Text

Submittal of 2016 Radioactive Effluent Release Report
	ML17138A005
Person / Time
Site:	Vermont Yankee
Issue date:	05/11/2017
From:	Chappell C; Entergy Nuclear Vermont Yankee
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	BVY 17-017
	Download: ML17138A005 (472)
	v • d • e

{{#Wiki_filter:'~ ~~E:ntergy Entergy Nuclear Vermont Yankee, LLC Vermont Yankee 320 Governor Hunt Rd. Vernon, VT 05354 (802) 257-7711 Coley C. Chappell Manager, Design and Programs BVY 17-017 May 11, 2017 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001

SUBJECT:

2016 Radioactive Effluent Release Report Vermont Yankee Nuclear Power Station Docket No. 50-271 License No. DPR-28 D1:iar Sir or Madam: In accordance with Vermont Yankee (VY) Technical Specifications (TS) 6.6.D, enclosed is a copy of the annual 2016 Radioactive Effluent Release Report (RERR). In addition, VY TS 6.7.B requires reporting of changes to the Off-Site Dose Calculation Manual (ODCM). Appendix H of the RERR provides a summary of two revisions to the ODCM involving the changes which occurred during 2016. Attachment 1 to Appendix H provides a copy of the affected pages from Revision 36. Attachment 2 to Appendix H provides a complete copy of Revision 37. There are no new regulatory commitments contained in this submittal. Should you have any questions or require additional information concerning this submittal, please contact me at (802) 451-3374. Sincerely, [C~~ t7r ell---

Enclosure:

Radioactive Effluent Release Report for 2016 cc listing (next page)

BVY 17-017 I Page 2 of 2 cc: Mr. Daniel H. Dorman Regional Administrator, Region 1 U.S. Nuclear Regulatory Commission 2100 Renaissance Blvd, Suite 100 King of Prussia, PA 19406-2713 Mr. Jack D. Parrott, Senior Project Manager Office of Nuclear Material Safety and Safeguards U.S. Nuclear Regulatory Commission Mail Stop T-8F5 Washington, DC 20555 Ms. June Tierney, Commissioner Vermont Department of Public Service 112 State Street - Drawer 20 Montpelier, VT 05602-2601 Vermont Department of Health Division of Radiological Health Attn: Bill Irwin P.O. Box 70 Burlington, VT 05402-0070 Massachusetts Department of Public Health Jack Priest, Director Radiation Control Program 529 Main Street, Suite 1M2A Charlestown, MA 02129 Augustinus Ong, Administrator Department of Health and Human Services Radiological Health Section 29 Hazen Drive Concord, NH 03301 ~6504 John Giarrusso Director of Nuclear Preparedness Massachusetts Emergency Management Agency 400 Worcester Road Framingham, MA 01702

Tony Honnellio Radiation Program Manager, Health and Safety Coordinator EPA, New England, Region 1 5 Post Office Square, Suite 100 (OSRR02-2}

Boston, MA 02109

RADIOACTIVE EFFLUENT RELEASE REPORT FOR2016 INCLUDING ANNUAL RADIOLOGICAL IMPACT ON MAN Entergy Nuclear Vermont Yankee, LLC Docket No. 50-271 License No, DPR-28

TABLE OF CONTENTS

1.0 INTRODUCTION

.............................................................................................................................................. 1 2.0 METEOROLOGICAL DATA ............................................................................................................................ 2 3.0 DOSE ASSESSMENT ....................................................................................................................................... 3 3 .1 DOSES FROM LIQUID EFFLUENTS ................................................................................................................... 3 3.2 DOSES FROM NOBLE GASES ........................................................................................................................... 4 3.3 DOSE FROM RADIONUCLIDES IN PARTICULATE FORM AND TRITIUM ............................................................ 4 3.4 WHOLE BODY DOSES IN UNRESTRICTED AREAS FROM DIRECT RADIATION ................................................. 5 3.5 DOSES FROM ON-SITE DISPOSAL OF SEPTIC WASTE, COOLING TOWER SILT AND SOIL ............................... 5 3.6 ON-SITE RECREATIONAL ACTIVITIES ............................................................................................................. 6 REFERENCES .......................................................................................................................................................... 7 APPENDIX A-SUPPLEMENTAL INFORMATION ......................................................................................... A-1 APPENDIX B -LIQUID HOLDUP TANKS ........................................................................................................ B-1 APPENDIX C - RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION ............................................................................................................... C-1 APPENDIX D -RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION ................................................................................................................ D-1 APPENDIX E -RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM ..................................... E-1 APPENDIX F - LAND USE CENSUS .................................................................................................................. F-1 APPENDIX G -PROCESS CONTROL PROGRAM ............................................................................................ G-1 APPENDIX H -OFF-SITE DOSE CALCULATION MANUAL ......................................................................... H-1 APPENDIX I - RADIOACTIVE LIQUID, GASEOUS AND SOLID WASTE TREATMENT SYSTEMS ............................................................................................................ I-1 APPENDIX J - ON-SITE DISPOSAL OF SEPTIC/SILT/SOIL WASTE ............................................................ J-1 11

LIST OF TABLES lA Gaseous Effluents - Summation of All Releases ............................................................................. 8 lB Gaseous Effluents - Elevated Releases .......................................................................................... 10 lC Gaseous Effluents - Ground Level Releases .................................................................................. 12 lD Gaseous Effluents - Non-routine Releases .................................................................................... 14 2A Liquid Effluents - Summation of All Releases .............................................................................. 15 2B Liquid Effluents - Routine Releases ............................................................................................. 17 3 Solid Waste and Irradiated Fuel Shipments ................................................................................... 19 4A Maximum Quarterly and Annual Off-Site Doses from Direct Radiation and Liquid and Gaseous Effluents for 2016 (10CFR50, Appendix I) .................................................................... 21 4B Maximum Annual Off-Site Doses from Direct Radiation and Liquid and Gaseous Effluents for 2016 (40CFR190) .................................................................................................................... 22 4C

Receptor Locations ........................................................................................................................ 23 4D Usage Factors for Environmental Pathways ................................................................................. 24 4E Environmental Parameters for Gaseous Effluents ........................................................................ 25 4F Environmental Parameters for Liquid Releases (Tritium) Via Groundwater ................................ 27 5A to 5H Annual (2016) Summary of Lower Level Joint Frequency Distribution .................................. 28-35 6A to 6H Annual (2016) Summary of Upper Level Joint Frequency Distribution .................................. 36-43 iii

Radiological Effluent Release Report for 2016 [Including Annual Radiological Impact on Man] Entergy Nuclear Vermont Yankee, LLC

1.0 INTRODUCTION

Tables 1 through 3 list the recorded radioactive liquid and gaseous effluents and solid waste shipments for the year, with data summarized on a quarterly basis for both liquids and gases. Table 4A summarizes the estimated radiological dose commitments from all radioactive liquid and gaseous effluents released during the year 2016 in response to the ALARA objectives of 10 CFR Part 50, Appendix I. Also included in Table 4A is the estimate of direct dose from fixed station sources along the limiting west site boundary line. Tables SA through 6H report the cumulative joint frequency distributions of wind speed, wind direction, and atmospheric stability for the 12-month period, January to December 2016. Radioactive effluents reported in Tables 1 and 2 were used to determine the dose to the maximum exposed individual member of the public for 2016. Dose commitments resulting from the release of radioactive materials in liquids and gases during the reporting period were estimated in accordance with the plant's Off-Site Dose Calculation Manual (ODCM), Section 10.l (Reference 1). These dose estimates were made using a "Method II" analysis as described in the ODCM, and as reported in Tables 4A and 4B of this report. A "Method II" analysis incorporates the methodology of Regulatory Guide 1.109 (Reference 2) and actual measured meteorological data recorded concurrently with the quarterly reporting period. As required by ODCM Section 10.1, this report shall also include an assessment of the radiation doses from radioactive effluents to member(s) of the public due to allowed recreational activities inside the site boundary during the year. As discussed in Section 3.6, there were no such recreational activities permitted and, therefore, there is no associated dose assessment. An assessment of radiation doses (including direct radiation) to the likely most exposed real member(s) of the public for the calendar year for the purposes of demonstrating conformance with 40 CFR Part 190, "Environmental Radiation Protection Standards for Nuclear Power Operations," is also required to be included in this report ifthe conditions indicated in ODCM 3/4.4, "Total Dose," have been exceeded during the year. Since the conditions indicated in the action statement under ODCM 3/4.4 were not entered into during the year, no additional radiation dose assessment is required. However, Table 4B does provide the combination of off-site doses and dose commitments from plant effluents and direct radiation sources for the limiting member of the public as a demonstration of compliance with the dose standards of 40 CFR Part 190. All calculated dose estimates for members of the public at the site boundary or beyond for the 2016 annual reporting period are below the dose criteria of 10 CFR Part 50, Appendix I, and 40 CFR Part 190. Appendices B through H indicate the status of reportable items per the requirements of ODCM Section 10.1. 1

2.0 METEOROLOGICAL DATA Meteorological data were collected in 2016 from the site's 300-foot meteorological tower located approximately 2,200 feet northwest of the reactor building, and about 1,400 feet from the plant stack. The 300-*foot tower is approximately the same height as the primary plant stack (308 feet) and is designed to meet the requirements of Safety Guide 23 (Reference 3) for meteorological monitoring. In mid-2009, the tower was moved to a location approximately 200 feet northwest of the original location. xJQ and D/Q values for elevated releases were derived for all receptor points from the site meteorological record for each quarter using a straight-line airflow model. All dispersion factors have been calculated employing appropriate source configuration considerations, as described in Regulatory Guide 1.111 (Reference 4). A source depletion model as described in "Meteorology and Atomic Energy - 1968" (Reference 5) was used to generate deposition factors, assuming a constant deposition velocity of 0.01 m/sec for all stack (elevated) releases. Changes in terrain elevations in the site environment were also factored into the meteorological models as appropriate. For any batch or discrete gas volume releases, the meteorological conditions concurrent with the time of release of radioactive materials in.gaseous effluents shall be used in determining the gaseous pathway doses. For 2016 there were no reported discrete or batch gas releases. In the event of a ground-level release, xJQ and D/Q values would be derived for the site boundary receptor points from the site meteorological record for each quarter using a straight-line airflow model. During this reporting period, there were no ground-level releases and therefore no associated dose impact. Table 4C lists the distances from the plant stack to the nearest site boundary, resident, and milk animal in each of the 16 principle compass directions as determined during the 2016 land use census. These locations were used in the calculation of atmospheric dispersion factors. The meteorological model was also executed for each calendar quarter to determine the location of the predicted maximum ground level air concentration from elevated releases from the plant's primary vent stack. These locations were included in the assessment of effluent doses along with identified points of interest from the annual land use census. 2

3.0 DOSE ASSESSMENT 3 .1 Doses From Liquid Effluents ODCM 3/4.2.2 limits total body doses (1.5 mrem per quarter, and 3 mrem per year) and organ doses (5 mrem per quarter, and 10 mrem per year) from liquid effluents to a member of the public to those specified in 10 CFR Part 50, Appendix I. By implementing the requirements of 10 CFR Part 50, Appendix I, ODCM 3/4.2.2 assures that the release of radioactive material in liquid effluents will be kept "as low as is reasonably achievable." There were no recorded routine liquid radioactive waste discharges during the report period. However, an abnormal release to the Connecticut River is postulated due to a past leak in an underground pipe tunnel that runs between the Advanced Offgas (AOG) system building and other plant buildings which allowed accumulated piping system leakage to enter the subsurface groundwater adjacent to the plant structures. The existence of the leak was first recognized in January 2010, when a river shoreline Protected Area Boundary monitoring well sample was reported to have detectable tritium. The addition of other monitoring wells and subsequent analysis defmed the extent of the affected groundwater plume moving toward the river and helped locate the source of the leak, which was stopped in February 2010. Estimates of tritium-contaminated ground water released from the site are based on Protected Area Boundary monitoring well data collected throughout 2016, and hydrological modeling of ground water movement in the affected zone impacted by the pipe tunnel leak. Using a conservative estimate of groundwater flow through the affected area toward the river on a quarterly basis, an estimate of the total potential tritium released from the site during each quarter of2016 was generated and reported in Table 2A. For the projected ground water flow into the Connecticut River in 2016, the dose impact to the maximum exposed individual (MEI) assumed the following exposure pathways: (1) ingestion of fish (taken from Vernon Pond), (2) ingestion of vegetables and fresh leafy produce irrigated by water taken from the river below Vernon Dam, (3) ingestion of milk and meat from animals that were fed irrigated crops and drinking water taken from the river below Vernon Dam, and (4) potable water for a hypothetical individual drawing drinking water fed by the river below Vernon Dam. For Vernon Pond (river area adjacent to the plant property), the near shore mixing zone associated with the fish ingestion pathway is conservatively taken as 1% of the minimum recorded monthly river flow (1379 cfs in October 2016) for dilution. All irrigation exposure pathways for the consumption of food products grown with irrigated water occur below Vernon Dam and assume the lowest 2016 quarterly average growing season river flow value (2939 cfs in the third quarter) for environmental mixing. For the drinking water pathway, river flow mixing is assumed to occur below Vernon Dam and uses the lowest annual quarterly average river flow (2939 cfs in the third quarter) as a conservative estimate of river dilution for all four quarters of the year. The dose models are taken from Regulatory Guide 1.109 (Reference 2) and use environmental parameters for exposure pathways listed in Tables 4D and 4F. The maximum estimated quarterly and annual whole body and organ doses to the limiting age group from liquid releases are reported in Table 4A. These estimated doses are well below the 10 CFR Part 50, Appendix I dose criteria of ODCM 3/4.2.2. Table 4B provides an estimate of the total annual dose impact (including contribution from liquids) associated with the highest exposed member of the public for demonstration of compliance to the dose standard contained in 40 CFR Part 190 for the uranium fuel cycle. 3

3.2 Doses From Noble Gases ODCM 3/4.3.2 limits the gamma air dose (5 mrad per quarter, and 10 mrad per year) and beta air (10 mrad per quarter, and 20 mrad per year) dose from noble gases released in gaseous effluents from the site to areas at and beyond the site boundary to those specified in 10 CFR Part 50, Appendix I. By implementing these, ODCM 3/4.3.2 assures that the releases of radioactive noble gases in gaseous effluents will be kept "as low as is reasonably achievable." Dose estimates due to the release of noble gases to the atmosphere are typically calculated at the site boundary, at the nearest resident in each of the sixteen principal compass directions, at the point of highest off-site ground level air concentration of radioactive materials, and at each of the milk animal locations located within five miles of the plant. For 2016, there were no noble gases detected in effluents released from the plant stack. 3.3 Dose From Radionuclides in Particulate Form and Tritium ODCM 3/4.3.3 limits the organ dose to a member of the public from tritium and radionuclides in particulate form in gaseous effluents released from the site to areas at and beyond the site boundary to those specified in 10 CFR Part 50, Appendix I (7.5 mrem per quarter and 15 mrem per year). By implementing the requirements of 10 CFR Part 50, Appendix I, ODCM 3/4.3.3 assures that the releases of any tritium and particulates in gaseous effluents will be kept "as low as is reasonably achievable." During 2016, four frac tanks were used on the Vermont Yankee site to temporarily store (outdoors) tritium-contaminated water extracted from onsite groundwater wells. The quantity of tritium released to the atmosphere through the evaporation of water from this frac tank was estimated, and the dose consequence to the maximally exposed individual was calculated. Exposure pathways that could exist as a result of the planned (routine) release of particulates to the atmosphere include external irradiation from activity deposited onto the ground surface, inhalation, and ingestion of vegetables, meat and milk. Dose estimates for 2016 were made at the site boundary and nearest resident in each of the sixteen principal compass directions, as well as all milk animal locations within five miles of the plant. The nearest resident and milk animals in each sector were identified by the most recent annual land use census as required by ODCM 3/4.5.2 prior to Revision 37 in December, 2016 (see Table 4C). Conservatively, a vegetable garden was assumed to exist at each milk animal and nearest resident location. Furthermore, the meat pathway was assumed to exist at each milk cow location since this data category is not part of the annual land use census. Doses were also calculated at the point of maximum ground level air concentration of radioactive materials in gaseous effluents and included the assumption that the inhalation, vegetable garden, and ground plane exposure pathways exist for an individual with a 100 percent occupancy factor. It is assumed that milk and meat animals are free to graze on open pasture during the second and third quarters with no supplemental feeding. This assumption is conservative since most of the milk anin1als inventoried in the site vicinity are fed stored feed throughout the entire year with only limited grazing allowed during the growing season. During the non-growing season (first and fourth quarters), the milk animals are assumed to receive only stored feed. During the growing season (second and third quaiters), all animal feed is assumed to be derived from fresh pasture. Usage factors for gaseous effluents 4 L_

are listed by age group and pathway in Table 4D. Table 4E provides other dose model parameter assumptions used in the dose assessments. In previous years when the plant was operating, Carbon-14 was an important nuclide to consider in the effluent dose calculations. However, with the plant permanently shut down since December of 2014, there is no longer any plant-related Carbon-14 production and, therefore, no associated dose impact for 2016. 3.4 Whole Body Doses in Unrestricted Areas From Direct Radiation As opposed to prior years before the permanent shut down when the majority of the dose in the unrestricted area consisted of direct and skyshine radiation from N-16 decay in the Turbine Building steam cycle during power operations, there was no such source during 2016 due to the elimination of its production and its short half-life. The other potential fixed sources of direct and scatter radiation to the site boundary are the Independent Spent Fuel Storage Installation (ISFSI), the low level radioactive materials stored in the North Warehouse, the Low Level Waste Storage Pad Facility (no radioactive waste material stored on the pad in 2016), and old turbine rotors and casings in the Turbine Storage Facility. The annual dose is based on dose rate measurements in these storage facilities and is projected to impact the same most restrictive site bom1dary dose location. The estimated direct radiation dose from all major sources combined for the most limiting site bom1dary location is listed in Table 4A. These site boundary doses assume a 100 percent occupancy factor, and take no credit for the shielding effect of any residential structure. Table 4B lists the combination of direct radiation doses at the limiting site boundary location and the max,imum offsite dose from gaseous and liquid effluents for the purpose of demonstrating compliance with the dose standards contained in 40 CFR Part 190. For 2016, this annual dose was below the 25 mrem total body and organ limit, as well as the 75 mrem thyroid limit, of 40 CFR Part 190. 3.5 Doses From On-Site Disposal of Septic Waste, Cooling Tower Silt and Soil ODCM Appendices B, F, and I require that all septic waste, cooling tower silt, and sand/soil applied within the approved designated disposal areas be controlled to ensure the dose to a maximally exposed individual during the period of Vermont Yankee site control is limited to less than 1 mrem/year to the whole body and any organ. After the period associated with Vermont Yankee operational control, the dose to the inadvertent intruder is to be limited to 5 mrem/year. The projected dose from on-site disposals of septic waste, cooling tower silt, and sand/soil mixes is given in Appendix J of this report. During 2016 there was no septage sludge, soil or cooling tower silt spread. The last spreading occurred on October 20, 2015. The dose limits applicable to the on-site spreading of materials were met for the dose associated with past spreading activities. 5

3.6 On-Site Recreational Activities During 2016, no access to the on-site boat launching ramp located north of the intake structure was pennitted for employees, their families, and guests. As such, there was no associated dose impact to members of the public due to any recreational activities on-site. 6

REFERENCES

1. Off-Site Dose Calculation Manual (ODCM), Revision 37, Entergy Nuclear Vermont Yankee, LLC, dated December 1, 2016.

2. Regulatory Guide 1.109, "Calculation of Annual Doses to Man From Routine Release of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I," U.S.

Nuclear Regulatory Commission, Office of Standards Development, Revision 1, October 1977.

3. Safety Guide 1.23, "Onsite Meteorological Programs," U.S. Atomic Energy Commission, February 17, 1972.

4. Regulatory Guide 1.111, "Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors," U.S. Nuclear Regulatory Commission, Office of Standards Development, March 1976.

5. Meteorology and Atomic Energy, 1968, Section 5-3.2.2, "Cloud Depletion," page 204, U.S. Atomic Energy Commission, July 1968.

6. Regulatory Guide 1.21, "Measuring, Evaluating, and Reporting Radioactive Material in Liquid and Gaseous Effluents and Solid Waste," U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, Revision 2, June 2009.

7

TABLElA Entergy Nuclear Vermont Yankee Effluent and Waste Disposal Annual Report for 2016 Gaseous Effluents -Summation of All Releases Quarter Quarter Est. Total Unit 1 2 Error,% A. Fission and Activation Gases

1. Total release Ci ND ND +/-2.30E+Ol

2. Average release rate for period µCi/sec ND ND NIA

3. Percent ofODCM limit (1) % ND ND NIA B. Iodines

1. Total Iodine Ci ND ND +/-1.SOE+Ol

2. Average release rate for period µCi/sec ND ND NIA

3. Percent ofODCM limit (3) % (3) (3) NIA

c. Particulates

1. Particulates with T-112>8 days Ci ND ND +/-1.SOE+Ol

2. Average release rate for period µCi/sec ND ND NIA

3. Percent of ODCM limit (3) % (3) (3) NIA

4. Gross alpha radioactivity Ci ND ND NIA D. Tritium (4)

1. Total release Ci 7.57E-02 1.88E-Ol +/-1.SOE+Ol

2. Average release rate for period µCi/sec l.05E-02 2.49E-02 NIA

3. Percent ofODCM limit % 4.85E-04 1.20E-03 NIA E. Carbon-14

1. Total release Ci ND ND NIA

2. Percent ofODCM limit (2) % (3) (3) NIA ND= Not Detected, or in the case ofC-14, no power operations in 2016 leads to a zero estimate ofC-14 production/release.

(1) ODCM Control 3.3.2. for the most limiting of beta air or gamma air dose. Percentage ofODCM limit calculated using Method I dose results. (2) Percentage ofODCM limit calculated using Method I dose results based on the limits ofODCM Control 3.3.3. (3) With respect to the form of Table IA from Regulatory Guide 1.21, Revision 1, any dose contribution from Carbon-14, Iodines, and particulates are included with Tritium in Part D. (4) Tritium released through evaporation from the on-site frac tank is included in these totals. 8

TABLE IA (Continued) Entergy Nuclear Vermont Yankee Effluent and Waste Disposal Annual Report for 2016 Gaseous Effluents - Summation of All Releases Quarter Quarter Est. Total Unit 3 4 Error,% A. Fission and Activation Gases

1. Total release Ci ND ND +/-2.30E+Ol

2. Average release rate for period µCi/sec ND ND NIA

3. Percent ofODCM limit (1) % ND ND NIA B. Iodines

1. Total Iodine Ci ND ND +/-1.80E+Ol

2. Average release rate for period µCi/sec ND ND NIA

3. Percent ofODCM limit (3) % (3) (3) NIA

c. Particulates

1. Particulates with T-1/2>8 days Ci ND ND +/-1.80E+Ol

2. Average release rate for period µCi/sec ND ND NIA

3. Percent of ODCM limit (3) % (3) (3) NIA

4. Gross alpha radioactivity Ci ND ND NIA D. Tritium (4)

1. Total release Ci 2.29E-01 l.18E-01 +/-l.80E+Ol

2. Average release rate for period µCi/sec 2.84E-02 2.28E-02 NIA

3. Percent of ODCM limit % l.47E-03 l.16E-03 NIA E. Carbon-14

1. Total release Ci ND ND NIA

2. Percent ofODCM limit (2) % (3) (3) NIA ND= Not Detected, or in the case ofC-14, no power operations in 2016 leads to a zero estimate ofC-14 production/release.

(1) ODCM Control 3.3.2. for the most limiting of beta air or gamma air dose. Percentage ofODCM limit calculated using Method I dose results. (2) Percentage ofODCM limit calculated using Method I dose results based on the limits ofODCM Control 3.3.3. (3) With respect to the form of Table lA from Regulatory Guide 1.21, Revision 1, any dose contribution from Carbon-14, Iodines, and particulates are included with Tritium in Part D. (4) Tritium released through evaporation from the on-site frac tank is included in these totals . 9

TABLElB Entergy Nuclear Vermont Yankee Effluent and Waste Disposal Annual Report for 2016 Gaseous Effluents - Elevated Releases Continuous Mode Batch Mode (1) Quarter Quarter Nuclides Released Units 1 2 1 2

1. Fission Gases Argon-41 Ci ND ND Krypton-85 Ci ND ND Krypton-85m Ci ND ND Krypton-87 Ci ND ND Krypton-88 Ci ND ND Xenon-133 Ci ND ND Xenon-133m Ci ND ND Xenon-135 Ci ND ND Xenon-135m Ci ND ND Xenon-138 Ci ND ND Unidentified Ci ND ND Total for Period Ci ND ND (1) (1)

2. Iodines Iodine-131 Ci ND ND Iodine-133 Ci ND ND Iodine-135 Ci ND ND Total for Period Ci ND ND (1) (1)

3. Particulates Strontium-89 Ci ND ND Strontium-90 Ci ND ND Cesium-134 Ci ND ND Cesium-137 Ci ND ND Barium-Lanthanum-140 Ci ND ND Manganese-54 Ci ND ND Chromium-51 Ci ND ND Cobalt-57 Ci ND ND Cobalt-58 Ci ND ND Cobalt-60 Ci ND ND Cerium-141 Ci ND ND Zinc-65 Ci ND ND Total for Period Ci ND ND (1) (1)

ND Not Detected at the plant stack (1) There were no batch mode gaseous releases for this reporting period. 10

TABLElB (Continued) Entergy Nuclear Vermont Yankee Effluent and Waste Disposal Annual Report for 2016 Gaseous Effluents -Elevated Releases Continuous Mode Batch Mode (1) Quarter Quarter Nuclides Released Units 3 4 3 4

1. Fission Gases Krypton-85 Ci ND ND Krypton-85m Ci ND ND Krypton-87 Ci ND ND Krypton-88 Ci ND ND Xenon-133 Ci ND ND Xenon-133m Ci ND ND Xenon-135 Ci ND ND Xenon-135m Ci ND ND Xenon-138 Ci ND ND Unidentified Ci ND ND Total for Period Ci ND ND (1) (1)

2. Iodines Iodine-131 Ci ND ND Iodine-133 Ci ND ND Iodine-135 Ci ND ND Total for Period Ci ND ND (1) (1)

3. Particulates Strontium-89 Ci ND ND Strontium-90 Ci ND ND Cesium-134 Ci ND ND Cesium-137 Ci ND ND Barium-Lanthanum-140 Ci ND ND Manganese-54 Ci ND ND Chromium-51 Ci ND ND Cobalt-58 Ci ND ND Cobalt-60 Ci ND ND Cerium-141 Ci ND ND Cerium-144 Ci ND ND Zinc-65 Ci ND ND Total for Period Ci ND ND (1) (1)

ND Not Detected at the Plant Stack (1) There were no batch mode gaseous releases for this reporting period. 11

TABLE IC Entergy Nuclear Vermont Yankee Effluent and Waste Disposal Annual Report for 2016 Gaseous Effluents - (Routine) Ground Level Releases <2> Continuous Mode Batch Mode Quarter Quarter Nuclides Released Units 1 (1) 2 (1) 1 (1) 2 (1)

1. Fission Gases Krvoton-85 Ci Krypton-85m Ci Krypton-87 Ci Krvoton-88 Ci Xenon-133 Ci Xenon-135 Ci Xenon-135m Ci Xenon-138 Ci Unidentified Ci Total for Period Ci

2. Iodines Iodine-131 Ci '

Iodine-133 Ci Iodine-I 35 Ci Total for Period Ci

3. Particulates Strontium-89 Ci Strontium-90 Ci Cesium-134 Ci Cesium-137 Ci Barium-Lanthanum-140 Ci Manganese-54 Ci Chromium-51 Ci Cobalt-58 Ci Cobalt-60 Ci Cerium-141 Ci Zinc-65 Ci Iron-55 Ci Total for Period Ci (1) There were no routine ground level gaseous releases for this reporting period.

(2) No radioactively contaminated used oil was burned during 2016. 12

TABLElC (Continued) Entergy Nuclear Vermont Yankee Effluent and Waste Disposal Annual Report for 2016 Gaseous Effluents - (Routine) Ground Level Releases(2) Continuous Mode Batch Mode Quarter Quarter Nuclides Released Units 3 (1) 4 (1) 3 (1) 4 (1)

1. Fission Gases Krypton-85 Ci Krypton-85m Ci Krypton-87 Ci Krypton-88 Ci Xenon-133 Ci Xenon-135 Ci Xenon-135m Ci Xenon-138 Ci Unidentified Ci Total for Period Ci

2. Iodines Iodine-131 Ci Iodine-133 Ci Iodine-135 Ci Total for Period Ci

3. Particulates Strontium-89 Ci Strontium-90 Ci Cesium- 134 Ci Cesium-137 Ci Barium-Lanthanum- 140 Ci Manganese-54 Ci Chromium-51 Ci Cobalt-58 Ci Cobalt-60 Ci Cerium-141 Ci Zinc-65 Ci Iron-55 CI Total for Period Ci (1) There were no ground level gaseous releases for this reporting period.

(2) No radioactively contaminated used oil was burned during 2016. 13

TABLElD Entergy Nuclear Vermont Yankee Effluent and Waste Disposal Annual Report for 2016 Gaseous Effluents -Non-routine Releases Quarter Quarter Nuclides Released Units 1(1) 2(1) 3(1) 4(1)

1. Fission Gases Krypton-85 Ci Krypton-85m Ci Krypton-87 Ci Krypton-88 Ci Xenon-133 Ci Xenon-135 Ci Xenon-135m Ci Xenon-138 Ci Unidentified Ci Total for Period Ci

2. Iodines Iodine-131 Ci Iodine-133 Ci Iodine-I 35 Ci Total for Period Ci

3. Particulates Strontium-89 Ci Strontium-90 Ci Cesium-134 Ci Cesium- 137 Ci Barium-Lanthanum-140 Ci Manganese-54 Ci Chromium-51 Ci Cobalt-58 Ci Cobalt-60 Ci Cerium-141 Ci Zinc-65 Ci Iron-55 Cl Total for Period Ci (1) There were no non-routine ground level gaseous releases for this reporting period.

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TABLE2A Entergy Nuclear Vermont Yankee Effluent and Waste Disposal Annual Report for 2016 Liquid Effluents - Summation of All Releases Est. Total Units Quarter 1 Quarter 2 Error,% A. Fission and Activation Products

1. Total Release not includin tritium, ha Ci ND ND N/A

2. Ci/ml ND ND

3. % ND ND B. Tritium

1. Total Release Ci 7.06E-03 6.65E-03 +/-2.00E+01 I

2. Averaoe Diluted Concentration.Durino Period uCi/ml 1.30E-06 1.23E-06

3. Percent of Applicable Limit (1) % 1.09E-04 1.04E-04 C. Dissolved and Entrained Gases

1. Total Release Ci ND ND N/A I

2. Averaoe Diluted Concentration Durino Period uCi/ml ND ND

3. Percent of Aoolicable Limit % ND ND D. Gross Al ha Radioactivi

1. Total Release Ci ND ND N/A E. Volume of Waste Release (prior to dilution) Liters (2) (2) N/A IF. Volume of Dilution Water Used During Period Liters 3.89E+06 3.89E+06 (3)

ND Not detected in liquid effluents. (1) The percent of limit is based on the ODCM Control 3.2.2 limiting dose (1.5 mrem/quarter to the total body) from liquid effluents and is related to the abnormal leakage of tritiated plant water into the underground environment. The percent of the concentration limits specified in Appendix B to 10CFR20.1001 - 20.2402, Table 2, Column 2 (ODCM Control 3. 2.1) were estimated to be 0.13%, 0.12%, 0.12%, and 0.12% for the first, second, third, and fourth quarters, respectively. (2) Leakage of contaminated plant water to subsurface areas was stopped in February 2010. The release of contaminated ground water to the Connecticut River is based on site boundary monitoring well data collected during 2016. (3) Dilution due to groundwater flow through the affected subsurface plume area toward the Connecticut River was estimated to be 7.83 gpm (or 3.89E+06 liters per quarter) during 2016. An estimated total error is not applicable. 15

TABLE2A (Continued) Entergy Nuclear Vermont Yankee Effluent and Waste Disposal Annual Report for 2016 Liquid Effluents - Summation of All Releases Est Total Units Quarter 3 Quarter4 Error,% A. Fission and Activation Products

1. Total Release (not includinq tritium, oases, aloha) Ci ND ND NIA I

2. Averaqe Diluted Concentration Durinq Period uCi/ml ND ND

3. Percent of Aoolicable Limit (1) % ND ND B. Tritium

1. Total Release Ci 6.35E-03 6.19E-03 +/-2.00E+01 I

2. Averaqe Diluted Concentration Durinq Period uCi/ml 1.16E-06 1.21 E-06

3. Percent of Applicable Limit (1) % 9.93E-05 9.67E-05 C. Dissolved and Entrained Gases

1. Total Release Ci ND ND NIA I

2. Averaoe Diluted Concentration Durino Period uCi/ml ND ND

3. Percent of Aoolicable Limit % ND ND D. Gross Al ha Radioactivi

1. Total Release Ci ND ND NIA E. Volume of Waste Release rior to dilution Liters NIA F. Volume of Dilution Water Used During Period Liters 3.89E+06 3.89E+06 (3)

ND Not detected in liquid effluents. The percent of limit is based on the ODCM Control 3.2.2 limiting dose (1.5 mrem/quarter to the total body) from liquid effluents and is related to the abnormal leakage of tritiated plant water into the underground environment. (1) The percent of the concentration limits specified in Appendix B to 10CFR20.1001 - 20.2402, Table 2, Column 2 (ODCM Control 3. 2.1) were estimated to be 0.13%, 0.12%, 0.12%, and 0.12% for the first, second, third, and fourth quarters, respectively. Leakage of contaminated plant water to subsurface areas was stopped in February 2010. The release (2) of contaminated ground water to the Connecticut River is based on site boundary monitoring well data collected during 2016. Dilution due to groundwater flow through the affected subsurface plume area toward the Connecticut (3) River was estimated to be 7.83 gpm (or 3.89E+06 liters per quarter) during 2016. An estimated total error is not applicable. 16

TABLE2B Entergy Nuclear Vermont Yankee Effluent and Waste Disposal Annual Report for 2016 Liquid Effluents - Routine Releases Continuous Mode Batch Mode I Nuclides Released Units Quarter 1 I Quarter 2 Quarter 1 I Quarter 2 Strontium-89 Ci - - - - Strontium-90 Ci - - - - Cesium-134 Ci - - - - Cesium-137 Ci - - - - lodine-131 Ci - - - - Cobalt-58 Ci - - - - Cobalt-60 Ci - - - - lron-59 Ci - - - - Zinc-65 Ci - - - - Manaanese-54 Ci - - - - Zirconium-Niobium-95 Ci - - - - Molybdenum-99 Ci - - - - Technetium-99 Ci - - - - Barium-Lanthanum-140 Ci - - - - Cerium-141 Other (specify) Ci Ci Ci Unidentified I Ci I Total for Period (above) Ci Xe-133 Ci Xe-135 Ci ND Not detected in liquid effluents. Dash indicates no release of this type. 17

TABLE2B (Continued) Entergy Nuclear Vermont Yankee Effluent and Waste Disposal Annual Report for 2016 Liquid Effluents - Routine Releases Continuous Mode Batch Mode I Nuclides Released Units Quarter 3 I Quarter 4 Quarter 3 I Quarter 4 Strontium-89 Ci - - - - Strontium-90 Ci - - - - Cesium-134 Ci - - - - Cesium-137 Ci - - - - lodine-131 Ci - - - - Cobalt-58 Ci - - - - Cobalt-60 Ci - - - - lron-59 Ci - - - - Zinc-65 Ci - - - - Mani:ianese-54 Ci - - - - Zirconium-Niobium-95 Ci - - - - Molybdenum-99 Ci - - - - Technetium-99 Ci - - - - Barium-Lanthanum-140 Ci - - - - Cerium-141 Other (specify) Ci Ci Ci Unidentified I Ci I Total for Period (above) Ci Xe-133 Ci Xe-135 Ci ND Not detected in liquid effluents. Dash indicates no release of this type. 18

Table.3 Entergy Nudear Verrn_ont'Yarikee Efflue1'lt arid'Wa~~l'! Dls1iosal Annual Report First and Second Quarters for 2016 Solid Waste *and Irradiated Fuel Shipments A. -Solid W~ste Shipped Off-5ite for Surra I er Disp0i9I (rrct irradiated fuel)*

            -l.         Type o;tf Waste t-'-S~n~lp~pe~d~**~r.~~~m~\/V~~fo~r~B~u~rl~*l'--~~~~~~~~~~~-1-~~U~ni~t~-+~~~*~O.~u~a~rt~e~rs~*1;;;....;;;&~2=--~~-J-~--'E~st~"~T~o~t*~l~E~tr~o~r~~~*~--i'
                *.Scent resin< ftlter&luifo:es* et~.                                          m'                    '.9Ji7 E~oo                             +/-25%

Cl *.2..2.1 E:>Ol !:25%

              ~h;Drv Compre.sslbJe Waste        eauioment etc.                                m *                                                          +/-253

!:25.'Xt c:. J~diata.d ~ampl)nc:nt=~ control rod::-, ate, Ci

                                                                                                                                                            +/-25%
               *d.otn~r (Olfl
                                                                                                                                                           .+/-25%

Unit Quarters l*& 2. Est Total £rror %

               *a.spsnt resins, filter sludges, Mc~

er b.Dr\i O:lmpressible Waste, eciuipment, etc. 6.W E+Ol +/-2.5% 133£-02

c. Irradiated c.ornpone-ntS. control rods.*et-c:. m' Ci

                >tOthsr(Oll)                                                                                          l.i2 E+OO er                      3.75 E*04

Z. Estimate of Major Nuclide*composillon {By Type of Waste) ..

SpAnt resim., fHtiu slud~~s Dry' co111presslb~ Waste, l!quipment, lrraqi~*componi>nt$, control rod~, Othl11r Wa'ilB etc. etc. . Nuclide* Percentlll Nuclide Percentill Nuclide Percentlll Nuclide Percentf1\ C-H 3.13% *ca-60 62:24% Fa-55 48.44% Co-60 62.72% Mn-54 '1.45% Fe-55 20.87% Co-60 42.00% Fe-55 20.74% Fe-55 20.27% Zn-65 :12.09% Ni-63 5.35% z:n-6s 11.97,. Co-60 41.6491, Mn*.54 3A9!'>1, Ta-182 *t.B6% Mo-54 3.4891. Ni-63 *z:UJ53 CD*5o 0.77% Mn*.54 "l.73~' c.,:~s 0.13% Zri-65 7.40% Cs'l37 0.23% Cs-137 0;39% C~*137 o.23% C:**t<i7 2.9"% 'Tc-9!) .o.1s% *rc-99 O.J;\!~. (1) lncludes only these nuclidesthat *re.greater than 0.1% of the total ad;1vity 3, Disposition of SolidWasre*shipmenrs (i' :&2""Q.uaners) No. of Shio111ents From VY f!' rom ProcMs:or Mode Ta Processor* To Burial Resin~WCS2 2 0 Truck .0 2 DAW-BC02 .2 *o Truck 2 0 DAW-" Clive UT 9 .0 9 Truck 0 9 lrrHdw2* :z 0 Trhck 0 2: Other2 .0 2 Tl'llc!( 0 2 B. lm1d.l;ited F1telS~lpments(Dispoiltion): None C. Arldition*I Oat~ (1'1& 2"'0.u~rl:i!rs) Suppletnental Information. VY toPtocessor VY ta Burial Proces.ors ta Burial

     *crass of Solid Waste Shipped                                                         AU                          AU,C                             AU Typ@ of Containers u..,,9                                                            GOC                    Ty?" A, Tvl"' B              GC-C, 11/p" {'., T\'I"' B Solidmtaticn i'.gerot:or ~,brot!>.o.it Uaed                                         NonB                         '.None                          none l:;R.-:: Gallah~l'!l.o;,d i!CO*" l>wrc.r.,eli*Opaio:ti~ni
             ¥.l-CS:= -ritpS~e Co,ntli"olSpecP!~fzts
           *~$= E"l>r;;*r:Sr.>k1t!o0t<.

Goi:::: """"'"' O,,,;i*g1t Contaim,r 19

Table..3 E.ntergy Nuclear Vermont Yarik.;;oe tffh:1ent and'Wa5t~.Dl:;posal Armual Report Third and FOurth Quartets for2016* Solid Waste and Irradiated Fuel Shipments A .S?liilWaste Shipped *Off-Site tor.auria) or.O~p=I inm_lr.rndi~fed Fuel), 1; Type<!fW;ute Shippad frum VYfot Burial Unit Est. Tote! Ertar% ci 1,62 Ef.OO

                  "'Dry Compressible Waste,.equlpment, etc*.

Cl-C:.Jnadiatetfcomponents;.control rods, etc~- Cl d.OtheT {6iO Shipped from Pr<ita:;sol{*) for Burial Unit m ouarters3 & 4 1.3.!i E+Oi +/-:25% b;I)'ry Cnmpras:sible *wasle;:aquiprrient, Btc. 7.17 E+OO 1:25% Ci ci i:1

2. _Estimate.of Major NucJi\Je C~mpcisition jSy i:vpe '?f Waste!

Dr,;.Cnmpr;,..lbt;.W"8te,.equipment, lrradfa1t.ed comp,011e-rit~* .toOtro\ rods. . -~ OtherWaste etc .. *'Ste> Pe11:ent(l) Nudide Perci:ntjl) Nuclide P,eri;entll)

Nuclld_il

  .C-14                        D36%                        Co,60                 :52..54%

2;749~ r...s.s la.71.%

                             *1s,00%                     :znc6s                  12.00%

1.59% Mn-54 355% 60.S9% cco:.sa'. tl.80% Ni-1>3 3,71% °"'137 U;i3%.

  -Zn-65

('l) 'iricludes orilythose:iiuclid1>< that aregmat!lrth~ri.0.1~* of.the tqtalactiVity 3; Disp¢1ti<>~ Pl'iolid _WaMe S~ir"1""ts (3'" & 4~" QJ;s11ersJ No, of_ S!iipmentx *framvV From ProeessOr Mods Io Proc,.ssor .Tollurlsl Resin -wcs -.1 -1 Q_ Trutk :o. l Recin-.cf!Ve lff"-'a 0 3 Truck 0 3 DAW-*4 a 4 Tr.uck *O 4* Truck Truck B. . ltn1fiia!:..d Fuel Shipms_nt_s .(Disposilfrm): Non.,. c, Additional Oat* {l" &-2"" Quaitei<<;) none

              'Gil'= Gallabar limi!
              ;lll;'O= !le~rC...,,,kO~tioJ\~
               \fJCS.*:::*v~fa:st@ C::m:t;,~f '5pEc'fa1fm l."S.F~~r.ie.rgy so~r.umo~-s
              *5m: -=~oo;,,.l!Jeiign,Coiitalner 20

TABLE4A Entergy Nuclear Vermont Yankee Maximum+ Quarterly and Annual Off-Site Doses from Direct Radiation and Liquid and Gaseous Effluents for 2016 (10CFR50, Appendix I) Dose (mrem)(a) 1st 2nd 3rd Source uarter Quarter Quarter Total Body Dose 1.64E-06 l .56E-06 1.49E-06 1.45E-06 6.13E-06 Footnotes (c) (c) (c) (c) (c) Organ Dose l.64E-06 l.56E-06 1.49E-06 Footnotes (c) (c) (c) Beta Air (mrad) Footnotes (d) (d) (d) (d) Gamma Air (mrad) Footnotes (d) (d) 0.72 0.71 0.71 2.86 (e)

"Maximum" means the largest fraction of the corresponding 10CFR50, Appendix I dose design objective.

(a) The lettered footnotes indicate the age group, organ, and location of the dose receptor, where appropriate. (b) The yearly dose is the sum of the doses for each quarter, or a full annual assessment. (c) The critical age group/organ for the Maximum Exposed Individual (MEI) is the Adult/Total Body and all organs (except Bone) from the release ofH-3 to groundwater. (d) There were no noble gas releases in this quarter. (e) Maximum direct dose point located on the old west site boundary, approximately 208 meters from the Turbine Building. (t) The critical age group/organ for the MEI is the Child/all organs (except Bone), at a location NW, 2600 meters from the stack. (g) The critical age group/organ for the MEI is the Child/all organs (except Bone), at a location WSW, 2400 meters from the stack. 21

TABLE4B Entergy Nuclear Vermont Yankee Maximum.. Annual Off-Site Doses from Direct Radiation and Liquid and Gaseous Effluents for 2016 (40CFR190) Total Body Maximum Organ Thyroid Pathway (mrem) (mrem) (mrem) Direct External (a) (b) 2.86 2.86 2.86 Liquids (c) 6.13E-06 6.13E-06 6.13E-06 Gases (c) 1.84E-04 1.84E-04 1.84E-04 Annual Total (d) 2.86 2.86 2.86

The location of the projected maximum individual doses from combined direct radiation plus liquid and gaseous effluents correspond to residences at the southwest boundary relative to the Turbine Hall.

(a) No residential shielding credit or occupancy time fraction (i.e., occupancy is assumed to be 100%) is used. Expected direct external radiation doses would be reduced by approximately 54% with a realistic residential shielding credit and occupancy time (i.e., by using a 0.7 shielding factor from Regulatory Guide 1.109 (Reference 2) and an annual occupancy time of 6760 hours). (b) The direct dose reported here was calculated using the current ODCM methodology and represents the dose to the former nearest residence, which was located in the South sector at 385 meters from the stack prior to the vacancy of this residence in 2008 and the purchase of land by Vermont Yankee. (c) Maximum dose to any organ over all age groups for each release. (d) Annual dose limits contained in 40 CFR Part 190 are 25 mrem to the total body and any organ, and 75 mrem to the thyroid for any real member of the public. 22

TABLE4C Receptor Locations Entergy Nuclear Vermont Yankee Nearest Milk Sector Site Boundary (t) Nearest ResidentC 2> Animal <2>C3> (meters) (meters) (meters) N 400 1400 -- NNE 350 1384 5520 (cows) NE 350 1255 -- ENE 400 966 -- E 500 933 -- ESE 700 1915 -- SE 750 1963 6670 (cows) SSE 850 2044 -- s 385 644 -- SSW 300 451 -- SW 250 418 -- WSW 250 451 9730 (cows) w 300 628 820 (cows) WNW 400 1062 -- NW 550 2253 -- NNW 550 1738 -- (1) Site boundary locations taken from Table 6.10.2 of the ODCM. (2) The location(s) given are based on information from the Vermont Yankee 2016 Land Use Census and Table 7.1 of the ODCM and are relative to the plant stack. Gardens are assumed to be present at all resident locations. (3) Although milk collection has been discontinued due to the permanently shutdown and defueled status of the plant, receptor locations were conservatively retained in the dose analysis. 23

TABLE4D Usage Factors for Environmental Pathways Entergy Nuclear Vermont Yankee* e Fish Potable Veg. Leafy Veg. Milk Meat Inhalation Group (kg/yr) Water (kg/yr) (kg/yr) (I/yr) (kg/yr) (m3/yr) (l/vr) Adult 21 730 520 64 310 110 8,000 Teen 16 510 630 42 400 65 8,000 Child 6.9 510 520 26 330 41 3,700 Infant 0 330 0 0 330 0 1,400

Regulatory Guide 1.109, Table E-5 (Reference 2).

24

TABLE4E Environmental Parameters for Gaseous Effluents

Entergy Nuclear Vermont Yankee I Vegetables Cow Milk Goat Milk Meat r=

I I I I I I

Variable Stored Leafy Pasture Stored Pasture Stored Pasture Agricultural Productivity 2 YV 2 2 0.70 2 0.70 2 0.70 (kg/m2) p 240 240 240 240 240 240 240 240 Soil Surface Densitv (k!!/m2) T Transport Time to User (hrs) -- -- 48 48 48 48 480 480 TB Soil Exposure Time<*>(hrs) 131,400 131,400 131,400 131,400 131,400 131,400 131,400 131,400 TE Crop Exposure Time to Plume 1,440 1,440 720 1,440 720 1,440 720 1,440 (hrs) TH Holdup After Harvest (hrs) 1,440 24 0 2,160 0 2,160 0 2,160 QF Animals Dailv Feed (kg/dav) -- -- 50 50 6 6 50 50 FP Fraction of Year on Pasture -- -- (b) -- (b) -- (b) -- FS Fraction Pasture Feed When on -- -- 1 -- 1 -- 1 -- Pasture<cl Note: Footnotes on following page. 25

TABLE 4E (Continued) Environmental Parameters for Gaseous Effluents Entergy Nuclear Vermont Yankee Vegetables Cow Milk Goat Milk Meat Variable Stored Leafy Pasture Stored Pasture Stored Pasture Stored FG Fraction of Stored Vegetables 0.76 -- -- -- -- -- -- -- Grown in Garden FL Fraction of Leafy Vegetables -- 1.0 -- -- -- -- -- -- Grown in Garden FI Fraction Elemental Iodine= 0.5 -- -- -- -- -- -- -- -- H Absolute Humidity= 5.6<d> -- -- -- -- -- -- -- --

From VY ODCM, Table 6.9.l (Reference 1).

(a) For Method II dose/dose rate analyses of identified radioactivity releases of less than one year, the soil exposure time for that release may be set at 8,760 hours (one year) for all pathways. (b) For Method II dose/dose rate analyses performed for releases occurring during the first or fourth calendar quarters, the fraction of time animals are assumed to be on pasture is zero (non-growing season). For the second and third calendar quarters, the fraction of time on pasture (FP) will be set at 1.0. FP may also be adjusted for specific farm locations if this information is so identified and reported as part of the land use census. (c) For Method II analyses, the fraction of pasture feed while on pasture may be set to less than 1.0 for specific farm locations if this information is so identified and reported as part of the land use census. (d) for all Method II analyses, an absolute humidity value equal to 5.6 (gm/m3) shall be used to reflect conditions in the Northeast (

Reference:

Health Physics Journal, Volume 39 (August), 1980; Pages 318-320, Pergammon Press). 26

TABLE4F Environmental Parameters for Liquid Releases (Tritium) Via Groundwater Entergy Nuclear Vermont Yankee Variable Potable Water Aquatic Food Stored Veg. Leafy Veg. Meat Cow Milk Name (Units) Mixing Ratio 5.94E-06 1.27E-03 5.94E-06 5.94E-06 5.94E-06 5.94E-06 Transit Time (hrs)* 12 24 0 0 0 0 Water Uptake** (animal) (L/day) -- -- -- -- 50.0 60.0 Feed Uptake** (animal) (kg/day) -- -- -- -- 50.0 50.0

Values are from Regulatory Guide 1.109, Table E-15 (Reference 2)
- Values are from Regulatory Guide 1.109, Table E-3 (Reference 2) 27

TABLE SA VERMONT YANKEE JAN 16 - DEC 16 METEOROLOGICAL DATA JOINT FREQUENCY DISTRIBUTION 3S. 0 FT WIND DATA STABILITY CLASS A CLASS FREQUENCY (PERCENT) = 2. 42 WIND DIRECTION FROM SPEED N NNE NE ENE E ESE SE SSE SSW SW WSW w WNW NW NNW VRBL TOTAL MPH CALM 3 1 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 4 (1) 1. 4 9 .so .00 . 00 .oo

00 .00
00 .oo . 00 .00 .00
00 .00 .00 .00 .00 1. 99 (2) .04
01 .00 .00 .00
00 .00 .00 .00 .oo .oo .00 .00 .oo .oo .oo .oo .OS C-3 4 0 2 1 2 2 0 0 0 0 1 4 0 0 0 2 0 18 (1) 1. 99 .00 1. 00 .so 1. 00 1. 00 .oo .oo .00 .00 .so 1. 99 .oo .oo .00 1.00 .00 8. 96 (2)
OS .oo .02
01 .02 . 02 .00 .oo .oo .00 .01 .OS .oo .oo .oo .02
00 .22 4-7 13 5 7 3 14 9 6 3 1 0 0 2 2 0 3 11 0 79 (1) 6. 47 2. 49 3. 48 1. 49 6. 97 4. 48 2.99 1. 49 . so .oo .00 1. 00 1.00 .00 1. 49 s. 47 .oo 39.30 (2) .16 .06
08
04 .17 .11
07
04
01 .00 .00 .02 .02 .00 .04 .13
00 .9S 8-12 9 1 0 0 s 13 s 13 8 2 0 1 2 1 7 22 0 89 (1) 4. 48 .so .00
00 2. 49 6. 47 2. 49 6. 47 3. 98 1.00 .00 .so 1.00 .so 3. 48 10. 9S
00 44.28 (2) .11 .01 .oo .00 .06 .16 .06 .16 .10 .02 .00 .01 .02 .01 .08 .27 . 00 1. 07 13-18 0 0 0 0 0 1 0 2 1 0 0 0 0 0 1 s 0 10 (1) .oo .00 .00
00
00 .so .oo 1. 00 . so
00 .oo .00 .00 .00 .so 2. 49 .00 4. 98 (2) .oo .00 .00
00
00 .01 .oo .02 .01
00
00 .oo .00 .00 .01 .06 . 00 .12 19-24 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 (1) .oo .oo .00 .oo
00 . 00 .00 .oo
00
00 . 00 .oo .oo .oo .00 .00
00 .00 (2) .oo .00 .00 .oo
00 .00 .00 .oo .00
00 . 00 .oo .00 .00 .00 .00 . 00 .00 GT 24 0 0 0 0 0 0 0 0 0 0 0 0 0 1 0 0 0 1 (1) .oo
00 .00 .oo
00 .00 . 00 .oo .00 .00
00 .00 .oo .so .00 .00 .00 .so (2) .oo .00 .00 .oo .00 .00
00 .oo .00 .00 .oo .oo .oo .01 .00 .00 .oo .01 ALL SPEEDS 29 7 9 4 21 2S 11 18 10 2 1 7 4 2 11 40 0 201 (1) 14. 43 3. 48 4. 48 1. 99 10. 4S 12. 44 s. 47 8. 96 4.98 1. 00 .so 3. 48 1. 99 1. 00 S.47 19. 90 .00 100. 00 (2) .3S .08 .11 .OS . 2S .30 .13
22 .12 . 02 .01
08 .OS
02 .13 .48 .oo 2. 42

( 1) =PERCENT OF ALL GOOD OBSERVATIONS FOR THIS PAGE (2)=PERCENT OF ALL GOOD OBSERVATIONS FOR THIS PERIOD C= CALM (WIND SPEED LESS THAN OR EQUAL TO . 9S MPH) 28

TABLE SB VERMONT YANKEE JAN 16 - DEC 16 METEOROLOGICAL DATA JOINT FREQUENCY DISTRIBUTION

35. 0 FT WIND DATA STABILITY CLASS B CLASS FREQUENCY (PERCENT) = 3.47 WIND DIRECTION FROM SPEED N NNE NE ENE E ESE SE SSE SSW SW WSW w WNW NW NNW VRBL TOTAL MPH CALM 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 (1) .00 .oo .00 .00 .00 .oo

00 . 00 .00
00 .oo .oo . 00 .00
00 .00 .oo .oo (2) .00 .00 .oo .00 .00 .00
00 .oo .00 . 00 .oo .oo .00 .00 . 00 .00 .oo .oo C-3 2 3 4 4 1 0 0 0 0 0 0 0 2 1 0 19 (1) .69 1. 04 .69 1.39 1. 39 .35
00 .00
00 .oo
00 .oo .00 .00 .69
35 .oo 6. 60 (2) .02 .04 .02 .05
05
01 . 00 .00 .00 .oo .00 .00 .00 .00 .02 .01 .oo .23 4-7 27 6 11 7 13 23 15 7 5 2 0 0 3 2 B 21 0 150 (1) 9. 3B 2. OB 3. B2 2.43 4. 51 7. 99 5. 21 2. 43 1. 74 .69 . 00 .00 1. 04 .69 2. 7B 7. 29 .oo 52. OB (2) .33 . 07 .13
OB .16 . 2B
lB .OB .06 .02 . 00 .00
04
02 .10
25 .00 1. Bl B-12 9 5 0 0 2 5 B 20 7 1 0 2 3 4 5 20 0 91 (1) 3.13 1. 74 .00 .00 .69 1. 74 2. 7B 6. 94 2. 43 .35
00 .69 1. 04 1.39 1. 74 6. 94 . 00 31. 60 (2) .11 .06 .00 .00
02
06 .10
24
OB
01 .oo .02
04 .05 .06
24 .oo 1.10 13-lB 1 0 0 0 0 0 0 0 1 0 0 0 4 2 7 10 0 25 (1) .35 . 00 .00 .oo
00
00 .00 .oo .35 .00
00 . 00 1. 39 .69 2. 43 3. 47 .00 B. 6B (2) .01 . 00 .00 .oo . 00 .00
00 .oo
01 .00
00 .00 .05 .02
OB .12 .00 .30 19-24 0 0 0 0 0 0 0 0 0 0 0 0 0 0 2 1 0 3 (1) .00
00 .oo .oo . 00
00 .oo
00 .00 .00 .oo .oo .oo .oo .69 .35 . 00 1. 04 (2) .00 .oo .00 .00 . 00
00 .oo
00 .00 .00 .oo
00 .00 .00
02 .01 .00
04 GT 24 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 (1) .00 .00 .00
00 .00
00 .oo .00 .00 .00 .00 .00 .00 .oo .oo .oo .00 .00 (2) .oo .oo .00 .00 .oo .00 .00 .00 .00 .oo .oo .00 .00 .00 .00 .00 .00 .oo ALL SPEEDS 39 14 13 11 19 29 23 27 13 3 0 2 10 B 24 53 0 2BB (1) 13.54 4. B6 4.51 3. B2 6. 60 10. 07 7. 99 9. 3B 4.51 1. 04 .00
69 3. 47 2. 7B B. 33 lB. 40 .oo 100. 00 (2) .47 .17 .16 .13 .23
35 . 2B .33 .16 . 04 .oo .02 .12 .10 .29 . 64 .00 3.47

( 1 ) =PERCENT OF ALL GOOD OBSERVATIONS FOR THIS PAGE (2)=PERCENT OF ALL GOOD OBSERVATIONS FOR THIS PERIOD C= CALM (WIND SPEED LESS THAN OR EQUAL TO

95 MPH) 29

TABLE SC VERMONT YANKEE JAN 16 - DEC 16 METEOROLOGICAL DATA JOINT FREQUENCY DISTRIBUTION

35. 0 FT WIND DATA STABILITY CLASS C CLASS FREQUENCY (PERCENT) 5. 09 WIND DIRECTION FROM SPEED N NNE NE ENE E ESE SE SSE SSW SW WSW w WNW NW NNW VRBL TOTAL MPH CALM 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 o o o (1) .oo .00 .00

00 .oo .00 . 00 .oo .00 .00 .00
00 .oo .oo .oo .oo .oo .oo (2) .00 .00 .oo
00 .00 .00
00 .00 .00 .00 .oo
00 .oo .00 .00 .00 .00 .oo C-3 3 B 5 4 B 6 2 2 o 2 1 2 1 1 3 4 o 52

( 1)

71 1. 90 l. lB
95 1. 90 1. 42 .47
47
00 .47 . 24
47
24 . 24 . 71
95 .oo 12. 32 (2)
04 .10 .06 .05 .10 .07 .02
02
00 . 02 .01 .02 .01
01
04 .05 .oo .63 4-7 30 B 6 12 27 34 21 12 11 6 3 l l 7 13 26 o 21B (1) 7 .11 1. 90 1. 42 2. B4 6. 40 B. 06 4. 9B 2. B4 2.61 1. 42 . 71
24 .24 1. 66 3. OB 6.16 .oo 51. 66 (2) .36 .10 .07 .14 .33 . 41 .25 .14 .13 . 07 .04 .01 .01 .OB .16 .31
00 2.63 B-12 14 4 o 0 o 7 2 15 13 6 o 2 5 10 13 19 0 110 (1) 3.32 .95 .00
00
00 1. 66
47 3. 55 3.0B 1. 42 .oo
47 l.18 2.37 3. 08 4.50 .00 26.07 (2) .17 . 05 .00 . 00 . 00
OB
02 .lB .16 .07 .oo
02 .06 .12 .16 .23 .oo l. 33 13-lB o o o o o l o l 1 o o o 5 7 12 11 o 3B (1) .00 .00 .oo .oo .00
24 .oo . 24
24 .00 .oo .oo l. lB 1. 66 2. B4 2.61 .00 9. 00 (2) .00 .00 .00 .oo .oo
01 .00 . 01
01 .00 .00 .00 .06
OB .14 .13 .00 . 46 19-24 o o o o o o o o o 0 o o o l 2 o o 3 (1) .oo .00 .00 .00 .oo
00 .00 .00 .00
00
00 .00 .00
24
47 .oo .00 . 71 (2) .oo .00 .00 .oo .00 . 00 .oo .oo
00
00
00 .00 .oo
01 .02 .00 .oo . 04 GT 24 o o o o o o o o o o 0 o o 1 o o o 1 (1) .00 .00 .00 .00 .00 .00 .00 .00 .00 .00
00 .00 .00
24 .oo .00 .oo .24 (2) .oo .oo .00 .00 .oo
00 .00 .00 .00 .oo .oo .00 .00 .01 .00 .oo .00 .01 ALL SPEEDS 47 20 11 16 35 4B 25 30 25 14 4 5 12 27 43 60 o 422 (1) 11.14 4. 74 2. 61 3. 79 B. 29 11. 37 5. 92 7 .11 5. 92 3.32 .95 l. lB 2 .B4 6. 40 10.19 14. 22 .00 100. 00 (2) .57 .24 .13 .19
42 .5B
30 .36 .30 .17
05 .06 .14 .33 .52
72 . 00 5. 09

( 1) ~PERCENT OF ALL GOOD OBSERVATIONS FOR THIS PAGE (2) ~PERCENT OF ALL GOOD OBSERVATIONS FOR THIS PERIOD c~ CALM (WIND SPEED LESS THAN OR EQUAL TO .95 MPH) 30

TABLE SD VERMONT YANKEE JAN 16 - DEC 16 METEOROLOGICAL DATA JOINT FREQUENCY DISTRIBUTION 3S. 0 FT WIND DATA STABILITY CLASS D CLASS FREQUENCY (PERCENT) 4B. 20 WIND DIRECTION FROM SPEED N NNE NE ENE E ESE SE SSE s SSW SW WSW w WNW NW NNW VRBL TOTAL MPH CALM 0 1 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 1 (1) .00 .03 .oo

00
00
00 .00 .00 .00 .oo
00 .oo .00 .00
00 . 00 .00 .03 (2) .oo .01 .00 .00 . 00
00 .00 .oo .oo
00 .00 .00 .00 .00 . 00
00 .oo .01 C-3 9S Sl 4S 39 44 66 B2 60 44 47 3S 39 44 44 79 104 0 91B (1) 2. 3B 1. 2B 1.13 .9B 1.10 1. 6S 2. OS 1. so 1.10 1. lB
BB
9B 1.10 1.10 1. 9B 2.60 .oo 22. 97 (2) 1. lS
62
S4 . 47 .S3 .BO .99
72
S3 .S7
42
47 .S3 .S3 .9S 1.2S .oo 11. 07 4-7 1S2 SS 29 20 S3 BO 190 19B Bl 32 3S so 103 Bl lBO 313 0 16S2 (1) 3. BO 1. 3B
73 .so 1. 33 2.00 4. 7S 4. 9S 2.03 .BO
BB 1. 2S 2.SB 2. 03 4. so 7. B3 .00 41.33 (2) 1. B3 .66 .3S
24
64 .96 2. 29 2.39 .9B .39 .42 .60 1. 24 .9B 2.17 3. 77 .oo 19. 92 B-12 BB 10 4 0 2 2S 19 9S 6S 12 14 lB 176 1B2 1S6 213 0 1079 (1) 2. 20 .2S .10 .00 .OS . 63
4B 2. 3B 1. 63 .30 .3S
4S 4. 40 4. SS 3. 90 S.33 .00 27. 00 (2) 1. 06 .12 .OS .00
02 .30 . 23 1. lS .7B .14 .17 .22 2.12 2.19 1. BB 2.S7 .00 13.01 13-lB 9 0 0 0 0 0 1 7 13 l 0 0 60 109 7S 39 0 314 (1) .23 .00 .00 .oo
00
00 .03 .lB .33 .03
00
00 1.SO 2. 73 1. BB .9B .oo 7. B6 (2) .11 .oo .oo .oo
00
00 .01
OB .16
01
00 .oo . 72 1. 31
90 .47 .00 3.79 19-24 0 0 0 0 0 0 0 0 0 0 0 0 4 21 s 2 0 32 (1) .oo .00 .00
00 .00
00 .oo
00 .00
00 .oo .00 .10 .S3 .13
OS .oo .BO (2) .oo .00 .00 .00 .00
00 .oo .oo .00
00 .00 .00 .OS .2S .06 .02 .oo .39 GT 24 0 0 0 0 0 0 0 0 0 0 0 0 0 l 0 0 0 1 (1) .oo .00 .00 .00 .00 .00
00 .00 .00 .oo .00 .00 .00 .03 .00 .00 .00 .03 (2) .00 .00 .00 .00 .oo .oo
00 .oo .00 .00 .00 .00 .00 .01 .00 .oo .oo .01 ALL SPEEDS 344 117 7B S9 99 171 292 360 203 92 B4 107 3B7 43B 49S 671 0 3997 (1) B. 61 2.93 1. 95 1. 4B 2. 4B 4. 2B 7. 31 9. 01 s. OB 2.30 2 .10 2. 6B 9.6B 10. 96 12.3B 16. 79 .00 100. 00 (2) 4 .15 1.41 .94
71 1.19 2. 06 3. S2 4. 34 2. 4S 1.11 1. 01 1. 29 4. 67 S.2B 5. 97 B. 09 .oo 4B. 20

( 1) ~PERCENT OF ALL GOOD OBSERVATIONS FOR THIS PAGE (2) =PERCENT OF ALL GOOD OBSERVATIONS FOR THIS PERIOD C= CALM (WIND SPEED LESS THAN OR EQUAL TO

95 MPH) 31

TABLE5E VERMONT YANKEE JAN 16 - DEC 16 METEOROLOGICAL DATA JOINT FREQUENCY DISTRIBUTION 3S. 0 FT WIND DATA STABILITY CLASS E CLASS FREQUENCY (PERCENT) ~ 24.2B WIND DIRECTION FROM SPEED N NNE NE ENE E ESE SE SSE s SSW SW WSW w WNW NW NNW VRBL TOTAL MPH CALM 1 0 0 0 1 1 2 0 1 1 1 2 2 0 0 0 0 12 (1) .OS .00 .00 .00

OS .OS .10 .oo .OS . OS
OS .10 .10 .oo .oo .00 .oo .60 (2) .01
00 .00 .00 .01 .01
02 .oo
01
01 .01 .02 .02 .oo .oo .oo
00 .14 C-3 3B 2S 19 14 17 30 41 6B 104 121 170 1S3 136 9S 99 64 0 1194 (1) 1. B9 1. 24 . 94
70
B4 1. 49 2. 04 3 .3B s .17 6. 01 B. 4S 7.60 6. 76 4. 72 4.92 3. lB . 00 S9.31 (2)
46 .30 .23 .17 .21 .36
49
B2 1. 2S 1. 46 2.0S 1. BS 1. 64 1. lS 1.19
77
00 14. 40 4-7 31 3 0 1 7 16 44 S6 46 22 lS 40 S2 4S 100 147 0 62S

( 1) 1. S4 .lS .00 .OS . 3S .79 2.19 2. 7B 2. 29 1. 09

7S 1. 99 2.SB 2. 24 4. 97 7 .30 .00 31. OS (2) .37
04 .oo
01 .OB .19 .S3 .6B . S5 .27 .lB . 4B . 63 .54 1. 21 1. 77 .00 7. 54 B-12 5 0 0 4 1 9 2B 7 4 2 14 23 2S 36 0 158 (1) .2S .oo .00 .00 .OD .20 .05 .45 1. 39
35 .20 .10
70 1.14 1. 24 1. 79 .oo 7. 85 (2) .06 .00 .00 .oo .oo
05
01 .11 .34 .00
05 .02 .17
2B .30
43 .oo 1.91 13-lB 0 0 0 0 0 0 0 0 10 0 0 0 3 3 4 3 0 23 (1) .oo .oo .oo . 00 .00 .oo .00 .00
50 .oo .oo .00 .lS .15 .20 .lS .00 1.14 (2) .00 .00 .00 . 00 .00
00 .oo .oo .12 .oo
00 .oo . 04
04 .OS .04
00 .2B 19-24 0 0 0 0 0 0 0 0 0 0 0 0 0 0 1 0 0 1 (1) .00 .00 .oo
00
00
00 .00 .00 .00
00
00 .00 .00 .00 . OS
00 .00 .OS (2) .oo .oo .00 .00
00 .00 . 00 .00 .00
00
00 .00 .00 .00
01 . 00 .00 .01 GT 24 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 (1) .oo .oo .00 .00 .00 .00 .00 .00
00 .00 .00 .00 .00 .00 .00 .00 .00 .00 (2) .00 .00 .00 .oo .00 .oo .oo .00
00 .00 .00 .00 .00 .oo .00
00 .oo .00 ALL SPEEDS 7S 28 19 15 2S Sl 88 133 1B9 151 190 197 207 166 229 250 0 2013 (1) 3. 73 1.39 . 94 . 75 1. 24 2. S3 4. 37 6. 61 9. 39 7. so 9. 44 9. 79 10. 28 B. 25 11. 3B 12. 42 .00 100. 00 (2) .90
34 .23 .18 .30
62 1. 06 1. 60 2. 28 1. B2 2. 29 2. 3B 2. so 2. 00 2. 76 3.01 .00 24. 2B

( 1) - PERCENT OF ALL GOOD OBSERVATIONS FOR THIS PAGE (2)-PERCENT OF ALL GOOD OBSERVATIONS FOR THIS PERIOD C- CALM (WIND SPEED LESS THAN OR EQUAL TO

95 MPH) 32

TABLE SF VERMONT YANKEE JAN 16 - DEC 16 METEOROLOGICAL DATA JOINT FREQUENCY DISTRIBUTION

35. 0 FT WIND DATA STABILITY CLASS F CLASS FREQUENCY (PERCENT) 12. 99 WIND DIRECTION FROM SPEED N NNE NE ENE E ESE SE SSE SSW SW WSW w WNW NW NNW VRBL TOTAL MPH CALM 2 0 0 0 0 1 0 1 1 0 0 1 0 0 0 0 0 6 (1) .19

00 .00 .00
00
09 .oo .09
09 .00 .00 .09 .00 .00 .00 .oo .00 .56 (2) .02
00 .00 .00
00
01 .oo .01 .01 .00 .00 .01 .00 .00 .oo .00 .00 .07 C-3 17 9 7 8 7 5 18 29 66 126 249 162 120 59 30 20 0 932 (1) 1. 58
84 . 65
74
65
46 1. 67 2.69 6.13 11. 70 23 .12 15. 04 11.14 5. 48 2. 79 1. 86 .oo 86. 54 (2) .21 .11 .08 .10 .08 .06 .22 .35 .80 1. 52 3. 00 1. 95 1. 45 . 71 .36
24 .oo 11.24 4-7 2 1 0 0 1 1 4 12 13 8 18 5 16 10 22 21 0 134 (1) .19 .09 .00 .oo .09 .09 . 37 1.11 1. 21 . 74 1. 67 . 46 1. 49 .93 2. 04 1. 95 .00 12. 44 (2) .02 .01 .00 .00 .01 .01
05 .14 .16 .10 . 22 .06 .19 .12 .27 .25 .00 1.62 8-12 0 0 0 0 0 0 0 0 1 0 0 0 0 0 0 3 0 4 (1) .00 .00 .00 .00 .oo .00 .00 .00
09 .oo .00
00 .oo .00 .oo .28 .00 .37 (2) .00 .00 .00 .00 .oo .00 .00 .00 .01 .00 .00 .00 .oo .oo
00
04
00 .OS 13-18 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 (1) .oo .00 .00 .00
00 .00 .oo .oo .00 .00
00
00 .00 .oo
00 .oo
00 .oo (2) .oo .00 .00 .00
00
00 .00 .00 .00 .00 .oo .oo .oo .oo
00 .00
00 .00 19-24 0 0 0 0 0 0 0 0 0 0 0 0 0 0 1 0 0 1 (1) .00 . 00 .oo .oo . 00
00 .00 .00
00 .oo
00 .oo .00 .00
09 .00
00 .09 (2) .00 . 00 .00
00 .oo
00 .00 .00 .00 .00
00 .oo . 00 .oo .01 .00
00 .01 GT 24 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 (1) .oo .OD .00
00 .00 . OD .DO . 00 .00 .00
00 .00 .00 .00 .00 . 00 .00 .oo (2) .DO .oo .00 .00 .00 . 00 .DO .oo .00 .DO .00 .00 .00 .00 .00 .00 .00 .00 ALL SPEEDS 21 10 7 8 8 7 22 42 81 134 267 168 136 69 53 44 0 1077 (1) 1. 95 .93
65
74
74
65 2. 04 3. 90 7. 52 12. 44 24. 79 15. 60 12. 63 6. 41 4. 92 4. 09 . 00 100 .00 (2) .25 .12
08 .10 .10 . 08 .27 .51 .98 1. 62 3.22 2.03 1. 64 .83 . 64 .53 .00 12. 99

( 1) ~PERCENT OF ALL GOOD OBSERVATIONS FOR THIS PAGE ( 2) ~PERCENT OF ALL GOOD OBSERVATIONS FOR THIS PERIOD c~ CALM (WIND SPEED LESS THAN OR EQUAL TO . 95 MPH) 33

TABLESG VERMONT YANKEE JAN 16 - DEC 16 METEOROLOGICAL DATA JOINT FREQUENCY DISTRIBUTION

35. 0 FT WIND DATA STABILITY CLASS G CLASS FREQUENCY (PERCENT) = 3.55 WIND DIRECTION FROM SPEED N NNE NE ENE E ESE SE SSE SSW SW WSW w WNW NW NNW VRBL TOTAL MPH CALM 0 0 0 0 0 0 0 0 0 0 0 1 0 0 0 0 0 1 (1) .00 .oo .oo .00 .00 .00

00 .oo .oo .00 .00 . 34
00 . 00 .oo
00 .00 . 34 (2) .oo
00 .00
00 .oo .00 . 00 .00
00
00 .oo .01 . 00 . 00 .00 .00 .00 .01 C-3 8 7 2 4 4 6 9 11 24 32 52 32 27 13 10 9 0 250 (1) 2. 72 2.38 .68 1. 36 1. 36 2. 04 3.06 3. 74 8 .16 10. 88 17. 69 10. 88 9 .18 4. 42 3. 40 3.06 .oo 85. 03 (2) .10
08 .02 .05
05 .07 .11 .13 .29 .39
63 .39 .33 .16 .12 .11 .00 3. 01 4-7 0 1 0 0 0 3 0 1 1 1 10 2 4 5 7 5 0 40 (1) .00
34 .oo
00
00 1. 02 .00
34 .34
34 3. 40 .68 1.36 1. 70 2. 38 1. 70 .00 13. 61 (2) .00 .01 .00
00
00
04 .00 .01
01
01 .12 .02 .05 .06
08 .06 .00 . 48 8-12 0 0 0 0 0 0 0 0 0 0 0 0 1 0 0 2 0 3 (1) .oo .oo .00 .00 . 00
00 .oo .00 .00
00 .oo .oo
34 .oo .00
68 .oo 1. 02 (2) .oo .oo .00 .00 .00
00 .oo .00
00
00 .00 .00 .01 .oo .oo .02 .00
04 13-18 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 (1) .00 .00 .oo .00 .00 .00
00 .oo .oo .00
00 .00 .00 .00
00 .oo
00 .oo (2) .00 .00 .00 .oo .00 .00 . 00
00 .00 .oo
00 .00 .00 .00 . 00 .00 . 00 .00 19-24 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 (1) .oo .00 .00 .00 .oo .00 . 00 .00 .00 .00 .oo .oo .oo .oo
00 .00 . 00 .00 (2) .oo .00 .00 .00 .oo .00 . 00 .00 .00 .00 .oo .oo .00
00 .00 . 00 .00 .00 GT 24 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 (1) .00 . 00 .oo .00 .00 . 00 .oo
00 .00 . 00 .00 .00 .oo .00 .oo .oo .00 .oo (2) .00 . 00 .00 . 00 .00 . 00 .00 .oo .oo
00 .00 . 00 .00 .00 .00 .oo .oo .00 ALL SPEEDS 8 8 2 4 4 9 9 12 25 33 62 35 32 18 17 16 0 294 (1) 2. 72 2. 72 .68 1. 36 1. 36 3.06 3. 06 4. 08 8. 50 11. 22 21. 09 11. 90 10.88 6.12 5. 78 5. 44 .oo 100. 00 (2) .10 .10 .02 .OS
05 .11 .11 .14
30 .40
75 . 42 .39 .22 .21 .19 .00 3. 55

( 1) =PERCENT OF ALL GOOD OBSERVATIONS FOR THIS PAGE (2)=PERCENT OF ALL GOOD OBSERVATIONS FOR THIS PERIOD C= CALM (WIND SPEED LESS THAN OR EQUAL TO

95 MPH) 34

TABLE5H VERMONT YANKEE JAN 16 - DEC 16 METEOROLOGICAL DATA JOINT FREQUENCY DISTRIBUTION

35. 0 FT WIND DATA STABILITY CLASS ALL CLASS FREQUENCY (PERCENT) 100. 00 WIND DIRECTION FROM SPEED N NNE NE ENE E ESE SE SSE s SSW SW WSW w WNW NW NNW VRBL TOTAL MPH CALM 6 2 0 0 l 2 2 l 2 l l 4 2 0 0 0 0 24

11) .07 . 02 .00 .00 .01 .02 . 02 .01 . 02 .01 .01 .05 .02 .00 .00 .00

00 .29 (2) .07 . 02 .00 .00 .01 .02
02 .01 .02 .01 .01 .05 . 02 . 00 .00 .00 .oo .29 C-3 167 103 82 74 86 116 152 170 238 328 508 392 328 212 223 204 0 3383 (1) 2. 01 l. 24 .99 .89 l. 04 l. 40 l. 83 2. 05 2. 87 3.96 6 .13 4. 73 3.96 2. 56 2. 69 2. 46 .00 40.80 (2) 2. 01 l. 24 .99 .89 l. 04 l. 40 l. 83 2. 05 2. 87 3. 96 6 .13 4. 73 3.96 2. 56 2. 69 2. 46 .oo 40. 80 4-7 255 79 53 43 115 166 280 289 158 71 81 100 181 150 333 544 0 2898 (1) 3. 08 .95 .64
52 l. 39 2. 00 3. 38 3. 49 l. 91 .86 .98 l. 21 2.18 l. 81 4. 02 6.56
00 34.95 (2) 3. 08 .95 . 64
52 l. 39 2.00 3. 38 3. 49 l. 91 .86
98 l. 21 2.18 l. 81 4.02 6.56
00 34. 95 8-12 125 20 4 0 9 54 35 152 122 28 18 25 201 220 206 315 0 1534 (1) l. 51 . 24 .05 .00 .11 . 65
42 l. 83 l. 47 . 34 .22 .30 2. 42 2. 65 2. 48 3. 80 .oo 18. 50 (2) l. 51 .24 .05 .00 .11 . 65
42 l. 83 l. 47 . 34
22 .30 2. 42 2. 65 2. 48 3. 80
00 18.50 13-18 10 0 0 0 0 2 l 10 26 l 0 0 72 121 99 68 0 410 (1) .12 .00 .00 .00 .00 .02 .01 .12 . 31 .01 .00 . 00 . 87 l. 46 l.19 . 82 .oo 4. 94 (2) .12 .00 .00 .oo .00 .02 .01 .12 . 31 . 01 .00 .00 . 87 l. 46 l.19 .82 .oo 4. 94 19-24 0 0 0 0 0 0 0 0 0 0 0 0 4 22 11 3 0 40 (1) .oo .oo .00 . 00 .00
00 .oo .oo
00 .oo .oo .00
05 .27 .13
04
00 . 48 (2) .00 .00 .00 .00 .00 . 00 .oo
00 .00 .00 .00 .00 .05 . 27 .13 . 04 .oo .48 GT 24 0 0 0 0 0 0 0 0 0 0 0 0 0 3 0 0 0 3 (1) .00 .00 .00 .00 .00 .oo . 00 .00 .00 . 00 .00 .00 .00 . 04 . 00 .00 .00 . 04 (2) .00 .bo .00
00 .oo .00 . 00 .00 .00 .oo . 00 .oo .00
04 . 00 .00
00 . 04 ALL SPEEDS 563 204 139 117 211 340 470 622 546 429 608 521 788 728 872 1134 0 8292 Ill 6. 79 2. 46 l. 68 l. 41 2. 54 4 .10 5. 67 7 .50 6.58 5.17 7 .33 6.28 9 .50 8. 78 10.52 13. 68 .oo 100. 00 (2) 6. 79 2. 46 l. 68 l. 41 2. 54 4 .10 5. 67 7. 50 6. 58 5.17 7 .33 6. 28 9. 50 8. 78 10. 52 13. 68 .00 100. 00 I l) =PERCENT OF ALL GOOD OBSERVATIONS FOR THIS PAGE (2)=PERCENT OF ALL GOOD OBSERVATIONS FOR THIS PERIOD C= CALM (WIND SPEED LESS THAN OR EQUAL TO .95 MPH) 35

TABLE6A VERMONT YANKEE JAN 16 - DEC 16 METEOROLOGICAL DATA JOINT FREQUENCY DISTRIBUTION 297. 0 FT WIND DATA STABILITY CLASS A CLASS FREQUENCY (PERCENT) =

46 WIND DIRECTION FROM SPEED N NNE NE ENE E ESE SE SSE SSW SW WSW w WNW NW NNW VRBL TOTAL MPH CALM 1 1 0 0 0 0 0 0 0 0 0 0 0 0 0 1 0 3 (1) 2.SO 2.SO .oo .oo .00 .00
00 .oo
00 .00 .00 .00 .oo .oo .oo 2. so .00 7 .so (2) .01 .01 .oo .oo
00 .00
00 .oo .00 .00 .00 .00 .oo .oo .oo .01 .00 .03 C-3 0 1 1 1 0 l 2 3 0 0 0 0 0 0 0 0 0 9 (1) .00 2.SO 2. so 2. so
00 2.SO s. 00 7 .so .oo
00 .00 .00 .00 .oo .oo .oo .00 22. so (2) .00 .01 .01
01
00
01 . 02
03 .oo
00 .00 .00 .00 .00 .00 .oo .oo .10 4-7 l l 1 l 0 1 0 l 0 0 0 0 0 0 0 3 0 9 Ill 2. so 2.50 2. 50 2. so
00 2. 50 .oo 2. 50 .00 .oo .oo
00 .oo .oo .oo 7. 50 .00 22.SO (2) .01 .01 .01 .01
00
01
00 .01 .00 .00 .00
00 .oo .oo .oo .03 .oo .10 8-12 5 0 l 0 0 2 0 l 0 0 0 1 0 0 1 2 0 13 (1) 12.SO .00 2.50 .oo
00 5.00 .00 2 .50 .oo .00 .00 2. so .00 .00 2. so S.00 .oo 32.50 (2) .06 .00 .01 .oo .00 .02 .00
01 .oo . 00 .00 .01 .oo .oo .01 .02 .oo .15 13-18 0 0 0 0 0 l 0 2 0 0 0 0 l 0 0 2 0 6 (1) .00 .oo .oo .oo .00 2. so .00 s. 00 . 00 .00 . 00 .00 2 .so .oo .00 5. 00
00 lS. 00 (2) .00 .00 .oo .oo .oo . 01 .00 .02 . 00 .00 . 00 .00 .01 .oo .00
02
00 .07 19-24 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 (1) .oo .00 .00 .00 .00
00 .00
00 .00 .oo
00 .oo . 00 .oo .00 . 00
00 .00 (2) .oo .00 .00 .00 .oo .00 .00 .00
00 .oo
00 .oo .00 .00 .oo
00 . 00 .00 GT 24 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 (1) .oo .00 .00 .00 .oo .00 .00
00 .00 .oo .oo .00 .00 . 00 .00 . 00 .oo .oo (2) .00 .oo .oo .00 .00 .00 .oo .oo
00 .00
00
00 .oo .oo .oo .00
00 .00 ALL SPEEDS 7 3 3 2 0 s 2 7 0 0 0 l 1 0 l 8 0 40 (1) 17. 50 7 .50 7. so s. 00 . 00 12. so 5. 00 17. 50 .00 .00 .oo 2. 50 2. 50 .oo 2. 50 20.00 .00 100. 00 (2) .08 . 03 .03
02
00 .06 . 02 . 08 .00 .00 .oo
01 .01 .00 .01 .09 .00 . 46

( 1) ~PERCENT OF ALL GOOD OBSERVATIONS FOR THIS PAGE (2)~PERCENT OF ALL GOOD OBSERVATIONS FOR THIS PERIOD c~ CALM (WIND SPEED LESS THAN OR EQUAL TO . 95 MPH) VERMONT YANKEE JAN 16 - DEC 16 ME:TE:OROLOGICAL DATA JOINT FREQUENCY DISTRIBUTION 36

TABLE6B VERMONT YANKEE JAN 16 - DEC 16 METEOROLOGICAL DATA JOINT FREQUENCY DISTRIBUTION 297. 0 FT WIND DATA STABILITY CLASS B CLASS FREQUENCY (PERCENT) 1.36 WIND DIRECTION FROM SPEED N NNE NE ENE E ESE SE SSE SSW SW WSW w WNW NW NNW VRBL TOTAL MPH CALM 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 (1) .00 .00 .00 .00 .oo

00 .00 .00
00 .00 .00 .00 .00 .00 . 00
00 .00 .oo (2) .00 .00 .oo .oo .00 . 00 .oo .oo
00 .oo
00 .oo .00 .oo
00 .oo .oo .00 C-3 0 0 l 0 0 l l l 0 0 0 0 0 0 0 0 0 4 (1) .oo .oo .84 . 00 .00 . 84 .84 .84 .00 .00 .00 .00 .00 .00 .00 .00
00 3.36 (2) .00 .00 .01 .00 .oo .01 .01 .01 .00 .00 .00 .00 .00 .oo .00 .oo . 00 .OS 4-7 5 l 0 2 l 11 l l 0 0 0 0 0 0 0 4 0 26 (1) 4. 20
84 .00 1. 68
84 9. 24
84
84
00 . 00 .00
00 . 00
00 .00 3.36 .00 21. 85 (2)
06
01 .00 .02 .01 .13 .01
01 .00 . 00 .oo .00 . 00 .00 .00
05 .oo .30 8-12 9 5 6 1 1 10 5 2 0 0 0 1 2 0 2 17 0 61 (1) 7 .56 4. 20 5 .04
84 .84 8. 40 4. 20 1. 68 .00 .00 .oo
84 1. 68 .00 1. 68 14. 29 .oo 51. 26 (2) .10 .06 .07 .01 .01 .11 .06 .02
00 .00 . 00
01 .02 .oo .02 .19 .00
70 13-18 1 2 0 0 0 1 2 2 3 0 0 1 1 1 1 9 0 24 (1)
84 1. 68 .oo .00 . 00
84 1. 68 1. 68 2 .52 .oo .oo
84
84
84
84 7 .56 .00 20.17 (2) .01 . 02 .oo .oo .oo
01 .02
02
03 .00 .00 .01 .01 .01 .01 .10 .00 .27 19-24 0 0 0 0 0 0 0 0 0 0 0 0 l 0 0 3 0 4 (1) .oo .oo .00 . 00 .00 .00
00 .oo
00 .oo
00 .00
84 .oo .00 2. 52 .oo 3.36 (2) .oo .00 .00 . 00 .oo .oo
00 .oo .00 .00 .00 .00 .01 .00 .00 .03 .oo .05 GT 24 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 (1) .00 .00 .00 .00 .oo .00 .00 .00 .00 .00 .00 .00
00 . 00 .00 .00 .oo .oo (2) .00 .00 .00 .oo .oo .00 . 00 .oo .00
00 .oo .oo
00
00 .oo .oo .00 .oo ALL SPEEDS 15 8 7 3 2 23 9 6 3 0 0 4 1 3 33 0 119 (1) 12. 61 6. 72 5. 88 2. 52 1. 68 19 .33 7. 56 5. 04 2. 52 .00
00 1. 68 3.36
84 2. 52 27. 73 .oo 100. 00 (2) .17
09 .08 .03
02 .26 .10 .07 .03 .00
00 .02 .05
01 .03 .38 .oo 1.36

( 1 ) =PERCENT OF ALL GOOD OBSERVATIONS FOR THIS PAGE ( 2) =PERCENT OF ALL GOOD OBSERVATIONS FOR THIS PERIOD C= CALM (WIND SPEED LESS THAN OR EQUAL TO

95 MPH) 37

TABLE6C VERMONT YANKEE JAN 16 - DEC 16 METEOROLOGICAL DATA JOINT FREQUENCY DISTRIBUTION 297.0 FT WIND DATA STABILITY CLASS C CLASS FREQUENCY (PERCENT) 3.13 WIND DIRECTION FROM SPEED N NNE NE ENE E ESE SE SSE SSW SW WSW w WNW NW NNW VRBL TOTAL MPH CALM 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 (1) .oo .oo .00

00 .00 .00
00 .00
00 .00 .00 .00 .00 .oo .oo .oo
00 .00 (2) .oo
00 .00
00 .00 .00
00 .oo .00 . 00 .00 .00 .00
00 .oo .oo .oo .oo C-3 1 2 1 1 1 4 6 0 0 0 0 0 0 0 0 1 0 17 (1) .36
73 .36 .36 .36 1. 46 2.19
00 .00 .00 .oo .oo .00 .00 .oo .36 .00 6. 20 (2) .01 .02 .01 . 01 .01 .05 . 07 .00 .00 .oo .oo .00 .00 .00
00 .01 .00 .19 4-7 11 3 4 6 6 17 10 0 3 0 1 0 0 2 6 17 0 86 (1) 4.01 1. 09 1. 46 2 .19 2.19 6. 20 3. 65 .00 1. 09 .00 .36 .00 .00 .73 2.19 6.20 .00 31. 39 (2) .13 .03 . 05 .07
07 .19 .11 .oo .03 .oo .01 .oo .00 .02 . 07 .19
00 .98 8-12 13 9 2 3 4 12 16 15 2 0 1 1 4 2 3 25 0 112 (1) 4. 74 3. 28
73 1. 09 1. 46 4. 38 5. 84 5. 47 . 73
00. .36 .36 1. 46 .73 1.09 9.12 .00 40. 88 (2) .15 .10 .02 .03
05 .14 .18 .17 .02 .00
01 .01
05 .02 .03 .29 .00 1. 28 13-18 7 1 2 0 1 0 1 5 3 0 0 2 0 5 7 8 0 42 (1) 2.55 .36 .73
00 .36 .00 .36 1. 62 1. 09 .oo
00
73 . 00 1. 82 2. 55 2. 92 . 00 15. 33 (2) .OB .01 .02
00
01 .00 .01 .06 .03 .oo .oo
02 . 00 .06
08 .09 . 00 .48 19-24 0 0 0 0 0 0 0 0 0 0 0 0 2 1 2 9 0 14 (1) .00
00 .oo .00 .00 .00 . 00 .oo
00 .00 .00 .00 . 73 .36
73 3.28 .00 5.11 (2) .00 .oo .00 .00 .00 .00 . 00 .oo
00 . 00 .00 .00 .02
01 .02 .10 .oo .16 GT 24 0 0 0 0 0 0 0 0 0 0 0 0 0 0 1 2 0 3 (1) .00 .00 .00 .00
00 .00 . 00 .00 .00 .00 .00 .00 .oo .oo .36
73 .oo 1.09 (2) .00 .oo .oo .00 . 00 .00 .oo .oo .00 .00 . 00 .00 .00 .oo .01 .02 .oo .03 ALL SPEEDS 32 15 9 10 12 33 33 20 8 0 2 3 6 10 19 62 0 274 (1) 11. 68 5. 47 3.28 3. 65 4. 38 12. 04 12. 04 7. 30 2. 92 .00
73 1. 09 2.19 3. 65 6. 93 22. 63
00 100. 00 (2) .37 .17 .10 .11 .14
38
38
23 . 09 .00 .02 . 03 .07 .11
22 . 71 . 00 3 .13

( 1 ) =PERCENT OF ALL GOOD OBSERVATIONS FOR THIS PAGE (2)=PERCENT OF ALL GOOD OBSERVATIONS FOR THIS PERIOD C= CALM (WIND SPEED LESS THAN OR EQUAL TO . 95 MPH) 38

TABLE6D VERMONT YANKEE JAN 16 - DEC 16 METEOROLOGICAL DATA JOINT FREQUENCY DISTRIBUTION 297.0 FT WIND DATA STABILITY CLASS D CLASS FREQUENCY (PERCENT) 51. 07 WIND DIRECTION FROM SPEED N NNE NE ENE E ESE SE SSE SSW SW WSW w WNW NW NNW VRBL TOTAL MPH CALM 0 1 0 0 1 2 0 0 0 0 0 0 0 0 0 8 0 12 (1) .oo

