Text

Annual Radioactive Effluent Release Report for 2019
	ML20118C140
Person / Time
Site:	Arkansas Nuclear
Issue date:	04/27/2020
From:	Keele R; Entergy Operations
To:	; Document Control Desk, Office of Nuclear Material Safety and Safeguards, Office of Nuclear Reactor Regulation
References
	OCAN042001
	Download: ML20118C140 (137)
	v • d • e

Entergy Operations, Inc.

1448 S.R. 333 Russellville, AR 72802 Tel 479-858-7826 Riley D. Keele, Jr.

Manager, Regulatory Assurance Arkansas Nuclear One 0CAN042001 April 27, 2020 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555

Subject:

Annual Radioactive Effluent Release Report for 2019 Arkansas Nuclear One, Units 1 and 2 NRC Docket Nos. 50-313, 50-368, and 72-13 Renewed Facility Operating License Nos. DPR-51 and NPF-6 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.

The information which fulfills this reporting requirement for ANO-1 and ANO-2 for the 2019 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 limits. The data reveals that radioactive effluents had an overall minimal dose contribution to the surrounding environment.

This letter contains no new commitments.

If there are any questions or if additional information is needed, please contact me.

Respectfully, ORIGINAL SIGNED BY RILEY D. KEELE, JR.

RDK/rwc

Enclosure:

Annual Radioactive Effluent Release Report Enclosure

Attachment:

Offsite Dose Calculation Manual

0CAN042001 Page 2 of 2 cc: NRC Region IV Regional Administrator NRC Senior Resident Inspector - Arkansas Nuclear One NRC Project Manager - Arkansas Nuclear One

Enclosure 0CAN042001 Annual Radioactive Effluent Release Report

Page 1 of 29 Plant: Arkansas Nuclear One YEAR: 2019 Document Number: 0CAN042001 Annual Radioactive Effluent Release Report

Plant: Arkansas Nuclear One Year: 2019 Page 2 of 29 Annual Radioactive Effluent Release Report TABLE OF CONTENTS

1.0 INTRODUCTION

................................................................................................................ 3 2.0 SUPPLEMENTAL INFORMATION ..................................................................................... 3 3.0 GASEOUS EFFLUENTS .................................................................................................. 13 4.0 LIQUID EFFLUENTS ........................................................................................................ 17 5.0 SOLID WASTE

SUMMARY

.............................................................................................. 21 6.0 RADIOLOGICAL IMPACT TO MAN ................................................................................. 25 7.0 METEORLOGICAL DATA ................................................................................................ 28 ATTACHMENT - OFFSITE DOES CALCULATION MANUAL

Plant: Arkansas Nuclear One Year: 2019 Page 3 of 29 Annual Radioactive Effluent Release Report

1.0 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. ANO-1 capacity factor for 2019 was 0.88 and ANO-2 was 0.83.

2.0 SUPPLEMENTAL INFORMATION 2.1 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 ANO is considerably lower than 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:

1. Fission and activation gases:

a. Noble gases dose rate due to radioactive materials released in gaseous effluents from the areas at and beyond the site boundary shall be limited to the following:

Less than or equal to 500 mrem/year to the total body Less than or equal to 3000 mrem/year to the skin

b. Noble gas air dose due to noble gases released in gaseous effluents to areas at and beyond the site boundary shall be limited to the following:

1) Quarterly Less than or equal to 5 mrads gamma Less than or equal to 10 mrads beta

2) Yearly Less than or equal to 10 mrads gamma Less than or equal to 20 mrads beta

2. Iodine, tritium, and all radionuclides in particulate form with half-lives greater than 8 days.

a. The dose rate for Iodine-131, Iodine-133, tritium, and all radionuclides in particulate form with half-lives greater than 8 days in gaseous effluents released to areas at and beyond the site boundary shall be limited to the following:

Less than or equal to 1500 mrem/yr to any organ

Plant: Arkansas Nuclear One Year: 2019 Page 4 of 29 Annual Radioactive Effluent Release Report

b. The dose to a MEMBER OF THE PUBLIC from Iodine-131, Iodine-133, tritium, and all radionuclides in particulate form with half-lives greater than 8 days in gaseous effluents released to areas at and beyond the site boundary shall be limited to the following:

1) Quarterly Less than or equal to 7.5 mrem to any organ

2) Yearly Less than or equal to 15 mrem to any organ

3. Liquid Effluents Dose

a. The dose or dose commitment to a MEMBER OF THE PUBLIC from radioactive materials in liquid effluents released to unrestricted areas shall be limited to the following:

1) Quarterly Less than or equal to 1.5 mrem total body Less than or equal to 5 mrem critical organ

2) Yearly Less than or equal to 3 mrem total body Less than or equal to 10 mrem critical organ

4. Total Dose (40CFR190)

a. The annual (calendar year) dose or dose commitment to any MEMBER OF THE PUBLIC due to releases of radioactivity and to radiation from uranium fuel cycle sources shall be limited to the following:

Less than or equal to 25 mrem, Total Body or any Organ except Thyroid Less than or equal to 75 mrem, Thyroid 2.2 Maximum Permissible Concentrations

1. Fission & Activation Gases, Iodines, and Particulates with Half Lives > Eight (8) Days For gaseous effluents, maximum permissible concentrations are not directly used in release rate calculations since the applicable limits are expressed in terms of dose rate at the site boundary.

Plant: Arkansas Nuclear One Year: 2019 Page 5 of 29 Annual Radioactive Effluent Release Report

2. Liquid Effluents The concentration of radioactive material released shall be limited to the concentration specified in 10 CFR 20, Appendix B, Table 2, Column 2, for radionuclides other than dissolved or entrained noble gases. For dissolved or entrained noble gases, the total concentration released shall be limited to 2.0E-4 microcuries/ml.

2.3 Measurements & Approximations of Total Radioactivity

1. Gaseous Effluents

a. Fission & activation gases Gas samples are collected weekly and are counted on a high purity germanium detector (HPGe) for principal gamma emitters. The containment vent, auxiliary building, piping penetration room, and spent fuel pool areas for ANO-1 and ANO-2.

Additionally, the Aux. Bldg. Extension and low level radwaste storage are sampled.

All effluent waste streams are only sampling when the ventilation is active.

Technical Specification release points are continuously monitored and the average release flow rates for each release point are used to calculate the total activity released during a given time period.

b. Iodines Iodine is continuously collected on F&J TE2C cartridge filter via an isokinetic sampling assembly from each release point. Filters are exchanged once per week and then analyzed on an HPGe system. The flow rates for each release point are averaged over the duration of the sampling period and these results, along with specific isotopic concentrations, are then used to determine the total activity released during the time period in question.

c. Particulates (half-lives > 8 days)

Particulates are continuously collected on a filter paper via an isokinetic sampling assembly on each release point. Filters are exchanged once per week and then analyzed on an HPGe system. The flow rates for each release point are averaged over the duration of the sampling period and these results, along with specific isotopic concentrations, are then used to determine the total activity released during the time period in question.

d. Tritium Tritium is collected by passing a known volume of the sample stream through a calcium chloride desiccant on a weekly basis from each release point. The collected samples are distilled and analyzed by liquid scintillation. The tritium released was calculated for each release point from the measured tritium concentration, the volume of the sample, the tritium collection efficiency, and the respective stack exhaust flow rates.

