Text

Effluent & Waste Disposal Semiannual Rept for Third & Fourth Quarters 1986
	ML20212C860
Person / Time
Site:	Vermont Yankee
Issue date:	12/31/1986
From:	; VERMONT YANKEE NUCLEAR POWER CORP.
To:	;
Shared Package
ML20212C837	List: ML20212C836; ML20212C860;
References
	NUDOCS 8703040017
	Download: ML20212C860 (143)
	v • d • e

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I EFFLUENT AND WASTE DISPOSAL I SEMIANNUAL REPORT FOR THIRD AND FOURTH QUARTERS, 1986 I l I ;

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I G703040017 B70301 PDR ADOCK 05000271 R PDR

I TABLE 1A I,

Vermont Yankee Effluent and Waste Disposal Semiannual Report Third and Fourth Quarters, 1986 Gaseous Effluents - Summation of All Releases t Unit Quarter Quarter Est. Total A. Fission and Activation Gases

1. Total release Ci ND ND i 2. Average release rate for period uCi/sec ND ND

3. Percent of Tech. Spec. limit (1) % - -

I B. Iodines i 1. Total Iodine-131

2. Average release rate for period

3. Percent of Tech. Spec. limit (2)

Ci uCi/see ND ND

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4.35E-05 5.54E-06

+

I C. Particulates

1. Particulates with T-1/2 > 8 days Ci 9.89E-04

i 2. Average release rate for period uCi/see 1.26E-04 *

3. Percent of Tech. Spec. limit % -

(3) (3)

4. Gross alpha radioactivity Ci 2.44E-06

3 D. Tritium

1. Total release Ci 8.26E-01 1.27E+00

2. Average release rate for period uCi/see 1.05E-01 1.62E-01

3. Percent of Tech. Spec. limit % (3) (3)

+ Percent of Technical Specification limits will be provided in the supplemental report which will include the annual dose summary.

(1) Technical Specification 3.8.F.1.a for gamma air dose. Percent values for Technical Specification 3.8.F.1.a for beta air dose are approximately the same.

I (2) Technical Specification 3.8.G.1 for dose from I-131. I-133, Tritium, and radionuclides in particulate form.

I (3) Per Technical Specification 3.8.G.1, dose contribution from Tritium and particulates are included with I-131 above in Part B.

ND - Not Detected.

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- Awaiting outside laboratory analysis results.

I TABLE IB r

[ Vermont Yankee Effluent and Waste Disposal Semiannual Report Third and Fourth Quarters 1986 Gaseous Effluents - Elevated Release (1)

Continuous Mode Batch Mode Nuclides Released Unit Quarter Quarter Quarter Quarter 3 4 3 4

1. Fission Gases 5 Krypton-85 Ci ND ND Krypton-85m Ci ND ND Krypton-87 Ci ND ND II Krypton-88 Ci ND ND Xenon-133 Ci ND ND Xenon-135 Ci ND ND

,I Xenon-135m Xenon-138 Ci Ci ND ND ND ND l

Ci Unidentified Ci

'a Total for period Ci ND ND i

2. Iodines E Iodine-131 Ci ND 4.35E-05 l Iodine-133 Ci 8.21E-04 1.99E-04 Iodine-135 Ci ND 1.80E-03 Total for period Ci 8.21E-04 2.04F-03

3. Particulates Strontium-89 Ci 3.06E-05

Strontium-90 Ci 2.87E-07

Cesium-134 Ci ND ND Cesium-137 Ci 4.67E-05 7.14E-05 Barium-Lanthanum-140 Ci ND ND i Manganese-54 Ci 7.43E-05 6.31E-05 Cobalt-60 Ci 7.17E-04 5.01E-03 I Zine-65 Unidentified Ci Ci 1.51E-04 2.59E-04 (1) There were no batch mode gaseous releases for this reporting period.

ND - Not detected in the off-gas mix.

- Awaiting outside laboratory analysis results.

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I TABLE IC I Vermont Yankee Effluent and Waste Disposal Semiannual Report Third and Fourth Quarters 1986_

Gaseous Effluents - Ground Level Releases I There were no routine measured ground level continuous or batch mode gaseous releases during the third or fourth quarters of 1986.

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TABLE ID Vermont Yankee i

Effluent and Waste Disposal Semiannual Report Third and Fourth Quarters 1986 Caseous Effluents - Nonroutine Releases i There were no nonrontine or accidental gaseous releases during the third or fourth quarters of 1986.

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I TABLE 2A Vermont Yankee Effluent and Waste Disposal Semiannual Report Third and Fourth Quarters 1986 Liquid Effluents - Nonroutine Releases I There were no liquid releases during the third or fourth quarters of 1986.

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TABLE 2B Vermont Yankee Effluent and Waste Disposal Semiannual Report Third and Fourth Quarters 1986 Liquid Effluents - Nonroutine Releases

l There were no nonroutine or accidental releases during the third or fourth quarters l

I of 1986.

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TABLE 3 Vermont Yankee v

m Effluent and Waste Disposal Semiannual Report .

Third and Fourth Quarters, 1986 Solid Waste and Irradiated Fuel Shipments .,

A. Solid Waste Shipped Off-Site for Burial or Disposal (Not Irradiated Fuel) -

H Unit 6-Month Est. Total Period Error, %

I 1. Type of Waste m3 h3

a. Spent resins, filter sludges, evaporator bottoms, etc. Ci 2.27E+01 1.78E+01 27.50E+01

b. Dry compressible waste, contaminated m3 2.94E+01 .

Ci 6.00E+00 e7.50E+01 I

equipment, etc.

c. Irradiated components, control rods, etc. (Safe ends) m3 Ci 2.24E+00 6.94E+01 27.50E+01 ]

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

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l a. Cesium-137 % 3.55E+01 b. (Cont'd) Cesium-137 % 2.60E-01 l

I Cobalt-60

. Zinc-65 Manganese-54

% 2.49E+01

% 2.15E+01

% 6.08E+00

c. Cesium-137 Cobalt-60 Iron-55

% 4.10E+01

% 3.41E+01

% 1.29E+01 Iron-55 % 5.78E+00 Zine-65 % 5.58E+00

/ Cesium-134 % 1.15E+00 Manganese-54 % 2.49E+00 Nickel-63 % 1.70E+00 Cesium-134 % 2.35E+00 Chromium-51 % 1.40E+00 Plutonium-241 % 2.10E-01 i

j Hydrogen-3 Carbon-14

b. Iron-55 Cobalt-60

% 8.30E-01

% 1.80E-01

% 7.31E+01

% 2.18E+01 Strontium-90 % 2.10E-01 1 Manganese-54 Zine-65

% 2.76E+00

% 1.93E+00

3. Solid Waste Disposition Number of Shipments Mode of Transportation Destination 7 Truck Barnwell, SC ,

B. Irradiated Fuel Shipments (Disposition): None C. Supplemental information

1) Class of solid waste containers shipped: 2A (unstable), 3A, 1B, 1C

2) Types of containers used: 6 Type A, 1 Strong-tight-container i 3) Solidification agent or absorbant: None 4436R ,

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l APPENDIX A I .

I EFFLUENT AND WASTE DISPOSAL SEMIANNUAL REPORT l

Supplemental Information Third and Fourth Quarters, 1986 Facility: Vermont Yankee Nuclear Power Station I Licensee: Vermont Yankee Nuclear Power Corporation

~ 1A. Technical Specification Limits - Dose and Dose Rate Technical Specification and Category Limit I a. Noble Gases 3.8.E.1 Tctal 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

c. Liquids 3.8.B.1 1.5 mrem in a quarter i

Total body dose 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 I A-1 4436R

q 2A. Technical Specification Limits - Concentration

{ Technical Specification and Category Limit

a. Noble Cases No MPC limits

b. Iodine-131.

o Iodine-133. Tritium and Radionuclides in Particulate Form With Half-Lives Greater Than 8 Days: No MPC limits l

c. Liquids l

3.8.A.1 Total fraction of MPC excluding noble gases (10CFR20, Appendix B, as Table II, Column 2): 11.0 3.8.A.1 Total noble gas concentration: 12E-04 uCi/cc

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

a. Average gamma energy: 3rd Quarter 1.09E+00 MeV/ dis 4th Quarter 1.22E+00 MeV/ dis

b. Average beta energy: Not Applicable

4. Measurements and Approximations of Total Radioactivity Provided below are the me'. hods used to measure or approximate the total radioactivity in effluents and the methods used to determine radionuclide composition.

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a. Fission and Activation Gases I Daily samples are drawn at the discharge of the air ejector.

Isotopic breakdown of the releases are determined from these samples. A logarithmic chart of the stack gas monitor is read I daily to determine the gross release rate. At the very low release rates normally encountered during operation with the augmented ,

off-gas system the error of release rates may be approximately 100 percent.

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b. Iodines Continuous isokinetic samples are drawn from the plant stack through a particulate filter and charcoal cartridge. The filters I and cartridge are removed weekly (if releases are less than 4 percent of the Tech Spec limit), or daily (if they are greater than 4 percent of the limit), and are analyzed for radioiodine 131, 132, 133, 134, and 135. The iodines found on the filter are added to those on the charcoal cartridge. The error involved in these steps may be approximately 50 percent.

.e T c. Particulates The particulate filters described in b. above are also counted for particulate radioactivity. The error involved in this sample is also approximately 150 percent. g. '

d. Liquid Effluents I

Radioactive liquid effluents released from the facility are continuously monitored. Measurements are also made on a i

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 I ~

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[ 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 analyzed for I

gross gamma and gamma isotopic radioactivity. A monthly proportional composite sample, comprising an aliquot of each batch released during a month, is also analyzed for tritium, SR-89, SR-90, gross beta and gross alpha radioactivity, in addition to L gamma spectroscopy.

[ There were no liquid releases during the reporting period.

S. 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 1 period.

6. Abnormal Releases

a. Liquid i

There were no nonroutine liquid releases during the reporting period.

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

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B APPENDIX B I

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LIQUID HOLDUP TANKS Requirement: Technical Specification 3.8.D.1 limits the quantity of I

radioactive material contained in any outside tank. With the quantity of radioactive material in any outside tank exceeding the limits of Technical Specification 3.8.D.1, a description of the events leading to this cot.dition is required in the next Semiannual Effluent Release Report per Technical Specification 6.7.C.1.

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

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APPENDIX C I RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION I Requirement: Radioactive liquid effluent monitoring instrumentation channels are required to be operable in accordance with Technical I 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 l.

I Response: Since the requirements of Technical Specification Table 3.9.1 governing the operability of radioactive liquid effluent monitoring instrumentation were met for this reporting period, i no response is required.

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APPENDIX D r

RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION d

Requirement: Radioactive gaseous effluent monitoring instrumentation channels are required to be operable in accordance with Technical l Specification Table 3.9.2. If inoperable gaseous effluent monitoring instrumen*.ation is not returned to operable status within 30 days pursuant to Note 5 of Table 3.9.2, an explanation L

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.l.

Response: Since the requirements of Technical Specification Table 3.9.2 governing the operability of radioactive gaseous effluent f monitoring instrumentation were met for this reporting period, no response is required.

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a APPENDIX E s

RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM Requirement: 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 L included in the next Semiannual Effluent Release Report:

(1) identify the cause(s) of the sample (s) no longer being L available, (2) identify the new location (s) for obtaining available replacement samples and (3) include revised ODCM figure (s) and table (s) reflecting the new location (s).

7 Response: All required milk samples were available during this reporting period. Changes made to the milk sampling program and ODCM, as pat- of the Land Use Census, are described in Appendix H.

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I APPENDIX F LAND USE CENSUS I Requirement: 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 I Technical Specification 4.8.G.1, Technical Specification 6.7.C.1 requires the identification of the new location (s) in the next I Semiannual Effluent Release Report.

Response: No locations were identified by the 1986 land use census that would yield at least a 20 percent greater dose or dose commitment than the values currently being calculated pursuant to Technical Specification 4.8.G.I.

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I APPENDIX G I

PROCESS CONTROL PROGRAM Requirement: 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.

I Response: There was no licensee initiated change (s) to the Process Control Program during this reporting period.

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I APPENDIX H I 0FF-SITE DOSE CALCULATION MANUAL I Requirement: Technical Specification 6.13.A.1 requires that licensee initiated changes to the Off-Site Dose Calculation Manual (ODCM)

I be submitted to the Commission in the Semiannual Radioactive Effluent Release Report for the period in which the change (s) was made effective.

Response: There were three licensee initiated change (s) to the Off-Site Dose Calculation Manual during this reporting period.

I 1. The changes proposed and implemented for Section 4.0 (Environmental Monitoring) of the ODCM in the Semiannual I Radioactive Effluent Release Report for the first and second quarters of 1986 were approved as Revision 1 to the ODCM during the reporting period for this report. These changes were as follows:

I a. Milk sampling locations TM-11 (Miller Farm 0.8 km, WNW) and TM-13 (Newton Farm, 5.1 km, SSE were replaced with TM-14 I (Brown Farm, 2.1 km SSW) and TM-15 (Coombs Farm, 4.7 km, NW) based on a dosimetric analysis done pursuant to Technical I Specification 3.9.D.2 (Land Use Census);

b. Milk sampling location TM-12 (Whitaker Farm, 2.6 km, S) was also replaced with location TM-13 (Newton Farm, 5.1 km, SSE) when IM-12 went out of business. (The choice of TM-13 as a replacement was based on the dosimetric ranking of available I locations done as part of the Land Use Census);

c. Milk sampling location TM-11 (Miller Farm, 0.8 km WNW) and air sampling location AP/CF-15 (Tyler Hill Road, 3.2 km, WNW) were also added to the ODCM as non-Technical Specification locations.

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L Based on the 1986 Land Use Census, the required milk sampling locations were identified as TM-14 (Brown Farm, 2.1 km, SSW),

TM-15 (Coombs Farm, 4.7 km, NW), and TM-16 (Tall Oaks Farm, 4.7 km, WNW), (TM-21, Moore Farm, remains the control f location). Location TM-16 was, therefore, added to Section 4.0 as part of Revision 1, along with the other changes discussed above.

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4 2. Additional pages of text were also added to Section 4.0, as

( Revision 2, to identify the Intercomparison Program which is being participated in and discusses the basis of the Airborne

{ Pathway Monitoring Program.

3. The ODCM was also revised in Revision 3 to reflect additional liquid dose pathways, and the clarification of existing basic assumptions involved in gaseous dose and dose rate calculations. These changes are identified as follows:

a. Section 1 of the ODCM has been expanded to clarify the specific plant and corporate office responsibility assignments and requirements for making changes to the'0DCM.

It also includes corrected Table of Contents and summary tables of dose equations, variable definitions, and dose factors necessary to reflect changes made in Chapters 3 and I 5. It should be noted that new liquid dose factors (Table 1.1-11) have been included resulting from the recent identification of additional exposure pathways not considered in the original ODCM.

I b. Sections 3.1, 3.4, 3.5, and 3.6 have been reworded to clarify the concepts of dose and dose rate as used in the ODCM for implementation of the Technical Specification requirements.

It should also be noted that in addition to the rewording of I

l the explanation of where the noble gas dose rate equations apply, the dose rate equations have also been modified to change the assumption of credit taken from residential i _

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I building shielding factor (S p) from the value of 0.7 to 1.0. This additional conservatism to be used in the dose rate calculations is applied on an instantaneous basis since no assurances can be given that the critical receptor will be inside residential structures at all times. For accumulated doses, the use of a 0.7 shielding factor is still applicable.

c. Sections 3.2 and 3.3 (liquid doses) identifies changes in I dose factors to reflect the inclusion of additional exposure pathways not previously identified. These new pathways include 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 irtigated land, and the direct exposure from the ground plane associated with radioactivity deposited by the water pathway.

