ML20246F005

Effluent & Waste Disposal Semiannual Rept for First & Second Quarters,1989

ML20246F005
Person / Time
Site:	Vermont Yankee
Issue date:	06/30/1989
From:	Tremblay L VERMONT YANKEE NUCLEAR POWER CORP.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
BVY-89-78, NUDOCS 8908300105
Download: ML20246F005 (45)

Text

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1. ERMONT YANKEE

NUCLEAR POWER CORPO' RATION ]

Ferry Road, Brattleboro,'VT 05301-7002 y

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ENGINEERING OFFICE 580 MAIN STREET DOL 1CN, MA 01740 August 25, 1989 <s0s> n9-6m BVY 89-78 United States Nuclear Regulatory Commission Document Control Desk Washington, DC 20555.

References:

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

Subject:

' Vermont' Yankee Nuclear Power Station Effluent and Waste Disposal Semiannual Report for First and Lecond Quarters, 1989

Dear Sir:

Enclosed herewith please find one copy of the subject Vermont Yankee Nuclear Power Corporation semiannual report. This report covers the period beginning January 1, 1989 and ending June 30, 1989, and is submitted in accordance with our Technical Specification 6.7.C.1.a.

We trust that'the enclosed information is satisfactory; however, should you have any questions, please contact us.

Very truly yours.

VERMONT YANKEE NUCLEAR POWER CORPORATION QWA - A i Leonard A. Tremblay, Jr.

Licensing Engineer LAT/b11/0593w Enclosure cc: USNRC Region I l

USNRC Resident Inspector, VYNPS i

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EFFLUENT AND WASTE DISPOSAL SEMIANNUAL REPORT FOR FIRST AND SECOND QUARTERS, 1989-Vermont Yankee Nuclear Power Station 4827R

s TABLE 1A Varmont Yankee Eff.luent and Waste Disp 2 sal Seminnnual Report First and Second'Ouariars. 1989

- Gasenus Effluents - Summation of All Releagea Unit Quarter Quarter Est. Total 1 2 Error. %

A. Fission and Activation Gases Of _fu.85E+02 1.58E+01 1.00E+02

1. Total release

2. Averace release rate for period uCi/sec 8.71E+01 2.01E+00

.I 3. Percent of Tech. Spec. limit (1)  % 1.25E-01 4.34E-03 B. Iodines C1 1.43E-03 2.53E-04 5.00E+01

1. Total Indine-131 I 2. Average release rate for period

3. Percent of Tech. Spec. limi1_(2)(4) %

uCi/sec 1.82E-04 9.79E-01 3.22E-05 2.24E-01 I. C. Particulate

1. Particulate with T-1/2 > 8 days Ci 8.44E-04 3.48E-04 5.00E+01 I 2. Average release rate for period

3. Percent of Tech uSpec. limit

4. Gross _ Alpha radioactivity uCi/see Ci 1.07E-04 (3) 5.92E-06 4.43E-05 (3) 4.15E-06 D. Tritium

1. Total release Ci 1.21E+01 1.50E+01 5.00E+01

2. Average release rate for period uCi/see 1.54E+00 1.91E+00

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

(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.

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

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

1 (4) The first and second quarter percent of Technical Specification limits are based on conservative plant quarterly dose determinations.

ND Not detected at the plant stack.

4827R

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TABLE IB i

= Vermont Yankee f E.ffluent and Waste Disposal Semiannual Report Eirst and Second Ouarters. 1989 Sargnus Effluents - Elevated Re1 ease (1)

I Nuclides_. Released Unit _

Continuous Mode Quarter 1

Quarter 2 1 Batch Mode Quarter Quarter 2

1. Fission Gases EIyp.tDn-85 Ci ND ND Kryplon-85m C1 1.00E+01 ND Krypton-87 Ci 3.00E+00 ND Kryp_tj;m-33_ Ci ND ND

_ Xenon-133 Ci 4.8.lE+02 6.80E+00 I Xenon-135

___ Xenon-135m Xenon-138 Ci Ci Ci 1.91E+02 ND ND 9.02E+00 ND ND Ci I Unid. e ntified Total for period Ci Ci 6.85Ef02 1.58E+01 I 2. Iodines

. . . Iostine-131 lodine-133 Ci Ci 1.43E-03 1 19E-04 2.53E-04 2.68E-04 Iodine-135 Ci ND ND I Talal for oerind Ci 1.60E-03 5.21E-04 3 _ Articulates Strontium-89 Ci 3.95E-06 2.64E-05 I- . Strontium-90 Ci ND ND Cesimp-134 Ci 1.64E-06 ND Cesium-137 Ci 2.33E-05 1.84E-05 Barium-Lanthanum-140 Ci NP ND

__flanganegg-54 Ci 1.98E-04 7.78E-05 Zine-65 C1 1.27E-04 4.21E-05 I Cobalt-53 Cphal.t-60 Ci Ci Ci 2.07E-05 3.75E-04 9.42E-05 ND 1.83E-04 ND Iron-59 (1) There were no batch mode reseous releases for this reporting period.

xD Net eetect.e at the ,1 ant st.ch.

