but may not exceed a site total of 1.00.

Monitor Number Monitor Name Fractional Allocation RX-9820 Reactor Building Purge and Ventilation 0.1000 RX-9825 Auxiliary Building Ventilation 0.2000 RX-9830 Spent Fuel Pool Area Ventilation 0.1500 RX-9835 Emergency Penetration Room Ventilation 0.0001 Monitor Number Monitor Name Fractional Allocation 2RX-9820 Containment Building Purge and Ventilation 0.1000 2RX-9825 Auxiliary Building Ventilation 0.2000 2RX-9830 Spent Fuel Pool Area Ventilation 0.1500 2RX-9835 Emergency Penetration Room Ventilation 0.0001 2RX-9840 PASS Building Ventilation 0.0100 2RX-9845 Auxiliary Building Extension Ventilation 0.0100 2RX-9850 Radwaste Storage Building Ventilation 0.0100 Note: The setpoints to be used during a batch release (i.e., Reactor Building

[Containment] Purge or Waste Gas Holdup System) will be calculated for each release before it occurs.

3.1.2.b SPING monitor setpoints may be calculated as follows:

Xe-133 eq (µCi/cc)

Setpoint (Ci/cc) =A F(9.4390E-9)(TMPC) where:

A = allocation fraction (the fraction of an MPC at the site boundary (of noble gas Xe-133 eq activity) assigned to the particular release point);

Xe-133 eq = Xenon-133 equivalent concentration; F = discharge flow of the particular release point in cubic feet per minute (cfm) 2.0E-5(sec/m3) 9.4390E-9 = 2.8317E-2(cm/cf) 60(sec/min)

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ARKANSAS NUCLEAR ONE ODCM where:

2.0E-5 = the annual average gaseous dispersion factor (corrected for radioactive decay) as defined in Section 2.3 of the ANO-1/ANO-2 Safety Analysis Report (SAR); and TMPC = total MPCs at site boundary.

3.2 Airborne Release Dose Rate Effects 3.2.1 Noble Gas Release Rate 3.2.1.a To calculate the noble gas release dose rate, the average ground-level concentration of radionuclide i at the receptor location must first be determined from the following equation (see RG 1.109-20 equation B-4).

xi () = (3.17 x 104)(Qi)[D1X/Q()]

where:

xi () = average ground level concentration in Ci/m3 of nuclide i at the user-specified controlling distance in sector (1.05 km);

() = one of the sixteen 22.5° sectors surrounding the reactor site, designated N, NNE, NE, etc. (WNW);

3.17 x 104 = number of Ci per Ci divided by the number of seconds/year; Qi = release rate of nuclide i in curies/yr and D1X/Q() = annual average gaseous dispersion factor (corrected for radioactive decay) in the sector at angle at the receptor location in sec/m3. This value is 2.0E-5 sec/m3 for short term releases.

The annual dose to the total body and skin due to noble gas can be calculated according to Sections 3.1.2.b and 3.2.1.c.

3.2.1.b Annual Total Body Dose Rate The annual average total body dose rate to the maximally exposed individual is calculated as follows:

DT() = (RBPF)(SF)(i [xi()

• DFBi]

where:

DT() = total body dose rate due to immersion in a semi-infinite cloud of gas at the controlling distance in sector , in mrem/yr. The program computes one total body dose rate value for each sector in which the user has specified a controlling distance and reports only the maximum value;

= one of sixteen 22.5° sectors surrounding the reactor site, designated N, NNE, NE, etc. (WNW);

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ARKANSAS NUCLEAR ONE ODCM RBPF = Reactor Building (Containment) Purge Factor - This factor is used to calculate the length of time (fractional duty cycle) that the purge fans will be in operation. It is calculated by comparing the highest dose rate (DOSER) to its applicable release limit, taking into account the allocation factor for the release point (RBPF = Allocation

• Limit/DOSER). This factor is calculated only for ANO-1 and ANO-2 Reactor Building (Containment) purges. For all other releases, this factor is set to 1.0; SF = dimensionless attenuation factor accounting for the dose reduction due to shielding by residential structures. The NRC recommended value is 0.7 (for maximum individual) xi() = average ground-level concentration of nuclide i at the receptor location in the sector at angle from the release point, as defined in Section 3.2.1.a; and DFBi = total body dose factor for a semi-infinite cloud of radionuclide i, which includes the attenuation of 5 g/cm² of tissue, in mrem-m3/Ci-yr 3.2.1.c Annual Skin Dose Rate The annual dose rate to the skin of the maximally exposed individual due to noble gases is calculated as follows (see RG 1.109-20 equation B-9):

DS() = RBPF[(1.11)(SF)(i(xi())(DFi) + i(xi())(DFSi)]

where:

DS() = skin dose due to immersion in a semi-infinite cloud of gas at the user-specified controlling distance in sector , in mrem; Note: The program computes a skin dose value for each sector in which the user as specified a controlling distance, but prints out only the maximum value.

RBPF = Reactor Building [Containment] Purge Factor as defined in Section 3.2.1.b.

1.11 = average ratio of tissue to air energy absorption coefficient; SF = dimensionless attenuation factor accounting for the dose reduction due to shielding by residential structures. The value is 0.7 (for maximum individual);

xi() = is the average ground-level concentration of nuclide i at the receptor location in the sector at angle from the release point, as defined in Section 3.2.1;

= one of sixteen 22.5° sectors surrounding the reactor site, designated N, NNE, NE, etc. (WNW);

DF = gamma air dose factor for a semi-infinite cloud of radionuclide i, in mrad-m3/Ci-yr; and DFSi = beta skin dose factor for a semi-infinite cloud of radionuclide i, which includes the attenuation by the outer dead layer of skin, in mrem-m3/Ci-yr.

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ARKANSAS NUCLEAR ONE ODCM 3.2.2 I-131, Tritium and Particulate Release Dose Rate Effects The annual dose rate to the maximally exposed individual for I-131, tritium and radionuclides in particulate form with half-lives greater than eight days is calculated as follows:

DRTOT = (RBPF)(DRI + DRG + DRM) where:

RBPF = Reactor Building (Containment) Purge Factor as defined in Section 3.2.1.b; I

DR = dose rate to the controlling age group (infant) associated with the inhalation of radioiodines and particulates, as calculated in Section 3.4.1.b; DRG = dose rate from direct exposure to activity deposited on the ground plane, as calculated in Section 3.4.1.a; and DRM = dose rate to the controlling age group (infant) and the controlling organ for ingestion of food (milk), as calculated in Section 3.4.1.d.

Calculation of the annual dose rate considers the infant as the most restrictive age group. The organs that are considered as contributing to the dose rate are: skin, bone, liver, total body, thyroid, kidney, lung, and GI-LLI. The food pathway for the infant is considered to be from milk only. All three pathways will contribute to the total body dose, while the skin will be affected by only the ground plane pathway. The other organs are affected only by the inhalation and food pathways.

3.3 Dose Due to Noble Gases The air dose in unrestricted areas due to noble gases released in gaseous effluents shall be less than or equal to 5 mrad for gamma radiation and 10 mrad for beta radiation for any calendar quarter for each unit. The objective of less than or equal to 10 mrad of gamma radiation and 20 mrad of beta radiation for a calendar year per unit (2.5 mrad and 5 mrad respectively per quarter) should be used for planning releases.

Note: The following equations have been simplified from equations in NUREG-0133, Revision 0, in that there are no free-standing stacks at ANO. The equations were further simplified in that there are no long term (i.e., continuous) releases. The individual stack vents are sampled weekly, or are assigned a release period of 168 hours0.00194 days <br />0.0467 hours <br />2.777778e-4 weeks <br />6.3924e-5 months <br /> per sample (i.e., considered as short term (batch) releases). Individual samples are to be taken for each waste gas release and Reactor (Containment) Building purge.

3.3.1 Beta and Gamma Air Doses from Noble Gas Releases Using the average ground level concentration of radionuclide i at the receptor location calculated in Section 3.2.1.a, the associated annual gamma or beta air dose may be calculated by the following equation (see RG 1.109-20 equation B-5).

D() or D() = i [(xi())(DFi or DFi)]

where:

D() or D() = the gamma or beta air dose for the controlling distance in sector (only the maximum value is reported), and DFi or DFi = gamma or beta air dose factors for a uniform semi-annual infinite cloud of nuclide i, in mrad-m3/Ci-yr.

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ARKANSAS NUCLEAR ONE ODCM 3.4 Dose Due to I-131, Tritium, and Particulates in Gaseous Effluents The calculational methodology for determining the dose to an individual from I-131, tritium, and radionuclides in particulate form with half-lives greater than 8 days in gaseous effluents released to unrestricted areas as specified in the Limitations is in this section.

The child is the controlling age group unless stated otherwise.

The inhalation and ground plane pathways are considered to exist at all locations. The grass-cow-milk, grass-cow-meat, and vegetation pathways are used where applicable.

It is assumed that iodines are in the elemental form.

A dispersion parameter of 2.0E-5 sec/m3 (per ANO-1/ANO-2 SAR, Section 2.3) is used for w in the inhalation pathway since the majority of gaseous activity released from the site is within the 8 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> time frame (i.e., Reactor Building [Containment] purges and Waste Gas Decay tanks).

The equation is:

DTOT = DG + DI + DV + DL + DM + DF where:

DTOT = total dose; DG = dose contribution from ground plane deposition as calculated in Section 3.4.1.a; I

D = dose contribution from inhalation of radioiodines, tritium, and particulates (> 8 days) as calculated in Section 3.4.1.b; DV = dose contributions from consumption of vegetation (defined as produce) for humans and stored feed for cattle. See Section 3.4.1.c for calculations; DL dose contributions from consumption of fresh leafy vegetables (defined as garden products) for humans and pasture grass for cattle. See Section 3.4.1.c for calculations; DM = dose contribution from consumption of cow's milk; and Note: Consumption by the cow of both stored feeds and pasture grasses is taken into account when calculating this dose contribution. Concentration factors for both food sources are calculated.

DF = dose contribution from consumption of meat.

Note: Consumption by the cow of both stored feeds and pasture grasses is taken into account when calculating this dose contribution. Concentration factors for both types of animal are calculated.

Revision 25 16

ARKANSAS NUCLEAR ONE ODCM 3.4.1 Total Dose from Atmospherically Released Radionuclide After the calculation of the concentration factors from the applicable parts of Section 3.4.1, the maximum individual dose as calculated for controlling age group a and controlling organ j, in sector at the controlling distance r is given from:

DG(r,,j,a) (Section 3.4.1.a) for ground plane deposition DI(r,,j,a) (Section 3.4.1.b) for inhalation V V DV(r,,j,a) = DFIijaUaCi(r,) for produce i

L L DL(r,,j,a) = DFIijaUaCi(r,) for leafy vegetables i

M M DM(r,,j,a) = DFIijaUaCi(r,) for cow's milk i

F F DF(r,,j,a) = DFIijaUaCi(r,) for meat i

where:

a = controlling age group (infant, child, teen, or adult);

j = controlling organ (bone, liver, total body, thyroid, kidney, lung, or GI-LLI);

r = user-selected distance from the release point to the receptor location in a particular sector, in kilometers (the controlling distance is the same for all airborne pathways, 1.05 km);

= one of sixteen 22.5° sectors surrounding the reactor site, designated N, NNE, NE, etc. (WNW);

DFIija = dose conversion factor for ingestion of nuclide i, organ j, and age group a, in mrem/Ci; Note: Values used in these tables are taken from Tables E-11 through E-14 of RG 1.109. DFIija is selected according to the controlling organ and age group as specified in the database.

UVa, UaL, UM a, Ua F

= ingestion rates for produce, leafy vegetables, cow's milk, and meat, respectively, for individuals in age group a. Values used are taken from Table E-5 of RG 1.109.);

CVi, DLi, CiM, DiF = concentration of nuclide i for produce, leafy vegetables, cow's milk, and meat, respectively, in Ci/kg or Ci/liter.

The program calculates that maximum individual dose for each sector surrounding the plant in which the user has specified a controlling distance for each of the following pathways: A) ground plane deposition; B) inhalation and the ingestion of; C) produce; D) leafy vegetables; E) cow's milk; and F) meat. Only the receptor point receiving the maximum dose value is printed.

Revision 25 17

ARKANSAS NUCLEAR ONE ODCM 3.4.1.a Dose from Ground Plane Deposition The dose DG from direct exposure to activity deposited on the ground plane is calculated as follows (see RG 1.109-24, equations C-1 and C-2):

- it b DG(R,,j,a) = {(SF)(1.0 x 1012)(i[(i-1)(1 - e )]}(DOQ(r,))(Qi)(DFGij) where:

r = user-selected distance from the release point to the receptor location in a particular sector, in kilometers. The controlling distance is the same for all airborne pathways (1.05 km);

= one of sixteen 22.5° sectors surrounding the reactor site, designated N, NNE, NE, etc. (WNW);

a = user-selected age group (infant, child, teen, adult) which is the same controlling age group used for all airborne pathways (child);

j = user-selected organ (bone, liver, total body, thyroid, kidney, lung, GI-LLI) which is the same controlling organ used for all airborne pathways;

{} = represents the concentration factor stored in the database; SF = dimensionless attenuation factor accounting for the dose reduction due to shielding by residential structures. The value is 0.7 (for maximum individual);

1.0 x 1012 = number of Ci per Ci; i = decay constant of nuclide i in hr-1; tb = length of time over which the accumulation is evaluated (nominally 15 years which is the approximate midpoint of facility operating life or 1.31 x 105 hours0.00122 days <br />0.0292 hours <br />1.736111e-4 weeks <br />3.99525e-5 months <br />);

DOQ(r,) = average relative deposition of the effluent at the receptor location r in sector

, considering depletion of the plume during transport, in m2 (1.7E-8/m2);

Qi = release of nuclide i in curies, and DFGij = open field ground plane dose conversion factor for organ j (total body or skin) from radionuclide i, in mrem-m2/Ci-hr. The dose factor is selected according to the user-specified controlling age group a and controlling organ j.

3.4.1.b Dose from Inhalation of Radionuclides in Air The dose DI to organ j of age group a associated via inhalation of radioiodines and particulates is (see RG 1.109-25, Equations C-3 and C-4):

DI(r,,j,a) = (3.17 x 104)(Ra)(i[(Qi)(D2DPX/Q(r,))(DFAija)]

where:

r = user-selected distance from the release point to the receptor location in a particular sector, in kilometers. The controlling distance is the same for all airborne pathways (1.05 km);

= one of sixteen 22.5° sectors surrounding the reactor site, designated N, NNE, NE, etc. (WNW);

Revision 25 18

ARKANSAS NUCLEAR ONE ODCM j = user-selected organ (bone, liver, total body, thyroid, kidney, lung, GI-LLI) and is the same controlling organ as that used for all airborne pathways; a = user-selected age group (infant, child, teen, adult) and is the same controlling age group as that used for all airborne pathways; 3.17 x 104 = number of Ci/Ci divided by the number of seconds/year; Ra = annual air intake for individuals in age group a (in m3/year). The air intake factor is selected in accordance with the user-specified controlling age group; Qi = release of nuclide i in curies; D2DPX/Q(r,) = annual average atmospheric dispersion factor of the radionuclide at the receptor location r in sector (in sec/m3) as calculated; and Note: This includes depletion (for radioiodines and particulates) and radioactive decay of the plume.

DFAija = inhalation dose factor for radionuclide i, organ j, and age group a.

The inhalation dose factor is selected in accordance with the user-specified controlling age group a and controlling organ j.

3.4.1.c Dose from Nuclide Concentrations in Vegetation Note: To reduce the computational overhead of the computer, the calculations for dose resulting from nuclide concentrations in forage, produce and leafy vegetables is performed in three steps.

First, the concentration factors (CF) are computed and stored in the database. The concentration factor includes all the parameters that are considered constant for each nuclide and agricultural activity, such as the radioactive decay constant, removal rate constant, exposure time, etc.

Second, the deposition rate from the plume is multiplied by the concentration factor and the nuclide activity to produce the nuclide concentration as follows:

V Ci(r,) = (CFi)(DOQ(r,))(Qi) where:

V Ci(r,) = concentration of nuclide i at the receptor location (r,);

CFi = concentration factor of nuclide i; DOQ(r,) = relative deposition of nuclide i. For the short term dispersion option, DOQ is replaced by (F x DOQ), where F is the short term dispersion correction factor; Qi = quantity of nuclide i released in curies.

For carbon-14 and tritium, the nuclide concentration is calculated from the concentration factor times the decayed and depleted X/Q for radioiodines and particulates (D2DPX/Q), times the quantity of nuclide i released in curies. For the short term dispersion option, D2DPX/Q is replaced by F x D2DPX/Q, where F is the short term dispersion correction factor.

Revision 25 19

ARKANSAS NUCLEAR ONE ODCM V

CT(r,) = (CFT)(D2DPX/Q(r,))(QT) for tritium, and CFV14(r,) = (CF14)(D2DPX/Q(r,))(Q14) for carbon-14 Third, the nuclide concentrations for a particular pathway (produce, leafy vegetables, cow's milk, and meat) are summed and multiplied by: 1) the ingestion rate for a particular age group and

2) the dose conversion factor:

V D(r,,j,a) = i [(DFIija)(Ua)(Ci(r,))]

where:

r = user-selected distance from the release point to the receptor location in a particular sector, in kilometers (1.05 km);

= one of sixteen 22.5° sectors surrounding the reactor site, designated N, NNE, NE, etc. (WNW);

j = user-selected organ (bone, liver, total body, thyroid, kidney, lung, GI-LLI), and is the same controlling organ as that used for all airborne pathways; a = user-selected age group (infant, child, teen, adult), and is the same controlling age group as that used for all airborne pathways; DFIija = dose conversion factor for ingestion of nuclide i, organ j, and age group a, in mrem/Ci, according to the controlling organ and age group; Ua = annual ingestion rate of food in a particular pathway (kilograms/year or liters/year) for individuals in age group a, according to the controlling age group; and V

Ci(r,) = concentration of nuclide i at the receptor location (r,).

3.4.1.c.1 Calculating Vegetation Concentration Factors NUREG-0133 calculations for radioiodines and particulate radionuclides (except tritium and carbon-14), the concentration factor of nuclide i in and on vegetation is estimated as follows:

V r - t CFi = (CONST)( )(e i h)(f)

(Yv)(i) where:

V CFi = concentration factor of radionuclide i in vegetation (forage, produce, or leafy vegetables), in m2-hr/kg; CONST = 1.14 x 108 number of Ci per Ci (1012) divided by the number of hours per year (8760);

r = is the fraction of deposited activity retained on crops, leafy vegetables, or pasture grass, from airborne radioiodine and particulate deposition:

r = 1.00 for radioiodines r = 0.20 for particulates Yv = agricultural productivity (yield or vegetation area density), in kg (wet weight)/m2:

Ys = 2.0 kg/m2 for stored animal feed for grass-animal-man pathways Y = 0.7 kg/m2 for pasture grass for grass-animal-man pathways Revision 25 20

ARKANSAS NUCLEAR ONE ODCM Y1 = 2.0 kg/m2 for leafy vegetation (fresh) for crop/vegetation-man pathways Yg = 2.0 kg/m2 for garden produce (stored vegetables) for crop/vegetation-man pathways i = is the decay constant of nuclide i in hr-1; th = is a holdup time that represents the time interval between harvest and consumption of the food, in hours:

th = 0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> for pasture grass consumed by animals th = 2160 hours0.025 days <br />0.6 hours <br />0.00357 weeks <br />8.2188e-4 months <br /> for stored feed consumed by animals th = 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for leafy vegetables consumed by humans th = 1440 hours0.0167 days <br />0.4 hours <br />0.00238 weeks <br />5.4792e-4 months <br /> for produce consumed by humans f = is the fraction of leafy vegetables or produce grown in garden of interest:

f = 0.76 for the fraction of produce ingested, grown in garden of interest (this is fg in equation C-13 of RG 1.109) f = 1.00 for the fraction of leafy vegetables grown in garden of interest (this is f1 in equation C-13 of RG 1.109) f = 1.00 for all other pathways 3.4.1.c.2 Concentration Factor for Carbon-14 For carbon-14, the concentration factor in and on vegetation is estimated as follows (see RG 1.109-26, equation C-8):

CFV14 = (2.2 x 107)()

where:

CFV14 = concentration factor of carbon-14 in and on vegetation, in m2-hr/kg; and

= is defined as the ratio of total annual release time (for C-14 atmospheric releases) to the total annual time during which photosynthesis occurs (taken to be 4400 hours0.0509 days <br />1.222 hours <br />0.00728 weeks <br />0.00167 months <br />), under the condition that the value of should never exceed unity.

For continuous C-14 releases, is taken to be unity (thus, the value of 2.2 x 107 is stored for CFV14 in lieu of a site specific value for ).

3.4.1.c.3 Concentration Factor for Tritium The concentration factor for tritium in vegetation is calculated from the tritium concentration in air surrounding the vegetation (see RG 1.109-27, equation C-9):

V 1.2 x 107 CF T =

H where:

V CFT = concentration factor for tritium in vegetation (in m2-hr/kg); and H = absolute humidity at the location of the vegetation, in g/m3 (the regulatory default value for H is 8.0 grams/m3).

V Thus, the value 1.5 x 106 is stored for CFT in lieu of a site specific value for H.

Revision 25 21

ARKANSAS NUCLEAR ONE ODCM 3.4.1.c.4 Nuclide Concentrations in Produce and Leafy Vegetables The concentrations in and on produce and leafy vegetables of all radioiodine and particulate nulcides i (except carbon-14 and tritium) are calculated as follows:

CVi(r,) = (CFVi)(DOQ(r,))(Qi) for produce; and CLi(r,) = (CFLi)(DOQ(r,))(Qi) for leafy vegetables where:

CFVi = concentration factor of nuclide i in produce; CFiL = concentration factor of nuclide i in leafy vegetables; L

Note that the difference between CFVi and CFi are the values for th and f1.

DOQ(r,) = relative deposition of the radionuclide i at the receptor (r,); and Qi = release of nuclide i (in curies).