02 .00 .oo
02
04 .00 .oo .oo .00 .00 .00 .00 .oo . 00 .18 .00
27 (2) .00 .01 .00 .00 .01 . 02 .00 .00 .00 .oo .00 .00 .oo .oo .oo
09 .00 .14 C-3 41 32 35 41 38 52 69 50 22 16 8 13 18 18 45 68 0 566 (1) .92
72
78 .92 .85 1.16 1. 54 1.12 .49 .36 .18 .29
40
40 1. 01 1. 52 .00 12. 67 (2)
47 .37 .40
47 .43
59 .79 .57
25 .18 .09 .15 .21 .21 .51
78 .00 6. 47 4-7 99 22 26 22 40 72 153 139 63 25 23 16 28 32 65 216 0 1041 (1) 2. 22
49 .58
49
90 1. 61 3. 42 3.11 1. 41 .56
51 .36
63 . 72 1. 45 4. 83 .00 23. 30 (2) 1.13 .25 .30
25
46 . 82 1. 75 1. 59
72 .29 .26 .18 .32 . 37
74 2. 47 .00 11. 90 8-12 120 26 17 9 25 29 86 202 155 30 35 48 137 160 80 297 0 1456 (1) 2.69 .58
38 .20 .56 . 65 1. 92 4. 52 3. 47 . 67 .78 1.07 3.07 3.58 1. 79 6. 65 .00 32 .S9 (2) 1.37 .30 .19 .10 .29 .33 .98 2. 31 1. 77 . 34 .40 .55 1.57 1. 83 .91 3.39 .00 16. 64 13-18 75 12 9 0 2 10 12 46 112 14 11 19 lSl 181 147 236 0 1037 (1) 1. 68 .27 .20 .00
04
22
27 1. 03 2. 51 .31 .2S
43 3 .38 4. OS 3. 29 s. 28 .00 23. 21 (2)
86 .14 .10 . 00
02 .11 .14 .53 1. 28 .16 .13 .22 1. 73 2.07 1. 68 2. 70 .00 11. BS 19-24 15 0 0 0 0 1 2 2 22 1 0 1 40 72 60 87 0 303 (1)
34 .00 .oo . 00 .00
02
04 .04 .49
02 .00 .02 .90 1. 61 1. 34 1. 9S .00 6. 78 (2) .17 .oo .00 .oo .oo
01 . 02 .02
25
01 .00 .01
46 .82 .69
99 .00 3. 46 GT 24 1 0 0 0 0 0 0 1 3 0 0 0 3 19 12 14 0 53 (1) .02 .00 .00 .00
00 .00
00
02 .07 . 00 .00 .00 .07
43 .27 .31 .oo 1.19 (2) .01 .00 .oo .00 .00 .00
00 .01 .03 .00 . 00 .oo .03 .22 .14 .16 .00 .61 ALL SPEEDS 3Sl 93 87 72 106 166 322 440 377 86 77 97 377 482 409 926 0 4468 (1) 7. 86 2. 08 1. 95 1. 61 2. 37 3. 72 7. 21 9. 85 8. 44 1. 92 1. 72 2.17 8. 44 10. 79 9. lS 20. 73
00 100. 00 (2) 4.01 1.06 .99
82 1. 21 1.90 3.68 5. 03 4. 31 .98
88 1.11 4. 31 s. Sl 4. 67 10. 58 .oo Sl. 07

( 1 ) =PERCENT OF ALL GOOD OBSERVATIONS FOR THIS PAGE (2) =PERCENT OF ALL GOOD OBSERVATIONS FOR THIS PERIOD C= CALM (WIND SPEED LESS THAN OR EQUAL TO . 95 MPH) 39

TABLE6E VERMONT YANKEE JAN 16 - DEC 16 METEOROLOGICAL DATA JOINT FREQUENCY DISTRIBUTION 2 97. 0 FT WIND DATA STABILITY CLASS E CLASS FREQUENCY (PERCENT) = 31.26 WIND DIRECTION FROM SPEED N NNE NE ENE E ESE SE SSE s SSW SW WSW w WNW NW NNW VRBL TOTAL MPH CALM 1 3 s 3 4 1 3 3 1 0 0 1 0 2 0 6 0 33 (1) .04 .11 .18 .11

lS
04 .11 .11
04 .oo .00
04 .00 .07 .oo .22 .oo 1. 21 (2) .01 .03 .06
03
OS
01 .03 .03 .01 .00 . 00 .01 .00 .02 .00 .07 .oo .38 C-3 119 62 6S 61 74 71 91 60 30 16 10 12 13 2S 43 92 0 844 (1) 4. 3S 2. 27 2 .38 2. 23 2. 71 2. 60 3. 33 2.19 1.10 .S9 .37 .44
48 .91 l.S7 3.36 .00 30.86 (2) 1. 36 . 71
74
70 .es .81 1. 04 .69 .34 .18 .11 .14 .lS .29 .49 l.OS .00 9. 6S 4-7 lOS 10 8 6 lS se 1S6 110 41 lS lS 23 18 16 94 321 0 1011 (1) 3. 84 . 37 .29 .22 .SS 2 .12 s. 70 4. 02 1. so .SS .SS
84 .66 .S9 3.44 11. 74 .oo 36.97 (2) 1. 20 .11 .09 .07 .17 .66 1. 78 1. 26
47 .17 .17 .26 .21 .18 1.07 3. 67 .00 ll.S6 8-12 so 11 0 0 1 7 4S S9 61 18 18 20 48 S9 47 19S 0 639 (1) 1. 83
40 .00 .00 . 04 .26 1. 6S 2 .16 2.23 .66 .66
73 1. 76 2.16 1. 72 7.13 .00 23. 36 (2)
S7 .13 .00 .oo .01
08 .Sl . 67 . 70 .21
21 .23
SS .67
S4 2.23 .oo 7 .30 13-18 4 0 0 0 0 2 1 3 31 lS 2 4 21 23 16 60 0 182 (1) .lS .00 .00 .00 .oo .07
04 .11 1.13 .SS .07
lS
77
84 .S9 2.19 .00 6. 6S (2) .OS .00 .00 .00 .00 .02 .01 .03 .3S .17 .02 .OS
24 .26 .18
69 .00 2. 08 19-24 0 0 0 0 0 0 0 0 4 3 0 1 1 2 2 11 0 24 (1) .oo .oo .00
00
00 . 00 .00 .oo
lS .11
00
04 . 04 .07 . 07 .40
00 .88 (2) .00 .oo .00 .oo
00 .oo .00 .oo .OS .03 .00 .01 .01 .02 . 02 .13
00
27 GT 24 0 0 0 0 0 0 0 0 1 0 0 0 0 0 0 1 0 2 (1) .00
00 .00 .00 .oo .00 .oo .00
04 .00 .oo
00 .oo .oo .oo
04 .00 .07 (2) .oo .oo .00 .oo .oo .00 .00 .00 .01 .00 .00 .00 .00 .oo .00 .01 .oo .02 ALL SPEEDS 279 86 78 70 94 139 296 23S 169 67 4S 61 101 127 202 686 0 273S (1) 10. 20 3.14 2. es 2. S6 3. 44 s. 08 10. 82 8. S9 6 .18 2. 4S 1. 6S 2. 23 3. 69 4. 64 7.39 2S.08 .00 100. 00 (2) 3 .19 .98 .89 .80 1. 07 l.S9 3. 38 2.69 1. 93 . 77 .Sl
70 l.lS 1. 4S 2.31 7. 84 .oo 31. 26 (l)=PERCENT OF ALL GOOD OBSERVATIONS FOR THIS PAGE (2)=PERCENT OF ALL GOOD OBSERVATIONS FOR THIS PERIOD C= CALM (WIND SPEED LESS THAN OR EQUAL TO
9S MPH) 40

TABLE6F VERMONT YANKEE JAN 16 - DEC 16 METEOROLOGICAL DATA JOINT FREQUENCY DISTRIBUTION 297.0 FT WIND DATA STABILITY CLASS F CLASS FREQUENCY (PERCENT) 10. 72 WIND DIRECTION FROM SPEED N NNE NE ENE E ESE SE SSE SSW SW WSW w WNW NW NNW VRBL TOTAL MPH CALM 2 0 0 1 0 1 2 1 0 0 0 0 1 0 0 1 0 9 (1) .21

00 .00 .11 .00 .11 .21 .11 .00 .oo
00 .00 .11 .00 .00 .11
00 .96 (2) .02 .00 .00 .01 .oo .01
02 .01 .00 .oo . 00 .00 .01 .oo .oo .01 .oo .10 C-3 56 26 25 27 29 17 34 25 10 10 2 7 8 16 18 34 0 344 (1) 5. 97 2. 77 2. 67 2. 88 3. 09 1. Bl 3. 62 2.67 1. 07 1.07 .21 . 75
85 1. 71 1. 92 3. 62 .oo 36. 67 (2) .64 .30 .29 .31 .33 .19 .39 .29 .11 .11 .02
08 .09 .18 .21 .39
00 3. 93 4-7 44 6 3 2 7 31 64 48 22 12 13 10 8 12 34 90 0 406 (1) 4. 69
64 .32 .21
75 3.30 6. 82 5.12 2. 35 1. 28 1. 39 1.07
85 1. 28 3. 62 9. 59 .00 43. 28 (2) .50 .07 .03 .02
08
35
73 .55 . 25 .14 .15 .11 .09 .14 .39 1. 03 .00 4. 64 8-12 5 1 0 0 0 4 27 16 14 6 4 1 11 10 16 49 0 164 (1) .53 .11 .00
00 .oo
43 2. 88 1. 71 1. 49
64
43 .11 1.17 1. 07 1. 71 5.22 .oo 17. 48 (2) .06 .01 .00
00
00 .05 .31 .18 .16 .07 .05 .01 .13 .11 .18 .56 .oo 1. 87 13-18 3 0 0 0 0 0 0 0 1 0 1 0 3 2 3 1 0 14 (1) .32 .00 .oo .00
00 .oo .oo .oo .11 .00 .11
00 .32 .21 .32 .11 .00 1. 49 (2) .03 .00 .00
00
00 .00 .oo .oo
01 .00 .01 .oo .03
02 .03
01 .00 .16 19-24 0 0 0 0 0 0 0 1 0 0 0 0 0 0 0 0 0 1 (1) .oo .oo .00 .00
00 .oo .00 .11 .00 .oo .oo .00 .00 .00 . 00 . 00 .00 .11 (2) .oo .00 .00 .oo
00 .00 .00 .01 .00 .oo
00 .00 .00 .00 .00
00 .oo .01 GT 24 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 Ill .00 .oo .oo .00 . 00 .oo .oo .00 .00 .00 .oo .oo .oo .oo .00 .DO .00 .00 (2) .oo .00 .oo .oo
00 .00 .00 .oo .oo .00 .00 .00 .00 .oo .oo .oo .oo .00 ALL SPEEDS 110 33 28 30 36 53 127 91 47 28 20 18 31 40 71 175 0 938 (1) 11. 73 3.52 2. 99 3.20 3. 84 5. 65 13. 54 9. 70 5. 01 2. 99 2 .13 1. 92 3.30 4 .26 7 .57 18. 66 .oo 100. 00 (2) 1. 26 .38 .32 . 34 .41 .61 1. 45 1. 04
54 .32 .23 .21 .35
46 .81 2. 00
00 10. 72 I 1 ) ~PERCENT OF ALL GOOD OBSERVATIONS FOR THIS .PAGE (2)~PERCENT OF ALL GOOD OBSERVATIONS FOR THIS PERIOD c~ CALM (WIND SPEED LESS THAN OR EQUAL TO . 95 MPH) 41

TABLE6G VERMONT YANKEE JAN 16 - DEC 16 METEOROLOGICAL DATA JOINT FREQUENCY DISTRIBUTION 297.0 FT WIND DATA STABILITY CLASS G CLASS FREQUENCY (PERCENT) = 2. 00 WIND DIRECTION FROM SPEED N NNE NE ENE E ESE SE SSE SSW SW WSW w WNW NW NNW VRBL TOTAL MPH CALM 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 (1) .00 .00 .00

00 . 00 .00 .00 .oo
00
00 . 00 .00 .oo
00 .00 .00 .00 .oo (2) .oo .oo .00 .00
00 .00 .00 .00 .oo .oo . 00 .00 .oo .oo .oo .oo .oo .00 C-3 6 1 3 2 2 0 2 6 2 3 1 4 1 3 5 4 0 45 (1) 3. 43 .57 1. 71 1.14 1.14
00 1.14 3. 43 1.14 1. 71
57 2 .29
57 1. 71 2.86 2.29 .oo 25. 71 (2)
07 .01 .03 .02
02
00
02 . 07 .02 .03
01 .05 .01 .03 .06 .05 .00 .51 4-7 5 0 0 1 0 1 12 11 8 4 1 5 10 9 5 12 0 84 (1) 2. 86 .00 .00
57 .00 .57 6. 86 6. 29 4. S7 2.29
57 2. 86 s. 71 s .14 2.86 6. 86 .oo 48. 00 (2) .06 .00 .00
01 . 00
01 .14 .13
09 .05
01
06 .11 .10 .06 .14 .oo .96 8-12 1 0 0 0 0 0 3 4 4 4 4 1 6 3 5 6 0 41 (1) .S7 . 00 .oo .00 .00 .oo 1. 71 2. 29 2. 29 2.29 2. 29
S7 3. 43 1. 71 2. 86 3. 43 .oo 23. 43 (2) .01 . 00 .oo .00 .00 .oo . 03 . 05 .05
OS .OS
01 .07 .03 .06 .07 .00 .47 13-18 0 0 0 0 0 0 0 0 1 1 0 0 1 1 0 0 0 4 (1) .00 .00 .oo .00 .00 .00
00 .00 .S7 .S7 .00 .00 .57 .57 . 00 .00 .00 2. 29 (2) .00 .00 .oo .00 .00 .00
00
00 .01 .01 .00 .oo .01 .01
00 . 00
00 .as 19-24 0 0 0 0 0 0 0 0 1 0 0 0 0 0 0 0 0 1 (1) .oo .oo .00 .00 .oo .00 . 00 .00 . S7
00 .oo .00 .00 .00 .00 . 00
00 .S7 (2) .oo .oo .00 .00 .00 .00
00 .00 .01
00 .oo .00 .00 .00 . 00 . 00 .00 .01 GT 24 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 (1) .00 .00 .00 .00 .oo .00 .00 .00 .00 .00 .oo .00 .00 .00 .00 .00 .00 .oo (2) .00 .00 .00 .oo .00 .00 .oo .oo .00 .00 .00 .00 .oo .oo .oo .00 .oo .00 ALL SPEEDS 12 3 3 2 1 17 21 16 12 6 10 18 16 lS 22 0 175

( 1) 6. 86

S7 1. 71 1. 71 1.14
S7 9. 71 12. 00 9.14 6. 86 3. 43 5. 71 10. 29 9 .14 8. 57 12. S7 .oo 100. 00 (2) .14 .01 .03 .03
02
01 .19
24 .18 .14 .07 .11 .21 .18 .17 .25 .00 2.00

( 1) =PERCENT OF ALL GOOD OBSERVATIONS FOR THIS PAGE (2)=PERCENT OF ALL GOOD OBSERVATIONS FOR THIS PERIOD C= CALM (WIND SPEED LESS THAN OR EQUAL TO . 9S MPH) 42

TABLE6H VERMONT YANKEE JAN 16 - DEC 16 METEOROLOGICAL DATA JOINT FREQUENCY DISTRIBUTION 297. 0 FT WIND DATA STABILITY CLASS ALL CLASS FREQUENCY (PERCENT) = 100. 00 WIND DIRECTION FROM SPEED N NNE NE ENE E ESE SE SSE s SSW SW WSW w WNW NW NNW VRBL TOTAL MPH CALM 4 5 5 4 5 4 5 4 1 0 0 1 1 2 0 16 0 57 (1) .05 .06 .06 .05 .06 .05 .06

05
01 . 00
00
01 .01 .02 .oo .lB .00 .65 (2) .05 .06 .06 .05 .06 .05 .06 .05 .01 .oo
00 .01 .01 .02 .oo .lB .00 .65 C-3 223 124 131 133 144 146 205 145 64 45 21 36 40 62 111 199 0 1B29 (1) 2. 55 1. 42 1. 50 1. 52 1. 65 1. 67 2. 34 1. 66
73 .51
24 .41
46 . 71 1.27 2. 27
00 20. 91 (2) 2.55 1. 42 1. 50 1. 52 1. 65 1. 67 2. 34 1. 66
73 .51
24 . 41 . 46
71 1. 27 2. 27
00 20. 91 4-7 270 43 42 40 69 191 396 310 137 56 53 54 64 71 204 663 0 2663 (1) 3. 09
49 . 4B . 46 .79 2.lB 4. 53 3. 54 1. 57
64 .61
62
73 .Bl 2. 33 7. 5B .oo 30. 44 (2) 3. 09
49 . 4B . 46 .79 2.lB 4. 53 3. 54 1. 57
64 .61 . 62
73 .Bl 2. 33 7. 5B .oo 30. 44 B-12 203 52 26 13 31 64 1B2 299 236 5B 62 73 20B 234 154 591 0 24B6 (1) 2.32 .59 .30 .15 .35
73 2. OB 3. 42 2. 70 .66
71
B3 2. 3B 2. 67 1. 76 6. 76 .oo 2B. 41 (2) 2.32 .59 .30 .15 .35
73 2. OB 3. 42 2. 70 .66
71
B3 2. 3B 2. 67 1. 76 6. 76 .oo 2B. 41 13-lB 90 15 11 0 3 14 16 5B 151 30 14 26 17B 213 174 316 0 1309 (1) 1. 03 .17 .13 . 00
03 .16 .lB .66 1. 73
34 .16 .30 2.03 2. 43 1. 99 3. 61
00 14. 96 (2) 1. 03 .17 .13 .00
03 .16 .lB .66 1. 73
34 .16
30 2.03 2. 43 1. 99 3. 61
00 14. 96 19-24 15 0 0 0 0 1 2 3 27 4 0 2 44 75 64 110 0 347 (1) .17 .00 .oo .00 . 00
01 .02 .03
31 .05
00 .02 .50 .B6
73 1. 26 .00 3. 97 (2) .17 .00 .oo .oo
00
01
02 .03 .31 .05 .oo .02 .50 .B6
73 1. 26 .oo 3. 97 GT 24 1 0 0 0 0 0 0 1 4 0 0 0 3 19 13 17 0 5B (1) .01 .00 .oo .00
00
00
00 .01 .05
00 .oo .00 .03
22 .15 .19 .00 .66 (2)
01 .00 .00 .oo .oo .00 .oo .01 .05
00 .oo .oo .03 .22 .15 .19 .00 .66 ALL SPEEDS B06 239 215 190 252 420 B06 B20 620 193 150 192 53B 676 720 1912 0 B749 (1) 9. 21 2. 73 2. 4 6 2.17 2. BB 4. BO 9. 21 9 .37 7. 09 2.21 1. 71 2 .19 6.15 7. 73 B.23 21.B5 .oo 100. 00 (2) 9. 21 2. 73 2. 46 2 .17 2. BB 4. 80 9. 21 9 .37 7.09 2. 21 1. 71 2.19 6.15 7. 73 8.23 21.B5 .oo 100. 00

( 1) =PERCENT OF ALL GOOD OBSERVATIONS FOR THIS PAGE (2) =PERCENT OF ALL GOOD OBSERVATIONS FOR THIS PERIOD C= CALM (WIND SPEED LESS THAN OR EQUAL TO .95 MPH) 43

APPENDIX A SUPPLEMENTAL INFORMATION Facility: Vermont Yankee Nuclear Power Station Licensee: Entergy Nuclear Vermont Yankee IA. ODCM DOSE AND DOSE RATE LIMITS - ODCM Controls Dose Limit

a. Noble Gases 3/4.3.1 Total body dose rate 500 mrem/yr 3/4.3.1 Skin dose rate 3000 mrem/yr 3/4.3.2 Gamma air dose 5 mrad in a quarter 3/4.3.2 Gamma air dose 10 mrad in a year 3/4.3.2 Beta air dose 10 mrad in a quarter 3/4.3.2 Beta air dose 20 mrad in a year

b. Tritium and Radionuclides in Particulate Form 3/4.3.1 Organ dose rate 1500 mrem/yr 3/4.3.3 Organ dose 7 .5 mrem in a quarter 3/4.3.3 Organ dose 15 mrem in a year

c. Liquids 3/4.2.2 Total body dose 1.5 mrem in a quarter 3/4.2.2 Total body dose 3 mrem in a year 3/4.2.2 Organ dose 5 mrem in a quarter 3/4.2.2 Organ dose 10 mrem in a year 2A. ODCM LIMITS - CONCENTRATION ODCM Control Limit

a. Noble Gases No ECL Limits

b. Tritium and Radionuclides in Particulate Form No ECL Limits A-1

c. Liquids 3/4.2.l Sum of the fractions ofECL excluding noble gases (10CFR20, Appendix B, Table 2, Column 2): .:S l.OE+Ol 3/4.2.1 Total noble gas concentration: .::: 2E-04 µCi/cc

3. AVERAGE ENERGY Provided below are the average energy (E) of the radionuclide mixture in releases of fission and activation gases, if applicable.

a. Average gamma energy: Not Applicable

b. Average beta energy: Not Applicable

4. MEASUREMENTS AND APPROXIMATIONS OF TOTAL RADIOACTIVITY Provided below are the methods used to measure or approximate the total radioactivity in effluents and the methods used to determine radionuclide composition.

a. Fission and Activation Gases "

Continuous stack monitors monitor the gross Noble Gas radioactivity released from the plant stack. Because release rates are normally below the detection limit of these monitors, periodic grab samples are taken and analyzed for the gaseous isotopes present. These are used to calculate the individual isotopic releases indicated in Table lB and the totals of Table lA. The error involved in these steps may be approximately +/-23 percent.

b. Particulates Continuous isokinetic samples are drawn from the plant stack through a particulate filter. The filters are normally removed weekly and are analyzed for particulates. The error involved in these steps may be approximately

         +/- 18 percent.

A-2

d. Tritium ODCM Table 4.3 .1 requires as a minimum that grab samples from the plant stack be taken monthly and analyzed for tritium. The stack tritium collection has been upgraded with silica gel columns and continuous sampling of stack effluents. The error involved in this sample is approximately +/-18 percent.

e. Waste Oil Prior to issuing the permit to burn a drum of radioactively contaminated waste oil, one liter of the oil is analyzed by gamma spectroscopy to determine concentrations of radionuclides that meet or exceed the LLD for all of the liquid phase radionuclides listed in ODCM Table 4.2.1.

Monthly, samples from drums that were issued burn permits are sent to the contracted laboratory for compositing and analysis. The lab analyzes for tritium, alpha, Fe-55, Sr-89, and Sr-90 on the composite sample. The error involved in this sample is approximately +/-15 percent.

f. Liquid Effluents If radioactive liquid effluents are to be released from the facility, they are continuously monitored. Measurements are also required on a representative sample of each batch of radioactive liquid effluents released. For each batch, station records are retained of the total activity (mCi) released, concentration

(µCi/ml) of gross radioactivity, volume (liters), and approximate total quantity of water (liters) used to dilute the liquid effluent prior to release to the Connecticut River. Each batch of radioactive liquid effluents to be released is analyzed for gross gamma and gamma isotopic radioactivity. A monthly proportional composite sample, comprising an aliquot of each batch released during a month, is analyzed for tritium and gross alpha radioactivity. A quarterly proportional composite sample, comprising an aliquot of each batch released during a quarter, is analyzed for Sr-89, Sr-90, and Fe-55. A-3

5. BATCH RELEASES

a. Liquid There were no routine liquid batch releases during the reporting period.

b. Gaseous There were no routine gaseous batch releases during the reporting period.

6. ABNORMAL RELEASES

a. Liquid

1) In 2016 there was a continuous release due to the residual radioactivity in groundwater from a previously undetected leak from a subsurface structure.

The leak condition was identified through monitoring well data in January 2010. The leak was stopped in February 2010.

2) For 2016, the total Tritium radioactivity conservatively estimated to be released to the Connecticut River is 0.0262 Curies. No other plant-related radionuclides were detected in ground water.

3) During the NEI 5-year self-assessment for groundwater monitoring, it was determined that VY was not fully meeting the intent of reporting groundwater contamination results in annual reports. VY has consistently reported the curies of all liquid and gaseous effluents from the plant through regularly monitored discharge pathways as well as from the tritium leak of2010. The reporting of individual well results or trends has not been a part of the annual report and this was entered into the Corrective Actions Program as WT-WTVTY-2016-00018 CA-00009 to ensure inclusion in 2016 and future reports.

VY has installed 32 groundwater wells to monitor the 2010 leak event or to monitor additional at-risk structures, systems or components (SSCs) that could cause a release oflicensed material to the groundwater. One well (GZ-08) has been dry since installation and no samples were collected from it in 2016. A second well, GZ-24 was compromised by excavation activity in 2014 and is no longer able to be sampled. A summary of the remaining 30 wells is included in Table 1. There are only six (6) wells that have detectable activity (>MDC) in 2016 and all of these wells are well below the EPA limit of20,000 pCi/L for drinking water. None of the wells in this program supply drinking water, and no drinking water wells on-site or adjacent to VY have shown tritium at detectable levels in regular surveillance samples. A-4

Ta ble 1: VYGroundwat er T"f nmm Summary Concentration Number of Range1 Groundwater analyses Mean well Sampled performed Concentration1 Min Max GZ-01 4 < 617 < 583 < 656 GZ-02 4 < 608 < 582 < 643 GZ-03 4 < 613 < 587 < 645 GZ-04 4 < 613 < 584 < 643 GZ-05 4 < 613 < 580 < 643 GZ-06 4 < 616 < 576 < 646 GZ-07 4 < 605 < 571 < 637 GZ-09 4 < 616 < 571 < 645 GZ-10 4 < 615 < 587 < 647 GZ-11 4 < 612 < 578 < 643 GZ-12 4 < 616 < 584 < 649 GZ-12D 4 1539 746 2170 GZ-13 4 < 618 < 591 < 642 GZ-13D 4 < 614 < 579 < 652 GZ-14 4 1474 916 2440 GZ-14D 4 6568 5460 7600 GZ-15 4 1650 1350 2200 GZ-16 4 < 614 < 576 < 653 GZ-17 4 < 613 < 582 < 642 GZ-18 4 < 608 < 585 < 642 GZ-18D 4 < 612 < 580 < 631 GZ-19 4 < 614 < 578 < 651 GZ-19D 4 < 615 < 584 < 652 GZ-20 4 < 613 < 578 < 642 GZ-21 4 < 612 < 578 < 650 GZ-22D 12 4784 3470 5970 GZ-23 12 1378 737 1790 GZ-25 12 < 575 < 450 < 664 GZ-26 12 < 576 < 454 < 663 GZ-27 12 < 575 < 427 < 677 1 All concentrations are in units of pCi/L Required LLD for tritium == 2,000 pCi/L A-5 L__ __

b. Gaseous There were no non-routine gaseous releases (measured) during the reporting period.

A-6

APPENDIXB LIQUID HOLDUP TANKS Requirement Technical Specification 3.1.A.1 limits the quantity ofradioactive material contained in any outside tank. With the quantity of radioactive material in any outside tank exceeding the limits of Technical Specification 3.1.A.1, a description of the events leading to this condition is required in the next annual Radioactive Effluent Release Report per ODCM Section 10.1. Response: The limits of Technical Specification 3.1.A.1 were not exceeded during this reporting period. B-1

APPENDIXC RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION Requirement: Radioactive liquid effluent monitoring instrumentation channels are required to be functional in accordance with ODCM Table 3 .1.1. If a non-functional radioactive liquid effluent monitoring instrument is not returned to functional status prior to a release pursuant to Note 4 of Table 3 .1.1, an explanation in the next annual Radioactive Effluent Release Report of the reason(s) for delay in correcting the non-functionality are required per ODCM Section 10.1. Response: The Service Water Radiation Monitor was non-functional until 12/1/2016 due to the method of determining the set point. ODCM Rev. 37 updated the set point determination and returned this monitor to functionality. Compensatory actions were maintained while the monitor was non-functional. C-1

APPENDIXD RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION Requirement: Radioactive gaseous effluent monitoring instrumentation channels are required to be functional in accordance with ODCM Table 3.1.2. If a non-functional, gaseous effluent monitoring instrumentation is not returned to functional status within 30 days pursuant to Note 5 of Table 3.1.2, an explanation in the next annual Radioactive Effluent Release Report of the reason(s) for the delay in correcting the inoperability is required per ODCM Section 10.1. Response: Since the requirements of ODCM Table 3.1.2 governing the functionality of radioactive gaseous effluent monitoring instrumentation were met for this reporting period, no response is required. D-1

APPENDIXE RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM Requirement: The radiological environmental monitoring program is conducted in accordance with ODCM Control 3/4.5.1. With milk samples no longer available from one or more of the sample locations required by ODCM Table 3.5.1, ODCM (Rev. 36) 10.1 requires the following to be included in the next annual Radioactive Effluent Release Report: (1) identify the cause(s) of the sample(s) no longer being available, (2) identify the new location(s) for obtaining available replacement samples and (3) include revised ODCM figure(s) and table(s) reflecting the new location(s). Response: As part of Rev. 37 to the ODCM, milk sampling has been eliminated (See Appendix H). These changes are supported by RSCS Technical Support Document (TSD) 16-041 "Vermont Yankee Shut-Down Environmental Radionuclides of Concern and Off-Site Dose Calculation Manual Changes" and RSCS TSD 15-056 "Vermont Yankee Offsite Dose Calculation Manual (ODCM) Revisions for Decommissioning". E-1

APPENDIXF LAND USE CENSUS Requirement: A land use census is conducted in accordance with ODCM Control 3/4.5.2. With a land use census identifying a location(s) that yields at least a 20 percent greater dose or dose commitment than the values currently being calculated pursuant to ODCM Control 4.3.3, the new location(s) must be identified in the next Annual Radioactive Effluent Release Report. Response: The Land Use Census was completed during the third quarter of2016. No locations were identified which yielded a 20 percent greater dose or dose commitment than the values currently being calculated pursuant to ODCM Control 4.3.3. See Table 4C for a listing of nearest residents and milk animals in the site area as determined in the 2016 Land Use Census. F-1

APPENDIXG PROCESS CONTROL PROGRAM Requirement: ODCM Section 10.l requires that licensee initiated changes to the Process Control Program (PCP) be submitted to the Commission in the annual Radioactive Effluent Release Report for the period in which the change(s) was made. Response: There were no changes made to the Process Control Program during this reporting period. G-1

APPENDIXH OFF-SITE DOSE CALCULATION MANUAL Requirement: Technical Specification 6.7.B.1 requires that licensee initiated changes to the Off-Site Dose Calculation Manual (ODCM) be submitted to the Commission in the annual Radioactive Effluent Release Report for the period in which the change(s) was made effective. Response: Two revisions of the Offsite Dose Calculation Manual (ODCM) were enacted in 2016. The basis for each change is described below and the changes are presented in Attachments 1 and 2 to this Appendix. Summary of changes made in ODCM Revision 36 (Effective 3/14/2016) (See ) There are several stations maintained in the current REMP that are carried to monitor plant effluents while VY was online. Due to the permanent cessation of plant operation, these non-required control stations no longer provide any useful data and are being removed from the REMP on the basis that they are not required. This change does not affect the REMP in such a way as to reduce it to a level that does not meet regulatory requirements. The proposed change to discharge contaminated ground water via the storm drain system will add surveillance tasks for Chemistry to characterize each batch of water to be discharged and report the discharges in annual operating reports and effluents reports for the environmental monitoring program. All the changes included provide calculations required to comply with the limits and reporting requirements and provide necessary monitoring steps for collection, characterization and discharge of ground water. These steps are based on current requirements for discharges via the Radwaste Pathway as outlined in the ODCM, so the proposed change maintains the rigor and robustness of the currently permitted, although not utilized, discharge limits and conditions. The proposed change to discharge water that has intruded into plant buildings via the storm drain system will not challenge the system ability to remove excess surface run-off water during precipitation events. Discharges will be performed as batch discharges vice continuous discharges and, as such, can be released when the storm drain system is not challenged by excess water. Input from clean footing drains that has been added to storm drain discharge inputs include intrusion water from plant buildings. Summary of changes made in ODCM Revision 37 (Effective 12/1/2016) (See ) Revision 37 to the ODCM made changes to several tables used as references for calculation of off-site dose to simplify the tables by deleting isotopes that have a short H-1

half-life and have decayed away and are no longer_gc:mnane to dose calculations for this reason. The isotopes that are still present and capable of being discharged, either through routine or emergency releases are maintained in the appropriate tables. In some cases, radioactive decay and the greatly reduced risks for release of radiation have created changes to the Radiological Environmental Monitoring Program (REMP). These changes include the cessation of sampling and analysis for radioiodines in stack samples, environmental air samples and the cessation of milk sampling. These changes are supported by RSCS Technical Support Document {TSD) 16-041 "Vermont Yankee Shut-Down Environmental Radionuclides of Concern and Off-Site Dose Calculation Manual Changes" and RSCS TSD 15-056 "Vermont Yankee Offsite Dose Calculation Manual (ODCM) Revisions for Decommissioning". TSD 15-056 also supports the change in direct radiation measurement from the current TLD distribution to TLDs sited along the fence line and in other locations as deemed appropriate. The current arrangement of TLDs is in response to guidance following the TMI accident. This distribution is no longer appropriate to the risk associated with the station in its permanently defueled status. The change of the site boundary TLD is supported by "Dose Versus Distance From Hi-Storm lOOS Containing MPC-68 and MPC-68M for Vermont Yankee ISFSI" Holtec Report No: HI-2146076, Holtec Project No: 2347, Approved on 8/8/14. Additional changes include removing reference to equipment and systems that are no longer in service, including Advanced Off Gas (AOG), Steam Jet Air Ejector (SJAE) and other systems that were only associated with power generation. As these systems are permanently retired, their inclusion in the ODCM no longer serves a functional purpose. References to "reactor coolant" in Section 8.1.2 have been replaced with "spent fuel pool". The Service Water (SW) Discharge Monitor is designed to detect leakage of contaminated water into the Service Water system. The reactor coolant through the RHR Heat Exchangers was formerly the potential source of leakage but, since the plant has been shut down and the reactor and RHR systems have been drained, this is no longer a credible leakage pathway. This section describes the setpoint derivation methodology for the Service Water Discharge Monitor as it was formerly calibrated with a liquid source, reactor coolant. There is no longer any reactor coolant due to the draining of the reactor vessel and, therefore, the spent fuel pool is now the only credible source of possible leakage to service water making this the source of choice for calibrating the Service Water Discharge Monitor. The monitor calibration methodology is not changed, only the liquid source is changed. References to 1-131and1-133 are included in Technical Specifications {TS) 6.7.D.g.2 and 6.7.D.i and LBDCR 15-10 included revisions to these technical specifications, due to radioactive decay (see also TSD 15-056 and TSD 16-041). The proposed ODCM revision can be issued while the TS is waiting for implementation because the isotopes, if present would be detected by gamma isotopic analysis. The ODCM change will prevent the collection of samples for which there is no credible basis for being contaminated by these isotopes. H-2

Allachrncnl I ODCl\/I Revision,)() changed pages H-3

Section 7.2 and Table 7.1 were revised lo delete non-required 03/14/16 36 control stations from the REMP Sample Station listing for airborne, waterborne, mixed grasses, milk and sileage.
Section 3/4.2.1 was changed lo add intercepted groundwater releases to Surveillance 4.2.1.a to clarify that controls on liquid releases also apply to intercepted groundwater being released.
Section 3/4.2.1, Table 4.2.1 was changed to add intercepted groundwater requirements to Table 4.2.1. This establishes sampling and analysis protocols that are as rigourous as liquid waste releases and consistent with subsurface groundwater analysis. These protocols provide data sufficient to support dose estimates and assignments and allows comparison between interception wells and other subsurface groundwater.
Section 3/4.2. l, Table 4.2. l Notation (e) was revised to provide the same analytical specifications for intercepted groundwater and subsurface groundwater, enabling comparison of results for

ill gr:1:1ndwat*_*r ~**mpks.

Section 3/4.6 was revised to ensure that liquid effluent instrumentation is not req*.1ired for intercepted groundwater releases.
Section 3/4.2.1, Surveillance 4.~ _I .a was revised to ensure that the ODCM limit of 10 times IOCFR20 Appendix B applies to all water released to the Connecticut River.
Section 5. I was revised to include groundwater intercept release tank(s) as a potential release point and to clarify that intercepted groundwater releases are undiluted.
Section 5.2. l was revised to include intercepted groundwater tanks to the list of tanks from which batch releases are made.

Additionally, this section recognizes alternate sample point(s) ' other <han the rncl\vastc samjJle si1:!< ~i:*c :icceptaolc for collecting samples for analysis prior to release.

Section 5.2.4 is added to describe the new release pathway for intercepted groundwater and to describe it's use in a manner consistent with liquid waste treatment systems and subsurface groundwater.
Section 6.2 is revised to add dose calculation from intercepted groundwater and ensure it is performed the same as subsurface groundwater. This clarifies that total body dose estimates are only required prior to liquid waste releases.
Table 6.2. l is revised to add a footnote clarifying that the mixing ratio is based on a 20,000 gpm flow.
Section 6.2. l, Equation 6-1 is revised to add a flow correction term. This allows dose factors DFL to be adjusted for release flow rates if the flow rate is different than the rate used for Table 1.1.11.

Off-Site Dose Calculation Manual Rev.36 Page iv of xiii Vermont Yankee Nuclear Power Station

Section 6.2.2, Equation 6-2 is revised to add a tlow correction 03/14/16 36 (cont) term. This allows dose factors DFL to be adjusted for release flow rates if the !low rate is different than the rate used for Table I.I.II.
Section 6.3.1, Equation 6-3 is revised to add a tlow correction tenn. This allows dose factors DFL to be adjusted for release

                             !low rates if the How rate is different than the rate used for Table I.I.I I.

Section 6.3.2, Equation 6-4 is revised lo add a flow correction term. This allows dose factors DFL to be adjusted for release flow rates if the flow rate is different than the rate used for Table 1.1.11.

Off-Site Dose Calculation Manual Rev. 36 Page v of xiii Vermont Yankee Nuclear Power Station

7.0 ENVIRONMENTAL MONITORING PROGRAM The radiological environmental monitoring stations are listed in Table 7.1. The locations of the stations with respect to the Vcrmont Yankee plant are shown on the maps in Figures 7-1 to 7-6. 7.1 Intercomparison Program All routine radiological analyses for environmental samples are performed at offsite environmental laboratories. The laboratories participate in several commercial inter-comparison programs in addition to an internal QC sample analysis program and the analysis of client-introduced QC sample programs. The external programs may include the Department of Energy - Mixed Analyte Performance Evaluation Program (MAPEP), Analytics Cross-Check Program - Environmental Inter-laboratory Cross-Check Program, and Environmental Resources Association - Environmental R:'.dioactivity P*.:rfor111anc!.: Evnlu:ilion Program or other NRC-appnwcd sources. 7.2 Airborne Pathway Monitoring The environmental sampling progrnm is designed to achieve several major objectives, including sampling air in predominant up-valley and down-valley wind directions, and sampling air in nearby communities and at a proper control location, while maintaining continuity with two years of preoperational data and all subsequent years of operational data (post 1972.) The chosen air sampling locations are discussed below. To assure that an unnecessarily frequent relocation of samplers will not be required due to short-term or annual fluctuations in meteorology, thus incurring nccJlcss expense and dcstroyi1~g the co1~~inuity of the program, long t.!m1, site specific ground level D/Qs (five-year averages - 1978 through 1982) were evaluated in comparison to the existing air monitoring locations to determine their adequacy in meeting the above-stated objectives of the program and the intent of the NRC general guidance. The long-term average meteorological data base precludes the need for an annual re-evaluation of air sampling locations based on a single year's meteorological history. The Connecticut River Valley in the vicinity of the Vermont Yankee plant has a pronounced up- and down-valley wind flow. Based on five years of meteorological data, wind blows into the 3 "up-valley" sectors (N, NNW, and NW) 27 percent of the time, and the 4 "down-valley" sectors (S, SSE, SE, and ESE) 40 percent of the time, for a total "in-valley" time of 67 percent. Off-Site Dose Calculation Manual Section 7 Rev. 36 Page 1 of 12 Vermont Yankee Nuclear Power Station

Station J\P/CF-12 (NNW, 3.6 km) in North Hinsdale, New Hampshire, monitors the up-valley sectors. It is located in the sector that ranks fourth overall in terms of wind frequency (i.e., in terms of how often the wind blows into that sector), and is approximately 0.75 miles from the location of the calculated maximum ground level D/Q (i.e., for any location in any sector, for the entire Vennont Yankee environs). This station provides a second function by its location in that it also monitors North Hinsdale, New Hampshire, the community with the second highest ground level D/Q for surrounding communities, and it has been in operation since the preoperational period. The down-valley direction is monitored by two stations - at River Station Number 3.3 (AP/CF-11, SSE, 1.9 km) and at Northfield, Massachusetts (AP/CF-14, SSE, 11.6 km). They both reside in the sector with the maximum wind frequency and they bound the down-valley point of calculated maximum ground level D/Q (the second highest overall ground level D/Q for any location in any sector). Station AP/CF-11 is

!pproximak!y one mik !'mm thi.-; j~oint, between it and the plant. Statio:1 AP/CF-I-:.

also serves as a community monitor for Northfield, Massachusetts. Both stations have been in operation since the preoperational period. In addition to the up- and down-valley locations, two communities have been chosen for community sampling locations. The four nearest population groups with the highest long-term average D/Q values, in decreasing order, are Northfield, Massachusetts, North Hinsdale, New Hampshire, Brattleboro, Vem1ont, and Hinsdale, New Hampshire. The community sampler for Northfield is at Station AP/CF-14 (mentioned above). North Hinsdale is already monitored by the up-valley station (AP/CF-12, NNW, 3.6 km), which also indirectly monitors the city of Brattleboro, located further out in the same sector. The second sampler specifically designated for a community is at Hinsdale Substation (AP/CF-13, E, 3.1 km) in Hinsdale. The control air sampler was located at Spofford Lake (AP/CF-21, NNE, 16.4 km) due to its distance from the plant and the low frequency for wind blowing in that direction based on the long-term (five-year) meteorological history. Sectors in the general west to southwest direction, which would otherwise have been preferable due to lower wind frequencies, were not chosen since they approached the region surrounding the Yankee Atomic plant in Rowe, Massachusetts. Off-Site Dose Calculation Manual Section 7 Rev.36 Page 2of12 Vermont Yankee Nuclear Power Station

Table 7.1 Radiological Environmental Monitoring Stations< 11 Ex12ostire Pathway Sam12le Location Distance and/or Sam12le and Designated Code< 2l (kmf'il Direction< 5l I. AIRBORNE (Radioiodine and Particulate) AP/CF-I I River Station No. 3-3 1.88 SSE AP/CF-12 N. Hinsdale, NH 3.61 NNW AP/CF-13 Hinsdale Substation 3.05 E AP/CF-14 Northfield, MA l9 l 11.61 SSE 1111 AP/CF-21 Spofford Lake 16.36 NNE

2. WATERBORNE

a. Surface WR-I I River Station No. 3-3 1.88 Downriver WR-21 Rt. 9 Bridge< 9 l 11.33 Upriver

b. Ground WG-11 Plant Well 0.24 On-Site WG-12 Vernon Nursing Well 2.13 SSE WG-22 Copeland Well19 l 13.73 N

c. Sediment SE-11 Shoreline Downriver 0.57 SSE From SE-12 North Storm 0.13 E Shoreline Drain Outfall(Jl

3. INGESTION

a. Milk 18 l TM-11 Miller Farm 0.82 w TM-18 Blodgett Farm 3.60 SE TM-22 Franklin Farm(9 ) 9.73 WSW

b. Mixed TG-11 River Station No. 3-3 1.88 SSE Grasses TG-12 N. Hinsdale, NH 3.61 NNW TG-13 Hinsdale Substation 3.05 E TG-14 Northfield, MAl 9> 11.61 SSE TG-21 Spofford Lake< 9J 16.36 NNE Off-Site Dose Calculation Manual Section 7 Rev.36 Page 4of12 Vermont Yankee Nuclear Power Station

Ti\13LE 7.1 (Continued) Exposure Pathway Sample Location Distance and/or Sample and Designated Code( 21 (km{' 1 Direetion( 5 )

c. Silage TC-11 Miller Farm 0.82 w TC-18 Blodgett Fann 3.60 SE 1111 TC-22 Franklin Fann 9.73 WSW

d. Fish FH-11 Vernon Pond (6) (6)

FH-21 Rt. 9 Bridge( 91 11.83 Upriver

4. DIRECT RADII\ TION DR-l Rin:r Stati~m Nii. 3-J I .Ci I SSE DR-2 N. Hinsdale, NH 3.88 NNW DR-3 Hinsdale Substation 2.98 E DR-4 Northfield, MA(lJ) 11.34 SSE 1

DR-5 Spofford Lake 1' l 16.53 NNE DR-6 Vernon School 0.52 WSW DR-7 Site Boundary 17 ) 0.28 w DR-8 Site Boundary 0.25 SSW DR-9 Inner Ring 1.72 N DR-10 Outer Ring 4.49 N DR-11 Inner Ring 1.65 NNE DR-12 Outer Rfog 3.58 NNE DR-13 Inner. Ring 1.23 NE DR-1-t Oukr Ring 3.88 NE DR-15 Inner Ring 1.46 ENE DR-16 Outer Ring 2.84 ENE DR-17 Inner Ring 1.24 E DR-18 Outer Ring 2.97 E DR-19 Inner Ring 3.65 ESE DR-20 Outer Ring 5.33 ESE DR-21 Inner Ring 1.82 SE DR-22 Outer Ring 3.28 SE DR-23 Inner Ring 1.96 SSE DR-24 Outer Ring 3.89 SSE DR-25 Inner Ring 1.91 s Off-Site Dose Calculation Manual Section 7 Rev.36 Page 5of12 Vermont Yankee Nuclear Power Station

TABLE 7.1 (Continued) Exposure Pathway Sample Location Distance and/or Sample and Designated Code( 2l (km)( SJ Direction(SJ DR-26 Outer Ring 3.77 s DR-27 Inner Ring I. I 0 SSW DR-28 Outer Ring 2.23 SSW DR-29 Inner Ring 0.92 SW DR-30 Outer Ring 2.36 SW DR-31 Inner Ring 0.71 WSW DR-32 Outer Ring 5.09 WSW DR-33 Inner Ring 0.66 WNW DR-34 Outer Ring 4.61 w DR-35 Inner Ring 1.30 WNW DP.-36 Outc:* ~Zing 4.~3 \'/NW DR-37 Inner Ring 2.76 NW DR-38 Outer Ring 7.34 NW DR-39 Inner Ring 3.13 NNW DR-40 Out~r Ring 5.05 NNW ( 1) Sample locations are shown on Figures 7 .1 to 7 .6. (2) Station Nos. 10 through 19 are indicator stations. Station Nos. 20 through 29 are control stations (for ail except milk, silage aud the Jircct radiation stations). (3) To be sampled and analyzed semiannually. (4) Deleted (5) Distance and direction from the center of the Turbine Building for direct radiation monitors; from the plant stack for all others. (6) Fish samples are collected from anywhere in Vernon Pond, which is adjacent to the plant (see Figure 7-1). (7) DR-7 satisfies Control Table 3.5. l for an inner ring direct radiation monitoring location. However, it is averaged as a Site Boundary TLD due to its close proximity to the plant. (8) In accordance with Control Table 3.5.1, notation a, samples will be collected on the required schedule as availability of milk permits. All deviations from the sample schedule will be reported in the Annual Radiological Environmental Operating Report. (9) Control stations Off-Site Dose Calculation Manual Section 7 Rev.36 Page 6of12 Vermont Yankee Nuclear Power Station

3/4.2 RADIOACTIVE LIQUID EFFLUENTS 3/4.2. l Liquid Effluent Concentration CONTROLS 3 .2.1 The concentration of radioactive material in liquid effluents released in liquid waste effluents, intercepted groundwater released via storm drain or groundwater flowing to the Connecticut River from the site in radioactive concentrations above background (Unrestricted Areas for liquids is at the point of discharge from the plant discharge in Connecticut River) shall be limited to l 0 times the concentrations specified in Appendix B to l OCFR Part 20. l 00 l - 20.2402, Table 2, Column 2 for radionuclides other than noble gases and 2x l 0- 4 microcurie/milliter total activity concentration for all dissolved or entrained noble gases. APPLICABILITY: At all times. ACTION: With the concentration of radioactive material in liquid eft1uents released to Unrestricted Areas exceeding the limits of Control 3 .2.1, immediately take action to decrease the release rate of radioactive materials and/or increase the dilution flow rate to restore the concentration to within the above limits. SURVEILLANCE REQUIREMENTS 4.2.1.a Radioactive material in liquid waste, intercepted groundwater releases, and subsurface groundwater flows to the Connecticut River shall be sampled and analyzed in accordance with requirements of Table 4.2.1. 4.2.1.b The results of the analyses shall be used in accordance with the methods in the ODCM to assure that the concentrations at the point of release to Unrestricted Areas are limited to the values in Control 3 .2.1. Off-Site Dose Calculation Manual Section 3/4 Rev. 36 Page 11 of 48 Vermont Yankee Nuclear Power Station

TABLE 4.2.1 Radioactive Liquid Sampling and Analysis Program Lower Limit Minimum of Detection Liquid Release Sampling Analysis Type of Activity (LLD) Type Frequency Frequency Analysis (uCi/ml) (a) Batch Waste Prior to each Prior to each Principal Gamma 5 x 10-7 Release Tanks (h) release Each release Each Emitters (d) Batch Batch 1-131 1 x 10-6 One Batch per Once per Dissolved and 1 x 1o-5 month sampled month Entrained Gases prior to a (Gamma release Emitters) Prior to each Once per H-3 1 x 10-5 release Each month Batch Composite (cl Gross Alpha 1 x 10-7 Prior to each Once per Sr-89, Sr-90 5 x 1o-8 release Each quarter Batch Composite (cl Fe-55 1 x 10-6 Groundwater Prior to each Prior to each Principal Gamma Activity Interception release Each release Each Emitters Cd> Analysis LLDs (c) Release Tanks lbl Batch Batch H-3 .-~::- .. Prior to each Once per H-3<c> Activity release Each month Analysis LLDs Composite (cl Gross Alpha (e) Batch Off-Site Dose Calculation Manual Section 3/4 Rev. 36 Page 12 of 48 Vermont Yankee Nuclear Power Station

Lower Limit Minimum of Detection Liquid Release Sampling Analysis Type of Activity (LLD) Type Frequency Frequency Analysis (uCi/ml) (a) Prior lo each Once per H-3(cJ Activity release Each quarter Analysis LLDs Composite (c.:) Gamma (e) Batch Emitters1cl Ni-63 1" 1 Fe-55(cl Sr-89 Sr-901c) Alpha Spec(c.:l Subsurface Peri111eler l\.:ri1rn.:ll.!i' H-3 1'- Aci.;\.;Ly Groundwater Wells - Wells - Analysis LLDs I Quarteriy (gl Quarterly 1gl Gamma (e) Flows to the Emitters1cJ Connecticut River I Sentinel Wells Sentinel Wells Ni-63 1cl I 1

                       - Monthly° '   - Monthly°

Fe-55(c) Sr-89 Sr-90 1'-'l Alpha Spec(cl Off-Site Dose Calculation Manual Section 3/4 Rev.36 Page 13 of48 Vermont Yankee Nuclear Power Station

TABLE 4.2.1 NOTATION

a. The LLD is the smallest concentration of' radioactive material in a sample that will yield a net count, above system background, that will be detected with 95% probability with only 5%i probability of' falsely concluding that a blank observation represents a "real" signal.

For a particular measurement system (which may include radiochemical separation): LLD= 4.66*S 11 E*V*K*Y*e-*"*' 1 where: LLD= the lower limit of detection as defined above (microcuries or picocuries/unit mass or volume) Sb= the standard deviation of the background counting rate or of the counting rate of a blank sample as appropriate (counts/minute) E= the counting efficiency (counts/disintegration) V = the sample size (units of mass or volume) K= 2.22 x 106 disintegrations/minute/microcurie or 2.22 disintegration/minute/picocurie as applicable Y= the fractional radiochemical yield (when applicable)

             }. =    the radioactive decay constant for the particular radionuclide (/minute)
             ~t  =   the elapsed time between sample collection and analysis (minutes)

Typical values of E, V, Y and ~t can be used in the calculation. In calculating the LLD for a radionuclide determined by gamma-ray spectrometry, the background shall include the typical contributions of other radionuclides normally present in the samples. Analysis shall be performed in such a manner that the stated LLDs will be achieved under routine conditions. Occasionally, background fluctuations, unavoidably small sample sizes, the presence of interfering nuclides, or other uncontrollable circumstances may render these LLDs unavailable. It should be recognized that the LLD is defined as a "before the fact" limit representing the capability of a measurement system and not as an "after the fact" limit for a particular measurement. This does not preclude the calculation of an "after the fact" LLD for a particular measurement based upon the actual parameters for the sample in question and appropriate decay correction parameters such as decay while sampling and during analysis. Off-Site Dose Calculation Manual Section 3/4 Rev. 36 Page 14 of 48 Vermont Yankee Nuclear Power Station

TABLE 4.2.1 NOTATION (Cont'd) A batch release is the discharge of liquid wastes of a discrete volume. This volume can be in a tank, bladder or pillow tank or other collection devices. Prior to sampling for analysis, each batch shall be isolated and then thoroughly mixed to assure representative sampling.

c. A composite sample is one in which the quantity of liquid sampled is proportional to the quantity of liquid waste discharged and in which the method of sampling employed results in a specimen which is representative of the liquids released. Prior to analyses, all samples taken for the composite shall be thoroughly mixed in order for the composite sample to be representative of the effluent release.

cl. The principal gamma emitters for which the LLD specification will apply are exclusively the following radionuclides: Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, Cs-134, Cs-13 7, C*~* !4 ! , <md C'r.:-144. Thi:-: li~;t div.::; not :1~*::11~ ti."!( only 'hes'.' nuclides n:*,_. to be cktccted 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, but as "not detected." When unusual circumstances result in LLDs higher than required, the reasons shall be documented in the Radioactive Effluent Release Report. I e. At a minimum, each subsurface or intercepted groundwater sample shall be analyzed for the following analytes. Lower Limit of Detection for each analyte is as follows: STANDARD RADIONUCLIDES LLD (PCl/L) 3H 2000 s.;Mn 15 59Fe 30 58Co 15 6oCo 15 6szn 30 9szr 30 95Nb 15 1311 15 134cs 15 137 Cs 18 140Ba 60 140La 15 Off-Site Dose Calculation Manual Section 3/4 Rev. 36 Page 15 of 48 Vermont Yankee Nuclear Power Station

Ir tritium or plant-generated gamma adivity is positively detected, then a sample should be further analyzed for the presence of the following Hard-To-Detect (HTD) radionuclides, as a minimum, using the associated LLD values: HTD RADIONUCLIDES LLD (PCl/L) 55Fe 110 r,3Ni 530 89 Sr 23

                                   ') 0 Sr                           3.5
                             /\lpha Emitters                          15

r. ()f)rJ\/l Sect inn 9.J rurther ddines the location ai~d t!etcrmin:1tio!l Of flo'.\ llm>ugh the plant site groundwater streamtubes.

g. Perimeter wells used to measure flow and concentration of contaminants in streamtubes:

GZ-1 s, GZ-3s, GZ-4s, GZ-5s, GZ- I 3s, GZ-l 3d, GZ-14s, GZ-l 4d, GZ-18s, GZ-l 8d, GZ-l 9s and GZ- I 9d.

h. Sentinel used to measure flow and concentration of contaminants in streamtubes: GZ-27s, GZ-26s, GZ-25s, GZ-23s and GZ-22d.

Off-Site Dose Calculation Manual Section 3/4 Rev.36 Page 16 of48 Vermont Yankee Nuclear Power Station

3/4.6 EFFLUENT AND ENVIRONMENTAL CONTROL BASES INSTRUMENTATION Liquid Effluent lnstnunentation (3 .1.1) The radioactive liquid effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in liquid waste effluents during actual or potential releases of liquid waste ellluents. They are not required for intercepted groundwater releases. The alarm setpoints for these instruments are to ensure that the alam1 will occur prior to exceeding 10 times the concentration limits of Appendix B to 10CFR20.1001-20.2402, Table 2, Column 2, values. Automatic isolation li.mction is not provided on the liquid radwaste discharge line due to the infrequent nature of batch, discrete volume, liquid discharges (on the

n!cr nf once: wr yc~~r or less), and the~ administrat'.'!C controls provided !o ~nsurc that conservative discharge flow rates/dilution flows arc set such that the probability of exceeding the above concentration limits are low, and the potential off-site dose consequences are also low.

Gaseous Effluent Instrumentation (3.1.2) The radioactive gaseous effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in gaseous effluents during actual or potential releases of gaseous effluents. The alarm/trip setpoints for these instruments are provided to ensure that the alarm/trip will occur prior to exceeding design bases dose rates identified in Control 3.3.1. P...ADICACT!VE EFFLUENTS Liquid Effluents: Concentration (3 .2.1) This Control is provided to ensure that at any time the concentration of radioactive materials released in liquid waste effluents, intercepted groundwater released via storm drain or groundwater flowing to the Connecticut River from the site in radioactive contamination concentrations above background (Unrestricted Area for liquids is at the point of discharge from the plant discharge into Connecticut River) will not exceed 10 times the concentration levels specified in 10CFR Part

20. l 001-20.2402, Appendix B, Table 2, Column 2. These requirements provide operational flexibility, compatible with considerations of health and safety, which may temporarily result in releases higher than the absolute value of the concentration numbers in Appendix B, but still within the annual average limitation of the Regulation. Compliance with the design objective doses of Section II.A of Appendix I to 10CFR Part 50 assure that doses are maintained ALARA, and that annual concentration limits of Appendix B to 10CFR20.l001-20.2402 will not be exceeded.

Off-Site Dose Calculation Manual Section 3/4 Rev.36 Page 40 of48 Vermont Yankee Nuclear Power Station

5.0 METHOD TO CALCULATE OFF-SITE LIQUID CONCENTRATIONS Chapter 5 contains the basis for plant procedures that the plant operator requires to meet ODCM Control 3.2. I which limits the total fraction of combined effluent concentration in liquid pathways, excluding noble gases, denoted here as, F\'.NG at the point of discharge at any time (see Figure 9-1 ). F~:NG is limited to less than or equal to ten, i.e., F ENG<lO I - The total concentration of all dissolved and entrained noble gases at the point of discharge from all station sources, <lcnotcu C 1N(;' is iimited io 2t::-04 ~tCi/ml, i.e., C ~u ~ 2 E - 04 ~t Ci I ml. Evaluation of FTNG and c~G is required concurrent with the sampling and analysis program in Control Table 4.2.1. 5.1 Method to Determine FENG I and CNG I Determine the total fraction of combined effluent concentrations at the point of discharge in liquid pathways (excluding noble gases), denoted F\:NU, and determine the total concentration at the point of discharge of all dissolved and entrained noble gases in liquid pathways from all station sources, denoted C 1NG, as follows: FENG = L. Cpi ~ 10 I I ECL i (5-1)

               µCi/ml )

( ~tCi/ml and: Off-Site Dose Calculation Manual Section 5.0 Rev. 35 Page 1of5 Vermont Yankee Nuclear Power Station

(5-2) (~t Ci/ml) (~t Ci/ml) (µCi/ml) where: F\'N<i Total sum of the fractions of each radionuclide concentration in liquid effluents (excluding noble gases) at the point of discharge to an unrestricted area, divided by each radionuclide's ECL value. c 1*1 Concentration at point of discharge to an unrestricted area of radionuclide "i", except for dissolved and entrained noble gases, from any tank or other significant source, p, from which a discharge may be made (including the floor drain sample tank, the waste sample tanks, the detergent waste tank, groundwater interception release tanks and any other significant source from which a discharge can be made) (µCi/ml). This concentration can be calculated from: C 11 i = CrKI x FrK/[Fou_ +Fix] where: CrKI equals the concentration of radionuclide i in the tank to be discharged (µCi/ml); FDIL is equal to the dilution flow provided by the liquid radioactive waste dilution pumps (20,000 gpm); FTK equals the liquid waste discharge pump flow rate which regulates the rate at which liquid from s waste collection tank is discharged (gpm). For releases from Intercepted Groundwater tanks via storm drains, no dilution flow is credited and Cpi = CTKi* ECLi Annual average effluent concentration limits of radionuclide "i", except for dissolved and entrained noble gases, from l OCFR20. l 001-20.2402, Appendix B, Table 2, Column 2 (µCi/ml). c~ 0 Total concentration at point of discharge to an unrestricted area of all dissolved and entrained noble gases in liquid pathways from all station sources (µCi/ml). NG C Ii Concentration at point of discharge to an unrestricted area of dissolved and entrained noble gas "i" in liquid pathways from all station sources (µCi/ml). Off-Site Dose Calculation Manual Section 5.0 Rev. 35 Page 2of5 Vermont Yankee Nuclear Power Station

                                                                                                             ---1 5.2 Method to Determine Radionuclide Concentration for Each Liquid Effluent Pathway 5.2.1 Sample Tanks Pathways Cpi is detennined for each radionuclide above LLD from the activity in a representative grab sample of any of the sample tanks and the predicted flow at the point of discharge to an unrestricted area.

Most periodic batch releases are made from the two 10,000-gallon capacity waste sample tanks. These tanks serve to hold all the high purity liquid wastes after they have been filtered through the waste collector and processed by ion exchange in the fuel pool and waste liClllincra!izcr~. Uther pcriuJic t1<ih.:h n:kascs may aJsd COllll.: l"rnm ih 1.: dcierg...:llt \Vasi-: lank, intercepted groundwater tanks or the tloor drain sample tank. A batch release tank can be any collection device (e.g. bladder, pillow tank, etc.) that meets the discharge requirements of Table 4.2.1 Notation h. The liquid waste tanks are sampled from the radwaste sample sink and the contents analyzed for water quality and radioactivity. If the sample meets all the high purity requirements, the ~ontents of the tank me1y be re-used in the nuclear system. If the sample does not meet all the high purity requirements, the contents are recycled through the radwaste system or discharged. The groundwater intercept tank is sampled from the tank recirculation system, and the contents analyzed for water quality and rnclioactivity. If the snmple meets the requirements for discharge, it may be discharged to the storm drain system. Prior to discharge each sample tank is analyzed for tritium, dissolved noble gases and dissolved and suspended gamma emitters. 5.2.2 Service Water Pathway The service water pathway shown on Figure 9-1, flows from the intake structure through the heat exchangers and the discharge structure. Under normal operating conditions, the water in this line is not radioactive. For this reason, the service water line is not sampled routinely but it is continuously monitored with the service water discharge monitor (No. 17/351 ). The alarm setpoint on the service water discharge monitor is set at a level which is three times the background of the instrument. The service water is sampled if the monitor is out of service or if the alarm sounds. Off-Site Dose Calculation Manual Section 5.0 Rev.35 Page 3of5 Vermont Yankee Nuclear Power Station

Under expected or anticipated operating conditions, the concentration at any time of radionuclides at the point of discharge from the service water effluent pathway lo an unrestricted area will not exceed ten times the effluent concentration values of I OCFR20. I 001-20.2402, Appendix B, Table 2, Column 2. 5.2.3 Circulating Water Pathway The circulating water pathway shown on Figure 9-1, flows from the intake stmcture through the condenser and the discharge slmcture. Under nonnal operating conditions, the water in this line is not radioactive. For this reason, the circulating water line is not sampled routinely but it is monit('rcd continuously hy the disdwrgc process monitor (No. 17/359) located in Lhc discharge structure. The alarm setpoint on the discharge process monitor is set at a level which is three times the background of the instmment. The circulating water is sampled if the monitor is out of service or if the alann sounds. Under n01mal operating conditions, the average concentration of radionuclides at the point of discharge from the circulating water pathway to an unrestricted area will not exceed the annual effluent concentration limits in 10CFR20.1001-20.2402, Appendix B, Table 2, Column 2. 5.2.4 Intercepted Groundwater Concentrations in Flowpaths to the Connecticut River In order to minimize radioactive liquid waste, groundwater intruding into basement structures is intercepted from sub-slab groundwater wells. This water is environmentally derived and captured under basements to prevent additional cross contamination from plant systems and structures. The intercepted ground water is tanked, sampled and released to the storm system with outfalls to the Connecticut River. The Intercepted Groundwater Collection and Release System is isolated from all other plant systems containing liquids in order to prevent cross contamination and contains water collected from the plant environs, therefore continuous release monitoring is not required . Intercepted groundwater tanks are sampled and released as described in 5.2. l Sample Tank Pathway. Off-Site Dose Calculation Manual Section 5.0 Rev.35 Page 4of5 Vermont Yankee Nuclear Power Station

5.2.5 Subsurface Contaminated Groundwater Concentrations in Flowpaths to the Connecticut River The overall direction of groundwater flow at Vermont Yankee (VY) is towards the Connecticut River (west to east). Based on this understanding of site hydrogeologic conditions, the groundwater discharge rates from the developed portion of the site to the river are estimated using a streamtube approach based on Darcy's Law (see Section 9). To estimate the groundwater concentration in each of the designated streamtubes, samples will be collected and analyzed according to requirements specified in Section 3 14, Table 4.2.1. The concentrations shall then be detennined using methods provided in Section 5.1. Off-Site Dose Calculation Manual Section 5.0 Rev.35 Page 5 of 5 Vermont Yankee Nuclear Power Station

The plant has both elevated and ground level gaseous release points: the main vent stack (elevated release), and the North Warehouse waste oil burner (ground level release). Therefore, total dose calculations for skin, whole body, and the critical organ from gaseous releases will he the sum of the elevated and ground level doses.

     /\ppendix D provides an assessment of the surveillance needs for waste oil to ensure that off-site doses from its incineration is maintained within the ALAR/\ limits of the effluent Controls.

6.2 Method to Calculate the Total Bodv Dose from Liquid Releases Effluent Control 3.2.2 limits the total body dose commitment to a Member of the Public from radioactive material in liquid effluents to 1.5 mrem per quarter and 3 mrem per year. Control 3.2.3 requires liquid radwaste treatment when the total body dose estimate exceeds 0.06 mrem in any month. Control 3.4.1 limits the total body dose commitment to any real member of the public from all station sources (including liouids) to 25 mrem in a year. D0se evaluation is required at least once per month. If the liquid radwaste treatment system is not being used, dose evaluation is required before each release. Use Method I first to calculate the maximum total body dose from a liquid release to the Connecticut River as it is simpler to execute and more conservative than Method II. Use Method II if a more accurate calculation of total body dose is needed (i.e., Method I indicates the dose is greater than the limit), or if Method I cannot be applied. If the rad waste system is not operating, the total body dose must be estimated prior to a liquid radioactive waste release (Control 3 .2.3 ). To evaluate the total body dose. use Equation 6.1 to estimate the dose from the planned release and add this to the total body dose accumulated from prior releases during the month. To assess the dose contribution from intercepted and subsurface groundwater contaminated with plant-generated radionuclides, a dose evaluation shall be performed using Method I on a monthly basis. Radionuclide concentration averages and groundwater streamtube average flow rates shall be utilized to estimate the total plant-generated radioactive contaminants released subsurface for the previous monthly period. Off-Site Dose Calculation Manual Section 6 Rev. 36 Page 3 of56 Vermont Yankee Nuclear Power Station

6.2.1 Method I The increment in total body dose from a liquid release is: (6-1) (mrem ) = Cc 1") (mrem)

                         """"Ci" where:

FC Flow correction calculated by dividing the flow at the unrestricted area release point in gpm divided 20,000 gpm or release flow in rt3/sec divided hy 44.6 ft 3/sec. Site-specific total body dose factor (mrem/Ci) for a liquid release. See Table 1.1.11. Total activity (Ci) released for radionuclide "i." (For strontiums and Fe 55, use the most recent measurement available.) Equation 6-1 can be applied under the following conditions (otherwise, justify Method I or consider Method I I): I. Normal operations (not emergency event),

2. Liquid releases were to the Connecticut River, and

3. Any continuous or batch release over any time period.

6.2.2 Basis for Method I This section serves three purposes: (I) to document that Method I complies with appropriate NRC regulations, (2) to provide background and training information to Method I users, and (3) to provide an introductory user's guide to Method II. Method I may be used to show that the effluent Controls which limit off-site total body dose from liquids (3.2.2 and 3.2.3) have been met for releases over the appropriate periods. Control 3.2.2 is based on the ALARA design objectives in 10CFRSO, Appendix I Subsection II A. Control 3.2.3 is an "appropriate fraction," determined by the NRC, of that design objective (hereafter called the Objective). Control 3.4.1 is based on Environmental Standards for Uranium Fuel Cycle in 40CFR 190 (hereafter called the Standard) which applies to direct radiation as well as liquid and gaseous effluents. Off-Site Dose Calculation Manual Section 6 Rev. 36 Page 4 of56 Vermont Yankee Nuclear Power Station

Exceeding the Objective or the Standard does not immediately limit plant operation hut requires a report to the NRC within 30 days. In addition, a waiver may he required. Method I was developed such that "the actual exposure of an individual ... is unlikely to be substantially underestimated" (I OCFR50, Appendix I). The definition, below, of a single "critical receptor" (a hypothetical individual whose behavior results in an unrealistically high dose) provides part of the conservative margin to the calculation of total body dose in Method I. Method II allows that actual individuals, with real behaviors, be taken into account for any given release. In fact. Method I was based on a Method 11 analysis for the critical receptor with maximum exposure conditions instead of any real individual. That analysis was called the "base case;" it was then reduced to form Method I. The steps performed in the Method I derivation follow. First, in the base case, the do<;e impact to the critical receptor (in the form pf dose far:f_or<: OFLith* mrcm/C'i) for a I curie release of each radioisotope in liquid effluents was derived. The base case analysis uses the methods, data and assumptions in Regulatory Guide 1.109 Wciuations A-2. A-3, A-7, A-13 and A-16, Reference A). The liquid pathways identified as contributing to an individual's dose are the consumption of fish from the Connecticut River, the ingestion of vegetables and leaty vegetation which were irrigated by river water. the consumption of milk and meat from cows and beef cattle who had river water available for drinking as well as having feed grown on irrigated land, and the direct exposure from the ground plane associated with activity deposited by the water pathway. A plant discharge flow rate of 44.6 ft3/sec was used with a mixing ratio of 0.0356 which corresponds to a minimum regulated river flow of 1250 cfs at the Vernon Dam just below the plant discharge outfall.* Tables 6.2.1 and 6.2.2 outline human consumption and environmental parameters used in the analysis. The resulting, site-specific, totnl hody dnsc fa 1~tcrs appear !!1 Table 1.1 .11. For any liquid release, during any period, the increment in annual average total body dose from radionuclide "i" is: (6-2) (mrem) (Ci) (m~~m) Off-Site Dose Calculation Manual Section 6 Rev. 36 Page 5 of56 Vermont Yankee Nuclear Power Station

where: FC Flow wrrection calculated by dividing the flow at the unrestricted area release point in gpm divided 20.000 gpm or release flow in ll 3/scc divided by 44.6 ftJ/scc. DFI.i1h Site-specific total body dose fm:tor (mrem/Ci) for a liquid release. Sec Table I. I. I I. Total activity (Ci) released from radionuclide "i." An Mp equal to 1.0 for the fish pathway is assumed between the discharge structure and the dam. Method I is conservative because it is based on dose factors DFL 11 h which were chosen from the base case to be the highest of the four age groups for each radionuclide. as well as assuming minimum river dilution flow. 6.2.3 Method II If Method I cannot be applied, or if the Method I dose exceeds the limit or if a more exact calculation is required, then Method II should be applied. Method II consists of the models. input data and assumptions in Regulatory Guide I. I 09. Rev. I (Reference A). except where site-specific models. data or assumptions are more applicable, such as the use of actual .river flow at the time of actual discharge as opposed to the minimum river flow of l ,250 c!s that is as:;;_1med in the Method I dose factors (except for the fish pathway). The base case analysis, documented above. is a good example of the use of Method II. It is an acceptable starting point for a Method II analysis. Off-Site Dose Calculation Manual Section 6 Rev. 36 Page 6 of 56 Vermont Yankee Nuclear Power Station

6.3 Mdhod to Calculate Maximum Organ Dose from Liquid Releases Effluent Control 3.2.2 limits the maximum organ <lose commitment to a Member of the Public from radioactive material in liquid effluents to 5 mrem per quarter and I 0 mrem per year. Control 3.2.3 requires liquid ra<lwaste treatment when the maximum organ dose estimate exceeds 0.2 mrem in any month. Control 3.4.1 limits the maximum organ dose comniitment to any real member of the public from all station sources (including liquids) to 25 mrem in a year except for the thyroid. which is limited to 75 mrem in a year. Dose evaluation is required at least once per month if releases have occurred. If the liquid radwaste treatment system is not being used, dose evaluation is required before each release. Use Method I first to calculate the maximum organ dose from a liquid release to the Connecticut River as it is simpler to execute and more conservative than Method II. Use Method II if a more accurate calculation of organ dose is needed (i.e., i\k:huJ l i11c.h.::1tc:; ~hc 'h'..: is :;;*calcr tlrn:~ the limit), or i!'Mctho<l ! ca:1:10t he app!ie<l. If the rad\'I aste system is not operating. the maximum organ dose must be estimated prior to a release (Control 3.2.3). To evaluate the maximum organ dose, use Equation 6-3 to estimate the dose from the planned release and add this to the maximum organ dose accumulated from prior releases during the month. To assess the dose contribution from subsurface groundwater contaminated with plant-generated radionuclides, a dose evaluation shall be performed using Method I on a monthly basis. Radionuclide concentration averages and groundwater streamtube average flow rates shall be utilized to estimate the total plant-generated radioactive contaminants released for the monthly period. 6.3. I Method I The increment in maximum organ dose from a liquid release is: (6-3) (mrem) (Ci) (m~:m) (FC) where: FC Flow correction calculated by dividing the flow at the unrestricted area release point in gpm divided 20,000 gpm or release flow in ft3/sec divided by 44.6 ft 3/sec. DFLimo Site-specific maximum organ dose factor (mrem/Ci) for a liquid release. See Table 1.1.11. Off-Site Dose Calculation Manual Section 6 Rev. 36 Page 9of56 Vermont Yankee Nuclear Power Station

O* Total activity (Ci) released for radionuclide "i." (For strontiums and Fe-55, use the most recent measurement available.) Equation 6-3 can be applied under the t(.)llowing conditions (otherwise. justify Method I or consider Method II): I. Normal operations (not emergency event),

2. Liquid releases were to the Connecticut River, and

3. Any continuous or batch rckase over any time period.

Off-Site Dose Calculation Manual Section 6 Rev. 36 Page 10 of56 Vermont Yankee Nuclear Power Station

6.3.2 Basis for Method I This section serves three purposes: (I) to document that Method I complies with appropriate NRC n:gulations, (2) to provide background and training infi.mnation to Method I users, and (3) to provide an introductory user's guide to Method II. The methods to calculate maximum organ dose paralld the total body dose methods (see Section 6.2.2). Only the differences are presented here. For each radionuclide, a dose factor (mrem/Ci) was determined for each of seven organs and four age groups. The largest of these was chosen to be the maximum organ dose factor (DFLimo) for that radionuclide. This calculation used a release flow rate of 44.6 tl 3/sec or 20,000 gpm. The calculated dose is adjusted for dilution proportional to the flow rate by dividing the release flow by these values. For any liquid release, during any period, the increment in annual average dose from .aJi.:i1&ul.:fak "i" LJ tli.: ma:.:mL.m org<u; is: (6-4) (mrem) (Ci) . (mrem) Ci (FC) where: FC Flow correction calculated by dividing the flow at the unrestricted area release point in gpm divided 20,000 gpm or release flow in ft3/sec divided hy 44.6 ft3/sec. Site-specific maximum organ dose factor (mrem/Ci) for a liquid release. See Table 1.1.11. Total activity (Ci) released for radionuclide "i". Because of the assumptions about receptors, environment, and radionuclides; and because of the low Objective and Standard, the lack of immediate restriction on plant operation, and the adherence to I OCFR20 concentrations (which limit public health consequences) a failure of Method I (i.e., the exposure of a real individual being underestimated) is improbable and the consequences of a failure are minimal. Off-Site Dose Calculation Manual Section 6 Rev. 36 Page 11 of56 Vermont Yankee Nuclear Power Station L ODCl'd Revision J7 complete version H-4

VERMONT YANKEE NUCLEAR POWER STATION OFF-SITE DOSE CALCULATION MANUAL REVISION 37 This document contains Vcnnont Yankee proprietary information. This information may not be transmitted, in whole or in part, to any other organization without permission of Vermont Yankee. Effective Date: ----=-1=2/'""""0'-'-1"'-'/l'-'6'----

                                                                                            ;ti.~

Date Off-Site Dose Calculation Manual Rev. 37 Page i of xv Vermont Yankee Nuclear Power Station

REVISION

SUMMARY

D/\TE REVISION DESCRIPTION Three sections of the ODCM were modified with minor changes to 9/23/10 33 incorporate recommendations made in the 2009 RETS/REMP QA audit:

                             - /\typographical error in ODCM Table 6.10.1.was corrected.

The table listed a distance of 26,500 meters for the highest undepicted X/Q for skin dose calculations. The correct value is 2,650 meters. This table is descriptive in nature and is not used for calculation of doses.

                             - The error in distance measurement in the first paragraph on page 2of12 of Section 7 of the ODCM was corrected. The distance of 0.5 miles was revised to 0.75 miles. This value is descriptive in nature and is not used for calculation of doses.
                             - ODCM Table 3.5.1 provided distances in miles whereas other sections of this table contained distances in kilometers.

Additionally, there was no explanation as to why some "inner" ring TLDs are located further from the plant than some "outer" ring TLDs. This table was changed to include both kilometers and miles where distances are required. Additionally, the table footnotes were revised to explain the method used to determine proper location for the inner and outer ring TLDs in each of the 16 compass sectors.

                              - Footnote "e" to ODCM Table 4.5. l did not fully explain how the determination of Barium/Lanthanum 140 activity is determined using daughter ingrowth. Additional information was provided in footnote "e" to provide the reader with a better explanation.
                              - Table 7.1 was revised to clearly indicate the TLDs designated as "control" location TLDs.

Off-Site Dose Calculation Manual Rev. 37 Page ii of xv Vermont Yankee Nuclear Power Station

REVISION

SUMMARY

(Continued) Four main sections of the ODCM were modified with significant 7/08/2011 34 changes to incorporate the contaminated groundwater discharge pathway to the Connecticut River: 0 Section 3 I 4 was revised to include the subsurface groundwater pathway in the Liquids Discharge description. Groundwater monitoring wells used to determine the extent of these releases are listed. The Southwest Well was added as Ground (Potable Drinking) Water sample location in the REMP description of Section 3 I 4. 0 Section 5 was revised to include a description of the determination of plant generated radionuclide concentrations in groundwater discharges. 0 Section 6 was revised to include methods for calculating radiation dose from plant generated radionuclides in groundwater discharges. 0 Section 9 was revised to include the method for determination of groundwater flows in the 17 identified streamtubes flowing from the plant site to the Connecticut River. In addition to revisions of four main sections of the ODCM, the Table of Contents, Definitions and References Sections of the ODCM were revised to reflect the additional subsections, figures, tables, definitions and references in the ODCM. 0 Section 3/4 was revised to delete a reference requirement to 10/09/14 35 Note 5 in Section 2 of Table 3.1.2 (Gaseous Effluent Monitoring Instrumentation). It was determined that Note 5 was in conflict with Note 2 for Section 2 ~~ad th:::L;fore ::;hould be removed as a requirement for Section 2 (CR-VTY-2013-04078 CA-0002). 0 Also, in Sections 2, 314, 6, 8 and 10, the word "operable" was found to be inappropriate and should be replaced by the word "functional". Additionally, the word "inoperable" by the word "non-functional" and the word "operability" was replaced by "functionality" (EN-OP-104 and NRC Inspection Guide 9900) (CR-VTY-2013-04078 CA-0002). 0 In the tables of Section 7 of the Offsite Dose Calculation Manual, it was determined that the out-of-business dairy farms which had provided milk for the REMP but were no longer functional, should be eliminated from the description (WT-WTVTY-2011-00116). Section 9 of the Offsite Dose Calculation Manual was revised to eliminate the references to the Off Gas "30 minute" delay line. This was previously evaluated under CR-VTY-2010-1676 and it was determined that the reference to the "30 minute" should be removed. (WT-WTVTY-2009-00009 CA-0010) Off-Site Dose Calculation Manual Rev. 37 Page iii of xv Vermont Yankee Nuclear Power Station

0 Section 3/4.2.1 was changed to add intercepted groundwater 02/03/16 36 releases to Surveillance 4.2.1.a lo clarify that controls on liquid releases also apply to intercepted groundwater being released. 0 Section 3/4.2. 1, Table 4.2. 1 was changed to add intercepted groundwater requirements to Table 4.2.1. This establishes sampling and analysis protocols that arc as rigorous as liquid waste releases and consistent with subsurface groundwater analysis. These protocols provide data sufficient to support dose estimates and assignments and allows comparison between interception wells and other subsurface groundwater. 0 Section 3/4.2.1, Table 4.2.1 Notation (e) was revised to provide the same analytical specifications for intercepted groundwater and subsurface groundwater, enabling comparison of results for all groundwater samples. 0 Section 3/4.2.1, Surveillance 4.2.1.a _was revised to ensure that the ODCM limit of 10 times 10CFR20 Appendix B applies to all water released to the Connecticut River. 0 Section 3/4.6 was revised to ensure that liquid effluent instrumentation is not required for intercepted groundwater releases. 0 Section 5.1 was revised to include groundwater intercept release tank(s) as a potential release point and to clarify that intercepted groundwater releases are undiluted. 0 Section 5.2.1 was revised to include intercepted groundwater tanks to the list of tanks from which batch releases are made. Additionally, this section recognizes alternate sample point(s) other than the radwaste sample sink are acceptable for collecting samples for analysis prior to release. 0 Section 5.2.4 is added to describe the new release pathway for intercepted groundwater and to describe it's use in a manner consistent with liquid waste treatment systems and subsurface groundwater. 0 Section 6.2 is revised to add dose calculation from intercepted groundwater and ensure it is performed the same as subsurface groundwater. This clarifies that total body dose estimates are only required prior to liquid waste releases. 0 Table 6.2.1 is revised to add a footnote clarifying that the mixing ratio is based on a 20,000 gpm flow. 0 Section 6.2.1, Equation 6-1 is revised to add a flow correction term. This allows dose factors DFL to be adjusted for release flow rates if the flow rate is different than the rate used for Table 1.1.11. 0 Section 6.2.2, Equation 6-2 is revised to add a flow correction term. This allows dose factors DFL to be adjusted for release flow rates if the flow rate is different than the rate used for Table 1.1.11. 0 Section 6.3 .1, Equation 6-3 is revised to add a flow correction term. This allows dose factors DFL to be adiusted for release Off-Site Dose Calculation Manual Rev. 37 Page iv of xv Vermont Yankee Nuclear Power Station

flow rates if the flow rate is different than the rate used for Table I. I. I I. 0 Section 6.3.2, Equation 6-4 is revised to add a flow correction term. This allows dose factors DFL to be adjusted for release flow rates if the flow rate is different than the rate used for Table 1.1.11. 0 Section 7.2 and Table 7.1 were revised to delete non-required control stations from the REMP Sample Station listing for airborne, waterborne, mixed grasses, milk and sileage. 0 Delete Steam Jet Air Ejector (SJAE) Monitoring from Table 12/01/16 37 I. I. I 0 Delete "iodines" from Table 1.1.1 0 Delete Thyroid dose from Table 1.1. l 0 Delete "iodines" from Table 1.1.4 0 Delete iodine and N-16 terms from Table 1.1.6 0 Delete direct radiation calculations from Table 1.1.6 and replace with TLD readings from site boundary 0 Delete reference to North Warehouse from Table 1.1.6 0 Delete SJAE requirements from Table 1.1. 7 0 Delete Total body calculation for noble gas from Table 1.1. 7 0 Remove 1-131, I-133 and SJAE from Table 1.1.8 0 Delete all nuclides other than Kr-85 from Table 1.1.10 0 Delete all nuclides other than Kr-85 from Table 1.1. l OA 0 Delete nuclides with short half-life from Table 1.1.11 0 Delete nuclides with short half-life from Table 1.1.12 0 Delete Gaseous Radwaste Treatment system, Hot Standby and Refueling Outage definitions from Table 2.1. l 0 Delete variables from Table 2.1.2 that are no longer applicable 0 Delete iodines from Table 2.1 ~2 0 Delete SJAE and AOG terms from Table 2.1.2 0 Delete SJAE and AOG requirements from Table 3.1.2 0 Delete Notes 2,3,6,8 & 9 from Table 3.1.2 0 Delete references to Tech Spec 3.8.K and 3.8.J 0 Delete SJAE and AOG requirements from Table 4.1.2 0 Delete SJAE and AOG monitor Notes from Table 4.1.2 0 Delete I-131 analysis and LLD from Table 4.2.1 0 Delete I-131, Ba-140 & La-140 from Table 4.2.l Note (e) 0 Delete iodine requirements from 3/4.3. l 0 Delete SJAE, I-131 and short-lived noble gas requirements from Table 4.3.1 0 Delete start-up, shutdown, power change requirements, SJAE noble gas sample and short-lived gamma emitter list from Table 4.3.1 Notation 0 Delete iodines from 3/4.3.3 0 Delete 3/4.3.4 and Revise Figure 9.2 to reflect change 0 Delete AOG requirements from 3/4.3.5 0 Delete 3/4.3.6 Off-Site Dose Calculation Manual Rev. 37 Page v of xv Vermont Yankee Nuclear Power Station

0 Delete 3/4.3.7 0 Reduce the number of REMP air sample stations due lo reduced risk of release and remove iodine canisters 0 Reduce REMP TLDs lo site boundary and offsitc controls 0 Delete Table 3.5.1 4.a. Milk sampling 0 Delete Table 3.5.1 Notation (g) 0 Delete 1-131, Ba-140, La-140 and Notation (d) from Table 3.5.2 0 Delete 1-131, Ba-140, La-140, milk and Notes (e) and (g) from Table 4.5. l 0 Delete requirement to identify location of nearest milk animal in each meteorological sector from the land use census. 0 Delete infant thyroid dose from 3/4.6 Basis 3.3. l 0 Delete iodine dose from 3/4.6 Basis 3.3.3 0 Delete 3/4.6 Basis 3.3.4 0 Delete AOG from 3/4.6 Basis 3.3.5 0 Delete 3/4.6 Basis 3.3.6 0 Delete 3/4.6 Basis 3.3. 7 0 Delete reference to multiple reactor site from 3/4.6 Basis 3.4. l 0 Revise wording in 3/4.6 Basis 3.5. lto indicate the cessation of milk sampling. 0 Revise wording in 3/4.6 Basis 3.5.2 to indicate the cessation of milk sampling. 0 Revise wording of 5.2. l Sample Tank Pathway to reflect current operational status. 0 Delete Section 5.2.3, Circulating Water Pathway 0 Revise Section 6.1 to delete discussion of burning waste oil 0 Section 6.2.2, reformat to align terms correctly (editorial) 0 Delete thyroid dose from Section 6.3 0 Rc-fom1:it terms under Equation 6-3 (editorial) 0 Revise 6.4. l Method I to remove SJAE calculations and calculate based on stack gas grab samples. Due to decay, remove summation and calculate based solely on Kr-85 for stack and ground calculations 0 Revise 6.4.2 Method I equations to be specific for Kr-85 and remove reference to SJAE and AOG 0 Delete references to AOG and SJAE and make Section 6.5. l Method I specific to Kr-85 0 Revise 6.5.2 Method I to make Kr-85 specific (delete summation in Eq 6-8, 6-9, 6-12, 6-13, 6-7, 6-37, 6-38 and make all D and Q terms specific for Kr-85 0 Section 6.5.2 Basis for Method I delete references to the North Warehouse as a designated ground level release point and the waste oil burner. Designate the Reactor Building as a ground level release point. 0 Delete iodine from Section 6.6 0 Delete references to the North Warehouse waste oil burner and designate the Reactor Building as a ground level release point in Section 6.6 Off-Site Dose Calculation Manual Rev.37 Page vi of xv Vermont Yankee Nuclear Power Station

0 Delete Section 6. 7; beta air dose is bounding based on Kr-85 0 Revise Section 6.8 due to decay, remove summation and calculate based solely on Kr-85 for stack and ground calculations; make all D and Q terms specific for Kr-85 0 Section 6.8.2 Basis for Method I delete references to the North Warehouse as a designated ground level release point and the waste oil burner. Designate the Reactor Building as a ground level release point. 0 Delete iodine from Section 6. 9 0 Delete milk pathway, iodines and noble gas gamma air dose from Section 6.10 0 Revise the stack flow rate in Section 6.10 0 Table 6.10.1 was revised to include new atmospheric dispersion factors for the Reactor Building 0 Revise Table 6.10.2 to show distances to site boundary from the Reactor Building 0 Delete Section b. l l. l due to no dose from N-16 0 Delete Section 6.11.2 0 Delete Section 6.11.3 due to the Low Level Waste Pad being demolished 0 Delete calculation of ISFSI direct dose and measure dose using REMP TLDs 0 Revise Section 6.11.4 to replace calculations of direct dose with REMP TLDs 0 Delete Section 6.11.5 0 Revise Section 7.2 Airborne Pathway to delete sampling in nearby communities 0 Revise Section 7.3 to replace Turbine Bµilding with Reactor Building for distances to sample stations 0 Delete reference to Tech Spec 3.8.K. l and S.!AE in Section 8.2 0 Delete AOG from Section 8.2. l 0 Delete AOG from Section 8.2.1.1 0 Delete total body setpoint equation; beta skin dose is limiting for noble gas 0 Revise Section 8.2.1.2 to add Kr-85 values for Sg, revise the stack flow and make the example specific for Kr-85 and make set point based on Kr-85 skin dose 0 Revise Section 8.2.1.3 to remove reference to AOG and SJAE, make specific to Kr-85 and remove basis for total body set point 0 Delete Section 8.2.2, SJAE Noble Gas Activity Monitors 0 Revise Section 9.2 to delete references to equipment no longer in service (AOG, recombiner system, charcoal absorber system, gland seal system) 0 Delete met tower reporting requirement from Section 10.1 0 Revise Section 10.2 to delete milk sample reference. 0 Revise Section 10.4.2 to delete iodine from gaseous effluents Off-Site Dose Calculation Manual Rev.37 Page vii of xv Vermont Yankee Nuclear Power Station

ABSTRACT The VYNPS ODCM (Vermont Yankee Nuclear Power Station Off-Site Dose Calculation Manual) contains the effluent and environmental control limits, and approved methods to estimate the maximum individual doses and radionuclide concentrations occurring at or beyond the boundaries of the plant clue to normal plant operation. The effluent dose models are based on the U.S. NRC Regulatory Guide 1.109. Revision I. With initial approval by the U.S. Nuclear Regulatory Commission and the VYNPS Plant Management and approval of subsequent revisions by the Plant Management (as per the Technical Specifications) the methods contained in the ODCM arc suitable to demonstrate compliance with effluent controls. Off-Site Dose Calculation Manual Rev. 37 Page viii of xv Vermont Yankee Nuclear Power Station

TABLE OF CONTENTS REVISION

SUMMARY

.............................................................................................................. ii ABSTRACT ................................................................................................................................ ix TABLE OF CONTENTS .............................................................................................................. x LIST OF TABLES ..................................................................................................................... xiv LIST OF FIGURES ................................................................................................................... xvi Section-Page

1.0 INTRODUCTION

........................................................................................................ 1-1 1.1    Summary of Methods, Dose Factors, Limits, Constants, and Radiological Effluent Control Cross References ................................................................... 1-2 2.0      DEFINITIONS ............................................................................................................. 2-1 3/4.0    EFFLUENT AND ENVIRONMENT AL CONTROLS ...... :..................................... 3/4-1 3/4.1 Instrumentation .......................................................................................................... 3/4-1 3/4.1.1 Radioactive Liquid Effluent Instrumentation ....................................... :.................... 3/4-1 3/4.1.2 Radioactive Gaseous Effluent Instrumentation ......................................................... 3/4-6 3/4.2 Radioactive Liquid Effluents ................................................................................... 3/4-11 3/4.2.1 Liquid Effluents: Concentration .............................................................................. 3/4-11 3/4.2.2 Dose - Liquids .......................................................................................................... 3/4-17 3/4.2.3 Liquid Rad waste Treatment.. ................................................................................... 3/4-18 3/4.3 Radioactive Gaseous Effluents ................................................................................ 3/4-19 3/4.3.1 Gaseous Effluents: Dose Rate ................................................................................ 3/4-19 314.3.2 Gaseous Efl1uents: Dose from Noble Gases ........................................................... 3/4-22 3/4.3.3 Gaseous Effluents: Dose from Radioactive Material in Particulate Form and Tritium ..................................... ~ ........................................................................ 3/4-23 3/4.3.4 Deleted ..................................................................................................................... 3/4-24 3/4.3.5 Ventilation Exhaust Treatment ................................................................................ 3/4-25 3/4.3.6 Deleted ..................................................................................................................... 3/4-26 3/4.3.7 Deleted ..................................................................................................................... 3/4-27 3/4.4 Total Dose ................................................................................................................ 3/4-28 3/4.4. l Total Dose ................................................................................................................ 3/4-28 314.5 Radiological Environmental Monitoring ................................................................. 3/4-29 3/4.5.1 Radiological Environmental Monitoring Program .................................................. 3/4-29 3/4.5.2 Land Use Census ..................................................................................................... 3/4-38 3/4.5.3 Interlaboratory Comparison Program ...................................................................... 3/4-39 3/4.6 Effluent and Environmental Control Bases ............................................................. 3/4-40 Off-Site Dose Calculation Manual Rev. 37 Page ix of xv Vermont Yankee Nuclear Power Station

TABLE OF CONTENTS (Continued) Section-Page 5.0 METHODS TO CALCULATE OFF-SITE LIQUID CONCENTRATIONS .............. 5-1 5.1 Mel I1oci to Determine . F 1ENG an d c 1N<i ************************************************************************** 5- I 5.2 Method to Detennin'e Radionuclide Concentration for Each Liquid Effluent Pathway .......................................................................................................... 5-3 5.2.1 Satnple Tanks Pathways ............................................................................................... 5-3 5.2.2 Service Water Pathway ................................................................................................. 5-3 5.2.3 Deleted .......................................................................................................................... 5-4 5.2.4 Intercepted Groundwater Concentrations in Flowpalhs to the Connecticut River. ...... 5-4 5.2.5 Subsurface Contaminated Groundwater Concentrations in Flowpaths to the Connecticut River ................................................................................................... 5-4 6.0 OFF SITE DOSE CALCULATION METHODS ........................................................ 6-1 6.1 Introductory Concepts .................................................................................................. 6-1 6.2 Method to Calculate the Total Body Dose from Liquid Releases ................................ 6-3 6.3 Method to Calculate Maximum Organ Dose from Liquid Releases ............................ 6-9 6.4 Method to Calculate the Total Body Dose Rate from Noble Gases ........................... 6-11 6.5 Method to Calculate the Skin Dose Rate from Noble Gases ...................................... 6-16 6.6 Method to Calculate the Critical Organ Dose Rate Tritium and Particulates with T 1; 2 Greater Than 8 Days .................................................................................... 6-22 6.7 Deleted ........................................................................................................................ 6-26 6.8 Method to Calculate the Beta Air Dose from Noble Gases ........................................ 6-26 6.9 Method to Calculate the Critical Organ Dose from Tritium and Particulates ............ 6-29 6.10 Receptor Points and Long-Term Average Atmospheric Dispersion Factors for Important Expostire Pathways .............................................................................. 6-36 6.11 Method to Calculate Direct Dose from Plant Operation ........................................... 6-41 6.12 Cumulative Doses ....................................................................................................... 6-43 7.0 ENVIRONMENTAL MONITORING PROGRAM .................................................... 7-1 8.0 SETPOINT DETERMINATIONS ............................................................................... 8-1 8.1 Liquid Effluent Instrumentation Setpoints ................................................................... 8-1 8.2 Gaseous Effluent Instrumentation Setpoints .............................................................. 8-10 9.0 LIQUID AND GASEOUS EFFLUENT STREAMS, RADIATION MONITORS, AND RADW ASTE TREATMENT SYSTEMS .......................................................... 9-1 9.1 In-Plant Radioactive Liquid Effluent Pathways ........................................................... 9-1 9.2 In-Plant Radioactive Gaseous Effluent Pathways ........................................................ 9-4 9.3 Subsurface Groundwater Pathways to the Connecticut River. ..................................... 9-4 10.0 REPORTING REQUIREMENTS .............................................................................. 10-1 R. REFERENCES ............................................................................................................. R-1 Off-Site Dose Calculation Manual Rev.37 Page x of xv Vermont Yankee Nuclear Power Station

TABLE OF CONTENTS (Continued) Section-Page APPENDIX A: Method I Example Calculations ............................................................... Deleted APPENDIX B: Approval of Criteria for Disposal of Slightly Contaminated Septic Waste On-Site at Vennont Yankee ........................................................................................... B-1 APPENDIX C: Response to NRC/EG&G Evaluation of ODCM Update Through Revision 4 ......................................*.................................................................... *On File APPENDIX D: Assessment of Surveillance Criteria for Gas Releases from Waste Oil Incineration .............................................................................................................................. D-1 ,\PPENDIX E: NRC Safety Evaluation for Disposal of Slightly Contaminated Soil On-Site at VY (Below the Chem Lab Floor) -TAC No. M82152 ......... *On File APPENDIX F: Approval Pursuant to IOCFR20.2002 for On-Site Disposal of Cooling Tower Silt .......................................................................................... F-1 APPENDIX G: Maximum Permissible Concentrations (MPCs) in Air and Water Above Natural Background Taken from IOCFR20. l to 20.602, Appendix B ................................................................................................... : G-1 APPENDIX H: .................................................................... :................................................... H-1 I) "Request to Amend Previous Approvals Granted Under IOCFR20.302(a) for Disposal of Contaminated Septic Waste and Cooling Tower Silt to Allow for Disposal of Contaminated Soil," dated June 23,1999, BVY 99-80 .......................................................... H-2

2) "Supplement to Request to Amend Previous Approvals Granted Under 10CFR20.302(a) to Allow for Disposal of Contaminated Soil," dated January 4, 2000, BVY 00-02. H-19

3) "Vermont Yankee Nuclear Power Station, Request to Amend Previous Approvals Granted Under 10CFR20.302(a) to Allow for Disposal of Contaminated Soil (TAC No. MA5950)", dated June 15, 2000, NVY 00-58 ................................................................ H-37 Off-Site Dose Calculation Manual Rev. 37 Page xi of xv Vermont Yankee Nuclear Power Station

TABLE OF CONTENTS (Continued) Section-Page APPENDIX I: ........................................................................................................................... I-l I) "Request to Amend Previous Approval Granted Under 10CFR20.2002 for Disposal of Contaminated Soil", dated September 11, 2000 BVY 00-71 ............................................................................................. 1-2 "Vermont Yankee Nuclear Power Station - Safety Evaluation for an Amendment to an Approved l OCFR20.2002 Application (TAC No. MA9972)", dated June 26, 2001, NVY 01-66 ....................................... 1-5 APPENDIX J: ............................................................................................................................ .1-1

2) "Request to Amend Previous Approval Granted Pursuant to I OCFR20.200 I for Increase of the Annual Volume Limit and One-time Spreading of Current Inventory," dated October 4, 2004, BVY 04-110 ........................................................................................... J-2

3) "Safety Evaluation of Request to Amend Previous Approvals Granted Pursuant to I OCFR20.2002 - Vermont Yankee Nuclear Power Station (TAC No. MC5104)" dated July 19, 2005, NVY 05-090 .................... J-51

To access this document, go to the Electronic Document Management System. Search using ODCM.

