Text

Annual 2002 Radioactive Effluent Release Report
	ML031620372
Person / Time
Site:	Vermont Yankee
Issue date:	05/14/2003
From:	Devincentis J; Entergy Nuclear Operations, Entergy Nuclear Vermont Yankee
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	BVY 03-46
	Download: ML031620372 (402)
	v • d • e

Entergy Nuclear Vermont Yankee, LLC Entro y nter,~

w-OJ'~E Entergy Nuclear Operations, Inc.

185 Old Ferry Road Brattleboro, VT 05302-0500 May 14, 2003 BVY 03-46 U.S. Nuclear Regulatory Commission A'ITN: Document Control Desk Washingtol, D.C. 20555

Subject:

Vermont Yankee Nuclear Power Station License No. DPR-28 (Docket No. 50-271)

Annual 2002 Radioactive Effluent Release Report In accordance witl Vermont Yankee (VY) Technical Specification (TS) 6.6.D, attacled is a copy of the annual 2002 Radioactive EfLuent Release Report.

In addition, VY TS 6.7.B requires reporting of changes to the Off-site Dose Calculation Manual (ODCM). A summary of clanges made to the ODCM during 2002 is provided as Appendix H of the subject report and an uncontrolled copy of the current ODCM, Revision 30, is also included Onl CD ROM for your information.

We trust that tle information provided is adequate; hlowever, sliould you hiave questions or require additional information, please contact Mr. Sam A. Wender (802) 258-5650.

Sincerely, James M. DeVincentis Manager, Licensing Attachlments cc: USNRC Region I Administrator (v/o ODCM disk)

Jason C. Jang, USNRC Region I Inspector (ith ODCM disk)

USNRC Resident Inspector- VYNPS (w/o ODCM disk)

USNRC Project Manager - VYNPS (witli ODCM disk)

Vermont Department of Public Service (w/o ODCM disk)

Vermont Division of Occupational and Radiological Health (v/o ODCM disk)

Massaclusetts Metropolitan District Commission (v/o ODCM disk)

Massachusetts Department of Public HealtI (w/o ODCM disk)

A00 9

Docket No. 50-271 BVY 03-46 Attachment I Vermont Yankee Nuclear Power Station 2002 Radioactive Effluent Release Report

SUMMARY

OF VERMONT YANKEE COMMITMENTS BVY NO.: 03-46 "2002 Annual Radioactive Effluent Release Report" The followinig table identifies commitments made in this document by Vcrmont Yankee.

Any otler actions discussed in the submittal rcpresent intended or planned actions by Vermont Yankee. They are (lescribed to the NRC for te NRC's information and are not regulatory commitments. Please notify the Licensing Manager of any questions regarding thiis document or any associated commitments.

COMMITMENT COMMITTED DATE OR "OUTAGE'"

None N/A I1- _ _____ _ __

Vermont Yankee Nuclear Power Station Vernon, Vermont r ;L; '1w:

.-' _' t .A' ' . .N b _= e I P-.

2002 Radioactive Effluent Release Report

RADIOACTIVE EFFLUENT RELEASE REPORT FOR 2002 INCLUDING ANNUAL RADIOLOGICAL IMPACT ON MAN Entergy Nuclear Vermont Yankee, LLC Docket No. 50-271 License No. DPR-28 i

TABLE OF CONTENTS Pa e

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 ................................................... 3 3.3 Doses From Iodine-131, Iodine-133, Tritium, and Radionuclides in Particulate Form With Half-Lives Greater Than 8 Days ................................................... 3 3.4 Whole-Body Doses in Unrestricted Areas From Direct Radiation ..................................... 4 3.5 Doses From On-Site Disposal of Septic Waste and Cooling Tower Silt ........................ ....55 3.6 On-Site Recreational Activities ................................................... 5 REFERENCES .. 6 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 ii

LIST OF TABLES Number Title Page IA First and Second Quarters, 2002 Gaseous Effluents - Summation of All Releases 7 IA Third and Fourth Quarters, 2002 Gaseous Effluents - Summation of All Releases 8 1B First and Second Quarters, 2002 Gaseous Effluents - Elevated Releases 9 1B Third and Fourth Quarters, 2002 Gaseous Effluents - Elevated Releases 10 IC First and Second Quarters, 2002 Gaseous Effluents - Ground Level Releases 11 IC Third and Fourth Quarters, 2002 Gaseous Effluents - Ground Level Releases 12 ID Gaseous Effluents - Nonroutine Releases 13 2A Liquid Effluents - Summation of All Releases 14 2B Liquid Effluents - Nonroutine Releases 15 3 First and Second Quarters, 2002 Solid Waste and Irradiated Fuel Shipments 16 3 Third and Fourth Quarters, 2002 Solid Waste and Irradiated Fuel Shipments 18 4A Maximum Off-Site Doses/Dose Commitments to Members of the Public from 20 Liquid and Gaseous Effluents for 2002 (10CFR50, Appendix I) 4B Maximum Annual Dose Commitments from Direct External Radiation, 21 Plus Liquid and Gaseous Effluents for 2002(*) (40CFR190) 4C Receptor Locations for Vermont Yankee 22 4D Usage Factors for Various Gaseous Pathways at Vermont Yankee 23 4E Environmental Parameters for Gaseous Effluents at Vermont Yankee 24 SA to 5H Annual Summary of Lower Level Joint Frequency Distribution 26-33 6A to 6H Annual Summary of Upper Level Joint Frequency Distribution 34-41 iii

Radiological Effluent Release Report for 2002

[Including Annual Radiological Impact on Man]

1.0 INTRODUCTION

Tables 1 through 3 list the recorded radioactive liquid and gaseous effluents and solid waste 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 2002 in response to the ALARA objectives of I OCFR50, Appendix I. Also included on Table 4A is the estimate of direct dose from fixed station sources along the limiting west site boundary line. Tables 5A through 6H report the cumulative joint frequency distributions of wind speed, wind direction, and atmospheric stability for the 12-month period, January to December 2002. Radioactive effluents reported in Tables and 2 were used to determine the resulting doses for 2002.

As required by ODCM Section 10.1, dose commitments resulting from the release of radioactive materials in liquids and gases during the reporting period were estimated in accordance with the "Vermont Yankee Nuclear Power Station Off-Site Dose Calculation Manual" (ODCM).

These dose estimates were made using a "Method II" analysis as described in the ODCM. A "Method II" analysis incorporates the methodology of Regulatory Guide 1.109 (Reference 3) and actual measured meteorological data recorded during the 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. During this reporting period, no recreational activity was permitted and, therefore, there is no associated dose assessment as stated in Section 3.6.

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'*-

40CFR1 90, "Environmental Radiation Protection Standards for Nuclear Power Operations," are also required to be included in this report if the conditions indicated in ODCM 3/4.6, "Total Dose," have been exceeded during the year. Since the conditions indicated in the action statement under ODCM 3/4.6 were not entered into during the year, no additional radiation dose assessments are required. However, Table 4B does provide the combination of doses and dose commitments from plant effluents and direct radiation sources for the limiting member of the public off-site as a demonstration of compliance with the dose standards of 40CFR190.

All calculated dose estimates for this reporting period are below the dose criteria of I OCFR Part 50, Appendix I, and 40CFRI 90.

Appendices B through H indicate the status of reportable items per the requirements of ODCM Section 10.1.

I

2.0 METEOROLOGICAL DATA Meteorological data was collected during this reporting period from the site's 300-foot met 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 (94 meters) and is designed to meet the requirements of Regulatory Guide 1.23 for meteorological monitoring.

X/Q and D/Q values 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.i (Reference 1). A source depletion model as described in "Meteorology and Atomic Energy - 1968" (Reference 2) 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.

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 2002 land use census.

These locations were used in the calculation of atmospheric dispersion factors.

2

3.0 DOSE ASSESSMENT 3.1 Doses From Liquid Effluents ODCM 3/4.2.2 limits total body (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 10CFR Part 50, Appendix I. By implementing the requirements of 10CFR 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."

For periods in which liquid waste discharges actually occur, the exposure pathways that could exist are fish, direct exposure from river shoreline sedimentation, milk and meat via animal ingestion of the Connecticut River water, and meat, milk and vegetable pathways via crop irrigation with water withdrawn from the Connecticut River. The drinking water and aquatic invertebrate pathways do not exist down river of the Vermont Yankee plant.

There were no recorded liquid radwaste discharges during the report period, and therefore, no dose impact.

3.2 Doses From Noble Gases ODCM 3/4.3.2 limits the gamna 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 10CFR Part 50, Appendix I.

By implementing the requirements of 10CFR Part 50, Appendix I, 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, nearest resident in each of the sixteen principal compass directions, the point of highest off-site ground level air concentration of radioactive materials, and for each of the milk animal locations located within five miles of the plant.

3.3 Doses From Iodine-131. Iodine-133. Tritium. and Radionuclides in Particulate Form With Half-Lives Greater Than 8 Days ODCM 3/4.3.3 limits the organ dose to a member of the public from iodine-131, iodine-133, tritium and radionuclides in particulate form with half-lives greater than 8 days (hereafter called ;

iodines and particulates) in gaseous effluents released from the site to areas at and beyond the sit6`',

boundary to those specified in 10CFR Part 50, Appendix I (7.5 mrem per quarter, and 15 mrem per year). By implementing the requirements of 10CFR Part 50, Appendix I, ODCM 3/4.3.3 assures that the releases of iodines and particulates in gaseous effluents will be kept "as low as is reasonably achievable."

3

Exposure pathways that could exist as a result of the release of iodines and particulates to the atmosphere include external firrdiation from activity deposited onto the ground surface, inhalation, and ingestion of vegetables, meat and milk. Dose estimates 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 (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 animals inventoried in the site vicinity are fed stored feed throughout the entire year with only limited grazing allowed during the growing season. It has also been assumed that only 50 percent of the iodine deposited from gaseous effluent is in elemental form (12) and is available for uptake (see page -6, Reference 3). During the first and fourth quarters, the milk animals are assumed to receive-only stored feed. Usage factors for gaseous effluents are listed by age group and pathway in Tabl&1 1D.

Table 4E provides other dose model parameter assumptions used in the dose assessments.

The resultant organ doses were determined after adding the contributions from all pathways at each location. Doses were calculated for the vhole body, GI-tract, bone, liver, kidney, thyroid, lung and skin for adults, teenagers, children and infants. The maximum estimated quarterly and annual organ doses to any age group due to iodines and particulates at any of the off-site receptor locations are reported in Table 4A. These estimated organ doses are well below the 10CFR Part 50, Appendix I dose criteria of ODCM 3/4.3.3.

3.4 Whole-Body Doses in Unrestricted Areas From Direct Radiation The major source of dose at the site boundary, consisting of direct radiation and skyshine from the station, is due to N-16 decay in the Turbine Building. Because of the orientation of the Turbine Building on the site, and the shielding effects of the adjacent Reactor Building, only the seven westerly sectors (SSW to NNW) see any significant direct radiation.

A correlation method was derived that allows changes in the N-16 carryover in the main steam flow to be directly related to changes in the site boundary dose. This correlation is documented in section 6.1 1.1 (Equation 6-27a) of the ODCM. This method was used to calculate direct dose at the maximum site boundary location from radiation sources in the steam cycle.

The other fixed sources of dose, including direct radiation and skyshine, to the site boundary are from low level radioactive waste stored in the North Warehouse, the Low Level Waste Storage Pad Facility, and old turbine rotors and casings in the turbine storage facility. The annual dose is based on dose rate measurements in these three storage facilities and determined at the same most restrictive site boundary dose location as that for N-16 shine from the Turbine Building.

The estimated direct radiation dose from all major sources combined for the most limiting site 4

boundary location is listed on 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 and effluent release doses at the limiting nearest residence for the purpose of demonstrating compliance with the dose standards contained in 40CFR190. For direct radiation, no credit for actual occupancy time is taken (i.e., occupancy is equal to 100%).

3.5 Doses From On-Site Disposal of Septic Waste and Cooling Tower Silt Off-Site Dose Calculation Manual, Appendices B,F,H and I require that all applications of septic waste, contaminated soil and the cooling tower silt within the approved designated disposal areas be limited to ensure the dose to a maximally-exposed individual during the period of Vermont Yankee site control be maintained at 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 maintained at less than 5 mrem/year. The projected dose from on-site disposals of septic waste and the cooling tower silt is given in Appendix J of this report. al 3.6 On-Site Recreational Activities During 2002, no access for employees, their families and guests to the boat launching ramp located on-site just north of the intake structure was permitted. As such, no recreational activities were permitted on-site during the report period and, therefore, no associated dose impact to members of the public.

5

I:

REFERENCES

1. Regulatory Guide 1.11 1, "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.

2. Meteorology and Atomic Energy, 1968, Section 5-3.2.2, "Cloud Depletion", pg. 204. U. S.

Atomic Energy Commission, July 1968.

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

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

4. W. M. Lowder, P. D. Raft, and G. dePlanque 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.

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TABLE IA Vermont Yankee Effluent and Waste Disposal Annual Report First and Second Ouarters, 2002 Gaseous Effluents -Summation of All Releases I Unit j Quarter I

j Quarter 2

Est. Total Error, %

A. Fission and Activation Gases

1. Total release Ci 2.30 E+01 6.30 E+01 +/-23 %

2. Average release rate for period [tCi/sec 2.89 E+00 7.92 E+00

3. Percent of ODCM limit (1) % 4.58 E-02 7.66 E-02 B. Iodines

1. Total Iodine Ci 4.75 E-04 3.51 E-03 +/-18 %

2. Average release rate for period iLCi/sec 5.97 E-05 4.41 E-04

3. Percent of ODCM limit (2) % 6.68 E-01 2.90 E+00 C. Particulates

1. Particulates with T-1/2>8 days Ci 5.52 E-05 1.59 E-04 +/-18 %

2. Average release rate for period pCi/sec 6.95 E-06 2.01 E-05

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

4. Gross alpha radioactivity Ci N/A N/A D. Tritium

1. Total release Ci 2.67E+00 2.33E+00 +/-18%

2. Average release rate for period [iCi/sec 3.35 E-01 2.93 E-01

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

(I) ODCM Control 3.3.2. for the most limiting of beta air or gamma air dose.

(2) ODCM Control 3.3.3. for dose from 1-131, 1-133, Tritium, and radionuclides in particulate form.

(3) Per ODCM Control 3.3.3, dose contribution from Tritium and particulates are included with Iodine above in Part B.

7

TABLE A (Continued)

Vermont Yankee Effluent and Waste Disposal Annual Report Third and Fourth Quarters. 2002 Gaseous Effluents - Summation of All Releases Unit Quarter Quarter Est. Total J ~~~~3 4 Error, %

A. Fission and Activation Gases

1. Total release Ci 0.00 E+00 0.00 E+00 +/-23 %

2. Average release rate for period iCi/sec 0.00 E+00 0.00 E+00

3. Percent of ODCM limit (1) % 0.00 E+00 0.00 E+00 B. Iodines

1. Total Iodine Ci 4.93 E-05 4.14 E-05 +/-18 %

2. Average release rate for period jICi/sec 6.20 E-06 5.21 E-06

3. Percent of ODCM limit (2) % 1.08 E-01 8.59 E-02 C. Particulates

1. Particulates with T-1/2>8 days Ci 3.60 E-05 1.40 E-04 +/-18 %

2. Average release rate for period jiCi/sec 4.52 E-06 1.76 E-05

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

4. Gross alpha radioactivity Ci N/A N/A D. Tritium

1. Total release Ci 2.77E+00 2.74E+00 +/-15 %

2. Average release rate for period iICi/sec 3.48 E-01 3.44 E-01

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

(I) ODCM Control 3.3.2. for the most limiting of beta air or gamma air dose.

(2) ODCM Control 3.3.3. for dose from 1-131, 1-133, Tritium, and radionuclides in particulate form.

(3) Per ODCM Control 3.3.3, dose contribution from Tritium and particulates are included with Iodine above in Part B.

8

TABLE IB Vermont Yankee i 4,

Effluent and Waste Disposal Annual Report

.I First and Second Quarters, 2002 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 3.51 E-01 ND Krypton-87 Ci 2.53 E+00 3.10 E+00 Krypton-88 Ci ND ND Xenon-133 Ci 3.86 E-01 3.04 E+01 Xenon-133m Ci ND 5.50 E-01 _

Xenon-135 Ci 8.81 E+00 1.87 E+01 Xenon-135m Ci 1.09 E+01 1.02 E+01 Xenon-138 Ci ND ND Unidentified Ci ND ND Total for Period Ci 2.30 E+01 6.30 E+01 0.00E+00 0.OOE+00

2. lodines Iodine-131 Ci 3.60 E-04 2.79 E-03 Iodine-133 Ci 1.15 E-04 7.20 E-04 Iodine-135 Ci ND ND Total for Period Ci 4.75 E-04 3.51 E-03 O.OOE+00 0.OOE+00

3. Particulates Strontium-89 Ci 5.52 E-05 1.32 E-04 Strontium-90 Ci ND 1.75 E-07 Cesium-134 Ci ND ND Cesium-137 Ci ND 2.70 E-05 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 Zinc-65 Ci ND ND Total for Period ci 5.52 E-05 1.59 E-04 O.OOE+00 O.OOE+00 There were no batch mode gaseous releases for this reporting period.

N0 Not Detected at the plant stack 9

TABLE B (Continued)

Vermont Yankee Effluent and Waste Disposal Annual Report Third and Fourth Quarters, 2002 Gaseous Effluents -Elevated Releases Cotntinuous Mode Batch Mode (1)

Quarter Quarter Nuclides Released Units 3 4 3 4

. 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 0.00 E+00 0.00 E+00 0.00 E+00 0.00 E+00

2. lodines Iodine-131 Ci 4.93 E-05 4.14 E-05 Iodine-133 Ci ND ND Iodine-135 Ci ND ND Total for Period Ci 4.93 E-05 4.14 E-05 O.OOE+00 O.OOE+00

3. Particulates Strontium-89 Ci 3.60 E-05 5.78 E-05 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 5.12 E-06 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 5.07 E-05 Total for Period Ci 3.60 E-05 1.14 E-04 O.OOE+00 O.OOE+00 ND Not Detected at the Plant Stack (I) There were no batch mode gaseous releases for this reporting period. i.1 10

TABLE C Vermont Yankee Effluent and Waste Disposal Annual Report First and Second Quarters, 2002 Gaseous Effluents .Ground Level Releases (3 Continuous Mode Batch Mode Quarter Quarter Nuclides Released Units 1(1) 2(1) 1(1) 2(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 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00

2. lodines Iodine-131 Ci Iodine-133 Ci Iodine-I 35 Ci Total for Period Ci O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00

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 0.00 E+00 O.OOE+00 O.OOE+00 0.OOE+00 (1) There were no ground level gaseous releases for this reporting period.

11

TABLE C (Continued)

Vermont Yankee Effluent and Waste Disposal Annual Report Third and Fourth Quarters, 2002 Gaseous Effluents - Ground Level Releases Continuous Mode Batch Mode Quarter Quarter Nuclides Released Units 3 4 (1)(2) 3 4

1. Fission Gases Krypton-85 Ci ND Krypton-85m Ci ND Krypton-87 Ci ND Krypton-88 Ci ND Xenon-133 Ci ND Xenon-135 Ci ND Xenon-135m Ci ND Xenon-138 Ci ND Unidentified Ci ND Total for Period Ci O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00

2. lodines Iodine-i131 Ci ND Iodine-133 Ci ND Iodine-135 Ci ND Total for Period Ci O.OOE+00 0.OOE+00 O.OOE+00 O.OOE+00

3. Particulates Strontium-89 Ci ND Strontium-90 Ci ND Cesium- 134 Ci ND Cesium-137 Ci 7.21 E-06 Barium-Lanthanum- 140 Ci ND Manganese-54 Ci 2.61 E-07 Chromium-51 Ci ND Cobalt-58 Ci ND Cobalt-60 Ci 1.40 E-05 Cerium-141 Ci ND Zinc-65 Ci 2.27 E-07 _

Iron-55 CI 3.86 E-06 Total for Period Ci O.OOE+00 2.56 E-05 O.OOE+00 O.OOE+00 (I)Burning of used oil was treated as a continuous release for the fourth quarter. Used oil was burned only in the fourth quarter.

(2) The North Warehouse stack was used as a ground level release point for buming of used oil.

ND Not detected in the used oil sample.

12

TABLE D Vermont Yankee Effluent and Waste Disposal Annual Report for 2002 Gaseous Effluents -Nonroutine Releases There were no non-routine or accidental gaseous releases during this reporting period.

13

TABLE 2A Vernont Yankee Effluent and Waste Disposal Annual Report for 2002 Liquid Effluents .Summation of All Releases There were no liquid releases during this reporting period.

14

TABLE 2B Vermont Yankee Effluent and Waste Disposal Annual Report for 2002 Liquid Effluents .Nonroutine Releases There were no non-routine or accidental liquid releases during this reporting period.

15

TABLE 3 Vermont Yankee Effluent and Waste Disposal Annual Report First and Second Quarters. 2002 Solid Waste and Irradiated Fuel Shipments A. SOLID WASTE SHIPPED OFFSITE FOR BURIAL OR DISPOSAL (not irradiated fuel)

1. Type of Waste Shipped from VY for Burial or Disposal 1Unit 1 and 2ND 1~~~~i Quarters 2002 Est. Total Error, %

a. Spent resins, filter sludges, evaporator bottoms, etc. Ci 4.84 E+00 _ 25 %

b. Dry compressible waste, contaminated equipment, etc. ci None + 25 %

c. Irradiated components, control rods, etc.: rn None 25 %

Shipped from Processor(s) for Burial or Disposal IUnit and N0 1~~~~~s" Quarters 22002 Est. Total Error, %

a. Spent resins, filter sludges, evaporator bottoms, etc. m3 .27 E+01 +25 %

b. Dry compressible waste, contaminated equipment, etc. m3 4.62 E+00 +25 %

Ci 8.40 E-02

c. Irradiated components, control rods, etc.: Ci None +25 %

2. Estimate of Major Nuclide Composition (By Type of Waste)

a. Spent resins, filter sludges, evaporator bottoms, etc. b. Dry compressible waste, contaminated equipment, Isotope Percent(l) Isotope Percent(l)

Zinc-65 2.02 E+01 % Iron-55 6.41 E+01 %

Cesium-137 1.30 E+01 % Zinc-65 7.50 E+00 %

Cobalt-60 2.62 E+01 % Cobalt-60 1.62 E+01 %

Ni-63 1.42 E+01 % Manganese-54 5.30 E+00 %

Manganese-54 7.50 E+00 % Cesium-137 1.20 E+00 %

Iron-55 1.34 E+01 % Cr-51 3.30 E+00 %

(1) Includes only those nuclides that are greater than 1% of the total activity.

Note: Sections A. I. and A.2. above do not include the data for the waste shipments from VY to the processors. The data for this waste will be included in the report that covers the year that this waste is shipped from the processor for burial or disposal.

16

TABLE 3 (Continued)

Vermont Yankee Effluent and Waste Disposal Annual Report First and Second Quarters, 2002 Solid Waste and Irradiated Fuel Shipments

3. Disposition of solid waste shipments (Ist and 2nd Quarters)

Number of From From Mode of l Destination Shipments VY Processor T ranspor t at ion Processor Burial or Disposal I x Truck CNS, Inc.

Bamnwell, SC 15 x Truck Envirocare Clive, UT x2 x Truck Truck G~~~~~~~TS Oak Duratek Ridge, TN 2 x Truck CNS, Inc Bamwell, SC Studsvik, 2 x Truck Erw~in, TN B. Irradiated Fuel Shipments (Disposition): None C. Additional Data (Ist and 2nd Quarters)

Supplemental Shipments from l Shipments from VY Shipments from Information VY to Processors for Burial or Disposal Disposal Class of solid waste 4A A (quantity of containers shipped IB IA not required)

Type of containers used 3 Strong Tight I Type A Strong Tight (quantity of Type ofIcontainers used Strong TypeTight A I containers not required)

Solidification agyent or None None None absorbent 17

TABLE 3 (Continued)

Vermont Yankee Effluent and Waste Disposal Annual Report Third and Fourth Quarters, 2002 Solid Waste and Irradiated Fuel Shipments B. SOLID WASTE SHIPPED OFFSITE FOR BURIAL OR DISPOSAL (not irradiated fuel)

2. Type of Waste

1. 3dand4th Shipped from VY for Burial or Disposal Unit Quarters 2002 Est. Total Error, %

a. Spent resins, filter sludges, evaporator bottoms, etc. Ml None + 25 %

Ci m

3

b. Dry compressible waste, contaminated equipment, etc. Ci None + 25 %

c. Irradiated components, control rods, etc. Ci None +25 %

3rdand4h Shipped from Processor(s) for Burial or Disposal Unit Quarters 2002 Est. Total Error, %

a. Spent resins, filter sludges, evaporator bottoms, etc. 1.66 E+0 +25% *j Ci 4.35 E+01 m3 1.30 E+01

b. Dry compressible waste, contaminated equipment, etc. ci 1.81 E-01 + 25 %

c. Irradiated components, control rods, etc. None + 25 %

2. Estimate of Major Nuclide Composition (By Type of Waste)

a. Spent resins, filter sludges, b. Dry compressible waste, c. Irradiated components, control rods, evaporator bottoms, etc. contaminated equipment, etc. etc.

Isotope Percent (1) Isotope Percent (I) Isotope Percent (1)

Zinc-65 2.02 E+01 % Iron-55 6.41 E+01 % N/A N/A Cesium-137 1.30 E+01 % Zinc-65 7.50 E+00 %

Cobalt-60 2.62 E+01 % Cobalt-60 1.62 E+01 %

Ni-63 1.42 E+01 % Manganese-54 5.30 E+00 %

Manganese-54 7.50 E+00 % Cesium-137 1.20 E+00 %

Iron-55 1.34 E+01 % Cr-51 3.30 E+00 %

(I) Includes only those nuclides that are greater than 1% of the total activity.

Note: Sections A. 1. and A.2. above do not include the data for the shipments from VY to the processors. The data for this waste will be included in the report that covers the year that this waste is shipped from the processor for burial or disposal.

18

TABLE 3 (Continued)

Vermont Yankee Effluent and Waste Disposal Annual Report Third and Fourth Quarters, 2002 Solid Waste and Irradiated Fuel Shipments

3. Disposition of Solid Waste Shipments (3 rd and 4 th Quarters)

Number of From From Mode of Destination Shipments VY Processor Transportation Processor Burial or Disposal 3lx Truck CNS, Inc.

Barnwell, SC II x Truck Envirocare 11 X Truck Clive, UT 4 X Truck GTS Duratek 4 X Truck Oak Ridge, TN Truck BrCNS, Inc.

B. Irradiated Fuel Shipments (Disposition): None C. Additional Data (3 rd and 4 th Quarters)

Supplemental Shipments from VY to I Shipments from VY for ~~~~~Shipments Processors from for Burial or Information Processors Burial or Disposal Disposal Class of solid waste 5A Nonc A,B (quantity of containers shipped 2B not required)

Strong Tight, Type A Type of containers used SoTigh None (quantity of containers not Type A required)

Solidification agent or None None None absorbent 19

TABLE 4A Vermont Yankee Maximum* Off-Site Doses/Dose Commitments to Members of the Public from Liquid and Gaseous Effluents for 2002 (10CFR50, Appendix I)

Dose (mrem)(a)

Source l Quarter Quarter Quarter l Quarter Year(b)

Liquid Effluents Total Body Dose Organ Dose Footnotes i__________Footnotes j (c)

(c)

_______________ Airborne Effluents _

Iodines and 3.27E-03 1.74E-01 3.59E-03 8.08E-04 1.82E-01 Particulates Footnotes (1) (2) (2) (3)

Noble Gases Beta Air (mrad) 1.82E-03 3.33E-03 -- -- 5.15E-03 Footnotes (4) (4) (d) (d)

Gamma Air (mrad) 1.20E-03 1.68E-03 2.88E-03 Footnotes (6) (5) (d) (d)

Direct Radiation 3.55 3.08 3.66 2.57 12.85 e)

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

(a) The numbered 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) There were no liquid releases in this quarter.

(d) There were no noble gas releases in this quarter.

(e) Maximum direct dose point located on the west site boundary (1) CHILD/ THYROID/ NWI 2900 meters from stack (2) INFANT/ TIYROID/ NWI 4260 meters from stack (3) CHILD/THYROID NW/2600 meters from stack. (4) NW/2900 meters from stack (5) NNW/550 meters from stack (6) SSE/850 meters from stack 20

TABLE 4B Vernont Yankee Maximum Annual Dose Commitments from Direct External Radiation, Plus Liquid and Gaseous Effluents for 2002*

(40CFR190)

Pathway Total Body Maximum Organ Thyroid (mrem) (mrem) (mrem)

Direct External (a) 12.85 12.85 12.85 Liquids (b) N/A N/A N/A Gases (c) 1.29E-03 1.35E-03 1.31E-03 Annual Total (d) 12.85 12.85 12.85

(*) 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 occupancy time fraction (assumed 100%) or residential shielding credit is assumed which would reduce real doses below the calculated values. Expected direct external radiation doses would be reduced by about 54% with a realistic residential shielding credit and occupancy time (0.7 shielding factor from Regulatory Guide 1.109 and annual occupancy time 6760 hours0.0782 days 1.878 hours 0.0112 weeks 0.00257 months ).

(b) There was no liquid release in 2002.

(c) Maximum dose to any organ over all age groups for each release.

(d) Annual dose limits contained in the EPA Radiation Protection Standards (40CFRI90) equal 25 mrem to the total body and any organ, except 75 mrem to the thyroid of a real member of the public.

21

TABLE4C Receptor Locations for Vermont Yaikee Sector Site Boundary (i) Nearest Resident(2 ) Anim l(2)Within l (Meters) (Meters) 10 km I (Meters)

N 400 1470 --

NNE 350 1400 5520 l_________________ _ .(C ows)

NE 350 1250 ..

ENE 400 970 E 500 930 ESE 700 2830 --

SE 750 1970 3600 l_____________________ (cow s)

SSE 850 2050 5240*

_______________________ (cows)

S 385 450 2220 l_____________________ (cows)

SSW 300 450 --

SW 250 410 8200 (cows)

WSW 250 450 9590 l ___________________ _____________________ ____________________ (goats)

W 300 620 820 (cows)

WNW 400 1060 6980*

(cows)

NW 550 2600 4260*

(cows)

NNW 550 2600

Receptor locations were conservatively included although these farms have been classified as "out of business" (1) Vermont Yankee UFSAR Figure 2.2-5.

(2) The location(s) given are based on data from the Vermont Yankee 2002 Land Use Census and are relative to the plant stack. Gardens are assumed to be present at all resident locations.

22

TABLE 4D Usage Factors for Various Gaseous Pathways at Vermont Yankee (From Reference 1, Table E-5('))

Veg. Leafy Veg. Milk Meat Inhalation Age Group (kg/yr) (kg/yr) (11yr) (kglyr) (m 3 1yr)

Adult 520 64 310 110 8,000 Teen 630 42 400 65 8,000 Child 520 26 330 41 3,700 Infant 0 0 330 0 1,400 (1) Regulatory Guide 1.109.

23

TABLE 4E Environmental Parameters for Gaseous Effluents at Vermont Yankee Vegetables Cow Milk Goat Milk Meat l Variable Stored Leafy Pasture Stored l Pasture Stored Pasture Stored YV Agricultural Productivity 2 2 0.70 2 0.70 2 0.70 2 (kg/m2 )

P Soil Surface Density (kg/M2 ) 240 240 240 240 240 240 240 240 T Transport Time to User (hrs) 48 48 48 48 480 480 TB Soil Exposure Time(a)(hrs). 131,400 131,400 131,400 131,400 131,400 131,400 131,400 131,400 TE Crop Exposure Time to 1,440 1,440 720 1,440 720 1,440 720 1,440 Plume (hrs) _

TH Holdup After Harvest (hrs) 1,440 24 0 2,160 0 2,160 0 2,160 QF Animals Daily Feed (kg/day) 50 50 6 6 50 50 FP Fraction of Year on Pasture (b) (b) (b)

FS Fraction Pasture Feed When 1 1 1 on Pasture(c) _

Note:Footnotes on following page.

24

TABLE 4E (Continued)

Environmental Parameters for Gaseous Effluents at Vermont Yankee if 91 Vegetables Cow Milk GToat Milk Ment l Variable Stored Leafy Pasture Stored Stored Pasture Stored FG Fraction of Stored 0.76 Vegetables 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 l (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 hours0.0088 days 0.211 hours 0.00126 weeks 2.8918e-4 months (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 (nongrowing 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).

25

Table 5A VERMONT YANKEE JAN 02 - DEC 02 METEOROLOGICAL DATA JOINT REQUENCY DISTRIBUTION 35.0 FT WIND DATA STABILITY CLASS A CLASS FREQUENCY (PERCENT) - 1.38 WIND DIRECTION FROM SPEED N NNE NE ENE E ESE SE SSE S SSW SW WSW W WNW Nd 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 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 (2) .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 C-3 0 1 0 1 0 0 0 0 0 0 0 0 0 0 1 0 0 3 (1) .00 .83 .00 .83 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .83 .00 .00 2.48 (2) .00 .01 .00 .01 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .01 .00 .00 .03 4-7 11 1 1 0 0 0 1 1 0 0 0 0 0 1 0 10 0 26 (1) 9.09 .83 .83 .00 .00 .00 .83 .83 .00 .00 .00 .00 .00 .83 .00 8.26 .00 21.49 (2) .13 .01 .01 .00 .00 .00 .01 .01 .00 .00 .00 .00 .00 .01 .00 .11 .00 .30 8-12 20 0 0 0 0 0 1 1 6 0 0 1 7 13 6 21 0 76 (1) 16.53 .00 .00 .00 .00 .00 .83 .83 4.96 .00 .00 .83 5.79 10.74 4.96 17.36 .00 62.81 (2) .23 .00 .00 .00 .00 .00 .01 .01 .07 .00 .00 .01 .08 .15 .07 .24 .00 .87 13-18 1 0 0 0 0 0 0 0 5 0 0 0 2 2 1 4 0 15 (1) .83 .00 .00 .00 .00 .00 .00 .00 4.13 .00 .00 .00 1.65 1.65 .83 3.31 .00 12.40 (2) .01 .00 .00 .00 .00 .00 .00 .00 .06 .00 .00 .00 .02 .02 .01 .05 .00 .17 19-24 0 0 0 0 0 0 0 0 0 0 0 0 0 1 0 0 0 1 (1) .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .83 .00 .00 .00 .83 (2) .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .01 .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 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 (2) .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 ALL SPEEDS 32 2 1 1 0 0 2 2 11 0 0 1 9 17 8 35 0 121 (1) 26.45 1.65 .83 .83 .00 .00 1.65 1.65 9.09 .00 .00 .83 7.44 14.05 6.61 28.93 .00 100.00 (2) .37 .02 .01 .01 .00 .00 .02 .02 .13 .00 .00 .01 .10 .19 .09 .40 .00 1.38 (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) 26

Table 5B VERMONT YANKEE JAN 02 - DEC 02 METEOROLOGICAL DATA JOINT FREQUENCY DISTRIBUTION 35.0 r WIND DATA STABILITY CLASS B CLASS FREQUENCY PERCENr) - 2.67 WIND DIRECTION FROM SPEED N NNE NE ENE E ESE SE SSE S SSW SW WSW W WNW NW NNel VRBL TOTAL MPH CALM 0 0 0 0 0 0 0 0 0 a 0 0 0 0 0 0 0 0 (1) .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 (2) .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 C-3 1 2 2 2 0 0 1 0 0 0 0 0 0 0 0 1 0 9 (1) .40 .80 .80 .80 .00 .00 .40 .00 .00 .00 .00 .00 .00 .00 .00 .40 .00 3.59 (2) .01 .02 .02 .02 .00 .00 .01 .00 .00 .00 .00 .00 .00 .00 .00 .01 .00 , .10 4-7 28 8 0 0 0 6 0 8 7 2 2 1 3 6 6 13 0 90 (1) 11.16 3.19 .00 .00 .00 2.39 .00 3.19 2.79 .80 .80 .40 1.20 2.39 2.39 5.18 .00 35.86 (2) .32 .09 .00 .00 .00 .07 .00 .09 .08 .02 .02 .01 .03 .07 .07 .15 .00 1.03 8-12 30 2 0 0 0 2 0 2 22 5 1 3 4 18 9 24 0 122 (1) 11.95 .80 .00 .00 .00 .80 .00 .80 8.76 1.99 .40 1.20 1.59 7.17 3.59 9.56 .00 48.61 (2) .34 .02 .00 .00 .00 .02 .00 .02 .25 .06 .01 .03 .05 .21 .10 .27 .00 1.39 13-18 4 0 0 0 0 0 0 0 4 0 0 0 1 11 5 3 0 28 (1) 1.59 .00 .00 .00 .00 .00 .00 .00 1.59 .00 .00 .00 .40 4.38 1.99 1.20 .00 11.16 (2) .05 .00 .00 .00 .00 .o0 .00 .00 .05 .00 .00 .00 .01 .13 .06 .03 .00 .32 19-24 0 0 0 0 0 0 0 0 0 0 0 0 0 1 1 0 0 2 (1) .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .40 .40 .00 .00 .80 (2) .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .01 .01 .00 .00 .02 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 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 (2) .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 ALL SPEEDS 63 12 2 2 0 8 1 10 33 7 3 4 8 36 21 41 0 251 (1) 25.10 4.78 .80 .80 .00 3.19 .40 3.98 13.15 2.79 1.20 1.59 3.19 14.34 8.37 16.33 .00 100.00

)2) .72 .14 .02 .02 .00 .09 .01 .11 .38 .08 .03 .05 .09 .41 .24 .47 .00 2.87 (1)-PERCENT OF ALL GOOD OBSERVATIONS FOR THIS PAGE (2)-PERCENT OF ALL GOOD OBSERVATIONS FOR THrS PERIOD C- CALM (WIND SPEED LESS THAN OR EQUAL TO .95 MPH) 27

Table 5C VERMONT YANKEE JAN 02 - DEC 02 METEOROLOGICAL DATA JOINT FEUENCY DISTRIBUTION 35.0 FT WIND DATA STABILITY CLASS C CLASS FREQUENCY (PERCENT) - 5.19 WIND DIRECTION FROM SPEED N NNE NE ENE E ESE SE SSE S SSW SW WSW W WIN1J 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 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 (2) .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 C-3 1 7 2 4 3 7 1 1 0 1 0 0 0 0 0 8 0 35 (1) .22 1.54 .44 .8S .66 1.54 .22 .22 .00 .22 .00 .00 .00 .00 .00 1.76 .00 7.69 (2) .01 .08 .02 .05 .03 .08 .01 .01 .00 .01 .00 .00 .00 .00 .00 .09 .00 .40 4-7 37 20 5 1 7 1S 19 17 22 2 3 1 4 11 6 18 0 18 (1) 8.13 4.40 1.10 .22 1.54 3.30 4.18 3.74 4.84 .44 .66 .22 .88 2.42 1.32 3.96 .00 41.32 (2) .42 .23 .06 .01 .08 .17 .22 .19 .25 .02 .03 .01 .05 .13 .07 .21 .00 2.15 8-12 34 4 0 0 0 2 5 6 49 15 2 2 10 29 9 20 0 187 (1) 7.47 .88 .00 .00 .00 .44 1.10 1.32 10.77 3.30 .44 .44 2.20 6.37 1.98 4.40 .00 41.10 (2) .39 .05 .00 .00 .00 .02 .06 .07 .56 .17 .02 .02 .11 .33 .10 .23 .00 2.13 13-18 4 0 0 0 0 0 0 0 7 1 0 0 0 18 5 9 0 44 (1) .88 .00 .00 .00 .00 .00 .00 .00 1.54 .22 .00 .00 .00 3.96 1.10 1.98 .00 9.67 (2) .05 .00 .00 .00 .00 .00 .00 .00 .08 .01 .00 .00 .00 .21 .06 .10 .00 .50 19-24 0 0 0 0 0 0 0 0 0 0 0 0 0 1 0 0 0 1 (1) .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .22 .00 .00 .00 .22 (2) .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .01 .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 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 (2) .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 ALL SPEEDS 76 31 7 5 10 24 25 24 78 19 5 3 14 S9 20 55 0 455 (1) 16.70 6.81 1.54 1.10 2.20 5.27 5.49 5.27 17.14 4.18 1.10 .66 3.08 12.97 4.40 12.09 .00 100.00 (2) .87 .35 .08 .06 .11 .27 .29 .27 .89 .22 .06 .03 .16 .67 .23 .63 .00 5.19 (1)-PERCENT OF ALL GOOD OBSERVATIONS FOR THIS PAGE (27-PERCENT OF ALL GOOD OBSERVATIONS FOR THIS PERIOD C- CALM (WIND SPEED LESS THAN OR EQUAL TO .95 MPH) 28

Table 5D VERMONTYANKEE JAN 02 - DEC 02 METEOROLOGICAL DATA JOINT FREQUENCY DISTRIBUTION 35.0 F WIND DATA STABILITY CLASS D CLASS FREQUENCY (PERCENT) - 45.79 WIND DIRECTION FROM SPEED N NIE NE ENE E ESE SE SSE S SSW SW WSW W WNW NW NNW VRBL TOTAL MPH CALM 1 1 1 0 1 0 0 0 1 0 0 2 2 0 0 2 0 11 (1) .02 .02 .02 .00 .02 .00 .00 .00 .02 .00 .00 .05 .05 .00 .00 .05 .00 .27 (2) .01 .01 .01 .00 .01 .00 .00 .00 .01 .00 .00 .02 .02 .00 .00 .02 .00 .13 C-3 107 84 51 69 74 58 58 85 70 52 41 31 25 35 64 137 0 1051 (1) 2.92 2.09 1.27 1.72 1.84 1.45 1.45 2.12 1.75 1.30 1.02 .77 .62 .87 1.60 3.42 .00 26.20 (2) 1.34 .96 .se .79 .84 .66 .66 .97 .80 .59 .47 .35 .29 .40 .73 1.56 .00 12.00 4-7 228 107 21 22 34 72 114 238 254 52 37 37 68 82 72 353 0 1791 (1) 5.68 2.67 .52 .55 .85 1.80 2.84 5.93 6.33 1.30 .92 .92 1.70 2.04 1.80 8.80 .00 44.65 (2) 2.60 1.22 .24 .25 .39 .82 1.30 2.72 2.90 .59 .42 .42 .78 .94 .82 4.03 .00 20.45 8-12 132 32 0 0 1 3 6 38 173 64 22 6 58 173 102 132 0 942 (1) 3.29 .80 .00 .00 .02 .07 .15 .95 4.31 1.60 .55 .15 1.45 4.31 2.54 3.29 .00 23.49 (2) 1.51 .37 .00 .00 .01 .03 .07 .43 1.97 .73 .25 .07 .66 1.97 1.16 1.51 .00 10.75 13-18 13 0 0 0 0 0 0 0 34 8 0 5 2 65 50 29 0 206 (1) .32 .00 .00 .00 .00 .00 .00 .00 .85 .20 .00 .12 .05 1.62 1.25 .72 .00 5.14 (2) .15 .00 .00 .00 .00 .00 .00 .00 .39 .09 .00 .06 .02 .74 .57 .33 .00 2.35 19-24 0 0 0 0 0 0 0 0 2 0 0 0 0 2 3 3 0 10 (1) .00 .00 .00 .00 .00 .00 .00 .00 .05 .00 .00 .00 .00 .05 .07 .07 .00 .25 (2) .00 .00 .00 .00 .00 .00 .00 .00 .02 .00 .00 .00 .00 .02 .03 .03 .00 .11 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 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 (2) .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 ALL SPEEDS 491 224 73 91 110 133 178 361 534 176 100 81 155 357 291 656 0 4011 (1) 12.24 5.58 1.82 2.27 2.74 3.32 4.44 9.00 13.31 4.39 2.49 2.02 3.86 8.90 7.26 16.36 .00 100.00 (2) 5.61 2.56 .83 1.04 1.26 1.52 2.03 4.12 6.10 2.01 1.14 .92 1.77 4.08 3.32 7.49 .00 45.79 (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 5E VERMONTYANKEE JAN 02 - DEC 02 METEOROLOGICAL DATA JOINT FREQUENCY DISTRIBUTION 35.0 FT WIND DATA STABILITY CLASS E CLASS FREQUENCY (PERCENT) - 30.06 WIND DIRECTION FROM SPEED N NNE NE ENE E ESE SE SSE S SSW SW WSW W WNW I) W MWI VRBL TOTAL MPH CALM 0 0 0 0 0 0 1 0 1 1 1 0 1 1 1 2 0 9 (1) .00 .00 .00 .00 .00 .00 .04 .00 .04 .04 .04 .00 .04 .04 .04 .08 .00 .34 (2) .00 .00 .00 .00 .00 .00 .01 .00 .01 .01 .01 .00 .01 .01 .01 .02 .00 .10 C-3 66 28 14 12 22 25 45 62 111 144 156 190 208 183 215 171 0 1652 (1) 2.51 1.06 .53 .46 .84 .95 1.71 2.35 4.22 5.47 5.92 7.22 7.90 6.95 8.17 6.49 .00 62.74 (2) .75 .32 .16 .14 .25 .29 .51 .71 1.27 1.64 1.78 2.17 2.37 2.09 2.45 1.95 .00 18.86 4-7 53 7 2 2 4 8 24 51 79 60 30 24 69 90 97 170 0 770 (1) 2.01 .27 .08 .08 .15 .30 .91 1.94 3.00 2.28 1.14 .91 2.62 3.42 3.68 6.46 .00 29.24 (2) .61 .08 .02 .02 .05 .09 .27 .58 .90 .68 .34 .27 .79 1.03 1.11 1.94 .00 8.79 8-12 14 1 0 0 0 0 1 4 17 6 0 4 10 38 49 46 0 190 (1) .53 .04 .00 .00 .00 .00 .04 .15 .65 .23 .00 .15 .38 1.44 1.86 1.75 .00 7.22 (2) .16 .01 .00 .00 .00 .00 .01 .05 .19 .07 .00 .05 .11 .43 .56 .53 .00 2.17 13-18 0 0 0 0 0 0 0 0 1 1 0 0 1 3 4 1 0 11 (1) .00 .00 .00 .00 .00 .00 .00 .00 .04 .04 .00 .00 .04 .11 .15 .04 .00 .42 (2) .00 .00 .00 .00 .00 .00 .00 .00 .01 .01 .00 .00 .01 .03 .05 .01 .00 .13 19-24 0 0 0 0 0 0 0 0 1 0 0 0 0 0 0 0 0 1 (1) .00 .00 .00 .00 .00 .00 .00 .00 .04 .00 .00 .00 .00 .00 .00 .00 .00 .04 (2) .00 .00 .00 .00 .00 .00 .00 .00 .01 .00 .00 .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 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 (2) .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 ALL SPEEDS 133 36 16 14 26 33 71 117 210 212 187 218 289 315 366 390 0 2633 (1) 5.05 1.37 .61 .53 .99 1.25 2.70 4.44 7.98 8.05 7.10 8.28 10.98 11.96 13.90 14.81 .00 100.00 (2) 1.52 .41 .18 .16 .30 .38 .81 1.34 2.40 2.42 2.13 2.49 3.30 3.60 4.18 4.45 .00 30.06 (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 SF VERMONTYANKEE JAN 02 - DEC 02 METEOROLOGICAL DATA JOINT FREQUENCY DISTRIBUTION 35.0 FT WIND DATA STABILITY CLASS F CLASS FREQUENCY (PERCENT) - 12.01 WIND DIRECTION FROM SPEED N NNE NE ENE E ESE SE SSE S SSW SW WSW W WNN NW NW VRBL TOTAL MPH CALM 1 2 0 0 1 1 0 1 0 0 2 1 0 0 1 0 0 10 (1) .10 .19 .00 .00 .10 .10 .00 .10 .00 .00 .19 .10 .00 .00 .10 .00 .00 .95 (2) .01 .02 .00 .00 .01 .01 .00 .01 .00 .00 .02 .01 .00 .00 .01 .00 .00 .11 C-3 1S 13 4 7 4 7 10 23 49 93 159 182 164 109 83 49 0 971 (1) 1.43 1.24 .38 .67 .38 .67 .95 2.19 4.66 8.84 15.11 17.30 15.59 10.36 7.89 4.66 .00 92.30 (2) .17 .15 .05 .08 .05 .08 .11 .26 .56 1.06 1.82 2.08 1.87 1.24 .95 .56 .00 11.08 4-7 2 1 0 1 0 0 0 4 7 7 2 3 9 9 6 16 0 67 (1) .19 .10 .00 .10 .00 .00 .00 .38 .67 .67 .19 .29 .86 .86 .57 1.52 .00 6.37 (2) .02 .01 .00 .01 .00 .00 .00 .05 .08 .08 .02 .03 .10 .10 .07 .18 .00 .76 8-12 0 0 0 0 0 0 0 0 0 0 0 1 0 2 1 0 0 4 (1) .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .10 .00 .19 .10 .00 .00 .38 (2) .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .01 .00 .02 .01 .00 .00 .05 13-18 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 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 (2) .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .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) .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 (2) .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .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 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 (2) .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 ALL SPEEDS 18 16 4 8 5 8 10 28 56 100 163 187 173 120 91 65 0 1052 (1) 1.71 1.52 .38 .76 .48 .76 .95 2.66 5.32 9.51 15.49 17.78 16.44 11.41 8.65 6.18 .00 100.00 (2) .21 .18 .05 .09 .06 .09 .11 .32 .64 1.14 1.86 2.13 1.97 1.37 1.04 .74 .00 12.01 (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