Plant: Arkansas Nuclear One Year: 2019 Page 6 of 29 Annual Radioactive Effluent Release Report

e. Carbon-14

1) ANO-1 Carbon-14 release values were estimated using the methodology included in the EPRI Technical Report 1021106, using the 2019 normalized Carbon-14 production rate of 3.4 Ci/GWtyr, a gaseous release fraction of 98 percent, a Carbon-14 carbon dioxide fraction of 30 percent, a reactor power rating 2568 MWt, and equivalent full power operation of 320 days.

2) ANO-2 Carbon-14 release values were estimated using the methodology included in the EPRI Technical Report 1021106, using the 2019 normalized Carbon-14 production rate of 3.9 Ci/GWtyr, a gaseous release fraction of 98 percent, a Carbon-14 carbon dioxide fraction of 30 percent, a reactor power rating of 3026 MWt, and equivalent full power operation of 302 days.

2. Liquid Effluents

a. Batch Releases Each tank of liquid radwaste is sampled and analyzed for principal gamma emitters prior to release. Each sample tank is recirculated for a sufficient amount of time prior to sampling ensuring that a representative sample is obtained. Samples are then analyzed on an HPGe system and liquid release permits are generated based upon the values obtained from the isotopic analysis and the most recent values for H3, gross alpha, Fe-55, Sr-89, and Sr-90. An aliquot based on release volume is saved and added to composite containers. The concentrations of composited isotopes and the volumes of the releases associated with these composites establish the proportional relationships that are then utilized for calculating the total activity released for these isotopes.

b. Continuous Releases Samples are taken on ANOs two continuous liquid release streams (turbine building sumps) once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Samples are then analyzed on an HPGe system and liquid release permits are generated based upon the values obtained from the isotopic analysis and the most recent values for H3, gross alpha, Fe-55, Sr-89, and Sr-90. An aliquot based on release volume is saved and added to composite containers. The concentrations of composited isotopes and the volumes of the releases associated with these composites establish the proportional relationships that are then utilized for calculating the total activity released for these isotopes.

Plant: Arkansas Nuclear One Year: 2019 Page 7 of 29 Annual Radioactive Effluent Release Report

3. Estimated Total Error Present

a. Estimates of measurement and analytical error for gaseous and liquid effluents are calculated as follows:

Where:

ET = total percent error E1 ... En = percent error due to calibration standards, Laboratory analysis, instruments, sample flow, etc.

2.4 Batch Releases 2.4.1 Liquid

1. Unit 1

a. Number of batch releases: 86

b. Total time period for a batch release: 2.07E+04 minutes

c. Maximum time period for a batch release: 6.70E+02 minutes

d. Average time period for a batch release: 2.41E+02 minutes

e. Minimum time period for a batch release: 1.60E+01 minutes

2. Unit 2

a. Number of batch releases: 22

b. Total time period for a batch release: 8.60E+03

c. Maximum time period for a batch release: 6.59E+02 minutes

d. Average time period for a batch release: 3.91E+02 minutes

e. Minimum time period for a batch release: 2.00E+02 minutes

Plant: Arkansas Nuclear One Year: 2019 Page 8 of 29 Annual Radioactive Effluent Release Report 2.4.2 Gaseous

1. Unit 1

a. Number of batch releases: 98

b. Total time period for a batch release: 8.27E+05 minutes

c. Maximum time period for a batch release: 4.32E+04 minutes

d. Average time period for a batch release: 8.44E+03 minutes

e. Minimum time period for a batch release: 6.60E+01 minutes

2. Unit 2

a. Number of batch releases: 124

b. Total time period for a batch release: 1.01E+06 minutes

c. Maximum time period for a batch release: 6.13E+04 minutes

d. Average time period for a batch release: 8.13E+03 minutes

e. Minimum time period for a batch release: 1.00E+00 minutes 2.5 Continuous Releases 2.5.1 Liquid

1. Unit 1

a. Number of continuous releases: 5

b. Total time period for a continuous release: 2.54E+04 minutes

c. Maximum time period for a continuous release: 1.00E+04 minutes

d. Average time period for a continuous release: 5.08E+03 minutes

e. Minimum time period for a continuous release: 1.44E+03 minutes

2. Unit 2

a. Number of continuous releases: 12

b. Total time period for a continuous release: 1.11E+05 minutes

c. Maximum time period for a continuous release: 1.02E+04 minutes

d. Average time period for a continuous release: 9.23E+03 minutes

e. Minimum time period for a continuous release: 4.67E+03 minutes

Plant: Arkansas Nuclear One Year: 2019 Page 9 of 29 Annual Radioactive Effluent Release Report 2.5.2 Gaseous

1. Unit 1 There were zero continuous releases.

2. Unit 2 There were zero continuous releases.

2.6 Abnormal Releases 2.6.1 ANO-1

1. Liquid Number of releases: 0

2. Gaseous Number of releases: 0 2.6.2 ANO-2

1. Liquid Number of releases: 0

2. Gaseous Number of releases: 0 2.7 Non-routine, Planned Discharges 2.7.1 ANO-1 Non-routine Discharges There was one discharge point that is considered non routine because it is not a significant release point for the site (< 1 percent total dose) for ANO-1. The Intermediate Cooling Water (ICW) surge tank vent is routinely monitored for effluents because the system is a radioactively contaminated system that cools reactor related components and is vented to an uncontrolled area. The release is documented in a release permit.

Plant: Arkansas Nuclear One Year: 2019 Page 10 of 29 Annual Radioactive Effluent Release Report 2.7.2 ANO-2 Non-routine Discharges There were two discharge points that are considered non routine because it is not a significant release point for the site (<1 percent total dose) for ANO-2. The discharge point is the steam release off the Emergency Feedwater pump 2P-7A and the containment equipment hatch.

Later in the operating cycle for ANO-2, tritium concentrations are greater than the minimum detectable activity (MDA) in the secondary water. This is due to the U-tube designed Steam Generators where the tritium from the reactor coolant migrates across the tubes. When 2P-7A is run for surveillances there is a steam release with tritiated water to the atmosphere. A release permit is generated to capture the dose released.

The equipment hatch is opened during outages to move large equipment in and out of the containment building. During the period in which the hatch is opened it is routinely monitored for radioactivity and, in the event air were to exit the equipment hatch, a release permit is generated to capture the dose released.

2.8 Radioactive Waste Treatment System Changes There were zero changes to the radioactive waste treatment systems for either liquid or gases.

2.9 Land Use Census Changes There were zero changes to receptor locations or routes of exposure as a result of the 2019 land use census. ANO performs land use census once per every two years.

2.10 Effluent Monitor Instrument Inoperability

1. There was was one ventilation stack radiation monitor that was out-of-service greater than 30 days (ODCM L2.2.1 Action G) for ANO due to a sample pump failure which is a long lead time part. During the out-of-service time there was alternate sampling setup to ensure effluents released from the site were quantified and the appropriate dose calculated. Additionally, there is a second ventilation stack radiation monitor RX-9830 (Unit 1 Spent Fuel Area Ventilation SPING) that is monitoring the same area during the time 2RX-9830 was out-of-service.

2. Below is a table that shows the radiation monitor equipment identification (ID),

common name, time of inoperability, and the document deficiencies were corrected under.