I d. Sections 3.7 and 3.8 have been expanded to clarify concerns where beta and gamma air doses from noble gases are being calculated based on SJAE nuclide analysis for the determination of radionuclide mix ratios. The expanded Sections 3.7 and 3.8 identifies that the dose calculations should not use the SJAE noble gas release rate determinations directly for determination of off-site noble gas releases as this leads to overly conservative dose calculations which have no direct bearing on actual off-site doses. The I radionuclide mix ratios at the SJAE should be used with the plant stack gross release rate (in lieu of detactable gas mixes at the stack) to determine the off-site doses. During periods when the plant is shut down, alternate assumptions on where to obtain radionuclide release mix ratios are provided such that off-site doses being calculated are conservative, but still consistent with the gas inventory potentially I available for release.

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e. Section 3.11 has been expanded to address the determination of direct dose off-site from other than N-16 scatter sources in the turbine hall. Fixed sources of direct radiation, such as on-site solid waste storage, is also required to be considered.

I f. Section 5.2 addresses gaseous effluent monitor setpoint methodology and has been modified to reflect the change in the credit taken for residential building shielding factor as noted above in Item 6.

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g. A new Appendix A is added to the end of the ODCM to provide example calculations of Method I dose equations for reference and guidance information.

I None of the above changes will reduce the accuracy or reliability of dose calculations or setpoint determinations.

Revised ODCM pages reflecting the above changes are included as part of this appendix.

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I VERMONT YANKEE NUCLEAR POWER STATION OFF-SITE DOSE CALCULATION MANUAL REV. #1 I

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Reviened VJ V f d 2 ne ve.,-c.n / toi?h; Plant Ope'fational Review Date Committee I

Approved A 3 / /0// 6/86

[/ Plant Manager Dhte '

Approved [ M1 Jb / //[7[/b V e Pr'es le and dat'e nager rations i

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LIST OF AFFECTED PAGES '

~I Changes, deletions or additions in the most recent revision are indicated by a bar in

, the margin or by a dot near the page number if the entire page is affected.

Pace Revision Date

, i - 4.1 0 3/1/84 l 4.2 - 4.2a 1 10/23/86 4.3 - 4.4 0 3/1/84 4.5 1 10/23/86

'5 4.6 - 6.11 0 3/1/84

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l Table 4.1 L

Radioloaical Environmental Monitorina Stations

f Distance From Exposure Pathway Sample Location the Plant Direction From and/or Sample and Desianated Code ** Stack (km) the Plant

1. AIRBORNE (Radioiodine and Particulate)

L 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

{ *1AP/CF-15 Tyler Hill Rd.

AP/CF-21 Spofford Lake 3.2 16.1 WNW 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 i WG-21 Brattleboro CC 12.1 NNW

c. Sediment SE-11 Shorline Downriver 0.8 On-site I From Shorline SE-12 North Storm ***

Drain Outfall 0.15 On-site

3. INGESTION

a. Milk *1TM-11 Miller Farm 0.8 WNW I TM-14

1TM-13 Brown Farm Newton Farm 2.6 5.1 S

S3E TM-15 Coombs Farm 4.5 NW TM-16 Tall Oaks Farm 4.7 WNW TM-21 Moore Farm 15.9 N

b. Mixed TG-11 1.9 I Grasses TG-12 River Station No.3.3 N. Hinsdale, NH 3.6 SSE NNE TG-13 Hinsdale Substation 3.1 E I TG-14

1TG-15 Northfield, MA Tyler Hill Rd.

11.3 3.2 SSE WNW TG-21 Spofford Lake 16.1 NNE 4-2 Revision 1 DATE: 7 b Approved by: Md/> k g

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Table 4.1 (Continued)

Radioloaical Environmental Monitorina Stations

I Exposure Pathway and/or Sample Sample Location and Desianated Code **

Distance From the Plant Stack (km)

Direction From the Plant iI c. Silage *1TC-11 Miller Farm 0.8 WNW TC-14 Brown Farm 2.6 S

1TC-13 Newton Farm 5.1 SSE TC-15 Coombs Farm 4.5 NW TC-10 Tall Oaks Farm 4.7 WNW TC-21 Moore Farm 15.9 N l d. Fish FH-11 Vernon Pond --

On-site l

FH-21 Rt. 9 Bridge 12.8 Upriver

1 Non-Tech Spec Station i

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4-2a Revision 1 Date: hd Approved by: 8/w,0 t UV I

I Table 4.1 (continued)

Radiological Environmental Monitoring Stations

I Exposure Pathway and/or Samole Sample Location and Desianated Code Distance From the Plant Stack (km)

Direction From the Plant

4. DIRECT RADIATION DR-1 River Station 1.9 SSE No. 3.3 I DR-2 N. Hinsdale, NH 3.6 NNW DR-3 Hinsdale Substation 3.1 E 11.3 I DR-4 Northfield, MA SSE DR-5 Spofford Lake 16.1 NNE DR-6 Vernon School 0.6 SW DR-7 Site Boundary 0.32 SSW DR-8 Site Boundary 0.45 5 I DR-9 Inner Ring 1.8 N DR-10 Outer Ring 4.3 N DR-11 Inner Ring 1.8 NNE I DR-12 Outer Ring 3.4 NNE DR-13 Inner Ring 1.2 NE I DR-14 Outer Ring 4.2 NE l

DR-15 Inner Ring 1.4 ENE DR-16 Outer Ring 2.9 ENE DR-17 Inner Ring 1.3 E I DR-18 DR-19 DR-20 Outer Ring Inner Ring Outer Ring 3.1 3.6 5.5 E

ESE ESE pm DR-21 Inner Ring 2.1 SE g DR-22 DR-23 Outer Ring Inner Ring 3.5 2.1 SE SSE DR-24 Outer Ring 4.2 SSE DR-25 Inner Ring 2.3 S DR-26 Outer Ring 4.0 S DR-27 Inner Ring 1.2 SSW Outer Ring 2.4 I DR-28 SSW DR-29 Inner Ring 0.8 SW DR-30 Outer Ring 2.4 SW DR-31 Inner Ring 0.8 WSW DR-32 Outer Ring 5.0 WSW 5 DR-33 Inner Ring 0.8 W DR-34 Outer Ring 4.8 W I DR-35 DR-36 DR-37 Inner Ring Outer Ring Inner Ring 1.2 4.5 2.7 WNW WNW NW DR-38 Outer Ring 7.4 NW I DR-39 DR-40 Inner Ring Outer Ring 2.9 4.8 NNW NNW I

Sample locations are shown on Figures 4.1 to 4.6.

- Station 1Xs are indicator stations and Station 2Xs are control stations (for all but the Direct Radiation stations).

- - To be sampled and analyzed semi-annually.

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SEE ENtAMENENT IN FIGURE 4-1 I lVERNONDAM L_- _ _ _.

41-12 ' N* *

r VERNON, V.T. \ WR-ll 2n-ll w .. g AP M -ll Tr.. a 4 te-13 TM-13 0 1 2 3 w_ , _

a LILV PCHO E!L0t'ITERS ri, cure 4-2 Environemntal Sannline Locations Within Skm of Plant 4-5 Revision 1 Date: ~7 hb Approved by:

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VERMONT YANKEE NUCLEAR POWER STATION OFF-SITE DOSE CALCULATION MANUAL REV. #2

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Reviewed P~f lu W r (}-'*) PC,1c, / /E/alQ{

I Plant Operational Review Committee Date I Approved / /4// 6/86 flantManager Date I Approved ,c [s OR4

% Ice President and

/ l9/16lS Date Manager of Operations 4

LIST OF AFFECTED PAGES

, Changes, deletions or additions in the most recent revision are indicated by a bar in the margin or by a dot near the page number if the entire page is affected.

Page Revision Date i - 3.43 0 3/1/84 4.1 - 4.lb 2 12/9/86 4.2 - 4.2a 1 10/23/86 4.3 - 4.4 0 3/1/84 I 4.5 1 10/23/86 4.6 - 6.11 0 3/1/84 I

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I l 4.0 ENVIRONMENTAL MONITORING PROGRAM lE E 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 Proaram All routine radiological analyses for environmental samples are performed at the Yankee Environmental Laboratory. The Laboratory participates in the U.S. Environmental Protection Agency's Environmental Radioactivity Laboratory Intercomparison Studies Program for all the species and matri-ces routinely analyzed.

4.2 Airborne Pathway Monitorina The environmental sampling program is designed around several major objec-tives, 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 preopera-tional data and 13 years of operational data (as of 1985). The chosen air sampling locations are discussed below.

To assure that an unnecessarily frequent relocation of samplers will not be required due to short-term or annual fluctuations in meteorology, thus incurring needless expense and destroying the continuity of the program, long term, site specific ground level D/Qs (5-year averages - 1978 through 1982) were evaluated in comparison to the existing air monitoring loca-tions 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.

Revision 2 Date: U/Isl% Approved by: O(D

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I I 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, NW) 27 percent of the time, and the 4 "down-valley" sectors (S, SSE, SE, ESE) 49 percent of the time, for a total "in-valley" time of 76 percent.

Station AP/CF-12 (NNW, 3.6 km) in North Hinsdale, NH, monitors the up-valley sectors. It is located in the sector that ranks fourth overall in terms of wind frequency (i.e., in terms of how often the wind blows into that sector), and is approximately 0.5 miles from the location of the calculated maximum ground level D/Q (i.e., for any location in any sector, for the entire Vermont Yankee environs). This station provides a second function by its location in that it also monitors North Hinsdale, NH, the community with the secorid highest ground level D/Q for surrounding com-munities, and it has been in operation since the preoperational period.

I 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, MA (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/Q (the second highest overall ground level D/Q for any location in any I sector). Station AP/CF-11 is approximately one mile from this point, be-tween it and the plant. Station AP/CF-14 also serves as a community moni-tor for Northfield, MA. Both stations have been in operation since the preoperational period.

4-1A I Revision 2 Date: 19/l@@o Approved by:

m OO

_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . __ _ ____ _____________________ _ ______a

In cdditien to tha up- and down-vallsy 1ccaticns, two communitics h::va been chosen for community sampling locations. The four nearest population groups with the highest long-term average D/Q values, in decreasing order, are Northfield, MA; North Hinsdale, NH; Brattleboro, VT; and Hinsdale, NH.

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.

I 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 (5-year) meteorological history. Sectors in the general west to southwest direction, which would otherwise have been preferable due to lower wind frequencies, were not chosen since they approached the region surrounding the Yankee Atomic plant in Rowe, MA.

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 I

specific Technical Specification requirement.)

4-18 Revision 2 Date: l#!I,f/f 1/ b e Approved by: Re /

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l VERMONT YANKEE NUCLEAR POWER STATION OFF-SITE DOSE CALCULATION MANUAL REV. #3 I

I I

I Reviewed hU ih 'f ra fk % % / Mlhl I Plant Operational Review Committee Date' Approved -

[ / /2//8/84

1 ant Manager Oste '

I Approved (G 5 (ll@\ / \2l92lb W~ce President and Date Manager of Operations I

I I _ --

I LIST OF AFFECTED PAGES Changes, deletions or additions in the most recent revision are indicated by a bar in the margin or by a dot near the page number if the entire page is affected.

Pace Revision Date i-v 0 3/1/84 vi - vii 3 12/9/86 ix 0 3/1/84 <

l 1.1 3 12/9/86 1.2 - 1.5 0 3/1/84 1.6 - 1.11 3 12/9/86 1.12 0 3/1/84 1.13 - 1.19 3 12/9/86 1.20 - 2.4 0 3/1/84 3.1 - 3.46 3 12/9/86 4.1 - 4.18 2 12/9/86 4.2 - 4.2a 1 10/23/86 4.3 - 4.4 0 3/1/84 4.5 1 10/23/86 4.6 - 5.8 0 3/1/84 5.9 - 5.21 3 12/9/86 6.1 - 6.11 0 3/1/84 Appendix A 3 12/9/86

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TABLE OF CONTENTS Page g REVISION REC 0aD.................................................. 1,

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LIST Of EFFECTIVE PAGES.......................................... til DISCLAIMER OF RESPONSIBILITY..................................... iv ABSTRACT......................................................... v LIST OF FIGURES.................................................. vill LIST OF TABLES...................... ............................ ix

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 2.1 Method to Determine F ENG and C NG 2-1 2.2 MethodtoDetermineRddionuclide........................... Concentration for Each Liquid Effluent Pathway........................... 2-2 f 2.2.1 Sample Tanks Pathway............................... 2-2 2.2.2 Service Water Pathway.............................. 2-3 2.2.3 Circulating Water Pathway.......................... 2-3 3.0 0FF-SITE DOSE CALCULATION METH0DS................................ 3-1 3.1 Introductory Concepts...................................... 3-2 3.2 Method to Calculate Total Body Dose from Liquid I 3.3 Releases...................................................

Method to Calculate Maximum Organ Dose from Liquid 3-5 Releases................................................... 3-11 3.4 Method to Calculate the Total Body Dose Rate from Noble Gases................................................ 3-14 3.5 Method to Calculate the Skin Dose Rate from Noble Gases.... 3-18 I 3.6 Method to Calculate the Critical Organ Dose Rate from Iodines, Tritium and Particulates with T Than 8 Days.............................l/2 Greater

................... 3-22 3.7 Method to Calculate the Gamma Air Dose from Noble Gases.... 3-26 I 3.8 3.9 Method to Calculate the Beta Air Dose from Noble Gases.....

Method to Calculate the Critical Organ Dose from Tritium.

Iodines and Particulates...................................

3-29 3-32 Revision U Da e Approved By: h lir 4771R

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I TABLE OF CONTENTS (Continued)

(

Page l( . 3.10 Receptor Points ana Annual Average Atmospheric Dispersion Factors for Important Exposure Pathways.........

3.11 Method to Calculate Direct Dose from Plant Operation.......

3-38 3-42 I 4.0 3.12 Cumulative Doses...........................................

ENVIRONMENTAL MONITORING PR0 GRAM.................................

3-44 4-1 5.0 SETPOINT DETERMINATIONS.......................................... 5-1 5.1 Liquid Effluent Instrumentation Setpoints.................. 5-2 5.2 Gaseous Effluent Instrument. tion Setpoints................. 5-9 6.0 LIQUID AND GASEOUS EFFLUENT STREAMS, RADIATION MONITORS AND RADWASTE TREATMENT SYSTEMS................................... 6-1 6.1 In-Plant Liquid Effluent Pathways.......................... 6-1 6.2 In-Plant Gaseous Effluent Pathways......................... 6-3 REFERENCES....................................................... R-1 APPENDIX A: Method I Example Calculations....................... A-1 Ir I

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LIST OF TABLES b

Number Title Pace 1.1-1 Summary of Radiological Effluent Technical Specifications l

and Implementing Equations 1-3 1.1-2 Summary of Methods to Calculate Unrestricted Area l Liquid Concentrations 1-6 1.1-3 Summary of Methods to Calculate Off-Site Doses from l Liquid Releases 1-7 1.1-4 Summary of Methods to Calculate Dose Rates 1-8 l 1.1-5 Summary of Methods to Calculate Doses to Air from Noble Gases 1-9 l 1.1-6 Summary of Methods to Calculate Dose to an Individual 1 from Tritium, Iodine and Particulates 1-10 1.1-7 Summary of Methods for Setpoint Determinations 1-11 l

1.1-8 Summary of Var'lables 1-12 l 1.1-9 Definition of Terms 1-16 1.1-10 Dose Factors Specific for Vermont Yankee 1-18 1.1-11 Dose Factors Specific for Vermont Yankee for Liquid Releases 1-19 B

l 1.1-12 Dose and Dose Rate Factors Specific fer Vermont Yankee for Tritium, Iodine and Particulate Releases 1-20 l 3.2-1 Environmental Parameters for Liquid Effluents at Vermont Yankee 3-9 3.2-2 Usage Factors for Various Liquid Pathways at Vermont Yankee 3-10 8 3.9-1 Environmental Parameters for Gaseous Effluents at Vermont Yankee 3-38 3.9-2 Usage Factors for Various Gasecus Pathways at Vermont Yankee 3-39 3.10-1 Vermont Yankee Dilution Factors 3-43 4.1 Radiological Environmental Monitoriqq Stations 4-2 5.2-1 Relative Fractions of Core Inventory Noble Gases I After Shutdown '[ -20 ll '

Revision J Date? ' ' ' Approved By:

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1.0 INTRODUCTION

This ODCM (Off-Site Dose Calculation Manual) providas formal and approved methods for the calculation of off-site concentration, off-site doses and effluent monitor setpoints in order to comply with the Vermont Yankee 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 The required by the NRC on the dose consequences of plant operation.

methods contained herein follow accepted NRC guidance (Regulatory Guide 1.109) unless otherwise noted in the text.

