Text

Calculation PMP-6010-OSD-001, Rev. 19, Off-Site Dose Calculation Manual.
	ML061020294
Person / Time
Site:	Cook
Issue date:	12/31/2005
From:	; American Electric Power Co
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	AEP:NRC:6691-01, FOIA/PA-2010-0209
	Download: ML061020294 (182)
	v • d • e

OFF-SITE DOSE CALCULATION MANUAL The Off-Site Dose Calculation Manual, PMP-6010-OSD-001, was revised during this reporting period. Copies of Revisions 18 and 19 are included as part of the report. The reasons fo:: the changes and the PORC approval are documented on the Review and Approval trackdng fo:m These changes were determined to maintain the level of radioactive effluent control required by 10 CFR 20.1302, 40 CFR 190, 10 CFR 50.36a, and Appendix I to 10 CFR 50 and not adversely irrpact the accuracy or reliability of effluent, dose, or setpoint calculations.

A3.0-1

REVIEW AND APPROVAL TRACKING FOR1 ProcedureInformation:

Numbe: PMEP-6010-OSD-001 Rev. 18 Tile: OFF-STE DOSE CALCULATION MANUAL Alteration Caegory:

Cl MinorRcv/Correction wvReviewofAlteration Only _ Cancellation Cl Minor Rev/Correction w/Full Review ) Supersedod by (list sperseding procedures):

2l Change with Review of Alteration Onl El New Proccdure or Change with Full Review El Contractor Procedure T~mporawyProcedure/ Change:

121 N/A 0 TernporryProcedurm 0 TemporaryChange AR No.:

P prtionDate/Ending Ativity n (P Arsociated CO Impad Assessments:

Cfange Driver/CDITracking No(s). CDI-2005.00339 . O 10.

Rrviews:

DapaM~rxtrn eG nt ,

(Referto Figure 3, De oofRequiredReviews )

Ervirmnmnental 19 0 0 0 (]

Qerations (EE-:00 rocedures) 0 0 0 0 J Clernistry o O 0 13 M nMc) O0 0 0 0 . 13

___ 0 0 0 0' 0 lt)dated Revision Sumanary and Implementation Plan (if applicable) attached? 0 Yes hrplen=tationPlandeveloped? (Ref Step3.S.17) If yes,ARNo. 05101040 0Yea ONtA Ate there implernentalion actions to be complcted prior to the effective date? 0 Yes D No 10 CFR 50.59 Requiremnnts complete? TracldngNo.:_ _ _ Yes 0 N/A Qadified Technical Reviewer: C p 4 Date: O4:E,5 Ahiniiftrative Hold Statur 0 Released 0 Reissued 0 N/A CR No.:

W.iter . ) w "cs wF i /-i4?4 Date: ilrsa-O OpsManager Concurrence: N/A Date:

,yprowavrl ___

PORC ReviewRequired: Mtg lo.5 A

Apuroval Authority Review/Approval Date:

Effective Date: -

Followh-"Actions:

Prxedure Admistrative Review Conducted? (Data Sheet4 Complete) Ye, No Ccrnritmnmnt Database update requested in accordance with PMP-2350DCMS-0017 0 Yes 0 N/A NDM notified ornew records or changes to records that could affect record retention? .O Yes 0D N/A n SO This form is derived from 1lIc information in PMP-2010-PRC-002, Procedure Correction, Change, and Review, Rev. 15, Data Sheet 1, Review and Approval Tracking Form Page J or 7

(6-REVISION

SUMMARY

Number. PMP-6010 -OSD-001 Rievision: 18

Title:

OFF-SITE DOSE CALCULATION MANUAL Alteration Justification Changed level of use of this procedure from Per PA recommendation and more L.,

Information to Reference Use. appropriate with PMIP-2010-PRC-001, Figure 1, requirement. This is an administrative issue and has no impact on procedure content.

Added a Note to Section 1 as directed by. Provides explanation of [Current TS) and PM-2010-PRC-001, Att 2. Improved TS] revision methodology and applicability. This change is categorized as a correction per PMP-2010-PRC-001, Figure 1, Correction Criteria q andj.

L Added [Current TS] identifier to existing To support implementation of Improved Tech reference to Technical Specification 6.8.4 and Specs. These changes are categorized as updated for Improved Technical corrections per PMP-2010-PRC-001, Figure L Specifications [Improved TS) to reference 1, Correction Criteria n and j.

Technical Specification 5.5.3, Radioactive' Effluent Controls Ptogram to Section 1. - IL Moved information contained previously in. This describes a liquid release stream and step 3..1. j.8 to 3.1.2.c and added clarifying should be located under liquid effluent release information pertaining to sample flow. step, 3.1.2. The clarifying information is in response to CR 05039012 in an effort to avoid confusion. These changes~are categorized as corrections per PMP-2010-PRC-001, Figure 1, Correction Criteria p and q.

Added [Cuirent TS) identifier to existing references to Channel 'Functional'. Test and updated for Improved Tech Specs [rnproved To support implementation of Improved Tech Specs. These changes are categorized as corrections per PMP-2010-PRC-001, Figure L

TS] to reference Channel Operational in steps 1, Correction Criteria n andj.

3.2.1.f; 3.2.2.f, Att 3.3 Header and Notations and Att 3.5 Header and Notations.

Deleted reference to Eberline ESW radiation monitors WRA-350014500 and WRA- .

To support implementation of 12-CMM-50162. Westinghouse monitors (R-20 and R-L 3600/4600 in step 3.2.1 BASES, 3.3.1.b.2 as 28) continue to fulfill required applicability.

well as Att 3.2 Item 2.a, Action 3, and footnote +.

This change is categorized as corrections per PMP-2010-PRC-001, Figure 1, Correction L Criteria 1.

Added [Current TS] identifier to existing reference to Tech Spec 6.8.4.b and updated To support implementation of Improved Tech Specs. These changes are categorized as L

for Improved Tech Spec lImproved TS] to corrections per PMP-2010-PRC-001, Figure reference 5.5.1 .c to section 3.5 Bases. l, Correction Criterian and j.

This is a free-form as called out in PMP-201 O-PRC-002, Procedure Correction; Change, and Review, Rev. 15. Page -7of/

REVISION

SUMMARY

Nunber: PMP-6010-0SD-D01 Revision: I8

Title:

OFF-S1TE DOSE. CALCUIILATTIN MANUTAL Alteration Justification Added [Current TS] identifier to existing To support implementation of Improved Tech reference to Tech Spec 6.8.4.b and updated' Specs. These' changes are categorized as for Improved Tech Spec [Improved TS] to corrections per PMP-2010-PRC-001, Figure reference 5.5.1 .c to section 3.5 Bases. 1, Correction Criteria n and 3.

Changed name of Steam Generator Storage The steam generator lower assemblies from Facility to Radioactive Equipment Storage both units have been disposed of and are no Facility (Mausoleum) in Section-3.6 and Att longer stored in this facility, but it is being 3.19. used to store other radioactive equipment and will be in the future. This is a change.

Aided [Current TS] identifier to existing To support implementation of Improved Tech reference toTech Specs 6.8.1.e, 6.8.4.a, Specs. These changes are categorized as 6.8.4.b, 6.9.1.6, 6.9.1.7 and 6.14 and updated corredtons'perPMP-2010-PRC-O0l,Figure for Improved Tech Specs [Improved TS] to 1, Correction Criteria n and j.

reference 5.4.1.e, 5.5.1.c, 5.5.3, 5.6.2 and 5.6.3 to step 5.2.1 j -. ___

Added [and Gase6us]to 5.21.m' This document covers both liquid andgas:ous

. . . effluent nmonitors. Figure 1, Correction Added new r'efereiice 5.2.2.h for Unit Vent'

. Criteriain.; .

Position paper supplied by RETS-REMP Compensatory Sampling. workshop steering committee providing clarification on Unit Vent compensatory sampling descimbedinNURBG-1301. This is a change.

Added clarifying verbiage to ** note This is in response to CR 05039012 in an pertaining to steam generator sampling effort to avoid confusion. These changes are system in Att 3.2. categorized as corrections per PMP-2010-PRC-00l', Figure 1, Correction Criteria q.

Added [Current TS] identifier to existing To support implementation of Improved Tech re:cerence to .Tech Spec 1.6 and updated for Specs. These changes are categorized as Inmproved Tech Spec [Improved TS] to corrections per PMP-2010-PRC-001, Figtre re:erence section 1.1, under the term Operable 1, Correction Criteria n.

- Operability to Aft 3.2, footnote I.

This is a free-form as called out in PMP-201 O-PRC-002, Procedure Correction,.

Mjange, and Review, Rev. 15. Page 3 of

REVISION

SUMMARY

Numbert: PMP-6010-OSD-DDI Revision: 18 L

Title:

OFF-SITE DOSE CQiALCULATION IANUA L Alteration Justification Added [Current TSI identifier to existing Current Tech Spec 3/4.6.1 reference to L reference to 'Containment Integrity'. and CONTAINMENT INTERGRITY has been updated for Improved Tech Spec lImproved deleted since the definition in 1.8 is TSJ. to refer to 'Containment Operability' incorporated into Improved Tech Spec 3.6.1, under Att 3.4 Table Notation 2. 3.6.2 and 3.6.3 and is no longer maintained as a separate definition. To support the implementation of Improved Tech Specs. I.-

These changes are categorized as corrections per PM-2010-PRC-001, Figure 1, Correction Criteria n.

Added [Current TSI identifier to existing To support the implementation of Imrproved Tech Spec reference to 3.3.6-1 and updated for Improved Tech Spec [hnproved TS] to Tech Specs. These changes are categorized as corrections per PM-2010-PRC-O001, 3.3.6, Containment Purge Supply and Exhaust Figure 1, Correction Criteria n.

L System Isolation Instrumentation and 3.4.15, RCS Leakage Detection Instrumentation. L-.

under Att 3.4 Table Notation 3.

Added [Current TSJ identifier to existing To support the' iplementation of Improved Tech Spec reference to 3.4.6.1 and updated Tech Specs. These changes are categorized for Improved Tech Spec [Iaproved TS) to 'as corrections per PMP-2010-PRC-001, 3.4.15, RCS Leakage Detection Figure 1, Correction Criteria n.

Instrumentation under Att 3.4 Table Notation 10.

Added clarifying verbiage to notations c and d of Att 3.6 pertaining to steam generator This is in response to CR 05039012 in an effort to avoid confusion. Response to L

sampling system and pathway classification. Chemistry and Technical reviewer comments Also added clarifying information pertaining to the fact that this table reflects the minirnmrm for further clarification. These changes are categorized as corrections per PMP-201 0-L programmatic requirements, use of composite PRC-001, Figure 1, Correction Criteria q.

samples, and primary to secondary leak evidence.

L Added compensatory sample requirement for Sample requirement fromNUREG-1301 and unit vent gas sample (item J) and provided benchmarking against industry peers. Early details pertaining to early exit of iodine and exit criteria supplied through benchmarking L particulate compensatory sampling to Att 3.7. with peers also. This is a change. Response Also added clarifying information pertaining to the fact that this table reflects the minimum to Chemistry and Technical reviewer comments for further clarification and is L

programmatic requirements.' categorized as a correction per PMP-2010-PRC-00 1, Figure 1," Correction Criteria q.

This is a free-fonn as called out m PMP-2010-PRC-002, Procedure Correction, ge, and Review, Rev. 15. Page q ofL a -As:-.

-:"A. '

L

REVISION

SUMMARY

Ninber. PMP-601O-OSD-001 Revision: 18

Title:

OFF.SITE DOSE CALCULATION MANUAL Alteration Justification Added clarifying verbiage to Att 3.8 This is in response to CR 05020053 in an pertaining to applicability of MRP and the effort to provide clarification. This change is Turbine Room Sump. categorized as a correction per PMP-2010-PRC-001, Figure 1, Correction Criteria q.

Revised the X]Q and D/Q meteorological This is in response to CR 05068011 which in Formation in Att 3.16 and references to Att documents a higher 'Worst Case'. X/Q value 3.16 to reflect the latest 10 year average of that is used during gaseous radiation monitor 1995 -2004. setpoint calculations'. This is a change.

Deleted SWLII fron Att 3.19 Surface Water This is in response to FME issues associated sample location and reformatted one table. with CR 05048009. Ibis is a change. The:

formatting issue. reformatting is categorized as a correction per PMP-2010-PRC-001, Figure 1, Correction Criteria j.

Ri:vised maps on both Att 3.22 and 3.23. This was done in an effort to improve Circulating water inlet sample removed as readability and implement the change above welU as two inilk farms: Marginal marks were associated with the circulating water sanile not used on these maps due to the absence of point. These changes are categorized as any additions. corrections per PMP-2010-PRC-001, Figture

.__ 1, Correction Criteria L L

REVISION

SUMMARY

Numbcr: PMP-601&-0SD-OO1 Revision: 18

Title:

OFF-SITE DOSE CALCULATION MANUAL IMPLEkMENTATION PLAN Summary of Change.-

See Revision Summary.

Reasonfor Change:

Implement Improved TechnicalSpecifications, incorporatelatest meteorologicaldata, add one additionalunit vent compensatory sample, delefe one REMP sample location, reformat, and provide clarifications-Implementation Schedule 5/20105 - approved PvP-6010-OSD-001 5/27/05 -effective date for PMP-6010-OSD-001, PMP-4030-BIS-00.1, 12-6010-RPP-630, 12-6010-RPP-634,12-6010-RPP-643,12-THIP-6020-CHM-322,12-TH[-229i-ADMu012, 1-OHP-4021-001-002, ] -OHP-4021 -001-003, 1-OHP-4021-001-006,2-OHP-4021-001-002,2-OBP-4021-001-003, and 2-OHP-4021-001-006.

TrainingNeeds L N/A Effective Date 5127/05 L

ExpirationDate N/A RequiredBasis Documents Update .

Contemporary Design Bases Related ProcessesandProcedures PMP-4030-EIS-001, 12-6010-RPP-630,12-6010-RPP-634,12-6010-RPP.-643,12-TBP-L 6020-CHM-322,12-THT-2291-ADM-012, 1-OHP4021-001-002, 1-OHP-4021-001-003, 1-QHP-4021-001-006,;2-OBP-4021-001-002, 2-OHP-4021-001-003, and 2-OHP-4021-001-006..

L TlransitionPlan N/A L

Related Equipment Modifications N/A CommunicationPlan Discussed with B Hershberger and R Rose. Crew Event Notice for Chemistry and Ops prior

- ~kfi~1(J F~RAW i .-

This is a free-form as called out in PMP-2010-PRC-002, Procedure Correction, Change, and Review, Rev. 15. Page 6 of2..

.. J.

BUEVISION

SUMMARY

R.

Number: PMP-6010-OSD-001 Revis' ion: 18

Title:

OFF-SITE DOSE CALCULATION MANUAL -

to 5/27/05.

Special Tools, Aids, Permits, Etc.

N/A Condition Reports Relatedto ProcedureChangeImplementation 03147028, 05101040, 05039012, 05048009, 05020053, 05109008 and 05068011.

. 0 w .

Tis is a ree-fonn as called olt in PMP-201 COPRC-002, Pxoedure Corection, aiange~and Review, Rev. 15. Page ?.,f

rswr l PMP-6010-OSD-O01 l Rev. 18 Page l of 85 OFF-SITE DOSE CALCULATION MANUAL £/3I1oS

_ Reference Effective Dat e1 5140S~

Doue Foster John Carlson Environmental Writer Owner Cognizant Organization TABLE OF CONTENTS 1 I'URPOSE AND SCOPE. ........ ,. 4 2 DEFINITIONS AND ABBREVIATIONS. ................................... .................................... . 5

.3 l)ETAILS ..... . ;.5 2.1 Calculation of OII-Site Doses ......................... . .............................. 5 3.1.1 Gaseous Effluent Releases ;,... 5 3.1.2 Liquid Effluent Releases .. 10 3.2 Limits of Operation and Surveillances of the Effluent Release Points .... 13 3.2.1 . Radioactive Liquid Effluent Monitoring Instrmentation .. 13 3.2.2' Radioactive Gaseous Effluent Monitoring Instrunentation .. 14 3.2.3 Liquid Effluents . ..................................... . 15

a. ConcentrationExcluding Releases via the Turbine Room Sump CMRS) Discharge .15

b. Concentration of Releases from the TRS Discharge .16

c. Dose.16

d. Liquid Radwaste Treatment System .17 3.2.4 Gaseous Effluents .................... 19 a Dose Rate .............. 19

b. Dose-Noble Gases ............... 20

c. Dose - Iodine-131, Iodine-133, Tritium, and Radioactive Material in Particulate Form .............................. , 20

d. Gaseous Radwaste Treatment .21 3.2.S Radioactive Effluents - Total Dose .23 3.3 Calculation of Alaumilzp Setpoits .24 3.3.1 Liquid Monitors .. 25 a Liquid Batch Monitor Setpoint Methodology . .25

b. Liquid Continuous Monitor Setpoint Methodology . .26 3.3.2 Gaseous Monitors ..... ... ........... 28

a. Plant Unit Vent .. 28

b. Waste Gas Storage Tanks . .31

c. Contaiment Purge and Exhaust System . .31

d. Steam Jet Air Ejector System (SJAE) ; 32

e. Gland Seal Condenser Exhaust . . .32

ffLEC l PMP-6010-OSD.001 l Rev. 18 Page2 of 85 OFF-SITE DOSE CALCULATION MANUAL 5 L-513d lc5 4 Reference I l -Effective Date: 5 tK.

Doug Foster Rfe John Carlson Environmental . - - j ,

Wnter . Owner Cognizant Organization 3.4 Radioactive Effluents Total Dose ....... .. .. 33 3.5 Radiological Environmental Monitoring Program (REMP). ................................ 33 3.5.1 Purpose of the REMP . .... . . ... . .... 33 3.5.2 Conduct of the REMP . ........... .. ..... ; 33 3.5.3 Ainual Land Use Census ........................... . ; . 36 3.5.4 Interlaborator-yComparison Program . . . .......... 36 3.6 Radioactive Equipment Storage Facility (Mausoleum) Groundwater Monitoring Progranm ............................... 37 3.6.1 Purpose of the Radioactive Equipment Storage Facility (Mausoleum)

Groundwater Radiological Monitoring Programm.................................. 37 3.6.2 Conduct of the Radioactive Equipment Storage Facility (Mausoleum)

Groundwater Radiological Monitoring Programa................................... 37 L 3-7 Metornm1nkica1 Mndel ---- ------ - - --- 37---

3.8 Repoiting Requirements ... . ............ 37 L 3.8.1 3.8.2 Annual Radiological Environmental Operating Report (AREOR).....).. 37.

Annual Radiological Effluent Release Report (ARERR) . ..... 38 L

3.9 10 CER 50.75 (g) Implementation 3.10 Reportingh~anagement Re new........................

40 40 L

4 FINAL CONDITIONS ......................................... to.. .. to...o40 L

5 REFERENCES ..... ....... ... .........................................

... 41 SUPPLEMENTS .1 Dose Factors for Various Pathways ...................................... Pages 43 - 46 .2 Radioactive Liquid Effluent Monitorig Instruments .................. Pages 47 - 48 .3 Radioactive Liquid Effluent Monitoring Instrumentation Surveillance Requirements .......... ............................ Pages 49 - 50 .4 Radioactive Gaseous Effluent Monitoring Instrumentation ......... Pages 51 - 53 .5 Radioactive Gaseous Effluent Monitonng Instrumentation Surveillance Requiremnents ..... . . ............................ Pages 54 - 55

_" PMP 6010-OSD-001 Rev. 18 Page 3 of 85 OFF-SITE DOSE CALCULATION MANUAL - .t Reference IEffective Date._: 4s a-3110C Doug Foster John Carlson Environmental Writer Owner Cognizant Organization Attachment 3.6 Radioactive Liquid Waste Sampling and Analysis Program ........ Pages 56 - 57 Attachment 3.7 Radioactive Gaseous Waste Sampling and Analysis Program ..... Pages 58 - 59 Attachment 3.8 Multiple Release Point Factors for Release Points..; .............................. Page 60 .9 Liquid Effluent Release Systems ........... ........................ Page 61 Attachment 3.10 Plant Liquid Effluent Parameters ..................................... Page 62 .11 Volumetric Detection Efficiencies for Principle Ganuna Emitting Radionuclides for Eberline Liquid Monitors ....... Page 63 .12 Counting Efficiency Curves for R-19, and R-24 ... .... Pages 64 - 65 .13 Counting Efficiency Curve for R-20, and R-28..... Page 66

................................. .14 Gaseous Effluent Release Systems .Page 67 .15 Plant Gaseous Effluent Parameters .Page 68 .16 10 Year Average of 1995-2004 Data .Pages 69 --70 .17 Ainual Evaluation of - For All Sectors .Page 71 .18 Dose Factors ......................... ...: Pages 72 - 73 .19 Radiological Environmental Mcnitoring Program Sample Stations, Sample Types, Sample Frequencies ................................ Pages 74 - 77 .20 Maximum Values for Lower Limits of DetectionsAB- REMY ..... Pages 78- 79 .21. Reporting Levels for Radioactivity Concentrations in Environmental Samples ................ Page 80 .22 On-Site Monitoring Location - REMP ... Page 81 .23 Off-Site Monitoring Locations - RENM?................................................ Page; 82 .24 Safety Evaluation By The Office Of Nuclear Reactor Regulation..................................................I............................... Pages 83 - 85

Reference PMP-6010-OSD-001 Rev. 18 Page 4 of 85 OFF-SITE DOSE CALCULATION MANUAL NOTE: This procedure includes [Improved TS] information that is NOT applicable to Current Technical Speiication ([Current TS]) and [Current TS] informatio that will change when Improved Technical Specifications [Improved TS] is applicable. The [Improved TS] information shall NOT be used prior to the

[Improved TS] effective date. On and after the [Improved TS] effective date, the [Improved TS] information shall be used in lieu of the corresponding

[Current TS] information.

1 PURPOSE AND SCOPE NOTE: This is an Administrative procedure and only the appropriate sections need be performed per PMP-2010-PRC-003, step 3.2.7.

[Current TS] .

The Off-Site Dose Calculation Manual (ODCM) is the top tier document for the Radiological Environmental Monitoring Program (REMIP), the Radioactive Effluent Controls Program (RECP), contains criteria pertaining to the previous Radiological Effluent Technical Specifications (RETS) as L defined in NUREG-0472, and fully implements the requirements of Technical Specification 6.8.4. L

[Improved TS]

The Off-Site Dose Calculation Manual (ODCM) is the top tier document for the Radiological Environmental Monitoring Program (REMP), the L Radioactive Effluent Controls Program (RECP), contains criteria pertaining to the previous Radiological Effluent Technical Specifications (RETS) as defined in NUREG-0472, and fully implements the requirements of Technical Specification 5.5.3, Radioactive Effluent Controls Program.

L

The ODCM contains the methodology and parameters to be used in the calculation of off-site doses due to radioactive liquid and gaseous effluents and in theL calculation of liquid and gaseous monitoring instnrnentation alarm/trip setpoints.

The ODCM provides flow diagrams detailing the treatment path and the major X components of the radioactive liquid and gaseous waste management systems.

The ODCM presents maps of the sample locations and the meteorological model used to estimate the atmospheric dispersion and deposition parameters. L

The ODCM specifically addresses the design characteristics of the Donald C. L Cook Nuclear Plant based on the flow diagrams contained on the "OP L Drawings" and plant "System Description" documents.

L

Reference l PMP-6010-OSD-001 l .Rev. 18 l Page5offS -

OFF-SITE DOSE CALCULATION MANUAL 2 DEFVINITIONS AND ABBREVIATIONS Term: Meaning:

S or shiftly At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months D or.daily At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months W or weekly At least once per 7 days M or monthly At least once per 31 days Q or quarterly At least once per 92 days SA or semi-annually At least once per 184 days R At least once per 549 days.

StU Prior to each reactor startup P Completed prior to each release Sampling evolution Process of changing filters or obtaining grab samples 3 DETAILS 3.1 Calculation of Off-Site Doses 3.1.1 Gaseous Effluent Releases

a. The computer program MIDAS (Meteorological Information and Dose

. Assessment System) performs the calculation of doses from effluent releases. The site-specific parameters associated with MIDAS reside in the following subprograms:

M.DER MIDEX

MIDEL

MIDEG

MDEN'

b. The subprogram used to enter and edit gaseous release data is called MD 1EQ (EQ). The data entered in BQ can be used to calculate the accumulation of dose to individual land based receptors based on hourly meteorology and release data. The air dose from this data is calculated via the XDAIR subprogram in MIDAS. It computes air dose results for use in Reg. Guide 1.21 reports and 10 CFR 50 Appendix I calculations based on routine releases.

c. The formula used for the calculation of the air dose is generated from site specific parameters and Reg. Guide 1.109 (Eq 7):

DrD,,air= *Xf(MIOrN,)*Q,*3.17E-8J Where; Dy, Dp air = the gamma or beta air dose in mrad/yr to an individual receptor

Reference PMP-6010-OSD-001 Rev. 18 Page 6 of 85 OFF-SITE DOSE CALCULATION MANUAL X Q = the annual average or real time atnmspheric dispersion factor over land, sec/r 3 from Attachment 3.16, 10 Year Average of 1995-2004 Data m = the gamma air dose factor, mrad m 3 / yr sCi from Attachment 3.18, Dose Factors Ni = the beta air dose factor, mrad m 3 yr pCi from Attachment 3.18, Dose Factors

= the release rate of radionuclide, "i", in pCi/yr.

3.17E-8 = number of years in a second (years/second).

d. The value for the ground average I Q for each sector is calculated using equations shown below. Formula used for the calculation is generated from parameters contained in MIDAS Technical Manual, XDCALC (Eq 2).

~m 1 2.0 *3 L Where; L

minimumof 4 a2 or £,=i 3a,,L x = distance downwind of the source, meters. This information is found in parameter 5 of M[DEX-u,,, = wind speed for ground release, (meters/second).

h,,= vertical dispersion coefficient for ground release, (meters),

(Reg. Guide 14111 Fig.l)

Hc = building height (meters) fromparameter 28 of MIDERP -

(Containment Building = 49.4 meters) L Tf = terrain factor (= 1 for Cook Nuclear Plant) because we consider all our releases to be ground level (see parameter 5 in MIDEX).

2.03 = ; + 0.393 radians(22.50)

L

Reference PMP-6010-oSD-O01 l Rev. 18 Pagee7of .

L OFF-SITE DOSE CALCULATION MANUAL j

e. The dose due to gaseous releases, other than the air dose, is calculated by the MIDAS subprogram GASPRO. GASPRO computes the accumulation of dose to individual receptors based on hourly meteorology and release data. Calculations consider the effect of each important radionuclide for each pathway, organ, age group, distance and direction.

f. Calculations are based on the environmental pathways-to-man models in Reg. Guide 1.109. The program considers 7 pathways, 8 organs,. and 4 age groups in 16 direction sectors. The distances used are taken from the MLDEG file.

g. The formulas used for the following calculations are generated from site specific parameters and Reg. Guide 1.109:

1. Total Body Plume Pathway (Eq 10)

Dose (mrem/year) 3.17E -8*( Q. *X/Q* S* DFB)

Where; S= shielding factor that accounts for the dose reduction due lo shielding provided by residential structures during occupancy (maximum exposed individual = 0.7 per Table B-15 of Reg. Guide 1.109)

DFBi = the whole body dose factor from Table B-1 of Reg. Guide 1.109,mrem-m 3 perpCi-yr. SeeAttachment3.18, Dose Factors;

= the release rate of radionuclide 'T", in jiCi/yr

2. Skin Plwne Pathway (Eq 11)

Dose(mrem/yrJ=3.17E-8*S* I 7[(Q,

l

DF) + (Q,*DFSJ Where; 1.11 = conversion factor, tissue to air, mrem/mrad DF it = the gamma air dose factor for a uniform semi-infinite cloud of radionuclide "i", in mrad m3 ICi yr from Table B-1, Reg. Guide 1.109. See Attachment 3.18, Dose Factors.

Reference PMP-6010.05D-O01 Rev.18 Page 8 of 85 OFF-SITE DOSE CALCULATION MANUAL DFSi = the beta skin dose factor for a semi-infinite cloud of _

radionuclide Yi", in mremu 3 l/pCi yr from Table B-i, Reg. Guide 1.109. See Attachment 3.18, Dose Factors.

3. Radionuclide and Radioactive Particulate Doses (Eq 13 & 14) !

The dose, Dp in nremlyr, to an individual from radionuclides, other than noble gases, with half-lives greater than eight days in gaseous effluents released to unrestricted areas will be determined as follows:

D tp(mremyear)=3.17E- 8* 2(jR, *W

Qke)

Where; P,= the most restrictive dose factor for each identified L radionuclide 'i", in mi -rmem sec / yr yCi (for food and ground pathways) or mrem ni3 / yr pCi (for

inhalation pathway), for the appropriate pathway For sectors with existing pathways within five miles of the site, use the values of R, for these real pathways, otherwise use pathways distance of five miles. See Attachment 3.1, Dose Factors for Various Pathways, for the maximum Ri values for the most controlling age group for selected radionuclides. RP.

values were generated by computer code PARTS, see.

NUREG-0133, Appendix D. L W = the annual average or real timhe atnmspleric dispersion parameters for estimating doses to an individual at the worst case location, and where W is furfter defined as:

L

W/Qfor

-OR-the inhalation pathway,in sem 3 L Wf= D/IQ for thefoodandgroundpathwaysmin 1/ F. 2 L Qi- the release rate of those radioiodines, radioactive materials in particulate form and radionuclides other than noble gases withhalf-lives greaterthan eight days, in yCifyr

h. This calculation is made for each pathway. The maximum computed dose L at any receptor for each pathway is selected. These are summed together to get the dose to compare to the limits. Only the maximum of the cow milk or goat milk pathway (not both) is included in the totaL L L.

Reference l PMP-6010-SD-01 Rev. 18 -Page 9 of 5 OFF-SITE DOSE CALCULATION MANUAL j

i. In addition to the above routines, the QUICKG routine of the MIDAS system is used to provide data used in the monthly reports due to its ability to use annual average meteorological data rather than real time data, thus shortening the run time involved.

j. Steam Generator Blowdown System (Start Up Flash Tank Vent)

1. The amount of radioiodine and other radionuclides that are released via the start up flash tank and its vent are calculated through actual sample results while the start up flash tank is in service.

2. The following calculation is performed to determine the amount of curies released through this pathway. (Plant established formula.)

Curies= =Li *GPMV *time on flashtank (min)*3.785E- 3 mnl Where, 3.785B-3 = conversion factor, ml Ci/pCi gal

3. The flow rate is determined fiom the blowdown valve position and fie time on the start up tank Chemistry Department performs the sampling and analysis of the samles;

4. This data is provided to the MIDAS computer and dose calculations (liquid and gas) are performed to ensure compliance with Subsection 3.2, Limits of Operation and Surveillances of the Effluent Release Points, dose limits. MIDAS uses the formulas given in step 3.1.2, Liquid Effluent Releases, to calculate doses to members of the public.

NOTE: This section provides the minimum requirements to be followed at Donald C..

CookNuclear Plant. This would be used if actual sample data was not available each time the start up flash tank was in service.

5. The radioiodine release rate must be determined in accordance with the following equation every 31 day period whenever the specific activity of the secondary coolant systemis greater than 0.01 GCifg dose equivalent 1-131.

6. IF the specific activity of the secondary coolant system is less than 0.01 pCi/g dose equivalent I-131, THEN the release rate must be determined once every six months. Use the following plant established equation:

Q,=Ci* IPF* R5,b Where;

.Qy.= the release rate of 1-131 from the steam generator flash tank vent, in pCi/sec Ci = the concentration (pCi/cc) of 1-131 in the secondary coolant averaged over a period not exceeding seven days

I-.

I Reference I PMP-6010-OSD-001 OFF-SITE DOSE CALCULATION MANUAL l Rev. 18 l Page 10 of 85 I

IPF = the iodine partition factor for the Start Up Flash Tank, 0.05, in accordance with NUREG-0017 Rsgb = the steam generator blowdown rate to the start up flash ..

tank, i cc/sec

7. Use the calculated release rate in monthly dose projections until the next determination to ensure compliance with Subsection 3.2, Limits of Operation and Surveillances of the Effluent Release Points, dose limits. Report the release rate calculations in the Annual Radioactive Effluent Release Report.

3.1.2 Liquid Effluent Releases

a. The calculation of doses from liquid effluent releases is also performed by.

the MIDAS program. The subprogram used to enter and edit liquid release data is called MD1EB (EB).

b. To calculate the individual dose (mrem), the program DS L (ID) is used.

Lo It computes the individual dose forup to 5 receptors for 14 liquid pathways due to release of radioactive liquid effluents. The pathways can be selected using the MJDEL program and changing the values, in LI parameter 1. D.C. Cook Nuclear Plant uses 3 pathways: potable water, shoreline, and aquatic foods (fresh water sport fishing). L L

c. Steam Generators are sparged, sampled, and drained as batches usually early in outages to facilitate cooldown for entry into the steam generator.

This is typically'repeated prior to startup to improve steam generator chemistry for the startup; 'The sample stream, if being routed to the L

operating unit blowdown, is classified as a continuous release for quantification purposes to maintain uniformity with this defined pathway.

L.

d. The equations used are generated from site specific data and Reg. Guide 1.109. They are as follows:

1. PotableWater (Eq l) L R.~,j10*. UdiDpe" M*F*2.23E 3 Where; Rap; = the total annual dose to organ 'j" to individuals of age groups "a" from all of the nuclides "i" in pathway "p", in mrem/year 1100 = conversion factor, yr ft3 pCi / Ci sec L Uap = a usage factor that specifies the exposure time or intake rate for an individual of age group "a" associated with pathway '4p". Given in #29-84 of parameter 4 in M]DEL and Reg. Guide 1.109 Table E-5. See Attachment 3.1, Dose Factors for Various Pathways.

I .1..

Reference I PMP-6010-0SD-001 l Rev. 18 l Page 11 of t5 OFF-SITE.DOSE CALCULATION MANUAL j M = the dilution factor at the point of exposure (or the point oF withdrawal of drinking water orpoint of harvest of aquatic food). Given in parameter 5of MIDEL as 2.6.

F = the circulation water system water flow rate, in gpm, is used for evaluating dose via these pathways as dilution flow 2.23E-3 = conversion factor, ft3 min / sec gal

= the release rate of nuclide "i" for the time period of the nui input via MIDEB, Curies/year Dipj = the dose factor, specific to a given age group "a",

radionuclide 'T', pathway 'p", and organ 'J", which can be used to calculate the radiation dose from an intake of a radionuclide, in mrem/pCi These values are taken from tables E-II through E-14 of Reg. Guide 1.109 and are located within the MIDAS code.

Xi = the radioactive decay constant for radionuclide "i", in hours t' tp = the average transit time required for nuclides to reach the point of exposure, 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . This allows for nuclide transport through the water purification plant and the water distribution system For internal dose, to is the total elapsed time between release of the nuclides and ingestion of food or water, in hours. Given as #25 of parameter 4 in MIDEL (tp =12 hours)

2. Aquatic Foods (Eq 2)

.F Mp*F2.23E-3 Where, BP= the equilibriumbioaccunmulation factor for nuclide'T' in pathway 'pV', expressed as pCi L/kg pCi The factors are located within the MIDAS code and are taken from Table A-i ofReg. Guide l.109. SeeAttachment 3.1, Dose Factors for Various Pathways.

t = the average transit time required for nuclides to reach the point of exposure, 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . This allows for decay durug transit through the food chain, as well as during food*

preparation. Given as #26 of parameter 4 in MIDEL. (tp

=24 hours)

M,= the dilution factor at the point of exposure, 1.0 for Aquatic Foods. Given in parameter 5 of MIDEL as 1.0.

Reference PMP-6010-OSD001 Rev.18. Page12ofS8 OFF-SITE DOSE CALCULATION MANUAL

3. Shoreline Deposits (Eq 3).

MP*F*2.23E! Q D a Afpl-

Where; W theshoreline width factor. Given as an input of 0.3 when running the program, based on Table A-2 in Reg. Guide 1.109.

T = the radioactive half-life of the nuclide, Hi", in days 3 = the dose factor for standing on contaminated ground, in SDO mremr l hr pCi The values are taken fromtable E-6 of Reg. Guide 1.109 and are located within the MIDAS code.

See Attachment 3.1, Dose Factors for Various Pathways.

tb = the period of time for which sediment or soil is exposed to.

the contaminated water, 1.31E+5 hours. Given in MflEL as item 6 of parameter 4.

tp= the average transit time required for nuclides to reach the L point of exposure, 0 hours0 days 0 hours 0 weeks 0 months Given as #28 of parameter 4 in 110,000 = conversion factor yr ft3 pCi / Ci sec m2 day, this accounts for proportionality constant in the sediment radioactivity L

model M = the dilution factor'at the point of exposure (or the point of L withdrawal of drinking water orpoint of harvest of aquatic food). Given in parameter 5'of MIDEL as 2.6.

L

e. The MIDAS program uses the following planit specific parameters, which are entered by the operator. ' ' L

1. Irrigation rate=0 2.

3.

4.

Fractionoftimeonpasture=0 Fraction of feed on pasture = 0' Shore widthfactor= 0.3 (from Reg. Guide 1.109, Table A-2)

L L

f. The results of DS ILI are printed in LDRPT (LP). These results are used in the monthly report of liquid releases;L L

g. In addition, the program DOSUM (D1M) is used to search the results files of DSIlU to find the maximun liquid pathway individual doses The highest exposures are then printed in a summary table. Each line is L

conmpared with the appropriate dose limit. The table provides a concise summary of off-site environmental dose calculations for inclusion in Annual Radioactive Effluent Release Reports, required by Reg. Guide 1.21.

Reference PMP-6010-OSD-001 Rev. 18 Page 13 of 851 OFF-SITE DOSE CALCULATION MANUAL NOT'1E: The performance of each surveillance requirement must be within the specified time interval with a maximm allowable extension not to exceed 25% of tbe

.specified surveillance interval.

3.2 Limits of Operation and Surveillances of the Effluent Release Points 3.2.1 Radioactive Liquid Effluent Monitoing Instrumentation

a. The radioactive liquid effluent monitoring instrumentation channels shown in Attachment 3.2, Radioactive Liquid Effluent Monitoring Instruments, are operable with their alarm/trip setpoints set to ensure that the limits of step. 3.2.3a, Concentration Excluding Releases via the Turbine Room Surn (IRS) Discharge, are not exceeded.

b. The applicability of each channel is shown in Attachment 3.2, Radioactive Liquid Effluent Monitoring Instruments.

c. With a radioactive liquid effluent monitoring instrumentation channel alarm/trip setpoint less conservative than a value which will ensure the limits of step 3.2.3a, Concentration Excluding Releases via the Turbine Room Sump CMRS) Discharge, are met without delay, suspend the release of radioactive liquid effluents monitored by the affected channel and reset or declare the monitor inoperable.

d. With one or more radioactive liquid effluent monitoring instrumentation channels inoperable, take the applicable action shown in Attachment 3.2, Radioactive Liquid Effluent Monitoring Instruments, with a maximum allowable extension not to exceed 25% of the surveillance interval, excluding the initial performance.

e. Determine the setpoints in accordance with the methodology described in step 3.3.1, Liquid Monitors. Record the setpoints.

f. [Current TS]

Demonstrate each radioactive liquid effluent monitoring instrumentation channel is operable by performing the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION and CHANNEL FUNCIlONAL TEST at the frequencies shown in Attachment 3.3, Radioactive Liquid Effluent Monitoring Instrumentation Surveillance Requirements.

[Improved TS]

Demonstrate each radioactive liquid effluent monitoring instrumentation channel is operable by performing the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION and CHANNEL OPERATIONAL TEST at the frequencies shown in Attachment 3.3, Radioactive Liquid Effluent Monitoring Instrumentation Surveillance Requirements.

Reference I PMP-6010-OSD-OO1 Rev. 18 F Page 14 of 85 OFF-SITE DOSE CALCULATION MANUAL L BASES -LIQUID The radioactive liquid effluent instrumentation is provided to monitor and control, as.

applicable, the releases of radioactive materials in liquid effluents during actual or potential releases. The alarm/trip setpoints for these instruments shall be calculated in accordance with NRC approved methods in the ODCM to ensure the alarm/trip will occur prior to exceeding the limits of 10 CFR Part 20. The OPERABILITY and use of this instrumentation is consistent with the requirements of General Design Criteria specified in Section 11.3 of the Final Safety Analysis Report for the Donald C. Cook Nuclear Plant.

Due to the location of the Westinghouse ESW monitors, outlet line of containment spray heat exchanger (typically out of service), weekly sampling is required of the ESW system for radioactivity. This is necessary to ensure monitoring of a CCW to ESW systemleakL (Ref 5.2. 1gg]

3.2.2 Radioactive Gaseous Effluent Monitoring Instrumentation L

a. The radioactive gaseous process and effluent mnnitoring instrumentation channels shown in Attachment 3.4, Radioactive Gaseous Effluent L Monitoring Insumentation, are operable with their alarm/trip setpoints set to ensure that the limits of step 3.2.4a, Dose Rate, are not exceeded L

b. The applicability of each channel is shown in Attachment 3.4, Radioactive c.

Gaseous Effluent Monitoring Instrumentation.

With a radioactive gaseous process or effluent monitoring instrumntation L

channel alarm/trip setpoint less conservative than a value which will ensure that the limits of step 3.2.4a, Dose Rate, are met, without delay, suspend the release of radioactive gaseous effluents monitored by the L

affected channel and reset or declare the channel inoperable.

d. With less than the minimum number of radioactive gaseous efflumient L monitoring instrumentation channels operable, take the action shown in Attachment 34, Radioactive Gaseous Effluent Monitoring Instrumentation, with a maximum allowable extension not to exceed 25% L of the surveillance interval, excluding the initial performance.

NOTE: This surveillance requirement does not apply to the waste gas holdup system hydrogen and oxygen monitors, as their setpoints are not addressed in this L document.

e. Determine the setpoints in accordance withthe methodology as descnrbed L in step 3.3.2, Gaseous Monitors Record the setpoints.

Reference PMP-6010-OSD-001 Rev.18 l Pa e15of 8 OFT-SITE DOSE CALCULATION MANUAL

f. [Current TS]

Demonstrate each radioactive gaseous process or effluent monitoring instrumentation channel is operable by performing the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION, and CHANNEL FUNCTIONAL TEST operations at the frequencies shown ia Attachinent 3.5, Radioactive Gaseous Effluent Monitoring Instrumentation Surveillance Requirements.

[Improved IN]

Demonstrate each radioactive gaseous process or effluent monitoring instrumentation channel is operable by performing the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION,. and CHANNEL OPERATIONAL TEST operations at the frequencies shown in Attachntni 3.5, Radioactive Gaseous Effluent Monitoring Instrumentation Surveillance Requirements.

BASES - GASEOUS The radioactive gaseous effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in gaseous effluents during actual or potential.

releases. The alarm/trip setpoints for these instruments shall be calculated in accordance with NRC approved methods in the ODCM to ensure the alarm/trip will occur prior to exceeding the limits of 10 CFR Part 20. The OPERABILITY and use of this instrumentation is consistent with the requirements of General Design Criteria specified in Section 11.3 of the Final Safety Analysis Report for the Donald C. Cook Nuclear Plant.

3.2.3 Liquid Effluents

a. Concentration Excluding Releases via the Turbine Room Sump (IRS)

Discharge

1. Limit the conicentration of radioactive material released via the Batch Release Tanks or Plant Continuous Releases (excluding only TRS.

discharge to the Absorption Pond) to unrestricted areas to the concentrations in 10 CFR 20, Appendix B; Table 2, Column 2, for radionuclides other than dissolved or entrained noble gases. For dissolved or entrained noble gases, limit the concentration to 2E-4 pCihnI total activity.

2. With the concentration of radioactive material released from the site via the Batch Release Tanks orPlant Continuous Releases (other thaa the TRS to the Absorption Pond) exceeding the above limits, without delay restore the concentration to within the above limits.

3. Sample and analyze radioactive liquid wastes according to the sampling and analysis program of Attachment 3.6, Radioactive Liquid.Waste Sampling and Analysis Program.

Reference PMP-6010-OSD-0O1 Rev. 18 Page 16 of 85 OFF-SITE DOSE CALCULATION MANUAL

4. Use the results of radioactive analysis in accordance with the methods of this document to assure that all concentrations at the point of release are maintained within limits.

b. Concentration of Releases from the TRS Discharge

1. Limit releases via the TRS discharge to the on-site Absorption Pond to the concentrations specified in 10 CFR 20, Appendix B, Table 2, Column 2. For dissolved or entrained noble gases, limit the concentration to 2B-4 iCi/ml total activity.