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IAILLE IC Vermont Yankee Effluent and Waste Disposal Semiannual R*oort First and Second Ouart.grs. 1989-Gaseous Effluents - Ground Level Releaggs There were no routine measured ground level continuouc or batch mode gaseous releases during the first or second quarters of 1989.

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I- j TABLE ID Vermont Yankee Effluent and Waste Disposal Semiannual Report t

I.' Eirst and Second Ouarters. 1989 d

l-Gaseous Effluents - Nonroutine Releagg.a ID There were no nonroutine or accidental gaseoas releases during the first or second quarters of.1989.

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TABLE 2A Vermont Yankee

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Effluent and Waste Disposa)._ Semiannual Report Eirst and SeconiAnar.t.ers. 1982 Liquid Effluents - NQnroutine Releases I. .There were no liquid releases during the first or second quarters of 1989.

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TABLE'2B Vermont Yankee Effluent and Waste Disposal Semiannual Report First and Second Ouarters. 1989-Liquid Effluents - Nonroutine Releases

.I I There were no nonroutine or accidente.1 releases during the first or second quarters of 1989.

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IABLE 3 I, Eerraont Yankee Effinent and Wastg liaposal Semiannual _RepsIt First and Seconi Ouarters. 1989 Solid Waste and Irradiated Fuel Shioments A. Solid Waste Shipped Off-Site for Burial or Disposal (Not Irradiated Fuel)

I Unit 6-Month Est. Total Error. %

Period I 1. Type of Waste I a. Spent resins, filter sludges, evaporator bot 1Rmn a tt.

b. Dry compressible waste, contaminated m3 Ci m3 4.84E+00 2.15E+00 7.50E+01 e.2nipment. etc. Ci I c. Irradiated components, control rods, etc.

m3 C1

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

_ _a2_ lint-65  % L32E+01 I Cesi.um-137 Iron-55 Cobalt-60

% 2 2QE+01

% 1.97E+01

% 1.71E+01 Nickel-63  % 7.00E+00 I Cetium-134 Manganese-54 Hv.sirRgen-3

% 5.55E+00

% 3.11E+00

% 1.38E+00 I Iqdine-131 Pintonium-241 Carbon-14

% 5.33E-01

% 2.60E-01

% 1.49E-01 Transuranic  % 3.91E-03 Technetium-99  % 3.67E-03 I- Iodine-129  % 3.12E-03 Curium-242  % 9.30E-05 I 3. Solid Waste Disposition Number of Shipm_rnis Mode of Transportation Destination 1 Truck Barnwell, SC B. Irradiated Fuel Shipments (Disposition): None C. Supplemental information I 1) Class of solid waste containers shipped:

2) Types of containers used: 1 Type A

3) Solidification agent or absorbant: None lA 4827R I

I-APPENDIX A I

EFFLUENT AND WASTE DISPOSAL SEMIANNUAL REPORT

}. Supplemental Information First and Second Quarters, 1989 j Facility: Vermont Yankee Nuclear Power Station

" Licensee: Vannont_ Yankee Nuclear Power Corporation Technical Specification Limits - Dose and Dose Rate I 1A.

a.

Inchnical Specification and Catecory Hohle Gases Limit

'I 3.8.E.1 Total body dose rate 500 mrem /yr 3.8.E.1 Skin dose rate 3000 mrem /yr I. 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 I b. lo41ne-131. Iodine-133. Tritium and Radionuclist.ca in_ Ear _t.irulmi.e_.Entm With Half-Lives greater Than 8 Days I 3.8.E.1 Organ dose rate Organ dose 1500 mrem /yr 7.5 mrem in a quarter 3.8.G.1 3.8.G.1 Organ dose 15 mrem in a year

c. LLquists 3.8.E.1 Total body dose 1.5 mrem in a quarter 3.8.B.1 Total body dose 3 mrem in a year 3.8.B.1 Organ dose 5 mrem in a quarter I- 3.8.B.1 Organ dose 10 mrem in a year I

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Technical. Specification Limits - Concentration; I

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j Technical Specification _and Catenorv Limit 's

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'a.- l Noble Gases

.No MPC limits, i

g b. Iodine-131.' Iodine-133. Tritium and .