It shall be the responsibility of the Chemistry and Health Physics I a Supervisor to ensure that the ODCM is used in the performance of the sur-veillance requiremants and administrate controls of the appropriate por-tions of the Technical Specifications.

I All changes to the ODCM must be reviewed by PORC and approved by MOO in All accordance with Technical Specification 6.13 prior to implementation.

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.

e t Approved By: /W d c-

'M Revision O Date 1-1 3

I I

I Table i.1-2 Summary of Metnods to Calculate Unrestricted Area Liquid Concentrations i{ Equation Reference Number Category Ecuation Section C

2-1 Total Fraction of HPC in F ENG =

1 2.1 Liquids, Except Noble Gases j MPC I

, g g i

l 2-2 Total Activity of Dissolved O 'I 8 and Entrained Noble Gases C"1ml(b)" C"1 !

from all Station Sources < 2E-04 g

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t I

l <

,,,,ovee e ,l 2C

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Revision a cate -- -

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I Table 1.1-2 Summary of Methods to Calculate Unrestricted Area Liquid t'cncentrations

' g{ Equation Reference Number Category Ecuation Section C1 l 2-1 Tota' Fraction of MPC in E

=

2.1 Liquias, Except Noble Gases F)NG g MPC g I i t

I 2-2 Total Activity of 01ssolved and Entrained Noble Gases from all Station Sources CNG (uC1) 1 mi

" C

~< 2E-04 NG i

2.1 l i

I I

l If I

I I

t, 5

i Revision 3 Date Approved By:

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) Table 1.1-3 Summary of Methods to Calculate

( Off-Site Doses frcm Liquid Releases C

Equation Refercnce humber Category Equation Section_

r L

3-1 Total Body 3.2.1 (k Oose 0 tb (mrem)-{QDFL 1 itb i

3-3 Maximum 331 Organ Dose D

. (mrem).{QOFL , i.

g ,

(

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l I

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D Revision 3 Datf( ApprovedBy$fd/ ! -

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( 1-7 4771R

u Table 1.1-4 s

~

Summary of Methods to Calculate Cose Rates Equation Reference Number Category Ecuation Section w

3-5 Total Body Dose Rate g gmrem) - 0.72 3.4.1 from Noble Gases tb yr h1 DFB 1

3 L 3-7 Skin Dose Rate from Noble Gases j skin gmrem) ,

yr

{g ST 1

DF' 1

3.5.1 3

F 6

3-16 Critical Organ Dose .

T .STP 3.6.1 Rate from Iodines, R eg (*#'r}" I" 0 1 DFGjco Tritium 1

.and Particulates with T 1/2 Greater Than Eight Days 1

I l

l l

DEC 2 2 E5 d 7 Revision J Date ApprovedBy:[I -

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I Table 1.1-5 Summary of Methods to Calculate Dose: to Air from Noble Gases

{

Equation Reference Number Category Ecuation Section 3-21 Gamma Dose to Air ST 3*7*I DY Qi DFY i

from Noble Gases air (mrad) = 0.023 g t 3-23 Beta Dose to Air from Noble Gases B

Dair (mrad) - 0.02 ST Qi DF B

i 3.8.1 l I

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Revision 3 Dati" Approved By: D v

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Table 1.1-6 Summary of Methods to Calculate Dose to an Individual from Tritium, Iodine and Particulates

[

F Equation Reference L Number Category Eauation Section 3-25 Dose to Critical ST DFG 3 9'I Qi ico Organ from Iodines, Dco (mrem) - g Tritium and Particulates l

3-27 Direct Dose 3.11.1 Dd (mrem) - 1.29E-06 E u

F L

B(

L F

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Table 1.1-7 n

Summary of Methods for

- Setacint Determinations 3s Equation Reference s

Category Equation Section Number 5-1 Liquid Effluents:

Liquid Radwaste Discharge Monitor (17/350)

R L

spt (cps) = DF OF min

{ "'

1 5.1.1.1 I

Gaseous Effluents:

I- Plant Stack (RR-108-1A, RR-108-18) and A0G Offgas System (3127, L Activity Monitors 3128) Noble Gas

~

5-9 Total Body tb 11 5.2.1.1 Rspt (cpm) - 694 5g F OFB g

/ 5-10 Skin Rskin (cpm) - 3000gSF 1I 5.2.1.1 Of.

spt SJAE Noble Gas b^ 5.2.2.1 5-21 (cpm) = 1.6E+05 5 gF1 i Activity Monitors (17/150A, 17/1508)

R spt I

v I

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Revision 3 Date Approved By: r ! -

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q Table 1.1-8 (continued)

(( Summary of Variables P Variable Definition Units L

3 OF' - Composite skin dose factor mrem-m C

pC1-yr 0FB g - Total body gamma dose factor for nuclide "1"

[-*3 OFB g

- Composite total body dose factor OFL = 5 te-specific, total body dose factor for a mrem itb liquid release of nuclide "1" Ci

= Site-specific, maximum organ dose factor for a mrem I 0FLI

11guld release of nuclide "1" Ci t

0FG = Site-specific, critical organ dose factor for a mrem IC' gaseous release of nuclide."1"

.1 1

C1

= Site-specific, critical organ dose rate factor mrem-sec I- 0FG'CO I for a gaseous release of nuclide "1" pCi-yr DFS g - Beta skin dose factor for nuclide "1" ,

OFj = Combined skin dose factor for nuclide "1" *[,5fC OF{ = Gamma air dose factor for nuclide "1" {#f 0F = Beta air dose factor for nuclide%'1" f I

k - Critical organ dose rate due to iodines I CO and particulates YI ,

p A,,,, = Skin dose rate due to nobie gases =;;* .

kg - Total body dose rate due to noble gases 3 *f*

I % I 3 Dahc 2.2 $$ Approved By: [k f I Revision _ ,

1-13 4771R

Table 1.1-8 (con 81nued)

Summary of Variablec Variable Definition Units D/Q = Deposition factor for dry deposition of I elemental radiolodines and other particulates ,2 E = Gross electric output over the period of MW,h interest F

d

- Flow rate out of discharge canal gpm F, - Flow rate past liquid radwaste monitor gpm F - Flow rate past gaseous radwaste monitor ce sec ENG F - Total fraction of MPC in liquid pathways fraction (excluding noble gases)

MPC g = Maximum permissible concentration for uC1 radionuclide "1" (10CFR20, Appendix B, cc Table 2, Column 2)

( Qg - Release for radionuclide "1" from the curies point of interest Qg - Release rate for radionuclide "1" pcuries/sec at the point of interest

.ST

- The noble gas radionuclide "1" release pCi l Qg rate at the plant stack sec 8

.SJAE uC1 The noble gas radionuclide "1" release 01 rate at the steam jet air ejector 3'C

.A0G I Qg - The noble gas radionuclide "l" release rate at the exhaust of the augmented Off-Gas System uC1 see I .STP Qg - The lodine, tritium, and particulate radionuclide "i" release rate from the plant stack uC1 see ST - The release of noble gas radionuclide curies g1 "1" from the plant stack .

(

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1-14 4771R

Table 1.1-8 (ccntinued)

(

Sumcary of Varlaoles F

L Variable Definition Units

) curies '

STP - The release of todine, tritium, and L Q particulate radionuclides "1" from the plant stack R'spt = Liquid monitor response for the limiting cps concentration at the point of discharge 5 0 - Response of the noble gas monitor at the epm R

spt limiting skin dose rate tb '

R = Response of the noble gas monitor to cpm spt 11mittag total body dose rate Sp - Shielding factor Ratio l S - Detector counting efficiency com or mR/hr l 9

pC1/cc C1/ccl S = Detector counting efficiency for noble com mR/hr l[ 9g gas "1" pC1/cc 0' pct /cc S j = Detector counting efficiency from the most ces recent liquid monitor calibration pC1/mi S yg - Detector counting efficiency for ces radionuclide "1" pCi/mi X/Q = Annual or long-term average undepleted E .

3 atmospheric dispersion factor m l

[X/Q)Y = Effective annual or long-term average gamma atmospheric dispersion factor m I

r Revision 1 Date'i0 Approved By: (

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5 Table 1.1 9 Definition of Terms I

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I Table 1.1-9 (continued) r(

L Definition of Terms b.

k u

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I Table 1.1-10 Dose Factors Specific for Vermont Yankee for holle Gas Releases i

)

i Gamma Beta Air Gamma Air Total Body Beta Skin Combined Skin Dose Fact:

Dose Factor Dose Factor Dose Factor 1 Dose Factor ~

ad h) DFS g (((*~y*)DFj(*[5') Off( ) DF{(

Radionuclide DFB g({#

yr '

C I Ar-41 8.84E-03* 2.69E-03 9.13E-03 3.28E-03 9.30E-03 1.93E-C5

1.54E-05 2.88E-04 Kr-83m 7.56E-08 1.90E-03 1.97E-03 1.23E-03 1.17E-03 1.46E-03 Kr-85m 1.95E-03 1.72E-05 1.34E-03 8.58E-04 Kr-85 1.61E-05 1.11E-02 1.03E-02 6.17E-03 5.92E-03 9.73E-03 Kr-87 2.93E-03 1.52E-02 2.37E-03 1.37E-02 Kr-88 1.47E-02 1.06E-02 1.73E-02 1.01E-02 2.02E-02 Kr-89 1.66E-02 7.83E-03 1.63E-02 7.29E-03 1.76E-02 Kr-90 1.56E-02

[' 9.15E-05 4.76E-04 4.25E-04 1.11E-03 1.56E-04 Xe-131m 8.88E-04 1.48E-03 3.27E-04 2.51E-04 9.94E-04 Xe-133m 4.75E-04 1.05E-03 3.53E-Ca 2.94E-04 3.06E-04 Xe-133 7.39E-04 3.36E-03 7.11E-04 3.14E-03 Xe-135m 3.12E-03

' 2.46E-03 1.92E-03 1.86E-03 2.71E-03 Xe-135 1.81E-03 1.27E-02 1.51E-C3 1.42E-03 1.22E-02 8.89E-03 Xe-137 4.75E-03 9.21E-03 4.13E-03 9.97E-03 Xe-138 8.83E-03

8.84E 8.84 x 10-3 c

E P

--, .3 Approved By: A Revision J Date 1-18

/

4771R I

I

9 o er s>

dt IMAGE EVALUATION

////p%

O'e k,/77/%i @/ TEST TARGET (MT-3) 4 g((,,@g,/g, f,,, j,s

'd EE D=E m m -

C hN l,l i.e l

I.25 1.4 1.6 4 150mm >

4 6" >

d$ Oy -

?i;hhctEs' <$v

- ep j

- A.

9 o T @

se>h IMAGE EVAL.UATION

((/ , 8 ,,

g/ , . . , , - . , -

, '4 ,

l,0 'l M U LA

~

Li tu M d=1

==

ll 1.8 l

1.25 1.4 1.6 4 150mm >

4 6" >

k'?6/

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<>#+w f$

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x+ r

Table 1.1-11 (

Dose Factors Specific for Vermont Yankee

(_ for 5 Liquid Releases Total Body Maximum Organ Dose Factor Dose Factor

[ DFL DFL Radionuclide itb ("C1 ) imo ("C1 )

H-3 1.75E-04 1.75E-04 '

Na-24 1.66E-03 1.66E-03 Cr-51 2.11E-05 3.13E-03 Mn-54 1.79E-02 1.61E-01 l Mn-56 1.67E-06 1.89E-04 Fe-59 2.96E-02 1.34E-01 -

5.35E-02

[ Co-58 Co-60 1.41E-02 8.27E-02 2.67E-01 i

Zn-65 4.86E-01 7.97E-01 !

r' Sr-89 6.80E-02 2.38E+00 !

L Sr-90 2.07E+01 8.15E+01 Zr-95 4.09E-04 8.49E-02 Mo-99 5.32E-04 4.42E-03 6.48E-05 2.02E-03

( Tc-99m Sb-124 7.80E-03 1.92E-01 I-131 2.52E-03 1.46E+00 I-132 1.99E-06 5.82E-06 .

{(\ I-133 1.91E-04 9.34E-02 I-135 1.49E-05 2.38E-03 Cs-134 5.26E+00 -

7.01E+00

( Cs-137 Ba-140 3.15E+00 1.01E-03 6.07E+00 1.72E-02 Ce-141 2.27E-05 3.14E-02 .

W-187 4.23E-04 3.17E-01 c

E

,e.1 slo 2 Daaro 1-19 4,,,o ed 8, ENELT s

v c

(' 4771R

[

r' - _ - - - - - - - - - - -

7

(

3.0 0FF-SITE DOSE CALCULATION METH005

((

l Chapter 3 sr0Vides tha basis for plant procedures required to meet 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 each of the requirements 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 i 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 material with which the operator can calculate the needed doses and dose

{ rates. Setpoint methods for effluent monitor alarms are described in ,

Chapter 5.

[

Revision 3 Date Approvad By: ,, , h ~

3-1 J

4172R r -- -- -- -

(

3.1 Introduc9ery Concepts

( The Radiological Effluent Technical Specifications (RETS) either limit dose or dose rate. 1he term " Dose" for in1ested or inhaleo radioactivity pasns the dose commitment, measured in mrem, which results from the exposure to radioactive materials that, because of uptake and deposition in the body, will continue to expose the body to radiation for some period of time after the source of radioactivity is stopped. The time frame over which the dose commitment is evaluated is 50 years. The phrases " annual Dose" or " Dose in one year" then refers to the fifty-year dose commitment from one year's worth of releases. " Dose in a quarter" similarly means a fif ty-year dose <ommitment.

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.

t Gaseous effluents from the plant are also controlled such that the I. maximum " dose rates" at the site boundary at any time are limited to l( 500 mrem / year. The a'inual dose limits are the doses associated with the concentrations of Appendix B, Table II, Column 1 of 10CFR Part 20 l (10CFR20.106(a)). The use of the annual dose limits embodied in 10CFR Part 20 as plant " dose rate" values (to be applied at any time consistent with the capabilities of the monitoring instrumentation to determine) provides l

reasonable assurance that radioactive material discharged in gaseous effluents will not result in the exposure of member (s) of the public either within or outside the site boundary to annual average concentrations exceeding the I

l federal regulations.