The C-14 and H-3 nuclide concentrations are calculated from the concentration factors times the decayed and depleted radioiodine relative deposition D2DPX/Q times the fraction grown in the garden of interest (fg = 0.76, f1 = 1.0):

CVT(r,) = (CFVT)(D2DPX/Q(r,))(QT)(fg)

CTL(r,) = (CFLT)(D2DPX/Q(r,))(QT)(f1) for tritium V V C14 (r,) = (CF14 )(D2DPX/Q(r,))(Q14)(fg)

C14L (r,) = (CF14L )(D2DPX/Q(r,))(Q14)(f1) for carbon-14 3.4.1.d Nuclide Concentration in Cow's Milk The radionuclide concentration in cow's milk is dependent upon the quantity and contamination level of feed consumed by the animal. The concentration is estimated (see RG 1.109-27, equations C-10 and C-11) as follows:

m - i t f v v1 v1 Ci(r,) = {(Fm)(QF)(e )[(fp)(fs)(CFi) + (1 - fp)(CFi) + (fp)(1 - fs)(CFi)]}(D(r,)(Qi) where:

m Ci(r,) = is the concentration of nuclide i in cow's milk at the receptor location (r,), in Ci/liter;

{} = the expression in brackets represents the concentration factor (note that the concentration factor for cow's milk involves two different vegetation concentration factors (see below));

Fm = average fraction of the cow's daily intake of radionuclide i (which appears in each liter of milk), in days/liter; QF = amount of feed consumed by the cow per day, in kg/day (wet weight);

Revision 25 22

ARKANSAS NUCLEAR ONE ODCM i = decay constant of nuclide i in hr-1; tf = average transport time of the activity from the feed into the milk and to the receptor (a value of 2 days is assumed);

fp = fraction of the year that cows graze on pasture; fs = fraction of daily feed that is pasture grass when the cow grazes on pasture; v

CFi = vegetation concentration factor of nuclide i on pasture grass with the holdup time th = 0 days, in Ci/kg (refer to the explanation of the vegetation concentration factor calculation);

v1 CFi = vegetation concentration factor of nuclide i in stored feeds with the holdup time th = 90 days, in Ci/kg (refer to the explanation of the vegetation concentration factor calculations);

D(r,) = relative deposition DOQ(r,) of the radionuclides, except carbon-14 and tritium.

For carbon-14 and tritium, the decayed and depleted dispersion factor D2DPX/Q(r,) for radioiodines and particulates (in sec/m3) is used; and Qi = is the release of nuclide i in curies.

3.4.1.e Nuclide Concentration in Meat The radionuclide concentration in meat is dependent upon the quantity and contamination level of feed consumed by the animal. The concentration is estimated (see RG 1.109-27, equations C-11 and C-12) as follows:

f -i ts v v1 v1 Ci(r,) = {(Ff)(QF)(e )[(fp)(fs)(CFi) + (1- fp)(CFi) +(fp)(1 - fs)(CFi)]}(D(r,)(Qi) where:

Note: All parameters used in this pathway are for beef cattle.

f Ci(r,) = concentration of nuclide i in animal flesh at the receptor location (r,) in Ci/liter;

{} = the expression in brackets represents the concentration factor (note that the concentration factor for meat involves two different vegetation concentration factors);

Ff = average fraction of the animal's daily intake of radionuclide i which appears in each kilogram of flesh (in days/kg);

Qf = amount of feed consumed by the animal per day in kg/day (wet weight);

i = decay constant of nuclide i in hr-1; ts = average time from slaughter of the animal to consumption by humans (20 days);

fp = fraction of the year that animals graze on pasture; fs = fraction of daily feed that is pasture grass when the animal grazes on pasture; v1 CFi = vegetation concentration factor of nuclide i on pasture grass with the holdup time th = 0 days in Ci/kg (refer to the explanation of the vegetation concentration factor calculation);

v1 CFi = vegetation concentration factor of nuclide i in stored feeds with the holdup time th = 90 days, in Ci/kg (refer to the explanation of the vegetation concentration factor calculation);

Revision 25 23

ARKANSAS NUCLEAR ONE ODCM D(r,) = relative deposition DOQ(r,) of the radionuclides, except carbon-14 and tritium.

For carbon-14 and tritium, the decayed and depleted dispersion factor D2DPX/Q(r,) for radioiodines and particulates (in sec/m3) is used; Qi = is the release of nuclide i (in curies).

3.5 Gaseous Effluent Projected Dose Calculation 3.5.1 The quarterly projected dose is based upon the methodology of Sections 3.3 and 3.4, and is expressed as follows:

DQC + DRP DQP = ( )(92)

T where:

DQP = Quarterly projected dose (mrem);

DQC = cumulative dose for the quarter (mrem);

DRP = dose for current release (mrem);

T = current days into quarter; and 92 = number of days per quarter.

3.6 Dose to the Public Inside the Site Boundary 3.6.1 Liquid Releases Dose to the public inside the site boundary due to liquid releases will be due to ingestion of fish caught from the discharge canal and exposure to sediment along the discharge canal bank while fishing.

3.6.1.a Dose Due to Ingestion of Fish Dose due to ingestion of fish is calculated using the methodology given in Section 2.2, Liquid Dose Calculation.

3.6.1.b Dose Due to Exposure to Shoreline Sediments Dose from external exposure to shoreline sediments is calculated from equation A-7 of RG 1.109, Rev. 1, 10/77.

(Uap)(Mp)(W)

Rapj = 110,000( (i [(Qi)(Ti)(Daipj)(e-i tp)(1-e-i tb)]

F where:

Rapj = is the total annual dose to organ j of individuals of age group a from all of the nuclides i in pathway in mrem/yr; Uap = is the usage factor that specifies exposure time for the maximum individual of age group a in hours from Table E-5 of RG 1.109. Sixty-seven hours for shoreline recreation for a teen was chosen. Adult is the controlling age group for ingestion but the maximum usage factor (teen) was used rather than the adult factor to ensure a conservative dose estimate; Revision 25 24

ARKANSAS NUCLEAR ONE ODCM Mp = is the mixing ratio (reciprocal of dilution factor);

W = is the shoreline width factor from Table A-2 of RG 1.109. The discharge canal value of 0.1 was chosen; F = is the flow rate of the liquid effluent in ft3/sec. This was determined by:

.134 ft3 1 yr 1 hr F(ft3/sec) = waste volume (gal/yr) * *

• 1 gal 8760 hr 3600 sec where:

Qi = is the release of nuclide i in Ci/yr; Ti = is the radioactive half-life of nuclide i, in days, from Radioactive Decay Data Tables, Technical Information Center, U. S. Dept. of Energy, 1981; Daipj = is the dose factor specific to age group a, nuclide i, and organ j from Table E-6 of RG 1.109; i = is the radioactive decay constant of nuclide i in hr-1; tp = is the average transit time for nuclides to reach the point of exposure. A value of 0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> was chosen due to the proximity of the discharge canal to the plant; and tb = is the period of time for which sediment is exposed to the contaminated water in hours. The mid-point of plant operating life, 15 years was chosen per RG 1.109.

3.6.2 Airborne Release 3.6.2.a Dose Due to Noble Gases Dose to fisherman at the discharge canal can be calculated by the ratio of dispersion factor for the discharge canal (1.6E-4 sec/m3 from Table 2-45 SAR, Unit 1, 100 meters downwind in a southerly direction) and the usage factor of 67 hours7.75463e-4 days <br />0.0186 hours <br />1.107804e-4 weeks <br />2.54935e-5 months <br /> of shoreline recreation to the values used in Section 3.3 of this manual.

1.6E-4 67 hr Dose at discharge canal = DT() *

• 2.0E-5 8760 hr where DT() is the noble gas dose calculated by Section 3.3.

3.6.2.b Dose Due to Iodine, Tritium and Particulates from Gaseous Effluents Section 3.4 calculates total dose for iodine, tritium and particulates as the sum of:

DTOT = DG + DI + DV + DL + DM + DF where:

DG = ground plane deposition; DL = consumption of fresh leafy vegetables; DI = inhalation; Dm = consumption of milk; and Dv = consumption of vegetation; DF = consumption of meat and poultry Revision 25 25

ARKANSAS NUCLEAR ONE ODCM The only contributions relevant to fishing activities at the discharge canal are ground plane deposition and inhalation. As DG and DI are not independently available, a conservative estimate can be obtained by using the same correction factor developed for noble gas dose to the total dose calculated in Section 3.4 for iodine, tritium and particulates. Depletion of the plume as it travels downwind can be ignored since the fraction remaining in the plume at 100 meters (discharge canal) and 1046 meters (site boundary) are both greater than 90%

according to Figure 3 of RG 1.111.

The only activity inside the plant site by members of the public that might contribute a significant dose is fishing along the banks of the discharge canal. Travel along public roads would involve short exposure time and tours of the facility are conducted according to radiological control procedures enforced at the plant to control exposure. Fishing is the only uncontrolled activity.

4.0 ENVIRONMENTAL SAMPLING STATIONS - RADIOLOGICAL Section 1.0 of the ODCM provides reference to the Radioactivty Effluent Controls Program governed by ANO-1 TS 5.5.4 and ANO-2 TS 6.5.4. However, a Radiological Environmental Monitoring Program is also necessary to meet the intent of the purpose of the ODCM.

The Radiological Environmental Monitoring Program is established to provide radiation and radionuclide monitoring in the environs surrounding the site. The program provides a method for representative measurements of radioactivity in the highest potential exposure pathways. In addition, the program provides for verification of the accuracy of the effluent monitoring program and modeling of envronmental exposure pathways.

The Radiological Environmental Monitoring Program is established by the ODCM and conforms to the guidance contained in 10 CFR 50, Appendix I. The program also provides for:

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

2. A land use census to ensure that changes in the use of areas at and beyond the site boundary are identified and that modifications to the monitoring program are made, if required by the results of the census, and

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

Environmental samples are collected as specified in the Limitations. The approximate locations of selected sample sites are shown on Figures 4-1, 4-1A, and 4-1B for illustrative purposes.

Table 4-1 lists the approximate distances and directions of the sample stations from the plant.

Revision 25 26

ARKANSAS NUCLEAR ONE ODCM 5.0 REPORTING REQUIREMENTS 5.1 Annual Radiological Environmental Operating Report The Annual Radiological Environmental Operating Report is submitted by May 15 of each year and contains a summary of the Radiological Environmental Monitoring Program for the reporting period. This report meets the requirements of TS 5.6.2 (ANO-1) and TS 6.6.2 (ANO-2), and is consistent with the objectives outlined in the ODCM and 10 CFR 50, Appendix I, Sections IV.B.2, IV.B.3, and IV.C. The report is formatted consistent with RG 1.21, Revision 1, to the extent possible. A single submittal is normally prepared incorporating the data for both ANO units (common information is combined).

The Annual Radiological Environmental Operating Report includes the following:

1. Summarized and tabulated results of all radiological environmental samples and environmental radiation measurements required by the ODCM.

2. A summary description of the Radiological Environmental Monitoring Program.

3. A map of the sampling locations with concurrent table providing distances and directions from the Reactor (Containment) Building. Because the ODCM contains this information and the ODCM is submitted as part of the Radioactive Effluent Release Report, reference to the Radioactive Effluent Release Report submittal date and letter number may be included in the Annual Radiological Environmental Operating Report in lieu of submitting the sample location map and table.

4. A summary of the land use census results in accordance with Surveillance S 2.5.2.2.

5. A summary of the Interlaboratory Comparison Program in accordance with, Surveillance S 2.5.3.1.

As required by the Limitations, the report shall include the following for the conditions listed below:

1. A description of the condition or event and, if applicable, equipment involved.

2. The cause of the condition or event.

3. Actions taken to restore the condition and prevent/minimize recurrence.

4. The consequences of the condition or event.

Revision 25 27

ARKANSAS NUCLEAR ONE ODCM Required Limitation Description Action 2.5.1 A.2 Sample not taken at required location*

Sample equipment out-of-service (OOS)

Sample frequency not met Monitoring/analysis lower limit of detection (LLD) not met Concentration limits not met Dose from other radionuclides exceed concentration limits 2.5.1 B.1 New sample location identified 2.5.2 A.1 New sample location identified 2.5.3 A.1 Interlaboratory Comparison Program requirements not met NA NA Other harmful effects or evidence of irreversible damage detected

• The report shall include a summary of information not available for reporting at the time of submittal. Such missing information shall be submitted in a supplemental letter when data becomes available.

5.2 Radioactive Effluent Release Report The Radioactive Effluent Release Report is submitted prior to May 1 of each year, but not more than 12 months from the previous years submittal, and includes a summary of the quantities of radioactive liquid effluents, gaseous effluents, and solid waste released from the site. This report meets the requirements of TS 5.6.3 (ANO-1), TS 6.6.3 (ANO-2), 10 CFR 50.36a, and 10 CFR 50, Appendix I, Section IV.B.1. The report is formatted consistent with RG 1.21, Revision 1. A single submittal is normally prepared incorporating the data from both ANO units (common information is combined).

In general, the Radioactive Effluent Release Report includes the following:

1. A description of changes to the ODCM and PCP implemented during the reporting period.

TS 5.6.3 (ANO-1) and TS 6.6.3 (ANO-2) contain a description of the ODCM change process.

2. A summary of the hourly meteorological data collected over the previous calendar year. In lieu of including this information in the report, it is permissible to retain this summary available for NRC review, if so noted in the report.

3. A summary of radiation doses due to radiological effluents during the previous calendar year, calculated in accordance with the methodology specified in the ODCM.

4. The radiation dose to members of the public while performing activities inside the site boundary. The calculated dose includes only contributions directly attributed to operation of the units.

5. A description of major changes to the radioactive waste systems (liquid/gaseous/solid) during the previous calendar year, if not included in the cycle SAR update.

Revision 25 28

ARKANSAS NUCLEAR ONE ODCM As required by the Limitations, the report shall include the following for the conditions listed below:

1. A description of the condition or event and, if applicable, equipment involved.

2. The cause of the condition or event.

3. Actions taken to restore the condition and prevent/minimize recurrence.

4. The consequences of the condition or event.

Required Limitation Description Action 2.1.1 D.1 Liquid radioactive monitoring equipment OOS > 30 days 2.2.1 G.1 Gaseous radioactive monitoring equipment OOS > 30 days 2.3.1 A.2 Liquid radioactive release limits exceeded 2.3.1 F.1 Liquid radioactive monitor LLD exceeded 2.4.1 A.2 Gaseous radioactive release limits exceeded 2.4.1 E.1 Gaseous radioactive monitor LLD exceeded Revision 25 29

ARKANSAS NUCLEAR ONE ODCM FIGURE 4-1 RADIOLOGICAL SAMPLE STATIONS 1

2 0° 16 340° 20° US HWY 7 TO HARRISON 15 320° 40° INTERSTATE 40 3

TO FORT SMITH SR 5 PINEY BAY USE AREA Dover 125 SR 333 300° 60° 164 EAST TO 153 U.S.

MORELAND HWY 14 64 4

14 SR 24 TO MORELAND 280° ARKANSAS RIVER 116 INTERSTATE 40 80° 16 LONDON US 13 J I H G F E D C B A HWY 64 5

DELAWARE STATE PARK 127 260° DARDANELLE STATE PARK 111 ARKANSAS TECH UNIVERSITY 100° U.S. HWY 22 HWY 524 LAKE DARDANELLE DARDANELLE STATE PARK RUSSELLVILLE 6

12 DARDANELLE LOCK AND DAM HWY 22 DAM SITE EAST PARK HWY 7T 6 240° 120° HWY 155 HWY 7 SR 247 TO POTTSVILLE MT. NEBO HWY 27 137 STATE PARK DARDANELLE 11 220° HWY 28 140° 7

HWY 7 200° 160° 10 180° 8

HWY 7 TO HOT SPRINGS HWY 27 TO DANVILLE INSET DANVILLE (SEE INSET) 9 N W E HWY 154 S

HWY 27 55 HWY 10 Entergy Substation 7

Petit Jean River 57 HWY 10 Cowger Lake City of Danville Arkansas Nuclear One HWY 80 HWY 27 REMP Sample Locations (Far Field)

Revision 25 30

ARKANSAS NUCLEAR ONE ODCM FIGURE 4-1A RADIOLOGICAL SAMPLE STATIONS SR 333 152 3 108 Training 145 Center 146 109 147 13 1 West Access Rd. 10 56 2 8C 36 Scott Ln.

151 148 8S May Rd. Bunker Cemetery Bunker Hill Ln.

Hill Rd.

149 150 4 110 Arkansas Nuclear One REMP Sample Locations (Near Field)

Lake Dardanelle Revised 24May05 Revision 25 31

ARKANSAS NUCLEAR ONE ODCM FIGURE 4-1B RADIOLOGICAL SAMPLE STATIONS 62 58 STR-3 Switch STR-2 Yard STR-4 STR-6 West Access Road 64 STR-5 63 STR-1 N

Lake Dardanelle W E S

Arkansas Nuclear One REMP Sample Locations Site Map Revision 25 32

ARKANSAS NUCLEAR ONE ODCM FIGURE 4-2 MAXIMUM AREA BOUNDARY FOR RADIOACTIVE RELEASE CALCULATION (Exclusion Areas)

GASES - 1046 METER RADIUS LIQUIDS - END OF DISCHARGE CANAL (POINT A)

EMERGENCY

RESPONSE

FACILITY N EVACUATION ROUTE 2 HWY. 333 COOLING TOWER SWITCHYARD UNIT 2 EVACUATION ROUTE 3 UNIT 1 EVACUATION ROUTE 1 0.65 MILE RADIUS POINT A Revision 25 33

ARKANSAS NUCLEAR ONE ODCM TABLE 4-1 Environmental Sampling Stations - Radiological Approximate Sample Direction and Station Sample Types Sample Station Location Distance from Plant Airborne radioiodines The thermoluminescent dosimeter (TLD) is 1 88° - 0.5 miles Airborne particulates on a pole near the meteorology tower Direct radiation approx. 0.6 miles east of ANO.

Traveling from ANO, go approx. 0.2 miles west toward Gate 4. Turn left (at the east Airborne radioiodines end of the sewage treatment plant) and go 2 243° - 0.5 miles Airborne particulates approx. 0.1 miles. Turn right and go Direct radiation approx. 0.1 miles. The sample station is on the right.

If traveling west on Highway (Hwy) 333, go approx. 0.35 miles from Gate 2 at ANO.

TLD is located on utility pole on south side of Hwy 333 S.

3 5° - 0.7 miles Direct radiation If traveling east on Highway 333, go approx. 0.9 miles from junction of Hwy 333 and Flatwood Road. TLD is located on utility pole on south side of Hwy 333 S.

Go approx. 0.25 miles south from bridge over intake canal. Turn right onto May Road. Proceed approx. 0.1 miles west of 4 181° - 0.5 miles Direct radiation May Cemetery entrance. The TLD is located on a utility pole on the south side of May Road.

Go to the Entergy local office which is Airborne radioiodines located off Hwy 7T in Russellville, 6 111° - 6.8 miles Airborne particulates Arkansas (AR) (305 South Knoxville Direct radiation Avenue). The sample station is against the east wall of the back lot.

Turn west at junction of Hwy 7 and Hwy 27 in Dardanelle, AR. Proceed to junction of Hwy 27 and Hwy 10 in Danville, AR. Turn Airborne radioiodines 210° - right onto Hwy 10 and proceed a short 7 Airborne particulates 19.0 miles distance to the Entergy supply yard, which Direct radiation is on the right adjacent to an Entergy substation. The sample station is in the southwest corner of the supply yard.

166° - 0.2 miles Surface water (composite) 8 243° - 0.9 miles Shoreline sediment Plant discharge canal 212° - 0.5 miles Fish Revision 25 34

ARKANSAS NUCLEAR ONE ODCM TABLE 4-1 Environmental Sampling Stations - Radiological (continued)

Approximate Sample Direction and Station Sample Types Sample Station Location Distance from Plant 95° - 0.5 miles Surface water (grab) is collected at plant 10 Surface water (grab)

(intake canal) intake canal.

Traveling from Hwy 333, turn south onto Flatwood Road. Go approx. 1.0 miles.

13 273° - 0.5 miles Broad leaf vegetation The sample may be collected from either side of Flatwood Road.

From junction of Hwy 7 and Water Works Road, go approx. 0.8 miles west on Water 14 70° - 5.1 miles Drinking water Works Road. The sample station is on the left at the intake to the Russellville city water system from the Illinois Bayou.

Panther Bay, located on the south side of Shoreline sediment 16 287° - 5.5 miles the AR River across from the mouth of Fish Piney Creek.

The sample station is at the Wastewater 153° - Pond water 36 Holding Pond on the ANO site east of the 0.02 miles Pond sediment discharge canal.

From Dardanelle, travel south on Hwy 27.

208° - Go approx. 15.5 miles to the intersection of 55 Broad leaf vegetation 16.5 miles Hwys 27 and 154. The sample station is located at this intersection.

Traveling west from ANO, the sample Airborne radioiodines station is located at the west end of the 56 264° - 0.4 miles Airborne particulates sewage treatment plant near the facility Direct radiation blower building.

Go to Danville and turn left on Fifth Street.

208° - Go approx. three blocks. The Danville 57 Drinking water 19.5 miles public water supply treatment facility is located on the left.

GWM - 1; North of Protected Area on owner controlled area (OCA), west of north 58 22° - 0.3 miles Groundwater Security Check Point, east side of access road.

GWM - 101; North of Protected Area on 62 34° - 0.5 miles Groundwater OCA, east of outside receiving building.

Revision 25 35

ARKANSAS NUCLEAR ONE ODCM TABLE 4-1 Environmental Sampling Stations - Radiological (continued)

Approximate Sample Direction and Station Sample Types Sample Station Location Distance from Plant GWM - 103; South of Protected Area on 63 206° - 0.1 miles Groundwater OCA, northeast of Stator Rewind Building near woodline.

GWM - 13; South of Oily Water Separator, 64 112° - 0.1 miles Groundwater northwest corner of ANO-2 Intake Structure, inside the Protected Area.

If traveling from Hwy 333, turn south onto Flatwood Road and go approx. 0.4 miles.

The TLD is on a utility pole on the right.

108 306° - 0.9 miles Direct radiation If traveling north on Flatwood Road, go approx. 0.4 miles from sample station 109.

The TLD is on a utility pole on the left.

Traveling from Hwy 333, turn south onto Flatwood Road. Go approx. 0.8 miles. The 109 291° - 0.6 miles Direct radiation TLD is on a utility pole on the left across from the junction of Flatwood Road and Round Mountain Road.

From bridge over intake canal, go south approx. 0.25 miles. Turn left and go 110 138° - 0.8 miles Direct radiation approx. 0.25 miles. Turn right on Bunker Hill Lane. The TLD is on the first utility pole on the left.

From junction of Hwy 64 and Hwy 326 (Marina Road), go approx. 2.1 miles on 111 120° - 2.0 miles Direct radiation Marina Road. The TLD is on a utility pole on the left just prior to curve.

Go one block south of the west junction of Hwy 333 and Hwy 64 in London, AR. The 116 318° - 1.8 miles Direct radiation TLD is on a utility pole north of the railroad tracks.

Traveling north on Hwy 7, turn left onto Water Street in Dover, AR. Go one block and turn left onto South Elizabeth Street.

125 46° - 8.7 miles Direct radiation Go one block and turn right onto College Street. The TLD is on a utility pole at the southeast corner of the red brick school building, which is located on top of hill.

