Text

Effluent & Waste Disposal Semiannual Rept for Fist & Second Quarters,1994
	ML20072N038
Person / Time
Site:	Vermont Yankee File:NorthStar Vermont Yankee icon.png
Issue date:	06/30/1994
From:	; VERMONT YANKEE NUCLEAR POWER CORP.
To:	;
Shared Package
ML20072N032	List: ML20072N023; ML20072N038;
References
	NUDOCS 9409020295
	Download: ML20072N038 (187)
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{{#Wiki_filter:. _ _. _ _. -= t P EFFLUENT AND WASTE DISPOSAL b SEMIANNUAL REPORT FOR FIRST AND SECOND OUARTERS, 1994 i r i E i i l i VERMONT YANKEE NUCLEAR POWER STATION ( 1 R12\\f0 1 'sa an.. m. -m_. 0 44090?o245 9409: 4 D D C K 0 5 ):::C.. ? 1 PDR F 'iP R....,.# f t

TABLE lA Vermont Yankee Effluent and Waste Disposal Semiannual ReDort First and Second Quarters. 1994 Gaseous Effluents - Summation of All Releases Est. Quarter Quarter Total Unit 1 2 Error. % A. Fission and Activation Gases 1. Total release Ci . 7 7 E +02 <7.86E+02 +1.00E+02 2. Average release rate for period uCi/sec <9.88E+01 <1.00E+02 3. Percent of Tech. Spec. limit (1) B. lodines 1. Total Iodine-131 Ci 9.00E-04 8.91E-04 5.00E+01 2. Average release rate for period uCi/sec 1.14E-04 1.13E-04 3. Percent of Tech. Spec. limit (1) C. Particulates 1. Particulates with T-1/2 > 8 days Ci 9.83E-04 4.94E-04 15.00E+01 2. Average release rate for period uCi/sec 1.25E-04 6.28E-05 3. Percent of Tech. Spec. limit (1) 4 Gross alpha radioactivity Ci 3.27E-06 1.98E-06 D. Tritium 1. Total release Ci 5.54E+00 4.06E+00 5.00E+01 2. Average release rate for' period uCi/sec 7.05E-01 5.16E-01 3. Percent of Tech. Spec. limit (1) (1) Percent of Technical Specification limit will be provided in the Supplemental Effluent and Waste Di posal Report to be submitted per Technical Specification 6.7.C.I. All\\l3

TABLE IB Vermont Yankee Ef fluent and Waste Disposal Semiannual Report First and Second Ouarters 1994 Gaseous Effluents - Elevated Release Continuous Mode Batch Mode (l) Ouarter Quarter Quarter Qua rter Nuclides Released Unit 1 2 1 2 1. Fission Gases Krypton-85 Ci ND ND Krypton-85m Ci (2.99E+00 <3.06E+00 Krypton-87 Ci <2.21E+01 <2.23E+01 Krypton-88 Ci <1.08E+01 <l.12Et01 Xenon-133 Ci <1.40E+00 <1.41E+00 ~ ~ ~ ~ ~ ~ ~ Xenon-135 C1 <l.91E+01 <1.94E+01 ~~'~~~ Xenon-135m Ci <l.47E+02 <l.44E+02 Xenon-138 Ci (5.74E+02 <5.85E+02 Unidentified Ci ~ Total for period Ci <7.77E+02 <7.86E+02 2. Iodines Iodine-131 Ci 9.00E-04 8.91E-04 lodine-133 Ci 4.33E-03 4.57E-03 l Iodine-135 Ci ND ND Total for period Ci 5.23E-03 5.46E-03 3. Particulates Strontium-89 Ci 3.94E-04 3.48E-04 5trontium-90 Ci 6.71E-06 1.08E-05 Cesium 134 Ci ND ND Cesium-137 Ci 2.27E-06 1.21E-05 Barium-Lanthanum-140 Ci 5.60E-04 7.91E 05 Manganese-54 Ci 3.98E-06 ND Chromium-51 Ci ND NO Cobalt-58 Ci ND ND Cobalt-60 Ci 1.61E-05 4.40E-05 Cerium-141 Ci ND ND i Zinc-65 Ci ND ND Unidentified Ci Total for period Ci 9.83E-04 4.94E-04 (1) There were no batch mode gaseous releases for this reporting period. ND Not detected at the plant stack. RI2\\70

TABLE 1C Vermont Yankee Effluent and Waste Disposal Semiannual Report First and Second Quarters. 1994 Gaseous Effluents - Ground Level Releases III Continuous Mode Batch Mode Ouarter Quarter Quarter Quarter Nuclides Released Unit 1 2 1 2 1. Fission Gases Ci Krypton-85 Ci K ry pton -85m Ci Krypton-87 Ci Krypton-88 Ci Xenon-133 Ci Xenon-135 Ci Xenon-135m Ci Xenon-138 Ci Unidentified Ci Total for period Ci 2. Iodines Iodine-131 Ci Iodine-133 Ci Iodine-135 Ci Total for period Ci 3. Particulates Strontium-89 Ci Strontium-90 Ci Cesium-134 Ci Cesium-137 Ci Barium-Lanthanum-140 Ci i Manganese-54 Ci Chromium-51 Ci Tobalt-58 Ci j Cobalt-60 Ci l C e ri um-141 Ci Zinc-65 Ci Unidentified Ci Total for period Ci l (1) There were no ground level releases of gaseous effluents for this reporting period, l Rituo l

l l TABLE 10 Vermont Yankee Effluent and Waste Disposal Semiannual Report First and Second Ouarters, 1994 Gaseous Effluents - Nonroutine Releases") Continuous Mode Batch Mode Quarter Quarter Quarter Quarter Nuclides Released Unit 1 2 1 2 1. Fission Gases Ci Krypton-85 Ci Krypton 85m Ci Krypton-87 Ci Krypton-88 Ci Xenon-133 Ci Xenon-135 Ci Xenon-135m Ci Xenon-138 Ci Unidentified Ci Total for period Ci 2. Iodines Iodine-131 Ci Iodine-133 Ci Iodine-135 Ci Total for period Ci '3. Particulates Strontium 89 Ci Strontium-90 Ci Cesium-134 Ci Cesium-137 Ci Barium Lanthanum-140 Ci 1 Manganese-54 Ci Chromium-51 Ci Cobalt-58 Ci Cobalt-60 Ci Cerium-141 Ci Zinc-65 Ci l Unidentified Ci Total for period Ci (1) There were no nonroutine gaseous releases for this reporting period. i navo I l l

P TABLE 2A Vermont Yankee, Effluent and Waste Disposal Semiannual Report First and Second Quarters. 1994 Liquid Effluents - Summation of All Releases There were no liquid releases during the first or second quarters of 1994, t 1 i r b l (' r t 5 i 412\\10 t i

TABLE ?B Vermont Yankee Effluent and Waste Disposal Semiannual Report First and Second Quarters,1994 Liauid Effluents - Noncoutine Releases There were no nonroutine or accidental releases during the first or second-quarters of 1994 l 4 v 4 T a i I l l Alh70

TABLE 3 i Vermont Yankee Effluent and Waste Disposal Semiannual Report First and Second Ouarters, 1994 Solid Waste and Irradiated Fuel Shipments A. Solid Waste Shipped Off-Site for Burial or Disposal (Not Irradiated Fuel) 6-Month Est. Total Unit Period Error. % 1. Type of Waste a. Spent resins, filter sludges, evaporator m 4.35E+01 17.50E+01 3 bottoms etc. Ci 1.12E+02 b. Dry compressible waste, contaminated m 1.84E+01 i7.50E+01 3 equipment. etc. Ci 4.52E+01 Irradia'ed components. control rods. m t 3 etc. Ci 2. Estimate of Major Nuclide Composition (By Type of Waste): a. Zinc-65 2.75E+01 Cesium-137 2.16E+01 Cobalt 60 1.85E+01 Iron-55 7.59E+00 Cesium-134 7.45E+00 Manganese-54 % 7.39E+00 Nickel-63 4.13E+00 b. Iron-55 7.98E+01 Cobalt-60 9.04E+00 j Manganese-54 % 7.27E+00 3. Solid Waste Disposition. Number of Shipments Mode of Transportation Destination 9 Resin Shipments Truck Barnwell, SC 48 Partial Shipments From Truck Barnwell, SC Processor to Burial B. Irradiated Fuel Shipments (Disposition): None C. Supplemental Information -1) Class of solid waste container shipped: 110A (Unstable), 78

2) Types of containers used:

9 Type A, 108 Strong-Tight Container 3) Solidification ? gent or absorbent: None niac -- ^-

APPENDIX A EFFLUENT AND WASTE DISPOSAL SEMIANNUAL REPORT Supplemental Information First and-Second Quarters 1994 Facility: Vermont Yankee Nuclear Power Station Licensee: Vermont Yankee Nuclear Power Corporation 1A. Technical Specification Limits - Dose and Dose Rate Technical Specification and Category limit a. Noble Gases 3.8.E.1 Total body dose rate 500 mrem /yr 3.8.E.1 Skin dose rate 3000 mrem /yr 3.8 F.1 Gamma air dose 5 mrad in a quarter 3.8.F.1 Gamma air dose 10 mrad in a year 3.8 F.1 Beta air dose 10 mrad in a quarter 3.8.F.1 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 3.8.E.1 Organ dose rate 1500 mrem /yr 3.8.G.1 Organ dose 7.5 mrem in a quarter 3.8.G.1 Organ dose 15 mrem in a year i c. tiauids 3.8.B.1 Total body dose 1.5 mrem in a quarter 3.8.B.1 Total body dose 3 mrem in a year 3.8.B.1 Organ dose 5 mrem in a quarter 3.8.B.1 Organ dose 10 mrem in a year 4 2A. Technical Specification Limits - Concentration ( i Technical Specification and Catecory Limit l a. Noble Gases No MPC limits (No ECL Limits) b. lodine-131. lodine-133. Tritium and Radionuclides in Particulate Form With Half Lives Greater Than 8 Days No MPC limits (No ECL Limits) anoo A-1

c. Liauids 3.8.A.1 Total fraction of MPC (ECL) excluding noble gases (10CFR20, Appendix B. Table II, Column 2): 11.0 3.8.A.1 Total noble gas concentration: 12.00E-04 uC1/cc 3. Average Energy Provided below are the average energy ( E ) of the radionuclide mixture in releases of fission and activation gases, if applicable, i a. Average gamma energy: 1st Quarter 8.30E-01 MeV/ dis 2nd Quarter 8.30E-01 MeV/ dis 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 gross Noble Gas radioactivity released from the plant stack. Total Noble Gas release rates are calculated using this monitor. To determine the isotopic breakdown of the release, samples are taken of the Steam Jet' Air Ejector, which is the source gas.for the releases. These samples are analyzed by gamma spectroscopy to determine the isotopic composition. The isotopic composition is then proportioned to the gross releases determined from the stack monitor to quantify the individual isotopic releases. These are indicated in Table 18 and the totals of Table 1A. s The error involved in these steps may be approximately 1100 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 150 percent. A12\\10 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 ISO percent. d. Liquid Effluents Radioactive liquid effluents released from the f acility are continuously monitored. Measurements are also made 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 (uCi/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 effluent released is anal.yzed 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 propurtional composite sample, comprising an aliquot of each batch released during a quarter, is analyzed for Sr-89, Sr-90, and Fe-55. There were no liquid releases during the reporting period. 5. Batch Releases a. Liquid There were no routine liquid batch releases during the reporting period. b. Gaseous There were no routine gaseous batch releases during the reporting period. 6. Abnormal Releases a. Liquid There were no nonroutine liquid releases during the reporting

period, b.

Gaseous There were no nonrcutine gaseous releases during the reporting period. uzuo A-3 1 r

APPENDIX B LIQUID HOLDUP TANKS Requirement: Technical Specification 3.8,0.1 limits the-quantity of radioactive material contained in any outside tank. With the I quantity of radioactive material in any outside tank exceeding the limits of Technical Specification 3.8.D.1, a description of the events leading to this condition is required in the next Semiannual Effluent Release Report per Technical Specification 6.7.C.l.

Response

The limits of Technical Specification 3.8.D.1 were not' exceeded during this reporting period. l nzuo B-1

k APPENDIX C RADI0 ACTIVE LIOUID EFFLUENT MONITORING INSTRUMENTATION Requirement: Radioactive liquid effluent monitoring instrumentation channels are required to be operable in accordance with Technical . Specification Table 3.9.1. 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 9.1, an explanation in the next Semiannual Effluent Release Report of the reason (s) for delay in correcting the inoperability are required per Technical Specification 6.7.C.I.

Response

Since the requirements of Technical Specification Table 3.9.1 governing the operability of radioactive liquid effluent i monitoring instrumentation were met for this reporting period. no response is required. l Y I 1 I i J Ruvo C-1 ? l

t J APPENDIX D RADI0 ACTIVE GASE0US EFFLUENT MONITORING INSTRUMENTATION Reauirement: Radioactive gaseous effluent monitoring instrumentation channels are required to be operable in accordance with Technical Specification Table 3.9.2. If inoperable gaseous effluent I monitoring instrumentation is not returned to operable status within 30 days pursuant to Note 5 of Table 3.9.2, an explanation in the next Semiannual Effluent Release Report of the reason (s) for the delay in correcting the inoperability is required per Technical Specification 6.7.C.I.

Response

Since the requirements of Technical Specification Table 3.9.2 governing the operability of radioactive gaseous effluent monitoring instrumentation were met for this reporting period. no response is required. h t l l O f i L i i nuva D-1

APPENDIX E. RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM Recuirement: The Radiological Environmental Monitoring Program is conducted in accordance with Technical Specification 3.9.C. With milk samples no longer available from one or more of the sample locations required by Technical Specification Table 3.9.3. Technical Specification 6.7.C.1 requires the following to be included in the next Semiannual 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 samplea. and (3) include revised ODCM figure (s) and table (s) reflecting the new location (s).

Response

No changes were needed in the milk sampling locations specified in Technical Specification Table 3.9.3 due to sample unavailability. Riisio E1

APPENDIX F LAND USE CENSUS Recuirement: A land use census is conducted in accordance with Technical Specification 3.9,D, With a land use census identifying a location (s) which yields at least a 20 percent greater dose or dose commitment than the values currently being calculated in Technical Specification 4.8.G.1, Technical Specification 6.7.C.1 requires the identification of the new location (s) in the next Semiannual Effluent Release Report, i

Response

The 1994 land use census was not performed during this reporting period. It will be completed during the next reporting period l (second half of 1994). 4 i l i j i 7 t l l l i i i. i 4 i i i mue F-1 l l ,--.m,-,. .w,.,,., ,,,.-,,-,-,,.<n-n, .-.c-, ,n--.,-

i APPENDIX G PROCESS CONTROL PROGRAM Recuirement: Technical Specification 6.12. A.1 requires that licensee initiated changes to the Process Control Program (PCP) be submitted to the Commission in the Semiannual Radioactive Effluent Release Report for the period in which the change (s) was made.

Response

The licensee-initiated changes to the Process Control Program in this reporting period are: L 1, The title, the Radiation Protection Supervisor, was changed to the Radiation Protection Manager. 2. I.1 the Introduction Section, compliance with 10CFR20 was added. 3. In Section 2.0, the spent resin liners were replaced by casks for storing highly radioactive filters. 4. In Section 3.0, the Bird centrifuge for dewatering of depleted resins was replaced by the RDS-1000 Radioactive Waste Dewatering System. In addition, Class A resins that exceed 1.0 Ci/cc isotopes with greater than 5 year half-lives will be disposed of in an approved High Integrity Container (HIC). The RDS-1000 System has been tested by i Chem-Nuclear for certification in meeting the Barnwell Site i criteria and disposal requirement for freestanding liquid. This change of the dewatering system did not reduce the overall conformance of the dewatered spent resins / filter media waste product to existing criteria for solid waste shipments and disposal. 5. In Section 5.0, the use of a box compactor for compacting DAW was discontinued. 6. In Section 11.0, the Procedure OP-2527 Sampling and Analysis for Radioactive Classification, was added for determining 10CFR61 scaling factors. 7. The title for AP-0504 procedure was changed to " Shipment of Radioactive Materials." 8. In the list of procedures for the PCP implementation. AP-021, Maintenance Requests, was replaced by OP-2527, j Sampling and Analysis for Radwaste Classification. The revised Process Control Program is attached. n:aa G-1

l VERMONT YANKEE NUCLEAR POWER CORPORATION PROCESS CONTROL PROGRAM REV 3 1/15/94 i t Submitted [ 't REdiation Protsetion Manager Approved /w?" ash 2 94 cd-PORCJ Approved ' M/ M vu la Manager l Approved o l Manager of OperaTons

VERMONT YANKEE NUCLEAR POWER CORPORATION PROCESS CONTROL PROGRAM

== Introduction:== The Vermont Yankee Nuclear Power Corporation Process Control Program (PCP) describes the administrative and technical controls on the radioactive waste systems which provide assurance that Vermont Yankee meets federal shipping and burial site requirements. The PCP complies with Technical Specification 6.12 by describing process parameters, controls, tests, sampling and analysis to ensure compliance with 10 CFR 20,10 CFR 71, and 10 CFR 61 Energy,49 CFR 172-173 Transportation, state, and burial site regulation requirements. 1.0 Solidification Vermont Yankee Nuclear Power Corporation does not routinely solidify liquid waste. If the use of solidificadon to dispose of any liquid waste is required, it will be done by an outside vendor under the vendor's PCP. This PCP will be reviewed and approved by the Plant Health Physicist and the Radiation Protection Manager prior to implementation. This review is to identify that there is sufficient supponing documentation of the vendor's PCP to give assurance that the final product will meet all requirements for transport and burial, and that sufficient procedural controls exist to assure safe operations. 2.0 Cartridae Filter Elements Low activity cartridge filter elements will be air dried and handled as dry active waste. Filters that are too radioactive to be disposed ofin this manner will be placed in casks and shipped for disposal. 3.0 Resins Normal operations produce radioactive waste in the form of depleted resins. These resins are processed in the burial container using a rapid dewatering system (RDS-1000) manufactured by Chem-Nuclear Systems, Inc. Procedures provide detailed instrucdons for remote vacuum dewatering and air drying of these resins. The system has been tested, by Chem-Nuclear, for certification in meeting the Barnwell Site Criteria and disposal requirement for free standing liquid, These tests are described in Chem-Nuclear's Topical Report on the RDS-1000 Radioactive Waste Dewatering Syctem. In addition, to comply with the statement, "Any liquids present in waste packages shall be non-corrosive with respect to the l container", Vermont Yankee tested the pH of various resin mixtures used by the plant in solution with water. The range was found to be 4.2 - 8.4. A solution is not considered corrosive if the pH is greater than 4.0. ] 1

A resin sample is taken from each liner prior to shipment. The sample is counted to determine the activity and waste classification. Class A resins that exceed 1.0 pCi/cc ofisotopes with greater than 5 year half-lives and all Class B and C resins will be disposed ofin an approved High Integrity Container (HIC). Vendor supplied or temporary methods of processing resins may be used in lieu of the above process provided that the vendor or temporary process meets the requirements of quality described above and does not conflict with accepted burial criteria or safety requirements. Such methods will be reviewed and approved by the Plant Health Physicist and the Radiadon Protection Manager prior to implementation. 4.0 Filter Liners During refueling outages and normal operation, liquid radwaste processing may require use of a decanting filter on the condensate phase separators. A floating sucton is used to decant the water and resin into a filter liner. Filtered water is pumped from the liner. When use of the lineris completed, a vacuum pump is attached to dewater the resin in the liner. The liner is dewatered for a minimum of 48 hours and until no more water is viewed from the pump discharge. A resin sample is taken from the liner and counted to determine the activity and waste classification. 5.0 Dry Active Waste fDAW) DAW is compacted, as practcal, or shipped to a vendor that sorts the material for processing or recycling. All DAW is examined before being compacted or shipped. Any liquids or items found that would compromise the integrity of the package are removed and separated as specified by procedure. Containers used for DAW shipments meet the criteria of 49 CFR 173.425a. or b. "No leakage of radioachve material", as specified in 49 CFR 173.425.b.1 will be met provided that no radioactive materials in quantties equal to or exceeding those specified in 49 CFR 173.443 are detected on the external surfaces of the package at any dme during shipment. 6.0 Chelatina Aaents In order to comply with 10 CFR 20 Appendix F, chelating agents are controlled by the plant chemistry department using procedure AP 0620. 7.0 Explosive Waste No waste capable of detonation or of explosive decomposidon or reaction will be disposed as per 10 CFR 61.56(a)(4). 8.0 Toxic Waste No waste capable of generating toxic gases, vapors, or fumes will be disposed as per 10 CFR 61.56(a)(5). 2

9.0 PYroDhoric Waste No waste thatis pyrophoric will be disposed as per 10 CFR 61.56(a)(6). 10.0 Hiah Intearity Containers fHICs) Vermont Yankee Nuclear Power Corporation has contracted with various suppliers of approved HICS. South Carolina has approved PCPs for HICs used by Vermont Yankee. Any HIC Vermont Yankee may choose to use at some future time, will meet all applicable requirements. 11.0 Waste class determination Vermont Yankee periodically performs laboiatory analysis on all waste streams to determine the activity of radionuclides listed in Tables 1 and 2 of 10 CFR 61. Correlation analysis show that the relative concentration of each radionuclide, with respect to the overall activity in a given Vermont Yankee waste stream, remains constant over time. A set of scaling factors is determined which allows the activity of 10 CFR 61 radionuclides to be estimated using the results of gamma spectrometric analysis or direct gamma dose rate measurements. l For resin wastes, analysis is performed on samples of each source of resin compnsmg the contents of a burial container. Scaling factors are applied to the activity of radionuclides identified by gamma spectrometry analysis to determine the activity of those radionuclides which are not detected in the l gamma spectrurn. l For DAW, dose rate-to-curie conversion calculations are performed to determine the total activity present in a container. Scaling factors are applied to the container's total curie content to determine the activity ofindividual radionuclides. Specific procedures for determining 10 CFR 61 scaling factors are contained in OP 2527 Sampling and Analysis for Radwaste Classification. Once the activity of each radionuclide in a burial container is estimated, the waste classification is derived using methods required by 10 CFR 61. Specific procedures for waste class determination are contained in AP 0504, Shipment of Radioactive Material. 3

PROCEDURES WHICH IMPLEMENT THE PCP 1. AP 0504 Shipment of Radioactive Materials 2. OP 2511 Radwaste Cask, Drum, and Box Handling 3. OP 2527 Sampling and Analysis for Radwaste Classification 4. OP 2151 Liquid Radwaste 5. OP 2153 Solid Radwaste 6. AP 0620 Chemical Material Control i \\ l

APPENDI) H OFF-SITE DOSE CALCULATION MANUAL Requirement: Technical Specification 6.13.A.1 requires that licensee initiated changes to the Off-Site Dose Calculation Manual (ODCM) be submitted to the Commission in the Semiannual Radioactive Effluent Release Report for the period in which the change (s) was made effective.