2. With releases firom the TRS exceeding the above limits, perform a dose projection due to liquid releases to UNRESTRICTED AREAS to determine if the linits of step 3.2.3c. 1 have been exceeded. If the dose limits have been exceeded, follow the directions in step 3.2.3c.2, as applicable.

3. Sample and analyze radioactive liquid wastes according to the program in Attachment 3.6, Radioactive Liquid Waste Sampling and Analysis Prbgram.

4. Use the results of radioactive analysis in accordance with the methods L of this document to assure that all concentrations at the point of release are maintained within the limits stated above. -

c. Dose

1. Limit the dose or dose commitment to an individual from radioactive material in liquid effluents released to unrestricted areas during any calendar quarter to

1.5 mremto the totalbody and to < 5 mremto any organ, and during any calendar year to < 3 mrem to the total body and to *10 mrem to any organ.

L

2. With the calculated release of radioactive materials in liquid effluents exceeding ten times any of the limits in Steps 32.3a, 3.2.3b, or 3.2.3c. I above, prepare and submit a Written Report, pursuant to 10 CFR 20.2203, within 30 days after leaning of the event This report must describe the extent of exposure of individuals to radiation and radioactive material, including, as appropriate:

a) Estimate of each individual's dose, b) Levels of radiation and concentration of radioactive material involved, c) Cause of elevated exposures, dose rates or concentrations,

-AND-d) Corrective steps taken or planned to ensure against recurrence, .. L including schedule for achieving conformance with applicable limits.

These reports must be formatted in accordance with PMP-7030-001- X 002, Licensee Event Reports, Special and Routine Reports, even though this is not an lEPR L

Reference PMIP-6010-OSD-0O1 -7 Rev. 18 Pa e17of OFF-SITE DOSE CALCULATION MANUAL j

3. Determine cumulative and projected dose contributions from liquid effluents in accordance with this document at least once per 31 days.

Dose may be projected based on estimates from previous monthly projections and current or future plant conditions.

d. Liquid Radwaste Treatment System

1. Use the liquid radwaste treatment system to reduce the radioactive materials in liquid wastes prior to their discharge when the projected doses due to the liquid effluent from the site when averaged over 31 days, would exceed 0.06 mrem to the total body or 0.2 urem to any organ.

2. Project doses due to liquid releases to UNRESTRICTED AREAS at

least once per 31 days, in accordance with this document.

e. During times of primary to secondary leakage, the use of the startup flash tank should be minimized to reduce the release of curies from the secondary system and to maintain the dose to the public ALARA.

Operation of the North Boric Acid Evaporator (NBAE) should be done in a manner so as to allow the recycle of the distillate water to the Primary Water Storage Tank for reuse. This will provide a large reduction in liquid curies of tritium released to the environment, as there is approximately 40 curies of tritium released with every monitor tank of NBAE distillate.

Drainage of high conductivity water (Component Cooling Water and ice mnelt water containing sodium tetraborate) shall be evaluated to decide whether it should be drained to waste (small volumes only), the Turbine:

Room Sump (low activity water only) or routed without demineralizaticon processing to a monitor tank for release. This is necessary in order to inimie the detrimental affect that high conductivity water has on the radioactive wastewater demineralization system. The standard concentration and volume equation can be utilized to determine the impact on each method and is given here. The units for concentration and volume need to be consistent across the equation:

* ~(CX)Vj)+(Ca)(Va)=Ct)(Vt)

Where; C1 = the initial concentration of the system being added to Vi = the initial volume of the system being added to Ca = the concentration of the water that is being added to the system V. = the volume of the water that is being added to the system Ct = the final concentration of the system after the addition V, = the final volume of the system after the addition' The intent is to keep the:

Reference 7 PMP-6010-OSD0-1 lT Rev. 18 l Page 18 of 85l L

OFF-SITE DOSE CALCULATION MANUAL

WDS below 500 stmhdslcc.

TRS below lE-5 ptCcc.

Monitor Tank release ALARA to members of the public.

Wastewater leakage into the liquid waste disposal system will be monitored routinely. In the event the leak rate is determined to be over two gallons per minute (the assumed plant design leakage based on the original 2 gpm waste evaporator), increased scrutiny will be placed on locating inleakage, timeliness ofjob order activities, and/or activities L

causing increased production of waste water. L BASES - CONCENTRATION This specification is provided to ensure the concentration of radioactive materials released in liquid waste effluents from the site to unrestricted areas will be less than the concentration levels specified in 10 CFR Part 20, Appendix B, Table 2. This limitation provides additional L

assurance that the levels of radioactive materials in bodies of water outside the site will not result in exposures greater than 1) the Section i.A design objectives of Appendix 1, 10 CFR Part 50, to an individual and 2) the limits of 10 CFR Part 20. The concentration limit for noble gasses is based upon the assumption that Xe-135 is the controlling radionuclide and its

Effluent Concentration Unit in air (submersion) was converted to an equivalent concentration in water using the methods described in the International Commission on Radiological Protection (ICRP) Publication 2.

L

Reference T PMIP-6010-OSD-O01 Rev. 18 Page 19 of 85 OFF-SITE DOSE CALCULATION MANUAL I

DO'E This specification is provided to implement the requirements of Sections II.A, Il.A, and IV.A of Appendix I, 10 CFR-Part 50. The dose limits implement the guides set forth in Section ll.A of Appendix I. The ACTION statements provide the required operating flexibility and at the samba time; implement the guides set forth in Section IV.A of Appendix I to assure the releases of radioactive material in liquid effluents will be kept "as low as is reasonably achievable".

Also, for fresh water sites with drinking water supplies which can be potentially affected by plant operations, there is reasonable assurance that the operation of the facility will not result in radionuclide concentrations in the finished drinidg water that are in excess of the requirements of 40 CFR 141. The dose calculations in the ODCM implement the requirements in Section IILA of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data such that the actiial exposure of an individual through appropriate pathways is unlikely to be substantially underestimated. The equations specified in the ODCM for calculating the doses due to the actual release rates of radioactive materials in liquid effluents, will be consistent with the methodology provided in Regulatory Guide l.109, 'Calculation of Annual Doses to Man fromRoutine Releases of Reactor Effluents for the Pupo6se of Evaluating Compliance with 10 CFR Part 50, Appendix r', Revision 1, October 1977, and Regulatory Guide 1.113, 'Estimating Aquatic Dispersion of Effluents from Accidental and Routine Reactor Releases for the Purpose of Implementing

'Appendix r', April 1977. NUREG-0133 provides methods for dose calculations consistent with Regulatory Guide 1.109 and 1.113.

This specification applies to the release of liquid effluents from each reactor at the site. The liquid effluents from the shared system are proportioned among the units sharing the system.

LIQUID WASTE TREATMENT The operability of the liquid radwaste treatment system ensures that this system will be available for use whenever liquid effluents require treatment prior to release to the environment. The.requirements that the appropriate portions of this system be used when specified provide assurance that the releases of radioactive materials in liquid effluents will b.

kept "as low as is reasonably achievable". This specification implements the requirements of 10 CFR Part 50.36a, General Design Criteria Section 11.1 of the Fmal Safety Analysis Report for the Donald C. Cook Nuclear Plant, and design objective Section ll.D of Appendix 1 to 10 CFR Part 50. The specified limits governing the use of appropriate portions of the liquid radwaste treatment system were specified as a suitable fraction of the dose design objectives set f.rth in Section Il.A of Appendix 1, 10 CER Part 50, for liquid effluents.

3.2.4 Gaseous Effluents

a. Dose Rate

1. Limit the dose rate due to radioactive materials released in gaseous effluents from the site to

500 mrem/yr to the total body and

- 3000 mrern/yr to the skin for noble gases. Limit the dose rate due to all radioiodines and for all radioactive materials in particulate forn and radionuclides (other than noble gases) with half-lives greater than eight days to

1500 inremlyr to any organ.

2. With the dose rate(s) exceeding the above limits, without delay decrease the release rate to within the above limit(s).

Reference l PMP-6010-OSD-OO01 Rev. 18 . l Page 20 of 85 I OFF-SITE DOSE CALCULATION MANUAL l L

3. Determine the dose rate due to noble gases in gaseous effluents to be within the above limits in accordance with the methods and procedures described in this document 4.. Determine the dose rate due to radioactive materials, other than noble gases, in gaseous effluents to be within the above limits in accordance I-with the methods and procedures of this document by obtaining representative samples and performing analyses in accordance with the sampling and analysis program in Attachment 3.7, Radioactive Gaseous Waste Sampling and Analysis Program.

b. Dose - Noble Gases L

1. Limit the air dose in unrestricted areas due to noble gases released mi gaseous effluents during any calendar quarter, to 5 5 rnrad for gamma radiation and S 10 mrad for beta radiation and during any calendar year, to I10mrad for gamma radiation and* 20rnrad for beta radiation. i

2. With the calculated air dose fromradioactive noble gases in gaseous effluents exceeding ten times any of the above limits, prepare and submit a Written Report, pursuant to 10 CFR 20.2203 and addressed in step 3.2.3c.2, within 30 days after learning of the event

3. Determmie cumulative and projected dose contiibutions for the total time period in accordance with this document at least once every 31 days. IL

c. Dose-lodine-131, lodine-133, Tritium, andRadioactiveMaterialin Particulate Form

1. Limit the dose to a MEMBER OF THE PUBLIC fromradioiodine, radioactive materials in particulate form, and radionuclides other than noble gases with half-lives greater than eight days in gaseous effluents released to unrestricted areas (site boundary) to the following:

a) Duringanycalendarquartertolessthanorequalto7.Sniremto L any organ b) During any calendar year to less than or equal to 15 mrem to any organ. L

2. With the calculated dose fiom the release of radioiodines, radioactive materials in particulate form, orradionuclides other thannoble gases in gaseous effluents exceeding ten times any of the above limits, L prepare and submit a Written Report, pursuant to 10 CFR 20.2203 and

addressed in step 3.2.3c.2, within 30 days after learning of the event.

3. Determine cumulative and projected dose contributions for the total timc period in accordance with this document at least once every 31 days. L.

Reference I PMP-6010-OSD-OO1 I Rev. 18 l Page 21 of OFF-SITE DOSE CALCULATION MANUAL d Gaseous Radwaste Treatment

1. Use the gaseous radwaste treatment system and the ventilation exhaust treatment system to reduce radioactive materials in gaseous wastes fqior to their discharge when projected gaseous effluent air doses due to gaseous effluent releases to unrestricted areas when averaged over 31 days, would exceed 0.2 mrad for gamma radiation and 0.4 mrad for beta radiation. Use the ventilation exhaust treatment system to reduce radioactive-materials in gaseous waste prior to their discharge when the projected doses due to gaseous effluent releases to unrestricted areas.

when averaged over 31 days would exceed 0.3 mrem to any organ.

2. Project doses due to gaseous releases to UNRESTRICED AREAS at least once per 31 days in accordance with this document.

BASES - GASEOUS EFLLUENTS This specification is provided to ensure that the dose rate any time at the S]TE BOUNDARY from gaseous effluents from all units on the site will be within the annual dose limits of 10 CFR Part .20 for unrestricted areas. The annual dose limits are the doses associated with the concentrations of 10 CFR Part 20, Appendix. B, Table 2. These limits provide reasonable assurance that radioactive material discharged in gaseous effluents will not result in the exposure of xi individual in an unrestricted area, to annual average concentrations exceeding the limits specified in Appendix B., Table 2 of 10 CFR Part 20. For individuals who may at times be witlin the site boundary, the occupancy of the individual will be sufficiently low to compensate for any increase in the atnospheric diffusion factor above that for the site boundary. The specified instantaneous release rate limits restrict, at all times, the corresponding gamma and beta dose rates above background to an individual at or beyond the site boundary to 5 500 mreml3r to.

the total body or to *3000 mrem/yr to the skin. These instantaneous release rate limits also restrict, at all times, the corresponding thyroid dose rate above background to a child via the inhalation pathway to

1500 mrem/yr.

This specification applies to the release of gaseous effluents from all reactors at the site. The gaseous effluents from the shared system are proportioned among the units sharing that system.

DOS'E, NOBLE GASES This specification is provided to implement the requirements of Sections ll.B, lII.A, and IV.)k of Appendix I, 10 CFR Part 50. The dose limits implement the guides set forth in Section ll.B of App-.ndix I.

Reference PMP-6010-OSD-001 . Rev. 18 Page 22 of 85 OFF-SITE DOSE CALCULATION MANUAL The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in section IV.A of Appendix I to assure that the releases of radioactiveimaterial in gaseous effluents will be kept "as low as is reasonably achievable". The Surveillance Requirements implement the requirements in Section mIA of Appendix I that conform with the guides of Appendix I to be shown by calculational procedures based on models and data such that the actual exposure of an individual through the appropriate pathways is unlikely to be substantially underestimated. The dose calculations established in the ODCM for calculating the doses due to the actual release rates of radioactive noble gases in gaseous effluents will be consistent with the methodology provided in Regulatory Guide 1.109, '<Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix r', Revision 1, October 1977 and Regulatory Guide

1. 111, 'Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors", Revision 1, July 1977.. The ODCM equations provided for determining the air doses at the site boundary will be based upon the historical average atmospherical conditions. NUREG-0133 provides methods for dose -.

calculations consistent with Regulatory Guides 1.109 and 1.111.

DOSE, RADIOIOADA IOACIVE MATERIAL IN PARTICULATE FORM, A RADIONUCUDTES OTHER MHAN NOBLE GASES L

This specification is provided to implement the requirements of Sections ll.C, HILA, and IV.A of Appendix J, 10 CER Part 50. The dose limits are the guides set forth in Section IU.C of Appendix L The ACTION statements provide the required operating flexibility and at the sa=' time L

implement the guides set forth in section IV.A of Appendix I to assure that the releases of radioactive material in gaseous effluents will be kept "as low as is reasonably achievable". The L

ODCM.calculational methods specified in the surveillance requirements implement the requirements in Section Ill.A of Appendix I that conform with the guides of Appendix I to be shown by calculational procedures based on models and data such that the actual exposure of an L

individual through the appropriate pathways is unlikely to be substantially underestimated. The ODCM calculational methods approved by the NRC for calculating the doses due to the actual release rates of the subject materials are required to be consistent with the methodology provided L in Regulatory Guide 1.109, 'Calculation of Annual Doses to Man fromnRoutine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix r',

Revision 1, October 1977 and Regulatory Guide 1.111, 'Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled L Reactors", Revision 1, July 1977. These equations also provide the methodology for determining the actual doses based upon the historical average atmospheric conditions. The release rate specifications for radioiodines, radioactive material in particulate form, and radionuclides, other L than noble gases, are dependent on the existing radionuclide pathways to man, in the unrestricted area. The pathways which are examined in the development of these calculations are: 1) individual inhalation of airborne radionuclides, 2) deposition of radionuclides onto green leafy L vegetation with subsequent consumption by man, 3) deposition onto grassy areas where milk animals and meat producing animals graze with consumption of the milk and meat by man, and

4) deposition on the ground with subsequent exposure of man..

L.

Z Reference l PMP-6010-OSD-001 l Rev. 18 l Page 23 of 35 OFF-SITE DOSE CALCULATION MANUAL j GASEOUS WASTE TREATMENT The operability of the gaseous radwaste treatment system and the ventilation exhaust treatment systems ensures that the systems will be available for use whenever gaseous effluents require treatient prior tA release to the environment. The requirement that the appropriate portions of these systems be used when specified provides reasonable assurance that the releases of radibactive materials in gaseous effluents will be kept "as low as is reasonably achievable". 'his specification implements the requirements of 10 CFR Part 50.36a, General Design Criterion Section 11.1 of the Final Safety Analysis Report for the Donald C. Cook Nuclear Plant, and design objective Section ll.D of Appendix I to 10 CFR Part 50. The specified limits governing the use of appropriate portions of the. systems were specified as a suitable fraction of the guides fordt in Sections II.B and I.C of Appendii I, 10 CFR Part 50, for gaseous effluents.

3.2.5 Radioactive Effluents - Total Dose a The dose or dose commitment to a real individual from all uranium fuel cycle sources is limited to'< 25 mrem to the total body or any organ (except the thyroid, which is limited to; 75 mrem) over a period of 12 consecutive months.

b. With the calculated doses from the release of radioactive materials in liquid or gaseous effluents exceeding one half the annual limits of steps 3.2.3c (Dose), 3.2.4b (Dose - Noble Gases), or 3.2.4c (Dose - Iodine- 131, Iodine-133, Tritium, and Radioactive Material in Particulate Form) during any calendar quarter, perform the following:

Investigate and identify the causes for such release rates;

Define and initiate a program for corrective action;

Report these actions to the NRC within 30 days fiom the end of the quarter during which the release occurred.

IF the estimated dose(s) exceeds the limits above, and ]F the release condition resulting in violation has not already been corrected prior to violation of 40 CFR 190, THEN include in the report a request for a variance in accordance with the provisions of 40 CFR 190 and including the specified information of paragraph 190. 11(b). Submittal of the report:

is considered a timely request, and a variance is granted until staff action on the request is complete. The variance only relates to the limits of 40 CFR 190, and does not apply in any way to the requirements for dose limitation of 10 CER 50, as addressed in other sections of this document.

c. Determine cumulative dose contributions from liquid and gaseous effluents in accordance with this document (including steps 3.2.3c [Dose:,

3.2.4b [Dose - Noble Gases], or 3.2.4c [Dose - Iodine-131, Iodine-133, Tritium, and Radioactive Material in Particulate Form]).

Reference I PMP-6010-OSD-001 Rev. 18 l Page 24 of 85 OFF-SITE DOSE CALCULATION MANUAL BASES - TOTAL DOSE This specification is provided to meet the dose limitations of 40 CFR 190. The specification requires the preparation and submittal of a.Special Report whenever the calculated doses from.

plant-radioactive effluents exceed twice the design objective doses of Appendix I. For sites containing up to 4 reactors, it is highly unlikely that the resultant dose to a member of the public will exceed the dose limits of 40 CFR 190 if the individual reactors remain within the reporting requirement level. The Special Report will describe a course of action, which should result in the limitations of dose to a member of the public for 12 consecutive months to within the 40 CFR 190 limits. For the purposes of the Special Report, it may be assumed that the dose commitment to any member of the public from other uranium fuel cycle sources is negligible with the exception that dose contributions from other nuclear fuel cycle facilities at the same site or within a radius of 5 miles must be considered. If the dose to any member of the public is estimated to exceed the requirements of 40 CFR 190, the Special Report with a request for a variance (provided the release conditions resulting in violation of 40 CFR 190 have not already been corrected, in accordance with the provision of 40 CFR 190.11), is considered to be a timely request and fulfills the requirements of 40 CFR. 190 until NRC staff action is completed. An individual is not considered a member of the public during any period in which he/she is engaged in carrying out any operation, which is part of the nuclear fuel cycle.

3.3 Calculation of AlanrmTrip Setpoints L

.The alarm and trip setpoints are to provide monitoring, indication, and control of liquid and gaseous effluents. The setpoints are used in conjunction with sampling programs to assure that the releases are kept within the limits of 10 CFR 20, Appendix B, Table 2.

Establish setpoints for liquid and gaseous monitors. Depending on the monitor function, it would be a continuous or batch nxmnitor. T11 different types of monitors are subject to different setpoint methodologies.

L One variable uised in setpoint calculations is the multiple release point (MRP) factor. The MRN is a factor used such that when all the releases are integrated, the applicable LIMIT value will not be exceeded. The MRP is deternined such that the sum of the MRP's for that effluent type (liquid or gaseous) is less than or equal to 1. The value of the MRP is arbitrary, and it should be assigned based on operational performance. The values of the MRP's for each liquid release point are given in Attachment 3.8, Multiple Release Point L Factors for Release Points.

The Site stance on instrument uncertainty is taken fiom HPPOS-223, Consideration of Measurement Uncertainty When Measuring Radiation Levels Approaching Regulatory L

Limits, which states the NRC position is the result of a valid measurement obtained by a method, which provides a reasonable demonstration of compliance. This value should be accepted and the uncertainty in that measured value need not be considered.

L.

Reference I PAIP-6010 OSD-001 l Rev. 18 l Page 25 of OFF-SITE DOSE CALCULATION MANUAL 3.3.1 Liquid Monitors Establish liquid monitor setpoints for each monitor of the liquid effluent release systems. A schematic of the liquid effluent release systems is shown as Attachment 3.9, Liquid Effluent Release Systems. A list of the Plant Liquid Effluent Parameters is in Attachment 3.10, Plant Liquid Effluent Parameters. The details of each system design and operation can be found in the system descriptions. The setpoints are intended to keep releases within the limits of 10 CFR 20, Appendix B, Table 2, Cohimn 2. Determine setpoints using either the batch or the continuous methodology.

a. Liquid BatchMonitorSetpoint Methodology 1; There is only one monitor used on the Waste Disposal System for liquid batch releases. This monitor is identified as RRS-1000. Steam Gentrator Blowdown radiation monitors also can be used to monitor batch releases while dram ng steam generators. The function of these monitbrs is to act as a check on the sampling program. Ie sampling program determines the nuclides and concentrations of those nuclides prior to release. Thte discharge and dilution flow rates are then adjusted to keep the release within the limits of 10 CFR 20. Based on the concentrations of nuclides in the release, the count rate on the monitor can be predicted. The high alarm setpoint can then be set above the predicted value up to the maximum setpoint of the system

2. The radioactive concentration of each batch of radioactive liquid waste to be discharged is determined prior to each release by sampling and analysis in accordance with Attachment 3.6, Radioactive Liquid Waste Sampling and Analysis Program.

3. The allowable release flow rates are determined in order to keep the.

release concentrations within the requirements of 10 CER 20, Appendix B, Table 2, Column 2. The equation to calculate the flow rate is from Addendum AAl of NUREG-133:

[ l ]

LiMrnJ MRP Where; C; - the concentration of nuclide "i" in gtCi/ml LIMIT; = the 10 CFR 20, Appendix B, Table 2, Column2 limit of nuclide "i" in AiCi/ipl f = the effluent flow rate in gpm (Attachment 3.10, Plant Liquid Effluent Parameters).

F the dilution water flow rate as estimated prior to release. The dilution flow rate is a multiple of 230,000 gpm depending on the niumber of circulation pumps in operation.

MRP = the multiple release point factor. A factor such that when all the release points are operating at one time the limits of 10 CFR 20 will not be exceeded.

Reference PMP-6010-OSD-001 Rev. 18 Page 26 of 85 OFF-SITE DOSE CALCULATION MANUAL

4. This equation must be true during the batch release. Before the release is started, substitute the maximum effluent flow rate and the minimum dilution flow rate for f and F, respectively. If the 'equation is true, the release can proceed with those flow rates as the limits of operation. If the equation is not true, the effluent flow rate can be reduced or the dilution flow rate can be increased to make the equation true. This equation may be rearranged to solve for the maximum effluent release flow rate (f).

5. The setpoint is used as a quality check on the sampling program. The setpoint is tised to stop the effluent flow when the monitor reading is greater than the predicted value from the sampling program The predicted value is generated by converting the effluent 'concentration for each gamma emitting radionuclide to counts per unit of time as per Attachment 3. 11, Volumetric Detection Efficiencies for Principle

-,Gamma Emitting Radionuclides for Eberline Liquid Monitors, or Attachment 3.12, Counting Efficiency Curves for R-19, and R-24.

The sum of all the counts per unit of time is the predicted count rate. L The predicted count rate can then be multiplied by a factor to determine the bigh alarm setpoint that will provide a high degree of conservatism and eliminate spurious alaOms. L

b. Liquid Continuous Monitor Setpoint Methodology

1. There are eight monitors used as potential continuous liquid release monitors. These monitors are used in the steam generator blowdown (SGBD), blowdown treatment (BDT), and essential service water (ESW) systems.

2. These Westinghouse monitors (R) are being replaced by Eberline

. monitors (DRS) and are identified as:. IL

R-l9 or DRS 3100/4100 for SGBD R-24 or DRS 3200/4200 for BDT l The function of these monitors is to assure that releases are kept within the concentration limits of 10 CFR 20, Appen4ix B, Table 2, Column 2, entering the unrestricted area following dilution. L

3. The monitors on steam generator blowdown and blowdown treatment systems have trip functions associated with their setpoints. Essential service water monitors are equipped with an alarm function only and L monitor effluent in the event the Containment Spray Heat Exchangers are used.

Reference PMP-6010-OSD-001 Rev. 18 Page 27 of I OFF-SITE DOSE CALCULATION MANUAL j

4. The equation used to determine the setpoint for continuous monitors is fromAddendumAAl of NUREG-0133:

C*Eff*MkP*F*SF Sp<sr Where;

-S =setpoint ofmonitor (cpm).

C = 5E-7 iCi/nml, maximum effluent control limit from IO CFR 20, Appendix B, Table 2, Column 2 of a known possible nuclide in effluent stream. (Ibe limiting nuclide shall be evaluated annually by reviewing current nuclides against historical ones in order to determine if one with a more restrictive effluent concentration limit than Si90 is.

found. The concentration limit shall be adjusted appropriately.)

.OR-if a mixture is to be specified, CI.

IMIT, Eff Efficiency, this information is located in Attachment 3.1 Il, Volumetric Detection Efficiencies for Principle Gamma Emitting Radionuclides for Eberline Liquid Monitors, through Attachment 3.13, Counting Efficiency Curve for R-20, and R-28, for the specific monitors. For Eberline monitors the efficiency is nuclide specific and the calculation changes slightly to:

a(C,*Effr) replacesC*Eff MRP = multiple release point factor A factor such that when all the release points are operating at one time the Emits of 10 CFR 20 will not be exceeded (Attachment 3.8, Multiple Release Point Factors'for Release Points). The MRP for ESW monitors is set to 1.

F dilution water (circ water) flow rate in gpm obtained from Attachment 3. 10, Plant Liquid Effluent Parameters. For routine operation, the setpoint should be calculated using the minimum dilution flow rate of 230,000 gpim.

SF = Safety Factor, 0.9.

f = applicable effluent release flow rate in gpmw For routine operation, the setpoint should be calculated using maximum effluent flow rate (Attachment 3.10, Plant Liquid Effluent Parameters).

Reference I PMP-6010-OSD-001 I ;Rev. 18 Page 28 of 85 I OFF-SITE DOSE CALCULATION MANUAL 3.3.2 Gaseous Monitors For the purpose of implementing Step 3.2.2, Radioactive Gaseous Effluent Monitoring Instrumentation, and Substep 3.2.4a, Dose Rate, the alarm setpoints for gaseous effluents released into unrestricted areas will be established using the following methodology. In addition, the above steps do not apply to instantaneous alarm and trip setpoints for integrating radiation monitors sampling radioiodines, radioactive materials in particulate form and.

radionuclides other than noble gases. A schematic of the gaseous effluent release systems is presented. in Attachmeent 3.14, Gaseous Effluent Release Systems. Attachment 3.15, Plant Gaseous Effluent Parameters, presents the effluent flow rate parameter(s).

Gaseous effluent monitor high alarm setpoints will routinely be established at a fraction of the maxinmu allowable setpoint (typically 10% of the setpoint) for ALARA purposes. Alert alarms will normally be set to provide adequate indications of small changes in radiological conditions.

a. Plant Unit Vent

1. The gaseous effluents discharged fiom the plant vent will be monitored by the plant vent radiation monitor low range noble gas

channel [Tag No. VRS-1505 (Unit 1), VRS-2505 (Unit 2)] to assure L that applicable alarms and trip actions (isolation of gaseous release) will occur prior to exceeding the limits in step 3.2.4, Gaseous Effluents. The alarm setpoint values will be established using the following unit analysis equation:

SF *MRP*DLj Fp*/Q*j>(Wi*DCFQ) L Where: L Sp = the maximum setpoint of the monitor in p.Ci/cc for release point p, based on the most limiting organ SF. = an administrative operation safety factor, less than 1.0 MNP = a weighted multiple release point factor (5 1.0), such that when all site gaseous releases are integrated, the applicable dose will not be exceeded based on the release rate of each effluent poit.

The MRP is an arbitrary value based on the ratio of the release rate or the vohlmetric flow rate of each effluent point to the total respective flow rate value of the plant and will be consistent with past operational experience. The MRP is computed as follows:

Compute the average release rate, Q, (or the volumetric flow rate, fp) from each release point p.

Compute ZQp (or Xfp) for all release points.

Ratio QOpJQp (or fp/.Ef) for each release point. This ratio is the MRP for that specific release point L

Repeat the above bullets for each of the site's eight gaseous release points.

Reference PMP-6010-OSD-001 Rev. 18 Page 29 of OFF-SITE DOSE CALCULATION MANUAL Fp = the maximum volumrtric flow rate of release point "p", at the time of the release, in cc/sec. The maximum Unit Vcnt flow rate, by design, is 186,600 cfm for Unit 1 and 143,400 cfm for Unit 2.

DIj = dose rate limit to organ 'I" in an unrestricted area (mremlyr).

Based on continuous releases, the dose rate limits, DIj, from step 3.2.4a, Dose Rate, are as follows:

Total Body S 500 nrernlyear

Skin S 3000 mremlyear

Any OrganS 1500 mremfyear X/ Q = The worst case annual average relative concentration in the applicable sector or area, in sec/rn 3 (see Attachment 3.16, 10 Year Average of 1995-2004 Data);

WI =weighted factor for the radionuclide:

zCk Where, C; = concentration of the most abundant radionuclide "i" Ck = total concentration of all identified radionuclides in that release pathway. For batch releases, this value may be set to 1 for conservatism DCFjj = dose conversion factor used to relate radiation dose to organ 'j", from exposure to radionuclide "i" in mremm /yr ptCL See following equations.

The dose conversion factor, DCFI,, is dependent upon the organ of concern.

For the whole body: DCFjj =K Where; K. = whole body dose factor due to gamma emissions for each identified noble gas radionuclide in nremm3 / yr g1Ci See Attachment 3.18, Dose Factors.

Reference PMP-6010-OSD-001 Rev. 18 Page 30 of 85 OFF-SITE DOSE CALCULATION MANUAL For the skin: DCFij I, + 1.lML Where;

= skin dose factor due to beta emissions for each identified noble gas radionuclide, in nremer n r3 yrjCi See Attacbnent 3.18, Dose Factors.

1.1 = the ratio of tissue to air absorption coefficient over the energy range of photons of interest.

This ratio converts absorbed dose (mrad) to dose equivalent (mrem).

M= the air dose factor due to gamma

.emissions for each identified noble gas radionuclide in mrad m3 / yr IiCi See Attachment 3.18, Dose Factors.

For the thyroid, via inhalation: DCFj =Pi Where; PI = the dose.parameter, for radionuclides

other than noble gas for the inhalation pathway in nemm. / yr pCi (and the L food and ground path, as appropriate).

See Attachment 3.18, Dose Factors.

2. The plant vent radiation monitor low range noble gas high alarm L channel setpoint, S., will be set such tbhat the dose rate in unrestricted areas to the whole body, skin and thyroid (or any other organ),

whichever is most limiting, will be less than or equal to 500 mrem/yr, 3000 mremlyr, and 1500 mrerr/yr respectively.

L

3. The thyroid dose is limited to the inhalation pathway only.

4. The plant vent radiation monitor low range noble gas setpoint, Sp, will be recomputed whenever gaseous releases like Containment Purge, Gas Decay Tanks and CVCS HUTs are discharged through the plant.

vent to determine the most limiting organ.

L S. The high alarm setpoint, Sp, may be established at a lower value than the lowest computed value via the setpoint equation.

6. Contaiment Pressure Reliefs will not have a recomputed high alarm setpoint, but wi use the normal high alarm setpoint due to their randomness and the time constraints involved in recomputation. L

7. At certain times, it may be desirable to increase the high alarm setpoint, ff the vent flow rate is decreased. This may be accomplished in one of two ways.

Max Conc (uCi/cc)

Max Flowrate(cfm) =

New Max Concentration(,uCi/cc)

-OR-L

Reference. PMP-6010-OSD-O01 I Rev. 18 l Page 31 orS _

OFF-SITE DOSE CALCULATION MANUAL Max Conc (,uCicc)* Max Flowrate(in) NOfMa p i New Max Flowrate(cfm)

b. Waste Gas Storage Tanks 1; The gaseous effluents discharged fromthe Waste Gas System are m)nitored by the vent stack monitors VRS- 1505 and VRS-2505.

2. In the. event of a high radiation alarm, an automatic termination of the release from the waste gas system will be initiated from the plant vent radiation monitor low range noble gas channel (VRS-1505 or VRS-2505). Therefore, for any gaseous release configuration, which includes normal operation and waste gas system gaseous discharges, the ala=n setpoint of the plant vent radiation monitor will be recomputed to determine the most limiting organ based on all gaseous-effluent source terms..

Chemical and Volume Control SystemHold Up Tanks (CVCS HUT), containing high gaseous oxygen concentrations, may be released under the guidance of waste gas storage tank utilizing approved Operations' procedures.

3. It is normally prudent to allow 45 days of decay prior to releasing a Gas Decay Tank (GDT).. There are extenuating, operational circumstances that may prevent this from occurring. Under these' circumstances, such as high oxygen concentration creating a combustible atmosphere, it is prudent to waive the 45-day decay fu:r safety's sake.

c. Containment Purge and Exhaust System

1. The gaseous effluents discharged by the Containment Purge and Exbhau.A Systems and Instrumentation Room Purge and Exhaust System are monitored by the plant vent radiation monitor noble gas channels (VRS-1505 fbr Unit .1,VRS-2505 for Unit 2); and alarms

and trip actions will occur prior to'exceeding the limits in step 3.2.4a, Dose Rate.

2. For the Containment System, a continuous air sample fEom the containment atmosphere is drawn through a closed, sealed system to the radiation monitors (Tag No. ERS-1300/1400 for Unit 1 and ERS-2300t400 for Unit 2). During purges, these monitor setpoints will give a Purge and Exhaust Isolation signal upon actuation of high alarm setpoints for particulate and noble gas channels. The sample is then returned to containment. Grab sample analysis is performed for a Containment purge before release.

3. The Upper Containment area is monitored by normal range area gamma monitors (Tag No. VRS-101/1201 forUnit 1 and VRS-2101/2201 for Unit 2), which also give Purge and Exhaust Isolation Trip signals upon actuation of their high alarm

Reference PMIP-6010-OSD-001 Rev. 18 I Pa e 32 of 85 OFF-SITE DOSE CALCULATION MANUAL

4. For the Containment Pressure Relief System, no sample is routinely taken prior to release, but a sample is obtained twice per month

5. The containment airborne and area monitors, upon actuation of their high alarm, will automatically initiate closure of the Containment and Instrument Room purge supply and exhaust duct valves and containment pressure relief system valves. Complete trip of all isolation control devices requires high alarm of one of the two Train A monitors (ERS-1300t2300 or VRS-1 101Y2101) and one of the two Train B monitors (ERS-1400/2400 orVRS-120112201).

d. Steam Jet Air Ejector System (SJAE)

1. The gaseous effluents from the Steam Jet Air Ejector System discharged to the environment are continuously monitored by radiation monitor (Tag No. SRA-1900 for Unit 1 and SRA-2900 for Unit 2). The monitor will alarm prior to exceeding the limits of step 3.2.4a, Dose Rate. The alarm setpoint for the Condenser Air Ejector System monitor will be based on the maximun air ejector exhaust flow rate, (Attachment 3.15, Plant Gaseous Effluent Parameters).

The alami setpoint value will be established Using the following unit analysis equation: L SF*MRP *DLjL F P. >D ( WI *'DCFU)L Z Ssl = themaximumsetpointbasedonthemostlimiting organ, in p:i/cc and where the other terms are as previously defined

e. Gland Seal Condenser Exhaust

1. The gaseous effluents fromthe Gland Seal Condenser Exhaust discharged to the environnment are continuously monitored by L radiation monitor (Tag No. SRA-1800 for Unit 1 and SRA-2800 for Unit 2). The radiation monitor will alarm prior to exceeding the limits of step 3.2.4a, Dose Rate. The alarm setpoint for the GSCE monitor will be based on the maximum condenser exhaust flow rate (1260 L

CFM for Unit 1, 2754 CFM each for tbe two Unit 2 vents). The alarm setpoint value will be established using the following unit analysis equation SF *MRP*DL, Fp*XIQ*,(Wi*DCFqV)

Where;

Reference - PMP-6010-OSD-001 Rev. 18 l Pae 33 Of OFF-SITE DOSE CALCULA77ON MANUAL Sosm = the maximum setpoint, based on the most limiting organ, in pCi/cc and where the other terms are as previously defined 3.4 Radioactive Effluents Total Dose

.3.4.1 The cunmlative dose contributions from liquid and gaseous effluents will be determined by summing the cumulative doses as derived in steps 3.2.3c (Dost),

3.2.4b (Dose - Noble Gases), and 3.2.4c (Dose -Iodine-131, lodine-133, Tritium, and Radioactive Material in Particulate Form) of this procedure. Dose contrbution from direct radiation exposure will be based on the results of the direct radiation monitoring devices located at the REMP monitoring stations.

See NUREG-0 133, section 3.8.

3.5 Radiological Environmental Monitoring Program (REMP) 3.51 Purpose of the REMP

a. The purpose of the REMP is to:

Establish baseline radiation and radioactivity concentrations in the environs prior to reactor operations,

Monitor critical environmental exposure pathways,

Determine the radiological impact, if any, caused by the operation of the Donald C Cook NuclearPlant upon the local environment.

b. The first purpose of the REMP was completed prior to the initial operation of either of the two nuclear units at the Donald C. Cook Nuclear Plant Site.

The second and third purposes of the REP axe an on-going operation aad as such various environmental media and exposure pathways are examined. The various pathways and sample media used are delineated in Attachment 3.19, Radiological Environmental Monitoring Program Sample Stations, Sample Types, Sample Frequencies. Included is a list of the sample media, analysis required, sample stations, and frequency requirements for both collection and analysis. Attachment 3.19,

Radiological Environmental Monitoring Program Sample Stations, Sample Types, Sample Frequencies, defines the scope of the REMT for the Donald C Cook Nuclear Plant.

3.5.2 Conduct of the REMP [td.5.2-1u]

a. Conduct sample collection and analysis for the REM? in accordance with Attachment 3.19, Radiological Environmental Monitoring Program Sample Stations, Sample Types, Sample Frequencies, Attachment 3.20, Maximum Values for Lower Limits of DetectionsAB - REM?, and Attachment 3.21, Reporting Levels for Radioactivity Concentrations in.

Environmental Samples. These are applicable at all times. The on-site monitoring locations are shown on Attachment 3.22, On-Site Monitoring Location - REMP, and the off-site monitoring locations are shown on Attachment 3.23, Off-Site Monitoring Locations - REMN.

Reference PMP-6010-0SD-001 l Rev. 18 Pa e 34 of 85 OFF-SITE DOSE CALCULATION MANUAL

1. Perform each surveillance requirement within the specified time interval in Attachment 3.19, Radiological Environmental Monitoring Program Sample Stations, Sample Types, Sample Frequencies, with a maximum allowable extension not to exceed 25% of the surveillance interval.

2. If an environmental sample cannot be collected in accordance with step 3.5.2a, submit a description of the reasons for deviation and the actions taken to prevent a reoccurrence as part of the Annual Radiological Environmental Operating Report (AREOR).

Deviations from the required sampling schedule are permitted if specimens are unobtainable due to hazardous conditions, seasonal unavailability, or malfunction of automatic sampling equipment. If the deviation from the required sampling schedule is due to the malfuiction of automatic sampling equipment, make every effort to complete the corrective action prior to the end of the next sampling period.

3.. If a radionuclide is detected in any sample medium exceeding the limit established in Attachment 3.21, Reporting Levels for' Radioactivity Concentrations in EIivironmental Samples, or if more than one radionuclide is detected in any sample medium and the Total Fractional Level (TF-L), when averaged over the calendar quarter, is greater than or equal to 1, based on the following formula:

A=L Lv,LL Car +' Co0.2 ]

Where; Cal) = Concentration of Ig detected nuclide

, " nCncentrationof2". detectednuclde L(1) = Reporting Level of lI nuclide from Attachment 3.21, Reporting Levels for Radioactivity Concentrations in Environmental Samples.

L<Z) = ReportingLevelof2nanuclidefromAttachment 3.21, Reporting Levels for Radioactivity Concentrations in Environmental Samples.

Anid, if the activity is the result of plant effluents, evaluate the release conditions, environmental factors, or other aspects, which may have contributed to the identified levels for inclusion in the AREOR. If the radioactivity was not a result of plant effluents, descnibe the results mn the AREOR'

4. If a currently sampled milk farm location becomes unavailable, conduct a special milk farm survey within 15 days.

a) If the unavailable location was an indicator farm, an alternate sample location may be established within eight miles of the Donald C. Cook Nuclear Plant, if one is available.

F Reference l PMP-6010-OSDOO1 I Rev. 18 l Page 35 of 85 L OFF-SITE DOSE CALCULATION MANUAL I b) If the unavailable location was a background farm, an alternate

sample location may be established greater than 15 but less than 25 miles of the Donald C. Cook Nuclear Plant in one of the less.

prevalent wind direction sectors, if one is available.

c). If a replacement fam is unobtainable and the total number of indicator farms is less than three or the background farms is less; than one, performmonthly vegetation sampling in lieu of milc sampling.

BAISES - RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM (REM?)

.[Current TS]

The REMN provides measurements of radiation and radioactive materials in those exposure pathways and for those radionuclides, which lead to the highest potential radiation exposures of individuals resulting from the station operation. Thereby, this monitoring program supplements the radiological effluent monitoring program by verifying the measurable concentration of radioactive materials and levels of radiation are not higher than expected on the basis of the effluent measurements and modeling of the environmental exposure pathways.

The initially specified REMP will be effective for at least the first three years of commercial operation. Program changes may be initiated based on operational experience in accordance with the requirements of technical specification 6.8.4.b.

[Improved TS]

The REMP provides measurements of radiation and radioactive materials in those exposure pathways and for those radionuclides, which lead to the highest potential radiation exposures of individuals resulting from the station operation. Thereby, this monitoring program supplements the radiological effluent monitoring program by verifying the measurable concentration of radioactive materials and levels of radiation are not higher than expected on the basis of the effluent measurements and modeling of the environmental exposure pathways.

The initially specified REMP was effective for the first three years of commercial operation.

Program changes may be initiated based on operational experience in accordahce with the requirements of Technical Specification 5.5.1.c.

The detection capabilities, required by Attachment 3.20, Maximum Values for Lower Limits of D etectionsB--REMP,' are the state-of-the-art for routine environmental measurements in industrial laboratories, It should be recognized that the LLD is defined as a priori (before the fact) limit representing the capability of a measurement system and not as a posteriori (after the fact) limit for a particular measurement. Analyses shall be performed in such a manner that the stated L I)s will be achieved under routine analysis conditions. Occasionally, background fluctuations, unavoidably small sample sizes, the presence of interfering nuclides, or other uncontrollable circumstances may render these LLTDs unachievable. In such cases, the contributing factors will be identified and described in the Annual Radiological Environmental Operating Report

Reference I PMP-6010-OSD-001 l Rev. 18 l Page 36 of 85 OFF-SiTE DOSE CALCULATION MANUAL 3.5.3 Annual Land Use Census [Ref. 5.2.1u)

a. Conduct a land use census and identify the location of the nearest milk animal, the nearest residence and the nearest garden of greater than 500 square feet producing fresh leafy vegetables in each of the ten land sectors within a distance of five miles.

b. In lieu of ite garden census, grape and broad leaf vegetation sampling may be performed as close to the site boundary as possible in a land sector, containing samcple media, with the highest average deposition factor (D/Q) value.

c. Conduct this land use census annually between the dates of June l and October 1 by door-to-door survey, aerial survey, orby consulting local agricultural authorities.

1.. With a land use census identifying a location(s), which yields a calculated dose or dose commitment greater than the values currently being calculated in this document, make appropriate changes to incorporate the new location(s) within 30 days, if possible.

BASES -- LAND USE CENSUS L This is provided to ensure changes in the use of unrestricted areas are identified and modifications to the monitoring program are made in accordance with requirements of TS 6.8.4b, if required by the results of the census. This census satisfies the requirements of Section IV.B.3 of Appendix I to 10 CFR Part 50. Restricting the census to gardens of greater than 500 square feet provides assurance that significant exposure pathways via leafy vegetables will be identified and monitored since a garden of this size is the minimum required to produce the quantity (25 kg/yr) of leafy L

vegetables assumed in Regulatory Guide 1.109 for consumption of a child. To determine this minim garden size, the following assumptions were used: 1) that 20% of the garden was used for growing broad leaf vegetation (that is, similar to lettuce and cabbage), and 2) a vegetation field of 2- kg/square meter.