Off-Site Dose Calculation Manual Rev. 37 Page xii of xv Vermont Yankee Nuclear Power Station

LIST OFT ABLES Number Title Section-Page 1.1.1 Summary of Radiological Effluent Controls and Implementing Equations 1-3 1.1.2 Summary of Methods to Calculate Unrestricted Area Liquid 1-6 Concentrations 1.1.3 Summary of Methods to Calculate Off-Site Doses from Liquid I-7 Concentrations 1.1.4 Summmy of Methods to Calculate Dose Rates 1-8 1.1.5 Summary of Methods to Calculate Doses to Air from Noble Gases 1-9 1.1.6 Summary of Methods to Calculate Dose to an Individual from Tritium, 1-10 Iodine, and Particulates in Gas Releases and Direct Radiation I. I. 7 Summary or ivietho<ls for Sdpoint Detenninations 1-11 1.1.8 Effluent and Environmental Controls Cross-Reference 1-12 1.1.9 (Table Deleted) 1.1.10 Dose Factors Specific for Vem1ont Yankee for Noble Gas Releases I-I 5 1.1.1 OA Combined Skin Dose Factors Specific for Vermont Yankee Ground 1-I 6 Level Noble Gas Releases I. I. I I Dose Factors Specific for Vermont Yankee for Liquid Releases I-I 7 I .1.12 Dose and Dose Rate Factors Specific for Vermont Yankee for Tritium 1-18 and Particulate Releases

2. I. I Definitions 2-2 2.1.2 Summary of Variables 2-5

3. I. I Liquid Effluent Monitoring Instrumentation 3/4-2 4.1. l Radioactive Liquid Effluent Monitoring Instrumentation Surveillance 3/4-4 Requirements 3.1.2 Gaseous Effluent Monitoring Instrumentation 3/4-7

4. I .2 Gaseous Effluent Monitoring Instrumentation Surveillance Requirements 3/4-9 4.2. I Radioactive Liquid Sampling and Analysis Program 3/4-12 4.3.1 Radioactive Gaseous Waste Sampling and Analysis Program 3/4-20 3.5. l Radiological Environmental Monitoring Program 3/4-30 3.5.2 Reporting Levels for Radioactivity Concentrations in Environmental 3/4-35 Samples 4.5.1 Detection Capabilities for Environmental Sample Analysis 3/4-36 6.2. l Environmental Parameters for Liquid Effluents at Vermont Yankee 6-7 Off-Site Dose Calculation Manual Rev. 37 Page xiii of xv Vermont Yankee Nuclear Power Station

LIST OFT ABLES (Continued) 6.2.2 Usage Factors for Various Liquid Pathways at Vennont Yankee 6-8 6.9.1 Environmental Parameters for Gaseous Effluents at Vennont Yankee 6-33 6.9.2 Usage Factors for Various Gaseous Pathways at Vennont Yankee 6-35 6.10.1 Atmospheric Dispersion Factors 6-39 6.10.2 Site Boundary Distances 6-40 6.10.3 Recirculation Correction Factors 6-41 7.1 Radiological Environmental Monitoring Stations 7-3 8.2.1 Relative Fractions of Core Inventory Noble Gases After Shutdown 8-16 Off-Site Dose Calculation Manual Rev.37 Page xiv of xv Vermont Yankee Nuclear Power Station

LIST OF FIGURES Number 7-1 Environmental Sampling Locations in Close Proxi1nity to Plant ............................................................................................. 7-5 7-2 Environmental Sampling Locations Within 5 kin of Plant. .................................................................................................... 7-6 7-3 Environmental Sampling Locations Greater Than 5 kn1 from Plant ...................................................................................... 7-7 7-4 TLD Locations in Close Proximity to Plant .................................................... 7-8 7-5 TLD Locations Within 5 km of Plant ............................................................... 7-9 7-6 TLD Locations Greater than 5 km from Plant ............................................... 7-10 9-1 Radioactive Liquid Effluent Streams, Radiation Monitors, and Radwaste Treatment System at Vennont Yankee ............................................ 9-9 9-2 Radioactive Gaseous Effluent Streams, Radiation Monitors, and Radwaste Treatment System at Vermont Yankee .......................................... 9-10 9-3 Subsurface Shallow Groundwater Streamtubes from the Plant Site to the Connecticut River. .......................................................... 9-11 9-4 Subsurface Deep Groundwater Streamtubes from the Plant Site to the Connecticut River. .......................................................... 9-12 Off-Site Dose Calculation Manual Rev.37 Page xv of xv Vermont Yankee Nuclear Power Station

1.0 INTRODUCTION

The ODCM (Off-Site Dose Calculation Manual) provid1.:s formal and approv1.:d methods for the 1.:alculation of off-site concentration, off-site doses. and erfluent monitor sl!lpoints ii1 order to comply with th1.: Vermont Yankee Control Limits which implement the program requirements of Technical Specili1.:ation 6.7.D. The ODCM forms the basis for plant procedures and is designed for use by the procedure writer. In addition, the ODCM will be useful to the writer of periodic reports required by the NRC on the dose consequences of plant operation. The dose methods contained herein follow accepted NRC guidance for calculation of doses necessary to demonstrate compliance with the dose objectives of Appendix I to IOCFR50 (Regulatory Guide I. I 09) unless otherwise noted in the text. Demonstration of compliance with the dose limits of40CFRl90 (sec Control 3.4. l) will be considered as demonstrating compliance with the 0.1 rem limit of I OCFR20. l 30 I (a)( I) for members of the public in unrestricted areas (Reference 56 FR 23374, third column.) It shall be the responsibility of the RP/Chemistry Manager to ensure that the ODCM is used in the performance of the surveillance requirements of the appropriate portions ofODCM Controls. The administration of the program for the onsite disposal of slightly contaminated waste, as described in Appendices. is also the responsibility of the RP/Chemistry Manager. All changes to the ODCM must be reviewed by the Independent Safety Review and approved by the Senior Manager, Production, in accordance with Technical Specification 6.7.B, prior to implementation. All approved changes shall be submitted to the NRC for their information in the Radioactive Effluent Release Report for the period in wh:.:h the chuag.::(s) was mad::: dTectin. Tb: plant's Dcct::nent Control Center (DCC) shall maintain the current version of the ODCM and issue under controlled distribution all approved changes to it. Off-Site Dose Calculation Manual Section 1.0 Rev. 33 Page I of 18 Vermont Yankee Nuclear Power Station

I. I Su111111arv of" Md hods. I )ose Faclors. I .i111its. ( 'onstants. and Radiological l~ffluent Control Cross-lfrl'crcnces This section su111111ari/.cs the dose calculation 111dhods. The concentration and sdpoint 111dhods arc also summarized in Table 1.1.2 through Table 1.1.7. as well as the l\kthod I Dose equatillns. Where more accurate dose cakulatillns an.: needed use the l\kthod II for the appropriate dose as described in Sections (J.2 through 6.9 and <>.11. The dose factors used in the equations arc in Tables 1.1.10 through 1.1.12 and the Regulatory I .imits arc summarized in Table 1.1.1.

              /\cross-reference ol"old Technical Spccilication sections to the new ODCM sections containing the cquiv;tlcnl ( "onlrols is presented in Table 1.1.8.

Special definitions and equation variables used in the ODCM arc in Tables 2.1.1 and 2.1.2. Off-Site Dose Calculation Manual Section 1.0 Rev. 33 Page 2of18 Vermont Yankee Nuclear Power Station

TABLE 1.1.1 Summarv of Radiological Effluent Controls and lmplementine: Equations c ontro categorv 1\1'et hod 11 ) L. . llTilt 3.2.1 Liquid Effluent Concentration Sum of the Fractions of Eq. 5-1 ~IO Effluent Concentration Limits [Excluding Noble Gases] Total Noble Gas Concentration Eq. 5-2 ~2 x I 0 -\tCi/cc 3.2.2 Liquid Effluent Dose Total Body Dose Eq. 6-1 ~ 1.5 mrem in a qtr.

                                                                                                           ~3.0  mrem in a yr.

Organ Dose Eq. 6-3 ~5 mrem in a qtr.

                                                                                                           ~l 0  mrem in a yr.

3.2.3 Liquid Radwaste Treatment Total Body Dose Eq. 6-1 ~0.06 mrem in a mo. Operability Organ Dose Eq. 6-3 ~0.2 mrem in a mo. 3.3.l Gaseous Effluents Dose Rate Total Body Dose Rate from Eq. 6-5 ~500 mrem/yr. Noble Gases Eq. 6-39 Skin Dose Rate from Noble Eq. 6-7 ~3000 mrem/yr. Gases Eq. 6-38 Organ Dose Rate from Tritium Eq. 6-16 ~1500 mrem/yr. and Particulates with T 112>8 Eq. 6-40 Days Off-Site Dose Calculation ~vlanual Section 1.0 Rev.33 Page 3of18 Vermont Yankee Nuclear Power Station

TABLE 1.1.1 (Continued) Summary oJ Radiological Effluent Controls and Implementing Equations c ontro category M e th 0 d iii L'1m1't 3.3.2 Gaseous Effluents Dose from Noble Gamma Air Dose from Eq. 6-21 ~ 5mrad in a qtr. Gases Noble Gases Eq. 6-41 ~l 0 mrad in a yr.

Beta Air Dose from Noble Eq. 6-23 ~l 0 mrad in a qtr. Gases Eq. 6-43 ~ 20 mrad in a yr. I 3.3.3 Gaseous Effluents Dose from Tritium and Particulates Organ Dose from Tritium and Particulates with Tic Eq. 6-25 Eq. 6-44

                                                                                                          ~
                                                                                                          ~15 7.5 mrem in a qtr.

mrem in a yr.

                                                  >8 Days 3.3.5      Ventilation Exhaust Treatment         Organ Dose                  Eq. 6-25                    ~0.3   mrem in a mo.

3.4.1 Total Dose (from All Sources) Total Body Dose Footnote (:2) ~25 mrem in a yr. Organ Dose No Thyroid ~25 mrem in a vr. Oft:Site Dose Calculation Manual Section 1.0 Rev.33 Page 4of18 Vermont Yankee Nuclear Power Station

TABLE I. I. I (Continued) Summary of Radiological Effluent Controls and Implementing Equations c ontro c ate gory Met ho d111 11111 3.1.l Liquid Effluent Monitor Setpoint Liquid Radwaste Discharge Alarm Setpoint Eq. 8-1 Control 3.2.1 Monitor 3.1.2 Gaseous Effluent Monitor Setpoint Plant Stack Noble Gas Alarm/Trip Setpoint for Skin Eq. 8-10 Control 3.3.1.a Activitv Monitors Dose Rate (Skin) (I) More accurate methods may be available (see subsequent chapters). (2) Effluent Control 3 .4.1 requires this evaluation only if twice the limit of Equations 6-1. 6-3, 6-21, 6-23, or 6-25 is reached. If this occurs a Method II calculation shall be made considering available information for pathways of exposure to real individuals from liquid, gaseous, and direct radiation sources. Off-Site Dose Calculation Manual Section 1.0 Rev. 33 Page 5 of 18 Vermont Yankee Nuclear Power Station

TABLE 1.1.2 Summarv of Methods to Calculate Unrestricted Area Liquid Concentrations Equation Reference Number Category Equation Section 5-1 Sum of the Fractions of Combined Cpi 5.1 Effluent Concentrations in Liquids Fl '"N("

                                                      . ="--<10 I

I '7"ECL - I [Except Noble Gases] 5-2 CNci(~tCi)=ICNG Total Activity of Dissolved and 5.1 Entrained Noble Gases from all :<:;2E-04 I ml j Ii Station Sources Off-Site Dose Calculation Manual Section 1.0 Rev.33 Page 6of18 Vermont Yankee Nuclear Power Station

TABLE 1.1.3 Summary of Methods to Calculate Off-Site Doses from Liquid Concentrations Equation Reference Number Category Equation Section 6-1 Total Body Dose D1h (mrem) =~Qi DFL1h 6.2.1 l 6-3 Maximum Organ Dose Dmo(mrem)= I Qi DFLmo 6.3.1 i Off-Site Dose Calculation Manual Section 1.0 Rev. 33 Page 7of18 Vermont Yankee Nuclear Power Station

TABLE 1.1.4 Summary of Methods to Calculate Dose Rates Equation Reterence Number Category Equation Section 6-5 Total Body Dose Rate from

                                                                  = 0.75IQ~TDFBi 6.4.1 Noble Gases Released from           Rth{ mreml yr              i Stack 6-39     Total Body Dose Rate from Rth{ mrem l = 7.4IQ~'LDFBi 6.4.1 Noble Gases Released from yr            I Ground 6-7     Skin Dose Rate from Noble Gases Released from Stack          R . ( mrem

-i"\llS - -

yr J- L:(l.DF' i i is 6.5.1 6-38 Skin Dose Rate from Noble Gases Released from Ground Rsk111g

                                                      . (-   mrem yr J- L: (l"'DF' i
                                                                                 .i     j~

6.5.I 6-16 Critical Organ Dose Rate from 6.6.1 Stack Release of Tritium and R cos( mrem) =I Qi *m DFG ,sicu Particulates with T 112 >8 Days yr i 6-40 Critical Organ Dose Rate from Ground Level Release of Tritium and Particulates with Rcog yr J

                                                   .. ( mrem ~ L:Q""DFG'g1co i

1 6.6.I T112 >8 Days Off-Site Dose Calculation Manual Section 1.0 Rev.33 Page 8of18 Vermont Yankee Nuclear Power Station L_

TABLE 1.1.5 Summarv or Mdhods to Calculate Doses to Air from Nobk Gases Equation Category Equation Reference Number Section 6-21 Gamma Dose to Air from D!'111...s ( mrad ) = 0.024IQ* ST

DF.Y 6.7.I Noble Gases Released from . 1 1 I

Stack 6-41 Gamma Dose to Air from D~irg (mrad) = 0.23~ Q~ 1 LDJ~Y 6.7.1 Noble Gases Released from I Ground Level 6-23 Beta Dose to Air from Noble Gases Released from Stack fl ( Dairs mrad ) = 0.0502: Q;ST DFi0 6.8.1 i 6-43 Beta Dose to Air from Noble Gases Released from Ground D~irg (mrad) = 1.16~ Q~1 LDFt 6.8.1 I Level Off-Site Dose Calculation Manual Section 1.0 Rev. 33 Page 9of18 Vermont Yankee Nuclear Power Station

TABLE 1.1.6 Summary of Methods to Calculate Dose to an Individual from Tritium. Iodine, and Particulates in Gas Releases and Direct Radiation Equation Reference Number Cakgory Equation Section 6-25 Dose to Critical Organ 6.9.1 from Stack Release of Dens ( mrem ) = L: QiSTP DFG, sico I Tritium and Particulates 6-44 Dose to Critical Organ 6.9.1 from Ground Level Release Deng ( 1nre1n ) = "L. Qim.P DFG"' gico i - of Tritium and Particulates Direct Dose 6-27b II Ill 6.11.1 DMs1.R.11 = ~)~)Ri)I mt~ti i=I j=I . 6-27c 6.11.1 6-27d 6.11.1 Off-Site Dose Calculation Manual Section 1.0 Rev. 33 Page IO of 18 Vermont Yankee Nuclear Power Station

TABLE 1.1.7 Summary of Methods for Setpoint Detcrminations Equation Reference Number Category Equation Section 8-1 Liquid Effluents: Liquid Radwastc Discharge 8.1.1.1 Monitor ( 17/350) L ( Rsrt cps ) = OF. S1 L. Cmi OF ITIIO I Gaseous Effluents: Plant Stack (RR-108-IA, RR-I 08-1 B) Noble Gas Activity Monitors 8-10 Skin 8.2.1.1 R skui (cpm) = 3000 S J_ _l_ srt g F OF' c Off-Site Dose Calculation Manual Section 1.0 Rev. 33 Page 11of18 Vermont Yankee Nuclear Power Station

TABLE 1.1.8 Effluent and Environmental Controls Cross-Rcterence Original Technical Revised ODCM Control Topic Specification Section Control Section INSTRUMENTATION Radioactive Liquid Effluent 3/4.9.A 3/4.1.1 Instrumentation Effluent instrumentation list Table 3.9.1 Table 3.1.1 Instrumcnt surveillance requirements Table 4.9.1 Table 4.1.1 Radioactive Gaseous Effluent* 3/4.9.B 3/4.1.2 Instrumentation Enlucnt i11stn11nc:aaliu11 !is~ Tabic 3.9.2 Table 3.1.2 Instrumentation requirements Table 4.9.2 Table 4.1.2 RADIOACTIVE LIQUID EFFLUENTS Concentration 3/4.8.A 3/4.2.1 Liquid waste sampling & analysis Table 4.8. I Table 4.2.1 program Dose - Liquids 3/4.8.B 3/4.2.2 Liquid Radwaste Treatment 3/4.8.C 3/4.2.3 RADIOACTIVE GASEOUS EFFLUENTS Dose Rate 3/4.8.E 3/4.3.1 C<:seous wa~;h: sampling & a:mlysis Table 4.8.2 Table 4.3. i program Dose from Noble Gases 3/4.8.F 3/4.3.2 I Dose from Tritium and Radionuclides in 3/4.8.G 3/4.3.3 Particulate Form Ventilation Exhaust Treatment 3/4.8.1 3/4.3.5 I TOTAL DOSE Total Dose 3/4.8.M 3/4.4. l Off-Site Dose Calculation Manual Section 1.0 Rev. 33 Page12ofl8 Vermont Yankee Nuclear Power Station

TABLE 1.1.8 (Continued) RADIOLOGICAL ENVIRONMENTAL MONITORING Radiological Environmental Monitoring 3/4.9.C 3/4.5. I Program Listing of required monitorin~ criteria Table 3.9.3 Table 3.5. I Reporting levels for radioactivity in Table 3.9.4 Table 3.5.2 samples Detector capability for environmental Table 4.9.3 Table 4.5. I analysis Land U sc Census 3/4.9.D 3/4.5.2 lntercomparison Program 3/4.9.E 3/4.5.3 EFFLUENT CONTROL BASES Bases: 3.8 & 3.9 3/4.6 UNIQUE REPORTING REQUIREMENTS Annual Radioactive Effluent Release 6.7.C.I 10.1 Report Environmental Radiological Monitoring 6.7.C.3 10.2 Special Reports 6.7.C.2 10.3 Major Changes to Radioactive Liquid, 6.14 10.4 Gaseous, and Solid Waste Treatment Systems Off-Site Dose Calculation Manual Section 1.0 Rev. 33 Page 13of18 Vermont Yankee Nuclear Power Station

TABLE 1.1.9 (Deleted) Off-Site Dose Calculation Manual Section 1.0 Rev. 33 Page 14of18 Vermont Yankee Nuclear Power Station

TABLE 1.1.10 Dose Factors Specific for Vermont Yankee for Noble Gas Releases Radionuclide Gamma Total Body Beta Skin Combined Skin Beta Air Gamma Air Dose Factor Dose Factor Dose Factor Dose Factor Dose Factor (Stack Release) OF Bi OF Si DF,; DF~ on mre~1-m ) 3 mra~-m ) mra~-m ) 3 3 3 ( mrem-m pCi-yr

                                      )  (

pC1- yr ( mrem- sec p Ci- yr ) ( pC1- yr ( pC1- yr Kr-85 1.61 E-05 I .34E-03 2.15£-03 l .95E-03 1.72E-05

8.84E-03 = 8.84 x I 0-3 Off-Site Dose Calculation tv1anual Section 1.0 Rev. 33 Page 15 of 18 Vermont Yankee Nuclear Power Station

TABLE I. I. I OA Combined Skin Dose Factors Specific for Vermont Yankee Ground Level Noble Gas Releases Radionuclide , (-mrem-DF,g - . -sec-)

                                                          ~t Ci- yr Kr-85                           4.91E-02 Off-Site Dose Calculation Manual Section 1.0 Rev. 33 Page 16of18 Vermont Yankee Nuclear Power Station

TABLE 1.1.11 Dose Factors Specific for Vermont Yankee for Liquid Rckascs Rad ion ucl idc Total Body Maximum Organ Dose Factor Dose Factor

                                       . (----z.;-

I> I* L II h mrcm) D F L; mo (----z.;- mrcm) ll-3 2.06E-04 2.06E-04 Mn-54 2.08E-O I 3.00E+OO Fc-55 4. I 8E-02 2.54E-O I Co-60 2.13E-Ol 1.28E+OO Zn-(i5 8.06E+OO 1.64E+OI Sr-90 4.23E+O I I .67E+02 Zr-95 4.21 E-04 I .36E-O I Ag-I !Om 6.90E-03 7.02E-OI Sb-125 7.52E-03 1. l 5E-O I Cs-134 1.28E+02 1.60E+02 Cs-137 7.58E+O I 1.21E+02 Off-Site Dose Calculation Manual Section 1.0 Rev. 33 Page 17of18 Vermont Yankee Nuclear Power Station

TABLE 1.1.12 Dose and Dose Rate Factors Speci lie for Vcrmont Yankee for Tritium and Particulate Releases Stack Release Ground Level Release

Radio- Critical Organ Critical Organ Critical Organ Critical Organ nuclide Dose Factor Dose Rate Factor Dose Factor Dose Rate Factor

                                       ,    ( mrem-sec)                              , ( mrcm-scc)

DFGsirn ( ----cl mrem) DFG siw cI DFGgicu ( ----cl mrem) DFG '"'"" yr-~L c*I yr-~L 1-1-3 4.79E-04 1.51 E-02 1.IOE-02 3.47E-O I C-14 2.91E-01 9.18E+OO 6.68E+OO 2.1 IE+02 Mn-54 9.34E-Ol 3.69E+O I 4.97E+OO 1.95E+02 Fe-55 4.22E-O I 1.33L+O i 2.20L+OO 6.9..+E+OI Co-57 2.73E-01 1.03E+O I 1.42E+OO 5.30E+OI Co-60 l.02E+O I 4.54£+02 5.32E+O I 2.32E+03 Zn-65 4.96E+OO 1.60E+02 2.53E+OI 8.20E+02 Se-75 3.19E+OO 1.03E+02 1.63E+OI 5.27E+02 Sr-90 5.75E+02 1.81 E+04 3.03E+03 9.56E+04 Zr-95 9.23E-01 3.04E+O I 4.84E+OO 1.59E+02 Sb-125 l .67E+OO 6.53E+OI 8.59E+OO 3.34E+02 Cs-134 2.12E+OI 7.06E+02 1.09E+02 3.63E+03 Cs-137 2.17E+OI 7.41E+02 1.l IE+02 3.78E+03 Ce-144 5.14E+OO l.62E+02 2.69E+OI 8.52E+02 J

The release point reference is the Reactor Building. These dose and dose rate factors arc conservative for potentiai release applications associated with ground level effluents from other major facilities (i.e., Turbine Building and CAB).

Off-Site Dose Calculation Manual Section 1.0 Rev. 33 Page 18of18 Vermont Yankee Nuclear Power Station

2.0 DEFINITIONS This section lists definitions (Table 2.1. l) and dose equation variable names (Table 2.1.2) which arc utilized in the VY ODCM. Off-Site Dose Calculation Manual Section 2.0 Rev.36 Page I of9 Vermont Yankee Nuclear Power Station

TABLE 2.1.1 Definitions J I. Groundwater - For purposes of the ODCM, groundwater is defined as subsurface water which is either shallow, deep or in bedrock layers. Shallow and deep groundwater wells are sampled to determine the flow rate and contamination concentrations of groundwater flowing to the Connecticut River above or on the bedrock layer. Bedrock groundwater wells, utilized for drinking water purposes both on and off the plant site, are monitored for radioactive contamination as part of the REMP.

2. Immediate - Immediate means that the required action will be initiated as soon as practicable considering the safe operation of the unit and the importance of the required action.

3. lnstmment Calibration - An instmment calibration means the adjustment of an instrument signal output so that it corresponds, within acceptable range and accuracy, to a known value(s) of the parameter which the instrument monitors. Calibration shall encompass the entire instrument including actuation, alarm, or trip. Response time as specified is not part of the routine instrument calibration but will be checked once per operating cycle.

4. Instrument Check - An instrument check is qualitative determination of acceptable operability by observation of instrument behavior during operation. This determination shall include, where possible, comparison of the instrument with other independent instruments measuring the same variable.

5. Instrument Functional Test - An instrument functional test shall be:

a. Analog channels - the injection of a signal into the channel as close to the sensor as practi~ablc to verify uperability including alarm anJ/or Lrip furn..:lions.

b. Bistable channels - the injection of a signal into the sensor to verify the functionality including alarm and/or trip functions.

6. Intercepted Groundwater- For the purposes of the ODCM, intercepted groundwater is defined as groundwater which has infiltrated Turbine Building sumps or trenches which have no contact with water from plant systems and groundwater removed from the saturated zone soil adjacent to or underlying structures, equipment, excavations, etc. to reduce groundwater infiltration.

7. Liquid Waste Discharge - For the purposes of the ODCM, liquid waste discharges are plant process water treated with the liquid radwaste system and discharged through the liquid radwaste monitor, using the installed flow monitor and process radiation monitor instrumentation.

Off-Site Dose Calculation Manual Section 2.0 Rev.36 Page 2 of9 Vermont Yankee Nuclear Power Station

TABLE 2.1.1 (Continued)

8. Off-Site Dose Calculation Manual (ODCM) - A manual containing the current methodology and parameters used in the calculation of off-site doses due to radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring alann/trip setpoints, and in the conduction of the environmental radiological monitoring program. The ODCM shall also contain ( 1) the Radioactive Erlluent Controls (including the Radiological Environmental Monitoring) Program required by Technical Specification 6.7.D, and (2) descriptions of the infomrntion that should be included in the annual Radioactive Effluent Release Report and Annual Radiological Environmental Operating Report required by Technical Specifications 6.6.D and 6.6.E, respectively.

9. Site Boundary -The site boundary is shown in Plant Drawing 5920-6245.

10. Source Check - The qualitative assessment of channel response when the channel sensor is exposed to a radioactive source.

11. Streamtube(s) - Defined as flows of subsurface groundwater (having discrete width, depth and flow rate characteristics) in either the shallow or deep layers of permeable soils above the bedrock layer at the plant site. Streamtube flows are from west to east, towards the Connecticut River, and are assumed to discharge into the Connecticut River.

12. Surveillance Frequency - Unless otherwise stated in these specifications, periodic surveillance tests, checks, calibrations, and examinations shall be performed within the specified surveillance intervals. These intervals may be adjusted plus 25%. The operating cycle interval is considered to be 18 months and the tolerance stated above is applicable.

13. Surveillance Interval - The surveillance interval is the calendar time between surveillance tests, checks, calibrations, and examinations to be performed upon an instrument or component when it is required to be functional. These tests unless otherwise stated in these specifications may be waived when the instrument, component, or system is not required to be functional, but these tests shall be performed on the instrument, component, or system prior to being required to be functional.

14. Ventilation Exhaust Treatment System -The Radwaste Building and AOG Building ventilation HEP A filters are ventilation exhaust treatment systems which have been designed and installed to reduce radioactive material in particulate form in gaseous effluent by passing ventilation air through HEP A filters for the purpose of removing radioactive particulates from the gaseous exhaust stream prior to release to the environment. Engineered safety feature atmospheric cleanup systems, such as the Standby Gas Treatment (SBGT) System, are not considered to be ventilation exhaust treatment system components.

Off-Site Dose Calculation Manual Section 2.0 Rev.36 Page 3 of9 Vermont Yankee Nuclear Power Station

TABLE 2.1.1 (Continued)

15. Vent/Purging - Vent/purging is the controlled process of discharging air or gas from the primary containment to control temperature, pressure, humidity, concentration or other operating conditions.

Off-Site Dose Calculation Manual Section 2.0 Rev.36 Page 4 of9 Vermont Yankee Nuclear Power Station

Tl\BLE'.2.1.2 Summmy of Variables Variable Definition Units Concentration at point of discharge to an ~LCi/ml unrestricted area of dissolved and entrained noble gas "i" in liquid pathways from all station sources. Total activity of all dissolved and entrained noble ~LCi gases in liquid pathways from all station sources. ml Concentration of radionuclide "i" at the point of ~LCi liquid discharge to an unrestricted area. ml Concentration of radionuclide "i." cc Concentration, exclusive of noble gases, of µCi radionuclide "i" from tank "p" at point of discharge ml to an unrestricted area. Concentration of radionuclide "i" in mixture at the ~LCi monitor. ml fl Beta dose to air from stack release. mrad Dai rs 11 Beta dose to air from ground level release. mrad Dairg Gamma dose to air from stack release.

mrad D~irs y Gamma dose to air from ground level release. mrad Dairg Deas Dose to critical organ from stack release. mrem Dcog Dose to the critical organ from ground level mrem release.

Gamma dose to air, corrected for finite cloud. mrad Dhnitc Dose to the maximum organ. mrem Dose to skin from beta and gamma. mrem Off-Site Dose Calculation Manual Section 2.0 Rev.36 Page 5 of9 Vermont Yankee Nuclear Power Station

TABLE 2.1.2 (Continued) Variable Definition Units D11i Dose to the total body. mrcm DF Dilution factor. ratio DFmin Minimum allowable dilution factor. ratio DF'c Composite skin dose factor. mrem-sec pCi-yr DFBi Total body gamma dose factor for nuclide 11 i. 11 mrem-111 3 pCi-yr DFBc Composite Lola: budy dose faetor. mrem-111-1 pCi-yr DFLi1h Site-specific, total body dose factor for a liquid mrem release of nuclide 11 i. 11 Ci DFLimn Site-specific, maximum organ dose factor for a imem liquid release of nuclide 11 i. 11 Ci DFGsico Site-specific, critical organ dose factor for a stack mrem gaseous release of nuclide 11 i. 11 Ci DFG'sicn Site-specific, critical organ d0se rate factor for a mrem-sec stack gaseous release of nuclide 11 i. 11 µCi-yr DFGgico Site-specific, critical organ dose factor for a ground mrem level gaseous release of nuclide 11 i. 11 Ci DFG'gico Site-specific, critical organ dose rate factor for a mrem-sec ground level gaseous release of nuclide 11 i. 11 µCi-yr DFSi Beta skin dose factor for nuclide 11 i. 11 mrem-m 3 pCi-yr DF'is- Combined skin dose factor for nuclide "i" from a mrem-sec stack release. µCi-yr Off-Site Dose Calculation Manual Section 2.0 Rev.36 Page 6 of9 Vermont Yankee Nuclear Power Station

TABLE 2.1.2 (Continued) Variable Definition Units Combined skin dose factor for nuclide 11 i11 from a mrem-scc ground level release. ~LCi- yr OF{ Gamma air dose factor for nuclide 11 i. 11 mrad-m 3 pCi-yr Beta air dose factor for nuclide "i. 11 mrad-m 3 pCi-yr Critical organ dose rate due to particulates mrem released from stack. yr R1:og Critical organ dose rate due to particulates mrcm released from ground. yr Rskins Skin dose rate due to stack release of noble gases. mrem yr Rsking Skin dose rate due to ground release of noble mrem gases. yr Total body dose rate due to noble gases from stack mrem release. yr Total body dose rate due to noble gases from mrem ground level release. yr D/Q Deposition factor for dry deposition of particulates. Flow rate out of discharge canal. gpm Flow rate past liquid radwaste monitor. gpm F Flow rate past gaseous radwaste monitor. cc sec FENG Sum of the fractions of combined effluent fraction I concentrations in liquid pathways (excluding noble gases). Off-Site Dose Calculation Manual Section 2.0 Rev.36 Page 7 of9 Vermont Yankee Nuclear Power Station

TABLE 2.1.2 (Continued) Variable Definition Units Annual average effluent concentration limit for µCi radionuclide "i" (I OCFR20. l 001-20.240 I, cc Appendix B, Table 2, Column 2) Release for radionuclide "i" from the point of cunes interest. Release rate for radionuclide "i" at the point of ~tCi interest. sec The noble gas radionuclide "i" release rate at the ~tCi plant slack. sec The noble gas radionuclide "i" release rate from ~tCi ground level. sec The tritium and particulate radionuclide "i" release ~tCi rate from the plant slack. sec The tritium and pmticulate radionuclide "i" release ~tCi rate from ground level. sec The release of noble gas radionuclide "i" from the cunes plant stack. The release ;_)f noble gas radionuclide "i" from ground level. The release of tritium and particulate radionuclide cunes "i" from the plant stack. The release of tritium and particulate radionuclide cunes "i" from ground level. L Liquid monitor response for the limiting cps Rspt concentration at the point of discharge. skin Response of the noble gas monitor at the limiting cpm R spt skin dose rate. tb Response of the noble gas monitor to limiting total cpm Rspt body dose rate. Off-Site Dose Calculation Manual Section 2.0 Rev. 36 Page 8 of9 Vermont Yankee Nuclear Power Station

TABLE 2.1.2 (Continued) Variable Definition Units Detector counting efficiency from the most recent cpm mR/hr

                                                                                 --or--

gas monitor calibration. ~LCi/cc ~tCi/cc Sgi Detector counting efficiency for noble gas "i." cpm mR/hr

                                                                                 ---or---
                                                                                  ~LCi/cc  ~LCi/cc S1                      Detector counting efficiency from the most recent           cps liquid monitor calibration.                              ~tCi/ml S1i                      Detector counting efficiency for radionuclide ".i"         cps
                                                                                  ~tCi/ml I X/Qs                     Annual or long-term average undepleted atmospheric dispersion factor for stack release.

sec 3 m X!Qg Annual or long-term average undepleted sec atmospheric dispersion factor for ground level m 3 release. Effective annual or long-term average gamma sec [X/Q]~ - atmospheric dispersion factor. m 3 Effective annual or long-tcnn avcra6c gamma sec [X/Qj~ - atmospheric dispersion factor for a ground level m 3 release. Off-Site Dose Calculation Manual Section 2.0 Rev. 36 Page 9 of9 Vermont Yankee Nuclear Power Station

314.0 EFFLUENT AND ENVIRONMENTAL CONTROLS This section includes the effluent and environmental controls that were originally part of the V ennont Yankee Technical Speci ii cations. These controls were relocated into the ODCM without any substantial changes, in accordance with NRC Generic Letter 89-01. Text and tables were reformatted to the style of the ODCM. The various controls were renumbered from the original numbering scheme of the Technical Specifications. A cross-reference of the old Technical Specifications section to the new ODCM section is presented in Table 1.1.8. 314.1 INSTRUMENTATION 3/4.1.l Radioactive Liquid Effluent Instrumentation CONTROLS 3.1.1 The radioactive liquid effluent monitoring instrumentation channel shall be functional in accordance with Control Table 3.1.1 with their alarm setpoints set to ensure that the limits ofControl 3.2.1 are not exceeded. APPLICABILITY: During periods of release through monitored pathways as listed on Table 3.1.1. ACTION:

a. With a radioactive liquid effluent monitoring instrumentation channel alarm/trip setpoint less conservative than a value which will ensure that the limits of Control 3.2.1 are met, without delay suspend the release of radioactive liquid effluents monitored by the affected channel or change the setpoint so that it is acceptably conservative or declare the channel non-functional.

b. With one or more radioactive liquid effluent monitoring instrumentation channels non-functional, take the action shown in Table 3 .1.1.

SURVEILLANCE REQUIREMENTS 4.1.1.a Each radioactive liquid effluent monitoring instrumentation channel shall be tested and calibrated as indicated in Table 4.1.1. 4.1.1.b The setpoints for monitoring instrumentation shall be determined in accordance with the ODCM (Section 8.1 ). Off-Site Dose Calculation Manual Section 3/4 Rev. 37 Page 1of47 Vermont Yankee Nuclear Power Station

TABLE 3.1.1 Liquid Efl1uent Monitoring lnsln11nenlalion Minimum Channels Functional Notes I. Gross Radioactivity Monitors not Providing Automatic Termination of Release

a. Liquid Radwaste Discharge Monitor I* 1,4 (RM-17-350)

b. Service Water Discharge Monitor I 2,4 (RM-17-351)

2. Flow Rate Measurement Devices

a. Liquid Radwaste Discharge Flow Rate I* 3,4 Monitor

( FIT-20-485/442)

During releases via this pathway Off-Site Dose Calculation Manual Section 3/4 Rev. 37 Page 2 of 47 Vermont Yankee Nuclear Power Station

TABLE 3.1.1 NOTATION NOTE l - With the number of channels functional less than required by the minimum channels functional requirement, effluent releases may continue provided that prior to initiating a release:

a. At least two independent samples are analyzed in accordance with Control 4.2.1, and

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

Otherwise, suspend release of radioactive effluents via this pathway. NOT:: 2 - With th~ number or channels runctional less than n:quircd by t!1e minimum channels functional requirement, effluent releases via this pathway may continue provided that, at least once per 24 hours, grab samples arc collected and analyzed for gross radioactivity (beta or gamma) at a lower limit of detection of at least 10-7 microcurie/ml. NOTE 3 - With the number of channels functional less than required by the minimum channels functional requirement, effluent releases via this pathway may continue provided the flow rate is estimated at least once per 4 hours during actual releases. Pump performance curves may be used to estimate flow. NOTE 4 - With the number of channels functional less than required by the minimum channels functional requirement, exert reasonable efforts to return the imtrumem( s) io ;unctional status prior to the next release. Off-Site Dose Calculation Manual Section 3/4 Rev.37 Page 3 of47 Vermont Yankee Nuclear Power Station

TABLE4.1.l Radioactive Liquid Eflluent Monitoring Instrumentation Surveillance Requirements Instrument Instrument Source Instmment Functional lnstmment Check Check Calibration Test I. Gross Radioactivity Monitors not Providing Automatic Termination of Release

a. Liquid Radwaste Once each Prior to each Once each Once each Discharge Monitor (3) day* release, but 18 months ( l) quarter (2) no more than once each month

b. Service Water Once each Once each Once each Once each Discharge Monitor (3) day month 18 months ( l) quarter (2)

2. Flow Rate Measurement Devices

a. Liquid Radwaste Once each Not Not Once each Discharge Flow Rate day* Applicable Applicable quarter*

Monitor

During releases via this pathway.

Off-Site Dose Calculation Manual Section 3/4 Rev. 37 Page 4 of47 Vermont Yankee Nuclear Power Station

TABLE4.l.I NOTATION (I) The Instrument Calibration for radioactivity measurement instrumentation shall include the use of a known (traceable to National Institute for Standards and Technology) liquid radioactive source positioned in a reproducible geometry with respect to the sensor. These standards shall permit calibrating the system over its normal operating range of energy and rate. (2) The Instrument Functional Test shall also demonstrate the Control Room alann annunciation occurs if any of the following conditions exists: (a) Instrument indicate measured levels above the alarm setpoint. ( b) Circuit failure. (c) Instrnm*,'nt indic<~tcs a t!ownsc~ile failur.:. (d) Instrument controls not set in operate mode. (3) The alarm setpoints of these channels shall be determined and adjusted in accordance with the methodology and parameters in the Off-Site Dose Calculation Manual (see Section 8.1 ). Off-Site Dose Calculation Manual Section 3/4 Rev. 37 Page 5 of47 Vermont Yankee Nuclear Power Station

3/4. l INSTRUMENT J\ TION 3/4. l.2 Radioactive Gaseous Effluent Instnunenlation CONTROLS 3.1.2 The gaseous process and effluent monitoring instrumentation channels shall be functional in accordance with Control Table 3.1.2 with their alarm/trip setpoints set to ensure that the limits of Controls 3.3.1.a are not exceeded. APPLICABILITY: As shown in Table 3.1.2. ACTION:

a. With a gaseous process or effluent monitoring instrumentation channel alann/trip setpoint less conservative than a value which will ensure that the limits ofControl 3.3.1.a are met, immediately take actions to suspend the release of radioactive gaseous effluents monitored by the affected channel, or declare the channel non-functional, or change the setpoint so it is acceptably conservative.

b. With less than the minimum number of radioactive gaseous effluent monitoring instrumentation channels functional, take actions noted in Table 3 .1.2.

SURVEILLANCE REQUIREMENTS 4.1.2.a Each gaseous process or effluent monitoring instrumentation channel shall be tested and calibrated as indicated in Table 4.1.2. 4.1.2.b. The setpoints for monitoring instrumentation shall be determined in accordance with the ODCM (Section 8.2). Off-Site Dose Calculation Manual Section 3/4 Rev.37 Page 6 of47 Vermont Yankee Nuclear Power Station

TABLE 3.1.2 Gaseous Effluent Monitoring lnstmmentation Minimum Channels Instrument Functional Notes

3. Plant Stack

a. Noble Gas Activity Monitor I 5, 7, 10 (RM-17-156, RM-17-157)

b. Iodine Sampler Cartridge I 4, 5

c. Particulate Sampler Filter I 4,5

d. Sampler Flow Integrator I I, 5 (Fl-17-156/157)

e. Stack Flow Rate Monitor I 1, 5 (Fl-108-22)

Off-Site Dose Calculation Manual Section 3/4 Rev.37 Page 7 of47 Vermont Yankee Nuclear Power Station

TABLE 3.1.2 NOT J\TION NOTE 1 - With the number or channels functional less than required by the minimum channels functional requirement, effluent releases via this pathway may continue provided the flow ratt! is estimated at least once per 4 hours. NOTE 2 - Deleted NOTE 3 - Deleted NOTE 4 - With the number of channels functional less than required by the minimum channels functional requirement, effluent releases via the affected pathway may continue provided samples are continuously collected with auxiliary sampling equipment. NOTE 5 - With the number of channels functional less than required by the minimum channels functional requirement, exert reasonable efforts to return the instrument(s) to functional status within 30 days. NOTE 6 - Deleted NOTE 7 - The alann/trip setpoints of these channels shall be determined and adjusted in accordance with the methodology and parameters in the Off-Site Dose Calculation Manual (ODCM). NOTE 8 - Deleted NOTE 9 - Deleted NOTE 10 - With the number of channels fum:tional less than required by the ~1i1~imum channels functional requirement, effluent releases via this pathway may continue provided grab samples are taken at least once per 12 hours and these samples are analyzed for gross activity within 24 hours. Off-Site Dose Calculation Manual Section 3/4 Rev. 37 Page 8 of 47 Vermont Yankee Nuclear Power Station

TABLE 4.1.2 Gaseous Effluent Monitoring lnstrnmcntation Surveillance Requirements Instrument Instnunent Instrnment Functional lnstrnment Check Source Check Calibration Test

3. Plant Stack

a. Noble Gas Activity Once each day Once each Once each Once each Monitor month 18 months (3) quarter

b. Sampler Flow Once each Not Once each Not Integrator week Applicable 18 months Applicable

c. System Flow Rate Once each day Not Once c11ch Once c:1ch Monitor Applicable 18 months quarter Off-Site Dose Calculation Manual Section 3/4 Rev. 37 Page 9 of47 Vermont Yankee Nuclear Power Station

TABLE 4.1.2 NOTATION ( 1) Deleted (2) Deleted (3) The Instrument Calibration for radioactivity measurement instrnmentation shall include the use of a known (traceable to National Institute for Standards and Technology) radioactive source positioned in a reproducible geometry with respect to the sensor. These standards should pe1mit calibrating the system over its normal operating range of rate capabilities. (4) Deleted Off-Site Dose Calculation Manual Section 3/4 Rev. 37 Page 10 of 47 Vermont Yankee Nuclear Power Station

3/4.2 RADIOACTIVE LIQUID EFFLUENTS 3/4.2. I Liquid Efl1ucnt Concentration CONTROLS 3.2.1 The concentration of radioactive material in liquid efl1uents released in liquid waste effluents, intercepted groundwater released via stonn drain or groundwater flowing to the Connecticut River from the site in radioactive concentrations above background (Unrestricted Areas for liquids is at the point of discharge from the plant discharge in Connecticut River) shall be limited to I 0 times the concentrations specified in Appendix B to I OCFR Part 20. I 00 I - 20.2402, Table 2, Column 2.for radionuclides other than noble gases and 2x 10-~ microcurie/milliter total activity concentration for all dissolved or entrained noble gases. APPLICABILITY: At all times. ACTION: With the concentration of radioactive material in liquid effluents released to Unrestricted Areas exceeding the limits of Control 3.2.1, immediately take action to decrease the release rate of radioactive materials and/or increase the dilution flow rate to restore the concentration to within the above limits. SURVEILLANCE REQUIREMENTS 4.2.1.a Radioactive material in liquid waste, intercepted groundwater releases, and subsurface groundwater flows to the Connecticut River shall be sampled and analyzed in accordance with requirements of Table 4.2.1. 4.2.1.b The results of the analyses shall be used in accordance with the methods in the ODCM to assure that the concentrations at the point of release to Unrestricted Areas are limited to the values in Control 3 .2.1. Off-Site Dose Calculation Manual Section 3/4 Rev. 37 Page 11 of 47 Vermont Yankee Nuclear Power Station

TABLE 4.2.1 Radioactive Liquid Sampling and Analysis Program Lower Limit Minimum of Detection Liquid Release Sampling Analysis Type of Activity (LLD) Type Frequency Frequency Analysis ( uCi/ml) (a) Batch Waste Prior to each Prior to each Principal Gamma 5 x 10-7 Release Tanks (h) release Each release Each Emitters (d) Batch Batch I One Batch per Once per Dissolved and 1 x 10-5 month sampled month Entrained Gases prior to a (Gamma release Emitters) Prior to each Once per 1-1-3 1 x 10-5 release Each month Batch Composite (cl Gross Alpha 1 x 10-7 I Prior to each release Each Once per quarter Sr-90 5 x 10-8 Batch Composite (c) Fe-55 1 x 10-6 Gro11ndwatcr Prior to each Prior to each Principal Gamma Activity Interception release Each release Each Emitters (<lJ Analysis LLDs (<!) Release Tanks (b) Batch Batch H-3 Prior to each Once per H-3tc> Activity release Each month Analysis LLDs Composite (cJ Gross Alpha (e) Batch Prior to each Once per H-3(cl Activity release Each quarter Analysis LLDs Composite (c) Gamma (e) Batch Emitters(e) Ni-63(el Fe-55(c) Sr-90(e) Alpha Spec<e> Off-Site Dose Calculation Manual Section 3/4 Rev.37 Page 12 of 47 Vermont Yankee Nuclear Power Station

Lower Limit Minimum of Detection Liquid Release Sampling Analysis Type of Activity (LLD) Type Frequency Frequency Analysis ( uCi/ml) (a) Subsurface Perimeter Perimeter H-3(c) Activity Groundwater Wells - Wells - Analysis LLDs Gamma (c) Flows to the Quarterly (g) Quarterly Emitters(cJ Connecticut River( I) Sentinel Wells Sentinel Wells I - Monthly 01 > - Monthly Ni-63(cJ Fe-55(c) Sr-90(..:) I Alpha Spec(..:) Off-Site Dose Calculation Manual Section 3/4 Rev. 37 Page 13 of 47 Vermont Yankee Nuclear Power Station

TABLE 4.2.1 NOT J\ TION

a. The LLD is the smallest concentration of radioactive material in a sample that will yield a net count, above system background, that will be detected with 95% probability with only 5% probability of falsely concluding that a blank observation represents a "real" signal.

For a pmticular measurenwnt system (which may include radiochemical separation): 4.66*S1i LLD= - - - - - " - - - - -1 E

V
K *Y*e**<

where: LLD= the lower limit of detection as defined above (microcuries or picocuries/unit mass or volume) Sb= the standard deviation of the background counting rate or of the counting rate ofa blank sample as appropriate (counts/minute) E= the counting efficiency (counts/disintegration) V = the sample size (units of mass or volume) 6 K= 2.22 x 10 disintegrations/minute/microcurie or 2.22 disintegration/minute/picocurie as applicable Y= the fractional radiochemical yield (when applicable) le= the r:idio~!ctive dccc:y constant for the particular n~di(muc\ide (/minute)

            ~t =     the elapsed time between sample collection and analysis (minutes)

Typical values of E, V, Y and ~t can be used in the calculation. In calculating the LLD for a radionuclide determined by gamma-ray spectrometry, the background shall include the typical contributions of other radionuclides normally present in the samples. Analysis shall be performed in such a manner that the stated LLDs will be achieved under routine conditions. Occasionally, background fluctuations, unavoidably small sample sizes, the presence of interfering nuclides, or other uncontrollable circumstances may render these LLDs unavailable. It should be recognized that the LLD is defined as a "before the fact" limit representing the capability of a measurement system and not as an "after the fact" limit for a particular measurement. This does not preclude the calculation of an "after the fact" LLD for a particular measurement based upon the actual parameters for the sample in question and appropriate decay correction parameters such as decay while sampling and during analysis. Off-Site Dose Calculation Manual Section 3/4 Rev. 37 Page 14 of 47 Vermont Yankee Nuclear Power Station

TABLE 4.2. I NOT A TI ON (Cont'd)

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

c. A composite sample is one in which the quantity of liquid sampled is proportional to the quantity of liquid waste discharged and in which the method of sampling employed results in a specimen which is representative of the liquids released. Prior to analyses, all samples taken for the composite shall be thoroughly mixed in order for the composite sample to be representative of the effluent release.

d. The principal gamma emitters for which the LLD specification will apply are exclusively the following radionuclides: Mn-54, Co-60, Zn-65, Cs-134, Cs-137 and Ce-144. This list does not mean that only thc~;e nucliclcs arc to be detected and reported. Ot~1l~r pc:iks wh:ch 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, but as "not detected." When unusual circumstances result in LLDs higher than required, the reasons shall be documented in the Radioactive Effluent Release Report.

e. At a minimum, each subsurface groundwater sample shall be analyzed for the following analytes. Lower Limit of Detection for each analyte is as follows:

STANDARD RADIONUCLIDES LLD (PCI/L) 3H 2000 5.iMn 15 6oCo 15 6szn 30 95zr 30 134Cs 15 131Cs 18 Off-Site Dose Calculation Manual Section 3/4 Rev. 37 Page 15 of 47 Vermont Yankee Nuclear Power Station

Ir tritium or plant-generated gamma activity is positively detected, then a sample should be further analyzed for the presence of the following Hard-To-Detect (HTD) rndionuclidcs, as a minimum, using the associated LLD values: HTD RADIONUCLIDES LLD (PCl/L) 110 530 3.5 Alpha Emitters 15

f. ODCM Section 9.3 further defines the location and detennination of flow through the plant site groundwater streamtubes.

g. Perimeter wells used to measure flow and concentration of contaminants in stream tubes:

GZ-ls, GZ-3s, GZ-4s, GZ-5s, GZ-13s, GZ-13d, GZ-14s, GZ-14d, GZ-18s, GZ-18d, GZ-19s and GZ-1 9d.

h. Sentinel used to measure flow and concentration of contaminants in streamtubes: GZ-27s, GZ-26s, GZ-25s, GZ-23s and GZ-22d.

Off-Site Dose Calculation Manual Section 3/4 Rev.37 Page 16 of 47 Vermont Yankee Nuclear Power Station

3/4.2 RADIOACTIVE LIQUID EFFLUENTS 3/4.2.2 Dose - Liquids CONTROLS 3.2.2 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:

a. During any calendar quarter:

less than or equal to 1.5 mrem to the total body, and less than or equal to 5 mrem to any organ, and

b. During any cakm.iar year:

less than or equal to 3 mrem to the total body, and less than or equal to 10 mrem to any organ. APPLICABILITY: At all times. ACTION: With the calculated dose from the release of radioactive materials in liquid effluents exceeding any of the above limits, prepare and submit to the Commission within 30 days, pursuant to ODCM Section 10, a Special Report that identifies the cause(s) for exceeding the limit(s) and defines the corrective actions that have been taken to reduce the releases and the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with the above limits. SURVEILLANCE REQUIREMENTS 4.2.2 Cumulative dose contributions shall be determined in accordance with the methods in the ODCM at least once per month if releases during the period have occurred. Off-Site Dose Calculation Manual Section 3/4 Rev.37 Page 17 of 47 Vermont Yankee Nuclear Power Station

3/4.2 RADIOACTIVE LIQUID EFFLUENTS 3/4.2.3 Liquid Radwaste Treatment CONTROLS 3.2.3 The liquid radwaste treatment system shall be used in its designed modes of operation to reduce the radioactive materials in the liquid waste prior to its discharge when the projected doses clue to the liquid effluents released to Unrestricted Areas, when averaged with all other liquid releases over the last month, would exceed 0.06 mrem to the total body, or 0.2 mrem to any organ. APPLICABILITY: At all times. ACTION: With radioactive liquid waste being discharged without treatment and in excess of the above limits and any portion of the Liquid Radwaste Treatment System not in operation, prepare and submit to the Conunission within 30 days, a Special Report that includes the information detailed in ODCM Section 10.4.1. SURVEILLANCE REQUIREMENTS 4.2.3.a See Control 4.2.2. 4.2.3.b The liquid radwastc treatment system schematic is shown in ODCM Figure 9.1. Off-Site Dose Calculation Manual Section 3/4 Rev.37 Page 18 of 47 Vermont Yankee Nuclear Power Station

314.3 Rt\DIOACTIVE GASEOUS EFFLUENTS 3/4.3.1 Gaseous Eflluents Dose Rate CONTROLS 3.3.1 The dose rate due to radioactive materials released in gaseous effluents from the site to areas at and beyond the site boundary shall be limited to the following:

a. For noble gases; less than or equal to 500 mrem/yr to the total body and less than or equal to 3,000 mrem/yr to the skin, and

b. For tritium and radionuclides in particulate form with half-lives greater than 8 days; less than or equal to 1,500 mrem/yr to any organ.

t\PPLICABIUTY: At all times. ACTION: With the dose rate(s) exceeding the above limits, immediately take action to decrease the release rate to within the limits of Control 3.3.1. SURVEILLANCE REQUIREMENTS 4.3.1.a The dose rate due to noble gases in gaseous effluents shall be determined to be within the limits ofControl 3.3.1 in accordance with the methods in the ODCM. I 4.3.1.b The dose rate due to tritium and radionuclides in particulate form with half-lives greater than 8 days in gaseous effluents shall be determined to be within the limits of Control 3 .3 .1 in accordance with the methods in the ODCM by obtaining representative samples and performing analyses in accordance with the sampling and analysis program specified in Table 4.3 .1. Off-Site Dose Calculation Manual Section 3/4 Rev. 37 Page 19 of 47 Vermont Yankee Nuclear Power Station

TABLE 4.3.1 Radioactive Gaseous Waste Sampling And Analysis Program Lower Limit Minimum Type of of Detection Gaseous Release Sampling Analysis Activity (LLD) Type Frequency Frequency Analysis (uCi/ml) (aJ B. Containment Prior to each Prior to each Principal I x 10-9 (gl Purge release/ release/ Each Gamma Purge Emitters (d.gJ Each Purge Grab Sample for Particulates C. Main Plant Once per month Once per month Principal 1 x 10-.J (e) Stack (e) Grab Sample Gamma Emitters (dl H-3 Ix 10-6 Continuous (e) Once per week Principal 1x10-ll (b) Particulate Gamma Sample Emitters (d,g) Continuous le) Once per month Gross Alpha I x 10- 1( Composite Particulate Sample Continuous(.:) Once per quarter Sr-90 1x10- 11 Con:positc Particulate Sample Continuous Noble Gas Noble Gases 1 x 10-5 Monitor Gross Beta or Gamma Off-Site Dose Calculation Manual Section 3/4 Rev.37 Page 20 of 47 Vermont Yankee Nuclear Power Station

T /\BLE 4.3.1 NOT I\ TION

a. Sec ll.1otnotc a. of Table 4.2.1.

b. Samples shall be changed at least once per 7 days and analyses shall be completed within 48 hours after removal from samplers.

c. Deleted

d. The principal gamma emitters for which the LLD specification will apply arc exclusively the following radionuclidcs: Mn-54, Co-60, Zn-65, Cs-134, Cs-137 and Cc-144 for particulate emissions. This list docs not mean that only these nuclides arc to be detected and reported. Other peaks which arc measurable and identifiable, together with the above nuclidcs, shall also be identified and reported. Nuclides which arc below LLD for the analyses should not be reported as being present at the LLD level for that nudide, hut ;is "not detected." When un~1sual circumstances result in LLDs high~r than required, the reasons shall be documented in the Radioactive Effluent Release Report.

e. The ratio of the sample 11ow rate to the sampled stream flow rate shall be known for the time period covered by each dose or dose rate calculation made in accordance with Controls 3.3.1, 3.3.2, and 3.3.3.

f. Deleted

g. Lower Limit of Detection (LLD) applies only to particulate fonu radionuclides identified in Table Notation d. above.

Off-Site Dose Calculation Manual Section 3/4 Rev. 37 Page21 of47 Vermont Yankee Nuclear Power Station

3/4.3 RADIOACTIVE GASEOUS EFFLUENTS 3/4.3.2 Dose -- Noble Gases CONTROLS 3.3.2 The air <lose due to noble gases released in gaseous eff1uents from the site to areas at and beyond the site boundary shall be limited to the f()llowing:

a. During any calendar quarter:

less than or equal to 5 mrad for gamma radiation, and less than or equal to 10 mrad for beta radiation, and

b. During any calendar year:

less than or equal to I 0 mracl for gamma radiation, and less than or equal to 20 mracl for beta radiation. APPLICABILITY: At all times. ACTION: With the calculated air dose from radioactive noble gases in gaseous effluents exceeding any of the above limits, prepare and submit to the Commission within 30 days, pursuant to ODCM Section 10.4.2, a Special Report that identifies the cause(s) for exceeding the limit(s) and defines the corrective actions that have been taken to reduce the releases and the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with the above limits. SURVEILLANCE REQUIREMENTS 4.3.2 Cumulative dose contributions for the total time period shall be determined in accordance with the methods in the ODCM at least once every month. Off-Site Dose Calculation Manual Section 3/4 Rev. 37 Page 22 of 47 Vermont Yankee Nuclear Power Station

314.3 RADIOACTIVE GASEOUS EFFLUENTS 314.3.3 Dose --Radioactive Material in Particulate Form and Tritium CONTROLS 3 .3 .3 The dose to a member of the public from tritium and radionuclides in particulate fom1 with half-lives greater than 8 days in gaseous effluents released from the site to areas at and beyond the site boundary shall be limited to the following:

a. During any calendar quarter:

less than or equal to 7.5 mrem to any organ, and

b. During any calendar year:

less than or equal to 15 mrem lo any organ. APPLICABILITY: At all times. ACTION: With the calculated dose from the release of tritium and radionuclides in particulate form with half-lives greater than 8 days, in gaseous effluents exceeding any of the above limits, prepare and submit to the Commission within 30 days, pursuant to ODCM Section 10.4.2, a Special Report that identifies the causc(sj for exceeding the limitls) anu defines the corrective actions that have been taken to reduce the releases and the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with the above limits. SURVEILLANCE REQUIREMENTS 4.3.3 Cumulative dose contributions for the total time period shall be determined in accordance with the methods in the ODCM at least once every month. Off-Site Dose Calculation Manual Section 3/4 Rev. 37 Page 23 of 47 Vermont Yankee Nuclear Power Station

314.3 RADIOACTIVE GASEOUS EFFLUENTS 314.3.4 Deleted Off-Site Dose Calculation Manual Section 3/4 Rev. 37 Page 24 of 47 Vermont Yankee Nuclear Power Station

314.3 RADIOACTIVE GASEOUS EFFLUENTS 3/4.3 .5 V cnti lation Exhaust Treatment CONTROLS 3.3.5 The Radwaste Building Ventilation Filter (HEPA) Systems shall be used to reduce particulate materials in gaseous waste prior to their discharge from those buildings when the projected doses due to gaseous effluent releases from the site to areas at and beyond the site boundary would exceed 0.3 mrem to any organ over one month. APPLICABILITY: At all times. ACTION: With gaseous radwaste being discharged without processing through appropriate treatment systems as noted above, and in excess of the limits of Control 3.3.5, prepare and submit to the Commission within 30 days, a Special Report tliat includes the information detailed in ODCM Section I 0.4.2. SURVEILLANCE REQUIREMENTS 4.3.5 See Control 4.3.2 for surveillance related to Radwaste Building ventilation filter system operation. Off-Site Dose Calculation Manual Section 3/4 Rev. 37 Page 25 of 47 Vermont Yankee Nuclear Power Station

314.3 RADIOACTIVE GASEOUS EFFLUENTS 3/4.3.6 Deleted Off-Site Dose Calculation Manual Section 3/4 Rev. 37 Page 26 of 47 Vermont Yankee Nuclear Power Station

3/4.3 RADIOACTIVE GASEOUS EFFLUENTS 3/4.3.7 Deleted Off-Site Dose Calculation Manual Section 3/4 Rev. 37 Page 27 of47 Vermont Yankee Nuclear Power Station

314.4 TOTAL DOSE J/4.4.1 Total Dose ( 40 CFR 190) CONTROLS J.4.1 The Jose or dose commitment to a member of the public** in areas at and beyond the Site Boundary from all station sources is limited to less than or equal to 25 mrem to the total body or any organ over a calendar year. APPLICABILITY: At all times. ACTION: With the calculated dose from the release of radioactive materials in liquid or gaseous effluents exceeding twice the limits of Controls 3.2.2.a, 3.2.2.b, 3.3.2.a, 3.3.2.b, 3.3.3.a, or 3.3.3.b, calculations should be made, including direct radiation contributions from the station to determine whether the above limits of Control 3 .4.1 have been exceeded. If such is the case, prepare and submit to the Commission within 30 days a Special Report that includes the information detailed in ODCM Section 10.4.3. SURVEILLANCE REQUIREMENTS 4.4.1.a Cumulative dose contributions from liquid and gaseous effluents shall be determined in accordance with Controls 4.2.2, 4.3.2, and 4.3.3. 4.4.1.b Cumulative dose contributions from direct radiation from plant sources shall be determined in accordance with the methods in the ODCM. This requirement is applicable only under conditions set forth in Control 3.4.1 Action Statement.

Note: For this Control, a member of the public may be taken as a real individual accounting for his actual activities.

Off-Site Dose Calculation Manual Section 3/4 Rev. 37 Page 28 of47 Vermont Yankee Nuclear Power Station

314.5 RADIOLOGICAL ENVIRONMENT AL MONITORING 3/4.5.1 Environmental Monitoring Program CONTROLS 3.5.1 The radiological environmental monitoring program shall be conducted as specified in Table 3.5.1. APPLICABILITY: At all times. ACTION:

a. With the radiological environmental lfo.mitoring program llllt being conducted as specified in Tables 3.5.1 or 4.5.1, prepare and submit to the Commission, in the Annual Radiological Environmental Monitoring Report (per ODCM Section 10.2), a description of the reasons for not conducting the program as required and the plans for preventing a recurrence.

b. With the level of radioactivity as the result of plant effluents in an environmental sampling media at one or more locations specified in Control Table 3.5.1 exceeding the reporting levels of Control Table 3.5.2, prepare and submit to the Commission a Special Report within 30 days from receipt of the laboratory analysis (per ODCM Section 10.4.4).

SURVEILLANCE REQUIREMENTS 4.5. l The radiological environmental monitoring samples shall be collected pursuant to Table 3.5.1 from the locations given in the ODCM and shall be analyzed pursuant to the requirements of Table 3.5.1 and the detection capabilities required by Table 4.5.1. Off-Site Dose Calculation Manual Section 3/4 Rev. 37 Page 29 of 47 Vermont Yankee Nuclear Power Station

TABLE 3.5.1 Radiological Environmental Monitoring Program Exposure Pathway Number of Sample Sampling and Type and Frequency and/or Sample Locations (a) Collection of Analysis Frequency

l. AIRBORNE

a. Particulates Samples from Continuous Particulate sampler:

3 locations: operation of Gross beta radioactivity sampler with analysis on each sample 1 sample from up sample collection following lilter change. (~) valley, within 4 miles weekly or more Composite (by (-6.4 km) of Site frequently as location) for gamma (d) Boundary. (major required by dust isotopic at least once wind direction) loading. per quarter. l sample from clown valley, within 4 miles(-6.4 km) of Site Boundary. (major wind direction) I l sample from a control location. Off-Site Dose Calculation Manual Section 3/4 Rev.37 Page 30 of 47 Vermont Yankee Nuclear Power Station

TABLE 3.5.1 (Cont'd) Radiological Environmental Monitoring Program Exposure Pathway Number of Sample Sampling and Type and Frequency (a) and/or Sample Locations Collection Frequency of Analysis

..,   DIRECT          Routine monitoring         Quarterly.                Gamma dose, at RA DIATION 1b)  stations as follows:                                 least once per quarter.

10 incident response stations (one in each Incident response meteorological sector TLDs de-dose only on land) located along quarterly unless the site boumiary; gaseous release Controls were Additional stations to exceeded in period. be placed in special interest areas and control station areas. Off-Site Dose Calculation Manual Section 3/4 Rev.37 Page 31 of47 Vermont Yankee Nuclear Power Station

TABLE 3.5.1 (Cont'd) Radiological Environmental Monitoring Program Exposure Pathway Number of Sample Sampling and Type and Frequency (U) and/or Sample Locations Collection Frequency of Analysis

3. WATERBORNE

(~)

a. Surface I sample upstream. Monthly grab sample. Gamma isotopic (d) analysis of each sample. Tritium analysis of composite sample at least once per quarter.

I sample Composite sample Gamma isotopic (d) downstream. collected over a analysis of each period of one month sample. Tritium (I) analysis of composite sample at least once per quarter.

b. Ground 1 sample from within Quarterly. G . . (<ll amma 1sotop1c (potable - 8 km (5 miles) and tritium analyses drinking water distance. of each sample.

from bedrock wciis) 1 sample from a Quarterly. G . . (<lJ amma 1sotop1c control location. and tritium analyses of each sample.

c. Sediment from 1 sample from Semiannually. Gamma isotopic (d)

Shoreline downstream area with analysis of each existing or potential sample. recreational value. 1 sample from north Semiannually. Gamma isotopic (d) storm drain outfall. analysis of each sample. Off-Site Dose Calculation Manual Section 3/4 Rev. 37 Page 32 of47 Vermont Yankee Nuclear Power Station

TABLE 3.5.1 (Cont'd) Radiological Environmental Monitoring Program Exposure Pathway Number of Sample Sampling and Type and Frequency (a) and/or Sample Locations Collection Frequency of Analysis

4. INGESTION

a. Deleted

b. Fish I sample of two Semiannually. Gamma isotopic

                                                                                  . (d)        .

recreationally important analysis on echble species in vicinity of portions. plant discharge area. I sample (jH\.:(l!rably of Semiannually. Ga1nma isulopie (d) same species) in areas analysis on edible not influenced by plant portions. discharge. I c. Vegetation l grass sample at each Quarterly when Gamma isotopic (d) air sampling station. available. analysis of each sample. l silage sample at each Quarterly when Gamma isotopic

                                                                                  . (d)

I former milk sampling available. analysts of each station (as available). sample. Off-Site Dose Calculation Manual Section 3/4 Rev.37 Page 33 of 47 Vermont Yankee Nuclear Power Station

TABLE 3.5.1 NOTATION a Specific parameters of distance and direction sector from the centerline of the reactor and additional descriptions where pertinent, shall be provided for each and every sample location in Table 3.5.1 in a table and figure(s) in the ODCM (Section 7). Deviations arc permitted from ~l'.e required sampling _sch~~lule if speci1'.1ens <~re unobt'.1inable c~uc to hazardous cond1t1ons, seasonal unavailab1hty, malluncllon ot automatic samplmg equipment and other legitimate reasons. If specimens are unobtainable due to sampling equipment malfunction, every reasonable effort shall be made to complete corrective action prior to the end of the next sampling period. All deviations from the sampling schedule shall be documented in the Annual Radiological Environmental Operating Report pursuant to ODCM Section l 0.2. It is recognized that, at times, it may not be possible or practicable to continue to obtain samples of the media of choice at the most desired location or time. In these instances, suitable alternative media and locations may be chosen for the particular pathway in question and appropriate substitutions made within 30 days in the radiological environmental monitoring program. In lieu of a Licensee Event Report and pursuant to ODCM Section l 0.1, identify the cause of the unavailability of samples for that pathway and identify the new location(s) for obtaining replacement samples in the next Radioactive Effluent Release Repo11 and also include in the report a revised figure(s) and table for the ODCM reflecting tht.: new location(s). b One or more instmments, such as a pressurized ion chamber, for measuring and recording dose rate continuously may be used in place of, or in addition to, integrating dosimeters. For the purposes of this table, a Thermo luminescent Dosimeter (TLD) is considered to be one phosphor; two or more phosphors in a packet are considered as two or more dosimeters. Film badges shall not be used as dosimeters for measuring direct radiation. The 40 stations is not an absolute number. The frequency of analysis or readout for TLD systems will depend upon the characteristics of the specific system used ~nd should be selected to obtain optimum dose infomrntion with minimal fading. c Airborne particulate sample filters shall be analyzed for gross beta radioactivity 24 hours or more after sampling to allow for radon and thorium daughter decay. If gross beta activity in air pa~iculate samples is greater than ten times the yearly mean of control samples, gamma isotopic analysis shall be performed on the individual samples. d Gamm[! is0t6pic analysis means the identification aricl qu:int!ficr:ti.on of gamma-emittin::; radionuclides that may be attributable to the effluents from the facility. e The "upstream sample" shall be taken at a distance beyond significant influence of the di~c~arge. The "downstream" sample shall be taken in an area beyond but near the m1xmg zone. f Composite sample aliquots shall be collected at time intervals that are very short relative to the compositing period in order to assure obtaining a representative sample. g Deleted h Deleted. Off-Site Dose Calculation Manual Section 3/4 Rev.37 Page 34 of 47 Vermont Yankee Nuclear Power Station

TABLE 3.5.2 Reporting Levels For Radioactivity Concentrations In Environmental Samplcs(al Reporting Levels I Airborne Particulate Water or Gases Fish Vegetation Sediment I Analysis (pCi/l) (pCi/m3) (pCi/Kg, wet) (pCi/Kg, wet) (pCi/Kg, dry) H-3 2 x 10-l(h) 3 Mn-54 Ix 10 3 x 10-1 Co-60 3 x 10- I x 10-1 3 x 1031 cl 2 Zn-65 3 x 10 2 x 10-1 2 Zr-95 4 x 10 Cs-134 30 10 1 x 10 3 I x 10 3 Cs-137 50 20 2 x 10 3 2 x 10 3 (a) Reporting levels may be averaged over a calendar quarter. When more than one of the radionuclides in Table 3.5.2 are detected in the sampling medium, the unique reporting requirements are not exercised if the following condition holds: concentration ( 1) concentration (2)

                                 ------- +                                 + ...    < 1.0 reporting level ( l) reporting level (2)

When radionuclides other than those in Table 3.5.2 are detected and are the result of plant effluents, the potential annual dose to a member of the public must be less than or equal to the calendar year limits of Controls 3.2.2, 3.3.1, and 3.3.2. (b) Reporting level for drinking water pathways. For nondrinking water pathways, a value of 3 x 10-1 pCi/l may be used. (c) Reporting level for individual grab samples taken at North Storm Drain Outfall only. (d) Deleted Off-Site Dose Calculation Manual Section 3/4 Rev. 37 Page 35 of 47 Vermont Yankee Nuclear Power Station

T J\BLE 4.5.1 Detection Capabilities For Environmental Sample J\nalysis(a)(c) J\irbomc Water Particulate or Fish Vegetation Sediment Analysis(dl (pCi/I) Gas (pCi/m3) (pCi/Kg, wet) (pCi/Kg, wet) (pCi/Kg, dry) Gross beta 4 0.01 1111 H-3 2000 Mn-54 15 130 Co-60 15 130 Zn-65 30 260 Zr-95 15 Cs-134 15 0.05 130 60 150 Cs-13 7 18 0.06 150 80 180 Off-Site Dose Calculation Manual Section 3/4 Rev.37 Page 36 of47 Vermont Yankee Nuclear Power Station

TABLE 4.5.1 NOTATION (a) See Footnote (a) of Table 4.2.1. (b) Parent only. ( c) If the measured concentration minus the 5 sigma counting statistics is found to exceed the specified LLD, the sample docs not have to be analyzed to meet the specified LLD. (d) This list does not mean that only these nuclides are to be considered. Other peaks that arc identifiable, together with those of the listed nuclides, shall also be analyzed and reported in the Annual Radiological Environmental Operating Report pursuant to Technical Specification 6.6.E and ODCM Section 10.2. (e) Deleted ( t) Deleted. (g) Deleted (h) If no drinking water pathway exists, then a value of 3000 picocuries per liter may be used. Off-Site Dose Calculation Manual Section 3/4 Rev.37 Page 37 of47 Vermont Yankee Nuclear Power Station

314.5 RADIOLOGICAL ENVIRONMENTAL MONITORING 314.5.2 Land Use Census CONTROLS 3.5.2 A land use census shall be conducted to identify the location of the nearest residence in each or the 16 meteorological sectors within a distance of live miles. APPLICABILITY: At all times. ACTION:

a. With a land use census identifying one or more locations which yield a calculated <lose or dose commitment (via the same exposure pathway) at least 20 percent greater than at a location from which samples are currently being obtained in accordance with Control 3.5.1, add the new location(s) to the radiological environmental monitoring program within 30 days if pem1ission from the owner to collect samples can be obtained, and sufficient sample volume is available. The sampling location(s),

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

b. With the land census not being conducted as required above, prepare and submit to the Commission within 30 days a Special Report that includes information detailed in ODCM Section 10.4.5.

SURVEILLANCE REQUIREMENTS 4.5.2 The land use census shall be conducted at least once per year between the dates of June 1 and October 1 by either a door-to-door survey, aerial survey, or by consulting local agricultural authorities. The results of the land use census shall be included in the Annual Radiological Environmental Operating Report pursuant to Technical Specification 6.6.E and ODCM Section 10.2. Off-Site Dose Calculation Manual Section 3/4 Rev. 37 Page 38 of 47 Vermont Yankee Nuclear Power Station

3/4.5 RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.5.3 Intcrlaborato1y Comparison Program CONTROLS 3.5.3 Analyses shall be performed on referenced radioactive materials supplied as part of an lnterlaboratory Program which has been approved by NRC. APPLICABILITY: At all times. ACTION:

             'Ni~h analysis nut being p~rfonned as required abl>Ve, report lhe co1Tective actions taken to prevent a recu1Tence to the Commission in the Annual Radiological Environmental Operating Report pursuant to ODCM Section 10.2 SURVEILLANCE REQUIREMENTS 4.5.3         A summary of the results of analyses perfonned as part of the above required lnterlaboratory Program shall be included in the Annual Radiological Environmental Operating Report. NRC-approved interlaboratory programs utilized by environmental laboratories in processing Vermont Yankee samples shall be identified in the ODCM.

Off-Site Dose Calculation Manual Section 3/4 Rev. 37 Page 39 of 47 Vermont Yankee Nuclear Power Station

3/4.6 EFFLUENT AND ENVIRONMENTAL CONTROL BASES INSTRUMENTATION Liquid Effluent Instmmentation (3.1.1) The radioactive liquid effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in liquid effluents during actual or potential releases of liquid effluents. The alann setpoints for these instruments arc to ensure that the alam1 will occur prior to exceeding I 0 times the concentration limits of Appendix B to I OCFR20. I 001-20.2402, Table 2, Column 2, values. Automatic isolation function is not provided on the liquid radwaste discharge line due to the infrequent nature of batch, discrete volume, liquid discharges (on the order of once per year or less), and the administrative controls provided to ensure that conservative discharge How rates/dilution Hows are set such that the probability of exceeding the above concentration limits are low, and the potential off-site dose consequences are also low. Gaseous Effluent Instrumentation (3. I .2) The radioactive gaseous effluent instmmentation is provided to monitor and control, as applicable, the releases of radioactive materials in gaseous effluents during actual or potential releases of gaseous effluents. The alann/trip setpoints for these instruments are provided to ensure that the alann/trip will occur prior to exceeding design bases close rates identified in Control 3.3.1. RADIOACTIVE EFFLUENTS Liquid Effluents: Concentration (3.2.1) This Control is provided to ensure that at any time the concentration of radioactive materials released in liquid waste effluents or groundwater flowing to the Connecticut River from the site in radioactive contamination concentrations above background (Unrestricted Area for liquids is at the point of discharge from the plant discharge into Connecticut River) will not exceed 10 times the concentration levels specified in 10CFR Part 20.1001-20.2402, Appendix B, Table 2, Column 2. These requirements provide operational flexibility, compatible with considerations of health and safety, which may temporarily result in releases higher than the absolute value of the concentration numbers in Appendix B, but still within the annual average limitation of the Regulation. Compliance with the design objective doses of Section II.A of Appendix I to 10CFR Part 50 assure that doses are maintained ALARA, and that annual concentration limits of Appendix B to 10CFR20.l001-20.2402 will not be exceeded. Off-Site Dose Calculation Manual Section 3/4 Rev. 37 Page 40 of 47 Vermont Yankee Nuclear Power Station

3/4.6 EFFLUENT AND ENVIRONMENTAL CONTROL BASES (cont.) The concentration limit for noble gases is based upon the assumption that Xe-135 is the controlling radionuclide and that an effluent concentration in air (submersion close equal to 500 mrem/yr) was converted to an equivalent concentration in water. Liquid Effluents: Dose (3 .2.2) This Control is provided to implement the requirements of Sections II.A, III.A, and IV.A of Appendix I, IOCFR Part 50. The Limiting Condition for Operation implements the guides set forth in Section II.A of Appendix I. The requirements provide operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix l to assure that the releases of radioactive material in liquid effluents will be kept "as low as is reasonably achievable." The Surveillance Requirements imp!cmcnt the requirc:ncn~s in Sec~ion Ill.A of' Appendix I, i.e., tl:at conformance with the guides of Appendix I be shown by calculational procedures based on models and data such that the actual exposure of an individual through appropriate pathways is unlikely to be substantially underestimated. In addition, there is reasonable assurance that the operation of the facility will not result in radionuclide concentrations in potable drinking water that are in excess of the requirements of 40CFR14l. No drinking water supplies drawn from the Connecticut River below the plant have been identified. The appropriate dose equations for implementation through requirements of the Specification are described in the Vermont Yankee Off-Site Dose Calculation Manual. The equations specified in the ODCM for calculating the doses due to the actual release rates of radioactive materials in liquid effluents were developed from the methodology provided in Regulatory Guide 1.109, "Calculation of Annual Doses to Man from Routine Releases of Reactor Efflue11l:s for the Purpose of Evaluating Complian~e with 10CFR Part 50, Appendix I", Revision 1, October 1977 and Regulatory Guide 1.113, "Estimating Aquatic Dispersion of Effluents from Accidental and Routine Reactor Releases for the Purpose of Implementing Appendix I", Revision 1, April 1977. Off-Site Dose Calculation Manual Section 3/4 Rev. 37 Page 41of47 Vermont Yankee Nuclear Power Station

3/4.6 EFFLUENT AND ENVIRONMENTAL CONTROL BASES (cont.) Liquid Radwaste Treatment (3.2.3) The requirement that the appropriate portions of this system as indicated in the ODCM be used, when specified, provides assurance that the releases of radioactive materials in liquid effluents will be kept "as low as is reasonably achievable." This specification implements the requirements of 10CFR Part 50.36a and the design objective given in Section Il.D of Appendix I to 10CFR Part 50. The specified limits governing the use of appropriate portions of the liquid rad waste treatment system were specified as a suitable fraction of the dose design objectives set forth in Section II.A of Appendix I, 10CFR Part 50, for liquid effluents. Gaseous Effluents: Dose Rate (3.3.1) The specified limits as detem1ined by the methodology in the ODCM, restrict, at all times, the con-esponding gamma and beta dose rates above background to a member of the public at or beyond the site boundary to (500) mrem/year to the total body or to (3,000) mrem/year to the skin. This instantaneous dose rate limit allows for operational flexibility when off nonnal occun-ences may temporarily increase gaseous effluent release rates from the plant, while still providing controls to ensure that licensee meets the dose objectives of Appendix I to 10CFR50. Gaseous Effluents: Dose from Noble Gases (3.3.2) This Control is provided to implement the requirements of Sections 11.B, III.A, and IV.A of Appendix I, 10CFR Part 50. The Limiting Condition for Operation implements t:1e guidt::s si.:;L forth in Sec Lion il.B of Appendix I. The requi;*ements provide operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive material in gaseous effluents will be kept "as low as is reasonably achievable." The Surveillance Requirements implement the requirements in Section III.A of Appendix I, i.e., that conformance with the guides of Appendix I be shown by calculational procedures based on models and data such that the actual exposure of any member of the public through appropriate pathways is unlikely to be substantially underestimated. The appropriate dose equations are specified in the ODCM for calculating the doses due to the actual releases of radioactive noble gases in gaseous effluents. The ODCM also provides for determining the air doses at the site boundary based upon the historical average atmospheric conditions. Off-Site Dose Calculation Manual Section 3/4 Rev.37 Page 42 of47 Vermont Yankee Nuclear Power Station

3/4.6 EFFLUENT AND ENVIRONMENTAL CONTROL BASES (cont.) The equations specified in the ODCM for calculating the doses due to the actual release rates of radioactive noble gases in gaseous effluents were developed from the methodology provided in Regulatory Guide I. I 09, "Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with I OCFR Part 50, Appendix I", Revision I, October 1977 and Regulatory Guide 1.111, "Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water Cooled Reactors," Revision I, July 1977. Gaseous Effluents: Dose from Tritium and Radionuclides in Particulate Form

         !.1lll This Control is !1rovidecl to imrlcment the n.:quin~m.:nts of ~...:ction l!.C, lll.J\, and IV.A of Appendix I, I OCFR Part 50. The Limiting Condition for Operation are the guides set forth in Section 11.C of Appendix I. The requirements provide operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive materials in gaseous effluents will be kept "as low as is reasonably achievable." The Surveillance Requirements implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data such that the actual exposure of a member of the public through appropriate pathways is unlikely to be substantially underestimated. The equations specified in the ODCM for calculating the doses due to the actual release rates of the subject materials were also developed using the methodology provided in Regulatory Guide 1.109, "Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents frlr t!~c Purpos...: of Evaluating Complianc.:~ with 1OCf P, Pait 50, Appendix I," Revision 1, October 1977 and Regulatory Guide 1.111, "Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water Cooled Reactors," Revision 1, July 1977. These equations also provide for determining the actual doses based upon the historical average atmospheric conditions. The release rate specifications for tritium and radionuclides in particulate form with half-lives greater than 8 days are dependent on the existing radionuclide pathways to man, in areas at and beyond its site boundary. The pathways which were examined in the development of these specifications were: I) 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.