Radionuclides in Particulate Form With

.Eglf-LiYes_SIma.ter Than 8 Days: No MPC limits-  !

c.' Lionids

", '3.8.A.1' Total fraction of MPC g excluding noble gases (10CFR20, Appendix B, Table-11. Column.2): 11.0 3.8.A.in Total noole gas concentration: 12E-04 uCi/cc

3. Average-Energ'y Provided below are the average energy (E) of the radionuclides mixture'in releases of: fission and activation gases,. if applicable.

a. . ' Average gamma energy: 1st Quarter 4.29E-01 MeV/ dis 2nd' Quarter .8.81E-01 MeV/ dis l; b. . Average beta energy: Not Applicable 4.' ' Measurements and Approximations of Total Radioactivity Provided below are the methods used to measure or approximate the total .

1 radioactivity in effluents and the methods used to determine radionuclides

-composition.

A-2 4827R

a.. Fission and Activation Gases Stack-continuous monitors normally show less than detectable releases of fission and activation gases. Therefore, to determine the level of releases, gas grab samples are drawn at the stack mcnthly or when the noble gas monitors show an increase. The grab samples are analyzed by gamma spectroscopy to determine the isotopic composition and resultant release rate. The error I involved in these steps may be approximately 1100 percent.

b. Iodines Continuous isokinetic samples are drawn from the plant stack through a particulate filter and charcoal cartridge. The filters and 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, I 132, 133, 134, and 135. The error involved in these steps may be approximately 50 percent.

c. Particulate I The particulate filters described in b. above are also counted for particulate radioactivity. The error involved in this sample is also approximately 250 percent.

I d. Liquid Effluents Radioactive liquid effluents released from the facility are continuously monitored. Measurements are also made on a representative sample of each batch of radioactive liquid effluents released. For each batch, station records are retained of the total activity (mci) released, concentration (uCi/ml) of gross I

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he m g

fyg-l4 . ,

.,.c d, Z radioactivity... volume (liters), and approximate total quantity of'

'" water (liters) used to dilute the liquid effluent prior to. release.

'to the Connecticut River..

1 V- Each batch'of radioactive liquid' effluent' released is analyzed for

' gross' gamma and gamma isotopic radioactivity. A monthly-.

. proportional composite sample : comprising an' aliquot of each batch

. released during a month..is~also analyzedifor tritium, SR-89, SRi 90, gross beta'and gross alpha radioactivity,.in' addition.tol gamma spectroscopy.

l There were no liquid releases during the reporting period.

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5. Batch Releases

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

i b '.' Gaseous l

There were no routine gaseous batch releases during the reporting' )

. period.

6. Abnormal Releases L

a. Liquid l- ,

There were no nonroutine liquid releases during the reporting  !

period.

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

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APPENDIX B 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 ;

I the events leading to this condition is required in the next Semiannual Effluent Release Report per Technical Specification , 6.7.C.1.  ;

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

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APPENDIX C RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION Requirement : Radioactive liquid effluent monitoring instrumentation channels are required to be operable in accordance with Technical Specification Teble 3.9.1. If an inoperable radioactive liquid effluent monitoring instrument is not returned to operable I status prior to a release pursuant to Note 4 of Table 3.9.1, an explanation in the next Semiannual Effluent Release Report of I the reason (s) for delay in correcting the inoperability are required per Technical Specification 6.7.C.1.

Respontte: The service water radiation monitor checks for potential contamination within the service water system which is a nonradioactive system. The service water radiation monitor has been considered inoperable during closed and hybrid cycle I operation, and when the plant is not operating due to lack of sufficient dilution flow which affects system sensitivity.

Procedural changes are being made revising the method of  :

assessing the sensitivity requirements for the monitor. These changes will be completed by September 1, 1989, and the monitor will be considered operable. Daily grab samples of the service water have been taken and analyzed for specific activity, as required by the Technical Specifications, to ensure no radioactive contamination has entered the system.

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1 APPENDIX D RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION I 'Baguiremen.t: Radioactive gaseous effluent monitoring instrumentation channels are required to be operable in accordance with Technical Specification Table 3.9.2. If inoperable gaseous effluent monitoring instrumentation is not returned to operable status I. within 30 days pursuant to Note 5 of Table 3.9.2, an explanation in the next Semiannual Effluent Release Report of the reason (s) for the delay in correcting the inoperability is required per Tect.nical Specification 6.7.C.1.