3.5.4 Interlaboratory Comparison Program L

a. In order to comply with Reg. Guides 4.1 and 4.15, the analytical vendor participates in an Interlaboratory Comparison Program, for radioactive materials.

Address program results and identified deficiencies in the AREOR.

1. With analyses not being performed as required above, report the corrective actions taken to prevent a rectirence to the Commission in the AREOR.

BASES - INTERLABORATORY COMPARISON PROGRAM The requirement for participation in an Interlaboratory Comparison Program is provided to L ensure independent checks on the precision and accuracy of the measurements of radioactive material in environmental sample matrices are performed as part of the quality assurance program for environmental monitoring in order to demonstrate the results are reasonably valid. I

F Reference I PMP-6010-OSD-001 I Rev. 18 I Page37of I]

OFF-SITE DOSE CALCULATION MANUAL j 3.6 Radioactive Equipment Storage Facility (Mausoleum) Groundwater Monitoring Program 3.6.1 Purpose of the Radioactive Equipment Storage Facility (Mausoleum) Groundwater Radiological Monitoring Program

a. The purpose of the temporary on-site Radioactive Equipment Storage Facility (Mausoleum) Radiological Monitoring Program was to establish baseline radiological data for the groundwater surrounding the facility prior to the storage of the Unit 2 Steam Generator Lower Assemblies. Thereafter, the purpose is to monitor the groundwater through observation wells with locations as shown in Attachment 3.22, On-Site Monitoring Location - REMP, to determine the radiological impact, if any, caused by the use of the Storage Facility.

3.6.2 Conduct of the Radioactive Equipment Storage Facility (Mausoleum) Groundwater Radiological Monitoring Program

a. Collect and analyze groundwater samples in accordance with Attachment 3.19,

-Radiological Environmental Monitoring Program Sample Stations, Sarple Types, Sample Frequencies. Apply the values from Attachment 3.20, Maximum Values for Lower Limits of DetectionsA - REMP, (excluding I- 131) and Attachment 3.21, Reporting Levels for Radioactivity Concentrations in Environmental Samples, (excluding I-131).

3.7 Meteorological Model 3.7.1. Three towers are used to determine the meteorological conditions at Donald C.

Cook Nuclear Plant. One of the towers is located at the Lake Michigan shore[ine to determine the meteorological parameters associated with unmodified shoreline air.

The data is accumulated by microprocessors at the tower sites and normall y transferred to the central computer every 15 minutes.

3.7.2 The central computer uses a meteorological software program to provide atmospheric dispersion and deposition parameters. The meteorological mode:l used is based on guidance provided in Reg. Guide 1.1 11 for routine releases. AU calculations use the Gaussian plume model.

3.8 Reporting Requirements 3.8.1 Annual Radiological Environmental Operating Report (AREOR) a Submit routine radiological environmental operating reports coverig the.

operation oftheunits during the previous calendar yearprior to May 1 oi each year.

b. Include in the ARBOR:

Summaries, interpretations, and statistical evaluation of the results o:f the radiological environmental surveillance activities for-the reporting period.

Reference PMP-6010-OSD-OO1 Rev. 18 Page 38 of 85 OFF-SITE DOSE CALCULATION MANUAL .

A comparison with pre-operational studies, operational controls (as appropriate), and previous environmental surveillance reports and an assessment of the observed impacts of the plant operation on the environment.

The results of the land use censuses required by step 3.5.3, Annual Land Use Census.

If harmful effects or evidence of irreversible damage are detected by the monitoring, provide in the report an analysis of the problem and a planned course of action to alleviate the problem.

Summarized and tabulated results of all radiological environmental samples taken during the reporting period. In the event that some results are not available for inclusion with the report, submit the report noting and explaining the reasons for the missing results. Submit the missing data as soon as possible in a supplementary report.

A summary description of the REMP including sampling methods for each sample type, size and physical characteristics of each sample type, sample preparation methods, analytical methods, and measuring equipment used.

A map of all sample locations keyed to a table giving distances and directions from one reactor. L

The results of participation in the Interlaboratory Comparison Program L

3.8.2 b.

required by step 3.5.4, Interlaboratory Comparison Progra Annual Radiological Effluent Release Report (ARERR) a Submit routine ARERR covering the operation of the unit during the previous 12 months of operation within 90 days after January 1 of each year.

Include in the ARERR a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the units as outlined in L

L Reg. Guide 1.21, 'Measuring, Evaluating and Reporting' in Solid Wastes and Releases of Radioactive Materials in Liquid and Gaseous Effluents from Light-Water Cooled Nuclear Power Plants," with data summarized on a quarterly basis following the format of Appendix B, thereof. L

c. Submitin the ARERR 90 days after January 1 of each year and include a quarterly summary of hourly meteorological data collected during the reporting period.

L

This summary may be in the form of an hour-by-hour listing of wind speed, wind direction, a spheric stability, and precipitation (if measured) on magnetic tape, or in the form ofjoint frequency L

distributions of wind speed, wind direction and atmospheric stability.

Include an assessment of the radiation doses due to the radioactive liquid and gaseous effluents released from the unit or station during the previous calendar year.

L L.

Reference I PMP-6010-SD-O001 Rev. 18 l Page 39 of35 L OFF-SITE DOSE CALCULATION MANUAL

Include an assessment of the radiation doses from radioactive liquid and gaseous effluents to members of the public due to their activities inside the site boundary during the reporting period. Include all assumptions used in making these assessments (that is, specific

activity, exposure time and location) in these reports.

Use the meteorological conditions concurrent with the time of release of radioactive materials in gaseous effluents (as determined by sampling frequency and measurement) for determining the gaseous pathway doses.

Inoperable radiation monitor periods exceeding 30 continuous days, explain causes of inoperability and actions taken to prevent reoccurrence.

d. Submit the ARERR [Ref. 5221w] 90 days after January 1 of each year and include an assessment of radiation doses to the likely most exposed member of the public from reactor releases and other nearby uranium fuel cycle sources (including doses from primary effluent pathways and direct radiation) for the previous 12 consecutive months to show conformance with 40 CFR 190, Environmental Radiation Protection Standards for Nuclear Power Operation. Acceptable methods for calculating the dose contribution from liquid and gaseous effluents are given in Reg. Guide 1.109, Rev.1.

e. Include in the ARERR the following information for each type of solid waste shipped off-site during the report period.

Volume (cubic meters),

Total curie quantity (specify whether determined by measurement or-estimate),

Principle radionuclides (specify whether determined by measurement or estimate),

Type of waste (example: spent resi, compacted dry waste, evaporator

.bottorms),

Type of container (example: LSA, Type A, Type B, Large Quantity),

-AND-

Solidification agent (example: cement).

£ Include in the ARERR unplanned releases of radioactive materials in gaseous and liquid effluent from the site to unrestricted areas on a quarterly basis.

g. Include in the ARERR any change to this procedure made during the reporting period. c

Reference PMP-6010-OSD-001 Rev. 18 l Page 40 of 85 OFF-SITE DOSE CALCULATION MANUAL 3.9 10 CFR 50.75 (g) Implementation 3.9.1 Records of spills or other unusual occurrences involving the spread of contamination' in and around the site. These records may be limited to instances when significant contamination remains after decontamination or when there is a reasonable likelihood that contaminants may'have spread to inaccessible areas, as in the case of possible seepages.

3.9.2 These records shall include any known information orfidentification of involved nuclides, quantities, 'and concentrations.

3.9.3 This information is necessary to ensure all areas outside the radiological-restricted area are documented for surveying and remediation during decommissioning.

There is a retention schedule file'number where this information is filed in Nuclear Documents Management to ensure all required areas are listed to prevent their omission.

3.10 Reporting/Management Review 3.10.1 3.10.2 Incorporate any changes to this procedure in the ARERR.

Update this procedure when the Radiation Monitoring Systen, its instruments, or L

3.10.3 the specifications of instruments are changed.

Review or revise this procedure as appropriate based on the results of the land use L

census and RENMP.

3.10.4 Evaluate any changes to this procedure for potential impact on other related 3.10.5 Department Procedures.

Review this procedure during the first quarter of each year and update it if L

necessary. Review Attachment 3.16, 10 Year Average of 1995-2004 Data, and document using Attachment 3.17, Annual Evaluation of X/Q and D/Q Values For All Sectors. Thez'X/Q and D /Q values wll be evaluated to ensure all data is.

L within +/- 3 standard deviations of the 10 year annual average data and documented by completing Attachment 3.17, Annual Evaluation of x/Qand D/Q Values For All Sectors, and filed in accordance with the retention schedule.

L 4 FINAL CONDITIONS 4.1 None.

Reference I PMP-6010-OSD-O01 . I I Rev. 18 l Page 41 of 85

. OFF-SITE DOSE CALCULATION MANUAL 5 REFERENCES 5.1 Use

References:

5.1.1 .Implementation of Programmatic Controls for Radiological Effluent Technical Specifications in the Administrative Controls Section of the Technical Specifications and the Relocation of Procedural Details of RETS to the Off-Site Dose Calculation Manual or to the Process Control Program (Generic Letter 39-01)", United States Nuclear Regulatory Commission, January 31, 1989 5.1.2 12-THP-6010-RPP-601, Preparation of the Annual Radioactive Effluent Release Report' 5.1.3 12-THP-6010-RPP-639, Annual Radiological Environmental Operating Report (AREOR) Preparation And Submittal 5.2 Writing

References:

5.2.1 Source

References:

a. 10 CFR 20, Standards for Protection Against Radiation

b. 10 CER 50, Domestic Licensing of Production and Utilization Facilities

c. PMI-6010, RadiationProtection Plan

d. NUREG-0472

e. NUREG-0133

f. Regulatory Guide 1.109, non-listed parameters are taken fromthese data tables

g. Regulatory Guide 1.111

h. Regulatory Guide 1.113

i. Final Safety. Analysis Report (FSAR)

j. [Current TS]

Technical Specifications, Appendix A, Sections 6.8.1.e, 6.8.4.a, 6.8.4.b, 6.9.1.6, 6.9.1.7, anid 6.14, Off-Site Dose Calculation Manual

[Improved TS)

Technical Specifications 5.4;1.e, 5.5.1.c, 5.5.3,5.6.2, and 5.6.3 l

k. Final Environmental Statement Donald. C. Cook Nuclear Plant, August 1973 L NUREG-0017

m. ODCM Setpoints for Liquid [and Gaseous] Effluent Monitors (Bases),

ENGR 107-04 8112.1 Environs Rad Monitor System

n. HPPOS-223, Consideration of Measurement Uncertainty When Measuring Radiation Levels Approaching Regulatory Limiits.

o. Watts - Bar Jones (WBJ) Document, R-86-C-001, The Primary Calibration of Eberline Instrument Corporation SPING - 3/4 Low, Mid, and High Range Noble Gas Detectors

p. WBJ Document, R-86-C-003, The Primary Calibration of Eberline Instrument Corporation DAM-4 and Water Monitor

q. 40 CFR 190, Environmental Radiation Protection Standards for Nuclear Power Operations

r. NRC Comnitment 6309 (N94083 dated 11110/94)

Reference PMP-6010-OSD-001 Rev. 18 Pa e 42 of 85 OFF-SITE DOSE CALCULATION MANUAL

s. NRC Commitment 1151 L

L. NRC Commitment 1217

u. NRC Commitment 3240

v. NRC Commitment

w. NRC Commitment 3850, 4859 L

x. NRC Commitmenit 6442 y... NRC Commitment 3768

z. DlT-B-00277-00, HVAC Systems Design Flows aa. Regulatory Guide 1.21.

bb. Regulatory Guide 4.1 cc. 1-2-V3-02-Calc #4, Unit Vent Sample Flow rate for isokinetic '

particulates and Iodine sampling dd. HPS N13.30-1996, Appendix A Rationale for Methods of Determining Minimum Detectable Amount (MDA) and Minimum Testing Level (MDL ee. DIT-B-01971-00, DosePactors for Radioactive Particulate Gaseous Effluents Associated with the Child by the Inhalation Pathway L ff. DIT-B-01987-00, GroundPlane &FoodDoseFactorsPi for Radioiodines and Radioactive Particulate Gaseous Effluents gg. NRC Commitment 1010 L 5.2.2 General References

a. Cook Nuclear Plant Start-Up Flash Tank Flow Rate letter from D. L Boston dated January 21, 1997 L

b. Letter from B.P. Lauzau, Venting of Middle CVCS Hold-Up Tank Directly to Unit.Vent, May 1, 1992

c. AEP Design Information Transmittal on Aux Building Ventilation L

Systems d.. PMP-4030.EIS.O01, Event-Initiated Surveillance Testing

e. Environmental Position Paper, Fe Impact on Release Rates, approved L

3/14/00

f. Environmental Position Paper, Methodology Change from Sampling Secondary System Gaseous Effluents for Power Changes Exceeding 15%

within :hr to Responding to Gaseous Alert Alarms, approved 4/4/00 L

g. CR 02150078, RRS-1000 efficiency curve usage L h Environmental Position Paper, Unit Vent Compensatory Sampling, approved 4114/05 IL L

L

Reference PUP-6010-OSD-001 Rev. 18 PTa ge 43 of 35 OFF-SITE DOSE CALCULATION MANUAL kttaclm-nnt 3.1 l Dose Factors for Various Pathways l ages46 43 -46 R; Dose Factors PATHWAY Nuclide Ground Vegetable Meat Cow Milk Goat Milk Inhalation H-3 0.0E+00 4.0E+03 3.3E+02 2.4E+03 4.91E+03 1.31E3+(3 C-14 O.E013+00 3.513+06 5.3E+05 3.2E3+06 3.21B+06 3.61E+04 Cr-';1 5.4BE06 1.1I+07 1.5E1+06 6.913+06 . 8.313+05 2.1B3+04 Mn--54 - 1.63+-09 9.4E4+08 2.1E3+07 2.913+o7 3.5SI+06 2.0O3+06 Fe-59 3.2E+08 9.61E+08 1.7E+09 3.113+08 4.OE+07 1.51E+06.

Co-58 4.4AE08 6.0E+08 2.9E+08 8.41B+07 1.OE+07 1.3E1+06 Co-5O 2.5E+10 3.2E1+09 1.0Et+9 2.713+08 3.22+07 8.6E3+06 Zn-65 8.5E+08 2.7E+09 9.5E+08 1.6E+10 1.9E+09 1.2E3+06 Sr-F9 2.5iB+04 3.5B+10 3.8E+08 9.9E+09 2.1E+10 2.413+06 Sr-50 O.OE+00 1.4E}+12 9.6E+09 9.4E+10 2.0E+11 1.1E3+08 Zr-95 _ 2.9E+08 1.2B+09 1.5E+09 9.3B+05 1.1E+05 2.7E3+06 Sb-:24 6.9E+08 3.0BO+09 4.48t+08 7.28t+08 8.6E+07 3.8B+06 I-131 1.OE+07 2.4E1+10 2.58+09 *4.88+11 5.8E+11 1.6B+07 I-133 __1.5Bt+06 4.03+08 6.0Bt+01 4.413+09 5.3E.+09 3.813+06 Cs-:134 .7.9B+09 2.5E+10 1.1B+09 5.OE+10 1.5E+11 1.1E+06 Cs-:l36 1.7E+08 2.213+08 4.2E+07 5.1B+09 1.5E+10 1.9B+05 Cs-:l37 - 1.2E+10 2.5B+10 1.0E+09 4.5B+10 1.4E+11 9.01+05 Ba- 140 2.3E1+07 2.713+08 5.2E+07 2.1E1+08 2.6E+07 2.0E+06 Ce-:41 _ 1.5E+07 5.3B+08 3.0E+07 8.3E+07 1.0E+07 6.1B+05 Ce-144 - 7.9E+07 1.3E+10 3.6B+08 7.3E+08 8.7E+07 1.3E+07 3

hnitn fcr adl except inha~hicc pathway sr m2nmrsec/yT da3,inhadmn pathway urits are mr m yr pCi.

U p Values to be Used For the Maxinmim Exposed Individual Path ay ___ . ifant Child Teen Adult Fruits, vegetables and grain (kglyr) _ 520 630 520 Lealfy vegetailes (kg/yr) - 26 42 64 MM: (IJyr) 330 330 400 310 Meatand poulty (kg/yr) -- 41 65 110 Fishb (kg-/r) - 6.9 16 21 Drirding water (Iyr) 330 510 510 730 Shr-eline recreation (br/yr) 14 67 12 Inhb2ation (Wl/yr) 1400 3700 8000 8000 Tabl E-5 of R~g. Guide 1.109.

4-Reference . PMP-6010.OSD-OO1 Rev. 18 Page 44 of 85 OFF-SITE DOSE CALCULATION MANUAL

. .Pages: .1 Dose FPctors for Various Pathways 3-46 L

Bjp Factors for Aquatic Foods pCil/kg pCi '

L Element Flsh Invertebrate H 9.OE-1 9.OE-1 C 4.6E3 9.1E3 Na 1.0E2 2.0E2 P l.OE5 2.0E4 Cr 2.0E2 2.0E3 Mn 4.0E2 9.0E4

.Fe 1.OE2 3.2E3 L Co 5.OE1 2.0E2 Ni 1.OE2 1.OE2 Ch 5.OEI 4.0E2 Zn 2.0E3 1.0B4 Br 4.2E2 3.3B2 Rb 2.0E3 1.003 Sr . 3.011 1.0B2 Y

Zr Nb 2.51 3.3EO 3.0E4 1.0L3O 6.7E0 l.OE2 L

Mo 1.0EI 1.OE1 T Q Ru 1.521 1.0B1 5.0EO 3.0E2 L

Rh 1.0E1 3.0E2 Te I .

4.0E2 1.5E1

. 6.1E3 5.0E0 L

Cs 2.0E3 1.0O3 Ba La 4.0EO 2.5E1 2.0E2 1.03 L

Ce 1.00 1.103 Pr Nd

_ 2.521 2.521 1.0O3 l.OE3 L W 1.23 1.021 Np TableA-l of Reg. Gide 1.109.

1.OB1 4.0E2 L

L L

L

Reference PMP-6010-OSD-O001 Rev. 18 l Page 45 of 85 OFF-SllE DOSE CALCULATION MANUAL Attachrnmnt 3.1 Dose Factors for Various Pathways Pages:

I 43 -46

.D.p External Dose Factors for Standing on Contaminated Ground mrem m2 Ihr pCi Radionuclide Total Body Skin H-3 °0 0 C-14 0 0 Na-24 2.5B-8 2.9B3-8 P-32 0 0.

Cr-51 2.2E-10 2.6E-10 Mn-54 5.8E-9 6.8E-9 Mn-56 l.lE-8 1.3E-8 Fe-55 0 0 Fe-59 8.0E-9 9.4E-9 Co-58 7.OE-9 8.2E-9 Co-60 *1.71-8 2.OE-8 Ni-63 -0 0.

Ni-65 3.7B-9 4.3-9 Cu-64 1.5E-9 1.7E-9 Zn-65 4.01-9 4.6E-9 Zn-69 0 0 Br-83 6.4E-11 9.3B-11 Br-84 .1.21-8 1.4B-8 Br-85 0 0 Rb-86 6.3E-10 7.2B-10 Rb-88 3.5E-9 4.013-9 Rb-89 . 1.5E-8 1.81-8 Sr-89 5.6E-13 6.51-13 Sr-91 7.1E-9 8.3E-9 Sr-92 9.OE-9g 1.O-8 Y-90 2.2E-12 2.61-12 Y-91m 3.8E-9 4.4-9 Y-91 2.41-11 2.7B-11 Y-92 1.6E-9 1.91-9 Y-93 5.7E-10 7.8E-10 Zr-95 5.0E9 5.8E-9 Zr-97 5.51-9 6.4E-9 Nb-95 5.113-9 6.01-9 Mo-99 1.9E-9 2.2E-9 Tc-99m 9.6E1-10 1.1-9 Tc-101 2.7E-9 3.0E-9 Ru-103 3.6E1-9 4.2E-9 Ru-105 4.5E-9 5.1E-9 Ru-106 1.5E-9 1.8E-9 Ag-l110m 1.8E-8 2.1E-8 Te-125m 3.5E-11 4.8E-l1 Te-127m 1.1E132 1.3E1-12

Reference 7 PMP-6010-OSD-001 Rev. 18 Page 46 of 85 OFF-SITE DOSE CALCULATION MANUAL j Attachment 3.1 Dose Factors for Various Pathways l 4 Raionucdlide Total Body Skin Te-127 LOB-11 l.lE-11.

Te-129m 7.7E-10 9.0B-10 Te-129 7.1-10 E8.4-10 Te-131m 8.4E-9 9.9B-9 Te-131 2.2E-9 2.6E-6 Te-132 1.7E-9 2.OE-9 1-130 1.4E-8 1.7B-8 I-131 2.8E-9 3.48-9 1-132 1.7B-8 2.0E-8 1-133 3.7E-9 4.58-9 1-134 1.60-8 l.9E-8 I-135 1.2E-8 1.4E-8 Cs-134 1.28-8 1.4E-8 Cs-136 1.5E-8 1.7,-8 L Cs-137 4.2E-9 4.9E-9 Cs-138 2.12-8 2.4E-8 Ba-139 2.4-9 2.7E-9 L~

Ba-140 2.1B-9 2.4E-9 Ba-141 Ba-142 La-140 4.3E-9 7.9E-9 1.5E-8 4.9E-9 9.02-9 1.7E8 L

La-142 1.5E-8 1.8E-8 Ce-141 Ce-143 5.5E-10 2.2B-9 6.2E-10 2.5E-9 L

Ce-144 3.2E-10 3.71-10 Pr-143 Pr-144 0-2.0E-10 0

2.31-10 L

Nd-147 1.0L-9 1.22-9 W-187 Np-239 3.1E-9 9.5E-10 3.6E,9 1.1-9 L

Table E-6 of Reg. Guido 1.109.

L~

L

Reference. . PMP-6010-OSD-001 . I Rev. 18 Page 47 of 85 OFF-SITE DOSE CALCULATION MANUAL Pages:

Attachment 3.2 Radioactive Liquid EffluentMonitoring Instruments 4748 -

INSTRUVMENT Applicabilit Channels

.__ __ _ _ ___ . .Operable _

1. Gross Radioactivity Monitors Providing Automatic Release Termination

a. Liquid Radwaste (1)# At times of release 1 Effluent Line (RRS-1001) . _-

b. Steam Generator (1)# At times of release** 2 Blowdown Line (R-19, DRS 314100 +) _

c. Stem Generator (1)# At times of release 2 Blowdown Treatment Effluent (R-24, DRS 3/4200 +) _ _

2. Gro ss Radioactivity Monitors Not Providing Automatic Release Termination

a. Senrice Water (1) per At all times 3 System Bffluent Line(R-20, R-28) train

3. Continuous Composite Sampler Flow Monitor

a. Turbine Building Sump (1) At all times 3 Efflaent Line

4. Flow Rate Measurement Devices

a. .Liquid Radwaste Line (I) At times of release 4 (RF[-285) . ._._.

b. Discharge Pipes* . (1) At all times - TA

c. Steam Generator Blowdown (1) At times of release 4 Treatment Effluent (DFI-352) _

Pampcurves and valve settingp omy be utilized to estimate fow- in such ses, Action S t 4 is not aplicable. This is yiin efrnce to start up flash tank flow.

[OCbnent TS)

OPEFABIL1TV' of RRS-1001 includes OPERABULfLY of sample flow switch RPS 1010, which is an attendant instrurrMt as definid by Technical Spedfication 1.6. This item is also applicable for all Eberline liquid rnxitcrs (and their respecive flow switdies) listed here.

[Imnproved TS]

OFELABU. of RRS-1001 includes OPERAULlTIY of sanmle flow switch RFS-1010, which is an attendant instnmxnt as defmd minTechnical Speafiation sectioa 1.l, under the temOperable - Opeaility. Ihis itemis also applicable faall Eberline liquid rritors (and their respective flow switches) listed heme.

Since these monitors can be used for either batch cr continuous release the appmpriate action statcme nt of I or 2 shaild apply (that is, Action 1 if a steam generator drain is being perforrd in lieu of A on 2). It is possible, due to the steam genemzx sanpling systewnlineup. that BOTHaction statements areachtally enrted This wold be the case whensanplingforsteamganerator draining requires duplicate samples while the samrple system is lined up to discharge to the cperating units blowdowc. systemL In this case the steam generator drain samples can fulfill the sample requircnent for Action 2 also. Action 2 would be exited when samprling was terninated.

+ Some Westinghouse (R) radiation nrymitcrs are being replaced by Bberline (DRO)monitcrs. Either mnonitor can fuilfil the operability requirement

Reference I PMP-6010-OSDO001 Rev. 18 l Page 48 of 85 OFF-SITE DOSE CALCULATION MANUAL Attachment 3.2 Radioactive Liquid Effluent MonitoingInstruments Pages:

a IF an RMS mcnitor is inoperable solely as the result of the loss of its control room alarm annunciation, TEIN cme of the following actions is acceptable to satisfy the ODCM action statement compansatry surveillance rirerneent:

1. Collect grab samples and conduct laboratory analyses per the specific macitcrs action statement, OR-

2. Collect local monitcr readings at a frequency equal to or greater than (mnre frequently than) the action frequency.

IF the RMS mcnitcr is inoperable for reasons other than the loss of control room annunciation, TBEN the cnly acceptable action is taling grab samples and ccnducting laboratory analyses as the reading is equivalent to a grab sample when the mncnitcr is functional.

TABLE NOTATION Action I With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases may continue. provided that pricr to initiating a release:

1. At least two independent samples are analyzed in acccrdance with Step 3.23a and;

2. At least two technically qualified members of the Pacility Staff independently verify the discharge valving. Otherwise, suspend release of radioactive effluents via this pathway. L Action 2 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requireent, effluentreleases via this pathway may continue for up to 30 days provided grab samples are analyzed for gross radioactivity (beta or gamma) at a limit of detection of at least 10.7 pCi/gramr L

1. At least mcne per shift wen the specific activity of the seccndary coolant is> 0.01 Ci/gramDOSE EQUIVALENT 1-131.

2. At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when the specific activity of the secondary coolant is <0.01 pCi/gram DOSE EQUIVALENT .1-131.

L Action 3 With the number of channels OPERABLE less than required by the Minim= Channels OPERABLE

raiemart, effluent releases'via this pathway may continue for up to 30 days provided that at least once L

per shift, grab samples are collected and analyzed for gross radioactivity (beta or gamma) at a lower limit of detection of at least 10-7 pCl/xnl. Since the Westinghouse ESW mrnitcrs (R-20 and R-28) are only used for post LOCA leak detection and have no auto trip function associated with them, grab samples ae only needed if the Contanmnt Spray Heat Exchanger Is in service.

Action 4 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE rcquiremen, effluent releases via this pathway may continue for up to 30 days provided the flow rate is L

estimated at least once per 4 hrs during actual releases.

CQmpensatmy actions an: govrerd by PMPn 43301S.OO1, Event-Initiated Surveilance Testing L

Reference PMP-6010-OSD-001 Rev. 18 Page 49 of S5 OFF-SITE DOSE CALCULATION MANUAL t 3 Radioactive Liquid Effluent Monitoring Pages:

Instrumentati6n Surveillance Requirements 49 - 50 Instnument CHANNEL SOURCE CHANNEL CHANN13L CHECK CHECK' CALIBRATION ([Current TS]

FUNCTIONAL)

([Improved. TS]

OPERATIONAL)

._ :_ .TEST

1. Gross Radbactivity Monitors Providing Automatic Release Termination

a. Liquid. Radwaste D* P R(3) Q(5)

Efflueat Line (RRS-1001)

b. Steam Generator D* M R(3) Q(1)

Blowdown Effluent Line .

c. Steam Generator D* M R(3) Q(1)

Blowdown Treatment Effluent Line'

2. Gross Radioactivity Monitors Not Providing Automatic Release Termination

a. Service Water D M R(3) Q(2)

System Effluent Line 1

3. Continuous Composite Samplers

a. Turbine Building D* N/A N/A N/A Sump Effluent Line . .

4. Flow Rate Measurement Devices

a. Liquid Radwaste D(4)* N/A R Q Effluent

b. Steam Generator D(4)* N/A N/A. N/A Blowdown Treatment Line .

Duin grtleases lna this paxhway

Reference PMP-6010-OSD-OO1 Rev. 18 Page 50 of 85 OFF-SITE DOSE CALCULATION MANUAL AtahenR3~adio active Liquid Effluent Monitorn Pages:'

Instrumentation Surveillance Requirements I 49 - 50 J TABLENOTATION,

1. [Current TS) Demonstrate with the CHANNEL FU1NCIONAL TEST that automatic isolation of this pathway and control room alarm annunciation ocours if any of the following conditions exists:

[Improved TS) Demonstrate with the CHANNEL OPERATIONAL TEST that automatic isolaticn of this pathway and control room alarm annunciation ocours if any of the following conditions exists I. Istrurnlt indicates measured levels above the alarmip setpoint.

2. Circuit failur* .

3. Instrummit indicates a downscale failure.*

4. Instbument control not set in operating mode.*

5. Loss of sample flow. #.

2. [Current TS] Demonstrate with the CHANNEL FUNCIIONAL TEST that control room alarm annunciation occurs if any of the following conditions exists:

[Improved TS] Demcnstrate with the CHANNEL OPERATIONAL TEST that control room alarm annunciation ocrs if any of the following conditions exists: L

1. Instrument indicates measured levels above the alarm setpoint.

2. Circuit failure.

3. Instrument indicates a dowmscale failure.

4. Instnmiait controls not set in operating mode.

5. Loss of sampleflow. #

3. Perfor the initial CHANNEL CALIBRATION using mle ar mcre sources with traceability back to the Natial Institute of Standards ond Tenlogy (NIS). These sources permit calibrating the system over its intended range of energy and measurement range Fcr subsequent CHANNEL CALIBRATION, sources that have been related to the initial calibration may be used.

4. Veify indicaticn of flow during periods of release with the CHANNEL CHECK. Perfcrm the CHANNEL CHECK at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months on days on which continuous, periodic cr batch releases are mde L

5. [Current TS] Demcnstrate with the CHANNEL FUNCIIONAL TEST that automatic isolation of this pathway and ocntrol room alarm annunciation occurs if any of the follovnng conditions exists:

rmprovedTS Demonstrate with the CHANNEL OPERATIONAL TEST that automatic isolatin of this pathway and control room alarm annunciation occurs if any of the following conditions exists:

L

1. Instrument indicates measured levels above the alarrltrip setpoint..

2. Circuit failrne.*

3. Instrunent indicates a downscale failure.**

L

4. Instnrment control not set in bopating mode.*

5. Loss of sample flow.

L Instrument indicates, but does not provide fcr automatic isolation

- nstruent indicates, but does not necessarily cause automatic isolation. No eredit is taken fcr the automatic isolation on such occunrences.

Applicable cnly to Elbline sample flow instrumentation

[Current TS] Opceations uarently perfamns the routine channel checks and source checks. Maintenance and Radiation Protection perfom chamel calihaaions and channel fuctimal tests. Chenistry perfcmns the darmel check on the continuous coaposite L sampler. These responsibilities are subject to change without revision to this docuin=L

[Improved IS] Operations cu=tly perforrns the routine channel checks and source checks- Maintenance and RWdialin

F Protection perfcrm channel calibrations and chnel opatiional tests. Cemistryperfcms the channel check onthe continuous

composite sampler. These responibilities are subject to change without revision to this docun=L

Reference PMP-6010-OSDD001 Rev. 18 Page 51 Dr 85

. t OFF-SITE DOSE CALCULATION MANUAL Atvachm~nt 3.4 lRadioactive Gaseous Effluent Monitoring Instrumentation Pages:

Instrument (Irstrument Operable' l inimur Action Channels

_ Action

1. Condenser Evacuatict System .

a. Noble Gas Activity (1) 1 6 Maciuir (SRA-1905/2905)

b. Flow RateMcnitcr(SM401, (1) 1 5 1/2-MR-054 and/cr SRA- 1910/2910) OR (SFR-402 and I/2-MR-054) . _

2. Unit VatL Auxiliary Building Ventilatim System .

a. Noble Gas Activity (1)

6 Manitir (VRS-150512505) .

b. Iodine Saple (1)

8 Cartiidge for VRA-I50325303 c Particulate Srnpler Filtr (1)

8 for VRA-1501/2501 .

d. Effluent System Flow Rate (1)

5 Measamg Device (VFR-315, MR-054 and/cr VFR-1510/2510) c SarriplrlowRate (1)

5 Measu-ing Device (VFS-1 521/2521) .

3. CcntAint Purge and Ccntainment Pressure Relief (Vent)

a. Ccntaihmnt Noble Gas Adivity Mcnitcr b.

ERS-1:3/1405 CcotainmtParticulateSaplPilter (1)

(1)10

2 fRS-23 7 0

ERS-1:3/1401 (ERS-2312401)

4. Waste Gas :Eoldup System and CVCS HUT

a. Noble 3asActivity . (1) ****4 9 Alurm and Ttrmiinatiml.

ofWasteGasReleases (VRS-1505/2505) . .

5. Gland Seal.Exhaust ._.

a. Noble Gas Activity (1) *** 6 Monitcr (SRA-1805/2805)

b. FlowRateMcnitc (SFR-201, MR-054 or (1) 5 SFR-1810/=810)

At all tines

- - " During releases via this pathway

Reference PMP-6010-OSD-OO1 Rev. 18 Pa e 52 0f 85 OFF-SITE DOSE CALCULATION MANUAL Attaehunt 3.4 Radioactive Gaseous Effluent Monitoring Insturnentation l .51 ags:53 TABLE NOTATIONS

1. IF an RMS mcnitcr is incperable solely as the result of the loss of it's control roosn alarm,annunation, THEN one of the following actions is acceptable to satisfy the ODCM action statement compensatory surveillance requirement

1. Take grab samples and ecnduct Iaboratcry analyses per the specific mnitcs acticn statement,

OR-

2. Take local mrnitor readings at a freqncy equal to or greater than (more frequently than) the action frequency.

IF the RMS monitor is inoperable for reasons other than the loss of control room annunciation, THEN the only acceptable action Is taking grab samples and conducting labcratary analyses as the reading Is equivalent to a grab sanmple when the monitor is functianal.

2 Considerreeases as occuming 'via this pathway" under the following conditims:

[Current TS.

The Containment Purge System is in operation and Containment integrity is established/equired,

[Improved TS]. -I The Containment Purge System is in operation and Contnt Opeability is applcable,

Tbh Cantainmit Purge System is in operation and is being used as the vent path fcr the venting of contaminated systems within the contammnt building picr to completing both degas and depressurization of the RCS.

IF neither of the above are applicable, THEN the containnmnt purge system is acting as a ventilation systm and is covered by Item 2 of this Attachment.

-OR-A CctanmentPressureRelief (CPR) is beingperfcrmed.

L

3. [Current TS.

PForprgc Cmcluingpressure reliel)purposes cnly. SeeTechnical Specification table 3.3-6 for additional informatian

[Improved TS]

For purge (including pressure relief) purposes only.. ReferenceTS 3.6.1. Containment Purge Supply and Exhaust System Isolation Instrumentation and 3.4.15, RCS Leakage Detection Instrumentation for additional information.

L

4. For waste gas releases only, see Item 2 (Unit Vent, Auxiliary Building Ventilation System) for additional requiranents. L ACTIONS

5. With 'the number of channels OPERABLE less than required by the Minimurn Channels OPERABLE requirement, effluent rleases via this palhway may continue for up to 30 days provided the flowrate is estimated at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . After 30 days, IF the channels are not OPERABLE, THEN continue releases with estimation of the flowrate onceper 4 hclrs andprovide a.description of why the inpcerability was not corrected in, the next Annual Radiological Effluent Release Repart

6. With the number of channels OPERABLE less required by the Minimum Cbannels OPERABLE requirement, effluent releases via this pathway may continue for up to 30 days provided grab samples are taken at least once per shift and these samples are analyzed for gross activity within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . After 30 days, IF the channels are not L OPERABLE, THEN cntinue releases with grab samples once per shift and provide a description of why the inopfability was not ccrrected in thenext Annual Radiological Effluent release Rcxt L.

Reference l PMP-6010.OSDOO1 Rev. 18 Page 53 of 85 OFF-SITE DOSE CALCULATION MANUAL Attacbmeni. l Radioactive Gaseous Effluent Monitoring Instrurnentation 51-53

7. 'With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE

-equirements, inmmediately suspend PURGING or VENTIG (CPR) of radioactive efflueats via this pathway.

& With the Nunr of channels OPERABLE less than required by the Minimmn Channels OPERABLE raq .ement.

efhtent

,telass vi. thecaffectedpathwaymay continue fcr up to 30 days provided srles required for weeklyanalysis are continuiusly ollected with ariliaty sampling equipmrnt as reqird inAtttacsment 3.7. Radioactive Gaseous Waste Sampling andAwnlysis

.PRogran. JAfter.30 days. IF the channels are not OPERABLE, THEN continue releases with sample collection by auxiliary

-ampling equipmnt and provide a description of why the inoperability was not corrected in the next Annual Radioligical

WlrntRelease Report.

Sanyling evolutions are not an intemupti of a continuous release Cr sampling period.

9. With the numberof channels OPERABLE less than rquired by the Miniium. Channels OPERABLE requirmehnt. the caLtents

-ofthe tank(s) may be released to the envirournent fcr up to 14 d&ys provided that prior to initiating the release:

a. At least two indepnnt samples of the tank's contents are analyzed and,

b. Al Ienst two technically qualified mrnbers of the Facility Staff independently verify the release rate calculaticma and dischirge valve lineups; otherwise, suspend release of radioactive effluents via this pathway.

10. [Current IlS.

See Technical Specificaion 3.4.6.1.

UIrnproydi TS]

See Technical Spedficatim 3A.15, RCS leakage Detection System bstrunltion.

Corpesatory actions are governed byPMP-4W30fS.001, Event-Ihitiated Surveillance Testing.

Reference PMP-6010-OSD-001 I Rev. 18 I Page 54 of 85, OFF-SITE DOSE CALCULATION MANUAL I I Instrarentation I-Radioactive Gaseous Effluent Monitoring Pages:

Atacn~t3' Surveillance Requirements I 54 -55' Instrument CHANNEL SOURCE CHANNEL CHANNEL CHECK. CHECK CALIBRATnON [Qrrent TS] FJUNCTONAL JL.

. .. . .[Improved IS OPERATnONAL

. . ,TESTr

1. CoridenseEvacuation Alarm Only.

System ' __'_.. L.

a. Noble Gas Activity Manitr (SRA-19052905)

b. SystemEffluent FlowRate D**

D**.

J M NA R(2)

R Ql (SFR-401, SFR-402, MR-054; SRA-1910/2910) . . ._ .. _ ._.__

Z Auxiliary Building Unit Alarm Only I I Ventilation Systern ._ ._ ._l

a. NoblcGasActivityMonitcr . M R(2) Q(1)

(VRS-1505t2505) .,

b. Iodine Sample NA NA NA (1kwVRA-150312503): .

c. Particulate Sampler W NA NA NA (Fkw VRA-150112501) . ._ .

dL SystemEffluit~lowRZate D* NA R Q Measurement Device (VFR-3 15, MR-054, VRS-1510/2510)

e. Sampler Flow Rate Measuring D* NIA R Q L

Device (VFS-1521/2521)

3. Ccninment Purge System and Alarm and Trip

. L Ccmtainnmt Pressure Relief a Cztainmrnt Noble Gas Activity Mcnitcr (ERS-13/1405 and ERS-23/2405)

S** P R(2) Q L

b. CcrtaunenmtPardctlate Sampler (ERS-13/1401 and ERS-23/240I)

S** NA R Q L

4. Waste Gas Holdup System Alam and Trip.

.Including CVCS HUT

a. Noble Gas Activity Maiitcr I

P** P R(2) Q(3)

L Providing Alarm and

. Tenninaticn.

(VRS-1505/2S05) . l *L

5. Gland Seal Exhaust AlarmOnly

a. Noble Gas Activity D** M R(2) Q(l)

(SRA-1805/2805) L

b. SystemEffluentFlowRate D** NA R. Q (SFRR-201, MR-054, SRA-1810J810)

Atalltimes L

- During releases via thispathway

Refer!nce l PMP-6010-OSD-001 Rev. 18 l Page 55 of 815 J

OFF-SITE DOSE CALCULATION MANUAL jvtcont 35 l Radioactive Gaseous Effluent Monitoring Pages

_ '. .Inistruenr -n. vilance Requirements 54-55 TABLE NOTATIONS

1. [Currerit TS] Demonstrate with the CHANNEL FUNCTIONAL TEST that control room alarm annunciation occurs il any of the following coiditicns exists:

[Improved TS) Demonstrate with the CHANNEL OPERATIONAL TEST that control room alarm annunciation occurs if any of the following conditions exists:

1. Instrumrnt indicates measured levels above the alarm setpoint.

2. Circuit failure.

3. Instrument indicates a downscaie failure.

4. Instrument controls not set in operate mode.

2. Perform the initial CHANNEL CALIBRATION using one or moer sources with traceability back to the MST.

These s'nrces permit calibrating the system over its intended range of energy and measurcment range. FPr subsequnt CHANNEL CALIBRATION, saorces that have been rdated to the initial calibration may be usei

3. [Current TS] Demon strate with the CHANNEL FUNCTIONAL TEST that automatic isolation of this pathway and control romrn alarm annunciation occurs if any of the following conditions exists:

bnapro'-ed TS] Demonstrate -with the CHANNEL OPERATIONAL TEST that automatic isolation of this pathway and cctro room alarm annunciatici occurs if any of the following conditions exists:

1. Ibstruiment indicates measured levels above the alarmhtip setpoinL

2. Circuit failure.

3. Instnument indicates a downscale failure.*

4. bistrument controls not set in operate mnod&

In"trument indicates, but does not provide automatic isolation.

fCruTent T9I Operaiams cunently perfcamris the routine chmnel checks, and scurce checks. Maintenance and Radistion Prctectionrm channel calibmtions and channel functional tests. These respmsibiitiesmsuect j to dange without mvision to this docauarn rImproved TS) pertions crrently perfbans the rutine channel checks, and soirce checks. Maintenance and Raditc Prctection pezform cannel calitxaficns and chaine operational tests. flse ztsp sibilties arc subject to chage wilhout revision to thi: docun~nL

Reference PMP-6010-OSD-O001 Rev. 18 - Pa e 56 of 85 OFF-SITE DOSE CALCULATION MANUAL Attachment3.6 lRadioattive Liqwia Waste Sampling and Analysis Program Pges i,

.Ref. 5.2.Isl_

LIQUID SAMPLING MINIMUM TYPE OF LOWER RELEASE FREQUENCY ANALYSIS ACTIVITY LIMIT OF TYPE FREQUENCY ANALYSIS DETECTION (LLD)

A. Batch Waste P P Principal Gamnma SX10 7 Release Tanks' Each Batch Each Batch Emitters' ___.

. . - . I-131 1x104 P

Each Batch P

Each Batch Dissolved and Entrained Gases (Gamma ex104 5 L

P Each Batch M.

Compositeb 1x104 Emitters)

H-3 1x1OV L Gross Alpha 1x10 7 L

P Each Batch.

Q CoMpositebr Sr-89, Sr-90 Fe-55 5xie 1x10 4 L

B. Plant Continuous Daily W

Composite Principal Gamma Emitters0 Sxlf 7 L Releases* a M M 1-131 Dissolved and 1X10 4 L Grab Sample Entrained Gases i1x0i5 (Gamma Emitters)

L M H-3 lxlO Daily Compositeb Gross Alpha ixi0.

L

.Q Sr-89, Sr-90 5xlO-8 Daily Compositeb Fe-55 . l10 4 L

During releases via this pathway This table provdes the minimumr requirements far the liquid sampling program. If additinal sawling is pHfdrmed then those sampleresults can be used to quantifyreleases in lieu of composite data for a more accurate quantification. Eales of these samples are the 72 howr secrndary coolant activity and Mciitcr Tank tritium samples.

Compositing of samples from more than one ccntinuous stream into the same containe is acceptable as Iong as no evidence of primary to secondary leakage exists ond separate sample results exist for tritium quantificaticn. The presence of pinmary to secondary leakage is indicated by non-natural gamna emitting radionuclides being identified in the Condenser Air Ejectcr and/cr in SIG blowdown.