Off-Site Dose Calculation Manual Section 3/4 Rev. 37 Page 43 of 47 Vermont Yankee Nuclear Power Station

3/4.6 EFFLUENT AND ENVIRONMENTAL CONTROL BASES (cont.) Gaseous Radwaste Treatment (3.3.4) Deleted Ventilation Exhaust Treatment (3.3.5) The requirement that the Radwaste Building HEPA filters be used when specified provides reasonable assurance that the release of radioactive materials in gaseous en1uents will be kept "as low as is reasonably achievable." This specification implements the requirements of I OCFR Part 50.36a and the design objective of Appendix I to I OCFR Part 50. The requirements governing the use of the appropriate portions of the gaseous rad waste filter systems were specified by the NRC in NUREG-0473, Revision 2 (July 1979) as a suitable fraction of the guide set forth in Sections 11.B and II.C of Appendix I, IOCFR Part 50, for gaseous effluents. Primary Containment (MARK I) (3.3.6) Deleted Steam Jet Air Ejector (SJAE) (3.3.7) Deleted Total Dose (40CFR 190) (3.4.1) This Control is provided to meet the dose limitations of 40CFR Part 190 to Members of the Public in areas at and beyond the Site Boundary. The specification requires the preparation and submittal of a Specific Report whenever the calculated doses from plant radioactive effluents exceed twice the design objective doses of Appendix I. The Special Report will describe a course of action that should result in the limitation of the annual dose to a Member of the Public to within the 40CFR Part 190 limits. For th(: purposes of the Special Repon, it may be assumed thal the dose commitment to the Member of the Public is estimated to exceed the requirements of 40CFR Part 190, the Special Report with a request for a variance (provided the release conditions resulting in violation of 40CFR Part 190 have not already been corrected), in accordance with the provisions of 40CFR Part 190.11 and 10CFR Part 20.2203(a)(4), is considered to be a timely request and fulfills the requirements of 40CFR Part 190 until NRC staff action is completed. The variance only relates to the limits of 40CFR Part 190, and does not apply in any way to the other requirements for dose limitation of 10CFR Part 20. An individual is not considered a Member of the Public during any period in which he/she is engaged in carrying out any operation that subjects them to occupational exposures. For individuals in controlled areas who are considered Members of the Public per 10CFR20, the dose limits of 10CFR20.1301 apply since the licensee has the authority to control and limit access to these areas. Off-Site Dose Calculation Manual Section 3/4 Rev. 37 Page 44 of 47 Vermont Yankee Nuclear Power Station

3/4.6 EFFLUENT AND ENVIRONMENT AL CONTROL BASES (cont.) Radiological Environmental Monitoring Program (3.5. l) The radiological monitoring program required by this Control provides measurements of radiation and of radioactive materials in those exposure pathways and for those radionuclides which lead to the highest potential radiation exposures of mcmber(s) of the public resulting from the station operation. This monitoring program implements Section IV.B.2 of Appendix I to I 0 CFR Part 50 and thereby supplements the radiological effluent.monitoring program by verifying that the measurable concentrations of radioactive materials and levels of radiation are not higher than expected on the basis of the effluent measurements and modeling of the environmental exposure pathways. Ten years of plant ope:-ntion, including the years pr:nr Ll the implementation of the Augmented Off-Gas System, have amply demonstrated via routine effluent and environmental reports that plant effluent measurements and modeling of environmental pathways are adequately conservative. In all cases, environmental sample results have been two to three orders of magnitude less than expected by the model employed, thereby representing small percentages of the ALA RA and environmental reporting levels. This radiological environmental monitoring program has therefore been modified as provided for by Regulatory Guide 4.1 (C.2.b), Revision 1, April 1975. Evaluation of plant gaseous effluents, meteorological conditions and potential accident scenaria have concluded that milk sampling is no longer required and silage and grass sampling have been instituted as an indicator of radionuclide deposition. Because of this change, the frequency of silage collection has been increased from annual to quarterly. The detection capabilities required by Table 4.5.1 are considered optimum for routine environmental measurements in industrial laboratories. It should be recognized that the LLD is defined as a before-the-fact limit representing the capability of a measurement system and not as an after-the-fact limit for a particular measurement. This does not preclude the calculation of an after-the-fact LLD for a particular measurement based upon the actual parameters for the sample in question. Off-Site Dose Calculation Manual Section 3/4 Rev. 37 Page 45 of 47 Vermont Yankee Nuclear Power Station

3/4.6 EFFLUENT AND ENVIRONMENT AL CONTROL BASES (cont.) Land Use Census (3.5.2) This Control is provided lo ensure that changes in the use of areas al and beyond the site boundaries are identified and that modifications to the monitoring program arc made if required by the results of this census. This census satisfies the requirements of Section IV.B.3 of Appendix I to 10 CFR Part 50. The requirement of a garden census has been eliminated along with the food product monitoring requirement due lo the substantial and widespread occurrence of dairy farming in the surrounding area which dominates the food uptake pathway. The addition of new sampling locations to Control 3.5. l, based on the land use census, is limited to those locations which yield a calculated dose or dose commitment greater than 20 percent of the calculated dose or dose commitment al any location currently being sampled. This eliminates the unnecessary changing of the environmental radiation monitoring program for new locations which, within the accuracy of the calculation, contributes essentially the same to the dose or dose commitment as the location already sampled. The substitution of a new sampling point for one already sampled when the calculated difference in dose is less than 20 percent, would not be expected to result in a significant increase in the ability to detect plant effluent related nuclides. Due to the decay of I-131 and limited source terms available alter pennanent shutdown, milk sampling was discontinued, but sampling of vegetation and air sampling have remained in place. Interlaboratory Comparison Program (3.5.3) The requirement for participation in an intercomparison 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 for the purposes of Section IV.B.2 of Appendix I to 10 CFR Part 50. Off-Site Dose Calculation Manual Section 3/4 Rev.37 Page 46 of 47 Vermont Yankee Nuclear Power Station

The addition of new sampling locations to Control 3.5.1, based on the land use census, is limited to those locations which yield a calculated dose or dose commitment greater than 20 percent or the calculated dose or dose commitment al any location currently being sampled. This eliminates the unnecessary changing of the environmental radiation monitoring program f<lr new locations which, within the accuracy of the calculation, contributes essentially the same to the dose or dose commitment as the location already sampled. The substitution of a new sampling point for one already sampled when the calculated difference in dose is less than 20 percent, would not be expected to result in a significant increase in the ability to detect plant effluent related nuclides. Due to the decay of 1-131 after permanent shutdown, analysis for 1-131 in milk samples has been discontinued, but gamma isotopic analysis and sampling of vegetation, silage and air sampling have remained in place. Off-Site Dose Calculation Manual Section 3/4 Rev. 37 Page 47 of 47 Vermont Yankee Nuclear Power Station

5.0 METHOD TO CALCULATE OFF-SITE LIQUID CONCENTRATIONS Chapter 5 contains the basis for plant procedures that the plant operator requires to meet ODCM Control 3.2.1 which limits the total fraction of combined cft1uent concentration in liquid pathways, excluding noble gases, denoted here as, F'/Nc; at the point of discharge at any time (see Figure 9-1 ). F'/Nn is limited to less than or equal to ten, i.e., FJ-:N() < 10 I - The total concentration of all dissolved and entrained noble gases at the point of discharge from all station sources, denoted C 1NG, is limited to 2E-04 ~LCi/ml, i.e., C ~c; ~ 2 E - 04 ~LCi I ml. Evaluation of F\:Nn and C~c; is required concurrent with the sampling and analysis program in Control Table 4.2.1. 5.1 Method to Determine pENCi and cNG Determine the total fraction of combined effluent concentrations at the point of disch~rge in liquid pathways (excluding noble gases), denoted F~:No, and determine the total concentration at the point of discharge of all dissolved and entrained noble gases in liquid pathways from all station sources, denoted C 1NG' as follows: FENG I

                   = L. Cpi
                             ~  10 I ECL (5-1)
           µCi/ml )

( ~LCi/ml and: Off-Site Dose Calculation Manual Section 5.0 Rev.36 Page 1of4 Vermont Yankee Nuclear Power Station

(5-2) (~t Ci/ml) (~L Ci/ml) (~t Ci/ml) where: F\*Nli Total sum of the fractions of each radionuclide concentration in liquid effluents (excluding noble gases) at the point of discharge to an unrestricted area, divided by each radionuclide's ECL value. C. Concentration at point of discharge to an unrestricted area of radionuclide "i", except for dissolved and entrained noble gases, from any tank or other significant source, p, from which a discharge may be made (including the floor drain sample tank, the waste sample tanks, the detergent waste tank, groundwater interception release tanks and any other significant source from which a discharge can be made) (µCi/ml). This concentration can be calculated from: C11 i = CrK1 x FrK/(Fon. +Fix] where: CrK1 equals the concentration of radionuclide i in the tank to be discharged (µCi/ml); Fo11. is equal to the dilution flow provided by the liquid radioactive waste dilution pumps (20,000 gpm); FrK equals the liquid waste discharge pump flow rate which regt!lntes the rate at which liquid from ::i waste collection tank is discharged (gpm). ECLi Annual average effluent concentration limits of radionuclide "i", except for dissolved and entrained noble gases, from 10CFR20.1001-20.2402, Appendix B, Table 2, Column 2 (µCi/ml). C~G Total concentration at point of discharge to an unrestricted area of all dissolved and entrained noble gases in liquid pathways from all station sources (µCi/ml). NG C!i Concentration at point of discharge to an unrestricted area of dissolved and entrained noble gas "i" in liquid pathways from all station sources (µCi/ml). Off-Site Dose Calculation Manual Section 5.0 Rev.36 Page 2of4 Vermont Yankee Nuclear Power Station

5.2 Method to Determine Radionuclide Concentration for Each Liquid El'f1uenl Pathway 5.2.1 Sample Tanks Pathways C 11 ; is determined for each radionuclide above LLD from the activity in a representative grab sample of any of the sample tanks and the predicted 11ow at the point of discharge to an unrestricted area. Most periodic batch releases are made from the two I 0,000-gallon capacity waste sample tanks. These tanks serve lo hold all the high purity liquid wastes after they have been processed by ion exchange. Other periodic batch releases may also come from the detergent waste tank or the floor drain sample tank. A batch release tank can be any collection device (e.g. bladder, pillow tank, etc.) that meets the discharge requirements of Table 4.2.1 Notation b. The liquid waste tanks are sampled and the contents analyzed for water quality and radioactivity. If the sample meets all the high purity requirements, the contents of the tank may be re-used in the spent fuel pool or toms. If the sample does not meet all the high purity requirements, the contents arc recycled through the radwaste system or discharged. The groundwater intercept tank is sampled from the tank recirculation system, and the contents analyzed for vv~&ter qua!ity a11J radioactivity. If the sample meets th~ rcquircmer.ts for discharge, it may be discharged to the stom1 drain system. Prior to discharge each sample tank is analyzed for tritium, dissolved noble gases and dissolved and suspended gamma emitters. I 5.2.2 Service Water Pathway The service water pathway shown on Figure 9-1, flows from the intake structure through the heat exchangers and the discharge structure. Under normal operating conditions, the water in this line is not radioactive. For this reason, the service water line is not sampled routinely but it is continuously monitored with the service water discharge monitor (No. 17/351 ). Off-Site Dose Calculation Manual Section 5.0 Rev. 36 Page 3 of 4 Vermont Yankee Nuclear Power Station

The alarm setpoint on the service water discharge monitor is set at a level which is three times the background of the instmment. The service water is sampled if the monitor is out of service or if the alarm sounds. Under expected or anticipated operating conditions, the concentration at any time of radionuclides at the point of discharge from the service water effluent pathway to an unrestricted area will not exceed ten times the effluent concentration values of 10CFR20.l001-20.2402, Appendix B, Table 2, Column 2. 5.2.3 Deleted 5.2.4 Intercepted Groundwater Concentrations in Flowpaths to the Connecticut River In order to minimize radioactive liquid waste, groundwater intmding into basement stmctures is intercepted from sub-slab groundwater wells. This water is environmentally derived and captured under basements to prevent additional cross contamination from plant systems and stmctures. The intercepted ground water is tanked, sampled and released to the storm system with outfalls to the Connecticut River. The Intercepted Groundwater Collection and Release System is isolated from all other plant systems containing liquids in order to prevent cross contamination and contains water collected from the plant environs, therefore continuous release monitoring is not rel~uired . 1!1tercepted groundwater tanks :i.re sampled :md ;elcascd as described* in 5.2.1 Sample Tank Pathway. 5.2.5 Subsurface Contaminated Groundwater Concentrations in Flowpaths to the Connecticut River The overall direction of groundwater flow at Vermont Yankee (VY) is towards the Connecticut River (west to east). Based on this understanding of site hydrogeologic conditions, the groundwater discharge rates from the developed portion of the site to the river are estimated using a streamtube approach based on Darcy's Law (see Section 9). To estimate the groundwater concentration in each of the designated streamtubes, samples will be collected and analyzed according to requirements specified in Section 3 14, Table 4.2.1. The concentrations shall then be determined using methods provided in Section 5 .1. Off-Site Dose Calculation Manual Section 5.0 Rev. 36 Page 4of4 Vermont Yankee Nuclear Power Station

6.0 OFF-SITE DOSE C J\LCULJ\TION METI !ODS Chapter 6 provides the basis for plant procedures required to meet the I OCFR50, Appendix I. J\LJ\RJ\ dose objectives, and the 40CFR 190 total dose limits to members of the public in unrestricted areas, as stated in the Radiological Effluent Controls (implementing the requirements of Technical Specification 6.7.D). A simple, conservative method (called Method I) is listed in Tables 1.1.2 to I. I. 7 for each of the Control requirements. Each of the Method I equations is presented, along with their bases in Sections 6.2 through 6.9 and Section 6.11. In addition, reference is provided to more sophisticated hut still conservative methods (called Method 11) for use when more accurate results are needed. This chapter provides the methods, data, and reference material with which the operator can calculate the needed doses and dose rates. Setpoint methods for effluent monitor alarms are described in Chapter 8. Demonstration of compliance with the dose limits of 40CFR 190 is considered to he a demonstration of compliance with the 0.1 rem limit of JOCFR20.130l(a)(I) for members of the public in unrestricted areas (Reference 56 FR23374, 3rd column). 6.1 Introductory Concepts The Radiological Effluent Controls Program (Technical Specifications 6.7.D) either limit dose or dose rate. The term "Dose" for ingested or inhaled radioactivity means the dose commitment, measured in mrem, which results from the exposure to radioactive materials that, because of uptake and deposition in the body, will continue to expose the body to radiation for some period of time after the source of radioactivity is stopped. The time frame over which the dose commitment is evaluated is 50 years. The phrases "annual Dose" or "Dose in one year" then refers to the fifty-year dose commitment from one year's worth of releases. "Dose in a quarter" similarly means a fifty-year dose commitment from one quarter's rclea~ 1:s. The term "Dose," with respect to external exposures, such as to noble gas clouds, refer only to the doses received during the actual time period of exposure to the radioactivity released from the plant. Once the source of the radioactivity is removed, there is no longer any additional accumulation to the dose commitment. Gaseous effluents from the plant are also controlled such that the maximum "dose rates" at the site boundary at any time are limited to 500 mrem/year. This instantaneous dose rate limit allows for operational flexibility when off normal occurrences may temporarily increase gaseous effluent release rates from the plant, while still providing controls to ensure that licensees meet the dose objectives of Appendix I to I OCFRSO. Off-Site Dose Calculation Manual Section 6 Rev. 37 Page I of 43 Vermont Yankee Nuclear Power Station

It should also be noted that a dose rate due to noble gases that exceeds for a short time period (less than one hour in duration) the equivalent 500 mrem/year dose rate limit stated in Control 3.3.1.a, docs not necessarily, by itsclt: constitute a Licensee Evt:nt Report (LER) under I OCFR Part 50.73 unless it is determined that the air concentration of radioactive efl1uents in unrestricted areas has also exceeded 20 times applicable concentration limits specified in Appendix B to 20. I 00 I - 20.2402. Table 2. Column I (four-hour notification per I OCFR50.72, and 30-day LER per I OCFR50.73). The quantities D and Rare introduced to provide calculable quantities, related to off-site dose, or dose rate which demonstrates compliance with the effluent controls. The dose D is the quantity calculated by the Chapter 6 dose equations. The D calculated by "Method I" equations is not necessarily the actual dose received by a real individual but usually provides an upper bound for a given release because of the conservative margin built into the dose factors and the selection and definition of critical receptors. The radioisotope specific dose factors in each "Method I" dose equation represent the greatest dose to any organ of any age group accounting for existing or potential pathways of exposure. The critical receptor assumed by "Method I" equations is typically a hypothetical individual whose behavior - in terms of location and intake - results in a dose which is expected to be higher than any real individual. The Method I equations employ five-year historical average atmospheric dispersion factors to define receptors of maximum impact. Method II allows for a more exact dose calculation for real individuals, if necessary, by considering only existing pathways of exposure, or actual concurrent meteorology with the recorded release. Maximum receptor doses determined using quarterly meteorology may be greater than doses calculated with Method I due to short time period variability of meteorological conditions from the long-term average. Quarterly average dispersion values for maximum receptors have been observed to differ from five-year average values by as much as 54%. Ris the quantity calculated in the Chapter 6 dose rate equations. It is calculated using the plant's effluent monitoring system reading and an annual average or long-term atmospheric dispersion factor. Dispersion factors based on actual concurrent meteorology during effluent releases can also be used via Method II, if necessary, to demonstrate compliance with off-site dose rate limits. Each of the methods to calculate dose or dose rate are presented in separate sections of Chapter 6, and are summarized in Tables 1.1.1 to 1.1. 7. Each method has two levels of complexity and are called Method I and Method II. Method I is the simplest; generally a linear equation. Method II is a more detailed analysis which allows for use of site-specific factors and variable parameters to be selected to best fit the actual release conditions, within the bounds of the guidance provided. Off-Site Dose Calculation Manual Section 6 Rev. 37 Page 2 of 43 Vermont Yankee Nuclear Power Station

The plant has both elevated and ground level gaseous release points: the main vent stack {elevated release) and the Reactor Building (ground level release). Therd'ore, total dose calculations for skin, whole body, and the critical organ from gaseous releases will be the sum of the elevated and ground level doses. 6.2 Method to Calculate the Total Body Dose from Liquid Releases Effluent Control 3.2.2 limits the total body dose commitment to a Member of the Public from radioactive material in liquid effluents to 1.5 mrem per quarter and 3 mrem per year. Control 3.2.3 requires liquid rad\vaste treatment when the total body dose estimate exceeds 0.06 mrem in any month. Control 3.4.1 limits the total body dose commitment to any real member of the public from all station sources (including liquids) to 25 mrem in a year. Dose evaluation is required at least once per month. If the liquid radwaste treatment system is not being used, dose evaluation is required before each release. Use Method I first to calculate the maximum total body dose from a liquid release to the Connecticut River as it is simpler to execute and more conservative than Method II. Use Method II if a more accurate calculation of total body dose is needed (i.e., Method I indicates the dose is greater than the limit), or if Method I cannot be applied. If the radwaste system is not operating, the total body dose must be estimated prior to a release (Control 3.2.3). To evaluate the total body dose, use Equation 6.1 to estimate the dose from the planned release and add this to the total body dose accumulated from prior releases during the month. To asses<> the dose contrih11tion from suhsurfacc groundwnter contaminated v;ith plant-generated radionuclides, a dose evaluation shall be performed using Method I on a monthly basis. Radionuclide concentration averages and groundwater streamtube average flow rates shall be utilized to estimate the total plant-generated radioactive contaminants released for the previous monthly period. Off-Site Dose Calculation Manual Section 6 Rev. 37 Page 3 of 43 Vermont Yankee Nuclear Power Station

6.2.1 Mdhod I The increment in total body dose from a liquid release is: (6-1) D11i = FC IQ, DFL i1h (mrem) (C1) . ( mrem. ) Ct where: FC Flow Correction calculated by dividing the flow at the unrestricted area release point in gpm divided 20,000 gpm or release flow in tl 3/sec divided by 44.6 11 3/sec. Site-specific total body dose factor (mrem/Ci) for a liquid release. Sec Table I. I. I I. Total activity (Ci) released for radionuclide "i." (For strontiums and Fe 55, use the most recent measurement available.) Equation 6-1 can be applied under the following conditions (otherwise, justify Method I or consider Method II):

1. Normal operations (not emergency event),

2. Liquid releases were to the Connecticut River. and

3. Any continuous or batch release over any time period.

6.2.2 Basis for Method I This section serves three purposes: ( l) to document that Method l complies with appropriate NRC regulations, (2) to provide background and training information to Method I users, and (3) to provide an introductory user's guide to Method II. Method I may be used to show that the effluent Controls which limit off-site total body dose from liquids (3.2.2 and 3.2.3) have been met for releases over the appropriate periods. Control 3.2.2 is based on the ALARA design objectives in I OCFR50, Appendix I Subsection II A. Control 3.2.3 is an "appropriate fraction," determined by the NRC, of that design objective (hereafter called the Objective). Control 3.4. l is based on Environmental Standards for Uranium Fuel Cycle in 40CFRI 90 (hereafter called the Standard) which applies to direct radiation as well as liquid and gaseous effluents. Off-Site Dose Calculation Manual Section 6 Rev. 37 Page 4 of 43 Vermont Yankee Nuclear Power Station

Exceeding the Objective or the Standard docs not immediately limit plant operation but requires a report to the NRC within 30 days. In addition. a 'Naiver may be required. Method I was developed such that "the actual exposure of an individual ... is unlikely lo be substantially underestimated" (I OCFR50, Appendix I). The definition, below, of a single "critical receptor" (a hypothetical individual whose behavior results in an unrealistically high dose) provides part of the conservative margin to the calculation of total body dose in Method I. Method II allows that actual individuals, with real behaviors. be taken into account for any given release. In fact, Method I was based on a Method II analysis for the critical receptor with maximum exposure conditions instead of any real individual. That analysis was called the "base case:" it was then reduced to form Method I. The steps performed in the Method I derivation follm\*. First, in the base case. the dose impact to the critical receptor (in the form of dose factors DFLith* mrem/Ci) for a I curie release of each radioisotope in liquid effluents was derived. The base case analysis uses the methods, data and assumptions in Regulatory Guide I. I 09 (Equations

      !\-2, A-3, A-7, A-13 and A-16, Reference A). The liquid pathways identified as contributing lo an individual's dose are the consumption of fish from the Connecticut River. the ingestion of vegetables and leafy vegetation which were irrigated by river water, the consumption of milk and meat from cows and beef cattle who had river water available for drinking as well as having feed grown on irrigated land, and the direct exposure from the ground plane associated with activity deposited by the water pathway. A plant discharge flow rate of 44.6 ft3/sec was used with a mixing ratio of 0.0356 which corresponds to a minimum regulated river flow of 1250 cfs at the Vernon Dam just below the plant discharge outfall.* Tables 6.2. l and 6.2.2 outline human consumption and environmctital parameters used in the analysis. The resulting, site-specific, total body dose factors appear in Table l .1.11.

For any liquid release, during any period, the increment in annual average total body dose from radionuclide "i" is: (6-2)

        ~ D1b = FC 0;   DFL1b (mrem) (Ci) (  m~~m) where:

FC Flow Correction calculated by dividing the flow at the unrestricted area release point in gpm divided 20,000 gpm or release flow 3 in ft /sec divided by 44.6 ft 3/sec Off-Site Dose Calculation Manual Section 6 Rev.37 Page 5 of 43 Vermont Yankee Nuclear Power Station

Site-specific total body dose factor (mrem/Ci) for a liquid release. See Table 1.1.11. Total activity (Ci) released from radionuclide "i." An Mp equal to 1.0 for the fish pathway is assumed between the discharge structure and the dam. Method I is conservative because it is based on dose factors DFLitb which were chosen from the base case to be the highest of the four age groups for each radionuclide, as well as assuming minimum river dilution flow. 6.2.3 Method II If Method I cannot be app!icd, C)f" if the Method I dose exceeds the limit or if a more exact calculation is required. then Method II should be applied. Method II consists of the models. input data and assumptions in Regulatory Guide I. I 09, Rev. I (Reference A), except where site-specific models, data or assumptions are more applicable, such as the use of actual river flow at the time of actual discharge as opposed to the minimum river flow of 1,260 cfs that is assumed in the Method I dose factors (except for the fish pathway). The base case analysis, documented above, is a good example of the use of Method I I. It is an acceptable starting point for a Method II analysis. Off-Site Dose Calculation Manual Section 6 Rev.37 Page 6 of 43 Vermont Yankee Nuclear Power Station

TABLE 6.2.1 Environmental Parameters for Liquid Effluents at Vermont Yankee (Derived from Reference A) FOOD GROWN WITH CONT AMINA TED WATER POTABLE AQUATIC SHORELINE LEAFY cow VARIABLE WATER FOOD ACTIVITY VEGETABLES VEG. MEAT MILK MP Mixing Ratio 1.0 0.0356 0.0356 0.0356 0.0356 0.0356 TP Transit Time (HRS) 24.0 0.000 0.0000 0.0000 480.0 48.0 2 YV Agricultural (KG/M ) 2.0 2.0 2.0 2.0 Productivity p Soil Surface (KG/M 2 ) 240.0 240.0 240.0 240.0 Density IRR Irrigation Rate (L/M 2/HR) 0.152 0.152 0.152 0.152 TE Crop Exposure (HRS) 1440.0 1440.0 1440.0 1440.0 Time TH Holdup Time (HRS) 1440.0 24.0 2160.0 2160.0 QAW Water Uptake (LID) 50.0 60.0 Rate for Animal QF Feed Uptake Rate (KG/D) 50.0 50.0 for Animal FI Fraction of Year Crops Irrigated 0.5 0.5 0.5 0.5 Location of Critical Connecticut River Below Vernon Dam Receptor Off-Site Dose Calculation Manual Section 6 Rev.37 Page 7 of -B Vermont Yankee Nuclear Power Station

TABLE 6.2.2 Usage Factors for Various Liguid Pathwavs at Vermont Yankee (From Reference A, Table E-5. Zero Where No Pathway Exists) LEAFY POTABLE AGE VEG. VEG. MILK MEAT FISH INVERT. WATER SHORELINE (KG/YR) (KG/YR) (LITER/YR) (KG/YR) (KG/YR) (KG/YR) (LITER/YR) (HR/YR) Adult 520.00 64.00 310.00 110.00 21.00 0.00 0.00 12.00 Teen 630.00 42.00 400.00 65.00 16.00 0.00 0.00 67.00 Child 520.00 26.00 330.00 41.00 6.90 0.00 0.00 14.00 Infant 0.00 0.00 330.00 0.00 0.00 0.00 0.00 0.00 Off-Site Dose Calculation Manual Section 6 Rev.37 Page 8 of 43 Vermont Yankee Nuclear Power Station

6.3 Method to Calculate Maximum Organ Dose from Liquid Releases Eflluent Control 3.2.2 limits the maximum organ dose commitment to a Member of the Public from radioactive material in liquid cnluents to 5 mrem per quarter and 10 mrem per year. Control 3.2.3 requires liquid radwaste treatment when the maximum organ dose estimate exceeds 0.2 mrcm in any month. Control 3.4. I limits the maximum organ dose commitment to any real member of the public from all station sources (including liquids) to 25 mrem in a year. Dose evaluation is required at least once per month if releases have occurred. If the liquid rad waste treatment system is not being used, dose evaluation is required before each release. Use Method I first to calculate the maximum organ dose from a liquid release to the Connecticut River as it is simpler to execute and more conservative than Method II. Use Method II if a more accurate calculation of organ dose is needed (i.e., Method I indicates the dose is greater than the limit), or if Method I cannot be applied. If the radwaste system is not operating, the maximum organ dose must be estimated prior to a release (Control 3.2.3). To evaluate the maximum organ dose, use Equation 6-3 to estimate the dose from the planned release and add this to the maximum organ dose accumulated from prior releases during the month. To assess the dose contribution from subsurface groundwater contaminated with plant-generated radionuclides, a dose evaluation shall be performed using Method I on a monthly basis. Radionuclide concentration averages and groundwater streamtube average flow rates shall be utilized to estimate the total plant-generated radioactive contaminants released for the monthly period. 6.3. I Method I The increment in maximum organ dose from a liquid release is: (6-3) Dmo = L i FC Qi DFL imo (mrem) = (C1). ( mrem Ci ) (FC) where: FC Flow Correction calculated by dividing the flow at the unrestricted area release point in gpm divided 20,000 gpm or release flow in ft3 /sec divided by 44.6 ft 3/sec. Site-specific maximum organ dose factor (mrem/Ci) for a liquid release. See Table 1.1.11. Off-Site Dose Calculation Manual Section 6 Rev. 37 Page 9 of 43 Vermont Yankee Nuclear Power Station

Total activity (Ci) released for radionuclide "i." (For strontiums and Fc-55. use the most recent measurement available.) Equation 6-3 can be applied under the following conditions (otherwise.justify Method I or consider Method II): I. Normal operations (not emergency event),

2. Liquid releases were to the Connecticut River. and

3. Any continuous or batch release over any tin1e period.

6.3.2 Basis for Method I This section serves three purposes: (I) to document that Method I complies with appropriate NRC regulations. (2) to provide background and training information to Method I users. and (3) to provide an introductory user's guide to Method I I. The methods to calculate maximum organ dose parallel the total body dose methods (see Section 6.2.2). Only the differences are presented here. For each radionuclide, a dose factor (mrem/Ci) was determined for each of seven organs and four age groups. The largest of these was chosen to be the maximum organ dose factor (DFLimo) for that radionuclide. For any liquid release. during any period, the increment in annual average dose from radionuclide "i" to the maximum organ is: (6-4) L\ Ur- mo -- F' L* 0 Q i D"L 1~ 11110 (mrem) (Ci) ( 111

                                                ~~    ) (  FC )

where: FC Flow Correction calculated by dividing the flow at the unrestricted area release point in gpm divided 20,000 gpm or release flow in ft3/sec divided by 44.6 ft 3/sec Site-specific maximum organ dose factor (mrem/Ci) for a liquid release. See Table 1.1.11. Total activity (Ci) released for radionuclide "i". Off-Site Dose Calculation Manual Section 6 Rev. 37 Page 10 of 43 Vermont Yankee Nuclear Power Station

Because of the assumptions about receptors, environment, and radionuclides; and because of the low Objective and Standard. the lack of immediate restriction on plant operation. and the adherence lo I OCFR20 concentrations (which limit public health consequences) a failure of Method I (i.e., the exposure of a real individual being underestimated) is improbable and the consequences of a failure arc minimal. 6.3 .3 Method 11 If Method I cannot be applied, or if the Method I dose exceeds the limit or ifa more exact calculation is required, then Method II should be applied. Method II consists of the models, input data and assumptions in Regulatory Guide 1.109, Rev. I (Reference A), except where site-specific models. data or assumptions are more applicable. The base case analysis, documented above, is a good example of the use of Method II. It is an acceptable starting point for a Method II analysis. 6.4 Method to Calculate the Total Body Dose Rate From Noble Gases Enlucnt Control 3.3.1 limits the instantaneous dose rate at any time to the total body from all release sources of noble gases at any location at or beyond the site boundary equal to or less than 500 mrem/year. Use Method I first to calculate the Total Body Dose Rate from the peak release rate via both elevated and ground level release points. The dose rate limit of Control 3.3.1.a is the total contribution from both ground and elevated releases occurring during the period of interest. Use Method II if Method I predicts a dose rate greater than the Control limit (i.e., use of actual meteorology over the period of interest) to determine if, in fact, Control 3.3. l lrnd ~:~tualiy b,.:c;1 e;~cecd~d during a short time imerV<il. Compliance with the dose rate limits for noble gases are continuously demonstrated when effluent release rates are below the plant stack noble gas activity monitor alarm setpoint by virtue of the fact that the alarm setpoint is based on a value which corresponds to the off-site dose rate limit of Control 3.3.1, or a value below it, taking into account the potential contribution of releases from all ground level sources. Determinations of dose rates for compliance with Control (3.3.1) are performed when the effluent monitor alarm setpoint is exceeded and the corrective action required by Control 3 .3. I is unsuccessful, or as required by the notations to Control Table 3.1.2 when the stack noble gas monitor is non-functional. Off-Site Dose Calculation Manual Section 6 Rev. 37 Page 11 of 43 Vermont Yankee Nuclear Power Station

6.4.1 Method I The Total Body Dose Rate due to noble gases can be determined by multiplying the individual radionuclide release rates by their respective dose factors, summing all the products together, and then multiplying this total by a conversion constant (0.75), as seen in the following Equation 6-5: Ribs 0.75 (6-5) mrem) (pC'.-se~) (*~1Ci) (111re1~1-111J) ( yr ~1C1- m sec pC1-yr where: Q ~T In the ease of noble gases. the release rate from the plant stack (µCi/sec) is specifically for Kr-85 due to radioactive decay of other noble gasses. The release rate at the plant stack is based on measured radionuclide concentrations and distributions in periodic grab samples taken at the stack. l. Q~T = M F Sg (6-28) (cpm )( ~tCi/ccJ(cc) uCi sec cpm sec M Plant Stack Gas Monitor I or II count rate (cpm). Sg Appropriate or conservative plant stack monitor detector counting efficiency for the given nuclide mix (cpm/(µCi/cc)). F Stack flow rate (cc/sec). DFB Total body gamma dose factor for Kr-85, 1.61 E-05 (see Table I.I.IO). Off-Site Dose Calculation Manual Section 6 Rev. 37 Page 12 of 43 Vermont Yankee Nuclear Power Station

For ground level nohle gas releases, the lolal hody dose rate is calculated as t<Jllows: R tbg 7.4 I <i ~ii. DFB.I (6-39) ( pCi. - sc~ J c1Ci J ( mrc_'.'.1 - m J 1 pC1 -m sec pC1 - yr where: 0:;1. =Ground kvel release rate (pCi/sec) of nohle gas Kr-85. The total body dose rate for the site is equal to l~bs + l~hg* Kr-85 shall be used to determine off-site dose rate and monitor setpoints. Equations 6-5 and 6-39 can be applied under the following conditions (otherwise. justify Method I or consider Method II): I. Normal operations (not emergency event), and

2. Noble gas releases via either elevated or ground level vents to the atmosphere.

Off-Site Dose Calculation Manual Section 6 Rev. 37 Page 13 of 43 Vermont Yankee Nuclear Power Station

6.4.2 Basis for Method I Method l may be used to show that the Control limit for total body dose rate from noble gases released to the atmosphere has been met for the peak noble gas release rate. Method I for stack releases was derived from Regulatory Guide 1.109 as follows: R11is = IE+ 06 Sr [X/Q]~ L Q;

ST DFB;

( yr mrem ) = ( pC'.) (#) (

                          ~tC1 se~)

m.)

                                                       ~1Ci ) ( mre1~1 - 111

( sec pCt - yr

3) (6-6) where:

Shielding factor= 1.0 for dose rate determination. [X!Ql Maximum annual average gamma atmospheric dispersion factor for stack (elevated) releases;= 7.51 E-07 (sec!m\ Release rate from the plant stack of Kr 85 (µCi/sec). 3 Kr-85 Gamma total body dose factor, 1.61 E-05 mrem-m . ) ( pC1-yr See Table I. I. I 0. Equation 6-6 reduces to:

                                              *ST Ribs             0.75       Q;               DFB; m~:m)             ( pC'.-se~J (~tCi) (mre~-m J 3                        (6-5)

(

                                 µC1- m*           sec         pC1 -yr For ground level releases, the ground level maximum long-term average gamma atmospheric dispersion factor= 7.37E-06 sec/m3 , thus leading to:

Rtbg = 1E+06

7.37 E- 06
GL Q, . DFB; or (6-39)

R1bg

7.4 0;

GL DFB; Off-Site Dose Calculation Manual Section 6 Rev. 37 Page 14 of 43 Vermont Yankee Nuclear Power Station

The selection of critical receptor, outlined in Section 6.10, is inherent in Method I. as are the maximum expected off-site annual or long-term average atmospheric dispersion factors. It is then conservatively assumed that most of the noble gas activity at the plant stack is the result of in-plant leaks from fuel and evaporated out of the spent fuel pool which arc removed to the plant stack by building ventilation air flow, and that this air flow has an isotopic distribution consistent with that routinely measured in stack gas grab samples. The calculation of ground level release dispersion parameters arc based on the location of the Reactor Building with respect to the site boundary that would experience the highest exposure. In the case of noble gas dose rates, Method II cannot provide much extra realism because Ribs and r~bg are already based on several factors which make use of current plant parameters. However, should it be needed, the dose rate analysis for critical receptor can be performed making use of current meteorology during the time interval of recorded peak release rate in place of the default atmospheric dispersion factor used in Method I. 6.4.3 Method II If Method I cannot be applied. or if the Method I dose exceeds the limit, then Method II may be applied. Method II consists of the models. input data and assumptions in Regulatory Guide I. I 09, Rev. I (Reference A), except where site-specific models, data or assumptions are more applicable. The base case analysis. documented above, is a good example of the use of Method ll. It is an acceptable starting point for a Method fl analysis. Off-Site Dose Calculation Manual Section 6 Rev. 37 Page 15 of43 Vermont Yankee Nuclear Power Station

6.5 Method to Calculate the Skin Dose Rate from Noble Gases Effluent Control 3.3.1 limits the instantaneous dose rate at any time to the skin from all release sources of noble gases al any location at or beyond the site boundary to 3,000 mrem/year. Use Method I first to calculate the Skin Dose Rate from both elevated and ground level release points to the atmosphere. The dose rate limit of Control 3.3.1.a is the total contribution from both ground and elevated releases occurring during the period of interest. Method I applies at all release rates. Use Method II if Mdhod I predicts a dose rate greater than the Control limits (i.e., use of actual meteorology over the period of interest) to determine it: in fact, Control 3.3.1 had actually been exceeded during a short time interval. Co:npliam:c with the dose ratt: limits fo:* nobk gases arc continuously demonstrated when effluent release rates arc below the plant stack noble gas activity monitor alarm setpoint by virtue of the fact that the alarm setpoint is based on a value which corresponds to the off-site Control dose rate limit, or a value below it, taking into account the potential contribution releases from all ground level sources. Determinations of dose rate for compliance with Control (3.3.1) are performed when the effluent monitor alarm setpoint is exceeded and the corrective action required by Control 3.3.1 is unsuccessful, or as required by the notations to Control Table 3.1.2 when the stack noble gas monitor is non-functional. Off-Site Dose Calculation Manual Section 6 Rev. 37 Page 16 of 43 Vermont Yankee Nuclear Power Station

6.5. I Method I The skin dose rate dui.: to noble gasi.:s is determined by multiplying the rdease rates by dosi.: foctors seen in the following Equation 6-7:

ST R sk111s Q, DF'"

(6-7) (~rem J ( ~~c~i ) ( mrem.- sec J yr /I C1- yr whcri.::

      <Yr             In the case of noble gases, the noble gas release rate from the plant stack

(~tCi/sec) or Kr-85 is used. The release rate at the plant stack is based on measured radionuclide concentrations and distributions in periodic grab samples taken at the slack .

                       <),                       M                           F (6-28)
                     ,uCi                        (cpm)     (,uCi/cc)       (cc) sec                                    cpm           sec M                  Plant stack gas monitor I or II count rate (cpm).

Sg Appropriate or conservative plant stack monitor detector counting efficiency for the given nuclide mix (cpm/(µCi/cc)). F Stack flow rate (cc/sec). DFi.~ combined Kr-85 skin dose factor 2. l 5E-03 (see Table 1.1.10) for stack release. Off-Site Dose Calculation Manual Section 6 Rev.37 Page 17 of43 Vermont Yankee Nuclear Power Station

For ground level releases, the skin dose rate from nohle gases is ealculated hy Equation 6-38:

       .             . (ii.

1 R sking = Q i DF ig (6-38) where: Q \ii. The noble gas release rate from ground level (~tCi/sec) for Kr-85. DF[~ Combined Kr-85 skin dose factor for a ground level release 4.91 E-02 [see Table I. I. I OA J. The monitor alarm setpoints should be based on Kr 85 as representing the most prevalent high dose factor noble gas expected to be present. Monitor alarm setpoints which J1ave be*~n determined to be con<.:crvati\*e unde!* any plant conditions m::y be utilized at any time in lieu of the above assumptions. Equations 6-7 and 6-38 can be applied under the following conditions (otherwise, justify Method I or consider Method II): I. Normal operations (not emergency event), and

2. Noble gas releases via both elevated and ground level vents to the atmosphere.

Off-Site Dose Calculation Manual Section 6 Rev. 37 Page 18 of43 Vermont Yankee Nuclear Power Station

6.5.2 Basis For Method I The methods lo calculate skin dose rate parallel the total body dose rate methods in Section 6.4.3. Only the differences arc presented here. Method I may he used to show that the Control limit for skin dose rate from noble gases released lo the atmosphere (Control 3.3.1) has been met for the peak noble gas release rate. Method I was derived from Regulatory Guide I. I 09 as follows: I. I I o;;,, + 3. I 7E+04 DFS, (6-8) (mrcm) (mrem) mrad) se~) 111rc1~1-111 1 (II) ( pCi-yrJ Ci) ( ( ) yr mrad yr ( Ci-st:c ( yr 111 pC1-yr \vhere: 1.11 Average ratio of tissue to air absorption coefficients will convert mrad in air to mrem in tissue. o;.:, = 3 .11E + 04 Qi rx 1Q L o F r 1 (6-9) (m;~d) p~i - yr ) ( C1 - sec ( Ci yr l( se~ ) ( mra~ m

                                                                       - 111 pC1 - yr 3
                                                                               )

now D hnite

                               -- DYair [X /Q]X              I [XI Q ls                                    (6-10)

(mrad) (sec) (mrad) yr yr m3 (::J and Qi 31.54 *ST Qi (6-11) (~;) (Ci~ sec)

                             µC1 - yr

(µCi) sec Off-Site Dose Calculation Manual Section 6 Rev. 37

                                                                      . Page 19 of43 Vermont Yankee Nuclear Power Station

so ll skins = I. 11 SF I E + 06 [ X/Q t L ({r OF.rt I (6-12)

                                            ) (# ) ( /.p CC~i. J ( se~ ) ( ~1Ci ) ( mrn~ - m J J( mrem 11 3

mrcm ( yr mrad m* sec pCt - yr

        +IE+ 06            X/Q." S~

QSTI DFS I

                                      ~tCi        mrct~ -            J

( p~i) 3 sec ( 111 Ct m3 sec pCt -yr substituting 3 [X/QJX 7.51 E-07 sec/m , 3 XI Os I .59E-06 sec/m SF Shielding factor= 1.0 for dose rate determinations gives Rskins = 0.83 DP I + (6-13) mrem)(pCi.- sec- mrem) ( yr ~LC1 - m 3 - mrad (~tCi)(mra~ - sec m pCt - yr 3 J( ~LCt pC'.- sec)(~tCi)(mre~1-

                                                                            - 111 3

sec pCt - yr 111 3 J

        = L Q~T [0.83 OF i + 1.59 DFS i]                                                                      (6-14) i define 0 F'is   =                i 0.83 OF + 1.59 OFS i                                                                    (6-15) then
                                      *ST Rskins Qi                                                                       (6-7)

( 1LCi) ( mre~1 - sec J sec 1£1-yr Off-Site Dose Calculation Manual Section 6 Rev.37 Page 20 of 43 Vermont Yankee Nuclear Power Station

For determining combined skin doses frlr ground level releases. 3 a [X/QIJ = 7.37E-06 sec/m and an undepicted X/Qg = 3.65E-05 see/m:i have been substituted into Equation 6-12 to give: I'\.> sk111g = ') ci1. (8.18 DF 11 + 36.5 DF 1)

                          "...'. 1 then           DF'lg = 8 ' 18 DFYI + 36.5 DFSI.                                            (6-37) and           .

Rsking = Q<;i i . OF'1g (6-38) \Vhere: 6~;1. The noble gas release rate from ground level release points (~LCi/sec) for Kr-85. DF;g Combined Kr-85 skin dose factor for a ground level release 4.91 E-02 [see Table 1.1.IOA]. The selection of critical receptor. outlined in Section 6.10 is inherent in Method I, as it determined the maximum expected off-site atmospheric dispersion factors based on past long-term site-specific meteorology. The calculation of ground level release dispersion parameters are based on the location of the Reactor Building with respect to the site boundary that would experience the highest exposure. 6.5.3 Method II If Method I cannot be applied, or ifthe Method I dose exceeds the limit, then Method II may be applied. Method II consists of the models, input data and assumptions in Regulatory Guide I. I 09, Rev. 1 (Reference A), except where site-specific models, data or assumptions are more applicable. The base case analysis, documented above, is a good example of the use of Method II. It is an acceptable starting point for a Method II analysis. Off-Site Dose Calculation Manual Section 6 Rev. 37 Page 21 of43 Vermont Yankee Nuclear Power Station

I 6.6 Method to Calculate the Critical Organ Dose Rate from Tritium and Particulates with T112 Greater Than 8 Days Effluent Control 3.3.1.b limits the dose rate to any organ, denoted Rc 0 , from all release sources of H-3 and radionuclidcs in particulate form with half lives greater than 8 days to 1500 mrem/year to any organ. The peak release rate averaging time in the case of particulates is commensurate with the time the particulate samplers are in service between changeouts (typically a week). Use Method I first to calculate the critical organ dose rate from both elevated and ground level release points to the atmosphere. The dose rate limit of Control 3.3.1.b is the total contribution from both ground and elevated releases occurring during the period of interest. Method I applies at all release rates. Use Method II if Method I predicts a dose rate greater than the Control limits (i.e .. use of actual meteorology over the period of interest) to determine iC in fact, Control 3.3.1.b had actually been exceeded during the sampling period. Off-Site Dose Calculation Manual Section 6 Rev. 37 Page 22 of43 Vermont Yankee Nuclear Power Station

6.6.1 Method I The critical organ dose rate from stack releases can be determined by multiplying the individual radionuclide release rates by their respective dose factors and summing all their products together, as seen in the following Equation 6-16: Reos DFG'sico (6-16) mrem) ~LCi) ( llll'CI~ - sec) ( ( sec ~tC1 - yr yr where: Stack activity release rate determination of radionuclide "i" (particulates with half-lives greater than 8 days, and tritium) in ~tCi/sec. For i = Sr-90 or tritium, use the best estimates (such as most recent measurements).

                         . spec1'fi1c cnt1ca S 1te               .. 1organ d ose rate f'actor (mrem-sec).

DFG's1co

                                                                              ~LC1 - yr for a ground level gaseous release. See Table 1.1.12.

For ground releases the critical organ dose rate from Tritium and Particulates with T 1/2 greater than 8 days is calculated as follows:

GLP Rcog= ~ Qi DFG'gico (6-40) where:

Ground activity release rate determination of radionuclide "i" (particulates with half-lives greater than 8 days, and tritium) in µCi/sec. For i = Sr-90, Fe-55 or tritium, use the best estimates (such as most recent measurements).

                         . spec1'fi1c cnt1ca S 1te               . . 1organ d ose rate 1actor

c. (mrem- . sec)

DFG'gico

                                                                              µC1-yr for a ground level gaseous release. See Table 1.1.12.

The critical organ dose rate for the site is equal to Reos + Rcog . Off-Site Dose Calculation Manual Section 6 Rev.37 Page 23 of 43 Vermont Yankee Nuclear Power Station

Equations 6-16 and 6-40 can be applied under the following conditions (otherwise, justify Method I or consider Method II): I. Normal operations (not emergency event), and

2. Tritium and particulate releases via either elevated or ground level vents to the atmosphere.

6.6.2 Basis fr)r Method I The methods to calculate critical organ dose rate parallel the total body dose rate methods in Section 6.4.3. Onlv the differences are presented here. Method I may be used to show that the Control limit for organ dose rate from tritium and radionuclides in particulate form with half lives greater than 8 days (hereafter called Particulates or P) released to the atmosphere (Control 3.3.1.b) has 0..:..:il met for the peak P release rate. The equation for Reos and Rcog is derived by modifying Equation 6-25 from Section 6.9 as follows: (mrem) (Ci) ( m~~m) (6-17) applying the conversion factor, 31.54 (Ci-sec/~tCi-yr) and converting Q to Q in ~tCi/sec as it applies to the plant stack yields: Reos 31.54

STP DFGsico Qi (6-18)

(m~:m) Ci~sec) ( µC1-yr µCi) ( sec (mr~m) C1 Off-Site Dose Calculation Manual Section 6 Rev.37 Page 24 of 43 Vermont Yankee Nuclear Power Station L__ _.. _

Equation 6.8 is written in the fl)l"m: R 3 l.54 2:

STI' DFGsico Qi (6-19)

( mrem yr J (Ci-sec

                         ~tCi-yr J        ( ~tCi) sec (m~~m)

DFG~irn and DFG~ico ground releases incorporates the conversion constant of 31.54 and has assumed that the shielding factor (SF) applied to the direct exposure pathway from radionuclides deposited on the ground plane is equal to 1.0 in place of the Si: value of 0.7 assumed in the determination of DFG sicn and DFG gico for the integrated doses over time. The selection of critical receptor (based on the combination of exposure pathways which include direct dose from the ground plane, inhalation and ingestion of vegetables, meat, and milk) which is outlined in Section 6.10 is inherent in Method I, as are the maximum expected off-site atmospheric dispersion factors based on past long-term site-specific meteorology. The calculation of ground level release dispersion parameters are based on the location of the Reactor Building with respect to the site boundary that would experience the highest exposure. Should Method II be needed, the analysis for critical receptor critical pathway(s) and atmospheric dispersion factors may be performed with actual meteorologic and latest land use census data to identify the location of those pathways which are most impacted by these type of releases. Off-Site Dose Calculation Manual Section 6 Rev. 37 Page 25 of 43 Vermont Yankee Nuclear Power Station

6.6.3 Method 11 If Method I cannot be applied. or if the Method I dose exceeds the limit, then Method II may be applied. Method II consists of the models, input data and assumptions in Regulatory Guide I. I 09, Rev. I (Reference A). except where site-specific models. data or assumptions arc more applicable. The base case analysis. documented above, is a good example of the use of Method II. It is an acceptable starting point for a Method II analysis. 6.7 Deleted 6.8 Method to Calculate the Beta Air Dose from Noble Gases Effluent Control 3.3.2 limits the beta dose to air from all release sources of noble gases at any location at or beyond the site boundary to I 0 mrad in at1y quarter and 20 mrad in a:1y year. Dose cval~1ation :s r~quircd at li:ast on*;e per mcnth. Use Method I first to calculate the beta air dose for elevated and ground level vent releases during the period. The total beta air dose limit of Control 3.3.2 is the total contribution from both ground and elevated releases occurring during the period of interest. Use Method II if a more accurate calculation is needed or if Method I cannot be applied. Off-Site Dose Calculation Manual Section 6 Rev. 37 Page 26 of 43 Vermont Yankee Nuclear Power Station

6.8.1 Method I The beta air dose from plant vent stack releases is: fl Dair~ = 0.050 (6-23) (mrad) ( pCi- ,. yr) (Ci) ( mra~- m3 J 3 C.1-111 pC1- yr where: beta dose factor to air for Kr 85 (I .95E-03). See Table 1.1.10. Q~T total gas activity (curies) released to the atmosphere via the plant stack of Kr 85 during the period of interest. For ground level noble gas releases. the beta air dose is calculated as follows: Q~iL DFf (6-43) where: Q yt. = Total activity (curies) released to the atmosphere via the ground level vents of Kr 85 during the period of interest. The beta air dose for the site is equal to D~irs + D~irg . Equations 6-23 and 6-43 can be applied under the following conditions (otherwise justify Method I or consider Method II): I. Normal operations (not emergency event), and

2. Noble gas releases via either elevated or ground level vents to the atmosphere.

Off-Site Dose Calculation Manual Section 6 Rev.37 Page 27 of 43 Vermont Yankee Nuclear Power Station

6.8.2 Basis for Method I This section serves three purposes: (I) to document that Method I complies with appropriate NRC regulations, (2) to provide background an<l training information to Method I users, an<l (3) to provide an introductory user's gu i<le to Method 11. Method I may be used to show that the Control limit for off-site beta air dose from gaseous effluents (3.3.2) has been met for releases over appropriate periods. This Control is based on the Objective in I OCFR50, Appendix I, Subsection B. I, which limits the estimated annual beta air dose at unrestricted area locations. Exceeding the Objective docs not immediately limit plant operation but requires a report to the NRC within 30 days. For any noble gas release, in any period, the <lose is taken from Equations B-4 a1id B-5 of Regulatory Guitk I. I09: fl Da1rs

                   =    3.17E+04       X/Qs        Q~T          DF('

(6-24) (mra<l) ( pCi- yr) Ci-sec (s"n Ill (Ci) ( mrad-m' pC1-yr J substituting X/Os Maximum long term average undepleted atmospheric dispersion factor for a stack release. I .59E-06 sec/m 3 We have (6-23) D~rs 0.050 DFf ( p~i-y~) 3 (mrad) (Ci) (mra~-m J C1-m pC1-yr For the ground level release:

          /1 Dairg     3. l 7E + 04 (X/Q )g    Q~;L  DFf                                      (6-44)

Off-Site Dose Calculation Manual Section 6 Rev.37 Page 28 of43 Vermont Yankee Nuclear Power Station

wh1.:re: (X/Q)g Maximum long-term average u1Hkpleted atmospheric dispersion factor for a ground level release. 3 3.65E-05 sec/m leading to:

              /I        1.16 Dairo     =                      DF{1                                              (6-43)

The calculation of ground level release dispersion parameters arc based on the location of the Reactor Building with respect to the site boundary that would experience the highest exposure. 6.8.3 Method II If Method I cannot be applied. or if the Method I dose determination indicates that the Control limit may be exceeded. or if a more exact calculation is required, then Method II may be applied. Method II consists of the models, input data and assumptions in Regulatory Guide I. I 09. Rev. I (Reference A). except where sitc-specific models, data or assumptions are more applicable. I 6. 9 Method to Calculate the Critical Organ Dose from Tritium and Particulates Effluent Control 3.3.3 limits the critical organ dose to a Member of the Public from all release sources of Tritium and particulates with half-lives greater than 8 days (hereafter called "P") in gaseous et11uents to 7.5 mrem per quarter and 15 mrem per year. Use Method I first to calculate the critical organ dose from both elevated and ground level vent releases. The total critical organ dose limit of Control 3.3.3 is the total contribution from both ground level and elevated releases occurring during the period of interest Use Method II if a more accurate calculation of critical organ dose is needed (i.e., Method I indicates the dose is greater than the limit). Off-Site Dose Calculation Manual Section 6 Rev. 37 Page 29 of 43 Vermont Yankee Nuclear Power Station

6.9.1 Method I D . ,*.".

                =  L      Q S"TI' 1

DFG

                                    . SIW                                                (6-25)

(mrcm) (Ci) mrcm) ( Ci STI' Q Total activity (Ci) released from the stack to the atmosphere of radionuclide "i" during the period of interest. For strontiums and tritium. use the most recent measurement. Site-specific critical organ dose factor for a stack gaseous release of radionuclide "i" (mrem/Ci). For each radionuclide it is the age group and organ \Vith the largest dose factor. See Table 1.1.12. The critical organ dose is calculated for ground level releases as follows: Deng L Q~iLI' DFGgico (6-44) (mrem) (Ci) ( m~~m J Total activity (Ci) released from ground level vents to the atmosphere of radionuclide "i" during the period of interest. For tritium, strontiums, and Fe-55 use the most recent measure. DFGgico Site-specific critical organ dose factor for a ground level release of nuclide "i" (mrem/Ci). For each radionuclide it is the age group and organ with the largest dose factor. See Table 1.1.12. The critical organ dose for the site is equal to Dcos + Dcog . Equations 6-25 and 6-44 can be applied under the following conditions (otherwise, justify Method I or consider Method II):

1. Normal operations (not emergency event),

2. P releases via the plant stack, Reactor Building to the atmosphere, and

3. Any continuous or batch release over any time period.

Off-Site Dose Calculation Manual Section 6 Rev. 37 Page 30 of 43 Vermont Yankee Nuclear Power Station

6.9.2 Basis for Method I This section serves three purposes: (I) lo document that Method I complies with appropriate NRC regulations, (2) to provide background and training information to Method I users, and (3) to provide an introductory user's guide to Method II. Method I may be used to show that the Control limit for off-site organ dose from gases (3.3.3) has been met for releases over the appropriate periods. Method I was developed such that "the actual exposure of an individual ... is unlikely to be substantially underestimated" (I OCFR50, Appendix I). The use below of a single "critical receptor" provides part or the conservative margin to the calculation of critical organ dose in l'vlethod I. Method 11 allows that actual individuals, with real behaviors, be taken into account for any given release. In fact, Method I was based on a Method II analysis of the critical receptor for the annual average conditions. For purposes or complying with the Control 3.3.3, maximum annual average atmospheric dispersion factors arc appropriate for batch and continuous releases. That analysis was called the "base case"; it was then reduced to form Method I. The base case, the method of reduction, and the assumptions and data used are presented below. The steps performed in the Method I derivation follow. First, in the base case, the dose impact to the critical receptor in the form of dose factors (mrem/Ci) of I curie release of each P radionuclide to gaseous enluents was derived. Then Method I was determined using simplifying and further conservative assumptions. The base case analysis uses the methods, data and assumptions in Regulatory Guide 1.109 (Equations C-2. C-4 and C-13 in Reference A). Tables 6.9.1and6.9.2 outline human consumption and environmental parameters used in the analysis. It is conservatively assumed that the critical receptor lives at the "maximum on:.site atmospheric dispersion factor lcc<1tio;i" as ddincJ in Section c. I 0. 1-lov>cvcr, he i:; exposed. conservativt::ly, to all pathways (see Section 6.10). The resulting site-specific dose factors are for the maximum organ and the age group with the highest dose factor for that organ. These critical organ, critical age dose factors are given in Table 1.1.12. For any gas release, during any period, the increment in annual average dose from radionuclide "i" is: (6-26) where DFGico is the critical dose factor for radionuclide "i" and Qi is the activity of radionuclide "i" released in curies. Method I is more conservative than Method II in the region of the effluent dose Control limit because it is based on the following reduction of the base case. The dose factors DFGico used in Method I were chosen from the base case to be the highest of the set for that radionuclide. In effect each radionuclide is conservatively represented by its own critical age group and critical organ. Off-Site Dose Calculation Manual Section 6 Rev.37 Page 31 of 43 Vermont Yankee Nuclear Power Station

6.9.3 METHOD II If Method I cannot be applied. or if the Method I dose exceeds the Control limit or if a more exact calculation is required, then Method II should be applied. Method II consists of the models. input data and assumptions in Regulatory Guide I. I 09. Rev. I (Reference A). except where site-specific models. data or assumptions arc more applicable. The base case analysis, documented above. is a good example of the use of Method 11. It is an acceptable starting point for a Method II analysis. Off-Site Dose Calculation Manual Section 6 Rev. 37 Page 32 of 43 Vermont Yankee Nuclear Power Station

TABLE 6.9.1 Environmental Parameters for Gaseous Effluents at Vermont Yankee (Derived from Reference A)* Vegetables Cow Milk Goat Milk l'v1eat Variable Stored Leafy Pasture Stored Pasture Stored Pasture Stored YV Agricultural (Kg/m 2 ) 2 2 0.70 2 0.70 2 0.70 2 Productivity p Soil Surface (Kg/111 2 ) 240 240 240 240 240 240 240 240 Density T Transport Time to User( 51 (Hrs) 48 48 48 48 480 480 TB Soil* Exposure Time( 11 (Hrs) 131400 131400 131400 131400 131400 131400 131400 131400 TE Crop Exposure Time to Plume (Hrs) 1440 1440 720 1440 720 1440 720 1440 TH Holdup After Harvest (Hrs) 1440 24 0 2160 0 2160 0 2160 QF Animals Daily Feed (Kg/Day) 50 50 6 6 50 50 FP Fraction of Year on Pasture( 21 0.50 0.50 0.50 FS Fraction Pasture When on Pasturel 31 FG Fraction of Stored Veg. Grown in Garden 0.76 FL Fraction of Leafy Veg. Grown in Garden FI Fraction Elemental Iodine= 0.5 A Absolute Humidity= 5.6 (gm/m 3)' 4'

ReguJatory Guide 1.109, Revision I.

Off-Site Dose Calculation l'vlanual Section 6 Rev. 37 Page 33 of 43 Vermont Yankee Nuclear Power Station

TABLE 6.9. l (Continued) Notes: (I) For Method II dose/dose rate analyses of identi tied radioactivity releases of less than one year. the soil exposure time for that release may be set at 8760 hours (I year) for all pathways. (2) For Method II dose/dose rate analyses performed for releases occurring during the first or fourth calendar quarters. the fraction of time animals are assumed to be on pasture is zero (nongrowing season). For the second and third calendar qua11ers. the fraction of time on pasture (FP) will be set at 1.0. FP m1y also be adjusted for specific farm locations if this information is so identified and reported as part of the land use census. (3) For Method II analyses, the fraction of pasture teed while on pasture may be set to less than 1.0 for specific farm locations if this information is so identified and reported as part of the land use census. (4) For all Method II analyses, an absolute humidity value equal to 5.6 (gm/m 3 ) shall be used to reflect conditions in the Northeast (

Reference:

Health Physics Journal, Vol. 39 (August), 1980: Page 318-320, Pergammon Press). (5) Variable Tis a combination of variables TF anct TS in Regulatory Guide I. I 09, Revision 1. Off-Site Dose Calculation Manual Section 6 Rev.37 Page 34 of 43 Vermont Yankee Nuclear Power Station

TABLE 6.9.2 Usage Factors fix Various Gaseous Pathways al Vermont Yankee (from Regulatory Guide I. I 09, Table E-5) Leafy Age Vegetables Vegetables Meal Inhalation Milk Group (kg/yr) (kg/yr) (I /yr) (kg/yr) {m 3/yr) Adult 520.00 64.00 310.00 110.00 8000.00 Teen 630.00 42.00 400.00 65.00 8000.00 Child 520.00 26.00 330.00 41.00 3700.00 Infant 0.00 0.00 330.00 0.00 1400.00 Off-Site Dose Calculation Manual Section 6 Rev.37 Page 35 of 43 Vermont Yankee Nuclear Power Station

6.10 Receptor Point and Long-Term Average Atmospheric Dispersion Factors for Important Exposure Pathways The gaseous effluent dose methods have been simplified by assuming an individual whose behavior and living habits inevitably lead to a higher dose than anyone else. The following exposure pathways to gaseous effluents listed in Regulatory Guide I. I 09 (Reference A) have been considered tor tritium, and particulates with half lives greater than 8 days: I. Direct exposure to contaminated ground;

2. Inhalation of air; 3 Ingestion of vegetables;

4. Ingestion of meat.

Beta air doses have also been considered for noble gases in plant effluents along with whole body and skin dose rate calculati{)ns. Section 6.10.1 details the selection of important off-site locations and receptors. Section 6.10.2 describes the atmospheric model used to convert meteorological data into atmospheric dispersion factors. Section 6.10.3 presents the maximum atmospheric dispersion factors calculated at each of the off:*site receptor locations. 6.10.1 Receptor Locations Distances to the site boundary from the two evaluated gaseous release pathways (the Stack and Reactor Building) are provided in Table 6.10.2. Four important off-site receptor locations are considered in the dose and dose rate equations for gaseous radioactive effluents from these two release pathways. They are: I. The point of maximum gamma exposure (maximum gamma X/Q) from an overhead noble gas cloud for determining skin and whole body dose rates;

2. The point of maximum ground level air concentration (maximum undepicted X/Q) of noble gases for determining skin and beta air dose rates and doses;

3. The point of maximum ground level air concentration (maximum depleted X/Q) of particulates for determining critical organ dose from inhalation; and

4. The point of maximum deposition (maximum D/Q) of particulates for determining critical organ dose from ingestion.

Off-Site Dose Calculation Manual Section 6 Rev.37 Page 36 of 43 Vermont Yankee Nuclear Power Station

The Stack release pathway was evaluated as an elevated release assuming a constant nominal Stack flow rate of 72,000 din. The point of maximum gamma exposure from Stack releases (SSE sector, 700 meters) was determined by finding the maximum live-year average gamma X/Q at any off-site location. The location of the maximum ground level air concentration and deposition of particulates (WNW sector. 2150 meters) was determined by finding the maximum live-year average depleted X/Q and D/Q at any off-site location. For the purposes of determining the Method I dose factors for tritium, and particulates, an animal was assumed to exist at the location of highest calculated ground level air concentration and deposition particulates as noted above. This location then conservatively bounds the deposition of particulates at all real animal locations. The Reactor Building release pathway was evaluated as a ground level release using the same meteorological period-ot:.record as the stack. The highest long-term atmospheric dispersion factors at the site boundary were determined (see Table 6.10.1) and doses and dose rates to the critical oft:.site receptor were calculated assuming the highest site boundary atmospheric dispersion factors all occurred at the same location. 6.10.2 Vermont Yankee Atmospheric Dispersion Model The long-term average atmospheric dispersion factors are computed for routine releases using AEOLUS-2 Computer Code (Reference 8). AEOLUS-2 is based, in part, on the constant mean wind direction model discussed in Regulatory Guide 1.111 (Reference C). Since AEOLUS-2 is a straight-line steady-state model, site-specific recirculation correction factors were developed for each release pathway to adjust the AEOLUS-2 results to account for temporal variations of atmospheric transport and diffusion conditions. The applicable recirculation correction factors are listed in Table 6.10.3. AEOLUS-2 produces the following average atmospheric dispersion factors for each location: I. Undepleted X/Q dispersion factors for evaluating ground level concentrations of noble gases;

2. Depleted X/Q dispersion factors* for evaluating ground level concentrations of particulates;

3. Gamma X/Q dispersion factors for evaluating gamma dose rates from a sector averaged finite cloud (undepleted source); and

4. D/Q deposition factors for evaluating dry deposition of elemental particulates.

Off-Site Dose Calculation Manual Section 6 Rev. 37 Page 37 of 43 Vermont Yankee Nuclear Power Station

The Reactor Building depicted X/Q and D/Q factors were derived using the plume depletion and deposition curves provided in Regulatory Guide 1.111. 1-lowcwr, because the Regulatory Guide 1.111 depletion and deposition curves are limited to an effective release height of 100 meters or less and the Vermont Yankee Stack effective release height (stack height plus plume rise) can exceed I 00 meters, the Stack depleted X/Q and D/Q factors were derived using the deposition velocity concept presented in "Meteorology and Atomic Energy - 1968" (Reference E, Section 5-3.2), assuming a constant deposition velocity of I cm/sec. 6.10.3 Long-Term Average Atmospheric Dispersion Factors for Receptors Actual measured meteorological data for the five-year period, 2002 through 2006, were analyzed to determine all the values and locations of the maximum off-site long-term average atmospheric dispersion factors. Each dose and dose rate calculation incorporates the maximum applicable off-site long-term average atmospheric dispersion factor. The '.'alues tu.:d and their locations an; summarizcJ in Table 6. ! O. l. Table 6.1 O. l also indicates which atmospheric dispersion factors are used to calculate the various doses or dose rates of interest. Off-Site Dose Calculation Manual Section 6 Rev. 37 Page 38 of 43 Vermont Yankee Nuclear Power Station

TABLE 6.10.l Atmospheric Dispersion Factors Dose to Individual Dose to Air Release Dispersion Pathway Factor Total Body Skin Critical Organ Gamma Beta X/Q Depleted - - l .25E-06 - - (sec/m 3) (2150m \VN\V) XIQ Undepleted 3

                                          -            l .59E-06               -            -               1.59£-06 (sec/m )                      (2650 111 SW)                                         (2650m SW)

Stack DIQ - - 1.25£-08 - - (l/m2) (2150111 WNW) X/QY 7.51E-07 7.5 lE-07 - - (sec/m 3) (700m SSE) (700m SSE) XIQ Depleted - - 3.44E-05 - - (sec/m3 ) (477m ENE) XIQ Undepleted - 3.65E-05 - - 3.65E-05 Reactor (sec/m 3) (477m ENE) (477m ENE) Building D/Q - - 6.40E-08 - - (l/m 2) (419mS) X/QY 3 7.37E-06 7.37E-06 - - (sec/m ) (477m ENE) (477m ENE) Off-Site Dose Calculation l\fanual Section 6 Rev.37 Page 39 of 3 Vermont Yankee Nuclear Power Station

Tt\BLE 6.10.2 Site Boundary Distances Downwind Stack Reactor Building SeL:tor Releases Releases N 400111 529111 NNE 350 111 468111 NE 350 Ill 448 Ill ENE 400 Ill 477 m E 500111 499111 ESE 700 m 482 Ill SE 750111 512111 SSE ssr m 555 m s 385 111 419111 SSW 300 111 575 111 SW 250111 505 111 WSW 250111 418 m w 300111 402 m WNW 400 m 528111 NW 550 Ill 917111 NNW 550111 831 m Off-Site Dose Calculation Manual Section 6 Rev. 37 Page 40 of 43 Vermont Yankee Nuclear Power Station

T /\ALE 6.10.3 Recirculation Correction Factors A. Stack Releases Sector 0.5 Mi 1.5 Mi 2.5 Mi 3.5 Mi 4.5 Mi 7.5 Mi N 1.4 1.4 1.2 I. I 1.0 1.0 NNE 1.8 1.8 1.4 1.2 1.0 1.0 NE 1.8 1.8 1.3 I. I 1.0 1.0 ENE 2.1 2.1 1.4 1.2 1.0 1.0 E 1.7 1.7 1.2 1.0 1.0 1.0 ESE 1.5 1.5 1.3 I. I 1.0 1.0 SE 1.8 1.8 1.3 1.2 I. I 1.0 SSE 1.4 1.4 1.2 1.2 1.2 1.2 s 1.3 1.3 I. I I. I 1.2 1.2 SSW 1.8 1.8 1.5 1.4 1.4 1.2 SW 2.1 2.1 1.7 1.6 1.4 I. I WSW 2.4 2.4 I. 9 1.6 1.5 I. I w 1.8 1.8 1.5 1.4 1.3 1.0 WNW 1.8 1.8 1.7 1.5 1.4 1.3 NW 1.5 1.5 1.3 1.3 1.3 I. I NNW 1.5 1.5 1.2 1.2 I. I I. I B. Reactor Building Release Sector 0.5 Mi 1.5 Mi 2.5 Mi 3.5 Mi 4.5 Mi 7.5 Mi N I. I I. I I. I I. I I. I 1.0 NNE 1.2 1.2 1.2 I.I I. I 1.0 NE I. I 1.2 I. I I. I 1.0 1.0 ENE 1.2 1.3 1.4 1.4 1.4 1.3 E I. I 1.3 1.4 1.4 1.4 1.2 ESE I. I I. I 1.2 I. I I. I 1.0 SE 1.0 I. I I.I I. I I.I 1.1 SSE 1.2 1.2 1.2 1.2 1.2 1.2 s 1.0 1.0 1.0 1.0 1.0 1.0 SSW 1.0 I. I 1.0 1.0 1.0 1.0 SW 1.2 1.3 1.2 1.0 1.0 1.0 WSW 1.1 I. I 1.0 1.0 1.0 1.0 w 1.2 1.2 I. I 1.0 1.0 1.0 WNW 1.2 1.4 1.3 1.2 1.2 1.0 NW I. I I.I 1.0 1.0 1.0 1.0 NNW 1.1 1.2 1.2 1.2 1.2 1.1 Off-Site Dose Calculation Manual Section 6 Rev. 37 Page41 of43 Vermont Yankee Nuclear Power Station

6.11 Method to Calculate Direct Dose From Plant Operation Erlluent Control 3.4.1 (40CFRI 90) restricts the dose to the whole body or any organ to any member of the public from all station sources (including direct radiation from fixed sources on-site) to 25 mrem in a calendar year. 6.1 I. I Deleted 6.11.2 Deleted 6.11.3 Dclcted 6.1 1.4 Independent Spent Fuel Storage Installation Dose Contribution The lndependcnt Spent Fucl Storage Installation (ISFSI) has been constructcd to p!*ovidc *:secure locatiun for th1: long term storage of spent nudcar fud bundks. This installation is located just north of the Rad waste Building and just east of the North Warehouse within the Vermont Yankee Protected Area. The vendor, 1-Ioltec International. has prepared a study rcport to satisfy the requirements of I OCFR 72.104 and this document is included as Reference 1-1. During the first fuel storage evolution, five casks will be located on the installation pad. Over subsequent years, up to 35 casks will be stored on the pad. The shielding analysis of the Independent Spent Fuel Storage Installation is provided in Reference H. The report analyzes the dose generated from a single cask as well as the dose generated from 5 casks for both the first row (266 meters to the west site boundary) and the last row (300 meters from the west site boundary) on the ISFSI pad. The dose from each cask to the west site boundary (DR-53A Location) is monitored with RH.'!P TLDs. 6.11.4 Total Direct Dose Summary Estimates of direct exposure above background in areas at and beyond the site boundary can be determined from measurements made by environmental TLDs located as shown in Table 7.1 and Figure 7.4 that are part of the Radiological Environmental Monitoring Program. Alternatively, direct dose calculations from identified fixed sources on site can be used to estimate the off-site direct dose contribution where TLD information may not be applicable. I 6.11.5 Deleted Off-Site Dose Calculation Manual Section 6 Rev.37 Page 42 of43 Vermont Yankee Nuclear Power Station

6.12 Cumulative Doses Cumulative Doses for a calendar quarter and a calendar year must be maintained to demonstrate a compliance with Controls 3.2.2, 3.3.2. and 3.3.3 (I OCFRSO, Appendix I Jose objectives). In addition. if the requirements of the Action Statement of Control 3.4.1 dictate. cumulative doses over a calendar year must be determined (demonstration of compliance with total dose, including direct radiation per requirements of 40CFR 190). To ensure the limits arc not exceeded, a running total must be kept for each release. Demonstration of compliance with the dose limits of 40CFR 190 is considered as demonstrating compliance with the 0.1 rem limit of I OCFR20. I 30 I (a)( I) for members of the public in unrestricted areas. Off-Site Dose Calculation Manual Section 6 Rev.37 Page 43 of 43 Vermont Yankee Nuclear Power Station

7.0 ENVIRONMENTAL MONITORING PROGRAM The radiological environmental monitoring stations are listed in Table 7.1. The locations of the stations with respect to the Vermont Yankee plant are shown on the maps in Figures 7-1 to 7-7. 7.1 Intercomparison Program All routine radiological analyses for environmental samples are performed at offsite environmental laboratories. The laboratories participate in several commercial inter-comparison programs in addition to an internal QC sample analysis program and the analysis of client-introduced QC sample programs. The external programs may include the Department of Energy - Mixed Analyte Perfornrnnce Evaluation Program (MAPEP), Analytics Cross-Check Program - Environmental Inter-laboratory Cross-Chcck Program, and Environmental Resources Association - Environmental Radioactivity Perfomiancc Evaluntion Program or other NRC-approved s0un;es. 7.2 Airborne Pathway Monitoring The environmental sampling program is designed to achieve several major objectives, including sampling air in predominant up-valley and down-valley wind directions and at proper control locations, while maintaining continuity with two years of preoperational data and all subsequent years of operational data (post 1972.) The chosen air sampling locations are discussed below. To assure that an unnecessarily frequent relocation of samplers will not be required due to short-term or annual fluctuations in meteorology, thus incurring needless expense and destroying the continuity of the program, long term, site specific ground level D/Qs {five-yc~~r 2\*erages - 1978 throug!-11982) were evaluat.;d in comparison to the existing air monitoring locations to determine their adequacy in meeting the above-stated objectives of the program and the intent of the NRC general guidance. The long-term average meteorological data base precludes the need for an annual re-evaluation of air sampling locations based on a single year's meteorological history. The Connecticut River Valley in the vicinity of the Vermont Yankee plant has a pronounced up- and down-valley wind flow. Based on five years of meteorological data, wind blows into the 3 "up-valley" sectors (N, NNW, and NW) 27 percent of the time, and the 4 "down-valley" sectors (S, SSE, SE, and ESE) 40 percent of the time, for a total "in-valley" time of 67 percent. Off-Site Dose. Calculation Manual Section 7 Rev.37 Page 1 of 10 Vermont Yankee Nuclear Power Station

Station AP/CF-12 (NNW, 3.6 km) in North Hinsdale, New Hampshire, monitors the up-valley sectors. It is located in the sector that ranks fourth overall in terms of wind frequency (i.e., in terms of how ollen the wind blows into that sector), and is approximately 0. 75 miles from the location of the calculated maximum ground level D/Q (i.e., for any location in any sector, for the entire Vermont Yankee environs). This station provides a second function by its location in that it also monitors North Hinsdale, New Hampshire, the community with the second highest ground level D/Q for surrounding communities, and it has been in operation since the preoperational period. The down-valley direction is monitored by the River Station Number 3-3 (AP/CF-11, SSE, l.9 km). This station resides in the sector with the maximum wind frequency and they bound the down-valley point of calculated maximum ground level D/Q (the second highest overall ground level D/Q for any location in any sector). Station AP/CF-11 is approximately one mile from this point, between it and the plant. This station h;1s been in operation since the preoperational period. The control air sampler was located al Spofford Lake (AP/CF-21, NNE, 16.4 km) due to its distance from the plant and the low frequency for wind blowing in that direction based on the long-term (five-year) meteorological history. Sectors in the general west to southwest direction, which would otherwise have been preferable due to lower wind frequencies, were not chosen since they approached the region surrounding the Yankee Atomic plant in Rowe, Massachusetts. 7.3 Distances and Directions to Monitoring Stations It should be noted that the distances and directions for direct radiation monitoring locations in Table 7.1, as well as the sectors shown in Figures 7-5 and 7-6, arc keyed to the ce11ter of the Reactor Building due to the critical nature of the Reactor Building-to-TLD distance for close-in stations. For simplicity, all other radiological environmental sampling locations use the plant stack as the origin. Control Table 3.5.1, Footnote a, specifies that in the Annual Radiological Environmental Operating Report and ODCM, the reactor shall be used as the origin for all distances and directions to sampling locations. Vermont Yankee interprets "the reactor" to mean the reactor site which includes the plant stack and the Reactor Building. The distances to the plant stack and Reactor Building will, therefore, be used in the Annual Radiological Environmental Operating Reports and ODCM for the sampling and TLD monitoring stations, respectively. Off-Site Dose Calculation Manual Section 7 Rev. 37 Page 2of10 Vermont Yankee Nuclear Power Station

Table 7.1 Radiological Environmental Monitoring Stations( 1> Exposure Pathway Sample Location Distance and/or Sample and Designated Code( 2 l (km)! 5> Direction(Sl I. AIRBORNE (Radioiodine and Particulate) AP/CF-11 River Station No. 3-3 1.88 SSE AP/CF-12 N. Hinsdale, NH 3.61 NNW AP/CF-21 Spofford Lake( 9l 16.36 NNE

2. WATERBORNE

a. Surface WR-11 River Station No. 3-3 1.88 Downriver WR-21 Rt. 9 Bridge('i) 11.83 Upriver

b. Ground WG-11 Plant Well 0.24 On-Site WG-12 Vernon Nursing Well 2.13 SSE WG-22 Copeland Well( 9 l 13.73 N

c. Sediment SE-11 Shoreline Downriver 0.57 SSE From SE-12 North Storm 0.13 E Shoreline Drain Outfall(J)

3. INGESTION

a. Deleted

b. Mixed TG-11 River Station No. 3-3 1.88 SSE Grasses TG-12 N. Hinsdale, NH 3.61 NNW TG-21 Spofford Lake( 9l 16.36 NNE

c. Silage TC-11 Miller Farm 0.82 w TC-18 Blodgett Farm 3.60 SE TC-22 Franklin Farm( 9 > 9.73 WSW

d. Fish FH-11 Vernon Pond (6) (6)

FH-21 Rt. 9 Bridge( 9J 11.83 Upriver Off-Site Dose Calculation Manual Section 7 Rev. 37 Page 3of10 Vermont Yankee Nuclear Power Station

TABLE 7.1 (Continued) Exposure Pathway Sample Location Distance and/or Sample and Designated Code( 21 (km)( 5 > Direction( 5 l

4. DIRECT RADIATION DR-I River Station No. 3-3 1.61 SSE DR-2 N. Hinsdale, NH 3.88 NNW DR-5 Spofford Lake('JJ 16.53 NNE DR-6 Vernon School 0.52 WSW DR-7 Site Boundary(?) 0.28 w DR-8 Site Boundary 0.25 SSW

( l) Sample locations are shown on Figures 7.1 to 7.7. (2) Station Nos. 10 through 19 arc indicator stations. Station Nos. 20 through 29 are control stations (except silage and the direct radiation stations). (3) To be sampled and analyzed semiannually. (4) Deleted (5) Distance and direction from the center of the Reactor Building for direct radiation monitors; from the plant stack for all others. (6) Fish samples arc collected from anywhere in Vernon Pond, which is adjacent to the plant (see Figure 7-1). (7) Deleted. (8) Deleted. (9) Control stations Off-Site Dose Calculation Manual Section 7 Rev.37 Page 4of10 Vermont Yankee Nuclear Power Station

Figure 7-1 Environmental Sampling Locations in Close Proximity to the Plant Off-Site Dose Calculation Manual Section 7 Rev. 37 Page 5of10 Vermont Yankee Nuclear Power Station

I

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   '                                                                                NE Figure 7-2 Environmental Sampling Locations Within 5 Km of Plant Off-Site Dose Calculation Manual Section 7 Rev.37 Page 6of10 Vermont Yankee Nuclear Power Station

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                                  ~NNW
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KM II I 2 3 J Figure 7-3 Environmental Sampling Locations Greater than 5 Km from Plant Off-Site Dose Calculation Manual Section 7 Rev.37 Page 7of10 Vermont Yankee Nuclear Power Station

KM 0 .2 .4 Figure 7-4 TLD Locations in Close Proximity to Plant Off-Site Dose Calculation Manual Section 7 Rev. 37 Page 8of10 Vermont Yankee Nuclear Power Station

f

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                                         /l KM 0         .5         LS         2 Figure 7-5 TLD Locations Within 5 Km of Plant Off-Site Dose Calculation Manual Section 7 Rev. 37 Page 9of10 Vermont Yankee Nuclear Power Station

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f SSW 0- I 2- ) 4I Figure 7-6 TLD Locations Greater Than 5 Km from Plant Off-Site Dose Calculation Manual Section 7 Rev. 37 Page 10of10 Vermont Yankee Nuclear Power Station

8.0 SETPOINT DETERMINATIONS Chapter 8 contains the basis for plant procedures used to meet the setpoint requirements of the Radioactive Effluent lnstmmcntation Controls. They are Control 3.1.1 for liquids and c<mtrol 3.1.2 for gases. Each outlines the instmmentation channels and the basis for each sctpoint. 8.1 Liquid Effluent Instmmentation Setpoints Control 3.1.1.1 requires that the radioactive liquid effluent instmmentation in Control Table 3.1.1 have alarm setpoints in order to ensure that Control 3.2.1 is not exceeded. Control 3.2.1 limits the activity concentration at any time in liquid effluents to ten items or less the effluent concentration values in Appendix B, Table 2, Column 2 of 10CFR20.l00 I through 20.2402, and a total noble gas concentration limit of 2E-04

     ~LCi/ml.

8.1.1 Liquid Radwaste Discharge Monitor (RM-17-350) The sample tank pathways shown on Figure 9-1 are monitored by the liquid radwaste discharge monitor (RM-17-350). Periodic batch releases may be made from the waste sample tanks, detergent waste tank or floor drain sample tank. Off-Site Dose Calculation Manual Section 8 Rev. 36 Page 1 of 17 Vermont Yankee Nuclear Power Station

8.1.1.1 Method to Determine the Setpoint of the Liquid Rad waste Discharge Monitor (RM-17-350) The instrnment response (in counts per second) for the limiting concentration at the point of discharge is the setpoint, denoted Rsctpoint. and is determined as follows: (8-1) (cps) (#) cps - _ml) * (~tCi) ( ~tCt ml Where: Fd OF Dilution factor (as a conservative 1m:asurc, a OF of Fm at least 1000 is used) (dimensionless) (8-2) Flow rate past monitor (gpm) Flow rate out of discharge canal (gpm} DFmin Minimum allowable dilution factor (dimensionless) (8-3) EH1uent concentration values for radionudide i" frum I OCFR20. l 001-20.2402, Appendix B, Table 2, Column 2 (~tCi/ml) Activity concentration of radionuclide "i" in mixture at the monitor (~tCi/ml) St Detector counting efficiency from the most recent liquid radwaste discharge monitor calibration curve (cps/(µCi/ml)) Off-Site Dose Calculation Manual Section 8 Rev. 36 Page 2of17 Vermont Yankee Nuclear Power Station

8.1.1.2 Liquid Radwaste Discharge Monitor Setpoint Example The following alarm setpoint example is for a discharge of the floor drain sample tank. The liquid radwaste discharge monitor has a typical counting efficiency, S1, of 4.9E+06 cps per I ~LCi/ml of gamma emitters which emit one photon per disintegration. The activity concentration of each radionuclide, C 111 j, in the floor drain sample tank is dete1mined by analysis of a representative grab sample obtained at the radwaste sample sink. This setpoint example is based on the following data: Cmi (~LCi/ml) ECLi (~LCi/ml) Cs-134 2. I 5E-05 9E-07 Cs-13 7 7.48E-05 I E-06 Co-60 2.56E-U5 3E-06 I cmi = 2.15E-05 + 7.48E-05 + 2.56E-05 ( ~LCi) ( ~LCi) ( ~LCi) (~~i) ml ml ml

                 = l.22E-04

(µCi) ml (8-3)

                      µCi-ml)

( ml-µCi

                 =   O.l [2.15E-05 + 7.48E-05 + 2.56E-05]

9E-07 lE-06 3E-06

                            µCi-ml)       ~tCi- ml)       µCi-ml)

( ml-µCi ( ml-µCi ( ml-µCi

                 =   10.7 Off-Site Dose Calculation Manual Section 8 Rev. 36 Page 3of17 Vermont Yankee Nuclear Power Station

The minimum dilution factor, OF 111 j 11 , needed to discharge the mixture of radionuclides in this example is 10.7. As a conservative measure, an actual dilution factor, OF, or 1,000 is usually used. The release rate of the floor drain sample tank may be adjusted from 0 to 50 gpm and the dilution pumps can supply up to 20,000 gpm or dilution water. With the dilution flow taken as 18,000 gpm, the release rate from the floor drain sample tank may be detennined as follows: Fd (8-4) OF (gpm) (gpm) 18,000 gpm

                    =  18 gpm 1,000 Under these conditions, the setpoint of the liquid radwaste discharge monitor is:

(8-1) OF Rsclpoint = S1 OFmin (cps) (#) cps-.ml) (µCi) ( ~tC1 ml

                  =  l,OOO 4.9E + 06     l.22E-04 10.7 (cps)          (#)     ( cp:t~im!)    ( ~~i) 55,869 cps In this example, the calculated limiting count rate alarm point for the liquid radwaste discharge monitor would be 55,869 cps above background. Plant procedures apply administrative limits below the calculated limiting count rate to account for such elements as instrument uncertainty and early alarm warning before exceeding Control limits.

Off-Site Dose Calculation Manual Section 8 Rev.36 Page 4of17 Vermont Yankee Nuclear Power Station

8.1.1.3 Basis for the Liquid Radwaste Discharge Monitor Setpoint The liquid radwaste discharge monitor setpoint must ensure that Control 3.2.1 is not exceeded for the appropriate in-plant pathways. The liquid radwaste discharge monitor is placed upstream of the major source of dilution flow and responds to the concentration of radioactivity discharged in batch releases as follows: (8-5) R = L cmi (cps) ( ~tCi) ml (cps-.ml)

                                  ~tC1 Where:

R Response of the monitor (cps) Detector counting efficiency for radionuclide "i" (cps/(µCi/ml)) Activity concentration of radionuclide "i" in mixture at the monitor (~tCi/ml) The detector calibration procedure establishes a counting efficiency for a given mix of nuclides seen by the detector. Therefore, in Equation 8-5 one may substitute S1 for S1i, where S1 represents the counting efficiency determined for the current mix of nuclides. If the mix of nuclides changes significantly, a new counting efficiency should be c1etermined for calculating the setpoint. (8-6) R = (cps) ( cps-.ml)

                       ~tC1

(~i) Off-Site Dose Calculation Manual Section 8 Rev. 36 Page 5of17 Vermont Yankee Nuclear Power Station

The effluent concentration for a given radionuclide must not exceed I 0 times the I 0 CFR Part 20 ECL at the point of discharge to an unrestricted area at any time. When a mixture of radionuclides is present, the concentration at the point of discharge to an unrestricted area shall be limited as follows: (8-7) ( ~tCi-ml) ml- ~tCi Where: Activity concentration of radionuclide "i" in the mixture at the point of discharge to an unrestricted area (~tCi/ml) Effluent concentration limit for radionuclide "i" from I OCFR20.1001-20.2402, Appendix B, Table 2, Column 2 (~tCi/ml) The activity concentration of radionuclide "i" at the point of discharge is related to the activity concentration of radionuclide "i" at the monitor as follows: (8-8) Fm cdi cmi Fd ( ~tCi) (µCi) ( gpm) ml ml gpm Where: Activity concentration of radionuclide "i" in the mixture at the point of discharge (µCi/ml) Flow rate past monitor (gpm) Flow rate out of discharge canal (gpm) Off-Site Dose Calculation Manual Section 8 Rev.36 Page 6of17 Vermont Yankee Nuclear Power Station

Substituting the right half of Equation 8-8 for Cdi in Equation 8-7 and solving for Ft1/F 111 yields the minimum dilution factor needed to comply with Equation 8-7: (8-3)

Fd ;: :: cmi DFmin 2:

Fm ECLi

10

( gpm) (~tCi- ml) gpm ml- ~tCi Where: Fc1 Flow rate out of discharge canal (gpm) F111 Flow rate past monitor (gpm) Cmi Activity concentration of radionuclide "i" in mixture at the monitor (µCi/ml) EC Li Effluent concentration limit for radionuclide "i" from I OCFR20. l 001-20.2402, Appendix B, Table 2, Column 2 (µCi/ml) 10 The instantaneous concentration multiplier allowed by Control 3.2.1 If Fc1/Fm is less than DFmin, then the tank may not be discharged until either Fc1 or Fm or both are adjusted such that: (8-3) gpm) ( gpm Off-Site Dose Calculation Manual Section 8 Rev. 36 Page 7of17 Vermont Yankee Nuclear Power Station

Usually Fdlrm is greater than DF 111 in (i.e., there is more dilution than necessary to comply with Equation 8-7). The response of the liquid rad waste discharge monitor at the sctpoint is therefore: (8-1) DF Rsetpoint = s, I cmi DFmin (cps) (#) ( cps-.mll ( ~LCi)

                                ~LCt          ml 8.1.2 Service Water Discharge Monitor ( RM-17-351)

The service water pathway shown on Figure 9-1 is continuously monitored by the service water discharge monitor (RM-17-351 ). The water in this line is not radioactive under normal operating conditions. The alann sctpoint on the Service Water Monitor (SWM) is set in accordance with the monitor's ability to detect dilute concentrations of radionuclide mixes that are based on measured nuclide distributions in the spent fuel pool. From routine sample gamma isotopic analyses, a Composite Maximum Permissible Concentration (CMPC) is calculated as follows: or (8-22) where: c Total concentration of detected radioactivity in spent fuel pool sample (~LCi/ml) Fraction of total radionuclide concentration represented by the ith radionuclide in the mix Maximum Permissible Concentration limit for radionuclide "i" as listed in Appendix G (µCi/ml) The Composite Effluent Concentration Limit (CECL) is also calculated using the equation above by substituting a value of lOX the appropriate ECL value from 10CFR20.l001-20.2402, Appendix B, Table 2, Column 2, for MPC. If the SWM's minimum achievable alarm setpoint is higher than the required CMPC equivalent count rate (or the CECL equivalent count rate if it is lower than the CMPC count rate), the monitor is declared non-functional, and daily SWM grab samples are collected and analyzed until the calculated CMPC (or CECL) equivalent count rate is above the SWM's alarm setpoint. Off-Site Dose Calculation Manual Section 8 Rev.36 Page 8of17 Vermont Yankee Nuclear Power Station

For example, if the spent fuel pool radionuclide mix distribution is as listed below, then the corresponding CMPC is calculated as follows: r1 (cone/total I OCFR'.W MPCi fi/MPCi Nudides Cone (~tCi/ml) cone) (~LCi/ml) (~LCilml) 1-131 6.00E-6 6.59E-2 3.0E-7 2.20E+5* 1-133 5.00E-6 5.49E-2 l.OE-6 5.49E+4 Co-60 8.00E-5 8.79E-I 3.0E-5 2.93E+4 Totals 9.1 OE-5 1.00 3.04E+5 CMPC= 1/3.04E+5 = 3.29E-6 (~tCi/ml) The CECL is also calculated by using the above methodology and substituting a value of I OX the appropriate ECL listed in I OCFR20. I 001-20.2402, Appendix 8, Table 2, Column 2, f(x MPC values. For this example, the calculated CECL is equal to 2.73E-6 pCi/ml.) If the SWM alam1 is set at 5 CPS (300 CPM) above background, and the current calibration factor for this monitor is 1.17E+8 CPMhtCi/ml, then the SWM will alam1 if a concentration as low as 2.56E-6 ~LCilml above background passes by the monitor. Since the most limiting CMPC or CECL (calculated above to be 2.73E-6 µCi/ml) is above the alarm setpoint (equal to 2.56E-6 ~LCi/ml), the SWM will be capable of alarming if radioactivity in excess of limiting concentration values for release to unrestricted areas passes by the monitor. However, if the composite concentration (CMPC or CECL) for the service water was found to he less than the SWM alarm setpoint of 2.56E-6 ~LCi/ml, then daily service water grab samples would have to be collected and analyzed until the composite concentration becomes greater than the concentration corresponding to the SWM's alarm setpoint. Service water is sampled if the monitor is out of service or if the alarm sounds. Under normal operating conditions, the concentration of radionuclides at the point of discharge to an unrestricted area from the service water effluent pathway will not exceed the effluent concentration limits specified in 10CFR20.l001-20.2402, Appendix B, Table 2, Column 2. Off-Site Dose Calculation Manual Section 8 Rev. 36 Page 9of17 Vermont Yankee Nuclear Power Station

8.2 Gaseous Ert1uent Instrumentation Setpoints Control J.1.2 requires that the radioactive gaseous effluent instrumentation in Control Table 3.1.2 have their alann setpoints set to ensure that Control 3.3.1 is not exceeded. 8.2.1 Plant Stack Noble Gas J\divity Monitors {RM-17-156 and RM-17-157) The plant stack noble gas activity monitors are shown on Figure 9-2. 8.2.1.1 Method to Determine the Sctpoint of the Plant Stack Noble Gas Activity Monitors (RM-17-156 and RM-17-157) The setpoints of the plant stack noble gas activity monitors are detenninecl in the same manner. The plant stack noble gas activity monitor response in counts per minute at the limiting off-site noble gas dose rate to the skin is the setpoint, denoted Rspt* Rspt 1s: DELETED (8-9) (8-10) R ~~: 11

                     =  3,000 F         OF~

Off-Site Dose Calculation Manual Section 8 Rev.36 Page 10of17 Vermont Yankee Nuclear Power Station

where: IE+06 Number of pCi per ~LCi (pCi/~tCi) 7.51 E-07 [X/Q]Y, maximum five-year average gamma atmospheric dispersion factor (sec/m3) s .. 0 Plant Stack Kr85 detector counting efficiency from the most recent calibration (cpm/(~LCi/cc)) F Appropriate plant stack flow rate (cm3 /sec) The relative release rate Kr85 at the monitor (stack <ir) (~LCi/sec) R ....

             *I, 111 Response of the monitor at the limiting skin dose rate ( cpm) spt 3,000               Limiting skin dose rate (mrern/yr)

DF'c Kr85 Composite skin dose factor (2.15E-03 mrem-sec/µCi-yr, see Table 1.1.l 0) DELETED (8-12) DF[5 Kr85 Composite skin dose factor (2.15E-03 mrem-sec/µCi-yr, see Table 1.1.10) 8.2. i .2 Plant Stack Noble Gas Activity Monitor Setpoint Example The following setpoint example for the plant stack noble gas activity monitors demonstrates the use of Equations 8 - l 0 for determining setpoints. The plant stack noble gas activity monitors, referred to as "Stack Gas I" (RM-17-156) and "Stack Gas 11" (RM-17-157), consist of beta sensitive scintillation detectors, electronics, a ratemeter readout, and a digital scaler which counts the detector output pulses. A strip chart recorder provides a permanent record of the ratemeter output. The monitors have typical calibration factors, Sg, of!about 7 .9E+07 cpm per

         ~LCi/cc of noble gas. The nominal plant stack flow is 3.40E+07 cc/sec 3

((72,000 cfm x 28,300 cc/ft )/60 sec/min). Off-Site Dose Calculation Manual Section 8 Rev.36 Page 11of17 Vermont Yankee Nuclear Power Station

Next: 3,000 F DF'c (3,000) (7.9£ + 07) I I (3.4£ + 07) (2.15£ -03)

                 = 3,242,134 cprn The setpoint, Rspt' R~~:n. In this example the "Stack Gas I" and "Stack Gas II" noble gas activity monitors should each be set at some administrative value below 3,242, 134 cpm above background to provide conservatism for such issues as instrument uncertainty and secondary releases from other locations. As an example, a conservative value might be based on controlling release rates from the plant in order to maintain off-site air concentrations below 20 x ECL when averaged over an hour, or to account for other minor releases. For example, if an administrative limit of 70 percent of the Control skin dose limit 3000 mrem/yr (3,242, 134 cpm) is chosen, then the noble gas monitor alarms should be set at no more than 2,269,493 cpm above background (0.7 x 3,242,000 = 2,269,493).

Off-Site Dose Calculation Manual Section 8 Rev. 36 Page 12 of 17 Vermont Yankee Nuclear Power Station

8.2.1.3 Basis for the Plant Stack Noble Gas Activity Monitor Setpoints The setpoints of the plant stack noble gas activity monitors must ensure that Control 3.3. l .a is not exceeded. Section 6.5 shows that Equation 6-7 is an acceptable method for detcm1ining compliance with the Control limits. Skin dose is more limiting 85 for Kr . Therefore, only the skin <lose must be considered separately. The derivations of Equation 8-l 0 begins with the general equation for the response R of a radiation monitor: (8-13) R =I cmi 3 (cpm) cpm -~m J ( cm

                                        ~LC~ J

( ~LC! where: R = Response of the instmment (cpm) Sgi = Detector counting efficiency for Kr85 (cpm/(µCi/cm 3 )) at the noble gas activity monitor (~LCi/cm ) 85 3 Cmi = Activity concentration of Kr (8-14) F

         ~tCi)

( cm 3 (µCi) (~) 3 sec cm where: Qi The release rate of Kr85 (µCi/sec). 3 F Appropriate flow rate (cm /sec) Off-Site Dose Calculation Manual Section 8 Rev.36 Page 13of17 Vermont Yankee Nuclear Power Station

Substituting the right half of Equation 8- I 4 into Equation 8-13 for Cmi yields: (8-15) R = L: F (cpm) ( cpm-cm . 3 J(~tCi)

                                    -       ( -sec )
                      ~LC1           sec      cm  3 The detector calibr~~ion procedure establishes a counting efficiency for a re t.crence rac1*1onuc 1*1cIe, K r ) .

8 The skin dose rate due to noble gases is determined with Equation 6-7: (6-7) llskin =L I

                            ~tCi) ( mrc~ - sec).

( sec ~tC1 -yr Where: Rskin Skin dose rate (mrem/yr) Qi The release rate of Kr85 identified (µCi/sec) equivalent to Q;T for noble gases released at the plant stack). DF'is Combined skin dose factor for Kr85 of 2.15E - 03 (see Table 1.1.10) (mrem-sec/µCi-yr). Off-Site Dose Calculation Manual Section 8 Rev.36 Page 14of17 Vermont Yankee Nuclear Power Station

Control 3.3.1.a limits the dose rate to the skin from noble gases at any location at or beyond the site boundary to 3,000 mrem/yr. By setting Rskin equal to 3,000 mrcm/yr and substituting OF~ for OF[ in Equation 6- 7 one may solve for :L Qi at the I limiting skin noble gas dose rate: (}; = 3,000 DF'c (~~) mrem ( yr l( JtCi - yr ) mrem - sec Substituting this result for Q; in Equation 8-16 yields R skin spt

the response of the monit0r at thcJimiting noble gas skir. dose r:ite:

(8-10) R skin spt

             = 3,000 F         OF~

(cpm) ( mreml ( cpm-

yr

                                 ~m
                              ~tC1 3

l( sec ) ( µCi - yr ) cm 3 mrem - sec Off-Site Dose Calculation Manual Section 8 Rev.36 Page 15of17 Vermont Yankee Nuclear Power Station

TABLE 8.2.1 Relative Fractions of Core Inventorv Noble Gas.es After Shutdown Time Kr-83m Kr-85m Kr-85 Kr-87 Kr-88 Xe-131111 Xe-133111 Xe-133 Xe-135m Xe-35 Xe-138 t <24 h .02 .043 .001 .083 .118 .002 .010 .306 .061 .093 .263 24 hr ::=;t <48 h .003 .004 .001 .004 .022 .758 .010 .198 48 h ::=;t <5 d .005 .006 .024 .907 .001 .058 5 d ::=;t <IO d .007 .008 .016 .969 10 d ::=;t <15 d .014 .014 .006 .966 15 d::=;t <20 d .026 .022 .002 .950 20 d::=;t <30 d .048 .034 .001 .917 30 d ::=;t <40 d .152 .070 .777 40 d::=;t <50 d .378 .105 .517 50 d::=;t <60 d .652 .108 .240 60 d:::; t <70 d .835 .083 .082 t 2::70 d .920 .055 .024 Off-Site Dose Calculation Manual Section 8 Rev. 36 Page 16of17 Vermont Yankee Nuclear Power Station

I 8.2.2 Deleted Off-Site Dose Calculation Manual Section 8 Rev.36 Page 17of17 Vermont Yankee Nuclear Power Station L _ _ __ _ _ _ _ _

9.0 LIQUID AND GASEOUS EFFLUENT STREAMS, RADIATION MONITORS AND RADWASTE TREATMENT SYSTEMS Figure 9-1 shows the normal (design) radioactive liquid effluent streams, radiation monitors, and the appropriate Liquid Radwaste Treatment System. Figure 9-2 shows the normal (design) gaseous ef!luent systems, radiation monitors, and the appropriate Gaseous Raclwaste Treatment System. Figure 9-3 shows the normal subsurface shallow groundwater stream tube configuration for the plant site. Figure 9-4 shows the no1mal subsurface deep groundwater stream tube configuration for the plant site. 9.1 In-Plant Radioactive Liquid Efl1uent Pathways The Liquid Rad waste System collects, processes, stores, and disposes of all radioactive liquid wastes. Except for the cleanup phase separator equipment, the condensate backwash receiving tank and pump and waste sample tanks, lloor drain sample tank and waste surge tank, the entire Radwaste System is located in the Radwaste Building. The Radwaste System is controlled from a panel in the Radwaste Building Control Room. The Liquid Rad waste System consists of the following components:

1. Floor and equipment drain system for handling potentially radioactive wastes.

2. Tanks, piping, pumps, process equipment, instrumentation and auxiliaries necessary to collect, process, store, and dispose of potentially radioactive wastes.