Revision 25 36

ARKANSAS NUCLEAR ONE ODCM TABLE 4-1 Environmental Sampling Stations - Radiological (continued)

Approximate Sample Direction and Station Sample Types Sample Station Location Distance from Plant The TLD is located on Arkansas Tech Campus on N. Glenwood Street. If traveling south on Hwy 7 from I- 40, turn right on N. Glenwood. Follow N. Glenwood 127 100° - 5.2 miles Direct radiation for approx. 0.6 miles. The TLD is located on a utility pole (with a No Parking sign on it) across from the northeast corner of Paine Hall.

At junction of Hwy 7 and Hwy 28 in Dardanelle, AR, go approx. 0.2 miles on 137 151° - 8.2 miles Direct radiation Hwy 28. The TLD is on a speed limit sign on the right in front of the Morris R. Moore Arkansas National Guard Armory.

The TLD is located near the west entrance to the Reeves E. Ritchie Training Center 145 28° - 0.6 miles Direct radiation (RERTC) on a utility pole on the north side of Hwy 333.

The TLD is located on the south end of the 146 45° - 0.6 miles Direct radiation east parking lot at the RERTC. The TLD is located on a utility pole.

The TLD is located on the west side of 147 61° - 0.6 miles Direct radiation Bunker Hill Road, approx. 100 yards from the intersection with Hwy 333.

Traveling east from ANO, turn right on Bunker Hill Road. Travel south for approx.

148 122° - 0.6 miles Direct radiation 0.25 miles to the intersection with Scott Lane. The TLD is located on the county road sign post.

Traveling south on Bunker Hill Road, turn right on May Road. Travel approx.

149 156° - 0.5 miles Direct radiation 0.4 miles. The TLD is located on a Notice sign on the north side of May Road.

Traveling south on Bunker Hill Road, turn right on May Road. Travel approx.

150 205° - 0.6 miles Direct radiation 0.8 miles. The TLD is located just past the McCurley Place turn off on the north side of May Road on a utility pole.

Revision 25 37

ARKANSAS NUCLEAR ONE ODCM TABLE 4-1 Environmental Sampling Stations - Radiological (continued)

Approximate Sample Direction and Station Sample Types Sample Station Location Distance from Plant Traveling west from ANO, turn south on plant road along the east side of the 151 225° - 0.4 miles Direct radiation sewage treatment plant. The TLD is located at the end of this road, near the lake on a metal post.

Traveling west on Hwy 333 from the RERTC, travel approx. 0.7 miles. The TLD 152 338° - 0.8 miles Direct radiation is located on the south side of Hwy 333 on a utility pole.

Travel Hwy 64 west to Knoxville Elementary School. The TLD is located 153 304° - 9.2 miles Direct radiation near the school entrance gate on a utility pole.

120° - East side of GSB drainage ditch near lift STR - 1 Storm water runoff 0.33 miles station.

351° - Inside protected area near Sally Port from STR - 2 Storm water runoff

< 0.10 miles drainage ditch along fence.

Outside Protected Area near Sally Port STR - 3 0.2° - 0.13 miles Storm water runoff from drainage ditch along fence.

102° - East side of Oily Water Separator from STR - 4 Storm water runoff 0.10 miles storm drain.

170° - West side of discharge canal from storm STR - 5 Storm water runoff

< 0.10 miles drain.

90° - East side of chemistry chemical storage STR - 6 Storm water runoff

< 0.10 miles area storm drain.

Revision 25 38

ARKANSAS NUCLEAR ONE ODCM APPENDIX 1 RADIOLOGICAL EFFLUENT CONTROLS Revision 25 39

ARKANSAS NUCLEAR ONE ODCM 1.0 DEFINITIONS

---

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

The defined terms of this section appear in capitalized type and are applicable throughout these Limitations and Bases.

Term Definition ACTION(S) ACTIONS shall be that part of a Limitation that prescribes Required Actions to be taken under designated Conditions within specified Completion Times.

BATCH RELEASE A BATCH RELEASE is the discharge of liquid or gaseous wastes of a discrete volume.

CHANNEL CALIBRATION A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel output such that it responds within the necessary range and accuracy to known values of the parameter that the channel monitors. The CHANNEL CALIBRATION shall encompass all devices in the channel required for channel FUNCTIONALITY and the CHANNEL TEST. The CHANNEL CALIBRATION may be performed by means of any series of sequential, overlapping, or total channel steps.

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

CHANNEL TEST A CHANNEL TEST shall be the injection of a simulated or actual signal into the channel as close to the sensor as practicable to verify FUNCTIONALITY of all devices in the channel required for channel FUNCTIONALITY. The CHANNEL TEST may be performed by means of any series of sequential, overlapping, or total steps.

CONTINUOUS RELEASE A CONTINUOUS RELEASE is the discharge of liquid waste of a non-discrete volume, e.g. from a volume of a system that has an input flow during the continuous release.

EXCLUSION AREA The EXCLUSION AREA is that area surrounding ANO within a minimum radius of 0.65 miles of the Reactor (Containment) Buildings and controlled to the extent necessary by the licensee for purposes of protection of individuals from exposure to radiation and radioactive materials.

Revision 25 40

ARKANSAS NUCLEAR ONE ODCM 1.0 DEFINITIONS (continued)

Term Definition FUNCTIONAL-FUNCTIONALITY A system, subsystem, train, component, or device shall be FUNCTIONAL or have FUNCTIONALITY when it is capable of performing its specified function(s), as set forth in the current license basis (CLB) and when all necessary attendant instrumentation, controls, normal or emergency electrical power, cooling and seal water, lubrication, and other auxiliary equipment that are required for the system, subsystem, train, component, or device to perform its specified function(s) are also capable of performing their related support function(s).

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

LIQUID RADWASTE A LIQUID RADWASTE TREATMENT SYSTEM is a TREATMENT SYSTEM system designed and used for holdup, filtration, and/or demineralization of radioactive liquid effluents prior to their release to the environment.

MEMBER(S) OF THE PUBLIC MEMBER(S) OF THE PUBLIC shall include all persons who are not occupationally associated with the plant. This category does not include employees of the utility, its contractors or vendors. Also excluded from the category are persons who enter the site to service equipment or to make deliveries. This category does include persons who use portions of the site for recreational, occupational or other purposes not associated with the plant.

MODE(S) Refer to Definitions section of ANO-1 and ANO-2 TSs.

PURGE - PURGING PURGE or PURGING is the controlled process of discharging air or gas from a confinement to reduce the airborne radioactivity concentration in such a manner that replacement air or gas is required to purify the confinement.

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

Revision 25 41

ARKANSAS NUCLEAR ONE ODCM 1.0 DEFINITIONS (continued)

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

UNRESTRICTED AREA An UNRESTRICTED AREA shall be any area beyond the EXCLUSION AREA boundary.

Revision 25 42

ARKANSAS NUCLEAR ONE ODCM 2.0 LIMITITATION (L) APPLICABILITY L 2.0.1 Limitations shall be met during the specified conditions in the Applicability, except as provided in L 2.0.2.

L 2.0.2 Upon discovery of a failure to meet a Limitation, the applicable ACTIONS of the associated Limitation shall be met, except as provided in L 3.0.5. If the Limitation is met or is no longer applicable prior to expiration of the specified Completion Time(s),

completion of the ACTIONS is not required, unless otherwise stated.

L 2.0.3 When a Limitation is not met and the associated ACTIONS are not met, an associated ACTION is not provided, or if directed by the associated ACTIONS, immediately initiate a condition report to document the condition and determine any limitations for continued operation of the plant.

Exceptions to this Limitation are stated in the individual Limitations.

L 2.0.4 When a Limitation is not met, entry into a MODE or other specified condition in the Applicability shall only be made when the associated ACTIONS to be entered permit continued operation in the MODE or other specified condition in the Applicability for an unlimited period of time.

L 2.0.5 Equipment removed from service or declared non-functional to comply with ACTIONS may be returned to service under administrative control solely to perform testing required to demonstrate its FUNCTIONALITY or the FUNCTIONALITY of other equipment. This is an exception to L 2.0.2 for the system returned to service under administrative control to perform the testing required to demonstrate FUNCTIONALITY.

Revision 25 43

ARKANSAS NUCLEAR ONE ODCM 2.0 SURVEILLANCE (S) APPLICABILITY S 2.0.1 Surveillances shall be met during the specified conditions in the Applicability for individual Limitations, unless otherwise stated in the Surveillance. Failure to meet a Surveillance, whether such failure is experienced during the performance of the Surveillance or between performances of the Surveillance, shall be failure to meet the Limitation. Failure to perform a Surveillance within the specified Frequency shall be failure to meet the Limitation except as provided in S 2.0.3.

Surveillances are not required to be performed on non-functional equipment or variables outside specified limits.

S 2.0.2 The specified Frequency for each Surveillance is met if the Surveillance is performed within 1.25 times the interval specified in the Frequency, as measured from the previous performance or as measured from the time a specified condition of the Frequency is met. For Frequencies specified as "once," the above interval extension does not apply. If an Action completion time requires periodic performance on a "once per . . ." basis, the above Frequency extension applies to each performance after the initial performance.

S 2.0.3 If it is discovered that a Surveillance was not performed within its specified Frequency, then compliance with the requirement to declare the Limitation not met may be delayed, from the time of discovery, up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or up to the limit of the specified Frequency, whichever is greater. This delay period is permitted to allow performance of the Surveillance.

If the Surveillance is not performed within the delay period, the Limitation must immediately be declared not met, and the applicable ACTIONS must be entered.

When the Surveillance is performed within the delay period and the Surveillance is not met, the Limitation must immediately be declared not met, and the applicable ACTIONS must be entered.

S 2.0.4 Entry into a specified condition in the Applicability of a Limitation shall only be made when the Limitation's Surveillances have been met within their specified Frequency, except as provided by S 2.0.3. When a Limitation is not met due to Surveillances not having been met, entry into a specified condition in the Applicability shall only be made in accordance with L 2.0.4.

Revision 25 44

ARKANSAS NUCLEAR ONE ODCM L 2.1 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION L 2.1.1 The following Radioactive Liquid Effluent Monitoring Instrumentation shall be FUNCTIONAL:

a. Liquid Radwaste Effluent Radiation Monitor with alarm/trip function

b. Liquid Radwaste Effluent Flow Monitor

c. One Main Steam Line Radiation Monitor per Steam Generator (ANO-1 only)

APPLICABILITY: Liquid Radwaste Effluent Monitor - during releases via the associated pathway Main Steam Line Radiation Monitors - MODES 1, 2, 3, and 4 ACTIONS

---

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

Separate Condition entry is allowed for each instrument.

CONDITION REQUIRED ACTION COMPLETION TIME A. Required Liquid Radwaste A.1 Suspend the release of Immediately Effluent Radiation Monitor radioactive effluents non-functional. monitored by the affected channel.

AND A.2.1 Restore the monitor to a Prior to release of FUNCTIONAL status. radioactive effluents monitored by the OR affected channel A.2.2.1 Analyze two independent Prior to release of samples of the associated radioactive effluents tank contents. monitored by the affected channel AND A.2.2.2 Computer input data Prior to release of verified by two qualified radioactive effluents individuals. monitored by the affected channel AND Revision 25 45

ARKANSAS NUCLEAR ONE ODCM L 2.1.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.2.2.3 Correct discharge valve Prior to release of lineup independently radioactive effluents verified by two qualified monitored by the individuals. affected channel B. Required Liquid Radwaste B.1 Estimate flow. 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> Effluent Flow Monitor non-functional. OR B.2 Suspend the release of Immediately radioactive effluents monitored by the affected channel.

C. One or more required Main C.1 Establish pre-planned 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Steam Line Radiation alternate monitoring method Monitor non-functional. of monitoring.

AND C.2 Restore the affected Main 7 days Steam Line Radiation Monitor(s) to a FUNCTIONAL status.

D. Required Action(s) and/or D.1 Initiate a condition report to Immediately Completion Time(s) of document the condition and Conditions A, B, and/or C determine any limitations for not met. the continued effluent release operations.

E. Required Radioactive E.1 Initiate a condition report to Immediately Liquid Effluent Monitoring document and track the Instrument non-functional condition for inclusion in the for > 30 days. Radioactive Effluent Release Report pursuant to TS 5.6.3 (ANO-1) / TS 6.6.3 (ANO-2).

Revision 25 46

ARKANSAS NUCLEAR ONE ODCM L 2.1.1 SURVEILLANCES SURVEILLANCE FREQUENCY S 2.1.1.1 Perform a CHANNEL CHECK of required instrumentation. 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> S 2.1.1.2 -----------------------------------NOTE------------------------------------

Not applicable to Liquid Radwaste Effluent Flow Monitor.

Perform a CHANNEL TEST of the required instrumentation. 92 days S 2.1.1.3 Perform a CHANNEL CALIBRATION on the required 18 months instrumentation.

S 2.1.1.4 -----------------------------------NOTES-----------------------------------

1. SOURCE CHECK not required when background radioactivity is greater than the check source.

2. Not applicable to Liquid Radwaste Effluent Flow Monitor or Main Steam Line Radiation Monitors.

Perform a SOURCE CHECK on the required Within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> prior instrumentation. to release of radioactive effluents monitored by the channel Revision 25 47

ARKANSAS NUCLEAR ONE ODCM L 2.2 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION L 2.2.1 The following Radioactive Gaseous Effluent Monitoring Instrumentation shall be FUNCTIONAL:

---

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

Refer to ANO-2 Technical Specification (TS) 3.3.3.1 for ANO-2 Containment Building Purge System Process Monitor operability requirements and associated ACTIONS.

a. Waste Gas Holdup Systems

1. Gas Activity Process Monitor with alarm/trip function

2. Effluent Flow Process Monitor

b. Reactor (Containment) Building Purge and Ventilation, Auxiliary Building Ventilation, Spent Fuel Pool Area Ventilation, Emergency Penetration Room Ventilation, Low Level Radwaste Building Ventilation, and ANO-2 Auxiliary Building Extension Ventilation SPING Monitors

1. Noble Gas Activity Monitor

2. Iodine Sampler

3. Particulate Sampler

4. Effluent Flow Monitor

5. Sampler Flow Monitor APPLICABILITY:

1. SPINGS 4 and 8 - when Emergency Penetration Room Ventilation is capable of auto-start

2. All Radioactive Gaseous Effluent Monitoring Instrumentation - during releases via the associated pathway Revision 25 48

ARKANSAS NUCLEAR ONE ODCM ACTIONS

---

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

Separate Condition entry is allowed for each instrument.

CONDITION REQUIRED ACTION COMPLETION TIME A. -------------NOTE-------------- A.1 Suspend the release of Immediately Applicable to releases radioactive effluents associated with Waste Gas monitored by the affected Holdup Systems and channel.

PURGE of the ANO-1 Reactor Building. AND A.2.1 Restore the monitor to a Prior to release of Required Waste Gas FUNCTIONAL status. radioactive effluents Holdup and/or Reactor monitored by the Building Purge System OR affected channel Gas Activity Process and/or Noble Gas Activity A.2.2.1 Analyze two independent Prior to release of Monitor non-functional. samples of the Waste Gas radioactive effluents Holdup Tank and/or monitored by the Reactor Building contents. affected channel AND A.2.2.2 Computer input data Prior to release of verified by two qualified radioactive effluents individuals. monitored by the affected channel AND A.2.2.3 -------------NOTE-------------

Not applicable to Reactor Building Purge System.

Correct discharge valve Prior to release of lineup independently radioactive effluents verified by two qualified monitored by the individuals. affected channel Revision 25 49

ARKANSAS NUCLEAR ONE ODCM L 2.2.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B. Required Effluent or B.1 Estimate flow. Once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> Sampler Flow Monitor non-functional. OR B.2 Suspend the release of Immediately radioactive effluents monitored by the affected channel.

C. --------------NOTE-------------- --------------------NOTE------------------

1. Applicable to releases If ANO-1 Reactor Building Purge other than those and Ventilation required Noble Gas described in Condition A Activity Monitor inoperable and above. moving irradiated fuel within the ANO-1 Reactor Building, refer to

2. Applicable to SPINGS 4 ANO-1 TS 3.9.3.

and 8 only when -----------------------------------------------

pathway is in service.

---

C.1 Obtain sample of effluent. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Required Noble Gas Activity AND Monitor non-functional.

C.2 Analyze sample of effluent. Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following completion of Required Action C.1 D. --------------NOTE-------------- D.1 Verify effluent samples are 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> Applicable to SPINGS 4 continuously collected by and 8 only when pathway auxiliary sampling equipment.

is in service.

---

AND Required Iodine and/or D.2 Replace Iodine and/or 7 days Particulate Sampler non- Particulate cartridges (as functional. applicable).

AND D.3 Analyze Iodine and/or Within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> Particulate cartridges (as following replacement applicable).

Revision 25 50

ARKANSAS NUCLEAR ONE ODCM L 2.2.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME E. Required Action(s) and/or E.1 Suspend the release of Immediately Completion Time(s) of radioactive effluents monitored Condition C and/or by the affected channel.

Condition D not met.

F. Required Action(s) and/or F.1 Initiate a condition report to Immediately Completion Time(s) document the condition and Condition A, B, and/or E determine any limitations for not met. the continued effluent release operations.

G. Required Radioactive G.1 Initiate a condition report to Immediately Gaseous Effluent document and track the Monitoring Instrument condition for inclusion in the non-functional for Radioactive Effluent Release

> 30 days. Report pursuant to TS 5.6.3 (ANO-1) / TS 6.6.3 (ANO-2).

SURVEILLANCES SURVEILLANCE FREQUENCY S 2.2.1.1 -----------------------------------NOTE------------------------------------

Not applicable to Iodine and Particulate Samplers Perform a CHANNEL CHECK of required instrumentation. 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> S 2.2.1.2 Verify presence of required Iodine Sampler Cartridge and 7 days required Particulate Sample Filter.

S 2.2.1.3 Perform a CHANNEL TEST of the required Reactor Building 31 days prior to Purge and Ventilation System Gas Activity Process and initiating Reactor Noble Gas Activity Monitors. Building Purge and/or Ventilation activities Revision 25 51

ARKANSAS NUCLEAR ONE ODCM L 2.2.1 SURVEILLANCES (continued)

SURVEILLANCE FREQUENCY S 2.2.1.4 -----------------------------------NOTES-----------------------------------

SOURCE CHECK not required when background radioactivity is greater than the check source.

Perform a SOURCE CHECK on the required Noble Gas 31 days Activity Monitors.

S 2.2.1.5 -----------------------------------NOTES-----------------------------------

1. SOURCE CHECK not required when background radioactivity is greater than the check source.

2. Only applicable to Waste Gas Holdup and Reactor Building Purge Systems.

Perform a SOURCE CHECK on the required Gas Activity Within 14 days prior Process and Noble Gas Activity Monitors. to release of radioactive effluents monitored by the channel S 2.2.1.6 Perform a CHANNEL TEST of the required Noble Gas 92 days Activity Monitors.

S 2.2.1.7 -----------------------------------NOTE------------------------------------

Not applicable to Iodine and Particulate Samplers Perform a CHANNEL CALIBRATION on the required 18 months instrumentation.

Revision 25 52

ARKANSAS NUCLEAR ONE ODCM L 2.3 RADIOACTIVE LIQUID EFFLUENTS L 2.3.1 Radioactive material released to the discharge canal shall:

a. For dissolved or entrained noble gases, be limited to a total concentration of 2 x 10-4 µCi/ml.

b. For radioactive nuclides other than dissolved or entrained noble gases, be limited to the concentration specified in 10 CFR 20, Appendix B, Table II, Column 2.

c. During any calendar quarter, result in a dose commitment to a MEMBER OF THE PUBLIC of 1.5 mrem to the total body and 5 mrem to any organ.

d. During any calendar year, result in a dose commitment to a MEMBER OF THE PUBLIC of 3 mrem to the total body and 10 mrem to any organ.

e. Be processed by a LIQUID RADWASTE TREATMENT SYSTEM when accumulative dose during a calendar quarter is projected to exceed 0.18 mrem to the total body and/or 0.625 mrem to any organ.

APPLICABILITY: At all times.

ACTIONS

---

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

Separate Condition entry is allowed for each Limitation L 2.3.1.a through L 2.3.1.e above and for each BATCH RELEASE and CONTINUOUS RELEASE Surveillance requirement not met.

CONDITION REQUIRED ACTION COMPLETION TIME A. Any limit listed in L 2.3.1.a A.1 Initiate action to restore to Immediately through L 2.3.1.e not met. within limit.

AND A.2 Initiate a condition report to Immediately document the condition, determine any limitations for the continued effluent release operations, and track the condition for inclusion in the Radioactive Effluent Release Report pursuant to TS 5.6.3 (ANO-1) / TS 6.6.3 (ANO-2).

Revision 25 53

ARKANSAS NUCLEAR ONE ODCM L 2.3.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B. --------------NOTE-------------- B.1 Verify associated effluent Immediately Only applicable to BATCH release suspended.

RELEASE.

---

AND Sampling and/or analysis B.2 Initiate a condition report to Immediately requirements not met. document the condition and determine any limitations for the continued effluent release operations.

C. --------------NOTE-------------- C.1 Obtain a grab sample of the 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Only applicable to associated secondary CONTINUOUS RELEASE coolant.

of secondary coolant.

---

AND Secondary coolant dose C.2 Perform gamma isotopic and 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> following equivalent I-131 (DEI) I-131 analysis of sample. sample acquisition

> 0.01 µCi/ml.

D. Annual dose limits of D.1 Apply for a variance from the Prior to exceed L 2.3.1.d projected to NRC to permit releases in 40 CFR 190 limits exceed 40 CFR 190 limits. excess of 40 CFR 190 limits. Immediately E. Required Action(s) and/or E.1 Initiate a condition report to Immediately Completion Time(s) of document the condition and Conditions C and/or D not determine any limitations for met. the continued effluent release operations.

OR Sampling and/or analysis requirements not met.

Revision 25 54

ARKANSAS NUCLEAR ONE ODCM L 2.3.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME F. Lower Limit(s) of Detection F.1 Initiate a condition report to Immediately (LLD) not met. document and track the condition for inclusion in the Radioactive Effluent Release Report pursuant to TS 5.6.3 (ANO-1) / TS 6.6.3 (ANO-2).

SURVEILLANCES SURVEILLANCE FREQUENCY S 2.3.1.1 -----------------------------------NOTE------------------------------------

Only applicable to BATCH RELEASE.

Obtain representative sample of each batch. Prior to release AND Perform gamma isotopic and I-131 analysis of sample. Prior to release AND Perform dissolved and entrained gas analysis of sample. 31 days following sample acquisition AND Perform gross alpha composite and H-3 analysis of sample. 31 days following sample acquisition AND Perform Sr-89, Sr-90, and Fe-55 composite analysis of 92 days following sample. sample acquisition Revision 25 55

ARKANSAS NUCLEAR ONE ODCM L 2.3.1 SURVEILLANCES (continued)

SURVEILLANCE FREQUENCY S 2.3.1.2 -----------------------------------NOTE------------------------------------

Only applicable to CONTINUOUS RELEASE.