Reference - PMP 6010-0SD01 Rev. 18 Page 57 of 85 OFF-SITE DOSE CALCULATION MANUAL Attachment 3.6 Radioactive Liquid W te Sampling and Analysis Program Pages:

I I 56 -57.....

TABLE NOTATION

a. 'he lower limit of detection (LLD) is defined in Table Notation A. of Attachment 3.20, Maximnurn Values fir Lower Limits of Detections" - RETP

b. A composite sample is one in which the quantity of liquid sampled is proportional to the quantity of liquid waste discharged and in which the method of sampling employed results in a specimen which is zepresentEtive of the liquids released.

c. Ak batch release is the discharge of liquid wastes of a discrete volume. Prior to sampling for analysis, i;olate, recirculate or sparge each batch to ensure thorough mixing. Examples of these are Monitor Tank and Steam Generator Drains. Before a batch is released the tank is sampled and analyzed to determi:ae that it can be released without exceeding federal standards.

d. AL continu ous release is the discharge of liquid of a non-discrete volume; e.g. from a volune of systc ;.

tjLat has an input flow during the continuous release. This type of release includes the Turbine Roomr Sump, Steam Generator Blowdown and the Steam Generator Sampling System

e. The principal gamma enitters for which the LLD specification applies exclusively are the following radionuclides: Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, Cs-134, Cs-137, Ce-141 and Ce-144. This list does not mean that ony these nuclides are to be detected and reported. Identify and report other peaks, which are measurable and identifiable, together with the above nuclides.

Reference PMP-6010-OSD-001 Rev. 18 Page 58 of 85 OFF-SITE DOSE CALCULATION MANUAL Atahen 7Rdioactive Gaseous Waste Sampling and j Pages:'

. Analysis Program 58 -59 Gaseous Release Type Frequency Minlmumi Type of Lower Ltmit Analysis Activity of Detection Frequency Analysis (lCw/ml)'

a. Waste Gas StrageTanks .PP . Principal Gamma and CVCS HUTs Each Tank Each Tank Emittersd 1x10 4 Grab Sample . . -

.___.___ ._ H-3 1x104.

b. Ccntainneut Purge P P cial GamPn.

Eadh Pirge Each Purge Eniitters 1X1O4 Grab Sample CPR(vent)** Twiceper Twice per Month Month

.H-3 1 x 104

c. Ccndenser Evacuation W crM M. Principal Gamma System Grab Sample Particulate Sample Eittrsd 11 Gland Seal M .. H-3 1x

. .

Prinlciple Gamma l x lO Noble Gas M 1-131 Emitters d L

IodineAdsacbing I x l02

. cotinuaug Media Wr Noble Gas Monitcr Noble Gases

.oblexes L

d. AuxiliaryBuilding Unit Vent*IodineAdscrbing Continuous' wb M edia 1-131 1 x 10 L

. Cninuus' w b Particulate Sample Principal Gamma Emitte'sd I x 10.1

.L Cc Ctinuc . M Gross Alpha CompositeParticulate . x 10"

___. ___. ___Sample W

Grab Sample.

,h H-3 Sample H-3

.e1 XI.

L Czntinuous C WO Noble Gas Q

PrincipleGamma Emitters d Sr-89, Sr-90 1x104 IL Conposite Particulate 1 x 10

.__ _ Sample . I.-

Ccntinucus " Noble Gas Mcnitcr Noble Gases 1 xiof

e. Incinerated Oil' P P Principal Gamma Each Batchd Each Batch Emittersd 5 x1C7

During releases via this pathway

- Only a twice per month sampling program for containmefl noble gases and H3 is required This table provides the minimum requirements fcr the gaseous sampling program If additional sampling is perforaed then those sample results can be used to quantify releases in lieu of composite data for a more accurate quantificaticn.

Examples of these samples armverificaticn or compensatory action sample results.

= Reference PMP-6010.OSDO001 Rev. 18 Page 59 of 35

. OFF-SITE DOSE CALCULATION MANUAL Attachnt 3.7 Radioactive Gaseous Waste Sampling and . Pages:

A Analysis Program 58 -59 TABLE NOTA7ION

a. The lower limit of detection (LID) is defined in Table Notation A of Attachment 3.20, Maximum Values for Lower Limits of Detections" - REMP

b. Change samples at least once per 7 days and complete analyses within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months after changing. Perform iWalyses at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for 7 days following each shutdown, startup or THERMAL POWIR change greater than 15%perhour of RATED THEaMAL POWER. WHEN samples collected for24.

hours are analyzed, THE the corresponding LLDs may be increased by a factor of 10. This requirement docs not apply IF (1) analysis shows that DOSEQ 1131 concentration in the RCS has not increased more lhan a factor of 3; and (2) the noble gas nonitor shows that effluent activity has not increased more than a 1'actor of 3. IF the daily sample requirement has bee entred, THEN it canbe exited early onceboth the Jadiationmonitorreading and theRCS DOSEQI131 levels have returned to within the factor of 3 of the

]?rVcfvel 'normal'.[Ref. 5.2.ly]

c. :Know the ratio of the sample flow rateto the sanmled seam flow rateforthe time period coveredby iach dose or dose rate calculation made in accordance with steps 3.2.4a, 3.2.4b, and 3.2.4c of this document.

Sampling evolutions are not an interruption of a continuous release or sampling period.

d. The principal gamma emitters for which the LLD specification applies exclusively are the following

radionuclides: Kr-87, Kr-88, Xe-133, Xo-133M, Xe-135 and Xe-138 forgaseous emissions and Mn-'i4, Fe-59, Co-58, Co-60, Zn-65, Mo-99, Cs-134, Cs-137, Ce-141 and Ce-144 forparticulate emissions. Nhis 1ist does not mean that only these nuclides are to be detected and reported. Identify and report other peas, which are measurable and identifiable, together with the above nuctides.

c. :,eleases from incinerated oil are discharged through the Auxiliary Boiter System Accoun forreleases based on pre-release grab sample data.

$ Collect samples of waste oil to be incinerated from the container in which the waste oil is stored (example: waste oil storage tanks, 55 gal. druns) prior to transfer to the Auxiliary Boier System EDamre 3;amples are representative of container contents.

g. Obtain and analyze a gas marinelli grab sample weekly for noble gases effluent quantification.

h rake tritium grab samples at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when the refueling cavity is flooded.

i Grab sanyling of the Gland Seal Exhaust pathway need not be perfomrmd if the RMS low range channel (SRA-18051205) readings are less than lE-6pCIcc. Attach thc RMS daily averages in lieu of sampling.

'Isis based on operating experience indicating no.activity is detected inthe Gland Seal Exhaustbelow this value. Compensatory sampling for out of service monitor is still required in the event 1805/2805 is inoperable.

j. Sampling and analysis shall also be performed following shutdown, startup or THERMAL POWER chnge exceeding 15% ofRATED THERMAL POWER within a onehourperiod. This noble gas sample shall be performed within fourhours of the event Evaluation of the sanmple results, based on previous samples, will be performed to determine if any further sampling is necessary.

Reference I PMP-6010-OSD-0O1 I Rei. 18 I Paee 60 of 85 1 OFF-SITE DOSE CALCULATION MANUAL Attachment 3.8 Multiple Release Point Factors for 'elefae Points Page:

L liquid Factcrs Malitor Description Monitu Numbe MP #

U I SG Blowdown 1Rl9/24. DRS 3100/3200* 0.35 U 2 SG Blowdown 2RI9/24, DRS 4100/4200* 0.35 U I &2liquidWasteDisciarge RRS-1000 0.30 Sources of radioactivity released from the Turbine Room Sump '(IS)typically criginate from the seccidary cycle which is already being mrnitcred by instrumentation that utilizes multiple release point (MRE' factors. The MRP is an administrative value that is used to assist with maintaining releases ALARA. The TRS has no actual radiaticn monitcr, but utilizes an automatic compsCitcr fr miitcring what has been releasei Ihe bath release pat, through RRS-1000, is the predorninant release path by several magnitudes. Tritium is the predominant radionuclide released from the site and the radiatim mailtars do not respcnd to this low energy beta emitter. Based ai this infonnatima and the large degree of ccaservatism bMit into the radiaticn monitcr setpoint methodology it does not appear to warrant fitrther reduction fcr the TRS release path since its source is predominantly the secondary cycle which is adequately covered by L

this factcr.

Gaseous Factcxs L

Monitor Description Mccitcr Number FlowRate (cfn) MRP #.

Unit I Unit Vent Gland Seal Vent VRS-1500 SRA-1800 186,600 1,260 0.54 0.00363 L SteamJetAirEjectcr SRA-1900 3,600 (b) 0.01 Start Up Fir Vent 1,536 0.004 L

Total 192,996 Unit2 L Unit Vent VRS-2500 143,400 0.41 L Gland Seal Vent SRA-2800 5,508 (a) 0.02 Steam Jet Air Ejector Start Up FTIVent SRA-2900 3,600 (b) 1,536 0.01 0.004 L Total . 154,044

Either R-19, 24. DRS 3(4100 cr3/4200 can be used far blowdown manitcing as the Ebrline nmitas (DRS) art repladig the Westirniouse (R)mrnitcrs.

L Norrinal Vahles a Two release points of 2,754 cfm each arm totaled for Uis value.

b Thisis the total design nm m Dfthe Start Up AirEjectors. TIis is a cms trvaive value fcrunit .

Reference PMP.6010-OSD-OO1 Rev. 18 Page 61 of 85 OFF-;SITE DOSE CALCULATION MANUAL I ,. .. .9 Attachment 3.9 lPage: Liquid Effluent Release'Systems 6

Reference PMP-6010-OSD-001 Rev. 18 Page 62 of 85 OFF-SITE DOSE CALCULATION MANUAL Attachment 3.10 I..

Plant Liquid Effluent Parameters I Page 62 lSYSTEM l COMPONENTS CAPACITY FLOW RATE . .

l. TANKS PUMPS. (EACH) l (EACH)* l I Waste Disposal System

+ Chemical Drain Tank 1 1 600 GAL.. 20 GPMl

+ Laundry & Hot Shower Tanks 2 1 600 GAL 20 GPM

+ Monitor Tanks 4 2 21,600 GAL 150 GPM

+ Waste Holdup Tanks 2 25,000 GAL

+ Waste Evaporators 3 30 GPM I

+ Waste Evaporator Condensate 2 2 6,450 GAL 150 GPM Tanks II Steam Generator Blowdown and Blowdown Treatment Systems L I

+ Start-up Flash Tank (Vented)# 1 1,800 GAL. 580 GPM

+ Normal Flash Tank (Not Vented) 1 525 GAL 100 GPM L

+ Blowdown Treatment System 1 60 GPM HI Essential Service Water System L

+ Water Pumps [ 4 10,000 GPM Containment Spray Heat 4 3,300 GPM

+

Exchanger Outlet [ ; . -

L IV Circulating Water Pumps Unit 1_-____ 3 l_ ._ l_230,000 GPM Unit 2 l_l_4 4 l 230,000GPM Nominal Values The 580 gpm value is calculated from the Estimated Steam Generator Blowdown Plow vs. DRV Valve Position letter prepared by M. J. OXeefe, dated 9127193. This is 830 gpm times the 70% that remains as L

liquid while the other 30% flashes to steam and exhausts out the flash tank vent.

Reference l PMP-6010-OSD-OO1 Rev. 18 Page 63 of _5 OFF-SITE DOSE CALCULATION MANUAL g Volumetric Detection Efficiencies for Principle Gamma l Page:

l EmittingRadionuclidesforEberlineLiquidMonitors 63 This includes the followng monitors: RRS-1000, DRS 3100, DRS 3200, DRS 4100, DRS 4200, WRA 3500, WRA 3600, WRA 4500 and WRA 4600. [Ref. s 2 lp)

NUCLIDE EMFFCBENCY (cprwtLcvcc) 1-131 3.78 E7 Cs-137 3.00O .

Cs-134 7.93 E7 Co-60 5.75 E7 CD-S8 4.58 E7

. Cr-51 3.60 E6 Wkn-54 3.30R7 Zn-65 1.58 EL Ag-l1lM 9.93 E7 Ba-133 4.85 E7

. Ba-140 1.92 E7 Cd-109. 9.58 ES CX-139 3.28 E7 Co-141 1.92E8 Ce-144 4.83 E6 Co-57 3.80 E7 CS-136 1.017ES

- Pes59 . 2.83 E7 Sb-124 5.93 E7 1-133 3.40 E7 1-134 7.23 El 1-135 3.95 E7 MO-99 8.68 E6 Na-24 4.45 E7 Nb-95 3.28 E7 Nb-97 3.50E7 Rb-89 5.00 E7 Ru-103 3.48 E7 Ru-106

1.23 E7 Sb-122 2.55 E7

. Sb-125 3.15 E7 Sn-1 13 7.33 E5 Sr-85 3.70 E7 Sr-89 288 E3

. Sr-92 3.67 E7 Tc-99M 3.60 E7 Y-88 5.25 E7 Zr-95 3.38 E7 Zr-97 3.10E7 Kr-85 1.56 ES Kr-85M 3.53 E7

. Kr-88 4.10E7 Xe-131M S.15 E5 Xe-133 7.78 E76 Xe-133M 5.75 E6 Xe-135 3.83 E7

Reference PMP-6010-OSD-001 Rev. 18 Page 64 of 85 OFF-SiTE DOSE CALCULATION MANUAL Attachment 3.12 Counting Efficiency Curves for R-19, and R-24 Pages:

64- 65 Counting Efflciency Curve for R-19 Efficieccy Factor = 4.2 B6 cpm/uCilmI 01adm tsnpkkad&f taXmk d cks at* m tft&i wi& CS.1m 0

C U3 10 9

9 8_ 8 8 8 8.

nicrocurlsWml r- r r -. r-- r r- r- r- r-A ; r---. r - -1 r - r r----. r--I r --. r r

Cz 1z~-E F- EL E -- L rz~- -~ f. F- El. F r U~r 1-Reference I PMP.6010-OSD-001 l Rev. 18 Page 65 of 85 OFF-SITE DOSE CALCULATION MANUAL Attachment 3.12 l Counting Efficiency Curves for R-19, and R-24 Il -nfrfsc I CountlngEfficlency Curve for R-24 Efficiency Factor = 7.5E6 cpmnuCinil (Bacd ca data Ulm 6uag M ai ttdia Awthn.S4) 0 0

.0.

Ul ID to V X)

'aa 0

R0R 8 8 0*

mlcrocurleWI

Reference PMP-6010-OSD4O01 l Rev. 18_ Pate 66 of 85 OFF-SITE DOSE CALCULATION MANUAL Attachment 3.13 Counting Efficiency Curve for R-20, and R-28 66 Counting EfiMclency Curve for R-0 and R.28 Efflcicncy Factor 4.3 E6 cpmhuCi/ml (basd on emphpcal dam ukm fhtf pazupwgmzal ladW wldL Co.SS) t w0E+ol

-1.00E+04 1.OE+05. -

t.00E403 9 1.00E+03 1 .oo.0 1.0OE+00 r mloroourbo/fml r r . r. r ; rn r r. r7 r- r' r r- r -- r- r- r

Reference I PMP-6010-OSD-OO1 I Rev. 18 l Page 67 of B.

OFF-SITE DOSE CALCULATION MANUAL .14 l : Gaseous Effluent Release Systems Page:

. 67

-cu.

FModhq

Reference. PMP 6010-OSD-001 Rev. 18 Page 68 of 85 OFF-SITE DOSE CALCULATION MANUAL Attachment 3.15 1 Plant Gaseous Effluent Parameters Page:

I ,

SYSTEM UNIT EXHAUST CAPACITY i L

FLOW RATE (CFM) II pi PLANT AUXILIARY BUILDING 1 186,600 max L-1 UNITVENT 2 143,400 maxc WASTE GAS DECAY TANKS (8) AND 1 125 4082 P @ @1-00 psig CHEMICAL & VOLUME CONTROL .28,741 fmax SYSTEM HOLD UP TANKS (3) @ 8, 0 level

+ AUXILIARY BUILDING l1 72,660 EXHAUST 2 59,400

+ ENG. SAFETY FEATURES 1 &2 50,000 VENT

+ FUEL HANDLING AREA VENT 1 30,000 I SYSTEM . . _._.

CONTAINMENTPURGE SYSTEM 1 &2 32,000 CONTAINMENTPRESSURE 1 &2 1,000 L

RELIEF SYSTEM INSTRUMENT SYSTEM ROOM PURGE 1& 2

.1,000 L II CONDENSER AIREJECTOR SYSTEM .

2 Release Points One for Each Unit L

NORMAL STEAM JET AIR EJECTORS STARTUPSTEAMJETAIR 1&2 1&2 230 3,600 L

EJECTORS " . .

L HI TURBINE SEALS SYSTEM 1,260 2

.2 5,508 2 Release Points for Unit 2 L

IV STARTUPPFLASHTANK VENT Il 1,536 1 1 L

to2 f l fas 1,536

+ Designates tota flow for all fans.:

L L

L

Reference PMP-6010-OSD-OO1 Rev. 18 Page 69 of Fs OFI-SITE DOSE CALCULATION MANUAL AXttahment 3.16 10YearAverage of 1995-2004 Data Pages: 69-70 X/Q GROUND AVERAGE (sec/m3 )

DIRECIION DISTANCE (METERS)-

(WM FROM) 594 2416 ' 4020 5630 7240 N 4.17E-06 4.82E-07 2.25E-07 1.33E-07 9.32E-08 NN_ 3.02E-06 3.64E-07 1.73E-07 1.04E-07 7.29E-08 NE 4.54E-06 5.31E-07 2.602-07 1.59E-07 1.13E-07

-ENE_ 7.16E-06 7.99E-07 4.04 -07 2.52E-07 1.80B-07 E 1.04E-05 1.13E-06 5.82E-07 3.66E-07 2.63E-07 ESE 1.07E-05 1.18E-06 6.04E-07 3.782-07 2.72E-07 se 1.15E-05 1.24E-06 6.36E-07 4.00E-07 2.88E-07 SSE 1.30E-05 1.42E-06 7.27E-07 4.57e-07 3.29E-07 S 1.41E-05 1.57E-06 7.92E-07 4.93E-07 3.54E-07 SS'", 7.03E-06 7.81E-07 3.90E-07 -2.41E-07 1.72E-07 SW 4.12E-06 4.73E-07. 2.28E-07 1.38E-07 9.73E208 WSV. 3.29E-06 3.65E-07 1.76E-07. 1.06E-07 7.52E-08 w __ 3.63E-06 4.1IE-07 1.962-07 1.18E-07 8.31E-08 W NI 3.02E-06 3.43E-07 1.61E-07 9.59E-08 6.71 -08

_ _ 1 3.22E-06 3.61e-07 1.71E-07 1.02E-07 7.16E-08 NW 13.84E 4.29E-07 2.02E-07 1.20E-07 8.40B-08 DMECTION DSTANCE (E (WIND 12067 24135 40225 56315 80500 N 4.64E-08 1.79E-08 8.89E-09 5.68E-09 3.56E-09 NNE_ 3.66E-08 1.43E-08 7.13E-09 4.S6E-09 2.87E-09 NE 5.75E-08 2.30E-08 1.15E!08 7.41E-09 4.72E-09 ENE_ 9.30E-08 3.80E-08 1.91E-08 1.23E-08 7.90E-09 E 1.37E-07 5.65E-08 2.85E-08 1.83E-08 1.18E-08 ESE 1.41E-07 5.81E-08 2.93E-08 1.88E-08 1.22E-08 SE 1.50E-07 6.20E-08 3.12E-08 2.01E-08 1 .302-08 SSE 1.71E-07 7.06E-08 3.56E-08 2.29E-08 1.48E-08 S . 1.84Er-07 7.49E-08 3.77E-08 2.43E-08 1.56E-08 55WA a's6208 3.59-O 1.80E-OS 1.520 7.39E-09 SW 4 93E-08 1.96E-08 9.771-09 6.27E-09 3.98E-09 WSNV 3.80E-08 1.5 1E-08 7.53E-09 4.83E-09 3.07E-09 W . 4.17E-08 1.64E-08 8.13E-09 5.20E-09 3.28E-09 WNA 3.34E-08 1.29E-08 6.4 1E-09 4.10E-09 2.57E-09 NW 3.57E-08 1.39E-08 6.89E-09 4.41E-09 2.77E-09 NNW 4.19E-08 3.35E-08 8.lOE-09 5.19E-09 3.27E-09 DMECTION - SECTOR 1 =

N =A _E =E S J W N NNE __B I ESE =F SSW. = K NE =_C l SE =G SW =L NW =Q ENE =D JSSE =H WSW =M NNW= R Warst Case Z/Q = 2.04E-05 sec/rn in Sector H 2004

Reference PMP-6010-OSD-001 Rev. 18 Page 70 of 85 OFF-SITE DOSE CALCULATION MANUAL iL-Attachment 3.16 10 Year Averake of 1995-2004 Data Pages: 69 - 70 D/Q DEPOSMION (t/m2 )

s DIERECTION DISTANCE (METERS) .

(WIND IRONO 594 12416 14020 15630 17240 N 2.37E-08 2.29E-09 1.04E-09 5.442-10 3.47B-10 NNE 9.86E-09 9.52E-10 4.32E-10 2.27E-10 1.45E-10 NB 1.29E-08 1.25E-09 5.67E-10 2.972-10 1.90E-10 ENE 1.59E-08 1.54E-09 6.97E-10 3.662-10 2.33E-10 E 1.87E-08 1.812-09 8.20E-10 4.30E-10 2.75E-10 ESE 1.85E-08 1.79B-09 8.12B-10 4.26E-10 2.722-10 SE 1.90E-o0 1.83E-09 8.30E-10 4.36E-10 2.78E-10 SSE 2.40E-08 2.32E-09 1.05E-09 .5.522-10 3.52B-10 S 3.68E-08 3.56B-09 1.61B-09 8.46E-10 5.40E-10 SSW 2.30E-0S 2.22E-09 1.01E-09 5.28E-10 3.37E-10' SW 2.22E-08 2.15B-09 9.74B-10 5.11-10 3.26E-10 WSW 2.1 E-08 2.04E-09 9.23B-10 4.84B-10 3.09B-10 W 2.00E-08 1.93E-09 8.74E-10 4.59E-10 2.93E-10 WNW 1.75E.08 1.69B-09 7.64E.10 4.01E-10 2.56E-10 NW 1.58E-OS 1.53E-09 6.94E-10 3.64E-10 2.322-10 L NNW 2.30E-08 2.22E-09 1.01p-09 5.28E-10 3.37E-10 DIERECTION (WIND FROM) 12067 f24135 DISTANCE(MEER)-BE9 40225 156315 180500 L

N NNE NE 1.45B-10 6.36E-11 8.07B-11 4.72211 1.97E-11 2.582-11 1.74E-11 7.24B-12 9.512-12 9.27E-12 3.862012 5.07E-12 4.65E-12 1.942-12 2.54B-12 L

EVE B

ESE 9.77E-11.

1.14B-10 1.13E-10 3.17E-11 3.73E-11 3.70E-11 1.17E-11 1.37E-11 1.362-11 6.23-12Z 7.34B-12 7.262-12 3.132-12 3.68B-12 3.64E-12 L

SE 1.162-10 3.78E,11 1.39E-11 7.422-12 3.72E-12 SSE S

SSW 1.47E-10 2.25E-10 1.41E-10 4.79E-11 7.34E-11 4.592-11 1.76E-11 2.70E-11 1.69E-11

.9.412-12 1.44E-11 9.01B-12 4.722-12 7.23E,12 4.52E-12 L

SW 1.36B-10 4.43B-11 1.63E-11 8.712-12 4.37E-12 WSW W

WNW 1.29B-10 1.22E-10 1.07E-10 4.20E-11 3.9E2-11 3.48E-11 1.55E-11 1.47E-11 1.28E-11 8.26B-12 7.822-12 6.84E-12 4.142-12 3.922-12 3.43E-12 L

NW 9.70E-11 3.162-11 1.16E-11 6.20E 3.11-E12 NNW 1.41E-10 4.582-11 1.692-11 9.002-12 4.52B-12 L

DIRECTION -SECTOR i N =A.. . lB =E gS = 1_W_ =N NNE =B jESE = F SSW K. WNW p f I NE = C SE = G SW. =L MNVW - Q.

ENE = D SSE = H WSW =M NNW = R Worst CaseD/Q =4.46E2-0 l/rn 2in ScctocA2001

Reference l PMP-6010-OSD-O01 Rev. 18 -Page 71 of 35 OFF-SITE DOSE CALCULATION MANUAL Attachment 317 Annual Evaluation of 4/Q and D/QValues For Page:

_IAll Sectors . 71l

.J. .1. Performedor received annual update of xlQ andD/Q values. Provide a description of what has been received.

I Signature Date Environmental Department (print name, title)

2. Worst z/Qand D/Q value and sector determined PMP-601-OSD-O001 has been up&led, ifnecessary. Provide an evaluation.

Signature Date Environmental Depmltment (print name, title)

3. Review nuclide mix for gaseous and liquid release paths to determine if the dose conversion factor of total body is still applicable. Provide an evaluation Signature Date Environmental Department (print name, title)

4. Approved and verified by:

I Signature Date Environmental Dep artinent (print name, title)

5. Copy to NS&A for infonnation.

Signature Date

Environmental Departnent (print name, tile)

Reference PMP-6010-OSD-001 Rev. 18 Page 72 of 85 OFF-SITE DOSE CALCULATION MANUAL h.

Attachment 3.18 leDoseFactors * ) Pa72-7 I.

DOSE FACTORS FOR NOBLE GASES AND DAUGHTERS*

TOTAL BODY SKIN DOSE GAMMA AIR BETA AIR DOSE FACTOR FACTOR DOSE FACTOR DOSE FACTOR K1 (DFBD Li (DFSM Ml (DFV N1 (DF%

inremm 3 (rmnem 3 (mrad 3 (mrad 3 RADIONUCLIDE per Aa yr) per pCi yr) per pCi yi) per pCi yr)

Kr-83m 7.56B-02 . 1.93E+01 2.88E+02 Kr-85m 1.17E+03 1.46E+03 1.23E+03 1.97E+03 I Kr-85 1.61E+01 1.34E+03 1.72E+01 1.95E+03 Kr-87 5.92E+03. 9.73E+03 6.17E+03 1.03E+04 Kr-88 1.4713+04 2.3713+03 1.5213+04 2.93E+03 Kr-89 1.66E+04 11.OIE+04 1.73E+04 1.06E+04 Kr-90 1.5613+04 7.29E403 1.63E+04 7.83E+03 L Xe-131m 9.15E+01 4.76E+02 1.56E+02 1.11E+03 Xe-133m Xe-133 2.51E+02 2.94E+02 9.94E+02 3.06&+02 3.2713+02 3.5313+02 1.48E+03 1.05E+03 L

Xe-135m 3.12E+03 7.11E+02 3.3613+03 7.39E,+02 Xe-135 1.81B+03 1.86B+03 1.92E+03 2.46E+03 L Xe-137 1.4213+03 1.22E+04 .1.51E+03 1.27E+04 Xe-138 Ar-41 8.83E+03 8.84E+03 4.13E+03 2.69E+03 9.21E+03 9.301+03 4.75E+03 3.28E+03 L

L L

The fisted dose factrs e for radionucdides dtat maybe detected in gaseous cfhnts. from Rcg. Guide 1.109, Tablc B-1.

L L

Reference I nIP-6010-OSD-001 Rev. 18 Page 73 of5 OFF-SITE DOSE CALCULATION MANUAL

... ., _ags Attachient 3.18 l Dose Factors -, Pages:

72 -73 DOSEYACTORS FOR RADIOIODINES AND RADIOACTIVE PARTICULATE, IN GASEOUS EFFLUENTS FOR CHILD* Rd. s2ziccRnaff pi. Pi' INHALATION FOOD & GROUND PATHWAY PATHWXY (mrem nP (mrem rn2 sec RADIONUCLIDE per jiCI yr) 'per 1Ci yr)

H-3 1.12E+03 1.57E+03 P-32 2.60E+06 7.76E+10 Cr-SI 1.70E404 1.20E+07 Mn-54 1.58E+06 1.12E+09 Fe-59 1.27E+06 5.92E+08 Co-58 1.11E+06 5.97E+08 Co-60 7.07E+06 4.63E+09 Zn-65 9.95E+05 1.17E+10 Rb-86 1.98E+05 8.78E+09 Sr-89 2.16E+06 6.62E+09 Sr-90 1.01E+08 1.12E+ll Y-91 2.63E+06 6.72E+06 Zr-95 2.23E+06 3.448+08.

Nb-95 6.14E+05 4.24E+08 Ru-103 6.62E+05 1.55E+08 Ru-106 1.43E+07 3.01E+08 Ag-1orm 5.48E+06 1.99E+10 I-131 1.62E+07 4.34E+11 1-132 1.94E-05 1.78E+06 1-133 3.85E+06 3.95E+09 1-135 7.92E+05 1.22E+07 Cs-134 1.01E+06 4.008+10 Cs-136 1.71E+05 3.00E+09 Cs-137 9.07E+05 3.34E+10 Ba-140 1.74E+06 1.46E+08 Ce-141 5.44E+05 3.3 1E+07 Cc-144 1.20E+07 1.91E+08

As Sr9,RD-IO6 and l-131 nralyscs areperfcnmed, THEN useP pvcn inP-32 fcrnonlisrtdeAidaozmcldeL wbeurnitu forbohB3 facton arethe sme eranmperuCG yr

b..

Reference PMP-6010-OSD-001 Rev. 18 Page 74 of 85 OFF-SITE DOSE CALCULATION MANUAL a.-

1 RadiologicalEnvironmentalMonitoringProgram l Pages:

Sample Stations, Sample Types, Sample Frequencies 74 - 77 Ref.5S l..5.2.lx.5.2.1t_

SAMLE DESCRIPTIONY SAPLE SALE ANALYSIS ANALYSIS L STATION ' FLOCATION TYPE FREQUENCY TYPE FREQUENCY ON-SrM AIRBORNE AND DIRECT RADIATION (MID)STATIONS ONS-1 (Y-1) 1945 ft @ lrfromPlant Axis AirborneParticulate Weekly GrossBeta Weekly

.0Gam Isotopic dart Comr.

Airborne Radiciodine I-131 Weekly TLD Qcdy Direct Radiation Caiedy ONS2 (r-2) 2339St @ 48°fromPlant Axis Aizborne Paticulate Weekly GrossBeta Weekly Gamma Isotopic uart.Comp.

Airborne Radioiodinc 1-131 Weekly

_D Quartey Dired Radiation Qady ONS-3 (T-3) 2407 ft @90 from Plant Axis Airbcene Particulate Weekly Goss BeTa Weekly

.L AirboneRadioiodine

-Quarerly Gamma lsotopic 1-131 Dirett Radiaticu Quaa. Comp.

Weekly QaCdy L

ONS-4 (1t4) 13521n. i 11 from Plant Axi. Airborne Particulate Weekly GrossBeta Weekly

.,°Gamma lsotopic Quart Comp.

Airborne Radiojodine 1-131 Weekly TLD tedy Dirt Radiation Quay ONS.5 (T-5) 1395 ft ( I9°from Plant Axis Airborne Particulate Weekly °rossBeta Weekly Gamma Isotopic Quat Cap.

Airbome Radioiodine TLD QuL 1.131 Direct RAdiation Weekly Quartey L

0 ONS-6 (T-6) 19171 CO210 fromPlan Axis AirbonieParticulate Weekly GrossBeta Wekly L

Gamma Istoic Quart Comp.

Airborne Radioodine _ 1-131 Weekly TLD Oc Dired Radiaticn Ouar dv T.7 2103 ft( 36' fromPlant Axis TLD Qurtesrly Direct Radiation Quartedy T49 2208 It (P B fromPlant Axis TILD) Quartery DirectRadiation Quartedy T-9 T 10 T-1 1 1368 it ( 149°frmPlani Axis 1390ft (P l27°fmPlant Axis 1969 ft (P 11° frmsPlant Axis TLI)

TID TLD Qartry Quarterly Quartery DirectR~diation DirectRadiation Direct Radiation Quartey Quarterly Quartedy L

T-12 2292ft ( 63 fromPlant Axis ILD Quarterly DirectRadiation Quarterly CONIROL AIRBORNE AND DIRECr RADIATiON (MLD) STAIONS L NB? 15.6 miles SSW AirborneParticulate Weekly GrossBeta Weekly SBN NewBuffalo, ?

26.2 miles SE AirborneRadioiodine TLD Airborne Particulate Qartedy Weekly Gamma Isotopic 1.13 1 Dit Radition (rous Beta Quaa Comp.

Weekly Quaresy Weldy L

Soath Bend, IN Gamma Isotopic Quart Comp.

DOW 243 milesENB AirborneRadioiodine TLD AirboneParticulate Quartedy Weekly 1-131 Direct Radiation GrossBeta Weely Weekly L

Dowagiac. Ml ' Ga-ma Isotopic Quart Com AirborneRadiolodine 1-131 Weekly TLD Outd Direct Radiatico Qureay COL 18.9 miles NNE Airborne Particulate Weekly Gross Beta Weekly L Cdlome, _ Gamma Isotopic Quart Comp.

AirborneRadiolodine 1-131 Weekly TLD Qu Direct Radiation _Qoedy L

Reference PMP-6010.OSDO001 Rev. 18 Page 75 er 85 OFF-SITE DOSE CALCULATION MANUAL t

A- chmnt e 3 Radiological Environmental Monitoring Program.

Sa3mple Stations, Sample Types, Sample Frequencies

.:: 74-'77 Pages:

S 74 SAMPLE DESCRIPTION' SAMPLE SAMLE ANALYSIS ANALYSIS STATION LOCATION TYPE FREQUENCY TYPE FRIEQUNC OFF-PY DIRECT RADIATION MD) STATIONS OFr-1 4.5 miles N Pole #B294-44 TLD Quaftedy Dinrt Radiation Qartery OFT-2 3.6 miles, NE, Stevensville -1D Quadtedy Direct Radiation Quartedy

_ __ Su sl~ation _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

OFT-3 5.1 miles NE. Palo #B296-13 Quartey Direct Radiation OFI-4 4.1 miles, E.Pole #B350-72 TLD Quarterly DirctRadiation Quarterly OFI.5 4.2 miles ESE. Pole #B3387-32 LD 2a y Direct RAdiation Quarterly OFT-6 4.9 miles SE. Pole #B426-1 WD Quartedy Direct Radiation Qarterly OFr-7 2.5 miles S. Bridgman Substation TLD Quartely Direct Radiation Qartedy OFT-8 4.0 miles S. Pole #B424.20 TLD Oaatedv Direct Radiation Qarel OFI.9 4.4 miles ESE. Pole NB369-214 TLD Quartedy Direct Radiation OFT-1D 3.8 miles S. Pole #B422-99 TLD O-tld Direct Radiation OFT-I 3.8 miles S. Pole B423-12 1LD QuarWdy DirrcRadilion QoadDdy GROUNDWATER (WELL WATEIR) SAMPLE STATIONS W.1 1969 A O110 fromPant Axis Grmndwater O~Qartedy Gamma Isotopic Quarterly Tritium uarterl WV2 2302h 0 63fiaomPlant Axis Grunadwater Quarterly Gamma Isotopic . Quartedy TGiaium Quarterly W-3 3Z79atoI107° fmm Phnt Axis Givndwater Quarterly Griimm Isciopic Quaztdy Tritium Quartedv W-4 41 B S11 301 fxem Plant Axis Orunndwater Quarterly Oamma Isotopic Quarterly T__tim _ _ _Quarterly W-5 44 A 0 290 romPlant Axis Oroundwater Quarterly Gamma Isotopic Quarterly Tritium Quary W7 4241 273 from Plant Axis tndwAts O, Quarterly Gamma Isotopic Quaatedy

____ Tritium Qu y W-7 12895 An0 169fomPlazt Axis orondwater Quarterly Gamma Isotopic Quarterly Tritium Q y W-8 1274 A10540 from Plant Axis Ornxdwater Quarterly Gamma Isotopic Quartexly W-9144 21frm~lutAxi A GI~dwte Qurtely Gamma Isotopic Q y Tritium___ uartmu W-10 4216R O 2291om PlantAxis Grtndwater Quarterly Oamma Isotopic Quartedy

___ Tritium Quarkts W-11 3206 A ID 253°fom at Axis Onundwater Quarerly Gamma Isotopic Quarterly

_,_Tritium Qcay_

W.12 2631 A 0 1620from Plait Axis G.oundwater Quarterly Gamma Isotopic Quarterly

, Tritium Qua y W-13 2152Q t0 182° from PlantAyas Grotadwater Quarterly Gamma Isotopic Quarterly Tritium Q Y W.14 17B0ID@ 164tfromPlaztAxis Ground'Wer Quarterly Gamma Isotopic Quarterly

_ Tritium..

DRDKING WATER_

Sri St Joseph Public ntake Sta. Drinking water Once pcrcalendar OrossBeta 14 day Cor t Day Gamma Isotopic - 14 day Coim 1131 14 dayCo a.

Tridum Quat Comp LTW LakcTwp. PublicIntake Sta. DrinkJing water Onceper calendar GrossBcta 14 dayCora 0.6mi. S Day GammrIsotopic 14 dayCO

.-131 14 day Coz

. ntrim I Quart. Conm

Reference PMP-6010-OSD-001 Rev. 18 -r -Pa e 76 of 85 OFF-SITE DOSE CALCULATION MANUAL A .c' t3 Radiological Environmental Monitoring Program ., Pages: E t mewt 3.9 Sarrple Stations, Sample Types, Sample Frequencies 74 - 77 SAMPLE DESCRIPTIONI l SAMPLE i SAMPLE l ANALYSIS l ANALYSIS STATION LOCATION TYPE FREQUENCY TYPE FREQUENCY SURFACE WATER SWL-2 Plant Site Boundwy - South Sudace Water Once per calendar Gamm Itotopic Month. Comp.

-SOOA soutl of Plant Centedine _ DAy TAtium I QultComp SWL-3 Plant Site Bounday - North Surface Water Once per calendar Gamma Isotopic I Month. Comp.

-5SD0 ftnortn ofPant Centedine . Day Tritinm Quart Comp.

SEDIMENT *_ _

SL-2 Plant Site Bonday - South Sediment Semi-AnnL Gamma Isotopic Semi-Animal

-50D0fti.th ofPlant Centedine . . . _

SL-3 Plant Site Boundawy - North Sediment S rni-AnL Gam Isotopc Semi-Anmanl

-500 ft north of liant Centcdine . .

SL-4 SL5 Plant Site Boundary

South South storm drain culvert to lake Phnt Site Boundary - North Sediment Sediment Quartedy Quardey Gamma Isotopic Gamma Isotopic Quartedy Quartedy L

North-stam drain culvert to lake SL4 & S amrdata collection points only not actual REMP samples L

GROUNDWATER (RADIOACTIVE MATERIAL STORAG FAC T MAUSOLEUM]) SAMPLE STATIONS

.1 SO-I 0.8 nm. 95'from Plant Axis Groundwater Quartedy Gross Alpha GrossBea am-m Isotopic

Quaredy y

y L

SO-2 O.'7 mi. 92 from Piant Axis Grocndwater Quartedy Gross Alpha Quarterly S04 0.7 mi. 0 93-from lant Aids Grounbdater Quartedy GrossBeta 0Gam Isotoic Gross Alpha G.ossfi Beta QCy O y Quarterly Qudy L

S0-5

.Gamma 0.7 mi. (O92from Plnt Axis Groundwater Quarterly 0_ Isotopic Gross Alpha GrossBets Isotopic Qarterl Quarterly Qcutedy

.L SO-1, 2,4 and 5 au data cdlection point only not actual REMP samples INGESTION -MIIX Indicator Pam_

- L NMilk l Once every . 1-131 I ptrle W0 Millc 15 days 0 Once every 15 days * -

Gamma Isotopic 1-131-oam= Isotopic persilc perde Ipersmpe L

Mlki. Onceevery

1_131l

_r_-MP__e-INGESTION-MILK Bacigro lPatmnns I

15 days Gamma Isotopic perat L

IrMik lOnce every l _ _1-131 pers__e l_ l _ _days l_l__ Gamma isotopic I pertmple l per 0=306 Ia iI *i Once every IS days I 1-131 Gammalsotopc I persac. pl.

L L

L

Reference PMP-6010-OSD-001 Rev. 18 rPage 77 cf 85 OFF-SITE DOSE CALCULATION MANUAL Atach t 3.i9 l Radiological Environmental Monitoring Program l Pages: -&

- Sample Stations, Sample Types, Sample Frequencies 74- 77 SAMPLE l DESCRIPUONI l. SAMPLE SAMPLE. l ANALYSIS AMAL'SIS STATION LOCATION YE ' f UENCY MPE FQUNCY INGHM1ON-FISH _

ONS-N 0.3 mile N. Lake Michigan M5- -i edible pion T 2ye I Gamma isotopic M ONS-S OFS.

OPS;_

j 3.55.0 0.4 mile S,Lakc Micbig mile N. Lake Michigan mile S. le Michigan Fish-edible ponion fish-edible portion Fish-edible potion 2maear 2Jyeer Cu amma. Isotomic m Isotopic Omma Isotopic pci 5sample per sample pers-IC INOGISlON - FOOD PRODUCTS On Si EC ONS-. Neareds =pl to Plant ia the Grapes At time of Gamma Isotopic At time of highest DIQ land sector harrest Invest containing media.

ONS-V Broadleaf At time of Onm Isotopic At time of vegetation havest harvest Offste OS-i In a land sectorcontaining O.apes At dme of Omam Isotopic At lime of grapes approximately 20 dilei hafvest Harvest fromthe pant, in one of the less prcrvlent DJQ land sectors OP{Broadleaf 1 Altdmc of Gam Isotopic m

At timc Of

_..vegetation harrest harvest INGETION -BROADLEAF INLEUF MIK 3 indi cter samples of broad leaf vegetation Broadleaf Monthly Gamma Isotopic Monthly collected at differentlocations, within eight vegctation when available 1131 when aval able miles of the plant in thc highest annual avcrCeDAQland sectr.

1 bac:ground sample of similr vegetation Broadleaf Monthly Gamma Isotopic Monthly zow 15-25 niles distant in one of vegetation when available I131 when availible the less prvaleni wind directions.

CoDflt composite samples ofDnkg and Sufacewatoratleast dily. Analyzeparticulate sapEltenforg bs Jactivity24 ormn ehours fonlosg rSaitnoval. s will allow forrado and tbmn daugbter decay. If giss bets actvity in alr or wiat resatrt han 10 times the yealy mean of contml samples foranymedium. perform g Isotopic analysis onthe Idividual samples If atl ast tbreeindicaorilkamples and onebackground milk sample cannotbe obtand tcreindieatorboadleaf samples willbecolted at differmt locations within eight miles of thepla. in the land sectorwith thehighest DI (ers to the highcst annal av geDQ. Also one backgeusndbzad leaf samp will be collected I5 to2S mies from tbe p]ant in one of the less prealent D/Q land setors The three milkindicatorand twobackrund faems wilbe determined bythe Annual Land Use Census aMd thoseth are willingtopaicipate.

IFit is deterined itht the milk aimals arefd stord fed, THEN monhly smpling is appropaltefor that timezpeiod.

Reference lPMP-6010.SD-001 . l Rev. IS Page 78 of 85 OFF-SITE DOSE CALCULATION MANUAL Attachrient 3.20 Maximum Values for Lower Limits of DetectionsAB - REW -9 L

rRcf.5.2.1v)

Radionuclides Food Product Water Milk Air Filter Fish Sediment pCilkg, wet pCi/I pCi/I pCi/m3 pCilkg, wet pCi/kg, dry Gross Beta 4 0.01 H-3 .2000 Ba-140 ._ ._._ 60 60.

La-140 15 15 Cs-134 60 15 15 0.06 130 150 Cs-137 60 18 18 0.06 150 180 L

Zr-95 30 _ __ __ __ _ __ _ _ _ _

Nb-95 15 ___

Mn-54 15 130 L Fe-59 30 260 Zn-65 . 30 260 Co-58_. 15 130 Co-60 1-131 60 .

. 15 1 1 0.07 130 _

L his Data is diecftlyim plant-speificT=diica Spcci&ficm L

UD for driking w . L L

L L

Reference l PMP-6010-OSD-001 Rev. 18 l Page 79 of 85 OFF-SITE DOSE CALCULATION MANUAL Attachment 3.20 l Max'ii~uiif Values for Lower Limits of Detections& - REMP l P7-(, .