The liquid radwastes are classified, collected, and treated as either high purity, low purity, chemical or detergent wastes. "High" purity and "low" purity mean that the wastes have low conductivity and high conductivity, respectively. The purity designation is not a measure of the amount of radioactivity in the wastes. High purity liquid wastes are collected in the 25,000-gallon waste collector tank. They originate from the following sources:

1. Drywell equipment drains.

2. Reactor Building equipment drains.

3. Radwaste Building equipment drains.

4. Turbine Building equipment drains.

5. Decanted liquids from cleanup phase separators.

6. Decanted liquids from condensate phase separators.

7. Resin rinse.

Off-Site Dose Calculation Manual Section 9 Rev. 36 Page 1 of 12 Vermont Yankee Nuclear Power Station

Low purity liquid wastes are collected in the 25,000-gallon tloor drain collector tank. They originate from the following sources: I. Drywell floor drains.

2. Reactor Building floor drains.

3. Radwaste Building floor drains.

4. Turbine Building floor drains.

5. Other floor drains in RCA (e.g., AOG and Service Building, stack, etc.).

Chemical wastes are collected in the 4,000-gallon chemical waste tank and then pumped to the floor drain collector tank. Chemical wastes arise from the chemical laboratory sinks, the laboratory drains and samplc sinks. Radioactive decontamination solutions are classified as detergent waste anti collected in the 1,000-gallon detergent waste tank. Once the wastes are collected in their respective waste tanks, they are processed in the most efficient manner and discharged or reused in the nuclear system. From the waste collector tank, the high purity wastes are processed in one of three alternative filter demineralizers and then, if needed, in one "polishing" demineralizer. After processing; the liquid is pumped to a waste sample tank for testing and then recycled for additional processing, transferred to the condensate storage tank for reuse in the nuclear system or discharged. The low purity liquid wastes are normally processed through the floor drain filter demineralizer and collected in the floor drain sample tank for discharge or they are combined with high purity wastes and prvcessea as high purity wastes. Chemical wastes are neutralized and combined with low purity wastes for processing as low purity wastes. Although there is only one approved discharge pathway from the Radwaste System to the river, there are three locations within the Radwaste System from which releases can be made. They are: the detergent waste tank (detergent wastes), the floor drain sample tank (chemical and low purity wastes), and waste sample tank (high purity wastes). The contents of any of these tanks can be released directly to the river. The liquid wastes collected in the tanks are handled on a batch basis. The tanks are sampled from the radwaste sample sink and the contents analyzed for radioactivity and water purity. A release is allowed once it is determined that the activity in the liquid wastes will not exceed Control release limits. Off-Site Dose Calculation Manual Section 9 Rev.36 Page 2of12 Vermont Yankee Nuclear Power Station

A discharge from any of the tanks is accomplished by first starting the sample pumps, opening the necessary valves, and positioning the now controller. The release rate in the discharge line. is set between 0 and 50 gpm. The dilution pumps which supply 20,000 gpm of dilution water are then started. An interlock docs not allow discharge to the river when dilution water is unavailable. The effluent monitor (No. 17/350) in the discharge line provides an additional check during the release. The alann or trip setpoint on the monitor is set according to the effluent Control limits and an analysis of the contents of the tank. The monitor warns the operator if the activity of the liquid waste approaches regulatory limits. In response to a warning signal from the monitor, the operator may reduce the now rate or stop the discharge. Off-Site Dose Calculation Manual Section 9 Rev. 36 Page 3of12 Vermont Yankee Nuclear Power Station

9.2 ln-Plant Radioactive Gaseous Efllucnt Pathways The gaseous radwastc system includes subsystems that dispose of gases from the station ventilation exhausts. The processed gases are routed to the plant stack for dilution and elevated release to the atmosphere. The plant stack provides an elevated release point for the release of waste gases. Stack drainage is routed to the liquid radwaste collection system through loop seals. Particulate (HEPA) filters with flame suppressant prefilters are located at the exit side of the delay pipe ahead of the moisture removal subsystem to remove radioactive particulates generated in the delay pipe. 9.3 Subsurface Groundwater Pathways to the Connecticut River 9.3. l Overview As presented in the Hydrogeologic Investigation Report prepared by GZA GeoEnvironmental, Inc. (See ODCM Reference Section) the overall direction of groundwater flow at the Vermont Yankee plant site is towards the Connecticut River. Based on this understanding of site hydro geologic conditions, the groundwater discharge rates from the developed portion of the site to tlh~ fr:cr, has been estimated using a st~am tub'.:: approach basl:!J on Darcy's i.,aw. 9.3.2 Geology The site geology consists of a discontinuous layer of engineered fill material underlain by glaciolacustrine/glaciofluvial deposits. The total overburden thickness varies from approximately 30 to 80 feet. The upper-most soil deposit is fill, generally consisting of silty sand. Directly underlying the fill is an upper sand unit which consists of fine to medium sand with various amounts of silt. This unit is underlain by a confining silt layer which appears fairly continuous in areas east of the AOG Building (approximately 1to16 feet thick) but pinches out towards the north. Groundwater flux towards the river through this silt unit is likely to be negligible due to its low permeability. Consequently, the silt layer was not included in the groundwater flux calculation presented herein. Below the silt layer and in the vicinity of the Connecticut River, is a finer-grained lower sand unit. This lower sand deposit appears to be of limited lateral extent, extending from approximately the intake structure southerly to the discharge structure. Off-Site Dose Calculation Manual Section 9 Rev. 36 Page 4of12 Vermont Yankee Nuclear Power Station

The bedrock below the overburden is reported to be biotitc gneiss. Pre-construction seismic testing or the bedrock indicates that it is generally hard and massive. The depth or bedrock in the release area, cast or the AOG building, is approximately 55 to 80 rcct below the plant grade or approximately 251.5 rt NGVD, with bedrock generally sloping towards the river. Groundwater movement through the bedrock in the vicinity of the river appears to be in an upward direction, toward the river. 9.3.3 Streamtubc Method To estimate the groundwater flow through the developed area of the site containing systems, stmctures and components (SSCs) which may carry radionuclides, a series of twelve stream tubes in the shallow overburden and five streamtubes in the deep overburden were delineated. The streamtubes show groundwater flow direction and were drawn perpendicular to the groundwater contours, which were developed from synoptic groundwater elevation data collected on December 15, 20 I 0. The boundaries of the stream tubes represent groundwater tlow lines, with groundwater flow generally parallel to these lines within the streamtubes. Based on frequent groundwater level measurements recorded in the monitoring wells at the site, the overall groundwater flow pattern and hydraulic gradients do not vary significantly at the site, and thus the location and shape of the streamtubes arc anticipated to be relatively stable. The location and configuration of these streamtubes are shown on Figures 9-3 and 9-4. In the shallow overburden, stream-tubes were generally centered on individual downgradient monitoring wells (perimeter wells) extending across the site from the off-gas stack (well GZ-27) southerly to approximately the area of the discharge structure (well GZ-5). In the deep overburden, streamtubes extend over the lateral extent of the lower sand deposit (from well GZ-180 southerly to well GZ-190), and were centered on the deep downgradient wells. 9 .3 .4 Cross Sectional Flow Area The cross sectional area (A) of flow for each streamtube is based on the streamtube width and the thicknesses of the saturated transmissive overburden units within the streamtube. For the shallow overburden streamtubes, the saturated thickness is estimated based on the difference between the measured groundwater table elevation and the elevation of the bottom of the upper sand unit depending on the specific geological conditions present within the screened portion of the well. If a significant thickness of saturated fill is present, the cross sectional area must also include the thickness of any saturated upper sand present under the fill. For the deep overburden stream-tubes, the flow area was computed using the thickness of the fully saturated lower sand unit. Off-Site Dose Calculation Manual Section 9 Rev.36 Page 5of12 Vermont Yankee Nuclear Power Station

9.3.5 Hydraulic Conductivity As presented in the GZA's May 2011 Hydrogeologic Investigation report, hydraulic conductivity (K) testing was perfonned in monitoring wells at the Site to characterize the hydrogeologic properties of the overburden soils. Based on this data, the geometric mean K for each major hydrogeologic unit is presented in Table 5.2 of the GZA report. These values, combined with other input parameters presented herein, are used to calculate groundwater flow rates through each stream tube. A summary of the geometric mean K values for each major hydrogeologic unit is presented below. Hydrogcologic Unit Geometric Mean K (ft/datl ll========-*~==~~=====--~=======~====-T**=~==========,===:~=~==~1 Shallow Fill 1--~~~~~~~~~~~--1--~~~~~~~~~1 0.3 Overburden Upper Sand (East of Power 4 Block) Upper sand (North of Power 12 Block) Deep Lower Sand 1.2 Overburden Off-Site Dose Calculation Manual Section 9 Rev. 36 Page 6of12 Vermont Yankee Nuclear Power Station

9.3.6 Streamtube Parameters I\ summary of selected stream tube parameters used to calculate groundwater flow rates (as discussed further below) is presented in the following table: Key Well within Approximate Ilydrogeologic Unit(s) Streamtube II Stream tube Width (ft) Present within Streamtube K (ft/day) Shallow Overburden Strcamtubes Upper Sand - (North of ST-IS GZ-27S 76 12 Power Block) Upper Sand - (North of ST-2S GZ-26S 215 12 Power Block) Upper Sand - (North of ST-3S GZ-25S 260 12 Power Block) Upper Sand - (North of ST-4S GZ-IS 200 12 Power Block) ST-5S GZ-18S 160 Fill 0.3 ST-6S rill 0.3 GZ-13S 50 Upper Sand - (East of 4 Power Block) ST-7S Fill 0.3 GZ-23S 85 Upper Sand -(East of 4 Power Block) ST-8S Fill 0.3 GZ-3S 85 Upper Sand - (East of 4 Power Block) ST-9S Fill 0.3 GZ-14S 1U5 Upper Sarni - (East of ' 4 Power Block) ST-lOS Fill 0.3 GZ-4S 90 Upper Sand - (East of 4 Power Block) ST-1 lS Fill 0.3 GZ-19S 95 Upper Sand - (East of 4 Power Block) ST-l2S Fill 0.3 GZ-5S 125 Upper Sand - (East of 4 Power Block) Deep Overburden Streamtubes ST-10 GZ-180 75 Lower Sand l.2 ST-20 GZ-130 100 Lower Sand l.2 ST-30 GZ-220 110 Lower Sand l.2 ST-40 GZ-140 110 Lower Sand l.2 ST-5D GZ-190 150 Lower Sand l.2 Off-Site Dose Calculation Manual Section 9 Rev.36 Page 7of12 Vermont Yankee Nuclear Power Station

9.3.6 Hydraulic Gradient For each streamtube, average hydraulic gradients (i) are calculated based on quarterly groundwater elevation contours developed from quarterly groundwater level measurements in the monitoring wells. 9.3.7 Calculated Groundwater Flow Rate Based on the above parameter estimates, the quarterly groundwater flow in each streamtube is computed using the following fonn of the Darcy's equation for fluid flow through porous media: Q=KiA Where: Q =flow (ft3/day) K =hydraulic conductivity (ft/day)

               =hydraulic gradient (ft/ft)

A = cross-sectional area of flow (width times depth) ( ft 2) The quarterly groundwater flow rates are utilized to determine monthly estimates of flow rate for the dose calculations required in Section 6 of the ODCM. The quarterly groundwater flow rates are also averaged over the reporting year to account for variations in seasonal precipitation, and thus the associated seasonal variation groundwater flux to estimate annual radiation doses. To conservatively account for potential uncertainties and the heterogeneity inherent in geologic materials such as those at the plant site, a factor of safety may be applied to the estimated flow to provide a more conservative estimate for the groundwater flow through the subsurface. Off-Site Dose Calculation Manual Section 9 Rev. 36 Page 8of12 Vermont Yankee Nuclear Power Station

( *11~nuL~.1! l.:..b \\".1st.::. l.Jh Dram.; S;mtpl.: SinJ,." Sµ.:m ... \."fl,1..:n-....zh.* Ch:mi1.-.1l D.:l~rgenr 0 Pcol1~lUtl? }L*~ul....

                                  \\":i~tl!                     \\ &31~

Tanl: T;onl. D.:1.:17.:n1 l.(j()QpiJ 1.000 p*I \\'.i.,1.: F1lr ~r Fl*'~ Droiin Dr.1111

                                           - - - - - - - - - - ...-\.illl!\.'lh.l!l FJ. .., ... lJl.,,in                                                  s.i.r:1),'{:!

Tank F1llo!r n..:nun T.111k

                                                                          ~~.000         g;d                                 1):\      sq fl"!                                                  10.000 fJI S.*ihi '.\*..1 .. 1.:

[.~,..,* PLUll\ \\'a..;t~~ S~111 R.:-4:1 D1)wdl l:q1upnll!nl Dram:<1. D,:\\,11.:r.-d111 R..:.1ch."f' Jhuldinp. 11L""r Di:un.. Di~~, ...... N,;.: Ri1flwr1sh: Bmll~ng fh,., .... Dr.uns l.m..:r .. Turhinl! Bu.ildifl!? F1'-xv Dram.~ H1ch Pnritv\\".i:-.t1..":\*. Wu~.:

                                                                             <.. \.~Jir;!l.~tl'f                                                                                                                                                                 Torus Di: w~IJ E<11upm.::n1 DI ain...                                                                                                               \\ *J~I.!  t>-.:rm:i R~aCll'r Build!.~ F.qmpnk!nl            Drain.;;              Tanh                                                                                  iJ5    ~*al1i(    ft1                                                                                    P.**n..*~. /ru.*

Rad1w1~1i: Rrnldmg fa1u.ip111i:nt Drai1L.. :!:i.OOOS?.J ~\:->t.:m F"r T1ub!M RuiJchng Eq1upmc:n1 Drams 1,300.000 . R.:u...: CL'ttden.~I.? Pl1o1;..: &:p.imh.'1 Cl.:anup PhaS.? Sepuiall'I' gal R.::i.111Kr.n.,..,:

                                                                                 \\".1 .. 1.:                                      \\"JAf~     l"l-1.l.::1.*ll ..fi Surei!                                                Filtl!rf~mm T.:s1~:           llOgpm                                1):-' ""I tt1 3\000g.1.l S..:r;1..:.! \\"..ii~

H..:JI Exd'L.lm::.:r 01-h.tH?.: \J.._-..rJt...'f 10.000 jfl>m *~\I~! --~'I: GW Intercept Tank C1rL"t11o1ring\\".:st.:r 10,000 gal

                                                                                   )o0.000 g:p11      l~'t,1J I) JllDllpsJ..__ _ ___.

Tnt*1k.! Stnil' nm~ Dliuot'n \\'111~r

                                                                                      .:!O.OOOJ?rm

[).~ Pt~u....;

                                                                                                                                                                                                ~knu 1R\l-J-.3;i1, Rfri.:r \\hl~r   lntak.:                                                                                                                                                                                                100 gpm Figure 9.1: Radioactive Liquid Effluent Streams. Radiation Monitors, and Rad waste Treatment System at Vermont Yankee*

Normal (design) radioactive process streams only are shown.

Off-Site Dose Calculation Manual Section 9 Rev. 36 Page 9of12

'.\lermont-Yankee.J~Jiic.lead~.o_wer.Statio.n _____________________________________________

Release to Atmosphere Tritium Sampler Particulate Sampler Plant Stack Noble Gas Activity ,n, Monitors (RM-17-156, RM-17-157) i.....---11wLJ Building Ventilation (Including Turbine Building) -..__, Standby Gas Treatment (Containment Purgei--- L-l~I- Plant Stack Figure 9.2: Radioactive Gaseous Effluent Streams. Radiation Monitors, and Radwaste Treatment System at Vermont Yankee*

Normal (design) radioactive process streams only are shown.

Off-Site Dose Calculation Manual Section 9 Rev.36 Page 10of12 Vermont Yankee Nuclear Power Station

DECEMBER 15, 2010 SHALLOW OVERBURDEN WELLS

     *~  ;,

II

                                                                                                 .. ~-.-lmiWWW PUUIAflll'~~-INU!nmOG

_, Ill' l:la' - I

                                                +11-rn1rrn 111'       11 ENTERGY VERMONT VANl<EE
                                                                                                     ~.VERMONT SHALLOW OVERBURDEN STREAM 1\JBES Off~Site Dose Calculation t-.fanual Section 9 RcY. 36 Page II ofl2 Vermont Yankee Nuclear Power Station

DECEMBER 15, 2010 DEEP OVERBURDEN WELLS

                                    . ~ONNECTICUT RIVER.-
                                                                                         ~ ., .. _;__**'* -** .... - .... -
                                     . : . (VERN~'!f'~ND) . -        :
                   ; ,.*                                        .~.,-.
                                                                                                                                      *~.\    \. __    2. FLUClUATIONS '.NL~ c.R:>uhOtiA 1l:R EU.v.-.T-Ofrr.SMAYOCO..."<O'lf;J( TNt:. :v.:. ro RAlHFAll, SF~ C>tAJwGCS INTI-§= RATE Of
                                                                                                                              .. - ..  '~ ~;*_*c_~*        EVN'Ora.t..~TION. RELEAS!' RATES FROM THE LOCAL CAM. NoO On.ER VAA!OUS FACTORS.
                                                                                                                                   .. H

..*-.* t*

                                                                                                                                           '*          1 THE LOCAJIOPtS NfO R£FERENCE ELEVA.TIO.~ Of MONllOf.tlNCi WE.US NfO OHSIJE WA.TE.ff Sl..f't"LY WEUS\\'FRF.~.:JBYSOIJTl-IEAJ\IVERMC>"4T

__ ..3 EfiGINf.F.RNO OF BRA.ffi EBC'fiO VERlilONT. L.OCATICWS OF OTl-tER WATER W'P.Y WEUS BASED ON FIELD OOSERVATIOHS. 4.. HASE Pl.AN. "-0..UOlN:l SITE SURJ'ACI::. TOPOGRAPHY'. PRO-.ilnED TO GlA IN Ei..ECTRCfaC FORMAT BY~ YN<<EE.. 12!1 24!1 SCALE IN FEET ENTERGY VERMONT YANKEE VERNON, VERMONT DEEP OVERBURDEN STREAM TUBES Off-Site Dose Calculation .Manual Section 9 Rev. 36 Page 12 of 12 Vermont Yankee Nuclear Power Station

10.0 UNIOUE REPORTING REQUIREMENTS I 0.1 Annual Radioactive Effluent Release Report In accordance with I OCFR 50.36a, the Radioactive Effluent Release Report covering the operation or the unit shall be submitted by May 15 of each year. The Radioactive Effluent Release Report shall inclucle a summary of the quantities of radioactive liquid and gaseous enluents and solid waste released from the unit as outlined in Regulatory Guide 1.21, Revision I, June 1974, "Measuring, Evaluating and Reporting Radioactivity in Solid Wastes and Releases of Radioactive Materials in Liquicl and Gaseous Effluents from Light-Water-Cooled Nuclear Power Plants," with data summarized on a quarterly basis following the fonnat of Appendix B thereof. For solid wastes the format for Table 3 in Appendix B of Regulatory Guicle 1.21 shall be supplemented with three additional categories: class of solid wastes (as defined by l OCFR Part 61 ), type of container (e.g., LSA, Type A, Type 8, Large Quantity), ancl solidification agent or absorbent, if any. In addition, the Radioactive Effluent Release Report shall include an annual summary of hourly meteorological data collected over the previous year. This annual summary may be either in the form of an hour-by-hour listing on magnetic tape of wind speed, wind direction, atmospheric stability, and precipitation (if measured), or in the fonn of joint frequency distributions of wind speed, wind direction, and atmospheric stability.* This same report shall include an assessment of the radiation doses due to the radioactive liquid and gaseous effluents released from the unit during the previous calendar year. The Radioactive Effluent Release Report shall also include an assessment of the radiation doses from radioactive effluents to member(s) of the public due to any allowed recreational activities inside the site boundary during the previous calendar year. Ail assumptions used in making these assessments (e.g., specific activity, exposure time and location) shall be included in these reports. For any batch or discrete gas volume releases, the meteorological conditions concurrent with the time of release of radioactive materials in gaseous effluents (as determined by sampling frequency and measurement) shall be used for determining the gaseous pathway doses. For radioactive materials released in continuous effluent streams, quarterly average meteorological conditions concurrent with the quarterly release period shall be used for detennining the gaseous pathway doses. The assessment of radiation doses shall be performed in accordance with the Off-Site Dose Calculation Manual (ODCM).

   'In lieu of submission with the Radioactive Effluent Release Report, the licensee has the options of retaining this summary of required meteorological data in a file that shall be provided to the NRC upon request.

Off-Site Dose Calculation Manual Section 10 Rev. 36 Page 1 of7 Vermont Yankee Nuclear Power Station

                                                                                                                -~-------~-

With the limits of Control 3.4.1 being exceeded during the calendar year, the Radioactive Effluent Release Report shall also include an assessment of radiation doses to the likely most exposed real member(s) of the public from reactor releases (including doses from primary enluent pathways and direct radiation) for the previous calendar year to show conformance with 40CFR 190, Environmental Radiation Protection Standards for Nuclear Power Operation. The Radioactive Effluent Release Report shall include a list and description of unplanned releases from the site to site boundary of radioactive materials in gaseous and liquid effluents made during the reporting period. With the quantity of radioactive material in any outside tank exceeding the limit of Technical Specification 3.1.A. I, describe the events leading to this condition in the next Radioactive Effluent Release Report. If non-fi.1cntirrna! r~~dioact!\*c liquid dl1ucn! monitoring instnum:ntati:m is nul returned to fi.mctional status prior to the next release pursuant to Note 4 of Control Table 3.1.1, explain in the next Radioactive Effluent Report the reason(s) for delay in correcting the inoperability. If non-fi.mctional gaseous effluent monitoring instrumentation is not returned to functional status within 30 days pursuant to Note 5 of Control Table 3 .1.2, explain in the next Radioactive Effluent Release Report the reason(s) for delay in correcting the inoperability. With a land use census identifying one or more locations which yield at least a 20 percent greater dose or dose commitment than the values currently being calculated in Control 4.3.3, identify the new location(s) in the next Radioactive Effluent Release Rcpcit. Changes made during the reporting period to the Process Control Program (PCP) and to the Off-Site Dose Calculation Manual (ODCM), shall be identified in the next Radioactive Effluent Release Report. Off-Site Dose Calculation Manual Section IO Rev. 36 Page 2 of7 Vermont Yankee Nuclear Power Station

I 0.2 Environmental Radiological Monitoring The Annual Radiological Environmental Operating Report covering the operation of the unit during previous calendar year shall be submitted by May 15th of each year. The report shall include summaries, interpretations, and an analysis of trends of the results of the radiological environmental surveillance activities for the report period. The material provided shall be consistent with the objectives outlined in the ODCM and in 10CFR 50, Appendix I, Sections IV.B.2, IV.B.3, and IV.C. The Annual Radiological Environmental Operating Report shall include summarized and tabulated results of all radiological environmental samples taken during the report period pursuant to Table 7-1 and Figures 7-1 through 7-6. In the event that some results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results. The missing data shall be submitted as soon as possible in a supplementary report With the level of radioactivity in an environmental sampling media at one or more of the locations specified in Control Table 3.5.1 exceeding the reporting levels of Control Table 3.5.2, the condition shall be described in the next Annual Radiological Environmental Operating Report only if the measured level of radioactivity was not the result of plant effluents. With the radiological environmental monitoring program not being conducted as specified in Control Table 3.5.1, a description of the reasons for not conducting the program as required and the plans for preventing a recurrence shall be included in the next Annual Radiological Environmental Operating Report. The Annual Radiological Environmental Operating Report shall also include the results or the lanJ use census required by Co11troi 3.5.2. A summary description of che radiological environmental monitoring program including a map of all sampling locations keyed to a table giving distances and directions from the reactor shall be in the reports. If new environmental sampling locations are identified in accordance with Control 3.5.2, the new locations shall be identified in the next Annual Radiological Environmental Operating Report. The reports shall also include a discussion of all analyses in which the LLD required by Control Table 4.5.1 was not achievable. The results of license participation in the intercomparison program required by Control 3.5.3 shall be included in the reports. With analyses not being performed as required by Control 3.5.3, the corrective actions taken to prevent a recurrence shall be reported to the Commission in the next Annual Radiological Environmental Operating Report. Off-Site Dose Calculation Manual Section 10 Rev. 36 Page 3 of7 Vermont Yankee Nuclear Power Station

10.3 ISFSI Reporting Requirements In accordance with IOCFR72.44(d)(3), the Annual Independent (Interim) Spent Fuel Storage Installation Radiactivc Effluent Control Program Report (AISFSIRECPR) will be generated and issued by February 28th of each year. Since it has been cletennincd by Holtec International in their Final Safety Analysis Report (Reference I) that the Holtec HI-STORM 100 Cask System does not create any radioactive materials or have any radioactive waste treatment systems, specific operating procedures for the control of radioactive effluents are not required. Specification 3.1.1, Multi-Purpose Canister (MPC), provides assurance that there are no radioactive effluents from the SFSC. In light of the information presented in the previous paragraphs, the AISFSIRECPR, to be issued by February 28th of each year, shall state that no rndioacti\*e erl1uents were discharged from the lndqh:mlcnt (InLCrim) Spent Fud Storage Installation and therefore no ISFSl-specific monitoring program is in place at Vermont Yankee and there are no lSFSl-specific data to report for the previous calendar year reporting period. 10.4 Special Reports Special reports shall be submitted to the Director of the Office of Inspection and Enforcement Regional Office within the time period specified for each report. 10.4.1 Liquid Effluents (Controls 3.2.2 and 3.2.3) With the calculated dose from the release of radioactive materials in liquid efnuents exceeding any of the limits of Control 3.2.2, prepare and subnut to the Commission within 30 days a special report which identifies the cause(s) for exceeding the limit(s) and defines the corrective actions taken to assure that subsequent releases will be in compliance with the limits of Control 3.2.2. With liquid radwaste being discharged without processing through appropriate treatment systems and estimated doses in excess of Control 3.2.3, prepare and submit to the Commission within 30 days a special report which includes the following information: ( 1) explanation of why liquid rad waste was being discharged without treatment, identification of any non-functional equipment or subsystems, and the reasons for the non-functionality; (2) action(s) taken to restore the non-functional equipment to functional status; and (3) summary description of action( s) taken to prevent a recurrence. Off-Site Dose Calculation Manual Section 10 Rev.36 Page 4 of7 Vermont Yankee Nuclear Power Station

I 0.4.2 Gaseous Effluents (Controls 3.3.2, 3.3.3, 3.3.4 and 3.3.5) With the calculated air dose from radioactive noble gases in gaseous efl1uents exceeding any of the limits of Control 3.3.2, prepare and submit to the Commission within 30 days a special report which identities the cause(s) for exceeding the limit(s) and the corrective action(s) taken to assure that subsequent releases will be in compliance with the limits of Control 3.3.2. With the calculated dose from the release of tritium and/or radionuclides in particulate form exceeding any of the limits of Control 3.3.3, prepare and submit to the Commission within 30 days a special report which identifies the causc(s) for exceeding the limit(s) and the corrective action(s) taken to assure that subsequent releases will be in compliance with the limits of Control 3.3.3. With gaseous radwaste being discharged without processing through appropriate treatment systems as defined in Control 3.3.4 for more than seven (7) consecutive days, or in excess of the limits of Control 3.3.5, prepare and submit to the Commission within 30 days a special report which includes the following information: (I) explanation of why gaseous rad waste was being discharged without treatment (Control 3.3.4), or with resultant doses in excess of Control 3.3.5, identification of any non-functional equipment or subsystems, and the reasons for the non-functionality; (2) action(s) taken to restore the non-functional equipment to functional status; and (3) summary description of action(s) taken to prevent a recurrence. 10.4.3 Total Dose (Comrol 3.4.1) With the calculated dose from the release of radioactive materials in liquid or gaseous effluents exceeding the limits of Control 3 .4.1, prepare and submit to the Commission within 30 days a special report which defines the corrective action(s) to be taken to reduce subsequent releases to prevent recurrence of exceeding the limits of Control 3.4. l and includes the schedule for achieving conformance with these limits. This special report, required by 10CFR Part 20.2203(a)(4), shall include an analysis that estimates the radiation exposure (dose) to a member of the public from station sources, including all effluent pathways and direct radiation, for the calendar year that includes the release(s) covered by this report. It shall also describe levels of radiation and concentrations of radioactive material involved, and the cause of the exposure levels or concentrations. If the estimated doses exceed any of the limits of Control 3 .4.1, and if the release condition resulting in violation of 40CFR Part 190 has not already been corrected, the special report shall include a request for a variance in accordance with the provisions of 40CFR Part 190. Submittal of the report is considered a timely request, and a variance is granted until staff action on the request is complete. Off-Site Dose Calculation Manual Section 10 Rev. 36 Page 5 of7 Vermont Yankee Nuclear Power Station

I 0.4.4 Radiological Environmental Monitoring {Control 3.5.1) With the level of radioactivity as the result of plant effluents in an environmental sampling media at one or more of the locations specified in Control Table 3.5.1 exceeding the reporting levels of Control Table 3.5.2, prepare and submit to the Commission within 30 days from the receipt of the Laboratory Analyses a special report which includes an evaluation of any release conditions, environmental factors or other factors which caused the limits of Control Table 3.5.2 to be exceeded. This report is not required if the measured level of radioactivity was not the result of plant effluents, however, in such an event, the condition shall be reported and described in the Annual Radiological Environmental Operating Report. I 0.4.5 Land Use Census {Control 3.5.2) With a land use census not being conducted as required by Control 3.5.2, prepare ant! submit to the Comnfr:sion within 30 days a special rcpo11 which ic.kntifics the reasons why the survey was not conducted, and what steps are being taken to correct the situation. I 0.5 Major Changes to Radioactive Liquid, Gaseous, and Solid Waste Treatment Systems** Licensee-initiated major changes to the radioactive waste systems (liquid, gaseous, and solid): A. Shall be reported to the commission in the Radioactive Effluent Release Report for the period in which the evaluation was reviewed by the independent Safety Review (ISR). The discussion of each change shall contain: I. A summary of the evaluation that led to the determination that the change could be made in accordance with IOCFR Part 50.59;

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

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

4. An evaluation of the change, which shows the predicted releases of radioactive materials in liquid and gaseous effluents and/or quantity of solid waste that differ from those previously predicted in the license application and amendments thereto;

**Licensee may choose to submit the information called for in this reporting requirement as part of the annual FSAR update.

Off-Site Dose Calculation Manual Section 10 Rev.36 Page 6 of7 Vermont Yankee Nuclear Power Station

5. An evaluation of the change, which shows the expected maximum exposures to member(s) of the public at the site boundary and to the general population that differ from those previously estimated in the license application and amendments thereto;

6. J\ comparison of the predicted releases of radioactive materials, in liquid and gaseous effluents and in solid waste, to the actual releases for the period prior to when the changes are to be made;

7. An estimate of the exposure to plant operating personnel as a result of the change; and

8. Documentation of the fact that the change was reviewed and found acceptable by JSR.

8. Shall become effective upon review and acceptance by JSR and approval by the Senior Manager, Production.

Off-Site Dose Calculation Manual Section 10 Rev.36 Page 7 of7 Vermont Yankee Nuclear Power Station

REFERENCES A. Regulatory Guide I. I 09, "Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with I OCFR50, Appendix I," U.S. Nuclear Regulatory Commission, Revision I, October 1977. B. Hamawi, J. N., "AEOLUS A Computer Code for the Detennination of Continuous and Intermittent-Release Atmospheric Dispersion and Deposition of Nuclear Power Plant Effluents in Open-terrain Sites, Coastal Sites, and Deep-River Valleys for Assessment of Ensuing doses and Finite-Cloud Gamma Radiation Exposures," Entech Engineering, Inc., PIOOR13A, March 1988 (Mod 5, Revised by Yankee Atomic Electric Company, March 1992).

c. Regulatory Guide 1.111, "Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water Cooled Reactors," U.S. Nucknr Reguhi.tor:; CC' tr.mission, Rev. I, July 1977.

D. National Bureau of Standards, "Maximum Pem1issible Body Burdens and Maximum Permissible Concentrations of Raclionuclides in Air and in Water for Occupational Exposure," Handbook 69, June 5, 1959. E. Slade, D. H., "Meteorology and Atomic Energy - 1968, USAEC, July 1968. F. Lowder, W. M., P. D. Raft, and G. clePlanque Burke, "Determination of N-16 Gamma Radiation Fields at BWR Nuclear Power Stations," Health and Safety Laboratory, Energy Research and Development Administration, Report No. 305, May 1976. G. Letter from Charles L. Miller of the United States Nuclear Regulatory Commission to John F. Schmidt of the Nuclear Energy Institute, elated December 26, 1995. H. "Dose vs Distance at the Vermont Yankee ISFSI" Holtec Report No: HI-2073701, Holtec Project No: 90348, Approved on 12/10/07. I. Certificate of Compliance, Holtec International, Final Safety Analysis Report, Amendment 2, 06/07/05. J. AREVA Calculation #32-9068362-000 "Vermont Yankee Site Boundary Direct Dose From N-16 Methodology", April 4, 2008, Mark Strum and John Hamawi. K. "Hydrogeologic Investigation of Tritium in Groundwater, Vermont Yankee Nuclear Power Station, Vernon, Vermont" File No. 09.0025576.11, dated May, 2011. Prepared for Entergy Vermont Yankee by GZA, GeoEnvironmental, Inc. One Edgewater Drive, Norwood MA 02062 "Vermont Yankee Shut-Down Environmental Radionuclides of Concern and Off-Site Dose Calculation Manual Changes", RSCS TSO No. 16-041 Off-Site Dose Calculation Manual Section R Rev.34 Page I of I Vermont Yankee Nuclear Power Station

APPENDIX B Approval of Criteria for Disposal of Slightly Contaminated Septic Waste On-Site at Vermont Yankee Revision ~9~ Date 3/2/90 B-1

~ -- RECEIVED UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON. D. C. 20555 SfP-7t:m !i August 30. 1989 VERM0WT YANl<EE LICENSING Mr. L. A. Tremblay Licensing Engineering Vermont Yankee Nuclear Power Corporation Engineering Office 580 Main Street Bolton, Massachusetts 01740-1398

Dear Mr. Tremblay:

SUBJECT:

APPROVAL UNDER 10 CFR 20.302(a) OF PROCEDURES FOR DISPOSAL OF SLIGHTLY CONTAMINATED SEPTIC WASTE ON SITE AT VERMONT YANKEE (TAC NO. 73776)

REFERENCE:

(a) June 28, 1989 letter from R. W. Capstick to US NRC Document Control Desk, including Attachment I and Attachment II. ( b) Final Environmental Statement related to the operation of Vermont Yankee Nuclear Power Station, dated July 1972. In reference (a) ~ermont Yankee Nuclear Power Corporation (Vermont Yankee. or the licensee) submitted an application for disposal of licensed material on site. This disposal was not previously considered by the staff in the Vermont Yankee Final Environmental Statement CFES), reference (b). This extensive application, prepared in accordance with 10 CFR 20.302(a), contains a detailed description of the licensed material, thoroughly analyzes and evaluates the information pertinent to the effects on the environment of the proposed disposal of the licensed material. and commits the licensee to follow specific procedures to minimize the risk of unexpected 9r hazardous exposures. In the FES, the NRC- staff considered the potential effects on the environment of licensed material from operation of the plant and. in the assessment of the total radiological impact of the Vermont Yankee Station concluded that:

     " ... operation of the Station will contribute only an extremely small increment to the radiation dose that area residents receive from natural background.

Since fluctuations of the background dose may be expected to exceed the increment contributed by the plant, the dose will be immeasurable in itself and will constitute no meaningful risk to be balanced against the benefits of

    ;the plant."

Revision ~9~ Date 312190 B-2

Mr. L. A. Tremblay August 30, 1989 Since the disposal proposed by the licensee involves licensed material containing less than 0.1 percent of the radioactive materials, primarily cobalt-60 and cesium-137, already considered acceptable in the FES. and involves exposure pathways much less significant than those considered in the FES, we consider the site-specific application (Reference (a)) for Vermont Yankee Nuclear Power Station to have insignificant radiological impact. We accept the commitments and evaluations of the licensee, documented in reference (a). as further assurance th_i,!t the proposed disposal procedures wi 11 have a negligible effect on the environment and the general population in comparison to normal background radiation. LAI In conclusion. we find the licensee's procedures with commitments as 855 documented in reference (a) to be acceptable. provided that reference (a) is permanently incorporated into the licensee's Offsite Dose Calculation Manual CODCM) as an Appendix. and future modifications of reference (a) be reported to NRC in accordance with licensee commitments regarding ODCM changes. Pursuant to 10 CFR 51.22(c)(9). no environmental assessment is required. This completes our review under TAC No.73776. Sincerely, Morton B. Fairtile. Project Manager Project Directorate I-3 Division of Reactor Projects I/II Office of Nuclear Reactor Regulation cc: See next page Revision ~9~ Date 3/2/90 B-3

Mr. L. A. Tremblay cc: Mr. J. Gary Weigand G. Dean Weyman President & Chief Executive Officer Chairman, Board of Selectman Vermont Yankee Nuclear Power Corp. Post Office Box 116 R.D. 5, Box 169 Vernon, Vermont 05354 Ferry Road Brattleboro, Vermont 05301 Mr. Raymond N. McCandless Vermont Division of Occupational Mr. John DeVincentis, Vice President and Radiological Health Yankee Atomic Electric Company Administration Building 580 Main Street *Montpelier, Vermont 05602 Bolton, Massachusetts 01740-1398 Honorable John J. Easton New England Coalition on Nuclear Attorney General Pollution State of Vermont Hill and Dale Farm 109 State Street . R.D. 2, Box 223 Montpelier, Vermont 05602 Putney, Vermont 05346 Conner & Wetterhahn, P.C. Vermont Public Interest Research Suite 1050 Group, Inc. 1747 Pennsylvania Avenue, N.W. 43 State Street Washington, D.C. 20006 Montpelier, Vermont 05602 Diane Curran, Esq. Regional Administrator, Region I Harmon, Curran & Tousley U.S. Nuclear Regulatory Commission 2001 S Street, N.W., Suite 430 475 Allendale Road Washington, D.C. 20009 King of Prussia. Pennsylvania 19406 David J. Mullett. Esq. R. K. Gad, III Special Assistant Attorney General Ropes 6 Gray Vermont Department of Public Service 225 Franklin Street 120 State Street Boston, Massachusetts 02110 Montpelier, Vermont 05602 Mr. W. P. Murphy, Vice President Jay Gutierrez and Manager of Operations Regional Counsel Vermont Yankee Nuclear Power Corporation U.S. Nucle~r Regulatory Commission R.D. 5, Box 169 475 All~ndale Road Ferry Road King of Prussia, Pennsylvania 19406 Brattleboro, Vermont 05301 G. Dana Bisbee, Esq. Mr. George Sterzinger, Commissioner Office of the Attorney General Vermont Department of Public Service Environmental Protection Bureau 120 State Street, 3rd Floor State House Annex Montpelier, Vermont 05602 25 Capitol Street Concord, New Hampshire 03301-6397 Public Service Board State of Vermont Atomic Safety and Licensing Board 120 State Street U.S. Nuclear Regulatory Commission Montpelier, Vermont 05602 Washington, D.C. 20555 Revision ~9~ Date 312190 B-4

Mr. L. A. Tremblay cc: Mr. Gustave A. Linenberger.Jr. Administrative Judge Atomic Safety and Licensing Board U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Resident Inspector Vermont Yankee Nuclear Power Station U.S. Nuclear Regulatory Commission P.O. Box 176 Vernon, Vermont 05354 John Traficante, Esq. Chief Safety Unit Office of the Attorney General One Ashburton Place, 19th Floor Boston, Massachusetts 02108 Geoffrey M. Huntington, Esquire Office of the Attorney General Environmental Protection Bureau State House Annex 25 Capitol Street Concord. New Hampshire D3301-6397 Charles Bechhoefer. Esq. Administrative Judge Atomic Safety and Licensing Board U.S. Nuclear Regulatory Commission Washington. D.C. 20555 Dr. James H. Carpenter Administrative Judge Atomic Safety and Licensing Board U.S. Nuclear Regulatory Commi~sion Washington, D.C. 20555 Adjudicatory File (2) Atomic Safety and Licensing Board Panel Docket U.S. Nuclear Regulatory Commission Washington, D.C. 20555 (25) Revision ~9~ Date 3/2/90 B-5

V'ERMONT YANKEE N'UCLEAR POWER CORPORATION Ferry Road, Brattleboro, VT 05301-7002 ENGINEERING OFFICE S80 MAIN STAEET June 28, 1989 BOLTON. MA01740 BVY 69-59 (SO$J 77H7 II United States Nuclear Regulatory Commission Washington, DC 20555 Attention: Document Control Desk

Reference:

License No. DPR-28 (Docket No. 50-271).

Subject:

Request to Routinely Dispose of Slightly Contaminated Septic Waste in Accordance with 10 CFR 20.302(a)

Dear Sir:

In accordance with the criteria of the Code of Federal Regulations. Title 10, Section 20.302(a) (10CFR20.302(a)), enclosed please find the subject application for the disposal of very low level radioactive waste materials. Vermont Yankee Nuclear Power Corporation (Vermont Yankee) hereby requests NRC approval of the proposed procedures for the disposal of slightly contaminated septic waste generated at the Vermont Yankee Nuclear Power Plant in Vernon. Vermont. This application specifically requests approval to dispose of septic tank waste, contaminated at minimal levels, which have been or might be generated through the end of station operations at the Vermont Yankee Nuclear Power Plant. The proposed method of disposal is for the on-site land spreading in designated areas in compliance with State of Vermont health code requirements for septic waste. Disposal of this waste in the manner proposed. rather than at a 10 CFR Part 61 licensed facility would save Vermont Yankee not only substantial cost, but also valuable disposal site space which would then be available for wastes of higher radioactivity levels. Disposal as radioactive waste would require treatment of the biological aspects of the septage and solidification to a stable waste form, thereby increasing the volume substantially. A radiological assessment and proposed operational controls, based upon the continued on-site disposal of septic waste as presently contained in the plant's septic tanks, are detailed in Attachments 1 and 2. Based upon this analysis, Vermont Yankee requests approval to dispose of septic tank waste on-site by land spreading in such a manner that the radioactivity concentration limit in any batch of septage to be spread does not exceed one-tenth of the MPC values listed in 10 CFR 20, Appendix B. Table II; and the combined radiological impact for all disposal operations shall be limited to a total body or organ dose of a maximally exposed member of the public of less than one mrem/year (less than 5 mrem/year to an inadvertent intruder). Revision ~9~ Date 3/2/90 B-6

United St~tes Nuclear Regulatory Commission June 28. 1989 Page 2 Due to our expected need to utilize the proposed methodology of land application of septic waste on-site during the spring of 1990. we request your review and approval of this proposed disposal method by the end of the first quarter of 1990. We trust that the information contained in the submittal is sufficient; however, should you have any questions or require further information concerning this matter. please contact this office Very truly yours, VERMONT YANKEE NUCLEAR POWER CORPORATION Robert W. Capstick, Jr. Licensing Engineer MSS/emd Enclosures cc: USNRC - Region I USNRC - Resident Inspector, VTNPS Revision ~9~ Date 3/2/90 B-7

ATTACHMENT 1 VERMONT YANKEE NUCLEAR POWER PLANT APPLICATION FOR APPROVAL TO ROUTINELY DISPOSE OF SE?TIC WASTE WITH MINIMAL LEVELS OF RADIOACTiVITY Revision _9_ Date 3/2/90 B-8

ATTACHMENT 1 VERMONT YANKEE NUCLEAR POWER PLANT Application for Approval to Routinely Dispose of Septic Waste With Minimal Levels of Radioactivity

1.0 INTRODUCTION

Vermont Yankee Nuclear Power Corporation (Vermont Yankee) requests approval. pursuant to 10CFR20.302(al. of a method proposed herein for the routine disposal of slightly contaminated septic tank waste. Vermont Yankee proposes to dispose of this waste by spreading it on designated areas within the plant's site boundary fence. This application addresses specific information requested in 10CFR20.302(a). 2.0 WASTE STREAM DESCRIPTION The waste involved in this application consists of residual solids and water associated with the sewage collection system at Vermont Yankee; The plant's sewage systems are of the septic tank and disposal field type. The two systems servicing the majority of the plant's sanitary waste are identified as (1) main septic system and (2) the south sewage disposal system. The main septic system (design flow capacity 4,950 gallons/day) consists of a wastewater lift station, septic tank, and dual alternating disposal fields located on the north side of the plant. This system services the main complex of buildings central to the plant and processes approximately 3,500 gallons of wastewater per day. The septic tank, shown in Figure Cl), will typically contain 9,250 gallons of septage. The south sewage disposal system is a newly-installed (January 1989) pressurized mound system. which is used in lieu of the construction office building (COB) holding tank that had previously serviced the lavatory facilities on the south end of the plant. The new system is composed of a septic tank (5,700 gallon capacity, see Figure 2), pumping station, and pressurized mound disposal field. When dosing the field, a force main pressurizes the disposal field's piping system with the septic tank effluent, which distributes throughout the field. The south sewage disposal system has Revision ~9~ Date 3/2/90 B-9

the design flow capacity to process 4,607 gallons of wastewater per day. The system is typically loaded at approximately 2,500 gallons per day during normal plant operations. Figure (3) indicates diagrammatically the flow of both potable and wastewater throughout Vermont Yankee. Both the main septic system and the south sewage disposal system's septic tanks collect waste from the plant's lavatories, showers. kitchens. and janitorial facilities outside the Radiological Control Area (RCA). No radioactivity is intentionally discharged to either of the septic systems. However, plant investigations into the source of low levels of contamination found in septic waste have identified that very small quantities of radioactive materials. which are below detection limits for radioactivity releases from the RCA, are carried out of the control area on individuals and accumulate in the septic waste collection tanks by way of floor wash water, showers, and hand washing. As a means of minimizing'the transport of radioactive materials into the septic collection tanks. the primary source of the radioactivity Ci .e .. floor wash water) is now poured through a filter bag to remove suspended solids and dirt before the water is released into a janitorial sink. The majority of the radioactivity found in waste sludge has been associated with the main septic tank. Grab samples of sludge from the bottom of the COB and main septic tank were analyzed by gamma spectroscopy with the following results of plant-related radionuclides: Activity Concentration Isotope. +/-1 Sigma <pC1/kg Wet) COB Sludge Cs-137 10.3 +/- 1.8 (June 8, 1988) Co-60 45.4 +/- 3.1 Main Tank Sludge Mn-54 39.3 +/- 4.3 (June 8, 1988) Co-60 853.0 +/- 12.0 Zn-65 52.7 +/- 8.2 Cs-134 13.0 +/- 2.2 Cs-137 120.7 +/- 5.2 Revision ~9~ Date 3/2/90 B-10

Th~ principle radionuclide is Cobalt-60, which accounts for 79% of the plant related activity in the septage samples. In comparison to in-plant smear samples taken for 10CFR61 waste characterizations, the septage sample from the main tank correlates very close with the distribution of radionuclides identified in-plant as shown below: Relative Isotopic Distributions Isotope In-Plant Smears Main Tank Sludge Mn-54 3.6% 3.6% Co-60 81.5 79.1 Zn-65 3.8 4.9 Cs-134 0.4 1.2 Cs-137 10.3 11.2 Additional analyses of the main tank septage showed that the liquid portion of the collected sample did not contain any plant-related activation or fission products. and that essentially all of the activity in the waste was associated with the solid sludge fraction. The average density of the collected sludge was found to be approximately equal to that of water, with a wet to dry ratio of 25.4 to 1. Both the liquid and solid fractions of the main tank septic waste were also analyzed for strontium with no detectable activity found. The liquid portion of the waste sample was also analyzed for tritium with no activity above the minimum detectable levels found. Appendix A to Attachment 2 contains the laboratory analysis reports of the samples taken from the COB and main septic tanks. Prior to identification of the plant-related radioactivity in septage waste. the COB holding tank was being pumped on the average of twice per week, with the sludge and waste liquid transported off-site primarily to the Brattleboro, Vermont, sewage treatment facility. Waste from the main septic tank was being pumped and transported off-site for disposal on the average of twice per year. Revision ~9~ Date 3/2/90 B-11

With the replacement of the COB holding tank by the new south sewage disposal system. and the requested implementation of on-site land disposal of accumulated septic waste. the frequency of collection tank pump-outs with land application of the waste is expected to be once per year. With the past pump-out frequency of the main tank being every six months. the accumulation of sludge at the bottom of the tank was well below its design capacity. During the 1988 sample collections. it was estimated that the sludge thickness was less than l foot of its 6-foot depth. However, for conservatism in the radiological evaluations. it is assumed that the sludge layer in the main septic tank and south disposal tank occupies 30% of their combined design volume, and that the frequency of pump-outs is semiannual as opposed to the exoected annual cycle. Also. as noted above from laboratory analyses of the sludge layer taken from the bottom of the main tank. the average density of the tank contents is approximately equal to that of water. with a wet-to-dry ratio of 25.4 to 1. Hence. the weight of solids CW 501 ) being disposed of is estimated. for purposes of this bounding dose assessment. to be approximately: wsol = 14,950 [gal] x 3,785.4 [cc/gal] x 10" 3 [kg/cc] x 0.30 [solids fraction] x Cl/25.4) [dry/wet ratio]

                     - 700 [kg] per pump-out of both tanks or. 1,400 kg of dry solids per year.

3.0 DISPOSAL METHOD Approval of this application will allow Vermont Yankee to dispose of septage by utilization of a technique of land spreading or surface injection in a manner consistent with all applicable state of Vermont health regulations regarding disposal of septic waste. Details of the ~hemical and biological controls necessary to satisfy state health code requirements are provided in Reference 5. The septage will be spread or surface injected on land areas owned by Vermont Yankee and situated within the plant's site boundary. Transportation of the septage waste to the disposal areas will involve pumping from one of the septic waste collection tanks Ci .e .. main septic tank, COB holding tank. Revision ~9~ Date 3/2/90 B-12

new replacement COB septic tank, or from any other on-site septic waste collection point) into an enclosed truck-mounted tank. The enclosed tank truck is used to prevent spillage while in transit to the disposal areas. The septage will be transported to one of the two disposal sites designated for land application for septage from Vermont Yankee, and applied at a fixed rate based on either limitations imposed by the state of Vermont for heavy metals or organic content of the waste, or on the radioactivity content such that projected maximum individual doses will not exceed established dose objectives. 3.1 Septic Waste Disposal Procedure Gamma isotopic analysis of septic waste shall be made prior to each disposal by obtaining a representative sample from each tank prior to pump-out. At least two septic waste samples will be collected from each tank to be pumped by taking a volumetric column of sludge and waste water which allows for analysis of the solid's distribution and content from top to bottom of each tank. The weight percent of solid content of the collected waste will be determined and applied to the gamma isotopic analysis in order to estimate the total radioactivity content of each tank to be pumped and spread on designated disposal fields. These gamma isotopic analyses of the representative samples will be performed at the environmental Technical Specification lower limit of detection (LLD) requirements for liquids (see Technical Specification Table 4.9.3) in order to document the estimatton of radiological impact from septage disposal. The radionuclide concentrations and total radioactivity identified in the septage will be compared to the concentration and total curie limits established herein prior to disposal. The methodology and limits associated with determining compliance with the disposal dose and activity criteria are described in Attachment 2. If the concentration and total activity limits are met, compliance with the dose assessment criteria will have been demonstrated since the radiological analysis (Section 4.5 and Attachment 2) was based on evaluating the exposure to a maximally exposed individual and inadvertent intruder after the accumulation of twenty years of periodic semiannual Revision ~9~ Date 3/2/90 B-13

spreading of the septic waste on a single (2 acre) plot within one of the designated disposal areas. If the activity limit per d1sposal area is projected to be exceeded, the appropriate exposure pathways as described in Section 4.5 will be evaluated prior to each additional application, or a separate plot within the designated disposal area will be utilized. Annually, for years in which disposal occurs, the potential dose impact from disposal operations conducted during the year. including the impact from previous years. will be performed and results reported in the plant's Semiannual Radioactive Effluent Release Report which is filed after January 1. All exposures will be assessed utilizing the methodology described in Attachment 2. The established dose criteria requires that all applications of septage within the approved designated disposal areas shall be limited to ensure the dose to a maximally-exposed individual be maintained less than 1 mrem/year to the whole body and any organ, and the dose to the inadvertent intruder be maintained less than 5 mrem/year. The total activity based on the measured radionuclide distribution for any single disposal plot is not expected to exceed the following: Maximum Accumulated Radioactivity Allowed Per Acre Isotope --~O;~....C.:..i""]_ __ Mn-54 1.4 Co-60 120.0 Zn-65 1.4 Cs-134 0.7 Cs-137 46.5 If any of the above radionuclides are projected to exceed the indicated activity values, then dose calculations will be performed prior to spreading, in accordance with the methods detailed in Section 4.2.2 of Attachment 2. to make the determination that the dose limit criteria will not be exceeded. Revision _i_ Date 3/2/90 B-14

The concentration of radionuclides in any tank of septic waste to be disposed of will also be limited to a combined Maximum Permissible Concentration of Water CMPC) (as listed in 10CFR, Part 20, Appendix B. Table II. Column 2) ratio of less than or equal to 0.1. For radiological control, each application of septage will be applied on the designated land area by approved plant proced~re which adheres to the following assumptions which were used in developing the dose impact: o During surface spreading or injection, the septage, and any precipitation falling onto or flowing onto the disposal field. shall not overflow the perimeter of the designated area. o Septage shall not be surface spread or injected into the top 6-inch soil layer within 300 feet from any drinking water well supply. o Septage shall not be surface spread closer than 300 feet from the nearest dwelling or public building (or within 100 feet if injected into the top 6-inch surface layer). o Septage shall not be surface spread closer than 50 feet (or within 25 feet if injected into the top 6-inch surface layer> from any roads or site boundary adjacent to land areas. o Septage shall not be surface spread within 100 feet (or within 50 feet if injected into the top 6-inch surface layer) of any surface water (rivers, streams, drainage ditches). o Low areas of the approved fields. subject to seasonally high groundwater levels, are excluded from the septage application. In addition to the radiological controls to limit the total accumulation of radioactive materials released by septic waste spreading, state of Vermont health code requirements will be followed to ensure the protection of the public and environment from chemical and biological hazards. The application rate and acreage will be determined prior to each Revision ~9~ Date 3/2/90 8-15

disposal .operation. This will vary with the chemical composition of the septage. the percent solids, and the radioactive concentrations. 3.2 Administrative Procedures Complete records of each disposal will be maintained. These records will include the concentration of radionuclides in the septage, the total volume of septic waste disposed, the total activity in each batch as well as total accumulated on the disposal plot at time of spreading, the plot on which the septage was applied, and the results of any dose calculations required. The annual disposal of septage on each of the approved plot areas will be limited to within the established dose, activity, and concentration criteria noted above, in addition to limitations dictated by chemical and biological conditions. Dose guidelines, and concentration and activity limits, will be maintained within the appropriate values as detailed in Attachment 2. Any farmer using land which has been used for the disposal of septic waste will be notified of any applicable restrictions placed on the site due to the land spreading or injection of waste. 4.0 EVALUATION OF ENVIRONMENTAL IMPACT 4.1 Site Characteristics 4.1.1 Site Topography The proposed disposal sites consist of two fields located on the Vermont Yankee Nuclear Power Plant site, which is located on the west bank of the Connecticut River in southwestern Vermont at latitude 42 degrees, 47 minutes north and longitude 72 degrees 31 minutes west. Both fields are on plant property within the site boundary and surrounded by a chain link fence. Revision ~9~ Date 3/2/90 B-16

Site A contains an approximate eight-acre parcel of usable land centered approximately 2,200 feet northwest of the Reactor Building. Site B contains about two acres and is centered approximately 1,700 feet south of the Reactor Building. The usable acreage of both the north and south disposal fields is restricted to those areas which have no slopes greater than five percent to limit surface runoff. A radiological assessment based on the 1988 measured radioactivity concentrations in sludge has determined that a single two-acre plot would be sufficient for the routine disposal of septage for twenty years without exceeding the dose criteria to maximum exposed individual or inadvertent intruder. As a result, the eight-acre field to the northwest could be divided into four disposal plots, with the two-acre site at the south end of the plant site. providing a fifth plot. A portion of the United States Geological Survey topographic map (Brattleboro quadrangle), showing the plant site, is presented in the Final Safety Analysis Report CFSAR) as Figure 2.5-1. A plan map showing the plant site and the disposal sites is given on Figure 4. The sites are located along a glacial terrace on the west side of the Connecticut River. This terrace extends about 3,000 feet west rising gently and then more abruptly to a higher terrace and then to dissected uplands. Distance to the east from the disposal sites to the river is at least 100 feet if septage is disposed of by surface spreading within the designated areas, or 50 feet if septage is injected directly into the soil. Relief of the proposed disposal sites is low, with elevation ranging between 250 feet and 265 feet (msl ). Mean water surface elevation of the adjacent river is about 220 feet. The topographic character of the site and surrounding area is compatible with this use. The spreading of septage at these lo~ations will have no effect on the topography of the area. Revision ~9~ Date 3/2/90 B-17

4.1.2 Site Geology Profiles of site exploratory borings are shown in the FSAR in Figures 2.5-8 through 2.5-11. Current site characteristics as determined from a recent detailed site investigation can be found in Reference 5. Composition of surfacial materials is compatible with the proposed use of the site for septic waste disposal. 4.2 Area Characteristics 4.2.1 Meteorology The site area experiences a continental-type climate with some modification due to the marine climate which prevails at the Atlantic seacoast to the east. Annual precipitation averages 43 inches and is fairly evenly distributed in each month of the year. Potential impacts on septic waste disposal include occasional harsh weather: ice storms, severe thunderstorms. heavy rains* due to hurricanes, the possibility of a tornado, and annual snowfall of from 30 to 118 inches per year. In addition, frozen ground can occur for up to 4 months of the year. Septage spreading will be managed by written procedure such that material which is spread or a mix of that material with precipitation will not overflow the perimeter of the disposal site. Additional information on meteorology of the site can be found in Section 2.3 of the Final Safety Analysis Report. 4.2.2 Hydrology Hydrology of the site and local area is tied closely to flow in the adjacent Connecticut River. River flow is controlled by a series of hydroelectric and flood-control dams including the Vernon Dam which is about 3,500 feet downstream of the site. Revision ~9~ Date 3/2/90 B-18

All local streams drain to the Connecticut River and the site is in the direct path of natural groundwater flow from the local watershed easterly toward the river. Site groundwater level is influenced by both precipitation and changes in the level of ponding of the Connecticut River behind the Vernon Dam due to natural flow or dam operation. Flood flows on the Connecticut are controlled by numerous dams including five upstream of the site. Elevation of the 100-year flood is about 228 ft (msl); and, thus, well below the elevation of the proposed site which ranges from about 250 to 265 feet (msl). The 100-year flood level is based on information presented in References (1) and "(2). Septage disposal by means of land spreading on the proposed site will have no adverse impact on area hydrology. Further information about site hydrology is in Section 2.4 of the FSAR. 4.3 Water Usage 4.3.l Surface Water The adjacent Connecticut River is used for hydroelectric power, for cooling water for the Vermont Yankee plant. as well as for a variety of recreational purposes such as fishing and boating. The Connecticut River is not used as a potable water supply within 50 miles downstream of the plant. Locally, water from natural springs are used for domestic and farm purposes. FSAR Table 2.4.5 and Figure 2.4-2 show springs used within a 1-mile radius of the site. FSAR Table 2.4.4 and Figure 2.4~1 show water supplies with surface water sources which are within a ten-mile radius of the site. There will be no impact on surface water usage or quality as a result of septage disposal due to the required separation distances between surface waters and the disposal plots. Revision ~9~ Date 3/2/90 8-19

4.3.2 Groundwater Based on a review of groundwater measurements in various site borings presented in the FSAR and References 3 and 5, an upper estimate of groundwater levels at the plant is about 240 feet. Considering the proximity of the Connecticut River and Vernon Pond, with a mean water surface elevation of 220 feet, this estimate for the groundwater level appears to be reasonable. Given the topography of the proposed disposal sites, it is highly unlikely that the groundwater level will be within 3 feet of the disposal area surface elevation. Prior to each application of septic waste to a disposal plot. the groundwater level in nearby test wells will be determined and no application will be allowed if the groundwater level in the vicinity of the disposal plot is found to be less than 3 feet. Groundwater provides potable water for public wells as shown in FSAR Table 2.4.5 and Figure 2.4-1. Groundwater flow in the vicinity of the proposed disposal sites is towards the Connecticut River. There are no drinking-water wells located between the site and the river. Therefore. it is highly unlikely that any drinking water wells could be affected by septage disposal. FSAR Figure 2.4-2 and Table 2.4-5 present information on private wells near the plant. The Vermont Yankee on-site wells provide water for plant use. Th{s supply is routinely monitored for radioactive contamination. To quantify the impact of septage disposal on the Connecticut River, a conservative groundwater/radionuclide travel time analysis was performed. For an assumed average travel distance of 200 feet from the disposal site to the river, a groundwater travel time of 408 days was estimated from Darcy's Law. This estimate is based on a permeability for the glacial till of 10 gpd/ft 2 , a hydraulic gradient of 0.11 ft/ft, and a soil porosity of 0.3. This analysis conservatively assumed that the septage placed on the ground was immediately available to the groundwater. In practice, a minimum of 3 feet separation between groundwater and the surface will be required at time of application of the septic waste. Revision _j_ Date 3/2/90 B-20

Due to ionic adsorption of the radionuclides on solid particles in the groundwater flow regime, most radionuclides travel at only a small fraction of the groundwater velocity. For the radionuclides present in the sludge, retardation coefficients were developed from NUREG/CR-3130 (Reference 4). Retardation coefficients for Co-60, Cs-137, and Cs-134 were directly obtained from NUREG/CR-3130. The coefficients for Zn-65 and Mn-54 were conservatively estimated using NUREG/CR-3130 as a guide. The radionuclides, their half-lives, retardation coefficients, and their travel time to the river are summarized in Table 1. TABLE 1 Radionuclide Travel Times Retardation Travel Time Radionuclide Half Life Coefficient to River Co-60 5.3 years 860 961 years Cs-137 30.2 years 173 193 years Cs-134 2.1 years 173 193 years Zn-65 244 days 3 1. 224 days Mn-54 312 days 3 1. 224 days The radiological impact on the river for the radionuclides reaching the river under this conserv3tive analysis is discussed in Attachment 2. Water usage of the Connecticut River downstream from the disposal area is limited to drinking water for dairy cows, irrigation of vegetable crops, and irrigation of cow and cattle fodder. Based on the assessments noted above, it is concluded that groundwater sources will not be adversely impacted as a result of septage disposal on the proposed site. 4.4 Land Use Both the eight-acre and two-acre sites proposed for the disposal areas are currently part of the Vermont Yankee Nuclear Power Plant Site inside the plant's site boundary which is enclosed by a chain link fence. It is Revision ~9~ Date 3/2/90 B-21

undeveloped except for transmission line structures which traverse a portion of the northern disposal area. Development potential is under the control of Vermont Yankee. At present. the eight-acre site on the north end of the plant property is used by a local farmer for the growing of feed hay for use with his dairy herd. No curtailment of this activity as a result of the low levels of radioactivity in septage will be necessary. Utilization of the proposed sites for septic waste disposal will result in no impact on adjacent land or properties because of the separation of the disposal plot~ from off-site properties. the general movement of ground~ater toward the river and away from adjacent land areas, and the very low levels of radioactive materials contained in the waste. Administrative controls on spreading and the monitoring of disposal ared condicions will provide added assurance that this proposed practice will not impact adjacent properties. 4.5 Radiological Impact In addition to state of Vermont limits imposed on septage spreading, based on nutrient and heavy metal content, the amount of septage applied on* each of the proposed disposal plots will also be procedurally controlled to insure doses are maintained within the stated limits. These limits are based on NRC Nuclear Reactor Regulation CNRR> staff .Proposed guidance (described in AIF/NESP-037, August 1986). The proposed dose criteria require that the maximally exposed member of th2 general public ~eceive a dose less than 1 mrem/year to the whole body or any organ due to the disposal material, and less than 5 mrem/year to an inadvertent intruder. To assess the doses received by the maximally-exposed individual and the inadvertent intruder, six potential pathways have been identified. These include: (a) Standing on contaminated ~round, (b) Inhalation of resuspended radioactivity, Revision ~9~ Date 3/2/90 B-22

(c) Ingestion of leafy vegetables, (d) Ingestion of stored vegetables, (f) Ingestion of meat, and (g) Ingestion of milk. The liquid pathway was also evaluated and determined to be insignificant. Both the maximum individual and inadvertent intruder are assumed to be exposed to these pathways with difference between the two related to the occupancy time. The basic assumptions used in the radiological analys~s include: (a) Exposure to the ground contamination and to resuspended radioactivity is for a period of 104 hours per year during Vermont Yankee active control of the disposal sites, and continuous thereafter. The 104-hour interval being representative of a farmer's time on a plot of land (4 hours per week for 6 months). (b) The septic tanks are emptied every two to three years. (The assumed practice is to pump septic tanks once per year. The actual practice may be to pump septic tanks every two to three years.) (c) The tank radioactivity remains constant at the currently determined level. To account for the uncertainty associated with the counting statistics, the measured activity concentrations listed in Section 2 were increased by 3 sigmas. That is, the activity concentrations employed in dose assessment and the total radioactivity content per pump-out (at 700 kg of solids per batch) are as follows: Revision ~ Date 05/12/99 B-23

Upper-Bound Activity Upper-Bound Activity Isotope Concentration [pCi/kg dry] Content [Ci/tankful) Mn-54 1.348 9.436E-07 Co-60 23.060 . 1.614E-05 Zn-65 1,620 1.134E-06 Cs-134 322 2.254E-07 Cs-137 4 .100 2.870E-06 (d) The radiation source corresponds to the accumulation of radioactive material on a single plot (two-acre) within the proposed disposal sites over a period of 20 years (40 applications at 6-month intervals). (In actuality, the proposed sites will accommodate more than one disposal plot. and. in practice. more than one plot will most probably be used with an application frequency of once per year.) (e) For the analysis of the radiological impact during Vermont Yankee active control of the disposal sites. al 1 dispersed radioactive material remains on the surface and forms a source of unshielded radiation. (In practice. the septic waste will be either surface spread or directly injected withiri the top 6 inches of the disposal plot, in which case, the radioactive material will be mixed with the soil. This, in effect, would reduce the ground plane source of exposure by a factor of about four due to self-shielding.) (f) No radioactive material is dispersed directly on crops for human or animal consumption. crop contamination being only through root uptake. Cg> The deposition on crops of resuspended radioac~ivity is insignificantly sma 11 . Revision ~9~ Date 3/2/90 B-24

(h) Pathway data and usage factors used in the analysis are the same as those used in the plant's ODCM assessment of the off-site radiological impact from routine releases.with the exception that the fraction of stored vegetables grown on the disposal plots was conservatively increased from 0.76 to 1.0 (at present no vegetable crops for direct human consumption are grown on any of the proposed disposal plots). (i) It is conservatively assumed that Vermont Yankee relinquishes control of the disposal sites after the fortieth pump-out Ci .e .. the above source term applies also for the inadvertent intruder). (j) For the analysis of the impact after .Vermont Yankee control of the sites is relinquished. the radioactive material is plowed under and forms a uniform mix with the top six inches of soil: but. nonetheless. undergoes resuspension at the same rate as surface contamination. From radiological impact assessments associated with the disposal of septage on different plot sizes (Attachment 2), it was determined that a single two-acre plot within the disposal sites would accommodate the 1 mrem/year prescribed dose to the critical organ of the maximally exposed individual for a period of up to 20 years. as well as the 5 mrem/year prescribed dose to the inadvertent intruder after control is assumed to be relinquished. The calculated potential radiation exposures following the spread!ng of 40 combined (main septic system and south disposal system) tankfuls Cat six-month intervals) on a single two-acre plot are as follows: Control of Disposal Sites Radiation Exposure Individual/Organ Controlled by VYNPS 0.1 mrem/yr Child/Whole Body (Maximum Exposed Individual) 0.2 mrem/yr Maximum Child/Liver Uncontrolled 1.3 mrem/yr Adult/Whole Body (Inadvertent Intruder) 3.9 mrem/yr Maximum Teenager/Lung Revision ~9~ Date 3/2/90 B-25

The individual pathway contributions to the total dose at the end of the 20-year accumulation of waste deposited on a single two-acre plot are as listed below: Pathway-Dependent Critical Organ Doses Maximally Exposed Inadvertent Intruder Individual/Organ Critical Individual/Organ (Child/Liver) (Teenager/Lung) Pathway (mrem/year) (mrem/year) Ground Irradiation 0.0576 1.16 Inhalation 0.00122 2.74 Stored Vegetables 0.0913 0.00601 Leafy Ve£jetab1e 0.0046i 0.00040 Milk Ingestion 0.0421 0.00229 Meat Ingestion 0.00249 0.00012 TOTAL 0 .1994 3.909 In addition, an isotdpic breakdown of the critical or~an dose results listed above is shown in the following table: Isotopic Breakdown of Maximum Radiation Exposures Radioactivity Exposure Description Isotope [uCii2 Acres] [rnrem/yrJ During Vermont Yankee Mn-54 2.831 0.000436 control of the Co-60 235.3 0.0559 disposal sites. Zn-65 2.801 0.0230 Maximally Exposed Cs-134 1.457 0.00231 Individual/Organ: Cs-137 92.59 0.118 Child/Liver TOTAL 0.199 After Vermont Yankee Mn-54 2.831 0.0144 control of sites is Co-60 235.3 3.76 relinquished. Zn-65 2.801 0.00983 Inadvertent Intruder Cs-134 1.457 0.000505 Critical Individual/ Cs-137 92.59 0.1247 Organ: Teenager/Lung TOTAL 3.91 Revision ~9~ Date 3/2/90 B-26

Of interest are also derived dose conversion factors which provide a means of ensuring septage disposal operations within the prescribed radiological guidelines. The critical-organ (worst-case) all-pathway values per acre are as follows: All-Pathway Critical-Organ Dose Conversion Factors During Vermont Yankee Control of Disposal Sites Exposure Isotope Individual/Organ [mrem/yr-uCi/acreJ Mn-54 Adult /GE - LLI 3.74E-4 Co-60 Teenager/Lung 7.14E-4 Zn-65 Child/Liver l.64E-2 Cs-134 Child/Liver 3.18E-3 Cs-137 Child/Bone 2.66E-3 The calculational methodology and details of the radiological assessment and proposed operational controls on total activity and concentration of waste to be disposed are presented in Attachment 2. 5.0 RADIATION PROTECTION The disposal operation will follow the applicable Vermont Yankee procedures to maintain doses as low as reasonably achievable and within the specified dose and release concentration criteria. Revision ~9~ Date 3/2/90 B-27

REFERENCES

1. Flood Insurance Study, Vernon, Vermont. Windham County, FEMA, Community No. 500137, July 25. 1980.

2. Flood Insurance Study, Town of Hinsdale, New Hampshire, Cheshire County, FEMA, Community No. 330022. October 15, 1980.

3. Vermont Yankee Well Development Evaluation by Wagner. Heindel, and Noyes. Inc. July 10, 1986.

4. NUREG/CR-3130, Influence of Leach Rate and Other Parameters on Groundwater Migration, by Dames & Moore, February 1983.

5. Vermont Yankee Nuclear Power Corporation On-Site Septage Disposal Plan, by Wagner, Heindel, and Noyes, Incorporated, June 1989.

Revision ~9~ OAte 3/2/90 B-28 L __

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(This attachment to Appendix B is incorporated into the ODCM by reference due to size. A complete copy is on file with Vermont Yankee Document Control as part of Correspondence Letter BVY 89-59.) ATTACHMENT 2 VERMONT YANKEE NUCLEAR POWER PLANT RADIOLOGICAL ASSESSMENT OF ON-SITE DISPOSAL OF SEPTIC*WASTE and PROPOSED PROCEDURAL CONTROLS TO ENSURE COMPLIANCE WITH RADIOLOGICAL LIMITS Revision ~9~ Date 3/2/90 B-31

APPENDIX D ASSESSMENT OF SURVEILLANCE CRITERIA FOR GAS RELEASES FROM WASTE OIL INCINERATION Revision _1.§_ Date 10/28/93 D-1 L

APPENDIX D ASSESSMENT OF SURVE!LLANCL CRITERIA FOR GAS RLLEASES FROM WASTE OIL INCINERATION INTRODUCf!ON: fhe Nuclear Regulatory Commission amended its regulations (10CFR20) in a Federal Register Notice (Vol. 57, No. 235; page 57649 I Monday, December 7, 1992) that permitted the on-site lncineration of contaminated waste oil generated at licensed nuclear power plants without the need to ~mend existing operating licenses. This action will help to ensure that the limited capacity of licensed low level waste disposal facilities is used efficiently while maintaining releases from operating nuclear power plants at levels which are "as low as reasonably achievable." Incineration of this class of waste must be in f11ll compliance with the Commission'5 current regulations that restrict the release of radioactive materials to the envlronment. Any other applicable Federal. State. or local requirements that relate to the toxic or haLardous characteristics of the waste oil would also have to be satisfied. Incineration of waste oil is to be carried out under existing effluent limits, recordkeeping and reporting requirements. Specifically, licensees must comply with the effluent release limitations of 10CFR Part 20. and Part 50; Appendix I. This includes the site gaseous pathway dose and dose rate limits contained in the plant's Technical Spec~fications (Section 3.8). The dose contribution to members of the public resulting from the on-site incineration of contaminated waste oil must not cause the total dose or dose rate from all effluent sources to exceed the dose or concentration limits imposed by 10CFR20, 10CFR50; Appendix I. and the Radiological Effluent Technical Specifications (RETS). It is expected that the actual contribution to public exposures caused by waste oil burning will be a small fraction of the site's effluent limits. as well as a small portion of the total releases from the site. SOURCE DESCRIPTION Contaminated waste oil suitable for on-site incineration can be burned in the Waste Oil Burner located in the North Warehouse. The burner has its own exhaust stack situated on the roof of the warehouse. However. due to the short height of the exhaust stack above the roof line, this release point is considered to be a ground level point source for modeling discharges to the environment. In addition, the building wake effects from the North Warehouse are assumed to be independent of the larger Turbine Hall/Reactor Building complex due to its distance from these main structures. Consequently, the relatively small size of the North Warehouse leads to Revision _l§_ Date 10/28/93 D-2

meteorological dispersion factors that are conservative with respect to the dispersion factors for the main plant structures. The waste oil burner is rated to process oil at a 2 gal ./hour from a 500 gallon day tank. The offgas flow rate for the burner is rated as 199 cfm. This provides an air to oil dilution during the incineration of 44,800. WASTE OIL SAMPLING/SURVEILLANCE REQUIREMENTS The oil burner stack is not equipped with continuous air monitoring or sampling capability for the direct determination of radioloefical effluent releases during the incineration process. As a consequence. sampling and analysis of the waste oil prior to its incineration is necessary to project the dose and dose rate con5equcnces of burning cont~minated oil. Calculations of projected dose from the incineration of total quantity of oil to be added to the Waste Oil Burn Day Tank for each series of burns will be performed in accordance with the methods in the ODCM and compared to the accumulated site total dose for that period before initiation of incineration. Dose rate determinations will be determined by averaging the projected dose for-the quantity of radioactivity determined to be present in the oil over the expected duration of the burn necessary to incinerate the total volume to be added to the Day Tank. Inherent in this determination is the assumption that all radioactivity found to be present in each batch of oil will be released to the atmosphere during the incineration. No retention of activity in the combustion chamber is assumed in calculating the offsite radiological impact. Norn1al sampling and analysis methods for gaseous release streams cannot be applied directly to liquids (waste oil). Therefore, the sampling and analysis requirements for liquids as identified in Technical Specification Table 4.8.1 shall be used to determine the level of contamination in waste oil. The stated Lower Limits of Detection (LLD) given on Table 4.8.1 provide assurance that undetectable levels of contamination up to the LLD values will not result in a significant dose impact to the maximum offsite receptor. If waste oil was burned continuously for an entire calendar quarter. and the radionuclides listed in the ODCM Dose Conversion Factor Table 1.1-12 were assumed to be present in the oil at the LLD values specified in Technical Specification Table 4.8.1. the resultant maximum organ dose would amount to only 0.28% of the ALARA quarterly limit of 7.5 mrem. The principle limitation in the incineration of waste oil is that the site release limits contained in RETS, and implemented by the ODCM methodology, shall not be exceeded. The use of the liquid LLDs on waste oil sample analyses provide sufficient sensitivity to ensure that site dose limits will not be exceeded as a consequence of burning slightly contaminated oil. Revision __l§_ Date 10/28/93 D-3

APPENDIX F APPROVAL PURSUANT TO 10CFR20.2002 FOR ONSITE DISPOSAL OF COOLING TOWER SILT Revision _fl_ Date 08/14/97 F-1

UNITED STATES NUCLEAR REGULATORY COMMISSION WA.SHINOTON, 0,C. 2055.S-0001 June 18, 1997 w~.. -*, .r~*'.: NVY 97-85 ~-- 1

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                                                                                                                       ~ ...I Mr. Donald A. Reid
                                                                               **-,T **  ,*:-,-' **'  *. * ~:>..:'*'I Vice President, Operations Vermont Yankee Nuclear Power Corporation Ferry Road Brattleboro, VT 05301

SUBJECT:

REVISED SAFETY EVALUATION - APPROVAL PURSUANT TO 10 CFR 20.2002 FOR ONSITE DISPOSAL OF COOLING TOWER SILT - VERMONT YANKEE NUCLEAR POWER STATION (TAC NO. M96371)

Dear Mr. Reid:

By letter dated August 30, 1995, Vermont Yankee Nuclear Power Corporation (VYNPC) requested approval. pursuant to 10 CFR 20.2002, for the onsite disposal of slightly contaminated silt material removed from Vermont Yankee Nuclear Power Station's (Vermont Yankee's) cooling towers. In a safety evaluation (SE) dated March 4, 1996, the NRC staff approved the proposed silt disposal. However, because of discrepancies VYNPC identified between the safety evaluation and VYNPC's letter of August 30, 1995, VYNPC postponed implementation of the silt disposal until resolution of the discrepancies. By letter dated August 2. 1996, VYNPC informed the NRC staff of the discrepancies and requested that the SE be revised accordingly. Recognizing the discrepancies, the NRC staff has prepared the enclosed SE to resolve the discrepancies and to replace the SE of March 4, 1996. LAI The NRC staff's approval of VYNPC's silt disposal request is granted provided 1130 the enclosed replacement SE is permanently incorporated into Vermont Yankee's Offsite Dose Calculation Manual as an appendix. Any modification to the proposed action that may be considered in the future must have prior NRC staff approval Pursuant to the prov1s1ons of 10 CFR Part 51. the Commission has published in the Federal Register an Environmental Assessment and Finding of No Significant Impact (61 FR 6662). Revision _1.1_ Date 08/14/97 F-2

D. Reid If you have any further questions regarding this matter. please contact Mr. Kahtan Jabbour at (301) 415-1496. Sincerely, Patrick D. Milano, Acting Director Project Directorate I-3 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation Docket No. 50-271

Enclosure:

Safety Evaluation cc w/encl: See next page Revision _fl_ Date 08/14/97 F-3

Vermont Yankee Nuclear Power Vermont Yankee Nuclear Power Station Corporation cc: Mr. Peter LaPorte. Director Regional AcJministr*ator. Region I AfTN: James Muckerheide U. S. Nuclear Regulatory Commission Massachusetts Emergency Management 475 Allendale Road Agency King of Prussia, PA 19406 400 Worcester Rd. P.O. Box 1496 R. K. Gad, III Framingham. MA 01701-0317 Ropes & Gray One International Place Mr. Raymond N. McCandless Boston, MA 02110-2624 Vermont Division of Occupational and Radiological Health Mr. Richard P. Sedano. Commissioner Administration Building Vermont Department of Public Service Montpelier, VT 05602 120 State Street. 3rd Floor Montpelier, VT 05602 Mr. J. J. Duffy Licensing Engineer Public Service Board Vermont Yankee Nuclear Power State of Vermont Corpo*rat ion 120 State Street 580 Main Street Montpelier, VT 05602 Bolton. MA 01740-1398 Chairman. Board of Selectman Mr. Robert J. Wanczyk Town of Vernon Director of Safety and Regulatory P.O. Box 116 Affairs Vernon, VT 05354-0116 Vermont Yankee Nuclear Power Corp. Ferry Road Mr. Richard E. McCullough Brattleboro. VT 05301 Operating Experience Coordinator Vermont Yankee Nuclcdr Power Station t~r. Ross B. Barkhurst, President P.O. Box 157 Vermont Yankee Nuclear Power Governor Hunt Road Corporation Vernon, VT 05354 Ferry Road Brattleboro, VT 05301 G. Dana Bisbee. Esq. Deputy Attorney General Mr. Gregory A. Maret, Plant Manager 33 Capitol Street Vermont Yankee Nuclear Power Station Concord. NH 03301-6937 P.O. Box 157 Governor Hunt Road Resident Inspector Vernon, VT 05354 Vermont Yankee Nuclear Power Station U.S. Regulatory Commission P.O. Box 176 Vernon, VT 05354 Chief, Safety Unit Office of the Attorney General One Ashburton Place. 19th Floor Boston. MA 02108 Revision _fl_ Date 08/14/97 F-4

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20~1 SAFETY EVALUATION BY HIE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO ONSITE DISPOSAL OF SLIGHTLY CONTAMINATED COOLING TOWER SILT VERMONT YANKEE NUCLEAR POWER CORPORATION VERMONT YANKEE NUCLEAR POWER STATION DOCKET NO. 50-271

1.0 INTRODUCTION

By letter dated August 30, 1995, Vermont Yankee Nuclear Power Corporation (VYNPC) requested approval for the onsite disposal of slightly contaminated silt material removed from Vermont Yankee Nuclear Power Station's (Vermont Yankee's) cooling towers. In a safety evaluation (SE) dated March 4, 1996. the NRC staff approved the proposed silt disposal. However, because of discrepancies between the SE and VYNPC's letter of August 30, 1995, VYNPC postponed implementation of the silt disposal until resolution of the discrepancies. By letter dated August 2, 1996, VYNPC informed the NRC staff of the discrepancies and requested that the SE be revised accordingly. Recognizing the discrepancies, the NRC staff has prepared this SE to resolve the discrepancies and to replace the SE of March 4, 1996.

2.0 BACKGROUND

VYNPC has previously obtained NRC staff approval of the onsite disposal of very-low-level radioactive material similar to the proposed silt disposal. By letter dated June 28, 1989, VYNPC proposed the onsite disposal of slightly contaminated septic waste material by land application at Vermont Yankee. By letter dated August 30. 1989, the NRC staff approved this request pursuant to 10 CFR 20.302 (now 10 CFR 20.2002). The NRC staff considered this site-specific application for Vermont Yankee to have insignificant radiological impact because the proposed septic waste material disposal involved licensed material containing less than 0.1 percent of the radioactive material, primarily cobalt-60 and cesium-137, already considered acceptable in the Final Environmental Statement (FES) of July 1972, and involved exposure pathways much less significant than those in the FES. In addition, the proposed septic waste material disposal satisfied the following applicable boundary conditions for the disposal of licensed material:

a. The whole body dose to the hypothetical maximally exposed individual must be less than 1.0 mrem/year.

Revision _ll_ Date 08/14/97 F-5

                                             . I'. -

b. Doses to Lhe whole body and any organ of an inadvertent intruder from the probable pathways of exposure are less than 5 mrem/year.

c. ;The disposal musl be at the same site.

Following the NRC staff's approval on August 30, 1989, VYNPC implemented Lhe disposal of the contaminated septic waste material as proposed. By letter dated August 30, 1995, VYNPC requested that the previous authorization for the onsite disposal of very-low-level radioactive material be amended to permit the onsite disposal of slightly contaminated silt material, within the boundary conditions of the previously approved septic waste material disposal. 3.0 f:VALUArlON In its letter of August 30, 199S, VYNPC sLctteu ti1at ti1e proµosed sill. disposal method is the same as the previously approved septic waste disposal method. and utilizes land spreading in the same onsite areas approved for septic waste disposal. The volume of silt proposed for onsite disposal consists of 14,000 cubic feet (396 cubic meters) accumulated through August 1995 plus approximately 4,000 cubic feet (113 cubic meters) to be removed from the cooling towers during each 18-month operating cycle. The activity contained in the currently accumulated silt, based on samples taken by VYNPC in June 1995, is 0.193 millicuries, principally from 0.034 millicuries of cobalt-60 and 0.159 millicuries of cesium-137. The activity contained in the additional silt to be removed from the cooling towers each 18-month operating cycle is anticipated to be 0.059 millicuries, principally from 0.012 millicuries of cobalt-60 and 0.047 millicuries of cesium-137. VYNPC's radiological assessment enclosed with it~ August 30, 1995, letter demonstrates that the combined radiological impact for all onsite disposal operations, the proposed disposal of silt and the previously approved disposal of septic waste material, will continue to meet the applicable boundary conditions (given above) for the disposal of licensed material. Therefore, the proposed onsite disposal of slightly contaminated silt is acceptable . As discussed in VYNPC

s 1etter of August 2. 1996, i f the on site disposal of cooling tower silt or septic waste material would result in exceeding the applicable boundary conditions (given above). then VYNPC must obtain prior LAI NRC staff approval of the disposal. In addition, VYNPC made the following 1193 commitments:

1130

a. VYNPC wi 11 report in the Annual Radi ol ogi cal Effluent Rel ease Report a list of the radionuclides present and the total radioactivity associated with the onsite disposal activities at Vermont Yankee.

Revision _1l_ Date 08/14/97 F-6

Al b. VYNPC will maintain records of radionuclide concentrations and 1130 total activity associated with onsite disposal activities at Vermont Yankee in accordance with 10 CFR 50.75(g).

4.0 CONCLUSION

The NRC staff finds that the radiological conditions at the Vermont Yankee site (see attachment) that would result from the onsite disposal of slightly contaminated silt material. as proposed by VYNPC pursuant to 10 CFR 20.2002. and the previously approved onsite disposal of slightly contaminated septic waste material. are within the applicable boundary conditions (given above) for the disposal of licensed material. Therefore, the proposed onsite disposal of slightly contaminated silt removed from Vermont Yankee's cooling towers is acceptable. LAI VYNPC is required to permanently incorporate this SE into the Vermont Yankee 1130 Offsite Dose Calculation Manual as an Appendix to document the the radioactive material onsite disposal activities approved for Vermont Yankee. and VYNPC's related commitments regarding reporting and record keeping. Any additional modification of VYNPC's disposal activities which go beyond those proposed in the August 30, 1995, submittal, and are not addressed above must have prior NRC staff approval. In addition, any onsite disposal of cooling tower silt or septic waste material that would result in exceeding the applicable boundary conditions (given above). must also have prior NRC staff approval. Principal Contributors: J. Minns D. Dorman C. Harbuck Date: June 18. 199/

Attachment:

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VERMONT YANKEE RESPONSIBIL11Y NUCLEAR POWER CORPORATION F6rry Road, BranlebOro, VT 05301-7002 Rt,,_Y ro ENGi NEERING OFFICE 580 MAIN STREET OOLTON. MA0\740 (~Dal 77~711 August 30, 1995 BVY 95-97 United States Nuclear Regulatory Commission Washington, DC 20555 ATTN: Document Control Desk

References:

Ill Licr:nse Nil. DPR '.'8 IDl\Ckct. No. 50-271) ( 2) Letter from R. W. Capstick, Vermont Yankee, to USNRC. "Request to Routinely Dispose of Slightly Contaminated waste in Accordance with 10CFR20.302(a)". BVY 89-59, June 18, 1989. (3) Letter from M. B. Fairtile, USNRC, to L. A. Tremblay, Vermont Yankee, "Approval Under 10 CFR 20.302Ca) of Procedures for Disposal of Slightly Contaminated Septic Waste on Site at Vermont Yankee (TAC No. 73776)", dated August 30, 1989.

Subject:

Request to Amend Previous Approval Granted Under 10 CFR 20.302(a) for Disposal of Contaminated Septic Waste In accordance with the criteria of the Code of Federal Regulations, Title 10, Section 20.2002 (previously cited 10CFR20.302(a)l, enclosed please find the subject application to amend the previously granted approval (Reference 3) to dispose of slightly contaminated septic waste on site at Vermont Yankee by expanding the allowable waste stieam to include sli:;htly contaminated Cooling Towe;- silt mJterial. This application specifically requests approval to dispose of Cooling Tower silt deposits, contaminated at minimal levels, which have been or might be generated through the end of station operations at the Vermont Yankee Nuclear Power Plant. The proposed silt disposal method is the same as the septic waste disposal method requested in Reference 2 and approved in Reference 3. The disposal method utilizes on site land spreading in the same designated areas used for septic waste. Disposal of this waste in the manner proposed, rather than holding it for future disposal at a 10CFR Part 61 licensed facility when access to one becomes available, will save substantial costs and valuable disposal site space for waste of higher radioactivity levels. A radiological assessment and proposed operational controls based on continued on site disposal of accumulated river silt removed from the basins of the plant's mechanical draft cooling towers is contained in Enclosure A. The assessment demonstrates that the dose impact expected from the disposal of silt removed from the cooling towers during normal maintenance will not exceed the dose limits already imposed for septic waste disposal. The combined radiological impact for all on site disposal operations shall be limited to a total body or organ dose of a maximally exposed member of the public of less than one mrem/year during the period of active Vermont Yankee control of the site, or less than five mrem/year to an inadvertent intruder after termination of active site control. Enclosure B contains a copy of the original assessment and disposal procedures for Revision _fl_ Date 08/14/97 F-9

VERMONT YANKEE NUCLEAR POWER CORPORATION sepl\c waste (References 2 and 3) for your use and reference in evaluating the proposed amendment. Upon receipt of your approval, Enclosure A will be incorporated into the Vermont Yankee ODCM. LAI \We trust .that the information contained in the submittal \s sufficient. however, should 10~~ you have any questions or require further information concerning this matter, please Closed contact this office. Sincerely, VERMONT YANKEE NUCLEAR POWER CORPORATION James J. Duffy Licensing Engineer Enclosures A & B c: USNRC Region I Administrator (Letter and Enclosure Al USNRC Resident Inspector

VYNPS (Letter and Enclosure Al USNRC Project Manager
VYNPS (Letter and Enclosure Al Revision _11_ Date 08/14/97 F-10

ENCLOSURE A VERMONT YANKEE NUCLEAR POWER PLANT ASSESSMENT OF ROUTINE DISPOSAL OF COOLING TOWER SILT IN AREAS ON SITE PREVIOUSLY DESIGNATED IOR SEPTIC WASTE DISPOSAL Revision -11__ Date 08/14/97 F-11

Table of Contents Page

1.0 INTRODUCTION

  ..... .                         3 2.0   WASTE STREAM DESCRIPTION                        4 3.0   DISPOSAL METHOD . .                             9 3.1    Silt Disposal Procedure Requirements    9 3.2    Administrative Procedure Requirements   12 4.0    EVALUATION OF ENVIRONMENTAL IMPACTS            13
      ~-1    ~ite  Characteristics                   13 4.2    Radiological Impact                     13 5.0     RADIATION PROTECTION                          23

6.0 CONCLUSION

S 23

7.0 REFERENCES

24 2 Revision -1l._ Date 08/14/97 F-12

VERMONT YANKEE NUCLEAR POWER PLANT Assessment of Routine Disposal of Cooling rower Silt in Areas On-Site Previously Designated for Septic Waste Disposal

1.0 INTRODUCTION

Vermont Yankee Nuclear Power Corporation (Vermont Yankee) requested from the NRC in 1989 permission to routinely dispose of slightly contaminated septic waste in designated on-iite areas in accordance with 10CFR20.302(a). The NRC responded to this request on August 30, 1989 by granting approval of the proposed procedures for on-site disposal of septic waste concluding that the commitments. as documented in our request, were acceptable provided that our request and analysis b~ permanently incorporated into the plant's Offsite Dose Calculation Manual (ODCM). Revision 9 to the ODCM (Appendix Bl incorporated the assessment and approval of the methods utilized for on-site disposal of slightly contaminated sewage sludge. In addition to the previously identified solids content of septic waste as a source of environmental. low level radioactive contaminated material. cooling tower silt deposits resulting from the settling of solids from river water passing through the mechanical draft cooling tower system have been identified to also contain low levels of plant-specific radionuclides. Periodic removal of the silt from the cooling tower basins is a necessary maintenance practice to insure operability of the cooling system. However, due to the presence of by-prnduct mat~rials in the silt. proper dispos~l requirements must be ~pplied to insure that the potential radiological impact is within acceptable limits. This assessment of silt disposal expands the original septic waste disposal assessment to include earthen type materials (cooling tower silt deposits) while maintaining the original radiological assessment modeling and dose limit criteria that have been approved for septic waste spreading on site. This assessment demonstrates that cooling tower silt can be disposed of in the same manner and under the same dose limit criteria as previously approved for septic waste in Appendix B to the Vermont Yankee ODCM. Implementation of the following commitments as an amendment to the original 10 CFR 3 Revision _f.l_ Date 08/14/97 F-13

Part 20.302(a) approval for septic waste shall be incorporated into the Vermont Yankee ODCM upon approval by the NRC. 2.0 WASTE STREAM DESCRIPTION: The waste involved in this assessment is residual solids (silt) collected in the basins of the plant's mechanical draft cooling tower system. The silt consists of organic and inorganic sediments and earthen type materials that have settled from the cooling water flow taken from the Connecticut River as it passes through the towers. As a result of de-sludging the tower basins in 1993. an estimated 14,000 cubic feet of silt was accumulated on site. Clean-out operations will also occur periodically to ensure continued system operability. Sample analysis performed to the plant's environmental lower limits of detection requirements, as contained in Technical Specification Table 4.9.3., has identified Cobalt-60 and Cesium-137 in low concentrations as being present in silt collected in 1993. The cooling towers are located at the southern end of the plant facility complex but are not directly connected to any system in the plant that contains radioactivity. The postulated mechanism of how plant-related radionuclides have been introduced into the cooling system silt assumes that past routine effluents discharged from nearby plant gaseous release points were entrained in the large mechanically-induced air flow that is pulled through the towers as a heat exchange medium. The cooling water flow provides a scrubbing action as it is breaks up into fine water droplets due to the splash pans of the towers. This scrubbing action washes any airborne particulates out of the air. Over long periods of operation. any radioactivity removed from the air flow could buildup to measurable levels in silt that settles out in the basins at the bottom of the towers. Table 1 lists the analyses of twenty-one samples collected from the silt pile removed from the cooling tower basins. Radioactivity measurements. averaged over all_ the samples, indicate that the silt material can be characterized as containing app~oximately 50 pCi/kg (dry wt.) of Cobalt-60 and 198 pCi/kg (dry wt.) of Cesium-137. Eight of the samples indicated no positive Cobalt-60 above a minimum detectable level achieved for the analysis. 4 Revision _g_}_ Date 08/14/97 F-14

fable 1 Cooling Tower Silt Radioactivity (1993 sarnplesA) Sample II Co-60 Cs-137 (pCi/kg dry) (pCi/kg dry) Gl2759 53 144 G127 58 72 172 Gl2757 <14 201 G12756 <17 245 G12755 73 206 Gl2754 <16 165 G12753 <27 240 G12752 79 181 Gl2751 <29 180 G12750 35 107 G12749 59 171 G12748 <19 205 G12747 <38 209 G12746 < 7 218 Gl2745 50 241 Gl2744 40 220 G12743 68 264 G12742 45 195 Gl2724 71 115 Gl2723 104 264 G13940 126 218 Average: 50 198 Max. 126 264 Min. <7 107 Standard deviation: 30 42

Average wet to dry sample weight ratio equal to 1.6. Dry weight silt density equal to 1.3 gm/cc.