It should also be noted that a dose rate due to noble gases that exceeds for a short tisae period (less th'an one hour in duration) the equivalent 500 mrem /yaar dose rate limit stated in Technical Specification 3.8.E.1.a. does not necessartis, by itself, constitute a Licensee Event Report (LER) under 10CFR Part 50.73 unless it 13 determined that the Al r'

Revision J Date Approved By: N' k

( 4172R 3-2

I j concentration cf radioactive effluents in unrestricted areas has also exceeded two times MPC when averaged over one hour (four-hour notification per 10CFR50.72, and 30-day 1.ER per 10CFR50.73).

The quantitles D and R are introduced to provide calculable quantitles.

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 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 -

I 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. It should be noted that if plant release rates were such that an R equal to the Technical Specification (3.8.E.1) value was continued for one year, the annual dose limits of 10CFR20 would be reached. However, since maximum allowed release rates and the resulting dose rates in the range of the Technical Specification limits are very infrequent and are typically of short CEC 02 1$36 Approved By: , ,k C Revision U Date .

I[ 3-3 4172R I .

time duration, this approach of limiting dose rates equivalent to the annual dose limits then assures that 10CFR20.106 limits on an annual average air

( concentration in unrestricted areas will be met.

1 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.

l 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 11 15 a more detailed analysis I which allows for use of site-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 i the time of analysis under Method II.

l B

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I l

E I

I I

B Revision 3 DatF "'

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c j

-( 3-4 4172R I

3.2 Method to Calculate the Total Body Dose from Liould Releases Technical Specification 3.8.B.1 limits the total bocy dose cemmitment f

to a Meep.' of the Pubile from r adioactive meterial in liquid effluer.ts 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 arem 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 11gulds) to 25 mrem in a year. Dose evaluation is required at least once per month. If the 11guld radwaste treatment system is not being used, dose evaluation is required before each release. .

Use Method I first to calculate the maximum total body dose from a liquid release to the Connecticut River as it is simpler to execute and more conservative than Method II.

Use Method II if a more accurate calculation of total body dose is needed (i.e., Method I indicates the dose is greater than the limit), or if

{ Method I cannot be applied.

If the radwaste system is not operating, the total body dose must be estimated prior to a release (Specification 3.8.C.1). To evaluate the total b 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 11guld release is:

o,, <3-n 4 o, on.,,,

(.,em, m,m mam, C C-Revision J oatl ' ..2 Approved sy: K . T 61 -

3-5 C 4172R ,

c r - - - - - - - - - - - -

I where:

0FLitb = Site-specific total body dose factor (mrem /C1) for a liquid I

(

I release. See Table 1.1-11.

Q1

= Total activity (C1) released for radionuclide "1". (For strontiums and Fe-55, use the most recent measurement l 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),

2. Liquid releases were to the Connecticut River, and

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

l 3.2.2 8& sis for Method I E This section serves three purposes: (1) to document that Method I complies with appropriate NRC regulations, (2) to provide background and I( 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 l

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. These Technical Specifications are based on design objectives and standards in 10CFR and 40CFR. Technical Spectfication 3.8.B.1 15 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 i

(hereafter called the Objective). Technical Specification 3.8.M.) 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.

3 Reviston 3 Date Approved By:

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Exceeding the Objective er the Standard does not immediately licit plant operation but requires a report to the NRC within 30 days. In addition,

[ a waiver may be required. This is unlike exceed 19g 10CFR20 limits whicn could result in piant shutdown.

I 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 g dose) provides part of the conservative margin to the calculation of total 3

body dose in Method I. Method II allows that actual individuals, with real behaviors, be taken into account for any given release. In fact, Method I was based on a Method II analysis for the critical receptor with maximum exposure conditions instead of any real individual. That analysis was called the " base case"; it was then reduced to form Method I.

The steps performed in the Method I derivation follow. First, in the base case, the dose impact to the critical receptor (in the form of dose factors DFLitb, mrem /C1) 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 I

A-16, Reference A). The 11guld pathways identified as contributing to an in'dtvidual'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 I river water available for drinking as well as having feed grown on irrigated land, and the direct exposure from the ground plane associated with activity deposited by the water pathway. A plant discharge flow rate of 44.6 ft /sec was used with a mixing ratto 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.

5

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Fcr any 11guh1 release, during any period, the increment in annual average total body dose from radionuclide "1" is:

(

AD "O n DFL (3-2) tb ith (arem) (C1) ("y") '

where:

DFLitb = Site-specific total body dose factor (arem/Cl) for a liquid !'

release. See Table 1.1.11. ,

( Qt - Total activity (C1) released for radionuclide "1"..

Method I is conservative because it is based on dose factors DFL itb 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 II l 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 1,250 cfs that is assumed in the Method I dose factors. 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.

E C /

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M & 8 8 M M M M M M M M M M M M M M M m m s Table 3.2-1 Environmental Parameters for Liquid Effluents at Vermont Yankee (Derived from Reference A)

POTABLE AQUATIC SHORELINE FOOD CROWN WITH CONTAMINATED WATER VARIABLE WATER FOOD ACTIVITY VECETABLES LEAFY VEC. MEAT COW MILK MP Mixing Ratio - 0.0356 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 YV Agricultural (KC/MI ) 2.0 2.0 2.0 2.0 Productivity P Soil Surface (KC/M2 ) 240.0 240.0 240.0 240.0 Density IRR Irrigation Rate (L/M2/HR) 0.152 0.152 0.152 0.152 TE Crop Exposure (HRS) 1440.0 1440.0 1440.0 1440.0 Time TH Holdup Time (HRS) 1440.0 24.0 2160.0 2160.0 QAW Water Uptake Rate (L/D) 50.0 60.0 for Animal Feed Uptake Rate (KG/D) 50.0 50.0 QF for Animal ,

0.5 0.5 0.5 0.5 l FI Fraction of Year Crops Irrigated Location of Connecticut River below Vernon Dam.

Critical Receptor O1 OM Revision j Date_Di.t 3-9 Approved By )N i v

E E E E E E E E E E W W W W g g' g

" ,r Table 3.2-2 Usaae Factors for Various Liquid Fathways at Vermont Yankee (From Reference A. Table E-5. Zero where no pathway exists)

MILK MEAT FISH INVERT. POTABLE SHORELINE ACE VEC. LEAFY VEG. WATER (LITER /YR) (KG/YR) (KC/YR) (KG/YR) (LITER /YR) (HR/YR)

(KC/YR) (KC/ YR )

64.00 310.00 110.00 21.00 0.00 0.00 12.00 Adult 520.00 42.00 400.00 65.00 16.00 0.00 0.00 67.00 Teen 630.00 330.00 41.00 6.90 0.00 0.00 14.00 child $20.00 26.00 330.00 0.00 0.00 0.00 0.00 0.00 Infant 0.00 0.00 7-Ji Approved Bh DJ%,m ') '

3-10 Revision 3 Date t

i 3.3 Method to Calculate Maximum Organ Dose from Licuid Releases

(( Technical Specificatten 3.8.8.1 limits th9 maxir.um croer, dose commitment to a Member cf the Pubile from radioactive material in liquid effluents to 5 mres per quarter and 10 arem per year. Technical Specifica'tlon

{ 3.8.C.1 requires liquid radwaste treatment when the maximum organ dose estimate exceeds 0.2 stem in any month. Technical Specification 3.,8.M.1 k Ilmits 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

( thyrold, which is Ilmited to 75 arem 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 b conservative than Method II.

Use Method !! 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 te estimated prior to a release (Speciftcation 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 Itquid release is:

0, .

Og DFL g ,

(3-3)

Revislon 3 Da Approved By: -

k --

3-11

(. 4172R

(c. rem) (C1)(y'[)

l where:

OFLimo - S'te-specific maximum Organ dose f actor (crem/C) for a l 11guld release. See Table 1.1-11. i Qt - Total activity (C1) released for radionuclide "1". (For ,

strontiums and Fe-55, use the most recent measurement l available.)

Equation 3-3 can be applied under the following conditions (otherwise, justify iI 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 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 radionucilde, a dose factor (mrem /C1) was determined for each of seven organs and four age groups. The largest of these was chosen to be the maximum organ dose factor (OFLg ) for that radionucilde.

For any liquid release, during any period, the increment in annual average dose from radionuclide "1" to the maximum organ is:

I y b Revision d Date' Approved By:in 3(

(

3-12 4172R I

s AD,, =Q g DFLg ,, (3-4)

(mres)

(

(Cl) ("[)

where:

DFLt oo - Site-spectf1e maximum organ dose factor (arem/C1) for a

( liquid release. See Table 1.1-11.

Qt - Total activity (Cl) released for radionuclide "1".

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 !! !

b 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 !! 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 Methed !

!! calculations should be referred to YNSD to be performed and documented.

[

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Approved By: . ll (U Q 3-13 G

(l 4172R

[

r ----

I I I 3.4 Method to Calculate the Total Body Dose Rate From Noble Gases 1, Technical Specification 3.8.E.1 limits the dose rate at any time to the total body from noble gases at any location at or beyond the site boundary equal to or less than 500 mrem / year. By limiting the maximum Rtb to a rate equivalent to no more than 500 mrem / year, assurance 1s provided that the total body dose accrued in any one year by any member of the general public will be less than 500 mrem in accordance with the annual dose limits of 10CFR Part 20 to unrestricted areas.

Use Method I first to calculate the Total Body Dose Rate from the peak release rate via the plant stack. Method I applies at all release rates.

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 tine interval.

Compliance with the dose rate limits for noble gases are continuously II 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-stte dose rate limit of Technical Specification 3.8.E.1, or a value below it.

I Determinations of dose rates for compilance 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 Spectftcation Table 3.9.2 when the stack noble gas monitor is inoperable.

3.4.1 Method I The Total Body Dose Rate due to noble gases can be determined by multiplying the individual radionuclide release rates by their e pective dose Revision 3 Date Approved By: _ O _CU y

l 4172R I

I factors, summing all the products together, and then multiplying this total by ,

a conversion constant (0.72), as seen in the following Equation 3-5 (an !

g example calculation is p-ovided in Appendix A):

(3-5) it tb

= 0.72 hf DFB g 3

gmrem) goc 1-sec) ggCl) gerem-m )

yr Cl-m 3 sec pC1-yr where: l I h{T - In the case of noble gases, the release rate from the plant stack (pCl/sec) for each radionuclide, "1", identified. The I release rate at the plant stack is based on the measured radionuclide distributton in the off-gas at the Steam Jet Air Ejector (SJAE) during plant operation when the activity at the

}-

I 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:

i

'SJAE b M b F (3-28) hST I

=

39 ls {fSJAE 1

I gC1 sec

, gep,) guC1/cc) (cc_)

cpm sec H = Plant Stack Gas H'onitor I or II count rate (cpm).

Sg - Appropriate or conservative plant stack monitor detector counting efficiency for the given nuclide mix (cpm /(pC1/cc)).

F = Stack flow rate (cc/sec).

hyJAE - The last measured release rate at the steam jet air ejector of noble gas 1 (pC1/sec).

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

I Reviston '3 Caty C 2 3 N 0 Approved By: M.

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I 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. :n t!.is case, the ratio of eachhSJAE tothesumofallhSJAE in Equation 3-28 above is assumed to I reduce to a value of 1, and the total body gamma dose factor DFBg for Xe-133 (2.94 E-04 mrem-m 3

/ Ci-yr) is used in Equation 3-5. Alternately, a relative radionuclide "1" mix fraction (f ) gmay be taken from Table 5.2-1 as a function of time after shutdown, and substituted in place of the ratto of h5AE tothesumofallhSJAE 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 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. I Equation 3-5 can be applied under the following conditions (otherwise.

f justify Method I or consider Method II):

1. Normal operations (not emergency event), and

2. Noble gas releases via the plant stack to the atmosphere.

I 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.

I c

. . . ~ . .15 ! 1 Revision J Date Approved By: k Q k

( 3-16 4172R l

I Method I was derived from Regulatory Guide 1.109 as follows:

R = IE+06 5 7 (X/Q]Y hfT OF8 g (3-6) tb 3

I g mrem y yr g LCt) pc1 g)g sec

,3 )

gy .q) gerem-m )

sec pCl-yr where:

Sp = Shtelding factor - 1.0 for dose rate determination, j (X/Q)Y = Maximum annual average gamma atmospheric dispersion factor

= 7.2E-07 (sec/m3 )

I hfT - Release rate from the plant stack of noble gas "1" (pC1/sec).

OFB, . Gamma total body dose factor, ( ). See Table 1.1-10.

Equation 3-6 reduces to:

- 0.72 hfi DFB (3-5) tb I 3

gmrem) (DCl-sec) gpg) gmrem-m )

yr 3 sec pct-yr pCl-m The selection of critical receptor, outilned in Section 3.10, is inherent in f 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 i 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 butiding ventilation air flow, and that this air I flow has an isotopic distribution consistent with that routinely measured at '

the SJAE.

I i

/ Revision d Date

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C1 4172R I - .-. _ _ _. ._ . . .. - - - -

In the case Cf n ble gas d2se rates, Method II cannot provide cuch extra realism because R tb is already based on several factors which make use

( of current plant parameters. However, should it be needed, tne dose rate analysis for critical receptor can be performed making use of current meteorology during the time interval of recorded peak release rate in place of the default atmospheric dispersion factor used in Method I.

3.4.3 Method II 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 I 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 us'e of Method II.

It is an acceptable starting point for a Method II analysis. Anilyses requiring Method !! calculations should be referred to YNSO to be performed and documented.

l I

.I I

I Revision 3 Date Approved By: j nk -

3-18 4172R

f 3.5 Method to Calculate the Skin Dose Rate from Noble Gases I

( Technical Specification 3.8.E.1 limits the dose rate at any time to the l skin from noble gases at any location at or beyond the site boundary to 3.000 mrem / year. Bylimitingthemaximumit skin to a rate equivalent to no more l than 3,000 mrem / year, assurance is provided that the skin dose accrued in any I

one year by any member of the general public is much less than 3,000 mrem.

l .

1 Use Method I first to calculate the Skin Dose Rate from the peak l release rate via the plant vent stack. Method I applies at all release rates.

I 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 l l had actually been exceeded during a short time interval.

I l Compliance with the dose rate limits for noble gases are continuously j 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 l dose rate limit, or a value below it. I I

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 l Specification Table 3.9.2 when the stack noble gas monitor is inoperable. j 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): ,

Revision 1 DatIr1 Approved By:

'f g c -

3-19 4172R

.ST DF g (3-7)

R 01 skin " ,

( (m_rgmy yr (13g,!,) (mram-see) sec pCl-yr where:

hyT = In the case of noble gases, the noble gas release rate from the plant stack (pC1/sec) for each radionuclide, "1", identified.

The release rate at the plant stack is based on the measured radionuclide distributton in the off-gas at the Steam Jet Air 8 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 I release rate at the stack can also be stated as follows:

I

SJAE 01 L (3-28) g 7 fSTI 39 JAE I uCl_

sec (cpm) (uti/cc)_ (cc )

cp'n sec H - Plant stack gas monitor I or II count rate (cpm).

Sg - Appropriate or conservative plant stack monttor i Ig detector counting efficiency for the given nuclide mtx (cpm /(pCl/cc)).

F = Stack flow rate (cc/sec).

hyJAE = The last measured release rate at the steam jet air ejector of noble gas 1 (pC1/sec).

- combined skin dose factor (see Table 1.1-10).

OFl 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 eachofhfAE tothesumofallhSJAE in Equation 3-28 above is assumed to I ,

Reviston O Date " Approved By: 1b 3-20 l g' 4i72R i

I _

1 1

reduce to a value of 1, and the combined skin dose factor DF'y for Xe-133 ;

(3.90 E-04 mrem-sec/pC1-year) is used in Equation 3-7. Alternately, a 5.-l relative radio'1uclide "1" mix fraction (f ) gm3y be taken from Table as I(

i a function of time after shutdown, and substituted in place of the ratto of i

eachhSJAE tothesumofallhSJAE 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 '

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 I the above assumptions.