'78- 79 NOTES A. lhe Lower Limit of Detection lD)is defined asthe,smallest wnraitraticn of radioactive xaterial in a sampk that will it iteteded with 9596 probabllity and 5% probability of filsly condading that a blank obsrvation represents B R'al"sipd 1Tor a partala neasuirnemnt system (wh may include sadiodiemical separation), the LLD is given by the equation:

.U =

LL= 4.66a*S E*V* 2 .2 2 *Y*e(OAI)

Where LIM is dtc a onio lower limit of detection as defined above (as pCi per unit mass or volume) Peforn analysis in.such a.mame that the stated LLDs will bc adhievd ndr rouie cnditions Ocsionallybackgoimd flduatis, unavoidably small sample sizes, th prsence of intrfering radionuclides, or other uncontrollabe droates may render these ILDs umadiievable. lt should be furth clified that the LLDreprlts the capability of a mea ment system and not as an. aflter the fad Ei far a pariadrmeasuremnL S is the standard deviation of the badcgrmd couatingrate or of the cunting rate of ablansaple as appropriate (as counts per minute).

B is the counting effidency of the detection equi t as counts per transformation (tiht is, disintegratiom)

V is th sample size in appropriate mass cr volumne its 2.22 is te convesionfacor frnpicoaes (pCi) to tansfonmtions (disintegiratins) per minute Y is thefiniaairnadliochemical yield as appropriate X is fth radioactive decay constant for the particular radionuclide At is the apsed tie betwen the nidpoint of sample collectdk (or end of sample collection paeid) and tine of counting&

B. Identify and report offier peaks whid are measrablc and identifiable, together with die radionudides listed in Attachment

.;.20, Maimum Values for Lower Limts of DtcctdonsA.B - REMP.

a A 2.71 viln maybe added to th euaion to provide cotedionfor deviaionsintePoissm distributionatlow countgrats, ihdtis, 2.71 +4.66x S.

Reference PMP-6010-OSD-O001 Rev. 18 Page 80 of 85 OFF-SITE DOSE CALCULATION MANUAL

-Attachment3.21 Reporting Levels for Radioactivity Concentrations Page:

in Environmental Samples .0 l

Radionucides. Food Product Water Milk. Air Filter Fish pCi/kg, wet' pCi/l pCi/I pCi/m3 pCi/kg, wet H-3 20000 Ba-140 200. 300 X La-140 . 200 300 Cs-134 1000 30 60 10 1000 Cs-137 2000 50 70 20 2000 Zr-95 400 . . L Nb-95 Mn-54 .

400 1000 30000 L Fe-59 . 400 10000 Zn-65 Co-58

300 1000 20000 30000 L

Co-60 1-131 100 300 2 30.90 .-

10000 L

L L

L

Reference l - PMP-6010-OSDO001 F Rev. 18 Page 81 of 35 OFF-SITE DOSE CALCULATION MANUAL .22 On-Site1 ring Location -REMP Page:

TDT TLD T*5 ONSSouth ONS-Norlh Surface Water ArONS8 Ar ONS-S SWL'2 W3tsr iWelW 4 won Sedimant SL-2 / /

LEGEND ONS ONS-6: Air Sampling Station T-1-T-12: TLDSanpfingStation W-1 -W-14: REIAP GroundwaterWeas SWLZ2.3: Surface Water Sampling Stations

SL-2 SL-3: Sediment Sampling Stations ONS-N & S: Fish sampling locations SG-I. 2.4 and 5: Non REMP Information WeRs

Reference l PMP-6010 OSDO001, Rev. 18 Page 82 of 85 OFF-SITE DOSE CALCULATION MANUAL L Attacbrbent 3.23 Off-Site Mofiitoni4 Locations; REMI? P j L I

L L

L-

Reference PMP-6010-OSD-001 l Rev. 18 Page 83 of 85 OFF-SITE DOSE CALCULATION MANUAL Safety EvlainBy The.,Office OfNulear Pages:

Attachment 3.24 j- Reactor Regulation 83ages: _

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO DISPOSAL OF SLIGHTLY CONTAMINATED SLUDGE INDIANA MICHIGAN POWER COMPANY DONALD C. COOK NUCLEAR PLANT, UNIT NOS. 1 AND 2 DOCKET NOS. 50-315 AND 50-316 [Ref. 5.2.1r]

C(is is a 10 CFR 50.75 (g) item)

1. INTRODUCnION

.By Untts dated October 9, 1991, October 23, 1991. Sept=rb 3, 1993, and September 29. 1993, Indiana Michipn Power Company (l&M) requested approval pursuant to 10 CFR 20.2002 Ibr Oe on-site disposal of licensed material not prcvioosly mi zonsideed the Dmald C. Cook Nucler Plant Fal Envionmmtal Statement dated August 1973. Speciiafly. this rquest

ddresses actions taken in 1982 in which approximately 942 cubic meters of si ly contaminated sludge were r=moved from he turbine room sump absrption pond and pumped to the upper parking lot located widlin the exclusion area of die Dorald C.

ook Nuclear Plant The acntaminated sludge was spread over an area of pproxixnately 4.7 acs The sludge contained a oalradfiommlideinvant of 8.89 miffieruies (mCi) of Ccsium-137, CUsiu-136, Cesiun-134, Cobalt-60 and Iodine-13 1.

a its submit the liccseC addressed specific infration requested in aeorrdance with 10 CFR2.2002(a) provided a detailed desciption of the licensed material. ihorwghly analyzed and evaluatd information pertinent to the inpacts im th imvircuunrnt of the proposed disposal of licnsed mateial, and commitled to follow specific procedures to minimize the risk of imexpeetcd exposures.

2. ])ESCRwIPON OFWASTE

'[Me turbine moom sump absoiption pond is a collection place for water released fumni thc plant's trbine room sump. The amntaminaion was caused by a ptniny-to-seconday steam generator leak that entered the pond from the turbine building canp, a recognized relase pathway. Sladg cmistingmainly of leaves and roots mixed with send, built up in the pondL As a 3 sul the licensec dredged the pond in 1982. The radioactive sludge removed by the dredging adivities was pmnpcl to a watainmenaresa located withie e exclusion rea. IThe total volume of 942 cubic meters of the radioactive sludge kint was tiredged from the bottom of the turbine room absorption pond was subsequently spread and made into a gaveled road oervthe iffcrpakinglot area ofapproxirnately4.7 aces.

'he pzindpal adianim ides identifedin the dredged matai re fisted below.

TABLE 1 1 NUCLIDE ACTIVITY (mCi) ACTIVITY (mCi)

(half-life) 1982 1991

' 6Cs (13.2 d) 3 0.03 NA*

34Cs (2.1 y) 2.34 0.18 3 7 CS (30.2 y) 5.59 4.57 6 0Co (5.6 yj. 0.90 0.27 1311 (8.04 d) 0.03 NA*

TOTAL: 8.89 5.02

NA: not applicable due to deay

Reference l PMP-6010-OSD-001. Rev. 18 Page 84 of 85 OFF-SITE DOSE CALCULATION MANUAL Attachment 3.24 l SafetyEvaluationBy The Office Of.Nuclear Pages:

A 3 ReactorRegulation .83 - 85

3. RADIOLOGICAL IMPACIS Th; licnsee in 1982 evaluated the foblowing potental exposre pathways to members of the general publii from the radicriiddes in the sludge:

(I) external exposure caused by goudshine from the disposal site; (2) internal exposure caused by inialation ofre suspended radionuclide;

-AND-(3) internal exposure from ingesting ground water.

Tbe staff has reviewed te licensee's calculational methds and assumptions and finds dtat they are consistent with NUREG-l101, "OnsitC Disposal of Radioactive Wste," Volumes 1 and 2, Navembe 1986 and February 1987, respedively.

The staff inds te assessment methoddlogy acceptabe Table 2 lis the doses calculated by the license fo dre maximally exposed menber ofthe public basedt a total activity of 8.89mCi disposed in tat year.

TABLE 2 LI Pathway Whole Body:Dose Received by

.Maximally Exposed bnutvidual (mremfyear)

L Groundshine Inhalation 0.94 0.94 L Groundwater Ingestion 0.73 Total 2.61 L On July S. 1991, die licesee re-sarnpled th onsite disposal am to assure dWal no sigaificant impacts nd adverse effects bad occuLrr A atgproced basedon df ppm atenv nralowlevel doses was used by the lica ; howevez, no activity was detected during the re-samplinig. ITis is consistent with the oiginal adivity of the mroatial and the decay time The 1991 re-sampling process usedby th likunsee camfirms that the cnvirtnenta impact of the 1982 disposal was very smalL L

Ihe saff finds the licensee'smdehodology acceptable

4. ENVIRONM1AL INDING AND CONCLUSION Ihe staff has evaluated the enviranmental impac of he proposal to leave in place approximately 942 cubic meters of sligly L

contamanated sludge undereath Om upper parking lot on the Donald C. Cook Nuclear Plant site In 1982, the lioesee evaluated te potential exposure to mermers of the general public from the radionucides in the sludge and calculated the potential dose to the maximally exposed mcmbcr of the public. based on a total activity of E.89 mC disposed in that year, tobe 2.61 mrcmty. The staffuhas reviewed die licansee's cilaclatonal methods and assun ptions and fomud that they L

are consistent with NUREG-1101, Orsice Disposal of Radioactive Wastk Volumes 1 and 2. November 1986 and Febtary 1987, respectively. Mhe staff finds dhe assessment metbodology acceptable. For mparisn the radiation from the naturally occauingradionuides inmsoils and roCks plus cosmicradiation gives a pctso in mMidrg a viile-body dose rate of abowt 89 mren per yen outdoors. Subsequent licensee sampling in 1991 identified no detectable activity. The staff evaluated tc L

licensee's sampling and analysis methodology and finds it acceptable. he resus, of the 1991 resampling by the licsee, confirm that the eavirsamien impact of the. 1982 disposal'was very small.

L Based on tme above the staff finds that the potential envircra ental impacts of leaving the contaminazed sludge in place are insignificant. With regard to II= nona-diological inmacts, the staffbas detemined that leaving dhe sol in place rsents hie least impact to ae envirmunent.

L

Reference PMP-6010-OSD-O001 Rev. 18 IPae 85 oE 85 OFF-SITE DOSE CALCULATION MANUAL ttachment 3.24 l Safety Evaluation By The Office Of Nuclear: Nl Pages:

. *Reactor Regulation 83 - 85 S. CONCLUSION.

Based on the staffs review of the licensees discussion, the staff Snds the icensec's proposal to retain the material in its lesent location as documented in this Safety Evaluation acmptable. Also, this Safity Evaluation shall be pesmanently incorpoLted as an appendix to the licensees O ite Dose Calcuation Manual (ODCM), and any future modifications shall bereported toNRC maccrdance with the applicable ODCM change protoooL I&M letter from R . litzpatrick t6 the NRC Document Control Desk, September 29,1993 Tltcfor, .thc licersees proposal to consider the slightly c xntaminated sludge disposed by retentimn in place in the manner descibed in the Donald C. Cook Nuclear Plant submittals date October 9. 1991, October 23, 1991, Septcember 3, 1993, and Septeimber 29, 1993, is acceptable.

'Ih guidelines used by the NRC stafffor onsit disposal of licensed material and the staffs evaluation of how each guideline

'has been satisfied are given in Table 3.

?rsuant to 10 CFR 51.32, the Conunissian has determined that granting of this approval will have no significant inpact on the envirnent (October 31, 1994,59 FR 54477) -

jxincipal ContibutcE 3.Minns Dale: November 10 1994 I TABLE 3 . I 20.2002 GUDELINE FOR ONSM . STAFFSEVALUATION DISPOSAL

1. 1he xadicmive matoial should be disposed of in swh a 1. Due to the nature of the disposed material recycling tothe

an~nr tbatitis m ely that the material would be genemal public is not censidered likely.

iccydled.

2. Doses to tih totalbody and anybody crgan of i 2. Mis guideliae was addressed in Table 2. Alhough the naximally exposed indiiuals (a member of the general 2.61 mmutyr is greater than staffs guideine, the stafff nds it mbfic or a non-occupationally exposed worker) from the acceptable due to 9 yrs decay follow analysis and the Irobabie pathways of exposure to the disposed material expected lack of acdivity dctcdc in the 1991 survey.

should be less dtan 1mrrn/ymar. .

3. )oses to the totalbodyand anybody organ of an 3. Becuse lematcrial iwillberland-spad the staff amsidcas iansdvlv nt inrde from dprobable pathways of th maxinially exposed individual snmaio to also addres the exposure should be less than Snrem'year. intruder scenanio.

4. I)oses to the total body and anybody organ of an 4. Evenifrecycling wre to occur aferrlasc fromcregalaioy iadividual from assuned recycing of the disposed control, the dose to a maximally exposed member of the irsaterl at t tim the sposal site is released from public is not expected to exceed 1mrcm/year, based on regulatory control fion all ikdy pathways of exposure exposure scenarios considered in dis analysis.

should be less than 1 Men.

2 PLP. Branagan, Jr. and P.J. Congel, 'Visposal of Contaminated Radioactive Wastes from Nucar Power Plants," prsented at tfe Halt Thysics Societ s Mid-Year Symposium. on He1alth Physics Consideration in DecnaninationlDecsom issiwming.

lKnoxville, Tennessee Fruaryb 19S6, (CONF-860203)

REVIEW AND APPROVAL TRACING FORM Section I - ProcedureInformation:

Number: PMP-601 0-OSD-001 Rev. 19 Tile: OFF-SITE DOSE CALCULATION MANUAL Section 2-Alteration Cafgory:

'I Minor Editorial Correction E Cancellatibn CI Major Editorial Correction (Full Review) E Superseded by Qist superseding procedures)

El Minor Revision El Major Revision (Full Review) El New Procedure (Full Review)

Scctionu3-Tmp roced re/Revision: ------

El N/A E Temporary Procedure 0 Temporary Revision ARNo.:

Ei-piration Date / Ending Activity. _ C\ P_

'St river Co4 ur-oAssaocsiesmeli: -ted- -- -

Ciange Dver/CDITracldngNo(s). FCN-12-585-RO-04 0 N/A Section 5-Reviews:

Department ° e (Refer to Figure 6, Determination of Required Reviews)

R. Hersbberger 5 0D E D. Rupert ° 0 0 0 J. Hamer E 0 0 0

___ 0 0 0 Section 6 - Technical Review:

Updated Revision Summary and Implementation Plan (if applicable) attached? 0 Yes Implementation Plan developed? If yes, ARNo.: 5Zli('7 1 7 . Yes El N'A ArT there implementation actions to be completed prior to the effective date? El Yes 9l ND 10 CFR 50.59 Requirements complete? Tracking N o .: Yes 0N'A L .X) 0) Technical Reviewer - IL b / A/.. Date: 69-27-2 r Secrton 7-!Rea&yfor Apjrval: ___

0- Acministrativc Hold Status: El Released El Reissued 0 N/A CR No.: _

-9 Wliter: bJ iFoeSTit I .- - Date: Ir5 -

j I El Ops Manager Concurrence: N/A Date:

Sc tdon 8-ApprOvals A..... .____ .... _ _ ------

PCRC Review Required: 0 Y7 a No Mtg Approval Authority Review/Approval: - '1' Date: & O+Za S 12O 0 U Effective Date: _ _ __ ____ -

C ci I SeCtion 9 - FolowU p Actions:

Commitment Database.updaterequested in accordance with PMP-2350-CMS-001? 0 Yes 0 NM/A NI)Mnotificd ofnewrecords or changes to records that could affect record retention? E Yes . NA i a, Office Information ForForm Tracking Only -Not Partof Fori' This form is dcrived from the information in PMP-2010-PRC-002, q 2:

Procedure Alteration, Review, and Approval, Rev. 17, Data Q Sheet- , Review and Approval Tracking Form. Page _ of _:-.

REVISION

SUMMARY

Number: PMP-6010-OSD-001 Revision: 19

Title:

OFF-SITEFDOSE rATLCTLATION MANTIAT.

Alteration Justification Deleted references to Current Technical Improved Technical Specification became Specification items and removed heading for effective On 9125/05. This meets the criteria Improved Technical Specification. Marginal for Editorial Correction, Letters j and n.

markings were not utilized for these corrections.

Section 2, added definitions for Member of These definitions were in the Current Public, Purge, Source check and Venting. Technical Specifications, but are not in the Improved Technical Specifications. Change.

3.2.3.c.1, 3.2.3.d.1, 3.2.4.b.1. 3.2.4.c.1, and Recommendation associated with industry 3.2.4.d.I dose limits were all converted to per expert per 10 CFR 50, Appendix I, and unit. documented in CR 05217064. Change.

3.2.3.c.2, 3.2.4.b.2 and 3.2.4.c.2 clarified that Recommendation associated with industry the ten times only applied to concentration, expert and documented in CR 05217064. L not dose limits. Change.

3.2.3.c.2.a added reference to 40 CFR 141, Recommendation associated with industry L Safe Drinking Water Act even though this is the responsibility of the water treatment facility.

expert and documented in CR 05217064.

Change. L Bases - Gaseous Effluents changed verbiage to more accurately reflect the guidance after Recomnm endation associated with industry expert and documented in CR 05217064.

L the change of 10 CFR 20 concentration limits Change.

in the mid 1990's.

3.2.5.b Clarified that this is a two times Recommendation associated with'industry L

trigger that starts an evaluation of 40 CFR expert and documented in CR 05217064.

190 limits, 10 CFR 50, Appendix I limit Change.

exceedances are covered under steps 3.2.3.c. 1, 3.2.4.b. 1, and 3.2.4.c. I L 3.8.2.h added step clarify that even though we Resulting from implementing have identified dose limits are per unit we will still calculate doses per site for ALARA recommendation associated with industry expert and documented in CR 05217064.

L conservatism. Change. .3, Table Notation I removed 12-DCP-585 completing the installation of L automatic isolation restriction for steam Eberline radiation monitors and implementing generator blowdown and blowdown treatment Operations concerns. Change.

radiation monitors. This is indication only.

Table notation 2 removed loss of sample flow and # notation for ESW since the Eberline monitors have been abandoned.

Off ce Information ForForm Tracking Only - Not Partof Form This isafree-form as called out in PMP-2010-PRC-002, Proccdurc Alteration, Review, and Approval. Page Z of £ L.

REVISION

SUMMARY

Number: PMP-601 0-OSD-001 Revision: 19

Title:

OFF-SITE DOSE CALCULATION MANUAL Alteration Justification Attachment 3.11 ESW Eberline monitor 12-CMM-50162, Properly Abandon/Demo d signation was removed since the Eberline Eberline ESW monitors. These monitors monitors have been abandoned.. were never placed in service..

Attachment 3.6 Note was deleted that allowed Current plant practices and recommendation combining of continuous liquid stream associated with industry expert and composite samples plus the event trigger documented in CR 05217065. Change.

dscribing when combining of stream should not be done. No marginal markings.

The 50.59 process is not applicable for the ODCM since changes to this document are governed under Improved Technical Specification 5.5.1.

[ Office Information ForForm Tracking Only-Not Part of Form this is a free-form as called out in PMP-201 O-PRC-002, Procedure Alteration, Review,

[nd Approval. Page 3 of So

REVISION

SUMMARY

Number: PMP-601 0-OSD-001 Revision: 19

Title:

OFF-SITE DOSE CALCULATION MANUAL IMPLEMENTATION PLAN Summary of Change Deleted references to Current TS items and removed headings for Improved TS. Added definitions from original TS. Added per unit designation to 10 CFR 50 Appendix I dose guidelines, added link to 40 CFR 141 Safe Drinking Water Act, and clarified reporting requirements. Clarified annual dose rate limits and the link to 10 CFR 20 concentration limits. Removed automatic isolation restriction for steam generator blowdown and blowdown treatment radiation monitors. Deleted note that explicitly allowed combining of continuous liquid stream composite samples.

Reasonfor Change Improved TS have been fully and solely implemented. Some definitions that were in the original TS ire not supplied in the Improved TS. Based on contractor recommendation and to provide uniformity with industry standards. 12-DCP-585 is completing the installation of the Eberline blowdown radiation monitors and is driving the change made here and to other implementing procedures. Composite samples have been separated since 6/2/05 due to change in Chemistry Management philosophy and contractor recommendation. L Implementation Schedule The implementing procedures for Eberline radiation monitors are being tracked under 12-

DCP-585 implementation and need not be tracked here. The changes to these procedures are actually waiting on the cfhiage to the ODCM: Since the continuous liquid composite samples have already been separated and the implementing procedures state only that combining them is acceptable, this change can take place without changing the two L

implementing procedures since the change to the ODCM does not actually forbid combining composite streams. This plan and supporting Action Request merely track that the explicit acceptance is removed from the procedures to closely reflect ODCM guidance L

and portray current plant practices.

TrainingNeeds None.

L ExpirationDate N/A L

Required Basis Documents Update L N/A Related Processesand Procedures L 12-THP-6020-CHM-201, Steam Generator Chemistry Specifications, and 12-CHM-6020-CHM-3 1l, Turbine Room Sump.

IL Office Information ForForm Trackin Only-Not Partof Form This is a free-form as called out in PMP-2010-PRC-002, Procedure Alteration, Review, and Approval. Page . of.*.

REVISION

SUMMARY

Number: PMP-6010-OSD-001 Revision: 19

Title:

OFF-SITE DOSE CALCULATION MANUAL __

Transition Plan N/A Related EquipmentModifications 12-DCP-585 CommunicationPlan N/A Special Tools, Aids, Permits, Etc N/A Related ConditionReports 05060002, 05217064, 05217065, 05271019, 05276033, 05276040, and 05276043

_ Office Information ForForm Tracking Only -Not Partof Form This is a free-fonn as called out in PMP-201 O-PRC-002, Procedure Alteration, Review, Page 6j of and Approval.

fiai - .1.B-W PMP-6010-OSD-001 Rev. 19 Page 1 of 84 OFF-SITE DOSE CALCULATION MANUAL Lj ___ Reference EffectiveDate:1.A/oI Douie Foster John Carlson Environmental Writer Owner Cognizant Organization TABLE OF CONTENTS I PURPOSE AND SCOPE............................... 4 2 DEFINITIONS AND ABBREVIATIONS .............................. 4 3 DETAILS ...... ; ;...

3.1 Calculation of Off-Site Doses ........................ ';

3.1.1 Gaseous Effluent Releases ......................... . .';

3.1.2 Liquid Effluent Releases ................................................ . 10 3.2 Limits of Operation and Surveillances of the Effluent Release Points .................. 13 3.2.1 Radioactive Liquid Effluent Monitoring Instrumentation . . 13 I 3.2.2 Radioactive Gaseous Effluent Monitoring Instrumentation ................ 14I 3.2.3 Liquid Effluents ....... .................................. 1';

a. Concentration Excluding Releases via the Turbine Room Sump (RS) Discharge .......................................... 1'

b. Concentration of Releases from the TRS Discharge ........................... 15

c. Dose ......................................... 16 L d. Liquid Radwaste Treatment System ......................................... I 3.2.4 Gaseous Effluents ...................................... 1......

9

a. Dose Rate .......................................... 15 1b. Dose - Noble Gases ......................................... IC

c. Dose -lodine-131, Iodine-133, Tritium, and Radioactive Material in Particulate Form ...................... 15 j d. Gaseous Radwaste Treatment ........... .. . 20 3.2.5 Radioactive Effluents - Total Dose;............................ 22' 3.3 Calculation of Alarm/Tnrip Setpoints . . ............ . 27i 3.3.1 Liquid Monitors ...... ........... 24....................

V

a. Liquid Batch Monitor Setpoint Methodology .................................... 24

b. Liquid Continuous Monitor Setpoint Methodology ............................ 25 3.3.2 Gaseous Monitors ..................................... 27

a. Plant Unit Vent ..................................... 27 J b. Waste Gas Storage Tanks ..................................... 3CI

c. Containment Purge and Exhaust System...................................... 30

d. Steam Jet Air Ejector System (SJAE) ..................................... 31

e. Gland Seal Condenser Exhaust ..................................... 31

4.

fWRPMP.6010-OSD-001 Rev. 19 l Page 2 of 84 OFF-SITE DOSE CALCULATION MANUAL Reference I l Effective Date:ID 0 5~'f Doug Foster John Carlson Environmental Writer Owner Conzant Organizao 3.4 Radioactive Effluents Total Dose ........................................... 32 3.5 Radiological Environmental Monitoring Program (REMP) ............................... 32 3.5.1 Purpose of the REMP ................... ........................ 32 3.5.2 Conduct of the REMP .. ................................ 32 3.5.3 Aimual Land Use Census .. 34 3.5.4 Interlaboratory Comparison Program............. .................................. 35 3.6 Radioactive Equipment Storage Facility (Mausoleum) Groundwater Monitoring Program........ .................. . .... ;'.35 3.6.1 Purpose of the Radioactive Equipment Storage Facility (Mausoleum)

Groundwater Radiological Monitoring Program.................................. 35 3.6.2 Conduct of the Radioactive Equipment Storage Facility (Mausoleum)

Groundwater Radiological Monitoring Program ... 36 3.7 Meteorological Model ....... 36 3.8 Reporting Requirements........ . .. 36 3.8.1 Annual Radiological Environmental Operating Report (AREOR) ........ 36 3.8.2 Annual Radiological Effluent Release Report (ARERRY ....................... 37 3.9 10 CFR 50.75 (g) Implementation ...... 38 3.10 Reporting/Management Review . . . . . . 39 4 FINAL CONDITIONS ............................................................ 39 L

5S REFERENCES ................................................................................................................ 40 L

SUPPLEMENTS .1 Dose Factors for Various Pathways ........................................ Pages 42 - 45 .2 Radioactive Liquid Effluent Monitoring Instruments .................. Pages 46 - 47 .3 Radioactive Liquid Effluent Monitoring Instrumentation Surveillance Requirements ............ ............................ Pages 48 - 49 .4 Radioactive Gaseous Effluent Monitoring Instrumentation ......... Pages 50 - 52 Attachinent 3.5 Radioactive Gaseous Effluent Monitoring Instrumentation Surveillance Requirements ............ ............................ Pages 53 - 54

I r >UPMP-6010-OSD-001 Rev. 19 Page 3 of 84 OFF-SITE DOSE CALCULATION MANUAL

_ Reference _EffectiveDateloI0.2/.Q$

Doug Foster John Carlson Environmental Writer Owner Cognizant Organization Attahxhment 3.6 Radioactive Liquid Waste Sampling and Analysis Program ........ Pages 55 - 56 Attachment 3.7 Radioactive Gaseous Waste Sampling and Analysis Program ..... Pages 57 - 53 Attachment 3.8 Multiple Release Point Factors for Release Points ............................... Page 5)

Attauhment 3.9 Liquid Effluent Release Systems ............. ............................. Page 60 Attachment 3.10 Plant Liquid Effluent Parameters .............. ............................ Page 6:1 Attazhment 3.11 Volumetric Detection Efficiencies for Principle Gamma Emitting Radionuclides for Eberline Liquid Monitors .......................... Page 62 Atta:hment 3.12 Counting Efficiency Curves for R-19, and R-24 ......................... Pages 63 - 64 Attabihment 3.13 Counting Efficiency Curve for R-20, and R-28 .................................... Page 65 Atta,:lhment 3.14 Gaseous Effluent Release Systems .......................................... Page 66i Attacrhment 3.15 Plant Gaseous Effluent Parameters ................. ......................... Page 67 .16 10 Year Average of 1995-2004 Data .......................................... Pages 68 - 69 .17 Annual Evaluation of X/Q and D/Q Values For All Sectors ................. Page 70 Attac hment 3.18 Dose Factors............................................................................... Pages 71

- 7,2 .19 Radiological Environmental Monitoring Program Sample Stations, Sample Types, Sample Frequencies .............................. Pages 73 -76 .20 Maximum Values for Lower Limits of DetectionsAB - REMP..... Pages 77 - 78 .21 Reporting Levels for Radioactivity Concentrations in Environmental Samples ........................... ; Page 79 .22 On-Site Monitoring Location - REMP ...... .................... Page 80 .23 Off-Site Monitoring Locations - REMP .......................... Page 8:1 .24 Safety Evaluation By The Office Of Nuclear Reactor Regulation.................................................................................. Pages 82

- 84

Reference I PMP6010-OSD:,01 I Rev. 19 l Page 4 of 84 OFF-SITE DOSE CALCULATION MANUAL 1 PURPOSE AND SCOPE NOTE: Tfis is an Administrative procedure and only the appropriate sections need be performed per PMP-2010-PRC-003, step 3.2.7.

The Off-Site Dose Calculation Manual (ODCM) is the top tier document for the Radiological Environmental Monitoring Program (REM?), the Radioactive Effluent Controls Program (RECP), contains criteria pertaining to the previous Radiological Effluent Technical Specifications (RETS) as defined in NUREG-0472, and fully implements the requirements of Technical Specification 5.5.3, Radioactive Effluent Controls Program? L

The ODCM contains the methodology and parameters to be used in the calculation of off-site doses due to radioactive liquid and gaseous effluents and in the calculation of liquid and gaseous monitoring instrumentation alarm/trip setpoints.

L

.:* The ODCM provides flow diagrams detailing the treatment path and the major components of the radioactive liquid and gaseous waste management systems. L

The ODCM presents maps of the sample locations and the meteorological model used to estimate the atmospheric dispersion and deposition parameters. L

The ODCM specifically addresses the design characteristics of the Donald C.

Cook Nuclear Plant based on the flow diagrams contained on the "OP Drawings" and plant "System Description" documents.

L 2 DEFINITIONS AND ABBREVIATIONS L

Term:

S or shiftly D or daily Meaning:

At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months L

W or weekly At least once per 7 days M or monthly Q or quarterly SA or semi-annually At least once per 31 days At least once per 92 days At least once per 184 days L

R At least once per 549 days. L S/U Prior to each reactor startup P. Completed prior to each release Sampling evolution Process of changing filters or obtaining grab samples L.

Member(s) of Public AU persons who are not occupationally associated with the plant. Does not include employees of the utility, its contractors or its vendors. Also excluded from this category are persons who enter the site to service equipment or to make deliveries. This category does include persons who use portions of the site for recreational, occupational or other purposes not associated with the plant.

Purge/purging The controlled process of discharging air or gas from a

Reference I PMP-60100S-001 __ I Rev. 19 Page5of84 OFF-SITE DOSE CALCULATION MANUAL confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is required to purify the confinement.

Source check The qualitative assessment of Channel response when the Channel sensor is exposed to a radioactive source.

Venting' Controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is not provided or required. Vent, used in system names, does not imply a venting process.

3 DETAILS

... ...,. i. 1 3.1 *.Galculationof Off-Site Doses

. . dr. ..

3.1.1 Gaseous Effluent Releases

. ... "I .

a. The computer program MIDAS (Meteorological Information and Dose Assessment System) performs the calculation of doses from effluent releases. The site-specific parameters associated with MIDAS reside in the following subprograms:

MIDER

MIDEX

MIDEL

MIDEG

MIbEN

b. The subprogram used to enter and edit gaseous release data is called MDIEQ (EQ). The data entered in EQ can be used to calculate the accumulation of dose to individual land based receptors based on hourly meteorology and release data: The air dose from this data is calculated via the XDAIR subprogram in MIDAS. It computes air dose results for use in Reg. Guide 1.21 reports and 10 CFR 50 Appendix I calculations based on routine releases.

c. The formula used for the calculation of the air dose is generated from site specific parameters and Reg. Guide 1.109 (Eq 7):

Dr,Dpair=X*Z(MjorN,)*Q,*3.17E-81 Q

Where; Dy, Dp air = the gamma or beta air dose in nrad/yr to an individual receptor

ence PMP-6910-OSD-001 Rev. 19 Page 6 of 84 OFF-SITE DOSE CALCULATION MANUAL

/ 1Q the annual, average or real time atmospheric dispersion factor over land, sec/r 3 from Attachment 3.16, 10 Year Average of 1995-2004 L Data M; = the gamma air dose factor, nrad m 3 / yr tiCi, from Attachment 3.18, Dose Factors N = the beta air dose factor, mrad d 3 / yr pCi, from Attachment 3.18, Dose Factors the release rate of radionuclide, 'T', in gCi/yr.

3.17E-8 number of years in a second (years/second). L

d. The value for the ground average X/ Q for each sector is calculated using equations shown below. Formula used for the calculation is L generated from parameters contained in MIDAS Technical Manual, XDCALC (Eq 2). L 2.03 *TL Where; minim um ofor = L3 x = distance downwind of the source, meters. This information is found in parameter 5 of MIDEX L Ut = wind speed for ground release, (meters/second) L a:, = vertical dispersion coefficient for ground release, (meters),

(Reg. Guide 1.111 Fig.1)

H, = building height (meters) from parameter 28 of MIDER.

(Containment Building = 49.4 meters)

L Tf = terrain factor (= 1 for Cook Nuclear Plant) because we consider all our releases to be ground level (see parameter 5 in MIDEX).

2.03 = . +0.393 radians(22.50)

= Reference PMP-6010-OSD-001 ' Rev. 19 l Page 7 of 84 OFF-SITE DOSE CALCULATION MANUAL

e. The dose due to gaseous releases, other than the air dose, is calculated by the MIDAS subprogram GASPRO. GASPRO computes the accumulation of dose to individual receptors based on hourly meteorology and release data. Calculations consider the effect of each important radionuclide for each pathway, organ, age group, distance and direction.

f. Calculations are based on the environmental pathways-to-man models in Reg. Guide 1:109. The program considers 7 pathways, 8 organs, and 4 age groups in 16 direction sectors. The distances used are taken from the MIDEG file.

g. The formulas used for the following calculations are generated from site specific parameters and Reg. Guide 1.109:

1. Total Body Plume Pathway (Eq 10)

Dose (mrem/year)= 3.17E - 8 * ((Q. 1 Q

Sf

DFBd)

Where; Sf = shielding factor that accounts for the dose reduction due to shielding provided by residential structures during occupancy (maximum exposed individual = 0.7 per Table

-S15of Reg. Guide 1.109)

DFBi = the whole body dose factor froinTable B-1 of Reg. Guide 1.109, nrem - d 3 per jiCi - yr. See Attachment 3.18, Dose Factors.

= the release rate of radionuclide "Y', in tiCi/yr

2. Skin Plume Pathway (Eq 11)

Dose (mrem/yr)= 3.17E- 8

Sr * *a i DFr)+z(Q,* DFS,)J Q *1.11*

Q Where; 1.11 = conversion factor, tissue to air, mremfmrad DP j= = the gamma air dose factor for a uniform semi-infinite cloud of radionuclide "i", in mrad n 3 4/iCi yr from Table B-1, Reg. Guide 1.109. See Attachment 3.18, Dose Factors.

Reference I PMP-6010-OSD-001 I -Rev. 19 l Page 8 of 84 OFF-SITE DOSE CALCULATION MANUAL DFSi = the beta skin dose factor for a semf-infinite cloud of radionuclide "i", in mrem m/liCi yr from Table B-1, Reg. Guide 1.109. See Attachment 3.18, Dose Factors.

3. Radionuclide and Radioactive Particulate Doses (Eq 13 & 14) _

The dose, Drp in mrem/yr, to an individual from radionuclides, other than noble gases, with half-lives greater than eight days in gaseous effluents released to unrestricted areas will be determined as follows:

Dip(mremlyear)=3.17E- 8 *A(RS *W

Q)

'Where; _

Ri = the most restrictive dose factor for each identified radionuclide "i", in m2 mrem sec / yr ,iCi (for food and ground pathways) or mi-em m3 I yr piCi (for inhalation pathway), for the appropriate pathway For sectors with existing pathways within five miles L of the site, use the values of Ri for these real pathways, otherwise use pathways distance of five miles. See Attachment 3.1, Dose Factors for Various L Pathways, for the maximum R; values for the most controlling age group for selected radionuclides. Ri values were generated by computer code PARTS, see NUREG-0133, Appendix D.

W = the annual average or real time atmospheric dispersion parameters for estimating doses to an individual at the worst case location, and where W is L

further defined as:

WM, = Xl Q for the inhalation pathway, in sec/M3

-OR-Wfg = DIQ forthefoodandgroundpathwaysin I/m2

= the release rate of those radioiodines, radioactive materials in particulate form and radionuclides other than noble gases with half-lives greater than eight days, in pCi/yr h This calculation is made for each pathway. The maximum computed dose at any receptor for each pathway is selected. These are summed together to get the dose to compare to the limits. Only the maximum of the cow milk or goat milk pathway (not both) is included in the total.

Reference PMP-6010-OSD-001 Rev. 19 Page 9 of 84 OFT-SITE DOSE CALCULATION MANUAL

i. In addition to the above routines, the QUICKG routine of the MIDAS system is used to provide data used in the monthly reports due to its ability to use annual average meteorological data rather than real time data, thus shortening the run time involved.

j. Steam Generator Blowdown System (Start Up Flash Tank Vent)

I . The amount of radioiodine and other radionuclides that are released via the start up flash tank and its vent are calculated through actual sample results while the start up flash tank is in service.

2. The following calculation is performed to determine the amount of curies released through this pathway. (Plant established formula.)

Curies= -w

GPM *time on flash tank ( m )*3.785E - 3 Where; 3.785E-3 = conversion factor, ml CiltiCi gal. &t.

3. The flow rate is determined from the blowdown valve position and the:

time on the start up tank. Chemistry Department performs the sampling and analysis of the samples.

4. This data is provided to the MIDAS computer and dose calculations (liquid and gas) are performed to ensure compliance with Subsection 3.2, Limits of Operation and Surveillances of the Effluent Release Points, dose limits. MIDAS uses the formulas given in step 3.1.2, Liquid Effluent Releases, to calculate doses to members of the public.

NOTE: This section provides the minimum requirements to be followed at Donald C Cook Nuclear Plant. This would be used if actual sample data was not available each time the start up flash tank was in service.

5. The radioiodine release rate must be determined in accordance with the following equation every 31 day period whenever the specific activity of the secondary coolant system is greater than 0.01 [LCi/g dose equivalent 1-13 1.

6. IF the specific activity of the secondary coolant system is less than 0.01 tiCi/g dose equivalent 1-131, THEN the release rate mustbe determined once every six months. Use the following plant established equation:

QY = Ci

IPF* Rs5b Where; Qy = the release rate of 1-131 from the steam generator flash tank vent, in pCi/sec Ci = the concentration (tpCi/cc) of I-131 in the secondary coolant averaged over a period not exceeding seven days

Reference PMP-6010-OSJ01 Rev. 19 l Page 10 of 84 OFF-SITE DOSE CALCULATION MANUAL IPF = the iodine partition factor for the Start Up Flash Tank, 0.05, in accordance with NUREG-0017 Rsgb = the steam generator blowdown rate to the start up flash tank, in cc/sec

7. Use the calculated release rate in monthly dose projections until the next determination to ensure compliance with Subsection 3.2, Limits of Operation and Surveillances of the Effluent Release Points, dose limits. Report the release rate calculations in the Annual Radioactive Effluent Release Report.

L 3.1.2 Liquid Effluent Releases L

a. The calculation of doses from liquid effluent releases is also performed by the MIDAS program. The subprogram used to enter and edit liquid release data is called MDIEB (EB).

b. To calculate the individual dose (mrem), the program DS ILI (ID) is usedt It computes the individual dose for up to 5 receptors for 14 liquid L~

pathways due to release of radioactive liquid effluents. The pathways can be selected using the MIDEL program and changing the values in parameter 1. D.C Cook Nuclear Plant uses 3 pathways: potable water, shoreline, and aquatic foods (fresh water sport fishing).

c. Steam Generators are sparged, sampled, and drained as batches usually early in outages to facilitate cooldown for entry into the steam generator.

This is typically repeated prior to startup to improve steam generator chemistry for the startup. The sample stream, if being routed to the operating unit blowdown, is classified as a continuous release for quantification purposes to maintain uniformity with this defined pathway.

L

d. The equations used are generated from site specific data and Reg. Guide 1.109. Theyareasfollows:

1. Potable Water (Eq 1)

L R.Fi1100* (lap MP* F*2.23E-3 Where; Rap= the total annual dose to organ 'J" to individuals of age groups "a" from all of the nuclides "i" in pathway "p", in mrem/year 1100 - conversion factor, yr ft3 pCi I Ci sec L U.P = a usage factor that specifies the exposure time or intake rate for an individual of age group "a" associated with pathway "p". Given in #29-84 of parameter 4 in MIDEL and Reg. Guide 1.109 Table E-5. See Attachment 3.1, Dose Factors for Various Pathways.

Reference l PMP-6010-OSDOO1 I Rev. 19 l Page 11 of 84]

OFF-SITE DOSE CALCULATION MANUAL Mp = the dilution factor at the point of exposure (or the point of withdrawal of drinking water or point of harvest of aquatic food). Given in parameter 5 of MIDEL as 2.6.

F = the circulation water system water flow rate, in gpm, is used for evaluating dose via these pathways as dilution flow 2.23E-3 = conversion factor, f3 min / sec gal

= the release rate of nuclide "i" for the time period of the run input via MIDEB, Curies/year Daipj = the dose factor, specific to a given age group "a",

radionuclide 'i", pathway "p", and organ 'J", which can be used to calculate the radiation dose from an intake of a radionuclide, in mrem/pCi. These values are taken from tables E-I1 through E414 of Reg. Guide 1.109 and are located within the MIDAS code.

= the radioactive decay constant for radionuclide 'i", in hours~'

t = the average transit time required for nuclides to reach the point of exposure, 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . This allows for nuclide transport through the water purification plant and the water distribution system. For internal dose, tp is the total elapsed time between release of the nuclides and ingestion of food or water, in hours. Given as #25 of parameter 4 in MIDEL (tp =12 hours)

2. Aquatic Foods (Eq 2)

Rwj=11OO

U.7P

QiAP* Dw6..1a ,

M *F*2.23E-3 I Where, BjP = the equilibrium bioaccumulation factor for nuclide "i" in pathway "p", expressed as pCi L / kg pCi. The factors are located within the MIDAS code and are taken from Table A-1 of Reg. Guide 1.109. See Attachment 3.1, Dose Factors for Various Pathways.

t = the average transit time required for nuclides to reach the point of exposure, 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . This allows for decay during transit through the food chain, as well as during food preparation. Given as #26 of parameter 4 in MIDEL. (tp

= 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months )

M = the dilution factor at the point of exposure, 1.0 for Aquatic Foods. Given in parameter 5 of MIDEL as 1.0.

Reference I PMP-6010-OSD-OD1 l Rev. 19 l Page 12 of 84 OFF-SITE DOSE CALCULATION MANUAL

3. Shoreline Deposits (Eq 3)

UCPW* * .- vl* 1 Jr1 Rg~j= F P*E)

M,*F*2.23E-3

1,0 E fT

2PT3 *3D.o~~'][

[F;JPJLJ e-~

Cub Where; W = the shoreline width factor. Given as an input of 0.3 when running the program, based on Table A-2 in Reg. Guide 1.109.

Ti = the radioactive half-life of the nuclide, "i", in days D,,pj= the dose factor for standing on contaminated ground, in 1n1rnem m2 / hr pCi The values are taken from table E-6 of Reg. Guide 1.109 and are located within the MIDAS code.

See Attachment 3. 1, Dose Factors for Various Pathways.

tb the period of time for which sediment or soil is exposed io the contaminated water, 1.31E+5 hours. Given in MIDEL as item 6 of parameter 4. L tP = the average transit time required for nuclides to reach the point of exposure, 0 hours0 days 0 hours 0 weeks 0 months . Given as #28 of parameter 4 in MIDEL I 10,00. =conversion factor yr ft3 pCi / Ci sec m2 day, this accounts for proportionality constant in the sediment radioactivity model L

MP = the dilution factor at the point of exposure (or the point of withdrawal of drinking water or point of harvest of aquatic food). Given in parameter S of MIDEL as 2.6.

L

e. The MIDAS program uses the following plant specific parameters, which are entered by the operator.

L

1. Irrigation rate = 0

2. Fraction of time on pasture = 0 3.

4.

Fraction of feed on pasture = 0 Shore width factor = 0.3 (from Reg. Guide 1.109, Table A-2) L f The results of DS ILI are printed in LDRPT (LP). These results are used in the monthly report of liquid releases.

g. In addition, the program DOSUM (DM) is used to search the results files ofDSILI to find the maximum liquid pathway individual doses. The L

highest exposures are then printed in a summary table. Each line is compared with the appropriate dose limit. The table provides a concise summary of off-site environmental dose calculations for inclusion in L Annual Radioactive Effluent Release Reports, required by Reg. Guide 1.21.