Response

The ODCM was amended to include new meteorological dispersion modeling that takes into account changes to both plant design and environmental air flow affects. The first change incorporates a new estim3te of the plant Stack exit flow rate (175,000 cfm) resulting from the Turbine Building roof ventilation reroute to the plant stack, with the addition of two new 25,000 cfm fans to the system. It was assumed that the normal flow profile would have one of the two new fans in service, adding 25,000 cfm to the previous nominal plant stack flow of 150,000 cfm. This increases the effective height of the plant stack, and therefore, provides additional credit for air dispersion of effluents. The second modification in the meteorological model is the inclusion of site 'pecific recirculation correction factors. These factors provide adjustments to the straight-line steady-state dispersion models, and accounts for the potential return of effluents released from the plant to move back over the site area based on historical wind shift patterns. This adds additional conservatism to the existing dispersion estimates for both the plant stack and ground level (North Warehouse) release points. The recalculation of the dispersion estimates also utilized the most recent available five year site meteorological data, 1988 through 1992. The effect of the revised dispersion modeling causes all the gaseous pathway dose conversion factors to be changed since they have factored into them X/0 values from the dispersion modeling. This is reflected in Sections 1, 3, 5 and example calculations in Appendix A. The same methodology for deriving dose factors that have been used and approved in the past have been reapplied with the addition of the new X/O's. A second issue addressed in the amendment is in response to an internal audit finding that indicated that a footnote should be added to Table 4.1 for TLD monitoring location DR-8, explaining how the measured values are averaged with other TLD data. This amendment provides an explanation of how the requirements in Technical Specification Table 3.9.3 for an inner ring direct radiation monitoring are satisfied with DR-8, but that its measured value is averaged as a site boundary TLD due to its unvo H-1 \\

= close proximity to the plant. No change in the existing program is made by this additional editorial clarification. Section 6. including figures 6.1 and 6.2, have'also been amended to clarify that the system description and simplified one line drawing of effluent release pathways only address pathways that are, or likely to contain, radioactive materials discharged from the site. This section is not intended or required to reflect all types of site release, non-radioactive pathways. No change in the technical description or use of the radioactive system is affected by this editorial clarification. The revised ODCM pages for the above revisions (Amendment No.

17) are attached.

aizoo H-2

.t VERMONT YANKEE NUCLEAR POWER STATION OFF-SITE DOSE CALCULATION MANUAL REVISION 17 Reviewed: @.rylc[.//8 M 94 f/rf/44 Plant Operations (Aeview Committee Date Approved: /MW f/b6/9 V Pla anager [/ g ' Da'te N 7/'[h Approved: ~3 LA Vice President. Operations 'Da te R12\\84

THIS PAGE LEFT BLANK INTENTIONALLY .' l l I Revision 17 Date 5/31/94 uns. -ii-

LIST OF AFFECTED PAGES Page Revision Date i 17 05/31/94 11 13 05/01/92 iii 17 05/31/94 iv 0 03/01/84 v-x 17 05/31/94 1.1-1.2 15 07/08/93 1.3-1.4 17 05/31/94 1.5-1.6 15 07/08/93 1.7-1.9 17 05/31/94 1.10 15 07/08/93 1.11 17 05/31/94 1.12 16 10/28/93 1.13-1.18 15 07/08/93 1.19-1.20 17 05/31/94 1.21 15 07/08/93 1.22 17 05/31/94 2.1-2.4 15 07/08/93 3.1 15 07/08/93 3.2-3.7 17 05/31/94 3.8-3.12 15 07/08/93 3.13 3.17 17 05/31/94 3.18 16 10/28/93 3.19 15 07/08/93 3.20-3.50 17 05/31/94 3.51 15 07/08/93 3.52-3.58 17 05/31/94 l 3.59 15 07/08/93 j 4.1-4.lb 17 05/31/94 l 3 4.2 16 10/28/93 4.2a-4.3 17 05/31/94 4.4 10 04/04/91 4.5 16 10/28/93 l 4.6-4.9 10 04/04/91 5.1-5.6 15 07/08/93 5.7-5.19 17 05/31/94 5.20-5.21 15 07/08/93 6.1-6.10 17 05/31/94 R-1 17 05/31/94 A.1-A.2 15 07/08/93 A.3-A.24 17 05/31/94 l B.1-C.39 9 03/02/90 D.1-0.3 16 10/28/93 I 1 Revision 17 Date 05/31/94 -iii-a m a4 ) 'l .j

l DISCLAIMER OF RESPONSIBILITY This document was prepared by Yankee Atomic Electric Company on behalf of Vermont Yankee Nuclear Power Corporation. This document is believed to be completely true and accurate to the best of our knowledge and information. It is authorized for use specifically by Yankee Atomic Electric Company, Vermont Yankee Nuclear Power Corporation, and/or the appropriate subdivisions within the Nuclear Regulatory Commission only. With regard to any unauthorized use whatsoever, Yankee Atomic Electric Company, Vermont Yankee Nuclear Power Corporation and their officers, directors, agents and employees assume no liability nor make any warranty or representation with respect to the contents of this document or to its accuracy or completeness. l l l 1 uns4 -iv-

ABSTRACT The VYNPS ODCM (Vermont Yankee Nuclear Power Station Off-Site Dose i Calculation Manual) contains the 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. j i With initial approval by the U.S. Nuclear Regulatory Commission and the VYNPS Plant Management and approval of subsequent revisions by the Plant Management (as per the Technical Specifications)this ODCM is suitable to show compliance where referred to be the Plant Technical Specifications. Revision 17 Date 5/31/94 uns4 _y, 4

TABLE OF CONTENTS Page LIST OF AFFECTED PAGES iii DISCLAIMER OF RESPONSIBILITY .................... iv ABSTRACT .............................. v TABLE OF CONTENTS.......................... vi LIST OF TABLES ................ viii LIST OF FIGURES........................... x

1.0 INTRODUCTION

1-1 1.1 Summary of Methods. Dose Factors. Limits. Constants. Variables and Definitions.................. 1-2 2.0 METHOD TO CALCULATE OFF-SITE LIQUID CONCENTRATIONS 2-1 ENG NG 2.1 Method to Determine F and C 2-1 i y 2.2 Method to Determine Radionuclide Concentration for Each Liquid Effluent Pathway.. 2-3 2.2.1 Sampl e Tanks Pa thways................. 2-3 2.2.2 Service Water Pathway. 2-3 2.2.3 Circulating Water Pathway. 2-4 3.0 0FF-SITE DOSE CALCULATION METHODS................. 3-1 3.1 Introductory Concepts.................... 3-2 3.2 Method to Calculate the Total Body Dose from Liquid Releases 3-4 3 3.3 Method to Calculate Maximum Organ Dose from Liquid Releases. 3-10 3.4 Method to Calculate the Total Body Dose Rate From Noble Gases........................... 3-13 3.5 Method to Calculate the Skin Dose Rate from Noble Gases.. 3-19 3.6 Method to Calculate the Critical Organ Dose Rate from Iodines. Tritium and Particulates with T Greater Than 1/2 8 Days 3-26 l Revision 17 Date 5/31/04 l uns4 -vi-i

3.7 Method to Calculate the Gamma Air Dose from Noble Gases.. 3-30 3.8 Method to Calculate the Beta Air Dose from Noble Gases 3-34 i 3.9 Method to Calculate the Critical Organ Dose from Iodines, Tritium and Particulates 3-38 3.10 Receptor Points and Annual Average Atmospheric Dispersion Factors for Important Exposure Pathways.......... 3-45 3.11 Method to Calculate Direct Dose From Plant Operation . 3-51 3.12 Cumulative Doses 3-59 4.0 ENVIRONMENTAL MONITORING PROGRAM 4 - 1. 5.0 .SETPOINT DETERMINATIONS...................... 5-1 5.1 Liquid Effluent Instrumentation Setpoints.......... 5-2 5-9 5.2 Gaseous Effluent Instrumentation Setpoints 6.0 LIQUID AND GASE0US EFFLUENT STREAMS, RADIATION MONITORS. AND RADWASTE TREATMENT SYSTEMS 6-1 6.1 In-Plant Radioactive Liquid Effluent Pathways........ 6-1 6.2 In-Plant Radioactive Gaseous Effluent Pathways 6-4 REFERENCES R-1 APPENDIX A: Method I Example Calculations A-1 APPENDIX B: Approval of Criteria for Disposal of Slightly Contaminated Septic Waste On-Site at Vermont Yankee B-1 APPENDIX C: Response to NRC/EG&G Evaluation of ODCM Update Through Revision 4 C-1 APPENDIX D: Assessment of Surveillance Criteria for Gas Releases from Waste Oil Incineration 0-1 l i Revision 17 Date 5/31/94 -vii-emu l

i LIST OF TABLES Number Title Page 1.1-1 Summary of Radiological Effluent Technical Specifications and Implementing Equations 1-3 i 1.1-2 Summary of Methods to Calculate Unrestricted Area Liquid Concentrations 1-6 1.1-3 Summary of Methods to Calculate Off-Site Doses from Liquid Concentrations 1-7 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 Determinations 1-11 1.1-8 Summary of Variables 1-12 1.1-9 (Table Deleted) 1.1-10 Dose Factors Specific for Vermont Yankee for Noble Gas Releases 1-19 1.1-10A Combined Skin Dose Factors Specific for Vermont Yankee Ground Level Noble Gas Releases 1-20 1.1-11 Dose Factors Specific for Vermont Yankee for Liquid Releases 1-21 1.1-12 Dose and Dose Rate Factors Specific for Vermont Yankee for Iodines. Tritium, and Particulate Releases 1-22 3.2-1 Environmental Parameters for Liquid Effluents at Vermont Yankee '3-8 3.2-2 Usage Factors for Various Liquid Pathways at Vermont Yankee 3-9 3.9-1 Environmental Parameters for Gaseous Effluents at Vermont Yankee 3-42 3.9-2 Usage Factors for Various Gaseous Pathways at Vermont Yankee 3-44 3.10-1 Atmospheric Dispersion Factors 3-48 Revision 17 Date 5/31/94 -viti-urm

LIST OF TABLES (Continued) Number Title Page 3.10-2 Site Boundary Distances 3-49 3.10-3 Recirculation Correction Factors 3-50 4.1 Radiological Environmental Monitoring Stations 4-2 5.2-1 Relative Fractions of Core Inventory Noble Gases After Shutdown 5-20 l l 4 1 ) i Revision 17 Date 5/31 /94 -ix-ama.

LIST OF FIGURES t Number Title Page 4-1 Environmental Sampling Locations in Close Proximity to Plant 4-4 4-2 Environmental Sampling Locations Within 5 km of Plant 4-5 4-3 Environmental Sampling Locations Greater Than 5 km from Plant 4-6 4-4 TLD Locations in Close Proximity to Plant 4-7 4-5 TLD Locations Within 5 km of Plant 4-8 4-6 TLD Locations Greater Than 5 km from Plant 4-9 6-1 Liquid Effluent Streams, Radiation Monitors, and Radwaste Treatment System at Vermont Yankee 69 6-2 Gaseous Effluent Streams, Radiation Monitors, and Radwaste Treatment System at Vermont Yankee 6-10 Revision 17 Date 5/31 /94 une4 -x-

1.0 INTRODUCTION

This 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 Vankee Technical Specifications 3.8/4.8 and 3.9/4.9, hereafter referred to as the Radiological Effluent Technical Specifications. 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 Technical Specification 3.8.M) 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 Technical Specifications. The administration of the program for the disposal of slightly contaminated septic waste, as described in Appendix B, is the responsibility of the Senior Environmental Program Manager. All changes to the ODCM must be reviewed by PORC and approved by M00, in accordance with Technical Specification 6.13, prior to implementation. All approved changes shall be submitted to the NRC for their information in the Semiannual Radioactive Effluent 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 15 Date 07/08/93 1-1 une4

1.1 Summary of Methods. Dose Factors. Limits. Constants. Variables and Definitions This section summarizes the methods for the user. The first-time user should read Chapters 2 through 6. The concentration and setpoint methods are documented 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 Il for the appropriate dose as described in Sections 3.2,through 3.9 and 3.11. 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. The variables and special definitions used in this ODCM are in Tables 1.1-8 and 1.1-9. 1 l l Revision 15 Date 07/08/93 une4 1-2 1

TABLE 1.1-1 Summary of Radiological Effluent Technical SDecifications and implementino Ecuations Technical fl3 Specification Category Method Limit 3.8.A.1 Liquid Effluent Sum of the Fractions of Eq. 2-1 s 1.0 Concentration Effluent Concentration Limits [ Excluding Noble Gases] Total Noble Gas Eq. 2-2 s 2 x 10 'd pCi/cc Concentration 3.8.B.1 Liquid Effluent Dose Total Body Dose Eq. 3-1 s 1.5 mrem in a qtr. s 3.0 mrem in a yr. Organ Dose Eq. 3-3 s 5 mrem in a qtr. s 10 mrem in a yr. 3.8.C.1 Liquid Radwaste Total Body Dose Eq. 3-1 s0.06 mrem in a mo. Treatment Operability Organ Dose Eq. 3-3 s 0.2 mrem in a mo. 3.8,E.1 Gaseous Effluents Dose Total Body Dose Rate from Eq. 3-5 s 500 mrem /yr. Rate Noble Gases Eq. 3-39 l Skin Dose Rate from Noble Eq. 3-7 s 3000 mrem /yr. Gases Eq. 3-38 l 3.8.E.1 (Continued) Organ Dose Rate from Eq. 3-16 s 1500 mrem /yr. 1 Iodines. Tritium and Eq. 3-40 g I Particulates with T1/2 > 8 Days Revision 17 Date 5/31 /94 i 1-3 usu.

TABLE 1.1-1 (Continued) Summary of Radiological Effluent Technical Specifications and Implementing Ecuations Technical III Specification Category Method Limit 3.8.F.1 Gaseous Effluents Dose Gamma' Air Dose from Noble Eq. 3-21 s 5 mrad in a qtr. I from Noble Gases Gases Eq. 3-41 s 10 mrad in a yr. Beta Air Dose from Noble Eq. 3-23 s 10 mrad in a qtr. l Gases Eq. 3-43 l s 20 mrad in a yr. 3.8.G.1 Gaseous. Effluents Dose Organ Dose from I-131 Eq. 3-25 s 7.5 mrem in a qtr. from Iodines. Tritium. 1-133. Tritium, and Eq. 3-44 l and Particulates Particulates with T1/2 > 8 s 15 mrem in a yr. Days 3.8.1.1 Ventilation Exhaust Organ Dose Eq. 3-25 s 0.3 mrem in a mo. Treatment 3.8.M.1 Total Dose (from All Total Body Dose Footnote (2) s 25 mrem in a yr. Sources) Organ Dose s 25 mrem in a yr. Thyroid Dose s 75 mrem in a yr. Revision 17 Date 5/31/94 a m e. 1-4

TABLE 1.1-1 (Continued) Summary of Radiological Effluent Technical Specifications and Implementing Ecuations Technical Specification Category Method"3 Limit 3.9.A.1 Liquid Effluent Monitor Setpoint Liquid Radwaste Alarm Setpoint Eq. 5-1 T.S. 3.8.A.1 Discharge Monitor 3.9.B.1 Gaseous Effluant Monitor Setpoint Plant Stack and A0G Alarm / Trip Setpoint for Eq. 5-9 T.S. 3.8.E.la Of fgas System Noble Gas Total Body Dose Rate (Total Body) Activity Monitors Alarm / Trip Setpoint for Eq. 5-10 T.S. 3.8.E.la Skin Dose Rate (Skin) SJAE Noble Gas Activity Alarm Setpoint Eq. 5-21 T.S. 3.8.K.1 Monitors (1) More accurate methods may be available (see subsequent chapters). (2) Technical Specification 3.8.M.2 requires this evaluation only if twice the limit of Equations 3-1, 3-3, 3-21, 3-23 or 3-25 is reached. If this occurs a Method II calculation using actual meteorology and identified pathways for a real individual. shall be made. Revision 15 Date 07/08/93 1-5 one.

I y s TABLE 1.1-2 Summary of Methods to Calculate Unrestricted Area Liauid Concentrations Equation Reference Number Category Equation Section ENG C 2.1 2-1 Sum of the Fractions of 1 - }[ ECL s1 Combined Effluent F i g t Concentrations in Liquids [Except Noble Gases] r Ci ' NG 2.1 l-2-2 Total Activity of Dissolved us p C ~ i and Entrained Noble Gases i ml from all Station Sources 52E-04 Revision 15 Date 07/08/93 1-6 uns4 l

t TABLE 1.1-3 Summary of Methods to Calculate Off-Site Doses from Liouid Concentrations Equation Reference Number Category Equation Section 3-1 Total Body Dose Dtb (mrem) - { 03 DFLith i 3.3.1 l 3-3 Maximum Organ Dose D,o (mrem)- E Q DFlimo i 1 l i Revision 17 Date 5/31 /04 u na. 1-7

TABLE 1.1-4 Summary of Methods to Calculate Dose Rates Equation Reference Number Category Equation Section 3-5 Total Body Dose Rate ' mrem' from Noble Gases gbs -0.61 E 0,57 DFB l t i ' yp i Released from Stack 3-39 Total Body Dose Rate rmrem' 3.4.1 A = 6.4 0,GL DFB l from Noble Gases tbg i Released from Ground

yr 3-7 Skin Dose Rate from

' mrem' 3.5.1 0,sT m, k Noble Gases Released skins from Stack ' yr 3-38 Skin Dose Rate from ' mrem' g GL DF's Noble Gases Released sking i from Ground ' yp 3-16 Critical Organ Dose ' mrem' 3.6.1-b ~ E DsTP Of0'ico Rate from Stack cos i s Release of I-131, ' yr 1-133, Tri tium, and Particulates with T 1/2 >B Days 3-40 Critical Organ Dose ' mrem' 3.6.1 GLP k -Ei DFG Rate from Ground Level cos 91co Release of I-131 ' yr I-133, Tritium, and Particulates with T1/2 >8 Days f i Revision 17 Date 5/31/94 unu l-8

TABLE 1.1-5 Summary of Methods to Calculate Doses to Air from Noble Gases Equation Reference Number Category Equation Section 3.7.1 3-21 Gamma Dose to Air D[jrs (mrad) = 0.019 E OfI DF[ l from Noble Gases j Released from Stack I 0 r 3.7.1 3-41 Gamma Dose to Air airo (mrad) - 0.20 E. 0 ' DF ) D 1 1 from Noble Gases Released from Ground Level 3-23 Beta Dose to Air from 0 '0*1 D,0trs (mrad) = 0.033 E OfI DF l t Noble Gases Released from Stack 3-43 Beta Dose to Air from Dairg (mrad) - 1.12 E 0 DFp 3.8.1 p ct 4 3 l Noble Gases Released from Ground Level l Revision 17 Date 5/31 /94 e m e4 1-9 l

TABLE 1.1-6 Summary of Methods to Calculate Dose to an Individual from Tritium. Iodine, and Particulates in I Gas Releases and Direct Radiation Equation Reference Number Category Equation Section 3-25 Dose to Critical Organ STP 3.9.1 l Dco, (mrem) - Og DFG35co from Stack Release of I-131. 1-133 Tritium. and Particulates 3-44 Dose to Critical Organ cog (mrem) - E, OfLP D DFG9$co from Ground Level Release of I-131. I-133. Tritium, and Particulates Direct Dose 3-27 Turbine Building De (mrem) - Knis(L). E 3.11.1 North Warehouse 3-29 Shielded End Ds - 0.25 x Rs 3-30 Unshielded End Du - 0.53 x Ru 3.11.2 LLW Storace Pad 3-31 Direct Line (Module Dde - 0.28 x A X f d d Short Side Out) 3-32 Direct Line (Module 3.H.3 D s - 0.39 x R X fd d d Long Side Out) 3-33 Skyshine (Resin DsxR - 0.016 x AsKR X IsK Liners) 3-34 Skyshine (DAW) DsKD - 0.015 x Asgo xT sK 3-35 Resin Liner Transfer 3.11.3 DTran - 0.0025 x ATran xTTran (Unshielded) 1 3-36 Intermodular Gap Dose Dcop - 2.44E-2 x Wc., x Agt x fc., 3.11.3 Revision 15 Date 07/08/93 a m a4 1-10 i

l TABLE 1.1-7 j Summarv of Methods for i Setpoint Determinations Reference Equation Number Category Equation Section 4 5-1 Liquid Effluents: Liquid Radwaste L DF 5.1.1.1 Espt (cps) - I "I Discharge Monitor DF"i" (17/350) Gaseous Effluents: j Plant Stack (RR-108 1A, I RR-108-1B) and A0G Offgas System (3127. 3128) Noble Gas Activity Monitors 1 1 5.2.1.1 l 5-9 Total Body R,tbpt (cpm) - 818 S8 7 DFBc R,'p$" (cpm) - 3000 S o DF c 1 5.2.2.1 5-21 SJAE Noble Gas Activity R,sJAE (mR/hr) - 1.6E+05 S pg 97 Monitors (17/150A, 17/1508) i 1 i Revision 17 Date 5/31/94 1-11 a m s4

. ~. TABLE 1.1-8 Summary of Variables Variable Definition Units Agt Total gamma activity contained in a resin Ci liner in storage directly in line with a gap between adjacent storage modules. Concent' ration at point of discharge to an Ci/ml Ns C 11 unrestricted area of dissolved and entrained noble gas "i" in liquid pathways from all station sources. Total activity of all dissolved and entrained Ci Ns C 1 noble gases in liquid pathways from all ml station sources. Cdf Concentration of radionuclide "i" at the point Ci of liquid discharge to an unrestricted area. ml C Concentration of radionuclide "i". Ci i cc C, Concentration, exclusive of noble gases, of Ci p radionuclide "f" from tank "p" at point of ml discharge to an unrestricted area. C,i Concentration of radionuclide "i" in mixture pCi at the monitor. m) p Beta dose to air from stack release, mrad Dairs p Beta dose to air from ground level release. mrad De rg Oy Gamma dose to air from stack release. mrad airs Gamma dose to air from ground level release. mrad y Derg D Dose to critical organ from stack release. mrem co, D Dose to the critical organ from ground level mrem eo, l release. l Do Direct dose (Turbine Building). mrem i ( Revision 16 Date 10/28/93 unsi 1-12 L...