Equation 3-7 can be applied under the following conditions (otherwise.

justify Method I or consider Method II): ,

I 1. Normal operations (not emergency event), and

- 2. Noble gas releases via the plant stack to the atmosphere.

3.5.2 Basis For Method I The methods to calculate skin dose rate parallel the total body dose rate methods in Section 3.4.3. Only the differences are presented here.

Method I may be 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.

l Method I was derived from Regulatory Guide 1.109 as follows:

I l

l s o.e >

o . i.1, S, g,, . 3.i7E.04 ? o, <x,o0,S, g

g ovis,on o ate .. 4,,,o ed 8, 1.v C .

3-21 l g 4172R I _- - _

3 gmrem) _ g) g ) (trad) pCl-yr J C (sec) grrem-m )

yr yr gCl-sec) yr ,3 pct-yr where:

1.11 - Average ratto of tissue to air absorption coefficients will convert mrad in air to mrem in tissue. )

D - 3.17E+04 Qg [X/Q) DFgY (3-9) 1r 3

(mrad)

DCl-yr

( )gg ) mrad-m yr (Cl-sec) ,3 pC1-yr ~

i now D finite - D air

0 # #03 -

3 gmrad) gmrad) (m) m yr yr g sec)

,3 and Qg - 31.54 hST (3 33)

(Cl) (Cl-sec) guC1)

- yr pct-yr see I so 4,,,,.i.iis,iE.06n/QF{W Dq <3a2r 3

g (mrem) yr

() g) (pC1) pct g

,3 ) , (sec)

G (mrad-m pct-yr )

+ IE+06 X/Q i

hf DFS g 3

oCl sec fj. gmrem-m )

gC1 ) ,3 see pct-yr substituting (X/Q)Y - 7.2E-07 sec/m3 X/Q - 6.3E-07 sec/m3 Revision 3 Date- Approved By M A

Ic ,,,,,

3-22 I

I .

I SF - Shielding factor a 1.0 fcr dose rate determinations gives k skin 0.80 hf + 0.63 (2-13)

' hf 0FS 0F{ g :

E 3 3 (mrem) (DC1-sec-mrem) (g) gmrem-m ) goci-sec) ( g ) (mrem-m )

yr 3 sec pCi-yr 3 sec pCi-yr Cl-m -mrad Cl-m

- (3-14) hf(0.800F{+0.63DFS3 g define ;

DFj-0.80DF{+0.63DFS g (3-15) then R - DFj (3-7) skin I gmrem) yr gg) gmrem-sec) sec pCi-yr The selection of critical receptor, outlined in Section 3.10 is inherent in g

= Method I, as it determined the maximum expected off-site atmospheric dispersion I factors based on past long-term site-specific meteorology. ;

I 3.5.3 Method II 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 li 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 ,

l' calculations should be referred to YNSD to be performed and documented. l 4

I I

I Revision J DatpEC 2 2 1986 A roved By: k I 4172R 3-23 I __ -. -- --

I 3.6 Meth*d to Calculate the Critical Organ Dose Rate from Icdines. Tritium and Greater Thsn 8 Days Rrticulates with T1/2 Technical Specification 3.6.E.1.b limits the dose rate to any organ, I denoted Ren, from I-131. I-132, 3H, and radionuclides in particulate form with half lives greater than 8 days to 1500 mre:n/ year to any organ. The peak release rate averaging time in the case of todines and particulates is commensurate with the time the lodine and particulate samplers are in service between changeouts (typically a week). By limiting the maximum Re ,to a rate equivalent to no more than 1500 mrem / year, assurance is provided that the critical organ dose accrued in any one year by any member of a the general public will be less than 1500 mrem.

Use Method I first to calculate the critical organ dose rate from the peak release rate via the plant vent stack. Method I applies at all release rates.

use Method II if Method I predicts a dose rate greater than the I 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.b had actually been exceeded during the sampilng period.

3.6.1 Method I The critical organ dose rate 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 I (an example calculation is provided in Appendix A):

R, g

- hSTP OFGjg, (3 16) gmrem) guci) (mrem-sec) yr sec pCl-yr l Revision d Da h Approved By:

d/

GC .

G l l(. 4172R 3-24 I E

I

I there: ,

hSTP - Stack activity release rate dete:c.ination cf radio..eclide "i" (Iodine-131, Iodine-133, particui dtes with half-lives greater fk than 8 days, and tritium), in pC1/sec. For 1 - Sr89, Sr90 or tritium, use the best estimates (such as most recent measurements). l 0FGjg,-Sitespecificcriticalorgandosefactorrate(" f) fora 5-gaseous release. See Table 1.1-12.

I Equation 3-16 can be applied under the following conditions (otherwise, justify Method I or consider Method II):

1. Normal operations (not emergency event), and

2. Tritium, iodine, and particulate releases via the plant stack to the atmosphere.

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 presented here.

Method I may be used to show that the Technical Specification which limits organ dose rate from todines, tritium and radionuclides in particulate form wtth half lives greater than 8 days (hereafter called Iodines and i 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.

i The equation for Rg ,1s derived by modifying Equation 3-25 from I ,evision a Date

~

,,,,ovee.,:D it c v ,

I(' 3-25 4172R E

I

I Section 3.9 as follows:

k - Og DFG ggo (3-17) !

eo (mrem) (C1) ("y")

applying the conversion factor. 31.54 (Cl-sec/pci-yr) and converting Q to h in pC1/sec as it applies to the plant stack yleids:

I kg , = 31.54 hSTP DFG ggo (3-18) gmrem) (Cl-sec) gg) gmrem) yr pC1-yr sec C1 Eq. 3-18 is rewritten in the form:

k . hSTP DFGj g (3-19) i gmrem) gg) gmrem-sec) yr sec pC1-yr where DFGjgo - 31.54 DFG ggo (3-20)

(mrem-sec) pC1-yr (C1-sec) (mrem) pCl-yr Ci I The selection of critical receptor, outlined in Section 3.10 is inherent in Method I, as are the maximum expected off-site atmosphe.ric I dispersion factors based on past long-term site-specific meteorology.

Should Method II be needed, the analysis for critical receptor critical l

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

' Revision J Date - . Approved By: b -

l ( 3-26 i 4172R l

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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 II consists of the models, input t

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 .he use of Method II.

r L It is an acceptable starting point for a Method 11 analysis. Analyses requiring Method 11 calculations should be referred to YNSO to be performed p

L and documented.

r I

I I

I pevi,lon 2 o s 22 55 ,,,, edex: 2 V

d d ,

3-27 l 4172R g

I

I 3.7 Method to Calculate the Gamma Air Dose from Noble Gases Technical Specification 3.8.F.1 limtts the garr.ma dose to air frem 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 the plant stack j releases during the period. Method I applies at all dose levels.

Use Method II if a more accurate calculation is needed.

E Method I 3.7.1

~

B The gamma air dose from plant stack releases is:

0,Jr=0.023 Q, (3-21)

DFJ 1

(mrad) (DCl-w)3 (C1) (*#8d-* )

r Cl-m pCi-yr where:

Qg

- total noble gas activity (Curies) released to the atmosphere via the plant stack of each radionuclide "1" during the period of interest.

lI. See Table 1.1-10 OF{=gammadosefactortoairforradionuclide"i".

Equation 3-21 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 the plant stack to the atmosphere.

I a Revision O Date ApprovedBy:pypk p

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3.7.2 Basis for Method I I !

(

Method I may be used to show that the Technica'. Specification which limits off-site gamma air dose from gaseous effluents (3.8.F.1) has been met l

for releases over appropriate periods. This Technical Specification is based

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

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

For any noble gas release, in any period, the dose is taken from

! Equations B-4 and B-5 of Regulatory Guide 1.109 with the added assumption that O /

finite - 09 X/Q]Y [X/Q):

D 3.17E+04 [X/Ql Y Qg DFj (3-22) ar 1 a

(mrad) ( ) (sec/m3 ) (C1) ( )

s where:

[X/Q]Y - maximum annual average gamma atmospheric dispersion factor

- 7.2E-07 (sec/m3 )

J Qg - number of curies of noble gas "1" released l which leads to:

- 0.023 DFj (3-21)

D Q, ar 1 d

(mrad) (DCi- ) (C1) ( _

)

C1-m The main difference between Method I and Method 11 is that Method II would allow the use of actual meteorology to determine [X/Q]Y rather than '

use the maximum long-term average value obtained for the ye 1978 to 1982.

Approved By: /

Revision O Date J q 3-29 4172R

3.7.3 Method II l l 1 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 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. I Analyses requiring Method II calculations should be referred to YNSD to be ,

performed and documented.

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I Revision 3 Da a>k Approved By: V). u c>

( 3-30 I

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3.8 Method to Calculate the Beta Air Dose from Noble Gases F' Technical Specification 3.8.F.1 limits the Deta dose to air from 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.

L Use Method I first to calculate the beta air dose for the plant stack

[ releases during the period. Method I applies at all dose levels.

Use Method II if a more accurate calculation is needed or if Method I

( cannot be applied.

3.8.1 Method I F

L The beta air dose from plant vent stack releases is:

0 = 0.02 Qg DF (3-23) ir 1

[ (mrad) (D'-#)3 (C1) (mrad-m )

DU -YI Cl-m where:

DFf-betadosefactortoairforradionuclide"i". See Table 1.1-10.

Qg - total noble gas activity (Curies) released to the atmosphere via the plant stack of each radionuclide "1" during the period of interest.

Equation 3-23 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 the plant stack to the atmosphere.

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(. 3-31 l

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L 3.8.2 Basis for Method I

[g This section serves three purposes: (I) to document that Metnod I complies with appropriate NRC regulaticns, (2) to provide background and training information to Method I users, and (3) to provide an introductory 1

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 l presented here.

E Method I may be used to show that the Technical Specification which l

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.1, which limits the I

l estimated annual beta air dose at unrestricted area locations.

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

I l 7 For any noble gas release, in any period, the dose is taken from Equations B-4 and B-5 of Regulatory Guide 1.109:

6 Qg DF (3-24)

D, 7 - 3.17E+04 X/Q

3 I

f (mrad) (DCi-vr}

C1-sec (sec)

,3 (C1) (mrad-m pCi-yr I

substituting X/Q = Maximum annual average undepleted atmospheric dispersion factor.

l 3

- 5.3E-07 sec/m l

C

" ^} f Approved By: / ek_ Y Revision d Date 3-32 4172R

We have I Ofir = 0.02 Qg 0F B

(3-23) 3 (mrad) (DC1-w) 3 (C1) (mrad-m pC1-yr ) !

I 3.8.3 Method II C1-m 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 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.

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Revision O Dath Approved By: ,k Ir 4172R 3 33 v

I

I 3.9 Method to Calculate the Critical Organ Oose from Iodines. Tritium and I Particulates Technical Specification 3.8.G.1 limits the critical organ dose to a Member of the Public from radioactive Iodines, Tritium, and particulates with half-lives greater than 8 days (hereafter called "I+P") in gaseous effluents to 7.5 mrem per quarter and 15 mram per year. Technical Specification 3.8.M.1 limits the total body and organ dose to any real member of the public from all I station sources (including gaseous effluents) to 25 mrem in a year except for the thyroid, which is limited to 75 mrem in a year.

Use Method I first to calculate the critical organ dose from a vent stack release as it is simpler to execute and more conservative than Method II. Method I is conservative for total body, critical organ, and thyroid dose greater than 0.1 mrem.

Use Method II if a more accurate calculation of critical organ dose is I needed (i.e., Method I indicates the dose is greater than the limit), or if Method I cannot be applied, or if the majority of the release is Iodine and the 75 mrem limit for Specification 3.8.L.1 needs to be evaluated.

s I 3.9.1 Method I +

1 (3-25)

O, g

= Qg 0FG jgo (mrem) (C1) (*]*)

I Qy

- Total activity (C1) released to the atmosphere of radionuclide "1" during the period of interest. For strontiums, use the most recent measurement.

For each OFG jgg - Site-specific critical organ dose factor (mrem /C1).

radionuclide it is the age group and organ with the largest dose factor. See Table 1.1-12.

o Revision U Dath Approved By: 7' d, , k.

3-34 4172R I

L Equation 3-25 can be applied under the folloaing conditions (othercise, justify Method I or consider Method II):

e L.- 1. Normal operations (not emergency event),

2. I+P releases via the plant stack to the atmosphere, and

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

3.9.2 Basis for Method I

( l 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 organ dose from gases (3.8.G.1 and 3.8.L.1) have been met for releases over the appropriate periods. These Technical Specifications are

{ ,.

based on Objectives and Standards in 10CFR and 40CFR. Technical Specification 3.8.G.1 is based on the ALARA Objectives in 10CFR50, Appendix I, Subsection II C. 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.

These methods apply only to I+P in gaseous effluents contribution.

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

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 use below of a single " critical receptor" provides part of Date - Approved By:

Revision 3 3-35 4172R r _ - ---_

l the conservative margin to the calculation of critical organ dose in Method

- I. Method II allows'that actual individuals, with real behaviors, be taken i

H

( into account for any given release. In fact, Pethod I was based on a Metnod 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 I 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 P assumptions and data used are presented below.

L

~

The steps performed in the Method I derivation follow. First, in the i base case, the dose impact to the critical receptor in the form of dose factors (mrem /C1) of I curie release of each I+P radionuclide to gaseous effluents was derived. Then Method I was deteruined 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 anc 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 l

the highest dose factor for that organ. These critical organ, critical age dose factors are given in Table 1.1-12.

I For any gas release, during any period, the increment in annual average I

l dose from radionuclide "1" is:

(3-26)

A0gg,= QgDFG ggn I

is where DFG gC,is the critical dose factor for radionuclide "1" and Q g l

the activity of radionuclide "1" released in curies.

C Revision O Date Approved By: &'Rx J

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3-36

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M'thod I is more conservaeive than Meehod II in the region of tne L -

Technical Specificatten limits because it is based on the fcilowing re a ction f of the base case. The dose factors OFG ggg used in Method I were chosen from the base case to be the highest of the set for that radionuclide. In effect each radionuclide is conservatively represented by its own critical age group L and critical organ.

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.

3.9.3 METHOD II I If Method I cannot be applied, or if the Method I dose exceeds the limit or if a more exact calculation is required, then Method II should be applied. Method II consists of the models, input data and assumptions in l

Regulatory Guide 1.109, Rev. 1 (Reference A), except where site-specific I

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.

l I

l B

1 r~

, ,a~ GB6

\j (

Revision V Cafe , Approved By: L L 3,37

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m M M M W W W W W W W W W W W W W W W te 3.9-1 Environmental Parameters for Caseous Effluents at Vermont Yankee (Derived f rom Reference A)

Vegetables Cow Milk Coat Milk Meat' Variable Stored Leafy Pasture Stored Pasture Stored Pasture Stored YV #gricultural (Kg/M2 ) 2. 2. 0.75 2. 0.75 2. 0.75 2.

Productivity P Sail Surfqce (KC/M2 ) 240. 240. 240. 240. 240. 240. 240. 240.

Density T 'ransport Time (HRS) 48. 48, 48. 48. 480. 480.

to User TB Soil Exposure (HRS) 131400. 131400. 131400. 131400. 131400. 131400. 131400. 131400.

Time l TF Crop Exposure (HRS) 1440. 1440. 720. 720. 720. 720. 720. 720.

l Time to Plume TH Holdup After (HRS) 1440. 24. O. 2160. O. 2160. O. 2160.