Reference PMP-6010-OSDO00 I Rev. 19Page 23 of 84 OFF-SITE DOSE CALCULATION MANUAL NOTE: The performance of each surveillance requirement must be within the specified time interval with a maximum allowable extension not to exceed 25% of the specified surveillance interval.

3.2 Limits of Operation and Surveillances of the Effluent Release Points 3.2.1 Radioactive Liquid Effluent Monitoring Instrumentation a The radioactive liquid effluent monitoring instrumentation channels shown in Attachment 3.2, Radioactive Liquid Effluent Monitoring Instruments, are operable with their alarm/trip setpoints set to ensure that the limits of step 3.2.3a, Concentration Excluding Releases via the Turbine Room Sump (TRS) Discharge, are not exceeded.

b. The applicability of each channel is shown in Attachment 3.2, Radioactive Liquid Effluent Monitoring Instruments.

c. With a radioactive liquid effluent monitoring instrumentation channel alarmnltrip setpoint less conservative than a value which will ensure the limits of step 3.2.3a, Concentration Excluding Releases via the Turbine Room Sump CTRS) Discharge, are met without delay, suspend the release of radioactive liquid effluents monitored by the affected channel and reset or declare the monitor inoperable.

d. With one or more radioactive liquid effluent monitoring instrumentation channels inoperable, take the applicable action shown in Attachment 3.2, Radioactive Liquid Effluent Monitoring Instruments, with a maximum allowable extension not to exceed 25% of the surveillance interval, excluding the initial performance.

e. Determine the setpoints in accordance with the methodology described in step 3.3.1, Liquid Monitors. Record the setpoints.

f. Demonstrate each radioactive liquid effluent monitoring instrumentation channel is operable by performing the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION and CHANNEL OPERATIONAL TEST at the frequencies shown in Attachment 3.3, Radioactive Liquid Effluent Monitoring Instrumentation Surveillance Requirements.

BASES - LIQUID The radioactive liquid effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in liquid effluents during actual or potential releases. The alarm/trip setpoints for these instruments shall be calculated in accordance with NRC approved methods in the ODCM to ensure the alarm/trip will occur prior to exceeding the limits of 10 CFR Part 20. The OPERABILITY and use of this instrumentation is cor.sistent with the requirements of General Design Criteria specified in Section 21.3 of the Final Safety Analysis Report for the Donald C. Cook Nuclear Plant.

Reference - PMP-6010-OSD-001 Rev. 19 Page 14 of 84 L OFF-SITE DOSE CALCULATION MANUAL Due to the location of the Westinghouse ESW monitors, outlet line of containment spray heat exchanger (typically out of service), weekly sampling is required of the ESW system for radioactivity. This is necessary to ensure monitoring of a CCW to ESW system leak. [Ref 5.2. 1gg3 3.2.2 Radioactive Gaseous Effluent Monitoring Instrumentation

a. The radioactive gaseous process and effluent monitoring instrumentation channels shown in Attachment 3.4, Radioactive Gaseous Effluent Monitoring Instrumentation, are operable with their alarm/trip setpoints L

set to ensure that the limits of step 3.2.4a, Dose Rate, are not exceeded.

b. The applicability of each channel is shown in Attachment 3.4, Radioactive Gaseous Effluent Monitoring Instrumentation.

L

c. With a radioactive gaseous process or effluemkt'monitoring instrumentation channel alarmntrip setpoint less conservative thana value which will ensure that the limits of step 3.2.4a, Dose Rate; are met, without delay, suspend the release of radioactive gaseous effluents monitored by the affected channel and reset or declare the channel inoperable.

d. With less than the minimum number of radioactive gaseous effluent monitoring instrumentation channels operable, take the action shown in Attachment 3.4, Radioactive Gaseous Effluent Monitoring IL Instrumentation, with a maximum allowable extension not to exceed 25%

of the surveillance interval, excluding the initial performance.

L~

NOTE: This surveillance requirement does not apply to the waste gas holdup system hydrogen and oxygen monitors, as their setpoints are not addressed in this L document.

e. Determine the setpoints in accordance with the methodology as described in step 3.3.2, Gaseous Monitors. Record the setpoints.

L

f. Demonstrate each radioactive gaseous process or effluent monitoring instrumentation channel is operable by performing the CHANNEL L CHECK, SOURCE CHECK, CHANNEL CALIBRATION, and CHANNEL OPERATIONAL TEST operations at the frequencies shown in Attachment 3.5, Radioactive Gaseous Effluent Monitoring Instrumentation Surveillance Requirements.

L I.-

Reference lP - 'P.MP-6010-OSD-001 Rev. 19 Page 15 of 841 l OFF-SITE DOSE CALCULATION MANUAL BASES - GASEOUS ThAradioactive gaseous effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in gaseous effluents during actual or potential LI relmases. The alarmi/trip setpoints for these instruments shall be calculated in accordance with NR.C approved methods in the ODCM to ensure the alarm/trip will occur prior to exceeding the limits of ID0 CFR Part 20. The OPERABILITY abd use of this instrumentation is I . consistent with the requirements of General Design Criteria specified in Section 11.3 of the Final Safety Analysis Report for the Donald C. Cook Nuclear Plant.

3.2.3 Liquid Effluents

a. Concentration Excluding Releases via the Turbine Room Sump (TRS)

Discharge

1. Limit the concentration of radioactive material released via the Batch Release Tanks or Plant Continuous Releases (excluding only TRS discharge to the Absorption Pond) to unrestricted'areas to the L *concentrations in 10 CFR 20, Appendix B, Table 2, C6lumn 2, for radionuclides other than dissolved or entrained nob1'6 gases. For dissolved or entrained noble gases, limit the concentration to i 2E-4 l+/-Ci/ln total activity.

2. With the concentration of radioactive material released from the site via the Batch Release Tfilcs or Plant Continuous Releases (other than Li the TRS to the Absorption Pond) exceeding the above limits, without delay restore the concentration to within the above limits.

.3. Sample and analyze radioactive liquid'wastes according to the L~ sampling and analysis program of Attachment 3.6, Radioactive Liquid Waste Sampling and Analysis Programr.

4. Use the results of radioactive analysis in accordance with the methods of this document to assure that all concentrations at the point of release are maintained within limits.

Li b. Concentration of Releases from the TRS Discharge I. Limit releases via the TRS discharge to the on-site Absorption Pond to

,J the concentrations specified in 10 CFR 20, Appendix B, Table 2, Column 2. For dissolved or entrained noble gases, limit the concentration to 2E-4 ICi/mil total activity.

2. With releases from the TRS exceeding the above limits, perform a dose projection due to liquid releases to UNRESTRICTED AREAS to determine if the limits of step 3.2.3c. 1 have been exceeded. If the dose limits have been exceeded, follow the directions in step 3.2.3c.2, as applicable.

3. Sample and analyze radioactive liquid wastes according to the program in Attachment 3.6, Radioactive Liquid Waste Sampling and Analysis Program.

Ijr Jr-i Ia I JVU3r lI-JA.ULA I IUINNI&AINUALl

4. Use the results of radioactive analysis in accordance with the methods of this document to assure that all concentrations at the point of release are maintained within the limits stated above.

c. Dose '

1. Limit the dose or dose commitment to an individual from radioactive material in liquid effluents released to unrestricted areas during any calendar quarter to c1.5 mremlunit to the total body and to < 5 mremn/unit to any organ, and during any calendar year to < 3 rmrem/unit to the total body and to ID0 mrem/unit to any organ.

2. With the calculated release of radioactive materials in liquid effluents exceeding ten times any of the limits in Steps 3.2.3a or 3.2.3b, or exceeding 3.Z.3c.1 above, prepare and submit a Written Report, pursuant to 10 CFR 20.2203, within 30 days after learning of the event. This report must describe the extent of exposure of inidividuals -

to radiation and radioactive material, including,.as appropriate:

a) tstimate of each individual's dose. This is to include the radiological impacts on finished drinking water supplies with L

regard to the requirements of 40 CFR 141, Safe Drinking Water Act (applicable due to Lake Township water treatment facility),

b) Levels of radiation and concentration of radioactive material L

involved, c) Cause of elevated exposures, dose rates or concentrations,

-AND- L d) Corrective steps taken or planned to ensure against recurrence, including schedule for achieving conformance with applicable limits.L L

These reports must be formatted in accordance with PMF-7030-001-002, Licensee Event Reports, Special and Routine Reports, even though this is not an LER L

3. Determine cumulative and projected dose contributions from liquid effluents in accordance with this document at least once per 31 days.

Dose may be projected based on estimates from previous monthly L

projections and current or future plant conditions.

d. Liquid Radwaste Treatment System L

1. Use the liquid radwaste treatment system to reduce the radioactive materials in liquid wastes prior to their discharge when the projected L doses due to the liquid effluent from the site when averaged over 31 days, would exceed 0.06 mrem/unit to the total body or 0.2 mrernmunit to any organ.

2. Project doses due to liquid releases to UNRESTRICTED AREAS at least once per 31 days, in accordance with this document.

L

= Reference I PMP-6010:OSD-001- I Rev. 19 l Page 17 of 84 OFF-SITE DOSE CALCULATION MANUAL

e. During times of primary to secondary leakage, the use of the startup flash tank should be minimized to reduce the release of curies from the secondary system and to maintain the dose to the public ALARA.

Operation of the North Boric Acid Evaporator (NBAE) should be done in a manner so as to allow the recycle of the distillate water to the Primary Water Storage Tank for reuse. This.will provide a large reduction in liquid curies of tritium released to the environment, as there is approximately 40 curies of tritium released with every monitor tank of NBAE distillate.

Drainage of high conductivity water (Component Cooling Water and ice melt water containing sodium tetraborate) shall be-evaluated to decide whether it should be drained to waste (small volumes only), the Turbine Room Sump (low activity water only) or routed without demineralization processing to a monitor tank for release. This is necessary in order to minimize the detrimental affect that high conductivity water has on the radioactive wastewater dernineralization system. The standard concentration and volume equation can be utilized to determine the impact on each method and is given here. The units for concentration.

and volume need to be consistent across the equation:

(CX)(Vi)+(Ca)(Va)Q(C)(Vt)

Where; C = the initial concentration of the system being added to V; = the initial volume of the system being added to C. = the concentration of the water that is being added to the system V. = the volume of the water that is being added to the system Ci = the final concentration of ihe system after the addition Vt = the final volume of the system after the addition The intent is to keep the:

WDS below 500 gtmhos/cc.

TRS below IE-5 pC/cc.

Monitor Tank release ALARA to members of the public.

Wastewater leakage into the liquid waste disposal system will be monitored routinely. In the event the leak rate is determined to be over two gallons per minute (the assumed plant design leakage based on the original 2 gpm waste evaporator), increased scrutiny will be placed on locating inleakage, timeliness ofjob order activities, and/or activities .

causing increased production of waste water.

Reference I PMP-6010-OSD-OO1 I

Rev. 19 l Page 18 of 84 OFF-SITE DOSE CALCULATION MANUAL L

BASES - CONCENTRATION This specification is provided to ensure the concentration of radioactive materials released in liquid waste effluents from the site to unrestricted areas will be less than the concentration levels specified in 10 CFR Part 20, Appendix B, Table 2. This limitation provides additional assurance that the levels of radioactive materials in bodies of water outside the site will Dot result in exposures greater than 1) the Section ll.A design objectives of Appendix I, 10 CFR Part 50, to an individual and 2) the limits of 10 CFR Part 20. The concentration limit for noble gasses is based upon the assumption that Xe-135 is the controlling radionuclide and its Effluent Concentration Unit in air (submersion) was converted to an equivalent concentration in water using the methods described in the International Commission on Radiological Protection (ICRP) Publication 2.

DOSE This specification is provided to implement the requirements of Sections IL.A, U1I.A, and IV.A-:

of Appendix 1,^10 CFR Part 50. The dose limits implement the guides set forth in Section fl.A of Appendix 1. The ACTION statements provide the required operating flexibility and agtthe same time,' implement the guides set forth in Section lV.A of Appendix I to assure the releases .:

of radioactive'inaterial in liquid effluents will be kept "as low as is reasonably achievable',';.-

Also, for fresh water sites with drinking water supplies which can be potentially affected by plant operations, there is reasonable assurance that the operation of the facility will not result in radionuclide concentrations in the finished drinking water that are in excess of the L

requirements of 40 CFR 141. The dose calculations in the ODCM implement the requirements in Section Mfl.A of Appendix Xthat conformance with the guides of Appendix I be shown by calculational procedures based on models and data such that the actual exposure of an L

individual through appropriate pathways is unlikely to be substantially underestimated. The equations specified in the ODCM for calculating the doses due to the actual release rates of radioactive materials in liquid effluents, will be consistent with the methodology provided in Regulatory Guide 1.109, "Calculation of Annual Doses to Man from Routine Releases of L

Reactor Effluents for the Purpose of Evaluating Compliance with 10 CER Part 50, Appendix I", Revision 1, October 1977, and Regulatory Guide 1.113, "Estimating Aquatic Dispersion of Effluents from Accidental and Routine Reactor Releases for the Purpose of Implementing L

Appendix I", April 1977. NURBG-0133 provides methods for dose calculations consistent with Regulatory Guide 1.109 and 1.113.

This specification applies to the release of liquid effluents from each reactor at the site. The L

liquid effluents from the shared system are proportioned among the units sharing the system LIQUID WASTE TREATMENT L The operability of the liquid radwaste treatment system ensures that this system will be available for use whenever liquid effluents require treatment prior to release to the environment. The requirements that the appropriate portions of this system be used when L specified provide assurance that the releases of radioactive materials in liquid effluents will be kept "as low as is reasonably achievable". This specification implements the requirements of 10 CFR Part 50.36a, General Design Criteria Section 11.1 of the Final Safety Analysis Report for the Donald C. Cook Nuclear Plant, and design objective Section LI.D of Appendix I to 10 CFR Part 50. The specified limits governing the use of appropriate portions of the liquid radwaste treatment system were specified as a suitable fraction of the dose design objectives set forth in Section MILA of Appendix I, 10 CFR Part 50, for liquid effluents.

Reference I PMP-6010-OSD-001 '- Rev.199 Page 19 of 84 OFF-SITE DOSE CALCULATION MANUAL 3.2.4 Gaseous Effluents

a. Dose Rate

l. Limit the dose rate due to radioactive materials released in gaseous effluents from the site to 5 500 mrem/yr to the total body and

< 3000 mrern/yr to the skin for noble gases. Limit the dose rate due to all radioiodines and for all radioactive materials in particulate form and radionuclides (other than noble gases) with half-lives greater than eight days to

1500 mrem/yr to any organ.

2. With the dose rate(s) exceeding the above limits, without delay decrease the release rate to within the above limit(s).

3. Detennine the dose rate due to noble gases in gaseous effluents to be within the above limits in accordance with the methods and procedures described in this document.

4. Determine the dose rate due to radioactive materials, other than noble '-

gases, in gaseous effluents to be within the above limits in accordance with the methods and procedures of this document by obtaining representative samples and performing analyses in accordance with the sampling and analysis program in Attachment 3.7, Radioactive Gaseous Waste Sampling and Analysis Program.

b. Dose - Noble Gases

1. Limit the air dose in unrestricted areas due to noble gases released in gaseous effluents during any calendar quarter, to c 5 mrad/unit for gamma radiation and*5 10 mrad/unit forbetaradiation and during any calendar year, to : 10 mrad/unit for gamma radiation and *20 mrad/unit for beta radiation.

2. With the calculated air dose from radioactive noble gases in gaseous effluents exceeding any of the above limits, prepare and submit a Written Report, pursuant to 10 CFR 20.2203 and addressed in step 3.2.3c.2, within 30 days after learning of the evenL

3. Determine cumulative and projected dose contributions for the total tirae period in accordance with this document at least once every 31 days.

c. Dose -Jodine-l31, Iodine-133, Tritium, and Radioactive Material in Particulate Form

1. Limit the dose to a MEMBER OF THE PUBLIC from radioiodine, radioactive materials in particulate form, and radionuclides other than noble gases with half-lives greater than eight days in gaseous effluents released to unrestricted areas (site boundary) to the following:

a) During any calendar quarter to less than or equal to 7.5 arem/unit to any organ b) During any calendar year to less than or equal to 15 rnemlunit to any organ.

Reference I PMP-6010-OSD-001 I Rev; 19 I Page 20 of 84 OFF-SITE DOSE CALCULATION MANUAL 2: With the calculated dose from the release of radioiodines, radioactive materials in particulate form, or radionuclides other than noble gases in gaseous effluents exceeding any of the above limits, prepare and submit a Written Report, pursuant to 10 CFR 20.2203 and addressed in step 3.2.3c.2, within 30 days after learning of the event.

3. Determine cumulative and projected dose contributions for the total time period in accordance with this document at least once every 31 days.

d. Gaseous Radwaste Treatment

1. Use the gaseous radwaste treatment system and the ventilation exhaust treatment system to reduce radioactive materials in gaseous wastes prior to their discharge when projected gaseous effluent air doses due to gaseous effluent releases to unrestricted areas when averaged over 31 days, would exceed 0.2 mradfunit for gamma radiation and 0.4 mrad/unit ,

for beta radiation. Use the ventilation exhaust treatment system to reduce radioactive materials in gaseous waste prior to their discharge

.when the projected doses due to gaseous effluent releases to unrestricted-areas when averaged over 31 days would exceed 0.3 mrem/unit to any L organ.

2. Project doses due to gaseous releases to UNRESTRICIED AREAS at least once per 31 days in accordance with this document. L BASES - GASEOUS EFFLUENTS L This specification provides reasonable assurance that radioactive material discharged in gaseous effluents will not result in the exposure of a Member of the Public in an unrestricted area, either at or beyond the site boundary in excess of the design objectives of appendix I to 10 CFR 50.

This specification is provided to ensure that gaseous effluents from all units on the site will be L appropriately controlled. It provides operational flexibility for releasing gaseous effluents to satisfy the Section H.A and ILC design objectives of appendix I to 10 CFR 50. For individuals who may at times be within the site boundary, the occupancy of the individual will be sufficiently L low to compensate for any increase in the atmospheric diffusion factor above that for the site boundary. The specified instantaneous release rate limits restrict, at all times, the corresponding gamma and beta dose rates above background to an individual at or beyond the site boundary to <

500 nrem/yr to the total body or to < 3000 mremfyr to the skin. These instantaneous release rate L

limits also restrict, at all times, the corresponding thyroid dose rate above background to a child via the inhalation pathway to < 1500 mrem/yr. Limitations on the dose rate resulting from radioactive material released in gaseous effluents to areas beyond the site boundary conforming to the doses associated with 10 CFR 20, Appendix B, Table 2, Column 1.

This specification applies to the release of gaseous effluents from all reactors at the site. The gaseous effluents from the shared system are proportioned among the units sharing that system.

. L DOSE, NOBLE GASES This specification is provided to implement the requirements of Sections II.B, UI.A, and IV.A of

.Appefidix I, 10 CFER Part 50. The dose limits implement the guides set forth in Section ll.B of Appendix I. i

Reference I PMP-6010-OSD-OO1 I Rev. 197 l Page 21 of 841 OFF-SITE DOSE CALCULATION MANUAL The ACTION statements provide the required operating flexibility and at the same time imnlement the guides set forth in section IV.A of Appendix I to assure that the releases of radioactive material in gaseous effluents will be kept "as low as is reasonably achievable". Th-Surveillance Requirements implement the requirements in Section III.A of Appendix I that conform with the guides of Appendix I to be shown by calculational procedures based on models and data such that the actual exposure of an individual through the appropriate pathways is unlikely to be substantially underestimated. The dose calculations established in the ODCM for calculating the doses due to the actual release rates of radioactive noble gases in gaseous effluents wil be consistent with the methodology provided in Regulatory Guide 1.109, "Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I", Revision 1, October 1977 and Regulatory Guide 1.1 11, 'Methods forEstimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors", Revision 1, July 1977. The ODCM equations provided for determining the air doses at the site boundary will be based upon.the historical average atmospherical conditions. NUREG-0133 provides methods for dose calculations consistent with Regulatory Guides 1.109 and 1.111.

DC)SE, RADIOIODINES, RADIOACTIVE MATERIAL IN PARTICULATE FORM, AND RIDIONUCLIDES OTHER THAN NOBLE GASES This specification is provided to imiplenent the requirements of Sections II.C, lI.A, and IV.A of Appendix I, 10 CFR Part 50. The dose limits are the guides set forth in Section ll.C of Appendlix I.

The ACTION statements provide the required operating flexibility and at the same time imnlement the guides set forth in section WV.A of Appendix I to assure that the releases of radioactive material in gaseous effluents will be kept "as low as is reasonably achievable". The ODCM calculational methods specified in the surveillance requirements implement the recuirements in Section III.A of Appendix I that conform with the guides of Appendix I to be sho'wn by calculational procedures based on models and data such that the actual exposure of en individual through the appropriate pathways is unlikely to be substantially underestimated. The ODCM calculational methods approved by the NRC for calculating the doses due to the actual release rates of the subject materials are required to be consistent with the methodology provided in Regulatory Guide 1.109, 'Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CER Part 50, Appendix 1",

Revision 1, October 1977 and Regulatory Guide 1.111, 'Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors", Revision 1, July 1977. These equations also provide the methodology for determining the actual doses based upon the historical average atmospheric conditions. The release rate specifications for radioiodines, radioactive material in particulate form, and radionuclides, other than noble gases, are dependent on the existing radionuclide pathways to man, in the unrestricted area. The pathways which are examined in the development of these calculations are: 1) individual inhalation of airborne radionuclides, 2) deposition of radionuclides onto green leafy vegetation with subsequent consumption by man, 3) deposition onto grassy areas where milk animals and meat producing animals graze with consumption of the milk and meat by man, anl

4) deposition on the ground with subsequent exposure of man.

. Reference. PMP-6010-OSD-001 I Rev. 19 l' Page 22 of84 OFF-SITE DOSE CALCULATION MANUAL GASEOUS WASTE TREATMENT The operability of the gaseous radwaste treatment system and the ventilation exhaust treatment systems ensures that the systems will be available for use whenever gaseous effluents require treatment prior to release to the environment. The requirement that the appropriate portions of these systems be used when specified provides reasonable assurance that the releases of radioactive materials in gaseous effluents will be kept "as low as is reasonably achievable". This specification implements the requirements of 10 CFR Part 50.36a, General Design Criterion Section 11.1 of the Final Safety Analysis Report for the Donald C. Cook Nuclear Plant, and design objective Section ll.D of Appendix I to 10 CFR Part 50. The specified limits governing the use of appropriate portions of the systems were specified as a suitable fraction of the guides

'forth in Sections II.B and II.C of Appendix I, 10 COR Part 50, for gaseous effluents.

3.2.5 Radioactive Effluents,.;T6tal Dose

a. The dose or dose commitment to a real -individual from all uranium fuel cycle sources is lifinted 1to 25 nirem to the total body or any organ (except the thyroid; which is limited to < 75 rnrem) over a period of 12 consecutive mronths ;',

b. With the calciilated'disis from the release of radioactive materials in L liquid or gaseous effluents exceeding twice the limits of steps 3.2.3c (Dose), 3.2.4b (Dose- Noble Gases), or3.2.4c (Dose -Iodine-131, Iodine-133, Tritium, and Radioactive Material in Particulate Form) during any calendar quarter, perform the following:

L

Investigate and identify the causes for such release rates;

Define and initiate a program for corrective action; L

Report these actions to the NRC within 30 days from the end of the quarter during which the release occurred.

IF the estimated dose(s) exceeds the limits above, and IF the release condition resulting in violation has not already been corrected prior to violation of 40 CFR 190, THEN include in the report a request for a variance in accordance with the provisions of 40 CER 190 and including L

the specified information of paragraph 190.11(b). Submittal of the report is considered a timely request, and a variance is granted until staff action on the request is complete. The variance only relates to the limits of 40 CFR 190, and does not apply in any way to the requirements for dose limitation of 10 CFR 50, as addressed in other sections of this document.

c. Determine cumulative dose contributions from liquid and-gaseous L effluents in accordance with this document (including steps 3.2.3c [Dose],

3.2.4b [Dose - Noble Gases], or 3.2.4c [Dose - Iodine- 131, Iodine-133, Tritium,.and Radioactive Material in Particulate Form]). L

Refe rec - PMP-6010-OSD-001 Rev. 19 -7 Pie2-f8 Page 23of 84 OFF-SITE DOSE CALCULATION MANUAL BASES -- TOTAL DOSE Tlhi.s specification is provided to meet the dose limitations of 40 CFR 190. The specification requires the preparation and submittal of a Spe~cial Report whenever the calculated doses from plant radioactive effluents exceed twice the design objective doses of Appendix I. For sites containing up to 4 reactors, it is highly unlikely that the resultant dose to a member of the public will exceed the dose limits of 40 CFR 190 if the individual reactors remain within the reporting requirement level. The Special Report will describe a course of action, which should result in the limitations of dose to a member of the public for 12 consecutive months to within the 40 CFR 190 limits. For the purposes of the Special Report, it may be assumed that the dose commitment to any member of the public from other uraniurm fuel cycle sources is negligible with the exceptio a that dose contributions from other nuclear fuel cycle facilities at the same site or within a radius of 5 miles must be considered. If the dose to.pnyuerber of the public is estimated to exceed the requirements of 40 CFR 190, the Special Report with a request for a variance (provided the rele ase conditions resulting in violation of 40 CF;R 19O0 have not already been corrected, in accordance with the provision of 40 CFR 190.1 1), is considered to be a timely request and fulfills the requirements of 40 CFR 190 until NRC staff actiofn is completed.- An individual is not considered a member of the public during any period .in-which helshe is engaged in carrying out any operation, which is part of the nuclear fuel,cyclp. 1 3.3. Calculation of AlarmnTrip Setpoints The alarm and trip setpoints are to provide monitoring, indication, and control of liquid and gaseous effluents. The setpoints are used in conjunction with sampling programs to assure that the releases are kept within the limits of 10 CFR 20, Appendix B, Table 2.

Establish setpoints for liquid and gaseous monitors. Depending on the monitor function, it would be a continuous or batch monitor. The different types of monitors are subject to different setpoint methodologies.

One variable used in setpoint calculations is the multiple release point (MEP) factor. The MRP is a factor used such that when all the releases are integrated, the applicable LIMIT value will not be exceeded. The MRP is determined such that the sum of the MRP's for that effluent type (liquid or gaseous) is less than or equal to 1. The value of the ME? is arbitrary, and it should be assigned based on operational performance. The values of the MRP's for each liquid release point are given in Attachment 3.8, Multiple Release Point:

Factors for Release Points.

The Site stance on instrument uncertainty is taken from BHPOS-223, Consideration of Measurement Uncertainty When Measuring Radiation Levels Approaching Regulatory Limits, which states the NRC position is the result of a valid measurement obtained by a method, which provides a reasonable demonstration of compliance. This value should be accepted and the uncertainty in that measured value need not be considered.

6L Reference I PMP-6010-OSD-001 I Rev. 19 l Page 24 bf .4 OFF-SITE DOSE CALCULATION MANUAL 3.3.1 Liquid Monitors Establish liquid monitor setpoints for each monitor of the liquid effluent release _

systems. A schematic of the liquid effluent release systems is shown as Attachment 3.9, Liquid Effluent Release Systems. A list of the Plant Liquid Effluent Parameters is in Atiacmhment 3.10, Plant Liquid Effluent Parameters. The details of each system design and operation can be found in the system descriptions. The setpoints are intended to keep releases within the limits of 10 CFR 20, Appendix B, Table 2, Colurm 2. Determine setpoints using either the batch or the continuous methodology. ,

a. Liquid Batch Monitor Setpoint Methodology

1. There is only one monitor used on the Waste Disposal System for liquid batch releases. lTis monitor iidentified as RRS-1000. Steam Generator L

Blowdown radiation monitors also can be used to monitor batch releases while draining steam generators. The function of these monitors is to act as a check on the sarnpliUg proigraim The sampling program determines the nuclides and concentratiods of those nuclides prior to release. The discharge and dilution~flov rates are then adjusted to keep the release within the limits of 10CFR 20. Based on the concentrations of nuclides in L the release, the count rate on the monitor can be predicted. The high alarm setpoint can then be set above the predicted value up to the maximum setpoint of the system. L

2. The radioactive concentration of each batch of radioactive liquid waste to be discharged is determined prior to each release by sampling and analysis in accordance with Attachment 3.6, Radioactive Liquid Waste Sampling and Analysis Program.

L

3. The allowable release flow rates are determined in order to keep the release concentrations within the requirements of 10 CFR 20, Appendix B, Table 2, Column 2. The equation to calculate the flow L

L rate is from Addendum AA1 of NUREG-0133:

C LIMIT,

f MRP

<F+f L Where; L Q = the concentration of nuclide "i" in jACi/rnl LIMiT; = the 10 CFR 20, Appendix B, Table 2, Column 2 limit of nuclide "i" in ACi/inl L f = the effluent flow rate in gpm (Attachment 3.10, Plant Liquid Effluent Parameters)

F = the dilution water flow rate as estimated prior to release. The L dilution flow rate is a multiple of 230,000 gpm depending on the number of circulation pumps in operation.

MRP = the multiple release point factor. A factor such that when L

all the release points are operating at one time the limits of 10 CFR 20 will not be exceeded.

L

Reference I PAIP-6010-OSD.001. . l Rev. 19 l Page 25 of 841 L OFF-SITE DOSE CALCULATION MANUAL J

4. This equation must be true during the batch release. Before the release is started, substitute the maximum effluent flow rate and the minimum dilution flow rate for f and F, respectively. If the equation is true, the release can proceed with those flow rates as the limits of operation. If the equation is not true, the effluent flow rate can be reduced or the dilution flow rate can be increased to make the equation true. This equation may be rearranged to solve for the maximum effluent release flow rate (f).

5. The setpoint is used as a quality check on the sampling program The setpoint is used to stop the effluent flow when the monitor reading is greater than the predicted value from the sampling program. The predicted value is generated by converting the effluent concentration for each gamma emitting radionuclide to counts per unit of time as per Attachment 3.11, Volumetric Detection Efficiencies for Principle Gamma Emitting Radionuclides for Eberline Liquid Monitors, or Attachment 3.12, Counting Efficiency Curves for R-19, and R-24.

The sum of all the counts per unit of time is the predicted count rate.

The predicted count rate can then be multiplied by a factor to determine the high alarm setpoint that will provide a high degree of conservatism and eliminate spurious alarms.

b. Liquid Continuous Monitor Setpoint Methodology

1. There are eight monitors used as potential continuous liquid release monitors. These monitors are used in the steam generator blowdown (SGBD), blowdown treatment (BDII), and essential service water (ESW) systems.

2. These Westinghouse monitors (R) are being replaced by Eberline monitors (DRS) and are identified as:

R-19 orDRS 3100/4100 for SGBD

R-24 or DRS 3200/4200 for BDT The function of these monitors is to assure that releases are kept within the concentration limits of 10 CFR 20, Appendix B, Table 2, Column 2, entering the unrestricted area following dilution.

3. The monitors on steam generator blowdown and blowdown treatment systems have trip functions associated with their setpoints. Essential service water monitors.are equipped with an alarm function only and monitor effluent in the event the Containment Spray Heat Exchangers are used.

I.-

Reference ' I ' PMP-6010-OSD-001 Rev. 19 l Page 26 of 84 OFF-SITE DOSE CALCULATION MANUAL

4. The equation used to determine the setpoint for continuous monitors is from Addendum AA1 of NUREG-0133:

SW eC*Eff*MRP*F*SF, P ~f.

Where; Sp = setpoint of monitor (cpm)

C = 5E-7 90Ci/ml, maximum effluent control limit from 10 CFR 20, Appendix B, Table 2; Columnn 2 of a known possible nuclide in effluent stream. (The limiting nuclide shall be evaluated annually-by.rqviewing current nuclides against historical ones in ordck tb determine if one with a more restrictive effluent. concentration limit than Si90 is found. The concentrationiiifiit. shall be adjusted

,,i+appropately.) , .

-OR-if a mixture is to be specified, X z C, C,

ULMIT, Eff = Efficiency, this information is located in Attachment 3.1 1, L

Volumetric Detection Efficiencies for Principle Gamma Emitting Radionuclides for Eberline Liquid Monitors, through Attachment 3.13, Counting Efficiency Curve for R-20, and R-28, for the specific monitors. For Eberline monitors the efficiency is nuclide specific and the calculation changes slightly to:

ZICfC,*Effi) C*C

Eff f

L L

d replaces L

LIMIT; MRP = multiple release point factor. A factor such that when all the release points are operating at one time the limits of L

10 CER 20 will not be exceeded (Attachment 3.8, L Multiple Release Point Factors for Release Points). The MRP for ESW monitors is set to 1.

F = dilution water (circ water) flow rate in gpm obtained from Attachment 3.10, Plant Liquid Effluent Parameters. For' routine operation, the setpoint should be calculated using the minimum dilution flow rate of 230,000 gpm.

SF = Safety Factor, 0.9.

f = applicable effluent release flow rate in gpm. For routine operation, the setpoint should be calculated using maximum effluent flow rate (Attachment 3.10, Plant Liquid Effluent Parameters).

a Reference I " PINIP,6010-OSDp001 _ Rev. 19 l Page 27 of 84 L OFF-SITE DOSE CALCULATION MANVAL J 3.3.2 Gaseous Monitors For the purpose of implementing Step 3.2.2, Radioactive Gaseous Effluent Monitoring Instrumentation, and Substep 3.2.4a, Dose Rate, the alarm

setpoints for gaseous effluents released into unrestricted areas will be established using the following methodology. In addition, the above steps do not apply to instantaneous alarm and trip setpoints for integrating radiation monitors sampling radioiodines, radioactive materials in particulate form and radionuclides other than noble gases. A schematic of the gaseous effluent release systems is presented in Attachment 3.14, Gaseous Effluent Release Systems. Attachment 3.15,Plant Gaseous Effluent Parameters, presents the effluent flow rate parameter(s).

L Gaseous effluent monitor high alarm setpoints will routinely be established at a fraction of the maximum allowable setpoint (typically 10% of the setpoint) for

'S..

ALARA purposes. Alert alarms will normally be set to provide adequate indications of small changes in radiological conditions.

a. Plant Unit Vent

L I 1. The gaseous effluents discharged from the plant vent will be monitored by the plant vent radiation monitor low range noble gas channel [Tag No. VRS- 1505 (Unit 1), VRS-2505 (Unit 2)] to assure that applicable alarms and trip actions (isolation of gaseous release) will occur prior to exceeding the limits in step 3.2.4, Gaseous Effluents. The alarm setpoint values will be established using the following unit analysis equation:

SF *MRP*DLj Sp = . , _ .

Fp*7/Q*Z(W,*DCFu)

Where; Sp = the maximum setpoint of the monitor in p.Cifcc for release point p, based on the most limiting organ SF = an administrative operation safety factor, less than 1.0 MRP = a weighted multiple release point factor (< 1.0), such that when all site gaseous releases are integrated, the applicable dose will not be exceeded based on the release rate of each effluent point.

The MRP is an arbitraty value based on the ratio of the release rate or the volumetric flow rate of each effluent point to the total respective flow rate value of the plant and will be consistent with past operational experience. The MRP is computed as follows:

Compute the average release rate, Qp, (or the volumetric flow rate, fQ) from each release point p.

Compute ZQp (or Z1p) for all release points.

Ratio Qp/ZQp (or fp/Efp) for each release point. This ratio is the MRP for that specific release point

Repeat the above bullets for each of the site's eight gaseous release points.

Reference PMP-601DoS-OO1 I Rev. 19 l Page.28 of 84 OFF-SITE DOSE CALCULATION MANUAL Fp = the maximum volumetric flow rate of release point "p", at the time of the release, incc/sec. The maximum Unit Vent flow rate, by design, is 186,600 cfin for Unit 1 and 143,400 cfn for Unit 2.

DLj = dose rate limit to organ 'j" in an unrestricted area (mrem/yr).

Based on continuous releases, the dose rate limits, DLj, from step 3.2.4a, Dose Rate, are as follows:

Total Body* 500mremfyear

Skin

3000 mremlyear

Any Organ< 1500 mreminyear

X/2 = The worst case annual average relative cIncentration in the applicable sector or area, in sec/r 3 (see Attachment 3.16, 10 Year Average of 1995-2004 Data), ii

- WI weighted factor for the radionuclide: - .-

,W W= SC, E Ck L Where,

= concentration of the most abundant radionuclide 'T' Ck = total concentration of all identified L

radionuclides in that release pathway. For batch releases, this value may be set to I for conservatism DCFj = dose conversion factor used to relate radiation dose to organ '7", from exposure to radionuclide "i" in mrem m' / yr jxCi. See following equations.

The dose conversion factor, DCFjj, is dependent upon the organ of concern.

L For the whole body: DCF.j =

Where;

= whole body dose factor due to gamma L

emissions for each identified noble gas L radionuclide in mrem m3 / yr jiCi See Attachment 3.18, Dose Factors. L L

= Reference l MPRP-6O1SD:0O1. lT Rev. 19 l Page 29 of 84 OFF-SITE DOSE CALCULATION MANUAL For the skin: DCFjj = L; + LA.M; Where; I, = skin dose factor due to beta emissions for each identified noble gas radionuclide, in mrem m3 / yr pCi. See Attachment 3.18, Dose Factors.

1.1 = the ratio of tissue to air absorption coefficient over the energy range of photons of interest.

This ratio converts absorbed dose (mrad) to dose equivalent (mrem).

M; = the air dose factor due to garnma emissions for each identified noble gas radionuclide in mrad rn3 / yr pCi. See Attachment 3.18, Dose Factors..

For the thyroid, via inhalation: DCFj = Pi Where; Pi = the dose parameter, for radionuclides other than noble gas, for the inhalation pathway in mrem. / yr fLCi (and the food and ground path, as appropriate).

See Attachment 3.18, Dose Factors.

2. The plant vent radiation monitor low range noble gas high alarm channel setpoint, Sp, will be set such that the dose rate in unrestricted areas to the whole body, skin and thyroid (or any other organ),

whichever is most limiting, will be less than or equal to 500 mremTyr, 3000 mremlyr, and 1500 mremlyr respectively.

3. The thyroid dose is limited to the inhalation pathway only.

4. The plant vent radiation monitor low range noble gas setpoint, Sp, will be recomputed whenever gaseous releases like Containment Purge, Gas Decay Tanks and CVCS HUTs are discharged through the plant.

vent to determine the most limiting organ.

5. The high alarm setpoint, Sp, may be established at a lower value than the lowest computed value via the setpoint equation.

6. Containment Pressure Reliefs will not have a recomputed high alarm setpoint, but will use the normal high alarm setpoint due to their randomness and the time constraints involved in recomputation.

7. At certain times, it may be desirable to increase the high alarm setpoint, if the vent flow rate is decreased. This may be accomplished in one of two ways.

Max Conc ( /Cilcc)

Max Flowrate(cfin) = New Max cfin New Max Concentration(,uCi/cc)

-OR-

Reference l PMP-6010-OSD-01. . Rev. 19 l Page 30 of 84 OFF-SITE DOSE CALCULATION MANUAL L

Max Conc (,uCi/cc)

Max Flowrate(cfin) = New Maxu-ilcc New Max Flowrate (cfin) L

b. Waste Gas Storage Tanks

1. The gaseous effluents discharged from the Waste Gas System are monitored by the vent stack monitors VRS-1505 and VRS-2505.

L

2. In the event of a high radiation alarm, an automatic termination of the release from the waste gas system will be initiated from the plant vent radiation monitor low range noble gas channel (VRS-1505 or VRS-2505). Therefore, for any gaseous release configuration, which includes normal operation and waste gas system gaseous discharges, the alarm setpoint of the plant vent radiation monitor will be recomputed to determine the most limiting organ based on all gaseous effluent source terms.

I-Chemical and Volume Control System Hold Up Tanks (CVCS

, -, HUT), containing high gaseous oxygen concentrations, may be released unider the guidance of waste gas storage tank utilizing approved Operations' procedures.

L

3. It is normally prudent to allow 45 days of decay prior to releasing a Gas Decay Tank (GDT). There are extenuating, operational circumstances that may prevent this from occurring. Under these L

circumstances, such as high oxygen concentration creating a combustible atmosphere, it is prudent to waive the 45-day decay for L c.

safety's sake.

Containment Purge and Exhaust System L

1. The gaseous effluents discharged by the Containmen~t Purge and Exhaust Systems and Instrumentation Room Purge and Exhaust System are monitored by the plant vent radiation monitor noble gas L

channels (VRS-1505 forUnit 1, VRS-2505 forUnit 2); and alarms and trip actions will occur prior to exceeding the limits in step 3.2.4a, Dose Rate.

L

2. For the Containment System, a continuous air sample from the containment atmosphere is drawn through a closed, sealed system to L the radiation monitors (Tag No. ERS-1300/1400 for Unit 1 and ERS-2300/2400 for Unit 2). During purges, these monitor setpoints will give a Purge and Exhaust Isolation signal upon actuation of high L alarmnsetpoints for particulate and noble gas channels. The sample is then returned to containment. Grab sample analysis is performed for a Containment purge before release. L.

3. The Upper Containment area is monitored by normal range area gamma monitors (Tag No. VRS-I 101/1201 for Unit 1 and VRS-2101/2201 for Unit 2), which also give Purge and Exhaust Isolation Trip signals upon actuation of their high alarm.

Reference l PMP-6010-OSD0-51 -7 Rev. 19 l Page 31 of 84 OFF-SITE DOSE CALCULATION MANUAL

4. For the Containment Pressure Relief System, no sample is routinely taken prior to release, but a sample is obtained twice per month

5. The containment airborne and area monitors, upon actuation of their high alarm, will automatically initiate closure of the Containment and Instrument Room purge supply and exhaust duct ialves and containment pressure relief system valves. Complete trip of all isolation control devices requires high alarm of one of the two Train A monitors (ERS-1300f2300 or VRS-1 101/2101) and one of the two Train B monitors (ERS-1400t2400 orVRS-1201/2201).

d. Steam Jet Air Ejector System (SJAE)

1. The gaseous effluents from the Steam Jet Air Ejector System discharged to the environment are continuously monitored by radiation monitor (Tag No. SRA-l 900 for Unit 1 and SRA-2900 for Unit 2). The monitor will alarm prior to exceeding the limits of step 3.2.4a, Dose Rate. The alarm setpoint for the Condenser Air Ejector System monitor will be based on the maximum air ejector exhaust flow rate, (Attachment 3.15, Plant Gaseous Effluent Parameters).

The alarm setpoint value will be established using the following unit analysis, equation:

SF*MRP*DL1 Fp*/Q**(Wt,*DCFv,)

Where; SsIAE = the maximum setpoint, based on the most limiting organ, in tCi/cc and where the other terms are as previously defined

e. Gland Seal Condenser Exhaust

1. The gaseous effluents firn the Gland Seal Condenser Exhaust discharged to the environment are continuously monitored by radiation monitor (Tag No. SRA-1800 for Unit I and SRA-2800 for Unit 2). The radiation monitor will alarm prior to exceeding the limits of step 3.2.4a, Dose Rate. The alarm setpoint for the GSCE monitor will be based on the maximum condenser exhaust flow rate (1260 CFM for Unit 1, 2754 CFM each for the two Unit 2 vents). The alarm setpoint value will be established using the following unit analysis equation:

SF *MRP

DLj Sasc-r Fp*/Q *(W.* DCF#)

Where;

Reference I PMP-6010-SD-O01 I -Rev. 19 l Page 32 of 84 OF7-SITE DOSE CALCULATION MANUAL SscE = the maximum setpoint, based on the most limiting organ, in gxCi/cc and where the other terms are as previously defined 3.4 Radioactive Effluents Total Dose 3.4.1 The cumulative dose contributions from liquid and gaseous effluents will be determined by summing the cumulative doses as derived in steps 3.2.3c (Dose),

3.2.4b (Dose - Noble Gases), and 3.2.4c (Dose -Iodine-131, Iodine-133, Tritium, and Radioactive Material in Particulate Form) of this procedure. Dose contribution from direct radiation exposure will be based on the results of the direct radiation monitoring devices located at the REMP monitoring stations.

See NUREG-0133, section 3.8.

3.5 Radiological Environmental Monitoring Program (REMP) 3.5.1 Purpose of the REMP *

a. The purpose of the REM? is to:

Establish baseline radiation and radioactivity concentrations in the L environs prior to reactor operations,

Monitor critical environmental exposure pathways, L Determine the radiological ipact, if any, caused by the operation of the Donald C. Cook Nuclear Plant upon the local environment. -

b. The first purpose of the REMP was completed prior to the initial operation of either of the two nuclear units at the Donald C. Cook Nuclear Plant Site.