Table 5G VERMONT YANKEE JAN 02 - DEC 02 METEOROLOGICAL DATA JOINT FREQUENCY DISTRIBUTION 35.0 FT WIND DATA STABILITY CLASS G CLASS FREQUENCY (PERCENT) - 2.71 WIND DIRECTION FROM SPEED N NNE NE ENE E ESE SE SSE S SSW SW WSW W WNW NW NNW4 VRBL TOTAL MPH CALM 0 a 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 (1) .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 (2) .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 C-3 13 1 4 1 2 1 5 7 15 21 24 23 33 26 23 15 0 214 (1) 5.49 .42 1.69 .42 .84 .42 2.11 2.95 6.33 8.86 10.13 9.70 13.92 10.97 9.70 6.33 .00 90.30 (2) .15 .01 .05 .01 .02 .01 .06 .08 .17 .24 .27 .26 .38 .30 .26 .17 .00 2.44 4-7 2 0 0 0 0 0 0 2 0 1 0 1 2 3 6 6 0 23 (1) .84 .00 .00 .00 .00 .00 .00 .84 .00 .42 .00 .42 .84 1.27 2.53 2.53 .00 9.70 (2) .02 .00 .00 .00 .00 .00 .00 .02 .00 .01 .00 .01 .02 .03 .07 .07 .00 .26 8-12 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 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 (2) .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 13-18 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 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 (2) .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .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) .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00

12) .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .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 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 (2) .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 ALL SPEEDS Is 1 4 1 2 1 5 9 15 22 24 24 35 29 29 21 0 237 (1) 6.33 .42 1.69 .42 .84 .42 2.11 3.80 6.33 9.28 10.13 10.13 14.77 12.24 12.24 8.86 .00 100.00 (2) .17 .01 .05 .01 .02 .01 .06 .10 .17 .25 .27 .27 .40 .33 .33 .24 .00 2.71 (1)-PERCENT OF ALL GOOD OBSERVATIONS FOR THIS PAGE (21-PERCENT OF ALL GOOD OBSERVATIONS FOR THIS PERIOD C- CALM (WIND SPEED LESS THAN OR EQUAL TO .95 MPH) 32

Table 5H VERMONT YANKEE JAN 02 - DEC 02 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 2 3 1 0 2 1 1 1 2 1 3 3 3 1 2 4 0 30 (1) .02 .03 .01 .00 .02 .01 .01 .01 .02 .01 .03 .03 .03 .01 .02 .05 .00 .34 (2) .02 .03 .01 .00 .02 .01 .01 .01 .02 .01 .03 .03 .03 .01 .02 .05 .00 .34 C-3 213 136 77 96 105 98 120 178 245 311 380 426 430 353 386 381 0 3935 (1) 2.43 1.S5 .88 1.10 1.20 1.12 1.37 2.03 2.80 3.55 4.34 4.86 4.91 4.03 4.41 4.35 .00 44.92 (2) 2.43 1.55 .88 1.10 1.20 1.12 1.37 2.03 2.80 3.55 4.34 4.86 4.91 4.03 4.41 4.35 .00 44.92 4-7 361 144 29 26 45 101 158 321 369 124 74 67 155 202 193 586 0 2955 (1) 4.12 1.64 .33 .30 .51 1.15 1.80 3.66 4.21 1.42 .84 .76 1.77 2.31 2.20 6.69 .00 33.73 (2) 4.12 1.64 .33 .30 .51 1.15 1.80 3.66 4.21 1.42 .84 .76 1.77 2.31 2.20 6.69 .00 33.73 8-12 230 39 0 0 1 7 13 51 267 90 25 17 89 273 176 243 0 1521 (1) 2.63 .45 .00 .00 .01 .08 .15 .58 3.05 1.03 .29 .19 1.02 3.12 2.01 2.77 .00 17.36 (2) 2.63 .45 .00 .00 .01 .08 .15 .58 3.05 1.03 .29 .19 1.02 3.12 2.01 2.77 .00 17.36 13-18 22 0 0 0 0 0 0 0 51 10 0 5 6 99 65 46 0 304 (2) .25 .00 .00 .00 .00 .00 .00 .00 .58 .11 .00 .06 .07 1.13 .74 .53 .00 3.47 (2) .25 .00 .00 .00 .00 .00 .00 .00 .58 .11 .00 .06 .07 1.13 .74 .53 .00 3.47 19-24 0 0 0 0 0 0 0 0 3 0 0 0 0 5 4 3 0 25 (1) .00 .00 .00 .00 .00 .00 .00 .00 .03 .00 .00 .00 .00 .06 .05 .03 .00 .17 (2) .00 .00 .00 .00 .00 .00 .00 .00 .03 .00 .00 .00 .00 .06 .05 .03 .00 .17 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 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 (2) .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 ALL SPEEDS 828 322 107 122 153 207 292 551 937 536 482 518 683 933 826 1263 0 8760 (1) 9.45 3.68 1.22 1.39 1.75 2.36 3.33 6.29 10.70 6.12 5.50 5.91 7.80 10.65 9.43 14.42 .00 100.00 (2) 9.45 3.68 1.22 1.39 1.75 2.36 3.33 6.29 10.70 6.12 5.50 5.91 7.80 10.65 9.43 14.42 .00 100.00 (l)-PERCENT OF ALL GOOD OBSERVATIONS FOR THIS PAGE (2)-PERCENT OF ALL GOD OBSERVATIONS FOR TIS PERIOD C- CALM (WIND SPEED LESS THAN OR EQUAL TO .95 MPH) 33

Table 6A VERMONT YANKEE JAN 02 - DEC 02 METEOROLOGICAL DATA JOINT FREQUENCY DISTRIBUTION 297.0 FT WIND DATA STABILITY CLASS A CLASS FREQUENCY (PERCENT) - .24 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 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 (1) .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 (2) .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 C-3 4 0 0 1 0 2 2 0 0 0 0 0 0 0 0 0 0 9 (1) 19.05 .00 .00 4.76 .00 9.52 9.52 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 42 .86 (2) .05 .00 .00 .01 .00 .02 .02 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .10 4-7 2 0 0 0 0 -1 0 0 0 0 0 0 0 0 0 3 0 6 (1) 9.52 .00 .00 .00 .00 4.76 .00 .00 .00 .00 .00 .00 .00 .00 .00 14.29 .00 28.57 (2) .02 .00 .00 .00 .00 .01 .00 .00 .00 .00 .00 .00 .00 .00 .00 .03 .00 .07 8-12 2 0 0 0 0 0 1 0 0 0 0 0 0 0 0 1 0 4 (1) 9.52 .00 .00 .00 .00 .00 4.76 .00 .00 .00 .00 .00 .00 .00 .00 4.76 .00 19.05 (2) .02 .00 .00 .00 .00 .00 .01 .00 .00 .00 .00 .00 .00 .00 .00 .01 .00 .05 13-18 0 0 0 0 0 0 0 0 0 0 0 0 0 1 0 1 0 2 (1) .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 4.76 .00 4.76 .00 9.52 (2) .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .01 .00 .01 .00 .02 19-24 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 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 (2) .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 OT 24 0 0 0 0 0 0 0 0 0 0 0 0 0 0 O 0 0 0 (1) .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 (2) .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 ALL SPEEDS 8 0 0 1 0 3 3 0 0 0 0 0 0 1 0 5 0 21 (1) 38.10 .00 .00 4.76 .00 14.29 14.29 .00 .00 .00 .00 .00 .00 4.76 .00 23.61 .00 100.00 (2) .09 .00 .00 .01 .00 .03 .03 .00 .00 .00 .00 .00 .00 .01 .00 .06 .00 .24 (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

Table 6B VERMONT YANKEE JAN 02 - DEC 02 METEOROLOGICAI DATA JOINT FREQUENCY DISTRIBUTION 297.0 FT WIND DATA STABILITY CLASS B CLASS FREQUENCY (PERCENT) - .34 WIND DIRECTION FROM SPEED N NNE NE ENE E ESE SE SSE S SSW SW WSW W WNW h'W NNW VRBL TOTAL MPH CALM 00 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 (1) .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 (21 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 C-3 2 0 0 0 0 0 0 0 0 0 0 0 0 1 0 1 0 4 (1) 6.67 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 3.33 .00 3.33 .00 13.33 (2) .02 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .01 .00 .01 .00 .05 4-7 0 0 0 0 0 0 1 0 0 0 0 0 0 1 1 0 0 3 (1) .00 .00 .00 .00 .00 .00 3.33 .00 .00 .00 .00 .00 .00 3.33 3.33 .00 .00 10.00 (2) .00 .00 .00 .00 .00 .00 .01 .00 .00 .00 .00 .00 .00 .01 .01 .00 .00 .03 8-12 0 0 1 0 0 0 0 0 0 1 0 0 0 0 0 5 0 7 (1) .00 .00 3.33 .00 .00 .00 .00 .00 .00 3.33 .00 .00 .00 .00 .00 16.67 .00 23.33 (2) .00 .00 .01 .00 .00 .00 .00 .00 .00 .01 .00 .00 .00 .00 .00 .06 .00 .08 13-18 0 0 0 0 0 0 0 0 1 0 0 0 3 1 0 4 0 9 (1) .00 .00 .00 .00 .00 .00 .00 .00 3.33 .00 .00 .00 10.00 3.33 .00 13.33 .00 30.00 (2) .00 .00 .00 .00 .00 .00 .00 .00 .01 .00 .00 .00 .03 .01 .00 .05 .00 .10 19-24 0 0 0 0 0 0 0 0 2 0 0 0 2 0 0 2 0 6 (1) .00 .00 .00 .00 .00 .00 .00 .00 6.67 .00 .00 .00 6.67 .00 .00 6.67 .00 20.00 (2) .00 .00 .00 .00 .00 .00 .00 .00 .02 .00 .00 .00 .02 .00 .00 .02 .00 .07 GT 24 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 1 0 (1) .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 3.33 .00 3.33 (2) .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .01 .00 .01 ALL SPEEDS 2 0 1 0 0 0 1 0 3 1 0 0 5 3 1 13 0 30 (1) 6.67 .00 3.33 .00 .00 .00 3.33 .00 10.00 3.33 .00 .00 16.67 10.00 3.33 43.33 .00 100.00 (2) .02 .00 .01 .00 .00 .00 .01 .00 .03 .01 .00 .00 .06 .03 .01 .15 .00 .34 (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) 35

Table 6C VERMONT YANKEE JAN 02 - DEC 02 METEOROLOGICAL DATA JOINT FREQENCY DISTRIBUTION 297.0 FT WIND DATA STABILITY CLASS C CLASS FREQUENCY (PERCENT) - 1.47 WIND DIRECTION FROM SPEED N NNE NE ENE E ESE SE SSE S SSW SW WSW w WNX 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 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 (2) .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 C-3 1 0 0 0 0 0 0 1 0 0 0 0 0 0 1 0 0 3 (1) .78 .00 .00 .00 .00 .00 .00 .78 .00 .00 .00 .00 .00 .00 .78 .00 .00 2.33 (2) .01 .00 .00 .00 .00 .00 .00 .01 .00 .00 .00 .00 .00 .00 .01 .00 .00 .03 4-7 3 1 0 0 1 0 0 0 0 0 0 0 0 0 2 13 0 20 (1) 2.33 .78 .00 .00 .78 .00 .00 .00 .00 .00 .00 .00 .00 .00 1.55 10.08 .. 00 15.50 (2) .03 .01 .00 .00 .01 .00 .00 .00 .00 .00 .00 .00 .00 .00 .02 .15 .00 .23 8-12 5 2 0 1 0 1 2 5 3 0 0 0 3 2 1 8 0 33 (1) 3.88 1.55 .00 .78 .00 .78 1.55 3.88 2.33 .00 .00 .00 2.33 1.55 .78 6.20 .00 25.58 (2) .06 .02 .00 .01 .00 .01 .02 .06 .03 .00 .00 .00 .03 .02 .01 .09 .00 .38 13-18 11 0 0 0 0 0 0 0 4 0 0 1 9 7 6 19 0 57 (1) 8.53 .00 .00 .00 .00 .00 .00 .00 3.10 .00 .00 .78 6.98 5.43 4.65 14.73 .00 44.19 (2) .13 .00 .00 .00 .00 .00 .00 .00 .05 .00 .00 .01 .10 .08 .07 .22 .00 .65 19-24 0 0 0 0 0 0 0 0 2 0 0 0 1 2 0 9 0 14 (1) .00 .00 .00 .00 .00 .00 .00 .00 1.55 .00 .00 .00 .78 1.55 .00 6.98 .00 10.85 (2) .00 .00 .00 .00 .00 .00 .00 .00 .02 .00 .00 .00 .01 .02 .00 .10 .00 .16 GT 24 0 0 0 0 0 0 0 0 0 0 0 0 1 1 0 0 0 2 (1) .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .78 .78 .00 .00 .00 1.55 (2) .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .01 .01 .00 .00 .00 .02 ALI LSPEEDS 20 3 0 1 1 1 2 6 9 0 0 1 14 12 10 49 0 129 (1) 15.50 2.33 .00 .78 .78 .78 1.55 4.65 6.98 .00 .00 .78 10.85 9.30 7.75 37.98 .00 100.00 (2) .23 .03 .00 .01 .01 .01 .02 .07 .10 .00 .00 .01 .16 .14 .11 .56 .00 1.47 (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) 36

Table 6D VERMONT YANXEE JAN 02 - DEC 02 METEOROLOGICAL DATA JOINT FREQUENCY DISTRIBUTION 297.0 FT WIND DATA STABILITY CLASS D CLASS FREQUENCY (PERCENT) - 52.40 WIND DIRECTION FROM SPEED N NNE HE ENE E ESE SE SSE S SSW SW WSW W WIN NW NNW VRBL TOTAL MPH CALM 0 0 0 0 0 0 1 0 0 0 0 0 0 0 0 0 0 1 (1) .00 .00 .00 .00 .00 .00 .02 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .02 (2) .00 .00 .00 .00 .00 .00 .01 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .01 C-3 70 33 39 32 35 43 61 39 16 9 4 5 8 10 22 59 0 484 (1) 1.53 .72 .83 .70 .76 .94 1.33 .85 .35 .20 .09 .11 .17 .22 .48 1.29 .00 10.54 (2) .80 .38 .43 .37 .40 .49 .70 .45 .18 .10 .05 .06 .09 .11 .25 .67 .00 5.53 4-7 120 62 32 27 41 86 175 152 101 18 19 13 9 14 30 213 0 1112 (1) 2.61 1.35 .70 .59 .89 1.87 3.81 3.31 2.20 .39 .41 .28 .20 .31 .65 4.64 .00 24.23 (2) 1.37 .71 .37 .31 .47 .98 2.00 1.74 1.15 .21 .22 .15 .10 .16 .34 2.43 .00 12.69 8-12 174 55 5 1 3 22 61 207 309 49 46 36 107 111 69 300 0 1555 (1) 3.79 1.20 .11 .02 .07 .48 1.33 4.51 6.73 1.07 1.00 .78 2.33 2.42 1.50 6.54 .00 33.88 (2) 1.99 .63 .06 .01 .03 .25 .70 2.36 3.53 .56 .53 .41 1.22 1.27 .79 3.42 .00 17.75 13-18 131 9 0 0 1 1 7 34 186 28 15 14 108 205 89 246 0 1074 (1) 2.85 .20 .00 .00 .02 .02 .15 .74 4.05 .61 .33 .31 2.35 4.47 1.94 5.36 .00 23.40 (2) 1.50 .10 .00 .00 .01 .01 .08 .39 2.12 .32 .17 .16 1.23 2.34 1.02 2.81 .00 12.26 19-24 17 0 0 0 0 0 0 1 33 5 2 3 17 84 36 107 0 305 (1) .37 .00 .00 .00 .00 .00 .00 .02 .72 .11 .04 .07 .37 1.83 .78 2.33 .00 6.64 (2) .19 .00 .00 .00 .00 .00 .00 .01 .38 .06 .02 .03 .19 .96 .41 1.22 .00 3.48 GT 24 5 0 0 0 0 0 0 0 4 0 0 1 2 12 6 29 0 59 (I .11 .00 .00 .00 .00 .00 .00 .00 .09 .00 .00 .02 .04 .26 .13 .63 .00 1.29 (2) .06 .00 .00 .00 .00 .00 .00 .00 .05 .00 .00 .01 .02 .14 .07 .33 .00 .67 ALL SPEEDS 517 159 75 60 80 152 305 433 649 109 86 72 251 436 252 954 0 4590 (1) 11.26 3.46 1.63 1.31 1.74 3.31 6.64 9.43 14.14 2.37 1.87 1.57 5.47 9.50 5.49 20.78 .00 100.00 (2) 5.90 1.82 .86 .68 .91 1.74 3.48 4.94 7.41 1.24 .98 .82 2.87 4.98 2.89 10.89 .00 52.40 (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

Table 6E VERMONT YANKEE JAN 02 - DEC 02 METEOROLOGICAL DATA JOINT FREQUENCY DISTRIBUTION 297.0 FT WIND DATA STABILITY CLASS E CLASS FREQUENCY (PERCENT) - 33.52 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 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 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 (2) .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 C-3 98 87 70 57 60 83 87 61 22 20 11 4 11 6 26 69 0 772 (1) 3.34 2.96 2.38 1.94 2.04 2.83 2.96 2.08 .75 .68 .37 .14 .37 .20 .89 2.35 .00 26.29 (2) 1.12 .99 .80 .65 .68 .95 .99 .70 .25 .23 .13 .05 .13 .07 .30 .79 .00 8.81 4-7 163 21 11 10 20 27 197 194 83 27 25 31 24 33 62 304 0 1232 (1) 5.55 .72 .37 .34 .68 .92 6.71 6.61 2.83 .92 .85 1.06 .82 1.12 2.11 10.35 .00 41.96 (2) 1.86 .24 .13 .11 .23 .31 2.25 2.21 .95 .31 .29 .35 .27 .38 .71 3.47 .00 14 .06 8-12 40 3 0 0 2 3 29 64 71 42 7 13 54 91 42 202 0 663 (1) 1.36 .10 .00 .00 .07 .10 .99 2.18 2.42 1.43 .24 .44 1.84 3.10 1.43 6.88 .00 22.58 (2) .46 .03 .00 .00 .02 .03 .33 .73 .81 .48 .08 .15 .62 1.04 .48 2.31 .00 7.57 13-18 23 0 0 0 1 0 0 0 24 13 1 9 19 52 29 69 0 240 (1) .78 .00 .00 .00 .03 .00 .00 .00 .82 .44 .03 .31 .65 1.77 .99 2.35 .00 8.17 (2) .26 .00 .00 .00 .01 .00 .00 .00 .27 .15 .01 .10 .22 .59 .33 .79 .00 2.74 19-24 2 0 0 0 0 0 0 0 4 0 1 0 2 4 1 11 0 25 (1) .07 .00 .00 .00 .00 .00 .00 .00 .14 .00 .03 .00 .07 .14 .03 .37 .00 .85 (2) .02 .00 .00 .00 .00 .00 .00 .00 .05 .00 .01 .00 .02 .05 .01 .13 .00 .29 GT 24 0 0 0 0 0 0 0 0 2 1 0 1 0 0 0 0 0 4 (1) .00 .00 .00 .00 .00 .00 .00 .00 .07 .03 .00 .03 .00 .00 .00 .00 .00 .14 (2) .00 .00 .00 .00 .00 .00 .00 .00 .02 .01 .00 .01 .00 .00 .00 .00 .00 .05 ALL SPEEDS 326 111 81 67 83 113 313 319 206 103 45 58 110 186 160 655 0 2936 (1) 11.10 3.78 2.76 2.28 2.83 3.85 10.66 10.87 7.02 3.51 1.53 1.99 3.75 6.34 5.45 22.31 .00 100.00 (27 3.72 1.27 .92 .76 .95 1.29 3.57 3.64 2.35 1.18 .51 .66 1.26 2.12 1.83 7.48 .00 33.52 (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

Table 6F VERMONT YANKEE JAN 02 - DEC 02 ETEOROLOGICAL DATA JOINT FREQUENCY DISTRIBUTION 297.0 FT WIND DATA STABILITY CLASS F CLASS FREQUENCY (PERCENT) - 12.00 WIND DIRECTION FROM SPEED N NNE NE ENE E ESE SE SSE S SSW SW WSW W WNW NW NNW VRBL TOTAL MPH CALM 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 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 (2) .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 C-3 38 24 28 20 32 38 46 43 25 12 12 16 10 12 14 30 0 400 (1) 3.62 2.28 2.66 1.90 3.04 3.62 4.38 4.09 2.38 1.14 1.14 1.52 .95 1.14 1.33 2.85 .00 38.06 (2) .43 .27 .32 .23 .37 .43 .53 .49 .29 .14 .14 .18 .11 .14 .16 .34 .00 4.57 4-7 53 7 S 2 12 24 72 89 38 19 23 23 19 18 34 87 0 525 (1) 5.04 .67 .46 .19 1.14 2.28 6.85 8.47 3.62 1.81 2.19 2.19 1.81 1.71 3.24 8.28 .00 49.95 (2) .61 .08 .06 .02 .14 .27 .82 1.02 .43 .22 .26 .26 .22 .21 .39 .99 .00 5.99 8-12 3 0 0 0 0 0 10 12 10 11 8 3 9 e 10 37 0 121 (1) .29 .00 .00 .00 .00 .00 .95 1.14 .95 1.05 .76 .29 .86 .76 .95 3.52 .00 11.51 (2) .03 .00 .00 .00 .00 .00 .11 .14 .11 .13 .09 .03 .10 .09 .11 .42 .00 1.38 13-18 0 0 0 0 0 0 0 0 0 0 0 1 2 0 0 2 0 5 (1) .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .10 .19 .00 .00 .19 .00 .48

12) .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .01 .02 .00 .00 .02 .00 .06 19-24 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 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 (2) .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 GT 24 0 0 0 0 a 0 0 0 0 0 0 0 0 0 0 0 0 0 (1) .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 (2) .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 ALL SPEEDS 94 31 33 22 44 62 128 144 73 42 43 43 40 38 58 156 0 1051 (1) 8.94 2.95 3.14 2.09 4.19 5.90 12.18 13.70 6.95 4.00 4.09 4.09 3.81 3.62 5.52 14.84 .00 100.00 (2) 1.07 .35 .38 .25 .50 .71 1.46 1.64 .83 .48 .49 .49 .46 .43 .66 1.78 .00 12.00 (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) 39

Table 6G VERMONT YANKEE JAN 02 - DEC 02 METEOROLOGICAL DATA JOINT FREQUENCY DISTRIBUTION 297.0 FT WIND DATA STABILITY CLASS G CLASS FREQUENCY (PERCENT) .03 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 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 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 (2) .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 C-3 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 (13 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 (2) .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 4-7 0 0 0 0 0 0 0 0 0 0 0 0 1 0 0 1 0 2 (13 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 33.33 .00 .00 33.33 .00 66.67 (2) .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .01 .00 .00 .01 .00 .02 6-12 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 1 0 1 (1) .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 33.33 .00 33.33 (2) .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .01 .00 .01 13-18 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 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 (23 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .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) .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 (2) .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .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 (3) .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 (23 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 ALL SPEEDS 0 0 0 0 0 0 0 0 0 0 0 0 1 0 0 2 0 3 (1) .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 33.33 .00 .00 66.67 .00 100.00 (2) .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .01 .00 .00 .02 .00 .03 (l)PERCENT OF ALL GOOD OBSERVATIONS FOR THIS PAGE (23-PERCENT OF ALL GOOD OBSERVATIONS FOR THIS PERIOD C- CALM (WIND SPEED LESS THAN OR EQUAL TO .95 MPH) 40

Table 6H VERMONTYANKEE JAN 02 - DEC 02 METEOROLOGICAL DATA JOINT FREQUENCY DISTRIBUTION 297.0 FT WIND DATA STABILITY CLASS ALL CLASS FREQUENCY (PERCENT) - 100.00 WIND DIRECTION ERON SPEED N NNE NE ENE E ESE SE SSE S SSW SW WSW W WNW NW NNW VRBL TOTAL MPH CALM 0 0 0 0 0 0 1 0 0 0 0 0 0 0 0 0 0 1 (1) .00 .00 .00 .00 .00 .00 .01 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .01 (2) .00 .00 .00 .00 .00 .00 .01 .00 .00 .00 .00 .00 .00 .00 .00 .00 .00 .01 C-3 213 144 136 110 127 166 196 144 63 41 27 25 29 29 63 159 0 1672 (1) 2.43 1.64 1.55 1.26 1.45 1.89 2.24 1.64 .72 .47 .31 .29 .33 .33 .72 1.82 .00 19.09 (2) 2.43 1.64 1.55 1.26 1.45 1.89 2.24 1.64 .72 .47 .31 .29 .33 .33 .72 1.82 .00 19.09 4-7 341 91 48 39 74 138 445 435 222 64 67 67 53 66 129 621 0 2900 (1) 3.89 1.04 .55 .45 .84 1.58 5.08 4.97 2.53 .73 .76 .76 .61 .75 1.47 7.09 .00 33.11 (2) 3.89 1.04 .55 .45 .84 1.58 5.08 4.97 2.53 .73 .76 .76 .61 .75 1.47 7.09 .00 33.11 8-12 224 60 6 2 5 26 103 288 393 103 61 52 173 212 122 554 0 2384 (11 2.56 .68 .07 .02 .06 .30 1.18 3.29 4.49 1.18 .70 .59 1.97 2.42 1.39 6.32 .00 27.21 (2) 2.56 .68 .07 .02 .06 .30 1.18 3.29 4.49 1.18 .70 .59 1.97 2.42 1.39 6.32 .00 27.21 13-18 165 9 0 0 2 1 7 34 215 41 16 25 141 266 124 341 0 1387 (1) 1.88 .10 .00 .00 .02 .01 .08 .39 2.45 .47 .18 .29 1.61 3.04 1.42 3.89 .00 15.83 (2) 1.88 .10 .00 .00 .02 .01 .08 .39 2.45 .47 .18 .29 1.61 3.04 1.42 3.89 .00 15.83 19-24 19 0 0 0 0 0 0 1 41 5 3 3 22 90 37 129 0 350 (1) .22 .00 .00 .00 .00 .00 .00 .01 .47 .06 .03 .03 .25 1.03 .42 1.47 .00 4.00 (2) .22 .00 .00 .00 .00 .00 .00 .01 .47 .06 .03 .03 .25 1.03 .42 1.47 .00 4.00 GT 24 5 0 0 0 0 0 0 0 6 1 0 2 3 13 6 30 0 66 (1) .06 .00 .00 .00 .00 .00 .00 .00 .07 .01 .00 .02 .03 .15 .07 .34 .00 .75 (2) .06 .00 .00 .00 .00 .00 .00 .00 .07 .01 .00 .02 .03 .15 .07 .34 .00 .75 ALL SPEEDS 967 304 190 151 208 331 752 902 940 255 174 174 421 676 481 1834 0 8760 (1) 11.04 3.47 2.17 1.72 2.37 3.78 8.58 10.30 10.73 2.91 1.99 1.99 4.81 7.72 5.49 20.94 .00 100.00 (2) 11.04 3.47 2.17 1.72 2.37 3.78 8.58 10.30 10.73 2.91 1.99 1.99 4.81 7.72 5.49 20.94 .00 100.00 (1)-PERCENT OF ALL GOOD OBSERVATIONS FOR THIS PAGE (2)-PERCENT OF ALL GOOD OBSERVATIONS FOE THIS PERIOD C- CALM (WIND SPEED LESS THAN OR EQUAL TO .95 MPH 41

APPENDIX A EFFLUENT AND WASTE DISPOSAL ANNUAL REPORT Supplemental Information for 2002 Facility: Vermont Yankee Nuclear Power Station Licensee: Entergy Nuclear Vermont Yankee, LLC 1A. 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. Iodine-131, Iodine-133. Tritium and Radionuclides in Particulate Form With Half-Lives Greater Than 8 Days 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. Iodine-131. Iodine-133, Tritium and Radionuclides in Particulate Form With Half-Lives Greater Than 8 Days No ECL Limits A-1

c. Liquids 3/4.2.1 Sum of the fractions of ECL excluding noble gases (10CFR20, Appendix B, Table 2, Column 2): < 1.OE+O 3/4.2.1 Total noble gas concentration: < 2E-04 pCi/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 B and the totals of Table IA. The error involved in these steps may be approximately +/-23 percent.

b. Iodines Continuous isokinetic samples are drawn from the plant stack through a particulate filter and charcoal cartridge. The filters and cartridges are normally removed weekly and are analyzed for Iodine-131, 132, 133, 134, and 135. The error involved in these steps may be approximately +/-18 percent.

A-2

c. Particulates The particulate filters described in b. above are also counted for particulate radioactivity. The error involved in this sample is also approximately +/-18 percent.

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 sampling design included a cold trap collection device for this sample collection. The error involved in this sample is approximately +/-15 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 ([iCi/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 Waste oil was burned during the fourth quarter and was considered to be a continuous release.

The gaseous releases from burning waste oil are treated as either batch or continuous releases based on the total hours of burning in a calendar quarter.

6. ABNORMAL RELEASES

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

b. Gaseous There were no nonroutine gaseous releases (measured) during the reporting period.

A-4

APPENDIX B LIQUID HOLDUP TANKS Requirement Technical Specification 3.8.D.1 limits the quantity of radioactive material contained in any outside tank. With the quantity of radioactive material in any outside tank exceeding the limits of Technical Specification 3.8.D. , a description of the events leading to this condition is required in the next annual Radioactive Effluent Release Report per 10.1.

Response: The limits of Technical Specification 3.8.D.1 were not exceeded during this reporting period.

B-1

APPENDIX C RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION Requirement: Radioactive liquid effluent monitoring instrumentation channels are required to be operable in accordance with ODCM Table 3.1.1. If an inoperable radioactive liquid effluent monitoring instrument is not returned to operable 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 inoperability are required per ODCM Section 10.1.

Response: Since the requirements of ODCM Table 3.1.1 governing the operability of radioactive liquid effluent monitoring instrumentation were met for this reporting period, no response is required.

C-1

APPENDIX D RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION Requirement: Radioactive gaseous effluent monitoring instrumentation channels are required to be operable in accordance with ODCM Table 3.1.2. If inoperable gaseous effluent monitoring instrumentation is not returned to operable 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 operability of radioactive gaseous effluent monitoring instrumentation were met for this reporting period, no response is required.

D-1

APPENDIX E 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 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: No changes were needed in the milk sanpling locations as specified in ODCM Table 3.5.1 and implemented in ODCM Table 7.1 during the reporting year.

E-1

APPENDIX F 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 of 2002. 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.

F-I

APPENDIX G PROCESS CONTROL PROGRAM Requirement: ODCM Section 10.1 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

APPENDIX H 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: During the reporting period, two changes, Revision No. 29 and Revision No. 30 were made to the ODCM.

The major changes included in Revision 29 to the ODCM were:

(29.A) Additional Definitions added to Section 2 Terms such as "Functional Test" and "Source Check" were added in response to suggestions made during a previous inspection.

(29.B) Frequency of Air Sample Collections was restored to a weekly interval The collection frequency had been increased in August, 2001. The ODCM was revised to reflect the increase in frequency.

(29.C) LLD for Iodine in Drinking Water Samples Added Page 3/4-35 was revised to reflect the addition of Low Level Iodine Lower Limit of Detection changes as a result of suggestions from a previous inspection.

(29.D) NRC Correspondence added to the ODCM as New Appendices Appendices H and I were added to include copies of the correspondence with the NRC concerning the spreading of slightly contaminated radioactive materials on designated locations on the Vermont Yankee site. The creation of these appendices fulfilled an NRC requirement to include this correspondence in the ODCM.

(29.E) Elimination of certain minor inconsistencies between the ODCM and the UFSAR Section 9.2.5 A number of minor changes were made to ensure fidelity between the ODCM and UFSAR Section 9.2.5.

H-1

The major changes included in Revision 30 to the ODCM were:

(30.A) Revision of Section 6.1.1. "Method to Calculate Direct Dose from Plant Operation".

Based upon empirically-derived correlations between west fence-line dose and dose measured by the Main Steam Line Radiation Monitors, a change was made in the method for calculating the accumulated dose at the site boundary (west fence-line). Equation 6-27 was expanded into four individual calculations which together result in a more appropriate assessment of site boundary dose emanating from the turbine building during operation.

The above-noted ODCM changes were determined to maintain the level of protection in the calculation of off-site doses resulting from radioactive gaseous and liquid effluents since no changes have been made to either the dose calculation or set-point methodologies. It is therefore concluded that these revisions will maintain the level of radioactive effluent control required by 10CFR20.1302, 40CFR190, OCFR50.36a, and Appendix I to OCFR Part 50, and not adversely impact the accuracy or reliability of effluent dose or set-point calculations.

Revision 30 (including changes from Revision 29) of the ODCM is submitted to the Nuclear Regulatory Commission separately but concurrently with this report.

H-2

APPENDIX I RADIOACTIVE LIQUID, GASEOUS, AND SOLID WASTE TREATMENT SYSTEMS Requirement: ODCM Section 10.4 requires that licensee initiated major changes to the radioactive waste systems (liquid, gaseous, and solid) be reported to the Commission in the annual Radioactive Effluent 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.

I-1

APPENDIX J ON-SITE DISPOSAL OF SEPTIC/SILT/SOIL WASTE Requirement: Off-Site Dose Calculation Manual, Appendices B and F require that the dose impact due to on-site disposal of septic waste and the cooling tower silt during the reporting year and from previous years be reported to the Nuclear Regulatory Commission in the annual Radioactive Effluent Report if disposals occur during the reporting year.

Vermont Yankee Nuclear Power Station will report in the Annual Radioactive Effluent Release Report a list of the radionuclides present and the total radioactivity associated with the on-site disposal activities at Vermont Yankee.

Response: There was one on-site disposal of septic waste and one on-site disposal of construction soil during the reporting year. The total volume of the septage spread was approximately 11,000 gallons and the total volume of soil spread was approximately 28 m3 . The total activity spread on the 1.9 acres (southern) on-site disposal field from 2002 spreadings and from previous years was as follows:

Activity from 2002 and Activity from Activity Spread in 2002 Past Disposals Decayed to 2002 Nuclide (Ci) (ci)

Mn-54 1.90E-08 9.56E-08 Co-60 7.60E-08 1.75E-05 Zn-65 0.OOE+00 7.52E-07 Cs-134 1.90E-09 1.90E-09 Cs-137 2.62E-06 6.84E-05 Ce-141 0.OOE+00 6.89E-1 I The maximum organ (or whole body) incremental dose from material spread in 2002 was estimated to be 3.71 E-03 mrem/yr. The maximum organ dose from all past spreading operations, including the material spread in 2002, totaled 1.09E-01 mrem/yr. These calculated values are within the 1 rnrem/yr limit applied during the period of operational control of the site. The projected hypothetical dose for the period following the loss of operational control of the site area due to all spreading operations to-date is 2.92E-01 mrem/yr versus a 5 mrem/yr dose limit.

J-1

VERMONT YANKEE NUCLEAR POWER STATION OFF-SITE DOSE CALCULATION MANUAL I REVISION 30 This document contains Vermont Yankee proprietary information. This information may not be transmitted, in whole or in part, to any other organization without permission of Vermont Yankee.

Originator:_ /0/Z30a PriitName - ignhature Date Reviewed: =An M. t-IQIe. nl-Jm aaou-( 03 Print Name Signature Date Plant Operations Review Committee Approved: /JWw / :iJ Print Name Signature Date Plant Manager

LIST OF AFFECTED PAGES Pages Revision i to ix 30 1-1 to 1-18 30 2-1 to 2-10 29 314-1 to 3/447 29 5-1 to 54 29 6-1 to 6-59 30 7-1 to 7-12 27 8-1 to 8-23 29 9-1 to 9-10 27 10-1 to 10-7 27 R-1 27 Al A B-i to B-22 9 B-23 23 B-24 to B-33 9 All C D-1 to D-3 16 AlE F-I to F-34 21 G-l to G-20 21 H-1 to l-41 29 I-l to I-9 29

Appendix A deleted with Revision 26. Appendix C and E removed and archived with Revision 26 I Revision 30 Date 10/30/02 -ii-

ABSTRACT The VYNPS ODCM (Vernont 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 due to normal plant operation. The effluent dose models are based on the U.S. NRC Regulatory Guide 1.109. Revision 1.

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 are suitable to demonstrate compliance with effluent controls.

Revision 30 Date 10/30/02 -iii-

TABLE OF CONTENTS LIST OF AFFECTED PAGES ................... ,i ABSTRACT .. iii TABLE OF CONTENTS ................ iv LIST OF TABLES .. vii LIST OF FIGURES .. ix 1.0 INTRODUCTIN .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 ENVIRONMENTAL CONTROLS 3/4-1 3/4.1 Instrumentation .3/4-2 3/4.1.1 Radioactive Liquid Effluent Instrumentation .3/4-2 314.1.2 Radioactive Gaseous Effluent Instrumentation .3/4-7 3/4.2 Radioactive Liquid Effluents ........................... ,... 3/4-12 3/4.2.1 Liquid Effluents: Concentration . 3/4-12 3/4.2.2 Dose - Liquids . . . ... 3/4-16 3/4.2.3 Liquid Radwaste Treatment ......................... 3/4-17 3/4.3 Radioactive Gaseous Effluents . 3/4-18 3/4.3.1 Gaseous Effluents: Dose Rate .3/4-18 3/4.3.2 Gaseous Effluents: Dose from Noble Gases .3/4-21 3/4.3.3 Gaseous Effluents: Dose from I-131, I-133, Tritium, and Radionuclides in Particulate Form ,, 3/4-22 3/4.3.4 Gaseous Radwaste Treatment .................................. 3/4-23 314.3.5 Ventilation Exhaust Treatment ..................... ...................................... 3/4-24 3/4.3.6 Primary Containment .3/4-25 3/4.3.7 Steam Jet Air Ejector (SJAE) .3/4-26 3/4.4 Total Dose .............................. 3/4-27 3/4.4.1 Total Dose ............ ,.,... 3/4-27 3/4.5 Radiological Environmental Monitoring .3/4-28 3/4.5.1 Radiological Environmental Monitoring Program ............................. 3/4-28 3/4.5.2 Land Use Census .. 3/4-37 3/4.5.3 Intercomparison Program .. 3/4-38 3/4.6 Effluent and Environmental Control Bases .. 3/4-39 5.0 METHODS TO CALCULATE OFF-SITE LIQUID CONCENTRATIONS... 5-1 5.1 Method to Determine F ENO and Cl NG .5-1 5.2 Method to Determine Radionuclide Concentration for Each Liquid Effluent Pathway .......... 5-3 Revision 30 R Date 10/30/02 -Iv-

5.2.1 Sanple Tanks Pathways ........................ 5-3 5.2.2 Service Water Pathway .5-3 5.2.3 Circulating Water Pathway .------------.----;------- . .-...

...-.. 5-4 6.0 OFF SITE DOSE CALCULATION METHODS . . ...........................6-1 6.1 Introductory Concepts .............................................. 6-2 6.2 Method to Calculate the Total Body Dose from Liquid Releases ........................ 6-4 6.3 Method to Calculate Maximum Organ Dose from Liquid Releases .................. 6-10 6.4 Method to Calculate the Total Body Dose Rate from Noble Gases .................. 6-13 6.5 Method to Calculate the Skin Dose Rate from Noble Gases ............................. 6-18 6.6 Method to Calculate the Critical Organ Dose Rate from Iodines, Tritium and Particulates with T1/2 Greater Than 8 Days .............................. 6-25 6.7 Method to Calculate the Ganma Air Dose from Noble Gases .......................... 6-29 6.8 Method to Calculate the Beta Air Dose from Noble Gases ............................... 6-33 6.9 Method to Calculate the Critical Organ Dose from Iodines Tritium and Particulates .............................................. 6-37 6.10 Receptor Points and Annual Average Atmospheric Dispersion Factors for Important Exposure Pathways .............................................. 6-44 6.11 Method to Calculate Direct Dose From Plant Operation ................................... 6-50 6.12 Cumulative Doses .............................................. 6-59 7.0 ENVIRONMENTAL MONITORING PROGRAM .7-1 8.0 SETPOINT DETERMIATIONS .8-1 8.1 Liquid Effluent Instrumentation Setpoints .. 8-2 8.2 Gaseous Effluent Instrumentation Setpoints . 8-11 9.0 LIQUID AND GASEOUS EFFLUENT STREAMS, RADIATION MONITORS, AND RADWASTE TREATMENT SYSTEMS .9-1 9.1 In-Plant Radioactive Liquid Effluent Pathways .9-1 9.2 In-Plant Radioactive Gaseous Effluent Pathways ...................................... 9-4 10.0 REPORTING REQUIREMENTS ....................................... 10-1 R. REFERENCES ......... .. R-1 I Revision 30 Date 10/30/02

APPENDIX A: Method I Example Calculations ................................... D,.,..

Deleted APPENDIX B: Approval of Criteria for Disposal of Slightly Contaminated Septic Waste On-Site at Vermont Yankee .........-. i 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 APPENDIX 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 10CFR20.2002 for On-Site Disposal of Cooling Tower Silt .F-I APPENDIX G: Maximum Permissible Concentrations PCs) in Air and Water Above Natural Background Taken from 10CFR20.1 to 20.602, Appendix B ........ , .. G-1 APPENDIX H: . H-1

1) "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.,,,,,,,,,,,,,,,,,..........

H-37 APPENDIX I:.I-

1) "Request to Amend Previous Approval Granted Under 10CFR20.2002 for Disposal of Contaminated Soil", dated September 11, 2000 BVY 00-71 ....... . . . ... I-2

2) "Vermont Yankee Nuclear Power Station - Safety Evaluation for an Amendment to an Approved 10CFR20.2002 Application (TAC No.