TABLE 1.1-8 (Continued) Summary of Variables Variable Definition Units Dose rate at 3' from unobstructed side of mrem g storage module facing site boundary. hr Direct dose at site boundary per unobstructed mrem D dt storage module (short end). y r-modul e Direct dose at site boundary per unobstructed mrem D ds storage module (long side). yr-module Gamma dose to air. corrected for finite cloud. mrad r O finite Og,p Intermodular gap dose projected to the maximum mrem site boundary location from resin waste not yr directly shielded by DAW modules. Dose the maximum organ. mrem D, s Dose to skin from beta and gamma. mrem D Dose rate at 1 meter from source in shielded mrem g, end of North Warehouse, hr Annual dose at site boundary from fixed mrem D s sources in shielded end of North Warehouse. yr Maximum dose rate at 3' over top of DAW in a mrem gsKD storage module. hr I gsKR Maximum dose rate at 3' over top of each resin mrem liner in a Storage module. hr D xD Skyshine dose at the site boundary from DAW in mrem s storage modules (unobstructed top surfaces). y r-modul e Skyshine dose at the site boundary from resin mrem D 3xg liners in storage modules (unobstructed top y r-l i ne r surfaces). I l i Revision 15 Date 07/08/93 i 1-13 l l ams4 )

TABLE 1.1-8 (Continued) Summary of Variables Variable Definition Units Dose to the total body. mrem D tb The direct dose conversion factor for N-16 K mrem N/s scatter from the turbine hall to Location (L) MW,h gI'*" Dose rate at contact from the unshielded top rad surface of resin liner, hr Dyr,n Dose at the site boundary from unshielded meem movement of resin liner between transfer cask and storage module. Au Dose rate at 1 meter from source in unshielded mrem end of North Warehouse. hr The annual dose at the site boundary from mrem D u fixed sources in the unshielded end of North yr Warehouse. Dilution factor. ratio DF Minimum allowable dilution factor. ratio DF min Composite skin dose factor, mrem-sec DF* pCi-yr Total body gamma dose factor for nuclide "i". 3

DFB, mrem-m pCi-y r Composite total body dose factor.

3 DFB c mrem-m pCi-yr Site-specific, total body dose factor for a mrem DFl itb liquid release of nuclide "1". Ci DFLj, Site-specific, maximum organ dose factor for a mrem liquid release of nuclide "i". Ci Revision 15 Date 07/08/91 uns4 1-14

TABLE 1.1-8 (Continued) Summary of Variables Variable Definition Units D FG,j e, Site-specific, critical organ dose factor for mrem a stack gaseous release of nuclide "i". Ci Site-specific critical organ dose rate factor mrem-sec DFG51co for a stack gaseous release of nuclide "i". C i -y r Site-specific, critical organ dose factor for mrem DFG oico a ground level gaseous release of nuclide "i". Ci l Site-specific, critical organ dose rate factor mrem-sec DFGaico for a ground level gaseous release of nuclide C i -y r l .i. l Beta skin dose factor for nuclide "i". 3

DFS, mrem-m pC i -y r Combined skin dose f actor for nuclide "i" from mrem-sec DF15 a stack release.

Ci -y r Combined skin dose f actor for nuclide "i" from mrem-sec j DFis a ground level release. pCi-y r l amma air dose factor for nuclide "i". 3 ~ DF[ mrad-m pC i-yr Beta air dose factor for nuclide "i". mrad-m 3 D F, pC i-y r Critical organ dose rate due to iodines and mrem g* 5 particulates released from stack. yr Critical organ dose rate due to iodines and mrem g*8 particulates released from ground. yr Skin dose rate due to stack release of noble mrem g, gases. yr g, Skin dose rate due to ground release of noble mrem gases. yr Revision 15 Date 07/08/93 1-15 ons4

.- = TABLE 1.1-8 (Continued) Summary of Variables Variable Definition Units gtbs Total body dose rate due to noble gases from mrem stack release. yr gtbg Total body dose rate due to noble gases from mrem ground level release. yr D/0 Deposition factor for dry deposition of I elemental radiciodines and other particulates. -{y E Gross electric output over the period of HW,h interest. fd Fraction of a year that a storage module is in fraction use with an unobstructed side oriented toward west site boundary. fc, Fraction of a year that the intermodular gap fraction is not shielded. fsK Fraction of a year that a storage module is in fraction use with an unobstructed top surface. Fd Flow rate out of discharge canal. gpm F, Flow rate past liquid radwaste monitor. gpm Flow rate past gaseous radwaste monitor, F cc sac Sum of the fractions of combined effluent fraction ENG F 1 concentrations in liquid pathways (excluding noble gases).

ECL, Annual average effluent concentration limit Ci for radionuclide "i" (10CFR20.1001-20.2401, cc Appendix 8. Table 2 Column 2) 0, Release for radionuclide "i" from the point of curies interest.

l i 1 Revision 15 Date 07/08/93 l a m s4 1-16

i i TABLE 1.1-8 (Continued) Summary of Variables l Variable Definition Units Release rate for radionuclide "i" at the point Ci g of interest. sec The noble gas radionuclide "i" release rate at Ci sT gi the plant stack. .sec The noble gas radionuclide "i" release rate Ci GL 4 i from ground level. sec The noble gas radionuclide "i" release rate at Ci sJAE gi the steam jet air ejector. sec The noble gas radionuclide "i" release rate at Ci A0G gi the exhaust of the augmented Off-Gas System sec The iodine, tritium, and particulate Ci sTP gi radionuclide "i" release rate from the plant sec stack. The iodine, tritium, and particulate Ci GLP gi radionuclide "i" release rate from ground sec level. The release of noble gas radionuclide "i" from curies sT O i the plant stack. The release of noble gas radionuclide "i" from curies GL Oi ground level. The release of iodine, tritium, and curies sTP 0 1 particulate radionuclide "i" from the plant stack. The release of iodine, tritium, and curies GLP 0 1 particulate radionuclide "i" from ground level. Liquid monitor response for the limiting cps L Rspt Concentration at the point of discharge. I Revision 15 Date 07/08/93 1-17 mme

TABLE 1.1-8 (Continued) Summary of Variables Variable Definition Units Respo..se of the noble gas monitor at the cpm skin Rspt limiting skin dose rate. g Response of the noble gas monitor to limiting cpm tb spt t' body dose rate. S Sh. ig factor. Ratio p i S' Detector counting efficiency from the most cpm mR/hrI recent gas monitor calibration. Ci/cc Ci/ccl Sgt Detector count'$ efficiency for noble gas cpm mR/hri "i"- pCi/cc pCi/cc: 5 3 Detector counting efficiency from the most cps recent liquid monitor calibration. pC1/ml SH Detector counting efficiency for radionuclide cps "I"- pCi/ml Tiran Time that an

iielded resin liner is exposed hours in the storat, ad area.

W,p Intermodule gap width between adjacent DAW inches g storage modules which shield resin liner storage modules from the west site boundary. X/0 Annual or long-term average undepleted 3 sec atmospheric dispersion factor for stack ,3 release. X/0 Annual or long-term average undepleted g sec atmospheric dispersion factor for ground level ,3 release. D/03 Effective annual or long-term overage gamma r sec 5 atmospheric dispersion factor. ,3 gfg)y Effective annual or long-term average gamma sec 9 atmospheric dispersion factor for a ground level release. ,3 Revision 15 Date 07/08/93 a m a4 1-18

TABLE 1.1-10 Dose Factors Specific for Vermont Vankee for Noble Gas Releases Gamma Combined Skin Total Body Beta Skin Dose Factor Beta Air Gamma Air Radionuclide Dose Factor Dose Factor (Stack Release) Dose Factor Dose Factor O

DFB, DFS, DF\\

DF DF[ i i 3 3 3 3 ' mrem-sec' mrad-m mrad-m mrem-m mrem-m pC i -y r, pCi-yr, pCi-yr, pCi-yr, pCi-yr, ( Ar-41 8.84E-03* 2.69E-03 9.12E-03 3.28E-03 9.30E-03 Kr-83m 7.56E-08 1.31E-05 2.88E-04 1.93E-05 Kr-85m 1.17E-03 1.46E-03 2.35E-03 1.97E-03 1.23E-03 Kr-85 1.61E-05 1.34E-03 1.41E-03 1.95E-03 1.72E-05 Kr-87 5.92E-03 9.73E-03 1.43E-02 1.03E-02 6.17E-03 Kr-88 1.47E-02 2.37E-03 1.28E-02 2.93E-03 1.52E-02 Kr-89 1.66E-02 1.01E-02 2.23E-02 1.06E-02 1.73E-02 Kr-90 1.56E-02 7.29E-03 1.87E-02 7.83E-03 1.63E-02 Xe-131m 9.15E-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 3.06E-04 5.58E-04 1.05E-03 3.53E-04 Xe-135m 3.12E-03 7.11E-04 3.02E-03 7.39E-04' 3.36E-03 Xe-135 1.81E-03 1.86E-03 3.24E-03 2.46E-03 1.92E-03 Xe-137 1.42E-03 1.22E-02 1.37E-02 1.27E-02 1.51E-03 Xe-138 8.83E-03 4.13E-03 1.06E-02 4.75E-03 9.21E-03

8.84E 8.84 x 10-3 Revision _12_ Date 5/31/04 1-19 nne.

i i TABLE 1.1-10A Combined Skin Oose Factors Specific for Vermont Yankee Ground level Noble Gas Releases DF' mrem-s ec ' pCi -y r Radionuclide s AR-41 1.61E-01 KR-83M 1.38E-04 KR-85M 6.02E-02 KR-85 4.73E-02 t KR-87 3.86E-01 KR-88 1.92E-01 I KR-89 4.79E-01 KR-90 3.73E-01 t XE-131M 1.79E-02 f XE-133M 3.72E-02 XE-133 1.33E-02 XE-135M 4.90E-02 R XE-135 7.92E-02 XE-137 4.40E-01 XE-138 2.11E-01 1 Revision 17 Date 5/31/94' uru. 1-20

TABLE 1.1-11 Dose Factors Specific for Vermont Yankee for Liocid Releases Total Body Maximum Organ Dose Factor Dose Factor mrem' DFl 9 D F Lj,o ith bl C1 Radionuclide ( s L H-3 2.06E-04 2.06E-04 Na-24 3.38E-02 3.38E-02 Cr-51 3.10E-04 6.96E-02 Mn-54 2.08E-01 3.00E+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+00 Zn-65 8.06E+00 1.64E+01 Sr-89 2.55E-01 8.91E+00 Sr 90 4.23E+01 1.67E+02 Zr-95 4.21E-04 1.36E-01 Mo-99 4.79E-03 4.51E-02 Tc-99m 5.04E-06 2.33E-04 Ag-110m 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 1-132 3.10E-06 1.29E-04 I-133 3.31E-03 1.63E+00 1-135 3.16E-04 5.90E-02 Cs-134 1.28E+02 1.60E+02 Cs-137 7.58E+01 1.21E+02 Ba-140 4.08E-03 9.72E-02 Ce-141 2.31E-05 4.10E-02 W-187 1.18E-02 8.90E+00 Revision 15 Date 07/08/93 t m e4 1-21

TABLE 1.1-12 Dose and Dose Rate Factors Soecific for Vermont Yankee for Iodines Tritium, and Particulate Releases Stack Release Ground Level Release

Critical Organ Critical Organ Critical Organ Critical Organ Dose Factor Dose Rate Factor Dose Factor Dose Rate Factor mrem' mrem-s ec '

rem' Radio-

DFGste, DFG,ico OfG 0FG oico Ci 1

9 co Ci yr-Ci yr-pC1 nuclide ( 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 1.43E+00 Mn-54 7.01E-01 2.77E+01 4.71E+00 1.84E+02 Fe-55 3.17E-01 1.00E+01 2.05E+00 6.47E+01 Fe-59 6.99E-01 2.32E+01 4.60E+00 1.52E+02 Co-58 3.62E-01 1.30E+01 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 Sr-89 1.14E+01 3.60E+02 7.27E+01 2.29E+03 Sr-90 4.31E+02 1.36E+04 2.82E+03 8.89E+04 Zr-95 6.91E-01 2.28E+01 4.51E+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 I-131 7.71E+01 2.43E+03 5.02E+02 1.58E+04 I-133 8.21E-01 2.59E+01 8.29E+00 2.61E+02 Cs-134 1.58E+01 5.27E+02 1.02E+02 3.37E+03 Cs-137 1.63E+01 5.55E+02 1.04E+02 3.53E+03 Ba-140 1.13E-01 3.66E+00 2.18E+00 6.94E+01 Ce-141 1.70E-01 5.42E+00 1.19E+00 3.78E+01 Ce-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, A0G, and CAB). Revision 17 Date 5/31/94 unu l-22

3.0 0FF-SITE DOSE CALCULATION METHODS Chapter 3 provides the basis for plant procedures required to meet the 10CFR50, Appendix I, ALARA dose objectives, and the 40CFR190 total dose limits i to members of the public in unrestricted areas, as stated in the Radiological Effluent Technical Specifications (hereafter called RETS). A simple, conservative method (called Method I) is listed in Tables 1.1-2 to 1.1-7 for l each of the reqLirements of the RETS. Each of the Method I equations is presented, along with their bases in Sections 3.2 through 3.9 and Section 3.11. Appendix A provides example calculations for all Method I dose equations as guidance to their use. In addition, reference is provided to more sophisticated methods (called Method II) for use when more accurate results are needed. This chapter provides the methods, data, and reference l material with which the operator can calculate the needed doses and dose l rates. Setpoint methods for effluent monitor alarms are described in Chapter 5. 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). i l Revision 15 Date 07/08/93 3-1 a m s4

3.1 introductory Concepts The Radiological Effluent Technical Specifications (RETS) 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 stcpped. 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 external exposures, such as to noble gas clouds, refer only to the doses received during the actual time period of exposure to the radioactivity released from the plant. Once the source of the radioactivity is removed, there is no longer any additional accumulation to the dose commitment. Gaseous effluents from the plant are also controlled such that the maximum " dose rates" at the site boundary at any time are limited to 500 mrem / year. This instantaneous dose rate limit allows for operational flexibility when off normal occurrences may temporarily increase gaseous effluent release rates from the plant, while still providing controls to ensure that licensees meet the dose objectives of Appendix I to 10CFR50. l 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 Technical Specification 3.6.E.1.a. does not necessarily, by itself, constitute a Licensee Event Report (LER) under 10CFR 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 - 20.2401 Table 2 Column 1 (four-hour notification per 10CFR50.72, and 30-day LER per 10CFR50.73). The quantities D and R are introduced to provide calculable quantities, related to off-site dose, or dose rate which demonstrates compliance with the RETS. The dose D is the quantity calculated by the Chapter 3 dose equations. The D calculated by " Method I" equations is not necessarily the actual dose received by a real individual but usually provides an upper bound for a given release because of the conservative margin built into the dose factors and the Revision 17 Date s /31/o4 uns4 3-2

l

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. 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. R is the quantity calculated in the Chapter 3 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 3, and are summarized in Tables 1.1-1 to 1.1-7. Each method has two levels of complexity and conservative margin and are called Method I and Method II. Method I has the greatest margin and is the simplest: generally a linear equation. Method II is a more detailed analysis which allows for use of sita-specific factors and variable parameters to be selected to best fit the actual release. Guidance is provided but the appropriate margin and depth of analysis are determined in each instance at the time of analysis ur. der Method 11. The plant has both elevated and ground level gaseous release points: the main vent stack (elevated release), and the North Warehouse waste oil burner (ground level release). Therefore, total dose calculations for skin, whole body, and the critical organ from gaseous releases will be the sum of the elevated and ground level doses. Appendix 0 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 Technical Specifications. Revision 17 Date 5/31/94 3-3

nnu

A4MA4 .a e A A-6 D -44 J-EM-4 4-tA.Ja.M.en e34-A E M'*M 4 'u4.3.M J1*4M M3 8-+-.a 4-4 4 - *4-94 khhJ .4-hA. -AM aammAu.--aw A A_._wAan.4-. .a-m i I 1 l t ? h F h i l f b 3 I f \\ l i, h I i-I F J p h I l 6 m 1 P +-- m w- - ~ r-r m-w ,v,,<e .-me

m-

--a--~ ~ ~. +-~-. --

t 3.2 Method to Calculate the Total Body Dose from Liouid Releases Technical Specification 3.8.8.1 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. Technical Specification 3.8.C.1 requires liquid radwaste treatment when the total body dose estimate exceeds 0.06 mrem in any month. Technical Specification 3.8.M.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 11 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 (Specification 3.8.C.1). To evaluate the total body dose, use Equation 3.1 to estimate the dose from the planned release and add this to the total body dose accumulated from prior releases during the month. 3.2.1 Method I The increment in total body dose from a liquid release is: (3-1) Otb - E 04 0Flitb 1 mrem' (mrem) (Ci) C1 where: Site specific total body dose factor (mrem /Ci) for a liquid DFl itb release. See Table 1.1 11. Revision 17 Date 5/31 /94 3-4 nmu 1 I

1 1 0, Total activity (Ci) released for radionuclide "i". (For strontiums and Fe-55, use the most recent measurement available.) Equation 3-1 can be applied under the following conditions (otherwise, justify Method I or consider Method II): 1. Normal operations (not emergency event). I 2. Liquid releases were to the Connecticut River, and I 3. Any continuous or batch release over any time period. 3.2.2 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 II. Method I may be used to show that the Technical Specifications which limit off-site total body dose from liquids (3.8.B.1 and 3.8.C.1) have been ~ met for releases over the appropriate periods. Technical Specification 3.8.8.1 is based on the ALARA design objectives in 10CFR50, Appendix I Subsection II A. Technical Specification 3.8.C.1 is an " appropriate fraction", determined by the NRC, of that design objective (hereafter called the Objective). Technical Specification 3.8.M.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" (10CFR50, 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 basedDon a Method II analysis for the critical receptor with maximum exposure Revision 17 Date 5/31/94 nwu 35* 1

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 DFlitb, mrem /Cl) 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 3 deposited by the water pathway. A plant discharge flow rate of 44.6 f t /sec was used with a mixing ratio of 0.0346 which corresponds to a minimum regulated river flow of 1250 cfs at the Vernon Dam just below the plant discharge outfall.* Tables 3.2-1 and 3.2-2 outline human consumption and environmental parameters used in the analysis. The resulting, site-specific, total body dose factors appear in Table 1.1-11. For any liquid release, during any period, the increment in annual average total body dose from radionuclide "i" is: ADtb " Og DFlitb (3-2) mrem' (mrem) (Ci) Ci ( where: Site-specific total body dose factor (mrem /Ci) for a liquid DFl itb release. See Table 1.1.11. ^l Total activity (Ci) released from radionuclide "i". 0,

An Mp equal to 1.0 for the fish pathway is assumed between the discharge structure and the dam.