Obtain representative sample of effluent. 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> AND Perform gamma isotopic and I-131 analysis. 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following sample acquisition AND Perform dissolved and entrained gas analysis. 31 days following sample acquisition AND Perform gross alpha composite and H-3 analysis. 31 days following sample acquisition AND Perform Sr-89, Sr-90, and Fe-55 composite analysis. 92 days following sample acquisition S 2.3.1.3 Using data acquired by performance of S 2.3.1.1 and Within 7 days S.2.3.1.2, verify Limitations L 2.3.1.a through L 2.3.1.e following completion continue to be met. of each required analysis S 2.3.1.4 Using data acquired by performance of S 2.3.1.1 and 31 days S.2.3.1.2, verify the limits of 40 CFR 190 are not projected to be exceeded.

S 2.3.1.5 Verify the following LLDs are met: 12 months Gamma isotopic 5 x 10-7 µCi/ml I-131 and Fe-55 1 x 10-6 µCi/ml Dissolved/entrained gases (gamma emitters) 1 x 10-5 µCi/ml H-3 1 x 10-5 µCi/ml Gross alpha 1 x 10-7 µCi/ml Sr-89 and Sr-90 5 x 10-8 µCi/ml Revision 25 56

ARKANSAS NUCLEAR ONE ODCM L 2.4 RADIOACTIVE GASEOUS EFFLUENTS L 2.4.1 Radioactive Gaseous Effluent releases to unrestricted areas shall:

---

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

Dose rates associated with Reactor (Containment) Building Purge operations may be averaged over a one hour interval.

a. For noble gases, be limited to:

1. A total body dose rate of 500 mrem/yr.

2. A skin dose rate of 3000 mrem/yr.

3. A dose commitment to a MEMBER OF THE PUBLIC in any calendar quarter of 5 mrads gamma and 10 mrads beta radiation.

4. A dose commitment to a MEMBER OF THE PUBLIC in any calendar year of 10 mrads gamma and 20 mrads beta radiation.

b. For I-131, H-3, and for all radionuclides in particulate form having a half life of

> 8 days, be limited to:

1. An organ dose rate of 1500 mrem/yr.

2. A dose commitment to a MEMBER OF THE PUBLIC in any calendar quarter of 7.5 mrem to any organ.

3. A dose commitment to a MEMBER OF THE PUBLIC in any calendar year of 15 mrem to any organ.

c. Be processed by a VENTILATION EXHAUST TREATMENT SYSTEM when:

1. For noble gases, the dose over a calendar quarter is project to exceed 0.625 mrads gamma and/or 1.25 mrads beta radiation.

2. For I-131, H-3, and for all radionuclides in particulate form having a half life of > 8 days, the dose over a calendar quarter is project to exceed 1.0 mrem to any organ.

d. Be processed by the GASEOUS RADWASTE TREATMENT SYSTEM when degasifying the Reactor Coolant System (RCS), if projected dose would exceed 0.625 mrads gamma and/or 1.25 mrads beta radiation over a calendar quarter.

APPLICABILITY: At all times.

Revision 25 57

ARKANSAS NUCLEAR ONE ODCM L 2.4.1 ACTIONS

---

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

Separate Condition entry is allowed for each Limitation L 2.4.1.a through L 2.4.1.d above and for each Surveillance requirement not met.

CONDITION REQUIRED ACTION COMPLETION TIME A. Any limit listed in L 2.4.1.a A.1 Initiate action to restore to Immediately through L 2.4.1.d not met. within limit.

AND A.2 Initiate a condition report to Immediately document the condition, determine any limitations for the continued effluent release operations, and track the condition for inclusion in the Radioactive Effluent Release Report pursuant to TS 5.6.3 (ANO-1) / TS 6.6.3 (ANO-2).

B. Sampling and/or analysis B.1 Verify associated effluent Immediately requirements of S 2.4.1.1 release suspended.

not met.

AND B.2 Initiate a condition report to Immediately document the condition and determine any limitations for the continued effluent release operations.

C. Annual dose limits of C.1 Apply for a variance from the Prior to exceed L 2.4.1.a.4 and/or NRC to permit releases in 40 CFR 190 limits L 2.4.1.b.4 projected to excess of 40 CFR 190 limits. Immediately exceed 40 CFR 190 limits.

Revision 25 58

ARKANSAS NUCLEAR ONE ODCM L 2.4.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME D. Required Action(s) and/or D.1 Initiate a condition report to Immediately Completion Time(s) of document the condition and Condition C not met. determine any limitations for the continued effluent release OR operations.

Sampling and/or analysis requirements of S 2.4.1.2 not met.

E. Lower Limit(s) of Detection E.1 Initiate a condition report to Immediately (LLD) not met. document and track the condition for inclusion in the Radioactive Effluent Release Report pursuant to TS 5.6.3 (ANO-1) / TS 6.6.3 (ANO-2).

SURVEILLANCES SURVEILLANCE FREQUENCY S 2.4.1.1 -----------------------------------NOTE------------------------------------

Only applicable to Waste Gas Storage Tank and Reactor Building Purge release.

Obtain representative sample of gas to be released. Prior to release AND Analyze sample for principal gamma emitters. Prior to release AND

---

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

Only applicable to Reactor Building Purge release.

Perform H-3 analysis of sample. Prior to release Revision 25 59

ARKANSAS NUCLEAR ONE ODCM L 2.4.1 SURVEILLANCES (continued)

SURVEILLANCE FREQUENCY S 2.4.1.2 -----------------------------------NOTE------------------------------------

Only applicable to Auxiliary Building, Spent Fuel Pool Area, Auxiliary Building Extension Area (ANO-2), Low Level Radwaste Building, Emergency Penetration Room, and Reactor (Containment) Building Ventilation systems.

The following effluent samples shall be obtained to support the radioactive analysis specified:

a. ------------------------------------NOTE-------------------------------

Only applicable to Reactor Building Ventilation when Reactor Vessel Head is removed.

Representative sample for H-3 analysis. 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />

b. ------------------------------------NOTE-------------------------------

Only applicable to Spent Fuel Pool Area Ventilation.

Representative sample for H-3 analysis. 7 days

c. Charcoal sample for I-131 analysis. 7 days

d. Particulate sample for principal gamma emmiters 7 days analysis.

e. Particulate sample for composite gross alpha analysis. 31 days

f. Representative sample for principal gamma emmiters 31 days analysis.

g. Representative sample for H-3 analysis. 31 days

h. Particulate sample of for Sr-89 and Sr-90 composite 92 days analysis.

AND (continued) (continued)

Revision 25 60

ARKANSAS NUCLEAR ONE ODCM S 2.4.1.2 (continued)

Complete analysis of above samples:

i. Samples a, b, c, and d 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> following sample acquisition

j. Samples e, f, and g 31 days following sample acquisition

k. Sample h 60 days following sample acquisition S 2.4.1.3 Record SPING Noble Gas activity. 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> S 2.4.1.4 Using data acquired by performance of S 2.4.1.1 and 31 days S.2.4.1.2, verify Limitations L 2.4.1.a through L 2.4.1.d continue to be met.

S 2.4.1.5 Using data acquired by performance of S 2.4.1.1 and 31 days S.2.4.1.2, verify the limits of 40 CFR 190 are not projected to be exceeded.

S 2.4.1.6 Verify the following LLDs are met: 12 months Principal gamma emitters (gaseous) 1 x 10-4 µCi/ml Principal gamma emitters (particulate) 1 x 10-11 µCi/ml I-131 1 x 10-12 µCi/ml H-3 1 x 10-6 µCi/ml Gross alpha 1 x 10-11 µCi/ml Sr-89 and Sr-90 1 x 10-11 µCi/ml Noble gas (dose equivalent Xe-133) 1 x 10-6 µCi/ml Revision 25 61

ARKANSAS NUCLEAR ONE ODCM L 2.5 RADIOLOGICAL ENVIRONMENTAL MONITORING L 2.5.1 The following environmental sample locations shall be designated and maintained:

---

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

Other instruments may be used in place of, or in addition to, integrating dosimeters.

Pathway / Sample Type # Location Samples close to site boundary in or near different 3 sectors having the highest calculated annual average ground-level D/Q Airborne Radionuclide and Sample from the vicinity of a community having the Particulate 1 highest calculated annual average ground-level D/Q Background information sample from a control location 1

10-20 miles from one reactor building Inner ring stations with 2 or more dosimeters in each 16 meteorological sector in the general area of the site boundary Direct Radiation Stations with 2 or more dosimeters in special interest 8 areas such as population centers, nearby residences, schools, and in 1-2 areas to serve as control locations.

Surface 1 Indicator location influenced by plant discharge Water 1 Control location uninfluenced by plant discharge Drinking 1 Indicator location influenced by plant discharge Water 1 Control location uninfluenced by plant discharge Waterborne Shoreline 1 Indicator location influenced by plant discharge Sediment 1 Control location uninfluenced by plant discharge Ground 1 Indicator location influenced by plant discharge Water 1 Control location uninfluenced by plant discharge Indicator location within 5 miles of one reactor, if 1

commercially available Milk Control location > 5 miles from one reactor when an 1

indicator exists Sample of commercially and/or recreationally important 1

species in vicinity of plant discharge Fish Ingestion Sample of same species in area not influenced by plant 1

discharge Sample of broadleaf (edible or inedible) near the site 1 boundary from one of the highest anticipated annual Food average ground-level D/Q sectors Products Sample location of broadleaf vegetation (edible or 1 inedible) from a control location 10-20 miles from one reactor Revision 25 62

ARKANSAS NUCLEAR ONE ODCM L 2.5.1 APPLICABILITY: At all times.

ACTIONS

---

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

Separate Condition entry is allowed for each sample location and Surveillance requirement.

CONDITION REQUIRED ACTION COMPLETION TIME A. Sample location A.1 Initiate action to restore to Immediately requirement not met. within limits.

OR AND Required sample A.2 Initiate a condition report to Immediately equipment non-functional. document and track the condition for inclusion in the OR Annual Radiological Environmental Operating Sample Frequency not met. Report pursuant to TS 5.6.2 (ANO-1) / TS 6.6.2 (ANO-2).

OR Sample analysis Frequency not met.

OR One or more Lower Limit(s) of Detection (LLD) listed in Table 2.5-1 not met.

OR One or more limits listed in Table 2.5-2 not met.

OR Dose to a MEMBER OF THE PUBLIC from radionuclides other than those listed in Table 2.5-2 projected to exceed calendar year limits of L 2.3.1 and/or L 2.4.1.

Revision 25 63

ARKANSAS NUCLEAR ONE ODCM L 2.5.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B. Sample(s) from required B.1 Identify and add to the 30 days sample location(s) Radiological Environment unavailable. Monitoring Program, locations for obtaining replacement samples.

SURVEILLANCES SURVEILLANCE FREQUENCY S 2.5.1.1 -----------------------------------NOTE------------------------------------

Only applicable to Airborne Radionuclide and Particulate.

Collect sample from continuous sampler. 14 days AND Perform I-131 analysis of radioiodine canister. 14 days following sample acquisition AND Perform gross beta analysis of particulate sampler. 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and 14 days following filter change S 2.5.1.2 -----------------------------------NOTE------------------------------------

Only applicable to Direct Radiation locations.

Collect sample from required location. 92 days AND Perform gamma dose analysis of sample. 60 days following sample acquisition Revision 25 64

ARKANSAS NUCLEAR ONE ODCM L 2.5.1 SURVEILLANCES (continued)

SURVEILLANCE FREQUENCY S 2.5.1.3 -----------------------------------NOTE------------------------------------

Only applicable to Surface Water samples.

Collect sample from required location. 92 days AND Perform gamma isotopic analysis of sample. 21 days following sample acquisition AND Perform H-3 analysis of sample. 31 days following sample acquisition S 2.5.1.4 -----------------------------------NOTE------------------------------------

Only applicable to Drinking and Ground Water samples.

Collect sample from required location. 92 days AND Perform gamma isotopic analysis of sample. 21 days following sample acquisition AND Perform H-3 analysis of sample. 31 days following sample acquisition AND Perform I-131 analysis of sample. 21 days following sample acquisition AND Perform gross beta analysis of sample. 31 days following sample acquisition Revision 25 65

ARKANSAS NUCLEAR ONE ODCM L 2.5.1 SURVEILLANCES (continued)

SURVEILLANCE FREQUENCY S 2.5.1.5 -----------------------------------NOTE------------------------------------

Only applicable to Waterborne Shoreline Sediment samples.

Collect sample from required location. 12 months AND Perform gamma isotopic analysis of sample. 60 days following sample acquisition S 2.5.1.6 -----------------------------------NOTE------------------------------------

Only applicable to Milk samples.

Collect sample from required location. 92 days AND Perform gamma isotopic analysis of sample. 21 days following sample acquisition AND Perform I-131 analysis of sample. 21 days following sample acquisition S 2.5.1.7 -----------------------------------NOTE------------------------------------

Only applicable to edible portions of Fish samples.

Collect sample from required location. 12 months AND Perform gamma isotopic analysis of sample. 60 days following sample acquisition Revision 25 66

ARKANSAS NUCLEAR ONE ODCM L 2.5.1 SURVEILLANCES (continued)

SURVEILLANCE FREQUENCY S 2.5.1.8 ----------------------------------NOTES-----------------------------------

1. Only applicable to Food Product samples.

2. Only applicable if Milk sampling not performed.

Collect sample from required location. 12 months AND Perform gamma isotopic analysis of sample. 21 days following sample acquisition AND Perform I-131 analysis of sample. 21 days following sample acquisition S 2.5.1.9 Verify the LLDs listed in Table 2.5-1 are met. 12 months S 2.5.1.10 Verify radioactivity concentrations are less than or equal to 92 days the limits listed in Table 2.5-2, when averaged over a calendar quarter.

Revision 25 67

ARKANSAS NUCLEAR ONE ODCM L 2.5.1 TABLE 2.5-1 MAXIMUM VALUES OF THE LOWER LIMITS OF DETECTION (LLD)

Water Airborne Particulate or Gas Fish Food Products Sediment Analyses Milk (pCi/l)

(pCi/l) (pCi/m3) (pCi/kg, wet) (pCi/kg, wet) (pCi/kg, dry)

Gross Beta 4(a) 1 x 10-2(b)

H-3 2000(c)

Mn-54 15 130 Fe-59 30 260 Co-58, 60 15 130 Zn-65 30 260 Zr-95 30 Nb-95 15 I-131 1(d) 7 x 10-2(e) 1 60 Cs-134 15 5 x 10-2(f) 130 15 60 150

-2(f)

Cs-137 18 6 x 10 150 18 80 180 Ba-140 60 60 La-140 15 15 (a)

LLD for drinking water.

(b)

Only applicable to particulate.

(c)

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

(d)

LLD for drinking water. When no drinking water pathway exists, a gamma isotopic analysis LLD value of 15 pCi/l may be used.

(e)

Only applicable to gas.

(f)

Only applicable to particulate gamma isotopic analysis.

Revision 25 68

ARKANSAS NUCLEAR ONE ODCM L 2.5.1 TABLE 2.5-2 REPORTING LEVELS FOR RADIOACTIVITY CONCENTRATIONS IN ENVIRONMENTAL SAMPLES Airborne Particulate or Gas Analyses Water (pCi/l) Fish (pCi/kg, wet) Milk (pCi/l) Food Products (pCi/kg, wet)

(pCi/m3)

H-3 2 x 104(a)

Mn-54 1 x 103 3 x 104 Fe-59 4 x 102 1 x 104 Co-58 1 x 103 3 x 104 Co-60 3 x 102 1 x 104 Zn-65 3 x 102 2 x 104 Zr-95, Nb-95 4 x 102(b)

I-131 2(c) 0.9 3 1 x 102 Cs-134 30 10(d) 1 x 103 60 1 x 103 Cs-137 50 20(d) 2 x 103 70 2 x 103 Ba-140, La-140 2 x 102(b) 3 x 102(b)

(a)

Drinking water samples.

(b)

Total for parent and daughter.

(c)

LLD for drinking water. When no drinking water pathway exists, a value of 20 pCi/l may be used.

(d)

Applicable when performing a gamma isotopic analysis of individual particulate samples with gross beta activity more than 10 times the yearly mean of control samples.

Revision 25 69

ARKANSAS NUCLEAR ONE ODCM L 2.5 RADIOLOGICAL ENVIRONMENTAL MONITORING L 2.5.2 -----------------------------------------------------NOTE---------------------------------------------------

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

The location of the nearest milk animal, the nearest residence, and the nearest garden of greater than 500 ft2 producing fresh leafy vegetables in each of the 16 meteorological sectors within a 5 mile distance from one reactor (containment) building shall be identified.

APPLICABILITY: At all times.

ACTIONS

---

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

Separate Condition entry is allowed for each sample location.

CONDITION REQUIRED ACTION COMPLETION TIME A. New sample location A.1 Initiate a condition report to Immediately identified which yields a document and track the calculated dose due to condition for inclusion in the I-131, H-3, and/or Annual Radiological particulates projected to Environmental Operating exceed 40 CFR 190 limits. Report pursuant to TS 5.6.2 (ANO-1) / TS 6.6.2 (ANO-2).

OR AND New sample location identified which yields a A.2.1 Identify and add the new 30 days calculated dose via the sample location to the same exposure pathway in Radiological Environment excess of values calculated Monitoring Program.

at sample locations of Limitation L 2.51. AND A.2.2 Delete the previous Within 90 days sample location via the following October 31 associated exposure of the year in which pathway from the the new sample Radiological Environment location was Monitoring Program. identified.

Revision 25 70

ARKANSAS NUCLEAR ONE ODCM L 2.5.2 SURVEILLANCES

---

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

S 2.0.2 is not applicable to the Surveillances of this Limitation.

SURVEILLANCE FREQUENCY S 2.5.2.1 A land use census to identify the locations described in 24 months between Limitation L 2.5.2 shall be performed by door-to-door survey, June 1 and aerial survey, or by consulting local agricultural authorities. October 1 S 2.5.2.2 Include the results of S 2.5.2.1 in the Annual Radiological 12 months Environmental Operating Report pursuant to TS 5.6.2 (ANO-1) / TS 6.6.2 (ANO-2).

Revision 25 71

ARKANSAS NUCLEAR ONE ODCM L 2.5 RADIOLOGICAL ENVIRONMENTAL MONITORING L 2.5.3 Radioactive materials supplied as part of the Interlaboratory Comparison Program shall be analyzed.

APPLICABILITY: At all times.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Limitation not met. A.1 Initiate a condition report to Immediately document and track the condition for inclusion in the Annual Radiological Environmental Operating Report pursuant to TS 5.6.2 (ANO-1) / TS 6.6.2 (ANO-2).

SURVEILLANCES

---

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

S 2.0.2 is not applicable to the Surveillances of this Limitation.

SURVEILLANCE FREQUENCY S 2.5.3.1 Include the results of analyses performed as part of the 12 months Interlaboratory Comparison Program in the next Annual Radiological Environmental Operating Report pursuant to TS 5.6.2 (ANO-1) / TS 6.6.2 (ANO-2).

Revision 25 72

ARKANSAS NUCLEAR ONE ODCM B 2.0 LIMITITATION (L) APPLICABILITY BASES Limitations L 2.0.1 through L 2.0.5 establish the general requirements applicable to all Limitations and apply at all times, unless otherwise stated.

B 2.0.1 L 2.0.1 establishes the Applicability statement within each individual Limitation as the requirement for when the Limitation is required to be met (i.e., when the unit is in the MODES or other specified conditions of the Applicability statement of each Limitation).

B 2.0.2 L 2.0.2 establishes that upon discovery of a failure to meet a Limitation, the associated ACTIONS shall be met. The Completion Time of each Required Action for an ACTIONS Condition is applicable from the point in time that an ACTIONS Condition is entered. The Required Actions establish those remedial measures that must be taken within specified Completion Times when the requirements of a Limitation are not met. This Limitation establishes that:

a. Completion of the Required Actions within the specified Completion Times constitutes compliance with a Limitation; and

b. Completion of the Required Actions is not required when a Limitation is met within the specified Completion Time, unless otherwise specified.

Completing the Required Actions is not required when a Limitation is no longer applicable, unless otherwise stated in the individual Specification.

B 2.0.3 L 2.0.3 establishes the Required Actions that must be implemented when a Limitation is not met and the condition is not specifically addressed by the associated Conditions. It is not intended to be used as an operational convenience that permits routine voluntary removal of redundant systems or components from service in lieu of other alternatives that would not result in redundant systems or components being inoperable. This requirement is intended to provide assurance that plant management is aware of the condition and to ensure that the condition is evaluated for its affect on continued operation of the plant.

B 2.0.4 L 2.0.4 establishes Limitations on changes in MODES or other specified conditions in the Applicability when a Limitation is not met. It allows placing the unit in a MODE or other specified condition stated in that Applicability (e.g., the Applicability desired to be entered) when unit conditions are such that the requirements of the Limitation would not be met, in accordance with Limitation L 2.0.4.a, L 2.0.4.b, or L 2.0.4.c.

Revision 25 73

ARKANSAS NUCLEAR ONE ODCM BASES LIMITATION APPLICABILITY (continued)

B 2.0.4 L 2.0.4 allows entry into a MODE or other specified condition in the (continued) Applicability with the Limitation not met when the associated ACTIONS to be entered permit continued operation in the MODE or other specified condition in the Applicability for an unlimited period of time. Compliance with Required Actions that permit continued operation of the unit for an unlimited period of time in a MODE or other specified condition provides an acceptable level of safety for continued operation. This is without regard to the status of the unit before or after the MODE change. Therefore, in such cases, entry into a MODE or other specified condition in the Applicability may be made in accordance with the provisions of the Required Actions. The provisions of this Limitation should not be interpreted as endorsing the failure to exercise the good practice of restoring systems or components to FUNCTIONAL status before entering an associated MODE or other specified condition in the Applicability.

Upon entry into a MODE or other specified condition in the Applicability with the Limitation not met, L 2.0.1 and L 2.0.2 require entry into the applicable Conditions and Required Actions until the Condition is resolved, until the Limitation is met, or until the unit is not within the Applicability of the Limitation.

Surveillances do not have to be performed on the associated inoperable equipment (or on variables outside the specified limits), as permitted by S 2.0.1.

Therefore, utilizing L 2.0.4 is not a violation of S 2.0.1 or S 2.0.4 for any Surveillances that have not been performed on equipment. However, Surveillances must be met to ensure FUNCTIONALITY prior to declaring the associated equipment FUNCTIONAL (or variable within limits) and restoring compliance with the affected Limitation.