MA9972)", dated June 26, 2001, NVY 01-66,.............,,,,,,,,,,...,I-5

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

I Revision 30 Date 10/30/02 -vi-

LIST OF TABLES Number Title Pape 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 1-7 Concentrations 1.1.4 Summary 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, Iodine, and Particulates in Gas Releases and Direct Radiation 1-10 1.1.7 Summary of Methods for Setpoint Deterninations 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 Vermont Yankee for Noble Gas Releases 1-15 1.1.10A Combined Skin Dose Factors Specific for Vermont Yankee Ground Level Noble Gas Releases 1-16 1.1.11 Dose Factors Specific for Vermont Yankee for Liquid Releases 1-17 1.1.12 Dose and Dose Rate Factors Specific for Vermont Yankee for lodines, Tritium, and Particulate Releases 1-18 2.1.1 Definitions 2-2 2.1.2 Summary of Variables 2-3 3.1.1 Liquid Effluent Monitoring Instrumentation 3/4-3 3.1.2 Gaseous Effluent Monitoring Instrumentation 3/4-8 4.1.1 Liquid Effluent Monitoring Instrumentation Surveillance Requirements 3/4-5 4.1.2 Gaseous Effluent Monitoring Instrumentation Surveillance Requirements 3/4-10 4.2.1 Radioactive Liquid Waste Sampling and Analysis Program 3/4-13 Revision 30 Date 10/30/02 -vii-

LIST OF TABLES (continued)

Number Title Page 4.3.1 Radioactive Gaseous Waste Sampling and Analysis Program 3/4-19 3.5.1 Radiological Environmental Monitoring Program 3/4-29 3.5.2 Reporting Levels for Radioactivity Concentrations in Environmental 3/4-34 Samples 4.5.1 Direction Capabilities for Environmental Sample Analysis 314-35 6.2.1 Environmental Parameters for Liquid Effluents at Vermont Yankee 6-8 6.2.2 Usage Factors for Various Liquid Pathways at Vermont Yankee 6-9 6.9.1 Environmental Parameters for Gaseous Effluents at Vermont Yankee 6-41 6.9.2 Usage Factors for Various Gaseous Pathways at Vermont Yankee 643 6.10.1 Atmospheric Dispersion Factors 647 6.10.2 Site Boundary Distances 648 6.10.3 Recirculation Correction Factors 649 7.1 Radiological Environmental Monitoring Stations 74 8.2.1 Relative Fractions of Core Inventory Noble Gases After Shutdown 8-22 I Revision 30 Date 10/30/02 -viii-

LIST OF FIGURES Number Title 7-1 Environmental Sampling Locations in Close Proximity to Plant 7-7 7-2 Environmental Sampling Locations Within 5 km of Plant 7-8 7-3 Environmental Sampling Locations Greater Than 5 km from Plant 7-9 7-4 TLD Locations in Close Proxinity to Plant 7-10 7-5 TlD Locations Within 5 km of Plant 7-11 7-6 TLD Locations Greater than 5 km from Plant 7-12 9-1 Liquid Effluent Streams, Radiation Monitors, and Radwaste Treatment System at Vermont Yankee 9-9 9-2 Gaseous Effluent Streams, Radiation Monitors, and Radwaste Treatment System at Vermont Yankee 9-10 I Revision 30 Date 10/30/02 -ix-

1.0 INTRODUCTION

The ODCM (Off-Site Dose Calculation Manual) provides formal and approved methods for the calculation of off-site concentration, off-site doses, and effluent monitor setpoints in order to comply with the Vermont Yankee Control Limits which implement the program requirements of Technical Specification 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 10CFR50 (Regulatory Guide 1.109) unless otherwise noted in the text.

Demonstration of compliance with the dose limits of 40CFR190 (see Control 3.4.1) will be considered as demonstrating compliance with the 0.1 rem limit of 10CFR20.1301(a)(1) for members of the public in unrestricted areas (Reference 56 FR 23374, third column.)

It shall be the responsibility of the Chemistry Manager and Radiation Protection Manager to ensure that the ODCM is used in the performance of the surveillance requirements of the appropriate portions of ODCM Controls. The administration of the program for the onsite disposal of slightly contaminated waste, as described in Appendices, is also the responsibility of the Chemistry Manager.

All changes to the ODCM must be reviewed by PORC and approved by the Plant Manager, 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 which the change(s) was made effective. The plant's Document Control Center (DCC) shall maintain the current version of the ODCM and issue under controlled distribution all approved changes to it.

Revision 30 Date 10/30/02 1-1

1.1 Summary of Methods, Dose Factors, Limits, Constants, and Radiological Effluent Control Cross-References This section summarizes the dose calculation methods. The concentration and-setpoint methods are also summarized in Table 1.1.2 through Table 1.1.7, as well as the Method I Dose equations. Where more accurate dose calculations are needed use the Method II for the appropriate dose as described in Sections 6.2 through 6.9 and 6.11. The dose factors used in the equations are in Tables 1.1.10 through 1.1.12 and the Regulatory Limits are summarized in Table 1.1 .1.

A cross-reference of old Technical Specification sections to the new ODCM sections containing the equivalent Controls is presented in Table 1.1.8.

Special definitions and equation variables used in the ODCM are in Tables 2.1.1 and 2.1.2.

Revision 30 Date 10/30/02 1-2

TABLE 1.1.1 Summarv of Radiological Effluent Controls and Implementing Equations Control Categorv Method (l) TLirnit 3.2.1 Liquid Effluent Concentration Sum of the Fractions of Eq. 5-1 *10 Effluent Concentration Limits

[Excluding Noble Gases]

Total Noble Gas Concentration Eq. 5-2 *2 x 10 4 liCi/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 c5 mrem in a qtr.

<10 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.

3.3.1 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 3.3.1 (Continued) Organ Dose Rate from Iodines, Eq. 6-16 *1500 mrem/yr.

Tritium and Particulates with Eq. 6-40 T 1 2>8 Days I Revision 30 Date 10/30/02 1-3

TABLE 1.1.1 (Continued)

Summary of Radiological Effluent Controls and Implementing Equations Control Category Method () Limit 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 <10 mrad in a yr.

Beta Air Dose from Noble Eq. 6-23 *10 mrad in a qtr.

Gases Eq. 6-43 < 20 mrad in a yr.

3.3.3 Gaseous Effluents Dose from lodines, Organ Dose from 1-13 1, Eq. 6-25

7.5 mrem in a qtr.

Tritium, and Particulates 1-133, Tritium, and Eq. 6-44 <15 mrem in a yr.

Particulates with T1/2>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 <25 mrem in a yr.

Thyroid Dose *75 mrem in a yr.

. . i~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~

I Revision 30 Date 10/30/02 1-4

TABLE 1.1.1 (Continued)

Summary of Radiological Effluent Controls and Implementing Equations Control Categorv Method(l) Limit 3.1.1 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 and AOG Offgas Alarm/Trip Setpoint for Total Eq. 8-9 Control 3.3.1.a System Noble Gas Activity Body Dose Rate (Total Body)

Monitors Alarm/Trip Setpoint for Skin Eq. 8-10 Control 3.3.1.a Dose Rate (Skin)

SJAE Noble Gas Activity Alarm Setpoint Eq. 8-21 T.S. 3.8.K.1 and Monitors Control 3.3.7 (1) 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.

I Revision 30 Date 10/30/02 1-5

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 5.1 Effluent Concentrations in Liquids FENG = Cpi <10

[Except Noble Gases]

5-2 Total Activity of Dissolved and 5.1 Entrained Noble Gases from all (Nm]L)=CN 2E-04 Station Sources I Revision 30 Date 10/30/02 1 1-6

TABLE 1.1.3 Summarv of Methods to Calculate Off-Site Doses from Liquid Concentrations Equation

Reference Number Category Equation Section 6-1 Total Body Dose Dtb(mrem) = X Qi DFLit 6.2.1 6-3 Maximum Organ Dose Dmo (mrem) = Qi DFLm 6.3.1 I Revision 30 Date 10/30/02 1-7

TABLE 1.1.4 Sumrarv of Methods to Calculate Dose Rates Equation Reference Number Category Equation Section -

6-5 Total Body Dose Rate from 6.4.1 Noble Gases Released from R r )= I061YQ&DPB Stack 6-39 Total Body Dose Rate from 6.4.1 Noble Gases Released from R tbg( = 6.4Z QjGLDFB-Ground 6-7 Skin Dose Rate from Noble 6.5.1 Gases Released from Stack R sns( = YlQ frDFi' R~~~~~)=sGDI sEg(

6-38 Skin Dose Rate from Noble 6.5.1 Gases Released from Ground

'dSt (m ) = s p Sk 6-16 Ciitical Organ Dose Rate from 6.6.1 Stack Release of I-131, 1-133, Tritium, and Particulates with T 112 >8 Days 6-40 Critical Organ Dose Rate from cog (n ) = FQ DFGP 6.6.1 Ground Level Release of 1-13 1, I-133, Tritium, and Particulates with Tmr >8 Days I Revision 30 Date 10/30/02 1-8

TABLE 1.1.5 Summary of Methods to Calculate Doses to Air from Noble Gases Equation Reference Number Category Equation Section 6-21 Gamma Dose to Air from D (mrad)= 0.019 Q TD 6.7.1 Noble Gases Released from I ~~~~~~~~~~~~~~~~~~~i Stack 6-41 Gamma Dose to Air from D, (mrad)= 0.20ZQOLDjT 6.7.1 Noble Gases Released from Ground Level 6-23 Beta Dose to Air from Noble D. (mrad) = 0.033EQs T DFI 6.8.1 Gases Released from Stack 6-43 Beta Dose to Air from Noble DO (mrad) = 1.12EQGLDFP 6.8.1 Gases Released from Ground Level Revision 30 Date 10/30/02 1-9

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 1'lumber Category Equation Section 6-25 Dose to Critical Organ from Dcos(Iirein) = XQsPDFGS 6.9.1 Stack Release of 1-131, I-133, Tritium, and Particulates 6-44 Dose to Critical Organ from Dcog(rnrem)= XQtI2DFG 6.9.1 Ground Level Release of I-131, 1-133, Tritium, and Particulates Direct Dose Turbine Building I 6-27a 6.11.1 Dd = KN-16 Ktissue Kcalib DMSLRM I 6-27b n 6.11.1 DMs, = E[E(Rj,)Im],Ati 1=1 j-1 I 6-27c At = t - ti 6.11.1 I 6-27d 6.11.1 AT =ZAt i=1 North Warehouse 6-28 Shielded End Ds = 0.25xR5 6.11.2 6-29 Unshielded End D= 0.53xRu 6.11.2 LLW Storage Pad 6-30 Direct Line (Module Short Dd = 0.28xRd Xfd 6.11.3 Side Out) 6-31 Direct Line (Module Long Dd = 0.39xRd Xfd 6.11.3 Side Out) 6-32 Skyshine (Resin Liners) DsxR = 0.016xSKR XfSK 6.11.3 6-33 Skyshine (DAW) DsKD= 0.015xRsKD XfSK 6.11.3 6-34 Resin Liner Transfer DTm, = 0.0025xRTf XTT,= 6.11.3 (Unshielded) 6-35 Intermodular Gap Dose DGap = 2.44E-2 x WGSP XARLXfGap 6.11.3 I Revision 30 Date 10/30/02 _ 1-10

TABLE 1.1.7 Summars of Methods for Setpoint Determinations Equation Reference Number Category Equation Section 8-1 Liquid Effluents:

Liquid Radwaste Discharge RL Rsp DF S i Cmi (CpS) = DFm 8.1.1.1 Monitor (17/350)

Gaseous Effluents:

Plant Stack (RR-108-1A, RR-108-1B) and AOG Offgas System (3127, 3128)

Noble Gas Activity Monitors 8-9 Total Body 8.2.1.1 R'bpt (cpm) = 818S 1 1 F DFB, 8-10 Skin 8.2.1.1 R": spt (cpm) =300S1,3000' 8-21 SJAE Noble Gas Activity 8.2.2.1 RsAE(mR/hr)

Rspt = 1.6E 05 F1 FE+OSg Monitors (17/150A, 17/150B)

I Revision 30 Date 10/30/02 1-11

TABLE 1.1.8 Effluent and Environmental Controls Cross-Reference Control Topic Original Technical Revised ODCM Specification Section Control Section INSTRUMENTATION Radioactive Liquid Effluent Instrumentation 3/4.9.A 3/4.1.1 Effluent instrumentation list Table 3.9.1 Table 3.1.1 Instrument surveillance requirements Table 4.9.1 Table 4.1.1 Radioactive Gaseous Effluent Instrumentation 3/4.9.B 3/4.1.2 Effluent instrumentation list Table 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 program Table 4.8.1 Table 4.2.1 Dose - Liquids 3/4.8.B 314.2.2 Liquid Radwaste Treatment 3/4.8.C 3/4.2.3 RADIOACTIVE GASEOUS EFFLUENTS Dose Rate 3/4.8.E 3/4.3.1 Gaseous waste sampling & analysis program Table 4.8.2 Table 4.3.1 Dose from Noble Gases 3/4.8.P 3/4.3.2 Dose from I-13 1, I-133, Tritium and Radionuclides 3/4.8.G 3/4.3.3 in Particulate Form Gaseous Radwaste Treatment 3/4;8.H 3/4.3.4 Ventilation Exhaust Treatment 3/4.8.1 3/4.3.5 Primary Containment 3/4.8.L 3/4.3.6 Steam Jet Air Ejector 3/4.8.K* 3t4.3.7*

TOTAL DOSE Total Dose 3/4.8.M 3/4.4.1 RADIOLOGICAL ENVIRONMENTAL MONITORING Radiological Environmental Monitoring Program 3/4.9.C 3/4.5.1 Listing of required monitoring criteria Table 3.9.3 Table 3.5.1 Reporting levels for radioactivity in samples Table 3.9.4 Table 3.5.2 Detector capability for environmental analysis Table 4.9.3 Table 4.5.1 Land Use Census 3/4.9.D 3/4.5.2 Specification 3/4.8.K remains in plant Technical Specifications and is duplicated in ODCM Control 3/4.3.7.

I Revision 30 Date 10/30102 1-12

TABLE 1.1.8 (Continued)

Effluent and Environmental Controls Cross-Reference Control Topic Original Technical *Revised ODCM Specification Section Control Section Intercomparison 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 Report 6.7.C.1 10.1 Environmental Radiological Monitoring 6.7.C.3 10.2 Special Reports 6.7.C.2 10.3 Major Changes to Radioactive Liquid, Gaseous, and 6.14 10.4 Solid Waste Treatment Systems I Revision 30 Date 10/30/02 1-13

TABLE 1.1.9 (Deleted)

I Revision 30 Date 10/30/02 1-14

TABLE 1.1.10 Dose Factors Specific for Vermont Yankee for Noble Gas Releases Radionuclide Gamma Beta Skin Combined Skin Beta Air Gamma Air Total Body Dose Factor Dose Factor Dose Factor Dose Factor Dose Factor (Stack Release)

Cnirem-DFBi 3

pCi- yr)J DFS; (mrem-m 3 t pCi-yr )

DF,,

(mrem- sec tt Ci- yr J DFi' (mrad-m3) t pci- yr )

DF (mrad-t m3 pCi- yr)

Ar-41 8.84E-03k 2.69E-03 9.12E-03 3.2813-03 9.301-03 Kr-83m 7.56E-08 1.31E-05 2.8813-04 1.93E-05 Kr-85m 1.17E-03 1.46E-03 2.35E-03 1.97E-03 1.2313-03 Kr-85 1.6113-05 1.34F303 1.41E-03 1.95E-03 1.72E-05 Kr-87 5.92E-03 9.73E-03 1.431-02 1.03E-02 6.1713-03 Kr-88 1.47E-02 2.37F,03 1.28E-02 2.93E-03 1.5213-02 Kr-89 1.6613-02 1.01E-02 2.23E-02 1.0613-02 1.7313-02 Kr-90 1.5613-02 7.29E-03 1.8713-02 7.83E-03 1.63E-02 Xe-131m 9.1513-05 4.76E-04 6.01E-04 1.11E-03 1.56E-04 Xe-133m 2.51E-04 9.94E-04 1.26E-03 1.48E-03 3.27E-04 Xe-133 2.94E-04 3.0613-04 5.58E-04 1.05E-03 3.53E-04 Xe-135m 3.12E-03 7.11E-04 3.0213-03 7.39E-04 3.36E-03 Xe-135 1.81E-03 1.86E-03 3.241-03 2.4613-03 1.92E'03 Xe-137 1.42E-03 1.22E-02 1.3713-02 1.27E-02 1.5113-03 Xe-138 8.8313-03 4.13E-03 1.06E-02 4.75E-03 9.21E-03

8.84E-03 = 8.84 x 10-3 I Revision 30 Date 10/30/021 1-15

TABLE 1.1.1OA Combined Skin Dose Factors Specific for Vermont Yankee Ground Level Noble Gas Releases Radionuclide DF n rem- see Fci-R y J Ar-41 1.61E-01 Kr-83M 1.38E04 Kr-85M 6.02E-02 Kr-B5 4.73E-02 Kr-87 3.86E-01 Kr-88 1.92E-01 Kr-89 4.79E-01 Kx-90 3.73E-01 Xe-13 1M 1.79E-02 Xe-133M 3.72E-02 Xe-133 1.33E-02 Xe-135M 4.90E-02 Xe-135 7.92E-02 Xe-137 4.40E-01 Xe-138 2.11E-01 I Revision 30 Date 10/30/02 1-166

TABLE 1.1.11 Dose Factors Specjfic for Vernont Yankee for Liquid Releases Total Body Maximum Organ Dose Factor Dose Factor Radionuclide DiLb( nrem ) DFLim4_r _

H-3 2.06E-04 2.06E-04 Na-24 3.38E-02 3.38E-02 Cr-51 3.1OE-04 6.96E-02 Mn-54 2.08E-Ol 3.OOE+00 Mn-56 8.53E-06 5,29E-03 Fe-55 4.18E-02 2.54E-01 Fe-59 2.49E-01 1.84E+00 Co-58 5.97E-02 4.34E-01 Co-60 2.13E-01 1.28E+OO Zn-65 8.06E+OO 1.64E+0 1 Sr-89 2.55E-01 8.91E+O0 Sr-90 4.23E+O1 1.67E+02 Zr-95 4.21E-04 1.36E-01 Mo-99 4.79E-03 4.51E-O2 Tc-99m 5.04E-06 2.33E-04 Ag- 10 6.90E-03 7.02E-01 Sb-124 8.44E-03 2.22E-01.

Sb-125 7.52E-03 1.15E-01 1-131 2.57E-02 1.47E+01 I-132 3.1OE-06 1.29E-04 1-133 3.3 1E-03 1.63E+00 1-135 3.16E-04 5.90E-02 Cs-134 1.28E+02 1.60E+02 Cs-137 7.5SE+0 1.21E+02 Ba-140 4.08E-03 9.72E-02 Ce-141 2.3 lE-05 4. 1OE-02 W-187 1.18E-02 8.90E+OO I Revision 30 Date 10/30/02 1-17

TABLE 1.1.12 Dose and Dose Rate Factors Specific for Vermont Yankee for lodines, 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 DFGsi,o( Ci ) DFG(mrem-sec3 DVG (rfrem DFG& mrem-sec)

DF yr- tCi H-3 3.13E-04 9.87E-03 1.06E-02 3.34E-01 C-14 1.90E-01 5.99E+00 6.43E-+00 2.03E+02 Cr-51 6.11E-03 2.11E-01 4.16E-02 1.43EI+00 hM-54 7.01E-01 2.77E+01 4.71E+00 1.84E+02 Fe-55 3.17E-01 1.OOE+01 2.05E+00 6.47E+01 Fe-59 6.99E-01 2.32E1+01 4.60E+00 1.52E+02 Co-57 2.18E-01 8.23E+00 1.41E+00 5.33E+/-01 Co-58 3.62E-01 1.30E+0l 2.39E+00 8.52E-+01 Co-60 7.63E+00 3.41E+02 4.99E+01 2.16E+03 Zn-65 3.71E+00 1.20E+02 2.36E+01 7.63E+02 Se-75 2.41E+00 7.76E+01 1.53E+01 4.92E t02 Sn-1 13 I.03E+00 3.25E+01 6.58E+00 2.08E+02 Sr-89 1.14E+01 3.60E+02 7.27E+01 2.29E+03 Sr-90 4.3 1E+02 1.36E+04 2.82E+03 8.89E+04 Zr-95 6.91E-01 2.28E+01 4.5 IE+00 1.49E+02 Sb-124 1.26E+/-00 4.23E+01 8.35E+00 2.79E+02 Sb-125 1.25E+00 4.89E+01 8.01E+00 3.13E+02 1-131 7.71E+01 2.43E+03 5.02E+02 1.58E+04 I-133 8.22E-01 2.59E+01 8.30E+00 2.62E+02 Cs-134 1.58E+01 5.27E+02 1.02E+02 3.37E+03 Cs-137 1.63E+0 1 5.55E+02 1.04E+02 3.53E+03 Ba-140 1.13E-01 3.66E+0 2.1SE+00 6.94E+01 Ce-141 1.70E-01 5.42Ei+00 1.19E+00 3.78E+01 Cc-144 3.85E+00 1.22E+02 2.52E+01 7.98E+02

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

I Revision 30 Date 10130/02 1-18

2.0 DEFINMIONS This section lists definitions (Table 2.1.1) and dose equation variable names (Table 2.1.2) l which are utilized in the VY ODCM.

I Revision 29 Date 1/11/02 2-1

TABLE 2.1.1 Definitions

1. Gaseous Radwaste Treatment System - The Augmented Off-Gas System (AOG) is the gaseous radwaste treatment system which has been designed and installed to reduce radioactive gaseous effluents by collecting primary coolant system off-gases from the primary system and providing for delay or holdup for the purpose of reducing the total radioactivity prior to release to the environment.

2. Hot Standby - Hot standby means operation with the reactor critical and the main steam line isolation valves closed.

3. 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.

4. Instrument Calibration - An instrument 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.

5. 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.

6. 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 practicable to verify operability including alarm and/or trip functions.

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

7. 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 alarm/trip setpoints, and in the conduction of the environmental radiological monitoring program. The ODCM shall also contain (1) the Radioactive Effluent Controls (including the Radiological Environmental Monitoring) Program required by Technical Specification 6.7.D, and (2) descriptions of the information 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.

I Revision 29 Date 1/11/02 2-2

8. Refueling Outage - Refueling outage is the period of time between the shutdown of the unit prior to a refueling and the startup of the plant subsequent to that refueling. or the purpose of designating frequency of testing and surveillance, a refueling outage shall mean a regularly scheduled refueling outage; however, where such outages occur within 8 months of the completion of the previous refueling outage, the required surveillance testing need not be performed until the next regularly scheduled outage.

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. Surveillance Frquency 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.

12. 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 operable. These tests unless otherwise stated in these specifications may be waived when the instrument, component, or system is not required to be operable, but these tests shall be perfonned on the instrument, component, or system prior to being required to be operable.

13. Ventilation Exhaust Treatment System - The Radwaste Building and AOG Building ventilation HEPA 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 HEPA 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.

14. 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.

I Revision 29 Date 1/11/02 2-3

TABLE 2.1.2 Summarv of Variables Variable Definition Units ARL = Total gamma activity contained in a resin liner in Ci storage directly in line with a gap between adjacent storage modules.

liG Concentration at point of discharge to an jiCi/ml unrestricted area of dissolved and entrained noble gas "i" in liquid pathways from all station sources.

CING- Total activity of all dissolved and entrained noble ptCi gases in liquid pathways from all station sources. ml Cdi = Concentration of radionuclide "i" at the point of ACi liquid discharge to an unrestricted area. ml Ci = Concentration of radionuclide "i". pCi cc Cpi = Concentration, exclusive of noble gases, of pCi radionuclide "i" from tank "p" at point of discharge ml to an unrestricted area.

Cmi = Concentration of radionuclide "i" in mixture at the pCi monitor. ml DL = Beta dose to air from stack release. mrad Dati g Beta dose to air from ground level release. mrad Dy = Gamma dose to air from stack release. mrad DYatrg = Gamma dose to air from ground level release. mrad Dcos = Dose to critical organ from stack release. mrem Dcog = Dose to the critical organ from ground level release. mremn Dd = Direct dose Curbine Building). mrem Revision 29 Date 1111102 2-4

TABLE 2.1.2 (Continued)

Summary of Variables Variable Definition Units Rd Dose rate at 3 feet from unobstructed side of storage mrem module facing site boundary. hr Direct dose at site boundary per unobstructed storage mrem module (short end). yr- module Dds Direct dose at site boundary per unobstructed storage mrem module (long side). yr - module Gamma dose to air, corrected for finite cloud. mmd D0 ap Intermodular gap dose projected to the maximum site mrem boundary location from resin waste not directly yr shielded by DAW modules.

Din Dose to the maximum organ. rrem Ds Dose to skin from beta and gamma. mrem Rs Dose rate at 1 meter from source in shielded end of rnrem North Warehouse. hr Ds Annual dose at site boundary from fixed sources in rrem shielded end of North Warehouse. yr RsKD Maximum dose rate at 3 feet over top of DAW in a mrem storage module. hr RSKR Maximum dose rate at 3 feet over top of each resin mrem liner in a storage module. hr DsD Skyshine dose at the site boundary from DAW in mrem storage modules (unobstructed top surfaces). yr - module Skyshine dose at the site boundary from resin liners in mrem storage modules (unobstructed top surfaces). yr-liner I Revision 29 Date 1/11/02 2-5

TABLE 2.1.2 (Continued)

Summary of Variables Variable Definition Units Dtb = Dose to the total body mrem KN6 (L) = The direct dose conversion factor for N-16 scatter from the nrem turbine hall to Location (L) MWCh RTran = Dose rate at contact from the unshielded top surface of rad resin liner. hr DT= = Dose at the site boundary from unshielded movement of mrem resin liner between transfer cask and storage module.

Ru = Dose rate at 1 meter from source in unshielded end of mrem North Warehouse. hr Du = The annual dose at the site boundary from fixed sources in nrem the unshielded end of North Warehouse. hr DF = Dilution factor. ratio DFmin = Minimum allowable dilution factor. ratio DFc = Composite skin dose factor. mrem - sec pci -yr DFBi = Total body gamma dose factor for nuclide "i". mrem -m 3 pCi - yr DFBC - Composite total body dose factor. rnrem -m 3 pCi - yr DFLitb Site-specific, total body dose factor for a liquid release of mrem nuclide "i". ci DFLmo = Site-specific, maximum organ dose factor for a liquid mrem release of nuclide "i". ci I Revision 29 Date 1/11/02 2-6

TABLE 2.1.2 (Continued)

Summary of Variables Variable Definition Units DFGsico = Site-specific, critical organ dose factor for a stack rnrem gaseous release of nuclide "i". ci DFG'sico = Site-specific, critical organ dose rate factor for a stack mrem - sec gaseous release of nuclide "i". Ci - yr DFGgico = Site-specific, critical organ dose factor for a ground nrem level gaseous release of nuclide "i". ci DFG'gico = Site-specific, critical organ dose rate factor for a rnrem-sec ground level gaseous release of nuclide "i". Ci- yr DFSi = Beta skin dose factor for nuclide "i". mrem - m 3 pCi - yr

= Combined skin dose factor for nuclide "i" from a stack mrem - sec release i - yr DF'ig = Combined skin dose factor for nuclide "i" from a mrem -sec ground level release. jiCi - yr DPy = Gamma air dose factor for nuclide "i". mrad-m 3 pCi - yr DFi = Beta air dose factor for nuclide "i". rnrad - m3 pCi - yr Rcos = Critical organ dose rate due to iodines and particulates inrem released from stack. yr Rcog = Critical organ dose rate due to iodines and particulates rnrem released from ground. yr

= Skin dose rate due to stack release of noble gases. rnrem yr Rsking = Skin dose rate due to ground release of noble gases. mrem yr I Revision 29 Date 1/11/02 2-7

TABLE 2.1.2 (Continued)

Summary of Variables Variable Definition Units R tbs = Total body dose rate due to noble gases from stack release. mrem yr tbg = Total body dose rate due to noble gases from ground level mrem release. yr D/Q = Deposition factor for dry deposition of elemental radioiodines 1 and other particulates. 2 E = Gross electric output over the period of interest. MWeh fd Fraction of a year that a storage module is in use with an fraction unobstructed side oriented toward west site boundary.

fGap = Fraction of a year that the intermodular gap is not shielded. fraction fSK = Fraction of a year that a storage module is in use with an fraction unobstructed top surface.

Fd = Flow rate out of discharge canal. gpm Fm = Flow rate past liquid radwaste monitor. gpm F = Flow rate past gaseous radwaste monitor. cc sec FENG = Sum of the fractions of combined effluent concentrations in fraction liquid pathways (excluding noble gases).

ECLI = Annual average effluent concentration limit for radionuclide "i" xCi (10CFR20.1001-20.2401, Appendix B, Table 2, Column 2) cc Qi = Release for radionuclide "i" from the point of interest. curies I Revision 29 Date 1/11/02 2-8

TABLE 2.1.2 (Continued)

Summary of Variables Variable Definition Units Q,

= Release rate for radionuclide "i" at the point of interest. [tCi sec ST = The noble gas radionuclide "i" release rate at the plant stack. jiCi sec QGL

= The noble gas radionucide "i" release rate from ground level. jICi sec QsJAE = The noble gas radionuclide "i" release rate at the steam jet air puCi ejector. see QAoc = The noble gas radionucide "i" release rate at the exhaust of the pCi Advanced Off-Gas System sec (TP = The iodine, tritium, and particulate radionuclide "i" release rate jiCi from the plant stack. sec

&GLP = The iodine, tritium, and particulate radionuclide "i" release rate jCi from ground level. sec QsT = The release of noble gas radionuclide "i" from the plant stack. curies QGL The release of noble gas radionuclide "i" from ground level. curies QsTP = The release of iodine, tritium, and particulate radionuclide "i" curies from the plant stack.

QGLP = The release of iodine, tritium, and particulate radionuclide "i" curies from ground level.

R Liquid monitor response for the limiting concentration at the cps spt point of discharge.

I Revision 29 Date 1/11/02 2-9

TABLE 2.1.2 (Continued)

Summarv of Variables Variable Definition Units kint Response of the noble gas monitor at the limiting skin dose rate. cpm Rt t Response of the noble gas monitor to limiting total body dose rate. cpm SF Shielding factor. Ratio Sg Detector counting efficiency from the most recent gas monitor cpm ormR/hr calibration. jlCi/cc jiCi/cc Sgi Detector counting efficiency for noble gas "i". cpm or mR/hr ptCi/cc p.Ci/cc Si Detector counting efficiency from the most recent liquid monitor cps calibration. plCiIrl Sli Detector counting efficiency for radionuclide "i". cps iLCi/ml TTM Time that an unshielded resin liner is exposed in the storage pad hours area.

WGap Intermodule gap width between adjacent DAW storage modules inches which shield resin liner storage modules from the west site boundary.

X/QS Annual or long-term average undepleted atmospheric dispersion sec factor for stack release. m3 X/Qg Annual or long-term average undepleted atmospheric dispersion sec factor for ground level release. m3

[XJQ]Y Effective annual or long-term average gamma atmospheric sec dispersion factor. m3

[xJQ]Y Effective annual or long-term average gamma atmospheric sec

£ dispersion factor for a ground level release. 3 Revision 29 Date 1/11/02 2-10

3/4.0 EFFLUENT AND ENVIRONMENTAL CONTROLS This section includes the effluent and environmental controls that were originally part of the Vermont Yankee Technical Specifications. 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.

I Revision 29 Date 1/11/02 3/4-1

3/4.1 INSTRUMENTATION 3/4.1.1 Radioactive Liguid Effluent Instrumentation CONTROLS 3.1.1 The radioactive liquid effluent monitoring instrumentation channel shall be operable in accordance with Control Table 3.1.1 with their alarm setpoints set to ensure that the limits of Control 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 reIease of radioactive liquid effluents monitored by the affected channel or change the setpoint so that it is acceptably conservative or declare the channel inoperable.

b. With one or more radioactive liquid effluent monitoring instrumentation channels inoperable, 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).

Revision 29 Date 1/11/02 3/4-2

TABLE 3. 1.1 Liquid Effluent Monitoring Instrumentation Minimum Channels Operable Notes Gross Radioactivity Monitors not Providing Automatic Termination of Release

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

b. Service Water Discharge Monitor 1 2,4 (RM-17-351)

2. Flow Rate Measurement Devices

a. Liquid Radwaste Discharge Flow Rate 1* 3,4 Monitor (FT-2048544 __

During releases via this pathway I Revision 29 Date 1/11/02 3/4-3

TABLE 3.1.1 NOTATION NOTE 1 - With the number of channels operable less than required by the minimum channels operable 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 valving.

Otherwise, suspend release of radioactive effluents via this pathway.

NOTE 2 - With the number of channels operable less than required by the minimum channels operable requirement, effluent releases via this pathway may continue provided that, at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , grab samples are collected and analyzed for gross radioactivity (beta or gamma) at a lower limit of detection of at least 10-7 rnicrocurie/ml.

NOTE 3 - With the number of channels operable less than required by the minimum channels operable requirement, effluent releases via this pathway may continue provided the flow rate is estimated at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months during actual releases.

Pump performance curves may be used to estimate flow.

NOTE 4 - With the number of channels operable less than required by the minimum channels operable requirement, exert reasonable efforts to return the instrument(s) to operable status prior to the next release.

I Revision 29 Date 1/11/02 3/4-4

TABLE 4.1.1 Radioactive Liquid Effluent Monitoring Instrumentation Surveillance Requirements Instrument Instrument Instrument Functional Instrument Check Source Check Calibration Test

1. 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 no 18 months (1) quarter (2) more than once each month

b. Service Water Discharge Once each Once each Once each Once each Monitor (3) day month 18 months (1) quarter (2)

2. Flow Rate Measurement Devices

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

Monitor

During releases via this pathway.

I Revision 29 Date 1/11/02 314-5

TABLE 4.1.1 NOTATION (1) 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 alarm annunciation occurs if any of the following conditions exists:

(a) Instrument indicate measured levels above the alarm setpoint.

(b) Circuit failure.

(c) Instrument indicates a downscale failure.

(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).

J Revision 29 Date 1/11/02 3/4-6

3/4.1 INSTRUMENTATION 3/4.1.2 Radioactive Gaseous Effluent Instrumentation CONTROLS 3.1.2 The gaseous process and effluent monitoring instrumentation channels shall be operable 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, and Technical Specifications 3.8.J.1 and 3.8.K.1 (Control 3.3.7) are not exceeded.

APPLICAB3IY:

As shown in Table 3.1.2.

ACTION:

a. With a gaseous process or effluent monitoring instrumentation channel alarm/trip setpoint less conservative than a value which will ensure that the limits of Control 3.3.1.a and Technical Specification 3.8.K.1 are met, immediately take actions to suspend the release of radioactive gaseous effluents monitored by the affected channel, or declare the channel inoperable, or change the setpoint so it is acceptably conservative..

b. With less than the minimum number of radioactive gaseous effluent monitoring instrumentation channels operable, 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).

I Revision 29 Date 1/11/02 3/4-7

TABLE 3.1.2 Gaseous Effluent Monitoring Instrumentation Minimum .

Channels Instrument Operable Notes

1. Steam Jet Air Ejector (SJAE)

a. Noble Gas Activity Monitor

1 7, 8, 9 (RM-17-15OAB)

2. Augmented Off-Gas System

a. Noble Gas Activity Monitor Between the Charcoal 1 2, 5, 6, 7 Bed System and the Plant Stack (Providing Alarm and Automatic Termination of Release)

(RAN-OG-3127, RAN-OG-3128)

b. Flow Rate Monitor 1 1,5, 6 (FI-OG-2002, FI-OG-2004, FI-OG-2008)

c. Hydrogen Monitor 1 3,5, 6 (H2AN-OG-2921A/B, H2AN-OG-2922A/B)

3. Plant Stack

a. Noble Gas Activity Monitor 1 5,7,10 (RM-17-156, RM-17-157)

b. Iodine Sampler Cartridge 1 4,5

c. Particulate Sampler Filter 1 4,5

d. Sampler Flow Integrator 1 1,5 (FI-17-156/157)

e. Stack Flow Rate Monitor 1 1,5

(-108-22)___

This instrumentation channel(s) is required to support compliance with Technical Specification 3.8.K.

I Revision 29 Date 1/11/02 3/4-8

TABLE 3.1.2 NOTATION NOTE 1 - With the number of channels operable less than required by the minimum channels operable requirement, effluent releases via this pathway may continue provided the flow rate is estimated at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

NOTE 2 - With the number of channels operable less than required by the minimum channels operable requirement, effluent releases via this pathway may continue for a period of up to 7 days provided that at least one of the stack monitoring systerns is operable and off-gas system temperature and pressure are measured continuously.

NOTE 3 - With the number of channels operable less than required by the rninimum channels operable requirement, operation of the AOG System may continue provided gas samples are collected at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and analyzed within the fQllowing 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , or an orderly transfer of the off-gas effluents from the operating recombiner to the standby recombiner shall be made.

NOTE 4 - With the number of channels operable less than required by the minimura channels operable 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 operable less than required by the minimum channels operable requirement, exert reasonable efforts to return the instrument(s) to operable status within 30 days.

NOTE 6 - During releases via this pathway..

NOTE 7 - The alarm/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 - Minimum channels operable required only during operation of the Steam Jet Air Ejector.

NOTE 9 - With the number of channels operable less than required by the minimum channels operable requirement, gases from the SJAE may be released to the environment for up to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months provided:

I.The AOG System is not bypassed; and 2.The AOG System noble gas activity monitor is operable.

NOTE 10- With the number of channels operable less than required by the minimum channels operable requirement, effluent releases via this pathway may continue provided grab samples are taken at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and these samples are analyzed for gross activity within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Revision 29 Date 1/11/02 314-9

TABLE 4.1.2 Gaseous Effluent Monitoing Instrumentation Surveillance Requirements Instrument Instrument Instrurment Instrument Check Source Check Calibration Functional Test

1. Steam Jet Air Ejector (SJAE)

a. Noble Gas Activity Once each Once each Once each Once each Monitor + day** month 18 months (3) quarter (2)

2. Augmented Off-Gas System

a. Noble Gas Activity Once each Once each Once each Once each Monitor day* month 18 months (3) quarter (1)

b. Flow Rate Monitor Once each Not Once each Not Applicable day* Applicable 18 months

c. Hydrogen Monitor *' Once each Not Once each Once each day* Applicable quarter (4) month

3. Plant Stack

a. Noble Gas Activity Once each Once each Once each Once each Monitor day month 18 months (3) quarter (2)

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

c. System Flow Rate Once each Not R(a) Once each Monitor day Applicable quarter

During releases via this pathway.

- During operation of main condenser SJAE.

+ This instrumentation channel(s) is required to support compliance with Technical Specification 3.8.K (same as Control 3.3.7).

+- This instrumentation channel(s) is required to support compliance with Technical Specification 3.8.J.

(a) R = once each refueling cycle.

Revision 29 Date 1/11/02 3/4-10

TABLE 4.1.2 NOTATION (1) The Instrument Functional Test shall demonstrate that the instrument will provide an isolation signal to the system logic under the following conditions:

(a) Instrument indicates measured levels above the alarm setpoint.

(b) Circuit failure.

(c) Instrument indicates a downscale failure.

(d) Instrument controls not set in operate mode.

(2) The Instrument Functional Test shall also demonstrate that Control Room alarm annunciation occurs when any of the following conditions exist:

(a) Instrument indicates measured levels above the alarm setpoint.

(b) Circuit failure.

(c) Instrument indicates a downscale failure.

(d) Instrument controls are not set in operate mode.

(3) The Instrument Calibration for radioactivity measurement instrumentation 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 permit calibrating the system over its normal operating range of rate capabilities.

(4) The Instrument Calibration shall include the use of standard gas samples (high range and low range) containing suitable concentrations, hydrogen balance air, for the detection range of interest per Technical Specification 3.8.J.1.

I Revision 29 Date 1/11/02 3/4-11

314.2 RADIOACTIVE LIOIQUD EFFLUENTS 3/4.2.1 Liguid Effluent Concentration CONTROLS 3.2.1 The concentration of radioactive material in liquid effluents released to Unrestricted Areas shall be limited to 10 times the concentrations specified in Appendix B to OCFR Part 20.1001 - 20.2402, Table 2, Column 2 for radionuclides other than noble gases and 2x10 4 uCi/ml total activity concentration for all dissolved or entrained noble gases.

APPLICABTY:

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 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.

I Revision 29 Date 1/11/02 3/4-12

TABLE 4.2.1 Radioactive Liquid Waste 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 (b) release Each release Each Emitters (d)

Batch Batch I-131 1 x 10-6 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 H-3 1 x 10-5 release Each month Batch Composite (C)

Gross Alpha 1 x 10-7 Prior to each Once per Sr-89, Sr-90 5 x i-8 release Each quarter Batch Composite (C)

Fe-55 1 x 10-6 I Revision 29 Date_1/11/02 3/4-13

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% probability of falsely concluding that a blank observation represents a "real" signal.

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

LLD= 4.66*Sb E*V*K*Y*e'*At 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 disintegrationslninutelmicrocurie or 2.22 disintegration/minute/picocurie as applicable Y = the fractional radiochemical yield (when applicable)

X = the radioactive decay constant for the particular radionuclide (minute)

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

Typical values of E, V, Y and At 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 nucIides, 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.

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.

I Revision 29 Date 1/11/02 3/4-14

TABLE 4.2.1 NOTATION (Cont'd)

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 eritters 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-137, Ce-141, and Ce-144. This list does not mean that only these nuclides are to be detected and reported. Other peaks which are measurable and identifiable, together with the above nuclides, shall also be identified and reported. Nuclides which are below the LLD for the analyses should not be reported as being present at the LLD level, but as "not detected". When unusual circumstances result in LLDs higher than required, the reasons shal be documented in the Radioactive Effluent Release Report.

I Revision 29 Date 1/11/02 3/4-15

3/4.2 RADIOACTIVE LIOUID EFFLUENTS 3/4.2.2 Dose - Liquids CONIROLS 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 lirited 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 calendar year:

less than or equal to 3 mrem to the total body, and less than or equal to 10 mrem to any organ.

APPLICABILYTY:

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 deternined in accordance with the methods in the ODCM at least once per month if releases during the period have occurred.

I Revision 29 Date 1/11/02 3/4-16

314.2 RADIOACTIVE LIOUID 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 I discharge when the projected doses due 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 nrem to any organ.

APPIJCABILITY:

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 Commission within 30 days, a Special Report that includes the information detailed in ODCM Section 10.3.1.

SURVEILANCE REQUIREMENTS 4.2.3.a See Control 4.2.2.

4.2.3.b The liquid radwaste treatment system schematic is shown in ODCM Figure 9.1.

I Revision 29 Date 1/11/02 3/4-17

314.3 RADIOACTIVE GASEOUS EFFLUENTS 3/4.3.1 Gaseous Effluents 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 mren/yr to the total body and less than or equal to 3,000 mrem/yr to the sldn, and

b. For Iodine-131, Iodine-133, tritium and radionuclides in particulate form with half-lives greater than 8 days; less than or equal to 1,500 mrenlyr to any organ.

APPLICABILITY:

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.

SURVEILLFANCE REQUIREMENTS 4.3.1.a The dose rate due to noble gases in gaseous effluents shall be determined to be within the lirits of Control 3.3.1 in accordance with the methods in the ODCM.

4.3.1.b The dose rate due to Iodine-131, Iodine-133, 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.

I Revision 29 Date 1/11/02 3/4-18

TABLE 4.3.1 Radioactive Gaseous Waste Sampling And Analysis Progrm Lower Limit Minimum of Detection Gaseous Release Sampling Analysis Type of Activity (LLD)

Type Frequency Frequency Analysis (uCi/ml) (a)

A. Steam Jet Air Once per week Once per week Xe-138, 1 x 104 Ejector Grab Sample Xe-135, Xe-133, Kr-88, Kr-87, Kr-85M B. Containment Prior to each Prior to each Principal 1 x 10 Purge release/ release/ Each Gamma Each Purge Grab Purge Emitters ( and Sample for 1-131 Particulates C. Main Plant Once per month (c) Once per month (c) Principal 1 x 104 Stack Grab Sample Garnma Emitters (d)

H-3 x1lo, Continuous Once per week °b I-131 ( 1 x 10-12 Charcoal Sample Continuous (e) Once per week (b) Principal 1 x 10-,

Particulate Gamma Sample Emitters (cg) and 1-131 Continuous (e) Once per month Gross Alpha 1 x 10-1 Composite Particulate Sample Continuous (e Once per quarter Sr-89, Sr-90 1 x o-Composite Particulate Sample Continuous Noble Gas Noble Gases 1 x lo Monitor Gross Beta or Gamma I Revision 29 Date 1/11/02 3/4-19

TABLE 4.3.1 NOTATION

a. See footnote a. of Table 4.2.1.

b. Samples shall be changed at least once per 7 days and analyses shall be completed within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months after removal from samplers. Sampling shall also be performed at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for at least 7 days following each shutdown, startup or thermal power change exceeding 25% of rated thermal power in one hour, and analyses shall be completed within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months of changing the samples. When samples collected for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months are analyzed, the corresponding LLD s may be increased by a factor of 10. This requirement to sample at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for 7 days applies only if: (1) analysis shows that the dose equivalent I-131 concentration in the primary coolant has increased more than a factor of 3 and the resultant concentration is at least 1 x 10-1 pCi/ml; and (2) the noble gas monitor shows that effluent activity has increased more than a factor of 3.

c. Sampling and analyses shall also be performed following shutdown, startup, or a thermal power change exceeding 25% of rated thermal power per hour unless: (a) analysis shows that the dose equivalent I-131 concentration in the primary coolant has not increased more than a factor of 3 and the resultant concentration is at least 1 x 10-1 j.Ci/rnl; and (2) the noble gas monitor shows that effluent activity has not increased more than a factor of 3.

d. The principal gamma emitters for which the ILD specification will apply are exclusively the following radionuclides: Kr-87, Kr-88, Xe-133, Xe-133m, Xe-135 and Xe-138 for gaseous emissions, and Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, Cs-134, Cs-137, Ce-141 and Ce-144 for particulate emissions. This list does not mean that only these nuclides are to be detected and reported. Other peaks which are measurable and identifiable, together with the above nuclides, shall also be identified and reported.