Revision 17 Date 5/31/94 una4 3-6

Method I is conservative because it is based on dose factors DFlitb which were chosen from the base case to be the highest of the four age groups for each radionuclide, as well as assuming minimum river dilution flow.

3.2.3 Method If 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, 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. Analyses requiring Method Il calculations should be referred to YNSD to be performed and documented. 1 1 Revision 17 Date s/r /o4 uru4 3-7

j l

TABLE 3.2-1 Environmental Parameters for Liould Effluents at Vermont Yankee (Derived from Reference A) FOOD GROWN WITH CONTAMINATED WATER POTABLE AQUATIC SHORELINE VARIABLE WATER FOOD ACTIVITY VEGETABLES LEAFY VEG. MEAT COW MILK MP Mixing Ratio 1.0 0.0356 0.0356 0.0356 0.0356 0.0356 TP Transit Time (HRS) 24.0 0.0 0.0 0.0 480.0 48.0 l 2 YV Agricultural (KG/M ) 2.0 2.0 2.0 2.0 Productivity 2 P Soil Surface (KG/M ) 240.0 240.0 240.0 240.0 Density 2 IRR I.'rigation Rate (L/M /HR) 0.152 0.152 0.152 0.152 TE Crop Exposure (HRS) 1440.0 1440.0 1440.0 1440.0 Time TH Holdup Time (L/D) 1440.0 24.0 2160.0 2160.0 OAW Water Uptake Rate (KG/0) 50.0 60.0 for Animal 0F Feed Uptake Rate 50.0 50.0 for Animal FI Fraction of Year Crops 0.5 0.5 0.5 0.5 Irrigated Location of Connecticut River Below Vernon Dam Critical Receptor Revision 15 Date 07/08/93 3-8 H258'

TABLE 3.2-2 Usage Factors for Various Licuid 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 (KG/YR) (KG/YR) (LITER /YR) (KG/YR) (KG/YR) (KG/YR) (LITER /YR) (HR/YR) Adul t 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 l 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 l Revision 15 Date 07/08/93 3-9 nnes

3.3 Method to Calculate Maximum Organ Dose from Liauid Releases Technical Specification 3.8.B.1 limits the maximum organ dose commitment to a Member of the Public from radioactive material in liquid effl' ents to u 5 mrem per quarter and 10 mrem per year. Technical Specification 3.8.C.1 requires liquid radwaste treatment when the maximum organ dose estimate exceeds 0.2 mrem in any month. Technical Specification 3.8.M.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 limit), or if Method I cannot be applied. If the radwaste system is not operating, the maximum organ dose must be estimated prior to a release (Specification 3.8.C.1). To evaluate the maximum organ dose, use Equation 3-3 to estimate the dose from the planned release and add this to the maximum organ dose accumulated from prior releases during the month. 3.3.1 Method 1 The increment in maximum organ dose from a liquid release is: 0,o - E 0 D F L,o (3-3) 4 i mrem' (mrem (Ci) Ci ( Revision 15 Date 07/08/93 3-10 mns. -

where: DFL,o i Site-specific maximum organ dose factor (mrem /Ci) for a liquid release. See Table 1.1-11. Total activity (Ci) released for radionuclide "1". (for Og strontiums and Fe-55, use the most recent measurement available.) Equation 3-3 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. 3.3.2 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 3.2.2). 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 (DFl,,) for that radionuclide. i For any liquid release, during any period, the increment in annual average dose from radionuclide "i" to the maximum organ is: AD,o - Oi DFLima (3-4) (mrem) (Ci) < l' s Revision 15 Date 07/08/93 uns4 3-11

) i where: i Site-specific maximum organ dose f actor (mrem /Ci) for a D F L,,o liquid release. See Table 1.1-11. Total activity (C1) released for radionuclide "i". 0, 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. 3.3.3 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 11 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. Analyses requiring Method II calculations should be referred to YNSD to be performed and documented. d Revision 15 Date 07/08/93 3-12 uns4

3.4 Method to Calculate the Total Body Dose Rate From Noble Gases Technical Specification 3.8.E.1 limits the instantaneous dose rate at any time to the total body from all release sources of noble gases at any location at or beyond the site boundary equal to or less than 500 mrem / year. Use Method I first to calculate the Total Body Dose Rate from the peak release rate via both elevated and ground level release points. The dose rate limit of Technical Specification 3.8.E.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 Technical Specification limit (i.e.. use of actual meteorology over the period of interest) to determine if, in fact. Technical Specification 3.8.E.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 alarm setpoint is based on a value which corresponds to the off-site dose rate limit of Technical Specification 3.8.E.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 Technical Specifications (3.8.E.1) are performed when the effluent monitor alarm setpoint is exceeded and the corrective action required by Specification 3.8.E.2 is unsuccessful, or as required by the notations to Technical Specification Table 3.9.2 when the stack noble gas monitor is inoperable. 3.4.1 Method I 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 3-5 (an I example calculation is provided in Appendix A): i l Revision 17 Date 5/31/94 nns4 3-13 l

(3-5) l 0.61 E 0 si

0FB, k

tbs 1 / 5 r 3 3 ' mrem' pCi-sec 'pCi ' rarem-m yr Ci-m

sec, pCi-yr,

3 s where: dfT In the case of noble gases, the release rate from the plant stack (pCi/sec) for each radionuclide, "i", identified. The release rate at the plant stack is based on the measured radionuclide distribution in the off-gas at the Steam Jet Air Ejector (SJAE) during plant operation when the activity at the stack is below detectable levels, and the recorded total gas effluent count rate from the Stack Gas Monitor I or II. The release rate at the stack can also be stated as follows: o,sJAE i bsi ~ i sdAE Sg (3-28) 1 uCi 'pCi / cc ' (cc) (cpm) sec cpm sec i Plant Stack Gas Monitor I or II count rate (cpm). M = S, Appropriate or conservative plant stack monitor detector counting efficiency for the given nuclide mix (cpm /( Ci/cc)). Stack flow rate (cc/sec). F 6fJAE The last measured release rate at the steam jet air ejector of noble gas i (pCi/sec).

OFB, Total body gamma dose factor (see Table 1.1-10).

1 Revision 17 Date 5/31/94 3-14 man

For ground level noble gas releases, the total body dose rate is calculated as follows: Atbg - 6.4 E Of DFB (3-39) i t f r 3 3 pCi-sec ' Ci ' mrem-m 3 pCi-m, (s ec,, pC i -y r, where: 6f Ground level release rate ( Ci/sec) of noble gas. The total body dose rate for the site is equal to Rtbs + ktbg-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 6,JAE to the sum of all @g in Equation 3-28 above is assumed to f JAE reduce to a value of 1, and the total body gamma dose factor DFB, for Xe-133 3 (2.94 E-04 mrem-m /pCi-yr) is used in Equation 3-5. Alternately, a relative radionuclide "i" mix fraction (f ) may be taken from Table 5.2-1 as a s function of time after shutdown, and substituted in place of the ratio of dJAE i to the sum of all @,JAE in Equation (3-28) above to determine the relative fraction of each noble gas potentially available for release to the total (example calculations can be found in Appendix A). 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 i b 1 Revision 17 Date 5/31/94 l amu 3-15 i

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 3-5 and 3-39 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. 3.4.2 Basis for Method I Method I may be used to show that the Technical Specification which limits total body dose rate from noble gases released to the atmosphere (Technical Specification 3.8.E.1) has been met for the peak noble gas release rate. Method I for stack releases was derived from Regulatory Guide 1.109 as l follows: Atds - 1E+06 Sr [X/0][ }) 0si

DFB, (3-6) i f

5 r 3 3 ' mrem','pCi'(p) sec Ci ' mrem-m (sec, ( pCi-yr, yr, Ci, 3 m where: Shielding factor - 1.0 for dose rate determination. S F [X/0][ Maximum annual average gamma atmospheric dispersion factor for stack (elevated) releases; - 6.11E-07 (sec/m ). l 3 Revision 17 Date 5/31/94 3-16 amu

l dst Release rate from the plant stack of noble gas "i" ( Ci/sec). 3 mrem-m 0F8 i Gamma total body dose factor, See Table 1.1-10. pC i -y r, Equation 3-6 reduces to: ktbs - 0.61 E O sT DFBI (3-5) l i r e 3 3 ' mrem' pC i -s ec 'pC i ' mrem-m 3 1r C1-m

sec, p C 1 -y r,

r u For ground level releases, the ground level maximum long-term average gamma 3 atmospheric dispersion f actor - 6.42E-06 sec/m. thus leading to: Atbg - 1E+06 + 6.42E-06 E of' 0FB i or-Atbg - 6.4 { Of' DFB, 1 The selection of critical receptor, outlined in Section 3.10, 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 A0G, off-gas concentrations at the plant stack during routine plant operations are usually too low for determination of the radionuclide mix 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. Revision 17 Date 5/31/94 amu 3-17

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 11 cannot provide much extra realism because R and R are already based on several factors which make tbs tbg use of current plant parameters. However, should it be needed, the dose rate f analysis for critical receptor can be performed making use of current meteorology during the time interval of recorded peak release rate in place of the default atmospheric dispersion factor used in Method I. 3.4.3 Method I1 i 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. Analyses requiring Method II calculations should be referred to YNSD to be performed and documented. Revision 16 Date 10/28/93 3-18 mm

3.5 Method to Calculate the Skin Dose Rate from Noble Gases Technical Specification 3.8 E.1 limits the instantaneous dose rate at any time to the skin from all release sources of noble gases at any location at or beyond the site boundary to 3,000 mrem / year. l 1 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 Technical Specification 3.8.E.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, l l Use Method II if Method I predicts a dose rate greater than the Technical Specification limits (i.e., use of actual meteorology over the period of interest) to determine if, in fact. Technical Specification 3.8.E.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 alarm setpoint is based on a value which corresponds to the off-site Technical Specification 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 Technical Specifications (3.8.E.1) are performed when the effluent monitor alarm setpoint is exceeded and the corrective action required by Specification 3.8.E.2 is unsuccessful, or as required by the notations to Technical Specification Table 3.9.2 when the stack noble gas monitor is inoperable. 3.5.1 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 3-7 (an example calculation is provided in Appendix A): P P l Revision 15 Date 07/08/91_ neu4 3-19 1

D F's Askins " E 01 i (3-7) 1 ' mrem' 'pCi ' ' mrem-s e c ' yr

sec, pC i -y r ",

where: 6pT In the case of noble gases. the noble gas release rate from the plant stack ( Ci/sec) for each radionuclide. "i", identified. The release rate at the plant stack is based on the measured radionuclide distribution in the off-gas at the Steam Jet Air Ejector (SJAE) during plant operation when the activity at the stack is below detectable levels, and the recorded total gas effluent count rate from the Stack Gas Monitor I or II. The release rate at the stack can also be stated as follows: gsJAE b ~ sJAE (3-28) i i pCi (nCi/cc) (cc) (cpm) sec cpm sec Plant stack gas monitor I or 11 count rate (cpm). M Appropriate or conservative plant stack monitor 5 9 detector counting efficiency for the given nuclide mix (cpm /(pCi/cc)). Stack flow rate (cc/sec). F 6fJAE The last measured release rate at the steam jet air ejector of noble gas i ( Ci/sec). D F',, combined skin dose factor (see Table 1.1-10) for stack release. Revision 17 Date 5/31/94 3-20 ma.

For ground level releases, the skin dose rate from noble gases is calculated by Equation 3-38: Asking " E D U F 's (3 38) 1 i I where: 6" The noble gas release rate from ground level (pCi/sec) for each = radionuclide "i" identified. D F'i g Combined skin dose factor for a ground level release [see Table 1.1-10A]. The skin dose rate for the site is equal to $ skins + Nsking-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 6,JAE to the sum of all gi in Equation 3-28 above is assumed to JAE reduce to a value of 1, and the combined skin dose factor DF',3 for Xe-133 (5.58 E-04 mrem-sec/ Ci year) is used in Equation 3-7. Alternately, a l relative radionuclide "i" mix fraction (f ) may be taken from Table 5.2-1 as a i function of time after shutdown, and substituted in place of the ratio of each gi to the sum of all hg in Equation 3-28 above to determine the relative JAE JAE fraction of each noble gas potentially available for release to the total (example calculations can be found in Appendix A). 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 3-7 and 3-38 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, l l Revision 17 Date 5/31 /o4 u m. 3-21 j

i 3.5.2 Basis For Hetnod i The methods to calculate skin dose rate parallel the total body dose rate methods in Secti(n 3.4.3. Only the differences are presented here. Method I may De used to show that the Technical Specification which limits skin dose rate from noble gases released to the atmosphere (Technical Specification 3.8.E.1) has been met for the peak noble gas release rate. i Method I was derived from Regulatory Guide 1.109 as follows: Ofir + 3.17E+04 E 03 [X/0]s DFS (3-8) S 1.11 SF D i 1 3 mrem'. ' mrem' g) ' mrad' 'pC i -y r ' 'Ci ' sec mrem-m f yr

mrad, yr Ci-sec W,

3 pCi-yr, m j ( s where: Average ratio of tissue to at-absorption coefficients will 1.11 convert mrod in air to mres 7 tissue. D[ir - 3.17E+04 E Oi[X/0],DF[ (3-9) i f 3 e s 3 'm r a d ' 'pC i -y r ' 'C i ' sec mrad-m yr Ci -s e c, (W, , ( p C i -y r, 3 m s j i I D[ir [X/0][/[X/0], D[$ nit, now (3-10) f T r 3 3 ' mrad' ' mrad' sec m 3 yr yr m (sec, Revision 17 Date 5/31 /94 3-22 a m s4 I

= sT and 01 31.54 0 j (3-11) 'Ci ' 'C i -s e c ' 'pC i ' f, Ci-yr, ,sec, 1E+06 [X/0]f S of DFf (3-12) so Askins " l 11 SF 1 ~ r s 3 ' mrem' ' mrem ' g g) 'pci ' sec' fpci ' mrad-m , yr

mrad, Ci,

3 (sec, pCi-yr, m s a + 1E+06 X/0s [ Of DFS t 1 3 'p C i ' sec pCi mrem-m Ci, 3 sec pCi-yr, m ( substituting 3 [X/Q]I 6.11E-07 sec/m 3 3 1.04E-06 sec/m X/O s Shielding factor - 1.0 for dose rate determinations S p gives ST b DFf + 1.04 $0$ Askins - 0.68 E0g DFSi (3-13) i i l r 3 3' 'pci -s ec ' 'pC i ' mrem-m 3 ' mrem' 'pCi-sec-mrem 'pCi ' mrad-m 70 pCi-m -mrad sec, ( pCi -y r pCi-m (sec,, pCi-yr, 3 3 ( s s s j Revision 17 Date 5/31 /94 3-23 uns4

~ - E Of [0.68 DFf + 1.04 DFSj] (3-14) i define - 0.68 DFf + 1.04 DFSj (3-15) DF s then Askins - 0 DF (3-7) is ' mrem' 'nCi ' ' mrem-sec ' , yr

sec, C i -y r For determining combined skin doses for ground level releases, a

[X/0]g - 6.42E-06 sec/m and an undepleted X/0 - 3.52E-05 sec/m have been l 3 3 g substituted into Equation 3-12 to give: sking - E of' (7.13 DFf + 35.2 DFS ) A i i then DFg g - 7.13 DFj + 35.2 DFSj (3-37) sking - E Of' DFhg (3-38) and A 1 where: 6f - The noble gas release rate from ground level release points ( 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 3.10 is inherent in Method I, as it determined the maximum expected off-site atmospheric dispersion factors based on past long-term site-specific meteorology. Revision 17 Date 5/31/94 j 3-24 n m s4 1

I 4 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. 3.5.3 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. Analyses requiring Method II calculations should be referred to YNSD to be performed and documented. r i i t E l Revision 17 Date 5 / 31 /94 a m a. 3 j

~- 3.6 Method to Calculate the Critical Organ Dose Rate from lodines, Tritium Greater Than 8 Days and Particulates with T1/2 Technical Specification 3.8.E.1.b limits the dose rate to any organ, denoted R from all release sources of I-131.1-133. H-3. and radionuclides eg, in particulate form with half lives greater than 8 days to 1500 mrem / year to any o.gan. 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 Technical Specification 3.8.E.1.b is the total contribution from both ground and elevated releases occurring during the period of interest. Method I applies at all release rates. Use Method II if Method I predicts a dose rate greater than the Technical Specification limits (i.e., use of actual meteorology over the period of interest) to determine if, in f act. Technical Specification 3.8.E.1.b had actually been exceeded during the sampling period. i 3.6.1 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 3-16 (an example calculation is provided in Appendix A): STP ACOS " 23 D i sico DFG i i (3-16) ' mrem' 'pC1 ' ' mrem-s ec ' yr

sec,

( C i -y r, ( l 17 5/31/94 Revision Date 3-26

ntu'

where: 6pTP Stack activity release rate determination of radionuclide "i" (Iodine-131 Iodine-133, particulates with half-lives l greater than 8 days, and tritium), in Ci/sec. For i - Sr89, Sr90 or tritium, use the best estimates (such as most recent measurements). mrem-sec' DFG'sice Site specific critical organ dose rate factor pCi -y r, ror 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: P Acog, S g OfG (3-40) gico I where: 6fLP 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 Sr90, 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 0 for surveillance criteria on waste oil burning. Sitespecificcriticalorgandoseratefactor[(*#8*-3b OFG'gjto pCi-fr for a ground level gaseous release. See Table 1.1-12. The critical organ dose rate for the site is equal to Rcos + bcog-Equations 3-16 and 3-40 can be applied under the following conditions (otherwise, justify Method I or consider Method II): 1. Normal operations (not emergency event), and l 2. Tritium, iodine, and particulate releases via either elevated or grt,Jnd level Vents to the atmosphere. Revision 17 Date 5 / 31/ot. l cas4 3 P

l 3.6.2 Basis for Method i The methods to calculate critical organ dose rate parallel the total body dose rate methods in Section 3.4.3. Only the differences are Dresented here. Method I may be used to show that the Technical Specification which limits organ dose rate from iodines, tritium and radionuclides in particulate form with half lives greater than 8 days (hereafter called lodines and Particulates or "I+P") released to the atmosphere (Technical Specification 3.8.E.1.b) has been met for the peak I + P release rates. The equation for $cos and 5 is derived by modifying Equation 3-25 cog from Section 3.9 as follows: (3-17) bcos - E Oj DFG co i i mrem' (mrem) (C1) Ci 1 1 applying the conversion factor, 31.54 (Ci-sec/ Ci-yr) and converting 0 to 6 in pCi/sec as it applies to the plant stack yields: STP 31.54 EO DFGSico (3-18) A os c 1 ' mrem' 'Ci -s e c ' ' Ci ' ' mrem' yr, pC i -y r,

sec, Ci Equation 3.8 is written in the form:

STP 31.54 D DFGSiC0 (3-19) k ' mrem' 'C i -s e c ' ' Ci ' ' mrem' ( yr, pC i -y r, (sec, Ci Revision 17 Date 5/31/94 3-28 *

uns,

OFG'si co and'DFG (North Warehouse waste oil burner vent releases)- gico incorporates the conversion constant of 31.54 and has assumed that the shielding factor (S ) applied to the direct exposure pathway from p radionuclides deposited on the ground plane is equal to 1.0 in place of the S p value of 0.7 assumed in the determination of DFG and DFG for the sico gico 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 i ingestion of vegetables, meat, and milk) which is outlined in Section 3.10 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 Wareh'ouse 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. I 3.6.3 Method II If Method I cannot be applied, or if the Method I dose exceeds the limit, then Method II may be applied. Method 11 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. Analyses requiring Method II calculations should be referred to YNSD to be performed and documented. ? Revision 17 Date 5 /31 /94 annu 3-29 ) t