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

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7 I APPENDIX E RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM I Requirement: The Radiological Environmental Monitoring Program is conducted

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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,

_g W Technical Specification 6.7 C.1 requires the following to be included in the next Semiannual Effluent Release Report:

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

I Ruponse: The Coombs Farm (TM-15 and TC-15), a milk and silage sampling location in Vernon, Vermont, went out of business during January

.g 1989. The Coombs Farm was replaced with the Newton Farm (IM-13 3

and TC-13). The revised ODCM figure and table are attached in Apr>endix H.

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l APPENDIX F I 1AND USE CENSUS lI 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 6.7,0.1 requires the identification of the new location (s) in the next I Semiannual Effluent Release Report.

Responae: The 1989 3and use census was not performed during this reporting period. It will be completed during the next reporting period (second half of 1989).

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I I APPDiDIX G

'I PROCESS CONTROL PROGRAM

'I Requirf:meni: Technical Specification 6.12.A.1 requires that licensee 1

initiated changes to the Process Control Program (PCP) be submitted to the Commission in the Semiannual Radioact5ve Effluent Release Report for the period in which the ch.tnge(s) was made.

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

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I' APPENDIX H OFF-SITE DOSE CALCULATION MANUAL I Requirement: Technical Specification 6.13.A.1 requires that licensee l initiated changes to the Off-Site Dose Calculation Manual (ODCM) i be submitted to the Com:nission in the Semiannual Radioactive Effluent Release Report for the period in which the change (s)

I was made effective.

Responge: There were two licensee initiated changes to the ODCM during this reporting period.

I The Coombs milk and silage sampling location (TM-15 and TC-15) was replaced with a new location (TM-13 and TC-13) in January 1989. Full details may be found in Appendix E.

I Section 3.11 of the ODCM was expat.ded to include the dose impact at the site boundary from radioactive materials and low level waste stored in the north warehouse, and the dose due to interim storage of packaged Dry Active Waste (DAW) and spent ion exchange and filter media in modular concrete storage overpacks on the low level waste storage pad facility adjacent to the north warehouse.

This change to the ODCM adds to the caps.bility to determine of f-site doses from fixed sources of radioactivity not previously addressed in the manual, and therefore, does not reduce the accuracy or reliability of dose calculations or setpoint determinations that are being calculated.

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The expanded dose calculational techniques which are added to g-B the ODCM are necessary to address changes in plant operations which now include the storage of solid radioactive vaste on-site in areas that previously have not been used for such activities. These changes are justified since the plant is not presently permitted to ship solid waste to existing licensed burial grounds for disposal. The design changes to the plant i facilities and implementation procedures to control the storage of solid waste on-site have been evaluated in accordance with 10CFR50.59.

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The revised and new ODCM pages for the above ch:nges are attached.

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VERMONT YANKEE NUCIIAR POL'ER STATION OFT-SITE DOSE CALCULATION MANUAL REV. # 6

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I I Reviewed F A c a s i see a / z/ I>lg 8

I vtRNON DAM t n gg tw es

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VE.D.NOX , V. T. K *n mu r, AP M - H I

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M-12 "J9 13 I E!L0r(TER$

LlLY Nur rigure 4-2 Environemntal Samnline Locations hithin Skm of Plant l 4-5 Revision 6 n..

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.g VERMONT YANKEE NUCLEAR POWER STATION OFF-SITE DOSE CALCULATION MANUAL REV. #7 I

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I Reviewed 20u,e. 793I Pl' ant ~ Opera (Lions Review Committee

/ $21/19 '

Approved Pl#nt Manager x

/3/29/89_ ~

Approved /N[0 M Vice PresidentVand Manager

/ 4,39 I of Operations I

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t.1ST OF AFFECTED PAGES Changes, deletions, or additions in the most recent revisions are indicated by a e