Nuclides which are below LLD for the analyses should not be reported as being present at the LLD level for that nuclide, 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. The ratio of the sample flow 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. The gaseous waste sampling and analysis program does not explicitly require sampling and analysis at a specified LID to determine the 1-133 release. Estimates of I-133 releases shall be determined by counting the weekly charcoal sample for I-133 (as well as I-13 1) and assume a constant release rate for the release period.

g. Lower Limit of Detection (I D) applies only to particulate form radionuclides identified in Table Notation d. above.

I Revision 29 Date 1/11/02 3/4-20

314.3 RADIOACTIVE GASEOUS EFFLUENTS 3/4.3.2 Dose - Noble Gases CONTROLS 3.3.2 The air dose due to noble gases released in gaseous effluents 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 5 mrad for gamma radiation, and less than or equal to 10 mrad for beta radiation, and

b. DuTing any calendar year:

less than or equal to 10 mrad for gamma radiation, and less than or equal to 20 mrad for beta radiation.

APPLICABUIrY:

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 10.3.2, 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.

I Revision 29 Date 1111/02 3/4-21

3/4.3 RADIOACTIVE GASEOUS EFFLUENTS 3/4.3.3 Dose - Iodine-131, Iodine-133, Radioactive Material in Particulate Form. and Tritium CONTROLS 3.3.3 The dose to a member of the public from Iodine-131, Iodine-133, tritium, and radionuclides in particulate forn 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 nrem to any organ, and

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

APPLICABILTY:

At all times.

ACTION:

With the calculated dose from the release of Iodine-131, Iodine-133, 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 10.3.2, 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.

SURVERlLANCE 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.

I Revision 29 Date 1/11/02 3/4-22

314.3 RADIOACI'IVE GASEOUS EFFLUENTS 3/4.3.4 Gaseous Radwaste Treatment CONTROLS 3.3.4 The Augmented Off-Gas System (AOG) shall be used in its designed mode of operation to reduce noble gases in gaseous waste prior to their discharge whenever the main condenser steam jet air ejector (SJAE) is in operation.

APPLICABIL1TY:

At all times.

ACTION:

With gaseous radwaste from the main condenser air ejector system being discharged without treatment for more than 7 days, prepare and submit to the Commission within 30 days, a Special Report that includes the information detailed in ODCM Section 10.3.2.

SURVEILLANCE REQUIREMENTS 4.3.4.a The readings of the relevant instrument shall be checked every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the main condenser SJAE is in use to ensure that the AOG is functioning.

4.3.4.b The gaseous effluent treatment system schematic is shown in ODCM Figure 9.2.

I Revision 29 Date 1/11/02 314-23

314.3 RADIOACTIVE GASEOUS EFFLUENTS 3/4.3.5 Ventilation Exhaust Treatment CONTROLS 3.3.5 The AOG and 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.

APPLICABILTY:

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 that includes the information detailed in ODCM Section 10.3.2.

SURVEILLANCE REQUIREMENTS 4.3.5 See Control 4.3.2 for surveillance related to AOG and Radwaste Building ventilation filter system operation.

I Revision 29 Date 1/11/02 314-24

314.3 RADIOACTIVE GASEOUS EFFLUENTS 3/4.3.6 Primary Containment CONIROLS 3.3.6 When the primary containment is to be Vented/Purged, it shall be Vented/Purged through the Standby Gas Treatment System whenever the airborne radioactivity levels in containment of Iodine-131, Iodine-133 or radionuclides in particulate form with half-lives greater than 8 days exceed the levels specified in Appendix B to 10C0FR2.1001 - 20.2402, Table 1, Column 3.

APPLICABILITY:

At all times.

ACTION:

a. With the requirements of Control 3.3.6 not satisfied, immediately suspend all Venting/Purging of the containment.

b. During normal refueling and maintenance outages when primary containment is no longer required, then Control 3.3.3 shall supersede Control 3.3.6.

SURVEILANCE REQUIREMENTS 4.3.6 The primary containment shall be sampled prior to venting/purging per Table 4.3.1, and if the results indicate radioactivity levels in excess of the limits of Control 3.3.6, the containment shall be aligned for venting/purging through the Standby Gas Treatment System. No sampling shall be required if the venting/purging is through the Standby Gas Treatment (SBGT) System.

I Revision 29 Date 1/11/02 3/4-25

314.3 RADIOACTIVE GASEOUS EFFLUENTS 314.3.7 Steam Jet Air Ejector (SJAE) [Duplication of Technical Specification 3/4.8.K.)

CONIROLS 3.3.7 Gross radioactivity release rate from the SJAE shall be linited to less than or equal to 0.16 Cilsec (after 30 minutes decay).

APPLICABILITY:

At all times.

ACTION:

a. With the gross radioactivity release rate at the SJAE exceeding the above limit, restore the gross radioactivity release rate to within its limit within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least Hot Standby within the subsequent 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

b. With the gross radioactivity release rate at the SJAE greater than or equal to 1.5 Ci/sec (after 30-minute decay), restore the gross radioactivity release rate to less than 1.5 Ci/sec (after 30-minute decay), or be in Hot Standby within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

SURVEILLANCE REQUIREMENTS 4.3.7 a. The gross radioactivity release rate shall be continuously monitored in accordance with Control 3.1.2

b. The gross radioactivity release rate of noble gases from the SlAE shall be determined to be within the limit of Control 3.3.7 at the following frequencies by performing an isotopic analysis (for Xe-138, Xe-135, Xe-133, Kr-88, Kr-85m, Kr-87) on a representative sample of gases taken at the discharge.

1. Once per week.

2. Within the 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months following an increase of 25% or 5000 microcuries/sec, whichever is greater, in steady-state activity levels during steady-state reactor operation, as indicated by the SJAE monitor.

I Revision 29 Date 1/11/02 314-26

314.4 TOTAL DOSE 3/4.4.1 Total Dose (40 CFR 190)

CONTROLS 3.4.1 The dose 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 (except the thyroid, which is limited to less than or equal to 75 mrem) over a calendar year.

APPLICABIIlTY:

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 10.3.3.

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.

I Revision 29 Date 1/11/02 314-27

314.5 RADIOLOGICAL ENVIRONMENTAL 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 monitoring program not 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 10.3.4).

SURVEILLANCE REQUIREMENTS 4.5.1 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.

R Revision 29 Date 1/11/02 314-28

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

1. AIRBORNE

a. Radioiodine Samples from 5 Continuous operation Radioiodine canister:

and Particulates locations: of sampler with sample Analyze each sample for collection weekly or I-131.

1 sample from up more frequently as Particulate sampler: Gross valley, within 4 miles required by dust beta radioactivity analysis of Site Boundary. loading. on each sample following (major wind direction) filter change. Composite (by location) for gamma I sample Ifrom from down sample downisotopic . d)at least once per valley, within 4 miles quarter.

of Site Boundary.

(major wind direction) 1 sample each from the vicinity of two nearby communities, within 10 miles of Site Boundary.

1 sample from a control location.

I Revision 29 Date 1/1 1/02 314-29

TABLE 3.5.1 (Cont'd)

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

2. DIRECT 40 routine monitoring Quarterly. Gamma dose, at least RADIATIONb stations as follows: once per quarter.

16 incident response Incident response TLDs stations (one in each in the outer monitoring meteorological sector) locations, de-dose only within a range of 0 to quarterly unless gaseous 4 lag; release Controls were exceeded in period.

16 incident response stations (one in each meteorological sector) within a range of 2 to 8 kmg; the balance of the stations to be placed in special interest areas and control station areas.

I Revision 29 Date 1/11/02 314-30

TABLE 3.5.1 (Cont'd)

Radiological Environmental Monitoring Program Exposure Pathway Number of Sample Sampling and Type and Frequency andfor Sample Locations (a) Collection Frequency of Analysis I

3. WATERBORNE a Surface 1 sample upstream. Monthly grab sample. Gamma isotopic analysis (d)of each sample. Tritium analysis of composite sample at least once per quarter.

1 sample downstream. Composite sample collected over a period of one month

b. Ground 1 sample from within Quarterly. Gamma isotopic(d) and 8 km distance. tritium analyses of each sample.

1 sample from a control Quarterly.

location.

c. Sediient from 1 sample from Semiannually. Gamma isotopic I Shoreline downstream area with analysis (d) feach existing or potential sample.

recreational value.

1 sample from north Semiannually.

storm drain outfall.

I Revision 29 Date 1/11/02 314-3 1

TABLE 3.5.1 (Cont'd)

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

4. INGESTION

a. Milk Samples from milking Semimonthly if rnilking Gamma isotopic and 1-131 animals in 3 locations animals are identified on analysis of each sample.

within 5 km distance pasture; at least once per having the highest dose month at other times.

potential. If thele are less than 3 primary locations available then 1 or more secondary sample from milking animals in each of 3 areas between 5 to 8 Icm distance where doses are calculated to be greater than I nrem per year.

1 sample from milking animals in a control location.

b. Fish 1 sample of two Semiannually. Gamma isotopic analysis (d recreationally important on edible portions.

species in vicinity of plant discharge area.

I sample (preferably of same species) in areas not influenced by plant discharge.

(d)

c. Vegetation 1 grass sample at each air Quarterly when available. Gamma isotopic analysis sampling station. of each sample.

tdl 1 silage sample at each At time of harvest. Gamma isotopic analysis milk sampling station (as of each sample.

available).

I Revision 29 Date 1/11/02 3/4-32

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 are permitted from the required sampling schedule if specimens are unobtainable due to hazardous conditions, seasonal unavailability, malfunction of automatic sampling equipment and other legitimate reasons. If specimens are unobtainable due to sampling equipment malfunction, every 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 10.2. 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 enviromnental monitoring program. In lieu of a Licensee Event Report and pursuant to ODCM Section 10.1, 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 Report and also include in the report a revised figure(s) and table for the ODCM reflecting the new location(s).

b One or more instruments, such as a pressurized ion chamber, for measuring and recording dose rate continuously may be used in place of, or in addition to, integrating dosimeters. For the purposes of this table, a Thermoluminescent Dosimeter (TLD) is considered to be one phosphor; two or more phosphors in a packet are considered as two or more dosimeters. Film badges shall not be used as dosimeters for measuring direct radiation. The 40 stations is not an absolute number. The frequency of analysis or readout for TLD systems will depend upon the characteristics of the specific system used and should be selected to obtain optimum dose information with minimal fading.

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

d Gamma isotopic analysis means the identification and quantification of gamma-emitting 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 discharge. The "downstream" sample shall be taken in an area beyond but near the mixing 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 Each meteorological sector'shall have an established "inner" and an "outer" monitoring location based on ease of recovery (i.e., response time) and year-round accessibility.

l h Deleted.

I Revision 29 Date 1/11/02 314-33

TABLE 3.5.2 Reporting Levels For Radioactivity Concentrations In Environmental Sam lesCa)

Reporting Levels Airborne Particulate or Water Gases Fish Milk Vegetation Sediment Analysis (pCi/l) (pCi/m3) (pCi/Kg, wet) (pCiAl) (pCi/Kg, wet) (pCiJKg, dry)

H-3 2 x 104()

Mn-54 1x10 3 3x104 Fe-59 4 x 102 x o, Co-58 1x10 3 3x10 4 Co-60 3x 2 1 x 104 3 x 103e Zn-65 3 x 102 2 x 104 Zr-Nb-95 4 x 10 I-131 2 (d) 0-9 3 1 x 10 2 Cs-134 30 10 1 x 103 60 l X10 3 Cs-137 50 20 2x 10 3 70 2x 103 Ba-La- 2x 102 3x 102 140 (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 (1) 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 104 pCill may be used.

(c) Reporting level for individual grab samples taken at North Storm Drain Outfall only.

l (d) If no drinking water pathway exists, a value of 20 pciAiter may be used.

I Revision 29 Date 1/11/02 314-34

TABLE 4.5.1 I Detection Capabilities For Environmental Sample Analvsis(a)(c)

Airborne Fish Water Particulate or (pCi/Kg, Milk Vegetation Sediment Analysis (pCi/i) Gas (pCi/m3) wet) (pCi/l) (pCi/Kg, wet) (pCi/Kg, dry)

Gross beta 4 0.01 H-3 3000 Mn-54 15 130 Fe-59 30 260 Co-58, 60 15 130 Zn-65 30 260 Zr-Nb-95 15 (b)

I I-131 1( 0.07 1 60 Cs-134 15 0.05 130 15 60 1S0 Cs-137 18 0.06 150 18 80 180 Ba-La-140 15 (b)(e) 15 ( _

I Revision 29 Date 1/11/02 314-35

TABLE 4.5.1 NOTATON (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 does 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 are 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) The Ba-140 LLD and concentration can be determined by the analysis of its short-lived daughter product La-140 subsequent to an 8 day period following collection. The calculation shall be predicted on the normal ingrowth equations for a parent-daughter situation and the assumption that any unsupported La-140 in the sample would have decayed to an insignificant amount (at least 3.6 percent of its original value). The ingrowth equations will assume that the supported La-140 activity at the time of the collection is zero.

(f) Deleted.

(g) If no drinking water pathway exists, a value of 15 pci/liter may be used.

I Revision 29 Date 1/11/02 314-36

314.5 RADIOLOGICAL ENVIRONMENTAL MONlTORING 3/4.5.2 Land Use Census CONTROLS 3.5.2 A land use census shall be conducted to identify the location of the nearest milk animal and the nearest residence in each of the 16 meteorological sectors within a distance of five miles. The survey shall also identify the nearest milk animal (within 3 miles of the plant) to the point of predicted highest annual average D/Q value in each of the three major meteorological sectors due to elevated releases from the plant stack.

APPLICABILITY:

At all times.

ACTION:

a. With a land use census identifying one or more locations which yield a calculated dose 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 permission 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 10.3.5.

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.

I Revision 29 Date 1/11/02 3/437

314.5 RADIOLOGICAL ENVIRONMENTAL MONTrORING l 3/4.5.3 Interlaboratory Comparison Program CONTROLS 3.5.3 Analyses shall be performed on referenced radioactive materials supplied as part of an Interlaboratory Program which has been approved by NRC.

APPLICABILITY:

At all times.

ACTION:

With analysis not being performed as required above, report the corrective actions taken to prevent a recurrence 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 performed as part of the above required Interlaboratory Program shall be included in the Annual Radiological Environmental Operating Report. The identification of the NRC approved Interlaboratory Program which is being participated in shall be stated in the ODCM.

I Revision 29 Date 1/11/02 3/4-38

3/4.6 EFFLUENT AND ENVIRONMENTAL CONTROL BASES INSTRUMENTATION Liquid Effluent Instrumentation (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 alarm setpoints for these instruments are to ensure that the alarm 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 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 flow rates/dilution flows 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 (3.1.2)

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.

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 from the site 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 I Revision 29 Date 1/11/02 314-39

3/4.6: EFFLUENT AND ENVIRONMENTAL CONTROL BASES (cont.)

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 ILA of Appendix I to 10CFR Part 50 assure that doses are maintained ALARA, and that annual concentration limits of Appendix B to OCFR20.1001-20.2402 will not be exceeded.

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 dose 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 l.A, I.A, and IV.A of Appendix I, OCFR Part 50. The Limiting Condition for Operation implements the guides set forth in Section lA 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 material in liquid effluents will be kept "as low as is reasonably achievable". The Surveillance Requirements implement the requirements in Section l.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 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 40CFR 141. 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 Effluents for the Purpose of Evaluating Compliance 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.

I Revision 29 Date 1/1 1/02 3/4-40

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 l.D of Appendix I to 10CFR Part 50. The specified limits governing the use of appropriate portions of the liquid radwaste treatment system were specified as a suitable fraction of the dose design objectives set forth in Section II.A of Appendix I, 10CFR Part 50, for liquid effluents.

Gaseous Effluents: Dose Rate (3.3.1)

The specified limits as determined by the methodology in the ODCM, restrict, at all times, the corresponding gamma and beta dose rates above background to a member of the public at or beyond the site boundary to (500) mremyear to the total body or to (3,000) mrem/year to the skin. This instantaneous dose rate limit allows for operational flexibility when off nornal occurrences 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.

Control 3.3. .b also restricts, at all times, comparable with the length of the sampling peiods of Table 4.8.2 the corresponding thyroid dose rate above background to an infant via the cow-milk-infant pathway to 1500 mrem/year for the highest impacted cow.

Gaseous Effluents: Dose from Noble Gases (3.3.2)

This Control is provided to implement the requirements of Sections I.B, M.A, and IV.A of Appendix I, 10CFR Part 50. The Limiting Condition for Operation implements the guides set forth in Section l.B 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 material in gaseous effluents will be kept "as low as is reasonably achievable". The Surveillance Requirements implement the requirements in Section fIA of Appendix L 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 I Revision 29 Date 1/11/02 3/4-41

314.6: EFFLUENT AND ENVIRONMENTAL CONTROL BASES (cont.)

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 detennining the air doses at the site boundary based upon the historical average atmospheric conditions.

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 1.109, "Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10CFR Part 50, Appendix I", Revision 1, October 1977 and Regulatory Guide 1.1 1, "Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water Cooled Reactors," Revision 1, July 1977.

Gaseous Effluents: Dose from Iodine-131. Iodine-133, Tritium, and Radionuclides in Particulate Form (3.3.3)

This Control is provided to implement the requirements of Section I[.C, f.A, and IV.A of Appendix I, 10CFR Part 50. The Limiting Condition for Operation are the guides set forth in Section 1l.C of Appendix I. The requirements provide operating flexibility and at the same time implement the guides set forth in Section 1V.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 1J.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 for the Purpose of Evaluating Compliance with 10CER Part 50, Appendix I", Revision 1, October 1977 and Regulatory Guide 1.111, "Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water Cooled Reactors," Revision 1, July 1977. These equations also provide for determining the actual doses based upon the historical average atmospheric conditions. The release rate specifications for Iodine 131, Iodine-133, tritium, and radionuclides I Revision 29 Date 1/11/02 3/4-42

3/4.6: EFFLUENT AND ENVIRONMENTAL CONTROL BASES (cont.)

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: 1) individual inhalation of airborne radionucides, 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.

Gaseous Radwaste Treatment (3.3.4)

The requirement that the appropriate portions of the Augmented Off-Gas (AOG)

System be used whenever the SJAE is in operation provides reasonable assurance that the releases of radioactive materials in gaseous effluents will be kept "a low as is reasonably achievable". This specification implements the requirements of lOCER Part 50.36a and the design objectives of Appendix I to 10CFR Part 50.

Ventilation Exhaust Treatment (3.3.5)

The requirement that the AOG Building and Radwaste Building HEPA filters be used when specified provides reasonable assurance that the release of radioactive materials in gaseous effluents will be kept "as low as is reasonably achievable". This specification implements the requirements of 10CFR Part 50.36a and the design objective of Appendix I to IOCFR Part 50. The requirements governing the use of the appropriate portions of the gaseous radwaste 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 II.B and II.C of Appendix L 10CER Part 50, for gaseous effluents.

Primary Containment (MARK I) (3.3.6)

This Control provides reasonable assurance that releases from containment purging/venting operations will be filtered through the Standby Gas Treatment System (SBGT) so that the annual dose limits of 10CFR Part 20 for Members of the Public in areas at and beyond the Site Boundary will not be exceeded. The dose objectives of Control 3.3.3 restrict purge/venting operations when the Standby Gas Treatment System is not in use and gives reasonable assurance that all releases from the plant will I Revision 29 Date 1/11/02 3/4-43

314.6: EFFLUENT AND ENVIRONMNTAL CONTROL BASES (cont.)

be kept "as low as is reasonably achievable". The specification requires the use of SBGT only when Iodine-131, Iodine-133 or radionucides in particulate form with half-lives greater than 8 days in containment exceeds the levels in Table 1, Column 3, to Appendix B of 10CFR 20.1001-20.2401 since the filter system is not considered effective in reducing noble gas radioactivity from gas streams.

The use of the 18" purge and vent flow path isolation valves AC-7A (16-19-7A), AC-7B (16-19-7B), AC-8 (16-19-8), AC-10 (16-19-10) has been restricted to 90 hours0.00104 days 0.025 hours 1.488095e-4 weeks 3.4245e-5 months per year. Normal plant operations (other than inerting and de-inerting) will have AC-8 and AC-10 closed and nitrogen will be supplied to the drywell via the 1" nitrogen makeup supply. The differential pressure maintained between the drywell and torus will allow the nitrogen to "bubble over'" into the suppression chamber. A normally open AC-6B (3") allows for venting. A normally closed AC-6A (3") is periodically opened for performance of surveillances such as monthly torus to drywell vacuum breaker tests. Procedurally, when AC-6A is open, AC-6 and AC-7 are closed to prevent overpressurization of the SBGT system or the reactor building ductwork, should a LOCA occur. For this and similar analyses performed, a spurious opening of AC-6 or AC-7 (one of the closed containment isolation valves) is not assumed as a failure simultaneous with a postulated LOCA. Analyses demonstrate that for normal plant operation system alignments, including surveillances such as those described above, that SBGT integrity would be maintained if a LOCA was postulated.

Therefore, during normal plant operations, the 90 hour0.00104 days 0.025 hours 1.488095e-4 weeks 3.4245e-5 months clock does not apply.

Accordingly, opening of the 18 inch atmospheric control isolation valves AC-7A, AC-7B, AC-8 and AC-10 will be limited to 90 hours0.00104 days 0.025 hours 1.488095e-4 weeks 3.4245e-5 months per calendar year (except for performance of the subject valve stroke time surveillances - in which case the appropriate corresponding valves are closed to protect equipment should a LOCA occur). This restriction will apply whenever primary containment integrity is required. The 90 hour0.00104 days 0.025 hours 1.488095e-4 weeks 3.4245e-5 months clock will apply anytime purge and vent evolutions can not assure the integrity of the SBGT trains or related equipment.

Steam Jet Air Eiector (SJAE) (3.3.7)

Restricting the gross radioactivity release rate of gases from the main condenser SJAE provides reasonable assurance that the total body exposure to an individual at the exclusion area boundary will not exceed a small fraction of the limits of 10CFR Part 100 in the even this effluent in inadvertently discharged directly to the environment without treatment. This specification implements the requirements of General Design Criteria 60 and 64 of Appendix A to OCFR Part 50. (his basis is a duplicate of that for plant Technical Specification 3.8.K.)

I Revision 29 Date 1/11/02 3/4-44

3/4.6: EFFLUENT AND ENVIRONMENTAL CONTROL BASES (cont.)

Total Dose (40CFR190) (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.

For sites containing up to 4 reactors, it is highly unlikely that the resultant dose to a Member of the Public will exceed the dose limits of 40CFR Part 190 if the individual reactors remain within the reporting requirement level. 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 the purposes of the Special Report, it may be assumed that 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 40CER 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 IOCFR Part 20. An individual is not considered a Member of the Public during any period in which he/she is engaged in canying 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.

Radiological Environmental Monitoring Program (3.5.1)

The radiological monitoring program required by this Control provides measurements of radiation and of radioactive materials in those exposure pathways and for those radionucides which lead to the highest potential radiation exposures of member(s) of the public resulting from the station operation. This monitoring program implementsSection IV.B.2 of Appendix I to 10 CFR Part 50 and thereby supplements the radiological effluent monitoring program by verifying that the measurable concentrations of radioactive materials and levels of radiation are not higher than expected on the basis of the effluent measurements and modeling of the environmental exposure pathways.

Ten years of plant operation, including the years prior to the implementation of the Augmented Off-Gas System, have amply demonstrated via routine effluent and I Revision 29 Date 1/11/02 3/445

3/4.6: EFFLUENT AND ENVIRONMENTAL CONTROL BASES (cont.)

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 ALARA 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. Due to the large local population of cows and the ready availability of milk samples, food product sampling has been eliminated from the program in lieu of milk sampling. Since milking cows in the area spend very little time on pasture, silage and grass sampling have been instituted as an indicator of radionuclide deposition.

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 LU) for a particular measurement based upon the actual parameters for the sample in question.

Land Use Census (3.5.2)

This Control is provided to ensure that changes in the use of areas at and beyond the site boundaries are identified and that modifications to the monitoring program are 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 to the substantial and widespread occuirence of dairy farming in the surrounding area which dominates the food uptake pathway.

I Revision 29 Date 1/11/02 3/4 -46

3/4.6: EFFLUENT AND ENVIRONMENTAL CONTROL BASES (cont.)

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 of the calculated dose or dose commitment at 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.

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.

I Revision 29 Date 1/11/02 314-47

5.0 METHOD TO CALCUJLATE 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 effluent concentration in liquid pathways, excluding noble gases, denoted here as, FCNG at the point of discharge at any time (see Figure 9-1). FENGis limited to less than or equal to ten, i.e.,

FENG <10 The total concentration of all dissolved and entrained noble gases at the point of NG -4[C/l ..

discharge from all station sources, denoted Cl , is limited to 2E-0 jCi/rl, i.e.,

CNG1

<2E-04 uCi /ml.

Evaluation of FENG and CN0 is required concurrent with the sampling and analysis program in Control Table 4.2.1.

5.1 Method to Determine F,G and C'G Determine the total fraction of combined effluent concentrations at the point of discharge in liquid pathways (excluding noble gases), denoted F NG, 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 C1 NG, as follows:

ENG _ __ _

ECL; (5-1)

(pCi/m p.Ci/ml )

and:

Revision 29 Date 1/11/02 5-1

CNG = cG < 2E- 04 ii C1~

~ <2E04 (5-2)

( Ci!ml) (ILCiIl) ( Ci/mI) where:

FENG = 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.

Cp; = 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 and any other significant source from which a discharge can be made)

(gCiml). This concentration can be calculated from: Cpi = CTKI x FT/[FDuL + PT) where: CTKI equals the concentration of radionuclide i in the tank to be discharged (Ci/ml); FDL 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 a waste collection tank is discharged (gpm).

ECI = 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 (Ci/ml).

CNG = Total concentration at point of discharge to an unrestricted area of all dissolved and entrained noble gases in liquid pathways from all station sources (.Ci/ml).

CiG = Concentration at point of discharge to an unrestricted area of dissolved and entrained noble gas "i" in liquid pathways from all station sources (pCimi).

I Revision 29 Date 1/11/02 5-2

5.2 Method to Determine Radionuclide Concentration for Each Liquid Effluent Pathway 5.2.1 Sample Tanks Pathways Cpi is determined 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 demineralizers. Other periodic batch releases may also come from the detergent waste tank or the floor drain sample tank.

The tanks are sampled from the radwaste sarnple sink 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 nuclear system. If the sample does not meet all the high purity requirements, the contents are recycled through the radwaste system or discharged.

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 alann setpoint on the service water discharge monitor is set at a leyel 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.

Revision 29 Date' 1/11/02 5-3

Under expected or anticipated operating conditions, the concentration at an3time 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 10CFR20.1001-20.2402, Appendix B, Table 2, Column 2.

5.2.3 CirculatingWaterPathwav The circulating water pathway shown on Figure 9-1, flows from the intake structure through the condenser and the discharge structure. Under normal operating conditions, the water in this line is not radioactive. For this reason, the circulating water line is not sampled routinely but it is monitored continuously by the discharge process monitor (No. 17/359) located in the discharge structure.

The alarm setpoint on the discharge process monitor is set at a level which is three times the background of the instrument. The circulating water is sampled if the monitor is out of service or if the alarm sounds.

Under normal 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.

I Revision 29 Date 1/11/02 5-4

6.0 OFF-SrrE DOSE CALCULATION METHODS Chapter 6 provides the basis for plant procedures required to meet'the-10CFR5,O Appendix I, ALARA dose objectives, and the 40CFR190 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 1.1.7 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 but still conservative methods (called Method II) 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 40CFR190 is considered to be a demonstration of compliance with the 0.1 rem limit of 10CFR20.1301(a)(1) for members of the public in unrestricted areas (Reference 56 FR23374, 3rd column).

Revision 30 Date 10/30/02 6-i

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 releases. The term "Dose," with respect to extemal 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 10CFR50.

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, does not necessarily, by itself, constitute a Licensee Event Report (LER) under IOCFR Part 50.73 unless it is determined that the air concentration of radioactive effluents in unrestricted areas has also exceeded 20 times applicable concentration limits specified in Appendix B to 20.1001 - 20.2402, Table 2, Column 1 (four-hour notification per 10CFR50.72, and 30-day IER per 10CFR50.73).

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 r' 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 Revision 30 Date 10/30/02 6-2

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%.

I. is 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.

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 citical organ from gaseous releases will be the sum of the elevated and ground level doses. Appendix D provides an assessment of the surveillance needs for waste oil to ensure that off-site doses from its incineration is maintained within the ALARA limits of the effluent Controls.

I Revision 30 Date 10/30/02 6-3

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 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 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.

6.2.1 Method I The increment in total body dose from a liquid release is:

(6-1)

D t = Z Q DL itb (mrem) (Ci) (mCi J where:

DFLtb = Site-specific total body dose factor (mrem/Ci) for a liquid release. See Table 1.1.11.

Revision 30 Date 10/30/02 64

Qi = 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: (1) 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 HI.

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 10CFR50, Appendix I Subsection II A. Control 3.2.3 is an "appropriate fraction", determined by the NRC, of that designi objective (hereafter called the Objective). Control 3.4.1 is based on Environmental Standards for Uranium Fuel Cycle in 40CFR190 (hereafter called the Standard) which applies to direct radiation as well as liquid and gaseous effluents.

Exceeding the Objective or the Standard does not immediately limit plant operation but requires a report to the NRC within 30 days. In addition, a waiver may be required.

Method I was developed such that "the actual exposure of an individual ... is unlikely to be substantially underestimated" (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 I Revision 30 Date 10/30/02 6-5

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 dose impact to the critical receptor (in the form of dose factors DFI,tb, mrem/Ci) for a 1 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 (Equations 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 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 flpw rate of 44.6 ft3 /sec was used with a mixing ratio of 0.0356 which corresponds to a ninimum 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, total body dose factors appearin 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)

ADtb=QiDFLitb (mrem) (Ci) (mrim) where:

DFIztb = Site-specific total body dose factor (mrem/Ci) for a liquid release. See Table 1.1.11.

Qi = 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.

Revision 30 Date 10/30/02 6-6

Method I is conservative because it is based on dose factors DFL,t 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 I 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 1I should be applied. Method II consists of the models, input data and assumptions in Regulatory Guide 1.109, Rev. 1 (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 II. It is an acceptable starting point for a Method II analysis.

Revision 30 Date 10130/02 6-7

TABLE 6.2.1 Environmental Parameters for Liquid Effluents at Vermont Yankee (Derived from Reference A)

FOOD GROWN WITH CONTAMINATED WATER POTABLE AQUATIC SHORELINE LEAFY COW VARIABLE WATER FOOD ACTIVITY VEC3ETABLES VEG. MEAT MILK MP IVflixing 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 YV Agricultural (KGIlVl 2) 2.0 2.0 2.0 2.0 Productivity P Soil Surface Density (KGPM2 ) 240.0 240.0 240.0 240.0 2

IRR Irrigation Rate (L/N4 /11R) 0.152 0.152 0.152 0.152 TE Crop Exposure Time (HR S) 1440.0 1440.0 1440.0 1440.0 TH Holdup Time (HRS) 1440.0 24.0 2160.0 2160.0 QAW Water Uptake Rate for (IE 50.0 60.0 Animal QF Feed Uptake Rate for (KG/D) 50.0 50.0 Animal Fl Fraction of Year Crops Irrigated 0.5 0.5 0.5 0.5 Location of Critical Connecticut River Below Vemon Dam Receptor I Revision 30 Date 10/30/02 6-8

TABLE 6.2.2 Usage Factors for Various Liquid Pathways 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 -

(KGYR) (KG/YR) (LITER/YR) (KG/YR) (KG/YR) (KGfYR) (LlERYR) (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 j

I Revision 30 Date 10/30/02 6-9

6.3 Method to Calculate Maximum Organ Dose from Liquid Releases Effluent Control 3.2.2 limits the maximum organ dose commitment to a Member of the Public from radioactive material in liquid effluents 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 mrem in any month. Control 3.4.1 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 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., Method I indicates the dose is greater than the linit), 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.

6.3.1- MethodI The increment in maximum organ dose from a liquid release is:

(6-3)

Dmo = X Qi DFL imo i

(mrem) (Ci) ( Ci -)

Revision 30 Date 10/30/02 R 6-10

where:

DFIjmo = Site-specific maximum organ dose factor (rem/Ci)-for a liquid release.

See Table 1.1.11.

Qi = 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 following conditions (otherwise, justify Method I or consider Method I):

1. Normal operations (not emergency event),

2. Liquid releases, were to the Connecticut River, and

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

6.3.2 Basis for Method I This section serves three purposes: (1) 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. 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 (DFLj) 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)

A Dro= QiDFL imo (mrem) (Ci) (Ci')

I Revision 30 Date 10/30/02 6-11

where:

DFL,= = Site-specific maximum organ dose factor (mrem/Ci)for a liquid release.

See Table 1.1.11.

Qi = 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 10CFR20 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.

6.3.3 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 1.109, 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.

I Revision 30 Date 10/30/02 6-12

6.4 Method to Calculate the Total Body Dose Rate From Noble Gases Effluent Control 3.3.1 linits 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 3.3.1 had actually been exceeded during a short lime interval.

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 lirnit 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 3.3.1 is unsuccessful, or as required by the notations to Control Table 3.1.2 when the stack noble gas monitor is inoperable.

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 (0.61), as seen in the following Equation 6-5:

Revision 30 Date 10/30/02 6-13

t,= 0.61 z QST DEBi (6-5)

I~~~~~~~~~

(mrenq (pCi-sec) (g Ci mrem-ms) t yr ) (pi-in 3 ) sec) pCi-yr )

where:

Q = In the case of noble gases, the release rate from the plant stack (puCilsec) for each radionuclide, "i", identified. The release rate at the plant stack is based on measured radionuclide concentrations and distributions in periodic grab samples taken at the stack. As an altemative method, the radionuclide distribution in the off-gas at the Steam Jet Air Ejector (SJAE) can be used during plant operations, along with the Stack Gas Monitor effluent count rate, to estimate stack radionuclide releases. The release rate at the stack when using SJAE samples can be stated as follows:

SJAE 1 (6-28) uCi (M)(j Cifcc (cc) see 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 (cpml(,Cifcc)).

F = Stack flow rate (cc/sec).

= The last measured release rate at the steam jet air ejector of noble gas i (LCi/sec).

DFBi Total body gamma dose factor (see Table 1.1.10).

I Revision 30 Date 10/30/02 6-14

For ground level noble gas releases, the total body dose rate is calculated as follows:

Rtb8 = 6.4 i Q DFB i (6-39)-

(pCi - sec ) (AIci ( em - 3 )

U'ci - m sec pci - yr where:

G9L = Ground level release rate (Ci/sec) of noble gas.

The total body dose rate for the site is equal to Ztbs + Abg.

During periods (beyond the first five days) when the plant is shutdown and no radioactivity release rates can be measured at the SJAE, Xe-133 may be used in place of the last SJAE measured mix as the referenced radionuclide to determine off-site dose rate and monitor setpoints. In this case, the ratio of each QJAE to the sum of all QsJAE in Equation 6-28 above is assumed to reduce to a value of 1, and the total body gamma dose factor DF.Bi for Xe-133 (2.94 E-04 mrem-m 3 /pCi-yr) is used in Equation 6-5. Alternately, a relative radionuclide "i" mix fraction (f.) may be taken from Table 8.2.1 as a function of time after shutdown, and substituted in place of the ratio of Q'IAE to the sum of all QS' in Equation (6-28) above to deterimine the relative fraction of each noble gas potentially available for release to the total. Just prior to plant startup before a SJAE sample can be taken and analyzed, the monitor alarm setpoints should be based on Xe-138 as representing the most prevalent high dose factor noble gas expected to be present shortly after the plant returns to power. Monitor alarm setpoints which have been determined to be conservative under any plant conditions may be utilized at any time in lieu of the above assumptions.

Equations 6-5 and 6-39 can be applied under the following conditions (otherwise, justify Method I or consider Method ID:

1. Normal operations (not emergency event), and

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

I Revision 30 Date 10/30/02 6-15

6.4.2 Basis for Method I Method I 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:

j,5 =1E + 06 [X/Q]' E ( T DFB (mrem) (pCi) sc) n (lLCi mrem -m3) (6-6) yr m3 MCi sec ) pCi - yr )

where:

SF Shielding factor = 1.0 for dose rate determination.

[XIQ = Maximum annual average gamma atmospheric dispersion factor for stack (elevated) releases; = 6.1 1E-07 (sec/m 3).

QS - Release rate from the plant stack of noble gas "i" (Ci/sec).

DFBi = Gamma total body dose factor,( Ci m) j See Table 1.1.10.

Equation 6-6 reduces to:

bs= 0.61 Q DFBi Inrem pCi-sec (.Ci) mrer-m 3 (6-5) pCi-m 3) tsec pCi-yr For ground level releases, the ground level maximum long-term average gamma atmospheric dispersion factor = 6.42E-06 sec/m3 , thus leading to:

ktbg = 1E+06

6.42E-06 '

QL DFBi or (6-39)

Rtbg = 6.4 Q DFBi i

I Revision 30 Date 10/30102 6-16

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. Due to the holdup and decay of gases allowed in the AOG, off-gas concentratioffs at the plant-stack during routine plant operations are usually too low for determination of the radionuclide nix at the plant stack. It is then conservatively assumed that most of the noble gas. activity at the plant stack is the result of in-plant steam leaks which are removed to the plant stack by building ventilation air flow, and that this air flow has an isotopic distribution consistent with that routinely measured at the SJAE.

The calculation of ground level release dispersion parameters are based on the location of the North Warehouse with respect to the site boundary that would experience the highest exposure. The North Warehouse contains a waste oil burner that can be used for the incineration of low level contaminated waste oil, and is designated as a ground level release point to the atmosphere. Due to differences in building cross sectional areas and resulting building wake effects, the North Warehouse atmospheric dispersion factors are conservative in comparison to those associated with the main plant facilities, such as the Turbine Building. As a consequence, any potential or unexpected ground level release from the Turbine Building or adjoining structures can utilize the above ground release dose assessment equations.

In the case of noble gas dose rates, Method II cannot provide much extra realism because RIbs and 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 cai 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 1.109, 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.

Revision 30 Date 10/30/02 6-17

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 slin from all release sources of noble gases at 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 Method I predicts a dose rate greater than the Control limits (i.e., use of actual meteorology over the period of interest) to determine if, in fact, Control 3.3.1 had actually been exceeded during a short time interval.

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 alann 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 inoperable.

6.5.1 Method I The skin dose rate due to noble gases is determined by multiplying the individual radionuclide release rates by their respective dose factors, and summing all the products together as seen in the following Equation 6-7:

I Revision 30 Date 10/30/02 6-18

Rskins = E QS DF'h (mrem A Ci li (mrem- sec - . (67) yr) sec J ,uCi- yr )

where:

= In the case of noble gases, the noble gas release rate from the plant stack (gCi/sec) for each radionuclide, "i", identified. The release rate at the plant stack is based on measured radionuclide concentrations and distributions in periodic grab samples taken at the stack. As an alternative method, the radionuclide distribution in the off-gas at the Steam Jet Air Ejector (SJAE) can be used during plant operations, along with the Stack Gas Monitor effluent count rate, to estimate stack radionuclide releases. The release rate at the stack when using SJAE samples can be stated as follows:

ST =JAE M 1 F Qi Qi SJAE EQsle ~Sg (6-28) p Ci = (cpm) (ACi/cc) (cc) sec cpm sec M = Plant stack gas monitor I or II count rate (cpm).

S9 = Appropriate or conservative plant stack monitor detector counting efficiency for the given nuclide mix (cpm/(,Ci/cc)).

F Stack flow rate (cc/sec).

Q PJAE = The last measured release rate at the steam jet air ejector of noble gas i (Ci/sec).

DV = combined skin dose factor (see Table 1.1.10) for stack release.

I Revision 30 Date 10/30/02 6-19

For ground level releases, the skin dose rate from noble gases is calculated by Equation 6-38:

Rskin = Q iG DF'ig (6-38) where:

Q = The noble gas release rate from ground level (Ci/sec) for each radionuclide "i" identified.

D,g = Combinedskin dosefactorforagroundlevelrelease [seeTable 1.1.10A].

The skin dose rate for the site is equal to RI. + Rskig During periods (beyond the first five days) when the plant is shutdown and no radioactivity release rates can be measured at the SJAE, Xe-133 may be used in place of the last SJAE measured mix as the referenced radionuclide to determine off-site dose rate and monitor setpoints. In this case, the ratio each of QJAE to the sum of all QJAE in Equation 6-28 above is assumed to reduce to a value of 1, and the combined skin dose factor DF'is for Xe-133 (5.58 E-04 mrem-sec/[tCi-year) is used in Equation 6-7. Alternately, a relative radionuclide "i" mix fraction (fi) may be taken from Table 8.2.1 as a function of time after shutdown, and substituted in place of the ratio of each to the sum of all Q in Equation 6-28 above to determine the relative fraction of each noble gas potentially available for release to the total. Just prior to plant startup before a SJAE sample can be taken and analyzed, the monitor alarn setpoints should be based on Xe-138 as representing the most prevalent high dose factor noble gas expected to be present shortly after the plant returns to power. Monitor alarm setpoints which have been determined to be conservative under any plant conditions may 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):

1. Normal operations (not emergency event), and

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

Revision 30 Date 10/30/02 6-20

6.5.2 Basis For Method I The methods to calculate skin dose rate parallel the total body dose rate methods-in Section 6.4.3. Only the differences are presented here.

Method I may be used to show that the Control limit for sldn dose rate from noble gases released to the atmosphere (Control 3.3.1) has been met for the peak noble gas release rate.

Method I was deived from Regulatory Guide 1.109 as follows:

DS = 1.11 SF DI + 3.17E+ 04 E Qi [X/Qls DFSi (6-8)

Cire yr)J

) = (rem) mradJ)

(#) (rad" y (Ci (pci- yr)

- sec) rci ()

tyr) m fmrem- m3 )

pci-yr where:

1.11 = Average ratio of tissue to air absorption coefficients will convert mrad in air to mrem in tissue.

D'. = 3.17E +04 : Qi [X/Q], DFY (6-9)

(mrad (pCi-yr (Ci (sec (mrad -M 3 yr ) ci -sec ) yr) m pCi -yr now Dy = D [X/Q]l I[X/Q], (6-10) irad) _mrad sec rm3)

C yrD) yrd (m3J (sec)

I Revision 30 Date 10/30/02 6-21

and Qi = 31.54 *ST (6-11)

Qi (ciJ (Ci - sec ) ( pCi >

t ) yr tCi - yrJ sec sO Rskins = 1.11 S 1E + 06 [X/Q . Qs DWT (6-12)

Cmrem (mrem )

(#) PLO )" se 3 ) C() (a * )

yr J mrad) loci m sec ) pC - yr

+1E+06 x/QsZ O Q DFSi 3

(pCi) sec jci (mrem-m I ci ) m sec t pci - yr )

substituting 3

7/QY = 6.11E-07 SeC/m X/Qs = 1.04E-06 sec/m 3 SF = Shielding factor = 1.0 for dose rate determinations gives Rskins = 0.68 E Q T DF + 1.04 QST DFS, (6-13) rmremr pCi - sec- nrem' fliCi (mriad-m3)(pCi-secr jCi) (nrem-m 3 yr.jCi-m sec I s I Revision 30 Date 10/30/02 6-22

X,QST[ 0 .6 8 DFY +1.04 DFSil (6-14) define DF' 1 s= 0.68 DF + 104 DFSi (6-15) then Rs .s = ST Dlj (6-7)

Qi 67 (Inremj (PLCi mrem - sec) sec Ci - yr )

For determining combined skin doses for ground level releases, a [X/Q]Y = 6.42E-06 sec/m and an undepleted X/Qg = 3.52E-05 sec/m3 have been substituted into Equation 6-12 to give:

R5ki,g = z QGL (7.13 DF1 I + 35.2 DFi) then DFig = 7.13 DFJY + 35.2 DFS (6-37) and Askig =I Q?L DFi'g (6-38) where:

(Q9L = The noble gas release rate from ground level release points (p.Ci/sec) for each radionuclide "i" identified.

DFjg = Combined skin dose factor for a ground level release [see Table 1.1.10A].

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.

I Revision 30 Date 10/30/02 6-23

The calculation of ground level release dispersion parameters are based on the location of the North Warehouse with respect to the site boundary that would experience the highest exposure. The North Warehouse contains a waste oil burner that can be used for the-incineration of low level contaminated waste oil, and is designated as a ground level release point to the atmosphere. Due to differences in building cross sectional areas and resulting building wake effects, the North Warehouse atmospheric dispersion factors are conservative in comparison to those associated with the main plant facilities, such as the Turbine Building. As a consequence, any potential or unexpected ground level release from the Turbine Building or adjoining structures can utilize the above ground release dose assessment equations.

6.5.3 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 1.109, 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.