Table 1: Effluent Monitor Instrument Inoperability Start Date End Date Component ID Component Noun Name Condition Report Inoperable Inoperable Unit 2 Spent Fuel Pool CR-ANO-2-2019-2RX-9830 4-23-2019 8-12-2019 Area Ventilation SPING 1056

Plant: Arkansas Nuclear One Year: 2019 Page 11 of 29 Annual Radioactive Effluent Release Report 2.11 Offsite Dose Calculation Manual Changes

1. The ODCM was revised once on 9/17/2019 under Revision 029 to include the following changes:

Performed editorial fixes to the Step 3.1.1.b setpoint calculation. Changed units for gas to µCi/cc instead of µCi/ml since SPING data is provided in µCi/cc.

Changed S monitor setpoint from "cpm" to "cpm or µCi/cc". These changes resolved preexisting issues with units for the subject calculation with respect to implementation in the plant.

Editorial change removed 2RX-9840 PASS Building Ventilation from list of SPING allocations. This should have been removed from ODCM with Revision 028 where the SPING was removed from the rest of the document in accordance with engineering changes EC-71778 and EC-74229.

Editorial changed resolved a reference in L2.4.1 Action C. This Action previously referenced L2.4.1.b.4 which does not exist. Reference was appropriately changed to L2.4.1.b.3.

2. The most recent version of the ODCM is attached to this enclosure.

2.12 Process Control Program (PCP) Changes There were no changes to Entergy's PCP procedure EN-RW-105 in 2019.

2.13 NON-REMP Groundwater Monitoring Results (NEI 07-07)

1. ANO has a total of 16 Non-REMP wells as part of the NEI 07-07 Ground Water Protection Program. There are 11 wells that are sampled on an annual basis and 5 that are performed quarterly. The 5 wells that are performed quarterly have a higher potential to become contaminated due to the proximity to the plant.

2. There were a total of 31 samples obtained in 2019 for predominant gamma emitters (Mn-54, Co-58, Fe-59, Co-60, Zn-65, Nb-95, Zr-95, I-131, Cs-134, Cs-137, Ba-140, and La-140) with zero samples indicating results greater than the minimum detectable concentration.

3. There were a total of 31 samples taken in 2019 for tritium (H-3) with 3 samples indicating a positive result from Monitoring Well 17. Due to the well location in proximity to the Unit 1 containment building it is most likely tritium recapture from rain water. The sample results are below any reporting criteria or exceedance of any federal limit.

Plant: Arkansas Nuclear One Year: 2019 Page 12 of 29 Annual Radioactive Effluent Release Report Table 2: Positive Well Results Well Sample Date Analysis Result Error MDC Units MW-17 3/13/2019 14:28 H-3 506 244 335 pCi/L MW-17 9/11/2019 08:41 H-3 441 260 361 pCi/L MW-17 12/11/2019 08:57 H-3 480 270 374 pCi/L

[MDC - Minimum Detectable Concentration]

4. There were no spills that required entry into ANOs 10 CFR 50.75(g) document for future decommissioning.

2.14 LLD Levels In accordance with ODCM Appendix 1 lower limits of detection (LLDs) higher than required shall be documented in the Annual Radioactive Effluent Release Report (ARERR):

There were no instances in which ANO did not meet their required sample LLDs.

2.15 Errata/Corrections to Previous ARERRs There is no errata to be submitted with the 2019 ARERR.

Plant: Arkansas Nuclear One Year: 2019 Page 13 of 29 Annual Radioactive Effluent Release Report 3.0 GASEOUS EFFLUENTS 3.1 Gas Effluent and Waste Disposal Report 3.1.1 ANO-1 Data Table 3: Gaseous Effluents - Summation of All Releases (ANO-1)

A. Fission & Activation Est. Total Unit Quarter 1 Quarter 2 Quarter 3 Quarter 4 Gases Error %

1. Total Release Ci 0.00E+00 0.00E+00 1.03E-02 6.27E-03 24

2. Average release rate Ci/sec 0.00E+00 0.00E+00 1.30E-03 7.96E-04 for the period B. Iodine

1. Total Iodine - 131 Ci 0.00E+00 0.00E+00 0.00E+00 0.00E+00 20

2. Average release rate Ci/sec 0.00E+00 0.00E+00 0.00E+00 0.00E+00 for the period C. Particulates

1. Particulates with Ci 0.00E+00 0.00E+00 0.00E+00 0.00E+00 22 half-lives > 8 days

2. Average release rate Ci/sec 0.00E+00 0.00E+00 0.00E+00 0.00E+00 for the period D. Tritium

1. Total Release Ci 6.65E+00 4.68E+00 1.97E+00 7.74E+00 21

2. Average release rate Ci/sec 8.43E-01 5.93E-01 2.50E-01 9.82E-01 for the period E. Gross Alpha

1. Total Release Ci 0.00E+00 0.00E+00 0.00E+00 0.00E+00 31

2. Average release rate Ci/sec 0.00E+00 0.00E+00 0.00E+00 0.00E+00 for the period F. Carbon-14

1. Total Release Ci 2.11E+00 2.13E+00 2.16E+00 1.10E+00

2. Average release rate Ci/sec 2.77E-07 2.77E-07 2.77E-07 1.41E-07 for the period

% of limit is on the Radiological Impact to Man Table (reference Section 6.0)

Plant: Arkansas Nuclear One Year: 2019 Page 14 of 29 Annual Radioactive Effluent Release Report Table 4: Gaseous Effluents - Ground Level Release - Batch Mode (ANO-1)

Nuclides Released Unit Quarter 1 Quarter 2 Quarter 3 Quarter 4 Total Fission Gases Kr-85 Ci 0.00E+00 0.00E+00 1.03E-02 0.00E+00 1.03E-02 Xe-133 Ci 0.00E+00 0.00E+00 0.00E+00 6.27E-03 6.27E-03 Total for Period Ci 0.00E+00 0.00E+00 1.03E-02 6.27E-03 1.65E-02 lodines I-131 Ci 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Total for Period Ci 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Particulates Total for Period Ci 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Tritium H-3 Ci 6.65E+00 4.68E+00 1.97E+00 7.74E+00 2.10E+01 Gross Alpha Alpha Ci 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Carbon-14 C-14 Ci 2.11E+00 2.13E+00 2.16E+00 1.10E+00 7.50E+00

Plant: Arkansas Nuclear One Year: 2019 Page 15 of 29 Annual Radioactive Effluent Release Report 3.1.2 ANO-2 Data Table 5: Gaseous Effluents - Summation of All Releases (ANO-2)

A. Fission & Activation Est. Total Unit Quarter 1 Quarter 2 Quarter 3 Quarter 4 Gases Error %

1. Total Release Ci 0.00E+00 5.12E-02 0.00E+00 0.00E+00 24

2. Average release rate Ci/sec 0.00E+00 6.50E-03 0.00E+00 0.00E+00 for the period B. Iodine

1. Total Iodine - 131 Ci 0.00E+00 0.00E+00 0.00E+00 0.00E+00 20

2. Average release rate Ci/sec 0.00E+00 0.00E+00 0.00E+00 0.00E+00 for the period C. Particulates

1. Particulates with Ci 0.00E+00 0.00E+00 0.00E+00 0.00E+00 22 half-lives > 8 days

2. Average release rate Ci/sec 0.00E+00 0.00E+00 0.00E+00 0.00E+00 for the period D. Tritium

1. Total Release Ci 3.95E+00 6.90E+00 8.47E+00 5.42E+00 21

2. Average release rate Ci/sec 5.01E-01 8.75E-01 1.07E+00 6.87E-01 for the period E. Gross Alpha

1. Total Release Ci 0.00E+00 0.00E+00 0.00E+00 0.00E+00 31

2. Average release rate Ci/sec 0.00E+00 0.00E+00 0.00E+00 0.00E+00 for the period F. Carbon-14

1. Total Release Ci 2.85E+00 1.81E+00 2.00E+00 2.92E+00

2. Average release rate Ci/sec 3.74E-07 2.34E-07 2.56E-07 3.74E-07 for the period

% of limit is on the Radiological Impact to Man Table (reference Section 6.0)