3.7 Method to Calculate the Gamma Air Dose from Noble Gases Technical Specification 3.8.F.1 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 Technical Specification 3.8.F.1 is the total contribution from both ground f and elevated releases occurring during the period of interest. j Use Method II if a more accurate calculation is needed. ) 3.7.1 Method 1 The gamma air dose from plant stack releases is: (3-21) l 5T DFf 0.019 E0 $ D airs 1 3' pCi-y r mradm (Ci) (mrad) C i --m pCi-yr, 3 r where: OjT total noble gas activity (Curies) released to the atmosphere via the plant stack of each radionuclide "i" during the period of interest. DF} gamma dose f actor to air for radionuclide "i". See Table 1.1-10 For ground level noble gas releases, the gamma air dose is calculated as follows: (3-41) l airg - 0.20 { Of DFf D 1 Revision 17 Date 5/31/94 3-30 R12\\84

where: Of - Total noble gas activity (curies) released to the atmosphere via ground level vents of each radionuclide, "i", during the pericJ of interest. The gamma air dose for the site is equal to D irs +0}$p9 Equations 3-21 and 3-41 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. 3.7.2 ftpis for Method I Method I may be used to show that the Technical Specification which limits off-site gamma air dose from gaseous effluents (3.8.F.1) has been met for releases over appropriate periods. This Technical Specification is based on the Objective in 10CFR50, Appendix I, Subsection 8.1, 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 B-4 and B-5 of Regulatory Guide 1.109 with the added assumption that Dfinite T - D [X/0]T [X/0]: / 2 airs - 3.17E+04 [X/0][ E 0 DFf D j (3-22) i e 3 3 pC i-y r ' 3 mrad-m (mrad) Ci-sec, (sec/m ) (Ci) yr-pCi, 8 Revision 17 Date 5/31/94 amu 3 where: [X/0][ maximum long term average gamma atmospheric dispersion factor for a stack release. 3 6.11E-07 (sec/m ) OgT number of curies of noble gas "i" released from the plant stack which leads to: (3-21) l E oft DFf 0.019 Dairs 1 PCi-yr ' mead M (Ci) (mrad) C i -m ( pC i -y r, 3 A / For the ground level release: airg - 3.17E+04 [X/0]g E Of' DFf (3-42) D i where: (X/0)g Maximum long-term average gamma atmospheric dispersion factor for a ground level release 3 l 6.42E-06 sec/m leading to: (3-41) l airg - 0.20 E Of DFf D 1 Revision _12. Date 5/31 /04 3-32 nuss4

I 1 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. The main difference between Method I and Method II is that Method II would allow the use of actual meteorology to determine [X/0]T rather than use the maximum long-term average value obtained for the years 1981 to 1985. 3.7.3 Method II If the Method I dose determination indicates that the Technical Specification limit may be exceeded, or if a more exact calculation is required, then Method Il 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. Analyses requiring Method II calculations should be referred to YNSD to be performed and documented. f I l Revision 17 Date 5/31/94 uns4 3-33 +

3.8 Method to Calculate the Beta Air Dose from Noble Gases i Technical Specification 3.8.F.1 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 Technical Specification 3.8.F.1 is the total contribution from both ground and elevated releases occurring during the period of interest. Use Method 11 if a more accurate calculation is needed or if Method I cannot be applied. 3.8.1 Method i The beta air dose from plant vent stack releases is: T DFf Dairs - 0.033 J) 0 $ (3-23) 1 3' e s pCi -y r mra d-m (Ci) (mrad) 3 pCi-yr, CiHn ( r where: DFf beta dose factor to air for radionuclide "i". See Table 1.1-10. 0}T total noble gas activity (Curies) released to the atmosphere via the plant stack of each radionuclide "i" during the period of interest. Revision 17 Date 5/31/94 nas4 3-34

For ground level noble gas releases. the beta air dose is calculated as follows: a i rg - 1.12 : Of' 0Ff D (3-43) where: Ofl 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 to Dairs + Dairg - Equations 3-23 and 3-43 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. 3.8.2 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 ir.troductory user's guide to Method II. The methods to calculate beta air dose parallel the gamma air dose methods in Section 3.7.3. Only the differences are presented here. Method I may be used to show that the Technical Specification which limits off-site beta air dose from gaseous effluents (3.8.A.1) has been met-for releases over appropriate periods. This Technical Specification is based on the Objective in 10CFR50, Appendix I. Subsection B.I. 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. t Revision 17 Date 5/31/94 amu 3-35

For any noble gas release, in any period, the dose is taken from Equations B-4 and B-5 of Regulatory Guide 1.109: 0 X/0s E oft DFf (3-24) D - 3.17 E 44 airs i pCi-yr ' 'sec ' mrad + 3' (Ci) (mrad) ,C i -s e c, ,3 ( pC i -y r, substituting Maximum long term average undepleted atmospheric dispersion X/0 s factor for a stack release. 3 l 1.04E-06 sec/m We have (3-23) l E oft DFf 0.033 D airs 1 mrad + 3' p i-y r (mrad) (Ci) 3 pCi-y r Ci-m j for the ground level release: airg - 3.17E44 (X/0)g E Of DFf (3-44) D i where: (X/0)g - Maximum long-term average undepleted atmospheric dispersion factor for a ground level release, 17 5/31/94 Revision Date 3-36 uns4

3.52E-05 sec/m l 3 leading to: (3-43) l ai rg - 1.12 { 0 DFf D i 1 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. 3.8.3 Method II i If Method I cannot be applied, or if the Method I dose determination indicates that the Technical Specification limit may be exceeded, or if a more exact calculation is required, then Method II may be applied. Method 11 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. Analyses requiring Method 11 calculations should be referred to YNSD to be performed and documented. Revision 17 Date 5/31/04 nas4 3-37

I l 3.9 Method to Calculate the Critical Organ Oose from Todines. Tritium and Particulates Technical Specification 3.8.G.1 limits the critical organ dose to a Member of the Public from all release sources of I-131, 1-133, Tritium, and particulates with half-lives greater than 8 days (hereaf ter 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 Technical Specification 3.8.G.1 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). 3.9.1 Method I STP Dcos - ][ O g DFGsico (3-25) i "I'* (mrem) (Ci) C1 ( 0}TP 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. i DFGsico - 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 17 Date 5/31/94 3-38

cann

The critical organ dose is calculated for ground level releases as follows: GLP Ocog -J O 0FGgico (3-44) g 1 mrem' (mrem) (Ci) Ci 0ILP - 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. OFGgico - Site-specific critical organ dose factor for a ground level release of nuclide "i" (mrem /C1). For each radionuclide it is the age group and organ with the largest dose factor. See Table 1.1-12. The critical organ dose for the site is equal to Dcos + Dcog - Equations 3-2; a'.d 3-44 can be applied under the following conditions (otherwise, justify Method I or consider Method II): 1. Normal operations (not emergency event), 2. I+P releases via the plant stack, Turbine Building, and waste oil burner (see Appendix 0 for surveillance criteria on waste oil burning), to the atmosphere, and 3. Any continuous or batch release over any time period. 3.9.2 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. Revision 17 Date 5/31/94 uns. 3-39 + )

Method I may be used to show that the Technical Specifications which limit off-site organ dose from gases (3.8.G.1) have 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 a single " critical receptor" provides part of the conservative margin to the calculation of critical organ 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 of the critical receptor for the annual average conditions. For purposes of complying with the Technical Specifications 3.8.G.2 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. 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 (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, C-4 and C-13 in Reference A). Tables 3.9-1 and 3.9-2 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 3.10. However, he is exposed, conservatively, to all pathways (see Section 3.10). 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 - Oj DFG (3-26) ico where DFG is the critical dose f actor for radionuclide "i" and Og is the ico activity of radionuclide "i" released in curies. Method I is more conservative than Method II in the region of the Technical Specification limits because it is based on the following reduction 17 5/31/94 Revision Date 3-40 " neu4

l i of the base case. The dose factors DFG used in Method I were chosen from ico 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. 3.9.3 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 11 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. Analyses requiring Method 11 calculations should be referred to YNSD to be performed and documented. W Revision 17 Date 5/31/94 mm4 3-41 * - + - - - -

TABLE 3.9-1 Environaental Parameters for G6seous Effluents at Vermont Yankee (Derived from Reference A)" Vegetables Cow Milk Goat Milk Meat variable Stored Leafy Pasture Stored Pasture Stored Pasture Stored YV Agricultural (Kg/M ) 2. 2. 0.70 2. 0.70 2. 0.70 2. 2 Productivity Soil Surface (KG/M ) 240. 240. 240. 240. 240. 240. 240. 240. 2 Density T Transport Time to (HRS) 48. 48, 48. 48. 480. 480. User TB Soil Exposure Time"3 (HRS) 131400. 131400. 131400. 131400. 131400. 131400. 131400. 131400. TF Crop Exposure Time to (HRS) 1440. 1440. 720. 1440. 720. 1440. 720. 1440. Plume TH Holdup After Harvest (HRS) 1440. 24. O. 2160. O. 2160. O. 2160. OF Animals Daily Feed (KG/ DAY) 50. 50. 6. 6. 50, 50. FP Fraction of Year on 0.50 0.50 0.50 Pa sture(2) FS Fraction Pasture When 1. 1. 1. I3) on Pasture FG Fraction of Stored 0.76 Veg. Grown in Garden FL Fraction of Leafy 1. l Veg. Grown in Garden FI Fraction Elemental lodine - 0.5 A Absolute Humidity - 5.6 (gm/M )I'3 3

Regulatory Guide 1.109. Revision 1.

Revision 17 Date 5/31/94 i 3-42 j neu. l

TABLE 3.9-1 (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 hours (1 year) for all pathways. (2) For Method II dose / dose rate analyses performed for releases occurring during the first or fourth calendar quarters, the fraction of time animals are assumed to be on pasture is zero (nongrowing season). For the second and third calendar 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. 3 (4) For all Method II analyses, an absolute humidity value equal to 5.6 (gm/m ) shall be used to reflect conditions in the Northeast (

Reference:

Health Physics Journal, Vol. 39 (August), 1980: Page 318-320, Pergammon Press). Revision 17 Date 5/31/94 12m .3-43. 8

TABLE 3.9-2 Usage Factors for Various Gaseous Pathways at Vermont Yankee (from Regulatory Guide 1.109. Table E-5) Leafy Vegetables Vegetables Milk Meat Inhalation 3 Age (kg/yr) (kg/yr) (1/yr) (kg/yr) (m fyp) Group 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 l l 17 5/31/94 Revision Date 3-44

utse4

4 3.10 Receptor Point and Long-Term Averace 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 radiciodines, tritium, and particulates with half lives greater than 8 days: 1. Direct exposure to contaminated ground: 2. Inhalation of air: 3 Ingestion of vesotables: 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 3.10.1 details the selection of important off-site locations and receptors. Section 3.10.2 describes the atmospheric model used to convert meteorological data into atmospheric dispersion factors. Section 3.10.3 presents the maximum atmospheric dispersion factors calculated at each of the off-site receptor locations. 3.10.1 Receptor locations Distances to the site boundary from the two evaluated gaseous release pathways (the Stack and North Warehouse) are provided in Table 3.10-2. Four important off-site receptor locations are considered in the dose and dose rate equations for gaseous radioactive effluents from these two release pathways. They are: i 1. The point of maximum gamma exposure (maximum gamma X/0) from an overhead noble gas cloud for determining skin and whole body dose rates and gamma air doses: 2. The point of maximum ground level air concentration (maximum undepleted X/0) of noble gases for determining skin and beta air dose rates and doses: 5/31/94 Revision Date unu 3-45 4

\\ 3. The point of maximum ground level air concentration (maximum depleted X/0) of radiciodines and other particulates for determining critical organ dose from inhalation; and 1 4. The point of maximum deposition (maximum D/0) 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 X/0 at any off-site location. The location of the maximum ground level air concentration and deposition of radiciodines and other particulates (NW sector. 2700 meters) was determined by finding the maximum five-year average depleted X/0 and D/0 at any off-site location. For the purposes of determining the Method I dose factors for radiciodines, tritium, and particulates, a milk 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 evalua+:o a ground level release using the same meteorological period-of-record as "ne stack. The highest long-term atmospheric dispersion factors at the site boundary were determined (see Table 3.10-1) 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. 3.10.2 Vermont Yankee Atmospheric Dispersion Model The long-term average atmospheric dispersion factors are computed for routine releases using Yankee Atomic Electric Company's (YAEC) AE0LUS-2 Computer Code (Reference B). AEOLUS-2 is based, in part, on the constant mean wind direction model discussed in hegulatory Guide 1.111 (Reference C). Since AE0LUS-2 is a straight-line steady-state model, site-specific recirculation correction f actors were developed for each release pathway to adjust the AE0LUS-2 results to account for temporal variations of atmospheric transport and diffusion conditions. The applicable recirculation correction factors are listed in Table 3.10-3. Revision 17 Date 5/31/94 3-46 nw

AE0LUS-2 produces the following average atmospheric dispersion factors for each location: 1. Undepleted X/0 dispersion factors for evaluating ground level concentrations of noble gases; 2. Depleted X/0 dispersion factors for evaluating ground level concentrations of radiciodines and other particulates: 3. Gamma X/0 dispersion factors for evaluating gamma dose rates from a sector averaged finite cloud (undepleted source); and 4. D/0 deposition factors for evaluating dry deposition of elemental radiciodines and other particulates. The North Warehouse depleted X/0 and D/0 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 X/0 and D/0 factors were derived using the deposition velocity concept presented in " Meteorology and Atomic Energy - 1968" (Reference E Section 5-3.2), assuming a constant deposition velocity of I cm/sec. Gamma dose rate is calculated throughout this ODCH 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/0T] (Reference B. Section 4), and the replacement of X/0 in infinite cloud dose equations by the [X/07). 3.10.3 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 f actors. 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 3.10-1. Table 3.10-1 also indicates which atmospheric dispersion factors are used to calculate the various doses or dose rates of interest. 17 /31/94 Revision Dat ams4 3-47

TABLE 3.10-1 Atmospheric Dispersion Factors Dose to Individual Dose to Air Release Dispersion Pathway Factor Total Body Skin Critical Gamma Beta Organ 9.40E-07 X/0Depigted (2700m NW) (sec/m ) 1.04E-06 1.04E-06 X/0 Undepleted (2200m WNW) l (2200m WNW) d (sec/m ) Stack D/g ~ 9.40E-09 (1/m ) (2700m NW) ~ ~ X/07 6.11E-07 6.11E-07 6.11E-07 (sec/m ) (750m SSE) (750m SSE) (750m SSE) 3 3.32E-05 X/0Depigted (417m NE) (sec/m ) 3.52E-05 3.52E-05 X/0Undep}eted (417m NE) (417m NE) (sec/m ) North Warehouse D/g ~ 5.97E-08 (1/m ) (357m S) ~ ~ X/07 6.42E-06 6.42E-06 6.42E-06 (sec/m ) (417m NE) (417m NE) (417m NE) 3 Revision 17 Date 5/31/94 3-48 n m.

TABLE 3.10-2 Site Boundary Distances Downwind Stack North Warehouse Sector Releases Releases N 400 m 459 m NNE 350 m 417 m NE 350 m 417 m ENE 400 m 451 m E 500 m 570 m ESE 700 m 561 m SE 750 m 612 m SSE 850 m 663 m S 385 m 357 m SSW 300 m 238 m SW 250 m 213 m WSW 250 m 213 m W 300 m 221 m WNW 400 m 281 m NW 550 m 697 m NNW 550 m 680 m 17 5/31/94 Revision Date atra. 3-49

TABl.E 3.10-3 Recirculation Correction Factors A. Stack Releases Sector 0.5 Mi 1.5 Mi 2.5 Mi 3.5 Mi 4.5 Mi 7.5 Mi N 1.4 1.4 1.2 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 Mi 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 t Revision 17 Date 5/31 /94 3-50 unu

3.11 Method to Calculate Direct Dose From Plant Operation Technical Specification 3.8.H.1 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). I 3.11.1 Turbine Ls11 ding j The maximum contribution of direct dose to the whole body or to any organ due to N-16 decay from the turbine is: where: (3-27) The dose contribution from N-16 decay at either the site D d boundary of maximum impact (west site boundary) or closest off-site residence - (mrem). l I Gross electric output over the period of interest (MW h). E e KN16(L) - The N-16 dose conversion factor for (L) equal to either: (1) 3.23E-06 for the maximum west site boundary; or (2) 1.29E-06 for the closest residence (mrem /MW h). e 3.11.2 North Warehouse i Radioactive materials and low level waste can be stored in the north ( warehouse. The maximum annual dose contributions to off-site receptors (west I site boundary line) from sources in the shielded (east) end and the unshielded (west) end of the north warehouse are: I (3-28) and DU 0.53 x A for the unshielded end (3-29) = U fmrem' ' mrem /yr' ' mrem' yr mrem /hr, hr where: l Revision 15 Date 07/08/93 amu 3-51 E ----_ _ _ _

l D3 0.25 x R3 for the shielded end (3 28) ' mrem' ' mrem /y r ' ' mrem' yr, mrem /hr, hr and DU 0.53 x A for the unshielded end (3-29) U 4 ' mrem' ' mrem /y r ' ' mrem' yr , mrem /hr, hr where: The annual dose contribution at the maximum site boundary D 3 location from fixed sources of radiation stored in the mrem', shielded east end of the North Warehouse yr The annual dose contribution at the maximum site boundary D g location from fixed sources of radiation stored in the mrem', unshielded west end of the North Warehouse yr ( Dose rate measured at 1 meter from the source in the shielded 5 3 mrem', end of the north warehouse hr, i Dose rate measured at 1 meter from the source in the 8 0 "I'* unshielded end of the north warehouse hr Dose rate to dose conversion factor which relates mrem /yr at 0.25 the west site boundary per mrem /hr measured at 1 meter from Revision 17 Date 5/31/94 3-52 + uns4

1 the source in the shielded end of the warehouse assuming it is full to capacity for one year mrem /yr', 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 assuming it is full to capacity for one year mrem /yr', , mrem /hr, 3.11.3 Low Level 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 sight) 0.28 x A DdE d x fd (3-30) mrem ' mrem /yr' Pmrem' (g) yr-modul e mrem /hr, hr j or 0.39 x Rd fd (3-31) DdS mrem ' mrem /yr' ' mrem' (g) yr-modul e, mrem /hr, hr, where: Revision 17 Date 5/31/94 ) nms' 3-53

t The annual direct dose contribution at the maximum site D = dE 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 mrem yr-modul e [ The annual direct dose contribution at the maximum site D dS 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 mrem y r-modu l e, Maximum dose rate measured at 3' from the side of the R d storage module whose unobstructed f ace (i.e., a side or end surface which is not shielded by other waste modules) is toward the west site boundary. The fraction of a year that a storage module is in use on f d 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' from the narrow end of the rectangular storage module when that i face is orientated toward the west boundary. 0.39 - Dose rate to dose conversion factor which relates mrem /yr at the west site boundary per mrem /hr measured at 3' from the long side of the rectangular storage module when that face is orientated toward the west boundary. Revision 17 Date 5/31/94 3-54 ons4

1 (b) Scatter From Skyshine DSKR 0.016 x ASKR X isk (3-32) mrem ' mrem /yr' fmrem ' (g) y r-l i ner mrem /hr, hr j and 0.015 x ASKD X fSK (3-33) OSKD mrem ' mrem /yr' ' mrem' (g) yr-mocul e mrem /hr, hr, j 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-li ner [ t RSK0 The annual skyshine scatter contribution to the dose at the maximum site boundary from a rectangular storage module containing DAW whose top surface,is not mrem obstructed due to stacking of modules y r-mod ul e,, R'SKR For Resins, the maximum dose rate measured at 3' over the top of each liner in a storage module (mrem /hr). RSKD For DAW, the maximum dose rate measured at 3' over the top surface of a storage module with DAW (mrem /hr). Revision 17 Date 5/31/94 unu 3-55

The fraction of a year that a storage module is in use f = 3g on the storage pad. Dose rate to dose conversion factor for the scatter dose 0.016 = from each resin liner source in storage which relates ~ mrem /yr at the west site boundary per mrem / hour at 3' from the top of the module. Dose rate to dose conversion factor for the scatter dose 0.015 from DAW boxes in storage which relates mrem /yr at the west site boundary per mrem /hr at 3* from the top of the module. (c) Dose From Resin liners Durino Trensfer 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: x T rans (3-34) x A ran D rans = 0.0025 t t t ' mrem /hr' ' rad' (mrem) (br) rad /hr, hr, where: The dose contribution to maximum site boundary D trans resulting from the unshielded movement of resin liners between a transfer cask and a storage module (mrem). $ rans Dose rate measured at contact (2") from the t unshielded trans top surface of the resin liner in R/hr. The time (in hours) that an unshielded resin liner is T tran exposed in the storage pad area. Revision 17 Date 5/31 /94 3-56 ' unu

The dose rate to dose conversion factor for an 0.0025 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) Intermodular 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 (3-30 and 3-31) 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 (3-35) accounts for any physical spacing between the outside storage modules which have not been covered by additional external shielding. DGap

2. 4 4 E-2 xWGap x ARL X fGap (3-36)

' mrem' mrem S (in) (Ci) (#) yr y r-i n -C i, where: O The annual dose contribution at the maximum site Gap boundary (west) from radiation streaming.through the intermodular gap between DAW storage modules used to shield resin modules from direct radiation (mrem /yr). Gap The intermodular gap width (inches) between adjacent W DAW storage modules facing the west site boundary. RL The total gamma activity contained in a condensate A resin liner stored directly to line with the intermodular gap adjacent DAW modules ( C1). Revision 17 Date 5 /11 /04 unu 3-57