5 Revision -11._ Date 08/14/97 F-15

For the purpose of estimating the total activity in the silt pile, the less than values in Table l are included as positives in the calculation of the average radioactivity concentration. Cobalt-60, due to its relatively short half life, is typically associated with plant operations when measured in the near environment. However. Cesium-137 when measured in the environment may have a background component that is not related to power plant operations. Past weapons testing fallout has imposed a man made background level of Cesium-137 in New England soils and sediments that can vary over several hundred pCi/kg. The plant's Environmental Monitoring Program has shown that Connecticut River sediment in the vicinity of Vermont Yankee averages about 123 pCi/kg (dry wt.) of Cesium-137 (Table 2) with no plant related detectable level of Cobalt-60. The value of 123 pCi/kg may represent an estimate of background level of Cesium-137 in sediment that would be subject to entrainment in cooling water flow that enters the plant. In comparing the measured levels of Cesium-137 on Table 1 with the past river sediment level, the average concentration in the cooling tower silt is higher than that of the river sediment data but does fall within the observed range of recorded sediment Cs-137. The river sediment Cesium-137 concentration averages about 62% of the concentration value detected in the tower silt. For purposes of this assessment of plant-related dose impact from the on-site disposal of silt material, it is conservatively assumed that all detectable Cs-137 in cooling tower silt is directly related to plant operations. No background component is subtracted from the measured values for this case study since only a single sampling location (down stream) is included in the E1vironmental Monitoring Progr2m which may not fully describe the true background levels in the region. The total radioactivity for the current 14,000 cubic feet of silt collected on site can be estimated by multiplying this volume by its "as is" density of 2.1 gm/cc Ci .e. 1.3 gm/cc dry weight density x 1.6 wet/dry weight ratio) and then conservatively assume that the measured average dry weight radioactivity concentrations for Cobalt and Cesium would be the same as in the collected silt. Multiplying the average Cobalt-60 and Cesium-137 concentration in silt by the mass of the collected material produces estimates (Table 3) of total radioactivity that was removed from the cooling tower basin in September 1993. 6 Revision _l1_ Date 08/14/97 F-16

Table 2 Cesium-137 in Connecticut River Sediments* Cs-137 Date (pCi/kg dry) 05/24/94 61 10/13/93 85 06/02/93 60 10/15/92 137 05/20/92 176 10/24/91 178 05/16/91 230 10/25/90 84 05/16/90 62 10/04/89 <174 05/26/89 179 10112/88 115 05/12/88 62 Average: 132 Max. 230 Min. 60 Standard deviation: 56

Samples collected as part of the Vermont Yankee Radiological Environmental Monitoring Program CREMP) for river sediment sample location SE-11.

7 Revision 21 Date 08/14/97 F-17

Table 3 Estimated Total Radioactive Material for 1993 Tower Clean-Out Volume of Mass Average Total Act. Decayed Act. Silt Concentration (as of 11/93) (as of 6/95) (ft3) (kg) (pCi/kg) (uCi) ( uC i l Co-60 14,000 8.32E+5 50 42 34 Cs-137 14,000 8.32E+5 198 165 159 In addition to 14.000 cubic feet of silt already accumulated, it is anticipated U1at periouic maintenance work in cleaning out the cooling tower basins will generate approximately 4,000 cubic feet of new silt material over each succe~sive 18 month operating cycle. Assuming the same level of plant-related radioactivity concentration that was originally observed, the additional amounts of radioactivity that will require on site disposal following each refueling cycle can also be estimated. Table 4 lists an estimate of the total radioactivity that might be present at each 18 month clean-out cycle. Table 4 lstimated Total Radioactive Material for Each 18 Month Maintenance Cycle Volume of Mass Average Total Silt Concentration Activity ( ft3) (kg) (pCi/kg) ( uC i ) Co-60 4,000 2.38E+5 50 12 Cs-137 4,000 2.38E+5 198 57 8 Revision _fl_ Date 08/14/97 F-18

3.0 DISPOSAL METHOD: The method of silt disposal shall uLilize a technique of land spreading in a manner consistent with the current commitments for the on-site disposal of septic waste as approved by the NRC and implemer1ted as Appendix 8 of the Vermont Yankee ODCM (Reference 1). The same land areas designated and approved for septic waste disposal shall be used for the placement of silt removed from the cooling tower basins. Determination of the radiological dose impact shall also be made based on the same models and pathway assumptions as indicated in Appendix 8 of the ODCM. 3.1 Silt Disposal Procedure Requirements: Gamma isotopic analysis of silt samples shall be made prior to earh disposal by obtaining representative composite samples in sufficient numbers to characterize the material removed from the cooling tower basins. Each gamma isotopic analysis shall be required to achieve the environmental lower limits of detection as indicated for sediment on fable 4.9.3 of the Vermont Yankee Technical Specifications. The estimation of total radioactivity to be disposed of shall be made based on the average of all composite sample analyses. The estimation of total radioactivity and projected dose impact shall be made prior to placing the collected silt on the designated disposal plots. The dose impact from each disposal operation shall be included with all past septic waste and silt spreading operatio~s to ensure that the appropriate dose limits ar2 r.ot exceeded on any waste disposal area for the combination of all past operations. The established dose criteria requires that all applications of earthen type materials within the approved designated disposal areas shall be limited to ensure that dose to a maximally exposed individual (during the Vermont Yankee control period) be maintained less than 1 mrem/year to the whole body and any organ, and the dose to an inadvertent intruder following termination of site control be maintained less than 5 mrem/year to the whole body and any organ. 9 Revision _11.__ Date 08/14/97 F-19

The limits on concentrations of radionucl ides as addressed in Appendix B to the ODCM for septic waste (i.e., each tank of septic sludge to be disposed are limited to a combined MPC ratio of less than 0.1) were included to ensure proper control was in place to address the situation of small quantities of relatively high concentration material. This limitation does not directly apply to silt deposits since the silt is handled as dewatered sediments as opposed to liquid slurries of septic waste. For dry, earthen type material such as silt, a specific radionuclide concentration limit shall be applied in place of the septic waste liquid MPC ratio. No soil associated with a sample analysis that identifies a plant-related radionuclide in excess of the concentration limits of Table 5 will be permitted regardless of the tolal pathway dose assessment determined for the cp1antit.y maLer*i,11 under cnnsidcriltinn. For th7: case 1t1h~re more than one radionuclide is detected, the sum of the ratio rule will be applied. The measured concentration of each radionuclide divided by its limiting concentration value shall be added with the sum of all fractions equal to or less than 1. This limiting condition will prevent small volumes of relatively high specific radioactivity from being spread on the disposal plots, and therefore reduce the potential for creating unexpected hot spots of concentrated material. Table 5 lists, by radionuclide, soil concentration values that would generate an annual external effective dose equivalent of 25 mrem/year if it were assumed that an individual continuously stood on an infinite plane of soil contaminated to a depth of 15 cm. The essumptions of a~ infinite plane and continuous occupancy are conservative for situations where the amount of contaminated soil identified would not provide for a 15 cm soil depth over an extended surface area and where disposal site access is limited. Twenty-five mrem/year was selected as a reference value based on the fact that it was a suitable fraction of the NRC annual dose limit (100 mrem/year per 10 CFR Part 20.1301) applied to members of the public from all station sources. The 25 mrem/year also equals the EPA dose limit from 40 CFR Part 190 which would apply to real members of the public offsite and allow for credit to be taken in accounting for actual usage patterns such as occupancy time. The external dose factors provided on Table 5 were derived from Table E-2 of NUREG/CR-5512 (Reference 3). 10 Revision -1.!_ Date 08/14/97 F-20

Table 5 Dry Soil Maximum Concentration Values Soil Concentration pCi/kg Radionuclide (equal to 25 mrem/yr) Cr-51 1. 51E+05 Mn-54 5.50E+03 Fe-59 3.83E+03 Co-58 4.70E+03 Co-60 l.82E+03 Zn- 6'i 7.3SE+03 Zr-95 6.18E+03 Ag - 11 Om l.66E+03 Sb-124 2.51E+03 Cs-134 2.95E+03 Cs-137 8.13E+03 Ce-141 7.85E+04 Ce-144 8.75E+04 Assumptions include infinite planar distribution, uniform depth distribution to 15 cm, soil density at 1.625E+06 gm/m3 and external direct dose pathway only with a 100% occupancy factor. 11 Revision _fl..._ Date 08/14/97 F-21

3.2 Administrative Procedure Requirements: Dry silt material shall be dispersed using typical agriculturul dry bulk surface spreading practices only in upproved disposal areas on site. Complete records of each disposal will be maintained. These records will include the concentration of radionuclides detected in the silt, an estimate of the total volume of silt disposed of, the total radioactivity in euch disposal operation as wel.l as the total accumulated on each disposal plot at the time of the spreading, the plot on which the silt was applied, and the results of any dose calculations or maximum allowable accumulated activity determinations required to demonstrate that the dose limits imposed on these land spreading operations have not been exceeded. The determination of the tota1 r*cHlioac~ivily and d0'.:ie cu~culJtioils shull aiso inc..lude ull pasL seµtic wuste und silt disposal operations that placed low level radioactive material on the designated disposal plots. The periodic disposal of silt on each of the approved land spreading areas will be limited to within the same established dose and radioactivity criteria that have been approved for septic waste disposal. Concentration limits that are applied to the disposal of earthen type materials (dry soil) shall restrict the placement of small volumes of material that have relatively high concentrations of radioactivity such that direct exposure could not exceed a small proportion (25%) of the annual dose limits to members of the pui;l ic that is contilined in 10 CFR Part 20.1301. Any farmer leasing land used for the disposal of silt deposits will be notified of the applicable restrictions placed on the site due to the land spreading of low level contaminated material. These restrictions are the same as detailed for septic waste spreading as given in Reference 1. 12 Revision _f.l_ Date 08/14/97 F-22

4.0 EVALUATION OF ENVIRONMENTAL IMPACTS: 4.1 Site Churacteristics The designated disposal sites consist of two fields located on the Vermont Yankee Nuclear Power Plant site. Both fields are on the plant property within the site boundary security fence. Site A contains an approximate ten-acre parcel of land centered approximately 2,000 feet northwest of the Reactor Building. Site B consist of approximately two acres and is centered approximately 1,500 feet south of the Reactor Building. These are the same land parcels approved by the NRC for the land disposal of septic waste and are described in detail in Reference 1 along with the boundary restrictions for the placement of contaminated material. Radiological assessments of septic waste disposal have determined that a single two-acre plot would be sufficient for the routine disposal of that waste stream over a 20-year period without exceeding the dose criteria to a maximum exposed individual or inadvertent intruder. As a result, the ten-acre field to the northwest can be divided into five disposal plots, with the two-acre site at the south end of the plant site providing a sixth plot. It is therefore concluded that there is sufficient space within the already approved disposal plots to accommodate additional material from the cooling tower basins along with the septic waste without the likelihood of exceeding the approved dose limit criteria. Since the residual organic and inorganic solids associated with river sediment (silt) are similar to the sand and residual organic material remaining after decomposition of septic waste that is removed from the plant's septic tanks, the conclusions of no significant environmental Cn~n-radiological) impact associated with the disposal of septic waste are not changed by the addition of another earthen type material, namely silt. 4.2 Radiological Impact: The amount of cooling tower silt, in combination with any septic waste disposals, will be procedurally controlled to insure doses are maintained within the prior approved limits (Reference 1). These limits are based on 13 Revision __gi__ Date 08/14/97 F-23

past NRC proposed guidance (described in AIF/NESP-037), August 1986). The dose criteria require that the maximally exposed member of the general public receive a dose less than 1 mrem/year to the whole body or any organ due to the disposed material and less than 5 mrem/year to the whole body or any organ of an inadvertent intruder. To assess the doses received by the maximally exposed individual and inadvertent intruder resulting from silt spreading, the same pathway modeling, assumptions and dose calculation methods as approved for septic disposal are used. These dose models implement the methodologies and dose conversion factors as provided in Regulatory Guide 1.109 (Reference 2). Six potential pathways have been identified and include: (a) Standing on contaminated ground. (b) Inhalation of resuspended radioactivity, (c) Ingestion of leafy vegetables. (d) Ingestion of stored vegetables, (f) Ingestion of meat, and (g) Ingestion of cow's milk Based on the septic waste evaluations. the liquid pathway was determined to be insignificant. Both the maximum individual and inadvertent intruder are assumed to be exposed to these pathways with the difference between them related to occupancy time. The basic assumptions used in the radiological analyses include: (a) Exposure to ground contamination and resuspended radioactivity is for a period of 104 hours per year during the Vermont Yankee active control of the disposal sites and continuous thereafter. The 104-ha*ur interval is representative of a farmer's time spent on a plot of land (4 hours per week for 6 months). (b) For the purpose of projecting and illustrating the magnitude of dose impacts over the remaining life of the plant. it is assumed that the current concentration levels of activity detected in silt remain 14 Revision _f.l_ Date 08/14/97 F-24

constant. Table 1 indicates the measured radioactivity levels for Cobalt-60 and Cesium-137 first noted in silt material. (c) The maximum radiation source corresponds to Lhe accumulation of radioactive material on a single plot (two acre) within the approved disposal sites over a period of 13 operating cycles. This extends over the next 18 years until after the operating license expires in 2012. The initial application (referenced to June 1995) consists of 14,000 cubic feet of silt collected in 1993 along with the first periodic clean out of the tower basins that adds an additional 4,000 cubic feet. All subsequent applications of 4,000 cubic feet occur at 18-month intervals. (dl For the analysis of the radiological impact during the Vermont Yankee active control of the disposal sites, no plowing is assumed to take place and all dispersed radioactive material remains on the surface forming a source of unshielded direct radiation. (e) No radioactive material is dispersed directly on crops for human or animal consumption. Crop contamination is only through root uptake. (f) The deposition on crops of suspended radioactivity is insignificantly small. (g) Pathway data and usage factors used in the analysis are the same as those used in the plant's OOCM assessment of off-site radiological impact from routine releases with the exception that the fraction of stored vegetables grown on the contaminated land was conservatively increased from 0.76 to 1.0 (at present no vegetable crops for human consumption are grown on any of the approved disposal plots). (h) It is conservatively assumed that Vermont Yankee relinquishes control of the disposal sites after the operating license expires in 2012 (i.e., the source term accumulated on a single. 2-acre disposal plot applies also for the inadvertent intruder). 15 Revision _fl_ Date 08/14/97 F-25

( i) For the analysis of Lhe impact after Vermont Yankee control of the site is relinquished. the radioactive material is plowed under and Forms a uniform mix with the top six inches of soil; but nonetheless. undergoes resuspension in air at the same rate as the unplowed surfate contaminalion. Fo~ direct ground plane exposure the self shielding due to the six-inch plow layer reduces lhe surface dose rate by about a factor of four. The dose models and methods used to generate deposition values and accumulated activity over the operating life of the plant are documented in Attachment 2 to Reference 1. Based on the measured concentrations and silt volumes noted in Section 2.0 above, the total radioactivity that remains on the disposal plots after the operating license expires is estimated on Table 6. Table 6 Projected Radioactivity Buildup Due to Silt Spreading Nuclide Contribution Accu~ulation from Total Remaining from initial 13 cycles at in year 2013 14,000 ft3 (uCi) 4,000 ft3/ea. (uCi) ( uC i ) Cobult-60 3.2 61. 9 65.l Cesium-137 104.9 500.5 605.4 The calculated potential radiation exposure following the spreading of all silt material anticipated to be generated through the remainder of the operating license on a single, two-acre plot is provided on Table 7. 16 Revision _fl_ Date 08/14/97 F-26

Table 7 Dose Impact Due to Continued Spreading to End of License Disposill Site Access Radiation Exposure Individual/Organ Controlled by VYNPS 0.228 mrem/yr adult/whole body (max. exposed individual) 0.820 mrem/yr max. child/bone Uncontro 11 ed by VYNPS 1.46 mrem/yr adu 1t/who le body (inadvertent intruder after 2. '11 mrem/yr max. child/bone license termination) The individual pathway contributions to the total dose due to continued silt spreading are shown on Table 8. Table 8 Pathway-Dependent Critical Organ Doses Maximally exposed Inadvertent Intruder Individual/Organ Critical Individual/Organ (Child/Bone) (Chi l ct/Bone) Pathway (mrem/year) Cmrem/year) Ground Irradiation 0.0474 0.957 Inhalation 0. 00814 0.685 Stored Vegetables 0.528 0.528 Leafy Vegetables 0.0265 0.0265 Milk Ingestion 0.201 0.201 Meat Ingestion 0.00833 0.00833 Total: 0.82 2.41 17 Revision _11_ Date 08/14/97 F-27

In addition, the isotopic breakdown of the critical organ doses listed above (fable 8) for the two detected radionuclides is seen to be: Table 9 Isotopic Breakdown of Maximum Radiation Exposures Radioactivity Dose Percent of Description Isotope (uCi/2 acres) (mrem/yr) total During Cs-137 605.4 0.805 98.2 control of Co-60 65.l 0.0144 1. 8 disposal 0.82 sites Max. organ: child/bone Termination Cs-137 605.4 2. 12 88.0 of disposal Co-60 65.1 0.29 12.0 site Max. 2.41 organ: child/bone For comparison to the total dose calculated assuming the continued disposal of silt removed from the tower basins through the end of the operating license. the dose from just the original 14,000 cubic feet collected is shown on Table 10. 18 Revision_£!_ Date 08/14/97 F-28

Tdllle 10 Dose Impact Due to Simile (14,000 ft3) Silt Spreading Disposal Site Access Radiation Exposure Individual/Organ Controlled by VYNPS ( rna x. 0.064 mrern/yr adult/whole body exposed individual in 1995) 0.219 rnrern/ yr max. child/bone Uncontrolled by VYNPS 0.224 rnrem/yr adult/whole body (inadvertent intruder a ft er 0. 39,3 mrem/yr max. child/bone license termination) Table 10 shows that the application ot the siit material inilially collected (14,000 cubic feet) accounts for about 27 percent of the maximum individual organ dose during the control period as compared to the scenario of continued periodic silt spreading over the balance of the operating license. This illustrates that dose impacts from the material currently collected are well below the acceptance criteria of limiting the dose from any two-acre plot to no more than 1 mrem/year during the control period and 5 mrem/year after termination of the license, and is expected to remain below the acceptance criteria throughout the plant life. If unexpected buildup of radioactivity in future silt clean-out operations were to occur, the option for use of alternate disposal plots remains available to ensure that the impact from any single, two-acre plot stays within the acceptance criteria. Also of interest are derived dose conversion factors which provide a means of ensuring that septic and silt disposal operations rema~n within the prescribed radiological guidelines noted above. The critical organ (worst case) and whole body dose factors for all pathways on a per acre bases are 19 Revision -11.__ Date 08/14/97 F-29

given on Table 11 for periods during Vermont Yankee control of the disposal site and on fable 12 for post control periods associated with the lnadvertent intruder scenario. The dose conversion factors have been expanded to include other potential radionuclide beyond the original five that were addressed in Reference 1. This provides a means to assess other nuclides if future disposal operations identify additional radionuclides not previously observed. The development of these additional nuclide dose conversion factors utilize the same modeling and pathway assumptions as used to derive the factors for the original flve radionuclides identified in septic waste. The models for these site and pathway-specific dose factors are those in Regulatory Guide 1.109 (Reference 2) and are described in detail in Attachmerit II to Reference 1. 20 Revision ...11_ Date 08/14/97 F-30

Table 11 All-Pathway Critical Organ I Whole Body lJose Conversion Factors During Vermont Yankee Control of Disposal Sites Critical Organ Whole Body Dose Dose Factor Factor Nuclide Inclividual/Organ (mrem/yr per uCi/acre) Cr-51 Teen/Lung l.14E-05 5.76E-06 Mn-51\ Adult/GI - LLI 3.75E-04 l .93E-04 Fe-55 Child/Bone 6.45E-06 l.06E-06 Fe-5SI Teen/Lung 4.GlE-04 2.13E-04 Co-58 Teen/Lung 3.27E-04 2.0lE-04 Co-60 Teen/Lung 7.17E-04 5.31E-04 Zn-65 Child/Liver l.64E-02 l.03E-02 Zr-95 Teen/Lung 4.47E-04 l.34E-04 AgllOm Teen/GI - LLI l.32E-02 5.24E-04 Sb-124 Teen/Lung 8.34E-04 3.54E-04 Cs-134 Child/Liver 3.18E-03 l.28E-03 Cs-137 Child/Bone 2.66E-03 7.02E-04 Ce-141 Teen/Lung l.54E-04 l.50E-05 Ce-144 Teen/Lung 6.00E-04 2.44E-05 21 Revision _ll_ Date 08/14/97 F-31

Table 12 All-Pathway Critical Organ I Whole Body Dose Conversion Factors Post Vermont. Yankee Control of Disposal Sites (Inadvertent Intruder) Critical Organ \oJhol e Body Dose Dose Factor Factor Nuclide Individual/Organ (mrem/yr per uCi/acre) Cr-51 Teen/Lung 5.89E-04 l.19E-04 Mn-54 Teen/Lung l.02E-02 3.12E-03 Fe-55 Teen/Lung 3.50E-04 2.27E-05 Fe-59 Teen/Lunq l:.~~l-02 1'i .*DE-03 Co-58 Teen/Lung l .59E-02 3.72E-03 Co-60 Teen/Lung 3.19E-02 9.09E-03 Zn-65 Child/Liver l.89E-02 l.25E-02 Zr-95 Teen/Lung 2.93E-02 2.99E-03 Agl lOm Teen/Lung 3.59E-02 9.53E-03 Sb-124 Teen/Lung 4.73E-02 7.04E-03 Cs-134 Child/Liver l.21E-02 9.36E-03 Cs-137 Child/Bone 6.98E-03 3.85E-03 Ce-141 Teen/Lung l.21E-02 3.44E-04 Ce-144 Teen/Lung 5.00E-02 1.52E-03 22 Revision -1..!_ Date 08/14/97 F-32

5.0 RADIATION PROTECTION: The disposal operation of silt material from the cooling tower basins will follow the applicable Vermont Yankee procedures to maintain doses as low as reasonably achievable and within the specific dose criteria as previously approved for septic waste disposal (Reference 1).

6.0 CONCLUSION

S: Silt collected from the cooling tower basins is an earthen type material that is similar in characteristics to septic waste residual solids with respect to the radiological pathway behavior and modeling and can be disposed of through on-site land spreading on the same disposal plots as previously evaluated and approved for septic waste disposal. Th~ radiolngiral assessment of low level contaminated silt shows that the projected dose from the on-site periodic spreading of this material will have no significant dose impact to members of the public and can be maintained below the approved dose limitations already in place for septic waste. 23 Revision _fl_ Date 08/14/97 F-33

7.0 REFERENCES

(1) Vermont Yankee ODCM. Appendix B: "Approval of Criteria for Disposal of Slightly Conta1ninated Septic Water On-Site at Vermont Yankee". (Included NRC approval letter dated August 30, 1989, VY request for approval dated June 28, 1989 with Attachments I and II). (2) USNRC Regulatory Guide 1.109, Rev. 1; "Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix !", dated October 1977. (3) NUREG/CR-5512. Vol. l, "Resldual Radioactive Contamination From Decommissioning", Final Report, dated October 1992. file: vsilt.mss 24 Revision _11_ Date 08/14/97 F-34

APPENDIX G MAXIMUM PERMISSIBLE CONCENTRATIONS (MPCs) IN AIR AND WATER ABOVE NATURAL BACKGROUND TAKEN FROM 10CFR20.l TO 20.602. APPENDIX B Revision -1.l_ Date 08/14/97 G-1

APPENDIX G With the implementation of the revised Part 20 to Title 10 of the Code of Federal Regulations (10CFR20.1001-20.2401), the Maximum Permissible Concentrations (MPCs) that were part of the old 10CFR20 were replaced by a new Appendix B to 10CFR20 for the limits that apply to effluents released to unrestricted areas. However, MPC values were also used and accepted as licensing conditions for the control of radioactive materials in situations other than those directly covered by the requirements of the regulations. One example is the on-site disposal of septic waste which used the MPC values as one criteria of acceptability for land spreading. Appendix B to the ODCM references 10CFR20.1-20.601 Appendix B MPCs as concentration criteria for this disposal option. With the final publication of the revised 10CFR20.1001-20.2401, the original MPC tables of the old 10CFR20 are no longer in print. As such, this appendix to the ODCM is added to provide a reference source for the MPC values contained in the original Appendix B to 10CFR20.1-20.601 for those conditions that still refer to these requirements. Revision ..1.1._ Date 08/14/97 G-2

APPENDIX G MAXIMUM PERMISSIBLE CONCENTRATIONS (MPCs) (FROM 10CFR20.l TO 20.602, APPENDIX 8) APPENDIX B TO §§20.1 - 20.602 CONCENTRATIONS IN AIR AND WATER ABOVE NATURAL BACKGROUND [See footnotes at end of Appendix BJ Table I Table II Col. 1 Col. 2 Col. 1 Col. 2 Air Water A1r Water Element (atomic number) I sotope 1 (µCi /ml) CµCi /ml) (µCi/ml) CµCi /ml) Actinium (89) Ac 227 s 2x10- 12 6x10-s Bxio- 14 2xl0- 6 I 3x10- 11 9xl0- 3 9x10- 13 3x10- 4 Ac 228 s Bxl0- 8 3x10* 3 3x10- 9 9x10-s I 2x10-s 3x10* 3 6x10-lO 9xlo-s Americ.ium (95) Am 241 s 6x10* 12 lxlo- 4 2x10- 13 4xlo- 6 I lxlO*lO Bx10- 4 4x10- 12 3xl0- 5 Am 242m s 6x10- 12 1x10- 4 2xl0- 13 4x10- 6 I 3xlO*lO 3x10- 3 9x10- 12 9x10*S Am 242 s 4xl0- 8 4x10- 3 1x10- 9 1x10- 4 I Sxl0- 8 4x10- 3 2x10- 9 lxl0- 4 Am 243 s 6xl0-lZ lx10- 4 2x10- 13 4x10- 6 I lxlO-lO 0x10- 4 4x10- 12 3x10- 5 Am 244 s 4xlo- 6 lxl0- 1 lxlo- 7 sx10- 3 I 2x10- 5 lxlo- 1 8x10- 7 sx10- 3 Antimony Sb 122 s 2x10- 1 8x10- 4 6x10- 9 3x10- 5 I lxl0- 7 Bxl0- 4 Sxl0- 9 3x10* 5 Sb 124 s 2x10- 7 7x10- 4 sxio- 9 2x10* 5 I 2x10-s 7xl0- 4 7x10- 10 2xl0-s Sb 125 s sx10* 1 3xl0- 3 2x10- 8 lxl0- 4 I 3x10-s . 3x10- 3 9x10- 10 lxl0- 4 Argon (18) . A37 Sub 2 6x10- 3 . ' ...... lxl0- 4 . ....... A41 Sub 2xl0- 6 I I I I I I I I 4xio- 8 ........ Arsenic (33) .. As 73 s 2xl0- 6 4x10- 7 1x10- 2 1x10- 2 7xl0" 8 lxl0- 8 Sxl0- 4 Sxl0- 4 I As 74 s 3x10- 1 2x10- 3 lxlo- 8 sx10- 5 I 1x10- 7 2x10- 3 4x10- 9 sx10- 5 As 76 s lxl0- 7 6x10- 4 4x10- 9 2x10-s I lxl0- 7 6x10- 4 3x10- 9 2x10-s As 77 s Sxl0- 7 2x10- 3 2x10- 8 Bxl0- 5 I 4x10- 7 2x10- 3 lxl0- 8 Bxl0- 5 Revision _£1_ Date 08/14/97 G-3

APPENDIX B TO §§20.l - 20.602 CONCENTRATIONS IN AIR AND WATER ABOVE NATURAL BACKGROUND [See footnotes at end of Appendix BJ Table I Table I I Col. 1 Col. 2 Col. 1 Col. 2 Afr Water Afr Water Element (atomf c number) lsotope 1 (µCf/ml) (µCf/ml) CµCi/mll (µCf/ml) Astat1ne (BS) At 211 s 7x10- 9 SxlO-s 2x10* 10 2x10* 6 I 3xlo-s 2x10* 3 lx10* 9 7x10-s Barium (56) Ba 131 s lxl0- 6 Sx10* 3 4xlo-s 2x10* 4 I 4xl0- 7 5x10* 3 lxl0" 8 2x10* 4 Ba 140 s 1x10* 7 8xl0- 4 4xl0" 9 3x10- 5 I 4xl0-s 7x10- 4 lxl0- 9 2xl0- 5 Berkelium (97) Bk. i.'.49 5 9xiU *to 2x10* 2 3x10* 11 Gx10* 4 I 1x10* 7 2x10* 2 4x10* 9 6xl0- 4 Bk 250 s lxl0- 7 Gxl0- 3 5x10" 9 2x10* 4 I lxl0- 6 6x10* 3 4x10* 8 2x10* 4 Beryllium ( 4) Be 7 s 6x10* 6 5x10* 2 2x10* 7 2x10* 3 I lxl0" 6 sx10* 2 4xio-s 2x10* 3 Bf smuth (83) . Bi 206 s 2x10* 7 1x10* 3 6x10* 9 4x10- 5 I 1x10* 7 1x10* 3 5x10* 9 4x10* 5 Bf 207 s 2x10* 7 2x10* 3 6x10" 9 6xl0" 5 I lxl0" 8 2x10* 3 sx10* 10 6xl0" 5 Bi 210 s 6x10* 9 lxl0- 3 2x10* 10 4x10* 5 I 6x10* 9 lxl0- 3 2x10* 10 4x10" 5 ai 2l2 s 1x10* 7 lx10* 2 3xl0" 9 4x10- 4 I 2x10* 1 lx10* 2 7xl0" 9 4x10* 4 Bromine (35) . Br 82 s lx10* 6 ax10* 3 4x10* 8 3xl0- 4 I 2x10* 1 lx10* 3 6xl0- 9 4x10- 5 Cadmium (48) . Cd 109 s 5x10-B sx10* 3 2x10* 9 2x10* 4 I 7x10" 8 sx10* 3 3x10" 9 2x10* 4 Cd 115m s 4x10" 8 7xl0- 4 lx10* 9 3x10* 5 I 4xl0- 8 7x10* 4 lxl0- 9 3xl0- 5 Cd 115 s 2x10* 7 lxl0" 3 Bxl0- 9 3xl0" 5 I 2x10* 7 lxl0- 3 6x10* 9 4x10" 5 Ca 45 s 3xl0" 8 3x10* 4 lx10* 9 9xl0- 6 Calcium (20) - I lx10* 7 5x10- 3 4x10* 9 2x10* 4 Ca 47 s 2xio* 1 lx10* 3 6x10* 9 sx10* 5 I 2x10* 1 lxl0- 3 6x10* 9 3x10* 5 Revision~ Date 08/14/97 G-4

APPENDIX B TO §§20.1 - 20.602 CONCENTRATIONS IN AIR AND WATER ABOVE NATURAL BACKGROUND [See footnotes at end of Append1x BJ Table I Table II Col. 1 Col. 2 Col. 1 Col. 2 Air Water A1r Water Element (atomic number) Isotope 1 (µC1/ml) (µCi/ml) CµC1/ml) (µCi/ml) Californium (98) Cf 249 s 2x10* 12 lx10* 4 5x10*l 4 4xl0" 6 I lxlO*IO ?x10* 4 3x10* 12 2Xl0-S Cf 250 s 5x10" 12 4x10* 4 2x10* 13 lxlO*S I lxlO*lO 7x10* 4 3xlO*lZ 3x10*S Cf 251 s 2xlO*IZ lx10* 4 6x10* 14 4x10* 6 I lx10* 10 Bx10* 4 3x10* 12 3x10* 5 Cf 252 s 6x10* 12 2x10* 4 2x10* 13 7xl0" 6 I 3x10* 11 2x10

4 lx10* 12 ?x10* 6 Cf 253 s Bx10* 10 4x10* 3 3x10* 11 lx10* 4 I 0x10* 10 4x10* 3 3x10* 11 lx10* 4 Cf 254 s 5xl0" 12 4xl0* 6 2x10* 13 1x10* 7 I 5x10* 12 4x10* 6 2x10* 13 1x10* 7 Carbon (6) . c 14 s 4x10* 6 2xl0~ 2 1x10* 7 8x10" 4

( C0 2 ) Sub 5xlO*S ........ Ix10* 6 ........ Cerium (58) Ce 141 s 4x10* 7 3xl0" 3 2x10* 8 9x10* 5 I 2x10* 7 3x10* 3 5x10* 9 9x1o*s Ce 143 s 3x10* 7 lxl0" 3 9x10* 9 4xl0* 5 I 2x10* 7 lx10* 3 7x10* 9 . 4x10* 5 Ce 144 s lxlO *S 3x10* 4 3x10* 10 lxlO*S I 6x10* 9 3x10* 4 2xlO*lO lxlO*S Cesium (55) Cs 131 s lxlO*S 7x10* 2 4xl0" 7 2x10* 3 I 3x10* 6 3x10* 2 lx10* 7 9x10* 4 Cs 134m s 4x10* 5 2x10* 1 lx10* 6 6x10* 3 I 6x10* 6 3x10* 2 2x10* 7 1x10* 3 Cs 134 s 4x10* 8 3xl0" 4 lx10* 9 9x10* 6 I lx10* 8 lxl0" 3 4xlO*lO 4x10* 5 Cs 135 s 5x10* 7 3x10" 3 2x10* 8 lx10* 4 I 9x10* 8 7x10* 3 3x10* 9 2x10* 4 Cs 136 s 4x10* 7 2x10* 3 lx10* 8 9x10* 5 I 2xl0" 7 2x10* 3 6x10* 9 6x10* 5 Cs 137 s 6x10* 9 4x10* 4 2x10* 9 2x10* 5 I 1x10* 0 lxl0" 3 5x10* 10 4x10" 5 Revision _fl_ Date 08/14/97 G-5

APPENDIX B TO §§20.1 - 20.602 CONCENTRATIONS IN AIR AND WATER ABOVE NATURAL BACKGROUND [See footnotes at end of Appendix BJ Table I Table I I Col. 1 Col. 2 Col. 1 Col. 2 Air Water Air Water Element (atomlc number) Isotope 1 (µCi/ml) (µCi/ml) (µCi/ml) (µCi/ml) Chlorine (17) Cl 36 s 4x10* 7 2x10* 3 lxl0- 8 Bxl0- 5 I 2xl0" 8 2x10* 3 sx10* 10 6xl0" 5 Cl 38 s 3xl0- 6 1x10* 2 9x10* 8 4x10* 4 I 2x10* 6 lx10* 2 7x10- 8 4xl0- 4 Chromium (24) . Cr 51 s lxl0- 5 5x10* 2 4xl0" 7 2x10* 3 I 2x10* 6 5xl0" 2 sx10* 0 2x10* 3 Cabal t ( 27) Co 57 s 3x10* 6 2x1o*z lx10* 7 5x10* 4 I 2x10* 7 lx10* 2 6x10* 9 4xl0- 4 Co 58m s 2x10* 5 8x10* 2 6x10* 7 3xl0- 3 I 9xl0" 6 6x10* 2 3x10* 7 2x10* 3 Co 58 s Bxl0- 7 4x10* 3 3xl0" 8 lx10* 4 I Sx10* 8 3x10* 3 2x10* 9 9x10* 5 Co 60 s 3x10* 7 lx10* 3 lxl0- 8 5x10* 5 I 9x10* 9 lx10* 3 3xlO*lO 3x10* 5 Copper (29) Cu 64 s 2xl0" 6 lx10* 2 7xl0" 8 3xl0" 4 I lxl0- 6 6x10* 3 4xl0" 8 2xl0" 4 Curium (96) . . Cm 242 s lxlO*lO 7x10* 4 4x10" 12 2x10" 5 I 2x10* 10 7x10* 4 6x10* 12 2Xl0-S Cm 243 s 6xic- 12 lxl0- 4 2x10" 13 5;(i.0" 6 I ix10* 10 7x10* 4 3x10* 12 2x10* 5 Cm 244 s 9x10* 12 2x10* 4 3x10* 13 7x10- 6 I lx10* 10 Bxl0- 4 3x10* 12 3x10* 5 Cm 245 s sx10* 12 lx10* 4 2x10* 13 4xl0- 6 I lxlO*lO ax10* 4 4x10* 12 3x10" 5

                            . Cm 246          s       Sxl0- 12       1x10* 4  2x10* 13       4xl0- 6 I      1x10* 10       ax10* 4  4x10* 12       3x10* 5 Cm 247          s       sx10* 12       lx10* 4  2x10* 13       4xl0" 6 I      ix10* 10       6x10* 4  4xl0*12        2xl0" 5 Cm 248          s       6x10* 13       lxlO-S    2x10* 14      4x10* 7 I      1x10* 11       4x10" 5  4x10* 13       lxl0- 6 Cm 249          s        lxlO-S        6x10* 2    4xl0- 7      2x10* 3 I       ix10* 5       6x10* 2    4x10* 1      2x10* 3 Revision ..11_ Date      08/14/97 G-6

APPENDIX B TO §§20.1 - 20.602 CONCENTRATIONS IN AIR AND WATER ABOVE NATURAL BACKGROUND [See footnotes at end of Appendix BJ Table I Table II Col. 1 Col. 2 Col. 1 Col. 2 Afr Water Afr Water 1 (µCf/ml) (µCi/ml) Element (atomic number) lsotope (µCf/ml) CµCf/ml) Dysprorfum (66) Dy 165 s 3xlo- 6 lx10- 2 9x10" 8 4xlo- 4 I 2xl0- 6 lx10* 2 7x10- 8 4xl0- 4 Dy 166 s 2x10- 1 lxl0" 3 8x10- 9 4xl0" 5 I 2x10- 7 lxl0" 3 7x10- 9 4xto-s Einsteinium (99) . Es 253 s 0x10- 10 7x10- 4 7x10- 4 3x10- 11 2xl0" 5 I 6xl0-lO 2x10* 11 2x10- 5 Es 254m s 5x10- 9 5x10- 4 2x10- 10 2x10- 5 I 6x10" 9 5x10- 4 2x10-lO 2x10" 5 Es 254 s 2x10* 11 4x10* 4 6x10* 13 lxlO-s I 1x10* 10 4x10" 4 4x10* 12 lxl0" 5 Es 255 s 5x10- 10 8xl0- 4 2x10- 11 3x10- 5 I 4x10- 10 8x10" 4 1x10- 11 3x10* 5 Erbium (68) Er 169 s 6x10- 1 3x10- 3 2x10* 8 9xlo-s I 4xl0" 7 3xl0" 3 lx10" 8 9x10* 5 Er 171 s 7x10* 1 3x10* 3 2x10" 8 lxl0" 4 I 6x10* 7 3x10" 3 2x10" 8 lx10* 4 Europium (63) Eu 152 s 4xl0" 7 2x10* 3 lxl0" 8 6x10" 5 CT/2*9-2 hrs) I 3xl0- 7 2xl0" 3 lxl0- 8 6xl0-s Eu 152 s lxl0- 8 2xl0" 3 4xlO*lO 8x10" 5 (T/2*13 yrs) I 2x10" 8 2x10- 3 6xlO*lO 8xl0" 5 Eu 154 s 4x10* 9 6x10* 4 1x10* 10 2x10- 5 I 7x10- 9 6xl0" 4 2x10* 10 2x10* 5 Eu 155 s 9xl0" 8 6x10* 3 3x10" 9 2x10* 4 I 7x10* 8 6xl0" 3 3x10" 9 2x10" 4 Fermium (100) Fm 254 s 6xl0- 8 4x10* 3 2x10* 9 lxl0" 4 I 7x10" 8 4xl0" 3 2x10* 9 lxl0" 4 Fm 255 s 2x10* 8 lxl0" 3 6x10* 10 3xl0" 5 I lxl0- 8 1x10* 3 4x10* 10 3x10- 5 Fm 256 s 3xl0" 9 3xl0" 5 lxlO*lO 9xl0" 7 I 2xl0" 9 3xlo-s 6x10- 11 9xl0- 7 Fluorine (9) F 18 s 5xl0 "6 2x10* 2 2x10* 1 8x10* 4 I Jx10- 6 lxl0- 2 9x10* 8 5x10" 4 Revision ..11,_ Date 08/14/97 G-7

APPENDIX B TO §§20.1 - 20.602 CONCENTRATIONS IN AIR AND WATER ABOVE NATURAL BACKGROUND [See footnotes at end of Appendix BJ Table I Table I I Col. 1 Col. 2 Col. 1 Col.* 2 Air Water Air Water Element (atomic number) Isotope 1 (µCi /ml) (µCi/ml) (µCi /ml) (µCi/ml) Gadolinium (64) Gd 153 s 2x10* 7 6x10* 3 sx10* 9 2x10* 4 I 9xl0 *B 6x10* 3 3x10* 9 2x10* 4 Gd 159 s Sx10* 7 2x10* 3 2x10* 0 Bxl0" 5 I 4xl0* 7 2x10* 3 lx10* 8 8x10* 5 Gallium (31) Ga 72 s 2x10* 7 1x10* 3 Bx10* 9 4x10* 5 I 2x10* 7 1x10* 3 6xl0" 9 4x10* 5 Germanium (32) Ge 71 s 1x10* 5 5xl0 2 4x10* 7 2x10* 3 I 6x10* 6 5x10* 2 2x10* 7 2x10* 3 Gold (79) Au 196 s 1x10* 6 Sx10* 3 4x10* 8 2x10* 4 I 6x10* 7 4x10* 3 2x10* 8 lx10* 4 Au 198 s 3x10* 7 2x10* 3 lxl0" 8 sx10* 5 I 2x10* 7 1x10* 3 sx10* 9 sx10* 5 Au 199 s lx10* 6 sx10* 3 4xlo* 8 2x10* 4 I Bxl0" 7 4x10* 3 3x10* 8 2x10* 4 Hafnium (72) . Hf 181 s 4xl0- 8 2x10* 3 lx10* 9 7x10" 5 I 7x10" 8 2x10* 3 3x10* 9 7x10* 5 Holmium (67) Ho 166 s 2x10* 7 9x10* 4 1x10* 9 3x10* 5 I 2x10* 7 9x10* 4 6x10" 9 3x10* 5 Hydrogen (1) H3 s 5x10" 6 lx10* 1 2x10* 1 3x10" 3 I sx10* 6 1x10* 1 2x10* 7 3x10* 3 Sub 2x10* 3 . . .. .. . . 4xl0" 5 ........ Indium (49) . In l13m s sx10* 6 4x10* 2 3x10" 7 1x10* 3 I 7x10* 6 4x10* 2 2x10* 7 1x10* 3 In 114m s ix10* 7 sx10* 4 4x10" 9 2x10* 5 I 2x10* 8 5x10* 4 7x10* 10 2x10* 5 In 115m s 2x10* 6 1x10* 2 ax10* 0 4x10" 4 I 2x20* 6 1x10* 2 6x10* 8 4x10* 4 In 115 s 2x20* 7 3xl0" 3 9xl0" 9 9x10" 5 I 3x20" 8 3x10* 3 1x10* 9 9x10* 5 Iodine (53) I 125 s 5x3o* 9 4x10* 5 8x10* 11 2x10* 7 I 2xso* 1 6x10* 3 6x10* 9 2x10" 4 I 126 s Bx20* 9 sx10* 5 9x10* 11 3xl0" 7 I 3xao* 1 3xl0" 3 lx10" 8 9x10* 5 I 129 s 2x30* 9 1x10* 5 2x10* 11 6x10" 8 I 7x20* 8 6x10* 3 2x10* 9 2x10* 4 Revision~ Date 08/14/97 G-8

APPENDIX B TO §§20.1

20.602 CONCENTRATIONS IN AIR AND WATER ABOVE NATURAL BACKGROUND

[See footnotes at end of Appendix BJ Table I Table II Col. 1 Col. 2 Col. 1 Col. 2 Air Water Air Water Element (atomic number) Isotope 1 CµCi/ml) (µC1/ml) (µC1/ml) (µC1/ml) Iodine (53) I 131 s 9x7o* 9 6x10* 5 ix10* 10 3x10* 7 (Continued) I 3x9o* 7 2x10* 3 lxl0" 8 6xl0" 5 I 132 s 2x10* 1 2x10* 3 3x10* 9 ax10* 6 I 9xl0* 7 5x10* 3 3x10* 8 2x10* 4 I 133 s 3x10* 8 2x10* 4 4xlO*lO 1x10* 6 [ 2x10* 1 lx10* 3 7x10* 9 4x10* 5 I 134 s 5x10* 7 4x10" 3 6x10* 9 2x10* 5 [ 3x10* 6 2x10* 2 lx10* 7 6xl0" 4 I 135 s lx10* 1 7x10* 4 lxl0" 9 4xl0" 6 I 4x10* 7 2x10* 3 lxl0- 8 7x10* 5 Ir 190 s lx10* 6 6x10* 3 4x10* 8 2x10* 4 Iridium (77) - I 4x10* 7 Sxl0" 3 lx10* 8 2x10* 4 Ir 192 s lx10* 7 lx10* 3 4x10* 9 4x10* 5 I 3x10* 8 1x10* 3 9x10* 10 4x10* 5 Ir 194 s 2x10* 7 lx10* 3 ax10* 9 3xlO*S I 2x10* 7 9x10* 4 5x10* 9 3xlO*S Iron (26) Fe 55 s 9x10* 7 2x10* 2 3x10* 8 ax10* 4 I 1x10* 6 1x10* 2 3x10* 8 2x10* 3 Fe 59 s 1x10* 7 2x10* 3 5x10* 9 6x10* 5 I 5x10* 8 2x10* 3 2x10* 9 5x10*S Krypton (36) Kr 85m Sub 6x10* 6 ........ lx10* 7 ........ Kr 85 Sub lx10* 5

                                                                    ........ 3x10* 7     ........

Kr 87 Sub lx10" 6 ........ 2x10* 8 ........ Kr 88 Sub 1x10* 6 ........ 2x10* 8 ........ Lanthanum (57) La 140 s 2x10* 1 7x10* 4 5x10" 9 2x10* 5 I lx10* 1 1x10* 4 4x10* 9 2x10* 5 Lead (82) . . Pb 203 s 3x10* 6 1x10* 2 9x10* 8 4x10* 4 I 2x10* 6 1x10* 2 6x10* 8 4x10* 4 Pb 210 s 1x10* 10 4x10* 6 4x10* 12 1x10* 7 l 2xlO*lO 5xl0" 3 8x10* 12 2x10* 4 Pb 212 s 2x10* 8 6x10* 4 6xlo* 10 2x10* 5 l 2x10* 8 5x10* 4 7x10* 10 2x10* 5 Revision ...11_ Date 08/14/97 G-9

APPENDIX B TO §§20.1

20.602 CONCENTRATIONS IN AIR AND WATER ABOVE NATURAL BACKGROUND

[See footnotes at end of Appendix BJ Table I Table I I Col. 1 Col. 2 Col. 1 Col. 2 Air Water Air Water Element (atomic number) Isotope 1 (µCi/ml) (µCi/ml) CµCi/ml) (µCi/ml) Lutetium (71) . Lu 177 s 6x10* 7 3x10* 3 2xl0- 8 lxl0- 4 I 5x10- 7 3x10- 3 2x10- 8 1x10- 4 Manganese (25) Mn 52 s 2x10- 7 lxl0- 3 7x10- 9 3xl0- 5 I lxl0- 7 9x10- 4 5xl0- 9 3xl0-s Mn 54 s 4x10- 7 4xl0- 3 lxl0- 8 lxl0- 4 I 4xl0-a 3xl0- 3 1x10- 9 lxl0- 4 Mn 56 s Bx10- 7 4x10- 3 3xlo- 8 lxl0- 4 I 5x10- 7 3xl0- 3 2x10-s lxl0- 4 Mercury. (BO) Hg 197m s 7x10- 7 6x10- 3 3xl0- 8 2x10- 4 I Bx10- 7 5xl0- 3 3xlo- 8 2x10- 4 Hg 197 s lxl0- 6 9xl0- 3 4x10- 8 3x10- 4 I 3x10- 6 1x10- 2 9x10- 8 5x10- 4 Hg 203 s 7x10- 8 5x10- 4 2x10- 9 2x10- 5 I lxl0- 7 3x10- 3 4xlo- 9 lxl0- 4 Molybdenum (42) Mo 99 s 7x10" 7 5x10- 3 3x10- 8 2x10- 4 I 2x10- 7 lxl0- 3 7x10- 9 4xlo-s Neodymium (60) Nd 144 s Bxlo- 11 2x10- 3 3xl0-12 7x10" 5 I 3x10- 10 2xl0- 3 lxl0- 11 BxlO-s Nd 147 s 4x10- 7 2x10- 3 lxl0" 8 . 6xl0-s I 2x10- 7 2xl0- 3 Bxl0- 9 6xl0" 5 Nd 149 s 2xl0- 6 Bxl0- 3 6xl0- 8 3xl0" 4 I 1x10- 6 Bx10" 3 5xl0- 8 3x10* 4 Neptunium (93) Np 237 s 4x10* 12 9x10-s lxl0" 13 3x10- 6 I lXlO-lO 9xl0- 4 4x10* 12 3x10* 5 Np 239 s Bx10- 7 4xl0- 3 3xl0" 8 lx10" 4 I 7x10* 7 4x10- 3 2x10- 8 lxl0- 4 Nickel (29) . Ni 59 s 5xl0- 7 6x10- 3 2xl0- 8 2x10* 4 I Bxl0- 7 6x10- 2 3xl0" 8 2x10- 3 N1 63 s 6x10" 8 Bxl0- 4 2x10- 9 3x10* 5 I 3xl0" 7 2x10- 2 ixio- 0 7x10- 4 Ni 65 s 9xl0" 7 4x10" 3 3x10-e lxl0- 4 I Sx10- 1 3x10" 3 2x10- 9 lx10- 4 Revision _.11_ Date 08/14/97 G-10

APPENDIX B TO §§20.1 - 20.602 CONCENTRATIONS IN AIR AND WATER ABOVE NATURAL BACKGROUND [See footnotes at end of Appendix B] Table I Table I I Col. 1 Col. 2 Col. 1 Col. 2 Alr Water Alr Water Element (atomic number) !sotope 1 (µCi /ml l (µCi/ml) (µCl/ml) (µCi/ml) Nioblum (Columbium) (41) Nb 93m s ix10* 1 1x10* 2 4xl0- 9 4x10* 4 I 2x10- 7 lxl0- 2 5x10- 9 4x10- 4 Nb 95 s 5x10* 1 3x10* 3 2x10* 8 1x10* 4 I lx10* 1 3x10* 3 3x10* 9 lxl0- 4 Nb 97 s 6xl0- 6 3x10- 2 2x10- 7 9x10- 4 I 5xl0- 6 3x10* 2 2x10* 7 9x10* 4 I Osmium (76) Os 185 *s 5x10* 7 2x10* 3 2xl0- 8 7xio-s I 5x10* 8 2x10* 3 2x10* 9 7xl0-s Os 191m s 2x10* 5 7x10* 2 6x10* 7 3x10* 3 I 9x10- 6 7x10* 2 3xl0- 7 2x10" 3 Os 191 s lxl0- 6 sxio- 3 4x10* 8 2x10* 4 I 4xlD- 7 Sxl0- 3 1x10* 8 2x10* 4 Os 193 s 4x10* 7 2x10* 3 lxl0- 8 6x10-s I 3xl0- 7 2x10* 3 9x10* 9 Sxl0" 5 Palladium (46) Pd 103 s lxl0- 6 lx10* 2 5x10" 8 .3x10* 4 I 7x10- 7 Bx10" 3 3xl0" 8 3xl0" 4

                            - Pd 109           s         6x10* 1     3x10* 3   2xl0- 8        9x10* 5 I         4x10* 7     2x10* 3   lxl0" 8        7x10* 5 Phosphorus ( 15)              p 32             s         7x10" 8     Sx10* 4   2x10* 9        2x10* 5 I         8x10- 8     7x10" 4   3x10* 9        2x10* 5 Platinum (78)           . Pt 191           s         Bxl0- 7     4x10* 3   3Xl0-B         1x10* 4 I         6x10* 7     3x10* 3   2x10* 8        lxl0- 4 Pt 193m          s         7x10" 6     3x10* 2   2x10* 7        lx10* 3 I         5x10" 6     3x10* 2   2x10* 1        lxl0- 3 Pt 193           s         lxl0- 6     3x10* 2   4xl0- 8        9xl0- 4 I         3x10* 1      5x10* 2   lxl0- 8       2x10* 3 Pt 197m          s         6x10* 6      3x10* 2   2x10* 1       lxl0- 3 I        Sxl0- 6      3x10* 2   2x10* 7       9x10* 4 Pt 197           s         sx10* 1      4x10* 3   3xl0-B        1x10* 4 I        6x10* 1      3x10* 3   2xl.0" 8      1x10* 4 Revision _1.1_ Date      08/14/97 G-11

APPENDIX B TO §§20.1

20.602 CONCENTRATIONS IN AIR AND WATER ABOVE NATURAL BACKGROUND

[See footnotes at end of Appendfx BJ Table I Table I I Col. 1 Col. 2 Col. 1 Col. 2 Air Water Air Water Element (atomic number) Isotope 1 (µCi /ml l (µCi/ml) CµCi/ml) (µCi/ml) Plutonium (94) Pu 238 s 2x10* 12 1x10* 4 7x10* 14 Sxl0- 6 I 3x10* 11 8xl0" 4 lxl0- 12 3xl0-s Pu 239 s 2x10* 12 lx10* 4 6x10* 14 Sxl0- 6 I 4x10* 11 8xl0- 4 lx10* 12 3xl0" 5 Pu 240 s 2x10* 12 lx10* 4 6x10* 14 Sxl0- 6 I 4xl0-ll 8x10* 4 lxl0*12 3x10* 5 Pu 241 s 9x10* 11 7x10* 3 3x10* 12 2x10* 4 I 4x10" 8 4x10* 2 1x10* 9 lxl0- 3 Pu 242 s 2x10- 12 lxl0- 4 6x10* 14 Sxl0- 6 I 4x10* 11 9x10* 4 lx10* 12 3x10* 5 Pu 243 s 2x10* 6 1x10* 2 6x10* 8 3x10* 4 I 2x10* 6 lx10* 2 Bxl0- 8 3x10* 4 Pu 244 s 2x10* 12 lxl0- 4 6x10* 14 4xl0" 6 I 3x10* 11 3x10* 4 lx10* 12 lxl0- 5 Polonium (84) Po 210 s SxlO*lO 2x10* 5 2x10* 11 7x10* 7 I 2x10* 10 0x10* 4 7x10* 12 3x10* 5 Potassium (19) K42 s 2xl0- 6 9xl0- 3 7xl0-s 3x10" 4 I lx10* 7 6x10* 4 4xl0- 9 2x10* 5 Praseodymium (59) Pr 142 s 2x10* 7 9x10* 4 7xl0" 9 3x10* 5 I 2x10* 7 9x10* 4 Sxl0- 9 3x10* 5 Pr 143 s 3x10* 1 lx10* 3 lx10* 8 Sx10" 5 I 2x10* 7 lx10* 3 6x10* 9 sx10* 5 Promethf um ( 61) Pm 147 s 6x1o*s 6x10* 3 2x10* 9 2x10* 4 I lxl0- 7 6x10* 3 3x10* 9 2x10* 4 Pm 149 s 3xl0- 7 lxl0- 3 lxlO-s 4x10* 5 I 2x10* 7 1x10* 3 8x10* 9 4x10* 5 Protoactinium (91) Pa 230 s 2x10* 9 7x10* 3 6x10* 11 2x10* 4 I sx10* 10 7x10* 3 3x10* 11 2x10* 4 Pa 231 s 1x10* 12 3x10* 5 4x10* 14 9x10* 7 I lxlO*lO Bxl0- 4 4x10* 12 2xlO*S Pa 233 s 6x10* 7 4x10* 3 2x10* 8 lx10* 4 I 2x10* 7 3x10* 3 6x10* 9 1x10* 4 Revision _fl_ Date 08/14/97 G-12

APPENDIX B TO §§20.1 - 20.602 CONCENTRATIONS IN AIR AND WATER ABOVE NATURAL BACKGROUND [See footnotes at end of Appendix BJ Table I Table I I Col. 1 Col. 2 Col. 1 Col. 2 Air Water Air Water Element (atomic number) I sotope 1 (µCi /ml) (µCi/ml) (µCi /ml) (µCi/ml) Radium (88) Ra 223 s 2x10* 9 2x10* 5 6x10* 11 7x10* 7 I 2x10* 10 lx10* 4 8x10* 12 4x10* 6 Ra 224 s 5x10* 9 7x10* 5 2x10* 10 2x10* 6 I 7x10* 10 2x10* 4 2x10* 11 5xl0" 6 Ra 226 s 3x10* 11 4x10* 7 3x10" 12 3xl0- 8 I 5x10* 11 9x10* 4 2x10* 12 3x10* 5 Ra 228 s 7x10* 11 8x10* 7 2x10* 12 3x10* 8 I 4x10* 11 7x10* 4 lx10* 12 3x10* 5 Radon (86) Rn 220 s 3x10* 7 ........ lxlo* 0 . ....... Rn 222 3 3x10* 0 ........ 3x10* 9 ........ Rhenium (75) Re 183 s 3x10* 6 2x10* 2 9x10* 0 6x10* 4 I 2x10* 7 8x10* 3 5x10* 9 3xl0" 4 Re 186 s 6x10* 7 3x10* 3 2x10* 0 9x.10* 5 I 2x10* 7 1x10* 3 Bx10* 9 Sx10* 5 Re 187 s 9x10* 6 7x10* 2 3xio* 7 3x10" 3 I Sx10* 7 4x10* 2 2x10* 0 2x10* 3 Re 188 s 4x10- 7 2x10* 3 lx10* 0 6x10* 5 I 2x10* 7 9x10* 4 6x10* 9 3xl0-s Rhodium (45) Rh 103m s 8x10* 5 4x10* 1 3x10" 6 lx10* 2 I 6x10* 5 3x10* 1 2x10* 6 1x10* 2 Rh 105 s 8x10* 7 4x10* 3 3x10* 0 1x10* 4 I sx10* 7 3x10* 3 2x10* 0 1x10* 4 Rubidium (37) Rb 86 s 3x10* 7 2x10* 3 lxlO*B 7x10* 5 I 7x10* 0 7x10* 4 2x10* 9 2x10* 5 Rb 87 s 5x10" 7 3x10" 3 2x10* 0 lx10* 4 I 7x10* 8 sx10* 3 2x10* 9 2x10* 4 Ruthenium (44) . Ru 97 s 2x10* 6 lx10* 2 ax10* 0 4x10* 4 I 2x10" 6 1x10* 2 6x10* 0 3x10* 4 Ru 103 s Sx10* 7 2x10* 3 2x10* 0 ax10* 5 I Bx10* 8 2x10* 3 3x10" 9 8x10* 5 Ru 105 s 7x10* 7 3x10* 3 2x10" 8 1x10* 4 I 5x10* 7 3x10* 3 2x10* 0 1x10* 4 Ru 106 s 8x10* 8 4x10* 4 3x10* 9 1x10* 5 I 6x10* 9 3x10* 4 2x10* 10 1x10* 5 Revision _11_ Date 08/14/97 G-13

APPENDIX B TO §§20.1 - 20.602 CONCENTRATIONS IN AIR AND WATER ABOVE NATURAL BACKGROUND [See footnotes at end of Appendix BJ Table I Table I I Col. 1 Col. 2 Col. 1 Col. 2 Air Water A1r Water Element (atomic number) Isotope 1 (µC1/ml) (µC1 /ml) (µC1/ml) (µCi/ml) Samarium (62) . Sm 147 s 7x10- 11 2xl0- 3 2x10- 12 6x10-s l 3x10- 10 2x10- 3 9x10- 12 7x10" 5 Sm 151 s 6xl0- 8 1x10* 2 2x10- 9 4x10- 4 I 1x10* 7 1x10* 2 5x10- 9 4x10- 4 Sm 153 s 5x10" 7 2x10* 3 2xl0- 8 BxlO-s l 4x10* 7 2x10- 3 lxl0- 8 ax10-s Scandium ( 21) Sc 46 s 2x10- 1 1x10* 3 ax10- 9 4x10- 5 I 2x10" 8 1x10- 3 ax10* 1° 4x10*S Sc 47 s 6x10- 1 3x10* 3 2x10- 8 9x10-s I Sx10* 7 3x10- 3 2x10* 8 9x10* 5 Sc 48 s 2x10- 7 8x10" 4 6x10* 9 3x10* 5 l 1x10* 7 ax10* 4 5x10" 9 3xl0" 5 Selenium (34) . Se 75 s 1x10* 6 9x10* 3 4xl0- 8 3x10* 4 I 1x10* 7 ax10* 3 4x10* 9 3x10* 4 Silicon (14) S1 31 s 6xl0- 6 3x10* 2 2x10- 7 9xl0- 4 I lxl0- 6 6x10- 3 3x10" 8 2x10* 4 Silver (47) . Ag 105 s 6x10* 7 3x10- 3 2x10* 8 1x10* 4 I ax10-s 3x10* 3 3x10* 9 lxl0- 4 Ag llOm s 2x10* 7 9x10" 4 7x10" 9 3x10" 5 I 1x10* 8. 9x10* 4 3x10* 10 3x10* 5 Ag 111 s 3xl0- 7 1x10- 3 lx10" 8 4x10" 5 I 2x10* 7 lx10" 3 Bx10" 9 4xio-s Sodium ( 11) . Na 22 s 2x10" 7 lxl0- 3 6xl0" 9 4x10- 5 I 9x10- 9 9x10* 4 3x10* 10 3xlo- 5 Na 24 s lxl0- 6 6x10* 3 4x10- 8 2x10- 4 l 1x10* 7 Bx10" 4 Sx10* 9 3x10* 5 Strontium (38) Sr 85m s 4x10-s 2x10* 1 ixio- 6 7x10* 3 I 3xl0" 5 2x10- 1 lx10" 6 7x10" 3 Sr 85 s 2x10- 7 3xl0- 3 Bxl0" 9 1x10* 4 I lxl0- 7 sx10- 3 4xl0" 9 2x10* 4 Sr 89 s 3x10" 8 3xl0" 4 3x10* 10 3xl0" 6 I 4xlo-s Bxl0- 4 lxl0" 9 3xl0" 5 Sr 90 s lx10" 9 1x10* 5 3x10- 11 3x10* 1 I sx10* 9 lx10* 3 2x10- 10 4x10* 5 Revision _11._ Date 08/14/97 G-14

APPENDIX B TO §§20.1 - 20.602 CONCENTRATIONS IN AIR AND WATER ABOVE NATURAL BACKGROUND [See footnotes at end of Appendix Bl Table I Table II Col. 1 Col. 2 Col. 1 Col. 2 Afr Water Afr Water Element (atomic number) Isotope 1 CµCf/ml) (µCf/ml) CµC1/ml) (µCf/ml) Strontium (38) Sr 91 s 4x10- 7 2x10* 3 2x10-a 7xl0-s (Continued) I 3x10* 7 lx10* 3 9x10* 9 5x10* 5 Sr 92 s 4x10* 7 2x10* 3 2x10* 8 7xlO*S I 3x10* 7 2x10* 3 lx10* 8 6xlO*S Sulfur ( 16) . . s 35 s I 3x10* 7 3x10* 7 2x10* 3 sx10* 3 9x10* 9 9x10" 9 6x10" 5 3xl0" 4 Tantalum (73) Ta 182 s 4x10* 8 1x10* 3 lx10* 9 4x10* 5 I 2x10* 8 lx10* 3 7xlO*lO 4x10* 5 Technetium (43) . Tc 96m s sx10* 5 4x10* 1 3x10* 6 lx10* 2 I 3xl0" 5 3x10* 1 1x10* 6 lx10* 2 Tc 96 s 6x10* 7 3x10* 3 2x10" 8 1x10* 4 I 2x10* 7 lx10* 3 8x10- 9 Sx10- 5 Tc 97m s 2x10* 6 lx10* 2 Sx10* 8 4x10* 4 I 2x10* 7 Sxl0" 3 5xl0* 9 2x10* 4 Tc 97 s lx10* 5 Sx10* 2 4xto- 7 2x10- 3 I 3x10* 7 2x10* 2 lxlO-a 8x10* 4 Tc 99m s 4x10* 5 2x10* 1 lxl0* 6 6x10* 3 I lx10* 5 sx10* 2 Sx10* 7 3x10* 3 Tc 99 s 2x10* 6 1x10* 2 7x10* 8 3x10- 4 I 6x10* 8 sx10* 3 2x10* 9 2x10- 4 Tellurium (52) Te 12Sm s 4x10* 7 sx10* 3 lx10* 8 2x10* 4 I 1x10* 7 3x10* 3 4xto- 9 1x10* 4 Te 127m s lx10* 7 2x10* 3 sx10* 9 6xl0" 5 I 4x10-a 2x10* 3 lx10* 9 Sxl0- 5 Te -127 s 2x10* 6 8x10- 3 6x10" 8 3x10* 4 I 9x10* 7 sx10* 3 3xl0" 8 2x10- 4 Te 129m s 8x10* 8 1x10* 3 3x10- 9 3x10* 5 I 3x10* 8 6x10* 4 lxl0- 9 2x10- 5 Te 129 s Sx10* 6 2x10- 2 2x10- 7 sx10* 4 I 4x10* 6 2x10* 2 lx10* 7 8x10" 4 Te 131m s 4x10* 7 2x10* 3 1x10* 8 6xl0" 5 I 2x10* 7 1x10* 3 6x10* 9 4x10-s Te 132 s 2x10* 7 9x10* 4 7x10* 9 3x10* 5 I lx10* 7 6x10* 4 4x10* 9 2x10* 5 Revision ...11_ Date 08/14/97 G-15

APPENDIX B TO §§20.1 - 20.602 CONCENTRATIONS IN AIR AND WATER ABOVE NATURAL BACKGROUND [See footnotes at end of Appendix B] Table I Table I I Col. 1 Col. 2 Col. 1 Col. 2 Air Water Air Water Element (atomic number) Isotope 1 (µCi/ml) CµC1/ml) (µCi/ml) CµC1/ml) Terbium (65) Tb 160 s 1x10* 7 1x10* 3 3x10* 9 4x10* 5 I 3x10* 8 1x10* 3 lx10* 9 4x10* 5 Thalfum ( 81) Tl 200 s 3x10* 6 lx10* 2 9x10" 8 4x10* 4 I lxl0" 6 7x10* 3 4xl0" 8 2x10* 4 Tl 201 s 2x10* 6 9x10* 3 7xl0* 8 3x10* 4 I 9x10* 7 sx10* 3 3x10* 8 2x10* 4 Tl 202 s ax10* 7 4x10* 3 3x10* 8 lx10* 4 I 2x10* 7 2x10* 3 ax10* 9 7x10* 5 Tl 204 s 6xl0* 7 3x10* 3 2x10* 8 1x10* 4 I 3x10* 8 2x10* 3 9xlO*lO 6x10* 5 Thorium (90) Th 227 s 3xlO*lO Sx10" 4 lx10* 11 2x10* 5 I 2x10* 1° Sx10* 4 6x10* 12 2x10* 5 Th 228 s 9x10* 12 2x10* 4 3x10* 13 7x10* 6 I 6x10* 12 . 4x10* 4 2x10* 13 ix10* 5 Th 230 s 2x10* 12 SxlO*S ax10* 14 2x10* 6 I lx10* 11 9x10* 4 3x10*l 3 3xl0" 5 Th 231 s 1x10* 6 7x10* 3 5x10* 8 2x10* 4 I lx10

6 7x10* 3 4xl0* 8 2x10* 4 Th 232 s 3x10* 11 5x10* 5 lx10* 12 2x10* 6 I 3x10* 11 1x10* 3 1x10* 12 4x10* 5 Th natural s 6x10* 11 6x10* 5 2x10* 12 2x10* 6 I 6x10* 11 6x10* 4 2x10* 12 2x10* 5 Th 234 s 6x10* 8 Sx10* 4 2x10* 9 2x10* 5 I 3x10* 8 . 5x10* 4 lx10* 9 2x10* 5 Thulium (69) Tm 170 s 4xl0* 8 1x10* 3 1x10* 9 sx10* 5 I 3x10* 8 lx10* 3 1x10* 9 sx10* 5 Tm 171 s 1x10* 7 1x10* 2 4x10* 9 5x10" 4 I 2x10* 7 lx10* 2 ax10* 9 5x10" 4 Tin (50) . Sn 113 s 4x10* 7 2x10* 3 1x10* 8 9x10* 5 I Sx1o*B 2x10* 3 2x10* 9 Bx1o*S Sn 125 s 1x10* 7 sx10* 4 4x10* 9 2x10* 5 I Bxlo*S 5x10" 4 3x10* 9 2x10* 5 Revision ..1..1_ Date 08/14/97 G-16

APPENDIX B TO §§20.1

20.602 CONCENTRATIONS IN AIR AND WATER ABOVE NATURAL BACKGROUND

[See footnotes at end of Appendfx BJ Table I Table II Col. 1 Col. 2 Col. 1 Col. 2 Afr Water Afr Water Element (atomic number) lsotope 1 (µCi/ml) (µCf/ml) (µCf/ml) (µCf/ml) Tungsten (Wolfram) (74) w 181 s 2x10* 6 lx10* 2 0x10* 8 4x10* 4 I lx10* 7 lx10* 2 4x10* 9 3x10* 4 w 185 s ax10* 1 4x10* 3 3x10* 8 lx10* 4 I lx10* 7 3x10* 3 4x10* 9 lx10* 4 w 187 s 4x10* 7 2x10* 3 2x10* 8 7x10* 5 I 3x10* 7 2x10* 3 lxl0- 8 6xlO*S Uranfum (92) u 230 s 3x10* 10 lx10* 4 lx10* 11 Sx10* 6 I lxlO*lO lx10* 4 4x10* 12 Sxl0- 6 u 232 s lx10* 10 0x10* 4 3x10* 12 3x10* 5 I 3x10* 11 0x10* 4 9x10* 13 3x10* 5 u 233 s 5x10* 10 9x10* 4 2x10* 11 3xlO*S I lx10* 10 9x10* 4 4x10* 12 3x10* 5 u 234 54 6x10* 10 9x10* 4 2x10* 11 3x10* 5 I lxlO*lO 9x10* 4 4x10* 12 3x10* 5 u 235 s4 5x10* 10 ax10* 4 2x10* 11 3x10* 5 I lxlO*lO Bxl0- 4 4x10* 12 3x10* 5 u 236 s 6x10* 10 lx10* 3 2x10* 11 3x10* 5 I lx10* 10 lx10* 3 4x10* 12 3x10* 5 u 239 54 7x10* 11 lx10* 3 3x10* 12 4x10* 5 I lx10* 10 1x10* 3 Sxl0" 12 4x10* 5 u 240 s 2x10* 7 lx10* 3 8x10* 9 3xl0* 5 I 2x10*.7 lx10* 3 6x10* 9 3xlO*S U*natural s4 lx10* 10 1x10* 3 sx10* 12 3x10* 5 I 1x10* 10 ix10* 3 Sxl0" 12 3x10*S Vanadium (23) v 48 s 2x10* 7 9x10* 4 6x10* 9 3x10* 5 I 6x10* 0 ax10* 4 2x10* 9 3x10* 5 Xenon (54) Xe 13lm Sub 2x10* 5 .. .. . . . . 4x10* 7 ........ Xe 133 Sub lxlO*S . . ...... 3x10* 7 ........ Xe 133m Sub 1x10* 5 .. . .. . . . 3x10* 7 ........ Xe 135 Sub 4xl0* 6 .. .. .. . . 1x10* 7 ........ Ytterbium (70) Yb 175 s 7x10* 7 3x10* 3 2x10* 8 lx10* 4 I 6x10* 1 3xl0" 3 2x10* 8 lx10* 4 Yttrium (39) . y 90 s lx10* 7 6x10* 4 4x10* 9 2xlO*S I lx10* 7 6x10* 4 3x10* 9 2x10* 5 Revision ...11._ Date 08/14/97 G-17

APPENDIX B TO §§20.l

20.602 CONCENTRATIONS iN AIR AND WATER ABOVE NATURAL BACKGROUND

[See footnotes at end of Appendix BJ Table I Table I I Col. 1 Col. 2 Col. 1 Col. 2 A1r Water A1r Water Element (atomic number) Isotope 1 CµC1/ml) (µC1/ml) (µCi/ml) (µC1/ml) y 91m s 2x10* 5 1x10* 1 Bx10* 7 Jx10* 3 I 2x10* 5 ix10* 1 6x10* 7 3x10* 3 y 91 s 4x10* 8 Bx10* 4 lx10* 9 3x10* 5 I 3x10* 8 Bx10* 4 lx10* 9 3x10*S y 92 s 4x10* 7 2x10* 3 1x10* 8 6x10*S I 3x10* 1 2x10* 3 lxlO*B 6x10* 5 y 93 s 2x10* 1 sx10* 4 6x10* 9 3x10* 5 I lx10* 1 Bx10* 4 Sx10* 9 3x10* 5 Zinc (30) Zn 65 s 1x10* 7 3x10* 3 4x10* 9 lx10* 4 I 6x10* 8 sx10* 3 2x10* 9 2x10* 4 Zn 69m s 4x10* 7 2x10* 3 lx10* 8 7x10* 5 I 3x10* 1 2x10* 3 lx10* 8 6x10* 5 Zn 69 s 7xl0" 6 sx10* 2 2x10* 7 2x10* 3 I 9x10* 6 sx10* 2 3x10* 7 2x10* 3 Zirconium (40) Zr 93 s lx10* 1 2x10* 2 4x10* 9 Bx10* 4 I 3xl0* 7 2x10* 2 lx10* 8 Bx10* 4 Zr 95 s lx10* 7 2x10* 3 4x10* 9 6x10* 5 I 3x10* 8 2x10* 3 lx10* 9 6x10* 5 Zr 97 s lx10* 7 sx10* 4 4x10* 9 2x10* 5 I 9xl0* 8 sx10* 4 3x10* 9 2x10* 5 Any single radionucl1de not listed above with

                               ........       Sub     lx10* 6     ........ 3x10* 8     .......

decay mode other than alpha emission or spontaneous fission and with radioactive half-life less than 2 hours. Any single radionucl1de ........ 3x10* 9 9x10* 5 lxlO*lO 3x10* 6 not listed above with decay mode other than alpha emission or spontaneous fission and with radioactive half-life greater than 2 hours. Any single rad1onuclide ........ 6x10* 13 4x10* 7 2x10* 14

3x10" 8 not listed above which decays by alpha emission or spontaneous fission.

Revision _11_ Date 08/14/97 G-18

1Soluble <SJ; Insoluble(!). 7 *sub" means that values given are for submersion in a semispherical infinite cloud of airborne material. 3 These radon concentrations are appropriate for protection from radon-222 combined with its short-lived daughters. Alternatively, the value in Table I may be replaced by one-third (l/3) "working level.* (A "working level" is defined as any combination of short-lived radon-222 daughters, polonium-218, lead-214. bismuth-214 and polonium-214, .in one ~it~r of air, with~ut regard to the degree of equilibrium, that will result 1n the ultimate emission of 1.3 x 10 MeV of alpha particle energy.) The Table II value may be replaced by one-thirtieth (1/30) of a "working level.* The limit on radon-222 concentrations in restricted areas may be based on an annual average. 4 For soluble mixtures of U-238. U-234 and U-235 in air chemical toxicity may be the limiting factor. If the percent by weight-enrichment of U-235 is less than 5, the concentration value for a 40-hour work week, Table I, is 0.2 milligrams uranium per cubic meter of air average. For any enrichment, the product of the average concentration and time of exposure during a 40-hour work week shall not exceed 8 x 10* 3 SA µCi-hr/ml. where SA is the specific activity of the uranium inhaled. fhe concentration value for Table II is 0.007 milligrams 7 uranium per cubic meter of air. The specific activity for natural urdnium is G.77 x io* curies per gram U. The specific activity for other mixtures of U-238, U-235 and U-234, if not known, shall be: SA = 3.6 x 10* 1 curies/gram U U-depleted SA= (0.4 1- 0.38 E + 0.0034 E7 J 10- 6 E ~ 0.72 where E is the percentage by weight of u=-235, expressed as percent. NOTE: In any case where there is a mixture in air or water of more than one radionuclide, the limiting values for purposes of this Appendix should be determined as follows:

1. If the identity and concentration of each radionuclide in the mixture are known, the limiting values should be derived as follows: Determine, for each radionuclide in the mixture, the ratio between the quantity present in the mixture and the limit otherwise established in Appendix B for the specific radionuclide when not in a mixture. The sum of such ratios for all the radionuclides in the mixture may not exceed "1" (i.e., "unity").

EXAMPLE: If radionuclides A, B, and Care present in concentrations CA, C8 , and Cc, and if the applicable MPC's are MPCA. and MPC 8 , and MPCc respectively, then the concentrations shall be limited so that the following relationship exists: (CA/MPCAJ + (C 8 /MPC 9l + !Cc/MPCcl i 1

2. If either the identity or the concentration of any radionuclide in the mixture is not known, the limiting values for purposes of Appendix B shall be:

a. For purposes of Table I, Col. 1 - 6x10- 3

b. For purposes of Table I, Col. 2 - 4x10* 1

c. For purposes of Table I I, Col. 1 - 2xl0- 14

d. For purposes of Table I I. Col. 2 - 3xl0 8

3. If any of the conditions specified below are met, the corresponding values specified below may be used in lieu of those specified in paragraph 2 above.

a. If the identity of each radionuclide in the mixture is known but the concentration of one or more of the radionuclides in the mixture is not known, the concentration limit for the mixture is the limit specified in Appendix "B" for the radionuclide in the mixture having the lowest concentration limit: or

b. If the identity of each radionuclide in the mixture is not known, but it is known that certain radionuclides specified in Appendix "B" are not present in the mixture, the concentration limit for the mixture is the lowest concentration limit specified in Appendix "B for any radionuclide which is not known to be absent from the mixture; or Revision _1.1_ Date 08/14/97 G-19

Table [ Table l I Col. 1 Cul. 2 Col. l Co I. 2 Air Water Air Water

c. Element Cu tomi c number l and isotopP. (PTlµCi/ml) (µCi /ml l (µCi /ml l CµC i /ml l l f it is known that Sr 90, I 125, I 126, I 129, I 131 (I 133. Table I I only),

Pb 210. Pu 210, At 211. Ra 23, Ra 224. Ra 226. Ac 227, Ra 228. Th 230, Pu 231. Th 232. Th-nat. Cm 248, Cf 75'1. and Fm 256 are not present . .. .. . . . 9xl0 5

                                                                                      ........           3xl0* 6 If it is known that. Sr 90. I 125, I 126, I 129 ( [ 131. [ 133, Table I l only),

Pb 210, Po ZI 0. Ra 223. Ra 226, Ra 228, Prl 231. Th-nat, Cm 248, Cf 2S4, and Fm 256 are not present ....... 6x10* 5 0 o Io o o o o 2xl0* 6 If it is known that Sr 90. I 129 (I 125, I 126, I 131. Table I I only l. Pb 210. Ra 226, Ra 228, r.m 248, and Cf 254 are not present ........ 2x10* 5 . ....... 6x10* 7 If it is kr1own that ( l 129. Table I I unlyl, Ra 226, and Ra 228 are not present ........ 3xl0* 6 . ....... lx 1o* 7 l f it is known that alpha-emitters and Sr 90. I 129, Pb 210, Ac 227. Ra 228. Pa 230, Pu 241. and Bl<. 249 are not present 3x10* 9 ....... lx10* 10 . ....... lf it is known that alpha-emitters and Pb 210, Ac 227, Ra 228. and Pu 241 are not present 3x10* 10 .. . . . ... lx10* 11 ........ l f it. is known that a; phu *emitters arid Ac 227 are not present 3x10* 11 . . . ... . . lxlO 12 If it i s known that Ac 227. Th 230. Pa 231. Pu 238, Pu 239. Pu 240, Pu 242. Pu 244. Cm 248, Cf 249 and Cf 251 are not present 3x10* 12 . . .. . . . . lx10* 13 ........

4. If a mixture of radionuclides consists of uranium and its _daughters in ore dust prior to chemical separation of the uranium from the ore. the values specified below may be used for uranium and its daughters through radium-266, instead of those from paragraphs 1, 2. or 3 above.

a. For purposes of Table I. Col. 1 - lx10* 10 µCi/ml gross alpha activity; or 5x10* 11 µCi/ml natural uranium or 75 micrograms per cubic meter of air natural uranium.

b. For purposes of Table II. Col. 1 - 3x10* 12 µCi/ml gross alpha activity; 2x10* 12 µCi/ml natural uranium; or 3 micrograms per cubic meter of air natural uranium.

5. For purposes of this note, a radionuclide may be considered as not present in a mixture if (a) the ratio of the concentration of that radionuclide in the mixture (CA) to the concentration limit for that radionuclide specified in Table II of Appendix "B" CMPCA) does not exceed 1/10, (i.e. CA/MPCA ~ 1/10) and (bl the sum of such ratios for all the radionuclides considered as not present in the mixture does not exceed 1/4 i.e.

(CA/MPCA + C9 /MPC 9 .** + i, 1/4). Revision _11._ Date 08/14/97 G-20

Appendix H

1. "Request to Amend Previous Approvals Granted Under 10CFR20.302(a) for Disposal of Contaminated Septic Waste and Cooling Tower Silt to Allow for Disposal of Contaminated Soil", dated June 23rd, 1.999, BVY 99-80

2. "Supplement to Request to Amend.Previous Approvals Granted Under 10CFR20.30(a) to Allow for Disposal of Contaminated Soil, dated January 4lh, BVY 00-02 .

3. "Vennont Yankee Nuclear Power Station, Request to Amend Previous Approvals Granted under 10CFR20.302~) to Allow for Disposal of Contaminated Soil (TAC No. MA5950), dated June 15 , 2000, NVY 00-58 Revision -1.2..._ Date 1/11/02 H-1

VERMONT YANKEE . N:.UCLEAR POWER CORPORATION 185 Old Ferry Road, Brattleboro, VT 05301-7002 (802) 257-5271 June 23, 1999 BVY99-80 United States Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555

References:

(a) Letter from R.W. Capstick, Vermont Yankee, to USNRC, "Request to Routinely Dispose of Slightly Contaminated Waste in Accordance with 10CFR20.302(a)", BVY 89-59, June 28, 1989 (b) Letter fror.-i M.B. Fa.W-ilc, US1\1RC, to L.A. Trem~ley, Vermont Y~lcee, "Approval Under 10CFR20.302(a) of Procedures for Disposal of Slightly Contaminated Septic Waste on Site at Vermont Yankee (fAC No. 73776)", NVY 89-189, dated August 30, 1989 (c) Letter from J.J. Duffy, Vermont Yankee, to USNRC, Request to Amend Previous Approval Granted Under 10CFR20.302(a) for Disposal of Contaminated Septic Waste", BVY 95-~7, dated August 30, 1995 (d)_ Letter from S.A. Varga, USNRC, to D. A. Reid, Vermont Yankee, "Approval Pursuant to 10CFR20.2002 for Onsite Disposal of Cooling Tower Silt - Vermont Yankee Nuclear Power Station (fAC No. M93420)", NvY 96-39, dated March 4, 1.996 (e) Letter from P.D~ Milano, USNRC, to D. A. Reid, Vermont Yankee, "Revised Safety Evaluation - Approval Pursuant to 10CFR202002 for Onsite Disposal of Cooling Tower Silt - Vermont Yankee Nuclear Power Statioc {T.1\C No. M9637!)", NV'! 97-85, dated June 18, 1997

Subject:

Vermont Yankee Nuclear Power Station License No. DPR-28 (Docket No. 50-271) Request to Amend Previous Approvals Granted Under 10 CFR 20.302(a) for Disposal of Contaminated Septic Waste and Cooling Tower Silt to Allow for Disposal of Contaminated Soil ln accordance with 10CFR 20.2002 (previously 10CFR20.302(a)), Vermont Yankee submits this application to amend the previously granted approvals to dispose of slightly contaminated septic waste and cooling tower silt on-site. 1bis application expands the alfowable waste stream. to

include slightly contaminated soil generated as a residual by-product of on:--Site construction activities.

1his application specifically requests approval to dispose of soil contaminated at minimal levels, which has been or might be generated through end of station operations at the Vermont Yankee Nuclear Power Plant. The proposed soil disposal method is the same as the septic .waste and cooling tower silt disposal methods requested in References (a) and (c}, and approved in References (b) and (e). The disposal method utilizes on-site land spreading in the same desipted areas used for septic waste and cooling tower silt. Disposal of this waste in the manner proposed, rather than holding it for future disposal at a 10CFR Part 61 licensed facility will save substantial costs and reserve valuable disposal site space for waste of higher radioactivity levels. Revision .12... Date 1/11/02 H-2

VERMONT YANKEE NUCLEAR POWER CORPORATION BVY 99-80 I Page 2 of 2 A radiological assessment' and proposed operatioruil c.onti'ols for the inclusion of additional earthen material (soil) for on-site disposal with septic waSte and cooling tower silt is provided in Attachment A. The assessment demonstrates that the dose impact expected from the existing accumulation of approximately 25.5 cubic meters of soil, in total with all past waste spreading operations, will not approach the dose limits already imposed for septic and cooling tower silt disposal. In addition to the existing accumulated soil, VY also requests that any future low level contaminated soil that might be generated as a by-product of plant construction and maintenance activities be allowed to be disposed of in the same manner provided the approved acceptance dose criteria are met. All soil analyses will be to environmental lower limits of detection. The results of all disposal operations will also be rcpllrted in the Ann~ Radioactive Effluent Release Report. The combined radiological impact for all on-site disposal operations will continue to be limited to a total body or organ dose of a maximally exposed member of the public of less than one mrem/year during the period of active* Vermont Yankee control of the site, or less than five rnrem/year to an inadvertent intruder after tennination of active site control. The Vermont Yankee Off-Site Dose Calculation Manual (ODCM) contains a.copy of the original assessment and NRC approval for septic waste disposal (References a and b) and the previous amendment for cooling tower silt (References c and e). Upon receipt of your approval, the information contained in Attachment l as well as the basis for approval Will be incorporated into

     .theODCM.

We tru....-i: that the infonnation contained in the submittal is sufficient. However, should you have any questions or require further information concerning this matter, please contact Mr. Jim DeVincentis at 802-253-4236.

Sincerely, VERMONT YANKEE NUCLEAR POWER CORPORATION

                                                          /Gautam Sen             /.

Licensing Manager Attachment cc: USNRC Region I Administrator

  *~              USNRC Resident Inspector - VYNPS USNRC Project Manager - VYNPS Vf Department of Public Service Revision.£2.. Date    1/11/02 H-3

SUMMARY

OF. VER1'10NT YANKEE COMMITMENTS BVY NO.: 99-80 The following table identifies commitn:ient.s made in this document by Vermont Yankee. Any other actions discussed in the submittal represent intended or planned actions by Vermont Yankee. They are described to the NRC for the NRC's information and are not regulatory commitments. Please notify the Licensing Manager of any questions regarding this document or any associated commitments. Illb=============c=o=MMI===T=ME==NT==========::::k::===ll~g=~=~~*o~ur~AE~G~=?=,A=T=E======!ll None VY.APF 0058.04 (Sample) AP 0058 Original Page 1of1 H-4 Revision _2.2...._ Date 1/11/02

Docket No. 50-271 BVY99-80 Attachment 1 Vermont Yankee Nuclear Power Station Assessment of On-site Disposal of Contaminated Soil by Land .Spreading Revision _I2_ Date l/l l/02 H-5

BVY 99-80 I Attachment l /Page 1 l"ABLE OF CONTENTS Page LIST OF TABLES 2 1.0 IN1RODUCTION 3 1.1 Background . . .. . . .. . . .. . .. .. . . .. .. . .. .. . * .. .. .. . .. . . . . . . .. . . . . . *. *.. . * .. . . . . . . . . 3 1*.2 Objeetive ...........................................-. ... . . . . . . . . . . . .. **.... . . ....

3 2.0 WASTE DESCRIPTION ............ ;................................................... " 4
3.0 SOIL DISPOSAL AND ADMINISTRATivE PROCEDURE REQlTIRE~S......................................................................... 5 4.0 EVALUATION OF ENVIRONMENTA;L IMPACTS ......**... ~ .... ~-*............ 6 4.1 Site Characteristics . .. . .. .. . . .. .. . *.. . .. .. .. . *.. .. .. .. . ** . . . .. . . .* .. . *.. . .. .. . . . .. . .. 6*

4.2 Radiological Impact.................................................................. 6 5.0 RADIOLOGICAL PROTECTION...................................................... It

6.0 CONCLUSION

S........................................................................... 9

7.0 REFERENCES

.................................... ~.........................................                                                       9 Revision .l2_  Date   l/11/02 H-6

l' BVY 99-80 I Attachment l /Page 2 LIST OF TABLES Table 1: Radioanalytical Results of Composite Soil Samples.......................... .1 O Table 2: Estimated Total Radioactivity in Soil Volume .................................. 10 Table 3: Total Activity on South Field After Last Spreading Event................. 11 Table 4: Total Projected Radioactivity Remaining en South Filed at License Expiration ........ : ............................................ \......................... 11 Table 5: All-Pathway Critical Organ/Whole Body Dose Conversion Factors During Vermont Yankee Control of Disposal Site ......................................... 11 Table 6: All-Pathway Critical Organ/Whole Body Dose Conversion Factors Post Vermont Yankee Control of Disposal Sites (Inadvertent Intruder) ............. 12 Table 7: Dose Contribution from Co-60 and Cs-137 in Soil Volume after Land Spreading . . ... . .. . .. . .. .. . . . . . . . . .. . . . .. . . .. .. . .. .. .. .. . .. .. . .. .. . . . ... .. . . . . .. .. . . 12 Table 8: Present and Future Dose Impact Due to the Soil Spreading for Two Cases .................. :....................................................... 13 Revision _l2.._ Date 1/11102 H-7

BVY 99-80 I Attachment l / Page 3 1.0 - INTRODUCTION 1.1 Background In 1989, Vermont Yankee Nuclear Power Corporation requested approval from the NRC to routinely dispose of slightly contaminated septic waste in designated on-site areas in accordance with 10CFR20.302(a). Approval from the NRC was granted on August 30, 1989 and the information was permanently incorporated into the Offsite Dose Calculation Manual (ODCM) as Appendix B. In 1995, Vermont Ymkee Nuclear Power Corporation requested that the rrevfous authoriz.ation for on-site disposal of very low-level radioactive material in septic waste be amended to permit the on-site disposal of slightly con~ted cooling tower silt material. Approval from the NRC was granted on June 18, 1997 and the information was permanently incorporated into the ODCM as Appendix F. lli 1994, approximately 25.5 m3 of excess soil was generated during on:.site construction activities. Sampling of the soil revealed lo.w levels ofrad.ioactivity that were similar in radionuclides and activity to the septic waste and cooling tower silts previously . e~countered. An evaluation determined that the soil could be managed in similar fashion as the septic waste and cooling tower silts; however, prior approval from the NRC wol,tld be required under 10 CFR 20.2002 (formerly 20.302(a)). 1.2 Object!ve The objective of this assessment is to present the data and radiological evaluation to demonstrate that the proposed disposition of the soil will meet the existing boundary conditions as approved by the NRC for septic waste and cooling tower silt. The boundary conditions established for disposal of the septic waste and cooling tower silts on the designated plots are: The dose to the whole body or any organ of a hypothetical maximally exposed individual must be less than 1.0 miem/yr. Doses to the whole body and any organ dose of an inadvertent intruder from the probable pathways of exposure are less than 5 mrem/yr. Disposal operations must be at one of the approv:ed on-site locations. Revision .22._ Date 1/11/02 H-8

T BVY 99-80 I Attachment l /Page 4 2.0 WASTE DESCRIPTION The soil that is the subject of this evaluation was derived from excavations resulting from activities associated with a new securicy fence along the plant's Protected Area boundary. The volume of soil generated was approximately 25.5 m, and is typical of fill material containing light to dark brown poorly sorted soils with some small stones, and includes small incidental pieces of asphalt. The soil was removed from its original location by shovel, backhoe and front-end loader, and placed into dump trucks for transport to the location between the cooling towers where it was deposited on the ground surface and covered to prevent erosion. Th.is location was selected because it was away from areas routinely occupied by plaut staff, and could easily be controlled. The most probable source of the low levels of radioactive contamination is the presence of below detectable removable contamination redistributed by foot traffic from inside the plant to walkways and parking areas. Subsequent surface runoff carries the contamination to nearby exposed soil near the Protected Area boundary where it bad accumulated over time to low level detectable concentrations. In April 1995, a total of20 composite soil samples were collected to characterize the volume. Composites were obtamed. by collecting a grab sample from one side, the top and the opposite side at equal distances along the length of the pile, then combining the iliree into one sample. Soil samples were sent to the Yankee Atomic Environmental. Laboratory for analysis and counted to environmental lower limits of detection required of environmental media. Results of the analyses are presented in Table 1. Analytical results are provided for when the samples were collected and decay corrected to the present. The results identified both Cs-137 and Co-60 in most of the composite samples, which verified that plant-related radioactivity was present in the soil. For the pmpose of estimating the total activity in the soil pile, the actual analytical result was used for those samples that were less than the ~C to calculate the average radioactivity concentration. The mass of soil (dry)_was estimated by multiplying_the total in-situ_volume (25.5 m3) by its wet density,-l.47E+03 kglm3, and then dividing by the wet:dry ratio of 1.12, thu5

yielding a mass of 3.35E+o4 kg (dry). The ma.ss of the soil was then multiplied by the average Co-60 and Cs-137 concentrations measured in the soil to obtain' the total activity of each radionuclide in the 25 .5 m3* Table 2 presents the estimated total radioactivity in the 25 .5 m3 of soil at the time of sample collection and analysis, and decay corrected to the present Revision _12..._ Date ---1L!..!.iQ£ H-9

                                                                                           -------~-

BVY 99-80 I Attachment 1 / Page 5 3.0 SOIL DISPOSAL AND ADl\llNISTRATIVE PROCEDURE REQUIREMENTS The method of soil disposal will use the technique of land spreading in a manner consistent with the current commitments for the on-site disposal of septic waste and cooling tower silts as approved by the NRC. The accumulation of radioactivity on the disposal plot for this soil spreading operation will be treated as if cooling tower silt or septic waste was being disposed of since the characteristics of all these residual solids are similar (earthen-type matter). The south field (approximately 1.9 acres in size) designated

          ?nd approved for septic waste ~nd coolhlg tower silts disposal has been used for all past disposal operations, and will be used for the placement of this soil. Determination of the radiological dose impact has been made based on the same models and pathway assumptions used in the previous subml:ttals.