Plant: Arkansas Nuclear One Year: 2019 Page 16 of 29 Annual Radioactive Effluent Release Report Table 6: Gaseous Effluents - Ground Level Release - Batch Mode (ANO-2)

Nuclides Released Unit Quarter 1 Quarter 2 Quarter 3 Quarter 4 Total Fission Gases Xe-133 Ci 0.00E+00 5.12E-02 0.00E+00 0.00E+00 5.12E-02 Total for Period Ci 0.00E+00 5.12E-02 0.00E+00 0.00E+00 5.12E-02 lodines I-131 Ci 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Total for Period Ci 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Particulates Total for Period Ci 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Tritium H-3 Ci 3.95E+00 6.90E+00 8.47E+00 5.42E+00 2.47E+01 Gross Alpha Alpha Ci 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Carbon-14 C-14 Ci 2.85E+00 1.81E+00 2.00E+00 2.92E+00 9.57E+00

Plant: Arkansas Nuclear One Year: 2019 Page 17 of 29 Annual Radioactive Effluent Release Report 4.0 LIQUID EFFLUENTS 4.1 Liquid Effluent and Waste Disposal Report 4.1.1 ANO-1 Table 7: Liquid Effluents - Summation of All Releases (ANO-1)

A. Fission & Activation Est. Total Unit Quarter 1 Quarter 2 Quarter 3 Quarter 4 Products Error %

1. Total Release (not including tritium, Ci 7.21E-03 6.32E-03 9.04E-03 8.62E-03 21 gases or alpha) 2 Average diluted concentration during Ci/mL 2.25E-11 1.91E-11 2.34E-11 3.56E-11 period B. Tritium

1. Total Release Ci 1.58E+02 1.69E+02 1.72E+02 7.10E+01 12

2. Average diluted concentration during Ci/mL 4.95E-07 5.11E-07 4.44E-07 2.93E-07 period C. Dissolved & Entrained Gases

1. Total Release Ci 2.88E-04 1.96E-04 3.28E-03 1.09E-03 22

2. Average diluted concentration during Ci/mL 9.01E-13 5.92E-13 8.46E-12 4.49E-12 period D. Gross Alpha Activity

1. Total Release Ci 0.00E+00 0.00E+00 0.00E+00 0.00E+00 27 E. Volume of Waste Released Liters 1.25E+06 2.52E+06 2.37E+06 2.31E+06 (prior to dilution)

F. Volume of Dilution Water Liters 3.20E+11 3.31E+11 3.87E+11 2.42E+11 Used During Period

% of limit is on the Radiological Impact to Man Table (reference Section 6.0)

Plant: Arkansas Nuclear One Year: 2019 Page 18 of 29 Annual Radioactive Effluent Release Report Table 8: Batch Mode Liquid Effluents (ANO-1)

Batch Mode Nuclides Released Unit Quarter 1 Quarter 2 Quarter 3 Quarter 4 Total Na-24 Ci 0.00E+00 0.00E+00 1.03E-03 2.26E-04 1.26E-03 Cr-51 Ci 0.00E+00 0.00E+00 1.03E-04 1.23E-03 1.33E-03 Mn-54 Ci 1.43E-05 2.88E-05 6.19E-06 0.00E+00 4.93E-05 Fe-55 Ci 3.50E-06 5.70E-04 4.38E-04 1.80E-05 1.03E-03 Fe-59 Ci 0.00E+00 0.00E+00 0.00E+00 1.46E-05 1.46E-05 Co-56 Ci 0.00E+00 0.00E+00 2.65E-06 0.00E+00 2.65E-06 Co-58 Ci 6.30E-04 1.07E-03 1.90E-03 9.02E-04 4.51E-03 Co-60 Ci 1.13E-03 2.49E-03 3.17E-03 1.37E-03 8.17E-03 Zr-95 Ci 1.95E-04 1.21E-04 9.97E-05 6.46E-05 4.80E-04 Nb-95 Ci 3.53E-04 3.18E-04 3.26E-04 1.65E-04 1.16E-03 Nb-97 Ci 0.00E+00 4.75E-06 4.35E-06 0.00E+00 9.09E-06 Mo-99 Ci 0.00E+00 0.00E+00 5.12E-05 2.27E-05 7.39E-05 Tc-99m Ci 0.00E+00 0.00E+00 5.21E-05 2.31E-05 7.52E-05 Ru-105 Ci 1.33E-05 2.80E-05 8.92E-06 7.22E-06 5.74E-05 Ag-110m Ci 3.16E-04 1.85E-04 3.83E-04 1.59E-03 2.47E-03 Sb-124 Ci 3.61E-04 9.75E-05 2.64E-04 1.17E-03 1.89E-03 Sb-125 Ci 4.48E-04 1.64E-04 8.72E-04 8.78E-04 2.36E-03 I-131 Ci 0.00E+00 1.92E-06 0.00E+00 0.00E+00 1.92E-06 I-133 Ci 0.00E+00 0.00E+00 0.00E+00 5.25E-06 5.25E-06 I-134 Ci 8.55E-06 0.00E+00 0.00E+00 0.00E+00 8.55E-06 Cs-134 Ci 1.20E-03 1.35E-04 9.71E-05 2.55E-04 1.69E-03 Cs-136 Ci 0.00E+00 0.00E+00 0.00E+00 2.85E-06 2.85E-06 Cs-137 Ci 2.52E-03 1.09E-03 2.25E-04 6.32E-04 4.47E-03 Cs-138 Ci 4.64E-06 0.00E+00 0.00E+00 0.00E+00 4.64E-06 Total for Period Ci 7.20E-03 6.31E-03 9.04E-03 8.58E-03 3.11E-02 H-3 Ci 1.58E+02 1.69E+02 1.72E+02 7.10E+01 5.70E+02 Total for Period Ci 1.58E+02 1.69E+02 1.72E+02 7.10E+01 5.70E+02 Kr-88 Ci 3.37E-05 0.00E+00 0.00E+00 0.00E+00 3.37E-05 Xe-133 Ci 2.55E-04 1.96E-04 3.23E-03 1.08E-03 4.76E-03 Xe-135 Ci 0.00E+00 0.00E+00 4.77E-05 1.27E-05 6.04E-05 Total for Period Ci 2.88E-04 1.96E-04 3.28E-03 1.09E-03 4.85E-03

Plant: Arkansas Nuclear One Year: 2019 Page 19 of 29 Annual Radioactive Effluent Release Report Table 9: Continuous Mode Liquid Effluents (ANO-1)

Continuous Mode Nuclides Released Unit Quarter 1 Quarter 2 Quarter 3 Quarter 4 Total Na-24 Ci 9.65E-06 1.01E-05 0.00E+00 0.00E+00 1.98E-05 Cs-134 Ci 0.00E+00 1.86E-06 0.00E+00 0.00E+00 1.86E-06 Cs-137 Ci 0.00E+00 0.00E+00 0.00E+00 4.24E-05 4.24E-05 Total for Period Ci 9.65E-06 1.20E-05 0.00E+00 4.24E-05 6.40E-05 H-3 Ci 0.00E+00 2.90E-03 2.27E-03 0.00E+00 5.17E-03 Total for Period Ci 0.00E+00 2.90E-03 2.27E-03 0.00E+00 5.17E-03 4.1.2 ANO-2 Table 10: Liquid Effluents - Summation of All Releases (ANO-2)