Revision 30 Date 10/30/02 6-24

6.6 Method to Calculate the Critical Organ Dose Rate from Iodines, Tritium and Particulates with T1/2 Greater Than 8 Days Effluent Control 3.3.1.b limits the dose rate to any organ, denoted 1o, from all release sources of I-131, I-133, H-3, and radionucides in particulate form with half lives greater than 8 days to 1500 mremlyear to any organ. The peak release rate averaging time in the case of iodines and particulates is commensurate with the time the iodine and 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.l.b is the total contribution from both ground and elevated releases occurring during the period of interest.

Method I applies at all releaser 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 if, in fact, Control 3.3.1.b had actually been exceeded during the sampling period.

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:

Rcos = Y ('S DFG' sico (mrem) ~ (RCi) (inrem-sec) (6-16) yr ) ~ sec JCi - yr I Revision 30 Date 10/30/02 6-25

where:

QjSTP = Stack activity release rate determination of radionuclide "i" (Iodine-131, Iodine-133, particulates with half-lives greater than 8 days, and7tritium), in p.Cilsec. For i = Sr89, Sr9O or tritium, use the best estimates (such as most recent measurements).

DFG'sico = Site specific critical organ dose rate factor (nrem secj for a ground level gaseous release. See Table 1.1.12.

For ground releases (North Warehouse waste oil burner) the critical organ dose rate from Iodine, Tritium, and Particulates with T 1/2 greater than 8 days is calculated as follows:

Rtcog X G DFG'co (6-40) where:

QtGLP = Ground activity release rate determination of radionuclide "i" (Iodine-131, Iodine-133, particulates with half-lives greater than 8 days, and tritium), in pCi/sec. For i = Sr89, Sr9O, Fe-55, or tritium, use the best estimates (such as most recent measurements). For waste oil, the release rate is the total activity by radionuclide divided by the estimated burn time. (See Appendix D for surveillance criteria on waste oil buming.

DFG'gic - Site specific critical organ dose rate factor (mrem sec) for a ground level gaseous release. See Table 1.1.12.

The critical organ dose rate for the site is equal to 1 cos + Rcog Equations 6-16 and 6-40 can be applied under the following conditions (otherwise, justify Method I or consider Method II):

1. Normal operations (not emergency event), and

2. Titium, iodine, and particulate releases via either elevated or ground level vents to the atmosphere.

I Revision 30 Date 10/30/02 6-26

6.6.2 Basis for Method I The methods to calculate critical organ dose rate parallel the totalbody dose rate7methods in Section 6.4.3. Only the differences are presented here.

Method I may be used to show that the Control limit for organ dose rate from iodines, tritium and radionuclides in particulate form with half lives greater than 8 days (hereafter called Iodines and Particulates or "I+P") released to the atmosphere (Control 3.3.1 b) has been met for the peak I + P release rate.

The equation for &, and 1&g is derived by modifying Equation 6-25 from Section 6.9 as follows:

Dcos = E Qi DFGico i

(6-17)

(mrem) (Ci) (mrem) applying the conversion factor, 31.54 (Ci-sec/p.Ci-yr) and converting Q to (0in jiCifsec as it applies to the plant stack yields:

R_ = 31.54 E STP DFGsico i

C mrem yr )

~( Ci- sec LyCi-yr CCi)

(sec)

(mrem)

CiJ (6-18)

Equation 6.8 is written in the form:

R = 31.54 I QTP DFGsico (6-19)

(mremj (Ci-sec) (liCi) (mremD t yr) tIc i - r) (Iseci J ECim I Revision 30 Date 10/30/02 6-27

DFG'i o and DFGBCo (North Warehouse waste oil burner vent 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 Sp value of 0.7 assumed in the determination of DFG5jc 0 and DFGgico 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 North Warehouse with respect to the site boundary that would experience the highest exposure. The North Warehouse contains a waste oil burner that can be used for the incineration of low level contaminated waste oil, and is designated as a ground level release point to the atmosphere. Due to differences in building cross sectional areas and resulting building wake effects, the North Warehouse atmospheric dispersion factors are conservative in comparison to those associated with the main plant facilities, such as the Turbine Building. As a consequence, any potential or unexpected ground level release from the Turbine Building or adjoining structures can utilize the above ground release dose assessment equations.

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.

6.6.3 Method II If Method I cannot be appRied, 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 1.109, 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 H analysis.

Revision 30 Date 10/30/02 6-28

6.7 Method to Calculate the Gamma Air Dose from Noble Gases Effluent Control 3.3.2 limits the gamma dose to air from all release sources of-noble gases at any location at or beyond the site boundary to 5 mrad in any quarter and 10 mrad in any year. Dose evaluation is required at least once per month.

Use Method I first to calculate the gamma air dose for elevated and ground level vent releases during the period. The total gamma 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.

6.7.1 Method I The gamma air dose from plant stack releases is:

Dr = 0.019 E QT i

(6-21)

(mrad) (pCi- yr) (Ci) (mrad- m3)

~c i-W3. t pci-yr where:

Qs T = total noble gas activity (Curies) released to the atmosphere via the plant stack of each radionuclide "i" during the period of interest.

DF,y = gamma dose factor to air for radionuclide "i". See Table 1.1.10.

For ground level noble gas releases, the gamma air dose is calculated as follows:

D g= 0.20 Q L DF (6-41)

I Revision 30 Date 10/30/02 6-29

where:

QGL - Total noble gas activity (curies) released to the atmosphere via ground-level vents of each radionuclide, "i", during the period of interest.

The gamma air dose for the site is equal to Dya +Dg.

Equations 6-21 and 6-41 can be applied under the following conditions (otherwise justify Method I or consider Method H):

1. Normal operations (not emergency event), and

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

6.7.2 Basis for Method I Method I may be used to show that the Control limit for off-site gamma air dose from gaseous effluents (3.3.2) has been met for releases over appropriate periods. This Control is based on the Objective in OCFR50, Appendix I, Subsection B.1, which limits the estimated annual gamma air dose at unrestricted area locations.

Exceeding the Objective does not immediately limit plant operation but requires a report to the NRC.

For any noble gas release, in any period, the dose is taken from Equations B4 and B-5 of Regulatory Guide 1.109 with the added assumption that DI , = DI [XIQfyQ[XIQ]:

Das 3.17E+04 [X/Q]s , QsrDFy (6-22)

(pci- yr) (mrad-m 3~

(mrad)( Y (sec/m 3 ) (Ci) y Cl- secyr C I Revision 30 Date 10/30/02 6-30

where:

[X/Qy = maximum long term average gamma atmospheric dispersion faetor for a stack release.

= 6.1IE-07 (sec/m3 )

Qsr = number of curies of noble gas "i" released from the plant stack which leads to:

DI,, = 0.019 L DFy (6-21)

(mrad) (pCi - yr (Ci) (mrad - m3 )

(LC i- m3) t pCi-yr)

For the ground level release:

Drg = 3.17E+04 [XQ E QGL DFy (6-42) where:

(X/Q)y = Maximum long-term average gamma atmospheric dispersion factor for a ground level release

= 6.42E-06 sec/m3 leading to:

Dy,rg = 0.20 z QL DFiY (641)

I Revision 30 Date 10/30/02 6-31

The calculation of ground level release dispersion parameters are based on the location of the North Warehouse with respect to the site boundary that would experience the highest exposure.

The North Warehouse contains a waste oil burner that can be used for theincineration.of low level contaminated waste oil, and is designated as a ground level release point to the atmosphere.

Due to differences in building cross sectional areas and resulting building wake effects, the North Warehouse atmospheric dispersion factors are conservative in comparison to those associated with the main plant facilities, such as the Turbine Building. As a consequence, any potential or unexpected ground level release from the Turbine Building or adjoining structures can utilize the above ground release dose assessment equations.

The main difference between Method I and Method II is that Method II would allow the use of actual meteorology to deternine [X/Q]Y rather than use the maximum long-term average value obtained for the years 1981 to 1985.

6.7.3 Method [

If the Method I dose determination indicates that the Control limit may be exceeded, or if a more exact calculation is required, then Method I may be applied. Method II consists of the models, input data and assumptions in Regulatory Guide 1.109, Rev. 1 (Reference A), except where site-specific models, data or assumptions are more applicable.

Revision 30 Date 10/30/02 6-32

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 10 mrad in any quarter and 20 mrad in any year.

Dose evaluation is required at least once per month.

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.

6.8.1 Method I The beta air dose from plant vent stack releases is:

Dfi = 0.033 Qsr DFj (6-23)

(mrad) (I)Ci- mr) (Ci) pCmrad-y) pCi- Yr)

~

where:

DFi = beta dose factor to air for radionuclide "i". See Table 1.1.10.

QI T = total noble gas activity (curies) released to the atmosphere via the plant stack of each radionuclide "i" during the period of interest.

I Revision 30 Date 10/30/02 6-33

For ground level noble gas releases, the beta air dose is calculated as follows:

Dfiaig =-1. 12 E1 QiL DCe>: (6-43) where:

QL = Total noble gas activity (curies) released to the atmosphere via the ground level vents of each radionuclide "i" during the period of interest.

The beta air dose for the site is equal toDO,airs +D.aurg Equations 6-23 and 643 can be applied under the following conditions (otherwise justify Method I or consider Method II):

1. Normal operations (not emergency event), and

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

6.8.2 Basis for Method I This section serves three purposes: (1) 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 HI. The methods to calculate beta air dose parallel the gamma air dose methods in Section 6.7.3. Only the differences are presented here.

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 10CFR50, Appendix I, Subsection B.1, which limits the estimated annual beta air dose at unrestricted area locations.

Exceeding the Objective does not immediately limit plant operation but requires a report to the NRC within 30 days.

I Revision 30 Date 10/30/02 6-34

For any noble gas release, in any period, the dose is taken from Equations B4 and B-5 of Regulatory Guide 1.109:

Dirs = 317E+04 X/Q5 z QsT DF' (6-24)

(mrad) (PCi - yr) (secj (Ci) (mrad - m 3 M3) pCi-yrJ substituting X/Qs = Maximum long term average undepleted atmospheric dispersion factor for a stack release.

= 1.04E-06 sec/m 3 We have (6-23)

DL = 0.033 E QST D1i; (mrad) (pCi - yr) (Ci) (mrad-m3) tCi T3 pCi-yr-)

For the ground level release:

DFaugg = 3.17E+04 (X/Q)g QGL Dp (6-44).

where:

(X/Q)g = Maximum long-term average undepleted atmospheric dispersion factor for a ground level release.

I Revision 30 Date 10/30/02 6-35

= 3.52E-05 seclr3 leading to: . -

Dfirg = 1.12 E: QJL DF (643)

The calculation of ground level release dispersion parameters are based on the location of the North Warehouse with respect to the site boundary that would experience the highest exposure. The North Warehouse contains a waste oil burner that can be used for the incineration of low level contaminated waste oil, and is designated as a ground level release point to the atmosphere. Due to differences in building cross sectional areas and resulting building wake effects, the North Warehouseatmospheric dispersion factors are conservative in comparison to those associated with the main plant facilities, such as the Turbine Building. As a consequence, any potential or unexpected ground level release from the Turbine Building or adjoining structures can utilize the above ground release dose assessment equations.

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 1.109, Rev. 1 (Reference A), except where site-specific models, data or assumptions are more applicable.

Revision 30 Date 10/30/02 6-36

6.9 Method to Calculate the Critical Organ Dose from lodines, Tritium and Particulates Effluent Control 3.3.3 limits the critical organ dose to a Member of the Public-from all release sources of 1-131, I-133, Tritium, and particulates with half-lives greater than 8 days (hereafter called "I+P") in gaseous effluents 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).

6.9.1 Method I Dcos = E QsTp DFGsico (6-25)

(mrem) (Ci) (mrem)

STP 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.

DFGsic. = 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 with the largest dose factor. See Table 1.1.12.

Revision 30 Date 10/30/02 6-37

The critical organ dose is calculated for ground level releases as follows:

Dcog = QGL DFGi, 0 (644)

(mrem) (Ci) (nrem) c_i -

Q = 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" (mremlCi). 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 D + Dc,g -

Equations 6-25 and 6-44 can be applied under the following conditions (otherwise, justify Method I or consider Method 11):

1. Normal operations (not emergency event),

2. I+P releases via the plant stack, Turbine Building, and waste oil burner (see Appendix D for surveillance criteria on waste oil buming), to the atmosphere, and

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

6.9.2 Basis for Method I This section serves three purposes: (1) 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.

I Revision 30 Date 10/30/02 6-38

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" (10CFR50, Appendix I). The use below of asingle "critical receptor" provides part of the conservative margin to the calculation of critical organ dose in Method I. Method I 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 of complying with the Control 3.3.3, maximum annual average atmospheric dispersion factors are 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 perforned 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 1 curie release of each I+P radionuclide to gaseous effluents 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, C4 and C-13 in Reference A).

Tables 6.9.1 and 6.9.2 outline human consumption and environmental parameters used in the analysis. It is conservatively assumed that the critical receptor lives at the "maximum off-site atmospheric dispersion factor location" as defined in Section 6.10. However, he is exposed, conseivatively, 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:

ADico = QiDFGico (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 I in the region of the effluent dose Control limit because it is based on the following reduction Revision 30 Date 10/30/02 6-39

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.

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 1.109, 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.

Revision 30 Date 10/30/02 640

TABLE 6.9.1 Environmental Parameters for Gaseous Effluents at Vermnont Yankee (Derived from Reference A)*

Vegetables Cow Milk Goat Milk Meat Variable Stored Leafy Pasture Stored Pasture Stored Pasture Stored YV Agricultural (K gIm 2

) 2 2 0.70 2 0.70 2 0.70 2 Productivity P Soil Surface (K g/m 2 ) 240 240 240 240 240 240 240 240 Density T Trnsport Time to User5 (H rs) 48 48 48 48 480 480 TB Soil Exposure Time(l) (H rs) 131400 131400 131400 131400 131400 131400 131400 131400 TE Crop Exposure Time to Plume (H rs) 1440 1440 720 1440 720 1440 720 1440 TH Holdup After Harvest (H rs) 1440 24 0 2160 0 2160 0 2160 QF Animals Daily Feed (K g/Day) 50 50 6 6 50 50 FP Fraction of Year on Pasture(2) 0.50 0.50 0.50 FS Fraction Pasture When on 1 1' 1 Pasture(3)

FG Fraction of Stored Veg. 0.76 Grown in Garden FL Fraction of Leafy Veg. Grown 1 in Garden Fl Fraction Elemental Iodine =

0.5 A Absolute Humidity = 5.6 (gm/m3 )(4)

Regulatory Guide 1.109, Revision 1.

I Revision 30 Date 10/30/02 6-41

TABLE 6.9.1 (Continued)

Notes:

(1) 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 8760 hours0.101 days 2.433 hours 0.0145 weeks 0.00333 months (1 year) for all pathways.

(2) For Method H 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 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.

(3) 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.

(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 T is a combination of variables TF and TS in Regulatory Guide 1.109, Revision 1.

I Revision 30 Date 10/30/02 642

TABLE 6.9.2

Usage Factors for Various Gaseous Pathways at VermonrYankee : - -

(from Regulatory Guide 1.109, Table E-5)

Leafy Age Vegetables Vegetables Milk Meat Inhalation Group (kg/yr) (kg/yr) (l/yr) (kg/yr) (m 3 lyr)

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 I Revision 30 Date 10/30/02 6-43

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 1.109 (Reference A) have been considered for radioiodines, tritium, and particulates with half lives greater than 8 days:

1. Direct exposure to contaminated ground;

2. Inhalation of air; 3 Ingestion of vegetables;

4. Ingestion of.cow's milk; and

5. Ingestion of meat.

Beta and gamma air doses have also been considered for noble gases in plant effluents along with whole body and skin dose rate calculations.

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 ReceptorLocations Distances to the site boundary from the two evaluated gaseous release pathways (the Stack and North Warehouse) 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:

1. The point of maximum gamma exposure (maximum gamma X/Q) from an overhead noble gas cloud for determining sldn and whole body dose rates and gamma air doses;

2. The point of maximum ground level air concentration (maximum undepleted X/Q) of noble gases for determining skin and beta air dose rates and doses; I Revision 30 Date 10/30/02 6-44

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

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

The Stack release pathway was evaluated as an elevated release assuming a constant nominal Stack flow rate of 175,000 cfm. The point of maximum gamma exposure from Stack releases (SSE sector, 750 meters) was determined by finding the maximum five-year average gamma XIQ at any off-site location. The location of the maximum ground level air concentration and deposition of radioiodines and other particulates (NW sector, 2700 meters) was determined by finding the maximum five-year average depleted X/Q and D/Q at any off-site location. For the purposes of determining the Method I dose factors for radioiodines, tritium, and particulates, a rmilk animal was assumed to exist at the location of highest calculated ground level air concentration and deposition of radioiodines and other particulates as noted above. This location then conservatively bounds the deposition of radionuclides at all real milk animal locations.

The North Warehouse release pathway was evaluated as a ground level release using the same meteorological period-of-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 off-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 B). 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.

I Revision 30 Date 10/30/02 6-45

AEOLUS-2 produces the following average atmospheric dispersion factors for each location:

1. 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 radioiodines and other 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 radioiodines and other particulates.

The North Warehouse depleted X/Q and DIQ factors were derived using the plume depletion and deposition curves provided in Regulatory Guide 1.111. However, 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 100 meters, the Stack depleted XIQ 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 1 cm/sec.

Gamma dose rate is calculated throughout this ODCM using the finite cloud model presented in "Meteorology and Atomic Energy - 1968 (Reference E, Section 7 5.2.5). That model is implemented through the definition of an effective gamma atmospheric dispersion factor, [X/Q T] (Reference B, Section 4), and the replacement of X/Q in infinite cloud dose equations by the [X/Q].

6.10.3 Long-Term Average Atmospheric Dispersion Factors for Receptors Actual measured meteorological data for the five-year period, 1988 through 1992, 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 values used and their locations are summarized in Table 6.10.1. Table 6.10.1 also indicates which atmospheric dispersion factors are used to calculate the various doses or dose rates of interest.

I Revision 30 Date 10/30/02 6-46

TABLE 6.10.1 Atmospheric Dispersion Factors Release Dispersion Dose to Individual Dose to Air Pathway Factor Total Body Skin Critical Organ Gamma Beta X/Q Depleted 9.40E-07 -

(sec/m 3 ) (2700m NW)

XIQ Undepleted 1.04E-06 1.04E-06 (sec/n 3) (2200m WNW) (2200m WNW)

Stack D/Q 9.40E-09 (I/m2) (2700m NW)

X/QT 6.11E-07 6.1 E-07 - 6. IE-07

-_______(sec/m 3) (750m SSE) (750m SSE) (750m SSE)

X/Q Depleted 3.32E-05 l (sec/m 3 ) (417m NE)

X/Q Undepleted 3.52E-05 3.52E-05 North (sec/m 3 ) (417m NE) (417m I'jE)

Warehouse D/Q l 5.97E-08 (1/m 2) (357m S)

X/Q 6.42E-06 6.42E-06 6.42E-06 (sec/m 3 ) (417m NE) (417m NE) (417m NE)

I Revision 30 Date 10/30/02 6-47

TABLE 6.10.2 Site Boundary Distances Downwind Stack North Warehouse Sector Releases Releases N 400 m 459 m NNE 350m 417 m NE 350m 417m ENE 400 m 451 m B 500 m 570 m ESE 700 m 561 m SE 750 m 612 m SSE 850 m 663 m 385 m 357 m SSW 300 m 238 m SW 250 m 213 m WSW 250im 213 m Ww w 300 m 221 m WNW 400 m 281 m NW 550 m 697 m NNW 550 m 680 m I Revision 30 Date 10/30/02 6 6-48

TABLE 6.10.3 Recirculation Correction Factors A. Stack Releases Sector 0.5 M 1.5M i 2.5 Mi 3.5 Mi 4.5 Mi 7.5 Mi N 1.4 1.4 1.2 1.1 1.0 1.0 NNE 1.8 1.8 1.4 1.2 1.0 1.0 NE 1.8 1.8 1.3 1.1 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 1.1 1.0 1.0 SE 1.8 1.8 1.3 1.2 1.1 1.0 SSE 1.4 1.4 1.2 1.2 1.2 1.2 S 1.3 1.3 1.1 1.1 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 1.1 WSW 2.4 2.4 1.9 1.6 1.5 1.1 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 1.1 NNW 1.5 1.5 1.2 1.2 1.1 1.1 B. North Warehouse Release Sector 0.5 M 1.5 Mi 2.5 Mi 3.5 Mi 4.5 Mi 7.5 Mi N 1.1 1.1 1.1 1.1 1.1 1.0 NNE 1.2 1.2 1.2 1.1 1.1 1.0 NE 1.1 1.2 1.1 1.1 1.0 1.0 ENE 1;2 1.3 1.4 1.4 1.4 1.3 E 1.1 1.3 1.4 1.4 1.4 1.2 ESE 1.1 1.1 1.2 1.1 1.1 1.0 SE 1.0 1.1 1.1 1.1 1.1 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 1.1 1.0 1.0 1.0 1.0 SW 1.2 1.3 1.2 1.0 1.0 1.0 WSW 1.1 1.1 1.0 1.0 1.0 1.0 W 1.2 1.2 1.1 1.0 1.0 1.0 WNW 1.2 1.4 1.3 1.2 1.2 1.0 NW 1.1 1.1 1.0 1.0 1.0 1.0 NNW 1.1 1.2 1.2 1.2 1.2 1.1 I Revision 30 Date 10/30/02 6-49

6.11 Method to Calculate Direct Dose From Plant Operation Effluent Control 3.4.1 (40CFR190) 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 (except the thyroid, which is limited to 75 mrem).

6.11.1 Turbine Building The maximum contribution of direct dose to the whole body or to any organ due to N-16 decay from the turbine is:

De = KN-16 Ktusue Kcat DMsLM (6-27a) where:

Dd = Dose contribution (nrem) from N 16 decay to the site boundary critical receptor (West sector),

KIN-16 1.31E-05 [unitless correlation factor between the site boundary exposure (nR) and the time integral of the average MSLRM readings (mR)],

Ktissue = 0.71 [conversion factor, from radiation exposure in air to tissue dose (mrenmR)],

Kcailb = 478.8 (mR/hr) / [Average MSLRM reading (rnRJhr) for a calibration source of 500 mR/hr during the exposure interval of interest]

[Note: MSLRM calibrations would normally fall within the time interval of interest. In such cases, one approach would be to calculate the time integrals of the MSLRM readings separately for the periods before and after calibration, and then adjust each result by the corresponding calibration correction factor. A second (less complicated, though more conservative) approach would be to use the highest Kalib factor during the interval for both the pre- and post-calibration periods.]

DMSLM = Time integral of the MSLRM average reading (mR) during the interval AT.

I D. flf Th._ 4tflin^m - rn I evision u Late UI5U/U fi-DU

I The last variable is defined as:

n m DMs = [Z(Rj,)/m,At; (6-27b) 1=1I J=1 where Rji = MSLRM reading (mR/hr) by radiation monitor j at sequential time step ti, assumed to be constant during the time subinterval Ati (as defined below),

m = Number of functioning MSLRMs during the subinterval At, Ati = Time subinterval (hr), within AT, defined to span the interval between ti and the next monitor reading, i.e.:

At; = ti+i - ti (6-27c) n = Number of subintervals At1 within AT, such that n

AT = XA (6-27d) i-I During plant shutdown conditions, zero MSLRM readings should be assigned since there is no N16 production in the core. As a conservative altemative, actual readings (reflecting relatively-low background radiation) may be utilized. Also, archived monitor readings which get accidentally lost should be assigned estimated values; a single monitor reading may be provided for this, for any single time interval spanning the entire period with continuous loss in the archived data base.

The dose computed using the above equations may be conservatively applied to the nearest residence. However, it may be corrected for occupancy time (by multiplying the dose by the fraction of time typically spent by the resident at the location during the period of interest), if documented. Dose reduction by the shielding provided by the residential structure may also be considered.

6.11.2 North Warehouse Radioactive materials and low level waste can be stored in the North Warehouse. The maximum annual dose contributions to off-site receptors (west site boundary line) from sources in the shielded (east) end and the unshielded (west) end of the North Warehouse are:

I Revision 30 Date 10/30/02 l

6-51

Ds = 0.25 x ks for the shielded end (6-28)

(mremj (mrem/yr) mTem tyr J mrem/hrJ ( hr) and Du = 0.53 x Ru for theunshielded end (6-29)

Cmrem (mremhr/ CmremD yr ) rmreml/hr) hr J where:

Ds = The annual dose contribution at the maximum site boundary location from fixed sources of radiation stored in the shielded east end of the North Warehouse (

Du = The annual dose contribution at the maximum site boundary location from fixed sources of radiation stored in the unshielded west end of the North Warehouse

- Dose rate measured at 1 meter from the source in the shielded end of the north warehouse ( m Ru = Dose rate measured at 1 meter from the source in the unshielded end of the north warehouse (m) .

(hr 0.25 = Dose rate to dose conversion factor which relates mrem/yr at the west site boundary per mrem/hr measured at 1 meter from I Revision 30 Date 10/30/02 6-52

the source in the shielded end of the warehouse assuming it is full to capacity for one year ( mre ). -y mrem/hr) 0.53 = Dose rate to dose conversion factor which relates mrem/yr at the west site boundary per mrem/hr measured at 1 meter from the source in the unshielded end of the warehouse assuning it is full to capacity for one year E mremhr) 6.11.3 Low Lvel Waste Storage Pad Interim storage of packaged Dry Active Waste (DAW) and spent ion exchange and filter media is permitted in modular concrete storage overpacks on the LLW storage pad facility adjacent to the north warehouse. The arrangement of the storage modules is such that DAW is placed in modules which shield higher activity ion exchange media from the west site boundary.

The dose at the maximum site boundary receptor from both direct radiation and skyshine scatter can be calculated as follows:

(a) Direct Dose (line of si_ht)

Dds 0.28 X fd X fd mrem Emremlyr (#)

(6-30)

Cyr- module) mrem/hr (MrE) or DdS = 0.39 X Rd fd (6-31)

Cmrem ( mrem/yr) (mrem) (#)

yr- module) mremn/hr) 17hr where:

I Revision 30 Date 10/30/02 6-53

DdE = The annual direct dose contribution at the maximum site boundary from a single rectangular storage module which has an unobstructed short end surface (not shielded by other modules) orientated toward the west site boundary ( em J yr- module)

Dds = The annual direct dose contribution at the maximum site boundary from a single rectangular storage module which has an unobstructed long side surface (not shielded by other modules) orientated toward the west site boundary r yr- module)

Ad = Maximum dose rate measured at 3 feet from the side of the storage module whose unobstructed face (i.e., a side or end surface which is not shielded by other waste modules) is toward the west site boundary.

fd = The fraction of a year that a storage module is in use on the storage pad.

0.28 = Dose rate to dose conversion factor which relates mrem/yr at the west site boundary per mrem/hr measured at 3 feet from the narrow end of the rectangular storage module when that face is orientated toward the' west boundary.

0.39 = Dose rate to dose conversion factor which relates mrenlyr at the west site boundary per mrem/hr measured at 3 feet from the long side of the rectangular storage module when that face is orientated toward the west boundary.

I Revision 30 Date 10/30/02 6-54

(b) Scatter From Skyshine

.E

- (6-32)

DsKR = 0.016 x R x fsk C rem )

yr- liner)

(mreyr Irnrem/hr (em hr )

(#)

and (6-33)

DsKD 0.015 X RSKD X fSK C rnrem ' (mrem/yr) (mrem) hrJ

(#)

wyr-module) mrenhr where:

RsKR = The annual skyshine scatter contribution to the dose at the maximum site boundary from a single spent ion exchange media liner in a storage module whose top surface is not obstructed due to stacking of modules mrem yr- liner)

Rsya = The annual skyshine scatter contribution to the dose at the maxinum site boundary from a rectangular storage module containing DAW whose top surface is not obstructed due to stacking of modules (Imem

- For Resins, the maximum dose rate measured at 3 feet over the top of each liner in a storage module (mrenVhr).

A-SKD = or DAW, the maximum dose rate measured at 3 feet over the top surface of a storage module with DAW (mrem/hr).

I Revision 30 Date 10/30102 6-55

fsK The fraction of a year that a storage module is in use on the storage pad.

0.016 Dose rate to dose conversion factor for the scatter dose from each resin liner source in storage which relates mren/yr at the west site boundary per mrem/hour at 3 feet from the top of the module.

0.015 = Dose rate to dose conversion factor for the scatter dose from DAW boxes in storage which relates mrem/yr at the west site boundary per mrem/hr at 3 feet from the top of the module.

(c) Dose From Resin Liners During Transfer During the movement of resin liners from transfer casks to the storage modules, the liners will be unshielded in the storage pad area for a short period of time.

The maximum dose contribution at the site boundary during the unshielded movement of resin liners can be calculated from:

Dtan = 0.0025 x R Xaz x Ttrans (mrem) (mrenfhr) (a) (hr) (6-34) rad/hr hr where:

Dn = The dose contribution to maximum site boundary resulting from the unshielded movement of resin liners between a transfer cask and a storage module (mrem).

= Dose rate measured at contact (2") from the unshielded top surface of the resin liner in R/hr.

Ttan = The time (in hours) that an unshielded resin liner is exposed in the storage pad area.

I Revision 30 Date 10/30/02 6-56

0.0025 = The dose rate to dose conversion factor for an unshielded resin liner which relates mrem/hour at the west site boundary per rad/hr at contact (2") from the unshielded surface of the liner.: -

(d) Intennodular Gap Dose In addition to the above methods for determining doses at the west site boundary from the LLW storage pad, another dose assessment model has been included to address the possible condition of spaces or gaps existing between the placement of the DAW storage modules situated along the west facing side of the pad. This could result in a radiation streaming condition existing if ion exchange resin liners were placed in storage directly behind the gap. The direct dose equations (6-30 and 6-31) consider that the storage modules situated on the outside of the pad area provide a uniform shield to storage modules placed behind them. The intermodular gap dose equation (6-35) accounts for any physical spacing between the outside storage modules which have not been covered by additional external shielding.

(6-35)

DGap = 2.44E- 2 X WGap X ARL X f Gap (mremj f rnremJ (in) (Ci) (#)

where:

DGap The annual dose contribution at the maximum site boundary (west) from radiation streaming through the intermodular gap between DAW storage modules used to shield resin modules from direct radiation (mrern/yr).

Wcap = The intermodular gap width (inches) between adjacent DAW storage modules facing the west site boundary.

ARL = The total gamma activity contained in a condensate resin liner stored directly in line with the intermodular gap adjacent DAW modules ( Ci).

I Revision 30 Date 10/30/02 6-57

fGap = The fraction of a year that the intermodular gap is not shielded.

2.44E-2 The activity to site boundary dose conversion factor J

= -

_rem for a one-inch wide intermodular gap Kyrn The site boundary dose from waste materials placed into storage on the Low Level Waste Storage Pad Facility is determined by combining the dose contribution due to direct radiation (line of sight) from Part (a) above with the skyshine scatter dose from Part (b), resin liner transfer dose from Part (c), and any intermodular gap dose from Part (d).

6.11.4 Total Direct Dose Sunmary The dose contributions from the N-16 source in the Turbine Building, fixed sources in the North Warehouse, and fixed sources on the Low Level Waste Storage Pad Facility, shall be combined to obtain the estimate of total off-site dose to any member of the public from all fixed sources of radiation located on-site.

6.11.5 Other Fixed Sources In addition to the fixed sources noted above (Turbine Building, North Warehouse, and LLW Storage Pad), other identified temporary or fixed sources that are created due to plant operations will be included in the total direct summary of 6.11.4 if the projected annual dose contribution would add any notable addition to the reported total (i.e., > 0.1 mremlyr).

In 1995, turbine rotors and casings were replaced in the Turbine Hall with the old rotors and casings placed in storage sheds located on site west of the switchyard along the railroad spur.

Radiation surveys (December 1995) of low level contamination (principally Co-60) on the components led to a projected maximum west site boundary dose of 0.2 mrem/yr. This contribution will be added to the maximum site boundary total dose until the contribution is less than 0.1 mrem/yr, or the components are removed from storage location.

I Revision 30 Date 10/30/02 6-58

6.12 Cumulative Doses Cumulative Doses for a calendar quarter and a calendar year mustbe maintained-to demonstrate a compliance with Controls 3.2.2, 3.3.2, and 3.3.3 (10CFR5O, Appendix I dose 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 40CFR190). To ensure the limits are not exceeded, a running total must be kept for each release.

Demonstration of compliance with the dose limits of 40CFR190 is considered as demonstrating compliance with the 0.1 rem limit of 1 OCFR20.1301(a)(1) for members of the public in unrestricted areas.

I I

fl evision fAD_

l fl T A .

u dlate 14t%b.lt,%

UIiUIU n O-Z19

7.0 ENVIRONMENTAL MONrrORING PROGRAM I 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 7-6.

7.1 IntercomRarison Program All routine radiological analyses for environmental samples are performed at a contracted environmental laboratory. The contracted laboratory participates 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 include the Department of Energy - Environmental Measurements Laboratory Quality Assessment Program (EMLQAP),

Department of Energy - Mixed Analyte Performance Evaluation Program (MAPEP), Analytics Cross-Check Program - Environmental Inter-laboratory Cross-Check Program, and Environmental Resources Association - Environmental Radioactivity Performance Evaluation Program.

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 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 needless expense and destroying the continuity of the program, long term, 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.

Revision 27 Date 10/09/00 7-1 I

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 often the wind blows into that sector), and is approximately 0.5 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 DIQ 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 DIQ for any location in any sector). Station AP/CF-1 1 is approximately one mile from this point, between it and the plant. Station AP/CF-14 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, Vermont, 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 (APICF-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.

An additional air sampler is maintained at the Tyler Hill site (AP/CF-15, WNW, 3.1 km),

which is along the westem side of the valley in general proximity of historical dairy operations.

(The sixth location is not a specific Program requirement as detailed in Table 3.5.1 .) I Revision 27 Date 10/09/00 7-2 I

7.3 Distances and Directions to Monitoring Stations I 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, are keyed to the center of the Turbine Building due to the critical nature of the Turbine 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 Turbine Building; The distances to the plant stack and Turbine Building will, therefore, be used in the Annual Radiological Environmental Operating Reports and ODCM for the sampling and TLD monitoring stations, respectively.

Revision 27 Date 10/09/00 7-3 I

Table 7.1 I

Radiological Environmental Monitoring Stations(l)

Exposure Pathway Sample Location Distance and/or S121l and Designated Code(2 ) (km) 5)

Direction(

1. AIRBORNE (Radioiodine and Particulate)

AP/CF-1 1 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 APICF-14 Northfield, MA 11.61 SSE APICF-15 Tyler Hill Road(4) 3.14 WNW AP/CF-21 Spofford Lake 16.36 NNE

2. WATERBORNE

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

b. Ground WG-11 Plant Well 0.24 On-Site WG-12 Vernon Nursing Well 2.13 SSE WG-13 COB Well( 4) 0.26 On-Site WG-14 Plant Support Bldg Well(4) 0.27 On-Site WG-22 Skibniowsky Well 13.73 N

c. Sediment SE-1I Shoreline Downriver 0.57 SSE From SE-12 North Storm 0.13 E Shoreline Drain Outfall(3 )

3. INGESTION

a. Milk(8) TM-II Miller Farm 0.82 W I TM-14 Brown Farm 2.22 S TM-18 Blodgett Farm 3.60 SE I TM-22 Franklin Farm (4) 9.73 WSW TM-24 County Farm 21.64 N TM-25 Downey-Spencer(4 ) 6.90 W TM-26 Cheney Hill Farm 7.53 WNW

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, MA 11.61 SSE TG-15 Tyler Hill Rd. (4) 3.07 WNW TG-21 Spofford Lake 16.36 NNE Revision 27 Date 10/09/00 7-4 I

Table 7.1 (Continued)

Radiological Environmental MonitorinZ Stations"'

Exposure Pathway Sample Location Distance and/or S ale and Designated Code 2) (km Direction

c. Silage TC-11 Miller Farm 0.82 W TC-14 Brown Farm 2.22 S TC-18 Blodgett Fann 3.60 SE TC-22 Franklin Farm 4 ) 9.73 WSW TC-24 County Farm 21.64 N TM-25 Downey-Spencer( 4 ) 6.90 W TM-26 Cheney Hill Farm 7.53 WNW

d. Fish FH-1 1 Vemon Pond (6) (6)

FH-21 Rt. 9 Bridge 11.83 Upriver

4. DIRECT RADIATION DR-1 River Station No. 3-3 1.61 SSE DR-2 N. Hinsdale, NH 3.88 NNW DR-3 Hinsdale Substation 2.98 E DR-4 Northfield, MA 11.34 SSE DR-5 Spofford Lake 16.53 NNE DR-6 Vernon School 0.52 WSW DR-7 Site Boundary(7) 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-1I Inner Ring 1.65 NNE DR-12 Outer Ring 3.58 NNE DR-13 InnerRing 1.23 NE DR-14 Outer 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 Revision 27 Date 10/09/00 7-5 I

Table 7.1 (Continued)

Radiological Environmental Monitoring Stations' Exposure Pathway Sample Location Distance and/or Samp and Designated Code( 2 ) (km)(5)

Direction DR-26 Outer Ring 3.77 S DR-27 Inner Ring 1.10 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 DR-36 Outer Ring 4.43 WNW DR-37 Inner Ring 2.76 NW DR-38 Outer Ring 7.34 NW DR-39 InnerRing 3.13 NNW DR-40 Outer 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 all except milk, silage and the direct radiation stations).

(3) To be sampled and analyzed semiannually.

(4) Non-required Control station.

(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 3.5.1 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.

Revision 27 Date 10/09/00 7-6 I

SW SSW Figure 7-1 Environmental Sampling Locations in Close Proximity to the Plant Revision 27 Date 10/09/00 7-7 I

Figure 7-2 Environmental Sampling Locations Within 5 Km of Plant I

Revision 27 Date 10/09/00 7-8 I

Figure 7-3 Environmental Sampling Locations Greater than 5 Km from Plant I Revision 27 Date 10/09/00 7-9 I

i\ -

NW NE I

[I SW SSW Figure 7-4 TLD Locations in Close Proximity to Plant Revision 27 Date 10/09/00 7-10 I

Figure 7-5 TLD Locations Within 5 Km of Plant Revision 27 Date 10/09/00 7-11 I

Figure 7-6 TLD Locations Greater Than 5 Km from Plant I Revision 27 Date 10/09/00 7-12 I

8.0 SETPOINT DETERMINATIONS Chapter 8 contains the basis for plant procedures used to meet the setpoint requirements of the Radioactive Effluent Instrumentation Controls. They are Control 3.1.1 for liquids and Control 3.1.2 for gases. Each outlines the instrumentation channels and the basis for each setpoint.

I Revision 29 Date 1(11/02 8-1

8.1 Liquid Effluent Instrumentation Setpoints Control 3.1.1.1 requires that the radioactive liquid effluent instrumentation 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 10CFR20.1001 through 20.2402, and a total noble gas concentration limit of 2E-04 giCi/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.

8.1.1.1 Method to Determine the Setpoint of the Liquid Radwaste Discharge Monitor (RM-17-350)

The instrument response (in counts per second) for the lirniting concentration at the point of discharge is the setpoint, denoted R5 ,P 01,.,, and is determined as follows:

(8-1)

R ~ DF S, cm setpoint DFm n (cps) (#) (cps-ml J(ci Where:

DF F = Dilution factor (as a conservative measure, a DF of at least 1000 is used) (dimensionless) (8-2)

Flow rate past monitor (gpm)

Fd = Flow rate out of discharge canal (gpm)

DFd = Minimum allowable dilution factor (dimensionless)

i. ECL cnl(8-3)

Revision 29 Date 1/11/02 8-2

ECLI = Effluent concentration values for radionuclide "i" from I 10CFR20.1001-20.2402, Appendix B, Table 2, Column 2 (pci/mI)

Cr,n, = Activity concentration of radionuclide "i" in mixture at the monitor (pCi/ml)

Si = Detector counting efficiency from the most recent liquid radwaste discharge monitor calibration curve (cps/(.Ci/ml))

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, S, of 4.9E+06 cps per I ,Ci/ml of gamma emitters which emit one photon per disintegration.

The activity concentration of each radionuclide, C=, in the floor drain sample tank is determined by analysis of a representative grab sample obtained at the radwaste sample sink.

This setpoint example is based on the following data:

i Cm, (Ci/nd ECL (Ci/ml)

Cs-134 2.15E-05 9E-07 Cs-137 7.48E-05 IE-06 Co-60 2.56E-05 3E-06 I Revision 29 Date 1/1 1/02 8-3

= 2.15E-05 + 748E-05 + 2.56E-05 i

flci (pcq cci) t mlJ ml ) mlJ nil)

= 1.22E-04 9ci) mlJ (8-3)

DFmn = 0.12 m,CJ

ECLi RCi - ml) ml - ICiJ

= 01i r 2.15E-05 +

7.48E-05 2.56E-05 1

+ I L 9E-07 C

1E-06 3E-06 CjCi-ml Ici-ml CiCi -ml rml-.Ci) ml -[Ji ml -liCi )

= 10.7 The minimum dilution factor, DF7m, needed to discharge the mixture of radionuclides in this example is 10.7. As a conservative measure, an actual dilution factor, DF, of 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 of dilution water. With the dilution flow taken as 18,000 gpm, the release rate from the floor drain sample tank may be determined as follows:

Fd P = (8-4)

(gpm) (gpm) 18,000 gpm = 18 gpm 1,000 I Revision 29 Date 1/1 1/02 84

Under these conditions, the setpoint of the liquid radwaste discharge monitor is:

(8-I)

R sepit= DF DF Si z Cmi (cps) ( p) -nml cps (ci 1,000 4.9E + 06 1.22E - 04 10.7 (CpS) ( 11~~Ci ) MIrl

= 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 waming before exceeding Control limits.

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 = Cm, Sli (cps) (

Where:

R = Response of the monitor (cps)

Si = Detector counting efficiency for radionuclide "i" (cps/(Ci/rml))

Activity concentration of radionuclide "i" in mixture at the monitor (Aci/ml)

Revision 29 Date 1111/02 8-5

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 Si for SW 1, where S, represents the counting efficiency determined for the current mix of nuclides. If the mix of nuclides changes significantly, a new counting efficiency should be determined for calculating the setpoint.

(8-6)

R = S1 SC m i

(cps) cs Ci ) (ml)

The effluent concentration for a given radionuclide must not exceed 10 times the 10 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)

Cd 10 ECLi

("Ci- ml) ml - gCi)

Where:

Cd; = Activity concentration of radionuclide "i" in the mixture at the point of discharge to an unrestricted area (Cilml)

ECLi = Effluent concentration limit for radionuclide "i" from 10CFR20.1001-20.2402, Appendix B, Table 2, Column 2 (Cilml)

The activity concentration of radionucide "i" at the point of discharge is related to the activity concentration of radionuclide "i" at the monitor as follows:

(8-8) di Cmi Fd (uCi) (Ci gpm ml ( ml gpm Revision 29 Date 111/02 8-6

Where:

Cdi = Activity concentration of radionuclide "i" in the mixture at the point of discharge (iCi/ml)

F. = Flow rate past monitor (gpm)

Fd = Flow rate out of discharge canal (gpm)

Substituting the right half of Equation 8-8 for Cd; in Equation 8-7 and solving for FdFm yields the minimum dilution factor needed to comply with Equation 8-7:

(8-3)

DFmin 5 ECL 10 Cgpm r PCi-ml gpm ml-p.Ci J Where:

Ed Flow rate out of discharge canal (gpm)

Pm = Flow rate past monitor (gpm)

C:,,, = Activity concentration of radionuclide "i" in mixture at the monitor (Ci/ml)

ECL, = Effluent concentration limit for radionuclide "i" from 10CFR20.1001-20.2402, Appendix B, Table 2, Column 2 (tCi/hl) 10 = The instantaneous concentration multiplier allowed by Control 3.2.1 If Fd/Fm is less than DF., then the tank may not be discharged until either Pd or F. or both are adjusted such that:

(8-3) d - DFin Fm Cgpm gpmJ Revision 29 Date 1/11/02 8-7

Usually Fd[Fm is greater than DFi,, (i.e., there is more dilution than necessary to comply with Equation 8-7). he response of the liquid radwaste discharge monitor at the setpoint is therefore:

(8-1)

Rsetpoint = D S Cmi DFmin (cps) cpll ml )c 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 alarm setpoint 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 reactor coolant. From routine coolant sampIe gamma isotopic analyses, a Composite Maximum Permissible Concentration (CMPC) is calculated as follows:

C(f JMCI + fV'WC 2 ...) = C/CMPC or CMPC = /(f 1 IMPC1 + f2 /MPC 2 ...) (8-22) where:

C = Total concentration of detected radioactivity in reactor coolant sample (Pci/ml) fi = Fraction of total radionuclide concentration represented by the ith radionuclide in the mix PCi = Maximum Permissible Concentration limit for radionuclide "i" as listed in 10CFR20.106, Appendix B, Table II, Column 2 (iCi/ml)

The Composite Effluent Concentration Limit (CECL) is also calculated using the equation above by substituting the appropriate ECL value from OCFR20.1001-20.2402, Appendix B, Table 2, Column 2, for MPC.

Revision 29 Date 1111102 8-8

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 inoperable, and daily SWM grab samples are collected and analyzed until the calculated coolant CMPC (or CECL) equivalent count rate is above the SWM's alarm setpoint.