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

PAGE REVISION DATE iv-v 0 03/01/84 I vi-vii 1x 1.1 7

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TABLE OF CONTENTS W

Pace LIST OF EFFECTIVE PAGES iii' DISCLAIMER OF RESPONSIBILITY iv ABSTRACT v LIST OF TABLES viii LIST OF FIGURES ix

1.0 INTRODUCTION

1-1 1.1 Summary of Methods, Dose Factors, Limits, Constants, I- Variables and Definitions 1-2 2.0 METHOD TO. CALCULATE OFF-SITE LIQUID CONCENTRATIONS 2-1 2.1 Method to Determine F ENG 1 and C 1NG 2-1 I 2.2 Method to Determine Radionuclides Concer. ration for Each Liquid Effluent Pathway 2-2 2.2.1 Sample Tanks Pathway 2-2 I. 2.2.2 2.2.3 Service Water Pathway Circulating Water Pathway 2-3 2-3 3.0 0FF-SITE DOSE CALCULATION METHODS 3-1 3.1 Introductory Concepts 3-2 I  :, . 2 Method to Calculate Total Body Dose from Liquid 3.3 Releases Method to Calculate Maximum Organ Dose from Liquid 3-5 Releases I 3.4 3.5 Method to Calculate the Total Body Dose Rate from Noble Gases Method to Calculate the Skin Dose Rate from Noble Gases 3-11 3-14 3-19 3.6 Method to Calculate the Critical Organ Dose Rate from 8 Iollines, Tritium and Particulate with Tg Greater Than 8 Days 3-24 3.7 Method to Calculate the Gamma Air Dose from Noble Gases 3-28 3.8 Method to Calculate the Beta Air Dose from Noble Gases 3-31 1

3.9 Method to Calculate the Critical Organ Dose from Tritium, Iodines and Particulate 3-34 ol Revision 7 4/4/89 Date

1. -vi-

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TABLE OF CONTENTS (Continued) 3.10 Receptor Points and Annual Average Atmospheric Dispersion Factors for Important Exposure Pathways 3-40

'I 3.11 Method to Calculate Direct Dose from Plant Operation 3.12 Cumulative Doses 3-44 3-50 4.0 ENVIRONMENTAL MONITORING PROGRAM 4-1

[

5.0 SETPOINT DETERMINATIONS 5-1

! 5.1 Liquid Effluent Instrumentation Setpoints 5-2 l 5.2 Gaseous Effluent Instrumentation Setpoints 5-9 L

l 6.0 LIQUID AND GASEOUS EFFLUENT STREAMS, RADIATION MONITORS AND RADWASTE TREATMENT SYSTEMS 6-1 6.1 In-Plant Liquid Effluent Pathways 6-1 I 6.2 In-Plant Gaseous Effluent Pathways 6-3 R-1

. REFERENCES APPENDIX A: Method I Example Calculations A-1 I

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i LW Table 1.1-6 Summary of Methods to Calculate I 2 Dose to an Individual from Tritium, Iodine, and Particulate in Gas Releases and Direct Radiation I: Equation Reference Number Catecorv Ecuation Section.

'3-25 Dose to Critical 8 DFG 3.9.1 Organ from Iodines, Dco (mrem) =gI Q i ico 3

Tritium, and- i

.I' Particulate

)

Direct Dose 1

3-27 Turbine Buf'. ding D d

mrem = .29E-06 E 3.11.L North Warehouse l

3-28 Shielded End D = 0.25 x Rg 3.11.2 3

i

~3-29 Unshielded End Dg = 0.53 x RU 3.11.2 LLW Storace Pad 3-30 -Direct Line (Module .11.3 Short Side Out)

D dE

= 0.28 x dk x f d 3-31 Direct Line (Module D 3.11.3  !

Long Side Out) dS = 0.39 xdk x fd

-I I 3-32 Skyshine (Resin 3.11.3 Liners) D SKR=0.016xkSKR xf SK 3-33 Skyshine (DAW) D gg = 0.015 x kSKD

• fSK 3.11.3 3-34 Resin Liner Trans- 3.11.3 D

Tran=0.0025xkTm xT fer (Unshielded) Tm

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(

[

Revision 7 Date 4/4/89 1-10 I 4771R lE

Table 1.1-8 Summary of Variables I

Variable Definition Units I( = Concentration at point of discharge of dissolved and entrained noble gas "1" in liquid pathways from all station sources pCi/ml C = Total activity of all dissolved and entrained itCi nobic gases in liquid pathways from all mi station sources C = Concentration of radionuclides "i" at the point pfi di of liquid discharge ml Cg = Concentration of radionuclides "i" pfi cc C. = Concentration, exclusive of noble gases, of p_Ci P1 radionuclides "i" from tank "p" at point of ml di s cha'-:g e C. = Concentration of radionuclides "i" in mixture 11C1

• • ml at the monitor l

D = Beta dose to air mrad air Y

Dg = Gamma dose to air mrad D = Dose to the critical organ mrem D = Direct dose (Turbine Building) mrem d

I R d

= Dose rate at 3' from unobstructed side of storage module facing site boundary mrem hr D = Direct dose at site boundary per unobstructed mrem dE yr-module storage module (short end)

D = Direct dose at site boundary per unobstructed mrem I dS storage r.odule (long side) yr-module i Revision 7 Date 4/4/E L l-12 4771R u

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n _. . _ - . . _ _ _ .