B 2.0.5 L 2.0.5 establishes the allowance for restoring equipment to service under administrative controls when it has been removed from service or declared non-functional to comply with ACTIONS. The sole purpose of this Limitation is to provide an exception to L 2.0.2 (e.g., to not comply with the applicable Required Actions) to allow the performance of required testing to demonstrate:

a. The FUNCTIONALITY of the equipment being returned to service; or

b. The FUNCTIONALITY of other equipment.

The administrative controls ensure the time the equipment is returned to service in conflict with the requirements of the ACTIONS is limited to the time absolutely necessary to perform the required testing to demonstrate FUNCTIONALITY.

This Limitation does not provide time to perform any other preventive or corrective maintenance.

An example of demonstrating the FUNCTIONALITY of the equipment being returned to service is restarting a ventilation system that has been secured to comply with Required Actions and must be restarted to perform the required testing.

Revision 25 74

ARKANSAS NUCLEAR ONE ODCM B 2.0 SURVEILLANCE (S) APPLICABILITY BASES S 2.0.1 SRs shall be met during the MODES or other specified conditions in the Applicability for individual Limitations, unless otherwise stated in the individual Surveillance. Failure to meet a Surveillance, whether such failure is experienced during the performance of the Surveillance or between performances of the Surveillance, shall be failure to meet the Limitation. Failure to perform a Surveillance within the specified Frequency shall be failure to meet the Limitation except as provided in S 2.0.3. Surveillances are not required to be performed on non-functional equipment or variables outside specified limits.

S 2.0.2 The specified Frequency for each Surveillance is met if the Surveillance is performed within 1.25 times the interval specified in the Frequency, as measured from the previous performance or as measured from the time a specified condition of the Frequency is met.

For Frequencies specified as "once," the above interval extension does not apply.

If a Completion Time requires periodic performance on a "once per . . ." basis, the above Frequency extension applies to each performance after the initial performance.

Exceptions to this Limitation are stated in the individual Limitations.

S 2.0.3 If it is discovered that a Surveillance was not performed within its specified Frequency, then compliance with the requirement to declare the Limitation not met may be delayed, from the time of discovery, up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or up to the limit of the specified Frequency, whichever is greater. This delay period is permitted to allow performance of the Surveillance.

If the Surveillance is not performed within the delay period, the Limitation must immediately be declared not met, and the applicable Condition(s) must be entered.

When the Surveillance is performed within the delay period and the Surveillance is not met, the Limitation must immediately be declared not met, and the applicable Condition(s) must be entered.

S 2.0.4 Entry into a MODE or other specified condition in the Applicability of a Limitation shall only be made when the Limitation's Surveillances have been met within their specified Frequency, except as provided by S 2.0.3. When a Limitation is not met due to Surveillances not having been met, entry into a MODE or other specified condition in the Applicability shall only be made in accordance with L 2.0.4.

Revision 25 75

ARKANSAS NUCLEAR ONE ODCM B 2.1 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION BASES BACKGROUND 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.

LIMITATION The following Radioactive Liquid Effluent Monitoring Instrumentation is required to be FUNCTIONAL:

ANO-1: RE-4642 - Liquid Radwaste Monitor RE-2682 - A Main Steam Line Radiation Monitor RE-2681 - B Main Steam Line Radiation Monitor ANO-2: 2RE-2330 - Liquid Radwaste Monitor 2RE-4423 - Liquid Radwaste Monitor Both radiation monitoring and flow monitoring capability are required to be FUNCTIONAL for each Liquid Radwaste Monitor. With regard to Liquid Radwaste radiation monitoring, the alarm/trip function must also be FUNCTIONAL. The alarm/trip setpoints for these instruments are calculated in accordance with the methods contained in ODCM Section 2.1 to ensure that the alarm/trip will occur prior to potentially exceeding the limits of 10 CFR Part 20.

With regard to the Main Steam Line Radiation Monitors, these monitors must have a measurement range capability from 10-1 mR/hr to 104 mR/hr.

APPLICABILITY The Liquid Radwaste Monitors are required to be FUNCTIONAL during any release via the pathway in which the monitor is installed. The Main Steam Line Radiation Monitors are required to be FUNCTIONAL in MODES 1, 2, 3, and 4.

ACTIONS The following ACTIONS are generally applicable to the pathway in which a radioactive liquid release is in progress. Because more than one release could occur simultaneously, the ACTIONS are modified by a Note that permits separate Condition entry for each non-functional Radioactive Liquid Effluent Monitoring Instrument.

Revision 25 76

ARKANSAS NUCLEAR ONE ODCM B 2.1 ACTIONS (continued)

A.1 If the radiation monitoring feature of the Radioactive Liquid Effluent Monitoring Instrument is non-functional, any release via the associated pathway must be suspended immediately. This prevents the release of unmonitored effluents to the environment.

A.2.1 In addition to Required Action A.1, a non-functional radiation monitoring feature of a Radioactive Liquid Effluent Monitoring Instrument must be returned to a FUNCTIONAL status prior to the restart or subsequent release of effluents via the associated pathway. This prevents the release of unmonitored effluents to the environment. Exceptions to this requirement are included in Required Actions A.2.2.1 through A.2.2.3 below.

A.2.2.1 through A.2.2.3 In lieu of performing Required Action A.2.1 above, grab samples may be obtained and analyzed to provide a backup monitoring method for the effluent release. Because of the importance of monitoring radioactive liquid releases, two independent samples of the effluent must be obtained and analyzed. The independency required is with regard to obtaining and analyzing each sample separately. Two independent personnel are not required to obtain and analyze the two samples.

Notwithstanding the above, computer input data and the discharge valve lineup associated with the effluent release path must be verified by two independent, qualified individuals.

Integrity of independence is maintained by preventing interaction between personnel during the verification process. With regard to valve lineups, independent verification is conducted such that each check constitutes actual identification of the valve and a determination of both required and actual valve position.

B.1 and B.2 If the flow monitoring feature of the Radioactive Liquid Effluent Monitoring Instrument is non-functional, the flow rate may be estimated within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> of initial loss of the instrument and every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> thereafter, for the duration of the effluent release. Flow rate data is necessary to calculate the amount of radioactive released via the effluent discharge. The 4-hour Completion Time is reasonable because a significant change in flow rate over the course of an effluent release is unlikely.

S 2.0.2 is not applicable to the initial flow estimation, but may be applied to the flow estimations thereafter. Pump curves may be used to estimate flow.

Revision 25 77

ARKANSAS NUCLEAR ONE ODCM B 2.1 ACTIONS (continued)

C.1 If one or more Main Steam Line Radiation Monitors is non-functional, the pre-planned alternate monitoring method of monitoring must be established within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. The alternate method chosen should ensure continued monitoring of the Main Steam system for radiation while operating in MODES 1, 2, 3, or 4. In addition, the affected monitor(s) must be restored to a FUNCTIONAL status within 7 days.

D.1 If the Required Actions and associated Completion Times of Conditions A, B, and/or C cannot be met, then additional measures may be necessary to ensure continued safe operation or to reduce overall station risk. Therefore, a condition report must be initiated immediately to assess the impact on continued effluent release operations given the degraded condition.

E.1 Instrumentation installed to ensure radiological monitoring of effluent releases is expected to be normally available in accordance with the design function or purpose of the equipment.

During releases via a respective pathway, instrumentation that remains non-functional for greater than 30 days may indicate inappropriate importance placed on the equipment or over-reliance on the backup sampling method for effluent release monitoring. As an incentive to avoid either of these conditions, Radioactive Liquid Effluent Monitoring Instrumentation that remains non-functional for more than 30 days must be included in the Radioactive Effluent Release Report submitted pursuant to TS 5.6.3 (ANO-1) or TS 6.6.3 (ANO-2). In order to ensure inclusion, Required Action E.1 requires the condition to be tracked via a condition report.

Information to be provided in the respective Radioactive Effluent Release Report should include 1) the component number and noun name, 2) the failure mode, 3) the reason for continued inoperability, and 4) the expected return to service date.

SURVEILLANCES S 2.1.1.1 Performance of the CHANNEL CHECK every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> provides reasonable assurance for prompt identification of a gross failure of instrumentation. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels. Where parameter comparison is not possible, the CHANNEL CHECK will continue to identify gross instrument failure such as loss of power, unexpected upscale readings, failed-low indications, etc. The CHANNEL CHECK is key in verifying that the instrumentation continues to operate properly between CHANNEL CALIBRATIONs. The Frequency is based on unit operating experience that demonstrates channel failure is rare.

Revision 25 78

ARKANSAS NUCLEAR ONE ODCM B 2.1 SURVEILLANCES (continued)

S 2.1.1.2 A CHANNEL TEST is performed on the radiation monitoring portion of each required instrument channel to ensure the entire channel will perform the intended functions. The CHANNEL TEST demonstrates that automatic isolation of the associated pathway and Control Room alarm occur should the instrument indicate measured levels above the trip setpoint.

The channel test also demonstrates that alarm occurs when any of the following conditions exist:

A. Power to the detector is lost.

B. The instrument indicates a downscale failure.

C. Instrument controls are not set in the operate mode.

Any setpoint adjustment shall be consistent with Section 2.1 of the ODCM.

The Surveillance is modified by a Note clarifying that the CHANNEL TEST is applicable only to the radiation detection portion of the monitor function and is not applicable to the flow monitoring function. The Frequency of 92 days is based on unit operating experience, with regard to channel FUNCTIONALITY and drift, which demonstrates that failure of a channel in any 92-day interval is a rare event, especially in light of the infrequency of radioactive liquid releases.

S 2.1.1.3 CHANNEL CALIBRATION is a complete check of the instrument channel, including the sensor. The test verifies that the channel responds to a measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drift (as required) to ensure that the instrument channel remains FUNCTIONAL between successive tests. CHANNEL CALIBRATION shall find that measurement errors and setpoint errors are within the assumptions of the setpoint calculations. CHANNEL CALIBRATIONS must be performed consistent with the assumptions of the setpoint calculations. This Frequency is justified by the assumption of at least an 18 month calibration interval to determine the magnitude of equipment drift or deviation in the setpoint calculations.

Initial CHANNEL CALIBRATION is performed using one or more of the reference standards certified by the National Institute of Standards and Technology (NIST) or using standards that have been obtained from suppliers that participate in measurement assurance activities with NIST. These standards 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 are used.

Revision 25 79

ARKANSAS NUCLEAR ONE ODCM B 2.1 SURVEILLANCES (continued)

S 2.1.1.4 A SOURCE CHECK provides a qualitative assessment of channel response when the channel sensor is exposed to the radioactive source. This check is performed within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> prior to release of effluent via the associated flow path. When a SOURCE CHECK can be performed, it provides verification that the sensor will respond to an increase in radiation level. Note 1, however, does not require a SOURCE CHECK when the background radiation at the sensor is greater than the check source. This is acceptable because of the other required tests above (CHANNEL CHECK, CHANNEL TEST, CHANNEL CALIBRATION). The 8-hour restriction is reasonable because it is unlikely that the sensor will unexpectedly fail in any 8-hour period.

Note 2 provides clarification that the SOURCE CHECK applies only to the radiation detection portion of the Liquid Radwaste Monitor and is not applicable to the flow monitor portion or to the Main Steam Line Radiation Monitors.

Revision 25 80

ARKANSAS NUCLEAR ONE ODCM B 2.2 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION BASES BACKGROUND 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.

LIMITATION The following Radioactive Gaseous Effluent Monitoring Instrumentation is required to be FUNCTIONAL:

---

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

Refer to ANO-2 Technical Specification (TS) 3.3.3.1 for ANO-2 Containment Building Purge System Process Monitor (2RE-8233) operability requirements and associated ACTIONS.

ANO-1: RE-4830 - Waste Gas Holdup System Process Monitor*

RX-9820 - Reactor Building Purge and Ventilation SPING RX-9825 - Auxiliary Building Ventilation SPING RX-9830 - Spent Fuel Pool Area Ventilation SPING RX-9835 - Emergency Penetration Room Ventilation SPING ANO-2: 2RE-2429 - Waste Gas Holdup System Process Monitor*

2RX-9820 - Containment Building Purge and Ventilation SPING 2RX-9825 - Auxiliary Building Ventilation SPING 2RX-9830 - Spent Fuel Pool Area Ventilation SPING 2RX-9835 - Emergency Penetration Room Ventilation SPING 2RX-9845 - Auxiliary Building Extension Ventilation SPING 2RX-9850 - Radwaste Storage Building Ventilation SPING

• These monitors provide automatic isolation.

The radiation monitoring (process gas and SPING noble gas) and effluent flow monitoring capability are required to be FUNCTIONAL for each monitor. For SPING monitors the sample flow monitoring, the iodine sample, and the particulate sampler must also be FUNCTIONAL. With regard to Waste Gas Holdup System radiation monitoring, the alarm/trip function must also be FUNCTIONAL. The alarm/trip setpoints for specified instruments are calculated in accordance with the methods contained in ODCM Section 3.1 to ensure that the alarm/trip will occur prior to potentially exceeding the limits of 10 CFR Part 20.105. Note that the PURGE function of the ANO-1 and ANO-2 Reactor (Containment) Building is treated separately from the ventilation function.

Performance of a SOURCE CHECK on a given radiation monitor does not require the monitor to be declared non-functional due to the short period of time required to perform this test.

Revision 25 81

ARKANSAS NUCLEAR ONE ODCM B 2.2 APPLICABILITY The above monitors are required to be FUNCTIONAL during any release via the pathway in which the monitor is installed. Because SPINGs 4 and 8 monitor the Emergency Penetration Room Ventilation of ANO-1 and ANO-2, respectively, and because these ventilation systems are normally aligned for auto-start capability to aid in accident mitigation, these SPINGs must be FUNCTIONAL whenever the associated ventilation system is available for auto-start.

ACTIONS The following ACTIONS are applicable to the pathway in which a radioactive gaseous release is in progress. Because more than one release could occur simultaneously, the ACTIONS are modified by a Note that permits separate Condition entry for each non-functional Radioactive Gaseous Effluent Monitoring Instrument.

A.1 If the radiation monitoring feature, including the alarm/trip function for monitors having an automatic isolation feature, of the Waste Gas Holdup or ANO-1 Reactor Building Purge and Ventilation System Gas Activity Process or Noble Gas Activity Monitor(s) is non-functional, any release via the associated pathway must be suspended immediately. This prevents the release of unmonitored effluents to the environment.

A.2.1 In addition to Required Action A.1, a non-functional Waste Gas Holdup or ANO-1 Reactor Building Purge and Ventilation System Gas Activity Process or Noble Gas Activity Monitor, including the alarm/trip function for monitors having an automatic isolation feature, must be returned to a FUNCTIONAL status prior to the restart or subsequent release of effluents via the associated pathway. This prevents the release of unmonitored effluents to the environment. Exceptions to this requirement are included in Required Actions A.2.2.1 through A.2.2.3 below.

A.2.2.1 through A.2.2.3 In lieu of performing Required Action A.2.1 above, grab samples may be obtained and analyzed to provide a backup monitoring method for the effluent release. Because of the importance of monitoring radioactive gaseous releases, two independent samples of the effluent must be obtained and analyzed. The independency required is with regard to obtaining and analyzing each sample separately. Two independent personnel are not required to obtain and analyze the two samples.

Revision 25 82

ARKANSAS NUCLEAR ONE ODCM B 2.2 ACTIONS (continued)

A.2.2.1 through A.2.2.3 (continued)

Notwithstanding the above, computer input data and the discharge valve lineup associated with the effluent release path must be verified by two independent, qualified individuals.

Integrity of independence is maintained by preventing interaction between personnel during the verification process. With regard to valve lineups, independent verification is conducted such that each check constitutes actual identification of the valve and a determination of both required and actual valve position. Required Action A.2.2.3 is modified by a Note that excepts the valve lineup requirement from the ANO-1 Reactor Building Purge and Ventilation System since no manual valves are manipulated for this release path.

B.1 and B.2 If the flow monitoring features of the Radioactive Gaseous Effluent Monitoring Instrumentation is non-functional, the flow rate may be estimated within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> of initial loss of the instrument and every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> thereafter, for the duration of the effluent release. Flow rate data is necessary to calculate the amount of radioactive released via the effluent discharge.

Therefore, if flow cannot be estimated, it is necessary to suspend the release of radioactive effluents monitored by the affected channel. The 4-hour Completion Time is reasonable because a significant change in flow rate over the course of an effluent release is unlikely.

A Control Room RDACS trouble alarm is received when sample flows are not within predetermined limits (among other SPING conditions). With regard to SPINGs 4 or 8, procedures require a temporary sample pump to be installed when the sample flow channel is non-functional, which may be used to meet Required Action B.1, even if the flow path is in auto-standby status. With the temporary sample pump installed, Required Action D.1 will be met should the flow path auto start. Therefore, as indicated below, Condition D is not required to be considered while the SPING 4 and 8 flow paths are idle.

S 2.0.2 is not applicable to the initial flow estimation, but may be applied to the flow estimations thereafter. Pump curves may be used to estimate flow.

C.1 and C.2 Condition C is modified by two notes. Note 1 omits this Condition from being applicable to the Waste Gas Holdup or ANO-1 Reactor Building Purge and Ventilation System Gas Activity Process or Noble Gas Activity Monitors. These monitors are addressed in Condition A.

Note 2 requires the associated Required Actions and Completion Times of Condition C be applied to SPINGS 4 and 8 (Emergency Penetration Room Ventilation of ANO-1 and ANO-2, respectively) only when the pathway is in service, since noble gas activity sampling and analysis cannot be performed when the pathway is idle.

Revision 25 83

ARKANSAS NUCLEAR ONE ODCM B 2.2 ACTIONS (continued)

C.1 and C.2 (continued)

With the exception of Waste Gas Holdup System releases or during a PURGE of the ANO-1 Reactor Building, releases may continue via an associated pathway when the Noble Gas Activity Monitor(s) is non-functional, provided a sample of the effluent is obtained once every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and analyzed within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. This prevents the release of unmonitored effluents to the environment. ACTIONS C.1 and C.2 are modified by a note, referring to ANO-1 TS 3.9.3 for additional ACTIONS that may be necessary if the required ANO-1 Reactor Building Purge and Ventilation System Noble Gas Activity Monitor is inoperable.

S 2.0.2 is not applicable to the initial sample and analysis, but may be applied to the sample and analysis thereafter.

D.1, D.2, and D.2 Condition D is modified by a Note which requires the associated Required Actions and Completion Times of Condition D be applied to SPINGS 4 and 8 (Emergency Penetration Room Ventilation of ANO-1 and ANO-2, respectively) only when the pathway is in service, since iodine and particulate sampling and analysis cannot be performed when the pathway is idle.

If one or more required Iodine and/or Particulate Samplers are non-functional, auxiliary sampling equipment must be established within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. The backup Iodine and Particulate cartridges must be replaced every 7 days. Following replacement, the respective cartridge must be analyzed within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. This prevents the release of unmonitored effluents to the environment.

E.1 If the Required Actions and associated Completion Times of Condition C and/or D cannot be met, then releases via the associated pathway must be suspended. This prevents the release of unmonitored effluents to the environment.

F.1 If the Required Actions and associated Completion Times of Condition A, B, and/or E cannot be met, then additional measures may be necessary to ensure continued safe operation or to reduce overall station risk. Therefore, a condition report must be initiated immediately to assess the impact on continued effluent release operations given the degraded condition.

Revision 25 84

ARKANSAS NUCLEAR ONE ODCM B 2.2 ACTIONS (continued)

G.1 Instrumentation installed to ensure radiological monitoring of effluent releases is expected to be normally available in accordance with the design function or purpose of the equipment.

Instrumentation that remains non-functional for greater than 30 days may indicate inappropriate importance placed on the equipment or over-reliance on the backup sampling method for effluent release monitoring. As an incentive to avoid either of these conditions, Radioactive Gaseous Effluent Monitoring Instrumentation that remains non-functional for more than 30 days must be included in the Radioactive Effluent Release Report submitted pursuant to TS 5.6.3 (ANO-1) or TS 6.6.3 (ANO-2). In order to ensure inclusion, Required Action G.1 requires the condition to be tracked via a condition report.

Information to be provided in the respective Radioactive Effluent Release Report should include 1) the component number and noun name, 2) the failure mode, 3) the reason for continued inoperability, and 4) the expected return to service date.

SURVEILLANCES S 2.2.1.1 Performance of the CHANNEL CHECK every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> provides reasonable assurance for prompt identification of a gross failure of instrumentation. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels. Where parameter comparison is not possible, the CHANNEL CHECK will continue to identify gross instrument failure such as loss of power, unexpected upscale readings, failed-low indications, etc. The CHANNEL CHECK is key in verifying that the instrumentation continues to operate properly between CHANNEL CALIBRATIONs. The Frequency is based on unit operating experience that demonstrates channel failure is rare.

This Surveillance is modified by a Note the exempts the Iodine and Particulate Samplers from a CHANNEL CHECK since these components do not have electronic features or indications.

S 2.2.1.2 A local check must be made every 7 days to verify that required Iodine Sampler cartridges and Particulate Sample filters are in place. The 7-day Frequency is reasonable because it is unlikely a cartridge or filter could be inadvertently removed from the system.

Revision 25 85

ARKANSAS NUCLEAR ONE ODCM B 2.2 SURVEILLANCES (continued)

S 2.2.1.3 and S 2.2.1.6 A CHANNEL TEST is performed on required Gas Activity Process and Noble Gas Activity Monitors to ensure the entire channel will perform the intended functions. For the Waste Gas Holdup and ANO-2 Containment Building Purge Systems, the CHANNEL TEST demonstrates that automatic isolation of the associated pathway and Control Room alarm occur should the instrument indicate measured levels above the trip setpoint. The channel test also demonstrates that alarm occurs when any of the following conditions exist:

A. Power to the detector is lost.

B. The instrument indicates a downscale failure.

C. Instrument controls are not set in the operate mode.

Any setpoint adjustment shall be consistent with Section 3.1 of the ODCM.

Because the alarm/trip function and/or the importance of the release path, a CHANNEL TEST of the associated Gas Activity Process and Noble Gas Activity Monitors is required within 31 days prior to release via the Waste Gas Holdup or ANO-1 Reactor Building Purge and Ventilation Systems. This ensures the monitors are FUNCTIONAL within a reasonable period of time before such a release is commenced. All active pathway Gas Activity Process and Noble Gas Activity Monitors undergo a CHANNEL TEST once every 92 days. This Frequency is reasonable because each has a Control Room alarm function.

S 2.2.1.4 and S 2.2.1.5 A SOURCE CHECK provides a qualitative assessment of channel response when the channel sensor is exposed to the radioactive source. This check is performed within 14 days prior to release of effluent via the Waste Gas Holdup or ANO-1 Reactor Building Purge Systems. The 14-day restriction is reasonable because it is unlikely that the sensor will unexpectedly fail in any 14-day period. All active pathway Gas Activity Process and Noble Gas Activity Monitors must undergo a SOURCE CHECK every 31 days. This Frequency is reasonable because each has a Control Room alarm function.