For example, if the reactor coolant radionuclide mix distribution is as listed below, then the corresponding CMPC is calculated as follows:

fi I (conc/total 10CFR20 MPC; fj/MPC1 Nuclides Conc (Ci/nl) conc) (WCi/ml) (mI/pCi)

I-131 6.OOE-6 6.59E-2 3.OE-7 2.20E+5 I-133 5.OOE-6 5.49E-2 L.OE-6 5.49E+4 Co-60 8.OOE-5 8.79E-1 3.OE-5 2.93E+4 Totals 9.JOE-5 1.00 3.04E+5 CMPC = 1/3.04E+5 = 3.29E-6 (Ci/ml)

The CECL is also calculated by using the above methodology and substituting the appropriate ECL listed in 10CFR20.1001-20.2402, Appendix B, Table 2, Column 2, for MPC values. For this example, the calculated CECL is equal to 2.73E-6 gCi/fl.)

If the SWM alarm is set at 5 CPS (300 CPM) above background, and the current calibration factor for this monitor is 1.17E+8 CPM/,Ci/rnl, then the SWM will alann if a concentration as low as 2.56E-6 ,uCi/ml above background passes by the monitor. Since the most limiting CMPC or CECL (calculated above to be 2.73E-6 pCi/nil) is above the alarm setpoint (equal to 2.56E-6 IpCi/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 be 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.

Also, service water is sampled if the monitor is out of service or if the alarm sounds.

Revision 29 Date 1/11/02 8-9

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 I effluent concentration limits specified in OCFR20.1001-20.2402, Appendix B, Table 2, Column 2.

Revision 29 R Date 1/1 1/02 8-10

8.2 Gaseous Effluent Instrumentation Setpoints Control 3.1.2 requires that the radioactive gaseous effluent instrumentation in Control Table 3.1.2 have their alarm setpoints set to ensure that Technical Specification 3.8.K.1 and Control 3.3.1 are not exceeded. Technical Specification 3.8.K.1 (and Control 3.3.7) limits the gross radioactivity release rate at the steam jet air ejector (SJAE) to 0.16 Ci/sec.

8.2.1 Plant Stack Noble Gas Activity Monitors (RM-17-156 and RM-17-157) and Augmented Off-Gas System Noble Gas Activity Monitors (RAN-OG-3127 and RAN-OG-3128)

The plant stack and AOG noble gas activity monitors are shown on Figure 9-2.

8.2.1.1 Method to Determine the Setpoint of the Plant Stack Noble Gas Activity Monitors (RM-17-156 and RM-17-157) and the Augmented Off-Gas System Noble Gas Activity Monitors (RAN-OG-3127 and RAN-OG-3128)

The setpoints of the plant stack and AOG system noble gas activity monitors are determined in the same manner. The plant stack or AOG system noble gas activity monitor response in counts per minute at the limiting off-site noble gas dose rate to the total body or to the skin is the setpoint, denoted Rst. Rp is the lesser of:

(8-9)

Rtb- 1 1 Rspt - 818 S9 F DFBC cpm -cm3) (sec ( pCi-yr (cpm) y - Ci - )

cm3 )mrem-m 3 )

and: (8-10) 1 1 Rs1n spt = 3,000 S9 F DFS (Cpm) mrem (cpm -cm )( sec')( ,uCi-yr t FCi ) cm 3) Tmrem -secJ I Revision 29 Date 1/11/02 8-11

where:

Rtbspt = Response of the monitor at the limiting total body dose rate (cpm) 500 f rem - ,uCi - ml 818 (1E+06) (6.11E-07) yr-pCi-sec 500 = Lirniting total body dose rate (mrem/yr) 1E+06 = Number of pCi per gCi (pCi/4Ci) 6.11E-07 = [X/Q], maximum five-year average gamma atmospheric dispersion factor (sec/m3)

Sg = Appropriate (plant stack or AOG system) detector counting efficiency from the most recent calibration (cpm/(gCi/cc))

F = Appropriate (plant stack or AOG system) flow rate (cm3 /sec)

DFBc = Composite total body dose factor (mrem-m3/pCi-yr)

(8-1 1) z QDFBi fi Qi = The relative release rate of noble gas "i" in the miixture at the monitor (either the stack QsT or the AOG, QAOG ) for noble gases identified (Ci/sec)

DFB1 - Total body dose factor (see Table 1.1-10) (mrem-m3/pCi-yr)

Rskln spt = Response of the monitor at the limiting skin dose rate (cpm) 3,000 - Limiting skin dose rate (mrem/yr)

DF'c = Composite skin dose factor (inrem-sec/gCi-yr)

I Revision 29 Date 1/11/02 8-12

(8-12)

E i Dels

= i i

Des = Combined skin dose factor (see Table 1.1.10) (mrem-sec/gCi-yr) 8.2.1.2 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 8-9 and 8-10 for detennining setpoints.

The plant stack noble gas activity monitors, referred to as "Stack Gas I" (RM-17-156) and "Stack Gas II' (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 3E+07 cpm per pCi/cc of noble gas. The nominal plant stack flow is 7.32E+07 cc/sec ((155,000 cfm x 28,300 cc/ft)160 sec/min).

When monitor responses indicate that activity levels are below the LLDs at the stack (or AOG) monitors, the relative contribution of each noble gas radionuclide can conservatively be.

approximated by analysis of a sample of off-gas obtained during plant operations at the steam jet air ejector (SlAE). This setpoint example is based on the following data (see Table 1.1.10 for DFBi and DF'):

~JAE DFBi sci irem - m3 Cmrem- sec i secJ pci - yr ) tCi - yr Xe-138 1.03E+04 8.83E-03 1.06E-02 Kr-87 4.73E+02 5.92E-03 1.43E-02 Kr-88 2.57E+02 1.47E-02 1.28E-02 Kr-85m 1.20E+02 1.17E-03 2.35E-03 Xe-135 3.70E+-2 1.81E-03 3.24E-03 Xe-133 1.97E+01 2.94E-04 5.58E-04 I Revision 29 Date 1/11/02 8-13

(8-11)

I O§JAED DFBC = i .

i IQSJAEDFB = (1.03E+04) (8.83E-03) + (4.73E-02) (5.92E-03) i

+ (2.57E+02) (1.47E-02) + (1.20E+02) (1.17E-03)

+ (3.70E+02) (1.81E-03) + (1.97E+01) (2.94E-04)

= 9.83E+01 (Ci-mrem-m /sec-pCi-yr)

SiAE I I = 1.03E+04 + 4.73+02 + 2.57E+02 i

+ 1.20E+02 + 3.70E+02 + 1.97E+01

= 1.15E+04 p.Cilsec 9.83E + 01 1.15E+04

= 8.52E-03 (mrem-n/pCi-yr) 1 1 Rtb sin = 818 Sg F DFBC

= (818) (3E+07) 1 1 (7.32E + 07) (8.52E - 03)

= 39.348cpm Next: (8-11)

I &SJAE DF-D1C = '

(SJAE i

I Revision 29 Date 1/11/02 8-14

,QSJAEDpD = (1.03E+04) (1.06E-02) + (4.73E-02) (1.43E-02)

+ (2.57E+02) (1.28E-02) + (1.20E+02) (2.35E-03)

+ (3.70E+02) (3.24E-03) + (1.97E+01) (5.58E-04)

= 1.14E+02 (Ci-nrem-sec/sec-jCi-yr)

DF 1.14E+02 1.15E + 04

= 9.91E-03 (mrem-seclgCi-yr)

RkIdn = 3,000 Sg 1 DF

= (3,000) (3E + 07) 1 1 (7.32E + 07) (9.91E -03)

= 124,067 cpm The setpoint, Rs, is the lesser of Rtb and Rn. For the noble gas mixture in this example Rth is less than Rski , indicating that the total body dose rate is more restrictive.

Therefore, in this example the "Stack Gas I" and "Stack Gas I' noble gas activity monitors should each be set at some administrative value below 39,348 cpm above background to provide conservatism for such issues as instrment 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 rainor releases from the waste oil burner. For example, if an administrative limit of 70 percent of the Control whole body dose linit 500 mrem/yr (39,348 cpm) is chosen, then the noble gas monitor alarms should be set at no more than 27,543 cpm above background (0.7 x 39,348 = 27,543).

8.2.1.3 Basis for the Plant Stack and AOG System Noble Gas Activity Monitor Setpoints The setpoints of the plant stack and AOG system noble gas activity monitors must ensure that Control 3.3.1.a is not exceeded. Sections 6.4 and 6.5 show that Equations 6-5 and 6-7 are acceptable methods for determining compliance with the Control limits. Which equation (i.e., dose to total body or skin) is more limiting depends on the noble gas mixture. Therefore, Revision 29 Date 1/11/02 8-15

each equation must be considered separately. The derivations of Equations 8-9 and 8-10 begin with the general equation for the response R of a radiation monitor (8-13)

R- Sgi CnI

- cm3 j (cpm)cpm (cpm) t ci (P.CiA cm3 where:

R = Response of the instrument (cpm)

Sg = Detector counting efficiency for noble gas "i" (cpm/(PtCi/cm 3 ))

Cmi= Activity concentration of noble gas "i" in the mixture at the noble gas activity monitor (Ci/cm 3 )

The relative release rate of each noble gas (; jiCi/cm3 ), in the total release rate is normally determined by analysis of a sample of off-gas obtained at the Steam Jet Air Ejector (SJAE). Noble gas release rates at the plant stack and the AOG discharge are usually so low that the activity concentration is below the Lower Limit of Detection (LLD) for sample analysis. As a result, the release rate mix ratios measured at the SJAE are used to represent any radioactivity being discharged from the stack, such as may have resulted from plant steam leaks that have been collected by building ventilation. For the AOG monitor downstream of the charcoal delay beds, this leads to a conservative setpoint since several short-lived (high dose factor) noble gas radionuclides are then assumed to be present at the monitor, which in reality, would not be expected to be present in the system at that point. During periods when the plant is shutdown (after five days), and no radioactivity release rates can be measured at the SJAE, Xe-133 is the dominant long-lived noble gas and may be used as the referenced radionuclide to determine off-site dose rates and monitor setpoints. Alternately, a relative radionuclide, "i", mix fraction, (f;),

may be taken from Table 8.2.1 as a function of time after shutdown (including periods shorter than five days) to determine the relative fraction of each noble gas potentially available for release to the total. However, prior to plant startup before a SJAE sample can be taken and analyzed, the monitor alarm setpoints should be based on Xe-138 as representing the most prevalent high dose factor noble gas expected to be present shortly after the plant retums to power. Monitor alarm setpoints which have been deternined to be conservative under any plant conditions may be utilized at any time in lieu of the above assumptions. Ci, the activity Revision 29 Date 1/11/02 8-16

concentration of noble gas "i" at the noble gas activity monitor, may be expressed in terms of Q by dividing by F, the appropriate flow rate. In the case of the plant stack noble gas activity monitors the appropriate flow rate is the plant stack flow rate and for the AOG noble gas activity monitors the appropriate flow rate is the AOG system flow rate.

(8-14)

Crni = i 1 F

"Ci) lCi) (sec) cm3 sec cm 3) where:

Qi = The release rate of noble gas "i" in the mixture for each noble gas identified (jCi/sec).

F = Appropriate flow rate (cm /sec)

Substituting the right half of Equation 8-14 into Equation 8-13 for Cmi yields:

(8-15)

R =2 giQ F cpm) (cpm - cm3 )(Ci)(sec (c [LtCi sec cm3)

The detector calibration procedure establishes a counting efficiency for a reference radionuclide, Xe-133 (half life 5.24 days). For routine conditions where offgas is processed through the AOG, all short lived gases are decayed away before discharge leaving only long lived radionuclides as the significant contributors to the monitor response. In this case, Xe-133 as the reference radionuclide for the detector counting efficiency is representative of the expected release conditions. For off normal conditions that might lead to inclusion of short lived radioactivity in the gas stream being released, Xe-133 as the reference radionuclide is expected to lead to a conservative response factor for the detectors since the short lived noble gases tend to have higher energies that can cause them to over respond. Therefore, in Equation 8-15, one may substitute S, for Sg, where Sg represents the detector counting efficiency determined from the Xe-133 calibration. If necessary, the actual concentration and discharge rate of individual Revision 29 Date 1/11/02 8-17

gases being released from the stack (or AOG) can be determined by direct grab sample and laboratory analysis during specific periods of interest.

3-16)

Sg R- 1 Qi (cpm) t pCi

( sec-cpmcm cm 3 (sec

)

The total body dose rate due to noble gases is determined with Equation 8-5:

(8-5)

Rtbs = 0.61 Qi DFBi Umre

( pCi -sec Ci-m 3 (JCi i sec n

(mrem-m pCi-yr 3

Where:

R tbs = total body dose rate (nremlyr) due to noble gases from stack releas, ,

0.61 = (1.OE+06) x (6.I1E-07) (pCi-sec/,Ci-rn3 )

IE+06 = number of pCi per p,Ci (pCi/pCi) 6.1 IE-07 - X/OI maximum long term average gamma atmosDheric dispersio )n 3 aW, , w I factor (sec/m)

Qi = the release rate of noble gas "i" in the mixture for each noble gas identified (Cisec) (Equivalent to Q&T for noble gases released at the plant stack.)

DFBi = total body dose factor (see Table 1.1.10) (mrem-m /pCi-yr)

Revision 29 Date 1/11/02 8-18

A composite total body gamma dose factor, DFBC, may be defined such that:

(8-17)

DEBC = Z Qi = i DFBi

[mrem-m 3 pCi-yr )

(pCi )

1sec)

(Cic sec

[xnrem-m 3]

pCi - yr Solving Equation 8-23 for DFBC yields:

(8-11)

IQi DFB1 DFBC =

Qi Control 3.3.1.1.a limits the dose rate to the total body from noble gases at any location at

orbeyond the site boundary to 500 mrem/yr. By setting equal to 500 mrem/yr and substituting DFBC for DFBi in Equation 8-5, one may solve for F, Q at the limiting whole body noble gas i

dose rate:

(8-18) 1 818 i= DFBC CLCi sec)

' (mrem-,lCi-m3 yr- pCi - sec ) C pCi -

nrem-m3) yr Substituting this result for I Qi in Equation 8-16 yields Rtb the response of the monitor at the i t limiting noble gas total body dose rate:

(8-9) 1 1 Rtbspt - 818 Sg DFBC Lcpm-cm3 )

(cpm)

(mrem - [tCi - m3 yr - pCi - sec gcm )cm sec

) Cmrem pCi -yr i 3)

Revision 29 Date 1/1 1/02 8-19

The skin dose rate due to noble gases is determined with Equation 6-7:

(6-7) sdn= Qi D1s i

rnrem (gzCi( inrem-sec) yr sec gCi - yr Where:

Rskin = Skin dose rate (mrem/yr)

QI = The release rate of noble gas "i" in the mixture for each noble gas identified (Cilsec) equivalent to Q&T for noble gases released at the plant stack).

DF'. = Combined skin dose factor (see Table 1.1.10) (rem-sec/jCi-yr).

A composite combined skin dose factor, DFC, may be defined such that:

(8-19)

DFiCDF i Q =2 i~

Q1 DFs FI Cmrem -

tpLCi-yr sec')

)

(pCi Lsec

( iCi sec r(mrem - sec) pCi-yr )

Solving Equation 8-19 for DF' yields:

QiD}js i

Revision 29 Date 1/11/02 8-20

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 nrem/yr. By setting Rsldn equal to 3,000 mrem/yr and substituting DF: for De in Equation 6-7 one may solve for E Qj at the limiting skin noble gas i

dose rate:

1 Q = 3,000 i DFc' AlCi mrem Ci - yr sec yr mrem - sec Substituting this result for ,QI in Equation 8-16 yields RSbn , the response of the i Spt monitor at the limiting noble gas skin dose rate:

(8-10) 1 1 R sidn spt = 3,000 S9 P DPFc

[mrem~cpm - cm3) sec Ci - yr')

cm3 rarem-sec)

Revision 29 Date 1/11/02 8-21

TABLE 8.2.1 Relative Fractions of Core Inventory Noble Gases After Shutdown Time Kr-83m Kr-85m Kr-85 Kr-87 Kr-88 Xe-13 1m Xe-133m Xe-133 Xe-135m Xe-135 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 5d*t< lOd .007 .008 .016 .969 lOdct< 15d -- : --- .014 .014 .006 .966 _-

15 d<t<20d --- .026 .022 .002 .950 20 dct<30d --- .048 .034 .001 .917 30 d <t <40 d ---

-- .152 .070 .777 40 d*t<50d .378 .105 .517 50 d<t<60d .652 .108 .240 60 dct<70d --- .835 .083 .082 t >70 d -- --- .920 .055 .024 Revision 29 Date 1/11/02 8-22

8.2.2 Steam Jet Air Ejector (SJAE) Noble Gas Activity Monitors (RM-17-150A and RM-17-150B)

The SJAE noble gas activity monitors are shown in Figure 9-2.

8.2.2.1 Method to Determine the Setpoints of the Steam Jet Air Ejector Offgas Activity Monitors (RM-17-150A and RM-17-150B)

The SJAE noble gas activity monitor response in mR/hr at the limiting release rate is the setpoint, denoted, and is determined as follows:

(8-21)

RSJAE spt = 1.6E +05 S 1 g F (mR/hr)(R l (mR -cc sec tsec) hr - gCiJ cc RSJAE spt = Response of the monitor at the limiting release rate (mR/hr) 1.6E+05 = Limiting release rate for the SJAE specified in Technical Specification 3.8.K.1 (Ci/sec)

S9 = Detector counting efficiency from the most recent calibration

((mR/hr)/(tCi/cc))

F = SJAE gaseous discharge flow (cc/sec) 8.2.2.2 Basis for the SJAB Noble Gas Activity Monitor Setpoint The SJAE noble gas activity monitor setpoint must ensure that Technical Specification 3.8.K.1 is not exceeded. The derivation of Equation 8-21 is straightforward. Simply taking Equation 8-16 and substituting the limiting release rate at the SJAE for Q yields Equation 8-21, the setpoint equation for the SJAE noble gas activity monitor.

Revision 29 Date 1/1 1/02 8-23

9.0 LIOUID 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 effluent systems, radiation monitors, and the appropriate Gaseous Radwaste Treatment System.

9.1 In-Plant Radioactive Liquid Effluent Pathways The Liquid Radwaste 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, floor 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 Radwaste 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.

Revision 27 Date 10/09/00 9-1

Low purity liquid wastes are collected in the 25,000-gallon floor drain collector tank.

They originate from the following sources:

1. 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 sample sinks. Radioactive decontamination solutions are classified as detergent waste and 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 processed 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 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.

Revision 27 Date 10/09/00 9-2

I A discharge from any of the tanks is accomplished by first starting the sample pumps, opening the necessary valves, and positioning the flow 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 does 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 alarm 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 flow rate or stop the discharge.

Revision 27 Date 10/09/00 I 9-3

9.2 In-Plant Radioactive Gaseous Effluent Pathways I The gaseous radwaste system includes subsystems that dispose of gases from the main condenser air ejectors, the startup vacuum pump, the gland seal condenser, the standby gas treatment system and 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.

The air ejector Advanced Off-Gas Subsystem (AOG) reduces the ejector radioactive gaseous release rates to the atmosphere. The AOG System consists of a hydrogen dilution and recombiner subsystem, a dual moisture removal/dryer subsystem, a single charcoal absorber subsystem, and dual vacuum pumps. Equipment is located in shielded compartments to minimize the exposure of maintenance personnel.

Radioactive releases from the air ejector off-gas system consist of fission product noble gases, activation product gases, halogens, and particulate daughter products from the noble gases.

The particulates and halogens are effectively removed by the charcoal beds and high efficiency particulate filters in the AOG System. The activation product gases that are generated in significant quantities have very short half-lives and will decay to low levels in the holdup pipe, as well as in the absorber beds. The noble gases, therefore, are expected to provide the only significant contribution to off-site dose. The charcoal off-gas system is designed to provide holdup of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for krypton and 16.6 days for xenon at a condenser air inleakage rate of 30 scfm.

Steam dilution, process control, and instrumentation systems are designed to prevent an explosive mixture of hydrogen from propagating beyond the air ejector stages. An explosive mixture of hydrogen should never exist in the recombiner subsystem, "30-minute" delay pipe, condenser/dryer, or charcoal absorber beds. To prevent a hydrogen explosion in the recombiner/preheater and upstream lines during shutdown, the residual off-gas steam mixture containing hydrogen is purged with steam or air. Starting procedures insure sufficient steam is introduced upstream of the preheater to dilute any hydrogen entering the AOG System as the air Revision 27 Date 10/09/00 94

ejector line is prepared for operation. To prevent operating unsafely, instrumentation is used to detect an explosive mixture.

Hydrogen control is accomplished by providing redundant hydrogen analyzers on the outlet from the Recombiner System. These analyzers initiate recombiner system shutdown and switchover if the hydrogen concentration at the system outlet exceeds 2% by volume. During an automatic shutdown, two main air process valves close to isolate the recombiner system.

Additionally, the recombiner bed temperatures and recombiner outlet temperature provide information about recombiner performance to insure that inflammable hydrogen mixtures do not go beyond the recombiner.

Should a number of unlikely events occur, it would be hypothetically possible for a hydrogen explosion to occur in the off-gas system. Such an explosion within the recombiner system could propagate into the large "30-minute" delay pipe, through the condenser/dryer subsystem, and into the charcoal absorber tanks. However, the recombiner/adsorber subsystems, piping, and vessels are designed to withstand hydrogen detonation pressures of 500 psi at a minimum so that no loss of integrity would result. Furthermore, the seven tanks of charcoal would significantly attenuate a detonation shock wave and prevent damage to the downstream equipment.

During normal operation, the dryer/adsorber subsystem may be bypassed if it becomes unavailable provided the releases are within effluent Control limits. With the dryer/adsorber subsystem bypassed, the air ejector off-gas exhausts through the recombiner/condenser subsystems, and the 30-minute delay pipe.

The off-gas mixture combines with steam at the air ejector stage to prevent an inflammable hydrogen mixture of 4% by volume from entering the downstream hydrogen recombiners. Approximately 6,400 lb/hr of steam introduced at the second stage air ejector reduces the concentration of hydrogen to less than 3% by volume.

The recombiner subsystem consists of a single path leading from the hydrogen dilution steam jet ejectors to two parallel flow paths for hydrogen recombination. Each recombination subsystem is capable of operating independently of the other and each is capable of handling the condenser off-gas at a startup design flow of 1,600 lb/hr air and the normal off-gas design flow rate of 370 lb/hr. The major components of each recombiner flow path are a preheater, a hydrogen-oxygen recombiner, and a desuperheating condenser.

Revision 27 Date 10/09/00 9-5

The preheater assures that the vapor entering the hydrogen-oxygen recombiner is heated to approximately 300 0F. At this temperature, the water vapor in the stream becomes superheated steam, thereby, protecting the recombiner catalyst.

During passages through the recombiner, the recombination of H2 and °2 in an exothermic reaction increases the stream temperature to approximately 520 0F. This recombination results in a maximum effluent H2 concentration of 0.1% by volume.

The desuperheating condenser is designed to remove the heat of recombination and condense the steam from the remaining off-gas. The condensers discharge the off-gas through moisture separators into the initial portion of an underground 24-inch diameter delay pipe which allows for 40% of the total system holdup volume. The pipe slopes away from the off-gas particulate (HEPA) filters in both directions for drainage purposes. Loop seals prevent gas escaping through drainage connections. Shorter lived radionuclides undergo a substantial decrease in activity in this section of the system. The preheaters/recombiners operate at pressures slightly above atmospheric; the condenser and the subsystems that follow operate at subatmospheric pressures.

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.

In the moisture removal/dryer subsystem, the moisture of the gas is reduced to increase the effectiveness of the charcoal absorber beds downstream. The subsystem consists of two parallel cooling condensers and gas dryer units. Each condenser is cooled by a mechanical glycol/water refrigeration system that cools the off-gas to -400 F as it removes bulk moisture.

The dryer is designed to remove the remaining moisture by a molecular sieve desiccant to a dew point of less than 400 F (1% RH). One of the dryers absorbs moisture from the off-gas; the other desorbs moisture by circulating heated air through the bed in closed cycle.

The mixed refrigerant/dryer concept improves the reliability of the system. If the refrigerant system fails, the two dryer beds operate in parallel to remove the moisture and maintain the off-gas near the design dew point (-400F). If the dryer fails, the 40 0 F dew point air leaving the mechanical system can enter the guard bed for over 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months without affecting the performance of the charcoal beds downstream.

Revision 27 Date 10/09/00 9-6

The charcoal absorber subsystem consists of seven tanks of charcoal preceded by a smaller charcoal guard bed upstream. The guard bed protects the seven main tanks from excessive radioactivity levels or moisture in the event of a malfunction upstream in the moisture removal subsystem. The guard bed also removes compounds which might hinder noble gas delay. The seven tanks hold a minimum of approximately 90,000 pounds of charcoal.

The first two main tanks can be bypassed and used for storing a "batch of high activity" gas for static decay. The remaining five are all in series with no bypassing features so that the off-gas to the stack must be delayed.

Redundant particulate (HEPA) after-filters are used to remove charcoal fines prior to the vacuum pumps.

A water-sealed vacuum pump boosts the gas stream pressure to slightly over-atmospheric pressure before it is vented through the stack. To assure maintaining constant operating pressures in the system, a modulating bypass valve will recirculate process gas around the pump as required. During periods of high flow rates, both pumps can be operated in parallel.

Discharge of the vacuum pump then passes through the remaining 60% of the delay pipe prior to being vented through the station stack.

The gland seal off-gas subsystem collects gases from the gland seal condenser and the mechanical vacuum pump and passes them through a charcoal filter (if required) and then through holdup piping prior to release to the stack. The gases from the gland seal condenser system are discharged to the atmosphere via the ventilation stack after passing through the filter for iodine removal (if required) and then through the same 1-3/4 minute holdup piping that is used for the startup vacuum pump system. One automatic valve on the discharge side of each steam packing'exhauster closes upon the receipt of high level radiation signal from the main steam line radiation monitoring subsystem to prevent the release of excessive radioactive material to the atmosphere. The exhausters are shut down at the same time the valves close. In addition, the mechanical vacuum pump is automatically isolated and stopped by a main steam line high radiation signal. The filter assembly is located in the air ejector room.

The release of significant quantities of gaseous and particulate radioactive material is prevented by the combination of the design of the air ejector AOG system and automatic isolation of the system from the stack. Gas flow from the main condenser stops when the air Revision 27 Date 10/09/00 9-7

ejectors are automatically isolated from the main condenser by either a high radiation signal in the main steam line or by high temperature and/or pressure signals from the AOG System. The gland seal off-gas system is automatically isolated and stopped by a main steam line high radiation signal. In addition, monitoring the stack release provides a backup warning of abnormal conditions.

Revision 27 Date 10/09/00 I 9-8

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10.0 UNIOUE REPORTING REOUIREMENTS I 10.1 Annual Radioactive Effluent Release Report - l In accordance with 10CFR 50.36a, the Radioactive Effluent Release Report covering the operation of the unit shall be submitted by May 15 of each year.

The Radioactive Effluent Release Report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit as outlined in Regulatory Guide 1.21, Revision 1,June 1974, "Measuring, Evaluating and Reporting Radioactivity in Solid Wastes and Releases of Radioactive Materials in Liquid and Gaseous Effluents from Light-Water-Cooled Nuclear Power Plants", with data summarized on a quarterly basis following the format of Appendix B thereof. For solid wastes the format for Table 3 in Appendix B of Regulatory Guide 1.21 shall be supplemented with three additional categories:

class of solid wastes (as defined by 10CFR Part 61), type of container (e.g., LSA, Type A, Type B, Large Quantity), and 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 form of joint frequency distributions of wind speed, wind direction, and atmospheric stability. This same report shall include an assessment of the radiation doses due to the radioactive liquid and gaseous effluents released from the unit 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. All 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 determining the gaseous pathway doses. The assessment of radiation doses shall be performed in accordance with the Off-Site Dose Calculation Manual (ODCM).

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

Revision 27 Date 10/09/00 -

10-1

likely most exposed real member(s) of the public from reactor releases (including doses from primary effluent pathways and direct radiation) for the previous calendar year to show conformance with 40CFR190, 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 3.8.D.1, describe the events leading to this condition in the next Radioactive Effluent Release Report.

If inoperable radioactive liquid effluent monitoring instrumentation is not returned to operable 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 inoperable gaseous effluent monitoring instrumentation is not returned to operable 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 milk samples no longer available from one or more of the sample locations required by Control Table 3.5.1, identify the cause(s) of the sample(s) no longer being available, identify the new location(s) for obtaining available replacement samples, and include revised ODCM figure(s) and table(s) reflecting the new location(s) in the next Radioactive Effluent Release Report.

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 Report.

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.

10.2 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.

Revision 27 Date 1009/00 10-2l

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 of the land use census required by Control 3.5.2. A summary description of the 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 LL) 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.

Revision 27 Date 10/09/00 1 10-3l

10.3 Special Reports I 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.

10.3.1 Liquid Effluents (Controls 3.2.2 and 3.2.3)

With the calculated dose from the release of radioactive materials in liquid effluents exceeding any of the limits of Control 3.2.2, prepare and submit 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 radwaste was being discharged without treatment, identification of any inoperable equipment or subsystems, and the reasons for the inoperability; (2) action(s) taken to restore the inoperable equipment to operable status; and (3) summary description of action(s) taken to prevent a recurrence.

10.3.2 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 effluents exceeding any of the limits of Control 3.3.2, prepare and submit to the Commission within 30 days a special report which identifies 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 Iodine-131, Iodine-133, 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 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 3.3.3.

Revision 27 Date 10/09/00 14

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:

(1) explanation of why gaseous radwaste was being discharged without treatment (Control 3.3.4), or with resultant doses in excess of Control 3.3.5, identification of any inoperable equipment or subsystems, and the reasons for the inoperability; (2) action(s) taken to restore the inoperable equipment to operable status; and (3) summary description of action(s) taken to prevent a recurrence.

10.3.3 Total Dose (Control 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 3.4.1 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.

10.3.4 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 Revision 27 Date 10109/00 l 10-51

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.

10.3.5 Land Use Census (Control 3.5.2)

With a land use census not being conducted as required by Control 3.5.2, prepare and submit to the Commission within 30 days a special report which identifies the reasons why the survey was not conducted, and what steps are being taken to correct the situation.

10.4 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 commision in the Radioactive Effluent Release Report for the period in which the evaluation was reviewed by the PORC. The discussion of each change shall contain:

1. A summary of the evaluation that led to the determination that the change could be made in accordance with 10CFR 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;

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

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

Revision 27 Date 10/09/00 l 10-6

those previously estimated in the license application and amendments thereto;

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

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 PORC.

B. Shall become effective upon review and acceptance by PORC and approval by the Plant Manager.

Revision 27 Date 10/09/00 10-7

REFERENCES A. Regulatory Guide 1.109, "Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10CFR50, Appendix I,"

U.S. Nuclear Regulatory Commission, Revision 1, October 1977.

B. Hamawi, J. N., "AEOLUS A Computer Code for the Determination 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.,

P100R13A, 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.

Nuclear Regulatory Commission, Rev. 1, July 1977.

D. National Bureau of Standards, "Maximum Permissible Body Burdens and Maximum Permissible Concentrations of Radionuclides 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. dePlanque 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, dated December 26, 1995.

Revision 27 Date 10/09/00 R-1

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

UNITED STATES NUCLEAR REGULATORY COMMISSION , RCESING WASHINGTON, DC. 20555 August 30, 1989 WE'Mra AKEE Docket No. 50-271 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) Vermont 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 (FES), 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 or 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 3/2/90 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 that the proposed disposal procedures will have a negligible effect on the environment and the general population in comparison to normal background radiation.

LA I In conclusion, we find the licensee's rocedures with commitments as 855 documented in reference (a) to be acceptable, provided that reference (a) is Dermanently incorporated into the licensee's Offsite Dose Calculation Manual (ODCM) 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 3 -

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 & 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. Nuclear Regulatory Commission R.D. 5, Box 169 475 Allendale 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 3/2/90 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 Traficonte, 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 03301-6397 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 Commission 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

VERMONT YANKEE NUCLEAR POWER CORPORATION Ferry Road. Brattleboro. VT 05301-7002 ACPLV7 ENGINEERING OFFICE SSO WIN STREET June 28, 1989 BOLTON.MA01740 BVY 89-59 450017714711 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 States 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 SEPTIC 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(a), 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 3500 gallons of wastewater per day. The septic tank, shown in Figure (1), 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 radi-oactive 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 (i.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 (Ci/kq 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

The 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-li

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 1 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 expected 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 (Wsol) being disposed of is estimated, for purposes of this bounding dose assessment, to be approximately:

Wsoi = 14,950 [gal] x 3,785.4 [cc/gal] x 10-3 [kg/cc]

x 0.30 [solids fraction] x (1/25.4) [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 chemical 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 (i.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 estimation 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 disposal 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 .

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 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 0 1iim [uCi]

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 9 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 (MPC) (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 procedure 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 B-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 ToDociraDhy 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 (FSAR) 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 locations 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 4.3.1 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

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. This 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/ft2 ,

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 9 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 NUREGICR-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 Radi onucl ide 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 conservative 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 plots from off-site properties, the general movement of groundwater 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 area conditions 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 (NRR) staff proposed guidance (described in AIF/NESP-037, August 1986). The proposed 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 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 ground, (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 analyses include:

(a) Exposure to the ground contamination and to resuspended radioactivity is for a period of 104 hours0.0012 days 0.0289 hours 1.719577e-4 weeks 3.9572e-5 months 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 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months 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:

l Revision 23 Date 05/12/99 B-23

Upper-Bound Activity Upper-Bound Activity Isotope Concentration rCi/kq 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, all 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 within 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.

(g) The deposition on crops of resuspended radioactivity is insignificantly small.

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 (i.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 pr'escribed 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 spreading of 40 combined (main septic system and south disposal system) tankfuls (at 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/vear) (mrem/year)

Ground Irradiation 0.0576 1.16 Inhalation 0.00122 2.74 Stored Vegetables 0.0913 0.00601 Leafy Vegetable 0.00467 0.00040 Milk Ingestion 0.0421 0.00229 Meat Ingestion 0.00249 0.00012 TOTAL 0.1994 3.909 In addition, an isotopic breakdown of the critical organ dose results listed above is shown in the following table:

Isotopic Breakdown of Maximum Radiation Exposures Radioactivity Exposure Descrition Isotope [uCi/2 Acres] rmrem/yr]

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-Ci/acre]

Mn-54 Adult/GE-LLI 3.74E-4 Co-60 Teenager/Lung 7.14E-4 Zn-65 Child/Liver 1.64E-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 DAte 3/2/90 B-28

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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-33

APPENDIX D ASSESSMENT OF SURVEILLANCE CRITERIA FOR GAS RELEASES FROM WASTE OIL INCINERATION Revision 16 Date 10/28/93 D-1

APPENDIX D ASSESSMENT OF SURVEILLANCE CRITERIA FOR GAS RELEASES FROM WASTE OIL INCINERATION INTRODUCTION:

The Nuclear Regulatory Commission amended its regulations (10CFR20) in a Federal Register Notice (Vol. 57, No. 235; page 57649 / Monday, December 7 1992) that permitted the on-site incineration of contaminated waste oil generated at licensed nuclear power plants without the need toamend 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 full compliance with the Commission's current regulations that restrict the release of radioactive materials to the environment. Any other applicable Federal, State, or local requirements that relate to the toxic or hazardous 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 Specifications (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 16 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 REOUIREMENTS The oil burner stack is not equipped with continuous air monitoring or sampling capability for the direct determination of radiological 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 consequences of burning contaminated 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.

Normal 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 16 Date 10/28/93 D-3

APPENDIX F APPROVAL PURSUANT TO 10CFR20.2002 FOR ONSITE DISPOSAL OF COOLING TOWER SILT Revision 21 Date 08/14/97 F-I

,,At CC UNITED STATES 0 NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20655-CO 4-

%)****V #0 June 18, 1997 NV 97-85

,.i; JlN 2 9 .

Mr. Donald A. Reid .,.r% S ......

D .-;

Vice President, Operations Vermont Yankee Nuclear Power Corporation 1>A1; I)

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 provisions 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 21 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 21 Date 08/14/97 F-3

Vermont Yankee Nuclear Power Vermont Yankee Nuclear Power Station Corporation cc:

Mr. Peter LaPorte, Director Regional Administrator, Region I ATTN: 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 Corporation 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 Nuclear Power Station Mr. 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 21 Date 08/14/97 F-4

rb=s -,'- UNlTED STATES 0 r NUCLEAR REGULATORY COMMISSION WASHINOTON, D.C. 2OSSS-O1 SAFETY EVALUATION BY THE 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 21 Date 08/14/97 F-5

b. Doses to the whole body and any organ of an inadvertent intruder from the probable pathways of exposure are less than 5 mrem/year.

c. The disposal must be at the same site.

Following the NRC staff's approval on August 30. 1989, VYNPC implemented the 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 EVALUATION In its letter of August 30, 1995, VYNPC stated that the proposed silt 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 its 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 letter of August 2, 1996. if the onsite 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 will report in the Annual Radiological Effluent Release Report a list of the radionuclides present and the total radioactivity associated with the onsite disposal activities at Vermont Yankee.

Revision 21 Date 08/14/97 F-6

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, 1997

Attachment:

Vermont Yankee Site Area Map Revision 21 Date 08/14/97 F-7

rD Po 0

LA CD I-czi

VERMONT YANKEE RESPONSIBILITY NucLEAR POWER CORPORIATION 14 CLsc-Cfti C7~~~II .) Ferry Road, Brattleboro, VT 05301-7=0

'V ~~~~~~~~~~~~~~~ENGINEERING SW3 OFFCE MAW STREET I*~~~~~~~~.dXJj MA Dim0 B~~~~~~~~OLt4.

August 30. 1995 BVY 95-97 United States Nuclear Regulatory Commission Washington. DC 20555 ATTN: Document Control Desk

References:

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

(2) Letter from R. W. Capstick. Vermont Yankee, to USNRC. Request to Routinely Dispose of Slightly Contaminated waste n Accordance with IOCFR20.302(a)". BVY 89-59. June 18. 1989.

(3) Letter from M. B. Fairtile, USNRC. to L. A. Tremblay. Vermont Yankee,

'Approval Under 10 CR 20.302(a) 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 20.302Ca) for Disposal of Contaminated Septic Waste In accordance with the criteria of the Code of Federal Regulations. Title 10.

Section 20.2002 (previously cited OCFR20.302Ca)). 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.

stream to include slightly contaminated Cooling Tower silt material.

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 n Reference 3. The disposal method utilizes on site land spreading n the same designated areas used for septic waste. Disposal of this waste in the manner proposed.

rather than holding t for future disposal at a 1CFR 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 21 Date 08/14/97 F-9

VERMONT YANKEE NUCLEAR POWER CORPORATION septic 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.

LA'I We trust that the information contained in the submittal is sufficient. however, should 1044 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 A)

USNRC Resident Inspector - VYNPS (Letter and Enclosure A)

USNRC Project Manager - VYNPS (Letter and Enclosure A)

Revision 21 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 FOR SEPTIC WASTE DISPOSAL Revision 21 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 4.1 Site Characteristics . . . . . . . . . . . . . .. . . . 13 4.2 Radiological Impact . . . . . . . . . . . . . . .. . . . 13 5.0 RADIATION PROTECTION . . . . . . . . . . . . . . . . . . . . . . .-

23

6.0 CONCLUSION

S . . . . . . . . . . . . . . . . . . . . . . . . . . . 23

7.0 REFERENCES

. . . . . . . . . . . . . . . . . . . . . . . . . . . . 24 2

Revision 21 Date 08/14/97 F-12

VERMONT YANKEE NUCLEAR POWER PLANT Assessment of Routine Disposal of Cooling Tower 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-site 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 be permanently incorporated into the plant's Offsite Dose Calculation Manual (ODCM). Revision 9 to the ODCM (Appendix B) 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-product materials in the silt, proper disposal requirements must be applied 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 21 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 approximately 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 21 Date 08/14/97 F-14

Table 1 Cooling Tower Silt Radioactivity (1993 samples*)

Sample # Co-60 Cs-137 (pCi/kg dry) (pCi/kg dry)

G12759 53 144 G12758 72 172 G12757 <14 201 G12756 <17 245 G12755 73 206 G12754 <16 165 G12753 <27 240 G12752 79 181 G12751 <29 180 G12750 35 107 G12749 59 171 G12748 <19 205 G12747 <38 209 G12746 < 7 218 G12745 50 241 G12744 40 220 G12743 68 264 G12742 45 195 G12724 71 115 G12723 104 264 G13940 126 218

............... ............ .....*.............. ................................................................ .................................... b................................w...

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 21 Date 08/14/97 F-15

For the purpose of estimating the total activity in the silt pile, the less than values in Table 1 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 Environmental Monitoring Program 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 (i.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 21 Date 08/14/97 F-16

Table 2 Cesium-137 in Connecticut River Sediments*

Cs-137 Date (pCi/kg dry)

..................... .................................................... I....... ...... ...................................................................... -.

05/24/94 61 10/13/93 85 06/02/93 60 10/15/92 137 05/20/92 i 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 10/12/88 115 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 (REMP) for river sediment sample location SE-ll.

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) (pCi1kg) (uci ) (uci)

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 that periodic maintenance work in cleaning out the cooling tower basins will generate approximately 4,000 cubic feet of new silt material over each successive 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 Estimated Total Radioactive Material for Each 18 Month Maintenance Cycle Volume of Mass Average Total Silt Concentration Activity (ft3) (kg) (pCi/kg) (uci )

I...... I.................................. ............................................... ................................................

Co-60 4,000 2.38E+5 50 12 Cs-137 4,000 2.38E+5 198 57 8

Revision 21 Date 08/14/97 F-18

3.0 DISPOSAL METHOD:

The method of silt disposal shall utilize 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 implemented as Appendix B 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 B of the ODCM.

3.1 Silt Disposal Procedure Requirements:

Gamma isotopic analysis of silt samples shall be made prior to each 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 Table 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 operations to ensure that the appropriate dose limits are not 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 21 Date 08/14/97 F-19

The limits on concentrations of radionuclides 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 limi.t 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 total pathway dose assessment determined for the quantity material under consideration. For the case where 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 assumptions of an 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 21 Date 08114/97 F-20

Table 5 Dry Soil Maximum Concentration Values Soil Concentration pCi/kg Radionuclide (equal to 25 mrem/yr)

Cr-51 1.51 E+05 Mn-54 5.50E+03 Fe-59 3.83E+03 Co-58 4.70E+03 Co-60 1 .82E+03 Zn-65 7.85E+03 Zr-95 6.18E+03 Ag-liOm 1.66E+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 21 Date 08/14/97 F-21

3.2 Administrative Procedure Requirements:

Dry silt material shall be dispersed using typical agricultural dry bulk surface spreading practices only in approved 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 each disposal operation as well 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 total radioactivity and dose calculations shall also include all past septic waste and 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 public that is contained 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 21 Date 08/14/97 F-22

4.0 EVALUATION OF ENVIRONMENTAL IMPACTS:

4.1 Site Characteristics 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 (non-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 ImDact:

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 21 Date 08/14/97 F-23

past NRC proposed guidance (described in AIFINESP-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 hours0.0012 days 0.0289 hours 1.719577e-4 weeks 3.9572e-5 months 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 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months 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 21 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 the 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.

(d) 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 ODCM 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 21 Date 08/14/97 F-25

- (i) 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 soil; but nonetheless, undergoes resuspension in air at the same rate as the unplowed surface contamination. 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 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 Sreading Nuclide Contri buti on Accumulation from Total Remaining from initial 13 cycles at in year 2013 14,000 ft3 (uCi) 4,000 ft3/ea. (uCi) (uci)

Cobalt-60 3.2 61.9 65.1 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 21 Date 08/14/97 F-26

Table 7 Dose Impact Due to Continued Spreading to End of License Disposal 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 Uncontrolled by VYNPS 1.46 mrem/yr adult/whole body (inadvertent intruder after 2.41 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) (Child/Bone)

Pathway (mrem/year) (mrem/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 21 Date 08/14/97 F-27

In addition, the isotopic breakdown of the critical organ doses listed above (Table 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 65.1 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 21 Date 08/14/97 F-28

Table 10 Dose Impact Due to Single (14,000 ft3) Silt Spreading Disposal Site Access Radiation Exposure Individual/Organ Controlled by VYNPS (max. 0.064 mrem/yr adult/whole body exposed individual in 1995) 0.219 mrem/yr max.' child/bone Uncontrolled by VYNPS 0.224 mrem/yr adult/whole body (inadvertent intruder after 0.393 mrem/yr max. child/bone license termination)

Table 10 shows that the application of the silt material initially 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 remain 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 21 Date 08/14/97 F-29

given on Table 11 for periods during Vermont Yankee control of the disposal site and on Table 12 for post control periods associated with the inadvertent 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 five 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 Attachment II to Reference 1.