[:-

Table l '.1 -8 (continued)

Summary of Variables

' Variable Definition bcits Y = Gamma dose to air, corrected for finite cloud mrad D gg D = Dose to the maximum organ axem O

D = Dose to skin from beta and gamma mrem 4

R = Dose rate at 1 meter from source in shielded Erfln end of North Warehouse hr Dg = Annual dose at site boundary f rom fixed sources mIP.ln in shielded end of North Warehouse yr R = Maximum dose rate at 3' over top of DAW EnI cin

-3 SKD' g in a storage module hr R = Maximum dose rate at 3' over top of each mrem SKR hr resin liner in a storage module Dggp = Skyshine dose at the site boundary from DAW mrem in storage modules (unobstructed top surfaces) yr-module Dgg = Skyshine dose at the site boundary from resin wrem I liners in storage modules (unobstructed top surfaces) yr-liner D = se to de total body mrem tb IR Tran

= Dose rate at contact from the unshielded top surface of resin liner E_

hr

= Dose at the site boundary from unshielded mrem ID movement of resin liner between transfer cask and storage module Rg = Dose ra h at 1 meter from source in us M ded wrem end of North Warehouse hr I Revision 7 Date 4/4/89 1-13 4771R 1

/

~ Table 1.1-8 (continued) b Summary of Variables variable Definition Units D = The annual dose at site boundary from fixed mIem U

sources in the unshielded end of North Warehouse yr i DF = Dilution factor ratio DFg = Minimum allowable dilution factor ratio 3

DF' = Composite skin dose factor mrem-m c pCi-yr 3

DFB g = Total body gamma dose factor for nuclide "i" y I = Composite total body dose factor 3

DFB DFL g = Site-specific, total body dose factor for a mrem liquid release of nuclide "i" Ci I DFL g = Site-specific, maximum organ dose factor for a liquid release of nuclide "i" mI.cm Ci DFG, = Site-specific, critical organ dose factor for a mnem

• C gaseous release of nuclide "i" Ci

= Site-specific, critical organ dose rate factor mrem-see DFGj for a gaseous release of nuclide "i" pCi-yr m rem-<n DFS g = Beta sH n dose factor for u clide "i" pCi-yr

-E mrem-see E DFj = Combined skin dose factor for nuclide "i" pCi-yr I DF g Y

= Gamma air dose factor for nuclide "i" mrad-m_3 pCi-yr 3

mrad-m DF g = Beta air dose factor for nuclide "1" pCi-yr I Revision 7 Date 4/4/89 1-14 4771R I - _ _

'. Table 1.1-8 lL (continued)

W ' Summary of Variables j Variable Definition Units Ik #

= Critical organ dose rate due to iodines and particulate 7#

l I

k = Skin dose rate due to noble gases skin r I. R g. = Total body dose rate due to noble gases D/Q = Deposition factor for dry deposition of _1_

elemental radiciodines and other particulate ,2 I E = Gross electric output over the period of interest MW h f = Fraction of a year that a storage module is fraction d

in use with an unobstructed side oriented toward west site boundary f

gg

= Fraction of a year that a storage module is fraction in use with an unobstructed top surface IF d

= Flow rate out of discharge canal gpm F = Flow rate past liquid radwaste monitor gpm F. = Flow rate past gaseous radwaste monitor nc_ i I ENG sec

= Total fraction of MPC in liquid pathways fraction IF 7 (excluding noble gases)

= Maximum pemiss%1e concentration for M MPC g radionuclides "i" (10CFR20, Appendix B, cc Table 2, Column 2)

= Release for radionuclides "i" from the curies I -Q g point of interest I Revision 7 Date 4/4/89 1-15 4771R I

Table 1.1-8 (continu;d)

Summary of Variables l

'~ Variable Definition Units IQ .

g

= Release rate for radionuclides "i" at the point of interest pcurica sec

.ST Qg = The noble gas radionuclides "i" release yCi rate at the plant stack see I . AE 91

= The noble gas radionuclides "i" release 8"

rate at the steam jet air ejector

.A0G 1 Qg = The noble gas radionuclides "i" release rate uCi at the exhaust of the augmented Off-Cas System sec l

.STP

= The iodine, tritium, and particulate uCi I Q*.

radionuclides "i" release rate from the plant stack see ST = The release of noble gas radionuclides curies