The fraction of a year that the intermodular gap is f Gap not shielded. The activity to site boundary dose con, version f, actor 2.44E-2 mrem for a one inch wide intermodular gap y r-i n -C1 j 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). 3.11.4 Total Direct Dose Summary 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 j located on-site, j Revision 17 Date 5/31 /94 3-58 uru4

m. 3.12 Cumulative Doses Cumulative Doses for a calendar quarter and a calendar year must be maintained to demonstrate a compliance with Technical Specifications 3.8.B.1, 3.8.F.1 and 3.8.G.1 (10CFR50, Appendix I dose objectives). In addition, if the requirements of Technical Specification 3.8.M.2 dictate cumulative doses over a calendar year must be determined for Technical Specification 3.8.M.1 (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 10CFR20.1301(a)(1) for members of the public in unrestricted areas. Revision 15 Date 07/08/93 3-59 uns4

4.0 ENVIRONMENTAL MONITORING PROGRAM The radiological environmental monitoring stations are listed in Table 4.1. The locations of the stations with respect to the Vermont Yankee plant are shown on the maps in Figures 4-1 to 4-6. 4.1 Intercomparison Program All routine radiological analyses for environmental samples are performed at the Yankee Atomic Environmental Laboratory (YAEL). The YAEL participates in the U.S. Environmental Protection Agency's Environmental Radioactivity Laboratory Intercomparison Studies Program for all appropriate species and matrices offered by the agency. 4.2 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 18 years of operational data (as of 1990). The chosen air sampling locations are discussed below. s 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/0s (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. 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 Revision 17 Date 5/31/94 4-1 oru4

I maximum ground level D/0 (i.e. for any location in any sector, for the entire Vermont Yankee environs). This station provides a second function by its location in that it also monitors North Hinsdale, New Hampshire, the community with the second highest ground level D/0 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 11.3 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/0 (the second highest overall ground level D/0 for any location in any sector). Station AP/CF-11 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/0 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 (AP/CF-13 E, 3.1 km) in Hinsdale. The control air sampler was located at Spofford Lake (AP/CF-21, NNE, 16.1 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) 1 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, 3.4 km), which is along the western side of the valley in general proximity of historical dairy operations. (The sixth location is not a specific Technical Specification requirement.) i Revision 17 Date 5/31/94 i una4 4-la - i

r 4.3 Distances and Directions to Monitoring Stations It should be noted that the distances and directions for direct radiation monitoring locations in Table 4.1, as well as the sectors shown in Figures 4.5 and 4.6, 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. These distonces and directions are also used in the semiannual reports. Technical Specification 6.7 and Table 3.9.3, Footnote a, specify that in the Annual Radiological Environmental Surveillance 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 Surveillance Reports and ODCM for the l sampling and TLD monitoring stations, respectively. Revision 17 Date 5/31/94 unu 4-Ib

Table 4.1 Radiological Environmental Monitorino Stations ()) Exposure Pathway Sample Location Distance and/or Sample and Desionated Code (2) (km)(6s Direction (6) 1. AIRBORNE (Radiciodine and Particulate) AP/CF-11 River Station 1.9 SSE No. 3.3 AP/CF-12 N. Hinsdale. NH 3.6 NNW AP/CF-13 Hinsdale Substation 3.1 E AP/CF-14 Northfield, MA 11.3 SSE AP/CF-15 Tyler Hill Road (4) 3.2 WNW AP/CF-21 Spofford Lake 16.1 NNE 2. WATERBORNE a. Surface WR-11 River Station 1.9 Downriver No. 3.3 ^ WR-21 Rt. 9 Bridge 12.8 Upriver b. Ground WG-11 Plant Well On-Site 'WG-12 Vernon Nursing.Well 2.0 SSE WG-22 Skibniowsky Well 14.3 N c. Sediment SE-11 Shoreline Downriver 0.8 SSE From SE-12 North Storm 0.15 E Shoreline Drain Outfall(3) 3. INGESTION a. Milk TM-11 Miller Farm 0.8 WNW TM-12 Dominick (5) 5.2 E TM-14 Brown Farm 2.1 S TM-16 Meadow Crest Farm 4.4 WNW/NW TM-18 Blodgett Farm") 3.4 SE TM-24 County Farm 22.5 N b. Mixed TG-11 River Station 1.9 SSE Grasses No. 3.3 TG-12 N. Hinsdale NH 3.6 NNW TG-13 Hinsdale Substation 3.1 E TG-14 Northfield, MA 11.3 SSE TG-15 Tyler Hill Rd.(4) 3.2 WNW TG-21 Spofford Lake 16.1 NNE Revision 16 Date 10/28/93 anu4 4-2

Table 4.1 (Continued) fl) Radiological Environmental Monitoring Stations Exposure Pathway Sample Location Distggge Direction (6) and/or Sample and Designated Code (2) (km) 0.8 WNW c. Silage TC-11 MillerFggm TC-12 Dominick ) 5.2 E TC-14 Brown Farm 2.1 S 4.4 WNW/NW TC-16 MeadowCrestFggm TC-18 Blodgett Farm 3.4 SE TC-24 County Farm 22.5 N d. Fish FH 11 Vertion Pond (7) (7) FH-21 Rt. 9 Bridge 12.8 Upriver 4. DIRECT RADIATION DR-1 River Station 1.6 SSE No. 3.3 DR-2 N. Hinsdale, NH 3.9 NNW DR-3 Hinsdale Substation 3.0 E DR-4 Northfield. MA 11.0 SSE DR-5 Spofford Lake 16.3 NNE DR-6 Vernon School 0.46 WSW DR-7 Site Boundary 0.27 W DR-8 Site Boundary (8) 0.25 SW l DR-9 Inner Ring 2.1 N DR-10 Outer Ring 4.6 N DR-11 Inner Ring 2.0 NNE DR-12 Outer Ring 3.6 NNE DR-13 Inner Ring 1.4 NE DR-14 Outer Ring 4.3 NE DR-15 Inner Ring 1.4 ENE DR-16 Outer Ring 2.9 ENE DR-17 Inner Ring 1.2 E DR-18 Outer Ring 3.0 E DR-19 Inner Ring 3.5 ESE DR-20 Outer Ring 5.3 ESE DR-21 Inner Ring 1.8 SE i DR-22 Outer Ring 3.2 SE DR-23 Inner Ring 1.8 SSE DR-24 Outer Ring 3.9 SSE DR-25 Inner Ring 2.0 S Revision 17 Date 5/31/94 unu 4-2a i

Table 4.1 (Continued) Radiological Environmental Monitoring StationsII) Exposure Pathway Sample Location Distge Direction (6) and/or Sample and Designated Code (2) (km) DR-26 Outer Ring 3.7 S DR-27 Inner Ring 1.0 SSW DR-28 Outer Ring 2.2 SSW DR-29 Inner Ring 0.7 WSW DR-30 Outer Ring 2.3 SW DR-31 Inner Ring 0.8 W DR-32 Outer Ring 5.0 WSW DR-33 Inner Ring 0.9 WNW DR-34 Outer Ring 4.9 W DR-35 Inner Ring 1.4 WNW DR-36 Outer Ring 4.7 WNW DR-37 Inner Ring 3.0 NW DR-38 Outer Ring 7.7 NW DR-39 Inner Ring 3.2 NNW DR-40 Outer Ring 5.8 NNW (1) Sample locations are shown on Figures 4.1 to 4.6. (2) Station Nos. 10 through 19 are indicator stations. Station Nos. 20 through 29 are control stations (for all but the direct radiation stations). (3) To be sampled and analyzed semiannually. (4) Non-Tech Spec station. (5) Non-Tech Spec station. Sample will be collected as available. (6) Distance and direction from the center of the Turbine Building for direct radiation monitors; from the plant stack for all others. (7) Fish samples are collected from anywhere in Vernon Pond, which is adjacent to the plant (see Figure 4-1). (8) DR-8 satisfies Technical Specification Table 3.9.3 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. Revision 17 Date 5/31 /04 R12\\$4 4-3

l j 0 500 N METERS ] %D , ' ~. 's FENCELINE g 's 's i TC 11 's i I TM 11 A ( VERNOM POMO \\ 's STACK O' s s N s I g \\\\ w INTAXE 0 \\ t \\ itG-11 k \\ s A (N-11 p g%, \\- \\ N ' 01ScHARGE VERNON ELEMENTARY SCHOOL i l lt s / / A sg.11 HINSDALE. N.H. i t s ? W VERNON, V T. -0 conuEcTTcur / VERNON DAM \\/ ) Figure 4-1 Environmental Sampling Locations in Close Proximity to Plant 10 Date 4/4/91 Revision 4-4

l = \\ N ) o9'g@ TG 12 HINSDAl.E, N.H. AP/CF.12 A I i g TM 16 TC 16

tp,

'eq gt puY

e p

TG-13 Tc-15 1 8 AP/CF-13 3 A AP/CF-15 g 1 fll PLANT E ^ set inacz. verr xx rzo:mt 4-1 8 VERNON DAM i i I 1C 14 3 ipfer.33 VERNON, V.T. vf32 C;1

- A 3,.3, 3 i

Tc-18 o 1 2 3 L1LY PCMC KILOMETERS Environmental Sampling Locations Within 5 k=t of Plant Figure 4-2 Revision 16 Date 10/28/93-4-5

j l l i i N j ) \\ f A TM-24 TC 24 SP0FFORD LAKE TC 21 4 \\ A yg.pg A AP/CF-21 A FM-21 WR 21 HOGSACK 47. CHESTERFIELD $'E A MARLBORO BRATTLEBORO 9 in O rir r#MRCDErf a Elm 4-2 lS f _ _ _r_._ _ _, l s l l 9 TC-12 l HINSDALE I GUILFdRDe l A TM-12 e i WINCHESTER I YERNON e i e PLANT i i I 8 VERMONT NEW HAMPSHIRE ~ ~~ ~~ ~~ ~~ ~~ ~~ MAS 5ACHUSETTS t%SSACHUSETTS e NORTHFIELD A AP/CF-14 TG-14 { s GREENFIELD e 0 5 10 i KILOMETERS t 1 Figure 4-3 Environmental Sampling Locations Greater than 5 km from Plant i Revision 10 Date 4/4/91 4-6

I O 500 N METERS %z, i'g ^ ,s ~, e' 's 's f i FENCELINE ] ' i i s N l 's 's ~ g VERNCH PONO j \\ 's STACK Os A on-33 s g s N s 3 s g IHTAKE 09 s +g DR-7.A \\ DR 31 A A \\ ', j J DR-8 A

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CR 6 A k VERNON ELEMENTARY $CH00L h \\ OR 29 A d\\ j g [ HINSDALE. N.H. \\ g / -/ VERNON, V.T.

g g

O CONNECTTCUT / VERNON DAM OR 27 A R!vER Figure 4 4 TLD Locations in Close Proximity to Plant Revision 10 Date 4/4/91 1 47

} NNW

iNE N/

) NW A DR-40 A DR-10 NE N pt A DR-12 \\4DR-2 ' *A 4 DR-39 8""37 A A DR-36 +#, A U" A WNW ENE DR*16 si * ~~~] A DR-13 A F 4

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' VERNON DAM \\ A DR-21 A DR-19 A DR*32 g ca.30 DR-25 A FSt WSW VERNON, V.T. A A eR-1 DR-28 A DR-22 SW 4 DR-24 A DR 26 ) LILY PQFC >.--a SE KILOMETERS SSE SSW S Figure 4-5 TLD Locations Within 5 km or riant Revision 10 Date 4/4/91 4-8

.. ~. i I N } \\ l \\ I N""# "E I N CR-5

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CHESTERFIELD A MARLBORO I - 1 g e ( 71 BRATTLEB0 3 . Es em n m 45 I ~ ' ~ ~ ~ ~~ 8 r DR-38 h I i I HI SO 8 GUILFORD e 4, , I g CR-344lVERh e 1 1 I i / DR-20 I f VERMONT e - - /-, g, r - NEW HAMPSHIR i -- k - ~ MASSACHUSETTS 3 3 FMSSACHUSE N;*# / e NORTHFIELD \\ A CR-4 Jo r l, le T k g s l GREENFIELD e O S jo i KILOMETER $ Figure 4 6 TLD Locations Greater than 5 k n from Plant r Revision 10 Date 4/4/91 i. 4-9 i -n .w.

5.0 SETPOINT DETERMINATIONS Chapter 5 contains the basis for plant procedures that the plant operator requires to meet the setpoint requirements of the Radioactive Effluent Monitoring Systems Technical Specifications. They are Specification 3.9.A.1 for liquids and Specification 3.9.B.1 for gases. Each outlines the instrumentation channels and the basis for each setpoint. I j l ) i Revision 15 Date 07/08/93 5-1 urus

5.1 Licuid Effluent Instrumentation Setpoints Technical Specification 3.9.A.1 requires that the radioactive liquid effluent instrumentation in Table 3.9.1 of the Technical Specifications have alarm setpoints in order to ensure that Specification 3.8.A.1 is not exceeded. Specification 3.8. A.1 limits the activity concentration at any tir.e in liquid effluents to the appropriate effluent concentration values ic Appendix 8 Table 2 Column 2 of 10CFR20, and a total noble gas concentration limit of 2E-04 pCi/ml. 5.1.1 Licuid Radweste Discharge Monito' (17/350) The sample tank pathways shown on Figure 6-1 are monitored by the liquid radwaste discharge monitor (17/350). Periodsc batch releases may be made from the waste sample tanks, detergent waste tonk or floor drain sample tank. 5.1.1.1 Method to Determine the SetDoint of the Liouid Radwaste Discharge Monitor (17/350) The instrument response (in counts per second) for the limiting concentration at the point of discharge is the setpoint. denoted R and

setpoint, is determined as follows:

DF Rsetpoint " S1 { C,i DFmin 1 i '(5-1) (Cps) (#) ( ) ( ECi ) E~ pC1 ml Where: t' d DF Dilution factor (as a conservative measure, F a DF of at least 1000 is used) m (dimensionless). (5-2) F, Flow rate past monitor (gpm) Fd Flow rate out of discharge canal (gpm) DF,i n Minimum allowable dilution f actor (dimensionless) { (5-3) Revision 15 Date 07/08/93 unu 5-2

Effluent concentration values for radionuclide

ECL,

= "i" from 10CFR20.1001-20.2401. Appendix B. Table 2. Column 2 (pCi/ml) Activity concentration of radionuclide "i" in C,3 mixture at the monitor (pCi/ml) Detector counting efficiency from the most recent Sj liquid radwaste discharge monitor calibration curve (cps /(pCi/ml)) 5.1.1.2 Liauid Radwaste Discharoe 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 cos per 1 Ci/mi of gamma emitters which i emit one photon per disintegration. The activity concentration of each radionuclide, C,3, 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: 1 C,3 (uCi/ml) ECL, (pCi /ml) Cs-134 2.15E-05 9E-07 Cs-137 7.48E-05 1E-06 Co-60 2.56E-05 3E-06 Revision 15 Date 07/08/93 5-3 u ns.

E C,3 - 2.15E-05 + 7.48E-05 + 2.56E-05 1 (pCi) (pCi) (pCi) (pCi) m1 m1 m1 ml - 1. 2 2 E-04 ( pCi ) ml l E (5-3) D F,i n C (pCi-ml) ml-Ci l l l '2.15E-05 7.48E-05 2.56E-05 ' . 9E-07 1E-06 3E-06, Ci-ml Ci-ml ( pCi-ml ) ( ml-Ci) ( ml-pCi) ml-Ci - 107.2 The minimum dilution factor, OF,in, needed to discharge the mixture of radionuclides in this example is 108 (roundup). As a conservative measure, an actual dilution f actor 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: F<1 F, (5-4) g (gpm) (gpm) 18,000 gpm - 18 gpm 1,000 Revision 15 Date 07/08/93 a m s4 5-4

Under these conditions, the setpoint of the liquid radwaste discharge monitor is: (5-1) DF R,,tpoint - Si E C,5 0Fmin 1 (Cps) (#) ( N$) {Nml ) C1 1,000 4.9E46

1. 22 E-04 108 Ci (cps)

(#) ( cps-ml ( ml ) Ci - 5.535 cps In this example, the count rate alarm of the liquid radwaste discharge monitor should be set at 5.535 cps above background. 5.1.1.3 Basis for the Liauid Radwaste Discharae Monitor Setpoint The liquid radwaste discharge monitor setpoint must ensure that Specification 3.8. A.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 as follows: R - }) C,i Sit (5-5) 1 (cps) (Eml ) ( cps el) I pCi Where: Response of the monitor (cps) R Detector counting efficiency for radionuclide "i" S H (cps /(pCi/ml)) C, Activity concentration of radionuclide "i" in mixture at the monitor (pCi/ml) Revision 15 Date 07/08/93 5-5 aine4

The detector calibration procedure establishes a counting efficiency for a given mix of nuclides seen by the detector. Therefore, in Equation 5-5 one for Sii, where Si represents the counting efficiency may substitute Si 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. Si E C., (5-6) R 1 ' cps-el' ' Ci ' (cps) pCi , 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: C (5-7) di E s1 ECL i 5 Ci -m1 ( ml-gCi ) Where: Activity concentration of radionuclide "i" in the C di mixture at the point of discharge to an unrestricted area ( Ci/ml) Effluent concentration limit for radionuclide "1" ECL 5 from 10CFR20.1001-20.2401, Appendix B Table 2, Column 2 ( Ci /ml ) The activity concentration of radionuclide "i" at the point of discharge .is related to the activity concentration of radionuclide "i" at the monitor as follows: F Cet - C,5 ( ml ) (pCi) (gpm) (5 8) Ci ml gpm Revision 15 Date 07/08/93 5-6 nnu

1 Where: Activity concentration of radionuclide "i" in the C dt mixture at the point of discharge ( Ci/ml) Flow rate past monitor (gpm) F, Flow rate out of discharge canal (gpm) F d Substituting the right half of Equation 5-8 for C j in Equation 5-7 and o solving for F /F, yields the minimum dilution factor needed to comply with o Equation 5-7: i F C., d D F,i n s 2 (5-3) ECLj 'gpm' 'pCi-ml'

gpm, ml -pCi,

s Where: Flow rate out of discharge canal (gpm) F d Flow rate past monitor (gpm) F, C,, Activity concentration of radionuclide 'i" in mixture at the monitor ( Ci/ml) Effluent concentration limit for radionuclide "i"

ECL, from 10CFR20.1001-20.2401, Appendix B. Table 2 Column 2 (pCi/ml)

If F /f is less than DF then the tank may not be discharged until j d m

min, either F or F, or both are adjusted such that:

d 2 DF (5-3) min m (gpm) gpm Revision 17 Date 5/31/94 5-7 u ns. l

Usually F /F, is greater than DF,,, (i.e., there is more dilution than d necessary to comply with Equation 5-7). The response of the liquid radwaste discharge monitor at the setpoint is therefore: E, C,i R setpoint S (5-1) 1 DF,g, fcps-ml' 'pCi ' (cps) (#) pC1 , m1, 5.1.2 Service Water Discharge Monitor (17/351) The service water pathway shown on Figure 6-1 is continuously monitored by the service water discharge monitor (17/351). The water in this line is not radioactive under normal operating conditions. The alarm setpoint on this monitor is set at a level which is three times the background of the instrument. The service water is sampled if the monitor is out of service or if the alarm sounds. Under normal operating conditions, the concentration of radionuclides at the point of discharge to an unrestricted area from the service water effluent pathway will not exceed the effluent concentration limits specified in 10CFR20.1001-20.2401, Appendix B, Table 2, Column 2. Revision 17 Date 5/31/94 nmu 5-8 L

5.2 Gaseous Effluent Instrumentation Setooints Technical Specification 3.9.B.1 requires that the radioactive gaseous effluent instrumentation in Table 3.9.2 of the Technical Specifications have their alarm setpoints set to insure that Technical Specifications 3.8.E.1 and l 3.8.K.1 are not exceeded. Technical Sp c.:ification 3.8.K.1 limits the gross radioactivity release rate at the steam jet air ejector (SJAE) to 0.16 Ci/sec. 5.2.1 Plant Stack Noble Gas Activity Monitors (RR-108-1A and RR-108-1B) and Augmented Off-Gas System Noble Gas Activity Monitors (3127 and 3128) The plant stack and A0G noble gas activity monitors are shown on Figure 6-2. 5.2.1.1 Method to Determine the Setpoint of the Plant Stack Noble Gas Activity Monitors (RR-108-1A and RR-108-1B) and the Auamented Off-Gas System Noble Gas Activity Monitors (3127 and 3128) The setpoints of the plant stack and A0G system noble gas activity monitors are determined in the same manner. The plant stack or A0G 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 R R is the lesser of: 3pt. 3pt (5-9) l h 818 S R st g 3 3 (cpm) (yr m uCi-m ) (com-cm ) (sec) ( pCi-yr) mre pci-sec pti 3 3 cm mrem-m and: h R n - 3,000 S s (5-10) 3 g com-en,3 I uti-yr } sec (cpm) (mrem} I } Ip) mrem-sec yr u where: R - Response of the monitor at the limiting total body dose t rate (cpm) 3 500 (mrem-uCi m ) 818 (1E+06) (6.11E-07) yr-pC1-sec 500 - Limiting total body dose rate (mrem /yr) Revision 17 Date 5/31/94 5-9 ) uns4