! Harvest i

I QF Animals Daily (KC/ DAY) 50. 50. 6. 6. 50. 50.

Feed l

FP Fraction of Year 0.50 0.50 0.50 on Pasture I FS Fraction Pasture 1. 1. 1.

when on Pasture FG Fraction of Stored 0.76 Veg. Grown in Carden FL Fraction of Leafy 0.50 Veg. Crown in Carden FI Fraction Elemental Iodine = 0.5 H Absolute Humidity = 8.00 (gm/M3) f

, l* cs (~~

l<evision ? Date C" 1-38 Approved By. E OC-( __

l

. Table 3.9-2 Usage Factors for Various Gaseous Pathways at Ver:nont Yankee (from Regulatory Guide 1.109, Table E-5)

Age Leafy Group Vegetables Vegetables Milk Meat Inhalation (kg/yr) (kg/yr) (1/yr) (kg/yr) (m3 /yr)

Adu1t 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 9

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Revision 3 _ Date _a Approved By: .

s)

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4172R I - _ - . . _ .

V L .3.10 Rece dor Poirts and Annual Average Atmospheric Dispersion Factors for Important Exposure Pathways g

k The gaseous effluent dose methods have been simpitfled 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 f in Regulatory Guide 1.109 (Reference A) have been considered:

L e 1. Direct exposure to contaminated air; L

2. Direct exposure to contaminated ground; -

E '

3. Inhalation of air; C 4. Ingestion of vegetables;

5. Ingestion of goat's milk; and

6. Ingestion of meat.

l 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 meteorologic data into atmospheric dispersion factors. Section 3.10.3 l

presents the maximum atmospheric dispersion factors calculated at each of the off-site receptor locations.

l 3.10.1 Receptor Locations l

Three important receptor locations are considered in the dose and dose l rate equations for gaseous radioactive effluents. They are:

I l

1. The point of maximum gama exposure; Revision 3 Cate _

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Approved Bv: 6 _ d 3-40 l 4172R

I

2. The point of maxtrum ground level air concentration of radionuclides; and l

3. The point of maximum ground level air concentration of radionuclides where a real milk animal exists.

The point of maximum gamma exposure (S sector, 400 meters) was determined by finding the maximum annual average gamma X/Q at any off-site location. The location of the maximum ground level air concentration of radionuclides (WNW sector, 2415 meters) was determined by finding the maximum annual average undepleted X/Q at any off-site location. The point of maximum ground level air concentration of radionuclides where a real milk animal exists (WNW sector, 3000 meters) was determined by finding the maximum I depleted X/Q at a real milk animal location.

3.10.2 Vermont Yankee Atmospheric Dispersion Model The annual average atmospheric dispersion factors are computed for routine (long-term) releases using Yankee Atomic Electric Company's (YAEC)

AEOLUS Computer Code (Reference B). AEOLUS is based, in part, on the straight-line airflow model discussed in Regulatory Guide 1.111 (Reference C). The valley in which the plant is located is considered by the model.

I AEOLUS produces the following annual average atmospheric dispersion factors for each location:

1. Undepleted X/Q dispersion factors for evaluating ground level concentrations; 5 2. Depleted X/Q dispersion factors for evaluating ground level concentrations; I

I DayC 2 2 60O I Revision '3 3-41 Approved By o

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

and I

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

Gamma dose rate is calculated throughout this ODCM using the finite cloud model presented in " Meteorology and Atemic Energy - 1968" (Reference E.

Section 7-5.2.5). That model is implemented through the definition of an I effective gamma atmospheric dispersion factor, CX/QY ] (Reference 8, Section 6), and the replacement of X/Q in infinite cloud dose equations by the I [X/QY ),

3.10.3 Annual Averace Atmospheric Discersion Factors for Receptors Actual measured meteorological data for the five-year period, 1978 through 1982, were analyzed to determine all the values and locations of the maximum off-site annual average atmospheric dispersion factors. Each dose and I dose rate calculation incorporates the maximum applicable off-site annual 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 varicus doses or dose rates of interest.

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{ r l Ravi::en J oat? Appreved By: dD

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3-42 I 4172R I .-

m M M M M M M M M M M M M M M M Table 3.10-1 V_e rmon t Yankee Dilution Factors Dose to Critical Dose Rate to Individual Dose to Air Organ Total Body Skin . Cr-itical Organ Camma Beta _Thygold X/Q depleted (**3 ) - -

4.6E-07 ,

4.6E-07 m

X/Q undepleted (* ) -

6.3E-07( - -

6.3E-07 2) ,

a D/Q ( ) - -

8.9E-10 - -

8. 9E- 10(

m X/QI( ) 7.2E-07 I' 7.2E-07( ' -

7.2E-07 - -

(1) Maximum gamma exposure point: S sector 400 meters (0.25 miles).

(2) Maximum ground level concentration: WNW sector, 2415 meters (1.50 miles).

(3) Worst real milk animal concentratica: WNW sector 3000 meters (1.86 miles),

l' g,

3-43 s :

Approved Byif A Y k

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Date He vision _ _ . _ _ .

y

3.11 Method n Calculat9 Dose from Plant Operntion Technical Specification 3.8.M.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 the turbine and reactor building) to 25 mrem in a calendar year (except the thyroid, which is limited to 75 mrem).

I 3.11.1 Method The maximum contribution of direct dose to the whole body or to any organ due to N-16 decay from the turbine is:

Dd = 1.29E-06 E (arem) (mrem) (Web) where:

E = gross electric output over the period of interest (MWe b) 1.29E-06 = (2.14E-06 mR/MW,h) (0.6 mrem /mR)

In addition to the calculations of direct dose due to N-16 decay from the turbine hall, direct dose contribution to off-site receptors from iden-tified fixed sources, such as solid waste storage facilities, shall be included in the estimate of total direct dose off-site to any member of the public. Analytical design calculations of direct dose from these sources can be used in lieu of direct measurements, such as with a high pressure ion chamber.

Revision 3 Dahb2219$ Approved By: c 3-44 W

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I 3.11.2 Basis of Method The major scurce of direct radiation (including sky shine) from the station during normal operation is due to N-16 decay in the turbine building.

Because of the orientation of the turbine building on the site, and the shleiding effects of the adjacent reactor building, only the seven westerly sectors (SSW to NNW) see any significant direct radiation.

High pressure tonization chamber (HPIC) measurements have been made in the plant area in order to estimate the direct radiation from the station.

The chamber was located at a point along the west site boundary which has been determined to receive the maximum direct radiation from the plant. Using measurements of dose rate made while the plant operated at different power levels, from shutdown to 100 percent, a gross electric output to dose rate conversion factor of 2.14E-06 mR/MWh has been derived at the location of the I nearest resident based on measurements at the site boundary. Fleid measurements of exposure, in units of Roentgen, were modified by multiplying by 0.6 to obtain whole body dose equivalents in units of rem in accordance with recommendations of HASL Report 305 (Reference F) for radiation fields resulting from N-16 photons.

Therefore, knowing the gross megawatts generated during the period of interest, one may obtain an estimate of the maximum dose from direct radiation '

at the nearest residence through the use of Equation 3-27 due to N-16 decay.

I The direct dose contribution from other significant sources are accounted for by analytical calculations of source strength, geometry, and shielding factors i

as they relate to off-site receptors,

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lI Revision 3 Da Approved By:

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,3.12 Cumulative Doses Cumulative Doses fcr a calendar quarter and a calendar year must be

maintained to meet Technical Specifications 3.8.B.1, 3.8.F.1 and 3.8.G.I. In addition, if the requirements of Technical Specification 3.8.H.2 dictate, cumulative doses over a calendar year must be determined for Technical Specification 3.8.M.1. To ensure the limits are not exceeded, a running total must be kept for each release.

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I i I

, .Revtston 1 Date' Approved By:. t 3-46 4172R i

I 5.2 Gaseous Effluent Instrumentation Setpoint Technical Specification 3.9.B.1 require < that the radicactive 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 3.8.K.1 are not exceeded. Technical Specification 3.8.E.1.a limits the activity concentration in off-site gaseous effluents to well below the appropriate MPCs in 10CFR20 by limiting the whole body and skin dose rates to areas at or beyond the site boundary. Technical Specification 3.8.K.1 limits the gross radioactivity release rate at the steam jet air ejector (SJAE) to 0.16 C1/sec.

5.2.1 Plant Stack Noble Gas Activity Monitors (RR-108-1A and RR-108-18) and I 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 I Activity Monitors (RR-108-1A and RR-108-18) and the Augmented 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 l 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. ,

l denoted R spt. R spt is the lesser of:

R - 694 S g h DFB g (5-9) 3 3 see (cpm) (mrem-pC1-m } (com-cm } (l} ( pC1-yr' F -pCl-sec pCl cm mrem-m 3

l Reviston 3 Datei. Approved By: L, Shr _ _' -

5-9 4773R I

I

and:

R's t - 3,000 S g h b (5-10) 3 (cpm) see uCl (mrem) yr (com-cm pCl I D mrem-yr sec '

Cm where:

- Response of the monitor at the limiting total body dose I

R st rate (cpm)

$0 694 = -

c mrem-uCdm) yr-pCl-sec (1E+06) ( '/ . 2 E-07 ) i I 500 - Limtting total body oose rate (mrem /yr)

I IE+06 - Number of pCl per pCl (pCi/pCl) 7.2E-07 = [X/Q]Y, maximum annual average gamma atmospheric dispersion factor (sec/m3)

- Appropriate (plant stack or A0G system) detector I

S g

counting efficiency l (cpm /(pCl/cc))

F = Appropriate (plant stack or A0G system) flow rate (cm3/sec)

DFB g - Composite total body dose factor (mrem-m3 /pCl-yr) h,OFB,

. (5-11) bi I Revision U Date ' - ApprovedBy:]/~

l8

'O Q(

c

'Tg 5-10 4773R

l L

F u -

hg - The relative release rate of noble gas "1" in the mixture atthemoniter(eitherthestack,hST ortheACG,h )

l L for noble gases identified (pCl/sec)

DFB g - Total body dose factor (see Table 1.1-10) (mrem-m3 /pCl-yr) 3 "

R 3

= Response of the monitor at the ilmiting skin dose rate (cpm) ,

I L

3,000 - Limiting skin dose rate (mrem /yr) -

I 0F' = Composite skin dose factor (mrem-sec/pCl-yr) hg 0F{

= (5-12) bg i

I 0Fj - Combined skin dose factor (see Table 1.1-10)

(mrem-sec/pCl-yr) l 5.2.1.2 Plant Stack Noble Gas Activity Monitor Setpoint Example l

I The following setpoint example for the plant stack noble gas activity i

monitors demonstrates the use of equations 5-9 and 5-10 for determining setpoints.

I The plant stack noble gas activity monitors, referred to as " Stack Gas I" (RR-108-1A) and " Stack Gas II" (RR-108-18), consist of beta sensitive I 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 factors, S , gof IE+08 cpm per 1 pC1/cc of noble f

1 1e t Revision -.3 Date Approved By: _ ,ff_ i-5-11 4773R

lI '

gas. The ncminal plant stack flow is 7.5E+07 cc/sec ((160,000 cfm x 28,300 I 3 cc/ft )/60 sec/ min). l 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 simple 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 DFBg andDFj):

hSJAE OFB, 3 OFj I i guC1) sec (mrem-m DCi-yr ) (mrem-sec) uCl-yr

'-i 8 i.03E+04 8.83E-03 9.97E-03 iE 5.92E-03 1.11E-02

!E Kr-87 4.73E+02 Kr-88 2.57E+02 1.47E-02 1.37E-02 g

Kr-85m 1.20E+02 1.17E-03 1.90E-03 E

Xe-135 3.70E+02 1.81E-03 2.69E-03 Xe-133 1.97E+01 2.94E-04 4.75E-04 I DFB =

p .SJAE L., Q g 0FB 3

' (5-11) g .SJAE 01

~

.SJAE Q3 0FBg - (1.03E+04)(8.83E-03) + (4.73E+02)(5.92E-03)

I

! + (2.57E+02)(1.47E-02) + (1.20E+02)(1.17E-03)

+ (3.70E+02)(1.81E-03) + (1.97E+01)(2.94E-04) i 3

= 9.83E+01 (pC1-mrem-m /sec-pC1-yr) {

Revisten 3 Dah Approved By T 5-12 I 4773R

hf ^ = 1.03E+04 + 4.73E+02 + 2.57E+02

+ 1.20E+02 + 3.70E+02 + 1.97E+01

- 1.15E+04 pC1/sec 9.83E+01 0FB c = 1.15E+04

= 8.52E-03 (mrem-m3 /pC1-yr) i R t' 094 b g 0FB (5-9) c E

= 694) +08) '

(7.5E+07) (8.52E-03)

= 109,000 cpm -

I Next:

.SJAE Og 0Fj OF' = (5-12) t 7 .SJAE LQ 0Fj=(1.03E+04)(9.97E-03)+(4.73E+02)(1.11E-03) g tr + (2.57E+02)(1.37E-02) + (1.20E+02)(1.90E-03)

+ (3.70E+02)(2.69E-03) + (1.97E+01)(4.75E-04)

= 1.08E+02 (pCl-mrem-sec/sec-pCl-yr)

I 0F.

e 1.08E+02 1.15E+04

' Revision 3 Date Approved By: h -

5-13 I 4773R I

L E

- 9.39E-03 (mrem-secluct-yr) l 1

S'" - 3.C00 S (5-10)

R 3pg gh

- (3,000) (IE+08) (7.5E+07) (9.39E-03)

- 426,000 cpm 5 "

and R The setpoint, Rspt, is the lesser of t st*

For the noble gas mixture in this example R spt is less than .

R t , indicating that the 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 109.000 cpm above background or at some conservative value below this (such as that which might be based on I controlling release rates from the plant in order to maintain off-site air i concentrations below 2 x MPC when averaged over an hour).

5.2.1.3 Basis for the Plant Stack and AOG' 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 3-7 are acceptable methods for determining compliance with that Technical Specification. Which equation (i.e., dose to total body or skin) is more limiting depends on the noble gas 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 C ,g (5-13) gg 3

(cpm)- (com m) ( )

Cm '

m l C Revision 3 Date Acoroved By: L( /._ _

\

4773R I

t

L F -

where:

R - Response of the instrument (cpm) i 3

( 59 , = Detector counting efficiency for noble gas "1" (cpm /(pC1/cm ))

C ,g - Activity concentration of noble gas "1" in the mixture at the 3

noble gas activity monitor (pC1/cm )

The relative release rate of each noble gas, Qg (pC1/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 4 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, "1", mix fraction, (f ),g may be taken from Table 5.2-1 as a function of time after shutdown (including periods shorter than five days) to determine the relative fraction of each noble gas potentially !

i available for release to the total. However, prior to plant startup before a f 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 conditionsmaybeutilizedatanytimeinlieuoftheaboveassumptions.

Restsion 3 Date Approved Byi

~-

b '

5-15 4773R r

L-I 5 ,C ,g, the activity concentration of noble gas "1" at the noble gas activity ,

monitor, may be expressed in terms of Qg by dividing by F, the appropriate i 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.

(5-14)

C,g - hg h Ci <gC1)<sec) d '

L

($m>

C Cm I

where:

L Q, - The release rate of noble gas "1" in the mixture for each noble gas identified. ,

r' F - Appropriate flow rate (cm3 f3,c)

Substituting the right half of Equation 5-14 into Equation 5-13 for C ,3 yields:

(5-15)

R - S gg hg h (cpm) (CDm m) g )( )

Cm 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 59 for 59g, where Sg represents the counting efficiency determined for the current mix of nuclides. If the mix of nuclides changes significantly, a new counting efficiency should be determined for calculating the setpoint.