A. Fission & Activation Est. Total Unit Quarter 1 Quarter 2 Quarter 3 Quarter 4 Products Error %

1. Total Release (not including tritium, gases Ci 1.02E-04 2.67E-04 3.93E-04 1.35E-04 21 or alpha) 2 Average diluted concentration during Ci/mL 3.20E-13 8.08E-13 1.01E-12 5.59E-13 period B. Tritium

1. Total Release Ci 1.42E+01 3.54E+02 1.05E+02 3.70E+01 12

2. Average diluted concentration during Ci/mL 4.43E-08 1.07E-06 2.72E-07 1.53E-07 period C. Dissolved & Entrained Gases

1. Total Release Ci 5.95E-04 1.80E-03 0.00E+00 2.15E-04 22

2. Average diluted concentration during Ci/mL 1.86E-12 5.45E-12 0.00E+00 8.88E-13 period D. Gross Alpha Activity

1. Total Release Ci 0.00E+00 0.00E+00 0.00E+00 0.00E+00 27 E. Volume of Waste Released Liters 2.42E+05 1.44E+07 5.61E+06 1.76E+05 (prior to dilution)

F. Volume of Dilution Water Liters 3.20E+11 3.31E+11 3.87E+11 2.42E+11 Used During Period

% of limit is on the Radiological Impact to Man Table (reference Section 6.0)

Plant: Arkansas Nuclear One Year: 2019 Page 20 of 29 Annual Radioactive Effluent Release Report Table 11: Batch Mode Liquid Effluents (ANO-2)

Batch Mode Nuclides Released Unit Quarter 1 Quarter 2 Quarter 3 Quarter 4 Total Mn-56 Ci 0.00E+00 2.34E-06 0.00E+00 0.00E+00 2.34E-06 Co-56 Ci 0.00E+00 1.94E-06 0.00E+00 0.00E+00 1.94E-06 Co-58 Ci 0.00E+00 1.49E-05 2.20E-05 0.00E+00 3.69E-05 Co-60 Ci 0.00E+00 4.64E-05 7.21E-05 6.61E-06 1.25E-04 Y-91m Ci 0.00E+00 0.00E+00 5.73E-06 0.00E+00 5.73E-06 Sb-124 Ci 0.00E+00 5.17E-06 6.89E-05 1.02E-05 8.42E-05 Sb-125 Ci 1.67E-05 2.39E-05 1.49E-04 8.16E-05 2.71E-04 I-134 Ci 0.00E+00 3.39E-06 0.00E+00 0.00E+00 3.39E-06 Cs-137 Ci 8.58E-05 1.69E-04 7.51E-05 3.71E-05 3.67E-04 Total for Period Ci 1.02E-04 2.67E-04 3.93E-04 1.35E-04 8.98E-04 H-3 Ci 1.42E+01 3.54E+02 1.05E+02 3.70E+01 5.10E+02 Total for Period Ci 1.42E+01 3.54E+02 1.05E+02 3.70E+01 5.10E+02 Xe-133 Ci 5.95E-04 1.80E-03 0.00E+00 2.15E-04 2.61E-03 Total for Period Ci 5.95E-04 1.80E-03 0.00E+00 2.15E-04 2.61E-03 Table 12: Continuous Mode Liquid Effluents (ANO-2)

Continuous Mode Nuclides Released Unit Quarter 1 Quarter 2 Quarter 3 Quarter 4 Total H-3 Ci 0.00E+00 6.98E-02 1.31E-02 0.00E+00 8.29E-02 Total for Period Ci 0.00E+00 6.98E-02 1.31E-02 0.00E+00 8.29E-02

Plant: Arkansas Nuclear One Year: 2019 Page 21 of 29 Annual Radioactive Effluent Release Report 5.0 SOLID WASTE

SUMMARY

5.1 Solid Waste Shipped Offsite for Burial or Disposal (Not Irradiated Fuel) ANO-1 5.1.1 Types of Waste Summary Table 13: Types of Solid Waste Summary (ANO-1)

Total Total Est. Total Types of Waste Quantity (m3) Activity (Ci) Error (%)

a. Spent resins, filter sludges, evaporator bottoms, etc. 9.63E+00 8.78E-03 25

b. Dry compressible waste, contaminated equip, etc. 2.63E+02 3.18E-01 25

c. Irradiated components, control rods, etc. 0.00E+00 0.00E+00 -

d. Other 0.00E+00 0.00E+00 25 5.1.2 Estimate of major nuclide composition (by waste type) only > 1% (Note 1) are reported.

Table 14: Major Nuclides Summary (ANO-1)

Major Nuclide Composition % Curies

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

Co-60 1.33 1.17E-04 Cs-137 98.67 8.66E-03

b. Dry compressible waste, contaminated equip, etc.

C-14 25.73 8.18E-02 Mn-54 6.66 2.12E-02 Fe-55 20.06 6.38E-02 Co-58 10.73 3.41E-02 Co-60 24.63 7.83E-02 Ni-63 7.42 2.36E-02 Zn-65 1.18 3.74E-03 Tc-99 1.53 4.88E-03 Cs-137 1.90 6.03E-03

c. Irradiated components, control rods, etc.

None - -

d. Other None - -

Note 1 - "Major" radionuclide is equivalent to a "principle" radionuclide, i.e. greater than 1 percent of total activity.

Plant: Arkansas Nuclear One Year: 2019 Page 22 of 29 Annual Radioactive Effluent Release Report 5.1.3 Solid Waste Disposition Table 15: Solid Waste Disposition (ANO-1)

Number of Shipments Mode of Transportation Destination 8 Hittman Transport Bear Creek 2 Landstar Bear Creek Table 16: Irradiated Fuel Shipments Disposition (ANO-1)

Number of Shipments Mode of Transportation Destination 0 - -

5.2 Solid Waste Shipped Offsite for Burial or Disposal (Not Irradiated Fuel) ANO-2 5.2.1 Types of Waste Table 17: Types of Solid Waste Summary (ANO-2)

Total Total Est. Total Types of Waste Quantity (m3) Activity (Ci) Error (%)

a. Spent resins, filter sludges, evaporator bottoms, etc. 1.44E+01 1.22E-02 25

b. Dry compressible waste, contaminated equip, etc. 3.07E+01 8.96E-03 25

c. Irradiated components, control rods, etc. 0.00E+00 0.00E+00 -

d. Other 0.00E+00 0.00E+00 -

Plant: Arkansas Nuclear One Year: 2019 Page 23 of 29 Annual Radioactive Effluent Release Report 5.2.2 Estimate of major nuclide composition (by waste type) only > 1% (Note 1) are reported.

Table 18: Major Nuclides (ANO-2)

Major Nuclide Composition % Curies

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

Co-60 4.08 4.98E-04 Cs-137 95.92 1.17E-02

b. Dry compressible waste, contaminated equip, etc.

C-14 25.57 2.30E-03 Mn-54 6.68 6.00E-04 Fe-55 20.00 1.80E-03 Co-58 11.08 9.95E-04 Co-60 24.53 2.20E-03 Ni-63 7.38 6.62E-04 Zn-65 1.18 1.06E-04 Tc-99 1.53 1.37E-04 Cs-137 1.89 1.69E-04

c. Irradiated components, control rods, etc.