IE+06 - Number of pCi per Ci (pCi/pCi) = [X/03, maximum five-year average gamma atmospheric l 7 6.11E-07 3 dispersion factor (sec/m ) S - Appropriate (plant stack or A0G system) detector g counting efficiency from the most recent calibration (cpm /( Ci/cc)) 3 F - Appropriate (plant stack or A0G system) flow rate (cm /sec) 3 DFB - Composite total body dose factor (mrem-m /pCi-yr) c I d DFB j j (5-11) I d j i d - The relative release rate of noble gas "i" in the mixture j ST ortheA0G,dA0G) at the monitor (either the stack, for noble gases identified (pCi/sec) 3 DFB - Total body dose factor (see Table 1.1-10) (mrem-m /pti-yr) j R - Response of the monitor at the limiting skin dose rate (cpm) 3,000 - Limiting skin dose rate (mrem /yr) DF'c - Composite skin dose f actor (mrem-sec/pCi-yr) y0 4 DF'i s l (5-12) yO g i DF'i s - Combined skin dose factor (see Table 1.1-10) l (mrem-sec/pC1 yr) Revision I7 Date 5/M /94 amu 5-10

5.2.1.2 Plant Stack Noble Gas Activity Monitor Setcoint Example The following setpoint example for the plant stack noble gas activity monitors demonstrates the use of equations 5-9.and 5-10 for determining setpoints. The plant stack noble gas activity monitors, referred to as " Stack Gas I" (RR-108-1A) and " Stack Gas II" (RR-108-1B), consist of beta sensitive scintillation detectors, electronics, an analog 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 f actors 5, of IE+08 cpm per 1 pC1/cc of noble gas. 9 The nominal plant stack flow is 8.3E+07 cc/sec ((175,000 cfm x 28,300 l 3 cc/f t )/60 sec/ min). When monitor responses indicate that activity levels are below the LLDs at the stack (or A0G) 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 (SJAE). This setpoint example is based on the following data (see Table 1.1-10 for DFB, and DF',): 0,SJM DFBt D F'i, l 3 g Ci ) g ) { mrem-sec ) i mrem-m sec pCi -y r pCi -y r 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+02 1.81E-03 3.24E-03 Xe-133 1.97E+01 2.94E-04 5.58E-04 Revision 17 Date 5/31./4 4 5-11 a m e4 9 W 7

}}6fJAEDFBi (5-11) 0FB c 1 0f AE 1 };OfJAEDFBt = (1.03E+04)(8.83E-03) + (4.73E-02)(5.92E-03) 1 + (2.57E402)(1.47E 02) + (1.20E+02)(1.17E-03) + (3.70E+02)(1.81E-03) + (1.97E+01)(2.94E-04) 3 = 9.83E+01 ( Ci-mrem-m /sec-pCi-yr) }[ofJAE - 1.03E+04 + 4.73E+02 + 2.57E+02 1 + 1.20E+02 + 3.70E+02 + 1.97E+01 - 1.15E+04 Ci/sec 9.83E+01 DFB C 1.15E+04 3 - 8.52E-03 (mrem-m /pCi-yr) tb y y R - 818 5 spt U l~ DFB c I I - (818) (IE+08) (8.3E+07) (8.52E-03) - 115,674 cpm 1 l r Next: l };0lJAEDF'j s D F'e (5-11) 13Df i l 1 Revision 17 Date 5/31/94 amu 5-12

}]og##ID F'3 3 - (1.03E404)(1.06E-02) + (4.73E-02)(1.43E-02) S t + (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 (pCi-mrem-sec/sec-Ci-yr) DF'* 1.14 E +02 1.15E+04 - 9.91E-03 (mrem-sec/ Ci-yr) spt 7 y R - 3,000 S skin 8 _F_ D F'c I I - (3,000) (1E+08) ( 8. 3 E +07 ) (9.91E-03) - 364,728 cpm The setpoint, R is the lesser of R *pt a nd R "t ". For the noble D 5

3pt, s

sp D 5 kin gas mixture in this example R *pt is less than R indicating that the 3 3pt total body dose rate is more restrictive. Therefore, in this example the " Stack Gas I" and " Stack Gas II" noble gas activity monitors should each be set at 115.674 cpm above background or at some conservative value below this l (such as that which might be based on controlling release rates from the plant in order to maintain off-site air concentrations below 20 x ECL when averaged over an hour), or to account for other minor releases from the waste oil burner. For example, if an administrative limit of 70 percent of the Technical Specification whole body dose limit 500 rem /yr (115.674 cpm) is chosen, then the noble gas monitor alarms should be set at no more than 80,972 cpm above background (0.7 x 115,674 - 80.972). 5.2.1.3 Basis for the Plant Stack and A0G System Noble Gas Activity Monitor Setpoints The setpoints of the plant stack and A0G system noble gas activity monitors must ensure that Technical Specification 3.8.E.1.a is not exceeded. Sections 3.4 and 3.5 show that Equations 3-5 and 3-/ are acceptable methods for determining compliance with that Technical Specification. Which equation (i.e., dose to total body or skin) is mora limiting depends on the noble gas Revision 17 Date 5/31/94 5-13 unu

mixture. Therefore, each equation must be considered separately. The derivations of Equations 5-9 and 5-10 begin with the general equation for the response R of a radiation monitor: R - }[ S 91 C,i (5-13) i f 3 r 3 3 (cpm) cpmgm pCi pC1 3 cm, where: R - Response of the instrument (cpm) 3 S - Detector counting efficiency for noble gas "i" (cpm /(pCi/cm )) ol C,( - Activity concentration of noble gas "i" in the mixture at the 3 noble gas activity monitor (pCi/cm ) The relative release rate of each noble gas. 6, (pCi/sec), 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 A0G 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 present 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 A0G 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 5.2-1 3 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 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. Revision 17 Date 5/H /04 oru4 5-14 P

C,i, the activity concentration of noble gas "1" at the noble gas activity monitor, may be expressed in terms of Og 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 A0G noble gas activity monitors the appropriate flow rate is the A0G system flow rate. I (5-14) C,i - 0, F ( Ci ) ( Ci) (sec) 3 3 sec cm cm where: The release rate of noble gas *i" in the mixture for each 0, noble gas identified ( Ci/sec). 3 Appropriate flow rate (cm /sec) F Substituting the right half of Equation 5-14 into Equation 5-13 for C,i yields: S 0 (5-15) R gt 1 ' ' Ci ' r 3 3 sec cpm-cm (cpm) 3 pCi (sec cm, j The detector calibration procedure establishes a counting efficiency for a given mix of nuclides seen by the detector. Therefore, in Equation 5-15 one may substitute S for 591, where S represents the counting efficiency o o 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. R-5 0 9 1 3 (cpm) ( cpm-cm ) ( sec ) ( Ci ) C1 3 sec cm Revision 17 Date 5/31/94 5-15 nnu

l l The total body dose rate due to noble gases is determined with Equation 3-5: Atbs - 0.61 E D

DFB, (3-5) i 3

(mrem) ( pCi-sec) ( Ci ) ( mrem-m ) 3 sec pCi-yr yr Ci-m Where: total body dose rate (mrem /yr) due to noble gases A tbs from stack release (1.0E+06) x (6.11E-07) (pCi-sec/ Ci-m ) l 3 0.61 number of pCi per pCi (pCi/ C1) 1E + 06 [X/0F, maximum long term average gamma l 6.11E - 07 3 atmospheric dispersion factor (sec/m ) the release rate of noble gas "i" in the mixture 0 5 for each noble gas identified (pCi/sec) (Equivalent to QsT for noble gases released at i the plant stack.) total body dose factor (see Table 1.1-10) DFB i 3 (mrem-m /pCi-yr) A composite total body gamma dose factor. DFB, may be defined such c that: DFB E 0 -E d DFB e 5 i i (5-17) 1 1 r 3 r S 3 3 mrem-m 'pCi * ' Ci ' mrem-m pCi-yr, (sec, sec, y pCi-yr, Revision 17 Date 5 /31/ob 5-16 unu

Solving Equation 5-23 for DFBe yields: i E0 DFBt DFB I (5-11) e E01 1 Technical Specification 3.8.E.1.a limits the dose rate to the total body from noble gases at any location at or beyond the site boundary to 500 mrem /yr. By setting A equal to 500 mrem /yr and substituting DFB for DFB, tb e in Equation 3-5, one may solve for E Og at the limiting whole body noble gas i dose rate: E 0, - 818 DFB e 3 ( pCi ) ( mrem-$1Ci-m ) ( pCi-yr y 3 sec yr-pCi -s ec mrem-m D in Equation 5-16 yields R[pt, the Substituting this result for E 65 response of the monitor at the limiting noble gas total body dose rate: D R pt - 818 S g DFB e 3 ( mrem-11Cid ) ( cpm-cm ) (sec) ( pCi-yr ) (cpm) 3 3 y r-pCi-s ec pC) cm mrem-m The skin dose rate due to noble gases is determined with Equation 3-7: A "t "E 0 D F'$ $ (3-7) l p 1 i (mrem) ( Ci ) ( mrem-sec ) yr sec pCi -y r Where: j I" Alp't Skin dose rate (mrem /yr) = 1 Revi sion _12_ Date 5 / 31/ot 1 5-17 ains4

i ) 0, The release rate of noble gas "i" in the mixture for each noble gas identified (pCi/sec) i (Equivalent to 0,8I for noble gases released at the plant stack). D F'$, Combined skin dose factor (see Table 1.1-10) l (mrem sec/pCi-yr). A composite combined skin dose factor, DF'c, may be defined such that: DF[ E O -E 0 DF{3 (5-19) l i 3 1 1 ' mrem-sec ' fpCi ' 'pCi ' ' mrem-s ec ' pCi-y r

sec,

,sec,, pCi-yr, j Solving Equation 5-19 for DF' yields: E 0, DF l 3 DF' - ' E01 1 Technical Specification 3.8.E.1.a limits the dose rate to the skin from noble gases at any location at or beyond the site boundary to 3,000 mrem /yr. By setting A5M" equal to 3,000 mrem /yr and substituting DF for DF' in e 4 Equation 3-7 one may solve for E Q at the limiting skin noble gas dose rate: i 1 3 E 0, - 3,000 1 D F,e (pCi) (mrem) ( Ci-yr ) sec yr mrem-sec 3 Revision 17 Date 5/31/94-uns4 5-18

\\ Substituting this result for 0, in Equation 5-16 yields R[fti". the response of the monitor at the limiting noble gas skin dose rati: (5-10) i" Rlft 3,000 S o F DF,c 3 (cpm) { mrem) ( cpm-cm ) ( sec) ( pCi-yr ) 3 mrem-sec yr pC1 cm Revision 17 Date 5/31/94 5-19 uns4

TABLE 5.2-1 Relative Fractions of Core Inventory Noble Gases After Shutdown Time Kr-83m Kr-85m Kr-85 Kr-87 Kr-88 Xe-131m 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 i t < 48 h .003 .004 .001 .004 .022 .758 .010 .198 48 h I t < 5 d .005 .006 .024 .907 .001 .058 5 d I t < 10 d .007 .008 .016 .969 10 d I t < 15 d .014 .014 .006 .966 15 1 t < 20 d .026 .022 .002 .950 20 i t < 30 d .048 .034 .001 .917 30 i t < 40 d .152 .070 .777 40 1 t < 50 d .378 .105 .517 50 i t < 60 d .652 .108 .240 60 i t < 70 d .835 .083 .082 t 170 d .920 .055 .024 Revision 15 Date 07/08/93 unu 5-20

5.2.2 Steam Jet Air Ejector (SJAE) Noble Gas Activity Monitors (17/150A and 17/1508) The SJAE noble gas activity monitors are shown in Figure 6-2. 5.2.2.1 Method to Determine the Setpoints of the Steam Jet Air Ejector Offoas Activity Monitors (17/150A and 17/1508) The SJAE noble gas activity monitor response in mR/hr at the limiting release rate is the setpoint, denoted RfffE, and is determined as follows: RfffE 1.6E+05 5 (5-21) 9 (mR/hr) (pCi) mNc) (sec) ( br-pCi cc sec Where: Rf((E Response of the monitor at the limiting release rate = (mR/hr) Limiting release rate for the SJAE specified in 1.6E+05 Technical Specification 3.8.K.1 ( Ci/sec) Detector counting efficiency from the most recent S o calibration ((mR/hr)/( Ci/cc)) SJAE gaseous discharge flow (cc/sec) F 5.2.2.2 Basis for the SJAE 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 5-21 is straightforward. Simply taking Equation 5-16 and substituting the limiting release rate at the SJAE for 6 yields Equation 5-21, the setpoint equation for the SJAE notle gas activity monitor. + Revision 15 Date 07/08/93 5-21 amu

6.0 LIQUID AND GASE0US EFFLUENT STREAMS, RADIATION MONITORS AND RADWASTE TREATMENT SYSTEMS Figure 6-1 shows the normal (design) radioactive liquid effluent streams, radiation monitors, and the appropriate Liquid Radwaste Treatment System. Figure 6-2 shows the normal (design) gaseous effluent systems, radiation monitors, and the appropriate Gaseous Radwaste Treatment System, 6.1 In-Plant Radioactive liauid Effluent Pathways l 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 potenti&lly 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. 1 Revision 17 Date 5/31/94 nirw4 6-1

l 1 l 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., A0G 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 Revision 17 Date 5/31 /04 6-2

nma,

for radioactivity and water purity. A release is allowed once it is determined that the activity in the liquid wastes will not exceed Technical Specification release limits. 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 l additional check during the release. The alarm or trip setpoint on the l monitor is set according to the Technical Specification 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. l i Revision 17 Date 5 /31/o!.. Rl!\\84 6-3

6.2 In-Plant Radioactive Gaseous Effluent Pathways l 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 ( A0G) reduces the ejector radioactive gaseous release rates to the atmosphere. The A0G 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 A0G 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 hours 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 residsal 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 A0G System as the air ejector line is Revision 17 Date s /11 /04 6-4 onu

M EXAMPLE PROBLEM NO. 9 (Continued) The dose factor (DFG,,co) for each radionuclide detected in the plant stack charcoal and particulate filter sample (plus tritium) is taken from Table 1.1-12 of the ODCH. Therefore: Dco, - (5.42E-04)(1.14E+01) + (1.10E-02)(4.31E+02) + (2.30E-01)(7.63E+00) + (1.15E-02)(1.63E+01) + (2.60E-02)(3.71E+00) + (4.30E-03)(7.71E+01) + (1.12E-04)(8.21E-01) + (0.15)(3.13E-04) - Answer Deo, - 7.12 mrem maximum organ dose for the month. l l Revision 17 Date 5/31/94 uns4 A-23 l l

prepared for operation. To prevent opcrating 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/ absorber 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 / absorber subsystem may be bypassed if it becomes unavailable provided the releases are within Technical Specification limits. With the dryer / absorber 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 17 Date 5/31/94 6-5 nmu 1

<_m The preheater assures that the vapor entering the hydrogen-oxygen recombiner is heated to approximately 300*F. At this temperature, the water vapor in the stream becomes superheated steam, thereby, protecting the i recombiner catalyst. During passages through the recombiner, the recombination of H2 and 02 in an exothermic reaction increases the stream temperature to approximately 520*F. 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 -40*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 -40*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 b'eds operate in parallel to remove the moisture and maintain the off-gas near the design dew point (-40*F). If the dryer fails, the -40*F dew point air leaving the mechanical system can enter the guard bed for over 6 hours without affecting the performance of the charcoal beds downstream. Revision 17 Date 5/31/94 mm 6-6

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. t The release of significant quantities of gaseous and particulate radioactive material is prevented by the combination of the design of the air ejector A0G system and automatic isolation of the system from the stack. Gas flow from the main condenser stops when the air ejectors are automatically Revision 17 Date 5/31 /04 6-7

nna.

isolated from the main condenser by either a high radiation signal in the main l steam line or by high temperature and/or pressure signals from the A0G 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. i Revision 17 Date 5/31/94 R12\\84 b*O

4 v N k -f t, J t Chunkeltab Westes w - tab Dreim Cheadset Driertent Sample side W N -.) q Foiialung Rssins moi. Waste M M W $*3 IM N Waets p a%Deccatal'h N O h k

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,g -(t) 1 + M F-1SM gal US og R) 10,000 gal yp 13 g CL W JL JL i 3is se gym ? Soud Wasz D 1.e= Purity Wesies: spam men. Dry =eil r.. - Drains t - in Reerine adding Place Dans Dispeaubie Rad *este Bdding Floor Drains Liners 1r 37 W Building Floor Drains 3 t Jk jg Weste Waste M Fuet teel Condeamste High Pwity Weews-DryweB F4aspoent Drams r CoDector g. 8"'*8' M M Reacter Buildmg Pgtipnaret Desins (267 sq R) 05 cubic n) M sa kne >lo, coo y , i, IE Pad =am Baldina Pei,rnent Drain. -) 25A00 set g -+ $00,000 gal A Systein For Reuse Th Buildeg Egopment Drains !!9 gym 1P 17 Condensate Phase Separater Cleanup Phase Sepermar Resin anse p 4 M t Waste Chuecelon N Weste p M 110 m Surge Fither Denest N M i10 grea US at Al D 35,000 sal Liquid Pamf=ne's

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t Dis *=se FIGURE 6-1: Radioactive Liquid EfIluent Streams, Radiation Monitors, and Radwaste Treatment System at Vermont Yankee *

Normal (design) radioactive process streams only are shown.

CDiSTRUMODCMFIGS ., ~... _ -.. -.., .s .. ~.. .-.... m..-

.O Release to AnnosE M 4 JL I "A Particuten A.RA N h, Mau 08 Bulkfing Wen 1stion darleding Tutme Boudtag) TWbene Gland SenLNechankel vacues Puey g Staney Gas Trenunear (Cantainment Purge) j P.ana stack Wasie Od Bumer Startup Bypens a k n (Mmh Werthouse) Ji MasNre MaiMuft s se,,.op S s% Bed ,n-ce c-*d A Prehes er Condenser a Condenwr bhga l Hydrogen l l' 1r 1r 42-l l Catatync lll bd M0"i"' PJier Charecal Four Masmre 6 J Glycoi/waen MM B 3r d I'", Confensaw D 88d JL ><1 JL ^' A2 JL Air in Air la 1P Bleed Bleed I e7 I y No. 3 1' kg v4>4 4 Funer (One Per Trms) p 4, p AOG Byrass f* ~ ,r No. 4

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$yseem Dryer Bed 7% AOG Bypass C e HYDROGEN DS.UTION AND RECDMSINER SUBSYSW l MOISTURE REMOVAUDR)TJt St!B3Y37EM l OEMtCOAL AD5098ER SU35YITEM FIGURE 6-2: Radioactive Gaseous Efiluent Streams, Radiation Monitors, and Radwaste Treatment System at Vermont Yankee *

Normal (design) radioactive process streams only are shown.

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, Parch 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, March 1976. 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. 5/31/94 Revision 17 Date R-1 a:ru4

APPENDIX A METHOD I EXAMPLE CALCULATIONS Revision 17 Date 5/31/94 R12\\84

EXAMPLE CALCULATION N0. 1 Type Total Body Dose From Liquid Effluents Reference 1 a) 00CM Sections 3.2 and 3.3 (Method I). b) Technical Specifications 3.8.8.1 and 3.8.C.1. Problem Calculat'. the off-site dose to the total body and maximum organ resulting from the batch release of radioactive liquid effluents. Plant Data a) Analysis from a representative grab sample of the liquid waste volume to be discharged indicates the following radionuclide activities were released to the Connecticut River: 3. FL

(mrem /Ci)

DFl

- (mrem /Ci)

Rale s (C ) g t g. ime H-3 3.40E-01 2.60E-04 2.06E-04 Co-60 1.72E-05 2.13E-01 1.28E+00 Co-134 3.51E-05 1.28E+02 1.60E+02 1-131 7.20E-04 2.57E-02 1.47E+01

Total body dose factor from Table 1.1-11.
- Maximum organ dose factor from Table 1.1-11.

l Revision 15 Date 07/08/93 A-1 oru.