-I.'O i "i" from the plant stack uries

= The release of iodine, tritium, and particulate radionuclides "i" from the plant stack L = Liquid monitor response for the limiting cps R

8Pt concentration at the point of discharge R

sMn = Response of the noble gas monitor at the epm spt limiting skin dose rate tb = Response of the noble gas monitor to epm R

8Pt limiting total body dose rate S = Shielding factor Ratio F

Revision Date 1-16 I 4771R

p Igble-1.1-8

-(continued)  ;

Summary of Variables i

Variable Definition' __

Units S = Detector counting efficiency from the most com mR/hr E' #

recent gas monitor calibration pCi/cc pCi/cc  !

S' = Detector counting efficiency for noble epm 'l

1. mRZhr._

8 gas "i" pCi/cc pCi/cc S = Detector counting efficiency from the most eps y

recent liquid monitor calibration pC1/ml j l

S yg = Detector counting efficiency for ' cps  ;

radionuclides "i" pCi/mi j I  !

T Tun

= Time that an unshielded resin liner is hours exposed in the storage pad area a

1 X/Q = Annual or long-term average undepleted atmospheric dispersion factor m I- [X/Q] = Effective annual or long-term average f gamma atmospheric dispersion factor m I

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3.11 Method to Calculate Dose From Plant Operation Tachnical Specification 3.8.M.1 restricts the dose to the whole body or cny organ to any member of the public from all station sour::es (including direct radiation from fixed sources on-site) to 25 mrem in a calendar year (except the thyroid, which is limited to 75 mrem).

3.11.1 Turbine Buildine I The maximum contribution of direct dose to the whole body or to any

. organ due to N-16 decay from the turbine is:

D = . 6 E (3-27) d (mrem) (mrem) (MW* h)

MW, h where:

.I D = The dose contribution from N-16 decay at the site I d boundary off maximum impact (west site boundary) - mrem.

E = Gross electric output over the period of interest (MW , h).

I 1.29E-06 = (2.14E-06 mR/MWeh) (0.6 mrem /mR).

I 3.11.2 North Warehouse Radioactive materials and low level waste can be stored in the north warehouse. The maximum annual dose contributions to off-site receptors (west site boundary line) from sources in the shielded (east) end and the unshielded (west) end of the north warehouse are:

I m Revision 7 Date 4/4/_82 3-44 7272R I

.)

l Dg = 0.25 .x

~

Rg for the shielded end (3-28) l

{ J (mrem) (mrem /vr) mrem) t yr mrem /hr hr L

i l-t and l

Dg = 0.53 x R f r the unshielded end U

(3-29)

(torem) (prem/yI) (mIsm) yr mrem /hr hr I where:

i

'  ?

Dg = The annual dose contribution at the maximum site boundary l

location from fixed sources of radiation stored in the l I- shielded east end of the North Warehouse (E#**).

yr l

i >

Dg = The annual dose contribution at the maximum site boundary location from fixed sources of radiation stored in the unshielded west end of the North Warehouse (" **).

R g = Dose rate measured at 1 meter from the source in the shielded endofthenorthwarehouse.("(**).

~

k U-

= Dose rate measured at 1 meter from the source in the unshieldedendofthenorthwarehouse(*[*)* r 0.25 = Dose rate to dose conversion f actor which relates mrem /yr at the west site boundary per mrem /yr measured at 1 meter from the source in the shielded end of the warehouse assuming it is full to capacity for one year (""**

, ).

I Revision 7 Date 4/4/89 3-45 7272R

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L 0.53 = Doso rats to dosa conversion factor which relates mrem /yr at the west site boundary per mrem /yr measured at 1 meter from g.

5 the source in the unshielded end of the warehouse assuming it is full to capacity for one year ( ).

3.11.3 Low Level Waste Storace Pad Interim storage of packaged Dry Active Waste (DAW) and spent ion cxchange and filter media is permitted in modular concrete storage overpacks on the LLW storage pad facility adjacent to the north warehouse. The arrangement of the storage modules is such that DAW is placed in modules which chield higher activity ion exchange media frou the west site boundary. The dose at the maximum site boundary receptor from both direct radiation and I skyshine scatter can be calculated as follows: (

1 I

(a) Direct Dose (line of sicht)

D dE d * 'd (yr mteUL_)

module mInm/JI) mrem /hr (mrem) hr or D " * ' *

• dS d d mrem mrem /vr yr-module) (mrem /hry atem)

I- hr where:

D dE

= The annual direct dose contribution at the maximum site boundary from a single rectangular storage module which has an unobstructed short end surface (not shielded by other I modules) orientated toward the west site boundary mrem (yr-module),

Revision 7 Date 4/4/R9 3-46 7272R

D dS

= Th2 Ennual direct dose contribution at the maximum site i,. boundary from a single rectangular storage module which has

.an unobstructed long side surface (not shielded by other l:

modules) orientated toward the west site boundary (yrME--)

-module .