Dey soil material will be dispersed using typical agricultural dry bulk surface spreading practices in approved disposal areas on-site. Incidental pieces of asphalt and stones that were picked up with the soil from area where paving ran along the fence line will be screened out before the soil is spread and disposed of as radioactive tµaterial at an off-site

li.censed facility if detectable radioactivity is found.
Records of the disposal that will be maintained include the following:

(a) Th~ radionuclide concenuation3 detected b the soil (me~sured to environmental lower limits of detection). (b) The total volume of soil disposed of. (c) The total radioactivity in the disposal operation as well as the total accumulated on each disposal plot at the time of spreading. (d) The plot on which the soil was applied. (e) Dose calculations or maximum allowable accumulated activity determinations required to demonstrate that the dose limits imposed on the land spreading operations have not been exceeded. To ensrire that the addition of the soil containing the radioactivity will not.exceed the boundary conditions, the total radioactivity and dose calculation will include all past disposals of septic waste and cooling tower silt containing low-level radioactive material on the designated disposal plots. In addition, concentration limits applied to the disposal of earthen type materials (dry soil) restrict the placement of small volumes of materials that have relatively high radioactivity concentrations. Any farm.er leasing land used for the disposal of soil will be notified of the applicable restrictions placed on the site due to the spreading oflow level contaminated material. These restrictions are the same as detailed for the previously approved septic waste spreading application. Revision..1.2..._ Date 1/11/02 H-10

DVY 99-80 I Attachment I /Page 6 4.0 EVALUATION OF ENVIRONMENTAL IMPACTS 4.1 Site Characteristics The designated disposal site is located on the Vermont Yankee Nuclear Power Plant site and is within the site boundary security fence. The south field consists of approximately 1.9 acres and is centered approximately 1500 feet south of the Reactor Building. This parcel of land has been previously approved by the NRC for the land disposal of septic waste and cooling tower silt. 4.2 Radiological Impact The amount of radioactivity added to the* 5outh field soil is procedurally controlled to ensure that doses are maintained within the prior approved limits of the boundary

conditions.

To assess the dose received (after spreading the soil) by the maximally exposed individual during the period of plant controls over the property, and to an inadvertent intruder after plants controls of access ends, the same pathway modeling, assumptions and dose calculation methods as approved for septic and cooling tower silt disposal were used. These dose models implement the methods and dose conversion factors as provided in Regulatory Guide 1.109. The following six potential pathways were identified and included in the analysis: (a) Standing on contaminated ground. (b) Inhalation of resuspended radioactivity .

          . (c)     Ingestion ofleafy vegetables.

(d) Ingestion of stored vegetables. (e) Ingestion of meat. (f) Ingestion of cow's milk. Both the maximum individual and inadvertent intruder are assumed to be exposed to these pathways; the difference between them is due to the occupancy time. The basic assumptions used in the radiological arialyses include: * (a) Exposure to ground contamination and re-suspended radioactivity is for a period of 104 hours per year during the Vermont Yankee active control of the disposal sites and continuous thereafter. The 104-hour interval is representative of a farmer's time spent on a plot of land (4 hours per week for 6 months). Revision _l9_ Date 1/11102 H-11

BVY 99-80 I Attachment 1 I Page 7 (b) For the purpose .of projecting and *illustrating the magnitude of dose impacts over the remaining life" of the plant, it is assumed that the concentration levels of activity as of Aprill, 1999 remain constant. Table

1 indicates the measured radioactivity levels for Co-60 and Cs-137 first noted in the soil, and decay corrected to April 1, 1999.

(b) For the analysis of the radiological impact during the Vermont Yankee active control of the disposal sites until 2013, no plowing is assumed to take place and all dispersed radioactive material remains on the surface forming a source of unshielded direct radiation. (c) The crop exposure time was changed from 2160 hours to 0 hours to reflect the condition that no radioactive material is dispersed directly on crops for human or animal consumption. Crop contamination is only through root uptake. (d) The deposition on crops of re-suspended radioactivity is insignificant. (e) The pathway data and usage factors used in the analysis are the same as those used in the Vermont Yankee's ODCM assessment of off-site

radiological impact from routine releases, with the following exceptions.

The :fraction of stored vegetables grown on the contaminated land was conservatively increased from 0. 76 to 1.0 (at present no vegetable crops for human consumption are grown on any of the approved disposal plots). Also, the soil exposure time to account for buildup was changed from the

standard 15 years to 1 year. *

(t) It is conservatively assumed that Vermont Yankee relinquishes control of the disposal sites after the operating license expires in 2012 (i.e., the source tenn accumulated on a single disposal plot applies also for the inadvertent intruder). (g) For the analysis of the impact after Vermont Yankee control of the site is relinquished, the radioactive material is plowed under and fonns a uniform mix with the top six inches of the soil; but nonetheless, undergoes re-suspension in the air at the same rate as the unplowed surface contamination. However, for direct ground plane exposure the self-shielding due to the six-inch plow layer reduces the surface dose rate by about a factor of four.

Revision -62._ Date 1/11/02 H-12

BVY 99-80 I Attachment l /Page 8 As shown in the previous subnilttals, in which the concentrations of Co-60 and Cs-137 in septic waste exceed those identified in the soil, the liquid pathway was found to be an insignificant contributor to the dose. Therefore, the liquid pathway is not considered in this analysis. The dose models and methods used to generate deposition values and accumulated activity over the operating life of the plant are documented the ODCM. Table 3 presents the radioactivity that currently exists on the south field after the last spreading event which occurred on September 28, 1998 (total elapsed time from September 28, 1998 to April l, 1999 is l 84 days). fa addition, the total activity on the south field is presented assuming the addition of the 25.5 m 3 of soil subject of this evaluation. The total activity that would be present on south field at license termination (i.e., total elapsed time of 14 years post April 1, 1999, or 2013), assuming no future additions of material containing radioactivity after disposal of the proposed soil volwne was also evaluated and is presented in Table 4. In order to demonstrate compliance with the boundary conditions, the critical organ and whole body dose from all pathways to a maximally exposed individual during Vermont Yankee control, and to :the inadvertent intruder were calculated. The dose calculations were performed using the dose conversion factors presented in Table 5 and 6 below which were obtained from the ODCM. The contribution to dose from Co-60 and Cs~137 to the whole l-ody and organ at the present and at license expiration is presented in Table

7. The present and future total dose impact from the south field with and without spreading of the soil is presented in Table 8:

These results demonstrate that disposal of the approximately 25.5 m3 of accumulated soil will be well within the accepted dose limit criteria of 1 mrem/yr to any organ or whole body during the control period, and 5 mremlyr to an inadvertent intruder after control of the site is assumed to be relinquished. This analysis shows that significant dcrse margin still exists on the approved disposal plots to accommodate potential future spreading operations. ** 5.0 RADIOLOGICAL PROTECTION The disposal operation of soil piles will follow the applicable Vermont Yankee procedures to maintain doses as low as reasonably achievable and within the specific dose criteria as previously approved for septic waste and cooling tower silt disposal. Revision.22__ Date 1/11/02 H-13

BVY 99-80 I Attachment l /Page 9

6.0 CONCLUSION

S Soil generated from on-site constructiOn activ~ties reflects an earthen type material similar in characteristics to septic waste residual solids and cooling tower silt with respect to the radiological pathway behavior and modeling. Based on the similarity in characteristics between the proposed soil volume and waste streams that have already been approved for disposal, and the evaluation of the incremental dose impact, it is concluded that disposal of the approximately 25.5 m3 of existing soil through on-site land spreading will meet the boundazy conditions specified in the ODCM. That is, with respect to the addition of the approximately 25.5 m3 soil pile to the existing radioactivity already spread oil the south field: ..

1. Total doses to the whole body and critical organ to the hypothetically maximally exposed individual were estimated a3 3.00E-02 rnrem/yr and 1.04E-Ol, respectively, which are less than the prescribed 1.0 mrem/yr. *

2. Total doses to the whole body and critical organ of an inadvertent intruder from the probable pathways of exposure were estimated as l .13E-O 1 mrem/yr and 2.21E-01 mremlyr, respectively, which are less than 5 mrem/yr. *

3. The disposal is assumed to take place on the south field that is the same site approved for disposal of septic waste and cooling tower silts.

If the soil were spread on an approved plot which had not yet been used for disposal, the dose impact from the approximately 25.5 m3 of soil alone would at present be 4. l 7E-03 mrem/yr whole body and a maximum organ dose of 1.46E-02 mrem/yr. In addition, for the inadvertent intruder, the whole body dose would be l.60E-02 mrem/yr, and a maximum organ dose of 3. l lE-02 mremlyr. Each of these doses also meet the boundary conditions specified in the ODCM.

7.0 REFERENCES

(1) Vermont Yankee ODCM, Rev 23, Appendix B, "Approval of Criteria for Disposal of Slightly Contaminated .Septic Waste On-Site at Vermont Yankee". (2) Vermont Yankee ODCM, Rev 23, Appendix F, "Approval Pursuant to IOCFR20.2002 for On-Site Disposal of Cooling Tower Silt". (3) USNRC Regulatory Guide 1.109, Rev 1, "Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance \Vith 10CFR Part 40, Appendix I", dated October 1997. Revision -12_ Date II I I/02 H-14

BVY 99-80 I Attachment l /Page 10 Table 1 Radioanalytical Results of Composite Soil Samples Cs-137 Co-60 (pCi/kg) (pCi/kg) Sample ID April, 1995 April, 1999 April, 1995 April, 1999 G22716 234 213 49** 29 G22717 522 476 143 84 G22718 337 307 37** 22 022719 291 265 Ill 66 G2272,0 348 317 47"'* 28 G22721 135 123 73 43 G22722 107 98 82 48 G22723 222* 203 140 83 G22724 180 . 164 92 54 G22725 269 245 118 70 G22726 810 739 114 67 G22727 378 345 106 63 G22728 810 739 124 73

           .         G22729                                                343                  62               37
          .                                            376 G22730                            331                 302                  87                51 G22731                            253                 231                 5**                3 G22732                             150                 137                 58                34 G22733                            247                 225                 105                62 G22734                            326                 297                54**                32 G2273S                            235                 214                 100                59 Average              328                  299                 85                50 Maximum Value                   .810                  739                143                84 Minimum Value                   107                   98                  s                 3 Standard Deviation                  191                  174                 37                22
        .* Average*wet to dry sample weight ratio equal to 1.12. Av~ragc wet density equal to 1.47 gm/cc.
         ** The apparent response of' the g<µnma isotopic analysis which was less than Minimum Detectable Concentration.

Table2 Estimated Total Radioactivity in Soil Volume Volume Average Concentration Total Activity of Soil Mass (pCVkg-dry) (µCi) Radionuclide (m3) (kg-dry) April, 1995 April, 1999 April,1995 April, 1999 Cs-137 25.S 3.35E+04 328 299 11.0 10.0 Co-60 25.5 3.3SE+o4 85 50 2.8 1..7 Revision 29 Date 1/11/02 H-15

BVY 99-80 I Attachment l / Page 11 Table 3 Tqtal Activity on South Field After Last Spreading Event Total activity after Total activity decay Total activity after last spreading event corrected to April 1, proposed soil Radionuclide (µCi/acre) *1999 (µCi/acre) disposal (µCi/acre) Mn-54 0.17 0.11 0.11

             *co-60                      5.93                     5.55                  6.44 Zn-65                     0.074                    0.044                 0.044 Cs-137                    32.27                    31.90                 37.16 Table4

Total Projected Radioactivity on South Field Remaining at License Expiration Total Activity as of License Expiration Radionuclide (µCi/acre)

Mn-54 8.9E-07 Co-60 0.89 Zn..;65 7.6E-09 Cs-137 2633 Tables All Pathway Critical Organ/Whole Body Dose Conversion Factors During Vemi.ont Yankee Control of Disposal Sites Critical Organ Whole Body Dose Factor Dose Factor Radionuclide Individual/Organ (mrem/yr per µCf/acre) (mremlyr per µCi/acre) Mn-54 Adult/GI-LL! 3.75E-04 l.93E-04 Co-60 T~n/Lung 7.17E-04 5.31E-04 Zn-65 Child/Liv.er 1.64E-02 1.03E-02

Cs-137 Child/Bone 2.66E-03 7.02E;.04 Revision _l2._ Date 1/11102 H-16

BVY 99-80 I Attachment l /Page 12 Table 6 All Pathway Critical Organ/Whole Body Dose Conversion Factors Post Vermont

Yankee Control of Disposal Sites (Inadvertent Intruder)

Critical Organ Whole Body Dose Factor Dose Factor Radionuclide Individual/Organ (mrem/yr per µCl/acre) (mrem/yr per µCi/acre)

                                      ~-=-~+-..,,-~=-=-=---~~~~~+-=-==-=-=-~~----1 Mn-54                 Teen/Lung                l.02E-02                           3.12E-03
                                      ~----~-==------=~~~~~+----..-~~~~--1 Co-60                 Teen/Lung                3.19E.02                           9.09E-03 Zn-65                 Child/Liver---i--,,.1-=.s=9E=---=02-=--------+,1=--"".2="'s=E=--""-0=2=--------1
    .......,_--~~------=-"="""-=----

Cs-137 Child/Bone 6.98E-03 3.85E-03

                                                      . Table 7
        . Dose Contribution from Co-60 and Cs-137 in 25.5 m 3 Soil Volume after Land Spreading Present Dose Impact'               Future Dose lmpacf' (Muimally exposed                        (Inadvertent Intrud.er)

Case individual) Dose Individual/ Dose Individual/ (mrem/yr) Organ (mremlyr) Organ Cobalt-60 4.75E-04 Whole body 129E-03 Whole body_ 6.42E-04 Max.Organ 4.SlE-03 Max. Organ Cesium-137 3.69E-03 Whole body l.47E-02 Whole body IAOE-02 Max.Organ 2.66E-02 Max.Organ 1 Based on inventory ofCo-60 of0.895 µCi/acre and Cs-137 ofS.26 µCi/acre in April 1, 1999. 2 Based on inventory of Co-60 of0.141 µCi/acre and Cs-137of3.82 µ<;ila,cre in April 1,. 2013. Revision -12.... Date 1/11/02 H-17

BVY 99-80 I Attachment 1 /Page_ 13 Table 8 Present and Future Dos~flmpact. Due to the .Soil Spreading

                                                                      . for Two Cases Present Dose Impact             Future Dose Impact (Maximally exposed                   (Inadvertent Intruder)

Case individual) Dose Individual/ Dose Individual/ (mrem/yr) Organ (mrem/yr) Organ Case One South Field as it currently 2.58E-02 Whole body 9.70E-02 Whole body

     ~xists                          8.96E-02      Max. Organ        l.&9E-Ol        Max. Orrran Case Two South Field if disposal of      3.00E-02 . Whole body*       l.13E-Ol        Whole body soil volume is approved         1.04E-02      Max. Organ        2.21E-01         Child/Bone Increase in dose impact         4.17E-03      Whole body        l.60E-02        Whole body from disposal of soil           1.46E-02      Max.Organ         3.llE-02        Max. Organ Revision _12.._ Date  1/11/02 iI-18

             -.      V.ERMONT YANKEE NUCLEAR POWER CORPORATION 185 Old Ferry Road, Brattleboro, VT 05301-7002 (802) 257-5271 January 4, 2000 BVY 00-02 United States Nuclear Regulatory Commission ATIN: Document Control Desk Washington, DC 20555

References:

(a) Letter, VYNPC to USNRC, "Request to Amend Previous Approvals Granted Under 10 CFR 20.302(a) for Disposal of Contaminated Septic Waste and Cooling Tower Silt to Allow for Disposal of Contaminated Soil," BVY 99-80, dai:ed June 23, 1999

Subject:

Vermont Yankee Nuclear Power Station License No. DPR-28 (Dock~t No. 50-271) Supplement to Request to Amend Previous Approvals Granted Under 10 CFR 20.302(a) to Allow for Disposal of Contaminated Soil Reference (a) provided Vermont Yankee's application to amend the prev~ously granted approvals to dispose of slightly contaminated septic waste and cooling tower silt on-site to include slightly contaminated soil generated as a residual by-product of on-site construction activities. *Tue request was to allow the disposal of approximately 25.5 cubic meters of waste that has been accumulated to date and to allow for disposal of future waste from construction related activities. Based on discussions with USNRC staff, additional foformation related to the estimated volume and dose consequences of the proposed future material was needed to complete your review. Attachment 1 has been revised accordingly to include the *information requested. Attachment 1 supercedes the evaluation previously submitted. We trust that the infonnation will allow you to complete your review of our submittal. However; should you have any questions or require further information concerning this matter, please contact Mr. Jim DeVincentis at 802-258-4236. ~ Sincerely, Ve~ont Yankee Nuclear Power Corporation

                                                                 ""'Gautam Sen          /.

Licensing Manager Attachment cc: USNRC Region I Administrator USNRC Resident htspector - VYNPS USNRC Project Manager - VYNPS VT Department of Public Service H-19 Revision .12..._ Date II 11 /02

SUMMARY

OFVERMONTYANKEECO.MMITMENTS BVY NO.: 00-02 The following table identifies commitments made in this docnment by Vermont Yankee. Any other actions discussed in the submittal represent intended or planned actions by Vermont Yankee. They are described to the NRC for the NRC's information and are not regulatory commitments. Please notify the Licensing Manager of any questions regarding this document or any associated commitments. COMMITMENT ,.. COMMITTED DATE 11 OR "!JUTAGE" ..

jl None NIA VYAPF 0058.04 AP 005 8 Original Page 1 of I Revision .l2_ Date 1/11/02 H-20

Docket No. 50-27 l BVY00-02 Attachment 1 Vermont Yankee Nuclear Power Station Assessment of On-site Disposal of Contaminated Soil by Land Spreading Revision .12_ Date l/11102 H-21

BVY 00-02 I Attachment l / Page I of 17

                                      -* TABLE OF CONTENTS Page LIST OF TABLES           *...........**.....**....*......*........*.*******..***.*................... 2 1.0,. INTRODUCTION                                                                                       3 1.1 Background           *******~***************************************~*********************    3 1.2 Objective                                               . :.
                                *.*...*.**.*...................................................***.****    3.

2.0 WASTE DESCRIPTION ........... ~*............................................... :.... 4 3.0 SOIL DISPOSAL AND AO:MINISTRATIVE PROCEDURE REQUIRE:rvlENTS .................................................... :..................... 5 4.0 EVALUATION OF ENVIRONMENTAL IMPACTS ........................... .,... 6 4.1 Site Characteristics ................................... : ...*............... ~ ........ ~. 6 4.2 Radi0Iogical Impact.................................................................. 7 5.0 RADIOLOGICAL PROTECTIO~ .... ~ .............................................. _.. 9

6.0 CONCLUSION

S........................................................................... 9

7.0 REFERENCES

............................................................................. 11           ,.

Revision~ Date 1/11/02 H-22

BVY 00-02 /Attachment l /Page 2of17 LIST OF TABLES Table 1: Radioanalytical Results of Composite Soil Samples...........................

12 .

Table 2: Estimated Total Radioactivity in 25.5 m3 Soil Volume._. ................... :. 12 Table 3: Estimated Total Radioactivity in Future Soil Additions....................... 13 Table 4: Record of Septic and Sil~ Radioactive Material Spread each Year on South Field Disposal Plot.................................................. :**".................. 13. Table 5: Total Projected Radioactivity Remaining on Soutli"Filed at License* Exprration . .......................................................................

                                  .                         .                                            .       14 Table 6: All-Pathway Critical Organ/Whole Body Dose Conversion Factoci. .. . .....                            14 Table 7: Dose Impact from Past Septic and Silt Spreading Activity .. ~.................                       15 Table 8: Dose.Impact from Past Septic and Silt Spreading 'and Single 25.5~3 Soil Disposal .................................................................... :.:~~::*:.~*-::-*- *-i 5 Table 9: Dose Impact from Present and Future Soil Disposal Along with Past Septic and Silt Disposal........................................................................             16 Table 10: Dose Impact from Past Disposals through 7/15/99 Plus all Annual Projected Disposals of Septic, Silt and Soil......................................................              16 Table 11: Summary of Dose Imp.acts Associated with Different Disposal Scenarios .. ................................................................... _. ~ ............*   17 Revision .l2._  Date  1/11/02 H-23

BVY 00-02 /Attachment 1/Page3of17 1.0. INTRODUCTION i.1 Background In 1989, Vermont Yankee Nuclear Power Corporation requested approval from the NRC to routinely dispose of slightly contaminated septic waste in designated on-site areas in accordance with 10CFR20.302(a). Approval from the NRC was granted on August 30, 1989 and the information was pennanently incorporated into the Offsite Dose Calculation Manual (ODCM) as Appendix B. . In 1995, Vermont Yankee Nuclear Power Corporation requested that the previous authorization for on-site disposal of very low-level radioactive material in septic waste be amended to permit the on-site disposal of slightly contaminated cooling tower silt material. Approval from the NRC was granted ori June 18, 1997 and the information was pennanently ** incorporated into the ODCM as Appendix F.

In 1994, approximately 25.5 m3 of excess soil was generated during on-site construction activities. Subsequent sampling and analysis of the soil revealed low levels of radioactivity that were similar in radionuclides and activity to the septic waste and.cooling tower silts previously encountered. An evaluation determined that the soil could be managed in similar fashio.n as the septic waste and cooling tower Silts; however, prior approval from the NRC would be required unde.r 10 CFR 20.2002 (formerly 20.302(a)).

1.2 Objective The purpose of this assessment is to present the data and radiological evaluation to demonstrate that the proposed disposition of the soil (i.e., on-site disposal via. land spreading on designated fields) will meet the existing boundaiy dose conditions as approved by the NRC for septic waste and cooling tower silt. The boundary conditions established for disposal of the septic waste and cooling tower silt on designated plots are:

        *l. The dose to the whole body or any organ of a hypothetical maximally exposed individual must be less than 1.0 mrem/yr.

2. Doses to the whole body and any organ of an inadvertent intruder from the probable pathways of exposure are less than 5 mrem/yr.

3. Disposal operations must be at one of the approved on-site locations.

Revision 29 Date 1/11/02 H-24 .* *'

BVY 00-02 / Attachment l /Page 4 of 17 2.0 WASTE DESCRIPTION The existing accumulation of contaminated soil was derived from excavation* activities associated with the construction of a new security fence along the plant's Protected Area boundary. The volume of soil generated was approximately 25.5 m 3, and is typical of fill material containing light to dark brown poorly sorted soils with some small stones, and includes small incidental pieces of asphalt. The soil was removed from its original location by shovel, backhoe and front-end loader, and placed into dump trucks for transport to a location between the cooling tow~rs where it was deposited on the ground surface and covered to prevent erosion. This location was selected because it was away from areas routinely occupied by plant staff, and could easily be controlled. The most probable source of the low levels of radioactive contamination is the presence qfbe1ow detectable removable contamination redistributed by foot traffic from inside the plant to walkways and.parking areas. Subsequent surface ronoff carries the contamination to nearby exposed soil near the Protected Area boundary where it had accumulated over time to low level detectable concentrations. In April 1995, a total of20 composite soil samples were collected to characterize the accumulated volUm.e. Compo~tes were obtained by taking a grab sample from opposite sides of tl;ie pile and the top at equal distances along its length. These three grab samples were then combined into one composite sample. Soil samples were sent to the Yankee Atomic Environmental Laboratory for analysis and counted to environmental lower limits of detection reqwxed of enviranmental media Results of the analys-es are presented in Table ! . For estimating the total activity in the soil pile, the actual analytical result was used for those samples that were less than the :MDC to calculate the average radioactivity concentration: Analytical results are provided for both the times when the samples were collected as well as decay corrected to the present (7/15/99). The results identified both Cs-137 and Co-60 in most of the composite samples, which verified that plant-related radioactivity, was present in the soil. The mass of accumulated soil (dry) was estimated by multiplying the total in-situ volume (25.S m 3) by it's wet density, l.47E+03 kg/m3;and then dividing by the wet:drj ratio of l.12,

       *thtis yielding a mass of 3.35E+04 kg (dry). The mass of the soil was then multiplied by the*

average measured Co-60 and Cs-137 concentrations to obtain the total activity of each-' 3 radionuclide in the 25.5 m 3

Table 2 presents the estimated total radioactivity in the 25.5 m volume at the time of sample collection and analysis, and decay corrected to the date of the most recent disposal (septic waste) spreading operation (i.e., July 15, 1999).

In addition to the existing 25.5 m3 (900 ft3) of soil included in this request, it is anticipated that the need to dispose ofvecy low-level contaminated soil will occur in the future. Each spring, approximately 28.3 m 3 (1000 rt3) of road and walkway sand spread during the winter season is swept up from inside the Protected Areii. This material is subject to the same contamination mechanisms that are believed to have lead to the observed contamination in the construction fill removed from within the Protected Area in the past. For purposes of Re\*ision ~9 Date~ H-25

BVY 00-02 I Attachment l I Page 5 of 17 evaluating the radiological impact of potential future soil disposals, it is assumed that an 3 additional 28.3 m per year of sand I soil is contaminated at the same concentration levels as originally observed (April 1995) in the currently collected 25.5m3 of soil. It is also assumed that this material is placed on the same approved disposal field used for all past septic and cooling tower silt disposal operations. Table 3 shows the estimated amount of radioactivity associated with the annual disposal of the 28 .3m3 of soil. It is assumed that this material is disposed of each year for the next 14 years (until the end of plant operating license in 2013) on the same field (South Disposal Plot) along with the continued application of septic waste and cooling tower silt. Table 4 shows a* record of the actual amount of septic/silt.material that bas been spread on.* south field for the past 10 years. A review of the actual waste disposal operations show that the annual average radioactivity content placed on the 1.9 acre South field from septic and silt disposals are as follows:

Mn-54 0.147 uCi/year Co-60 2.58 uCi/year Zn-65 0.269 uCi/year Cs-134 *0.010 uCi/year Cs-137 6.21 uCi/year The maximum radioactivity inventory resulting from the accumulated buildup of past and projected future disposal operations (i.e., septic waste, cooling tower silt_.' pl~ the existing 25:5m3 of accumulated soil along with a projected annual addition of28.3m" of soil each year until the termination of the operating license) is shown on Table 5.

3.0 SOIL DISPOSAL AND ADMINISTRATIVE PROCEDURE REQUIREMENTS The method of soil disposal will use the technique of land spreading in a manner consistent ** with the *current commi1ments for the on-site disposal of septic waste and cooling tower silts as approved by the NRC. The acc:umulatlon of radioactivity on the disposal plot f<?r this soil . spreading operation will be treated as if cooling tower silt or septic waste was being disposed of since the characteristics.of all these residual solids are similar (earthen-type matter). *The south field (approximately 1.9 acres in size) has been designated and approved for septic waste and _cooling tower silt disposal and has been used for all past disposal operations, and is expected to be used for the placement of soil. Determination of the radiological dose impact has been made based on the same models and pathway assumptions used in the previous submittals.

Dry soil material will be dispersed using typical agricultural dry bulk surface spreading practices in approved disposal areas on-site. Incidental pieces of asphalt and stones that are picked up with the soil will be screened out before the soil is spread and disposed of as radioactive material at an off-site licensed. facility if detectable radioactivity is found.

Revision.12..... Date 1/11/02 H-26 *'

1' BVY 00-02 /Attachment l /Page 6 of 17 Records of the disposal that will be maintained mclude- the following: .* (a) The radionuclide concentrations detected in the soil (measured to environmental lower limits of detection). (b) The total volume of soil disposed of. (c) The total radioactivity in the disposal operation as well as the total accumulated on each disposal plot at the time of spreading. (d) The plot on which the soil was applied. (e) Dose calculations or ~um allowable accumulated activity determinations required to demonstrate that the dose limits imposed on fue land spreading operations have not been exceeded. To ensure that the addition of soil containing low levels of~dioactivity will not exceed the boundary dose limits, each new spreading operl:!-tion will require an estimate of total radioactivity that includes all past disposals of septic waste, cooling tower silt and soil material on the designated disposal plots. This will be compared with the boundary dose limits or equivalent radioactivity limits on a per acre basis: In addition, concentration limits applied to the disposal of earthen type materials (dry soil) restrict the placement of small volumes of materials that have relatively high radioactivity concentrations. Any fanner leasing land used for the disposal of soil (or other approved waste) will be notified of the applicable restrictions placed on the site due to the spreading oflow level contaminated material. These restrictions are the same as detailed for the j?reviously approved septic waste spreading application. 4.0 EVALUATION OF ENVIRONMENTAL IMPACTS 4.1 Site Characteristics All designated disposal sites are located on !}le Vermont Yankee Nuclear Power Plant site and are within the site boundary security fence. The south field consists ~f approximately 1.9 acres and is centered approximately I 500 feet south of the Reactor Building. This parcel of land has been previously approved by the NRC for the land disposal of septic waste* and cooling tower silt, and is the* only portion of the approved disposal areas which has been utilized to-date for the spreading of contaminated material. For estimating the maximum future radiological impact, it is assumed in the analysis that all future disposal qperations will continue to use the South field as the disposal plol I *1.___R_cv-is-io_n_2_9~D-a-tc--l/-ll-/0-2----' H-27

L. BVY 00-02 I Attachment I I Page 7 of 17 4.2 Radiological Impact The amount of radioactivity added to the south field is procedurally controlled to ensure that doses are maintained within the prior approved limits of the bowidary conditions. To assess the dose received by the maximally exposed individual during the period of plant controls over the property, and to an inadvertent intruder after it is assumed plant access controls end, the same pathway modeling, assumptions and dose calculation methods as approved for septic and cooling tqwer silt disposal were used. These dose models implement the methods and dose conversion factors as provided in Regulatory Guide 1.109, Revision 1 (1977). .. The following six potential pathways we;re identified and blcluded in the analysis; (a) Standing on contaminated *ground. (b) Inhalation of resuspended radioactivity. (c) Ingestion ofleafy vegetables. (d) Ingestion of stored vegetables. ( e) Ingestion of meat. * (f) Ingestion of cow's milk. . As shown.in the previous application for septic waste disposa!, the liquid pathway was found to be an insignificant contributor to the dose for tlie radionuclides identified fixed in the soil type matrixes associated these waste forms. Therefore, the liquid pathway is not considered in this analysis. Both the maximum individual and inadvertent intruder are assumed to be exposed to these pathways, the difference between them being due to the occupancy time. The basic assumptions used in the radiological analyses include: * (a) Exposure to ground contamination and re-suspended radioactivity is for a period of 104 _ho~ per year during the Vermont Yankee active control of the disposal sites and continuous thereafter. The 104-hour interval is representative of a farmer's time spent on a plot of land (4 hours per week for 6months). (b) For the purpose of projecting and illustrating the magnitude of dose impact over the remaining life of the plant, it is assumed that future disposals of

septic and silt material will be placed annually on the same field at the annual average radioactivity levels observed for these waste streams over the past ten years. The future disposals will also consist of the additional 28.3 m3 (1000 ft3) annual volume of new soil at the same radioactivity concentrations observed at the time of collection of the existing 25.5 m3 soil :volume. The maximwn individual dose impact from the buildup of disposed material Revision 29 Date 1/11/02 H-28

BVY 00-02 I Attachment l / Page 8 of 17 occurs at the s~.e time (2013) for both the Control Period and Intruder scenarios. (c) For the analysis of the radiological impact duri.D.g the Vermont Yankee active control of the disposal sites until 2013, no plowing is assumed to take place and all dispersed radioactive material remains on the surface fanning a source of unshielded direct radiation. (d) The crop exposure time was changed from 2160 hours to 0 hours to reflect the condition that no radioactive material is dispersed directly on crops for hum~ or animal consumption. Crop contaminati9n is only through root uptake. (e) The deposition on crops of.re-suspended radioactivity is insignificant. (f) Most of the pathway data and usage factors used in the analysis are the same as those used in the Vermont Yankee's ODCM assessment of off-site radiological impact from routine releases. The fraction of stored vegetables grown on the contaminated land was conservatively increased from 0. 76 to 1.0 (at present no vegetable crops for human consumption are grown on any

of the approved disposal plots). For each year's spreading operations, the soil exposure time to account for buildup was changed from the standard 15 years to 1 year.

(g) It is conservatively assumed that Vermont Yankee relinquishes control of the disposal sites after the operating license expires in 2013 (i.e., the somce term accumulated on a single disposal plot applies also for the inadvertent intruder). (h) For the analysis of the impact after Vermont Yankee control of the site is relinquished, the radioactive material is plowed under and forms a uniform ~ mix with the top six inches of the soil; but nonetheless, undergoes re-stispension in the air at the same rate as the unplowed surface contamination.

However, for direct ground plane exposure the self-shielding due to the six- *.

inch plow layer reduces the surface dose rate by about a factor of four. The dose models and methods used to generate deposition values and accumulated activity over the operating life of the plant are documented in the Vermont Yankee ODCM. The total activity that would be present on south field at the end of the operating period (i.e.* total elapsed time of 14 years post July 15, 1999, or 2013) from the buildup of all waste streams (i.e., septic, cooling tower silt and soil) is presented in Table 5. Revision 29 Date 1111102 H-29

BVY 00-02 /Attachment 1 /Page 9 of 17 In order to evaluate the dose i~pact associated with the different disposal streams, a dose

     *assessment was performed for the following four disposal scenarios:

(I) Impact from past septic and silt spreading only - Table 7 (II) Impact from past septic and silt spreading, plus a single 25.5m3 soil disposal operation for the existing accumulated soil - Table 8 (III) Impact from past septic and silt disposals along with the existing 25.5 m3 of accumulated soil and postulated future annual soil disposal volumes (28.3 m3

                       /yr). -Table 9         _

(IV) Impact from past septic and silt disposals plus annual projected disposals of septic, silt and soil. -Table 10 For each scenario, the critical organ and ~hole body dose fiom all pathways to a maximally exposed individual for both the Vermont Yankee control period and the inadvertent intruder situation were calculated. The dose calculatioris were performed using the dose conversion factors presented in Table 5, which were obtained from the Vermont Yankee ODCM, Appendix F, "Approval Pursuant to 10CFR20.2002 for On;.Site Disposal of Cooling Tower Silt"

      ./f.. summary of the calculated dose impact associated with the four different scenarios is shown in Table 11. These results demonstrate that disposal of the 25.5 m3 of accmnulated soil will be well within the accepted dose limit criteria of 1 mrem/yr to any organ or whole body during the control J?eriod, and 5 mrem/yr to an inadvertent intruder. In addition, if continued soil spreading is necessary, the resulting dose is expected to also remain below the established limits even assuming the annual application of already approved disposal media (i.e., septic waste and cooling tower silt).

5.0 RADIOLOGICAL PROTECTION The disposal operation of the soil will follow the applicable Vermont Yankee procedures to maintain doses as low as reasonably achievable and within: the* specific dose criteria as previolisly approved for septic waste and cooling.tower silt disposal. *

6.0 CONCLUSION

S Soil generated from on-site construction and maintenance activities constitutes an earthen type material similar in characteristics to septic waste residual solids and cooling tower silt with respect to the radiological pathway behavior and modeling. Based on the similarity in characteristics between the proposed soil volume and waste streams that have already been approved for disposal and the evaluation of the incremental dose impact, it is concluded that the disposal of the existing 25.5 m3 and the projected 28.3 m3/year of soil through on~site land spreading will meet the existing NRC approved boundary dose conditions specified in

                                                     . H-30 Revision .l2..._  Date  1111/02

BVY 00-02 /Attachment l /Page 10of17 the Vermont Yankee 0 DCM (see Appendix B for Septic Waste Disposal). That is, with respect to the addition of the initial 25.5 m 3 of soil along with the projected 28.3 m3/year of soil and the projected future disposal of septic and silt waste to the existing radioactivity . already spread on the south field:

1. Total doses to the whole body and critical organ of the hypothetically maximally exposed as individual were estimated l. l 5E-O I mrem/yr and 4.03E-O 1 mrem/yr, respectively,

      *which are less than the prescribed 1.0 mrem/yr limit during the period of active site control.                   *

2. Total doses to the whole body and critical organ of an inadvertent intruder from the probable pathways of exposure were estimated as 7.57E~Ol nirem/yr and 1.17 mrem/yr, respectively, which are less than 5 mrem/yr limit associated with an intruder scenario following assumed loss of site acce5s contro~ as the end of the operating license.

3. *For pUIJJoses of projecting maximum impact, all disposals (past and future) are assumed to take place on the south disposal plot.

Therefore, the disposition of the present 25.5 m 3 and the projected 28..3 m3/year of soil will

    ~ntinue to meet the existing boundary conditions as approved by the NRC for septic waste and cooling tower silt.

Revision 29 Date 1/11/02 H-31

BVY 00-02 I Attachment l /Page 11 of 17

7.0 REFERENCES

(I) Vermont Yankee ODCM, Rev 23, Appendix B, "Approval of Criteri?- for Disposal of Slightly Contaminated Septic Waste On-Site at Vermont Yankee". (2) Vermont Yankee ODCM, Rev 23, Appendix F, "Approval Pursuant to 10CFR20.2002

       -"for On-Site Disposal of Cooling Tower Si.It".

(3) USNRC Regwatory Guide 1.109," Rev I, "Cal~ation of Annuai Doses to Man fro!ll Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with : 10CFR Part 40, Appendix I", dated October 1977. .. . H-32 Revision -22._ Date l/ 11/02

T BVY 00-02 /Attachment I/ Page 12of17 Table 1 Rndioanalytical Results of Composite Samples Taken from 2S.5m3 Soil Pile Cs-137 Co-60 (pCilkg) (pCi/kg) Sample ID April. 1995July15, 1999 April. 1995 July 15. 1999 G22716 234 212.2 49 28.0

G22717 522 473.4 143 81.7 .

022718 : 337 305.7 37 21.1

022719 291 263.9 111 63.4 022720 022721 348 135 315.6. .. 47 26.8
122.4 ., 73 41.7 G22722 107 97.0 82 46.8 G22723 222 201.4 140 . 80.0 022724 180 163.3 92 52.6 022725 269 244.0 118 67.4 G22726 810 734.7 114 65.1.

G22727 378 342.8 10~ 60.6 022728 810 734.7 124 . 70.8 022729 376 341.0 62 35.4 . 022730 331 300.2 87 49.7 022731 253 229.5 5 2.9

022732 150 136.0 58 -33.1 022733 247 224.0 105 60.0 G22734 326 295.7 54 30.8 "'

                   *G22735                         235:       213.1                    100        57.1 Average       328          298                    85           49 Maxim.urn      810          735                    143          82 Minimum        107           97                     s            3 Std Dev.      186          169                    36           20
        *The apparent response of the gamma isotopic analysis was less than the fyfinimum Detectable Concentration.

Table2

Estimated Total Radioactivity in 25.5m3 Accumulated Soil Volume Soil Average Concentration Total Activity ofSoil Mass (pCi/kg - dry) (uCi)

Nuclide (m3) {kg - dOO April 1995 July 15, 1999 April 1995 July 15. 1999 Cs-137 25.5 3.35E+o4 328 298 11.0 10.0 Co-60 25.5 3.35E+o4 85 49 2.8 1.6 Revision .1.2... Date 1/11/02 H-33 ... ~.*.**

BVY 00-02 /Attachment l /Page 13of17 Table3 Estimated T~.tal.Radioactivity in Future Soil Additions Volume. Soil Average Concentration of soil Mass (pCi/kg-dry assuming) Total Activity Nuclide ill[}_ (kg-dry) (April. 1995 concentrations) (uCi/yr) Cs-137 28.3 3.72E+o4 328 12.8

         . Co-60           28.3           3.72E+o4                      85                           3.16 Table4 Rc~o::-d ofS.ej_Jtk :tnd Silt R.-..dfoncfr1c Mat.erutl Spread E;.lch Ye:4l" on the South Field Spreading        Material Mn-54 .* Co-60                Zn-65       Cs-134         Cs137        Ce-141 Year         Date            Type (uCi/acre) (uCi/acre) (uCi/acre) (uCi/acre)              (uCi/acre)   (uCi/acre) 1990       10/31/90         Septage         0.00       3.89        0.00         0.00          0.26           0.00 1990       11/20/90         Septage         0.17       2.03        0.41         0.00          0.29       1.40E-08 1991                           no           0.00       0.00        0.00         0.00          0.00           0.00 spreading 1992       10/19/92         septage         0.11       1.73        0.52         0.05          0.32         0.006 11993        10/14/93         septage         0.05       1.41        0.21         0.00          0.30           0.00 1994       06/14/94         septage         0.08      *0.43        0.00         0.00          0.09           0.00 1995       06129195         septage         0~00       0.88        0.00         0.00          0.00            0.00 1996                           no           0.00.      0.00         0.00        0.00          0.00         .."0.00 spreading 1997       06/18/97         septage         0.12       1.00         0.00        0.00          0.19            0.00 1998       07/30/98         septage         0.14       0.72         0.09        0.00          0.12            0.00 1998       09/28/98            Silt         0.00       0.00         0.00        0.00          30.87           0.00 1999 *07/1S/99              Septage         0.11        1.47        0.20        0.00           0.25           0.00 Average Activity/yr                 .

(uCi/acre): 0.08 1.36 0.14 0.01 327 0.01 Average Activity (uCi/yr) to be 0.147 2.58 0.269 0.010 6.21 0.001 not disposed ofon 1.9 significant acre field Revision -12._ Date 1111102 H-34

BVY 00-02 /Attachment I /Page 14of17 Table5 Total Projected RadioactiVity Remaining on South Field after.License Termination Accum. Activity Accum. Activity Accum. Activity Accum. Activity in Silt & Septic in Soil Total All Paths Total All Paths

                                  @Yeai2013             @year2013                     @year2013             @ Year2013

Nuclide CuCi) (uCi) (uCi) (uCi/acre)*

Mn-54 0.26 0.26 0.14 Co-60 19.68 21.83 41.51 21.85 Zn-65 0.42 0.42 0;22 Cs-137 119.03 154.74 273.78 144.09 Cs-134 0.04 .. 0.04 0.02

    ** The total activity is assumed to be spread on the 1.9 acre South field to generate the uCi/acre value.

Table6 All Pathway Critical Organ/Whole Body Dose Conversion Factors During VY Control Post VY Control Untrnder Scenario) Critical Organ Whole Body Critical Organ Whole Body

                                              *Dose Factor        Dose Factor                Dose Factor       Dose Factor (mrem/yr per       (mrem/yr per               (mrem/yr per      (mrem/yr per Radionuclide        JndividuaUOrgan        µCi/acre)          uCi/acre)                  uCi/acre)         uciJacre)

Mn-54 Adult/GI-LLI 3.75E-04 l.93E-04 l.02E-02 3.12E-03 Co-60 Teen/Lung - *

7.17E-04 . S.31E-04 3.19E-02 9.09E-03 Zn-65. Child/Liver . 1.62:E~o2 l.03E-02 l.89E-02 1.25E-02 Cs-137 Child/Bone 2.66E-03 7.02E-04 6.98E-03 3.SSE-03 I Revision~ Date 1/11/02 H-35

L BVY 00-02 I ~ttachment l / Page 15 of 17 Table 7 (Scenario D Dose Impact from Past Septic and Silt Spreading on South Field (as of 7/15/99) Control Sccmarlo: All Other Spreadings Maximum Organ Whole Body Maximum Whole Body Half.Life: To Date Dose Facuir Dose: Fa.c:tor Organ Dose Dose {Years) {uC"ilacrc:) Cmmnlvr/uC"u'acrcl (m:m'yr!uCilacrc) (mre~} (mrcm/11r1

        . Mn-54            0.86          0.20                               J.75E--04               l..93E-04      7.JSE.QS           3.78E-05 c~               S.21           6.86                              7.17E.Q4                S31E.Q4        4..92E.Q3          3.64E-03 Zn-OS          *.0.67           0.23                               l.62E-02               l.03E-02       3.77E.(J3          2.40E-03 Cs-137          30.17          31.92                               2.66E--03           *. 7.ll2E.Q4      8.49E-02           2.24E-02 Tot!! D'.l£":    9J7E-02      .1    2.35E-02 Dose Limit:            l                  l
                                                                                    .,             %ofllmit:          9.37%             2.85%

IDtruder Scenario: All Other Ai:tivity on Plot Spreadings Decayed to Maximum Org3n Whole Body Maximum Whole Body Half-Life To Date Ycar2013 Dosc:Fador Dose Fader OrganDosc Dose mml (uCVacre} (uCl/acre) (mrcm/yrlilC"u'a.cre) (mrern/yr/uCi/acre} !~mlYr} {mrem/11r) Mn-S4 O.M 020 2.31E-06 l.02E-02 3.12E..()J 2.36E-08 7.22E--09 Co-60 S.21 6.86 l.08E+-OO 3.19E-02 9.09E-03 3.46E-02 9.SSE--03 Za-05 0.67 0.23 l.15E47 l.89E-02 l.25E-02 2.17E-09 l.43E-09

Cs-137 30.17 31.92 2.31E+ol 6.98E.Q3 3.SSE-03 1.62E-01 B.91E-02
Total Dose: l.96E-Ol 9.!IOE-02 Dose Limit: s s

                                                                                                   %oCLimit:          3.92%              1..98%

Table 8 {Scenario ID Dose Impact from Past Septic/Silt Spreading and Single 25.Sm3 Soil Disposal Control Scenario: All Spreadings Maximwn Organ Whole Body Maximum Whole Body Half:*Life to Dau:. Dose Factor Dose Fir;tor Organ Dose Dose (Years) (uCl/acrc) (mR:m/yr/uCl/acrc) Cmrc:m/yr!uCi/acre) {mn:m/vr} [IJll"Cmfvr) Mn-S4 0.86 0.196 3.7SE--04 l.93E.Q4 7.3SE..OS 3.78E-O.S C4Hi0 S.21 7.70 7.17E--04 S.JIE--04 S..52.E-03 4.09E-03 Zn-OS 0.67 0.233 l.62.E-02 1.03E-02 3.77E.Q3 2.40E-03

         . C&-137            30.17          37.19.                            2.66E-03               7.02E.Q4 . 9.89E-02*          2.61E*Dl

Total Dose: l.08E-Ol 3.26E-02 DoscLimit: l : l o/e oCLimit: 10.83% 3-26%

Intruder Scenario: All Spreadings Activity on Plot Maximum Otgen Whole Body Maximum Whole Body Half-Ufc to Date Decayed to 2013 Dose Fa.ctor DoscFacror Organ Dose Dose ('[C31"S) (uCi/acrc} (uCi/acrel Cmrc:mfyr/uCi/z=l Cmrc:m/yr/uC"vacrc:l {mn:!!!b'.r} ~ (mremf~*r) Mn-S4 0.86 0.196 2.31E.Q6 l.02E-02 3.12E-o3 2.J6E.()8 7.22E-09 Co-00 S.21 7.70 l.22E+-OO 3.19E.Q2 9.09E-03 3.SSE-02 1.1 IE-02 Za-6S 0.67 0.233 1.ISE.07 l.89E-02 1.2SE.Q2 2.17E-09 1.43E.Q9 Cs-137 30.17 37,19 2.70E-+<ll 6.98E.Q3 3.85E.Q3 l.SSE-01 l.04E-01 Total Dose 2.27E-Ol l.lSE-01 Doscl.imit

s 5

                                                                                                    %ofLimit:          4.S4%             2.30%

Revision ..22_ Date 1111/02 H-36

UNITED STATES NUCCEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 June*lSl. 2000 NVY 00-58 Mr. Samuel L Newton Vice President, Operations Vermont Yankee Nuclear Power Corporation 185 Old Ferry Road Brattleboro, VT 05301

SUBJECT:

VERMONT YANKEE NUCLEAR POWER STATION, REQUEST TO AMEND PREVIOUS APPROVALS GRANTED UNDER 10 CFR 20.302(a) TO ALLOW FOR DISPOSAL OF CONTAMINATED SOIL (TAC NO. MA5950)

Dear Mr. Newton:

By letter dated June 23, 1999, as supplemented on January 4, 2000, you submitted a request to amend a previously approved application granted by the Nuclear Regulatory Commission (NRC) pursuant to 10 CFR 20.2002 (previously 10 CFR 20.302) to allow the addition of slightly contaminated soil and soiVsand material to the list of already approved materials*(Le., septic waste and cooling tower silt) for on-site disposal via land spreading on designated fields. We have completed our review of your proposal and find it to be acceptable because the previously approved bounding conditions will continue to be met. . Pursuant to the provisions of 10 CFR Part 51, the NRG has published an Environmental Assessment and Finding of No Significant Impact in the Federal Register on Jtme 15 , 2000 ( 65 FR 37583 ). . .. Since~+-~2-Q

                                                        *J~.       zw:nnski,  ~irector Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket No. 50-271

Enclosure:

Safety Evaluation cc w/encl: See next page Revision ..l2_ Date 1111/02 H-37

Vermont Yankee Nuclear Power Station cc: Regional Administrator, Region r' Mr. Raymond N. McCandless U. S. Nuclear *Regulatory Commission Vermont Department of Health 475 Allendale Road Division of Occupational King of Prussia, PA 19406 and Radiological Health 108 Cherry Street Mr. David R. Lewis Burlington, VT 05402 Shaw, Pittman, Potts & Trowbridge 2300 N Street, N.W. Mr. Gautam Sen

  -*     Washington, DC 20037-1128                Licensing Manager*

Vermont Yankee Nuclear Power Mr. Richard P. Sedano, Commissioner Corporation Vermont Department of Public Seivice 185 Old Ferry Road 112 State Street P .0. Box 7002 Montpelier, VT 05620-2601 Brattleboro, VT 05302-7002 Mr. Michael H. Dworkin, Chairman Resident Inspector . Public Seivice Board Vermont Yankee Nuclear Power Station t

State of Vermont U. S. Nuclear Regulatory Com.mission 112 State Street P.O. Box 176 .

Montpelier, vr ~5620-2701 Vernon, VT Os354 ** Chairman, Board of Selectmen Director, Massachusetts Emergency Town of Vernon Management Agency

P.O. Box 116 ATTN: James Muckerheide Vernon, VT 05354-0116 400 Worcester Rd.

Framingham, MA 01702-5399 Mr. Richard E. McCullough Operating Experience Coordinator Jonathan M. Block, Esq. Vermont Yankee Nuclear Power Station Main Street P.O. Box 157 . ' P. 0. __Box 566 qovernor Hunt Road *Putney, VT 05346-0566 Vernon, VT 05354 G. Dana Bisbee, Esq. Deputy Attorney General

33 Capitol" Street Concord, NH 03301-6937 Chief, Safety Unit Office of the Attorney General One Ashburton Place, 19th Floor Boston, MA 02108 Ms. Deborah 8. Katz Box83 Shelburne Falls, MA 01370 H-38 Revision..2.2.... Date 1/11/02

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 205$-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION VERMONT YANKEE NUCLEAR POWER STATION DOCKET NO. 50-271

1.0 INTRODUCTION

By letter dated June 23, 1999, as supplemented on January 4, 2000, Vermont Yankee Nuclear Power Corporation (the licensee), submitted a request to amend a previously approved

application granted by the Nuclear Regulatory Commission (NRC) pursuant to 1o CFR 20.2002

        *(previously 10 CFR 20.302) to allow the addition of slightly contaminated soil and soil/sand material to the list of already approved materials (i.e., septic waste and cooling tower silt) for on-site disposal via land ~µ.reading on designated fields.

In 1989, pursuant to 10 CFR 20.302 (current 10 CFR 20.2002), the licensee received approval from the NRC to routinely dispose of contaminated septic waste in designated on-site areas. In 1997, the NRC amended the approved on-site disposal application to also.include contaminated
cooling tower silt material.

In this 10 CFR 20.2002 amendment application, the licensee identified 25.5 cubic meters of soil to be disposed of on-site immediately, and approximately 28.3 cubic meters of soil/sand material to be disposed of on an annual basis until the expiration of the plant's operating ljcense in 2013. The 25.5 cubic meters of contaminated soil were generated as a result of on-site construction activities. The anticipated 28.3 cubic meters of soil/sand material will be generated from the annual winter spreading of sand on roads and wal.kways at the plant site. The licensee has performed a comprehensive radiological evaluation that includes all of the anticipated materials (i.e., the current 25.5 cubic meters and the 2a:a cubic meters generated annually tliereafter). The licensee's evaluation shows that the soiVsand can be managed on-site in the same manner as the. septic waste and cooling tower silt (i.e., by land spreading on designated fields). *

2.0 EVALUATION The licensee will dispose of the soil and future soil/sand material using a land spreading technique consistent with its current commitments for on-site disposal of septic waste and cooling tower silts previously approved by the NRC. The licensee will continue to use the designated and approved areas of their property (approximately 1.9 acres in size) which currently receives the septic waste and cooling tower silts. Determination of the radiological dose impact of the new material has been made based on the same dose assessment models and pathway assumptions used in the previously approved submittals.

Revision _l2_ Date~ H-39

*~

2 The licensee will procedurally control and maintain records of all disposals. The following information will be recorded:

1. The radionuclide concentrations detected in the material (rneasured to radiation levels consistent with the licensee's radiological environmental monitoring program).

2. The total volume of material disposed.

3. The total radioactivity ii') the disposal operation as well as the total radioactivity accumulated on each disposal plot at the time of spreading.

4. The plot on which the material was applied.

5. Dose calcul?tions or maximum allowable accumulated activity determinations required to aemonstrate that the dose condition values imposed (i.e., imposed by this 10 CFR 20.2002 application) on the land spreading operation have. not been exceeded.

The bounding dose conditions for the on-site disposals are as follows:

1. The annual dose fo* the whole body or any organ of a hypothetical maximally exposed individual must be less than 1.0 mrem:

2. Annual doses to the whole body and any organ of an inadvertent intruder from the probable pathways of exposure must be less than 5 mrem.

3. Disposal operations must be at one of the approved on-site locations.

To ensure that the addition of new material containing low levels of radioactivity will not exceed the bounding dose conditions, for each new spreading operation the licensee will calculate an estimate of the total radioactivity applied to the designated disposal plots. These calculated estimates will include all past disposals of septic waste, c0<;>ling tower silt, soil and soiVsand material on the designated disposal plots. This witl be compared with the bounding dose condition value or equivalent radioactivity value on a petacre basis. In addition, concentration limits will be applied to the disposed material to restrict the placement of small volumes of material that may have relatively high radioactivity concentrations.

T.he licensee assessed the dose that may be*received.by the maximally ~xpose<i'indMdual during the period of plant control over the property, and to an inadvertent intruder after plant access control ends using the same pathway modeling, assumptions, and dose calculation methods that were previously approved by the NRG for the septic waste and cooling tower silt disposals. The dose models are*based on the guidance in NRC Regulatory Guide 1.109, Revision 1 (1977).

The licensee's dose assessment is as follows:

1. Total annual doses to the whole body and critical organ of the hypothetically maximally exposed individual were estimated to be 0.115 mrem and 0.403 mrem, respectively.*

These values are less than the prescribed annual dose condition value of 1.0 mre~ for ,--~~~~~~~---, the time period of active site cont~ol. ' ~Rc-*vi-sio_n-29~0-at_e~l/l-1/0-2~--l' . H-40

3

2. Total annual doses to th~ whole body and critical organ of an inadvertent intruder from the probable pathways of exposure were estimated to be 0.757 mrem and 1.17 mrem.

These ~alues are *1ess than the prescribed annu~I dose condition value of 5.0 mrem for the time period after active site control. * *

3. The dose calculations are based on projecting the maximum potential impact of all disposals (past and future) on the designated disposal plot of land.

3.0 CONCLUSION

The staff finds the licensee's proposal to dispose of the low-level radioactive soil and soiVsand material, pursuant to 10 CFR 20.2002,* in the same nianner. location, and within the bounding dose conditions as the materials (i.e., septic waste and cooling tower silt) preyiously approved by the NRG to be acceptable because the bounding conditions will contimJe to be met. Principal Contributor: S. Klementowicz Date: June 15, 2000 Revision ..2.2_ Date 1111/02 H-41

Appendix I I. "Request to Amend Previous Approval Granted Pursuant to 10CFR20.2002 for 11 Disposal of Contaminated Soil, dated September 11 1, 2000, BVY 00-71

                    . 2. Vennont Yankee Nuclear Power Station - Safety Evaluation for an Amendment to an Approved I OCFR20.2002 Application (TAC No. MA9972)", dated June 26'h, 2001, NVY 01-66 I-1 Revision .1.2_ Date 1/11/02

VERMOtlT YANKEE NUCLEAR POWER CORPORATION 185 OLD FERRY ROAD, PO BOX 7002, BRATTLEBORO, VT 05302-7002 (802) 257-5271 September 11, 2000 BVY 00-71 United States Nuclear Regulatory Commission ATIN: Document Control Desk Washington, DC 20555

References:

(a) Letter, VYNPC to USNRC, "Request to Amend Previous Approvals Granted under 10 CFR 20.302(a) for Disposal of Contaminated Septic Waste and Cooling Tower Silt to Allow for Disposal of Contaminated Soil," BVY 99-80, dated June 23, 1999. (b) Letter, VYNPC to USNRC, "Supplement to Request to Amend Previous Approvals Granted under I 0 CFR 20.302(a) to Allow for Disposal of Contaminated Soil," BVY 00-02, dated January 4, 2000. (c) Letter, USNRC to VYNPC, "Vermont Yankee Nuclear Power Station, Request to Amend Previous Approvals Granted under 10 CFR 20.302(a) to Allow for Disposal of Contaminated Soil (TAC No. MA5950)," NVY 00-58, dated June 15, 2000. (d) Letter, USNRC to VYNPC, "Revised Safety Evaluation - Approval . Pursuant to 10 CFR 20.2002 for Onsite Disposal of Cooling Tower Silt - Vermont Yankee Nuclear Power St2.tion (TAC No. M96371)," NVY 97-85, dated June 18,1997.

Subject:

Vermont Yankee Nuclear Power Station License No. DPR-28 (Docket No. 50-271) Request to Amend Previous Approval Granted Pursuant to 10 CFR 20.2002 for Disposal of Contaminated Soil In accordance with 10 CFR 20.2002 (previously 10 CFR 20.302(a)), Vermont Yankee (VY) subm.its this application to amend the previously granted approval to dispose of slightly contaminated soil. This application expands the allowable wast~ stream* to *include slightly

contaminated soil generated as a residual by-product of other types of on-site construction activities.

In References (a) and (b), VY requested approval to dispose of approximately 25.5 m 3 of accumulated soil that was generated due to construction .activities. In addition, it was requested that VY be allowed to dispose of approximately 28.3 m3 of soil that is spread annually on station roads and walkways during the winter. NRC acceptance is documented in Reference (c). This application specifically requests approval to dispose of contaminated soil that is created due to other on-site construction related activities including but not limited to design change implementation and land maintenance. Revision 29 Date --1L!..l.iQ£ I-2

... VERMONT YANKEE NUCLEAR POWER CORPORATION BVY 00-71/ Page 2 of2 In addition, VY requests that NRC's review recogni:z;e that, although VY in9icated in Reference (b) that the south disposal field (approximately 1.9 acres in size) is currently expected to be used for disposal of the subj~t material, VY is also authorized to use the alternate.north disposal field (approximately 10 acres in size). Approval to use both the north and south fielqs for disposal was granted in Reference (d). VY's radiological impact assessments have conservatively assumed all of the disposal activities occur on the smaller south field to maximize potential calculated doses.

These assessments bound the situation where a portion of the land spreading occurs on the north field. VY will continue to limit the total activity spread, from approximately 28.3 m 3 of soil generated each year, to within the limits assumed in the radiological assessment previously submitted in Reference (b). A radiologicaJ .~sessment and proposed operational controls for the inclusion of the additional material for on-site disposal was provided in Reference (b). The assessment demonstrates that the dose impact expected from the proposed activity, in total with all past waste spreading operations, will not approach the dose limits already imposed for septic and cooling tower silt disposal. All soil analyses will be to environmental lower limits of detection. The results of all disposal operations will be reported in the Annual Radioactive Effluent Release Report. The combined radiological impact, for all on*site disposal operatiors, will continue to be limited to a total body or organ dose of a maximally exposed member of the public of less than one mrem/year during the period of active VY control of the site, or less than five mrem/year to an inadvertent intruder after termination ofactive site control. Upon receipt of your approval, this request as well as the basis for approval will be incorporated into the Off-Site Dose Calculation Manual. We trust that the infonnation contained in the submittal is sufficient. However, should you have any questions or require further information concerning this matter, please contact Mr. Jim . Devincentis at 802-258-4236. Sincerely,

                                                           /~Sen             .     /

Licensing Manager cc: USNRC Region I Administrator USNRC Resident Inspector - VYNPS USNRC Project Manager - VYNPS VT Department of Public Service Revision -12__ Date 1/11/02 I-3

SUMMARY

OF. VERMONT YANKEE COMMITMENTS BVY NO.: 00-71 The following tabJe identifies commitments made in t~is document by Vermont Yankee. Any other actions discussed in the submittal represent intended or pJanned actions by Vermont Yankee. They are described to the NRC for the NRC's information and are not reguJatory commitments. PJease notify the Licensing Manager of any questions regarding this document or any associated commitments. I COMMITMENT I COMMITTED DATE

OR "OUTAGE" I

                           .. -          None                                           NIA VYAPF 0058.04 AP 0058, Revision 1 Page 1of1 I-4 Revision .12_   Date    II l l /02

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 205~001 June 26, 2001 Mr. Michael A. Balduzzi NVY 01-66 Vice President, Operations Vermont Yankee Nuclear Power Corporation *

  . 185 Old Ferry Road .

P.O. Box 7002 Brattleboro, VT 05302-7002

SUBJECT:

VERMONT YANKEE NUCLEAR POWER STATION - SAFETY EVALUATION FOR AN AMENDMENT TO AN APPROVED 10 CFR 20.2002 APPLICATION (TAC NO. MA9972) .

Dear Mr. Balduzzi:

The U.S. Nuclear Regulatory Commission (NRC) staff ~s completed its review of the Vermont Yankee Nuclear Power Corporation (VYNPC) request dated September 11, 2000, to amend an approved 10 CFR 20.2002(previously10 CFR 20.302) application dated June 23, 1999, as supplemented on January 4, 2000. The licensee requested NRC approval to alloVf the addition of slightly contaminated soil resulting from on-site construction-related activities, including but not limited to, design change*lmplementation and land main~enance, to the list of already approved materials (i.e., septic waste, cooling tower silt and soiVsand from roads and walkways) for on-site disposal. Based on our review, we find the proposed changes to be acceptable because the previously. approved bounding conditions will continue to be met. The *enclosure to this letter provides* our safety evaluation of VYNPC's application.

Pu.-suant to the provisions of 1o Ci-R Part 51, ~he NRG has published 3n Environmental Assessment and finding of No Significant Impact in the Federal R~gisteron June 14, 2001

{66 FR 32399). . Sincilt~.

                                                    ~A
                                                    ~~~~on   Zwolinskl, Director of Licensing Project Management Office of. Nuclear Reactor Regulation.

Docket No. 50-271

Enclosure:

Safety Evaluation cc w/encl: See next page Revision .12.... Date 1/11/02 I-5

Vermont Yankee Nuclear Power Station cc: . Regional Administrator, Region I Ms. Deborah B.

Katz U. S. Nuclear Regulatory Commission Box83*

475 Allendale Road Shelburne Falls, MA 01370 King of Prussia, PA 19406 Mr. Raymond N. McCandless -r Mr. David R. Lewis Vermont Department of Health Shaw, Pittman, Potts & Trowbridge Division of Occupational 2300 N Street, N.W. and Radiological Health Washington, DC 20037-1128 108 Cherry Street Burlington, VT 05402 Ms. Christine S. Salembier, Commissioner Vermont Department of Public Service Mr. Gautam Sen 112 State Street Licensing Manager Montpelier, VT 05620-2601 Vermont Yankee Nuclear Power Corporation Mr. Michael H. Dworkin, Chairman 185 Old Ferry Road Public Service Board P~O. Box 7002 State of Vemiont

Brattleboro. VT 05302-7002 112 State Street Montpelier, VT 05620-2701 Resident Inspector Vermont Yankee Nuclear Power Station Chairman, Board of Selectmen U. S. Nuclear.Regulatory Commission Town of Vernon P.O.Box.176
P.O. Box 116 Vernon, VT 05354 Vernon, VT 05354-0116 Director, Massachusetts Emergency Mr. Richard E. McCullough Management Agency Operating Experience Coordinator ATTN: James Muckerheide Vermont Yankee Nuclear Power Station 400 Worcester Rd.

P.O. Box 157 . Framingham. MA 01702-5399 Governor Hunt Road Vernon, VT 05354 Jonathan M. Block. Esq. Main Street G. Dana Bisbee, Esq. P. 0. Box566 Deputy Attorney General Putney, vr '05346-0566 33 Capitol Street Concord, NH 03301-69~7 Chief, Safety Unit Office of the Attorney General One Ashburton Place, 19th Floor Boston, MA 02108 L.__R_ev_i_si_o_n_2_9~~D-a-te~-l-/1-1-/0-2~~~--11* I-6

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 2055S-0001 "SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION. VERMONT YANKEE NUCLEAR POWER CORPORATION VERMONT YANKEE NUCLEAR POWER STATION

~*

DOCKET NO. 50-271

1.0 INTRODUCTION

By letter dated September 11, 2000, Vermont Yankee Nuclear Power Corporation (VYNPC/ licensee) submitted a request to amend_.a Title 10 of the Code of Federal Regulations (1Q CFR} Section 20.2002(former10 CFR 20.302} application, dated June 23, 1999, as supplemented on January 4, 2000, that was approved by the U.S. Nuclear Regulatory Commission (NRG). This amendment will allow the addition of slightly contaminated soil resulting from on-site construction-related activities, including but not limited to, design change implementation and land maintenance, to the list of already approved materials (i.e., septic waste, cooling tower silt and soiVsand from roads and walkways) for on-site disposal via land spreading on designated disposal fields. In 1989, pursuarit to 1o CFR 20.302 (current 1 O CFR 20.2002), the licensee received approval from the NRC to routinely dispose of contaminated septic waste in designated on-site areas~ .. Jn 1997, the NRC amended the approved on-site disposal application to also include contaminated cooling tower silt materia!. In 2000. the NRC amended the approved on-site disi)osal application to also include a one-time disposal of slightly contaminated soil and an annual . disposal of 28.3 cubic meters of slightly cont~minated soiVsa'nd material. I In this 10 CFR 20.2002 amendment application. the licensee requested that slightly contaminated soil resulting from on-site construction-related activities be disposed of on-site on an annual basis until the end of the-plant's operating license in 2013. The anticipated annual volume of soil generated by on-site construction, as identified by the licensee, combined with the soiVsand generated from the annual winter spreading of sand on roads and walkways at the plant site will not exceed 28.3 cubic meters. This volume is tfle same volume that was approved in the January 4, 2000, request. The licensee pertormed a comprehensive radiological evaluation which included the annual disposal of 28.3 cubic meters of soil and soiVsand materials, and shows that these materials can be managed on-site in the same manner as the septic waste and cooling tower silt (i.e., land spreading on designated fields). 2.0 EVALUATION The licensee will dispose of the future soil material using a land spreading technique consistent with the current commitments for on-site disposal of septic waste, co.oling tower silts and sand/soil material previously approved by the NRG. The licensee will continue to use the I-7 Revision -12._ Date 1/11/02

designated and approved areas of their property which include approximately 1.9 acres, which currently receive the septic waste, cooling tower silts and soiVsand material, and approximately 1O acres which have not been previously used for disposal. Determination of the radiological dose impact of the new material has been made based on the same dose assessment models and pathway assumptions used in the previously approved submittals. The licensee will procedurally control and maintain records of all disposals. The following

information will be recorded:

-c.

1. The radionuclide concentrations detected in the material (measured to radiation levels consistent with the licensee's radiological environmental monitoring program);

2. The total volume of material disposed;

3. The total radioactivity in the disposal operation as well as the total radioactivity accumulated on each disposal plot at the time of spreading;

4. The plot on which the material was applied;

5. Dose calculations or maximum allowable accumulated activity determinations required to demonstrate that the do5e condition values imposed (i.e., imposed by the approved 10 CFR 20.2002 application dated June 23, 1999) on the land spreading operation have not been exceeded .

. .~**

'*;: ... The bounding dose conditions for the on-site disposals are as follows:

1. The annual dose to the whole body or any organ of a hypothetical maximally exposed indMdual must be less than 1.0 mrem.

2. Annual doses to the whole body and any organ of an inadvertent intruder from the .

probable pathways of expooure must be less than 5 mrem.

3. Disposal operations must be at one of the approved on-site locations.

4. Total annual combined volume of soil and soiVsand materials must not exceed 28.3 cubic meters.

To ensure that the addition of new material containing low levels of .radioactivity will not exceed the bounding dose ronditions, for each new spreading operation the licensee will calculate an estimate of the total radioactivity that includes all past disposals of septic waste, cooling tower silt, and soiVsand and soil material on the designated disposal plots. This will be compared with the bounding dose condition value or equivalent radioactivity value on a per acre basis. The licensee assessed the dose from soil and soiVsand material that may be received by the maximally exposed individual during the period of plant control over the property1 and to an inadvertent intruder after plant access control ends using the same pathway modeling, assumptions, and dose calculation methods that were previously approved by the NRC for the septic waste and cooling tower silt disposals. The dose mooels are based on the guidance in NRC Regulatory Guide 1.109, Revision 1 (1977). I-8 Revision .12.._ Date 1/11/02

L..

The licensee's dose assessment is as follows:

1. Total annual doses to the _whole body and critical organ of the hypothetically maximally

                  *exposed individual were estimated to be 0.1.15 mrem and 0.403 mrem respectively.
                   .These values are less than< the prescribed annual dose condition value of 1.0 mrem for the time period of active site control.                   *

2. Total annual doses to the whole body and critical organ of an inadvertent intruder from the probable pathways of exposure were estimated to be 0.757 mrem and 1.17 mrem.

Th~se values are less than the prescribed annual dose condition value of 5.9 mrem for the time period after active site control.

3. The dose calculations are based on projecting the maximum potential impact, of all disposals (past and future} on the approved disposal site.

3.0 CONCLUSION

The staff finds the licensee's proposal to dispose of the low-level radioactive soil material, pursuant to 10 CFR 20.2002. in the same manner, location,* and within the bounding dose conditions as the materials (i.e., septic waste, cooling tower silt and soil/sand from roads and walkways) previously approved by the NRC to be acceptable. The licensee has committed to permanently incorporate this modification into their Offsite Dose Calculation Manual. Principal Contnbutor: A. Hayes Date: June 26, 2001 Revision .12.... Date 1/11/02 I-9 I

APPENDIX 1 I. "Request to /\mend Previous Approva: Granted Pursuant tu I OCFR2U.20U I fur increase of the Annual Volume Limit and One-time Spreading of Current Inventory," dated October 4, 2004, BVY 04-110 ......................................................................................... 2

2. "Safety Evaluation of Request to Amend Previous Approvals Granted Pursuant to IOCFR20.2002 - Vermont Yankee Nuclear Power Station (TAC No. MC5104)" dated July 19, 2005, NVY 05-090 ............................................................................................. 51 Appendix J Original Off-Site Dose Calculation Manual Page I of57 Vermont Yankee Nuclear Power Station

APPENDIX J (Continued)

  *~Entergy Entergy Nudear Northe Entergy Nuclear Operations. Inc.

Vermont Yankee 185 Old Ferry Rd. P.O. Box 500 Brattleboro, VT 05302 Tel 802-257-5271 October 4, 2004 Docket No. 50-271 BVY 04-110 Attn: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001

References:

a) Letter, VYNPC to USNRC, "Request to Amend Previous Approval Granted Pursuant to 10 CFR 20.2002 for Disposal of Contaminated Soil", BVY 00-71, dated September 11, 2000. b) Letter, USNRC to VYNPC, "Vermont Yankee Nuclear Power Station - Safety Evaluation for an Amendment to an Approved 10CFR 20.2002 Application (TAC No. MA9972)", NVY 01-Q6, Dated June 26, 2001. c) Letter, VYNPC to USNRC, "Supplement to Request to Amend Previous Approvals Granted under 10 CFR 20.302(a) to Allow for Disposal of Contaminated Soil", BVY 00-02, dated January 4, 2000.

Subject:

Vermont Yankee Nuclear Power Station Request to Amend Previous Approval Granted Pursuant to 10CFR20.2002 for Increase of the Annual Volume Limit and One-time Sprei'ldinr of ~*-*rrent hwe~tory In accordance with 10CFR20.2002 (previously 10CFR20.302(a)), Entergy Nuclear Operations, Inc. (ENO) submits this application to amend the previously granted Vermont Yankee (VY) approval to dispose of slightly contaminated soil. This application requests an increase of the current annual volume limit of 28.3 cubic meters of soil as specified in the previous approval (Reference (b)) to a new volume limit of 150 cubic meters of soil. This application also requests permission to spread the current inventory of approximately 528 cubic meters of soil as described in Attachment A in a one-time spreading activity following receipt of your approval. ENO will continue to limit the total activity spread each year to remain within the limits specified in the radiological assessment previously submitted in Reference (c). A radiological assessment of the impact of spreading the current inventory of soils and sediments located at VY is provided in Attachment A. The assessment concludes that: a) There is significant capacity remaining in the South Disposal Plot to continue to accept additional earthen materials for land spreading without exceeding established dose limitations. Appendix J Original Off-Site Dose Calculation Manual Page 2 of57 Vermont Yankee Nuclear Power Station

APPENDIX J (Continued) BVY 04-110 I Page 2 b) The existing inventory of waste soils in storage can be placed on the South Disposal Plot without exceeding dose impact limits previously established for a single disposal field. c) The continued use of the South Disposal Plot will not exceed the limiting dose criteria established in the VY Offsite Dose Calculation Manual. d) The approved dose impact methodology used to detennine compliance with the on-site spreading dose limits are not driven by the volume of waste material disposed of, but by the total radioactivity content of the material that is spread over a fixed disposal plot area (1.9 acres for the South Disposal Plot). The dose modeling conservatively assumes that all radioactivity spread on the field remains in the top 15 centimeter surface layer, even after subsequent additions are placed on the same field area. Existing limits on the concentrc;cion of radioactivity in waste media provides protection from small volumes of "hot" or high specific activity materials from being spread qn the disposal field. The results of all disposal operations will continue to be reported in the Annual Radioactive Effluent Release Report. The combined radiological impact, for all on-site disposal operations, will continue to be limited to a total body or organ dose of a maximally exposed member of the public of less than one mrem/year during the period of active VY control of the site, or less than five mrem/year to an inadvertent intruder after termination of active site control. Upon receipt of your approval, this request as well as the ba:;is for approval will be incorporated into the VY Offsite Dose Calculation Manual. There are no new commitments being made in this submittal. We trust that the information c:mtained in the submitta! is sufficient. However, should you have any questions or require further infonnation concerning this matter, please contact me at (802) 258-4236. Sincerely,

                                              ~~~    Manager, Licensing Vermont Yankee Nuclear Power Station Attachment (1) cc:      USNRC Regional Administrator - Region 1 USNRC Resident Inspector - VYNPS USNRC Project Manager-VYNPS Vermont Department of Public Service Appendix J Original Off-Site Dose Calculation Manual Page 3 of 57 Vermont Yankee Nuclear Power Station

APPENDIX J (Continued) Attachment to BVY 04-110 Dockel No. 50-271 Assessment of On-Site DispQsa! of Contaminated Stored Soils by Land Spreading Entergy Nuclear Operations, Inc. Vermont Yankee Nuclear Power Station Appendix J Original Off-Site Dose Calculation Manual Page 4 of57 Vermont Yankee Nuclear Power Station

APPENDIX J (Continued) T ABLF. OF CONTENTS Page No. Table of Contents 2 List of Tables 3 1.0 Evaluation Objective 4 I . I Background 4 2.0 Summary of Results 6 3.0 Method of Evaluation 7 3.1 Waste Characterization 7 3.2 Soil Disposal and Administrative Procedure Requirements 8 3 .3 Disposal Plot Characteristics 9 3.4 Radiological Impact Methodology 9 4.0 Assumptions and Inputs 12 5.0 Evaluations 14 5.1 Case Study I (Past Spreading Impacts) 14 5.2 Case Study II (Stored Soil/Sand Inventory Impacts) 14 5.3 Case Study III (Projected Future Spreading Impacts) 15 6.0 Results I Conclusions 17 7.0 References 34 Appendix A: Security Fence Upgrade'Soil Pile #2002-01 A-1 Appendix B: Security Fence Upgrade Soil Pile #2002-02 B-1 Appendix C: Security Fence Upgrade Soil Pile #2002-03 C-1 Appendix D: 2001 protected Area Road Sweeping Pile D-1 Appendix E: 2001 HWC Soil Excavations E-1 Appendix F: 1996 Soil Remnants Analysis - Security Fence Upgrade F-1 Appendix G: Record of Most Recent Spreadings (2003) on South Disposal Field 0-1 2 Appendix J Original Off-Site Dose Calculation Manual Page 5 of 57 Vermont Yankee Nuclear Power Station

APPENDIX J (Continued) List of Tables 1 Inventory of Contaminated Soil Piles In Storage (December 2003) 8 2 Site Specific Control Period Dose Conversion Factors 12 3 Site Specific Intruder Dose Conversion Factors 13 4 Record of Septic/CT Silt/Construction Soil Radioactive Material Spreading Each Year on the South Disposal Field 19 5 Record of Septic Waste Only for Radioactive Material Spreading Each Year on the South Disposal Field 20 6 Record of Cooling Tower Silt Only for Radioactive Material Spreading Each Year on the South Disposal Field 21 7 Record of Construction Soil/Sand Waste Only for Radioactive Material Spreading Each Year on the South Disposal Field 21 8 C;;-137 in Storage Piles aite1 La.;r Spreading in 2vOJ 22 9 Co-60 in Storage Piles after Last Spreading in 2003 23 10 Zn-65 in Storage Piles after Last Spreading in 2003 24 11 Mn-54 in Storage Piles after Last Spreading in 2003 25 12 Soil/Sand Storage: Total Activity to be Spread on 6/1/04 and Decayed to 20.13 26 .. 13 Projection of Additional Septic/Silt/Soil-Sand at Current Generation Rates to 2013 26 14 Projection of Additional Sand/Soil Mix 27 15 Projection of additional Cooling Tower Silt 27 16 Projection of Additional Septic Waste 27 17 Projected I Yr Septage & Silt Spreading for 6/l/04 and decayed to 2013 28 18 Current Total Spreading as of 11/4/03 & How Much Remains at 6/1/04 and 2013 28 19 Ci.irrent Spreading Totals Plus Total Projected Future Spreadings to 2013 29 20 Radioactivity Content from Existing Materials 29 21 Past Spreading Control Period Dose Only (No Stored Material or Future Additions) 30 22 Past Spreading Only (No Stored Materials or Future Additions) Intruder Dose 30 at End of Plant Operations in 2013 23 Control Period Dose : Past Spreadings & Current Stored Soil Material Inventory 31 24 Past Spreading & Current Stored Soil Inventory Intruder Dose at End Plant Ops 31 25 Dose Impact from Spreading of Current Inventory of Stored Material Only 32 26 Past, Stored Material and Projected Future Disposal (all septage/silt/soils) Doses at End of Plant License 33 3 Appendix J Original Off-Site Dose Calculation Manual Page 6 of57. Vermont Yankee Nuclear Power Station

APPENDIX J (Continued) 1.0 EVALUATION OBJECTIVE Current restrictions on the annual volume of slightly contaminated soil ( 1000 ft3 or 28.3 m3 ) that can be disposed of on-site (ODCM, Appendix I, Reference 1), coupled with several plant facility construction projects in recent years, has resulted in the accumulation of a back-log oflow level contaminated earthen material that is awaiting to be dispositioned by land spreading on previously approved on-site disposal areas. The objective of this assessment is to present the data and formal evaluation to demonstrate that the proposed one time disposal of the existing accumulated backlog of soil I sand materials (as of November 2003) without regards to the annual soil volume limit, will meet the existing dose objective boundary conditions as approved by the NRC for septic waste, Cooling Tower silt and other earthen type materials (Reference I), even if use of the same disposal field for future spreading is assumed to continue over the remaining plant operating license. The established dose based boundary conditions (NRC approved) for disposal and accumulation of low-level contaminated septic waste, Cooling Tower silts and soil/sand mixes on designated plots within the VY site boundary will continue to be applied without change. These dose limit criteria are taken from Appendix B of Reference l, and are:

The dose to the whole body or any organ of a hypothetical maximally exposed individual must be less than 1.0 mrem/yr during the period that VY has active control over the disposal plots (plant operating life).
The doses to the whole body and any organ of an inadvertent intruder following the period of active plant control over the property from the probable pathways of exposure is less than 5 mrem/yr.
Disposal operations must be at one of the approved on-site locations.

1.1 Background In 1989, Vermont Yankee Nuclear Power Corporation requested from the NRC permission to routinely dispose of slightly contaminated septic waste in designated on-site areas in accordance with l OCFR20.302(a). Approval from the NRC was granted on August 30, 1989, provided that the request and analysis be pennanently incorporated into the plant's Offsite Dose Calculation Manual (ODCM). Revision 9 to the ODCM (Appendix B) incorporated the assessment and the approval of methods utilized for on-site disposal of slightly contaminated sewage sludge by land spreading. The approval allowed for the existing septic inventory to be disposed of on-site along with future quantities anticipated to be generated as part of routine system maintenance. For purposes of demonstrating that future addition of waste materials could be added to the disposal plots, the radiological analysis projected an annual generation rate of about 18,600 gallons of sewage containing about 1400 kg of solid materials that might require on-site spreading. NRC permission for these future disposals was granted as long as both the projected dose (for both current and all past disposal operations) and radionuclide concentration limits(:$ 10% of the IOCFR20, Appendix B, Table II, concentration values) are satisfied. No specific limit on annual volume of septic waste that could be disposed of was included in the approval. In 1995, Vennont Yankee requested from the NRC that the previous authorization for on-site disposal of septic waste be amended to permit the on-site disposal of slightly contaminated Cooling 4 Appendix J Original Off-Site Dose Calculation Manual Page 7 of57 Vermont Yankee Nuclear Power Station

APPENDIX J (Continued) Tower silt material. The application analyzed the expected radiological impact from both the existing inventory at that time of about 14,000 ft 3 (-396 m3 ) of accumulated silt, along with an operating cycle ( 18 months) generation rate of about 4000 ft 3 (-113 m3 ). The NRC returned their safety evaluation, dated March 4, 1996, granting approval for the proposed silt disposal. Similar to the sewage waste disposal, NRC acceptance required that all disposal operations be conducted such that both the projected dose (for both current and all past disposal operations) and radionuclide concentration limits are satisfied. The soil concentration limits (for any sample) were based on limiting external annual dose to 25 mrem assuming continuous occupancy on an infinite plot at that concentration unifonnly spread to a 15 cm depth. No specific limit on annual volume of Cooling Tower silt that could be disposed of was included in the approval. The NRC also required that any further modification to the proposed action have prior NRC staff approval. In 1999 (with a supplemental filing in 2000), Vermont Yankee filed a third request under 10 CFR 20.2002 with the NRC to amend the previously approved applications for on-site land disposal of slightly contaminated earth type materials (seotic slt1dge an~ Cooling Tower silt) to include approximately 900 ftJ (25.5 m3 ) of accumulated contaminated soil generated during construction activities within the VY Protected Area. Sampling of the soil revealed low levels of radioactivity that were similar in radionuclides and activity levels to the septic waste and Cooling Tower silts previously encountered .. The request to the NRC for this additional material also indicated that additional amounts of contaminated soil I sand associated with road sweepings following winter sanding of road and walkways in the Protected Area could result in an estimated I 000 ft 3 per year (28.3 m 3 per year) that might need to be disposed of as slightly contaminated materials. The NRC requested that the initial submittal (1999) of the soil spreading 20.2002 application include an analysis that evaluated projected future additions of an estimated annual volume of soil being added to the designated disposal plots. This information was required if Vermont Yankee intended to use the 20.2002 soil disposal application for approval to dispose of potential future volumes (i.e., not just a one-time disposal application) oflow level contaminated soil in the same manner as already approved for septic waste and Cooling Tower silt. Vermont Yankee revised its application by adding an analysis for a projected annual volume of 1000 ft3, or equivalently 28 J m3 , of contaminated soil starting in the year 2000 and continuing on a yearly basis until end of plant license in 2013. At the end of the projected disposal stream, the accumulated buildup of contamination from all .sources (septic waste, Cooling Tower silt and soil I sand mixes) on the disposal fie!~ was evaluated for both the dose impact to the critical receptor at the end of the control period and the assumed intruder. These dose impacts were found acceptable when compared against the original on-site spreading dose acceptance criteria of I mrem/yr (Control Period) and 5 mrem/yr (Intruder Scenario). The I 000 ft3 (28.3 m3 ) annual generation rate of soil was based on plant staff estimates that approximately that amount of soil and sand is collected from road and walkway sweepings inside the Protected Area following each year's winter clean-up as part of routine maintenance. This is the only type of earthen materials that has a specific annual volume limit associated with it in addition to the projected dose and concentration limits associated with the disposal of septic waste and Cooling Tower silt. No volume estimate for unidentified future site excavation and construction activities was provided. 5 Appendix J Original Off-Site Dose Calculation Manual Page 8 of 57 Vermont Yankee Nuclear Power Station

APPENDIX J (Continued) 2.0

SUMMARY

OF RESULTS The evaluation of the radiological impact of all past, accumulated storage inventory, and projected foture waste spreading operations on a single disposal plot (I .9 acres at South end of site) have indicated the folJowing results and conclusions:

All past spreading of septic waste, Cooling Tower silt, and soil/sand mixes through the end of2003 have resulted in a maximum organ dose to a critical receptor (control Period use) that accounts for only 13. 7% of the I mrem/year limit. With respect to the Intruder dose limit of 5 mrem/year, all past spreadings through the end of 2003 account for only 7.2% of the maximum organ dose to the limiting receptor at the end of plant license.
The impact from the projected spreading of the existing waste materials in storage is estimated to account for only 5.3% of the I mrem/yc;ar Control Period dose limit, or only 17.7% of the same limit when all past spreadings are combined with the materials currently in storage (as of the end of November 2003}. The maximum fntrnder organ dose from all past spreading and stored materials is 0.456 mrem/yr, or 9. l % of the 5 mrem/yr Intruder scenario dose limit. These results indicate that the existing inventory of waste materials in storage can be placed on the South disposal plot without exceeding dose impact limits previously established for a single disposal field.
Assuming the same annual average generation rate of radioactivity in waste materials (septagc/silt/soil) that has been observed over the last fourteen years is added to all past waste spreadings and stored soil commitments, the projected dose at the end of the current plant licensing period (year 2013) yields a limiting maximum critical receptor dose (for either the control period or inadvertent intruder) equivalent to only 25.3% of the most restrictive annual dose limit (associated with the I mrem/year limit for the maximum organ during the Control Period). This finding demonstrates that the continued use of the South disposal plot, even with the addition of J 8,653 ft 3 of slightly contaminated soil currently in storage, will not exceed the approved limiting dose criteria established in the ODCM.
The dose impact methodology used to determine compliance wilh the on-site spreading dose limits are not driven by the volwne of waste material disposed of, but by the total radioactivity content of the material. Existing limits on the concentration of radioactivity in waste media provides protection from small volumes of"hot" or high specific activity materials from being spread on the disposal field.