20 Revision 21 Date 08/14/97 F-30

Table 11 All-Pathway Critical Organ / Whole Body Dose Conversion Factors During Vermont Yankee Control of Disposal Sites Critical Organ Whole Body Dose Dose Factor Factor Nuclide Individual/Organ (mrem/yr per uCi/acre)

Cr-51 Teen/Lung 1.14E-05 5.76E-06 Mn-54 Adul t/GI - LLI 3.75E-04 1.93E-04 Fe-55 Child/Bone 6.45E-06 1.06E-06 Fe-59 Teen/Lung 4.61E-04 2.13E-04 Co-58 Teen/Lung 3.27E-04 2.O1E-04 Co-60 Teen/Lung 7.17E-04 5.31E-04 Zn-65 Child/Liver 1.64E-02 1.03E-02 Zr-95 Teen/Lung 4.47E-04 1.34E-04 Ag11Dm Teen/GI-LLI 1.32E-02 5.24E-04 Sb-124 Teen/Lung 8.34E-04 3.54E-04 Cs-134 Child/Liver 3.18E-03 1.28E-03 Cs-137 Child/Bone 2.66E-03 7.02E-04 Ce-141 Teen/Lung 1.54E-04 1.50E-05 Ce-144 Teen/Lung 6.OOE-04 2.44E-05 21 Revision 21 Date 08/14/97 F-31

Table 12 All-Pathway Critical Organ / Whole Body Dose Conversion Factors Post Vermont Yankee Control of Disposal Sites (Inadvertent Intruder)

Critical Organ Whole Body Dose Dose Factor Factor Nuclide Individual/Organ (mrem/yr per uCi/acre)

Cr-51 Teen/Lung 5.89E-04 1.19E-04 Mn-54 Teen/Lung 1.02E-02 3.12E-03 Fe-55 Teen/Lung 3.50E-04 2.27E-05 Fe-59 Teen/Lung 2.55E-02 4.43E-03 Co-58 Teen/Lung 1.59E-02 3.72E-03 Co-60 Teen/Lung 3.19E-02 9.09E-03 Zn-65 Child/Liver 1.89E-02 1.25E-02 Zr-95 Teen/Lung 2.93E-02 2.99E-03 Agl1Om Teen/Lung 3.59E-02 9.53E-03 Sb-124 Teen/Lung 4.73E-02 7.04E-03 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.OE-02 1.52E-03 22 Revision 21 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. The radiological 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 21 Date 08/14/97 F-33

7.0 REFERENCES

(1) Vermont Yankee ODCM, Appendix B; "Approval of Criteria for Disposal of Slightly Contaminated 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 I". dated October 1977.

(3) NUREG/CR-5512, Vol. 1, "Residual Radioactive Contamination From Decommissioning", Final Report, dated October 1992.

file: vsilt.mss 24 Revision 21 Date 08/14/97 F-34

APPENDIX G MAXIMUM PERMISSIBLE CONCENTRATIONS (MPCs)

IN AIR AND WATER ABOVE NATURAL BACKGROUND TAKEN FROM 10CFR20.1 TO 20.602. APPENDIX B Revision 21 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 21 Date 08/14/97 G-2

APPENDIX G MAXIMUM PERMISSIBLE CONCENTRATIONS (MPCs)

(FROM 10CFR20.1 TO 20.602. APPENDIX B)

APPENDIX B TO SS20.1 - 20.602 CONCENTRATIONS IN AIR AND WATER ABOVE NATURAL BACKGROUND

[See footnotes at end of Appendix B]

Table I Table II Col. 1 Col. 2 Col. 1 Col. 2 Air Water Air Water Element (atomic number) Isotopel (gdC/ml) (gCd/ml) (gCi/ml) (gCi/ml)

Actinium (89) . . . . . . Ac 227 S 2x 12 6x10' 5 8x10' 14 2x10' 6 I 3x10' 1 1 9x10' 3 9x10'1 3 3x10' 4 Ac 228 S 8x10' 8 3x10' 3 3x10 9 9xlo' I 2x10' 8 3x10' 3 6x40 1 0 9x10' 5 Americium (95) . . . . . Am 241 S 6x10' 12 1x1O4 2x10' 13 4xlO- 6 I 1x10' 10 8x10' 4 4x10 1 2 3x10-5 Am 242m S 6x10 12 1x10' 4 2x10'1 3 4x10- 6 I 3x10' 10 3x10' 3 9x10'11 9x10' 5 Am 242 S 4x10' 8 4x10' 3 1x10' 9 1x10' 4 I sx10 8 4x10' 3 2x10' 9 1x10' 4 Am 243 S 6x10' 12 1x10' 4 2x10' 13 4xlO- 6 I 1x10' 10 8x10'4 4x10' 12 3x1O5 Am 244 S 4x10' 6 1x10' 1 1x10 7 5xiO I 2x10' 5 1x10 1 8x10' 7 5x10' 3 Antimony Sb 122 S 2x10' 7 8x10' 4 6x10' 9 3x10' 5 I 1x10' 7 8x10' 4 Sx10' 9 3x10 5 Sb 124 S 2x10' 7 7x10' 4 5x10 9 2x10' 5 I 2x10' 8 7x10' 4 7x10 10 2x10'5 Sb 125 S 5x10' 7 3x10' 3 2xl0' 8 1x10' 4 I 3x10' 8 3x10' 3 9x10' 1 0 1x10-4 Argon (18) . . . . . . . A37 Sub 2 6x10 3 ........ 1x10 4 ........

A41 Sub 2x106 ........ 4x10'8 ........

Arsenic (33) . . . . . . As 73 S 2x10 6 1x10-2 7x10' 8 5x10' 4 I 4x10'7 1x10-2 1x10 8 5x1O- 4 As 74 S 3x10 7 2x10' 3 1x10 8 5x10 5 I 1x10' 7 2x10' 3 4x10'9 5x1O5 As 76 S 1x10' 7 6x10' 4 4xlO' 2x10' 5 I 1x10' 7 6x10' 4 3x10' 9 2x1o As 77 S 5x10-7 2xlO 2x10 8 8x10' 5 I 4x10' 7 2x10 3 1x10' 8 8x10'5 Revision 21 Date 08/14/97 G-3

APPENDIX B TO S20.1 - 20.602 CONCENTRATIONS IN AIR AND WATER ABOVE NATURAL BACKGROUND

[See footnotes at end of Appendix B)

Table I Table II Col. 1 Col. 2 Col. 1 Col. 2 Air Water Air Water Element (atomic number) Isotopel (pCi/ml) (VCi/ml) (jiCi/ml) (liCi/ml)

Astatine (85) . . . . . . At 211 S 7x10 98 5x10-53 2x10 -190 2xlo 65 I 3x10- 2x10' 1x10- 7x10 Barium (56) . . . . . . . Ba 131 S 1x10- 6 5x10- 3 4x10- 8 2x10-4 I 4x10-7 5x10-3 ix10-8 2x10-4 Ba 140 S 1x10- 7 8x10- 4 4x10-9 3x10-5 I 4x10 8 7x10-4 1x10' 9 2x10- 5 Berkelium (97) . . . . . Bk 249 S 9x10-1 2x10-2 3x10- 11 6x10-4 I lxl0- 7 2x10-2 4x10-9 6x10-4 Bk 250 S lx10-7 6x10- 3 5x10-9 2x10-4 I 1x10-6 6x10- 3 4x10- 8 2x10-4 Beryllium (4) . . . . . . Be 7 S 6x10-6 5x10-2 2x10 7 2x10-3 I 1x10-6 5x10- 2 4x10-8 2x10-3 Bismuth (83) . . . . . . Bi 206 S 2x10 7 1x1O3 6x10-9 4x10' 5 I 1x10-7 1x10- 3 Sx10-9 4x10-5 Bi 207 S 2x10- 7 2x10-3 6x10-9 6x10-5 I 1x10' 8 2x10-3 5xlO-10 6x10-5 8i 210 S 6x10-9 1x10-3 2x10-1 4x10-5 I 6x10-9 1x10- 3 2x0-10 4xlO5 Bi 212 S lx10' 7 1x10- 2 3x10-9 4x10-4 I 2x10- 7 1x10-2 7x10-9 4x10-4 Bromine (35) . . . . . . Br 82 S lx10 6 Bx10 3 4xlO 8 3x10-4 I 2x10- 7 1x10- 3 6x10-9 4x10-5 Cadmium (48) . . . . . . Cd 109 S 5x10-8 5x10- 3 2x10-9 2x10-4 I 7x10 8 5x10' 3 3x1O' 9 2x10' 4 8

Cd 115m S 4x10 7x10' 4 1x10' 9 3x10' 5 8

I 4x10 7x10' 4 1x10' 9 3x10' 5 Cd 115 S 2x10' 7 1x10- 3 8xlO-9 3x10-5 I 2x10-7 1x10- 3 6x10-9 4x10-5 Calcium (20) . . . . . . Ca 45 S 3x1O-8 3x10- 4 lxlO-9 9xlo-6 I lx10' 7 5x10' 3 4x1O' 9 2x10' 4 Ca 47 S 2x10' 7 1x10- 3 6x10' 9 Sx10-5 I 2x10-7 1x10- 3 6x10' 9 3x10 5 Revision 21 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 Appendix B]

Table I Table II Col. 1 Col. 2 Col. 1 Col. 2 Air Water Air Water Element (atomic number) Isotopel (ClA/ml) (jidl/ml) (lCi/ml) (pCi/ml)

Callfornium (98) . . . . Cf 249 S 2x10 1'2 1x10'4 5x10'14 4x1O-6 I 1x10-1 0 7x10' 4 3x10l 2 2x10' 5 Cf 250 S 5x10- 12 4x10' 4 2x10' 1 3 1x10' 5 I 1x10' 10 7x10' 4 3x10l 2 3x10' 5 Cf 251 S 2x10'1 2 1x10 4 6x10' 1 4 4xIO 6 I 1x10' 10 8x10' 4 3x10'1 2 3x10' 5 Cf 252 S 6x10' 12 2x10' 4 2x10 1 3 7x1O-6 I 3x10 1 1 2x10' 4 1x10 1 2 7x10-6 Cf 253 S 8x10' 1 o 4x10 3 3x10 1 1 1x10' 4 I 8x10' 10 4x10' 3 3x10 1 1 1x10' 4 12 Cf 254 S 5x10 4x1O 6 2x10 13 lx10 7 12 I 5x10 4x10 6 2x10 13 1x10 ?

Carbon (6) . . . . . . . C 14 S 4x10' 6 2x10: 2 1x10' 7 8x10' 4 (C02 ) Sub 5x10 5 ......... 1x10 6 ........

Cerium (58) . . . . . . . Ce 141 S 4x10-7 3x10 3 2x10' 8 9x1O 5 I 2x10' 7 3x10 3 5x10' 9 9x10' 5 Ce 143 S 3x10' 7 1x10' 3 9x10' 9 4x10' 5 I 2x10' 7 1x10' 3 7x10' 9 4x10 5 Ce 144 S 1x10' 8 3x10' 4 3x10-10 1x10 5 I 6x10' 9 3x10' 4 2x10' 1 1x10' 5 Cesium (55) . . . . . . . Cs 131 S 1x10' 5 7x10-2 4x1O-7 2x10-3 I 3x1O 6 3x10' 2 1x10' 7 9x10' 4 Cs 134m S 4x1O5 2x10' 1 1x10- 6 6x10 3 I 6x10-6 3x10' 2 2x10-7 1x10-3 Cs 134 S 4x10- 8 3x10-4 1x10- 9 9xlo-6 I 1x10-8 1x10-3 4x1O 1 0 4x10' 5 Cs 135 S 5x10- 7 3x10' 3 2x10' 8 1x10' 4 I 9x10' 8 7xlO 3 3x10' 9 2x10' 4 7

Cs 136 S 4x10 2x10 3 1x10' 8 9x10'5 7

I 2x10 2x10' 3 6x10' 9 6x10' 5 Cs 137 S 6x10' 8 4x10' 4 2x1O 9 2x10' 5 I 1x10' 8 1x10- 3 5x10' 10 4x10' 5 Revision 21 Date 08/14/97 G-5

APPENDIX B TO 520.1 - 20.602 CONCENTRATIONS IN AIR AND WATER ABOVE NATURAL BACKGROUND (See footnotes at end of Appendix B)

Table I Table II Col. 1 Col. 2 Col. 1 Col. 2 Air Water Air Water Element (atomic number) Isotopel {Ci/ml) (.Ci/ml) (iCiIml) (jCi/ml)

Chlorine (17) . . . . . . Cl 36 S 4x10' 7 2x10- 3 1x10 8 8x10-5 I 2x10' 8 2x10-3 8x10' 10 6x10-5 Cl 38 S 3x10 6 1x10 2 9x10 8 4x10 4 I 2x10' 6 1x10-2 7x10- 8 4x10-4 Chromium (24) . . . . . . Cr 51 S 1x10 5 5x10 22 4x10 87 2x10 3 I 2x10' 6 5xio0 8x10- 2x10 3 Cobalt (27) . . . . . . . Co 57 S 3x10 6 2x10' 2 1x10' 7 5x10' 4 I 2x10' 7 1x10 2 6x10' 9 4x10' 4 Co 58m S 2x10' 5 8x10' 2 6x1O7 3x10' 3 I 9x10' 6 6x10' 2 3x10' 7 2x10 3 Co 58 S 8x10' 7 4x10-3 3x10-8 1x10-4 I 5x10- 8 3x10- 3 2x10- 9 9x10-5 Co 60 S 3x10-7 1x10- 3 1x10- 8 5x10'5 I 9x10' 9 1x10- 3 3x10'10 3x10-5 Copper (29) . . . . . . . Cu 64 S 2x10-6 1x10-2 7x10-8 3x10-4 I ixio6 6x10-3 4x10' 6 2x10-4 Curium (96) . . . . . . . Cm 242 S 1x10 10 7x10' 4 4x10' 1 2 2x10' I 2x1O-1° 7x10-4 6x10 12 2x10-5 Cm 243 S 6x10-1 2 1x10-4 2x10-1 3 5x1o-6 I 1x10-10 7x10'4 3x1O- 12 2x10' 5 Cm 244 S 9x10 1 2 2x10-4 3x10' 13 7xlO- 6 I 1x1o-10 8x10-4 3x10-1 2 3x10-5 Cm 245 S Sx10-12 1x10-4 2x10- 13 4x1O-6 I 1x1o-10 8x10-4 4x10'1 2 3x10 5 Cm 246 S 5x10 1 2 1x10-4 2x10- 13 4x1O- 6 I 1x10'1 8x10- 4 4x10-1 2 3x10-5 Cm 247 S Sx10-1 2 1x10- 4 2x10- 13 4x1O- 6 I 1x1O-10 6x10-4 4x10 12 2x10' 5 Cm 248 S 6x10' 1 3 1x1O 2x10' 14 4x10' 7 I 1xio 11 4x10' 5 4x10 13 1x10' 6 Cm 249 S 1x10-5 6x10-2 4x10- 7 2x10-3 I 1x10' 5 6x10- 2 4x10- 7 2x10-3 Revision 21 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 Air Water Air Water Element (atomic number) Isotope 1 _ (pCi/ml) (pCi/ml) (pCi/ml) (pCi/ml)

Dysprorium (66) . . . . . Dy 165 S 3x10-6 1x10-2 gx1o08 4x10-4 I 2x10- 6 1x10-2 7x10- 8 4x10-4 Dy 166 S 2x10 7 1x10' 3 x10 9 4x10-5 3

I 2x10-7 lx10- 7x10- 9 4x1O05 Einsteinium (99) . . . . Es 253 S 8x 101° 7x10' 4 3x10-11 2x10-5 I 6x10-10 7x10-4 2xIO' 2x10 5' Es 254m S 5x10'9 5x10' 4 2x10 10 2x10-5 I 6x10'9 5x10' 4 2x10' 1 2x10'5 Es 254 S 2x10-1 1 4x10' 4 6x10' 1 3 1x1o-s I 1x10- 10 4x10-4 4x10-1 2 1x10-5 Es 255 S 5x10-10 8x10' 4 2x10 11 3x10-5 I 4x10'10 8x10' 4 1x1O' 11 3x10-5 Erbium (68) . . . . . . . Er 169 S 6x10-7 3x10'3 2x10'8 9x10'5 I 4x10 7 3x10-3 1x1O 8 9x10-5 Er 171 S 7x10-7 3x10-3 2x10-6 1x10' 4 I 6x10-7 3x10- 3 2x10-8 1x10' 4 Europlum (63) . ... .. Eu 152 S 4x10-7 2x10- 3 1x10- 8 6x10' 5 CT/2-9.2 hrs) I 3x10 7 2x10 3 1x10 8 6x10 5 Eu 152 S 1x10-8 2x1O- 3 4x10-10 8x10-5 (T/2-13 yrs) I 2x10 8 2x10-3 6x10 10 8xlO 5 Eu 154 S 4x1O- 9 6x1O-4 1x10-10 2x10-5 I 7x10-9 6x10- 4 2x10 10 2x10-5 Eu 155 S 9x10 8 6x10 3 3x10- 9 2x10 4 I 7x10- 8 6x10- 3 3x10-9 2x10-4 Fermium (100) ... ... Fm 254 S 6x10- 8 4x10- 3 2x10 9 1x10-4 I 7x10-8 4x10- 3 2x10- 9 1x10-4 Fm 255 S 2x10-8 1x10-3 6x10-10 3x10-5 I 1x10-8 1x10-3 4x10-1 0 3x10-5 Fm 256 S 3x10-9 3x1O- 5 lx 1 0-1 9x1O-7 I 2x10- 9 3x10-5 6x10 11 9x1O-7 Fluorine (9) ... .. . F 18 S 5x10-6 2x10-2 2x10- 7 8x10-4 I 3x10- 6 1x10' 2 9x10 8 5x10 4 Revision 21 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 B)

Table I Table II Col. 1 Col. 2 Col. 1 Col. 2 Air Water Air Water Element (atomic number) Isotope1 (,uC1/ml) (Ci /ml) (.tCi/ml) (I+/-Ci/ml)

Gadolinium (64) . . . . . Gd 153 S 2x10'7 6x10 3 8x10 9 2x10'4 I 9x10' 8 6x10' 3 3x10' 9 2x10- 4 Gd 159 S 5x10' 7 2x10' 3 2x10 8 8x1095 I 4x10' 7 2x10' 3 1x10 8 8x10 5 Gallium (31) . . . . . . Ga 72 S 2x10' 7 1x10 3 8x10 9 4x10' 5 I 2x10'7 1x10' 3 6x10' 9 4x10 5 Germanium (32) .Ge 71 S 1x10' 5 Sx10'2 4x10 7 2x10' 3 I 6x10' 6 5x10 2 2x10' 7 2x10' 3 Gold (79) . . . . . . . . Au 196 S 1x10' 6 5x10' 3 4x10' 8 2x10' 4 I 6x10' 7 4x10' 3 2x10' 8 1x10 4 Au 198 S 3x10' 7 2x10' 3 1x10' 8 5x1O I 2x10' 7 1x10' 3 8xlO'9 5x10' 5 Au 199 S 1x1o 5x10 3 4x10 8 2x10 4 I 8x10' 7 4x10' 3 3x10 8 2x10' 4 Hafnium (72) . . . . . . Hf 181 S 4x10' 8 2x10' 3 1xlO'9 7x10' 5 I 7x10 8 2x10 3 3xIO 7x10 5 Holmium (67) . . . . . . Ho 166 S 2x10' 7 9x10' 4 7x1O' 9 3x10 5 I 2x10' 7 9x10' 4 6x10 9 3x10' 5 Hydrogen (1) . . . . . . H3 S sx10 6 1x10 1 2x10' 7 3x10' 3 I 5x10' 6 1x10' 1 2x10' 7 3x10' 3 Sub 2x10 3 ........ 4x10 5 ........

Indium (49) . . . . . . . In 113m S 8x10' 6 4x10' 2 3x10' 7 1x10' 3 I 7x10 6 4x10' 2 2x10' 7 1x10' 3 In 114m S 1x10' 7 5x10' 4 4x10' 9 2x10o-I 2x10 8 5x10' 4 7x10 10 2x10o5 2

In 115m S 2x1O'6 1x10 8x10 8 4x10' 4 I 2x20' 6 1x10 2 6x10' 8 4x10' 4 In 115 S 2x20' 7 3x10' 3 gx10'9 9x10' 5 I 3x20' 8 3x10' 3 lx1o 9x10' 5 Iodine (53) . . . . . . . I 125 S 5x30' 9 4x10' 5 8x10' 1 1 2x10' 7 I 2x50' 7 6x10' 3 6x10' 9 2x10' 4 1 126 S 8x20' 9 5x10' 5 9x1O 1 1 3x10' 7 I 3x80' 7 3x10' 3 1x10' 8 9x10 5 1 129 S 2X30 9 1x10' 5 2x1 1 6x10 8 I 7x20' 8 6x10' 3 2x10' 9 2xlo14 Revision 21 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 B)

Table I Table II Col. 1 Col. 2 Col. 1 Col. 2 Air Water Air Water Element (atomic number) Isotope1 (Ci /ml) (jCi/ml) (pC /ml) (ACl/ml) 9 5 1 Iodine (53) . . . . . . . 1 131 S 9x70 6x10 lx10 l 3x10'7 7 3 (Continued) I 3x90 2x10' 1xiO18 6x10 5 1 132 S 2x10' 7 2x10' 3 3x10' 9 Bx10O6 I 9x10 7 5x10' 3 3x10' 8 2x10' 4 1 133 S 3x10 8 2x10' 4 4x10' 10 1x106 I 2x10- 7 *xl0 3 7x10' 9 4x10' 5 I 134 S 5x10' 7 4x10' 3 6x10' 9 2x10' 5 I 3x10-6 2x10' 2 1x10- 7 6x10' 4 I 135 S lx10' 7 7x10' 4 1xlO'9 4x1O' 6 7 3 8 5 I 4x10' 2x10' 1x10' 7x10' Iridium (77) . . . . . . Ir 190 S 1x10 6 6x10' 3 4x1O08 2x10 4 I 4x10'7 SX10-3 1x1O08 2x10 4 Ir 192 S 1x10- 7 1x10- 3 4x10-9 4x10-5 I 3xlO-8 1x10-3 9x1o- 0 4x10' 5 Ir 194 S 2x10' 7 1x10' 3 8x10' 9 3x10-5 I 2x10' 7 9x10'4 5xlO'9 3x10' 5 Iron (26) . . . . . . . . Fe 55 S 9x10-7 2x10 2 3xl0 8 8xlO14 I ixiO16 7x1lO 2 3x10' 8 2x10 3 Fe 59 S 1x1O 7 2x10' 3 5x10'9 6xlO5 I 5x10'8 2x10' 3 2x1O' 9 5x10' 5 Krypton (36) . . . . . . Kr 85m Sub 6x1O' ........ lx10'7 ........

Kr 85 Sub 1x1O'5 ........ 3xlO 7 ........

6 Kr 87 Sub lxlo0 , 2x10'8 ........

Kr 88 Sub 1x10'6 ........ 2xlO 8 Lanthanum (57) . . . . . La 140 S 2x10 7 7x10 4 5x10 9 2x1O05 I l x10 7 7x10 4 4x0 9 2x10 5 Lead (82) . . . . . . . . Pb 203 S 3x10' 6 1x10' 2 9xlo8 4X10-4 I 2x10 6 1x10 2 6x10 8 4x10' 4 Pb 210 S 1x10'10 4x10' 6 4xlO 12 lx10-7 I 2x10' 10 5x10' 3 8x10 1 2 2x1O 4 Pb 212 S 2x10' 8 6x10' 4 6x10' 10 2x10' 5 8 4 10 I 2x10' 5x10- 7x10' 2x10' 5 Revision 21 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 B)

Table I Table II Col. 1 Col. 2 Col. 1 Col. 2 Air Water Air Water Element (atomic number) IsotopeL (3PCI/ml) (C1I/ml) ("C1/ml) (jCi/ml)

Lutetium (71) . . . . . Lu 177 S 6x10'7 3x10'3 2x10'8 1X10'4 I 5x1O' 7 3x10' 3 2x10' 8 1xiO 4 Manganese (25) . . . . . Mn 52 S 2x1O 7 1x10 3 7x10' 9 3x10' 5 I 1x1o 7 9x10' 4 5x10' 9 3x10 5 Mn 54 S 4x10'7 4x10 3 1x1O-8 lx10' 4 I 4x10- 8 3x10-3 1x10-9 1x1O-4 Mn 56 S 8x10- 7 4x10-3 3x10-8 1x10-4 I 5x10- 7 3x10-3 2x10-8 1x10-4 Mercury (80) . . . . . . Hg 197m S 7x10 7 6x10 3 3x10- 8 2x10-4 I 8x10- 7 5x10 3 3x10- 8 2x10-4 Hg 197 S 1x1O-6 9x10-3 4x10-8 3x10-4 I 3x1O-6 1x10- 2 9x10- 8 5x10-4 Hg 203 S 7x10-8 5x1O-4 2xlO- 9 2xlO 5 I 1x10- 7 3x10-3 4x1O-9 1x10-4 Molybdenum (42) .Mo 99 S 7x10' 7 5x10-3 3x10- 8 2x10-4 I 2x10- 7 1x10 3 7x10-9 4x10-5 Neodymium (60) . . . . . Nd 144 S 8x10-11 2x10 3 3x10- 12 7x10-5 I 3x10-1 2x10' 3 1x10' 11 8x10-5 Nd 147 S 4x1O-7 2x10- 3 1x10- 8 .6x10- 5 I 2x10-7 2x10- 3 8x10-9 6x1O-5 Nd 149 S 2x10 6 8x10 3 6x10 8 3x10-4 3

I 1x1o- 8x10 5x10-8 3x10-4 Neptunium (93) .Np 237 S 4x10- 12 9x10-5 lx10 13 3x10 6 I 1x10-10 9x10-4 4x10-1 2 3x10-5 Np 239 S 8x10- 7 4x10- 3 3x10- 8 1x10-4 I 7x10-7 4x10- 3 2x10- 8 1x1O-4 Nickel (28) . . . . . . . Ni 59 S 5x10- 7 6x10-3 2x10-8 2x10' 4 I 8x10- 7 6x10- 2 3x10-8 2x10-3 Ni 63 S 6x10- 8 8x10- 4 2x10-9 3x10-5 I 3x10- 7 2x10-2 1x10-8 7x10-4 Ni 65 S 9x10- 7 4x10- 3 3x10- 8 1x10-4 I 5x10-7 3x10- 3 2x10-8 1x10-4 Revision 21 Date 08/14/97 G-10

APPENDIX B TO S20.1 - 20.602 CONCENTRATIONS IN AIR AND WATER ABOVE NATURAL BACKGROUND

[See footnotes at end of Appendix B)

Table I Table II Col. 1 Col. 2 Col. 1 Col. 2 Air Water Air Water Element (atomic number) Isotope1 (pCi/mil) (PClIml) (pCi/ml) (pCI/ml)

Nioblum (Columbium) (41) Nb 93m S 1x10'7 1x10 2 4x10'9 4x10'4 I 2x10' 7 1x10 2 5x1O 9 4x10' 4 Nb 95 S 5x10 7 3x10' 3 2x10 8 1x10 4 I 1x10' 7 3x10' 3 3x10- 9 1x10-4 Nb 97 S 6x10-6 3x10-2 2x10-7 9x10' 4 I 5x10-6 3x10-2 2x10 7 9x1O' 4 Osmium (76) . . . . . . . Os 185 S 5x10-7 2x10-3 2x10-8 7x10'5 I 5x10-8 2x10' 3 2x1O' 7x10' 5 Os 191m S 2x10' 5 7x10' 2 6x10' 7 3x10 3 I 9x10' 6 7x10' 2 3x10'7 2x10' 3 Os 191 S 1x1O'6 5x10' 3 4x10'8 2x10' 4 I 4x10' 7 5x10' 3 1x10 8 2x10- 4 Os 193 S 4x10' 7 2x10 3 1x10 8 6x10' 5 I 3x10' 7 2x10 3 9x10' 9 5x10' 5 Palladium (46) . . . . . Pd 103 S 1x10' 6 1x10, 5x108 3x1Qv 4 I 7x10' 7 8x10' 3 3x10' 8 3x10 4 Pd 109 S 6x10' 7 3x10' 3 2x10' 8 9x10' 5 I 4x10 7 2x10 3 1x10 8 7x10' 5 Phosphorus (15) .P 32 S 7x10' 5x10' 4 2x1O' 2x10' 5 I 8x10' 8 7x10' 4 3x10' 9 2xlO 5 Platinum (78) . . . . . . Pt 191 S 8x10 7 4x10 3 3x10' 8 1x10w4 I 6x10' 7 3x10- 3 2x1O0 8 lx10' 4 Pt 193m S 7x10' 6 3x10' 2 2x10' 7 1x10' 3 I 5x10- 6 3x10 2 2x10-7 1x10' 3 Pt 193 S 1x10' 6 3x10 2 4x10' 8 9x10' 4 I 3x1O-7 5x10' 2 1x10' 8 2x10' 3 Pt 197m S 6x10-6 3x10' 2 2x10- 7 1x10 3 6

I 5x10' 3x1O 2 2x10- 7 9x10' 4 Pt 197 S 8x10' 7 4x1O 3 3x10-8 1x10' 4 I 6x1O 7 3x10' 3 2x1O- 1x10' 4 Revision 21 Date 08/14/97 G-11

APPENDIX B TO 520.1 - 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 Air Water Element (atomic number) IsotopeL (gCi/ml) (jICi /ml 1* 1*

Plutonium (94) Pu 238 S 2x1O- 1 2 1x10 4 I 3x1O'1 Bx10 4 Pu 239 S 2x10- 12 1x10-4 I 4x10 11 8x10- 4 Pu 240 S 2x101 2 1x10 4 I 4x10 11 8x10' 4 Pu 241 S 9x10-l1 7x10-3 I 4x1O- 8 4x10-2 Pu 242 S 2x10 12 1x10-4 I 4x10' 11 9x10 4 Pu 243 2x10 6 1x10-2 I 2x10'6 1x10'2 Pu 244 S 2x10 1 2 1x10-4 I 3xlO1 3x10-4 10 5 Polonium (84) . . . . . . Po 210 S 5x10 2x10 I 2x10 1 ° 8x10' 4 Potassium (19) . . . . . K42 S 2x10 6 9x10 3 I lx10' 7 6x10- 4 Praseodymium (59) . . . . Pr 142 S 2x10'7 9x10 4 I 2x10'7 9x10-4 7 3 Pr 143 S 3x10 1x10 7 3 I 2x10 1X10-8 3 Promethium (61) . . . . . Pm 147 S 6x10 6x10 I lx10'7 6x10 3 7 3 Pm 149 S 3xjO 1x10 I 2xl10 7 1X10-3 Protoactinium (91) . . . Pa 230 S 2x10 9 7xlo 3 I 8x0 1 0 7x10 3 12 5 Pa 231 S 1x10 3x10 10 4 I 1x10 8x10 7

Pa 233 S 6x10 4x10 3 7

I 2x10 3x10- 3 Revision 21 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 II Col. 1 Col. 2 Col. 1 Col. 2 Air Water Air Water Element (atomic number) Isotopel (Ci/ml) (PCi/ml) (PCi/ml) (iCI/ml)

Radium (88) . . . . .. . Ra 223 S 2x10 9 2x10 5 6x10 11 7x10-7 I 2xlO-1 1x1O' 4 8x1O 12 4x105 Re 224 S 5x10' 9 7x1O 5 2x10 10 2x10' 6 I 7xl0-1 ° 2x10- 4 2xlO1 1 5xio 6 Ra 226 S 3x10 1 1 4x1O 7 3x1O' 2 3x10-8 I Sx10' 1 1 9x10-4 2x10 1 2 3x10'5 Ra 228 S 7x10'11 8x10'7 2x10'12 3x10@8 I 4x10' 11 7x10 4 1x10' 12 3x10' 5 Radon (86) . . . . . . . Rn 220 S 3x10 7 ........ 1x10 8 ........

Rn 2223 3x10' 8 ........ 3x10'9 ........

Rhenium (75) . . . . . . Re 183 S 3x10'6 2x10 2 9x10' 8 6x10' 4 I 2xlo07 8x10'3 Sx10'9 3x10 4 Re 186 S 6x10'7 3x10'3 2x10'8 9x10'5 I 2x10 7 1x10'3 8x10-9 .5x10'5 Re 187 S 9x10 6 7x1O' 2 3x10' 7 3x10 3 I 5x1O'7 4x10'2 2x10'8 2x10'3 Re 188 S 4x10'7 2x10'3 1x1O 8 6x1O 5 I 2x10'7 9x10'4 6x10'9 3x10'5 Rhodium (45) . . . . . . Rh 103m S 8x10'5 4x10'1 3x10'6 1x10 2 I 6x10'5 3x1O 1 2x10' 6 1x10' 2 Rh 105 S 8x10' 7 4x10' 3 3x10' 8 1x1O, 4 7 8 I 5x10 3x1O'3 2x10' 1X10-4 Rubidium (37) . . . . . . Rb 86 S 3x10'7 2x10'3 1x10'8 7xlO15 I 7x1O08 7x1O'4 2x10'9 2xlO' Rb 87 S 5x10'7 3x10 3 2x10 8 1x10'4 I 7x10'8 5x10'3 2x10'9 2x10-4 Ruthenium (44) . . . . . Ru 97 S 2x10 6 1x10'2 8xlO 8 4x10 4 I 2x10 6 1x10 2 6x10'8 3x10'4 Ru 103 S 5x10'7 2x10'3 2x10'8 8x10' 5 I 8x10 8 2x10'3 3x10'9 8x10e5 Ru 105 S 7x10' 7 3x1O 3 2x10'8 1x1O'4 I 5x10 7 3x10'3 2x10'8 1x10'4 Ru 106 S 8x10'8 4x10'4 3x10'9 1x10 5 I 6x10'9 3x10'4 2x10'10 1x10'5 Revision 21 Date 08/14/97 G-13

APPENDIX B TO §520.1 - 20.602 CONCENTRATIONS IN AIR AND WATER ABOVE NATURAL BACKGROUND

[See footnotes at end of Appendix ]

Table I Table II Col. 1 Col. 2 Col. 1 Col. 2 Air Water Air Water Element (atomic number) Isotopel (pClIml) (+/-Ci/ml) (ICI/ml) (iCi/ml)

Samarium (62) . . . . . Sm 147 S 7x1O 1l 2x10-3 2x1O-12 6x1O 5 I 3x10' 1 _ 2x1,0 3 9x10' 12 7x10' 5 Sm 151 S 6x1O' 8 1x10 2 2x1O 9 4x10 4 I 1x10' 7 1x10' 2 5 x10'9 4x10' 4 Sm 153 S 5x10' 7 2x10' 3 2x10 8 8x10O5

} 4x10' 7 2x10' 3 1x10' 8 8x10' 5 Scandium (21) . . . . . . Sc 46 S 2x10' 7 1x10' 3 8x10' 9 4x10' 5 I 2x10 8 1x10' 3 8x101 0 4x10 5 Sc 47 S 6x10' 7 3x10' 3 2x10' 8 9x10' 5 I 5x10' 7 3x10' 3 2x10' 8 9x10' 5 Sc 48 S 2x10' 7 8x10' 4 6x10' 9 3x1O' I 1x10' 7 8x10' 4 Sx10' 9 3x10' 5 Selenium (34) . . . . . . Se 75 S 1x10' 6 9x10' 3 4x10 8 3x10 4 I lx1O 7 Bx10' 3 4x10' 9 3x10e4 Silicon (14) . . . . . . Si 31 S 6x10' 6 3x10' 2 2x10' 7 9x10 4 I 1x10' 6 6x10' 3 3xlO-8 2x10' 4 Silver (47) . . . . . . . Ag 105 S 6x10' 7 3x10' 3 2x10' 8 1x10 4 I 8x10 8 3x10' 3 3x10' 9 1x1O 4 Ag 110m S 2x10' 7 9x10' 4 7x10' 9 3x10' 5 I 1x1o' 9x10' 4 3x10 10 3xlO 5 Ag 111 S 3x10' 7 1x10' 3 1x1O' 8 4x10' 5 I 2x10' 7 1x10 3 8xlO 9 4x10' 5 Sodium (11) . . . . . . . Na 22 S 2x10-7 1x1O 3 6x10 9 4x10 5 I 9x10-9 9x10-4 3x10-10 3x10-5 Na 24 1x10- 6 6x10- 3 4xlO 8 2x10 4 I 1x10- 7 Bx10-4 5x10-9 3x10-5 Strontium (38) . . . . . Sr 85m S 4x10- 5 2x10- 1 1x10-6 7x10 3 I 3x1O- 5 2x10-1 1x10-6 7x10-3 Sr 85 S 2x10- 7 3x1O-3 8x10-9 1x10-4 I 1x10- 7 5x10 3 4x1O-9 2x10'4 Sr 89 S 3x10' 8 3x10' 4 3x10' 10 3x10-6 I 4xlO-" 8x10' 4 1x10' 9 3x10 5 Sr 90 S 1x1O 9 1x10' 5 3x10 11 3x10' 7 I 5x10' 9 1x10' 3 2x10 10 4x10 5 Revision 21 Date 08/14/97 G-14

APPENDIX B TO SS20.1 - 20.602 CONCENTRATIONS IN AIR AND WATER ABOVE NATURAL BACKGROUND

[See footnotes at end of Appendix B]

Table I Table II Col. 1 Col. 2 Col. 1 Col. 2 Air Water Air Water Element (atomic number) Isotope1 (jCi/ml) (gCi/ml) (iCi/ml) (VCiIml) 7 Strontium (38) .Sr 91 S 4x10 2x10-3 2x10'8 7x10'5 7 3 9 (Continued) I 3x10' 1x10 9x10 5x10 5 Sr 92 S 4x10'7 2x10' 3 2x10' 8 7x10 5 I 3x10 7 2x10' 3 1x10 8 6X10 5 Sulfur (16) . . . . . . . S 35 S 3x10' 7 2x10' 3 9x10- 9 6x10-5 I 3x10'7 8x10 3 9x10' 9 3x10'4 Tantalum (73) . . . . . . Ta 182 S 4x10' 8 1x10' 3 1x10' 9 4x10' 5 I 2x10 8 1x10'3 7x10 10 4x10' 5 Technetium (43) . . . . . Tc 96m S 8x10' 5 4x10' 1 3x10 6 1xio 2 I 3x10' 5 3x10 1 1x10' 6 1x10 2 Tc 96 S 6x10' 7 3x10' 3 2x10 8 1x10' 4 I 2x10' 7 1x10' 3 8x10' 9 5x10' 5 Tc 97m S 2x10 6 1x1O 2 8x10' 8 4x10-4 I 2x10' 7 5x10' 3 5x10' 9 2x10' 4 Tc 97 S 1x1o 5 5x10 2 4x10' 7 2x10 3 I 3x10' 7 2x10' 2 1x10 8 8x10' 4 Tc 99m S 4x10'5 2x10' 1 1x10 6 6x1093 I 1x10' 5 6x10' 2 5x10' 7 3x10 3 Tc 99 S 2x10' 6 1x10 2 7x10' 8 3x10' 4 I 6x10' 8 5x10' 3 2x10' 9 2x10-4 Tellurium (52) . . . . . Te 125m S 4x10 7 5x10' 3 1x10'9 2x10O4 I 1x10-7 3x10-3 4x10' 9 1x10' 4 Te 127m S 1x10-7 2x10- 3 5x1o- 9 6x10' 5 8 3 I 4x10' 2x10' 1x10' 9 5x10o' Te-127 S 2x10' 67 8x10 33 6x10 88 3x10 4 4

I 9x10 5x10' 3x10' 2x10 Te 129m S 8x10' 8 1x10' 3 3x10 9 3x105 I 3x10' 8 6x10' 4 1x10' 9 2x10' 5 Te 129 S Sx10' 6 2x10' 2 2x10' 7 8x10-4 I 4x10' 6 2x10' 2 .1x10 7 8x10 4 Te 131m S 4x10-7 2x10 3 1x10-8 6x10-5 I 2x10- 7 1x10-3 6x10-9 4x10-5 Te 132 S 2x10 7 9x10 4 7x10 9 3x10O 5 I 1x10' 7 6x10'4 4x10' 9 2x10' 5 Revision 21 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 II Col. 1 Col. Col. 1 Col. 2 Air Water Air Water Element (atomic number) Isotope_ (tCi/ml) (iCiIml) (ACi/ml) (Ci/ml)

Terbium (65) . . . . . . Tb 160 S 1x10 7 1x10 3 3x10 9 4x10-5 I 3x10-8 1x10-3 1x10-9 4x10-5 Thalium (81) .. ... . Tl 200 S 3x10-6 ix10-2 gx10 8 4x10-4 I 1x10 6 7x10 3 4x10 8 2x10-4 Ti 201 S 2x10-6 9x10-3 7x10 5 3x10-4 I 9x10 7 5x10 3 3x10 8 2xl0 4 Ti 202 S 8x10- 7 4x10- 3 3x10-8 1x10-4 I 2x10- 7 2x10- 3 8x10-9 7x10-5 Tl 204 S 6x10- 7 3x10- 3 2x10- 8 1x1O0 4 I 3x10- 8 2x10- 3 9x10-10 6x10-5 Thorium (90) . . . . . . Th 227 S 3x10 1 0 5x10-4 1x10-1 1 2x10-5 I 2x10-1 5x10- 4 6x10 12 2x10-5 Th 228 S 9x10 1O 2x10 4 3x10- 13 7xlO-6 I 6x10-12 4x10- 4 2x10- 1 3 1xio5 Th 230 S 2x10- 12 5x10-5 8x10 14 2x10 6 I 1x10-1 1 9x10- 4 3x10 13 3x10-Th 231 S 1x10-6 7x10-3 5x10 8 2x10 4 I 1x10-6 7x10-3 4x10-8 2x10-4 Th 232 S 3x10-11 5xlO- 1x10- 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 6xo-18 sx10-4 2x10- 9 2x10-5 I 3x10-8 Sx10-4 1x10- 9 2x10-5 Thulium (69) . ... .. Tm 170 S 4x10 8 lX10 3 1x10 9 5x10 5 I 3x10- 8 1x10- 3 1x10- 9 5x10-5 Tm 171 S 1x10- 7 1x10-2 4x10-9 Sx10, 4 I 2x10-7 1x10-2 8x1O-9 5x10-4 Tin (50) . . . . . . . . Sn 113 S 4x10-7 2x10- 3 1x10-8 9x10-5 I 5x10- 8 2x10- 3 2x10-9 8x10-5 Sn 125 S 1x10- 7 5x10-4 4x10-9 2x10-5 I 8x10- 8 5x10-4 3x10-9 2x10-5 Revi si on 21 Date 08/14/97 G-16

APPENDIX B TO §520.1 - 20.602 CONCENTRATIONS IN AIR AND WATER ABOVE NATURAL BACKGROUND

[See footnotes at end of Appendix B]

Table I Table II Col. 1 Col. 2 Col. 1 Col. 2 Air Water Air Water Element (atomic number) Isotopel (PCi/ml) (pCi/ml) (iCi/ml) (pCi/ml)

Tungsten (Wolfram) (74) . W 181 S 2x10 6 1x10 2 8x10-8 4x10 4 I lx10-7 1x10-2 4x10-9 3x10-4 W 185 S 8x10- 7 4x10-3 3x10' 8 lx10 4 I lx10'7 3x10' 3 4x10'9 lx10' 4 W 187 S 4x10' 7 2x10' 3 2x1O'8 7x10' 5 I 3x10' 7 2x1O 3 ix10 8 6x10' 5 Uranium (92) . . . . . . U 230 S 3x10 10 1x10' 4 1x10' 1 1 5xlO- 6 I 1x10' 10 lx10' 4 4x10 1' 2 5xio-6 U 232 S 1x10' 10 8x10' 4 3x10' 12 3x10' 5 I 3x1OT11 8x10' 4 9x10' 13 3x10'5 U 233 S 5x10- 10 9x10-4 2x10-11 3x10- 5 I 1x1O-1o 9x10- 4 4x10- 12 3x10-5 U 234 S4 6x10 10 qx10- 4 2x10 1l 3x10-5 I 1x10-10 9x10- 4 4x10 12 3x10-5 U 235 S4 5x10-1 0 8x10- 4 2x10-11 3x10-5 I 1x10-10 8x10- 4 4x10-12 3x1O-5 U 236 S 6x10 10 1x10-3 2x10-1 1 3x10- 5 I 1x101 0 1x10-3 4x10- 12 3x10-5 U 238 S4 7x10-1 1 1x10-3 3x10- 12 4x10-5 I 1x1O-10 1x10- 3 5x10- 1 2 4x10-5 U 240 S 2x10-7 1x10- 3 8x10- 9 3x10' 5 I 2x10' 7 1x10' 3 6x10' 9 3x1O' U-natural S4 1x10t 10 1x10' 3 5x10' 1 2 3x10'5 I 1x10' 1 1x10' 3 Sx10-1 2 3x10-5 Vanadium (23) . . . . . . V 48 S 2x10-7 9x10- 4 6x10-9 3x10-5 I 6x10 8 8x10 4 2x10' 9 3x10 5 Xenon (54) . . . . . . . Xe 131m Sub 2x10' 5 ........ 4x10'7 ........

Xe 133 Sub 1x1O'S ........ 3x10'7 ........

Xe 133m Sub 1x10'5 ........ 3x10'7 ........

Xe 135 Sub 4xl0 6 .X10.7 ........

Ytterbium (70) .Yb 175 S 7x10' 7 3x10' 3 2x10' 8 1x10' 4 I 6x10' 7 3x10' 3 2x10' 8 1x10' 4 Yttrium (39) . . . . . . Y 90 S 1x10' 7 6x10' 4 4x1O' 9 2x10' 5 I lxlo 7 6x10'4 3x10' 9 2x10 5 Revision 21 Date 08/14/97 G-17

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) IsotopeL (Cir/ml) (jiCi/ml) (tLCi/ml) (txCi/ml)

Y 91m S 2x10'5 1x10-1 8x10'7 3x10'3 I 2x10-5 1x10-1 6x10-7 3x10-3 Y 91 S 4x10-8 8x10-4 1xlO- 9 3x10-5 I 3x10-8 8xlo-4 1x10-9 3x10-5 Y 92 S 4x10-7 2x10-3 1x10'8 6x10 5 3x10 7 8 I 2x10'3 ix10- 6x1O5 Y 93 S 2x10-7 8x10-4 6x10-9 3x10-5 I 1x10-7 8x10-4 5x10-9 3x10-5 Zinc (30) . . . . . . . . Zn 65 S 1x10-7 3x10-3 4x10-9 1x10-4 I 6x10'8 5x10'3 2x10 9 2x10'4 Zn 69m S 4x10-7 2x10 3 1x10-8 7x10-5 I 3x10-7 2x10-3 1x10-8 6x10-5 Zn 69 S 7x10'6 5x10-2 2x10-7 2x10'3 I 9x10-6 5x10-2 3x10-7 2x10-3 Zirconium (40) .Zr 93 S 1x10-7 2x10-2 4x10-9 8x10-4 7 2 8 I 3xlo 2x10- 1xi0 8x10-4 Zr 95 S 1x10-7 2x10-3 4xl- 9 6x10-5 I 3x1O08 2x10-3 1x10-9 6xl-5 Zr 97 S 1x10-7 5x10-4 4x10-9 2x10'5 I 9x1O-8 5x10-4 3x10-9 2x10-5 Any single radionuclide ........ Sub 1x10-6 ........ 3x10-8 .......

not listed above with decay mode other than alpha emission or spontaneous fission and with radioactive half-life less than 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

Any single radionuclide ........ 3x10-9 9x10'5 xO-10 3x10-6 not listed above with decay mode other than alpha emission or spontaneous fission and with radioactive half-life greater than 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

Any single radionuclide ........ 6x10-13 4x10-7 2x1O-14 3x1O-8 not listed above which decays by alpha emission or spontaneous fission.