None - -

d. Other None - -

Note 1 - "Major" radionuclide is equivalent to a "principle" radionuclide, i.e. greater than 1 percent of total activity.

5.2.3 Solid Waste Disposition Table 19: Solid Waste Disposition (ANO-2)

Number of Shipments Mode of Transportation Destination 1 Hittman Transport Bear Creek 1 Hittman Transport Gallaher Road Table 20: Irradiated Fuel Shipments Disposition (ANO-2)

Number of Shipments Mode of Transportation Destination 0 - -

Plant: Arkansas Nuclear One Year: 2019 Page 24 of 29 Annual Radioactive Effluent Release Report 5.3 Solid Waste Shipped Offsite for Burial or Disposal (Not Irradiated Fuel) ANO-Common 5.3.1 Types of Waste Table 21: Types of Solid Waste Summary (ANO-Common)

Total Total Est. Total Types of Waste Quantity (m3) Activity (Ci) Error (%)

a. Spent resins, filter sludges, evaporator bottoms, etc. 2.46E+01 1.77E-02 25

b. Dry compressible waste, contaminated equip, etc. 1.66E+02 7.81E-02 25

c. Irradiated components, control rods, etc. 0.00E+00 0.00E+00 -

d. Other 1.44E+01 2.63E-06 25 5.3.2 Estimate of major nuclide composition (by waste type) only > 1% (Note 1) are reported.

Table 22: Major Nuclides (ANO-Common)

Major Nuclide Composition % Curies

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

Co-60 2.12 3.76E-04 Cs-137 97.88 1.74E-02

b. Dry compressible waste, contaminated equip, etc.

C-14 9.55 7.47E-03 Mn-54 1.77 1.38E-03 Fe-55 19.33 1.51E-02 Co-58 4.04 3.16E-03 Co-60 21.31 1.67E-02 Ni-63 11.91 9.32E-03 Zr-95 5.94 4.65E-03 Nb-95 10.93 8.55E-03 Ag-110m 1.40 1.09E-03 Cs-134 1.14 8.90E-04 Cs-137 10.36 8.11E-03

c. Irradiated components, control rods, etc.

None - -

d. Other Co-60 13.65 3.59E-07 Cs-137 86.35 2.27E-06 Note 1 - "Major" radionuclide is equivalent to a "principle" radionuclide, i.e. greater than 1 percent of total activity.

Plant: Arkansas Nuclear One Year: 2019 Page 25 of 29 Annual Radioactive Effluent Release Report 5.3.3 Solid Waste Disposition Table 23: Solid Waste Disposition (ANO-Common)

Number of Shipments Mode of Transportation Destination 11 Hittman Transport Bear Creek 1 Landstar Bear Creek 1 Landstar Gallaher Road Table 24: Irradiated Fuel Shipments Disposition (ANO-Common)

Number of Shipments Mode of Transportation Destination 0 - -

6.0 RADIOLOGICAL IMPACT TO MAN 6.1 10 CFR Part 50, Appendix I, Evaluation 6.1.1 ANO-1 Assessment Table 25: Dose Assessment (ANO-1)

Quarter 1 Quarter 2 Quarter 3 Quarter 4 Annual Liquid Effluent Dose Limit, 1.5 mrem 1.5 mrem 1.5 mrem 1.5 mrem 3 mrem Total Body Total Body Dose 2.35E-03 9.33E-04 3.93E-04 8.50E-04 4.53E-03

% of Limit 1.57E-01 6.22E-02 2.62E-02 5.67E-02 1.51E-01 Liquid Effluent Dose Limit, 5 mrem 5 mrem 5 mrem 5 mrem 10 mrem Any Organ Maximum Organ Dose 3.18E-03 1.24E-03 4.48E-04 1.09E-03 5.95E-03

% of Limit 6.35E-02 2.48E-02 8.95E-03 2.17E-02 5.95E-02 Gaseous Effluent Dose Limit, 5 mrad 5 mrad 5 mrad 5 mrad 10 mrad Gamma Air Gamma Air Dose 0.00E+00 0.00E+00 1.12E-07 1.40E-06 1.52E-06

% of Limit 0.00E+00 0.00E+00 2.24E-06 2.81E-05 1.52E-05 Gaseous Effluent Dose Limit, 10 mrad 10 mrad 10 mrad 10 mrad 20 mrad Beta Air Beta Air Dose 0.00E+00 0.00E+00 1.27E-05 4.18E-06 1.69E-05

% of Limit 0.00E+00 0.00E+00 1.27E-04 4.18E-05 8.44E-05

Plant: Arkansas Nuclear One Year: 2019 Page 26 of 29 Annual Radioactive Effluent Release Report Table 25: Dose Assessment (ANO-1)

Quarter 1 Quarter 2 Quarter 3 Quarter 4 Annual Gaseous Effluent Organ Dose Limit (Iodine, Tritium, 7.5 mrem 7.5 mrem 7.5 mrem 7.5 mrem 15 mrem Particulates with > 8-day half-life)

Gaseous Effluent Organ Dose (Iodine, Tritium, Particulates 2.92E-02 2.06E-02 8.68E-03 3.40E-02 9.26E-02 with > 8-day half-life)

% of Limit 3.90E-01 2.74E-01 1.16E-01 4.54E-01 6.17E-01 6.1.2 ANO-2 Assessment Table 26: Dose Assessment (ANO-2)

Quarter 1 Quarter 2 Quarter 3 Quarter 4 Annual Liquid Effluent Dose Limit, 1.5 mrem 1.5 mrem 1.5 mrem 1.5 mrem 3 mrem Total Body Total Body Dose 6.28E-05 6.44E-04 1.67E-04 1.42E-04 1.02E-03

% of Limit 4.19E-03 4.29E-02 1.12E-02 9.46E-03 3.39E-02 Liquid Effluent Dose Limit, 5 mrem 5 mrem 5 mrem 5 mrem 10 mrem Any Organ Maximum Organ Dose 8.42E-05 6.87E-04 1.83E-04 1.60E-04 1.11E-03

% of Limit 1.68E-03 1.37E-02 3.66E-03 3.19E-03 1.11E-02 Gaseous Effluent Dose Limit, 5 mrad 5 mrad 5 mrad 5 mrad 10 mrad Gamma Air Gamma Air Dose 0.00E+00 1.15E-05 0.00E+00 0.00E+00 1.15E-05

% of Limit 0.00E+00 2.29E-04 0.00E+00 0.00E+00 1.15E-04 Gaseous Effluent Dose Limit, 10 mrad 10 mrad 10 mrad 10 mrad 20 mrad Beta Air Beta Air Dose 0.00E+00 3.41E-05 0.00E+00 0.00E+00 3.41E-05

% of Limit 0.00E+00 3.41E-04 0.00E+00 0.00E+00 1.71E-04 Gaseous Effluent Organ Dose Limit (Iodine, Tritium, 7.5 mrem 7.5 mrem 7.5 mrem 7.5 mrem 15 mrem Particulates with > 8-day half-life)

Gaseous Effluent Organ Dose (Iodine, Tritium, Particulates 1.74E-02 3.03E-02 3.72E-02 2.38E-02 1.09E-01 with > 8-day half-life)

% of Limit 2.32E-01 4.05E-01 4.97E-01 3.18E-01 7.25E-01

Plant: Arkansas Nuclear One Year: 2019 Page 27 of 29 Annual Radioactive Effluent Release Report 6.2 40 CFR Part 190 Evaluation for an Individual in the Unrestricted Area Table 27: EPA 40 CFR Part 190 Evaluation Whole Body Thyroid Any Other Organ Dose Limit 25 mrem 75 mrem 25 mrem Dose 7.59E-01 7.55E-01 1.39E+00

% of Limit 3.04E+00 1.01E+00 5.56E+00 Liquid dose, gaseous dose including C14, direct shine from each unit, the Independent Spent Fuel Storage Installation (ISFSI), and any other nuclear power related facility within 5 miles of the station are considered when calculating dose compliance with 40 CFR 190.