When a SOURCE CHECK can be performed, it provides verification that the sensor will respond to an increase in radiation level. Note 1 of S 2.2.1.5 and the Note associated with S 2.2.1.4 does not require a SOURCE CHECK when the background radiation at the sensor is greater than the check source. This is acceptable because of the other required tests above (CHANNEL CHECK, CHANNEL TEST, and CHANNEL CALIBRATION).

Revision 25 86

ARKANSAS NUCLEAR ONE ODCM B 2.2 SURVEILLANCES (continued)

S 2.2.1.7 CHANNEL CALIBRATION is a complete check of the instrument channel, including the sensor. The test verifies that the channel responds to a measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drift (as required) to ensure that the instrument channel remains FUNCTIONAL between successive tests. CHANNEL CALIBRATION shall find that measurement errors and setpoint errors are within the assumptions of the setpoint calculations. CHANNEL CALIBRATIONS must be performed consistent with the assumptions of the setpoint calculations. This Frequency is justified by the assumption of at least an 18 month calibration interval to determine the magnitude of equipment drift or deviation in the setpoint calculations.

Initial CHANNEL CALIBRATION is performed using one or more of the reference standards certified by the National Institute of Standards and Technology (NIST) or using standards that have been obtained from suppliers that participate in measurement assurance activities with NIST. These standards 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 are used.

This Surveillance is modified by a Note the exempts the Iodine and Particulate Samplers from a CHANNEL CALIBRATION since these components do not have electronic features or indications.

Revision 25 87

ARKANSAS NUCLEAR ONE ODCM B 2.3 RADIOACTIVE LIQUID EFFLUENTS BASES BACKGROUND This Limitation is provided to ensure that the concentration of radioactive materials released in liquid waste effluents from the site to unrestricted areas will be less than the concentration levels specified in 10 CFR Part 20, Appendix B, Table II. This limit provides additional assurance that the levels of radioactive materials in bodies of water outside the site will not result in exposures greater than the Section II.A design objectives of 10 CFR 50, Appendix I, to a MEMBER OF THE PUBLIC.

LIMITATION The concentration limit for noble gases is based upon the assumption that Xe-133 is the controlling radioisotope and its maximum permissible concentration (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.

Radioactive nuclides other than dissolved or entrained noble gases must be maintained within the limits of 10 CFR 20, Appendix B, Table II, Column 2 values. The various dose limitations are conservative with regard to 10 CFR 20 requirements in order to provide a margin of safety through the use of as low as reasonably achievable (ALARA) practices.

Necessary portions of the LIQUID RADWASTE TREATMENT SYSTEM shall be used to reduce the radioactive materials in liquid waste prior to discharge when it is projected that the cumulative dose during a calendar quarter due to liquid effluent releases would exceed 0.18 mrem to the total body or 0.625 mrem to any organ. The provisions of this Limitation do not apply to the laundry tanks due to their incompatibility with the radwaste system.

The specified limits governing the use of appropriate portions of the LIQUID RADWASTE TREATMENT SYSTEM are a suitable fraction of the guide set forth in Section II.A of 10 CFR 50, Appendix I, for liquid effluents. The values of 0.18 mrem and 0.625 mrem are approximately 25% of the yearly design objectives on a quarterly basis. The yearly design objectives are provided in 10 CFR 50, Appendix I, Section II.

APPLICABILITY The Limitations are required to be met at all times.

ACTIONS Because more than one Limitation or Surveillance requirement may not be met at a given time, the ACTIONS are modified by a Note that permits separate Condition entry for each Limitation and/or Surveillance requirement that is not met.

Revision 25 88

ARKANSAS NUCLEAR ONE ODCM B 2.3 ACTIONS (continued)

A.1 and A.2 If any Limitation L 2.3.1.a through L 2.3.1.e is not met, action must be initiated immediately to restore the parameter within limits. This could require a reduction in offsite releases scheduled for the near future or further processing of effluents prior to release. In any event, a condition report must be initiated to determine whether additional actions are necessary to permit continued operations involving radioactive liquid effluent releases given the current circumstances. In addition, corrective action must be issued to identify and track the Limitation that was exceeded for inclusion in the annual Radioactive Effluent Release Report. However, the condition need not be reported in the annual Radioactive Effluent Release Report if reported otherwise (i.e., in accordance with reporting requirements of 10 CFR 20, 10 CFR 50.72, 10 CFR 50.73, or 40 CFR 190).

B.1 and B.2 If the sampling and/or analysis requirements of S 2.3.1.1 are not met, the release must be terminated. This action prevents or minimizes the potential for an unmonitored offsite radioactive liquid release. Such release may commence or be re-initiated once the sampling and analysis requirements of S 2.3.1.1 are met. Regardless, a condition report must be initiated to determine whether additional actions are necessary to permit continued operations involving radioactive liquid effluent releases given the current circumstances. If a condition report has already been initiated relevant to this Condition, then this assessment may be performed in conjunction with that condition report; a second condition report is not required.

C.1 and C.2 This ACTION is modified a Note, limiting its applicability to only a CONTINUOUS RELEASE of secondary coolant.

With elevated dose equivalent I-131 (DEI) activity in the secondary coolant, it is prudent to modify the frequencies for obtaining and analyzing grab samples. Therefore, with secondary coolant DEI > 0.01 µCi/ml, sample frequency is modified from once every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to once every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. The analysis of the sample must be completed with 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> of sample acquisition. More frequent monitoring of the secondary coolant will assist in detecting further increases in activity and provide personnel better opportunity for in developing corrective action plans, as necessary.

Revision 25 89

ARKANSAS NUCLEAR ONE ODCM B 2.3 ACTIONS (continued)

D.1 In accordance with 40 CFR 190, a variance must be received from the regulatory authority (NRC) if offsite dose to a member of the public will, or has exceeded, limits established in 40 CFR 190. Because Surveillance S 2.3.1.3 tracks the accumulated dose to members of the public over specified time periods (calendar quarter or calendar year), the dose may be projected and a determination made with regard to whether it is likely 40 CFR 190 limits will be exceeded. If 40 CFR 190 limits are projected to be exceeded, an application for a variance from the NRC must be submitted prior to the estimated date in which any 40 CFR 190 limit will be exceeded. The variance will allow continued offsite liquid and gaseous releases in excess of 40 CFR 190 limits. Note that the variance is normally expected to remain in effect until the end of the current calendar year since 40 CFR 190 limits only apply to the calculated annual dose to members of the public.

If application for variance cannot be made prior to exceeding any 40 CFR 190 limit, it may be prudent to notify the NRC by phone as soon as possible of the need for a variance, providing the expected date in which the application will be submitted. Note that the NRC may provide verbal approval for variance in situations where time is a factor.

E.1 If the Required Actions and associated Completion Times of Conditions C and/or D cannot be met or if the sampling and/or analysis requirements denoted in Surveillances S 2.3.1.1 and/or S 2.3.1.2 are not met, then additional measures may be necessary to ensure continued safe operation or to reduce overall station risk. Therefore, a condition report must be initiated immediately to assess the impact on continued effluent release operations given the requirements that are not being met.

F.1 Surveillance S 2.3.1.5 establishes required capability of various sample analyses. A given analysis must be capable of detecting respective radioactivity at a reasonably low threshold in order to ensure radioactive liquid releases to the public are carefully and accurately monitored.

If the stated thresholds can not be met, a condition report must be initiated and corrective action issued to ensure the condition is included and described in the annual Radioactive Effluent Release Report.

Revision 25 90

ARKANSAS NUCLEAR ONE ODCM SURVEILLANCES S 2.3.1.1 and S 2.3.1.2 All radioactive liquid effluent releases are required to be monitored. Because a BATCH RELEASE is of a known quantity and of finite duration, sampling of batch effluents must be performed prior to release. In addition, the sample must undergo a gamma isotopic and DEI analysis prior to the release to provide high confidence that radioactive release limits will not be exceeded. Remaining analyses may then be completed at the designated Frequency during or following the release.

For a BATCH RELEASE, a composite sample, one in which the quantity of liquid sampled is proportional to the quantity of liquid waste discharged and in which the method of sampling employed results in a specimen which is representative of the liquids released, is performed.

In order to ensure a representative sample, the batch shall be thoroughly mixed before the sample is obtained.

Unlike the BATCH RELEASE, a CONTINUOUS RELEASE must be monitored at a set Frequency. While gross activity monitoring is available for various release paths as is recommended by Regulatory Guide (RG) 1.21, Measuring, Evaluating, and Reporting Radioactivity in Solid Wastes and Releases of Radioactive Materials in Liquid and Gaseous Effluents from Light-Water-Cooled Nuclear Power Plants, such monitoring does not provide the necessary breakdown and quantification of radioactivities being discharged. Therefore, the ODCM requires grab samples and analyses of these effluents at a specified Frequency.

To be representative of the quantities and concentrations of radioactive materials in liquid effluents, a CONTINUOUS RELEASE sample must be proportional to the rate of flow of the effluent stream.

S 2.3.1.3 Limitation L 2.3.1 establishes limits on radioactive liquid concentrations discharged from the plant and the accumulative dose that may be received by a MEMBER OF THE PUBLIC as a result of such releases. In order to determine that these limits are met and being maintained, the results of analyses required by Surveillances S 2.3.1.1 and S 2.3.1.2 must be compared to the Limitation requirements on a specified Frequency. Therefore, analysis results obtained within a given 7-day period must be considered, in some cases along with previous analysis results of all liquid release over a specified period of time (calendar quarter or calendar year),

to ensure limits are not exceeded.

S 2.3.1.4 In accordance with 40 CFR 190, a variance must be received from the regulatory authority (NRC) is offsite dose to a member of the public will, or has exceeded, limits established in 40 CFR 190. Because Surveillance S 2.3.1.3 tracks the accumulated dose to members of the public over specified time periods (calendar quarter or calendar year), the dose may be projected and a determination made with regard to whether it is likely 40 CFR 190 limits will be exceeded. The 31-day Frequency is acceptable because associated ODCM limits for these releases are significantly less than those described in 40 CFR 190 and, therefore, it is unlikely any 40 CFR 190 limit would be exceeded in any 31-day period.

Revision 25 91

ARKANSAS NUCLEAR ONE ODCM B 2.3 SURVEILLANCES (continued)

S 2.3.1.5 The Lower Limit of Detection (LLD) is the smallest concentration of radioactive material in a sample that will be detected with 95% probability with 5% probability of falsely concluding that a blank observation represents a real signal. This Surveillance contains a list of isotopes and required LLD for each. Sample analysis sensitivity must be such that radioactivities can be detected and measured at the LLD value.

It should be recognized that the LLD is an a Priori (before the fact) limit representing the capability of measurement system and not an a Posteriori (after the fact) limit for a particular measurement.

For a particular measurement system (which may include radio-chemical separation):

4.66Sb LLD =

E

• V

• T

• 2.22

• Y

• e-t where:

LLD = lower limit of detection as defined above (as pCi per unit mass or volume)

Sb = standard deviation of the background or blank sample counts

= square root of either the background or the blank sample counts E = counting efficiency (as counts per transformation)

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

T = elapsed count time 2.22 = number of transformations per minute per picocurie Y = fractional radiochemical yield (when applicable)

= radioactive decay constant for the particular radionuclide t = elapsed time between sample collection (or end of the sample collection period) and time of counting Typical values of E, V, Y and t should be used in the calculation.

For certain mixtures of gamma emitters, it may not be possible to measure radionuclides in concentrations near their sensitivity limits when other nuclides are present in the sample in much greater concentrations. Under these circumstances, it will be more appropriate to calculate the concentration of such radionuclides using observed ratios with those radionuclides which are measurable.

Revision 25 92

ARKANSAS NUCLEAR ONE ODCM B 2.3 SURVEILLANCES (continued)

S 2.3.1.5 (continued)

The principal gamma emitters for which the LLD limitation will apply are exclusively the following radionuclides: Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, Cs-134, Cs-137, Ce-141, and Ce-144. This list does not mean that only these nuclides are to be detected and reported.

Other peaks which are measurable and identifiable, together with the above nuclides, shall also be identified and reported. Nuclides which are below the LLD for the analyses should not be reported as being present at the LLD level. When unusual circumstances result in LLD requirements not being met, the reasons shall be documented in the Radioactive Effluent Release Report as stated in Required Action F.1 of this Limitation, or the Annual Radiological Environmental Operating Report as stated in L 2.5.1, Required Action A.2.

Revision 25 93

ARKANSAS NUCLEAR ONE ODCM B 2.4 RADIOACTIVE GASEOUS EFFLUENTS BASES BACKGROUND This Limitation is provided to ensure that radioactive materials released in gaseous effluents from the site to unrestricted areas will be less than the limits specified in 10 CFR Part 20. This Limitation also implements the requirements of Sections II.C, III.A, and IV.A of 10 CFR 50, Appendix I.

Figure 4-2 illustrates the maximum area boundary for radioactive release calculations. For individuals who may at times be within the exclusion area boundary, the occupancy of the individual will be sufficiently low to compensate for any increase in the atmospheric diffusion factor above that for the exclusion area boundary.

LIMITATION Radioactive nuclides must be maintained within the limits of 10 CFR 20. The various dose rate and dose limitations are conservative with regard to 10 CFR 20 requirements in order to provide a margin of safety through the use of as low as reasonably achievable (ALARA) practices.

The necessary VENTILATION EXHAUST TREATMENT SYSTEMs shall be used to reduce the radioactive materials in gases prior to discharge when it is projected that the cumulative dose during a calendar quarter due to gaseous effluent releases would exceed values specified in this Limitation. The specified limits governing the use of the VENTILATION EXHAUST TREATMENT SYSTEMs are a suitable fraction of the dose design objectives set forth in Sections II.B and II.C of 10 CFR Part 50, Appendix I, for gaseous effluents.

APPLICABILITY The Limitations are required to be met at all times.

ACTIONS Because more than one Limitation or Surveillance requirement may not be met at a given time, the ACTIONS are modified by a Note that permits separate Condition entry for each Limitation and/or Surveillance requirement that is not met.

Revision 25 94

ARKANSAS NUCLEAR ONE ODCM B 2.4 ACTIONS (continued)

A.1 and A.2 If any Limitation L 2.4.1.a through L 2.4.1.d is not met, action must be initiated immediately to restore the parameter within limits. This could require a reduction in offsite releases scheduled for the near future or further processing of effluents prior to release. In any event, a condition report must be initiated to determine whether additional actions are necessary to permit continued operations involving radioactive gaseous effluent releases given the current circumstances. In addition, corrective action must be issued to identify and track the Limitation that was exceeded for inclusion in the annual Radioactive Effluent Release Report. However, the condition need not be reported in the annual Radioactive Effluent Release Report if reported otherwise (i.e., in accordance with reporting requirements of 10 CFR 20, 10 CFR 50.72, 10 CFR 50.73, or 40 CFR 190).

B.1 and B.2 If the sampling and/or analysis requirements of S 2.4.1.1 are not met, the release must be terminated. This action prevents or minimizes the potential for an unmonitored offsite radioactive liquid release. Such release may commence or be re-initiated once the sampling and analysis requirements of S 2.4.1.1 are met. Regardless, a condition report must be initiated to determine whether additional actions are necessary to permit continued operations involving radioactive liquid effluent releases given the current circumstances. If a condition report has already been initiated relevant to this Condition, then this assessment may be performed in conjunction with that condition report; a second Condition Report is not required.

C.1 In accordance with 40 CFR 190, a variance must be received from the regulatory authority (NRC) if offsite dose to a member of the public will, or has exceeded, limits established in 40 CFR 190. Because Surveillance S 2.4.1.3 tracks the accumulated dose to members of the public over specified time periods (calendar quarter or calendar year), the dose may be projected and a determination made with regard to whether it is likely 40 CFR 190 limits will be exceeded. If 40 CFR 190 limits are projected to be exceeded, an application for a variance from the NRC must be submitted prior to the estimated date in which any 40 CFR 190 limit will be exceeded. The variance will allow continued offsite liquid and gaseous releases in excess of 40 CFR 190 limits. Note that the variance is normally expected to remain in effect until the end of the current calendar year since 40 CFR 190 limits only apply to the calculated annual dose to members of the public.

If application for variance cannot be made prior to exceeding any 40 CFR 190 limit, it may be prudent to notify the NRC by phone as soon as possible of the need for a variance, providing the expected date in which the application will be submitted. Note that the NRC may provide verbal approval for variance in situations where time is a factor.

Revision 25 95

ARKANSAS NUCLEAR ONE ODCM B 2.4 ACTIONS (continued)

D.1 If the Required Actions and associated Completion Times of Condition C cannot be met or if the sampling and/or analysis requirements denoted in Surveillances S 2.4.1.2 are not met, then additional measures may be necessary to ensure continued safe operation or to reduce overall station risk. Therefore, a condition report must be initiated immediately to assess the impact on continued effluent release operations given the requirements that are not being met.

E.1 Surveillance S 2.4.1.5 establishes required capability of various sample analyses. A given analysis must be capable of detecting respective radioactivity at a reasonably low threshold in order to ensure radioactive gaseous releases to the public are carefully and accurately monitored. If the stated thresholds can not be met, a condition report must be initiated and corrective action issued to ensure the condition is included and described in the annual Radioactive Effluent Release Report.

SURVEILLANCES Continuous gaseous release paths are monitored by instrumentation denoted in Limitation L 2.2.1. Limitation L 2.2.1 provides Required Actions and Completion Times for circumstances when required instrumentation is out of service. Therefore, the Surveillances associated with this Limitation (L 2.4.1) envelop only required grab, charcoal, and particulate samples necessary to verify 10 CFR 20 limits will be met.

The Surveillance Limitations implement the requirements in 10 CFR 50, Appendix I, Section III.A, 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 calculations established in this manual for calculating the doses due to the actual release rates of radioactive noble gases in gaseous effluents are consistent with the methodology provided in RG 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 RG 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 equations in this manual provided for determining the air doses at and beyond the site boundary are based upon the historical average atmospheric conditions.

The release rate limitations for iodine-131, tritium, and radionuclides in particulate form with half-lives greater than 8 days are dependent on the existing radionuclide pathways to man in the areas at or beyond the site boundary. The pathways that were examined in the development of these calculations were: 1) individual inhalation of airborne radionuclides, 2) deposition of radionuclides onto green leafy vegetation with subsequent consumption by man, 3) deposition onto grassy areas where milk animals and meat producing animals graze with consumption of the milk and meat by man, and 4) deposition on the ground with subsequent exposure of man.

Revision 25 96

ARKANSAS NUCLEAR ONE ODCM B 2.4 SURVEILLANCES (continued)

S 2.4.1.1 and S 2.4.1.2 All radioactive gaseous effluent releases are required to be monitored. Because a Waste Gas Holdup Tank or Reactor (Containment) Building Purge release is of a known (or estimated) quantity and of finite duration, sampling of these effluents must be performed prior to release.

In addition, the sample must be analyzed for principal gamma emitters and tritium prior to the release in order to provide high confidence that radioactive release limits will not be exceeded.

S 2.4.1.3 To meet the intent of the continuous monitoring requirement for noble gases, the noble gas activity from each SPING operating on an activity flow path must be recorded at least once every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The current, highest, and average activity recorded from a particular SPING over the required grab sample period designated in other Surveillances associated with this Limitation are used to scale the noble gas and tritium activity obtained from the associated grab sample. The final resulting activity is used, in part, to support completion of S 2.4.1.4 and S 2.4.1.5 below.

S 2.4.1.4 Limitation L 2.4.1 establishes limits on radioactive gases discharged from the plant and the dose rates and accumulative dose that may be received by a MEMBER OF THE PUBLIC as a result of such releases. In order to determine that these limits are met and being maintained, the results of analyses required by Surveillances S 2.4.1.1 and S 2.4.1.2, as adjusted by readings taken in accordance with S 2.4.1.3 as appropriate must be compared to the Limitation requirements on a specified Frequency. Therefore, analysis results obtained within a given 31-day period must be considered, in some cases along with previous analysis results of all gaseous releases over a specified period of time (calendar quarter or calendar year), to ensure limits are not exceeded.

The ratio of the sample flow rate to the sampled stream flow rate must be known for the time period covered by each dose or dose rate calculation made in accordance with this Limitation.

S 2.4.1.5 In accordance with 40 CFR 190, a variance must be received from the regulatory authority (NRC) is offsite dose to a member of the public will, or has exceeded, limits established in 40 CFR 190. Because Surveillance S 2.4.1.3 tracks the accumulated dose to members of the public over specified time periods (calendar quarter or calendar year), the dose may be projected and a determination made with regard to whether it is likely 40 CFR 190 limits will be exceeded. The 31-day Frequency is acceptable because associated ODCM limits for these releases are significantly less than those described in 40 CFR 190 and, therefore, it is unlikely any 40 CFR 190 limit would be exceeded in any 31-day period.

Revision 25 97

ARKANSAS NUCLEAR ONE ODCM B 2.4 SURVEILLANCES (continued)

S 2.4.1.6 The Lower Limit of Detection (LLD) is the smallest concentration of radioactive material in a sample that will be detected with 95% probability with 5% probability of falsely concluding that a blank observation represents a real signal. This Surveillance contains a list of isotopes and required LLD for each. Sample analysis sensitivity must be such that radioactivies can be detected and measured at the LLD value. The Surveillance also contains the LLD for the Noble Gas Monitors associated with Limitation 2.2.1.

For an explanation of the LLD calculation, refer to the S 2.3.1.5 Bases.

For certain radionuclides with low gamma yield or low energies, or for certain radionuclides mixtures, it may not be possible to measure radionuclides in concentrations near the LLD.

Under these circumstances, the LLD may be increased inversely proportional to the magnitude of the gamma yield (i.e., (1 x 10-4/I)), where I is the photon abundance expressed as a decimal fraction), but in no case shall the LLD, as calculated in this manner for a specific radionuclide, be > 10% of the MPC value specified in 10 CFR 20, Appendix B, Table II, Column 1.

The principal gamma emitters for which the LLD limitation will apply are exclusively the following radionuclides: Kr-87, Kr-88, Xe-133, Xe-133m, Xe-135, and Xe-138 for gaseous emissions and Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, Cs-134, Cs-137, Ce-141 and Ce-144 for particulate emissions. This list does not mean that only these nuclides are to be detected and reported. Other peaks which are measurable and identifiable, together with the above nuclides, shall also be identified and reported. Nuclides which are below the LLD for the analyses should not be reported as being present at the LLD level for that nuclide. When unusual circumstances result in LLD's higher than required, the reasons shall be documented in the Radioactive Effluent Release Report.