6

                                                .J*q Appendix J Original Off-Site Dose Calculation Manual Page 9 of57 Vermont Yankee Nuclear Power Station

APPENDIX J (Continued) 3.0 METHOD OF EVALUATION The method of evaluating the impacts from the on-site spreading is the same as used and approved in the original exemption request made to the NRC under 10CFR20.302 for septic waste and which has been applied in all subsequent amendment requests for additional types of earthen materials to be disposed by land spreading on-site. The pathway and dose models found in Regulatory Guide 1.109 (Reference 2) are employed in performing the radiological dose impact assessment. The application of the dose models begins with the characterization of the waste materials that are to be subject to the dose evaluation. 3.1 Waste Charactcri7..ation TI1e existing accumulated soil I sand that has created a backlog of accumulated material over the last several years is identified on Table 1, along with the estimated volume and origination of material. The soil materials were primarily derived from excavation activities associated with the construction of new security fences aiong the plant's Protected Area boundary and the construction of new plant installations associated with the capability to perform hydrogen water chemistry treatment of the plant's coolant system. Also included arc road treatment sands used for winter traction inside the protected area. The soil I sand mix is typical of fill material containing light to dark brown poorly sorted soils with some small stones, and may include small incidental pieces of asphalt. The soil was removed from its original location by shovel, backhoe and front-end loader, and placed into dump trucks for transport to the temporary storage area located between the Cooling Towers where it was deposited on the ground surface and covered to prevent erosion. This location was selected because it was away from areas routinely occupied by plant staff, and could easily be controlled. The most probable source of the low levels of radioactive contamination is due to the presence of below detectable removable contamination redistributed by foot traffic from inside the plant to walkways and parking areas. Subsequent surface runoff carries the contamination to* nearby exposed soil near the Protected Area boundary where it accumulates over time to low-level detectable concentrations. Down wash and deposition ofparticuiate activity released from the plant's Primary Vent Stack as part of routine gaseous emissions may also have contributed to the low levels of detectable activity. For potential future disposal volumes of sand and soil, the current volume limit (I 000 rt3 [28.3 m 3], ODCM Appendix H, Reference I) was based on the expected rate of road sand used for winter road and walkway traction, but did not anticipate or reflect the potential for future site construction activities that could excavate soils on-site that also contain low levels of plant related radioactivity. The present inventory of stored soil I sand between the Cooling Towers includes approximately 18,653 ft3 (528 m 3) of material of which only 3.3% or 616 ft3 (17.4 m3) originated as roadway sweepings. In accordance with Appendixes B and F of the ODCM, disposal of septic waste and Cooling Tower silt material is not limited by an annual volume limit but by total dose impact related to the radioactivity content of the silt and the concentration of radioactivity contained within it. Currently, only soils I sand mixtures have an annual on-site disposal limit equal to I 000 ft3 (28.3 m3). 7 Appendix J Original Off-Site Dose Calculation Manual Page 10 of57 Vermont Yankee Nuclear Power Station

APPENDIX J (Continued) Table I Inventory of Contaminated Soil Piles ln Storage (December 2003) Overall Overdll Max Pile description Length width Height Estlmated Volume ft ft inches (1) 2002-01: Security Fence Upgrade 75 21 44 4,679 (2) 2002-02: Security Fence Upgrade 75 28 60 9,000 (3) 2002-03: Security Fence Upgrade 39 22 42 2,457 (4) Sand Sweepings (inside protected area) 16 11 42 616 (5) 2001 HWC Soils Excavation 38 11 42 931 (6) 1996 Soll Remnants from Fence Upgrade 24 6.4 84 970 total Vol. (ft"3) = 18,653 3.2 Soil Disposal and Administrative Procedure Requirements The method of soil/sand disposaJ of the existing bac~log inwntory will use the technique of land spreading in a manner consistent with the current commitments for the on-site disposal of septic waste and Cooling Tower silts as approved by the NRC and implemented in Appendices B and F of the Vermont Yankee ODCM (Reference I). The accumulation ofradioactivity on the disposal plot for this proposed soil spreading operation will be treated as if Cooling Tower silt or septic waste was being disposed of since the characteristics of aJI these residual solids are similar (earthen-type matter). The South field (approximately 1.9 acres in si:ze) has been used for all past disposal operations and is expected to be used for the placement of the existing backlog (see Table 1 above) of approximately 18,653 ft 3 (528 m3) of soil and all annual projected future disposals of septic waste, Cooling Tower silt, and low-level contaminated soil volumes through the end of the plant's current operating license (year 2013). Determination of the radiological dose impact has been made based on the same models and pathway assumptions as indicated in Appendix B of the Vermont Yankee ODCM and approved as part of the original disposal analysis application for septic waste. Both the existing accumulated and future potential soil material will be dispersed using typical agricultural dry bulk surface spreading practices in approved disposal areas on site. Incidental pieces of asphalt and large stones that are picked up with the soil will be screened out before the soil/sand is spread. Records of the disposal that will be maintained include the following (as prescribed in Reference I, Appendix B): (a) the radionuclide concentrations detected in the soil/sand (measured to environmental lower limits of detection) (b) the total volume of material disposed of (c) the total radioactivity in the disposal operation as well as the total accumulated on each disposal plot at the time of spreading (d) the plot on which the soil was applied, and 8 Appendix J Original Off-Site Dose Calculation Manual Page 11 of57 Vermont Yankee Nuclear Power Station

APPENDIX J (Continued) (e) dose calculations or maximum allowable accumulated activity detenninations required to demonstrate that the dose limits imposed on the land spreading operations have not been exceeded. To ensure that the addition of the soil containing the radioactivity will not exceed the boundary conditions, the total radioactivity and dose calculation will include all past disposal operations of septic waste, Cooling Tower silt and soil/sand decay-corrected to the date of the latest spreading placed on the designated disposal plots. In addition, concentration limits applied to the disposal of earthen type materials (dry soil) restrict the placement of small volumes of materials that have relatively high radioactivity concentrations. Any farmer leasing land used for the disposal of soil will be notified of the applicable restrictions placed on the site due to the spreading oflow level contaminated material. These restrictions are the same as detailed for septic waste spreading as given in Reference l. The disposal opt:ration of che soil pilt.!s wiil foliow the applicable Vermont Yankee procedures to maintain doses as low as reasonably achievable and within the specific dose criteria as previously approved for septic and Cooling Tower silt waste disposal. 3.3 Disposal Plot Characteristics All designated disposal sites (six different plots) are loc;ated on the Entergy Nuclear Northeast Vermont Yankee plant site and are within the site boundary security fence. The South field consists of approximately 1.9 acres and is centered approximately 1500 feet South of the Reactor Building. This field has been the only one of the NRC approved fields that has actually been utilized for this purpose to-date. It is anticipated that future disposal operations will also utilize the South field since sufficient margin in comparison to the approved dose limit criteria still exists for anticipated waste disposal of the existing backlog of soil now in storage, plus all expected future disposals of septic waste, Cooling Tower silt and soil I sand mixes assuming the same observed generation rates (see Tables 13 through 19) p;;rsist to the en3 cf the plant license in 2013. In addition to the South field, the north end of the site has an additional ten acre parcel centered approximately 2,000 feet northwest of the Reactor Building. Prior assessments have demonstrated that a single plot of about 2 acres is sufficient to meet routine or expected disposal needs. Therefore the northern site could be subdivided into 5 plots if additional capacity was needed. 3.4 Radiological Impact Methodology The amount of radioactivity added to any of the disposal fields is procedurally controlled to insure that doses are maintained within the prior approved limits of the boundary conditions (see Section l.O above). To assess the dose received (after the spreading of the existing 18,653 ft3 [528 m3] along with both past recorded disposal applications, plus projected future applications) by the maximally exposed individual during the period of plant control, and to an inadvertent intruder after plant control of access ends (reference year of2013), the same pathway modeling, assumptions and dose calculation methods as approved for septic waste, Cooling Tower silt and past soil I sand disposals are used. 9 Appendix J Original Off-Site Dose Calculation Manual Page 12 of 57 Vermont Yankee Nuclear Power Station

APPENDIX J (Continued)

"Ibese dose models implement the methods and dose converi.ion factors as provided in Regulatory Guide 1.109 (Reference 2);

The following six potential pathways were identified and included in the analysis: (a) Standing on contaminated ground, (b) Inhalation of resuspended radioactivity, (c) Ingestion of leafy vegetables, (d)

Ingestion of stored vegetables, (e) Ingestion of meat, and (t) Ingestion of cow's milk Both the maximum individual and inadvertent intruder are assumed to be exposed to these pathways, with the difference between them being the occupancy time. The basic assumptions used in the radiological analyses include:

(a) Direct exposure to ground contamination and inhalation of resuspended radioactivity from the ground by movement of air is for a period of I 04 hours per year during the Vermont Yankee active control of the disposal sites and continuous thereafter. The 104-hour interval is representative of a farmer's time spent on a plot of land (4 hours per week for 6 months). The resuspension factor for soil material on the ground back into the air is taken as l .OE-05 based on an assumption that the disposal field will display characteristics similar to semiarid grassland experimental results. (NUREG-75/014; WASH-1400,"Reactor Safety Study", Appendix VI, Table VI E-3; USNRC, October 1975] (b) For the purpose of projecting and illustrating the magnitude of dose impacts over the remaining life of the plant, it is assumed that future disposals of septic, silt and soil material will be placed annually on the same field at the annual average radioactivity levels observed for these waste streams over the past fourteen years. The future disposals will also consist of the annual average radioactivity content observed in the accumulated :::.Jillsw.d materials colh:ctcd over the last several years that involved site facility construction projects that has lead to the existing backlog. The maximum individual dose impact from the buildup of disposed material occurs at the same time (2013) for both the Control Period and Intruder scenarios. (c) For the analysis of the radiological impact during the Vermont Yankee active control of the disposal sites until 2013, no plowing is assumed to take place and all dispersed radioactive material remains on the surface forming a source of unshielded direct radiation. (d) The crop exposure time was changed from 2160 hours to 0 hours to reflect the condition that no radioactive material is dispersed directly on crops for human or animal consumption. Crop contamination is only through root uptake. (e) The deposition on crops ofresuspended radioactivity is insignificant. 10 r-11> Appendix J Original Off-Site Dose Calculation Manual Page 13 of 57 Vermont Yankee Nuclear Power Station

APPENDIX J (Continued) (f) Most of the pathway data and usage factors used in the analysis are the same as those used in the Vermont Yankee's ODCM assessment of off-site radiological impacts from routine releases. The fraction of stored vegetables grown on the contaminated land was conservatively increased from 0.76 to l.O (at present no vegetable crops for human consumption arc grown on any of the approved disposal plots). Also, the soil exposure time to account for buildup was changed from the standard 15 years (given in Reference 2) to 1 year. (g) It is conservatively assumed that Vermont Yankee relinquishes control of the disposal sites after the current operating license expires in 2013 (i.e., the source term accumulated on a single disposal plot applies also for the inadvertent intruder at that time). (h} For the analysis of the impact after Vermont Yankee control of the site is relinquished, the radioactive material is ploVr'.ed under and forms a uniform mix with the top six inches of the soil, but, nonetheless, undergoes resuspension in the air at the same rate as the unplowed surface contamination. Eowever, for Jirect ground plane exposure the self-shielding due to the six-inch plow layer reduces the surface dose rate by about a factor of four. As shown in Reference I (Appendix B) for the original analysis in septic waste, the liquid transport and exposure pathway was found to be an insignificant contributor to the dose. Restrictions on the placement of the disposal plots put them at significant distances from wet lands, potable well supplies, and surface waters (Connecticut River). Therefore, the liquid pathway is not considered in this analysis. The dose models and methods used to generate deposition values and accumulated activity over the operating life of the plant are documented in Reference I (Appendixes Band F). Table 18 presents the radioactivity that currently exists on the South field after the last spreading event, which occurred on November 4, 2003. Table 18 also indicates the residual radioactivity that would remain on South field at the end of the plant operating period (2013) if no additional disposals were to take place. II

                                                         .T-1~

Appendix J Original Off-Site Dose Calculation Manual Page 14 of 57 Vermont Yankee Nuclear Power Station

APPENDIX J (Continued} 4.0 ASSUMPTIONS AND INPUTS

1) The volume of the accumulated soil I sand currently in storage between the Cooling Towers (as of December, 2003) was estimated from field measurements of.length, width, height and general shape of each pile taken in 2003. The estimated volume of each pile is summarized on Table I.

2) The radioactivity content of each pile was determined by averaging the numerious grab samples (typically 30 samples per pile) collected for characterization of the soil material collected. Appendixes A through F provide the individual results of positive analysis for plant related radionuclides. Laboratory analyses were performed either by Vermont Yankee or the AREVA-Framatome (formerly the Yankee Atomic I Duke Engineering)

Environmental laboratory with samples counted with respect to the NRC environmental LLD requirements as indicated in the VY ODCM.

3) Appendix G provides the total accumulated radioactivity on the South disposal Plot (1.9 acres) decayed to the date of the last spreading of.waste materials of November 4, 2003.

4) Dose Conversion Factors (DCF) specific to the land spreading of materials at Vermont Yankee were taken from the VY ODCM, Appendix F, Tables l J and 12 (Reference 1).

These DCF's were based on the same dosimetric models and input parameters as used in the original analysis of septic waste spreading at VY and which was approved by the NRC for inclusion in the ODCM as Appendix B. Section 3.2 of this calculation provides an outline of the key aspects of the dose model and asswnptions used. The following (Tables 2 & 3) listing notes these site specific dose conversion factors for both the Control Period and the Intruder scenario for the nuclides detected as positive in one or more of the sample analyses. Table2 Site Specific Control Period Dose Conversion Factors Max Organ Whole Body Isotope lndlvldual/Organ DCF Control DCF Control Half-Life Decay Constant (A) (mremlyr- (mrem/yr-* days Yr-1

                                           µCi/acre)      µCi/acre)

Mn-54 AdulUGl-LLI 3.75E-04 1.93E-04 3.125E+02 8.113E-01 Co-60 Teenll..ung 7.17E-04 5.31E-04 1.925E+03 1.315E-01 Zn-65 Child/Liver 1.64E-02 1.03E-02 2.438E+02 1.038E+OO C&-134 Child/Liver 3.18E-03 1.28E-03 7.531E+02 3.356E-01 Cs-137 Child/Bone 2.66E-03 7.02E*04 1.102E+04 2.290E-02 Ce-141 Teen/Lung 1.54E-04 1.50E-05 3.250E+01 7.788E+OO Ce-144 Teen/Lung 6.00E-04 2.44E-05 284.6 8.888E-01 12 Appendix J Original Off-Site Dose Calculation Manual Page 15 of 57 Vermont Yankee Nuclear Power Station

APPENDIX J (Continued) Tnblc3 Site Specific Intruder Dose Conversion Factors Max Organ Whole Body Isotope DCF Intruder DCF Intruder (mremlyr-µCllacre) ( mrem/yr-µC i/acre) Mn-54 Teen/Lung 1.02E-02 3.12E-03 Co-QO Teen/Lung 3.19E-02 9.09E-03 Zn-65 Child/Liver 1.89E-02 1.25E-02 Cs-134 Child/Liver 1.21E-02 9.36E-03 Cs-137 Child/Bone 6.98E-03 3.85E-03 Ce-141 Teen/Lung 1.21E-02 3.44E-04 Ce-144 Teen/Lung 5.00E-02 1.52E-03 13 Appendix J Original Off-Site Dose Calculation Manual Page 16 of 57 Vermont Yankee Nuclear Power Station

APPENDIX J (Continued) 5.0 EVALUATIONS In order to demonstrate compliance with the boundary dose conditions as stated in the ODCM, the critical organ and whole body dose from all pathways to the maximally exposed individual during Vermont Yankee Control Period and to the Inadvertent Intruder (for time periods following the end of the current operating plant license scheduled for 2013) were calculated for several scenarios (case studies) or combinations of disposal options. The dose calculations were performed using the site-specific dose factors for detected radionuclides as presented in Tables 2 and 3 (obtain from the VY ODCM, Appendix F, Tables 11 and 12). The objective is to demonstrate that the addition of the existing soil/sand materials currently in storage will not cause the radiological dose limits for materials spread on the South disposal field to be exceeded, even if it is assumed that all projected future annual disposals of septic waste, cooling tower silt and excavated soils/roadway sand containing the observed historical levels of plant related radionuclides are also placed on the same disposal plot. 5.1 Case Study I (Past Spreading lmpactfil The first case study (Case I) evaluated the spreading related dose impact associated with the past septic, Cooling Tower silt and soil spreading activity only. Table 4 shows the annual history and total amount of radioactivity in septic, silt and soil/sand waste materials by radionuclide that h-as been spread on the South field for the past 14 years (last spreading on 1114/03 ). These radioactivity disposal values were taken from the past spreading records. Using (multiplying) the dose conversion factors listed on Tables 2 and 3 along with the total accumulated radioactivity content on the South disposal plot as of the last waste spreading in 2003 as shown on Table 18, the committed dose impact is found on Tables 21 and 22 for the Control Period dose as of the last spreading on November 4, 2003, as well as the Intruder Dose projected to 2013 from all materials currently spread on the South disposal plot. This assessment assumes no other material is spread on the disposal plot in the future, and therefore represents the existing dose commitments from all past spreadings. This establishes the dose margin in comparison to the Control Period and Intruder dose limits still available for the South disposal plot. 5.2 Case Study II (Stored Soil/Sand Inventory Impacts) The second case study (Case II) looks at the spreading dose impact (Control Period and the Intruder impact at the end of plant license) first from the radioactivity associated with the existing soil in storage (18,653 ft3 as indicated on Table I) between the Cooling Towers, and then in combination with all past spreadings in order to demonstrate that the single south disposal plot can accommodate the current backlog of soil now being stored if it were all spread on top of all past disposals. Table 18 indicates the total accumulated septic, silt and soil activity per radionuclide remaining on the disposal plot as of the last spreading (I 1/04/03), as well as decayed to the projected reference date (6/1/04) for the spreading of all existing soil being held in storage, and to the projected end of the current plant license in 2013. Tables 8, 9, 10 and I I provide the estimate of radioactivity content in each of the storage piles of soil/sand being held between the Cooling Towers for the detected nuclides Cs-I 3 7, Co-60, Zn-65, and Mn-54, respectively. The activity concentrations are based on multiple grab samples collected 14

J-1/

Appendix J Original Off-Site Dose Calculation Manual Page 17 of 57 Vermont Yankee Nuclear Power Station

APPENDIX J (Continued) from each of the six separate storage piles, and are taken from the grab sample laboratory radiological analyses for each of the storage piles provided in Appendixes A through F. The average Cs-137 concentration in each of the soil piles was determined by including in the average the minimum detectable concentration (MDC) of each radionuclide because most of the samples indicated a positive concentration for Cs-137. This slightly biases the assessment towards a conservative upper bound estimate of the potential activity in the pile since the use of the MDC does not represent the existence of positive measured activity in the sample, but is treated as such. For the other detected radionuclidcs (Co-60, Zn-65 and Mn-54), this biasing of the data was not applied since the occurrence of positive values only represented a small fraction of the total number of samples taken. The total activity determination for each pile is then calculated by taking the average measured concentration times the measured soil density times the estimated volume of the storage pile, correcting for decay time between the date of the sample analysis date to the estimated date of field disposal (6/1/04). This calculated total radionuclide activity value for each pile is assumed to be placed on the single 1. 9 acre South disposal plot with the resulting surface concentration (uCi/acre) for the projected disposal calculated for each pile and totaled for all six piles currently in storage. The estimated average concentration for each of the four detected radionuclides is decay corrected from the date of the sample collection to June 1, 2004, as the reference date for estimating dose impacts from the proposed disposal of the accumulated material in storage. Table 12 summarizes the radioactivity associated with the soil/sand in storage decayed to June 1, 2004, and to the estimated end of the current plant license in 2013 for dett!nnining the Intruder dose impact at that time. Table 20 combines the radioactivity content of all soil/sand materials stored between the Cooling Towers with all remaining activity previously spread on the South disposal plot, decayed to 2013 for use in determining the Intruder dose at the end of plant license. Table 23 applies the dose conversion factors from Table 2 for maximum organ and whole body doses with the projected activity on the South disposal plot from Tables 12, 18 and 20 to find the Control Period dose for all past material spreadings, stored material additions, and the sum total of past spreadings and proposed stored material additions. Table 24 illustrates the same dose impact combination of waste streams as applied to the site Intruder at the en9 of plant license. The combination of all past spreadings and the current stored soil material results in a maximum organ dose of only 17.7 % (0.177 mrem/yr) of the 1 mrem/yr Control Period dose limit. The maximum Intruder organ dose is estimated to be 0.456 mrem/yr or 9.1% of the 5 mrem/yr Intruder scenario dose limit. For comparison, Table 25 indicates that the maximum organ dose during the Control Period from only the inventory of 18,653 tt3 of soil/sand is estimated to be 0.0526 mrem/yr, or 5.3% of the 1 mrem/yr dose limit. 5.3 Case Study III (Projected Future Spreading Impacts) The third case study (Case III) projects what the likely annual spreading additions of earthen material from all sources (septic waste, Cooling Tower silt, soil/sand mixes) would be based on historical records, combined with the existing 18,653 ft3 (528 m3 ) of material in storage, in order to determine the long term acceptability of the South disposal plot to continue to be used for all waste spreading applications. Based on the historical spreading data listed in Table 4, Tables 5, 6 and 7 show the total accumulated septic waste, Cooling Tower silt and soil/sand annual average field spreading surface concentrations by radionuclide and waste source, respectively. This data breakdown is used in this disposal case study to predict future disposal rates to be applied to the South disposal plot. Table 13 provides a 15 Appendix J Original Off-Site Dose Calculation Manual Page 18 of57 Vermont Yankee Nuclear Power Station L

APPENDIX J (Continued) summary of the buildup of future spreadings over time after 2004 from all three waste streams (Septic waste, Cooling Tower silt, and soil/sand mixes), which could be projected to accumulate on the South disposal plot by 2013. The annual disposal quantity for each waste stream is based on the average annual disposal quantity observed for each stream since on-site disposal was originally approved in 1990. The total of all three streams (septic waste, Cooling Tower silt, and soil/sand) is taken to be representative of the future generation rate for each year from 2004 until 2013. The last column of Table 13 indicates the decay corrected accumulated activity on the South disposal plot from only additions to the South field from all future waste earthen materials. Tables 14, 15 and 16

provide additional detail of the projt."Cted annual and accumulated materials by waste stream (i.e.,

sand/soil mix, Cooling Tower silt and septic waste). The buildup equation used in Tables 13 through 18 accounts for both annual additions to the field as well as decay over this in-growth period is given by: Act; (t) = Act1 (a)"' (I- E£1* 1l)/(I - E) Where: Act; (t) = the total activity ofradionuclide "i" (uCi) remaining at the end of the buildup period, t (years). Act1 (a)= the annual radioactivity addition of nuclide "i" to the disposal plot in uCi. The values for projected future additions are based on the annual average value observed for that nuclide for the specific disposal stream (i.e., septic waste, cooling tower slit, soil/sand). E exp(-A.; L\t) is the decay constant for the selected radionuclide "i" (I/year) L\t time interval between applications= I year. In addition, Case Study II above evaluated the radiological impact from the disposal of the existing inventory of soil projected for mid-year, 2004 (including all past wa.sce spreading operations), the total impact for 2004 should also include a projected disposal of both septic waste and Cooling Tower silt from one year's operation. Table 17 combines one year's generation of both septage and silt for assumed spreading in 2004, plus subsequent decay to 2013. Table 26 combines the site-specific dose conversion factors from Tables 2 and 3 for the Control Period and Intruder scenario, respectively, with all previously spread radioactivity on the South disposal plot (Table 18) with both the proposed disposal of the existing soil stored inventory (Table

12) and projected annual additions of earth type waste materials based on the observed average annual generation rate (Tables 13 through 17). Table 19 provides a summary of accumulated radioactivity on the South field from the past spreading, materials in storage and future annual disposals on the same plot out to the assumed end of plant license, which corresponds to the Intruder dose scenario time frame. The resulting doses to the maximum organ and whole body of the maximum individual at the end of plant license (Table 26) reflects the maximum expected impact from all past and future disposals being placed on the South disposal field.

16 J-\q Appendix J Original Off-Site Dose Calculation Manual Page 19 of57 Vermont Yankee Nuclear Power Station

APPENDIX J (Continued) 6.0 RESULTS I CONCLUSIONS The evaluation of the radiological impact of all past, accumulated storage inventory, and projected future waste spreading operations on the 1. 9 acre South disposal plot have shown that the existing field is being operated within the previously approved dose limit criteria. The specific findings include: I. For Case Study I (Past Spreading Impacts), Table 21 shows that after 14 years of spreading septic waste, Cooling Tower sill, and soil/sand mixes on the a single, 1.9 acre disposal plot the committed dose impact results in a maximum organ dose to a critical receptor (Control Period use) that accounts for only 13.7% of the 1 mrem/year limit. With respect to the Intruder dose limit of 5 mrem/year at the end of assumed active property control (i.e., end of plant license assumed for dose projection purposes), Table 22 indicates that all past spreadings through the end of2003 account for only 7.2% of the maximum organ dose to the limiting receptor at the end of plant license. These finding illustrate that there is significant capacity remaining in the South disposal plot to continue to accept additional earthen materials that are suitable for iand spreading without exceeding established dose limitations.

2. Table 23 shows that the impact from the projected spreading of the existing 18,653 ft 3 of

soil/sand material in storage is estimated to account for only 5.3% of the 1 mrem/yeat Control Period dose limit, or only 17. 7% of the same limit when all past spreadings are combined with the materials currently in storage (as of the end ofNovember 2003). The maximum Intruder organ dose from all past spreading and stored materials is calculated to be 0.456 mrem/yr (Table 24), or 9.1 % of the 5 mrem/yr Intruder scenario dose limit. These results indicate that the existing inventory of waste materials in storage can be placed on the South disposal plot without exceeding dose impact limits previously established for a single disposal field, or using a significant proportion of the South disposal plot's capacity to receive additional materials for disposal in the future.

3. Assuming the same annual average generation rate ofradioa~tivity in waste materials (septage/silt/soil) that has been observed over the last 14 years is added each year through 2013 to all pasl was le spreadings (including the stored suiis invemory) already committed to the South disposal plot, Table 26 indicates that the projected dose at the end of the current plant licensing period (year 2013) yields a limiting maximum critical receptor dose (for either the control period or inadvertent intruder) equivalent to only 25.3% of the most restrictive annual dose limit (associated with the 1 mrem/year limit for the maximum organ during the Control Period). This finding demonstrates that the continued use of the South disposal plot, even with the addition of 18,653 ft 3 of slightly contaminated soil currently in storage, will not exceed the approved limiting dose criteria established in the ODCM.

4. The approved dose impact methodology used to determine compliance with the on-site spreading dose limits are not driven by the volume of waste material disposed of, but by the total radioactivity content of the material that is spread over a fixed disposal plot area ( 1. 9 acres for the South field). The dose modeling assumes that all radioactivity spread on the field remains in the top 15 cm surface layer of soil, even after subsequent additions are 17 j-lV Appendix J Original Off-Site Dose Calculation Manual Page 20 of57 Vermont Yankee Nuclear Power Station

APPENDIX J (Continued) placed on the same field area. Existing limits on the concentration of radioactivity in waste media provides protection from small volumes of"hot" or high specific activity materials from being spread on the disposal field. 18 Appendix J Original Off-Site Dose Calculation Manual Page 21 of57 Vermont Yankee Nuclear Power Station

APPENDIX J (Continued) Table4 Reaxd of Septic I Coollng T O\Wr Siii / Construction Soll Radioactive Metnrial Spreading Each Year on the South Disposal Field Spreading Material Mn-54 Co-60 Zn-65 Cs-134 Cs-137 Ce-141 Year Dale Type (uCilaae) (uCi/aae) (UCl/aae) (uCilaae) (uCilaae) (uCi/aae) 1990 10/31/90 Septage 0 3.89 0 0 0.26 0 11120/90 Seplage 0.17 2.03 0.41 0 0.29 1.40E-08 1991 none 0 0 0 0 0 0 1992 10/19192 Seplage 0.11 1.73 0.52 0.05 0.32 0.006 1993 10/14193 Septage 0.05 1.41 0.21 0 0.3 0 1994 06/14194 Sept age 0.08 0.43 0 0 0.09 0 1995 06/29195 Septage 0 0.88 0 0 0 0 1996 no no 0 0 0 0 0 0 1997 06/18197 Septage 0.12 1 0 0 0.19 0 1998 07'30198 Seplage 0.14 0.72 0.09 0 0.12 0 09/18198 CTSil 0 0 0 0 30.87 0 l~l G7/151!)9 s..pii'ge C.11 1.n 0.2 0 0.25 0 2000 08/09/00 Seplage" 0 0 0 0 0 0 10/24/00 CTSill 0.117 0.68 0 0 0 0 10/24/00 Soii/Send 0 0.602 0 0 3.698 0 2001 06-20-01 Septage 0 4.078 1.088 0 0.156 0.089 09/25/01 Soil/Sand 0 0 0 0 1.4 0 2002 06/21/02 Septage 0.01 0.04 0 0.001 0.01 0 11/11/02 Soil/Sand 0 0 0 0 1.37 0 2003 07/01/03 Septage 0 1.03 0 0 0 0 10/25/03 Seplage 0 0.12 0 0 0 0 11/04l03 Soil/Sand 0 0 0 0 1.34 0 11/04l'03 CTSill 0 0.256 0 0 0 0 Average Adtv'ity/yr (uCi/ecre): 0.06 1.45 0.18 0.0036 2.90 0.01 (Over 14 year sprnading hislllf)') Average ecliviy (uCl/yr) 0.123 2.76 0.342 0.007 5.52 C'.C13 disposed of on 1.9 aCl'e field each year

No radioactivity detected in septic waste 58111ples.

19 r-n-Appendix J Original Off-Site Dose Calculation Manual Page 22 of57 Vermont Yankee Nuclear Power Station

APPENDIX J (Continued) Table5 Reconf ol Septic Waste Only for Rndioadive Material Spreading Each Year on the Sooth Disposal Fleld Spreading Material Mn-54 Co-M Zn-65 Cs-134 Cs-137 Ce-141 Year Data Type (UCllacre) (UCVecre) (uCVacre) (uCVBa'e) (uCVacre) (uCilacre) 1990 10/31190 ~e 0 3.89 0 0 0.26 0 11120/90 Septage 0.17 2.03 0.41 0 0.29 1.40E-08 1991 none 0 0 0 0 0 0 1992 10/19192 Septage 0.11 1.73 0.52 0.05 0.32 0.006 1993 10/14193 Septage 0.05 1.41 0.21 0 0.3 0 1994 06/14194 Septage 0.08 0.43 0 0 0.09 0 1995 06/29/95 Septage 0 0.88 0 0 0 0 1996 none 0 0 0 0 0 0 1997 06/18197 Septage 0.12 0 0 0.19 0 1998 07/30/98 Sept-age 0.14 0.72 0.09 0 0.12 0 1999 07/15199 Septage 0.11 1.47 0.2 0 025 0 20CO C8/09IOO Septage* 0 0 0 0 0 0 2001 06-20-01 Septage 0 4.078 1.088 0 0.158 0.089 2002 06/21/02 Septage 0.01 0.04 0 0.001 0.01 0 2003 07/'01/03 Septage 0 1.03 0 0 0 0 10/25103 Septage 0 0.12 0 0 0 0 Average Adivitylyr (uCVacre): 0.06 1.34 0.18 0.004 0.14 0.01 (Over 14 year spreading history) Avarag11 edlvlty (UCi/yr) 0.107 2.58 0.342 0.007 0.27 0.013 disposed of on 1. 9 acre field each year

No radioactivity detected in septic waste samples.

Appendix J Original Off-Site Dose Calculation Manual Page 23 of 57 Vermont Yankee Nuclear Power Station

APPENDIX J (Continued) Table6 Reooro of Cooling Tower Sill Waste Only for Radioactive Material Spreading Each Year on the South Disposal Field Spreading Material M,.._54 Co-60 Zn-65 Cs-134 Cs-137 Ce-141 Year Date Type (uCl/aae) (uCi/acra) (uCl/aae) (uCi/acra) (uCilaae) (uCl/aae) 1998 09/18198 SHI 0 0 0 0 30.87 0 2000 10/24'00 Silt 0.117 0.68 0 0 0 0 2003 11/04/03 Silt 0 0.256 D 0 0 0 Average AdN"ify/yT (uCl/acre) 0.004 0.030 0.000 0.000 0.996 0.000 (31 yaar silt generation history) Average activiy (uCVyr) 0.007 0.057 0.000 0.000 1.892 o.ooo disposed of Oil 1.9 aae field Note: Coollng Tower yearly average Is over 31 y&ars of operation since the ftrst dlsoosal In 1998 included aQ accumulated matenal since plant startup Table? Recoro of Conslrucllon Soll/Sand Waste Only for Radioactive Material Spreading Each Year on th& South Disposal Field Spreading Mater.al Mn-54 Co-60 Zn-65 Cs-134 Cs-137 Ce-141 Year (UC I/acre) (uCl/aae) (uCi/aae) (uCl/acre) Date Type (uCilaae) (uCl/aae) 2000 10/24/00 SolllSalld 0 0.602 D 0 3.698 0 2001 09/25/01 Soil/Sand 0 0 0 0 f.4 0 2002 f f/11/02 Soll/Sand 0 0 0 0 1.37 0 2003 11/04J03 Soil/Sand 0 0 0 0 1.34 D AV9rag& Activity/yr (uCl/acra): 0.00 0.15 0.00 0.00 1.95 o.oo Average adlvly (uCUyr) disposed of 0.00 0.29 0.00 0.00 3.71 o.oo on 1.9 aae field per year Note: Soil/ road sand yearly average Is over only the 4 years of operation since that Is the period of material collection. 21 J-1-1 Appendix J Original Off-Site Dose Calculation Manual Page 24 of57 Vermont Yankee Nuclear Power Station

Table 8 Cesium -137 in storage Plies after Last Spreading 2003" Activity decayed to 06/01/04 Pile desaiptlon Estimated Date of Decay time Aver. Cs- Measured Total Cs- Aver. Cs- Cs-137 Volume Analysis to 6/1/04 % Cs-137 137 Conc.(w density 137 137 applied to of total

                                                                                       /LLD) No                   (decayed)       Conc.(w/        1.9 acre decay                    (w/LLD)          LLD)            field decayed 11"3                     ears               uCVgm         gm/cc         uCi           uCVgm         uCl!acre 2002-01: Security Fence Upgrade            4,679     OSJ00.'02       2.08            7.12E-08      1.07       9.62E+OO         6.79E-08       5.07E+-OO 2002-02: Security Fen<:e Upgrade            9,000     06/04/02        2.00                                                                                    31.46%

Park Lot Sweep (Inside protected 616 11/02/J1 2.58 4.18E-08 4.80E-08 1.11 1.13!:~1 3.99E-08 5.94E+OO 36.95% >

                                                                                                                                                                          '"O area)                                                                                              1.56       1.23E+-OO        4.52E-OB       6.48E-01        4.03%       '"O 2001 HWC Soils Excavation                     931     10l22!01                                                                                                             m 1996 Remnants( mixed contam. +               970     04/1;;,~5 2.58 8.13 5.41E-08 9.22E-08 1.057       1.42E+OO         5.10E*OB       7.48E..Q1       4.65%       z non-cont)                                                                                           1.7        3.57E+-OO       7.65E-08       1.88E+OO       11.69%        CJ 2002-03 Securily Fence Chunks &

soil mix 2,457 12/16/02 1.54 4.11E-08 1.24 3.42E+OO 3.97E-OB 1.SOE+OO 11.20% x totals= 18,653 n c Average" 5.81E-08 total= 3.C~E+01 5.34E-08

~-

1.61E+01 100.00% t: (";. 0...

Note: Soil analysis data provided in Appendixes A through F.

22

< (1) s0 Table 9 g Pl

i

 ?;"'                                                                Cobalt -60 in Storage Piles after Last Spreading in 2003*         Activi;y decayed to 06/01/04 (1)

(1) zs:: Pile desaiption Estimated Date of (") Decay Aver.Co-60 Measured Total Co-60 Aver. co-eo Co-60 % Co-60 0 Volume Analysis time to Cone., no density (decayed) Cone.. decayed applied to 1.9 Pl of total .... 6/1/04 decay acre field "'0

(1) ft-'3 years uCVgm gm/cc ut:i uCVgm uCi/acre .... 2002-01: Security Fence Upgrade 4,679 05/Ql'l/02 2.08 O.OOE+OO 1.07 0.00E+OO O.OOE+OO O.OOE+OO 0.00% (/) 2002-02: Security Fence Upgrade 9,000 06/04/02 2.00 O.OOE+OO 1:11 O.OOE+OO

6) 0.00E+OO O.OOE+OO 0.00%

..... Park Lot Sweep (Inside protected area) 616 11/02/01 2.58 B.72E-09 1.56 1.69E-01 B.22E-09 5* 2001 HWC Soils Excavation 931 10/22/01 2.58 O.OOE+OO 1.057 o.ooi:+oo 0.00E+-00 B.90E-02 O.OOE+OO 33.01% >

i w

I 1996 Remnants( mixed contam. + non-cont) 970 0411~!95 B.13 2.14E-08 1.7 3.43!:-01 1.7BE-08 1.81E-01 0.00% 66.99%

                                                                                                                                                                                                 '"O
                                                                                                                                                                                                 '"O tTl z
                         ~

2002-03 Security Fence Chunks & soil 2,457 12116/02 1.54 O.OOE+-00 1.24 O.OOE+OO mix O.OOE+OO O.OOE+OO 0.00% 0 totals= 18,653 average= 5.02E-09 total= 5.12E-01

Note: Soil analysis data provided in Appendixes A through F.

4.33E-09 2.70E-01 100.00% nc

                                                                                                                                                                                                 ~
                                                                                                                                                                                                 ~

r.c::...

      '1:::l 0     :>>
      ~
      ~
             ~'O I   'O
      "      (/)   (1)

N 0\ ct ::le: 0 0 ~

      ....., 0     ._

Ul Cll

      -...J  (1) 0....

(la'Q" Pl : :-i* (") Pl s:: - [ 5*

i

Pl

             ~                                                                                                    23 e.

...,n<

s 0 Table 10

="" Zinc -85 In Storage Piles after Last Spreading In 2003* Activity decayed to 06/01/04 n n Pile description Estimated

Date of Decay Aver. Zn- Measured Total Zn-65 Volume Analysts Aver. Zll-65 Zn-65 %Zn-65 lime to 65 Cone., density (decayed) (') Cone., applied to of total

~                                                                                                               611/04      no decay                                  decayed     1.9 acre

...."" f!A3 years uCVgm gm/cc uCI uCVgm field ~ 0 ~

2002--01

Security Fence Upgrade 4,679 05/06/02 2.08 O.OOE+OO 1.07 O.OOE+OO n

..., 2002-02: Security Fence Upgrade 9,000 0.00E+OO O.OOE+OO 0.00% 06104/02 2.00 3.87E-09 1.11 Peri< Lot Sweep (Inside protected area) 1.05E+OO 3.70E-Q9 5.SOE--01 95.04% C/:J 616 11/02/01 2.58 O.OOE+OO 1.56 O.OOE+OO g 2001 HWC Soils Excavation 1996 Remnants( mixed conlam. + non-931 10122/01 2.58 2.0SE-09 1.057 5.46E-02 O.OOE+OO 1.96E--09 O.OOE+OO 2.87E-02 0.00% 4.96% 5* 970 04/13195 8.13 O.OOE+OO 1.7 cont) O.OOE+OO O.OOE+OO = 2002-03 Security Fence Chunks & soil 2,457 12116/02 1.54 O.OOE+OO 1.24 O.OOE+OO 0.00E+OO 0.00%

                                                                                                                                                                                                       ""O mix                                                                                                                              O.OOE+OO    O.OOE+OO    0.00%     ""O
                            '-i                                                                                                                                                                         tT1 l

totals" 18,653 Average= 9.92E-10 total= 1.10E+OO 9.43E-10 5.79E-01 z 100.00% 0

Note: Soil analysis data provided in Appendixes A through F.

Q

                                                                                                                                                                                                       ~
        ~         0 ;l>                                                                                                                                                                                r.c...
      ~

n 7J:g cr.i n N

       ......,J CJ)
                         =
                  ...... 0.

0 0 >:

       ...., 0           ._

Ul"'

       -....!    n ...,  0 l.lcr'Q"
                 -=

(') "" e5* i:: - 24

                =
                ~
                =

E.

.... Table II E3 0 a

i

~ Mn-54 in Storage PHes after Last Spreading in 2003* Activity decayed to 06/01!04 0 0 z Pile description Estimated Date of Decay Aver. Mn-54 Mea&ured Total Mn-54 Aver. Mn-54 Mn-54 % Mn-54 {'; Volume Analysis Umeto Cone., no density (decayed) Cone., applied to of total ~ 611/04 decay decayed 1.9 acre '"ti f1*3 ~ears uCl/gm gm/cc uCi field 0 ___!!f.!!9.m uCi/acre 0 .... 2002-01: Security Fence Upgrade 4,679 05/06/02 2.08 2.09E-09 1.07 2.B2E.Q1 1.99E-09 1.49E-01 69.78% C/J 2002-02: Security Fence Upgrade 9,000 06/04/02 2.00 O.OOE+OO 1.11 O.OOE+OO O.OOE+OO ..... Park Lot Sweep Onside protected area) 0.00E+OO 0.00% ~ 616 11/02/01 2.58 4.77E-09 1.56 1.22E.Q1 4.50E-09 6.44E-02 30.22% 5*

i 2001 HWC Soils Excavation 931 10fl2/01 2.58 O.OOE+OO 1.057 O.OOE+OO O.OOE+OO O.OOE+OO 0.00% >......,

1996 Remnants (mixed contam. +non* 970 04/13195 8.13 O.OOE+OO 1.7 O.OOE+OO O.OOE+OO O.OOE+OO 0.00% "'O cont) [Tl 2002-03 Security Fence Chunks & soil mix 2,457 12116/02 1.54 O.OOE+OO 1.24 O.OOE+OO O.OOE+OO 0.00E+OO 0.00% z L~

totals" 18,653 Averagem 1.14E-09 total:: 4.05E-01 1.0BE-09 2.13E-01 100.00% 0

                           ,...._.I

-o

*~

Nore: Soil analysis data provided in Appendixes A through F.

                                                                                                                                                                                                    -n"-

0 as*

                                                                                                                                                                                                       ~
      '"ti 0 ;l>
      "' ......, 'O (IQ 'i" 'O 0        C/J 0 N
               ..... :::i
               -*     Q..

00 0 -* 0 tl ~

      ......, 0 ' -

Ul"'

      -.I 0 0

n C'l "' s:: ciQ.

               "'-:::i

[ 5*

i

25

              §"'

E..

APPENDIX J (Continued) Tobie 12 SoiVSand In Storage Total Activity lo be spread on 611/04 and decayed to 2013 As of6/1/04 Decay Time to As of2013 Isotope lamda 11yr Qs (uCUacre)" 611/2013 (yr.;) Qa

                                - - - - - - - - - i.----*-----*- ------*-*-                    __ (uCifaae)

Mn-54 0.8113 0.213 9 1.44E--04 Co-00 0.1315 0.270 9 8.27E-02 Zn-65 1.0382 0.579 9 5.07E-05 Cs-134 0.3356 0.000 9 O.OOE+OO Cs137 0.0229 16.1 9 1.31E+01 Ce-141 7.7883 0.000 9 O.OOE+OO

Note: Qa values from Table 8 (Cs- I 37), Table 9 (Co-60), Table I 0 (Zn-65), wid Tobie 11 (Mn-54)

Table 13 Projection of Additional Septic, sill. and soil/sand at Current Generation Rates to 2013 Annual Septic Annual Silt Annual Annual Additiona Addition Addition SoilfSand Total Accumulation at Addition Additions e_nd of9ye ars Isotope lamda 11yr Qa (uCifacre) Qa (uCifacre) Qa (uCifacre) Qa Qe (uCifacre) (uCilacrel Mn-54 0.8113 0.056 0.0038 0.043 0.103 0.185 Co-80 0.1315 1.345 0.0302 0.174 1.549 8.183 Zn-65 1.0382 0.180 0.0000 0.116 0.296 0.458 Cs-134 0.3356 0.004 0.0000 0.000 0.004 0.012 Cs137 0.0229 0.142 0.9958 4.782 5.919 43.767

  ~~~-- --------  7.7883      - -0.007 0.0000             0.000
                                                                       ---- --- - - - - - 0.007             O.OG7 26 Appendix J Original Off-Site Dose Calculation Manual Page 29 of57 Vermont Yankee Nuclear Power Station

APPENDIX J (Continued) Table 14 Projoction of Additional Sand/Soil Mix At Historical Generation Rates To Year 2013 (9 years after 2004) I Projected Annual Soil/Sand Additions Accumulated Sand/Soil to Field at end of 9 years lsotooe lamda '!!}!__ __ ----~~-(.':.1Cila£!:!'L ____ a~ (u~i/acre> Mn-54 0.8113 4.26E-02 7.65E-02 Co-60 0.1315 1.74E-01 9.21E-01 Zn-65 1.0382 1.16E-01 1.79E-01 Cs-134 0_3356 O.OOE+OO O.OOE-+00 Cs137 0.0229 4.78E+OO 3.54E+01 Ce-141 7.7883 O.OOE+OO O.OOE+OO Table 15 Projection of Additional Cooling Tower (CT) Silt At Historical Generation Rates To Year 2013 (9 years after 2004) Projocted Annual CT Accumulated CT Silt at Silt Additions to Field end of 9 years Isotope lamda 1/yr Qa (uCVucre) Qe (uCilacre) Mn-54 0.8113 3.77E-03 6.78E-03 Co-00 0.1315 3.02E-02 1.59E-01 Zn-65 1.0382 O.OOE+OO O.OOE+OO Cs-134 0.3356 O.OOE+OO O.OOE+OO Cs137 0.0229 9.96E-01 7.36E+oo Ce-141 7.7883 O.OOE+OO O.OOE+OO Table 16 Projection of Additional Septic Waste At Historical Generation Rates To Year 2013 (9 years after 2004) Projected Annual Septic Accumulated Septic Additions to Field Waste at end of 9 years lsolope lamda 1/yr Qa (uCilacre) Qe (uCi/acre) Mn-54 0.8113 5.64E-02 1.01E-01 Co-60 0.1315 1.34E+OO 7.10E+OO Zn-65 1.0382 1.80E-01 2.78E-01 Cs-134 0.3358 3.64E-03 1.19E-02 Cs137 0.0229 1.42E-01 1.05E+OO Ce-141 7.7883 6.79E-03 6.79E-03 27 Appendix J Original Off-Site Dose Calculation Manual Page 30 of57 Vermont Yankee Nuclear Power Station

APPENDIX J (Continued) Table 17 Projected 1 Yr Septage +Silt Spreading for6/1/04 ~!1~~eca:r:~_ to _2_()_1_~ _. ________ _ As of 611/04 D ecay Time to As of 2013 lsotoee lamda_ 1/yr Qa (uCilacre) 6/112013 --+--a_a_~Ci/acreL_ Mn-54 0.8113 0.060 9 4.00E-05 Co-00 0.1315 1.375 9 4.21E-01 Zn-65 1.0382 0.180 9 1.57E-05 Cs-134 0.3356 0.004 9 1.78E-04 Cs137 0.0229 1.138 9 9.26E-01 Ce-141 7.7683 0.007 9 2.45E-33 Table 18 Current Total Spreadings as of 11/4/03 and How much Remains at 6/1/04 and 2013 As of 11/41U3 Decay Time to As of6/1/04 Decay Time to As of2013 lamda Qa Qa Isotope Qa (uCi/acre)* 6/1/04 (yrs) 2013 (yrs) 1/vr iuCilacrel cuCilacrel Mn-54 0.8113 0.241 0.5753 0.151 9 1.02E-04 Co-{)() 0.1315 16.33 0.5753 15.140 9 4.64 Zn-65 1.0382 1.47 0.5753 0.809 9 7.08E-05 Cs-134 0.3356 0.00063 0.5753 0.001 9 2.53E-05 Cs137 0.0229 38.18 0.5753 37.68 9 30.66 Ce-141 7.7883 8.80E-10 0.5753 1.0E-11 9 3.60E-42

      *Data Taken from Plant Disposal Records 28 Appendix J Original Off-Site Dose Calculation Manual Page 31 of57 Vermont Yankee Nuclear Power Station

,~----------- APPENDIX J (Continued) Table 19 Current Spreading Totals Plus Total Projected Future Spreadings to 2013 All 1 Yr Future Current Field Act Stored SoiVsand Tolal al Activity Septage+sill Spreading Decayed to Decayed to *Decayed to Decayed lo 2013* Decayed to 2013 2013- 2013- 2013-*

               -                                                                 - - = - - - - - - * * . --*-*--- --* --*---* --- -****

Isotope (uCi/acre) (uCi/acre) (uCi/acre) (uCilacre) (uCi/acra) Mn--54 1.02E-04 1.44E-04 4.06E-05 0.185 1.85E-01

                      ~                      4.636                6.27E-02                0.421              6.163            1.33E+01 Zn-65                7.0BE-05               5.07E-05              1.57E-05             0.456            4.58E-01 Cs-134               2.53E-05               O.OOE+OO              1.78E-04             0.012            1.21E-02 Cs137                  30.66                  13.10                 0.926             43.767            8.65E+01 Ce-141               3.60E-42               O.OOE+OO              2.45E-33             0.007            6.79E-03 Notes:

Activity Concentration from Table 18.

                      .... Activity Concentration from 'i'ahle i 2 .
                      ***      Activity Concentration from Table 15, 16 and 17 .
                      ****     Activity Concentration from Table 13 and 14 .

Table20 Radioactivity Content from Existing Materials (Past Spreadings & Stored Materials Only) End Plant Operations for Intruder Dose: Last Application 6/112013 6/1/2004 End Date Date Decay duration to end of Plant 9 years Operations: Total Past+ Total Activity on Total Activity Past Material Stored Current Sou1h Field decayed to Spread only Material to be Material Spread plus storage year2013 up to 11 /4/03 decayed to Isotope decayed to piles (l!Cl/acre)" (uCi/ar-.re) d~yed to 2013 2013 (uCi/ar.re)

.l.013 1611/04\ (uCi/acra} (uCi/acre\

Mn-54 0.364 2.46E-04 1.02E-04 1.44E-04 2.46E-04 Co--60 15.41 4.72E+OO 4.64E+OO 8.27E-02 4.72E+OO Zn-65 1.388 1.21E-04 7.0SE-05 5.07E--05 1.21E-04 Cs-134 0.001 2.53E-05 2.53E-05 O.OOE+OO 2.53E-05 Cs-137 53.78 4.38E+01 3.06E+-01 1.31E+01 4.38E+01 L--~-111. ... --~*OE-11 3.60E--42 3.60E-42 O.Q_q~:+:QQ- ---~,.@_!;~~- - -

Includes a" material spread as of 11/4/03 decay corrected to the indicated date.

                                                                            . 29 Appendix J Original Off-Site Dose Calculation Manual Page 32 of57 Vermont Yankee Nuclear Power Station

APPENDIX J (Continued) Table 21 Past Spreading Control Period Doses As of I I /04/03 (No Stored Material or Future Additions Included) Isotope Total Activity Max Organ Whole Body Remaining on Existing Material Existing Material South Field as from Past from Past of11~3 Spreading Spreading uCi/aae Mremlyear Mremlyear Mn-54 0.241 9.04E-05 4.65E-05 Co-60 16.33 1.17E-02 8.67E-03 Zn-65 1.47 2.41E-02 1.51E-02 Cs-134 0.00063 2.00E-06 8.00E-07 Cs-137 38.18 1.02E-01 2.68E-02 Ce-141 8.8E-10 1.36E-13 1.32E-14 Total Dose= 1.37E-01 5.u7E-02 Dose Limit per 1 field=

                                 % of Dose 13.7%                 5.1%

limit Table 22 Past Spreading Intruder Period Doses At End of Pl8Jlt License in 2013 (No Stored Material or Future Additions Included after 1114/03) Isotope Total Activity Total Activity Max Organ Whole Body Remaining on Remaining on Existing Material Existing Material South Field as South Field from Past from Past of 11/04I03 decayed Spreading Spreading corrected to 2013 uCi/acre uCi/acre Mrem/year Mrem/year Mn-54 0_241 1.02E-04 1.04E--06 3.18E-07 Co-60 16.33 4.636 1.48E-01 4.21E-02 Zn-65 1.47 7.08E-05 1.34E--06 8.85E-07 Cs-134 0.00063 2.53E-05 3.07E-07 2.37E-07 Cs-137 38.18 30.66 2.14E-01 1.18E-01 Ce-141 8.8E-10 3.60E-42 4.36E-44 1.24E-45 Total Dose= 3.62E-01 1.60E-01 Dose Limit per 5 5 field=

                                       % of Dose limit              7.2%                  3.2%

30 Appendix J Original Off-Site Dose Calculation Manual Page 33 of 57 Vermont Yankee Nuclear Power Station

APPENDIX J (Continued) Table 23 Control Period Dose: Past Spreading & Current Stored Soil Inventory (as of 6/1/04) Isotope Mex Organ Whole Body MaxOrgan Whole Body Max Organ Whole Body Existing Existing All Past All Past Stored Material from Material from Stored Material Spreading Spreading Material to be Past Past to be Spread Plus Stored Plus Stored Sp read Spreading Spreading Inventory Inventory

                  ~-~m/ye_ar           Mrem/\'.ear__     Mremtye~-- --~~-~~r____ . _!!'Jrernlyear
                                                       ~--**--

_Mrer:nfY~ar Mn-54 5.67E-05 2.92E-05 7.99E-05 4.11E-05 1.37E-04 7.03E-05 Co-60 1.09E-02 8.04E-03 1.94E-04 1.43E-04 1.10E-02 8.18E-03 Zn-65 1.33E-02 8.33E-03 9.50E -03 5.96E-03 2.28E-02 1.43E-02 Cs-134 1.65E-06 6.65E-07 O.OOE +00 O.OOE+OO 1.65E-06 6.65E-07 Cs-137 1.00E-01 2.65E-02 4.2BE-02 1.. 13E-02 1.43E-01 3.78E-02 Ce-141 1.54E-15 1.SOE-16 0.00E +-00 O.OOE+-00 1.54E-15 1.50E-16

                                                                   --+---

Total Oo$e= 1.24E-01 4.29E-02 5.26E-02 1.75E-02 1.77E-01 6.03E-02

                                                                 ~-1_,_:% _J_,,~ _ L_~;:'_*__

Dose Limit 1 1 1 per field=

    %ofDose 12.4%              4.3%              5.3° limit                                                   -

Table24 Past Spreading & Stored Soil Inventory (No Future Additions) Intruder Dose at End of Plant License in 2013

                                                                                   ---~*---*-*     -***** **--* ****- -**-- --- --*-

Isotope Max Organ WholeBody Max Organ Whol e Body Max Organ Whole Body Existing Existing AH Past All Past Material from Material from Stored stored Material to be Malerial to be Spreading Spreading Past Past Spread Sp read Plus Stored Plus Stored Spreading Spreading Inventory Inventory Mrem/year Mrem/year M_r~ear __ M remlyea_r-11---M_re_~e_a_r_+-_M_~m/ye~ ... Mn-54 1.04E-06 3.18E-07 1.47E-06 4.48E-07 2.50E-06 7.66E-07 Co-60 1.48E-01 4.21E-02 2.64E-03 7.52E-04 1.51E-01 4.29E-02 Zn-85 1.34E-06 8.85E-07 9.58E-07 6.33E-07 2.30E-06 1.52E-06 Cs-134 3.07E-07 2.37E-07 O.OOE+OO 0. OOE+oo 3.07E-07 2.37E-07 Cs-137 2.14E-01 1.18E-01 9.14E-02 5. 04E-02 3.05E-01 1.SBE-01 Ce-141 4.36E-44 1.24E-45 O.OOE+-00

                                 -- --*--*---*---- --------                 a. OOE+-00 1------*

4.36E-44 1.25E-45 Total Dose= 3.62E-01 1.60E-01 9.41E-02 5.1 2E-02 4.56E-01 2.11E-01 Dose Limit per 5 5 5 5 5 5 field=

    %ofDose             7.2%                                                    1 0%                                    4.2%

3.2% 1.9% 9.1% limit --- ~. .. --**~--- 31 Appendix J Original Off-Site Dose Calculation Manual Page 34 of57 Vermont Yankee Nuclear Power Station

(]) 0 s Table 25 g Dose Impact from Spreading of Current Inventory of Stored Material Only" s:>l

i Total Stored Waste

~                                                                                  Max Organ          Whole Body          Max Organ         'Mlole Body     % Contribution % Contnbulion Max

(]) Activity Max organ (]) organ z~ fsotope As of 06/01 /04.. Control (06/01/04) Control (06/01/04) Intruder (2013) Intruder (2013) by isotope byi&0tope (") fP" (uCl/acre) mremlyear mremlyear mrem/year mrem/yr Intruder Dose Control Period Dose ....s:>l "'t:I 0 ~ Mn-54 2.13E-01 7.99E-05 4.11 E--05 1.47E..Q6 4.48E-07 (]) 0.00% 0.032% .... Co-60 2.70E-01 1.94E-04 1.43E-04 2.64E.Q3 u:i 7.52E-04 0.25% o.on% Zn-65 g Cs-134 5.79E-01 9.50E-03 5.96E.Q3 9.58E-07 6.33E-07 0.00% 3.760% c;* O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO 0.0000% 0.0000%

i Cs-137 1.61E+01 4.28E-02 1.13E-02 9.14E-02 5.04E-02 8.7% 17.0%

                                                                                                                                                                                                  "-:l i-/           Ce-141              O.OOE+OO           0.00E+oo             O.OOE+OO           O.OOE+OO          0.00E+OO                                             "-:l 0.0000%          0.0000%           m Ce-144 i

1 O.OOE+OO u-1 Total Dose= O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO 0.0000% z 5.26E-02 1.75E-02 0

                               --""*                           Dose Limit a             1                   1 9.41E.Q2 5

5.12E-02 5 0'%6.9% 20.8%

                                                             %ofOoseUmlt              5.3%                1.7%                1.9%              1.0%                                             '-
                                                                                                                                                                                               .0 n

Includes only Stored soil/sand materials collected as of end of 2003 (See Table 1)

                                                                                                                                                                                                 ~
                                     ** Total Activity values from Table 12.                                                                                                                     ;:l t::

r.

        '"O
        ~

0

l{'O

                      >-                                                                                                                                                                         c...
        ~        I   'O
        "       u:i (])

w -* ::i Vl 0 e: 0 0 ;.<

        ....., 0      .......

Ul rJ>

        -...) (])

0 n ciQ. s:>l: :-i* (")

e-s:>l c;*

i s;:: 32 s:>l

i

               ~
               ~

APPENDIX J (Continued) Tobie 26 Past, Stored Materials and Projected Future Disposal (all septagelsilt and soils) Doses at End of Plant License* Total Waste Max. Organ VVhote Body- Max. Organ-- "°Whole BodY --***3 ***-- % Conlributfon-Activity Contribution Max organ Max organ Isotope ln2013- Control Control Intruder Intruder by isotope by isotope (uCi/aae) mremlyear mremtyear mrem/year mrem/yT Intruder Dose Control Period Dose Mn-54 1.85E-01 6.94E-05 3.57E-05 1.89E-03 5.nE-04 0.18% 0.027% Co-60 1.33E+01 9.54E-03 7.00E-03 4.24E-01 1.21E-01 40.30% 3.776% Zn-65 4.58E-01 7.51E-03 4.72E-03 8.66E-03 5.73E-03 0.82% 2.974% Cs-134 1.21E-02 3.85E-05 1.55E-05 1.48E-04 1.13E-04 0.0139% 0.0152% Cs-137 8.85E+01 2.35E-01 6.21E-02 6.18E-01 3.41E-01 58.7% 93.2% Ca-141 6.79E-03 1.05E-06 1.02E-07 8.22E-05 2.34E-06 0.0078% 0.0004% Ce-144 O.OOE+-00 O.OOE+OO O.OOE-t-00 O.OOE+oo O.OOE+OO 0.0000% 0.0000%

                         -----*- -------------                         *------ -          ~------r--*--------            ---*-----

2.!J3E-U'l

                                                                                      ---4*-~_;_:_:~--1J_0_.0_"/._._c:_

Tolal Dose= 7.40E-02 1.05E**OO Dose Limit= 1 1 5

                    % of Dose Limit        25.3%           7.4%              21.1%

lnciudes an past spreadings, soil stored Between CooUng Towers (as of 11/4/03), and all annual projected additions of septage/silt/soil.

 -rotat Activity values from last column of Table 19.

33 Appendix J Original Off-Site Dose Calculation Manual Page 36 of57 Vermont Yankee Nuclear Power Station

APPENDIX J (Continued) REFERENCES {I) Vermont Yankee Off site Dose Calculation Manual (ODCM), Revision 30, including the following appendixes: (i) Appendix B, "Approval of Criteria for Disposal of Slightly Contaminated Septic Waste On-Site at Vermont Yankee" (Included NRC approval letter dated August 30, 1989, VY request for approval dated June 28, 1989 with Attachments I and II.) (ii) Appendix F, "Approval Pursuant to I OCFR20.2002 For Onsite Disposal of Cooling Tower Silt" (Included NRC approval letter dated March 4, 1996, VY Request for Approval dated August 30, 1995.) (iii) Appendix H, "Request to Amend Previous Approvals Granted Under IOCFR20.302(a) for Disposal of Contaminated Septic Waste and Cooling Tower Silt to Allow for Disposal of Contaminated Soil" dated June 23, 1999, with supplements dated January 4, 2000, and June 15, 2000. (2) USNRC Regulatory Guide 1.109, Rev.I; "Calculation of Annul Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with I 0 CFR Part 40, Appendix I," dated October 1997. 34

                                                     .J-37 Appendix J Original Off-Site Dose Calculation Manual Page 37 of57 Vermont Yankee Nuclear Power Station

APPENDIX J (Continued) Appendix A Security Fence Upgrade Soil Pile #2002-01 Soil Analysis Data Density from samples data Sample Cs137 Cs-137 LLD Mn-54 detected (uCi/gm) Volume (cc) wet weight location# detected reported (gm) (uCi/grn) (uCi/gm) 38 5.76E-08 1200 1266 3F 1.16E-07 1200 1062 4A 7.76E-08 1200 1393 48 5.62E-08 1200 1512 40 4.81E-08 1200 1483 4F 1.15E-07 1200 1137 SA 6.95E-08 1200 1390 50 1.IWE-07 1200 1194 SF 1.72E-07 5.86E-08 1200 1051 1A 3.93E-08 1200 985 18 5.97E-08 1200 1335 1C 2.92E-08 1200 1188 10 7.39E-08 1200 1355-1E 3.65E-08 1200 1686 1F 4.18E-08 1200 1294 2A 6.83E-08 1200 1240 28 8.53E-08 1200 1043 2C 4.88E-08 1200 1171 2D 4.16E-08 1200 1420 2E 3.43E-08 1200 1304 2F 4.75E-08 1200 1360 3A 8.35E-08 1200 1189 3C 9.59E-08 1200 1348 3E 5.65E-08 1200 1073 4C 4.44E-08 1200 1486 4E 9.37E~JS 1200 ~057 58 7.23E-08 1200 1428 SC 4.53E-08 1200 1551 averages 1200 1285.75 Wt* 4.00E-08 Density 1.07 gm/cc Positive Ave.= Density 1071.46 kg/rn3 Positive & LLD ave. = 7.12E-08 A-1 Appendix J Original Off-Site Dose Calculation Manual Page 38 of 57 Vermont Yankee Nuclear Power Station

APPENDIX J (Continued) Appendix 8 Security Fence Upgrade Soil Pile # 2002-02 Soil Analysis data Density from Sample Data Sample Cs137 Cs-137 LLD Zn-65 detected (uCi/gm) Volume wet location detected reported (ml) weight

          #        (uCi/gm)     (uCifgm)                                                     (gm)

A3 3.98E-08 1200 1388 A4 6.93E-08 1200 1221 AS 4.S3E-08 1200 1278 A6 4.69E-OB 1200 1249 A7 4.04E-08 1200 133S AS 7.29E-08 1200 1274 81 4.06E-08 7.73E-08 1200 1392 82 3.02E-08 1200 1404 83 4.0BE-08 1200 1313 84 3.34E-08 1200 1230 8S 3.38E-OB 1200 1241 86 2.13E-OB 1200 1299 87 4.24E-OB 1200 1297 88 3.73E-OB 1200 147S C3 3.90E-08 1200 1696 C4 4.25E-OB 1200 1329 cs 4.00E-08 1200 1318 cs 4.14E-08 1200 1230 C7 4.60E-08 1200 133S 07 3.22E-08 1200 1294 wtd 5.09E-09 average 1200 1329.9 positive ave. density 1.11 gm/cc Positive & LLD ave. = 4.18E-08 1108.25 kg/m3 B-1 Appendix J Original Off-Site Dose Calculation Manual Page 39 of57 Vermont Yankee Nuclear Power Station

APPENDIX J (Continued) Appendix C Security Fence Upgrade Soil Pile # 2002-03 Soil Analysis Data Density from samples data Sample Cs137 Cs-137 LLD Volume(ml) Wet weight location# detected reported (gm) (uCilgm) (uCllgm) 1 3.00E-08 1000 1348 2 3.46E-OB 1100 1424 3 2.58E-OB 1100 1410 4 3.72E-08 1100 1410 5 3.84E-08 1000 1034 6 4.58E-08 1200 1108 7 3.14E-IJ8 1000 1526 8 4.29E-OB 1000 1532 9 5.42E-OB 1000 1086 10 5.00E-08 1000 1289 11 5.68E-OB 1000 1185

       . 12                       4.62E-08                                 1000          1135 average         1041.67        1290.58 wt positive    3.2E-09                               density                 1.24 gm/cc aver.=

1238.96 kg/m3 Positive & LLD ave. = 4.11E-08 C-1 Appendix J Original Off-Site Dose Calculation Manual Page 40 of57 Vermont Yankee Nuclear Power Station

APPENDIX J (Continued) Appendix D 2001 Protected Area Road Sweeping Pile Soil Analysis Data Density from Sample Data Sample Cs137 Cs-137 LLD Co-60 Co-60 LLD Mn-54 Volume(ml) wet localio detected reported detected reported detected weight n# (uCi/gm) (uCi/gm) (uCi/gm) (uCi/gm) (uCilgm) (gm) 1 6.55E-08 5.36E-08 1000 1435 2 3.43E-08 4.24E-08 1100 1753 3 2.91E-08 3.03E-08 1000 1550 4 4.83E-08 4.95E-08 1000 1615 5 6.37E-08 7.83E-08 1000 1657 6 3.67E-08 4.10E-08 1000 1609 7 1.96E-08 3.13E-08 1000 1565 8 1000 1650 9 3.11E-OB 2.26E-08 1000 1439 10 6.09E-08 6.05E-08 1005 1762 11 4.07E-08 4.63E-08 1000 1421 12 4.75E--OB 2.35E--08 1.23E--08 1000 1454 13 7.14E-08 4.30E-OB 1000 1561 14 4.04E-08 3.22E-08 1000 1659 15 6.65E--08 6.47E--08 5.92E-08 1000 1503 16 4.20E-08 3.73E-08 1000 1607 17 4.69E-08 3.13E-08 1000 1536 18 5.05E-08 6.21E-08 1000 1594 19 5.45E-08 5.59E-08 1000 1474 20 6.25E-08 3.62E-08 2.38E-08 1000 1427 wt 3.17E-08 8.72E-09 4.77E-09 average 1005.2 1563.5 positive ave.= density 1.58 gm/cc Positive & LLD 4.80E-C3 4.43E-08 1555.4 kg/m3

  ~:!_e.= *****--*

D-1 Appendix J Original Off-Site Dose Calculation Manual Page 41of57 Vermont Yankee Nuclear Power Station

r----- APPENDIX .I (Continued) Appendix E 2001 HWC Soil Excavations Soil Analvsls Data Densitv from Samole data Cs137 Cs-137 LLD Zn-65 wet Sample detected reported detected Volume weight location# (uCi/gm) (uCi/gm) (uCi/gm) (ml) (gm) 1 4.09E-08 1000 1099 2 6.11E-08 1000 1043 3 6.04E-08 1000 1007 4 4.78E-08 1000 1008 5 5.30E-08 1000 1103 6 5.77E-08 1000 1025 7 5.94E-08 ; 1000 956 8 4.75E-08 1000 957 9 5.0SE--08 1000 976 10 6.16E-08 1000 1050 11 5.54E-08 1000 1043 12 6.50E-08 1000 1126 13 3.44E-08 1000 1141 14 5.67E-08 1000 997 15 5.26E-08 1000 1105 16 5.BOE-08 1000 1190 17 5.SBE-08 na 1000 1040 18 5.29E-08 1000 1141 19 5.85E-08 1000 1050 20 5.20E-08 1000 1073 wt positive ave.= 1.95E-08 average= 1000 1056.5 Positive & LLD ave.= 5.41E-08 density= 1.0565 gm/cc 1056.5 kg/m3 E-1 Appendix J Original Off-Site Dose Calculation Manual Page 42 of57 Vermont Yankee Nuclear Power Station

APPENDIX J (Continued) Appendix F 1996 Soil Remnants Analysis -Security Fence Upgrade llle attached report ("Radioactivity Analyses for Soil Piles SIOred Between Cooling Towers", REG-115/96, dated July 15, 1996) indicates that two soil piles totaling 4000 ft1 were coliected in 1995 and stored between the Cooling Towers. Even 1 though the larger of the two piles (3100 ft ) did not indicate any detectable plant related radioactivity, the two piles were eventually combined with portions disposed of by land spreading on the South disposal plot as annual disposal volume limits for soil/sand mixes permitted. The remnants of the piles currently contain about 970 ft 1 of material. For the estimated concenlration ofCs-137 in the combined piles, concentration values from Tables I and 3 were averaged in proportion to there volumes as shown: Cs-137: 900 ft' at an average concentration of328 pCi/kg 3100 ft with no detectable activity {average absolute value ~ I I pCi/kg) 1 Average concentration (weighted ave.) = 900 ft1

328oCi/kg + 3100 ft' '" 11 oCi/kg 900 ft1 + 3100 ft3 82.33 pCi/kg (dry)

Therefore, on a wet weight bosis applicable to the measured volume of the remaining pile of soil between the Cooling Towers, the weighted average concentration for Cs-137 is: 82.33 pCi/kg

I kg/ I 000 g
I E-06 uCi/pCi
I. 12 wet/dry volume ratio 9.22 E--08 uCi/gm (wet) for Cs-137 ForCo-60; Average concentration (weighted ave.)= 3 900 ft
85 oCi/kg + 3100 ft1 .. 0.034 oCi/kg 900 ft1 + 3 IOO ft' 19.15 pCi/kg On a wet weight basis applicable to the measured volume of the remaining pile'ofsoil between the Cooling Towers, the weighted average concentration for Co-60 is:

n.~ pCi/'1.g

I kg/I 000 g
i E--06 uCi/pCi
1.12 wet/dry volume ratio 2.14 E--08 uCi/gm (wet) for Co-60.

F-1 Appendix J Original Off-Site Dose Calculation Manual Page 43 of57 Vermont Yankee Nuclear Power Station

l APPENDIX J (Continued) MEMORANDUM YANKEE ATOMIC- BOLTON To 0. D. Weyman Date July 15, 1996 Group# REG-I 15196 From M. S. Strum W.0.# - - - - - - - Subject Radioactivily Analyses fur Soil Piles Stored Between d1C Cooling l.M.S..# TO\Vflr.il File# -vys--oi-lac-.doc---- REFERENCES (I) Envircnnental Lalxnatmy Analyais Rqxnu Sample Numbms 022686 throuah 02273S, soil - Fence and Rl:p&vin&- Rdamce Date 4113195. QACKGBOUND Site ~ conruuctioa activities ha"" gcnerated ITIU piles of soil fitlm the prolncllod lllflll thlrt wo:rc: placed be<wccn tho plana's cooling,_ pending r&diologiai.I asscssmcmt and final disposal dlspasitjon. One pile wu estimated at 3 I 00 cubic fCel and was initially matk.ed as havillg come ftOIU security fmice c><cavation. The sc.;a>d pile, a1tim111e0 at aliuu* 900 culnc feet. wu placed dir1'd1y scut!a of tho lim pile and cast oflbc 14,000 cubic fi:ct of cooling tower 1ilt also llton:d ~the towers. This *cc:ond pile ....... initiaJly- design.ued as coming from repaving aalviti:is. (Noll: thal dJ<:SO designations may have "-1 ~rsed with the n:pavma activity genenlting the 3100 cubic feet of material end lhe security knee c:u:avation g=a1ing the 900 cubic fi:et of material.) DISCQSSION Al ...., discussed last \\...ck. I'm fonvantiog copies of the labon.tllfy analyses (Rcfcn:nc:c I) fur two piles of dirt cum:ntly loa11ed immediately cast oflhe cooling tOM:r lilt pile °bet'Mell tho plant's cooling Tbc 3.1.IXl cubic foot pile wu S1D11Jlcd by collecting 30 COD1p<lSite p sanq>lcs (G226861fuouah 022715) taki:n at equal dimnca looa it& 82 fooc length (the pile is about 15.5 Rd wide and 4 feet high at its peak,). Each compoaito sample is consist of 3 ,pall ialiqi.'O!S talccn cm the left. top; aud rjpt side of the ptlc 111 cacb rctcrence distance starting ftam die pile's north c:nd. Too IXllDIDl:llt field oo each ana.lyiiis report indicates a ample location relalive to the north end oftbo pile. M an c:uinplo, SlUllPle 022691 baa a comment of6-15.0, indicating the 6rh sample taken ata dismnoe of 15.0 feef&c:m the aartb end of the pile. T&bles I and 2 summariz.o the results oftbo gamma isotopic analyses f<< Cs-137 ....i Co-60. Nmc of the 30 camposite sample. in:licaled any positive Co:sium or Ccbalt, or any olbcr plane ralated radionuclide. .iu a QQllSCqUCDCC, tl.U pilo appcan to be free ftom any radioactivity coalaminatkn. Tbo 900 cubic foo& pile was sampled in the same manner as above, with a toCal of'20 camposite samples collccfcd (022716 duough 022735). Taine J shows lfW bodi Cs-137 and Co-60 \\'D!I delected in all ex most oflbe grab ftllllilm, indicating that positive plaal related radioactivity exists in tbt. IOil, and diai I OCFR20.2002 appa-aval fbr dlspoa1 will bo needed. F-2

                                                                             .J-~cf Appendix J Original Off-Site Dose Calculation Manual Page 44 of57 Vermont Yankee Nuclear Power Station.*

APPENDIX J (Continued) R.EG-1 ISl96. July IS. 1996 Pago2 Please call a& your convenience to discua the ocxs neps necessary to handle du: 900 i:ubic fi>ot pile. Marl: S. Strum lea.cf Radiological Engineer Environmartal Engineering Depart. Arrachmcnts c R. Marcc.llo P. Llulcfield M. Mari.v> F-3 Appendix J Original Off-Site Dose Calculation Manual Page 45 of 57 Vermont Yankee Nuclear Power Station

APPENDIX J (Continued) Tabfo 1 Vennont Yankee Soil Analysis E-Lab soil data rrx 3100 ft3 soil marked""

                             *Fence* Hne construction m11terial Cs-137 (pCilkg d<yl Sample     Cone.        +*sigma     3 sigma     MOC 1.0.

1-1.4 9 21 63 78 2-4.1 33 22 66 *14 3-6.8 33 26 78 89 4-9.6 -9 23 69 91 5-12.3 20 21 63 73 6-16.0 16 19 67 69 7.17.8 -41 26 76 110 8.20.5 -6 21 63 86 9-23.2 *6 33 99 120 10-26.0 17 23 69 80 11-28.7 11 16 48 60 12-31.4 -14 21 63 82 13-34.2 -12 21 63 83 14-36.9 22 18 64 69 lii-39.G -12 2::! ao 87 16-42.4 15 18 64 64 17-46.1 12 28 84 99 18-47.8 17 21 63 77 19*60.6 -33 20 60 84 20-63.3 -42 23 69 94 21-66.0 26 36 108 130 22-68.8 26 27 81 93 23-61.6 30 26 76 86 24-64.2 18 25 75 89 25-67.0 28 26 78 89 26-69.7 46 23 69 73 27-72.4 33 23 69 76 28-76.2 48 23 69 70 29-77.9 37 26 75 84 30-80.8 9 29 87 100 Average: 11 23 70 86 Mex. vak.Ja: 48 36 108 130 Min. value: -42 16 48 69 F-4 Appendix J Original Off-Site Dose Calculation Manual Page 46 of57 Vermont Yankee Nuclear Power Station

APPENDIX J (Continued) Table 2 Vermon' Yanke .. Soil Analysis E-Lab soil data for 3100 ft3 soil ma.-ked as

                             *Fence" line construction matoria*

Co-60 fpCilkg dry) Sample Cunc~ f - sigma 3 mg,n. MOC 1.0. 1-1.4 10 18 64 70 2-4. 1 -12 23 69 100 3-6.B

11 21 63 97 4-9.6 39 30 90 100 5-12.3 1 21 63 86 6-16.0 32 18 64 57 7.17.8 -50 25 75 120 8.20.5 16 15 45 56 9-23.2 43 28 84 91 10-26.0 -17 23 69 97 11-28.7 -13 21 63 93 12-31.4 -4 20 60 84 j 3-34.2 -2 17 51 75 14-36.9 4 26 75 98 16-39.6 -16 24 72 100 16-42.4 -39 28 84 120 17-46.1 -17 23 69 100 18-47.8 -14 29 87 130 19-60.6 -9 18 64 82 20-53.3 37 22 66 72 21-66.0 17 27 81 140 22-68.B 5 30 90 120 23-61.6 7 28 84 110 24-64.2 -10 19 57 90 26-67.0 27 27 81 98 26-69.7 1 29 87 120 27-72.4 16 21 63 79 28-75.2 6 26 75 98 29-77.9 -8 32 96 130 30-80.6 -27 23 69 110 Average: 0 24 71 97 Max. value: 43 32 96 140 Min. value: -60 16 45 56 F-5 Appendix J Original Off-Site Dose Calculation Manual Page 47 of 57 Vermont Yankee Nuclear Power Station

§i a

 ~                                                                                 T*bl113

s V11rmont Yank1111 Soil Analysis Cl>

('1) E*Lab Analysis of soil from 900 ft3 pll11 Initial m11rkad "repaving dirt" zs:: Cs-137 Co-60 () Hmple lpCl/kg dry! Positive Act. ii'" 1.0. lpCi/\cg dryl Pogitive Act Cone. +* slgm* 3 sigm* >3 sigma

"'""d.....                                                                                             Cone.       +*sigma       3 sigma  > 3 eigm*

0 1-1 .1 234 63 159 positive 49 38 114 NO

('1) 2*3.3 622 67 171 posltlv* 143 29 87 positive

 .....                                 3*5.5        337          43        129      positive                 37             21         63      NO

(/J 4-7.7 291 29 87 positive 111 17 51 positive e 5* 6-9.9 6-12.1 348 136 61 26 153 78 positive positive 47 73 29 23 87 69 NO positive >

s 7*14.3 107 24 72 positive 8-Hl.5 82 15 45 positive .....,

222 44 132 positive 140 28 84 positive [Tl 9*18.7 10-20.9 180 269 37 61 111 153 positive potoltlve 92 21 63 positive z y 11*2.0 12*4.2 810 378 51 38 163 114 positive positive 1, 8 114 106 31 21 21 93 63 63 posltlvo positive posltlve Cl

-+

CJ:. 13-8.4 14-8.6 15-10.8 16-13 810 376 331 263 66 24 22 33 198 72 66 99 po*ltive po11ltlv11 positive positive 124 62 87 5 27 13 12 22 81 39 36 66 positive positive positive NO n 0 17-16.2 160 12 36 positive 67 9 27 posltiv* ::::. 18*17.4 247 30 90 positive 105 17 51 positive :::: 19-19.6 328 56 165 poaltlve 54 40 s:: 120 NO (')

            ""CO>
            "' ....., 'O (IQ     'i" 'O 20-21.8        236          32         96      po1ltlve                100             23         69   poaitiv*
                                                                                                                                                       -0..

('1) (/J ('1)

              " -* ::s
            ..... -     Q.

Av11rag11: 328 39 85 23 00 (1) -* Max. v1lua: 810 66 143 40 0 0;..:

            ....., 0 ~              Min. value:     107          12                                            5              9 VI"'
            -.I    ('1)

Q (')o'Q"

                   "'-::s"'

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                   §: -

5*

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g

e. F..Q

APPENDIX J (Continued) Appendix G Record for Last Spreadings (2003) on South Disposal Field G-1 Appendix J Original Off-Site Dose Calculation Manual Page 49 of57 Vermont Yankee Nuclear Power Station

APPENDIX J (Continued) Total Recorded Spreading Dota for 2003 Serial# Spreading Media Mn-54 Co-60 Zn-()5 Cs-134 Cs-137 Cc-141 Cc-144 Date Type (uCi/acre) (uCilacre) (uCi/acre) (uCilacrc) (uCi/acre) (uCilacre) (uCi/acrc) 2003-01 7-1-3 Septic - I.OJ - - - - - 2003--02 10-25-03 Septic - 0.12 - - - - - 2003-03 11-4-03 Sand/soil - - - - 1.34 - - 2003-04 11-4-03 CT Silt - 0.256 -- - - - -- Total - 1.41 -- - 1.34 - -- G-2

                                                        .:[-6C Appendix J Original Off-Site Dose Calculation Manual Page 50 of57 Vermont Yankee Nuclear Power Station

APPENDIX J (Continued) UNITED STATES NUCLEAR REGU~ATORY COMMISSION WASHINGTON, D.C. 20555.0001 July 19, 2005

                                                    /V'A/ oS - o 'Io Mr. Michael Kansler President Entergy Nuclear Operations, Inc.

440 Hamilton Avenue White Plains, NY 10601

SUBJECT:

SAFETY EVALUATION OF REQUEST TO AMEND PREVIOUS APPROVALS GRANTED PURSUANT TO 10 CFR 20.2002 - VERMONT YANKEE NUCLEAR POWER STATION (TAC NO. MC5104) Dclar Mr. Kansler: By letter dated October 4, 2004, as supplemented on January 17, 2005, Entergy Nuclear Operations, Inc. (Entergy) submitted a request to the Nuclear Regulatory Commission (NRC) to modify previous approvals granted pursuant to Title 10 of the Code of Federal Regulations (10 CFR) Section 20.2002 (previously 10 CFR 20.302(a)), for on-site disposal of slightly contaminated material at Vermont Yankee Nuclear Power Station. Specifically, Entergy requested an increase of the current approved annual volume limit of 28.3 cubic meters of soil/sand to a new annual_ volume limit of 150 cubic meters of soil/sand. In addition, Entergy has requested a one-time approval for on-site disposal of the current backlog inventory of approximately 528 cubic meters of soil/sand. The NRC staff has completed it~ review of the request and has determined that the proposed changes are acceptable as documented in the enclosed Safety Evaluation. Pursuant to the provisions of *10 CFR Part 51, the NRC has published an Environmental Assessment and Finding of No Significant Impact In the Federal Register on July 19, 2005 (70 FR 41440). Sincerely, Richard B. Ennis, Senior Project Manager, Section 2 Project Directorate I Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket No. 50-271

Enclosure:

As stated cc w/encl: See next page Appendix J Original Off-Site Dose Calculation Manual Page 51 of57 Vermont Yankee Nuclear Power Station

APPENDIX J (Continued) Vermont Yankee Nuclear Power Station cc: Regional Administrator, Region I Ms. Carla A. White, RAPT, CHP U.S. Nuclear Regulatory Commission Radiological Health 475 Allendale Road Vermont Department of Health King of Prussia, PA 19406-1415 P .0. Box 70, Drawer #43 108 Cherry Street Mr. David R. Lewis Burlington, VT 05402-0070 Pillsbury, Winthrop, Shaw, Pittman, LLP 2300 N Street, N.W. Mr. James M. DeVincentis Washington, DC 20037-1128 Manager, Licensing Vermont Yankee Nuclear Power Station Ms. Christine S. Salembier, Commissioner P .0. Box 0500 Vermont Department of Public Service 185 Old Ferry Road 11 2 State Street Brattleboro, VT 05302-0500 Montpelier, VT 05620-2601 Resident Inspector Mr. Michael H. Dworkin, Chairman Vermont Yankee Nuclear Power Station Public Service Board U.S. Nuclear Regulatory Commission State of Vermont P.O. Box 176 112 State Street Vernon, VT 05354 Montpelier, VT 05620-2701 Director, Massachusetts Emergency Chairman, Board of Selectmen Management Agency Town of Vernon ATIN: James Muckerheide P.O. Box 116 400 Worcester Rd. Vernon, VT 05354-0116 Framingham, MA 01702-5399 Operating Experience Coordinator Jonathan M. Block, Esq. Vermont Yankee Nuclear Power Station Main Street 320 Governor Hunt Road P.O. Box566 Vernon, VT 05354 Putney, VT 05346-0566 G. Dana Bisbee~ Esq. Mr. John F. McCann Deputy Attorney General Director, Nuclear Safety Assurance 33 Capitol Street Entergy Nuclear Operations, Inc. Concord, NH 03301-6937 440 Hamilton Avenue White Plains, NY 10601 Chief, Safety Unit Office of the Attorney General Mr. Gary J. Taylor One Ashburton Place, 19th Floor Chief Executive Officer Boston, MA 02108 Entergy Operations 1340 Echelon Parkway Ms. Deborah B. Katz Jackson, MS 39213 Box83 Shelburne Falls, MA 01370 Appendix J Original Off-Site Dose Calculation Manual Page 52 of57 Vermont Yankee Nuclear Power Station

APPENDIX J (Continued) Vermont Yankee Nuclear Power Station cc: Mr. John T. Herron Mr. Ronald Toole Sr. VP and Chief Operating Officer 1282 Valley of Lakes Entergy Nuclear Operations, Inc. Box R-10 440 Hamilton Avenue Hazelton, PA 18202 White Plains, NY 10601 Ms. Stacey M. Lousteau Mr. Danny L. Pace Treasury Department Vice President, Engineering Entergy Services, Inc. Entergy Nuclear Operations, Inc. 639 Loyola Avenue 440 Hamilton Avenue New Orleans, LA 70113 White Plains, NY 10601 Mr. Raymond Shadis Mr. Brian O'Grady New England Coalition Vice President, Operations Support Post Office Box 98 Entergy Nuclear Operations, Inc. Edgecomb, ME 04556 440 Hamilton Avenue White Plains, NY 10601 Mr. James P. Matteau Executive Director Mr. Michael J. Colomb Windham Regional Commission Director of Oversight 139 Main Street, Suite 505 Entergy Nuclear Operations, Inc. Brattleboro, VT 05301 440 Hamilton Avenue White Plains, NY 10601 Mr. William K. Sherman Vermont Department of Public Service Mr. John M. Fulton 112 State Street Assistant General Counsel Drawer20 Entergy Nuclear Operations,* Inc. Montpelier, VT 05620-2601 440 Hamilton Avenue White Plains, NY 10601 Mr. Jay K. Thayer Site Vice President Entergy Nuclear Operations, Inc. Vermont Yankee Nuclear Power Station P.O. Box 0500 185 Old Ferry Road Brattleboro, VT 05302-0500 Mr. Kenneth L. Graesser 38832 N. Ashley Drive Lake Villa, IL 60046 Mr. James Sniezek 5486 Nithsdale Drive Salisbury, MD 21801 Appendix J Original Off-Site Dose Calculation Manual Page 53 of57 Vermont Yankee Nuclear Power Station

APPENDIX J (Continued) UNITED STATES NUCLEAR REGU~ATORY COMMISSION WASHINGTON, D.C. 20555--0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION ENTERGY NUCLEAR OPERATIONS. INC. VERMONT YANKEE NUCLEAR POWER STATION DOCKET NO. 50-271

1.0 INTRODUCTION

By letter dated October 4, 2004, as supplemented on January 17, 2005, Entergy Nuclear Operations, Inc. (Entergy or the licensee) submitted a request to the Nuclear Regulatory Commission (NRC) to modify previous approvals granted pursuant to Title 10 of the Code of Federal Regulations (1 O CFR) Section 20.2002 (previously 10 CFR 20.302(a)), for on-site disposal of slightly contaminated material at Vermont Yankee Nuclear Power Station (VYNPS). Specifically, Entergy requested an increase of the current approved annual volume limit of 28.3 cubic meters of soiVsand to a new annual volume limit of 150 cubic meters of soiVsand. In addition, Entergy has requested a one-time approval for on-site disposal of the current backlog inventory of approximately 528 cubic meters of soil/sand.

2.0 REGULATORY EVALUATION

As described in 10 GFR 20.2002, "Method for obtaining approval of proposed disposal procedures," licensees are required to obtain NRG approval of proposed procedures, not otherwise authorized in the regulations, to dispose of licensed material generated in the licensee's activities. Previous NRG approval for VYNPS on-site disposal of various slightly contaminated waste materials is documented in letters dated August 30, 1989, March 4, 1996, June 18, 1997, June 15, 2000, and June 26, 2001. Based on these previous approvals, the licensee is currently authorized to dispose, in designated on-site areas, the following materials: (1) septic waste; (2) cooling tower silt; (3) soiVsand generated from the annual winter spreading on roads and walkways; and (4) soil resulting from on-site construction-related activi1ies. Disposal of septic waste and cooling tower silt material ls not limited by an annual volume, but by a total dose impact related to the radiological content of the material and the concentration of radioactivity contained within it. The combination of the soiVsand generated from the annual winter spreading on roads and walkways and the soil resulting from on-site construction-related activities is currently subject to an annual volume limit of 28.3 cubic meters. The licensee's application dated October 4, 2004, proposed to increase this annual volume limit for the same materials (i.e., soiVsand) to 150 cubic meters. In addition, the application proposed a one-time disposal of the current backlog inventory of approximately 528 cubic meters of soil/sand. The current restrictions on the annual volume of slightly contaminated soil/sand that can be disposed on-site coupled with several plant facility projects in recent years, has resulted in the Appendix J Original Off-Site Dose Calculation Manual Page 54 of57 Vermont Yankee Nuclear Power Station

APPENDIX J (Continued) accumulation of a backlog of low-level contaminated earthen material that is awaiting disposal by land spreading on previously-approved on-site disposal areas. The current approved annual volume limit of 28.3 cubic meters of soiVsand for disposal was based on licensee estimates of soil and sand collected from road and walkway sweepings inside the Protected Area following each year's winter cleanup (i.e., current annual limit does not account for future site excavation and construction activities).

3.0 TECHNICAL EVALUATION

The licensee proposes to dispose of the current backlog inventory of soil/sand and the future soil/sand material using a land spreading technique consistent with the current commitments for on-site disposal of slightly contaminated material previously approved by the NRC. The licensee will continue to use designated areas of its property approved for this waste material. Determination of the radiological dose impact of the new material has been made based on the same dose assessment models and pathway assumptions used in the previously-approved applications. The licensee will procedurally control and maintain records of all disposals. The following information will be recorded:

1. the radionuclide concentrations detected in the material;

2. the total volume of material disposed;

3. the total radioactivity in the disposal operation as well as the total radioactivity accumulated on each disposal plot at the time of spreading;

4. the plot of land on which the material was applied; and

5. dose calculations or maximum allowable accumulated activity determinations required to demonstrate that the dose values have not been exceeded.

The licensee's application states that the existing NRG-approved bounding dose conditions for the proposed on-site disposals will continue to be applied without change. The bounding dose conditions for the on-site disposals are as follows:

1. the annual dose to the whole body or any organ of a hypothetical maximally exposed individual will be less than 1.0 millirem (mrem) (during the period the licensee has active control over the disposal sites, i.e., during the current operating license period);

2. annual doses to the whole body and any organ of an inadvertent intruder from the probable pathways of exposure will be less than 5 mrem (following the period the licensee has active control over the disposal sites); and

3. disposal operations will be at an approved on-site location.

To ensure that the addition of new waste material will not exceed the bounqing dose conditions for each new spreading operation, the licensee's total radioactivity and dose calculation will include all past disposals of septic waste, cooling tower silt, soil and soil/sand material on the Appendix J Original Off-Site Dose Calculation Manual Page 55 of57 Vermont Yankee Nuclear Power Station \

APPENDIX J (Continued) designated disposal plots. In addition, concentration limits will be applied to the disposed material to restrict the placement of small volumes of material that may have relatively high radioactivity concentrations. VYNPS is currently authorized to dispose of licensed material, pursuant to 10 CFR 20.2002, in two designated locations both within the site boundary security fence. The South field (approximately 1.9 acres in size} is centered approximately 1500 feet south of the reactor building. The North field (approximately 10 acres in size) is centered approximately 2000 feet northwest of the reactor building. The South field has been the only field utilized for disposal of licensed material to date. The licensee's application states that it is anticipated that future disposal operations will also use the South field since sufficient margin in comparison to the approved dose limit criteria still exists for anticipated waste disposal of the existing backlog of material now in storage, plus all expected future disposals of material, assuming the same observed generation rates. The licensee performed an evaluation of the radiological impact of all past, accumulated storage inventory, dnd proj&cted future wasta spreaaing operations on the South field. The licensee assessed the dose that may be received by the maximally exposed individual during the period of plant control over the property, and to an inadvertent intruder after plant access control ends using the same pathway modeling, assumptions, and dose calculation methods that were previously approved by the NRC tor the waste material. The dose models are based on the guidance in Regulatory Guide 1.109, Revision 1. Table 26 in the attachment to the licensee's October 4, 2004, application provides the calculated maximum organ and whole-body doses, at the end of the current plant license period, based on the combination of all past disposal of waste materials, disposal of the current backlog inventory oi waste material and projected annual disposal of waste materials. The results are summarized as follows: Individual Individual Intruder Intruder Organ Dose Whole Body Dose Organ Dose Whole Body Dose {mrem/;*ear) (mrs rrJyear) (r:;rem/yea;) (mrem/year) Total Dose 0.253 0.074 1.05 0.468 Dose Limit 1 1 5 5

  % of Dose Limit           25.3%                7.4%                21.1%               9.4%

Based on the above calculated dose rates, the NRC staff concludes that the proposed increase of the current approved annual volume limit of 28.3 cubic meters of soil/sand to a new annual volume limit of 150 cubic meters of soil/sand and a one-time approval for on-site disposal of the current backlog inventory of approximately 528 cubic meters of soiVsand would result in dose rates within the bounding dose conditions for on-site disposal previously approved by the NRC. Appendix J Original Off-Site Dose Calculation Manual Page 56 of57 Vermont Yankee Nuclear Power Station

APPENDIX J (Continued)

4.0 CONCLUSION

The NRG staff finds the licensee's request for dispo.sal of a new annual volume limit of 150 cubic meters of soil/sand and a one-time approval for on-site disposal of the current backlog inventory of approximately 528 cubic meters of soiVsand, pursuant to 10 CFR 20.2002, In the same manner, location, and within the bounding dose conditions as previously approved by the NRC, to be acceptable. Principal Contributors: S. Klementowicz A. Ennis Date: July 19, 2005 1--f7 Appendix J Original Off-Site Dose Calculation Manual Page 57 of57 Vermont Yankee Nuclear Power Station

j i APPENDIX I RADIOACTIVE LIQUID, CiASEOUS, AND SOLID WASTE TREATMENT SYSTEMS Requirement: ODCM Se~lion 10.5 requires that licensee initiated major changes to the radioactive waste systems (liquid, gaseous, and solid) he reported to the Commission in the annual Radioactive Efl1uent Release Report for the period in which the evaluation was reviewed by the Plant Operation Review Committee. Response: There were no licensee-initiated major changes to the radioactive waste systems during this reporting period. 1-1

APPENDIX J ON-SITE DISPOSAL OF SEPTIC/SILT/SOIL WASTE Requirement: Off-Site Dose Cakulation Manual, Appendices B, F and I require that the dose impact dut to on-site disposal of septic waste, cooling tower silt, and sand/soil type materials during the reporting year and from previous years be reported to the Nuclear Regulatory Commission in the Annual Radioactive Effluent Release Report if disposals occur during the reporting year. Entergy Nuclear Vermont Yankee will report in the Annual Radioactive Enluent Release Report a list of the radionuclides present and the total radioactivity associated with the disposal activities on the Vermont Yankee site.

Response: There were no on-site disposal spreadings during 2016. The total radioactivity spread on the I. 9 acres (southern) 011-~itc Jis1H>sal ficll! li*.im pre* ious ycai s was as follows:

Activity from All Past Disposals Decayed to 1/1/2016 Radionuclide (Cil Mn-54 9.92E-09 Co-60 9.42E-06 Zn-65 2.98E-09 Cs-134 6.25E-IO Cs-137 7.47E-05 The maximum organ dose from all past and current spreading operations totaled l .08E-O 1 mrem/year. This calculated value is within the 1 mrem/year limit applied during the period of operational control of the site. The projected hypothetical "intruder" dose for the period following the loss of operational control of the site area, due to all spreading operations to-date, is 2.24E-01 mrem/year versus a 5 mrem/year dose limit. The "intruder dose" period begins on the date that the plant operating license is projected to expire, January 15\ 2030. J-1}}

BVY 17-017, Submittal of 2016 Radioactive Effluent Release Report

Text

SUBJECT:

Enclosure:

1.0 INTRODUCTION

1.0 INTRODUCTION

Reference:

SUMMARY

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1.0 INTRODUCTION

1.0 INTRODUCTION

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Dear Mr. Tremblay:

SUBJECT:

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Dear Mr. Reid:

Enclosure:

1.0 INTRODUCTION

2.0 BACKGROUND

4.0 CONCLUSION

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Subject:

1.0 INTRODUCTION

6.0 CONCLUSION

7.0 REFERENCES

1.0 INTRODUCTION

6.0 CONCLUSION

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SUMMARY

6.0 CONCLUSION

7.0 REFERENCES

6.0 CONCLUSION

7.0 REFERENCES

References:

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6.0 CONCLUSION

7.0 REFERENCES

6.0 CONCLUSION

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Dear Mr. Newton:

Enclosure:

1.0 INTRODUCTION

3.0 CONCLUSION

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Subject:

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Dear Mr. Balduzzi:

Enclosure:

1.0 INTRODUCTION

3.0 CONCLUSION

References:

Subject:

SUMMARY

SUBJECT:

Enclosure:

1.0 INTRODUCTION

2.0 REGULATORY EVALUATION

3.0 TECHNICAL EVALUATION

4.0 CONCLUSION

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