6.3 40 CFR Part 190 Calculation Table 28: EPA 40 CFR Part 190 Calculation Unit Total Body Thyroid Max Organ Routine Airborne Effluents[Note 1] Unit 1 8.92E-07 - 1.11E-05 Routine Airborne Effluents ITP Unit 1 9.26E-02 9.26E-02 9.26E-02 Routine Liquid Effluents Unit 1 4.53E-03 9.99E-04 5.95E-03 Airborne Releases of C14 Unit 1 3.20E-02 3.20E-02 1.71E-01 Routine Airborne Effluents [Note 1]

Unit 2 6.69E-06 - 1.89E-05 Routine Airborne Effluents ITP Unit 2 1.09E-01 1.09E-01 1.09E-01 Routine Liquid Effluents Unit 2 1.02E-03 8.30E-04 1.11E-03 Airborne Releases of C14 Unit 2 1.20E-01 1.20E-01 6.11E-01 Ground Water & Storm Drain Totals Site 0.00E+00 0.00E+00 0.00E+00 Direct Shine from areas such as dry cask storage, radwaste storage, Site 4.00E-01 4.00E-01 4.00E-01 Equipment Mausoleums[Note 2]

Total 40 CFR 190 Dose Site 7.59E-01 7.55E-01 1.39E+00 Note 1: Routine airborne dose in this table is mrad expressed as mrem. This addition does not represent a real dose and is listed here solely to help demonstrate compliance with 40 CFR 190.

Note 2: Average direct radiation control location TLD was compared to the average area of an "actual" person TLD monitoring area [Reeves E. Richie Training Center (RERTC)] to determine yearly dose from direct shine to an actual member of the public. Since constant occupation is unrealistic, a value of 52 weeks a year

40 hours4.62963e-4 days 0.0111 hours 6.613757e-5 weeks 1.522e-5 months in a week occupying results in a total of 2080 hours0.0241 days 0.578 hours 0.00344 weeks 7.9144e-4 months . Control Value = 28.4 mrem/yr. RERTC Value = 30.1 mrem/yr.

Weighted = 2080/8760 (hrs) = 23.7%. Actual Dose = (30.1 - 28.4)

0.237 = 0.40 mrem.

Plant: Arkansas Nuclear One Year: 2019 Page 28 of 29 Annual Radioactive Effluent Release Report 7.0 METEOROLOGICAL DATA 7.1 Joint Frequency Distributions

1. Period of Record: 01/01/2019 - 12/31/2019

2. Elevation: 57 m Table 29: Percentage of Each Wind Speed/Direction Wind Speed (mph)

Wind 0-2 2-4 4-6 6-8 8 - 10 10 - 15 15 - 20 > 20 Total Direction N 0.58 0.45 0.28 0.50 0.27 0.16 0.00 0.00 2.23 NNE 0.74 0.71 0.86 0.75 0.25 0.07 0.00 0.00 3.38 NE 1.38 1.20 1.49 1.15 0.34 0.16 0.02 0.00 5.74 ENE 2.00 2.13 4.23 3.32 1.70 0.91 0.12 0.00 14.41 E 1.92 2.05 3.51 3.18 2.66 2.40 0.42 0.00 16.14 ESE 1.39 1.52 1.88 1.71 1.02 0.76 0.16 0.03 8.48 SE 1.08 1.09 1.68 1.82 1.15 0.49 0.06 0.01 7.36 SSE 0.75 0.85 1.08 1.66 1.22 0.68 0.02 0.00 6.25 S 0.69 0.37 0.69 1.30 0.63 0.37 0.02 0.00 4.08 SSW 0.52 0.34 0.28 0.57 0.38 0.13 0.00 0.00 2.21 SW 0.78 0.41 0.39 0.35 0.25 0.13 0.01 0.02 2.34 WSW 0.71 0.58 0.53 0.21 0.10 0.15 0.13 0.13 2.54 W 1.09 1.03 1.33 1.04 0.79 1.16 0.80 0.56 7.79 WNW 1.26 1.08 1.24 1.29 1.44 3.07 0.94 0.17 10.48 NW 0.57 0.39 0.32 0.34 0.57 0.76 0.23 0.05 3.23 NNW 0.37 0.42 0.32 0.23 0.23 0.28 0.06 0.00 1.91

3. Variable

a. Total period of calm hours: 4.89%

b. Percentage of missing data: 2.4%

Plant: Arkansas Nuclear One Year: 2019 Page 29 of 29 Annual Radioactive Effluent Release Report 7.2 Stability Class Table 30: Classification of Atmospheric Stability Stability Condition Pasquill Categories Percentage Extremely Unstable A 0.08 Moderately Stable B 0.08 Slightly Unstable C 0.18 Neutral D 27.95 Slightly Stable E 58.57 Moderately Stable F 11.99 Extremely Stable G 1.14

Enclosure Attachment 0CAN042001 Offsite Dose Calculation Manual

ARKANSAS NUCLEAR ONE OFFSITE DOSE CALCULATION MANUAL REVISION 29 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 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 29 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 APPENDIX 1 RADIOLOGICAL EFFLUENT CONTROLS BASES Revision 29 3

ARKANSAS NUCLEAR ONE ODCM 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 29 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)-1(Qi)(Daij)]

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 0 hours 0 weeks 0 months (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 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , 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 29 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-1)(Qi)(Daij)]

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 0 hours 0 weeks 0 months (i.e., no decay correction) for the above equation to be consistent with the equation presented in Chapter 4 of NUREG-0133.

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; Revision 29 9

ARKANSAS NUCLEAR ONE ODCM 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.

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.

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ARKANSAS NUCLEAR ONE ODCM 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 or Ci/cc);

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

K = conversion factor determined from response curve of monitor (cpm per Ci/cc). 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 (or µCi/cc for RX-9820)).

1.2 = Safety Factor to correct for instrument uncertainties.

3.1.2 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-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; DRI = 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 0.0467 hours 2.777778e-4 weeks 6.3924e-5 months 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 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 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; DI = 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.

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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, ULa, UMa, UFa = 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, CM F i, Di = 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 29 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):

- t DG(R,,j,a) = {(SF)(1.0 x 1012)(i[(i-1)(1 - e i b)]}(DOQ(r,))(Q )(DFG )

i ij 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 0.0292 hours 1.736111e-4 weeks 3.99525e-5 months );

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 29 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:

CVi(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.