.l .

7 c C Revision 3 DathEt 2 21986 ApprovedBy:[ c_

E a ~

,,,3, d

f - - - - -

~

R = 5 9 h 1 hg (5-16) l (Cpm) ( Dm-cm ) gg) gg) pC1 g,3 sec L,

The total body dose rate due to noble gases is determined with Equation 3-5: . i r !

k = 0.72 0F8 g (3-5) tb Qg I 3 (mrem) (DCl-sec) guCl) gmrem-m )

yr Cl-m 3 5'" P '-Y' l

" where:

k = total body dose rate (mrem /yr)

L tb 0.72 . (i.0E.06) x (7.2E-07) (pCi-secipci-m3) p IE + 06 - number of pCl per pCl (pCl/pC1) 7.2E - 07 = (X/Q]Y, maximum annual average gamma atmospheric dispersion factor (sec/m 3) l.

I Qg - the release rate of noble gas "1" in the mixture for each noble gas identified (pC1/sec) (Equivalent to l .ST Qg for noble gases released at the plant stack.)

0FB, - total body dose factor (see Table 1.1-10)

(arem-m3 /pCl-yr)

Revision 3 Date

'.' . '4 Approved By: es m C t ~ _

5-17

, 4773R 4

A. composite total body gamma dose factor, OFB g

, may be defined such that; OFB g hg - hg 0FB, (5-17) 3 3 mrem-m guC1) guC1) gmrem-m )

y pC1-yr see sec pCl-yr Solving Equation 5-23 for DFB yleids:

1

? hg0FB g .

I OFB - (5-11) 1 o

Technical Specification 3.8.E.1.a limits the dose rate to the total f body from noble gases at any location at or beyond the site boundary to 500 mrem /yr. Bysettingk g equal to 500 mrem /yr and substituting DFBg for DFB g l In Equation 3-5, one may solve for 0, at the limiting whole body noble gas dose rate:

~

h- g 694 0FB (5-18) l

\ c .

L C (mrem-uci-m (v.Ci )

sec yr-pCl-sec 3 (oCi-vr mrem-m 3

I Substituting this result for Q in Equation 5-16 yields R to the response i

of the moedtor at the limiting noble gas total body dose rate:

l R 094 S t" g 0FB e

3 3 s (cpm) (mrem-pC1-m } com-cm ' (see' Ioct-yr '

yr-pti-sec pC1 3 3 cm mrem-m s Revision d Date Approved By: 'a v\ -Cr -

(-"y 5-18 4773R I

I The skin dose rate due to noble gases is determined with Equa21on 3-7:

" (3-7) k5 - hg 0Fj gmrem) (gCl) (mrem-sec) yr sec pCl-yr I where:

R

" = Skin dose rate (mrem /yr) hg -Thereleaserateofnoblegas"1"inthemixtureforeach g noble gas identified (pC1/sec) (Equivalent to Qg for noble gases released at the plant stack.)

DFj - Combined skin dose factor (see Table 1.1-10) (mrem-sec/pCl-yr)

A composite combined skin dose factor, OF, may be defined such that:

E' DF' hg = hg DFj (5-19) i 1 (mrem-sec) (uCl) (uCl) (mrem-sec) pCl-yr pCl-yr sec sec Solving equation 5-19 for DF' yields:

hj OFj I

DF' = (5-12)

Q, t '

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.

Bysettingk5 "squalto3,000 mrem /yrandsubstituting0F'forDFjin i Equation 3-7 one may solve for [ Qg at the limiting skin noble gas dose rate:

I

., CN

\

Revislon 3 Date Approved By: L _i f _,

5-19 4773R

b i

T 1 s . z_.6,.3,cco op. (s-20)

I c

( (fl y gmrem) uCl-yr gmrem-sec) sec yr d

i Substituting this result for hgin Equation 5-16 yields R I", the response t

of the monitor at the limiting noble gas skin dose rate: -

S " - 3,000 R, S g h 0 (5-10) ,

3 CP"I (mremI (com-cm } (see' I uCl-yr yr pct g,3 mrem-sec) c

[

Revision d Date Approved By: %

at / h d' c'

5-20 4773R

W WM-Q W W W W WT _F W 7 FW7 FL _ J U U U W~ '

TABLE 5.2-1 )

Relative Fractions of Core Inventory .

Noble Gases After Shutdown Kr-85m Kr-85 Kr-87 Kr-88 Xe-131m Xe-133m Xe-133 Xe-135m Xe-135 Xe- 138 Ilme Kr-83m l

.001 .083 .118 .002 .010 .306 .061 .093 .263 t < 24 h .02 .043

.004 .001 .004 .022 .758 .010 .198 --

24 hr 1 t < 48 h -- .003 --

.005 -- -- .006 .024 .907 .001 .058 --

l 48 h 1 t < 5 d -- --

.007 -- -- .008 .016 .969 -- -- . - -

5 d 5 t < 10 d -- --

-- .014 -- -- .014 .006 .966 -- -- --

10 d 1 t < 15 d --

l

.026 -- .002 .002 .950 -- -- --

l 15 < t < 20 d -- -- --

.048 -- -- .034 .001 .917 -- -- --

20 i t < 30 d -- --

-- .152 -- -- .070 -- .777 -- --

30 i t < 40 d --

.378 -- -- .105 -- .517 -- --

401 t < 50 d -- --

-- .652 -- -- .108 -- .240 -- -- --

50 1 t < 60 d

-- .835 -- -- .083 -- .082 -- -- --

60 1 t < 70 d

.920 -- -- .055 -- .024 -- --

t2 70 d -- --

\

" Approved By: l_b Date DEC 2 2 1986 Revision cgsm -

V 5-21 4783R

l L

I s

I b

I

)

h APPENDIX A METHOD I EXAMPLE CALCULATIONS i

Revision 3 Date ~ Approved By: ,

5103R

I EXAMPLE CALC'1LATION NO.1 Type Total Body Dose From Liquid Effluents I

I References E

F a) ODCM Sections 3.2 and 3.3 (Method I).

i b) Technical Specifications 3.8.B.1 and 3.8.C.1.

l Problem Calculate the off-site dose to the total body and maximum organ resulting from f the batch release of radioactive liquid effluents.

5 Plant Data

[

i 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:

f DFL

D R g ,o**

Activity Qg itb Released (C1) (mrem /Ci) (mrem /Ci) l 1 3.40 E+00 1.75 E-04 1.75 E-04 H-3 1.72 E-03 8.27 E-02 2.67 E-01 l Co-60 3.51 E-03 5.26 E+00 7.01 E+00 Co-134 7.20 E-02 2.52 E-03 1.46 E+00 I-131

Total body dose factor from Table 1.1-11.

- Maximum organ dose factor from Table 1.1-11.

( r c '

s A1 Approved By_ [f -

Rev!rlon 1 Date L '

'J 5121R

, y c,21ation

( The total body dose is calculated from Equation (3-1):

D "

tb S i "i tb (3-1)

(ares) (Ci) (arem/Ci)

H therefore:

Dtb = (3.40 E+00)(1.75 E-04) + (1.72 E-03)(8.27 E-02) +

(3.51 E-03)(5.26 E+00) + (7.20 E-02)(2.52 E-03) = ,

Answer (1)

Dtb = 1.94 E-02 mrem to the total body.

Next, the maximum organ dose is calculated from Equation (3-3):

D,, . = Qg Dn ,, g (3,3) i (ares) (Ci) (mrem /C1) r therefore:

D = (3.40 E+00)(1.75 E-04) + (1.72 E-03)(2.67 E-01) +

(3.51 E-03)(7.01 E+00) + (7.20 E-02)(1.46 E*40) =

Answer (2)

D , = 1.31 E-01 ares to maximum organ.

f .m R* vision 3 Date' t. - 2 Approved By L _%

5121R l

t i EXR4PLE CALCULATION NO. 2 Typ_e

! Total Body Dose Rate From Noble Cases References a) ODCM Section 3.4 (Method I).

b) Technical Specification 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 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): 7.55 E+07 cc/sec (160,000cfmx4.72E+02*[g'"= )

c) Plant stack monitor detector counting i E+08 cpm per efficiency (Sg): uCi/cc d) The last measured release rate six of f .SJAE

\ noble gas from the SJAE (Q g ),

and corresponding dose factor DFB g from Table 1.1-10.

Revision 3 n 1! yCr P Datt'_EG 3 2 1986

, Approved Byt ,k %__ ~v A-3 5103R

l B

i .SJAE DFB Oi i (uci/sec) (mrem-m 3 /pci-yr)

Xe-138 5.15 E+03 8.83 E-03 Kr-87 2.37 E+02 5.92 E-03 Kr-88 1.29 E+02 1.47 E-02 Xe-135 1.85 E+02 1.81 E-03 Calculation l

The dose rate is calculated from Equations (3-5) and (3-28): ,

i I

( -$}

l R

tb = 0.72 i Q DFB g mrem yr (pCi-see uCi-m 3 W uCi mrem-m pCi-yr l

and where the stack release rate is determined frcm:

I

.ST Qg (3-28) g

" " F I S i .SJAE Sg 01 g (cpm) (uCi/ce) (ce y cpm sec

(" )

..SJAE First,determinethesum([)ofallQ g and the fraction that each noble i

gas i represents in the total gas mix.

>t .SJAE Qg = (5.15 E+03) + (2.37 E+02) + (1.29 E+02) + (1.85 E+02)

= 5.70 E+03 uCi/see I

and the relat.ive fraction of each noble gas:

Revision 3 Dael " O approved ny:

~

Es -

A-4 5103R I

)

i

^

Relative Fraction

/5.70 E+03

{ g Qg of Total Xe-138 5.15 E+03/5.70 E+03 = 0.904 Kr-87 2.37 E+02/5.70 E+03 = 0.042 Kr-88 1.29 E+02/5.70 E+03 = 0.023 Xe-135 1.85 E+02/5.70 E+03 = 0.032 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:

i 1

M F DFB g ktb = 0.72 gg g f

g

= 0.72 80.000 1/1E+08 7.55 E+07 [f g g DFB g

= 4.35 E+04 ((0.904)(8.83 E-03) + (0.042)(5.92 E-03) +

(0.023)(1.47 E-02) + (0.032)(1.81 E-03)] =

Answer Rg = 375 mres/ year noble gas total body dose rate.

i

(

Revirion O

, c c Date, __

Approved By: ( --

A-5 V 5103R

EXAMPLE CALCULATION NO. 3 IZE1

/ Total Body Dose Rate From Noble Cases L

References I

a) ODCM Section 3.4 (Method I).

l b) Technical Specification 3.8.E.1.a.

Problem s

b. Calculate the off-site total body dose rate resulting from the release of 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): 7.55 E+07 ce/sec l

(160,000 cfm x 4.72 E+02 cc eg sec , )

c) Plant stack monitor detector counting 1 E+08 cpm per

, efficiency (Sg): uCi/cc d) The noble gas mix fractions f (t) g corresponding to 32 days taken from Table 5.2-1.

g - * ?, 2 666 C

]bN,y e,

Revision Date Approved By: S -

~

A-6 5103R r

day)* DFBg **

i i(

Kr-85 0.152 1.61 E-05 Xe-131m 0.070 9.15 E-05 Xe-133 0.777 2.94 E-04

[

Fraction of nuclide in six as function of time (see Table 5.2-1).

/

( ** Dose factors from Table 1.1-10.

t Calculation The dose rate is calculated from Equations (3-5) and (3-28):

(3-k 0.72 DFB g

( eb = i Q

(eres yr (DCi-sec uCi-e 3 . (uCiM (arem-m pci-yr and where the stack release rate is de*. ermined from:

.SJAE (3-28)

.ST Qg g Oi

F

.SJAE Sg i Qg i

uC (eps) e guci/cc) g _e_)

( g _s e e.1 ) cpm sec

.SJAE However, for a time (t) after shutdown, the ratio of Qg to the sum release rate of a11' noble gases can be replaced in Equation (3-28) by the relative fraction (f (t)] g of each noble gas available in the system; therefore, Equation (3-28) can be written:

/

Pevision 3 Date _

Approved By: _

A-7 5103R r

5 .ST '

Qg = f g(t) M F 3,

Therefore, using the above data for a time period 31 days after shutdown, the dose rate equation can also be written as:

M F DFB g ktb = 0.72 3, f g (t)

I = 0.72 80.000 1/1E+08 7.55 E+07 ((0.152)(1.61 E-05) +

(0.070)(9.15 E-05) + (0.777)(2.94 E-04)] =

Answer I ktb = 10.3 arem/ year noble gas total body dose rates at 32 days after shutdown.

I I

E ,

3 ,e isi,n a Oaen

.,,,o.e ,, 35

'J g $103R I . . .

l EXAMPLE CALCUI.ATION NO. 4 i

l Skin Dose Rate From Noble Cases References a) ODCM Section 3.5 (Method I).

l b) Technical Specification 3.8.E.1.a. -

1 Problem 1

Calculate the off-site skin dose rate resulting from the release of noble I gases from the plant stack during power operations.

Plant Data 1

a) Maximum plant stack gas monitor (I or II)

Count rate during period of interest (M): 80.000 cpm l

B b) Stack flow rate during release (F): 7.55 E+07 cc/sec (160.000 cfm x 4.72 E+02 * = )

l 1 E+08 cpm per c) Plant stack monitor detector counting efficiency (Sg): uCi/cc d) The last measured release rate mix of

.SJAE noble gas from the SJAE (Q g ).

I and corresponding dose factor DF from Table 1.1-10.

~~

Revision 3 Date Approved By: ,e.

5103R I

8

.SJAE DFj 5 Qg (mrem _,eeI I (uci/see) uCi/ year Xe-138 5.15 E+03 9.97 E-08 Kr-87 2.37 E+02 1.11 E-02 Kr-88 1.29 E+02 1.37 E-02 Xe-135 1.85 E+02 2.69 E-03 I Calculation The skin dose rate is calculated from Equations (3-7) and (3-28): .

.ST R Fg (3-7)

Oi akin " i mrem) guci) gerem-see)

I yr see uCi-yr and where the stack release rate is determined from:

.SJAE

.ST (3-28)

I Qg g Oi "

.SJAE E 01 (cpm) g guci/ce) gec_)

cpm sec I uCi) sec

.SJAE First,determinethesum([)ofallQ g and the fraction that each noble i

gas i represents in the total gas mix.

I .SJAE Qg = (5.15 E+03) + (2.37 E+02) + (1.29 E+02) + (1.85 E+02)

= 5.70 E+03 uC1/sec.

and the relative fraction of each noble gast I

C Ruvision 0 DEC 2 2 1986 s~,

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/5.70 E43 Relative Fraction i

91 ,, 7,,,1 Xe-138 5.15 E+03/5.70 E+03 = 0.904 Kr-87 2.37 E+02/5.70 E+03 = 0.042 Kr-88 1.29 E+02/5.70 E+03 = 0.023 Xe-135 1.85 E+02/5.70 E+03 = 0.032 Next, the stack release rate of each noble gas i fro. Equation (3-28) can be substituted into Equation (3-5) to give the skin dose rate as:

I F f DF g skin " Sg g g

= 80,000 1/1E+08 7.55 E+07 [(0.904)(9.97 E-03) +

(0.042)(1.11 E-02) + (0.023)(1.37 E-02) + (0.032)(2.69 E-03)]

= 6.04 E+04 (9.01 E-03 + 4.66 E-04 + 3.15 E-04 + 8.61 E-05)

= 6.04 E+04 (1.03 E-02) b Answer C .

E,,,, = 625 .,e.,, ear no,1e ..s s,in e se rate.