= Maximum d se rate measured at 3' f rom the side of the k

d storage module whose unobstructed face (i.e., a side or end surface which is not shielded by other waste modules) is toward the west site boundary.

f = The fraction of a year that a storage module is in use d

on the storage pad.

I 0.28 = Dose rate to dose conversion factor which relates mrem /yr at the west site boundary per mrem /hr measured at 3' from the narrow end of the rectangular storage module when that face is orientated toward the west boundary.

0.39 = Dose rate to dose conversion factor which relates prem/yr at the west site boundary per mrem /hr measured at 3' from the long side of the rectangular storage module when that face is orientated toward the west boundary.

(b) Scatter From Skyshine I. D = 0.016 x k gg xf gg (3-32)

SKR mrem mrmmlyr (yr-liner) (Ernm) mrem /hr) hr and I

Revision 7 Date 4/4/89 3-47 7272R I

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Dgg = 0;015 x R gg xf gg (3-33) mrem

! (yr-module) (mr3m/vr) nirem/hr Grem) hr i

~

where:

R = The annual skyshine scatter contribution to the dose at the SKR maximum site boundary from a single spent ion exchange media

~

liner in a storage module whose top surface is not obstructed due to stacking of modules (yr- i er}*

R gg = The annual skyshine scatter contribution to the dose at the maximum site boundary from a rectangular storage module containing DAW whose top surface is not obstructed due to stacking of modules ( 9 g ).

I '

= For Resins, the maximum dose rate measured at 3' over the R ggg top of each liner in a storage module (mrem /hr).

l R gg = For DAW, the maximum dose rate measured at 3' over the top surface of a storage module with DAW-(mrem /hr).

I f = The fraction of a year that a storage module is in use SK on the storage pad.

0.016 = Dose rate to dose conversion factor for the scatter dose from each resin liner source in storage which relates mrem /yr at the west site boundary per mrem / hour at 3' from the top of the module.

I Revision 7 Date 4/4/s9 3-48 I

0.015 = Do32 rate to doca convaraion fcctor for the scatter do n from DAW boxes in storage wl.ich relates mrem /yr at the west site boundary per mrem /hr at-3' from the top of the module.

I (c) Dose From Resin Liners Durine Transfer During the movement of resin liners from transfer casks to the storage modules, the liners will be unshielded in the storage pad area.for a short period of time. The maximum dose contribution at the site boundary during the unshielded movement of resin liners can be calculated from:

D trans

= 0.002E x k tran trans

(~

I.

(mrem) (*#**/h#)

R/hr h M where:

D = The dose contribution to maximum site boundary resulting from the unshielded movement of resin liners between a a transfer cask and a storage module (mrem).

-I R

tmns

= Dose rate measured at contact (2") from the unshielded top surface of the resin liner in R/hr.

I T = The time (in hours) that an unshielded resin liner 9

is exposed in the storage pad area.

I 0.0025 = The dose rate to dose conversion factor for an unshielded resin liner which relates mrem / hour at the west site boundary per R/hr at contact (2") from the unshielded surface of the liner.

I Revision 7 Date 4/4/89 3-49 7272R l

E_ - . _ . . . _ _ _ _ _ _ . . _ . _ _ _ _ _ . _ _ _ _ _ _ _ _ _ .

7_---__.

' Tha sito bound:ry dosa from w:sta materials pieced into stor g2 on the Low Level Waste Storage Pad Facility is determined by combining the dose contribution due to direct radiation (line of sight) from Part (a) above with the skyshine scatter dose and resin liner transfer dose from Parts (b) and (c)

- =

together.

3.11.4 Total Direct Dose summarv The dose contributions from the N-16 source in the Turbine Building, fixed sources in the north warehouse, and fixed sources on the Low Level Waste Storage Pad Facility, shall be combined to obtain the estimate of total off-site dose to any member of the public from all fixed sources of radiation located on-site. ,

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Revision 7 Date 4/4/89 3-50 7272R

I APPENDIX I I_

RADIOACTIVE LIQUID, CASECUS, AND SOLID WASTE TREATMENT SYSTEMS BaguiItm. cal: 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.

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

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