Revision 21 Date 08/14/97 G-18

'Soluble (S); Insoluble (I).

2 "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 (1/3)

"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 liter of air, without regard to the degree of equilibrium, that will result in the ultimate emission of 1.3 x 105 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 Ci-hr/ml. where SA is the specific activity. of the uranium inhaled. The concentration value for Table II is 0.007 milligrams uranium per cubic meter of air. The specific activity for natural uranium is 6.77 x 10 7 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 10-7 curies/gram U U-depleted SA - (0.4 + 0.38 E + 0.0034 E2) 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 C are present in concentrations CA, CB, and Cc. and if

_ the applicable MPC's are MPCA, and MPCB. and MPCc respectively, then the concentrations shall be limited so that the following relationship exists:

(CA/MPCA) + (CBIMPCB) + (CC/MPCC) < 1

2. If either the identity or the concentration of any radionuclide in the mixture is not known, the limiting values for purposes of Aendix B shall be:

a. For purposes of- Table I. Col. 1 - 6x1O

b. For purposes of Table I. Col. 2 - 4x10-7

c. For purposes of Table II. Col. 1 - 2x10 4

d. For purposes of Table II. Col. 2 - 3x10 a

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 21 Date 08/14/97 G-19

Table I Table II Col. I Col. 2 Col. 1 Col. 2 Air Water Air Water

c. Element (atomic number) and isotope (PT11,Ci/ml) (pCi/ml) (pCi/ml) (pCi/ml)

If it Is known that Sr 90, I 125, I 126, I 129. I 131 (I 133, Table II only),

Pb 210. Po 210. At 211, Ra 23, Ra 224, Ra 226, Ac 227, Ra 228, Th 230. Pa 231.

Th 232. Th-nat. Cm 248, Cf 254, and Fm 256 are not present . . . . . . . . ........ ........ 3x10 6 If it is known that Sr 90. 1 125, I 126.

I 129 (I 131. I 133. Table II only).

Pb 210. Po 210, Ra 223, Ra 226. Ra 228, Pa 231, Th-nat, Cm 248, Cf 254. and 5 6 Fm 256 are not present . . . . . . . . ....... 6x10 ........ 2xl0 If it is known that Sr 90, I 129 (I 125, I 126, 1 131. Table II only), Pb 210.

Ra 226, Ra 228. Cm 248. and Cf 254 are not present . . . . . . . . . . . . . . ........ 2x10 5 ........ 6x10-7 If it is known that (I 129, Table II only),

Ra 226. and Ra 228 are not present . ........ 3x10 6 ........ 1x10 7 If it is known that alpha-emitters and Sr 90, 1 129, Pb 210. Ac 227. Ra 228.

Pa 230. Pu 241, and Bk 249 are not 9

present . . . . . . . . . . . . . . . . 3x1O ....... 1x1-0 1 ........

If it is known that alpha-emitters and Pb 210. Ac 227. Ra 228. and Pu 241 are not present . . . . . . . . . . . . . . 3xlO10 ........ lxO11 1 ........

If it is known that alpha-emitters and Ac 227 are not present . . . . .. . . 3x10 1 1 ........ lx10 2 If it is known that Ac 227. Th 230, Pa 231, Pu 238, Pu 239. Pu 240, Pu 242, Pu 244, 12 13 Cm 248. Cf 249 and Cf 251 are not present 3x10 ........ 1x10 ........

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 - 1x10-10 Ci/ml gross alpha activity: or 5x1O- 1 jICi/ml natural uranium or 75 micrograms per cubic meter of air natural uranium.

b. For purposes of Table II, Col. 1 - 3x10-12 Ci/ml gross alpha activity: 2x10 12 pCi/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 8" (MPCA) does not exceed 1/10. (i.e. CA/MPCA 1/10) and (b) 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 + CB/MPCS ... + < 1/4).

Revision 21 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, 1999, BVY 99-80 2.. "Suppiement to Request to AriendPrevious Approvals Grahted Under IOCFR20.30(a) to Allow for Disposal of Contaminated Soil", dated January 4t, BVY 00-02

3. 'Vermont YankeeNuclear Power Station, Request to Amend Previous Approvals Granted under l OCFR20.302,a) to Allow for Disposal of Contaminated Soil (TAC No. MA5950), dated June 15, 2000, NVY 00-58 H-1

VERMONT YANKEE.

NUCLEAR POWER CORPORATION 185 Old Ferry Road, Bratleboro, VT 05301-7002 (802) 257-5271 June 23, 1999

. . .. . BVY 99-80 United States Nuclear Regulatory Commission ATIN: 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 from M.B. Fairtile, USNRC, to L. A. Trembley. Vermont Yankee,

'Approval Under 10CFR20.302(a) of Procedures for Disposal of Slightly Contaminated Septic Waste on Site at Vermont Yankee (TAC No.

73776)", NVY 89-189, dated August 30, 1989 (c) Letter from I.J. Duffir, Vermont Yankee, to USNRC, 'Request to Amend Previous Approval Granted Under IOCFR20.302(a) for Disposal of Contaminated Septic Waste", BVY 95-97, dated August 30, 1995 (d) Letter from S.A. Varga, USNRC, to D. A. Reid, Vermont Yank-ee, "Approval Pursuant to 10CFR20.2002 for Onsite Disposal of Cooling Tower Silt - Vermont Yankee Nuclear Power Station (TAC No.

M93420)", NVY 96-39, dated March 4, 1996 (e) Letter from P.D. Milano, USNRC, to D. A. Reid, Vermont Yankee,

'Revised Safety Evaluation - Approval Pursuant to 10CFR20.2002 for Onsite Disposal of Cooling Tower Silt - Vermont Yankee Nuclear Power Station (TAC No. M9637 )", 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 Approvals Granted Under 10 CFR 20302(a) for Disposal of Contaminated Septic Waste and Cooling Tower Silt to Allow for Disposal of Contaminated Soil In accordance with OCFR 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. This application expands the allowable waste. stream,to include slightly contaminated soil generated as a residual by-product of on-site construction activities.

This 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 designated 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 wull save substantial costs and reserve valuable disposal site space for waste of higher radioactivity levels.

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VERMONT YANKEE NUCLEAR POWER CORPORATION BVY 99-80 IPage 2 of 2 A. radiological assessment and proposed operationil controls 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 constmction 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 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 nember of the public of less than one mrem/year during the period of active Vermont Yankee control of the site, or less than five mremlyear to an inadvertent intruder after termination 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 1 as well as the basis for approval will be incorporated into

.the ODCM.

We trust that the information contained in the submittal is sufficient However, should you have any questions or require futher information concerning this matter, please contact Mr. Jim DeVincentis at 802-2584236.

Sincerely, VERMONT YANKEE NUCLEAR POWER CORPORATION

IatanSen Licensing Manager Attachment cc: USNRC Region I Administrator USNRC Resident Inspector - VYNPS USNRC Project Manager - VYNPS VT Department of Public Service H-3

SUMMARY

OF VERMONT YANKEE COMMITMENTS BVY NO.: 99-80 The following table identifies commitments 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.

COMEMITMENT CCOMTMIIE DATE g . . ~~~~~~~~~OR"OUTAGE" None. ..

VYAPF 0058.04 (Sample)

AP 0058 Original Page 1 of 1 H-4

Docket No. 50-271 BVY 99-80 Attachment 1 Vermont Yankee Nuclear Power Station Assessment of On-site Disposal of Contaminated Soil by Land Spreading H-5

BVY 99-8 0 1Attachment 1 /Page 1 TABLE OF CONTENTS Page LIST OF TABLES ............................................................................ 2

1.0 INTRODUCTION

.).U..... 3 1.1 Background .. . 3 1.2 bjective . . ..................................

.3................ 3 2.0 WASTEDESCI RON ..... 4

3.0 SOIL DISPOSAL AND ADMITNISRAtIVE PROCEDlRE EQrUIRENiEvINDTrS . . . 5 4.0 EVALUA T IONOFNlRONEINTAL IMPACTS ........... . 6...........

6 4.1 Site Characteristics ................................ ; 6 4.2 R adiological Impact. .. 6 5.0 RADIOLOGICAL PROTECION .. . .

6.0 CONCLUSION

S ... 9 7.0 EFERECES ........ 9 H-6

Z.

BVY 99-80 / Attachment I I Page 2 LIST OF TABLES Table 1: Radioanalytical Results of Composite Soil Samples . . 10 Table 2: Estimated Total Radioactivity in Soil Volume . .................................

10 Table 3: Total Activity on South Field After Last Spreading Event ................. Il Table 4: Total Projected Radioactivity Remaiing on South Filed at License Expiration ........ . ; ;ll Table 5: All-Pathway Critical OrganfWhole 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-I 37 in Soil Volume after Land Spreading ........................................... . 12 Table 8: Present and Future Dose Impact Due to the Soil Spreading for Two Cases .. 13 H-7

BVY 99-80 / Attachment 1 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 Yankee Nuclear Power Corporation requested that the previous authorization foi 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 on June 18,1997 and the information was permanently incorporated into the ODCM as Appendix F.

In 1994, approximately 25.5 m3 of excess soil was generated during on-site construction activities. Sampling of the soil revealed low levels ofradioactivity 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 fashion as the septic waste and cooling tower silts; however, prior approval fromthe NRC would be required under 10 CFR 20.2002 (formerly 20.302(a)).

1.2 Objective 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 rnrem/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 approved on-site locations.

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L.

BVY 99-80 1Attachment 1/ Page 4 2.0 WASTE DESCRIPTION The soil that is the-subject ofthis evaluation was derived.from excavations resulting from activities associated with a new security 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. 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 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 had 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 obtained 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 three into one sample. Soil samples were sent to the Yankee Atomic Ervironmental 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 purpose of 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.

The mass of soil (dry) was estimated.by multiplyingthe total in-situ.volume (25.5 m3 ) by its wet 4ensity, 1.47E+03 kg/m3, and then dividing by the wet:dry ratio of 1.12, thus yielding a mass of 3.35E+04 kg (dry). The mass 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 m 3. Table 2 presents the estimated total radioactivity in the 25.5 m 3 of soil at the time of sample collection and analysis, and decay corrected to the present.

H-9

BVY 99-80 / Attachment I / Page 5 3.0 SOIL DISPOSAL AND ADMINISTRATIVE PROCEDURE REQUREMENTS 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 and approved for septic waste and cooling tower silts disposal has been used for all past disposal operati6ns, 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 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 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 material at an off-site licensed facility if detectable radioactivity is found.

Records of the disposal that will be maintained include the following:

(a) The radionuclide concentrations detected in the soil (measured to enviromnental lower limits of detection).

(b) The total volume of soil disposed of.

(c) The total radioactivity in the disiosal 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 ensire 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 farmer leasing land used for the disposal of soil will be notified of the applicable restrictions placed on the site due to the spreading of low level contaminated material.

These restrictions are the same as detailed for the previously approved septic waste spreading application.

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BVY 99-80 / Attachment 1 / 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 'outh 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 giound.

(b) Inhalation of resuspended radioactivity.

(c) Ingestion of leafy 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 aialyses include:

(a) Exposure to ground contamination and re-suspended radioactivity is for a period of 104 hours0.0012 days 0.0289 hours 1.719577e-4 weeks 3.9572e-5 months 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 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months per week for 6 months).

H-11

BVY 99-80 / Attachment 1 /Page 7 (b) For the purpose of projecting and llustrating the magnitude of dose impacts over the remaining life of the plant, it is assumed'that the' concentration levels of activity as of April 1, 1999 remain-constant. Table' I 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 duing the Vermont Yankee active control of the disposal sites until'2013, no plowing is assumedto 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 hours0.025 days 0.6 hours 0.00357 weeks 8.2188e-4 months to 0 hours0 days 0 hours 0 weeks 0 months 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 factois 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 hunan 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.

(f) 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 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 forms 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.

H-12

BVY 99-80/ Attachment 1 / Page 8 As shown in the previous subiittals, in which the concentrations of Co-60 and Cs-137 ii 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 1, 1999 is 184 days). In addition, the total activity on the south field is presented assuming the addition of the 25.5 m3 of soil subject of this evaluation.

The total activity that would be present ofm south field at license termination (i.e., total elapsed time of 14 years post April 1, 1999, or.2013), assuming no fixture additions of material containing radioactivity after disposal of the proposed soil volume 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 body and organ at the present and at license expiration is presented in Table

7. The present and fiture 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 mrem/yr to an inadvertent intruder after control of the site is assumed to be relinquished. This analysis shows that significant dase margin still exists on the approved disposal plots to accormmodate 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.

H-13

BVY 99-80 / Attachment 1 / Page 9

6.0 CONCLUSION

S -

Soil generated from on-site construction activities reflects an earthen type matenial 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 boundary conditions specified in the ODCM. That is, with respect to the addition of the approximately 25.5 rn3 soil pile to the existing radioactivity already spread on the south field:

1. Total doses to the whole body and critical organ to the hypothetically maximally exposed individual were estimated as 3.OOE-02 mremlyr and 1.04E-01, 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 1.13E-01 mremlyr and 2.21E-01 mrem/yr, respectively, which are less than 5 mremlyr.

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 m 3 of soil alone would at present be 4.17E-03 mrem/yr whole body and a maximum organ dose of l.46E-02 mrem/yr. In addition, for the inadvertent intruder, the whole body dose would be 1.60E-02 mrem/yr, and a maximum organ dose of 3.1 lE-02 mrem/yr. Each of these doses also meet the boundary conditions specified in the ODCM.

7.0 REFERENCES

() 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 with 10CFR Part 40, Appendix I", dated October 1997.

H-14

BVY 99-80 /Attachunent I IPage 10 Table 1 Radioanalytical Results of Composite Soil Samples Cs-137 CO-60 (pCilkg) (pCilkg)

Sample ED April, 1995 April, 1999 April, 1995 April, 1999 G22716 234 213 49** 29 G22717 522 476 143 84 G22718 337 307 37** 22 _

G22719 291 265 111 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 376 343 62 . 37*

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 G22735 235 214 100 59 Average 328 299 85 50 Maximum Value .810 739 143 84 Minimum Value 107 98 5 3 StandardDeviation 191 174 37 22 Avcrage wet to dry sample weight ratio equal to 1.12. Avcrage wet density equal to 1.47 gnincc

$ The apparent response of the gamma isotopic analysis which was less than Miuimrn Detectable Concentration.

Table 2 Estimated Total Radioactivity in Soil Volume Volume Average Concentration Total Activity of Soil Mass (pCi/kg-dry) (ICi)

Radionuclide (m) (kg-drY) April, 1995 April, 1999 April,1995 April, 1999 Cs-137 25.5 3.35E+04 328 299 11.0 10.0 Co-60 25.5 3.35E+04 85 50 2.8 1.7 H-15

BVY 99-80 / Attachment 1 / Page 11 Table 3 Total Activity ow 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 (p.Ci/acre) *1999 (pCilacre) disposal (jiCi/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 Table 4 Total Projected Radioactivity on South Field Remainng at License Expiration Total Activity as of License Expiration Radionucide (gCacre)

Mn-54 8.9E-07 Co-60 0.89 Zn-65 7.6E-09 Cs-137 2633 Table 5 All Pathway Critical Organ/Whole Body Dose Conversion Factors During Vermont Yankee Control of Disposal Sites Critical Organ Whole Body Dose Factor Dose Factor Radionuclide 1ndividual/Organ (mremlyr per lpCUscre) (mremlyr per pMiCacre)

MNi-54 Adult/GI-LLI . 3.75E-04 . 1.93E-04 Co-60 Teenthing 7.17E-04. . . 5.3 IE-04 Zn-65 Child/Liver 1.64E-02 1.03E-02.

Cs-137 Child/Bone 2.66E-03 7.02E:04 H-16

BVY 99-80 / Attachment I Page 12 Table 6 All Pathway Critical Organ/Whole Body Dose Conversion Factors Post Vermont Yankee Control of Disposal Sites (Inadvertent Intriider)

Critical Organ Whole Body Dose Factor Dose Factor Radionuclide Indimdual/Organ (mremlyr per isCIIacre) (mremtyr per pCilacre)

Mn-54 Teen/Lung 1.02E-02 3.12E-03 Co-60 TeenfLung 3.19E-02 9.09E-03 Zn-65 Child/Liver 1.89E-02 1.25E-02 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 Impace Future Dose Impact' (Maximally exposed (Inadvertent Intruder)

Case individual)

Dose Individual/ Dose Individuall (mremlyr) Organ (mremnyr) Organ Cobalt-60 4.75E-04 Whole body 1.29E-03 Whole body 6.42E-04 Max. Organ 4.51E-03 Max. Organ Cesium-137 3.69E-03 Whole body. 1.47E-02 Whole body 1.40E-02 Max. Organ 2.66E-02 Max. Organ 1Based on inventory of Co-60 of 0.895 pCilacre and Cs-137 of 5.26 pCi/acre in April 1,

1999.

2 Based on inventory of Co-60 of 0.141 'iCilacre and Cs-137 of 3.82 pCi/acre in April 1, 2013.

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BVY 99-80 1 Attachment 1/ Page 13 Table 8.

Present and Future Dose Impact 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 (mremlyr) Organ Case One South Field as it currently 2.58E-02 Whole body 9.70E-02 Whole body exists 8.96E-02 Max. Organ 1.89E-01 Max. Organ Case Two South Field if disposal of 3.OOE-02 Whole body 1.13E-01 Whole body soil volume is approved 1.04E-02 Max. Organ 2.21E-01 Child/Bone Increase -indose impact 4.17E-03 Whole body 1.60E-02 Whole body from disposal of soil 1.46E-02 Max. Organ 3.1lE-02 Max. Organ I -18

VERMONT YANKEE NUCLEAR POWER CORPORATION 185 Old Ferry Road, Brattleboro, Vr 05301-7002 (802) 257-5271 January 4, 2000 BVY 00-02 United States Nuclear Regulatory Commission ATTN: 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

Subject:

Vermont Yankee Nuclear Power Station License No. DPR-28 (Docket No. 50-271)

Supplement to Request to Amend Previous Approvals Granted Under 10 CFR 20302(a) to Allow forlDisposal of Contaminated Soil Reference (a) provided Vermont Yankee's application to amend the previously 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. fhe 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 information related to the estinated volume and dose consequences of the proposed future material was needed to complete your reyiew.

Attachment I has been revised accordingly to include the information requested. Attachment I supercedes the evaluation previously submitted.

We trust that the information 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, Vermont Yankee Nuclear Power Corporation

'Gautam Sen Licensing Manager Attachment cc: USNRC Region I Administrator USNRC Resident Inspector - VYNPS USNRC Project Manager - VYNPS VT Departaent of Public Service H-19

SUMEARY OF VERMONT YANKEE COMMITMENTS BVY NO.: 00-02 I

The foUowing table identifies commitments made in this document byVermontYankee. 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.

COMMITNT I.COMMIED DATE

__ __ _ _ _ _ _ _ _ _ __ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ O R "'OU TA G E t None . : ~~~~~~~~N/A -

VYAPF 0058.04 AP 0058 Original Page 1 of I 1H-20

. I Docket No. 50-271 BVY00-02 Attachment 1 Vermont Yankee Nuclear Power Station Assessment of On-site Disposal of Contaminated Soi by Land Spreading H-21

BVY 00-021 Attachment I /Page 1 of 17 TABLE OF CONTENTS LIST OF TALES........................................................2........ .. 2 IS TRODUCTION . . . ....................................... 3......

1.1 Background ......... . 3 1.2 Objective .................................... ..................... 3 2.0 WASTE DESCRIPTION 4 4............

3.0 SOIL DISPOSAL AND ADMINISTRATIVE PROCEDURE EQULIRIEONFTS ................................ ....................... 5............

4.0 EVALUATION OF ENVIRONENTAL PACTS . .. 6 4.1 Site Characteristics .................. . 6 4.2 Radiological Impact ...................... 7 5.0 RAD;IOLOGICAL PR.OE CnfION .......................... .9

6.0 CONCLUSION

S .......................... 9

7.0 REFERENCES

.......................... 11 H-22

BVY 00-02 Attachment 1/Page 2 of 17 LIST OF TABLES Table 1: Radioanalytical Results of Composite Soil Samples . ;12 Table 2: Estimated Total Radioactivity in 25.5 m 3 Soil Volume ......................... 12 Table 3: Estimated Total Radioactivity in Future Soil Additions . . 13 Table 4: Record of Septic and Silt Radioactive Material Spread each Year on South Field Disposal Plot............... 13.

Table 5: Total Projected Radioactivity Remaining on SouthFiled at License Expiration ...... . . . . . . ...................... 14 Table 6: All-Pathway Critical Organ/Whole Body Dose C6nversion Factor .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 25.5m3 Soil Disposal ............................

,, ...... . ; 15 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 7/15199 Plus all Annual Projected Disposals of Septic, Silt and Soil ....................... . .............. 16.,

16 Table 11: Summary of Dose Impacts Associated with Different Disposal Scenarios ...... ,,,. 17 H-23

BVY 00-02 / Attachment 1I/Page 3 of 17 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 Yankee Nuclear Power Corporation requested that the previous authorization for on-site disposal of very low-level radioactive naterial in septic waste be amended to permit the on-site disposal of slightly contaminated 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.

In 1994, approximately 25.5 m 3 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 radionucides and activity to the septic waste and cooling tower silts previously encountered. 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 would be required under 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 boundary 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:

1. 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.

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BVY 00-02 / Attachment 1/ Page 4 of 17 2.0 WASTE DESCRIPTION The existing accumulation of contaminated-soil was derived from excavationactivities associated withthe construction of a new security fence along the plant's Protected Area boundary. The volume of soil generated was approximately 25.5 in3 , 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 towers where it was deposited on the ground surface and covered to prevent erosion. This location was selected because it was away fron 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 of below detectable removable contamination redistributed by foot traffio from inside the piant to walkways and.parling areas. Subsequent surface runoff 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 of 20 composite soil samples were collected to characterize the accumulated volime. Composites were obtained by taking a grab sample from opposite sfdes of the 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 required of envirorlnental media. Results of the analyses are presented in Table 1.

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 (25.5 m 3 ) by it's wet density, 1.47Et03 kglm3,'and then dividiig by the w'et:dri ratio of 1.12, thus 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 radionuclide in the 25.5 n3 . Table 2 presents the estimated total radioactivity in the 25.5 m3 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 m 3 (900 ft3 ) of soil included in this request, it is anticipated that the need to dispose of very low-level contaminated soil will occur in the future. Each spring, approximately 28.3 m3 (1000 ft3 ) of road and walkway sand spread during the winter season is swept up from inside the Protected Area 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 H-25

L.

BVY 00-02 Attachment I Page 5 of 17 evaluating the radiological impact,of potential future soil disposals, it is assumed that an additional,28.3 m3 per'year of sand / soil is contaminated at'the same concentration levels as onginally observed .(April 1995) in the currehtly colle'cted 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 28.3m 3 of soil. It is assuned 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 has 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 uCilyear Co-60 2.58 uCi/year Zn-65 0.269 uCi'year Cs-134 0.010 uCilyear Cs-137' 6.21 uCilyear The maximum radioactivity inventory resulting from the accumulated buildup of past and projected future disposal operations (i.e., septic waste, cooling towier silt; plus the existing 25;5m3 of accumulated soil along with a projected annual addition of 28.3m 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 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 soi1 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.

H-26

BVY 00-02 I Attachment 1 / Page 6 of 17 Records of the disposal that will be maintained iiclude 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 maximum allowable accumulated activity determinations required to demonstrate that the dose limits imposed on the land spreading operations have not been exceeded.

To ensure that the addition of soil containing low levels of radioactivity will not exceed the boundary dose limits, each new spreading operation 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 farmer 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 of low level contaminated material. These restrictions are the same as detailed for the previously approved septic waste spreading application.

4.0 EVALUATION OF ENVIRONMENTAL IMPACTS 4.1 Site Characteristics All designated disposal sites are located on the Vermont Yankee Nuclear Power 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 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 operations will continue to use the South field as the disposal plot.

H-27

BVY 00-02 / Attachment 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 boundary 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 tower 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 were identified and icluded 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.

(f) Ingestion of cow's milk.

As shown in the previous application for septic waste disposal, the liquid pathway was found to be an insignificant contributor to the dose for the 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 assulmed 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 contamEiation and re-suspended radioactivity is for a period of 104 hours0.0012 days 0.0289 hours 1.719577e-4 weeks 3.9572e-5 months per year during the Vermont Yankee active control of the disposal sites and continuous ther6after. The 104-hour interval is representative of a farmer's time spent on a plot of land (4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months per week for 6 months).

(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 ft) 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 maximum individual dose impact from the buildup of disposed material H-28

L..

BVY 00-02 Attachment I / Page Bof 17 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 fomiing a source of unshielded direct radiation.

(d) The crop exposure time was changed from 2160 hours0.025 days 0.6 hours 0.00357 weeks 8.2188e-4 months to 0 hours0 days 0 hours 0 weeks 0 months to reflect the condition that no radioactive material is dispersed difectly on crops for human or animal consumption. Crop contamination 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 source 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-

- 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.

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 Ci.e., septic, cooling tower silt and soil) is presented in Table 5.

H-29

z-.

BVY 00-02 / Attachnent I / Page 9 of 17 In order to evaluate the dose impact associated with the different disposal streams, a dose assessment was performed for the following four disposal scenarios:

() Impact from past septic and silt spreading only - Table 7 (II) Impact from past septic and silt spreading, plus a single 25.5m 3 soil disposal operation for the existing accumulated soil - Table 8 (E) 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 Iyr). -Table 9 (IV) Impact from past septic and silt disposals plus annual projected disposals of septic, silt and soil. -Table 1O For each scenario, the critical organ and whole body dose frm all pathways to a maximally exposed individual for both the Vermont Yankee control period and the inadvertent intruder situation were calculated. The dose calculations 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."

A 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 accumulated soil will be well within the accepted dose limit criteria of 1 mremlyr to any organ or whole body during the control period, 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 PROTECIION The disposal operation of the soil will follow the applicable Vermont Yankee procedures to naintain doses as low as reasonably achievable and within the specific dose criteria as previously 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 m 3 /year of soil through on-site land spreading will meet the existing NRC approved boundary dose conditions specified in H-30

1I BVY 00-02/ Attachment I Page 10 of 17 the Vermont Yankee ODCM (see Appendix B f6r Septic Waste Disposal). That is, with respect to.the additionofthe initial 25.5 m 3 of soil along withthe projected 28.3 m 31year o soil and the projected future disposal of septic and silt waste to the existing radi6activity already spread on the south field:

1. Total doses to the whole body and critical organ of the.hypothetically maximally exposed individual were estimated as l.15E-01 mrem/yr and 4.03E-01 mrem/yr, respectively, which are less than the prescribed 1.0 mremr/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 esdmated as 7.57E-01 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 access control as the end of the operating license.

3. For purposes 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 m3 and the projected 28.3 m 3/year of soil will continue to meet the existing boundary conditionsas approved by the NRC for septic waste and cooling tower silt H-31

BVY 00-02 /Attachment I /Page I 'of 17

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, "Calciuation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with:

10CFR Part 40, Appendix I", dated October 1977.

H-32

BVY 00-02 /Attachment 1/ Page 12'of 17 Table 1 RadioanalyticalResults of'CompositeSampls Takenfrom 25.5m3 SoiPile Cs-137 Co-60 (pCi/kg) (pCi/kg)

Samjle ID April. 1995 July 15. 1999. Anril. 1995 Julv 15. 1999 G22716 234 2122 49 28.0

G22717 522 473.4 143 81.7 G22718 337 305.7 37 21.1

G22719 291 263.9 11 63.4 22720 348 315.6- 47 26.8

G22721 135 122.4 : 73 41.7 G22722 1&7 97.0 82 46.&

G22723 222 201.4 140 80.0 G22724 180 163.3 92 52.6 G22725 269 24.0 118 67A G22726 810 734.7 114 65.1.

G22727 378 . 342.8 106 60.6 G22728 810 734.7 124 .70.8 G22729 376 341.0 62 35.4 G22730 331 300.2 87 49.7 G22731 253 229.5 5 2.9 G22732 150 136.0 58 33.1 G22733 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 Maximum 810 735 143 82 inEimum 107 97 5 3 StdDev. 186 .169 36 20 The apparent response of the gamma isotopic ailysis was less than theMinimum Detectable Concentration.

Table 2 Estimated Total Radioactivity in 25.5m 3 Accumulated Soil Yolume Soil Average Concentration Total Activity of Soil Mass .(pCi/kg - dy) (uci)

Nuclide 2m3) P - dny) April 1995 July 15. 1999 Apnl 1995 July 15, 1999 Cs-137 25.5 3.35E+04 328 298 11.0 . 10.0 Co-60 25.5 3.35E+04 85 49 2.8 1.6

- . -33.. .q.

. 1:

BVY 00-02 1Attachment I I Page 13 of 17 Table 3 Estimated Total Radioactivity in Future Soil Additions Volume Soil Aveage Conentration of soil Mass (pCi/kg-dry assuming) Total Activity Nuclide Lm6) (kg-dry) (April, 1995 concentrations) (uCifer)

Cs-137 28.3 3.72E+04 328 12.8

.Co-60 28.3 3.72E+04 85 3.16 Table 4 Record of Septic and Silt Radioactive Material Spread Each Year on the South Field Spreading Material Mn-54 Co60 Zn-65 Cs-134 Cs137 Ce-141 Year Date Type (uCi/acre) (uCi/acre) (uCiacre) (uCi/acre) uCilacre) (uCiUacre) 1990 10/31/90 Septage 0.00 3.89 0.00 0.00 026 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 10/19192 septage 0.11 1.73 0.52 0.05 0.32 0.006 1993 10114/93 septage 0.05 1.41 0.21 . 0.00 0.30 0.00 1994 06/14/94 septage 0.08 OA3 0.00 0.00 0.09 0.00 1995 06/29/95 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 c0.00 spreading 1997 0618/97 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 *07/119 Septage 0.11 1.47 020 . 0.00 025 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 0.010 621 0.001 not disposed of on 1.9 significant acre field H-34

BVY 00-02 / Attachment I / Page 14 of 17 Table 5 Total Projected Radioactiity Remaining on South Field after License Termination

.4 Accuin. Activity Accum- Activity Accum. Activity Accum. Activity in Silt & Septic in Soil Total All Paths Total All Paths

( Year 2013 @ year 2013 @ year 2013 @ Year 2013 Nuclide (uCfi) (uCi) (uCi) :(uCi/acre)*

Mn-54 20.26. 026 0.14 Co-60 19.68 21.83 41.51 21.85 Za-65 0.42 0.42 0,22 Cs-137 119.03 154.74 273.78 144.09 Cs-134 0.04 *;6 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.

Table 6 All Pathway Critical Organ/Whole Body Dose Conversion Factors Durine VY Control Post VY Control (Intruder ScenarIo)

Critical Organ Whole Body Critical Organ Whole Body

'Dose Factor Dose Factor Dose Factor Dose Factor (mremlyr per (mremlyr per (mremlvr per (mremlyr per Radionuclide Indindual/Orean iiCi/acre) uCi/acre) uiCVacre) UCi/acre)

M.n-54 AdultIGI-LLI 3.75E-04 1.93E-04 1.02E-02 3.12E-03 Co-60 Teen/Lung -. - 7.17EM04 5.31E-04 3.19E-02 9.09E-03 Zn-65s Child/Liver

l.62E 1.03E-02 1.89E-02 . 1.25E-02 Cs-137 Child/Bone 2.66E-03 7.02E-04 6.98E-03

3.85E-03 H-35

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BVY 00-02 / Attachment I Page 15 of 17 Table 7 (Scenario )

Dose Impact from Past Septic and Silt Spreading on South Field (as of 7115/99)

Control Scenario:

All Other Spreadings Maxin ium Organ Whole Body Maximum Whole Body Eidf-Lif To Date Dosa Factor Dose Factor Organ Dose Dose (uCuacre) (m vrhCttacrc) (sn Vie acrel I (mrernhr'

. Ma-54 0.S6 020 3.1,T5E-04 193E-04 7.35E-05 3.78E-05

' Co60 5.27 6.86 7.117E-04 S31E-04 4.92E-03 3.64E-03 Za-65 *0.67 0.23 1.462E-02 1.03E-02 3.77E-03 2.40E-03 CS-137 30.17 31.92 2.4s6E-03 . 7JflE 04 849E-02 224E-02 Total Do: 937E-02 2.SSE-02 Dose Limit

%ofimit 937%h 2.85%

Intruder Senario:

AU Other Activity on Plot Spreadings Decayed to Maximut Organ Whole Body Maximum Wholc Body HalfLife To Date Year2013 Dose Factor DoseFactor OrpnDose Dose (uClUre (uCVacre) (MMMMuCear) (mrem/fyr) (mremlvr)

M-54 0.86 020 2.3 IE-06 1.02E-02 3.12E-03 236E-08 722E-09 Co-60 5.27 6.86 1.08E+00 3.19E-02 9.09E-03 3.46E-02 9.SSE.03 Zn-65 0.67 0.23 15-07 1.89E-2 125E-0Z 2.17E-09 1.43E-09 Cs-137 30.17 31.92 2.315E+01 69SE-03 3.85E-03 1.625-01 891E-02 Total Dose: 1-96E-01 9.90E02 Dose Liwlt 5 5

% ofLimit 3.92% 1.98%

Table 8 (Scenario m)

Dose Impact from Past Septic/Silt Spreading and Single 25.5m 3 Soil Disposal Control Scenario: All Spradins Mamum.Organ Whole Body Mamum Whole Body Half-Life to Dae' Dose Factor Dose Factor Organ Dose Dose (years) (uCVacm) (mumeyrAzCVacre) (mrem/yruC/acrel (mrem1/Yr)

Mn-54 0.86 0.196 3.75E-04 193E504 735E-05 , 3.78E2-U Co-60 5.27 7.70 7.17E-04 5.31E-04 552-03 4.09E-03 Zn-65 0.67 0.233. 1.62E-02 1.03E-02 3.77E 03 2.40E-03 nonenr "etr Cs-137 30.17 37.9 2.66E-03 7.02E-04 Total Dose: l.08E-01 3.26E02 Dose Limit 1 l

% oflimit 10.83% 3.26' Intruder Seenuria All Spreadugs Activity on Plot Maximm Organ WholeBody Maximum Wholc Body Half-Llfe to Date Decayed to 2013 Dose Factor DoseFactor OrganDose Dose (Yea) (uCvacr) (mreMnbzCVacre) (mrcm1/4fuCVacre' rmrembr) - (mrem/vrl Mn-54 0.86 0.196 2.31E-06 1.02E42 3.12-03 2.36E-08 7.22E-09 Co-60 527 7.70 L22E+00 3.19E-02 9.09E-03 3.8SE-02 1.1 IE-02 Zn-65 0.67 0.233 1.15E-07 1.89E-02 1.251-02 2.17E-09 1.43E-09 Cs-137 30.17 37;19 2.70E+01 698E-03 3.85E-03 I SSEOI 1.04E-01 Total Dose 2.27E-01 1.15E01 Dose Limit 5 *5

% of Llnit: 4.54°/ 2-30%

H-36

UNITED STATES NUCCEAR REGULATORY COMMISSION WASHINGTON, D.C. 205-01 J une 15l 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 slighty contaminated soil and soiVsand material to the list of already approved materials-(i:e., septic waste and cooling tower ilt) 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 NRC has published an Environmental Assessment and Finding of No Significant Impact in the FederalRegister on June 15 , 2000 (65 FR 37583 3.

John>. Zwolinski, Director Division of Ucensing Project Management Office of Nuclear Reactor Regulation Docket No. 50-271

Enclosure:

Safety EvaluaUon cc w/encl: See next page H-37

Vermont Yankee Nuclear Power Station cc:

Regional Administrator; Region I 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 Service 185 Old Ferry Road 112 State Street P.O. Box 7002 Montpelier, VT 05620-2601 Bratleboro, VT 05302-7002 Mr. Michael H. Dworkin, Chairman Resident Inspector Public Service Board Vernont Yankee Nuclear Power Station State of Vermont U. S. Nuclear Regulatory Commission 112 State Street P.O. Box 176 Montpelier, VT 05620-2701 Vernon, VT 05354 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. O..Box566 Govemor 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 Attomey General One Ashburton Place, 19th Floor Boston, MA 02108 Ms. Deborah B. Katz Box 83 Shelbume Falls, MA 01370 H-38

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555C-O0 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 10 CFR 20.2002

(previously 10 CFR 20.302) to allow the additon of slightly contaminated soil and soiUsand material to the list of already approved materials (i.e., septic waste and cooling tower silt) for on-site disposal via land spreading 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 plants operating lcense 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 walkways 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 28.3 cubic meters generated annually tfiereafter). The licensee's evaluation shows that the soil/sand 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 matenal 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.

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 (rmeasured 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 radioactMty 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 dose condition values imposed (i.e., imposed by this 10 CFR 20.2002 application) on the land spreading operafion have. not been exceeded.

The bounding dose conditions for the on-site disposals are as follows:

1. The annual dose id 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 probabte 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, cooling tower silt, soil and soil/sand 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. 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.

The licerisee assessed the dose that may be received'by the maximally exposed 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 NRC 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 mrern for the time period of active site control.

H-40

3

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.

These values are less than the prescribed annual dose conditon 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 CONCWSION The staff finds the licensee's proposal to dispose of the low-level radioactive soil and soil/sand material, pursuant to 10 CFR 20.2002,in the same manner, locaUon, and within the bounding dose conditions as the materials (i.e., septic waste and cooling tower silt) previously approved by the NRC to be acceptable because the bounding conditions will continue to be met.

Principal Contributor: S. Klementowicz Date: June 15, 2000 1..

. .~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~

H-41

Anpendix I

1. "Request to Amend Previous Approval Granted Pursuant to IOCFR202002 for Disposal of Contaminated Soil", dated September 1', 2000, BVY 00-71

.2. Vei.n Yake m . n to

. 2. Vermont Yankee .Nuclear Power Station - Safety Evaluation for an Amendrnent to an Approved I OCFR20.2002 Application (TAC No. MA9972)", dated June 26h, 2001, NVY 01-66 I-1

VERMONT 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 ATTN: Document Control Desk Washington, DC 20555

References:

(a) Letter, VYNPC to USNRC, "Request to Amend Previous Approvals

- Granted under 10 CFR 20302(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 10 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 0 CFR 20.2002 for Onsite Disposal of Cooling Tower Silt -

Vermont Yankee Nuclear Power Station (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) submits this application to amend the previously. granted approval to dispose of slightly contaminated soil. This application expands the allowable waste 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 m3 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.

I-2

VERMONTYANKEE NUCLEAR POWER CORPORATION BVY 00-7 1/Page 2 of 2 In addition, VY requests that NRC's review recognize that, although VY indicated in Reference (b) that the south disposal field (approximately 1.9 acres in size) is currently expected to be used for disposal of the subject material, VY is also authorized to use the alternate north disposal field (approximately 10 acres in size). Approval to use boih the north and south fields 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 283 m3 of soil generated each year, to within the limits assumed in the radiological assessment previously submitted in Reference (b).

A rdiologicaj assessment 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 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 basis for approval will be incorporated into the Off-Site Dose Calculation Manual.

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 Mr. Jim.

DeVincentis at 802-258-4236.

Sincerely, Vermont Yankee Nuclear Power Coporation Gau Sen Licensing Manager cc: USNRC Region I Administrator USNRC Resident Inspector - VYNPS USNRC Project Manager - VYNPS VT Department of Public Service I-3

SUMMARY

OF VERMONT YANKEE COMMITMENTS BVY NO.: 00-71 I

The following table identifies commitments 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.

COMMITMENT COMMIITTED DATE OR "OlUTAGE"

'- None N/A VYAPF 0058.04 AP 0058, Revision I Page I of I I-4

{NUCLEAR REGULATORY COMMISSION rs. xRBy WASHINGTON, D.C. 20555-0001 aE)

June 26, 2001 Mr. Michael A. Balduzzi NvY 01-66 Vice President; Operations Vermont Yankee Nuclear Power CorporaUon 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 hs completed its review of the Vermont Yankee Nuclear Power Corporation (VYNPC) request dated September 11, 2000, to amend an approved 10 CFR 20.2002 (previously 10 CFR 20.302) appjication dated June 23, 1999, as*

supplemented on January 4, 2000. The licensee requested NRC approval to 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 soi/sand 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 VYNPO's application.

Pursuant to the provisions of 10 CFR Part 51, the NRC has published an Environmental Assessment and finding of No Significant Impact in the FederalRegisteronJune 14,2001 (66 FR 32399).

Sincerely, Joh A. Zwolinski, Director Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket No. 50-271

Enclosure:

Safety Evaluation cc w/encl: See next page

Vermont Yankee Nuclear Power Station cc:

Regional Administrator, Region I Ms. Deborah B. Katz U. S. Nuclear Regulatory Comrnission Box 83-475 Allendale R6ad Shelbume Falls, MA 01370 King of Prussia, PA 19406 Mr. Raymond N. McCandless

Mr. David R. Lewis Vermont Department of Health Shaw, Pittrnan, 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. Salemb7er, 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 Vermont 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 Vemon, 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. Box 566 Deputy Attomey General. Putney, Vr 05346-0566 33 Capitol Street Concord, NH 03301-6937 Chief, Safety Unit Office of the Attomey General One Ashburton Place, 19th Floor Boston, MA 02108 I-6

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20655-0001

'SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION VERMONT YANKEE NUCLEAR POWER CORPORATON 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 amenda Title 10 of the Code of FederalRegulations(10 CFR)

Section 20;2002 (former 10 CFR 20.302) application, dated June 23, 1999, as supplemented on January 4, 2000, that was approved by the U.S. Nuclear Regulatory Commission (NRC).

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 materals (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, 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. ln 1997, the NRC amended the approved on-site disposal application to also include contaminated cooling tower silt material. In 2000, the NRC amended the approved on-site disposal application to also include a one-time disposal of slightly contaminated soil and an annual.

disposal of 28.3 cubic meters of slightly contaminated soil/sahd material.

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 soillsand 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 the same volume that was approved In the January 4,2000, request. The licensee performed a comprehensive radiological evaluation which included the annual disposal of 28.3 cubic meters of soil and soiLlsand 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 materTial using a land spreading technique consistent with the current commiments for on-site disposal of septic waste, cooling tower silts and sand/soil material previously approved by the NRC. The licensee will continue to use the I-7

designated and approved areas of their property which include approximately 1.9 acres, which currently receive the septic waste, cooling tower silts and soilsand materal, and approximately 10 acres which have not been previously used for disposal. Deterrnination of the radiological dose impact of the new matenal has been made Wased 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:

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 rdioactivity 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 actiMty determinaliQns required to demonstrate that the dose 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 hypotheffcal 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.

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 'conditions, for each new spreading operation the licensee will calculatb an estimate of the total radioactivity that includes all past disposals 6f septic waste, cooling tower.

silt, and soil/sand and soil material on the designated disposal plots. This will be compared with the bounding dose condition value or equlivalent radioactivity value on a per acre basis.

The licensee assessed the dose from soil and soill/sand material 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 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).

I-8 '

L.

The licensee's dose assessment is as follows:

1. Total annual doses to the whole body and critical organ of the hypothetically rraximally

exposed indrvidual 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 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.

These values are less than the prescribed annual 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 potenbal Impact, of all disposals (past and future) on the approved disposal site.

3.0 CONCWSION 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 soiVsand 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 1-9

ML031620372

Text

Subject:

SUMMARY

1.0 INTRODUCTION

1.0 INTRODUCTION

Reference:

1.0 INTRODUCTION

Reference:

Dear Mr. Tremblay:

SUBJECT:

REFERENCE:

Reference:

Subject:

Dear Sir:

1.0 INTRODUCTION

SUBJECT:

Dear Mr. Reid:

Enclosure:

1.0 INTRODUCTION

2.0 BACKGROUND

4.0 CONCLUSION

Attachment:

References:

Subject:

1.0 INTRODUCTION

6.0 CONCLUSION

7.0 REFERENCES

1.0 INTRODUCTION

6.0 CONCLUSION

7.0 REFERENCES

References:

Subject:

SUMMARY

1.0 INTRODUCTION

6.0 CONCLUSION

1.0 INTRODUCTION

6.0 CONCLUSION

7.0 REFERENCES

References:

Subject:

6.0 CONCLUSION

7.0 REFERENCES

6.0 CONCLUSION

7.0 REFERENCES

SUBJECT:

Dear Mr. Newton:

Enclosure:

1.0 INTRODUCTION

References:

Subject:

SUMMARY

SUBJECT:

Dear Mr. Balduzzi:

Enclosure:

1.0 INTRODUCTION

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