The second and third purposes of the REMP are an on-going operation and as such various environmental media and exposure pathways are examined. The various pathways and sample media used are delineated in Attachment 3.19, Radiological Environmental Monitoring Program Sample Stations, Sample Types, Sample Frequencies. Included is a list of the sample media, analysis required, sample stations, and frequepcy requirements for both collection and analysis. Attachment 3.19, Radiological Environmental Monitoring Program Sample Stations, Sample Types, Sample Frequencies, defines the scope of the REMP for L

the Donald C Cook Nuclear Plant.

3.5.2 Conduct of the REMP ERef. 5.2.1u]

a. Conduct sample collection and analysis for the REMP in accordance with Attachment 3.19, Radiological Environmental Monitoring Program Sample Stations, Sample Types, Sample Frequencie's, Attachment 3.20, Maximum Values for Lower Limits of Detections" - REMP, and Attachment 3.21, Reporting Levels for Radioactivity Concentrations in Environmental Samples. These are applicable at all times. The on-site monitoring locations are shown on Attachment 3.22, On-Site Monitoring Location - REMP, and the off-site monitoring locations are shown on Attachment 3.23, Off-Site Monitoring Locations - REMP.

l= Reference T PMP-6D01-1OSD-0O1. 0 Rev. 19 7 Page 33 of 841 L OFF-SITE DOSE CALCULATION MANUAL J

1. Perform each surveillance requirement within the specified time interval in Attachment 3.19, Radiological Environmental Monitoring

- -Program Sample Stations, Sample Types, Sample Frequencies, with a maximum allowable extension not to exceed 25% of the surveillance interval.

2. If an environmental sample cannot be collected in accordance with step 3.5.2a, submit a description of the reasons for deviation and the actions taken to prevent a reoccurrence as part of the Annual Radiological Environmental Operating Report (AREOR).

Deviations from the required sampling schedule are permitted if specimens are unobtainable due to hazardous conditions, seasonal unavailability, or malfunction of automatic sampling equipment. If the deviation from the required sampling schedule is due to the malfunction of automatic sampling equipment, make every:effort to complete the corrective action prior to the end of the next sampling period.

3. If a radionuclide is detected in any sample medium exceeding the limit established in Attachment 3.21, Reporting Levels for Radioactivity Concentrations in Environmental Samples, or if more than one radionuclide is detected in any sample medium and the Total Fractional Level (TFL), when averaged over the calendar quarter, is greater than or equal to 1,based on the following formula:

TFL= -Ca')+ C(2) +... 1 4,1) 42)

Where; C(,) = Concentration of 1 detected nuclide C(2) = Concentration of 2'd detected nuclide 1<1) = Reporting Level of 1" nuclide from Attachment 3.21, Reporting Levels for Radioactivity Concentrations in Environmental Samples.

L() = Reporting Level of 2 ndnuclide from Attachment 3.21, Reporting Levels for Radioactivity Concentrations in Environmental Samples.

And, if the activity is the result of plant effluents, evaluate the release conditions, environmental factors, or other aspects, which may have contributed to the identified levels for inclusion in the AREOR. If the radioactivity was not a result of plant effluents, describe the results in the ARBOR.

4. If a currently sampled milk farm location becomes unavailable, conduct a special milk farm survey within 15 days.

a) If the unavailable location was an indicator farm, an alternate sample location may be established within eight miles of the Donald C. Cook Nuclear Plant, if one is available.

Reference PMP-6010-OSD-001 Rev. 19 - Page 34 of 84 OFF-SITE DOSE CALCULATION MANUAL b) If the unavailable location was a background farm, an alternate sample location may be established greater than 15 but less than 25 miles of the Donald C. Cook Nuclear Plant in one of the less prevalent wind direction sectors, if one is available.

c) If a replacement farm is unobtainable and the total number of indicator farms is less than three or the background farms is less than one, perform monthly vegetation sampling in lieu of milk sampling.

BASES - RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM (REMP)

The REM? provides measurements of radiation and radioactive materials in those exposure pathways and for those radionuclides, which lead to the highest potential radiation exposures of individuals resulting from the station operation. Thereby, this monitoring program

supplements the radiological effluent monitoring program by verifying the measurable concentration of radioactive materials, aid levels of radiation are not higher than expected on the basis of the effluent measurements annd modeling of the environmental exposure pathways.

The initially specified REM? was effective for the first three years of commercial operation. L Program changes may be initiated based on operational experience in accordance with the requirements of Technical Specification 5.5.1 c.

The detection capabilities, required by Attachment 3.20, Maximum Values for Lower Limits L of Detections-B - REMP, are the state-of-the-art for routine environmental measurements in industrial laboratories.

It should be recognized that the LELD is defined as a priori (before the fact) limit representing the capability of a measurement system and not as a posteriori (after the fact) limit for a L

particular measurement. Analyses shall be performed in such a manner that the stated LLDs will be achieved under routine analysis conditions. Occasionally, background fluctuations, unavoidably small sample sizes, the presence of interfering nuclides, or other uncontrollable L

circumstances may render these LLDs unachievable. In such cases, the contributing factors will be identified and described in the Annual Radiological Environmental Operating Report. L 3.5.3 Annual Land Use Census (Ref. 5.2.1u]

a. Conduct a land use census and identify the location of the nearest milk animal, the nearest residence and the nearest garden of greater than 500 L

square feet producing fresh leafy vegetables in each of the ten land sectors within a distance of five miles.

L

b. In lieu of the garden census, grape and broad leaf vegetation sampling may be performed as close to the site boundary as possible in a land sector, containing sample media, with the highest average deposition factor (D/Q) value. L

c. Conduct this land use census annually between the dates of June 1 and October 1 by door-to-door survey, aerial survey, or by consulting local agricultural authorities. L

. Reference l PMP-6010-OSD-001 7 Rev. 19 l Page 35 of 84]

L OFF-SITE DOSE CALCULATION MANUAL j

1. With a land use census identifying a location(s), which yields a calculated dose or dose commitment greater than the values currently being calculated in this document, make appropriate changes to incorporate the new.location(s) within 30 days, if possible.

BASES-LAND USE CENSUS This is provided to ensure changes in the use of unrestricted areas are identified and modifications to the monitoring program are made in accordance with requirements of TS 6.8.4b, if required by the results of the census. This census satisfies the requirements of Section lV.B.3 of Appendix I to 10 CFR Part 50. Restricting the census to gardens of greater than 500 square feet provides assurance that significant exposure pathways via leafy vegetables will be identified and monito red since a garden of this size is the minimum required to produce the quantity (25 kg/yr) of leafy vegetables assumed in Regulatory Guide 1.109 for consumption of a child. To determine this ronTimum garden size, the following assumptions were used: 1) that 20% of the garden was used for growing broad leaf vegetation (that is,.similar to lettuce and cabbage), and 2) a vegetation field of 2 kg/square meter.

3.5.4 Interlaboratory Comparison PSagram

a. In order to comply with Reg. Guides 4.1 and 4.15, the analytical vendor participates in an Interlaboratory Comparison Program, for radioactive materials.

Address program results and identified deficiencies in the AREOR

1. With analyses not being performed as required above, report the corrective actions taken to prevent a recurrence to the Commission in the AREOR.

BASES - INTERLABORATORY COMPARISON PROGRAM The requirement fbr participation in an Interlaboratory Comparison Program is provided to ensure independent checks on the precision and accuracy of the measurements of radioactive material in environmental sample matrices are performed as part of the quality assurance progrmun for environmental monitoring in order to demonstrate the results are reasonably valid.

3.6 Radioactive Equipment Storage Facility (Mausoleun) Groundwater Monitoring Program 3.6.1 Purpose of the Radioactive Equipment Storage Facility (Mausoleum) Groundwater Radiological Monitoring Program

a. The purpose of the temporary on-site Radioactive Equipment Storage Facility (Mausoleiim) Radiological Monitoring Program was to establish baseline radiological data for the groundwater surrounding the facility prior to the storage of the Unit 2 Steam Generator Lower Asserfiblies. Thereafter, the purpose is to monitor the groundwater through observation wells with locations as shown in Attachment 3.22, On-Site Monitoring Location - REM?, to determine the radiological impact, if any, caused by the use of the Storage Facility.

Reference - PMP-6010-OSD-001 Rev. 19 1 Pa e 36f 84 . .

OFF-SITE DOSE CALCULATION MANUAL 3.6.2 Conduct of the Radioactive Equipment Storage Facility (Mausoleum) Groundwater Radiological Monitoring Program

a. Collect and analyze groundwater samples in accordance with Attachmlent 3.19, Radiological Environmental Monitoring Program Sample Stations, Sample Types, Sample Frequencies. Apply the values from Attachment 3.20, Maximum Values for Lower Limits of Detections" - REMP, (excluding I-131) and Attachment 3.21, Reporting Levels for Radioactivity Concentrations in Envirornental Samples, (excluding 1-13 1).

3.7 Meteorological Model 3.7.1 Three towers are used to determine therneteorological conditions at Donald C CookNuclearPlant. One ofthe-tow'er'sis located it the Lake Michigan shoreline to determine the meteorological parameters associated with unmodified shoreline air.

The data is accumulated by ricroirocessors at the tower sites and normally transferred to the central computer Fve~ry 15 minutes.

3.7.2 The central computer uses a meteorological software program to provide atmospheric dispersion and deposition.p.arameters. The meteorological model used L is based on guidance provided in Reg. Guide 1.111 for routine releases. ADL calculations use the Gaussian plume model.

3.8 Reporting Requirements X 3.8.1 Ainual Radiological Environmental Operating Report (AREOR)

a. Submit routine radiological environmental operating reports covering the operation of the units during the previous calendar year prior to May 1 of each L

year.

b. Include in the AREOR: L

Summaries, interpretations, and statistical evaluation of the results of the radiological environmental surveillance activities for the reporting period.

L

A comparison with pre-operational studies, operational controls (as appropriate), and previous environmental surveillance reports and an assessment of the observed impacts of the plant operation on the environment.

L

The results of the land use censuses required by step 3.5.3, Annual Land Use Census.

If harmful effects or evidence of irreversible damage are detected by the monitoring, provide in the report an analysis of the problem and a planned 4 course of action to alleviate the problem.

I.-

Referente I PMP-6010-OSD-001 I Rev. 19 - Page.37 of 84 .;

OFF-SITE DOSE CALCULATION MANUAL

Summarized and tabulated results of all radiological environmental samples taken during the reporting period. In the event that some results are not available for inclusion with the report, submit the report noting and explaining the reasons for the missing results. Submit the missing data as soon as possible in a supplementary report.

A summary description of the REMP including sampling methods for each sample type, size and physical characteristics of each sample type, sample preparation methods, analytical methods, and measuring equipment used.

A map of all sample locations keyed to a table giving distances and directions from one reactor.

The results of participation in the Interlaboratory Comparison Program required by step 3.5.4, Interlaboratory Comparison Program.

3.8.2 Annual Radiological Effluent Release Report (ARERR).

a. Submit routine ARERR covering tOeopefation of the unit during the previous 12 months of operation within 9.0 ddays after January 1of each year.

b. Include in the ARERR a summaryo fthe:quantities of radioactive liquid and gaseous effluents and solid waste released from the units as outlined in Reg. Guide 1.21, '¶Measuring, Evaluating and Reporting in Solid Wastes and Releases of Radioactive Materials in Liquid and Gaseous Effluents from Light-Water Cooled Nuclear Power Plants," with data summarized on a quarterly basis following, the format of Appendix B, thereof.

c. Submit in the ARERR 90 days after January 1 of each year and include a quarterly summary of hourly meteorological data collected during the reporting period.

This summary may be in the form of anhour-by-hour listing of wind speed, wind direction, atmospheric stability, and precipitation (if measured) on magnetic tape, or in the form ofjoint frequency distributions of wind speed, wind direction and atmospheric stability.

Include an assessment of the radiation doses due to the radioactive liquid and gaseous effluents released from the unit or station during the previous calendar year.

Include an assessment of the radiation doses from radioactive liquid and gaseous effluents to members of the public due to their activities inside the site boundary during the reporting period. Include all assumptions used in making these assessments (that is, specific activity, exposure time and location) in these reports.

Use the meteorological conditions concurrent with the time of release of radioactive materials in gaseous effluents (as determined by sampling frequency and measurement) for determining the gaseous pathway doses.

Inoperable radiation monitor periods exceeding 30 continuous days; explain causes of inoperability and actions taken to prevent reoccurrence.

Reference PMP-6010-OSD-001 Rev. 19 Page 38 of 84 OFF-SITE DOSE CALCULATION MANUAL

d. Submit the ARERR tRef. 5.2.1w] 90 days after January 1 of each year and , . . L-I.

include an assessment of radiation doses to the likely most exposed

nmember of the public from reactor releases and other nearby uranium fuel cycle sources (including doses from primary effluent pathways and direct radiation) for the previous 12 consecutive months to show conformance with 40 CFR 190, Environmental Radiation Protection Standards for I-Nuclear Power Operation. Acceptable methods for calculating the dose contnibution from liquid and gaseous effluents are given in Reg. Guide 1.109, Rev.1.

e. Include in the ARERR the following information for each type of solid waste shipped off-site during the report period:

Volume (cubic meters),

Total curie quantity (specify whether determined by measurement or estimate),

Principle radionuclides (specify whether determined by measurement or estimate),

Type of waste (example: spent resin, compacted dry waste, evaporator bottoms),

L

Type of container (example: LSA, Type A, Type B, Large Quantity),

-A4ND- IL

Solidification agent (example: cement).

f. Include in the ARERR unplanned releases of radioactive materials in gaseous and liquid effluent from the site to unrestricted areas on a L g.

quarterly basis.

Include in the ARERR any change to this procedure made during the L

reporting period. L L

h Due to the site having shared gaseous and liquid waste systems dose calculations will be performed on a per site bases using the per unit values.

This is ALARA and will ensure compliance with 40 CFR 141, National Primary Drinking Water Regulations. Unit specific values are site values L divided by two.

3.9 10 CFR 50.75 (g) Implementation 3.9.1 Records of spills or other unusual occurrences involving the spread of contamination in and around the site. These records may be limited to instances when significant contamination remains after decontamination or when there is a reasonable likelihood that contaminants may have spread to inaccessible areas, as in the case of possible seepages.

3.9.2 These records shall include any known information or identification of involved nuclides, quantities, and concentrations.

Reference 7 -PrIP-6010-OSDO001 Rev. 19 Page 39 of 84 OFF-SITE DOSE CALCULATION MANUAL 3.9.3 This information is necessary to ensure all areas outside the radiological-restricted area are documented for surveying and rernediation during decommissioning.

There is a retention schedule file number where this information is filed in Nuclear Documents Management to ensure all required areas are listed to prevent their omission.

3.10 Reporting/Management Review 3.10.1 Incorporate any changes to this procedure in the ARERR.

3.10.2 Update this procedure when the Radiation Monitoring System, its instruments, or the specifications of instruments are changed.

3.10.3 Review or revise this procedure as appropriate based on the results of the land ttse census and REMP.

3.10.4 Evaluate any changes to this procedure for potential impact on other related Department Procedures.

I 3.10.5 Review this procedure during the first quarter of each year and update it if necessary. Review Attachment 3.16, 10 Year Average:of 1995-2004 Data, and document using Attachment 3.17, Annual Evaluation of x/Q and D/Q Values For All Sectors. The X IQ and D IQ values will be evaluated to ensure all data is within +/- 3 standard deviations of the 10 year annual average data and documented by completing Attachment 3.17, Annual Evaluation of x/Q and D/Q Values For All Sectors, and filed in accordance with the retention schedule.

4 FINAL CONDITIONS 4.1 None.

Reference I PMP*6010-OSDOOI , Rev. 19 1 Pa e 40 of 84 OFF-SITE DOSE CALCULATION MANUAL 5 REFERENCES I .,

5.1 Use

References:

5.1.1 'Implementation of Programmatic Controls for Radiological Effluent Technical Specifications in the Administrative Controls Section of the Technical Specifications and the Relocation of Procedural Details of RETS to the Off-Site Dose Calculation Manual or to the Process Control Program (Generic Letter 89-01)", United States Nuciear Regulatory Comniission, January 31, 1989 5.1.2 12-THP-6010-RPP-601, Preparation of the Annual Radioactive Effluent Release Report 5.1.3 12-THP-6010-RPP-639, Annual Radiological Environmental Operating

- gReport (AREOR) Preparation And Submittal 5.2 Writing

References:

5.2.1 Source

References:

L

. a. 10 CFR 20, Standards for Protection Against Radiation

b. 10 CFR 50, Domestic Licensing of Production and Utilization Facilities L

c. PMI-6010, Radiation Protection Plan

d. NUREG-0472 e.

f.

NUREG-0133 Regulatory Guide 1.109, non-listed parameters are taken from these data L

tables

g. Regulatory Guide,.III L

h. Regulatory Guide 1.113

i. Final Safety Analysis Report (FSAR) j.

k.

Technical Specifications 5.4.1.e, 5.5.1.c, 5.5.3, 5.6.2, and 5.6.3 Final Environmental Statement Donald. C Cook Nuclear Plant, August L

1973

1. NUREG-0017 L

n. ODCM Setpoints for Liquid [and Gaseous] Effluent Monitors (Bases),

ENGR 107-04 8112.1 Environs Rad Monitor System

n. HPPOS-223, Consideration of Measurement Uncertainty When Measuring Radiation Levels Approaching Regulatory Limits L

o. Watts - Bar Jones (WBJ) Document, R-86-C-001, The Primary Calibration of Eberline Instrument Corporation SPING - 3/4 Low, Mid, L and High Range Noble Gas Detectors

p. WBJ Document, R-86-C-003, The Primary Calibration of Eberline Instrument Corporation DAM-4 and Water Monitor

q. 40 CFR 190, Environmental Radiation Protection Standards for Nuclear Power Operations

r. NRC Commitment 6309 (N94083 dated 11/10/94)

s. NRC Commitment 1151

t. NRC Commitment 1217

u. NRC Commitment 3240 L.

Reference PMP-6010:OSD.OO1 Rev. 19 Page 41 of 84 OFF-SITE DOSE CALCULATION MANUAL

v. NRC Commitment 3850

w. NRC Commitment 4859

x. NRC Commitment 6442

y. NRC Commitment 3768

z. DIT-B-00277-00, HVAC Systems Design Flows aa. Regulatory Guide 1.21 bb. Regulatory Guide 4.1 cc. 1-2-V3-02-Calc #4, Unit Vent Sample Flow rate for isokinetic particulates and Iodine sampling ddL BPS N13.30-1996, Appendix A Rationale for Methodsbf Determining Minimum Detectable Amount (MDA) and Minimum Testing Level (MDL ee. DIT-B-01971-00, Dose Factors for Radioactive Particulate Gaseous Effluents Associated with the Child by the Inhalation Pathway

.. r -: ff. DlT-B-01987-00, Ground Plane & Food Dose Factors Pi for ;

Radioiodines and Radioactive Particulate Gaseous Effluents gg. NRC Commitment 1010 J 5.2.2 General References

a. Cook Nuclear Plant Start-Up Flash Tank Flow Rate letter from D. L.

Boston dated January 21, 1997

b. Letter from B.P. Lauzau, Venting of Middle CVCS Hold-Up Tank Directly to Unit Vent, May 1, 1992 j c. AEP Design Information Transmittal on Aux Building Ventilation Systems

d. PMP-4030.EIS.001, Event-Initiated Surveillance Testing

e. Environmental Position Paper, Fe Impact on Release Rates, approved 3/14/00

f. Environmental Position Paper, Methodology Change from Sampling Secondary System Gaseous Effluents for Power Changes Exceeding 15%

within 1 hr to Responding to Gaseous Alert Alarms, approved 4/4/00

g. CR 02150078, RRS-1000 efficiency curve usage h Environmental Position Paper, Unit Vent Compensatory Sampling, approved 4/14/05

I Reference PMP-6010-OSD-O01 Rev. 19 Page 42 of 84 OFF-S1TE DOSE CALCULATION MANUAL I.

Attachment 3.1 I Dose Factors for Various Pathways Pages:

42 -45 R. Dose Factors L

PATHWAY Nudlide Ground Vegetable Meat CowMilk Goat Milk Inhalation L H-3 O.OE+O0 4.0E+03 3.3E+02 2.4E+03 4.9E+03 1.3E+03 C-14 O.OE+00 3.5E+06 5.3E+05 3.2E+06 3.2E+06 3.6E+04 Cr-51 5.4E+06 1.11+07 1.5E+06 6.9E+06 8.3E+05 2.1E+04 Mn-54 1.6E+09 9.4E+08 2. 1E+07 2.9E+07 3.5E+06 2.OE+06 Fe-59 3.2E+08 9.6E+08 1.7E+09 3.IE+08 4.OE+07 1.5E+06 Co-58. -.4.4E-+08 6.OE+08 2.9E+08 8.4E+07 1.OE+07 1.3E+206 Co-60 2.5E+10 3.2E+O9 1.OE+09 2.7E+08 3.2E+07 8.6B+06 Zn-6S 8.5E+08 2.7E+09 9.5E+08 1.6E+10 1.9E+09 1.2E+06 Sr-89 2.52+04 3.5E+10 3.8B+08 9.9E+09 2.1E+10 2.4i+06 i-Sr-90 O.OE+OO 1.4E+12 9.6B+09 9.4E+10 2.0B+l L.E+08 .4 Zr-95 . 2.9E+08 1.2E+09 1.5E+09 9.3E+05 1.2E+05 2.7E+06 Sb-124 I-131

.6.9E+08 1.02+07 3.02+09 2.4E+10 4.4E+08 2.5E+09 7.2E+08 4.8E+1 1 8.61+07 5.8E+11 3.8E+06 1.6E+07 L

1-133 1.5E+06 4.OE+08 6.0E+01 4.4E+09 5.3B+09 3.8B+06 Cs-134 Cs-136 7.9E+09 1.7E+08 2.5E+10 2.22+08 1.2E+09 4.2E+07 5.OE+10 5.1E+09 1.5E+1 1.5E+10 l.1E+06 1.9E+05 L Cs-137 1.2E+10 2.5E+10 1.OE+09 4.5E+10 1.4E+11 9.OE+05 Ba-140 Ce-141 Ce-144 2.3E+07 1.5E+07 7.92+07 2.7E+08 5.3E+08 1.32+10 5.2E+07 3.0E+07 3.6E+08 2.1E+08 8.3E+07 7.3E+08 2.6_+07 l.02+07 8.7E+07 2.0E+06 6.1E+05 1.32+07 L

Units for all except inhalatim pathway are m2r sec / yr pCi, inhalaticm pathway uits are mr m 3

I y tCi. L U,, Values to.be Used For the Maximum Exposed Individual L Pathway Infant Child Teen Adult Fruits, vegetables and grain Qcg/yr)

Leafy vegetables (kglyr) 520 26 630 42 520 64 L

Milc (IJyr)

Meat and poultry (kgfyr) 330 330 41 400 65 310 110 L

Fish (kg/yr) - 6.9 - 16 21 Drinking water (Uyr) 330 510 510 730 L Shoreline recreation (hr/yr) - 14 67 12 Inhafation (n 3 /yr) 1400 3700 8000 8000 Table E-5 of Reg. Guide 1.109.

L

Reference PMP-6010-OSD.001 Rev. 19 Page 43 of 841 OFF-SITE DOSE CALCULATION MANUAL .1 Dose Factors for Various Pathways' l Pages:

Bip Factors for Aquatic Foods pCiI / kg pCi Element Fish Invertebrate H 9.0E-1 9.0E-1 C 4.6E3 9.1E3 Na 1.0E2 2.0E2 P 1.0E5 2.0E4 Cr 2.0E2 2.0E3 Mn, 4.0E2 9.0E4 Fe 1.0E2 3.2E3

Co 5.0E1 2.0E2 .

I

.Ni 1.0E2 1.0E2 . I .

, Cu 5.0E1 4.0E2

___Zn_*2.0E3 1.0E4 I ., -C.:

Br 4.2E2 3.3E2 Rb 2.0E3 1.0E3 Sr 3.0E1 1.0E2 Y 2.5E1 1.0E3 Zr 3.3E0 6.7E0 Nb 3.0E4 1.0E2 Mo 1.0E1 1.01 Tc 1.5E1 5.0E0 Ru 1.0E1 3.0E2 Rh 1.0EI 3.0E2 Te 4.01E2 6.1E3 I 1.5E31 5.0E1 Cs 2.0E33 1.01E3 Ba 4.OEO 2.0E2 La 2.51E1 1.0E3 Ce 1.0E0 1.0E3 Pr 2.5E1 1.013 Nd 2.5E1 1.0E3 W 1.2E3 1.01E Np 1.0E1 4.0E2 Table A- of Reg. Guiide 1.109.

Reference I PMP-6010-OSD-001 Rev. 19. _ Page 44 of 84 OFF-SITE DOSE CALCULAIION MANUAL .1 Dose Factors for Vanous Pathways 2 Pages DIpj External Dose Factors for Standing on Contaminated Ground mrem m2 / hr pCi Radionuclide Total Body Skin H-3 0 0 C-14 0 0 Na-24 2.5E-8 2.9E-8 P-32 . 0 0 Cr-51 2.2E-10 2.6E-10 Mn-54 - 5.8E-9 6.8E-9 Mn-56 l.IE-8 1.3E-8 Fe-55 0 0 Fe-59 8.0E-9 9.4E-9 Co-58 7.OE-9 8.2E-9 Co-60 1.7E-8 2.0E-8 Ni-63. 0 0 Ni-65 3.7E-9 4.3E-9 Cu-64 Zn-65 15E-9 4.0E-9 1.7E-9 4.6E-9

-L Zn-69 0 0 Br-83 Br-84 6.4E-11 1.2E-8 9.3E-1 I 1.4E-8 L Br-85 0 0 Rb-86 Rb-88 Rb-89 6.3E-10 3.5E-9 1.5E-8 7.2E-10 4.0E-9 1.8E-8 L

Sr-89 Sr-91 5.6E-13 7.1E-9 6.5E-13 8.3E-9 L Sr-92 Y-90 9.0E-9 2.2E-12 1.0B-8 2.6E-12 L Y-91m 3.8E-9 4.4E-9 Y-91 Y-92 Y-93 2.4E-11 1.6E-9 5.7B-10 2.72-l I 1.9E-9 7.8E-10 L

LW 5.0E-9 5.8E-9 Zr-95 Zr-97 Nb-95 5.5E-9 5.1E-9 6.4E-9 6.0E-9 L

Mo-99 1.9E-9 2.2E-9 Tc-99m 9.6E-10 1.12-9 Tc-101 2.7E-9 3.0E-9 Ru-103 3.6E-9 4.2E-9 Ru-105 4.5E-9 5.1E-9 Ru-106 1.5E-9 1.8E-9 Ag-lOkn 1.8E-8 2.1E-8 Te-125m 3.52-11 4.8-l1 Te-127m l.lE-12 1.3E-12 L

Reference PMP.6010-OSD-001 I Rev. 19+ l Page 45 of 84.

OFF-SITE DOSE CALCULATION MANUAL .1 Dose Factors for Various Pathways l Pages:

Radionuclide Total Body Skin Te124LE-8 L7E-81 Te_ ___ ___ ___8A.E-9 9.9E-9

____ ___131_ __2.2E-9 2.6OE-6 3e13

.7E-9 __ _.0SE-9

______ 1.41-8 1.9E-8 CI-131 .2-E-9 3.413-9 C1-132 1.7E-8 2.01E-8 Cs-133 3.71E-9 4.9E3-9 C1-134 216E-8 J4-9B8 Ba-135 124E-8 1.41B-8 Cs-134 12.E-8 1.413-8 Cs-137 4.23E-9 4.9E-9 Cs-132 7.9E-9 9.OE-8 Ba-139 2.135-9 2.7E-9 Ba-140 21.52-9 2.BE-9 Ce- 141 4.3E-9O 4.9E-1 Ba-142 7.913-9 2.5E3-9 Ce-144 3.2E3-10 3.713-10 Pr-143 0 0 Pr-144 2.OE-10 2.3E3-10 Nd-147 1.OE3-9 1.2E3-9 W-187 3.lE3-9 3.6E-9 Np-239 9.5E-10 1.IE3-9 Table E-6 of Reg. Guide 1.109.

L

Reference

.I1,_ l PMP.6010-OSDO001 I Rev. 19 Al Page 46 of 84 OFF-SITE DOSE CALCULATION MANUAL Att'acnent 3.2 Radioactive Liquid Effluent Monitoring Instruments l PagWe-*: l 46 -47 INSTRUMENT Minimum Applicability Action Channels 1.

Operable' Gross Radioactivity Monitors Providing Automatic Release Termination L

a. Liquid Radwaste (1)# At times of release 1 Effluent Line (RRS-1001)

b. Steam Generator (1)# At times of release** 2 Blowdown Line (R-19, DRS 314100 +)

c. Steam Generator Blowdown Treatment.

(1)# At times bf release 2 L Effluent (R-24, DRS 3/4200 +)_

2. Gross Radioactivity Monitors Not Providing Auitomatic Release Termination

a. Service Water (1) per At all times 3 System Effluent Line(R-20, R-28) ' train _

3. Continuous Composite Sampler Flow Monitor

a. Turbine Building Sump Effluent Line .

(1) At all times 3 L

4. Flow Rate Measurement Devices a Liquid Radwaste Line (RFI-285)

(1) At times of release 4 L

b. Discharge Pipes*

c. Steam Generator Blowdown (1)

(1)

At all times At times of release NA 4 L Treatment Effluent (DFI-352)

Pump curves and valve setting; may be utilized to estimate flow- in such cases, Action Statement 4 is not applicable. This is primarily in reference to start up flash tank flow.

L OPERABILITY of RRS-1001 includes OPERABILiTY of sample flow switch RFS-1010, which is an attendant instrument as defined inTechnical Specification section l.l. under the termOperable- Operability. This itemis alsoapplicable fcral Eberline liquid monitors (and tecirrespective flow switches) listed hrem Since these mixitors can be used for either batch or continuous release the apprqxiate action statement of 1 or 2 should apply (that L

is, Action I if a steam generator drain is being perfcarrd in lieu of Action 2). It is possible, due to the steam generator sampling system lineup, that BOTH action statements are actually entered. This would be the case when sampling for steam generator draining requires duplicate sanmles while te sample system is lined up to discharge to the operating uits blowdown system In this case the steam generator drain samples can fulfill the sample requirement 5cr Action 2 also. Action 2 would be exited when L

sampling was terminated.

+ Some Westinghoiuse (R)radiationmonitors are being replaced by berline (DRS) ronitors. Either monit can fulfill the operabilityrequiren=L L a IF an RMS monitor is inoperable solely as the result of the loss of its control room alarm annunciation, THEN one of the following actions is acceptable to satisfy the ODCM action statement compensatory surveillance requirement:

1. Collect grab samples and conduct laboratory analyses per the specific monitor's action statement, L

-OR-L

ReerencePMP-6010-OSD-001 Rev. 19 Paige 47 of 84 OFF-SITE DOSE CALCULATION MANUAL II I por ..

Attach~ieW t 3.2 Radioactive Liquid Effluent Monitoring Instruments l 46a* 2. Collect local mnonitcrreadings at a frequency equal to or greater than (more frequently than) the action frequency.

IF the RMS monitor is inoperable for reasons other than the loss of control room annunciation, THEN the only acceptable action is taldng grab samples and conducting laboratory analyses as the reading is equivalent to a grab sample when the monitor is functional.

TABLE NOTATION Acton I With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases may continue, provided that prior to initiating a release:

1. At least two independent samples are.analyzed in accordance with Step 3.2.3a and;

2. At least two technically qualified members of theFacility Staffindependentlyverify the discharge valving. Otherwise, suspend release of radioactive effluents via this pathway.

Action 2 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this'pathway riay continue for up to 30 days provided grab samples Ace analyzed for gross radioactivity (beta or gmrnia) at a limit of detection of at least 10-71 0Cgram:

1. At least once per shift when the specific activity of the secondary coolant is> 0.01 pCi/grain DOSE EQUIVALENT 1-131.

2. At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when the specific activity of the secondary coolant is

0.01 pCignrm DOSE EQUIVALENT 1-131.

Action 3 With ihe number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 30 days provided that at least on e per shift, grab samples are collected and analyzed for gross radioactivity (beta or gamma) at a lower limit of detection of at least 10-7 uCi/ml. Since the WestinghouseESW monitors (R-20 and R-28) are only used for post LOCA leak detection and have no auto trip function associated with them, grab samples Er6 only needed if the Containment Spray Heat Exchanger is in service.

Action 4 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 30 days provided the flow rate: is estimated at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months during actual releases.

Corpensatxoy actions are govemed by PMP4030EIS.001, Event-Initiazed Surveillance Testing

Reference PMP-6010-OSD-001 Rev. 19 l Page 48 of 84!

OFF-SITE DOSE CALCULATION MANUAL Attachment 33 i I

Radioactive Liquid Effluent Monitoring Instrumentation Surveillance Requirements Pages:

48 -49 II L Instrument CHANNEL CHECK SOURCE CHECK CHANNEL CALIBRATION CHANNEL OPERATIONAL L

. TEST

1. Gross Radioactivity Monitors Providing Automatic Release Termination

a. Liquid Radwaste D* P R(3) Q(5)

Effluent Line (RRS-100l)

b. Steam Generator D* M R(3) Q(1)

Blowdown Effluent Line .

c. Steam Generator D* M R(3) . Q(l)

Blowdown Treatment Effluent Line .. .. L

2. Gross Radioactivity Monitors Not Providing Aitutomatic Release Termination a Service Water D .M R(3) Q(2)

System Effluent L Line :

3.' Continuous Composite Samplers

a. Turbine Building D* NIA l N/A l N/A L

Sump Effluent Line

4. Flow Rate Measurement Devices

a. Liquid Radwaste Effluent D(4)* WA R Q L

b. Steam Generator D(4)* NIA N/A N/A Blowdown Treatment Line L

During reeases via this pathway L

L L

L

Reference l PMP-6010-OSD- Rev; 19 Page 49 of 8 I OFF-SITE DOSE CALCULATION MANUAL

-Radioactive Liquid Effuent Monitoring l Pages:

Instrumentation Surveillance Requirements 48 - 49 TABLE NOTATION

1. Demonstrate with the CHANNEL OPERATIONAL TEST that automatic isolation of this pathway and control room alarm annunciation occurs if any of the following conditions exists:

1. Instrument indicates measured levels above the alarm/trip setpoint.

2. Circuit failure.*

3. Instrument indicates a downscale failure.*

4. Instrument control not set in operating mcode

5. Loss of sample flow. *

2. Demonstrate with the CHANNEL OPERATIONAL TEST that control room alarm annunciation occurs if aiy of the following conditions exists:

1. Instrument indicates measured levels above the alarm setpoint.

2. Circuit failure.

3. Instrument indicates a downscale failure.

4. Instrument controls not set in operating mode.

3. Perform the initial CHANNEL CALIBRATION using one or more sources with traceability back to the National Institute of Standards and Technology (NIST). These sources permit calibrating the system over its intended range of energy and measurement range Far subsequent CHANNEL CALIBRATION, sources that have been related to the initial calibration may be usecL

4. Verify indication of flow during periods of release with the CHANNEL CHECK Perform the CHANNEL CHECK at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months on days on which continuous, periodic or batch releases are made.

5. Demonstrate with the CHANNEL OPERATIONALTEST that automaticisolation of thispathwayand control room alarm annunciation occurs if any of the following conditions exists:

1. Instrument indicates measured levels above the alarmrtrip setpoint.

2. Circuit failure.**

3. Instrument indicates a doumscale failure.**

4. Instrument control not set in operating mode.*

5. Loss of sample flow.

Instrument indicates, but does n ot provide for automatic isolation

- Instrument indicates, but does not necessarily cause automatic isolation. No credit is taken for the automatic isolation n such occurrences.

Op rations currently perfonrs the routine channel checks and source checks. Maintenance and Radiation Protection perform chaanel calibrations and channel operational tests. COernistry performs the channel check on the continuous conmosite sampler.

The se responsibilities are subject to change without revision to this docunen't

Reference ' PMP-6010-OSD-001 Rev. 19 Page 50 of 84.

OFF-SITE DOSE CALCULATION MANUAL Attachment 3.4 l Radioactive Gaseous Effluent Monitoring Instrumentation Pages:

I 1 50 -52 Insrument (Instrmnent#) Operable' Minimumn Action

. Channels

_ _ _Action

1. Condenser Evacuation System _ _ ___ _____

a. Noble Gas Activity (1) 6 Maiitor (SRA-190512905) .

b. Flow Rate Mcnitcr (SFR.401, (1) 5 1I/2-MR-054 and/or SRA- 191012910) OR L.

(SFR-402 and 1/2-MR-054) . _

2. Unit Vent. Auxiliary Building Ventilation System ._.

a. Noble Gas Activity (1) 6 Mcnitcr (VRS-1505/2505)

b. Iodine Sampler (1)

8 Cartridgefor VRA-1503t2503 . ._.

c. ParticulateSamplerFilter (1)

8 for VRA-150112501 : .

d. Effluent System Flow Rate (1) 5 Measuring Device (V;R-315, MR-054 and/or VFR-1510/2510) .

L

e. SamplerFlowRate (1) 5 L

Measuring Device (VPS-1 5212521) .-

3. Ccntaxament Purge and Ccntainment Pressure Relief (Vent)

a. Containment Noble Gas Activity Monitor ERS-13/1405 (ERS-2312405)

(1) 7 L

b. ContainmentParticulate SamplerFilter (1) **10 ERS-13/1401 (ERS-23/2401)

4. Waste Gas Holdup System and CVCS HUI

a. Noble Gas Activity (1) ****4 9 L

Alarm and Terminaticn I'.

of Waste Gas Releases (VRS-1505/2505) L

5. Gland Seal Exhaust a Noble Gas Activity Monitor (SRA-1805 I..

L

b. Flow Rate Mmnitcr (SlR-201, MR-054 or SER-IBD102810)

At all times L

- - - During releases via this pathway L

L

Reference l PMP-6010-OSD-O01 l Rev. 19 l Page 51 of 84 OFF-SITE DOSE CALCULATION MANUAL Attachment 3.4 Radioactive Gaseotii Effltient lt Monitoring Instrumentation1 Pages:

50 -52 TABLE NOTATIONS

1. IF an RMS monitor is inoperable solely as the result of the loss of it's control room alarm annunciation, THEN one of the following actions is acceptable to satisfy the ODCM action statement compensatory surveillance requirement:

1. Take grab samples and conduct laboratory analyses per the specific monitor's action statemnent,

-OR-

2. Take local monitor readings at a frequency equal to or greater than (more frequently than) the action frequency.

IF the RMS monitor is inoperable for reasons other than the loss of contrd room annunciation, THIEN the only acceptable action is taking grab samples and conducting laboratory analyses as the reading is equivalent to a grab sample when the monitor is functional.

2. Consider releases as occurring "via this pathway" under the following conditions:

The Containment Purge System is in operation and Containment Operability is applicable,

-OR- I The Containment Purge System is in operation and is being used as the vent path for the venting of contaminated systems within the containment building prior ti coripleting both degas end depressurization of the RCS.

IF neither of the above are applicable, TMEN the containment purge system is acting as a ventilation systera end is covered by Item 2 of this Attachment.

-OR-A Containment Pressure Relief (CPR) is being performed.

3. For purge (including pressure relief) purposes only. Reference TS 3.6.1, Containmnt Purge Supply and Exhaust System Isolation Instrumentation and 3.4.15, RCS Leakage Detection Instrumentation for additional informatirL.

4. For waste gas releases only, see Item 2 (Unit Vent, Auxiliary Building Ventilation System) for additional requirements.

ACTIONS

5. With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 30 days provided the flow rate is estimated at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . After 30 days, IF the channels are not OPERABLE, TBEN continue releases Keith estimation of the flow rate once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and provide a description of why the inoperability was not corrected in the next Annual Radiological Effluent Release Report.

6. With the number of channels OPERABLE less required by the Minimum Channels OPERABLE requirerrnvt, effluent releases via this pathway may continue for up to 30 days provided grab samples are taken at least once per shift and these samples ae analyzed for gross activity within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . After 30 days, IF the channels are not OPERABLE, THEN continue releases with grab samples once per shiR and provide a description of why the inoperability was not corrected in the next Annual Radiological Effluent release Reporl.

7. With the number of channels OPERABLE less than required by the Minimum Channels 'OPERABLE requirements, immediately suspend PURGING or VENTING (CPR) of radioactive effluents via this pathway.

Reference I PMP-6010 OSD-001 . I Rev. 19 Page 52 of 84 OFF-SITE DOSE CALCULATION MANUAL Attachmnent 3.4 c 34 Radioactive Gaseous Effluent Monitoring Iiis~aumentation I50Pages: -52

8. With the number of channels OPERABLE less than required by the Minhmnu Channels OPERABLE requirement. effluent releases via the affected pathway nsy continue fcr up to 30 days provided samples required fcr weekly analysis are continuously collected with awxiary sampling equipment as required in Attachmrnt 3.7. Radioactive Gaseous Waste Sampling and Analysis ProgramL After 30 days. IF the channels are not OPERABLE THEN continue releases with sample collection by auxliary sampling equipxment and provide a description of why the inoperability was not corrected in the next Annual Radiological Effluent Release Report Sampling evolutions are not an interruption of a continuous release or sampling period.

9. With the number of channels OPERABLE less than required by the Minimum Channels OPERABIE requirement, the contents of the tank(s) rmay be released to the environment for up to 14 days provided that prior to initiating the release:

a. At leastiwo independent samples of the tans contents are analyzed and, I-

b. Ahtleal t*o technically qualilied rrrrdbrs of the Facility Staff independently verify the release rate calculatiins and.

discharge valve lineups; otherwise, suspend release of radioactive effluents via this pathway.

10. SeeTechnical Specification 3A.15. RCS Leakage Detection Systemlnstumntation.

L L

Compensatory actions am governed by PMP4030EIS.001. Event-laitiated Surveillance Testing.

L

Reference PMP-6010-OSD-OD1 Rev. 19 Page 53 of 84_

QTF-SITE DOSE CALCULATION MANUAL Attachment 35 Radioactive Gaseous Effluent Monitoring l Pages:

A hI Instrumentation Surveillance Requirements 53 - S4 Instrument CHANNEL SOURCE CHANNEL CHANNEL OPERATIONAL CHECK CHECK CALIBRAT1ON TEST

1. 4Condenser Evacuation Alarm Only Systemrn a1.NobleGasAcdvityMonitor D** M R(2) Q(1)

(SRA-1 905/2905) .

1h. System Effluent Flow Rate D** NA R Q (SFR-401, SFR402, MR-054, SRA-191012910) .

2. Auxiliary Building Unit Alarm Only Ventilation System 1- Noble-Gas Activity Monitor D M R(2) Q(1)

(VRS-1505/2505)

D. Iodine Sampler W* NA NA NA (For VRA-150312503) ._.

Particulate Sampler W* NA NA NA (For VRA-1501/2501)

d. SystemEffluentFlowRate D* NA R Q Measurement Device (VFR-315, MR-054, VRS-1510/2510)

e. Sampler Flow RateMeasuring D N/A R Q Device (VFS-1521/2521) _

3. ContainmentPurgeSystemand Alarm and Tnp Containment Pressure Relief

a. ContainmentNoble.Gas S** P R(2)

Activity Monitcr (ERS-13/1405 and ERS-23/2405)

b. Containment Particulate S*NA R Q Sampler (ERS-1311401 and ERS-23/2401)

4. Waste Gas Holdup System Alarm and Thip Including CVCS HUT

a. Noble Gas Activity Monitor p** P R(2) Q(3)

Providing Alarm and Termination (VRS-1 50512505) _ _ _ _ _ _ _ _ _ _ _ _

5. Gland Seal Exhaust Alarm Only

a. Noble Gas Activity D** M R(2) Q(1)

(SRA-1805/2805)

b. System Effluent Flow Rate D** NA R. Q (SFR-201, MR-054, SRA.181012810)

' At all times

- During releases via thispathway

L_

Reference PMP-6010-OSD-0D 1 Rev. 19 Page 54 of 84 OFF-SITE DOSE CALCULATION MANUAL Radioactive Gaseous Effluent Monitoring . Pages:

A Instrumentation Surveillance Requirements 53 - 54 TABLE NOTATIONS L

1. Demonstrate with the CHANNEL OPERATIONAL TEST that control room alarm annunciation occurs.if any of the following conditions exists:

1. Instrument indicates measured levels above the alarm setpoint. *

2. Circuit failure.

3. Instrument indicates a downscale failure.

4. Instrument controls not set in operate mode.

2. Perform the initial CHANNEL CALIBRATION using one or more sources with traceability back to the NIST.

These sources permit calibrating the system over its intended range of energy and measurement range. For subsequent CHANNEL CALIBRATION, sources that have been related to the initial calibration may be used. - -

3. Demcsstrate withtheCHANNELOPERATIONALTEST that automatic isolation of this pathway and control ropm alarm annunciation occurs if any of the following conditions exists:

1. Instrument indicates measured levels above the alarnaltrip setpoint.

2. Ciracuit failure.*

3. Instrument indicates a downscale failure.* . *

4. Instrument controls not set in operate mode. *.

Instrument indicates, but does not provide automatic isolation. *, . -

Operations currently pcrfbums the routine channel checks, and source checks. Maintenance and Radiation Protection perfon channel calibrions and channel operational tests. These responsibilities ae suect to change withoutrevision to this docurnent.