Revision 25 98

ARKANSAS NUCLEAR ONE ODCM B 2.5 RADIOLOGICAL ENVIRONMENTAL MONITORING B 2.5.1 Environmental Sampling BASES BACKGROUND The ODCM includes, in tables and figures, specific parameters of distance and direction from the centerline of one reactor, and additional description where pertinent, for each sample location required by the Radiological Environmental Monitoring Program. NUREG-0133, "Preparation of Radiological Technical Specifications for Nuclear Power Plants," October 1978, and Radiological Assessment Branch Technical Position (BTP), Revision 1, November 1979, provide guidance with regard to environmental sampling.

With regard to the aforementioned BTP, one airborne sample location should be from the vicinity of a community having the highest calculated annual average ground-level D/Q.

Because community is undefined in the regulations and because ANO is located in a largely rural area, Russellville is conservatively considered the community of choice due to its much larger population than the surrounding rural towns. While Russellville may not be located in the highest D/Q zone, the three highest D/Q sample stations provide sufficient information to determine impact on the surrounding rural zones.

The approximate locations of selected sample sites are shown on ODCM Figures 4-1, 4-1A, and 4-1B for illustrative purposes. ODCM Table 4-1 lists the approximate distances and directions of the sample stations from the plant.

D/Q refers to a radiological deposition rate considering prevalent winds around the site and is used to determine natural settling of effluents from the atmosphere.

LIMITATION This Limitation specifies the sample locations and distances, sample analysis type and frequency, and parameters to be sampled as part of the Radiological Environmental Monitoring Program.

The Limitation is modified by a Note that permits other instrumentation to be used in place of, or in addition to, integrating dosimeters for measuring and recording dose rate continuously.

For the purposes of this Limitation, a thermoluminescent dosimeter may be considered to be one phosphor and two or more phosphors in a packet considered as two or more dosimeters.

Film badges should not be used for measuring direct radiation.

APPLICABILITY The Limitations are required to be met at all times.

Revision 25 99

ARKANSAS NUCLEAR ONE ODCM B 2.5.1 ACTIONS Because more than one Limitation or Surveillance requirement may not be met at a given time, the ACTIONS are modified by a Note that permits separate Condition entry for each Limitation and/or Surveillance requirement that is not met.

A.1 and A.2 Deviations are permitted from the required sampling schedule if specimens are unobtainable due to hazardous conditions, seasonal unavailability, malfunction of automatic sampling equipment and other legitimate reasons. If specimens are unobtainable due to sampling equipment malfunctions, every effort shall be made to complete corrective action before the end of the next sampling period. All deviations from the sampling schedule shall be documented in the Annual Radiological Environmental Operating Report.

This ACTION lists several items that would result in the intent of the Radiological Environmental Monitoring Program not being met. In addition, this ACTION provides guidance for conditions where radionuclides other than those listed in Table 2.5-2 could result in a noteworthy dose to a MEMBER OF THE PUBLIC. Immediate action is required to restore conditions needed to meet the intent of the Radiological Environmental Monitoring Program.

All deviations from the Limitations and Surveillances required to meet the intent of the Radiological Environmental Monitoring Program must be reported in the Annual Radiological Environmental Operating Report. However, the condition need not be reported in the Annual Radiological Environmental Operating Report if reported otherwise (i.e., in accordance with reporting requirements of 10 CFR 20, 10 CFR 50.72, 10 CFR 50.73, or 40 CFR 190).

With the level of radioactivity as the result of plant effluents in an environmental sampling medium at one or more required locations exceeding the limits of Table 2.5-2 when averaged over any calendar quarter, the condition must be reported in accordance with Required Action A.2. The report should include an evaluation of any release conditions, environmental factors or other aspects which caused the limits to be exceeded, and define the actions taken to reduce radioactive effluents so that the potential annual dose to a MEMBER OF THE PUBLIC will remain less than the calendar year limits of Limitations L 2.3.1 and L 2.4.1. When more than one of the radionuclides in Table 2.5-2 is detected in the sampling medium, the information shall be included in the report if:

Concentration 1 Concentration 2 etc.

+ + 1.0 Reporting Level 1 Reporting Level 2 etc.

B.1 In addition to the requirements of Required Actions A.1 and A.2, a new location must be identified and added to the Radiological Environmental Monitoring Program within 30 days when required samples cannot be obtained from designated locations. Note that broad leaf samples are only required when milk samples are unavailable, pursuant to S 2.5.1.8.

Revision 25 100

ARKANSAS NUCLEAR ONE ODCM B 2.5.1 ACTIONS (continued)

B.1 (continued)

The specific locations from which samples were unavailable may then be deleted from the monitoring program. The cause(s) of the unavailability of samples the new location(s) for obtaining replacement samples shall be identified in next Annual Radiological Environmental Operating Report. The report shall also include a revised Table 4-1 reflecting the new location(s).

SURVEILLANCES S 2.5.1.1 through S 2.5.1.8 These Surveillances ensure samples are collected and analyzed at specified frequencies of the parameters, and from the locations, designated in Limitation L 2.5.1. The approximate locations of selected sample sites are shown on ODCM Figures 4-1, 4-1A, and 4-1B for illustrative purposes. ODCM Table 4-1 lists the approximate distances and directions of the sample stations from the plant.

Note that the gross beta analysis of required particulate samplers should not be performed within the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following particulate filter change. This is to allow for radon and thoron daughter decay. If it is discovered that the particulate gross beta activity is more than 10 times the yearly mean of control samples for any medium, consideration should be given to performing a gamma isotopic analysis of the individual particulate samples. Also note that particulate samples may need to be collected more frequently than the specified 14-day Frequency due to dust or other accumulation of matter.

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

S 2.5.1.9 The Lower Limit of Detection (LLD) is the smallest concentration of radioactive material in a sample that will be detected with 95% probability with 5% probability of falsely concluding that a blank observation represents a real signal. Table 2.5-1 contains a list of isotopes and required LLD for each. Sample analysis sensitivity must be such that radioactivities can be detected and measured at the LLD value.

For an explanation of the LLD calculation, refer to the S 2.3.1.5 Bases.

S 2.5.1.10 With the level of radioactivity as the result of plant effluents in an environmental sampling medium at one or more required locations exceeding the limits of Table 2.5-2 when averaged over any calendar quarter, the condition must be reported in accordance with Required Action A.2.

Revision 25 101

ARKANSAS NUCLEAR ONE ODCM B 2.5 RADIOLOGICAL ENVIRONMENTAL MONITORING B 2.5.2 Land Use Census BASES BACKGROUND The surveys required by this Limitation ensure that changes in environmental conditions as they relate to radioactive effluent releases from the site are identified and accounted for in the overall dose commitment to the public.

LIMITATION This Limitation ensures changes in the use of unrestricted areas are identified and that modifications are subsequently included in the Radiological Environmental Monitoring Program. The census satisfies 10 CFR 50, Appendix I, Section IV.B.3.

Restricting the census to gardens of > 500 ft2 provides assurance that significant exposure pathway via leafy vegetables will be identified and monitored since a garden of this size is the minimum required to produce the quantity (26 kg/year) of leafy vegetables assumed in RG 1.109 for consumption by a child. This minimum garden size was determined assuming

1) that 20% of the garden was used for growing broad leaf vegetation (i.e., similar to lettuce and cabbage) and, 2) a vegetation yield of 2 kg/m2.

The Limitation is modified by a Note that permits broad leaf vegetation sampling to be performed at the site boundary in the directional sector having the highest D/Q in lieu of performing a garden census. D/Q refers to a radiological deposition rate considering prevalent winds around the site and is used to determine natural settling of effluents from the atmosphere.

APPLICABILITY The Limitations are required to be met at all times.

ACTIONS Because more than one new sample location may be identified during a given census, the ACTIONS are modified by a Note permit separate Condition entry for each new location identified.

Revision 25 102

ARKANSAS NUCLEAR ONE ODCM B 2.5.2 ACTIONS (continued)

A.1, A.2.1, and A.2.2 When new locations are discovered that indicate higher radioactivity levels than current locations being sample pursuant to Limitation L 2.5.1 or if radioactivity levels at a new location are projected to exceed 40 CFR 190 limits (with regard to I-131, H-3, and particulate sources),

a condition report must be immediately initiated. Initiating a condition report will ensure reporting criteria is evaluated for the given condition. Regardless of any other report, the new location must be included in the next Annual Radiological Environmental Operating Report.

In addition to the requirements of Required Action A.1, the new location must be added to the Radiological Environmental Monitoring Program within 30 days. Following October 31 of the year in which the census is taken, the old sample location in this same pathway may be deleted from the Radiological Environmental Monitoring Program. This is expected to be performed within 90 days following the October 31 limit.

SURVEILLANCES S 2.5.2.1 through S 2.5.2.2 The land use census must be performed every 24 months and between the dates of June 1 and October 1 of the given year. The results of the census must be reported in the next Annual Radiological Environmental Operating Report.

The Surveillance requirements are modified by a Note that prevents the use of S 2.0.2.

Therefore, the 25% Frequency extension associated with S 2.0.2 cannot be applied to the Surveillances associated with this Limitation. This is because the Frequencies are associated with strict performance and reporting dates which cannot be exceeded.

Revision 25 103

ARKANSAS NUCLEAR ONE ODCM B 2.5 RADIOLOGICAL ENVIRONMENTAL MONITORING B 2.5.3 Interlaboratory Comparison Program BASES BACKGROUND This Limitation refers to the off-site radiochemistry laboratory. The Limitation provides independent checks on the accuracy of the measurements of radioactive material in environmental samples.

LIMITATION 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 a quality assurance program for environmental monitoring in order to demonstrate that the results are reasonably valid.

APPLICABILITY The Limitations are required to be met at all times.

ACTION A.1 Failure to meet the requirements of the Interlaboratory Comparison Program requires initiating a condition report to ensure the circumstances are included in the next Annual Radiological Environmental Operating Report.

SURVEILLANCE S 2.5.3.1 The results of the Interlaboratory Comparison Program analyses must be reported in the next Annual Radiological Environmental Operating Report.

Revision 25 104

Attachment 2 to 0CAN041504 EN-RW-105, Process Control Program

NUCLEAR QUALITY RELATED EN-RW-105 REV. 4 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 1 OF 21 PROCESS CONTROL PROGRAM Procedure Contains NMM ECH eB REFLIB Forms: YES NO HQN Effective Procedure Owner: Donnie Marvel Governance Owner: David Moore Date

Title:

Manager, RP

Title:

Manager, Fleet RP Site: ANO Site: HQN 3/25/14 Site Site Procedure Champion Title ANO Donnie Marvel Manager, RP BRP N/A N/A CNS Bob Beilke Manager, RP GGNS Roy Miller Manager, RP IPEC Frank Mitchell Manager, RP JAF Robert Brown Manager, RP PLP Doug Watkins Manager, RP PNPS Steven Brewer Manager, RP RBS Jim Hogan Manager, RP (acting)

VY David Tkatch Manager, RP W3 Daniel Frey Manager, RP HQN David Moore Manager, Fleet RP For site implementation dates see ECH eB REFLIB using site tree view (Navigation panel).

Site and NMM Procedures Canceled or Superseded By This Revision None Process Applicability Exclusion: All Sites:

Specific Sites: ANO BRP CNS GGNS IPEC JAF PLP PNPS RBS VY W3 Change Statement Editorial revision to address the issue identified in CR-HQN-2013-00858, CA-02 (Develop a draft procedure that includes instructions for vendors processing waste still owned by Entergy to comply with the PCP program.)

Reworded Step 5.1[1](b) to improve clarity: inserted text processed on-site OR off-site by vendors Associated PRHQN #: PR-PRHQN-2014-00048

NUCLEAR QUALITY RELATED EN-RW-105 REV. 4 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 2 OF 21 PROCESS CONTROL PROGRAM TABLE OF CONTENTS Section Title Page 1.0 PURPOSE ................................................................................... 3

2.0 REFERENCES

............................................................................ 3 3.0 DEFINITIONS .............................................................................. 6 4.0 RESPONSIBILITIES .................................................................... 9 5.0 DETAILS .................................................................................... 10 6.0 INTERFACES ............................................................................ 20 7.0 RECORDS ................................................................................. 20 8.0 SITE SPECIFIC COMMITMENTS ............................................. 21 9.0 ATTACHMENTS ........................................................................ 21

NUCLEAR QUALITY RELATED EN-RW-105 REV. 4 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 3 OF 21 PROCESS CONTROL PROGRAM 1.0 PURPOSE The Process Control Program (PCP) requires formulas, sampling, analyses, test and determinations to be made to ensure that the processing and packing of solid radioactive wastes based on demonstrated processing of actual or simulated wet solid wastes will be accomplished in such a way as to assure compliance with 10 CFR Parts 20, 61 and 71, State Regulations, burial ground requirements, and other requirements governing the disposal of solid radioactive waste. The scope of a PCP is to assure that radioactive waste will be handled, shipped, and disposed of in a safe manner in accordance with approved site or vendor procedures, whichever is applicable. [GGNS UFSAR, Chapter 16B.1 / TRM - 7.6.3.8 paragraph 1]

1.1 The purpose of this document is to provide a description of the solid radioactive waste Process Control Program (PCP) at all the Entergy fleet sites. The PCP describes the methods used for processing, classification and packaging low-level wet radioactive waste into a form acceptable for interim on-site storage, shipping and disposal, in accordance with 10 CFR Part 61 and current disposal site criteria.

1.2 To ensure the safe operation of the solid radwaste system, the solid radwaste system will be used in accordance with this Process Control Program to process radioactive wastes to meet interim on-site storage, shipping and burial ground requirements.

1.3 This document addresses the process control program in the context of disposal criteria, on-site processing and vendor processing requirements.

1.4 The Process Control Program implements the requirements of 10CFR50.36a and General Design Criteria 60 of Appendix A to 10CFR Part 50. The process parameters included in the Process Control Program may include but are not limited to waste type, waste pH, waste/liquid/solidification agent/catalyst ratios, waste oil content, waste principal chemical constituents, and mixing and curing times.

1.5 This document does NOT address the requirements for 10CFR Part 61.56 (waste characteristics) for material sent to intermediate processors, because the final treatment and packaging is performed at the vendor facilities.

2.0 REFERENCES

[1] EN-QV-104, Entergy Quality Assurance Program Manual Control

[2] Title 49, Code of Federal Regulations

[3] Title 10, Code of Federal Regulations, Part 20

NUCLEAR QUALITY RELATED EN-RW-105 REV. 4 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 4 OF 21 PROCESS CONTROL PROGRAM 2.0 continued

[4] Title 10, Code of Federal Regulations, Part 61

[5] Title 10, Code of Federal Regulations, Part 71, Appendix H [QAPM, Section A.1.c]

[6] Low-Level Waste Licensing Branch Technical Position on Radioactive Waste Classification, 11 May 1983

[7] Disposal Site Criteria and License

[8] Waste Processor Acceptance Criteria

[9] EN-LI-100, Process Applicability Determination

[10] NRC Information and Enforcement Bulletins NRC Information Notice 79-19: Packaging of Low-Level Radioactive Waste for Transport and Burial.

NRC Information Notice 80-24: Low-Level Radioactive Waste Burial Criteria.

NRC Information Notice 80-32: Clarification of Certain Requirements for Exclusive-Use Shipments of Radioactive Materials.

NRC Information Notice 80-32, Rev. 1: Clarification of Certain Requirements for Exclusive-Use Shipments of Radioactive Materials.

NRC Information Notice 83-05: Obtaining Approval for Disposing of Very-Low-Level Radioactive Waste - 10CFR Section 20.302.

NRC Information Notice 83-10: Clarification of Several Aspects Relating to Use of NRC-Certified Transport Packages.

NRC Information Notice 83-33: Non-Representative Sampling of Contaminated Oil.

NRC Information Notice 84-50: Clarification of Scope of Quality Assurance Programs for Transport Packages Pursuant to 10CFR 50 Appendix B.

NRC Information Notice 84-72: Clarification of Conditions for Waste Shipments Subject to Hydrogen Gas Generation.

NRC Information Notice 85-92: Surveys of Wastes Before Disposal from Nuclear Reactor Facilities.

NRC Information Notice 86-20: Low-Level Radioactive Waste Scaling Factors, 10CFR 61.

NRC Information Notice 86-90: Requests to Dispose of Very Low-Level Radioactive Waste Pursuant 10CFR 20.302

NUCLEAR QUALITY RELATED EN-RW-105 REV. 4 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 5 OF 21 PROCESS CONTROL PROGRAM 2.0[10], continued NRC Information Notice 87-03: Segregation of Hazardous and Low-Level Radioactive Wastes NRC Information Notice 87-07: Quality Control of On-Site Dewatering/ Solidification Operations by Outside Contractors

[11] NRC Information and Enforcement Bulletins (continued)

NRC Information Notice 89-27: Limitations on the Use of Waste Forms and High Integrity Containers for the Disposal of Low-Level Radioactive Waste NRC Information Notice 92-62: Emergency Response Information Requirements for Radioactive Material Shipments NRC Information Notice 92-72: Employee Training and Shipper Registration Requirements for Transporting Radioactive Materials NRC 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.

[12] Nureg-0800 Standard Review Plan Section 11.4 Revision 2, Solid Waste Management Systems.

[13] NRC Waste Form Technical Position, Revision 1 Jan 24 1991.

[14] NRC SECY 94-198 Review of Existing Guidance Concerning the Extended Storage of Low-Level Radioactive Waste.

[15] EPRI TR-106925 Rev-1, Interim On-Site Storage of Low Level Waste: Guidelines for Extended Storage - October1996

[16] NRC Branch Technical Position On Concentration Averaging And Encapsulation Jan 17 1995

[17] Commitment Documents (U-2 and U-3)

IPN-99-079, Supplement to Proposed Changes to Technical Specifications Incorporating Recommendations of Generic Letter 89-01 and the Revised 10 CFR Part 20 and 10 CFR Part 50.36a.

Appendix B Technical Specifications, Section 4.5 [IP, RECS ODCM Part 1]

NUCLEAR QUALITY RELATED EN-RW-105 REV. 4 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 6 OF 21 PROCESS CONTROL PROGRAM 3.0 DEFINITIONS

[1] Batch - A quantity of waste to be processed having essentially consistent physical and chemical characteristics as determined through past experience or system operation knowledge by the Radwaste Shipping Specialist. A batch could be a waste tank, several waste tanks grouped together or a designated time period such as between outages as with the DAW waste stream. An isolated quantity of feed waste to be processed having essentially constant physical and chemical characteristics.

(The addition or removal of water will not be considered to create a new batch).

[2] Certificate of Compliance - Document issued by the USNRC regulating use of a NRC licensed cask or issued by (SCDHEC) South Carolina Department of Health and Environmental Conservation regulating a High Integrity Container.

[3] Chelating Agents - EDTA, DTPA, hydroxy-carboxylic acids, citric acid, carbolic acid and glucinic acid.

[4] Compaction - The process of volume reducing solid waste by applying external pressure.

[5] Confirmatory Analysis - The practice of verifying that gross radioactivity measurements using MCA are reasonably consistent with independent laboratory sample data.

[6] Dewatered Waste - Wet waste that has been processed by means other than solidification, encapsulation, or absorption to meet the free standing liquid requirements of 10CFR Part 61.56 (a)(3) and (b)(2).

[7] De-watering - The removal of water or liquid from a waste form, usually by gravity or pumping.

[8] Dilution Factor - The RADMAN computer code factor to account for the non-radioactive binder added to the waste stream in the final product when waste is solidified.

[9] Dry Waste - Radioactive waste which exist primarily in a non-liquid form and includes such items as dry materials, metals, resins, filter media and sludges.

[10] Encapsulation - Encapsulation is a means of providing stability for certain types of waste by surrounding the waste by an appropriate encapsulation media.

[11] Gamma-Spectral-Analysis - Also known as IG, MCA, Ge/Li and gamma spectroscopy.

[12] Gross Radioactivity Measurements - More commonly known as dose to curie conversion for packaged waste characterization and classification.

NUCLEAR QUALITY RELATED EN-RW-105 REV. 4 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 7 OF 21 PROCESS CONTROL PROGRAM 3.0 continued

[13] Homogeneous - Of the same kind or nature; essentially alike. Most Volumetric waste streams are considered homogeneous for purposes of waste classification.

[14] Incineration - The process of burning a combustible material to reduce its volume and yield an ash residue.

[15] Liquid Waste - Radioactive waste that exist primarily in a liquid form and is contained in other than installed plant systems, to include such items as oil, EHC fluid, and other liquids. This waste is normally processed off-site.

[16] Low-Level Radioactive Waste (LLW) - Those wastes containing source, special nuclear, or by-product material that are acceptable for disposal in a land disposal facility. For the purposes of this definition, low-level radioactive waste has the same meaning as in the Low-Level Waste Policy Act, that is, radioactive waste not classified as high-level radioactive waste, transuranic waste, spent nuclear fuel, or by-product material as defined in section 11e.(2) of the Atomic Energy Act (uranium or thorium tailings and waste).

[17] Measurement of Specific Radionuclides - More commonly known as direct sample or container sample using MCA data for packaged waste characterization and classification.

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

[19] Prequalification Program - The testing program implemented to demonstrate that the proposed method of wet waste processing will result in a waste form acceptable to the land disposal facility and the NRC.

[20] Processing - Changing, modifying, and/or packaging radioactive waste into a form that is acceptable to a disposal facility.

[21] Quality Assurance/Quality Control - As used in this document, "quality assurance" comprises all those planned and systematic actions necessary to provide adequate confidence that a structure, system, or component will perform satisfactorily in service.

Quality assurance includes quality control, which comprises those quality assurance actions related to the physical characteristics of a material structure, component, or system to predetermined requirements.

NUCLEAR QUALITY RELATED EN-RW-105 REV. 4 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 8 OF 21 PROCESS CONTROL PROGRAM 3.0, continued

[22] Reportable Quantity Radionuclides (RQ) - Any radionuclide listed in column (1) of Table 2 of 49CFR Part 172.101 which is present in quantities as listed in column (3) of Table 2 of 49CFR Part 172.101.

[23] Sampling Plan - A program to ensure that representative samples from the feed waste and the final waste form are obtained and tested for conformance with parameters stated in the PCP and waste form acceptance criteria.

[24] Scaling Factor - A dimensionless number which relates the concentration of an easy to measure radionuclide (gamma emitter) to one which is difficult to measure (beta and/or alpha emitters).

[25] Significant Quantity - For purposes of waste classification all the following radionuclide values SHALL be considered significant and must be reported on the disposal manifest.

Any value (real or LLD) for radionuclides listed in Appendix G to 10CFR20 (H-3, C-14, I-129, Tc-99).

Greater than or equal to 1 percent of the concentration limits as listed in 10CFR Part 61.55 Table 1.

Greater than or equal to 1 percent of the Class A concentration limits listed in 10CFR Part 61.55 Table 2.