EXAMPLE CALCULATION NO. 1 (Continued) Calculation The total body dose is calculated from Equation (3-1): Dtb-I Og 0Fl ith (3-1) (mrem) (Ci) (mrem /Ci) Therefore: Dtb - (3.40E-01)(2.06E-04) + (1.72E-05)(2.13E-01) + (3.51E-05)(1.28E+02) + (7.20E-04)(2.57E-02) - Answer (1) Dtb - 4.59E-03 mrem to the total body. Next, the maximum organ dose is calculated from Equation (3-3): D., - }] 0 3 D F Lj,, 1 (mrem) (Ci) (mrem /C1) Therefore: 0,, = (3.40E-01)(2.06E-04) + (1.72E-05)(1.28E+00) + (3.51E-05)(1.60E+02) + (7.20E-04)(1.47E+01) - Answer (2) D,, - 1.63E-02 mrem to maximum organ. \\ i i i Revision 15 Date 07/08/93 nruw A-2 )

EXAMPLE CALCULATION NO. 2 Type Total Body Dose Rate from Noble Gases References a) ODCM Section 3.4 (Method I). b) Technical Specificat; ion 3.8.E.1.a. Problem Calculate the off-site total body dose rate resulting from the release of noble gases from the plant stack during power operations. Plant Data f a) Maximum plant stack gas monitor (I or II) Count rate during period of interest (M): 80.000 cpm b) Stack flow rate during release (F): 8.26E+07 cc/sec cc/sec, ) (175.000 cfm x 4.72E+02 cfm c) Plant stack monitor detector counting 1E+08 cpm per efficiency (Sg): Ci/cc d) The last measured release rate mix of DsJAE i noble gas from the SJAE and corresponding dose f actor DFB, from Table 1.1-10. Q sJAE DFBj 9 3 (mrem-m /DCi-yr) (uCi/sec) Xe-138 5.15E+03 8.83E-03 Kr-87 2.37E+02 5.92E-03 Kr-88 1.29E+02 1.47E-02 Xe-135 1.85E+02 1.81E-03 Revision 17 Date 5/31/94 A-3 unu

EXAMPLE CALCULATION NO. 2 (Continued) Calculation The dose rate is calculated from Equations (3-5) and (3-28): Atbs - 0.61 E ofI DFB t (3-5) 1 3 g mrem) ( pCi-sec ) ( Ci ) ( mrem-m ) yr Ci-m3 sec pCi -yr and where the stack release rate is determined from: gsJAE i ~ E DsJAE (3-28) 1 1 (cpm) (pCi/cc) ( cc ) cpm sec ( Ci ) sec First, determine the sum (E ) of all OfJAE and the fraction that each noble 1 i gas i represents in the total gas mix. EOfJAE - (5.15E+03) + (2.37E+02) + (1.29E+02) + (1.85E+02) 1 - 5.70E+03 pCi/sec and the relative fraction of each noble gas: I Revision 17 Date 5/31 /04 uns4 A-4

EXAMPLE CALCULATION NO. 2 (Continued) f 1 Of*/5.70E+03 Relative fraction of Total 0.904 Xe-138 5.15E+03/5.70E+03 0.042 Kr-87 2.37E+02/5.70E+03 0.023 Kr-88 1.29E+02/5.70E+03 0.032 Xe 135 1.85E+02/5.70E+03 Next, the stack release rate of each noble gas i from Equation (3-28) can be substituted into Equation (3-5) to give the dose rate as: Atbs - 0.61 M F E f DFBi 5 - 0.61 80.000 1/1 E +08

8. 26 E +07 E f

DFB i i i 4 - 4.03E+04 [(0.904)(8.83E-03) + (0.042)(5.92E-03) + (0.023)(1.47E-02) + (0.032)(1.81E-03)] - Answer Atbs - 348 mrem / year noble gas total body dose rate. l l Revision 17 Date 5/31/94 A-5 uns4

i EXAMPLE CALCULATION NO 3 T_yp e Total Body Dose Rate From Noble Gases References a) 00CM Section 3.4 (Method I). i b)_ Technical Specification 3.8.E.1.a. Problem Calculate the off-site total body dose rate resulting from the release of i noble gases from the plant stack recorded to have occurred 32 days after plant shutdown. Plant Data a) Maximum plant stack gas monitor (I or II) Count rate during period of interest (M): 80.000 cpm b) Stack flow rate during release (F): 8.26E+07 cc/sec (175.000 cfm x 4.72E+02 cc/sec. ) cfm c) Plant stack monitor detector counting 1E+08 cpm per efficiency (Sg): Ci/cc d) The noble gas mix fractions f (t) t corresponding to 32 days taken from Table 5.2-1. i li (32 day)* DFB ** 3 Kr-85 0.152 1.61E-05 Xe-131m 0.070 9.15E-05 Xe-133 0.777 2.94E-04 Fraction of nuclide in mix as function of time (see Table 5.2-1). i Dose factors from Table 1.1-10. i 17 /1/% Revision Date a m s. A-6 h

l EXAMPLE CALCULATION NO. 3 (Continued) Calculation The dose rate is calculated from Equations (3-5) and (3-28): Atbs - 0.61 { OfI DFB i 3 (mrem) g pCi-sec ) ( Ci ) ( mrem-m pCi-yr ) 3 sec yr Ci-m and where the stack release rate is determined from: Df 1 (3-28) ST d M F i sat sg go 1 ( pCi ) (cpm) ( pCi/cc) ( cc ) sec cpm sec However, for a time (t) af ter shutdown, the ratio of of" to the sum release rate of all noble gases can be replaced in Equation (3-28) by the relative fraction [f (t)] of each noble gas available in the system: therefore, g Equation (3-28) can be written: ofI - fj(t) M 1F Sg Therefore, using the above data for a time period 32 days af ter shutdown, the dose rate equation can also be written as: Atbs - 0.61 M 1F { f (t) DFB 3 i Sg g = 0.61 80,000 1/1 E +08 8.26E+07 [(0.152)(1.61E-05) + (0.070)(9.15E-05) + (0.777)(2.94E-04)] - Answer 9.6 mrem / year noble gas total body dose rates at 32 days after l A tbs shutdown. Revision 17 Date 5 /31 /04 A-7 unn

EXAMPLE CALCULATION NO. 4 TyDe Skin Dose Rate From Noble Gases References a) ODCM Section 3.5 (Method I). b) Technical Specification 3.8.E.1.a. Problem Calculate the off-site skin dose rate resulting from the release of noble I gases from the plant stack during power operations. j Plant Data a) Maximum plant stack gas monitor (I or II) Count rate during period of interest (M): 80.000 cpm b) Stack flow rate during release (F): 8.26E+07 cc/sec (175.000 cfm x 4.72E+02 cc/sec, ) cfm c) Plant stack monitor detector counting IE+08 cpm per efficiency (Sg): pCi/cc d) The last measured release rate mix of noblegasfromtheSJAE(Of*). and corresponding dose factor DF' j from Table 1.1-10. Of 0 F'j, 4 m m m-sec/d i m ) (ttCi /sec) Xe-138 5.15E+03 1.06E-02 Kr-87 2.37E+02 1.43E-02 Kr-88 1.29E+02 1.28E-02 Xe-135 1.85E+02 3.24E-03 Revision 17 Date 5/31/94 unu A-8

EXAMPl.E CALCULATION N0. 4 (Continued) Calculation The skin dose rate is calculated from Equations (3-7) and (3-28): A uns "E 0 0F,', (3-7) s 1 1 ' mrem' ' Ci ' fmrem-s ec ' yr

sec,

( C i -y r t j and where the stack release rate is determined from: Df 1 (3-28) 6,si M F sJAE $9 1 ( pCi ) (cpm) ( Ci/cc) ( cc ) sec cpm sec First, determine the sum (E ) of all fAE and the fraction that each noble i gas i represents in the total gas mix. E6fJAE - (5.15E+03) + (2.37E+02) + (1.29E+02) + (1.85E+02) 1 - 5.70E+03 pCi/sec and the relative fraction of each noble gas: j, OpJAE /5.70E+03 Relative raction of Total 0.904 Xe-138 5.15E+03/5.70E+03 0.042 Kr-87 2.37E+02/5.70E+03 0.023 Kr-88 1.29E+02/5.70E+03 0.032 Xe-135 1.85E+02/5.70E+03 Revision 17 Date 5/31 /94 A-9 R12\\84

EXAMPLE CALCULATION NO. 4 (Continued) Next, the stack release rate of each noble gas i from Equation (3-28) can be substituted into Equation (3-5) to give the skin dose rate as: Askins - M F f 0 F's, i - 80.000 1/IE48 8.26E+07 [(0.904)(1.06E-02) + (0.042)(1.43E-02) + (0.023)(1.28E-02) + (0.032)(3.24E-03)] - l - 6.61E+04 (9.58E-03 + 6.01E-04 + 2.94E-04 + 1.04E-04) - 6.61E+04 (1.06E-02) Answer k - 699 mrem / year noble gas skin dose rate. skins l l I 17 5/31/94 Revision Date unn A-10

EXAMPLE CALCULATION NO. 5 Type Skin Dose Rate From Noble Gases References a) ODCH Section 3.5 (Method 1). b) Technical Specification 3.8.E.1.a. Problem Calculate the off-site skin dose rate resulting from the release of noble gases from the plant stack six days after plant shutdown. Plant Data a) Maximum plant stack gas monitor (I or II) Count rate during period of interest (M): 120,000 cpm b) Stack flow rate during release (F): 8.26E+07 cc/sec cc/sec _ ) (175,000 cfm x 4.72E+02 cfm c) Plant stack monitor detector counting 17-08 cpm per efficiency (59): Li/cc d) Since the plant is shut down for more than five days, Xe-133 may be used as the referenced radionuclide in place of the ratio of DsJAE to the sum i of all 0,5

in Equation (3-28).

D F'j, mrem-sec i fi (t >5 days) Ci-y r Xe-133 1. 5.58E-04 l 17 5/31/94 Revision Date A-11 ams.

EXAMPLE CALCULATION NO. 5 (Continued) Calculation The skin dose rate is calculated from Equations (3-7) and (3-28): kskins " b b si Of'i s (3-7) i i g mrem) g Ci ) ( mrem-sec) yr sec pCi-y r and, the stack release rate is determined from: sJAE oft, gi I M F gsJAE 5 (3-28) 1 However, for times greater than five days after shutdown, Xe-133 may be used as the referenced radionuclide alone. Therefore, in Equation (3-28) the ratio of OfJAE to the sum of all OfJAE can be replaced by a value of I which indicates that all tne contribution to the release is from Xe-133. Therefore: Of,T 133 - 1.0 x 120,000 x 1/1E+08 x 8.26E47 l 'pCi/cc) (cpm) (cc/sec) cpm J ) of,I_333 - 99,120 Ci/sec Therefore,replacingthisvalueofOfT into Equation (3-7) we find the skin dose rate as: k3 tins 99,120 x

5. 58E-04 l

' mrem' f Ci ' ' mrem-s ec ' yr,

sec, Ci -y r Revision 17 5/31/94 Date uns4 A-12

EXAMPLE CALCULATION NO. 5 (Continued) Answer l A uns - 55.3 mrem / year. s l l ( i l 1 I i I Revision 17 Date 5/31/04 A-13 ons4

EXAMPLE PROBLEM NO. 6 Type Critical Organ Dose Rate From Iodine Tritium, and Particulates References a) ODCM Section 3.6 (Method I). b) Technical Specification 3.8.E.1.b. Problem Calculate the critical organ dose rate due to measured effluent data taken from the plant stack for a seven-day sample collection period. Plant Data a) Stack particulate analysis for the seven-day period of interest. m O FG',, co Activi ty 0 sTP 1 j ( mrem-sec ) l ( Ci-sec) yr-pCi l l Sr-89* 1.42E-04* 3.60E+02 Sr-90* 3.50E-03* 1.36E+04 Co-60 4.89E-02 3.41E+02 l Cs-137 3.90E-03 5.55E+02 Zn-65 1.01E-02 1.20E+02 Na-24** 2.76E-03** Mn-54+ <2.87E-06+ 2.77E+01 Notes m D FG',, co dose rate factor for each radionuclide is taken from Table 1.1-12. For Sr-89/90, use the most recent'available measurement from quarters composite analysis. Na-24 has a half life of less than 8-1/2 days, and therefore is not included in the dose analysis per requirements of Technical Specification 3.8.E.1.b even though it was detected. Revision 17 Date 5/31/94 unu A-14

l I l EXAMPLE PROBLEM NO. 6 (Continued) Mn-54 is not included in the dose analysis since it was not detected-as + being present af ter counting to at least the LLD. b) Stack iodine (charcoal and particulate activities combined for the seven-day period of interest): l m O FG',j co Activity 0sTP 1 ( mrem-sec ) l i ( Ci-sec) y r-pCi l I-131 1.16E-03 2.43E+03 I-133* <6.35E-05* 2.59E+01 I-135** 7.21E-03** and H-3+ 3.17E-02 9.87E-03 Notes I-133 is not included in the dose analysis for this case since it was not detected is being present in the stack analysis. 1 I-135 is not included in the dose analysis because it has a half life less than 8-1/2 for particulates, and is not included as a required iodine in Technical Specification 3.8.E.1.b. + Tritium value based as latest available stack grab sample. Calculation The dose rate is calculated from Equation (3-16): A -E Q" D FG'sico (3-16) cos i 1 (mrem) ( pCi ) (mrem-sec) yr sec Ci-y r l l The dose rate factors (DFG',4c, ) for each of the radionuclides detected in the plant stack charcoal and particulate filter sample (plus tritium) is taken from Table 1.1-12 of the ODCM. Revision 17 Date 5/31/94 unn A-15

EXAMPLE PROBLEM NO. 6 (Continued) Therefore: A,- (1.42E-04)(3.60E+02) + (3.50E-03)(1.36E+04) + (4.89E-02) co (3.41E+02) + (3.90E-03)(5.55E+02) + (1.01E-02)(1.20E+02) + (1.16E-03)(2.43E+03) + (3.17E-02)(9.87E-03) - Answer ft - 70.5 mrem / year critical organ dose rate from iodine, tritium and l cos particulate. A Revision 17 Date 5/31 /04 airss4 A-16

1 EXAMPLE PROBLEM NO. 7 Type Gamma Air Dose from Noble Gases released from stack References a) ODCH Section 3.7 (Method I). b) Technical Specification 3.8.F.1. Problem Calculate the maximum gamma air dose resulting from noble gases released from the plant stack over a calendar month. Plant Data Based on the daily off-gas analysis, the total activity released during the month of interest is: si DF[* Activity 0i 3 (mrad-m (Cl) pCi-yr) Kr-88 3.55E-01 1.52E-02 Kr-85m 4.71E+00 1.23E-03 Xe-138 2.75E+00 9.21E-03 Xe-135 3.51E+01 1.92E-03 Xe-133 9.42E+01 3.53E-04

Gamma air dose f actors taken from Table 1.1-10.

Revision 17 Date 5/31/94 A-17 anu4

l 1 EXAMPLE PROBLEM NO. 7 (Continued) I Calculation The maximum gamma air dose off-site is calculated from Equation (3-21): sT (3-21) l D 1rs - 0.019 Q DF[ i 3 (mrad) ( pCi-yr ) (Ci) ( mrad-m ) 3 pCi-y r Ci-m Therefore: l l DJirs - 0.019 [(3.55E-01)(1.52E-02) + (4.71E+00)(1.23E-03) l l + (2.75E+00)(9.21E-03) + (3.51E+01)(1.92E-03) + (9.42E+01)(3.52E-04)] 0.019 (5.40E-03 + 5.79E-03 + 2.53E-02 + 6.74E-02 + l 3.31E-02) Answer DJir3 2.60E-03 mrad gamma air dose during the month. l Revision 17 Date 5 /11/o/i unu A-18

EXAMPLE CALCULATION NO. 8 Type Beta Air Dose from Noble Gases released from stack References a) ODCM Section 3.8 (Method I). b) Technical Specification 3.8.F.1. Problem Calculate the maximum beta air dose resulting from the same noble gas releases 1 l given in Example Calculation No. 7. Plant Data j ) from Example No. 7. the total activity determined to be released during the month is: P sT DFi Activity 0i 9 3 (mrad-m (Ci) pCi-yr) Kr-88 3.55E-01 2.93E-03 l f Kr-85m 4.71E+00 1.97E-03 Xe-138 2.75E+00 4.75E 03 Xe-135 3.51E+01 2.46E-03 Xe-133 9.42E+01 1.05E-03 l

Beta air dose factors taken from Table 1.1-10.

l l I ) l j Revision 17 Date 5/31/94 A-19 unu

-. =-. EXAMPLE CALCULATION NO. 8 (Continued) Calculation The maximum beta air dose off-site is calculated from Equation (3-23): Dfirs - 0.033 E oft DFf (3-23) l t 3 (mrad) { pCi-yr ) (C1) ( mrad-m ) 3 pCi-y r Ci-m Therefore: Ofirs - 0.033 [(3.55E-01)(2.93E-03) + (4.71E+00)(1.97E-03) l + (2.75E+00)(4.75E-03) + (3.51E+01)(2.46E-03) + (9.42E+01)(1.05E-03)] 0.033 (1.04E-03 + 9.28E-03 + 1.31E-02 + 8.63E-02 + l 9.89E-02) Answer D trs - 6.88E-03 mrad beta air dose during the month. l Revision 17 Date 5/31/94 uns4 A-20

EXAMPLE PROBLEM NO. 9 Type Critical Organ Dose From Iodine, Tritium, and Particulates i References a) ODCM Section 3.9 (Method I). ] b) Technical Specification 3.8.G.I. Problem Calculate the critical organ dose due to the total activity recorded as being released from the plant stack during a calendar month. 1 I Plant Data a) From the combined stack analyses during the month, the following activity released is: m DFG,co STP 3 0 l 1 (mrem) i (Ci) Ci Sr-89* 5.42E-04* 1.14E+01 Sr-90* 1.10E-02* 4.31E+02 Co-60 2.30E-01 7.63E+00 Cs-137 1.15E-02 1.63E+01-f Zn-65 2.60E-02 3.71E+00 Na-24** 7.11 E - 03** Mn-54 <2.76E+06+ 7.01E-01 I Notes for Plant Data a) Above m Critical organ dose factor taken from Table 1.1-12. For Sr-89/90, use the most recent available measurement from the quarterly composite analysis. Revision 17 Date 5/31/94 A-21 ama.

l EXAMPLE PROBLEM NO. 9 (Continued) Na-24 has a half life of less than 8-1/2 days, and therefore is not -. included in accordance with Technical Specification 3.8.G.I. + Hn-54 is not included in the dose analysis since it was not detected as being present after counting to at least the LLD. b) Total iodine release for the month based on the combined charcoal and particulate filter samples taken during the month: sTP DFG,,,, i i g mrem) (Ci) Ci 1-131 4.30E-03 7.71E+01 1-133* 1.12E-04* 8.21E-01 1-135** 2.01E-02** and H-3+ 0.15 3.13E-04 Calculation The dose is calculated from Equation (3-25): 0" DFG ,c, (3-25) 3 D E 4 3 eos 1 (mrem) (C1) (mrem /Ci) Notes for Plant Data b) Abov,e In this case.1-133 was found in one of the weekly stack samples to be present, and therefore based on that value is included in the dose analysis. I-135 11 not included in the dose analysis because it has a half life less than 8-1/2 days for particulates and is not included as a required iodine in Technical Specification 3.8.8.1. + Tritium value based on the monthly stack grab sample. Revision 17 Date 5/31/94 nas4 A-22 1 ~ --a

EXAMPLE CALCULATION NO. 10 i Type Releases Limited to 20 times the effluent concentration limits of 10CFR20.1001-20.2401, Table 2, Column 1 References a) 10CFR50.72 b) 10CFR50.73 Problem find the minimum stack gas monitor response which would require an assessment to determine if a four-hour notification to NRC is required per 10CFR50.72. Assumptions a) Maximum expected stack flow rate (F) 8.26E+07 cc/sec cc/sec, ) (175,000 cfm x 4.72E+02 cfm b) Plant stack monitor detector 1E+08 counting efficiency (Sg) cpm per pCi/cc c) Maximum off-site ground level 1.04-06 l 3 dispersion parameter (X/0 undepleted) sec/m (from ODCM Table 3.10-1) d) Most restrictive ECL value for noble 9E-09 gases (10CFR20, Appendix 8. Table 2 pCi/cc Column 1) for Kr-88) Calculation (Part I) for the stack monitor which would correspond to an The setpoint Rst instantaneous off-site air concentration of 20 x ELC for the most restrictive noble gas can be calculated by: 1 1 R - 20 ECL Sg 1E+06 sPt X/0 F 3 (cpm) (pCi) ( cpm ) (m ) ( cc ) (sec) 3 cc Ci/cc sec m cc - 20 x (9E-09)(1E+08)(1/1.04E-06)(1E+06)(1/8.26E+07) R - 209,536 cpm setpoint alarm value spt Revision 17 Date 5/31/94 A-24 a m s4

i APPENDIX ! RADI0 ACTIVE LIQUID, GASEOUS, AND SOLID WASTE TREATMENT SYSTEMS Reauirement: Technical Specification 6.14.A requires that licensee initiated major changes to the radioactive waste systems (liquid, gaseous, and solid) be reported to the Commission in the Semiannual Radioactive Effluent Release Report for the period in which the evaluation was reviewed by the Plant Operation Review Committee. l l

Response

There were no licensee initiated major changes to the radioactive waste systems (liquid, gaseous, and solid) during j this reporting period. 1 I anac 1-1

APPENDIX J ON-SITE DISPOSAL OF SEPTIC WASTE Requirement: Off-Site Dose Calculational hanual, Appendix B requires that the dose impact due to on site disposal of septic waste during the reporting year and from previous years be reported.to the Commission in the Semiannual Radioactive Effluent Report filed after January 1. if disposals occur during the reporting year.

Response

No response is required for this reporting period. l i i l l a m ra J-1 . _}}

ML20072N038

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