E

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c Revision _3 _ Date I' _'"

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EXAMPLE CALCULATION NO. 5 T_ype 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 gases from the plant stack six days after plant shutdown.

{

Plant Data 1

a) Maximum plant stack gas monitor (I or II) l Count rate during period of interest (M): 120,000 cpm I

l b) Stack flow rate during release (F): 7.55 E+07 cc/sec (160,000cfmx4.72E+02**g[**= ,

)

1 c) Plant stack monitor detector counting 1 E+08 epm per efficiency (Sg): uCi/cc d) Since the plant is shut down for more than five days Xe-133 may be used as the referenced radionuclide in place

.SJAE of the ratio of Q g to the sum

.SJAE of all Qg in Equation (3-28). ;

7 Approved By:_

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f i (t 25 days) ("Ci-year) u Xe-133 1. 4.75 E-04 Calculation The skin dose rate is calculated from Equations (3-7) and (3-28):

( "II k S skin

i i DF{ ,

(area) (uy) (arem-see) yr sec uCi-yr -

and, the stack release rate is determined from

.SJAE

.ST (3-28) i Qg g Si " "

( JAE Sg i

However, for times greater than five days af ter shutdown, Xe-133 may be used y as the referenced radionuclide alone. Therefore, in Equation (3-28) the ratio

( .SJAE .SJAE of Q g to the sum of all Q g can be replaced by a value of 1 which indicates that all the contribution to the release is from Xe-133.

Therefore:

.ST QXe-133 = 1.0 x 120,000 x 1/1E+08 x 7.55 E+07

( (cpe) (cc/sec)

(uci/ec) cpe h

.ST

{ Q,,_,,, . ,0. 00 uC u s.. .

Revision '3 Date, , Approved By: __ /S QC %_'

A-13 5103R I -- - _ - - - - -

.ST

( Therefore, replacing this value of Q g into Equation (3-7) we find the skin dose rate as:

?

L j k,gg, = 90,600 x 4.75 E-04 l

5 g erem y yr gy) u sec (arem-see) uCi-yr Answer kskin = 43.0 arem/ year. ,

B l

l l

1 l

1 0

1 I m Revision d Dste t.pproved By: t _J N N- --

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k EXAMPLE PROBLEM NO. 6 TY.E1 l

Critical Organ Dose Rate From Iodine, Tritium, and Particulates References 8 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 offluent 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.

.STP DFC'

^** "I 0 1 gerem s c) i (uci/see) yr-uci Sr-89* 1.42 E-04* 3.48 E+01 Sr-90* 3.50 E-03* 1.34 E+03 ,

Co-60 4.89 E-02 2.26 E+01 l

Co-137 3.90 E-03 8.45 E+01 Zr-65 1.01 E-02 3.55 E+00 Na-24** 2.76 E-03** -

Mn-54+ < 2.8 7 E-06 + 1.45 E+00 (1) DFCico dose rate facter for each radionuclide is taken from -

Table 1.1-12.

> / r

f Ravision 1 Dee $ 2 3 N Approved By: $# d- '"

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$otes

For Sr-89/90, use the most recent available measurement from quarters I 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 I even though it was detected.

+Mn-54 is not included in the dose analysis since it was not detected as being present above the LLD, I

b) Stack iodine (charcoal and particulate activities combined for the seven-day period of interest):

I .STP DFG' Activity Q i mrem c i (uci/sec) I yr-uci I-131 1.16 E-03 2.64 E+02 I-133* <6.35 E-05* 4.04 E+00 I-135** 7.21 E-03** 3.69 E-01 and H-3+ 3.17 E-02 5.36 E-03 I

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.

- 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 I Technical Specification 3.8.E.1.b.

+ Tritium value based as latest available stack grab sample.

S I ,e.i. i_ 2 o.te _ _.aa

.,, rov.d ,,

3db_1

,10>.

I -- - _ _ --

B Calculation I The dose rate is calculated from Equation (3-16):

I k,=

g h DFGy (3-16)

(arem) yr y) u sec arem-sec) uCi-yr I The dose rate factors (DFG' g ) for each of the radionuclides detected in the plant stack charcoal and particulate filter sample (plus tritium) is taken I from Table 1.1-12 of the ODCM.

Therefore:

kg, = (1.42 E-04)(3.48 E+01) + (3.50 E-03)(1.34 E+03) + (4.89 E-02)

(2.26 E+01) + (3.90 E-03)(8.45 E+01) + (1.01 E-02)(3.55 E+00) 8 + (1.16 E-03)(2.64 E+02) + (3.17 E-02)(5.36 E-03) =

Answer I kg, = 6.47 mrem / year critical organ dose rate f rom iodine, tritium, and particulate.

I I

8 9 ,. -

Revision 3 Date Approved By: Kz ,

J 5103R I

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EXAMPLE PROBLEM NO. 7 IIPS Casuna Air Dose From Noble Cases References a) ODCM Section 3.7 (Method I).

E b) Technical Specification 3.8.F.1.

b Problem L Calculate the maximum gamma air dose resulting from noble gases released from the plant stack over a calendar month.

Plant Data L

Based on the daily off-gas analysis, the total activity released during the l

month of interest is:

DF{*

I Activity Q i

(mead-m pci-yr )

1 (C1) l Kr-88 3.55 E-01 1.52 E-02 5 Kr-85m 4.71 E+00 1.23 E-03 Xe-138 2.75 E+00 9.21 E-03 Xe-135 3.51 E+01 1.92 E-03 Xe-133 9.42 E+01 3.53 E-04

Gasuna air dose f actors taken f rom Table 1.1-10.

.. - , }

Revis* n 3 Due Apptovei By: c_ j V

A-18 5103R

L Calculation The maximum gasuna air dose of f-site is calculated from Equation (3-21):

H Y Y

( D,g, = 0.023 g Qg DF g 3

(3-21)

(arad) goci-yr) (Ci) (arad-m )

Ci-e 3 PCi-yr Therefore:

Y D,g, = 0.023 [(3.55 E-01)(1.52 E-02) + (4.71 E+00)(1.23 E-03) -

F h + (2.75 E+00)(9.21 E-03) + (3.51 E+01)(1.92 E-03) +

(9.42 E+01)(3.52 E-04)]

= 0.023 (5.40 E-03 + 5.79 E-03 + 2.53 E-02 + 6.74 E-02 +

3.31 E-02)

Answer Y

D,g, = 3.15 E-03 stad gamma air dose during the month.

E C

Lt c

Revielon 3 Da:e DEC 2 2 1986 Approved By: SU(x A-19 5103R r - - - - - - - - - - -

L f

, EXAMP'_E CALOULATION NO. 8 T,,Ipe

( Beta Air Dose From Noble Gases 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 given in Example Calculation No. 7.

Plant Data From Example No. 7, the total activity determined to be released during the month is:

DF g

Activity Q i mead-m i (Ci) pCi-yr Kr-88 3.55 E-01 2.93 E-03 Kr-85m 4.71 E40 1.97 E-03 l

Xe-138 2.75 E40 4.75 E-03 Xe-135 3.51 E41 2.46 E-03 Xe-133 9.42 E41 1.05 E-03

Beta air dose factors taken from Table 1.1-10.

[ ,m Revistan d Cate Approved By: i -

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I Calculation I

The maximum beta air dose off-site is calculated from Equation (3-23):

I

=

/2 D

air

= 0.02 Qg DF g (3-23) 3 (arad) (pci-yr) (C1) (arad-m )

3 PCi-yr Ci-m Therefore:

I D air

= 0.02 [(3.55 E-01)(2.93 E-03) + (4.71 E+00)(1.97 E-03)

+ (2.75 E+00)(4.75 E-03) + (3.51 E+01)(2.46 E-03) +

(9.42 E+01)(1.05 E-03)]

= 0.02 (1.04 E-03 + 9.28 E-03 + 1.31 E-02 + 8.63 E-02 +

9.89 E-02)

Answer I D 4

air

= 4.17 E-03 mrad beta air dose during the month.

I I

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'l nevision 3 cate Approved sy:

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EXAMPLE PROBLEM NO. 9 IZE!

Critical Organ Dose From Iodine, Tritium. and Particulates E a) ODCM Section 3.9 (Method I).

b) Technical Specification 3.8.G.I.

l Problem 1

Calculate t% critical organ dose due to the total activity recorded as being released frem the plant stack during a calendar month.

Plant Data a) From the combined stack analyses during the month, the following activity released ist DFC ge, (L}

l 1

I i _ (Ci) (mreta)

C i ___

l St-89* 5.42 E-04* 1.10 E+00 Sr-90* 1.10 E-02* 4.25 E+01 Co-60 2.30 E-01 7.15 E-01 Cs-137 1.15 E-02 2.68 E+00 2n-65 2.60 E-02 1.13 E-01 Ns-24** 7.11 E-03** -

Mn-54 <2.76 E-06+ 4.60 E-02 5

Revision 3 , DateECCb $86 mi c' Approved ByI br> ,

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Notes for Plant Data a) Above (1) 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.

- Na-2t. has a half life of less than 8-1/2 days, and therefore is not included in accordance with Technical Specification 3.8.G.1.

I +Mn-54 is not included in the dose analysis since it was not detected as being present above the required 1.LD.

b) Total iodine release for the month based on the combined charcoal and particulate f11ter samples taken during the months DFG g ,,

I' '

i Ni (C1) (arem}

Ci I (

I-131 I-133 4.30 E-03 1.12 E-04*

8.36 E+00 1.28 E-01 I-135** 2.01 E-02**

H-3+ 0.15 1.70 E-04 I Calcula_ tion I The dose is calculated from Equation (3-25):

D,= g Qg DT0 ico (3-25) i <.r.., <0,, <.r..,0,)

I I

a ** l l * (

Revision d Date Aprroved By: 3e 5103R I

I

! Notes for Plant Data b) Above

In this case, I-133 vas found in one of the weekly stack samples to be present, and theretcre based on that value is included in the dose analysis.

- I-135 is 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.B.1.

+ Tritium value based on the monthly stack grab sample.

The dose factor (DFGg ) for each radionuclide detected in the plant stack charcoal and particulate filter sample (plus tritium) is taken from Table 1.1-12 of the ODCM. -

I

' Therefore:

I D = (5.42 E-04)(1.10 E+00) + (1.10 E-02)(4.25 E+01) +

(2.30 E-01)(7.15 E-01) + (1.15 E-02)(2.68 E+00) +

l (2.60 E-02)(1.13 E-01) + (4.30 E-03)(8.36 E+00) +

(1.12 E-04)(1.18 E-01) + (0.15)(1.70 E-04) =

Answer D , = 0.70 mrem maximum organ dose for the month.

I I

!I I

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l

$3 g Revislor. 3 Date ____

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Approved By: A T\ C(,'

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A-24 s

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J I

EXAMPLE CALCULATION NO. 10 IZPS P .

L Releases Limited to 2X MPC Average Over an Hour in Off-Site Air Concentrations I References L

F a) 10CFR50.72 L

- b) 10CFR50.73 L

Problem L

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.

Assucmtions a) Maximum expected stack flow rate (F) 8.73 E+07 cc/sec F

(185,000 cfm x 4.72 E+02 **g'*" = )

b) Plant stack monitor detector 1 E+08 counting efficiency (Sg) cpm per uCi/cc c) Maximum off-site ground level 6.3 E-07 3

dispersion parameter (X/Q undepleted) sec/m (from ODCM Table 3.10-1) d) Most restrictive MPC value for noble 2 E-08 gases (10CFR20, Appendix B. Table II. uCi/sec Column 1) for Kr-87 t q/-

~

Revision O Date Apprevt.o Ey: LI ;

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A-25 5103R c _- - - - - - - --

l L

f Calculntien (Part I)

L The setpoint R,pg for the stack mos.itor which would correspond to an L

instantaneous off-site air concentration of 2 x MPC for the most restrictive noble gas can be calculated by:

R =2 MFC Sg 1 E+06 I

E (cpe) (uC_i) coe sec cc (uCi/cc) ( g,3 ) (ce ) (cc )

,3 L -

= 2 x (2 E-08)(1E-08)(1/6.3 E-07)(1E +06)(1/8.73 E+07)

F L R = 72,700 cpm setpoint alarm value.

apt Now if the monitor alarmed at a setpoint of 72,700 cpm, an additional evaluation would be necessary to determine if the 2 x MPC limit for the actual

{ radionuclide gas six would be exceeded when averaged over one hour. As an example, assume the following:

[ Plant Data a) Recorded stack flow rate during release (F) 7.55 E+07 cc/sec (160,000 cfm x 4.72 E+02 e ec , )

b) Duration of release spike. 50 minutes c) Maximum recorded stack monitor 210,000 cpm

{ count rate.

[

[ . Revision 3 DateU C - ApprovedBy:os[ r IM C.e -

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d) Time trace of stack monitor response during release.

210,000 -

cpm

\

1,000 - - - - - = -

Background

t=0 t SO *1 hr I

l

~

Total counts = area under trace l

l Approximately 210,000 cpm x 50 minutes x 1/2 l

I Approximately 5,250,000 counts.

I Total counts averaged over one hour = '

= 8 , 00 cpm 6 mi ute I Since the average count rate over one hour still exceeds the instantaneous alarm setting of 72,700 cym, we must now look at the individual radionuclide concentrations off-site and see if the sum of the ratios of each nuclide concentrations over its MPC value is less than 2.0 in order to determine if the 2 x MPC notification rule had been reached.

Therefore, the six fractions (f )g that each noble gas represents of the total release must be determined f rom a representative of f-gas or stack release sample.

e) Assume the following fractions were determined:

I .

/. p Revision 3 Date -

Approved By: 1,3 -

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- Qg

+

guci) g a i ,

cc i (10CFR2Q Appendix B) e L Xe-133 3.50 E-04 0.40 3 E-07 uCi/cc Xe-133m 2.98 E-04 0.34 3 E-07

] Xe-135 Kr-88 1.23 E-04 1.04 E-04 0.14 0.12 1 E-07 2 E-08 Total 6.75 E-04 I

L Calculation (Part II) .

The air concentration off-site of each noble gas i can be found from:

L C g= f g

M gg F X/Q 10 guci) (cpm) guci/ce) 3 ce (c_e__

sec ) (sec) (" )

cpm ,3 -

-6 E .f 1 m .sooiu ns.o >n.sst.02>ce.3z-o23<1o 3 Cg=fg (4.16 E48)

I l

Therefore, using the fg from the table above we find:

I Cg i (uci/ce)

Xe-133 1.66 E-08 l Xe-133m 1.42 E-08 I Xe-135 Kr-88 5.83 E-09 4.99 E-09 I

I

..evision

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Finally, the sum of th2 ratio of cenc ntretiens to MPC ctn b3 fcund:

L -

Cg C C C3 2 3 ?

[ MPC g MPC 2

MPC 3

MPC g or

{

1.66 E-08 1.42 E-08 5.82 E-09 4.99 E-09 I * * + <2 3 E-07 3 E-07 1 E-07 2 E-08 L,

0.055 + 0.047 + 0.058 + 0.250 = 0.41 <2 l Answers F

L Since the sum of the ratios of the time averaged individual radionuclides air

{

concentrations over their MPC limits is less than 2, there is no reporting requirement under 10CFR50.72 or 10CFR50.73.

E E

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Revision 3 Dated 6C Approved By:

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I APPENDIX I RADIOACTIVE LIQUIIf, GASEOUS, AND SOLID WASTE TREAIMENT SYSTEMS Requirement: 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.

Response: There were no licensee initiated major changes to the radioactive waste systems (liquid, gaseous, and solid) during this reporting period.

I I

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I-l 4436R

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ML20212C860

Text

1.0 INTRODUCTION

1.0 INTRODUCTION

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