Greater than or equal to 1 percent of the total activity.

Greater than or equal to 1 percent of the Reportable Quantity limits listed on 49CFR Part 172.101 Table 2.

[26] Solidification - The conversion of wet waste into a free-standing monolith by the addition of an agent so that the waste meets the stability and free-standing liquid requirements of the disposal site.

[27] Special Radionuclides - The RADMAN computer code term for radionuclides listed in Appendix G to 10CFR20 (i.e., H-3, C-14, I-129 & Tc-99)

[28] Stability - Structural stability per 10CFR61.2, Waste Form Technical Position, and Waste Form Technical Position Revision 1. This can be provided by the waste form, or by placing the waste in a disposal container or structure that provides stability after disposal. Stability requires that the waste form maintain its structural integrity under the expected disposal conditions.

NUCLEAR QUALITY RELATED EN-RW-105 REV. 4 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 9 OF 21 PROCESS CONTROL PROGRAM 3.0, continued

[29] Training - A systematic program that ensures a person has knowledge of hazardous materials and hazardous materials regulations.

[30] Type A Package - Is the packaging together with its radioactive contents limited to A1 or A2 as appropriate that meets the requirements of 49CFR Part 173.410 and Part 173.412, and is designed to retain the integrity of containment and shielding under normal conditions of transport as demonstrated by the tests set forth in 49CFR Part 173.465 or Part 173.466 as appropriate.

[31] Type B Package - Is the packaging together with its radioactive contents that is designed to retain the integrity of containment and shielding when subjected to the normal conditions of transport and hypothetical accident test conditions set forth in 10CFR Part 71.

[32] Volume Reduction - any process that reduces the volume of waste. This includes but is not limited to, compaction and incineration.

[33] Waste Container - A vessel of any shape, size, and composition used to contain the waste media.

[34] Waste Form - Waste in a waste container acceptable for disposal at a licensed disposal facility.

[35] Waste Stream - A Plant specific and constant source of waste with a distinct radionuclide content and distribution.

[36] Waste Type - A single packaging configuration and waste form tied to a specific waste stream.

4.0 RESPONSIBILITIES

[1] The Vice President Operations Support (VPOS) is responsible for the implementation of this procedure.

[2] Each site Senior Nuclear Executive (SNE) is responsible for ensuring that necessary site staff implements this procedure.

[3] The Low Level RadWaste (LLRW) Focus Group is responsible for evaluating and recommending changes and revisions to this procedure.

NUCLEAR QUALITY RELATED EN-RW-105 REV. 4 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 10 OF 21 PROCESS CONTROL PROGRAM 4.0, continued

[4] Each site RP Department - Radwaste Supervisor / Specialist (title may vary at the sites respectively) has the overall responsibility for implementing the PCP and is responsible for processing and transportation is tasked with the day-to-day responsibilities for the following:

Implementing the requirements of this document.

Ensuring that radioactive waste is characterized and classified in accordance with 10CFR Part 61.55 and Part 61.56.

Ensuring that radioactive waste is characterized and classified in accordance with volume reduction facility and disposal site licenses and other requirements.

Designating other approved procedures (if required) to be implemented in the packaging of any specific batch of waste.

Providing a designated regulatory point of contact between the Plant and the NRC, volume reduction facility or disposal site.

Maintaining records of on-site and off-site waste stream sample analysis and Plant evaluations.

Suspending shipments of defectively processed or defectively packaged radioactive wastes from the site when the provisions of this process control program are not satisfied.

5.0 DETAILS An isotopic analysis SHALL be performed on every batch for each waste stream so that the waste can be classified in accordance with 10CFR61. The isotopic and curie content of each shipping container SHALL be determined in accordance with 49CFR packaging requirements. The total activity in the container may be determined by either isotopic analysis or by dose-rate-to-curie conversion.

5.1. Precautions and Limitations

[1] Precautions (a) Radioactive materials SHALL be handled in accordance with applicable radiation protection procedures.

(b) All radioactive waste processed on-site OR off-site by vendors must be processed or packaged to meet the minimum requirements listed in 10CFR Part 61.56 (a) (1) through (8).

NUCLEAR QUALITY RELATED EN-RW-105 REV. 4 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 11 OF 21 PROCESS CONTROL PROGRAM 5.1[1], continued (c) If the provisions of the Process Control Program are not satisfied, suspend shipment of the defectively processed or defectively packaged waste from the site. Shipment may be accomplished when the waste is processed / packaged in accordance with the Process Control Program.

(d) The generation of combustible gases is dependent on the waste form, radioactive concentration and accumulated dose in the waste. Changes to organic inputs (e.g. oil) to waste stream may change biogas generation rates.

[2] Limitations (a) Only qualified personnel will characterize OR package radioactive waste OR radioactive materials for transportation or disposal.

(b) All site personnel that have any involvement with radioactive waste management computer software SHALL be familiar with its functions, operation and maintenance.

5.2. Waste Management Practices

[1] Waste processing methods include the following:

(a) Present and planned practice is NOT to solidify or encapsulate any waste streams.

(b) Waste being shipped directly for burial in a HIC (High Integrity Container) is dewatered to less than 1 percent by volume prior to shipment.

(c) Waste being shipped directly for burial in a container other than a HIC is dewatered to less than 0.5 percent by volume prior to shipment.

(d) IF solidification is required in the future, THEN at least one representative test specimen from at least every 10th batch of each type of radioactive waste will be checked to verify solidification.

(1) IF any specimen fails to verify solidification, THEN the solidification of the batch under test SHALL be suspended until such time as additional test specimens can be obtained, alternative solidification parameters can be determined, and a subsequent test verifies solidification. If alternative parameters are determined, the subsequent tests shall be verified using the alternative parameters determined.

NUCLEAR QUALITY RELATED EN-RW-105 REV. 4 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 12 OF 21 PROCESS CONTROL PROGRAM 5.2[1](d), continued (2) IF the initial test specimen from a batch of waste fails to verify solidification, THEN provide for the collection and testing of representative test specimens from each consecutive batch of the same type of waste until at least 3 consecutive initial test specimens demonstrates solidification. The process SHALL be modified as required to assure solidification of subsequent batches of waste.

[2] Operation and maintenance of dewatering systems and equipment include the following:

(a) Present and planned practice is to utilize plant personnel supplemented by vendor personnel or contracted vendor personnel, to operate AND maintain dewatering systems and equipment (as needed to meet disposal site requirements).

(b) All disposal liners are manufactured by and purchased from QA-approved vendors.

[3] ALARA considerations are addressed in all phases of the processes involving handling, packaging AND transfer of any type OR form of radioactive waste (dewatered or dry).

Resin, charcoal media, spent filter cartridges AND sludges are typically processed within shields. Sluiceable demineralizers are shielded when in service. Radiation exposure and other health physics requirements are controlled by the issuance of a Radiation Work Permit (RWP) for each task.

5.3. Waste Stream Sampling Methods and Frequency

[1] The following general requirements apply to Plant waste stream sampling:

(a) Treat each waste stream separately for classification purposes.

(b) Ensure samples are representative of or can be correlated to the final waste form.

(c) Determine the density for each new waste stream initially or as needed (not applicable for DAW and filters).

(d) Perform an in-house analysis for gamma-emitting radionuclides for each sample sent to an independent laboratory.

(e) Periodically perform in-house analysis for gamma emitting radionuclides for comparison to the current data base values for gamma emitters. (The current database is usually based on the most recent independent laboratory results.)

(f) Resolve any discrepancies between in-house results AND the independent laboratory results for the same or replicate sample as soon as possible.

(g) Maintain records of on-site and off-site waste stream sample analysis and evaluations.

NUCLEAR QUALITY RELATED EN-RW-105 REV. 4 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 13 OF 21 PROCESS CONTROL PROGRAM 5.3, continued

[2] When required, waste stream samples should be analyzed, re-evaluated and if necessary, shipped to a vendor laboratory for additional analysis. The same is true when there is a reason to believe that an equipment or process change has significantly altered the previously determined scaling factors by a factor of 10.

Specific examples include but are not limited to:

Changes in oxidation reduction methods such as zinc, injection, hydrogen water chemistry, Changes in purification methods including media specialization, media distribution, ion/cation ratios, Changes in fuel performance criteria including fuel leaks Other changes in reactor coolant chemistry.

Sustained, unexplained, changes in the routinely monitored Beta/Alpha ratios, as determined by Radiation Protection, When there is an extended reactor shutdown (> 90 days).

When there are changes to liquid waste processing, such as bypassing filters, utilizing filters or a change in ion exchange media.

When there are changes to the waste stream that could change the biogas generation rate.

[3] The following requirements apply to infrequent or abnormal waste types:

(a) Infrequent OR abnormal waste types that may be generated must be evaluated on a case-by-case basis.

(b) The RP Department Supervisor / Specialist responsible for processing AND shipping will determine if the waste can be correlated to an existing waste stream.

(c) IF the radioactive waste cannot be correlated to an existing waste stream, THEN the RP Department Supervisor / Specialist responsible for processing and shipping SHALL determine specific off-site sampling and analysis requirements necessary to properly classify the material.

[4] Specific sampling methods and data evaluation criteria are detailed in EN-RW-104 for specific waste streams.

NUCLEAR QUALITY RELATED EN-RW-105 REV. 4 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 14 OF 21 PROCESS CONTROL PROGRAM 5.4. Waste Classification

[1] General requirements for scaling factors include the following:

(a) The Plant has established an inferential measurement program whereby concentrations of radionuclides which cannot be readily measured are estimated through ratio-ing with radionuclides which can be readily measured.

(b) Scaling factor relationships are developed on a waste stream-specific basis.

These relationships are periodically revised to reflect current independent lab data from direct measurement of samples. The scaling factor relationships currently used by the sites are as follows:

Hard to detect ACTIVATION product radionuclides and C-14 are estimated by using scaling factors with measured Co-60 activities.

Hard to detect FISSION product radionuclides and H-3, Tc-99 and I-129 are estimated by using scaling factors with measured Cs-137 activities.

Hard to detect TRANSURANIC radionuclides are estimated by using scaling factors with measured Ce-144 activities. Where Ce-144 cannot be readily measured, transuranics are estimated by using scaling factors with measured Cs-137 activities. Second order scaling of transuranics is acceptable when Cs-137 and Ce-144 are not readily measurable.

[2] General requirements for the determination of total activity and radionuclide concentrations include the following:

(a) The activity for the waste streams is estimated by using either Gross Radioactivity Measurement OR Direct Measurement of Radionuclides. Current specific practices are as follows:

DAW - Gross radioactivity measurement in conjunction with the RADMAN computer codes, other approved computer codes or hand calculation.

Filters - Gross radioactivity measurement in conjunction with the FILTRK computer code, other approved computer codes or hand calculation.

All Other Waste Streams - Direct measurement of radionuclides in conjunction with the RADMAN computer codes, other approved computer codes or hand calculation.

(b) Determination of the NRC waste classification is performed by comparing the measured or calculated concentrations of significant radionuclides in the final waste form to those listed in 10CFR Part 61.55.

NUCLEAR QUALITY RELATED EN-RW-105 REV. 4 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 15 OF 21 PROCESS CONTROL PROGRAM 5.5. Quality Control

[1] The RADMAN computer code provides a mechanism to assist the Plant in conducting a quality control program in accordance with the waste classification requirements listed in 10CFR Part 61.55. All waste stream sample data changes are written to a computer data file for future review and reference.

[2] Audits and Management Review includes the following:

(a) Appendix G to 10CFR20 requires conduct of a QC program which must include management review of audits.

(b) Management audits of the Plant Sampling and Classification Program SHALL be periodically performed to verify the adequacy of maintenance sampling and analysis.

(c) Audits and assessments are performed and documented by any of the following:

Radiation Protection Department Quality Assurance Department Qualified Vendors (d) Certain elements of the Entergy Quality Assurance Program Manual are applied to the Process Control Program. [QAPM, Section A.1.c]

5.6. Dewatering Operations

[1] Processing requirements during dewatering operations include the following:

(a) All dewatering operations are performed per approved Plant or vendor operating procedures and instructions.

(b) Dewatering limitations and capabilities are verified by vendor Topical Reports or Operating and Testing Procedures.

[2] Dewatered resin activity limitations include the following:

(a) Dewatered resins will not be shipped off-site that have activities which will produce greater than 1.0E+8 rads total accumulated dose over 300 years. This is usually verified by comparing the container specific activity at the time of shipment to the following concentration limits for radionuclides with a half-life greater than five years:

10 Ci (0.37 TBq) per cubic foot.

350 uCi (12.95 MBq) per cubic centimeter

NUCLEAR QUALITY RELATED EN-RW-105 REV. 4 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 16 OF 21 PROCESS CONTROL PROGRAM 5.7. Waste Packaging Waste in final form will be packaged in accordance with Title 10 and Title 49 of the Code of federal regulations and in accordance with current burial site criteria as is detailed in EN-RW-102.

5.8. Administrative Controls

[1] Information on solid radioactive waste shipped off-site is reported annually to the Nuclear Regulatory Commission in the Annual Radioactive Effluent Release Report as specified by the Offsite Dose Calculation Manual (ODCM) or Technical Specification.

[ANO1 Technical Specifications - 5.6.3] [ANO2 Technical Specifications - 6.6.3]

[WF3 Technical Specifications - 6.9.18] [GGNS ODCM - 5.6.3.c] [JAF Technical Specifications - 5.6.3] [PLP ODCM, Appendix A - IV. A].

[2] All changes to the PCP SHALL be documented. All records of reviews performed SHALL be retained as required by the Quality Assurance Program. The documentation of the changes SHALL [GGNS UFSAR, Chapter 16B.1 / TRM -

7.6.3.8 paragraph 2]:

(a) Contain sufficient information to support the change with appropriate analyses or evaluations justifying the change.

(b) Include a determination that the change will maintain the overall conformance of the solidified waste product (if applicable) to existing requirements of Federal, State or other applicable regulations.

[3] All changes in the Process Control Program and supporting documentation are included in each sites next Annual Radiological Effluent Release Report to the Nuclear Regulatory Commission. [ANO ODCM - L3.2.1.C] [VTY TRM 6.12]

[4] The changes to EN-RW-105 SHALL become effective upon review and acceptance by the sites General Plant Manager (equivalent title at Palisades is Plant Superintendent) except as listed below:

(a) For Grand Gulf Nuclear Station, the changes to RW-105 SHALL be accomplished as specified in Grand Gulf Nuclear Station Technical Requirements Manual (TRM) Section 7.6.3.8. The changes SHALL become effective upon review and acceptance by the On-site Safety Review Committee (OSRC) and the approval of the GGNS Plant General Manager. [GGNS UFSAR, Chapter 16B.1 / TRM - 7.6.3.8 paragraph 2]

NUCLEAR QUALITY RELATED EN-RW-105 REV. 4 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 17 OF 21 PROCESS CONTROL PROGRAM 5.8[4], continued (b) For River Bend Nuclear Station, the procedure approval along with changes to RW-105 SHALL be accomplished per the River Bend Nuclear Station Technical Requirements, Section 5.5.14.1. The changes SHALL become effective upon review and acceptance by approval from the River Bend Nuclear Station Plant Manager or Radiation Protection Manager. [RBS Technical Requirements -

5.5.14.1, 5.5.14.2 & 5.8.2]

(c) For Waterford 3, the procedure approval along with changes to RW-105 SHALL be accomplished per Waterford 3 Technical Specifications 6.13.2. The changes SHALL become effective upon review and acceptance by the Waterford 3 General Plant Manager. [WF3 Technical Specifications - 6.13.2.b]

(d) For James A. FitzPatrick Nuclear Station, the procedure approval along with changes to EN-RW-105 SHALL be accomplished per the James A. FitzPatrick Station Technical Specifications, Section 5.6.3. The changes SHALL become effective upon review and acceptance through approval from the James A.

FitzPatrick Nuclear Station On-Site Safety Review Committee. [JAF UFSAR, Chapter 11.3.5]

(e) For Vermont Yankee, Changes to the Process Control Program SHALL become effective after review and acceptance by the (OSRC) On-Site Safety Review Committee and the Site VP.

(f) For IPEC, Changes to the Process Control Program SHALL become effective after final review and acceptance by the On-Site Safety Review Committee (OSRC).

5.9. Vendor Requirements

[1] Vendors performing radwaste services under 10CFR61 and 10CFR71 requirements will be on the Entergy Qualified Suppliers List (QSL). [QAPM, Section A.1.c]

[2] Vendors performing radwaste services on-site are to comply with the following:

(a) Dewatering and solidification services SHALL have a NRC-approved Topical Report or other form of certification documenting NRC approval of the processes and associated equipment/containers.

(b) All vendor procedures utilized for performing on-site radwaste processing services (to assure compliance with 10 CFR Parts 20, 61 and 71, State Regulations, burial ground requirements, and other requirements governing the disposal of solid radioactive waste) will be reviewed per the requirements of EN-LI-100, technically by the applicable sites Radiation Protection organization and only be accepted per the approvals specified in Section 5.8 [4].

NUCLEAR QUALITY RELATED EN-RW-105 REV. 4 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 18 OF 21 PROCESS CONTROL PROGRAM 5.9[2], continued (c) All changes to vendor procedures for ongoing on-site radwaste services will be reviewed technically by the sites Radiation Protection organization and screened per the requirements of EN-LI-100. Significant procedural changes will require the approvals specified in Section 5.8 [4]. During screening, the level of significance for procedural changes on equipment and process parameters may warrant the full 10CFR50.59 documentation and approval process.

(d) Plant management SHALL review vendor(s) topical reports and test procedures per applicable requirements in Section 5.8.

NOTE The PCP does not have to include the vendor's Topical Report if it has NRC approval, or has been previously submitted to the NRC.

(e) Plant management review will assure that the vendor's operations and requirements are compatible with the responsibilities and operation of the Plant.

(f) Training requirements and records listed in Section 5.10 also apply to contracted vendors.

5.10. Miscellaneous

[1] Special tools and equipment (a) Frequency of Use and Descriptions Required tools and equipment will vary depending on the specific process and waste container that is used. The various tools and equipment which may be required are detailed in specific procedures developed to govern activities described in this document.

[2] Pre-requisites (a) Maintenance of Regulatory Material Ensure that a current set of DOT, NRC, EPA and applicable State regulations, vendor processing facility and disposal site regulations and requirements are maintained at the site and are readily available for reference. The use of web based regulations is acceptable.

NUCLEAR QUALITY RELATED EN-RW-105 REV. 4 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 19 OF 21 PROCESS CONTROL PROGRAM 5.10[2], continued (b) Representative Radionuclide Sample Data Ensure that representative radionuclide sample data is on file for each active waste stream. Unless operation conditions or changes in processing methods require increased sample frequency, data is considered to be current if it meets the requirements of EN-RW-104.

(c) Initial and Cyclic Training A training program SHALL be developed, implemented and maintained for all personnel involved in processing, packaging, handling and transportation of radioactive waste to ensure radwaste operations are performed within the requirements of NRC Information Bulletin 79-19 and 49CFR Part 172.700 through Part 172.704.

Training requirements and documentation also apply to contracted on-site vendors.

NOTE Cyclic training is defined as within three years for DOT, and two years for IATA (d) Specific employee training is required for each person who performs the following job functions [172.702(b)].

Classifies hazardous materials.

Packages hazardous materials.

Fills, loads and/or closes packages.

Marks and labels packages containing hazardous materials.

Prepares shipping papers for hazardous materials.

Offers or accepts hazardous materials for transportation.

Handles hazardous materials.

Marks or placards transport vehicles.

Operates transport vehicles.

Works in a transportation facility and performs functions in proximity to hazardous materials which are to be transported.

Inspects or tests packages.

NUCLEAR QUALITY RELATED EN-RW-105 REV. 4 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 20 OF 21 PROCESS CONTROL PROGRAM 5.10[2] continued (e) Cyclic training is defined as within three years for DOT & within two years for IATA.

Copies of training records are required for as long as a person is employed and 90 days thereafter. The records should include, as a minimum, the following:

Trainee's name and signature Training dates Training material or source reference Trainer's information 6.0 INTERFACES

[1] EN-LI-100, Process Applicability Determination

[2] EN-RW-104, Scaling Factors

[3] EN-QV-104, Entergy Quality Assurance Program Manual Control 7.0 RECORDS

[1] Documentation of pertinent data required to classify waste and verify solidification will be maintained on each batch of processed waste as required by approved procedures.

[2] Documentation will also be maintained to ensure that containers, shipping casks, and methods of packaging wastes meet applicable Federal regulations and disposal site criteria. The records of reviews performed and documents associated with these reviews will be maintained as QA records.

NUCLEAR QUALITY RELATED EN-RW-105 REV. 4 MANAGEMENT MANUAL INFORMATIONAL USE PAGE 21 OF 21 PROCESS CONTROL PROGRAM 8.0 SITE SPECIFIC COMMITMENTS Document Document NMM Procedure Site Applicability Section Section ANO ODCM L3.2.1.C 5.8 [3] ANO ANO1 Technical Specifications 5.6.3 5.8 [1] ANO ANO2 Technical Specifications 6.6.3 5.8 [1] ANO RBS Technical Requirements 5.5.14

• RBS RBS Technical Requirements 5.5.14.1 5.8 [3] RBS 5.8 [4] (b)

RBS Technical Requirements 5.5.14.2 5.8 [4] (b) RBS RBS Technical Requirements 5.8.2 5.8 [4] (b) RBS WF3 Technical Specifications 1.22

• WF3 WF3 Technical Specifications 6.9.18 5.8 [1] WF3 WF3 Technical Specifications 6.13.2.b 5.8 [4] (c) WF3 JAF ODCM 6.2.1 5.8 [1] JAF JAF Technical Specifications 5.6.3 5.8 [1], 5.8 [4] JAF JAF FSAR Chapters 7 and 5.8 [4] JAF 11 11759 - NRC IN 79-19 All

• WF3 GGNS UFSAR, Chapter 16B.1 / 7.6.3.8 1.0 GGNS TRM paragraph 1 GGNS ODCM 5.6.3.c 5.8 [1] GGNS GGNS FSAR 11.4.5.S2 5.9 [2](a) GGNS GGNS FSAR 11.4.2.3AS7 5.9 [2](a) GGNS IPN-99-079 All

• IPEC Appendix B Technical Section 4.5,

• Specifications RECS ODCM IPEC Part 1 PLP Technical Specifications 5.5.15 5.8 [4] PLP PLP ODCM Appendix A - 5.8 [1] PLP IV. A NRC Letter 1.98.091 All

• PNPS NRC Letter 1.88.078 All

• PNPS VY Technical Specifications 6.4.H

• VY VY ODCM 10.1 5.8 [1] VY VY TRM 6.12 5.8 (3) VY QAPM Section A.1.c

• All

• Covered by directive as a whole or by various paragraphs of the directive.

9.0 ATTACHMENTS None