V Revision 29 19

ARKANSAS NUCLEAR ONE ODCM 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:

D(r,,j,a) = i [(DFIija)(Ua)(CVi(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 CVi(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:

r - t CFVi = (CONST)( )(e i h)(f)

(Yv)(i) where:

CFVi = 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 29 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 0 hours 0 weeks 0 months for pasture grass consumed by animals th = 2160 hours0.025 days 0.6 hours 0.00357 weeks 8.2188e-4 months for stored feed consumed by animals th = 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for leafy vegetables consumed by humans th = 1440 hours0.0167 days 0.4 hours 0.00238 weeks 5.4792e-4 months 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 1.222 hours 0.00728 weeks 0.00167 months ),

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:

CFTV = 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).

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

Revision 29 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:

V CVi(r,) = (CFi)(DOQ(r,))(Qi) for produce; and L

CLi(r,) = (CFi)(DOQ(r,))(Qi) for leafy vegetables where:

V CFi = concentration factor of nuclide i in produce; L

CFi = concentration factor of nuclide i in leafy vegetables; V L Note that the difference between CFi 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):

V CVT(r,) = (CFT)(D2DPX/Q(r,))(QT)(fg)

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

L C14L (r,) = (CF14)(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 - t v v1 v1 Ci(r,) = {(Fm)(QF)(e i f)[(f p)(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 29 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 - t v v1 v1 Ci(r,) = {(Ff)(QF)(e i s)[(f p)(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 29 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) - t - t Rapj = 110,000( (i [(Qi)(Ti)(Daipj)(e i p)(1-e i b)]

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 29 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 0 hours 0 weeks 0 months 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 0.0186 hours 1.107804e-4 weeks 2.54935e-5 months 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 29 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 29 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 29 27

ARKANSAS NUCLEAR ONE ODCM Required Limitation Description Action Sample not taken at required location*

Sample equipment out-of-service (OOS)

Sample frequency not met 2.5.1 A.2 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 Other harmful effects or evidence of irreversible damage NA NA 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 29 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 29 29

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

0° 16 340° 20° 2

US HWY 7 TO HARRISON 320° 40° 15 INTERSTATE 40 TO FORT SMITH 3

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

HWY 64 14 4 SR 24 TO 14 MORELAND 116 280° ARKANSAS RIVER 80° INTERSTATE 40 LONDON 16 US HWY 64 13 J I H G F E D C B A 5 DELAWARE STATE PARK 127 111 ARKANSAS TECH DARDANELLE UNIVERSITY 260° STATE PARK 100° U.S. HWY 22 HWY 524 LAKE DARDANELLE RUSSELLVILLE DARDANELLE STATE PARK 6

DARDANELLE LOCK AND DAM 12 HWY 22 DAM SITE EAST PARK HWY 7T 6

240° 120° HWY 155 HWY 7 SR 247 TO MT. NEBO HWY 27 POTTSVILLE STATE PARK DARDANELLE 137 11 220° HWY 28 140° 7

HWY 7 200° 160° 10 180° 8 HWY 27 TO HWY 7 TO DANVILLE 9

HOT SPRINGS DANVILLE INSET (SEE INSET) N FS Rd 1618A W E 55 AR Hwy 307 FS Rd 36 Ozark National Forest boundary S AR Hwy 27 AR Hwy 10 Entergy Substation Petit Jean River 7

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

Revision 29 30

ARKANSAS NUCLEAR ONE ODCM FIGURE 4-1A RADIOLOGICAL SAMPLE STATIONS SR 333 152 3 108 Training 145 Center 146 147 109 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 29 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 29 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 29 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 29 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.

Travel south on Hwy 27 and west on Hwy 307 to the western edge of the Ozark National Forest. Hwy 307 becomes Forest 217° -

55 Broad leaf vegetation Service (FS) Rd 36; proceed ~ 3/4 mile on 13.1 miles FS Rd 36 to its intersection with FS Rd 1618A. 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 29 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 29 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 29 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.

STR - 351° - Inside protected area near Sally Port from Storm water runoff 2 < 0.10 miles drainage ditch along fence.

STR - Outside Protected Area near Sally Port 0.2° - 0.13 miles Storm water runoff 3 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 29 38

ARKANSAS NUCLEAR ONE ODCM APPENDIX 1 RADIOLOGICAL EFFLUENT CONTROLS Revision 29 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 29 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 29 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 29 42

ARKANSAS NUCLEAR ONE ODCM 2.0 LIMITATION (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 29 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 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 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 29 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 29 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 0.00111 hours 6.613757e-6 weeks 1.522e-6 months 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 0.02 hours 1.190476e-4 weeks 2.7396e-5 months 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 29 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 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 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 0.00222 hours 1.322751e-5 weeks 3.044e-6 months prior instrumentation. to release of radioactive effluents monitored by the channel Revision 29 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 29 48

ARKANSAS NUCLEAR ONE ODCM L 2.2.1 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 29 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 0.00111 hours 6.613757e-6 weeks 1.522e-6 months 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 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Required Noble Gas Activity AND Monitor non-functional.

C.2 Analyze sample of effluent. Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following completion of Required Action C.1 D. --------------NOTE-------------- D.1 Verify effluent samples are 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months 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 Particulate cartridges (as non-functional. applicable).

AND D.3 Analyze Iodine and/or Within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months Particulate cartridges (as following replacement applicable).

Revision 29 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 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 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 29 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 29 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 29 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 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 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 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 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.

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 29 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 29 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 0.00667 hours 3.968254e-5 weeks 9.132e-6 months AND Perform gamma isotopic and I-131 analysis. 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 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 29 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 29 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.3 projected to excess of 40 CFR 190 limits.

exceed 40 CFR 190 limits.

Revision 29 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 29 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 0.00667 hours 3.968254e-5 weeks 9.132e-6 months

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 29 60

ARKANSAS NUCLEAR ONE ODCM L 2.4.1 SURVEILLANCES (continued)

SURVEILLANCE FREQUENCY S 2.4.1.2 (continued)

Complete analysis of above samples:

i. Samples a, b, c, and d 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months 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 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 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 29 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 29 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 29 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 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 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 29 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 29 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 29 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 29 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 Cs-137 18 6x 10-2(f) 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 29 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 29 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 2.5.1. 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 29 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 29 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 29 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 29 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 29 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 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 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 29 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 29 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 0.00111 hours 6.613757e-6 weeks 1.522e-6 months of initial loss of the instrument and every 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months 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 29 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 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . 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 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 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 29 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 29 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 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 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 29 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 20.1301. 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 29 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 29 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 0.00111 hours 6.613757e-6 weeks 1.522e-6 months of initial loss of the instrument and every 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months 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 29 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 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and analyzed within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . 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 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . 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 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months . 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 29 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 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 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 29 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 29 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 29 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 29 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 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to once every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . The analysis of the sample must be completed with 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 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 29 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 cannot 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 29 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 29 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 29 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 29 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 29 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 29 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 cannot 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 29 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 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . 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 29 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 29 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.

Community as defined by Websters dictionary is people with common interests living in a particular area; broadly: the area itself. The local municipalities of London, Russellville, and Dardanelle are all part of the River Valley community located near Dardanelle Lake and the Arkansas River. The grouping of houses that reside within WSW (highest D/Q) sector are located in London which is part of the River Valley community. Air Station #2 per the above mentioned NRC BTP meets the requirements of being located within the highest D/Q and also in the vicinity of a community. Reference CR-ANO-C-2016-2732.

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 29 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 29 100

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 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 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 29 101

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 29 102

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 29 103

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 29 104

ML20118C140

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ML20118C140

Text

Subject:

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1.0 INTRODUCTION

SUMMARY

1.0 INTRODUCTION

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RESPONSE

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