L

. L

Reference l PMP-6010-OSD-001 t l Rev. 19 Page 55 of 84 OFF-SITE DOSE CALCULATION MANUAL

!*s" to

.Atlachment 3.6 lRadioactiveLiquid Waste Sampling and Analyglii Prgramnl 5~es

[IeL5.2.1s]

LIQU1D SAMPLING MINIMUM TYPE OF LOWER RELEASE FREQUENCY ANALYSIS ACTIVITY LIMIT OF TYPE FREQUENCY ANALYSIS DETECTIO1N (LLD)

A. ]3atch Waste P P Principal Gamma 5x17 Release Tanks. Each Batch Each Batch Emitters '

I-131 1x0 4 _

P P Dissolved and Entrained Gases Each Batch Each Batch (Gamma 1x10 5

__. E~e trs_

1.

P M H-3 1xie

.Each Batch Compositeb

. Gross Alpha lx10 7 P Q Sr-89, Sr-90 5xlO-Each Batch Com bposite_

Fe-55 lx104 B. Plant W Principal Gamma Continuous Daily Composite Emitters e 5x10 7 Releases* d 1-131 IxlO6 M M Dissolved and Grab Sample Entrained Gases 1x10 5 (Gamma Emitters)

M . H-3 1x10 5 Daily Compositeb Gross Alpha lx10 7 Q b Sr-89, Sr-90 5x10 '

Daily Composite_

Fe-55 lxlD1

Duirjg releases via this pathway This table provides the minimum requirements fcr the liquid sampling program. If additional sampling is perftrmec then those sample results can be used to quantify releases in lieu of composite data for a more accurate quantification. Examples of the;e samples are the 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months secondary coolant aclivity and Monitor Tank tritium samples.

Reference PMP-6010-OSD-001 lOI Rev. 19 l Pae 56of84L OFF-SITE DOSE CALCULATION MANUAL Attachrnt 3.6 Radioactive Liquid Waste Sampling and Analysis Program 5ges6 TABLE NOTATION ,_

a. The lower limit of detection (LLD) is defined in Table Notation A. of Attachment 3.20, Maximum.

Values for Lower Limits of Detectionsw - REMP _

b. A composite sample is one in which the quantity of liquid sampled is proportional to the quantity of liquid waste discharged and in which the method of sampling employed results in a specimen which is representative of the liquids released.

c. A batch release is the discharge of liquid wastes of a discrete volume. Prior to sampling for analysis, .

isolate, recirculate or sparge each batch to ensure thorough mixing. Examples of these are Monitor Tank and Steam Generator Drains. Before a batch is released the tank is sampled and analyzed to determine that it can be released without exceeding federal standards.

d. A continuous release, is the,discharge of liquid of a non-discrete volume; e-g. from a volume of system .

that has an input flow during the continuous release. This type of release includes the Turbine Room L Sump, Steam Generator Blowdown and the Steam Generator Sampling System.

e. The principal gamma emitters for which the LLD specification applies exclusively are the following radionuclides: Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, Cs-134, Cs-137, Ce-141 and Ce-144. This L

list does not mean that only these nuclides are to be detected and reported. Identify and report other peaks, which are measurable and identifiable, together with the above nuclides. L

Reference l_ PMP-6010-OSD-001 Rev. 19 l Page 57 of 8 OFF-SITE DOSE CALCULATION MANUAL Radioactive Gaseous Waste Sampling and *. ages:

Attachment 3.7 Analysis Program 57 - 58 Gaseous Release Type Frequency Minimum Type of Lower Lnit Analysis Activity omDetection

_ Frequency Analysis (tCi/*l)

a. 'Waste Gas Storage Tanks P P Principal Gamma and CVCS Htrs Each Tank Each Tank Emitters' 1 x i0 4 Grab Sample H-3 I x le

b. Containment Purge P P Principal Gamma Each Purge Each Purge Emiuers d 1X 104

. Grab Sample

~~~~... .. ....... ..........

CPTRwveni** Th~ceper Twiceper Manth Month H-3 IxlO'

c. oindenserEvacuation WarM M Principal Gamma System Grab Sample Particulate Sample Emitters" I x ll Gland Seal Exhaust* . . H-3 ix io W Principle Gamma Ix IO4 Noble Gas Emitters d M 1-131 Iodine Adsorbing I x Io-12 Media Continuous W' Noble Gases Noble N Gas Monitcr I x 10-6

d. Auxiliary Building Unit Continuous a b 1-131 Vent* Iodine Adsorbing 1x 32 Media Continuous" W b Principal Gamma Particulate Sample Emittersa x Io-,o Continuousc M Gross Alpha Composite Particulate I x 10 2 Sample W - h H-3 Grab Sample H-3 Sample I x 104 WSJ Principle Gamma IxID14 Noble Gas Emitters d Continuous" Q Sr-89, sr-go CompositeParticulate I x 10 Sample .

Continuous' Noble Gas Monitor Noble Gases I x 10' c Incinerated Oil' P P Principal Gamma Each Batchr Each Batch' Emitters d 5x 1077

During releases via this pathway

- OAly a tvice per month sampling program for containment noble gases and H3 is required This table provides the minimum requirements for the gaseous sampling program If additional sampling is perfonned than those sample results can be used to quantify releases in lieu of composite data for a more accurate quantification.

Examples of these samples are verification ar compensatory action sample results.

Reference PMP-6010-OSD-001 Rev. 19 P e 58 of 84 OFF-SITE DOSE CALCULATION MANUAL Attadhuent 37 Radioactive Gaseous Waste Sampling and. l Pags-,.

.tafinn3. Analysis Program 57 - 58 TABLE NOTATION a The lower limit of detection (LLD) is defined in Table Notation A. of Attachment 3.20, Maximum Values for Lower Limits of DetectionsAB -REw

b. Change samples at least once per 7 days and complete analyses within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months after changing. Perform analyses at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for 7 days following each shutdown, startup or THERMAL POWER change greater than 15% per hour of RATED THERMAL POWER. WHEN samples collected for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months are analyzed, THEN the corresponding LLDs may be increased by a factor of 10. This requirement does not apply IF (1) analysis shows that DOSEQ I131 concentration in the RCS has not increased more than a factor of 3; and (2) the noble gas moitfor shows that effluent activity has not increased more than a factor of 3. IF the daily sample requirement has been entered, THEN it can be exited early once both the radiation monitor reading and the RCS DOSEQ 1131 levels have returned to within the factor of 3 of the pre-event 'nonral'.[Ref. 5.2. ly ,

c. Know the ratio of the sample flow rate to the sampled stream flow rate for the time period covered by each dose or dose rate calculation made in accordance with steps 3.2.4a, 3.2.4b, and 3.2.4c of this document.

Sampling evolutions are not an interruption of a continuous release or sampling period.

L

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

e. Releases from incinerated oil are discharged through the Auxiliary Boiler System Account for releases L based on pre-release grab sample data.

f Collect samples of waste oil to be incinerated from the container in which the waste oil is stored (example: waste oil storage tanks, 55 gal. drums) prior to transfer to the Auxiliary Boiler System Ensure L

samples are representative of container contents.

g. Obtain and analyie a gas marinelli grab sample weekly for noble gases effluent quantification. L

h. Take tritium grab samples at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when the refueling cavity is flooded. L

i. Grab sampling of the Gland Sea] Exhaust pathway need not be performed if the RMS low range channel (SRA-1 805/2805) readings are less than IE-6 pC/cc. Attach the RMS daily averages in lieu of sampling.

This is based on operating experience indicating no activity is detected in the Gland Seal Exhaust below this value. Compensatory sampling for out of service monitor is still required in the event 180512805 is inoperable.

j. Sampling and analysis shall also be performed following shutdown, startup orTTHERMAL POWER.

change exceeding 15% of RATED THERMAL POWER within a one hour period. This noble gas sample shall be performed within four hours of the event. Evaluation of the sample results, based on previous samples, will be performed to determine if any further sampling is necessary.

Reference l PMP-6010-OSD-OO1 I Rev. 19 l Page 59 of 84 OFF-SITE DOSE CALCULATION MANUAL Attachirhef 3.8 l Multiple Release Point Factors for Release Points 5Pge: . .

Liquid Factors Maiitor Description Monitor Number M"#

U I SG Blowdown IR19/24. DRS 310013200* 0.35 U 2 SG Blowdown 2R19124, DRS 410014200* 0.35 UI&2I quid Waste Discharge RRS-1000 0.30 Sources of radioactivity released from the Turbine Room Sunp (IRS) typically originate from the secondary cycle whi :h is already being monitored by instrumentation that utilizes multiple release point (MRP) factors. The MRP is an admanistrative value that is used toassist ith maintaining releases ALARA. The TRS has noanta ,radiation ffonilcr, but atilizes an autornatic compositor for mnomitcringwvbatl has teen released. The batch release path, through RRS-10IOO, is tke predominant release path by several magnitudes. Tritiumis the predominant radionuclide released from the site and theradiation monitors do not respond to this low energy'beta emitter. Based on this information and the large degree of conservatism built into theradiation monitor setpoint.rrethodology it does not ppear to warrant further reduction fcr the 7RS release path since its source is predomifiantdy the secondary cycle which is adequately covered by this factcr.

Gaseous Factors MaiitorDescriptian Mmiitr Numbtter ow Rate (crn) MRP UnMI Unil: Vent VRS-1500 186,600 0.54 Gland Seal Vent SRA-1800 1,260 O.00363 SteamletAirEjector SRA-1900 3,600 (b) 0.01 StartUpFTVent 1,536 0.004 TotW] 192,996 Unil 2 Unit Vent VRS-2500 143,400 0.41 Gland Seal Vent SRA-2800 5,508 (a) 0.02 SteamJetAir Ejector SRA-2900 3,600 (b) 0.01 Star: Up FT Vent 1,536 0.004 Total 154,044 _

EitberR-19, 24. DRS 3/4100 cr3/4200 can be used for blowdoynmonitoringas teEberline rnnitors (DRS) are replacing the Westinghouse (R) rnaniters.

Norinal Values a TwD release points of 2,754 cfm each ar totaled for this value.

b This is the total design rrnxinrn of the Start Up Airxrecton. This is a conservative vaue for unit 1.

L ReTerence I PMP-6010-OSD-001 l Rev. 19 Pa e 60 of84. 4.-

OFF-SITE DOSE CALCULATION MANUAL .9 '

Liquid Effluent Release Systems age:

L L

S TEM 1 COMPONENTS JCAPACITY 1 FLOW RATE

_ TANKS PUMPS (EACH) (EACH)*

I Waste Disposal System

Chemical Drain Tank I 1 600 GAL. 20 GPM

Laundry & Hot Shower Tanks 2 1 600 GAL. 20 GPM

Monitor Tanks 4 2 21,600 GAL. 150 GPM

Waste Holdup Tanks 2 . 25,000 GAL.

Waste Evaporators 3 _ 30 GPM

Waste Evaporator Condensate 2 2 6,450 GAL 150 GPM Tanks , a.! _ ___._.

II Steam Generator Blowdown and Blowdown Treatment Systems

F Start-up Flash Tank (Vented)# 1. 1,800 GAL 580 GPM

Normal Flash Tank (Not 1 525 GAL. 100 GPM Vented)J.I_ _ _ _ _ -

+ Blowdown Treatment System . 1 60 GPM III Essential Service Water System

+ Water Pumps 4 10,000 GPM

+ Containment Spray Heat 4 3,300 GPM

Exchanger Outlet . -

IV Circulating Water Pumps_

Cnit 1 3 230,000 GPI Unit 2 4 230,000 GP]j

Nominal Values

he 580 gpin value is calculated from the Estimated Steam Generator Blowvdown Flow vs. DRV Valve Position letter prepared by K J. O'Keefe, dated 9/27193. This is 830 gpm times the 70% that remains as liquid wtile the other 30% flashes to steam and exhausts out the flash tank Vent.

Reference PMP-6010-OSD-001 Rev. 19 T Page 62 of 84 OFF-SITE DOSE CALCULATION MANUAL .11 Volumetric Detection Efficiencies.for Principle Garimma Page:

1 Emitting Radionuclides for Eberline Liquid Monitors 62 I-This includes the following monitors: RRS-1000, DRS 3100, DRS 3200, DRS 4100, and DRS 4200. [Ref 5.2.1p]

NUCLEDE EFFICIENCY (cpDrm'CVcc0 1-131 3.78 E7 Cs-137 3.00 E7 Cs-134 7.93 E7 CO-60 5.75 E7 Co-58 4.58 E7 Cr-51 3.60E6 Mn-54 3.30 E7 Zn-65 1.58 E7 Ag-llOM 9.93 E7 Ba-133 4.85 E7 Ba-140 1.92E7 Cd-109 9.58 ES Ce-139 3.28 E7 Ce-141 1.92E8S Ce-144 4.83 E6 Co-57 3.80 E7 L Cs-136 1.07 E8 Fe-59 Sb-124 1-133 2.83 E7 5.93 E7 3.40 E7 L

1-134 1-135 MO-99 7.23 E7 3.95E7 8.68E6 L

Na-24 4.45E7 Nb-95 NbM97 Rb-89 3.28E7 3.50E7 5.00E7 L

Ru-103 3.48 E7 Ru-106 Sb-122 Sb-125 1.23 E7 2.55 E7 3.15 E7 L

Sn-113 7.33 E5 Sr-85 Sr-89 Sr-92 3.70 E7 2.88E3 3.67 E7 L

Tc-99M Y-88 Zr-95 3.60E7 5.25 E7 3.38 E7 L

Zr-97 3.10E7 Kr-85 1.56 E5 Kr-85M 3.53 E7 Kr-88 4.10 E7 Xe-131M 8.15 ES Xe-133 7.78 E6 Xe-133M 5.75 E6 Xe-135 3.83 E7

Ei7 r £ ElF l El.- -- E-.-. EL: CT r: i l i. E7- El 1 C.-

Reference PMP-6010-OSD-001 Rev. 19 Page63of84 OFF-SITE DOSE CALCULATION MANUAL Attachment 3.12 Counting Efficiency Curves for R-19, and R-24 Counting Efficiency Curve for R-19 Efficiency Factor = 4.2 E6 cpnluCilml (szued an mpkical ta tdakeduf pterOWatlUal tWig with C,.137) 1.0WE+07 Ae.A.-

OoE+o05 c

6 1.OOE+04 /

C1 0

.1 1. OOE+03 -

1..OOE+02 1..OOE+01 V , -

nnr.nn I .uu:

. I w 8 0 R S a. 0 0 mlcrocurladml

Reference PMP-6010-OSD-OO1 Rev. 19 lPae 64 of 84 OFF-SITE DOSE CALCULATION MANUAL. ..

Attachment 3.12 Counting Efficiency Curves for R-19, and.R-24 63 - 64 Counting Efficlency Curve for R-24 Efficiency Factor = 7.5E6 cpm/uCi/ml (Based an aempircal dats, t dizg swith Mn-54) 1.00E+07

.tOOE+06 ,

1.00E+0S. '.

1.0E+04 -r__

1.00E2+03_

1.OOE+02-

__ .OO+.

1.OOE+00 Uj IC (4) t -> .e . .@

8 0 D a mtcrocurleslml r - -r [ F~ -7 r- - r YT 17 n7 r - r- [- - r r-r- r -: r

E- fL - L E= E=; E V EC= C7_ E7I. rI- , F, £7 r7-~

F£ Reference PMP.6010-OSD-001 Rev. 19 Page 65 of 84 OFF-SITE DOSE CALCULATION MANUAL .13 Counting Efficiency Curve for R-20, and R-28 r Counting Efriiency Curve for R-20 and R.28 Efficiency ltlwor= 4.3 E6 cpmruCi/ml (Sued vt eLpkicll Wae tk3 duriag prepusfioun tedq with Co.5j) 1.OOE+07 I

1.OOE+06 1.00E+05 Z

a

' 1.OOE+04 I.

.0

I6 1.00E+03, s

on;-n2 -

ItoOEtOt 1.OoE+oO 4

9 mlcrfcurleslml

Reference PMP-6010-OSD-001 Rev. 19 Page 66 of 84 L OFF-SITE DOSE CALCULATION MANUAL

,I . i Attachment 3.14 - Gaseous Effluent Release Systems l* age:

6 L-

- 1.

.. , ILN L-L L

L 1 +-

r-rmn D 4--

Reference PMP-6010-OSD-001 I' Rev. 19 Page 67 of 84 OFF-SITE DOSE CALCULATION MANUAL

'It Attachment 3.15 Plant Gaseous Effluent Parameters Pa67 SYSTEM . UNIT EXHAUST CAPACiTY FLOW RATE

. (CFM)

.I PLANT AUXILIARY BUILDING 1 186,600 max U NIT VENT 2 143,400 max WASTE GAS DECAY TANKS (8)AND 1 125 4082 FI 3 @00 psig CHEMICAL & VOLUME CONTROL 28,741 ft3 max SYSTEM HOLD UP-TANKS (3) @ 8#, 0 level

+ AUXILIARY BUILDING 1 72,660 EXHAUST 2 59,400 _

+ ENG. SAFTY FEATURES 1 &2 50,000 I I .. 1.

VENT A

+ FUEL HANDLING AREA VENT 1 30,000 o.0*s . .

SYSTEM ;

CONTAINMENT PURGE SYSTEM 1 &2 32,000 _

CONTAINMENTPRESSURE 1 &2 1,000 RELIEF SYSTEM . _

INSTRUMENT ROOM PURGE 1&2 1,000 SYSTEM II CONDENSER AIR EJECTOR 2 Release Points SYSTEM _ne forEach Unit NORMAL STEAM JET AIR 1 &2 230 EJECTORS START UP STEAM JET AIR I &2 3,600 EJECTORS III TURBINE SEALS SYSTEM 1 1,260 l _ _

2. 5,508 2 Release Points

__ for Unit 2 START UP FLASH TANK VENT I j 1,536 j E L Dn2 o w536 f s 1

+ Designates total flow for all fans.

Reference PMP-6010-OSD-001 ` Rev. 19 Pa e 68 of 84 OFF-SITE DOSE CALCULATION MANUAL Attachment 3.16 10 Year Average of 1995-2004 Data' Pages: l 68-69 X/Q GROUND AVERAGE (sec/m 3 )

DIRECTION DISTANCE (ME w (WIND FROM) 594 2416 4020 5630 7240 N 4.17E-06 4.82E-07 2.25E-07 1.33E.07 9.32E-08 NNE 3.02E-06 3.64E-07 1.73E-07 1.04E-07 7.29E-08 NE 4.54E-06 5.31E-07 2.60E-07 I.59E-07 1.13E-07 ENE 7.16E-06 7.99E-07 4.04E-07 2.52E-07 1.80E-07 E 1.04E-05 1.13E-06 5.82E-07 3.66E-07 2.63E-07 ESE

1.07E-05 1.18E-06 6.04E-07 3.78E-07 2.72B-07 SE * . 1.15E-05 . 1.24E-06 6.36E-07 4.OOE-07 2.88E-07 SSE 1.30E-05 1.42E-06 7.27E-07 4.57E-07 3.29E.07 S 1.4 IE-05 1.57E-06 7.92E-07 4.93E-07 3.54E-07 SSW 7.03E-06 7.81E-07 3.90E-07 2.41E-07 1.72E-07 SW . . 4.12E-06 4.73E-07 2.28E-07 1.38E-07 9.73E-08 WSW 3.29E-06 3.65E-07 1.76E.07 1.06E-07 7.52E-08

__3.63E,06 4.1 1E-07 1.96E-07 1.18E-07 8.31E-OS .. r . I WNW 3.02E-06 3.43E-07 1.61E-07 9.59E-08 6.71E-08 NW 3.22E-06 3.61E-07 1.71E-07 1.02E-07 7.16E-08 L NNW 3.84E-06 4.29E-07 2.02E-07 IM20E.07 8.40E-08 DIRECTION (WIND FROM) 12067 24135 DS TANCE (ME 40225 RS) 56315 80500 L

N NNE NE 4.64E-08 3.66E-0B 5.75E-08 .

1.79E-08 1.43E-08 2.30E-08 8.89E-09 7.13E-09 1.15E-08 5.68E-09 4.56E-D9 7.41E-09 3.56E-09 2.87E-09 4.72E-09 L

ENE 9.3E-O8 13.8 Ea 1I.91E-08 1.23E-08 7.90E-09 B

ESE SE 1.37-07 _

1.41E-07 1.50E-07

.65E-08 5.81E-08 6.20E-08 2.85-08 2.93E-08 3.12E-08 I.83E-08 1.88E-08 2.OIE-08 1.18E-08 1.22EI08 1.30E-08 L

SSE 1.71E-07 7.06E-08 3.56E-08 2.29E-08 1.48E-08 S

SSW SW I .84E-07 8.86E-48 4.93E-08 7.49E-08 3.59E-08 1.96E-08 3.77E-08 1.80E-08 9.77E-09 2.43E-08 1.15E-08 6.27E-09 1.56E-08 7.39E-09 3.98E-09 L

WSW 3.80E-08 1.51E-0S 7.53E-09 4.83E-09 3.07E-09 W

WNW NW 4.17E-08 3.34E-08 3.57E4-8 I.64E-08 1.29E-08 1.39E-08 8.13E-09 6.41E-09 6.89E-09 5.20E-09 4.1OE-09 4.41E-09 3.28E-09 2.57E-09 2.77E-09 L

NNW 4.19E-08 3.35E-S08 8.10E-09 5.19E-09 3.27E-09 DIRECTION - SECTOR l L.

N =A E S = W =N NNE -B ESE =F SSW =K WNW =P NE =C SE =G SW =L NW =Q ENE -D SSE =1H WSW =M NNW R Worst Case X/Q = 2.04E-05 sec/r in Sector H 2004

Reference PMP-6010-OSD-001 Rev. 19 Page 69 of 8I OFF-SITE DOSE CALCULATION MANUAL Attachment 3.16 10 Year Average of 1995-2004 Dafa Pages: l 68 - 69 DIQ DEPOSMON (I/r2)

FBI[ ECTION , l DISTANCE (METERS)

(WIND FROM) 594 2416 4020 5630 7240 N 2.37E-08 2.29E-09 1.04E-09 5.44E-10 3.47E-10 NNe 9.86E-09 9.52E-10 4.32E-10 2.27E-10 1.45E-10 NE 1.291E-08 1.25E-09 '5.67E1-10 2.97E-10 1.903E-10 EN]3 1.591308 1.54E-09 6.97E-10 3.66E-10 2.33E-10 E 1.87E-08 1.8 1E-09 8.20E-10 4.30E-10 2.75E-10 ESE .1.85-08 1.791-09 8.12E-10 4.26E-10 2.7213-10 SE . 1.903E-08 1.83E-09 . 8.30E-10 4.36E-10. 22.78E-10 -

SSE . 2.40E-08 2.32E-09 1.051E-09 5.52E-10 3.52E1-10 S 3.68E-08 3.56E-09 1.61E-09 8.46E-10 5.40E-10 ssA5 2.30e-08 2.22E -09 1.01E-09 5.28E-10 3.-37E-10 sW__ 2.22E-08 2.15E-09 9.74E-10 5.11E-10 3.26E-10 wSW 2:1 IE-08 2.04E-09 9.23E-10 4.84E-10 3809E-10 w __ . 2.00E-08 1.93E-09 8.74E-10 4.59E-310 *2-93E-10 WNW 1.75E-08

1.69E1-09 7.64E-10 4.011E-10 2.:56E3-10..

_ _ _ 1.58E-08 1.53E-09 6.94E-10 3.641-10 2.32E-10 NNW 2.30E-08 2.22E-09 1.013-09 5.28E-10 3.374-0 IRECTION _ DISTANCE (METERS)

R"IND FROM) 12067 724135 140225 156315. 180500 N 1.45E-10 4.72E-1 1 1.74E-11 9.27E-12 4.65E,-12 NNE 6.36E-11 1.9713-11 7.24E-12 3.86E-12 1.941312 NE . 8.07E-11 2.sBE-11 9.51E-12 5.07E1-12 2.54E-12 ENE 9.77E-11 3.17F1-11 1.17E-11 6.23E-12 3.13E-12 E 1.14E-10 3.73E- 1 1.37E1-11 7.34E-12 3.68E-12 ES]- 1.13E-10 3.70E-11 1.36E-11 7.2613-12 3.64E-12 SE 1.16E-10 3.78E-1 I 1.39E-11 7.42E-12 3.72E-12 SS]7 1.47E-10 4.79E-11 1.761-1 1 9.41F1-Z 4.72E-12 s 2.25E-10 7.34E-1 I 2.70E-1 I 1.44E-11 7.23E-12 SSWN 1.41E-10 4.59E-1 I 1.69E-11 9.013E-12 4.52E-12 sm_ 1.36E-10 4.43E 11 1.631-11 8.71E-12 4.37E-12 W<;w 1.29E-10 4.20E-lI 1.5513-11 8.26E-12 4.14E-12 w 1.221-10 3.98E-11 1.47E-11 7.82E-12 3.92E-12 WIAW 1.07E-10 3.48E-11 1.28E-11 6.84E-12 3.43E-12 NVW 9.7DE-11 3.16E-11 1.16E-11 . 6.20E-12 3.11E-12 NbW 1.411-10 4.58E-11 1.69E-11 9.00E-12 4.52E-12 DIRECTION . SECTOR N =_A E =E S - W N IE =B ESE =F SSW =K WNW = P NE =C SE =G SW =L. NW = 2 ENE =D SSE =H WSW =M NNW=R _

Wcrst Case D/Q = 4.463-08 Itr 2 in Sector A 2001

Reference I PMP-6010-OSDO01 l Rev. 19 l Page 70 of 84 OFF.SITE DOSE CALCULATION MANUAL Attachment 317 Annual Evaluation of z/Q and D/Q.alues For l Page:

I All Sectors 70 6..

1. -X.Perfoimed or received annual update of x/Q and D/Q values. Provide a description.of what has been received Signature Date Environmental Department (print name, title)

.4 .. .' .. ~.

2. Worst X/Q and D/Q value and sector determined. PMP-6010-OSD-O01 has been updated, jthne'es~ary. Provide an evaluation.

signature S ;, ; Date Environuientaj Lepa=mIent (punt name, title)

3. Review nuclide mix for gaseous and liquid release paths to determine if the dose conversion factor of total body is still applicable. Provide an evaluation.

Signature Date L

Environmental Department print name, title)

L

4. Approved and verified by:

Signature Date Environmental Department (print name, title)

L

5. Copy to NS&A for information. L Signature Date L

Environmental Department (print name, title)

Reference PMP-6010-OSD-001 'l Rev. 19 Page 71. of 84l OFF-SITE DOSE CALCULATION MANUAL i1iif Attachment 3.18 Dose Factors Pages:

DOSE FACTORS FOR NOBLE GASES AND DAUGHTERS*

TOTAL BODY SKIN DOSE GAMMA AIR BETA AIR DOSE PACTOR FACTOR DOSE FACTOR DOSE FACTOR K, (DFB) L (DFSD Ml (DF"I) N1 (DFD muremin 3 (rnrem n3 (irad 3 (mrad m3 RADIONUCLIDE per RLCi yr) per pCi yr) per ACi yr) per 9Ci yr)

Kr-83m 7.56E-02 --- 1.93E+01 2.88E+02 Kr-85m , . 1.17E+03 1.46E+03 1.23E+03 1.97E+03

.1 Kr- 85 - 1.61E+01 1* 1.34E+03 1.72E+01 1.95E+03

.I.

Kr-87 5.92E+03 9.73E+03 6.17E+03 1.03E+04 Kr- 88 ,; ;. 1.47E+04 2.37E+03 1.52E+04 2.93E+03 Kr- 89 1.66E+04 1.OIE+04 1.73E+04 l.06E+04 Kr-90 1.56E+04 7.29E+03 1.63E+04 7.83E+03 Xe-131m 9.15E+01 4.76E+02 1.56E+02 1.11E+03 Xe-133m 2.51E+02 9.94E+02 3.27E+02 1.48E+03 Xe-133 2.94E+02 3.06E+02 3.53E+02 1.05E-+03 Xe-135m 3.12E+03 7.11lE+02 3.36E+03 7.39E+02 Xe-135 1.81E+03 1.86E+03 1.92E+03 2.46E+03 Xe-137 1.42E+03 1.22E+04 1.51E+03 1.27E+04 Xe-138 8.83E+03 4.13E+03 9.21E+03 4.75E+03 Ar-.41 8.84E+03 2.69E+03 '9.30E+03 3.28E+03

Ihe listed dose factors are for radionucidcs that may be dcteded in gaseous effluents. firm Reg. Guide 1.109. Table B-l.

I Reference PMP-6010-OSD-001 i Rv. 19 Page 72 of 84 OFF-SITE DOSE CALCULATION MANUAL

, ..(... . - , Attachment 3.18 Dose Factors' I 7 I-DOSE FACTORS FOR RADIOIODINES AND RADIOACTIVE PARTICULATE, I*

IN GASEOUS EFFLUENTSFOR CHILD* RLef.5.1ecandff Pt Pi INHALATION FOOD & GROUND PATHWAY PATHWAY (mrem m3 (mrem m2 sec RADIONUCLIDE per gCi yr) per IiCi yr)

H-3 1.12E+03 1.57E+03 P-32 2.60E+06 7.76E+10 . , ... - .

Cr-Sl . 1.70E+04 1.20E+07 Mn-54 1.58E+06 1.12E+09 Fe-59 * . - 1.27E+06 5.92E+08 Co-S58 ', 1.11E+06 5.97E+08 ....I I S Co&60 : 7.07E+06 4.63E+09 Zn-65 9.95E+05 1.17E+10 Rb-86 1.98E+05 8.78E+09 L Sr-89 2.16E+06 6.62E+09 Sr-90 1.01E+08 1.12E+11 Y-91 2.63E+06 6.72E+06 Zr-95 2.23E+06 3.44E+08 Nb-95 6.14E+05 4.24E+08 Ru-103 6.62E+05 1.55E+08 Ru-106 1.43E+07 3.01E+08 Ag-1 10m 1-131 1-132 5.48E+06 1.62E+07 1.94E+05 1.99E+10 4.34E+ l1 1.78E+06 L

1-133 1-135 Cs-134 3.85E+06 7.92E+05 1.OlE+06 3.95E+09 1.22E+07 4.OOE+10 L

Cs-136 Cs-137 Ba-140 1.71E+05 9.07E+05 1.74R+06 3.00E+09 3.34E+10 1.46E+08 L

Cc-141 5.44E+05 3.31E+07 Ce-144 1.20E+07 1.91E+08 L

As Sr-90.Rn-106 and 1-131 analyses art pefornMd, THEN useP i given in P-32 for nonlisted radionuclides.

'Tbe tnits forboth H3 factors armthe same. raremn m per ACi yr L

L

Reference l PMP-6010-OSD-001 I Rev.'19 lT Page 73 of 84 OFF-SITE DOSE CALCULATION MANUAL Lttac Attachme 3.19 nt 3.9 Radiological Environmental Monitoring Program lSample Stations, Sample Types, Sample Frequencies viii, -Pages:

73 -76 Ref!2.1v.,52Ix, 5.21t)

SAMPLE DESCRII0TION( SAMPLE SAMPLE ANALYSIS ANALYSIS STATION LOCATION TYPE FREQUENCY TYPE FREQUE1CY ON-SITE AIRBORNE AND DIRECT RADIATION (MID)STATIONS ONS-I (T-1) 1945 ft @ I 8'from Phnt Axis AirbameParticulate Weekly GrossBcta Weekly Guam Isotopic QuarL Comr.

AirborneRadioiodine 1-131 Weekly

______ _Qurte rly Diltet Radiation ONS .2 (T-2) 2338 ft 480 from Placn Axis Aiiborne Paticolate Weekly GossBtea Weekly Gamma Isotopic Quati Cotn.

Airborne Radioiodine_ 1-131 Wedldy

._-_TLD Quartely DirectRadiation ON'-3 (T-3) 2407 ft 0 90°fromPlant Axis AisbornePasticulate Weekly GrossBeta Weekly Gamma Isotopic Quart Cor AiiborneRadioiodine 1-131 Weekly

. TLD Quay Direct Radiation Quarterly ONA4CrT4) .1852hL @ I IS fron Plant Axis. AiiborneParticulate Weekly Gross Beta Weekly

.OGin Isotopic 0uart. Comr.

. AisborneRadioiodine 1-131 Weekly

_ TID Ourtedv Direct Radiati on Quarery ON1.'-5 (T-5) 1895 It@ 189'fin mPlant Axis-' 'AirbornePaiticDlaet Weekly OrossBeta Weekly Gamma Isotopic Quat ComL AilborneRadlioiodine 1-131 Wekly TLD urtedy Direct Radiation QuariY O* i6 (T-6) 1917 ft @ 21D'fomPlant Axis AixbornePaziiculate Weekly GO1ssBcta Weekly Gamnma Isotopic Quait. Coto w AiiborneRadioiodinc 1.131 Weekly TLD Quanay Direct Radiation T-7 21031?036' from Plaat Axis TLD Quartedy Dirct Radiation Quatedy T-8 _ 2208 t La 82' from Plant Axis TLD Quartedy DirectRadiation Quariedy T-9 1368 ft @ 149'fom Plant Axis TLD Quartedy Direct Radiation Quantedy T.] 1390 f10 127'fmomPlant Axis TLD Qualedy DirectRadiation Quartery T-1 I 1969 h @ I1I1 from Plant Axis TLD Quatedy Diect Radiation Quastedy T l, _ 22921A0 63' from PJant Axis TLD Quartelly DirectRadiation Quarterly CONTROL AIRBORNE AND DIRECr RADIATION MlMD) STATIONS NBII 15.6 miles SSW AiibornelParticulate Weekly Gross Beta Weekly New Bffao Ml

?A7 Oamma Isotopic Quart Comri.

Airborne Radioiodine 1-131 Weekly TLD Quarry Direct Radiation Quanedy SBEN 26.2 les E Airborne Pasiculate Weekly Gross Beta Weekly SouthBend, Gamma Isotopic ON Quart Comp.

AisborneRadioiodine _ 1-131 Weekly TLD Qua_ Ty Direct Radiation Ouatudy DOW 24.3 miles ENE AirbornerParticulate Weely Gross Beta Weekly Dowaise. MNg Gamma Isotopic Quart. Comi.

AirborneRadioiodine 1-131 Weekly TLD Qualtedy Direct Radiation COI. 1S.9milesNNe AiiborneParticulate Weekly GrossBeta Weekly Coloas. M Gamma Isotopic Quasi Com, Aisborne Radioodine 1-131 Weekly I TLD Quarery Direct Radiation r uanuiY

. Reference PMP-6010-OSD-001 Rev. 19 *t Pa e 74 of 84 OFF-SITE DOSE CALCULATION MANUAL c t3 Radiological Environmental Monitoring Program ' Pages:

Atchnent 3.19 Sample Stations, Sample Types, Sample Frequencies 73 - 76 I.-

SAMPLE l DESCRIPTION/ l SAMPLE SAMPLE ANALYSIS ANALYSIS STATION l LOCAnON I TYPE FREQUENCY TYPE FREQUENCY OF-SrE DRECr RADIATION (TLD) STATIONS OFT-I 4.5 mil es NE, Pole #B29444 TLD redy Direct Radiation Qutd y OFT-2 3.6 miles. NE, Stevensville TLD Quaneldy Direct Radiation Quartedy LI Substation OFT-3 5.1 miles INE, Pole #B296.13 TLD utl Direct Radiation Quarterl OFT-4 4.1 miles, E, Pole #B350-72 TLD Quatery Direct Radiation Quaerl OFT-S 4.2 miles ESE, Pole #B337-32 LD .Quaedy Direct Radiation Q Y OFT-6 4.9 miles SE. Pole #3426-1 7u1nd TLD Direct Radiation Quaterl OFT-7 2.5 miles S. Buidgman Substation TLD QOuartdy DirectRadiation Quarterly OFTrS 4.0 miles S. Pole #B424-20 TLD Quarey Direct Radiation Quately OFT-9 4A4milesESEPole#B369-214 'TLD Quanuiy Direct Radiation Quarterly OFT-ic 3.5 miles S. Pole #B422-99 71u d Direct Radiation Odart OFT-II 3.3 miles S. Pde#B423-12 TLD Quterly Direct Radiation Quarterly GROUNDWATER (wELLwAER) SAMPLE STATIONS W-1 i1969 ftO I1°fomPlant Axis Grundwater Quarterly Gamma lsotopic Qartedy Tritium Quarterly W-2 2302 ft

63 from Plant Axisr . ..Gromndwater Quarterly Gamma Isotopic Quarterly Tntium Quatedy W-3 3279 ft 0101071 fro Plant Axis orundrwater Quarterly Gamma Isotopic Quantedy Tritium OuutedY W4 4181 S O 301 'from Plant Axis G°adwater Quarterly Gamma lsotopic Qurtely W-5 W-6 404 h @ 290' from Plant Axis 424 110 273- from Plant Axis GzOundwater Omundwatcr Quterly Quarterly Tritinum Gamma Isotopic Gamma Isotopic Qatedy Quanaedy Quartedy L

W-7 W-8 1895 A 0 1l91 frm Plant Axis 1274 ft 0 54'rom Plant Axis Oloundwater Groundwater Quarely Quarterly Tritinm Gamma Isotopic Titium _

Gamma Isotopic

_Qutdy Quarterly Quartely Quarterly L

- Tritium Quarterly W.9 W.10 1447 It0 2 from Plant Axis 4216 f @ 179'fromPlant Axis Groundwater Groundwater Quarterly Quarterly Gamma Isotopic Tritium Gamma Isotopic Quartedy Quarterly Quarterly L

Tritium Quarterly W1lI W-12 3206 1 2631 ft 153'fromPlant Axis 162 from Plant Axis Growndwater Groundwater Quarterly Quarterly Gamma Isotopic Tntinm Gamma Isotopic Quarterly

__Quarterly Quaftedy L

Tritium Quanedy W-13 W-14 2152 ft@ 182OfrmPlant Axis 1780 ft 0 164 from Plant Axis Groundwater Groundwater Quarterly Quarterly GammaIsotopic Tertium Gamma Isotopic Tdthum Quarterly Qutely Quateldy

__Quarterly L

DRINKIG WATER STJ St. Joseph Public Intake Sta. Drinking water Once per calendar Gross Beta 14 day Comp 9 mi. NE Day Oamma Isotopic 14 day Comp 1-131 14 day Comp.

Tritium Quat Comp.

LTW LakeTwp. Public IntakeSta- Drinking water Once per calendar Grnss Beu 14 day Comrp 0.6 ti. S Day Gamma Isotopic 14 day Comp.

1-131 14 day Comp.

Tditium Quart Cotup

' Reference PMP-6010-OSD-001 I Rev. 19 Page 75 of 84 OFF-SITE DOSE CALCULATION MANUAL Radiological Environmental Monitoring Program. Pages: :s,,

. ac n . Sample Stations, Sample Types, Sample Frequencies 73 - 76 SAMPLE I DESCRIPTJON/ SAMPLE SAMPLE ANALYSIS ANALY':IS STATION LOCATION TYPE -FREQUENCY TYPE FREQUENCY SURFACE WATER SWL-2 Plant Site Boadary - Soth Suface Water Oncepercalendar I Gammalsotopie Month Comt,.

. - 500 ft south of Plant Centedine Day ITritium I Quar. Comp SWL-3 Plant Site Boundary - North Surface Water Once pr calendar Gmm Isotopic Month. Comi.

_- S0Oft. nonh of Pianl Centedine Day TTntium IQut Comp=

SEDIMENT SL-2 Plant Site Boundary - South Sediment Semi-Ann. GOmn Isotopic Semi-Annal

______ - 500 f. south ofPlant Centedine .__ ._ . l SL-3 PlantSiteBoondary-Nosth -Sedicment

Semi-Ann. Gamma Isotopic Semi-Annal

-500k S nortbhofPlantCcntedine ' .

SL4 Plant Site Bondry -. South Sediment Quarterly Garmas Isotopic Quaeeiy South storm dnrin culvert to lake . .

SL-S Plant sitCeBondary-North .,Sedimcnt, Quarrly oamm Isotopic Qeaary North storm dain culveat to lake ; J.. *,, .r '_.

IFM&5 are data coflection points onlynot actual REMP samples' .

GROUNDWATER (RADIOACTIVE MATERIAL STORAGE PACITY [MAUSOLEUM)) SAMPLE STATIONS SO-I 0.8 ni. @ 950 from Plant Axis Groundwater Quaterly Gross Alpha Quanedy Gross Beta Quartry

_____ _ OGammaIsotopic Q L..

SG-,' 0.7 mi. @ 92 from Plant Axis Grimndwater Quartery Gloss Alpha Quareraly Gloss Beta Quarterly Gamma Isotopic audy SO2. 0.7 mi. O 93 * [remPant Axis Groundwater Quartrly Gross Alpha Quartery Gross Beta Quartely Gamma Isotopic Quartedy SG-5 0.7 mi. @ 920 from Plant Axis Groundwater Quanedy Gross Alpha Quartedy Gross Beta Quanedy

_ammtla Isotopic Quarely

[G.-. 2, 4 and 5 are data collection points onlynot actual REhP samples -

ING-ESTION- MILK KMTON .

Inditcator farrm'

_IOnce _

__l_

Milk K

l15 Onceevery 15 days every days Once every l_15 days 1-131 Gamrm Isotopic 1-131 Gamma Isotopic 1-13 1 Gamma Isotopic G

per amp]c peane per sample persamp]le per sammle per sample INCESTION - MIK Background Parns I_____

Milk L Once every l_15 days11-131 IGamma Isotopic I persampl lpersample l

Mfilk Once every 1-131 per smMPle I _ l 15 days j Gamma Isotopic Iper sample

L-Reference- PMP-6010-OSD-O01 Rev. 19 Page 76 of 84 OFF-SITE DOSE CALCULATION MANUAL Attachment 3.19'. Radiological Environmental Monitoring Program Pages:

Sample Stations, Sarnple Types, Sample Frequencies 73 - 76 SAMPLE DESCRIPTION/ SAM1LE l SAMPLE ANALYSIS ANALYSIS STATION LOCATION TYPE FREQUENCY TYPE FREQUENCY INGESTION - FISH ONS-N 0.3 mile N. Lake Michign Fish - edible pontion joamma G Isotopic per sample ONS-S 0.4 mile S, Lake Michigan Fish - edible ponion 2ear Gamma Isotopic per sample OPS-N 3.5 mileN,LakeMiFhigan Eisb -edibleposion near jGmmalsotopic persakmPl OFS-S 5.0 milec S.Lake Michigan Fjisb - edible portion Vycar Ganuma Isotopic Ir samp INGESTION - FOOD PRODUCTS On Site ONS-G Nearest sampleto Plantin the GOrpes i' *' Attimeof nGa Isotopic Al time of highestD/Q land sector habest harvest containing media.

ONS-V Broadleaf At time of Oamma Isotopic Al time of vegetation : hawvest harvest Off Site OPS-Ina land sector containing grapes,approximately20 miles Grapes r _

fAt time of harvest

. Gamma Isotopic At time of Harvest from the plant, in one of the OPS.V less prevalent D/Q land sectors -:

Bmadleaf vegetadSon Al time of harvest Gamm Isotopic Al time of harvest L

INGESTION - BROADLEAF IN LIEU OF MILK L

3 indicator samples of broad leaf vegetation Broadleaf MonAhly lamma Isotopic Monthly colected at'differcnt locations, Within eight vegetation when available 1131 when available miles of the plant in the highest annual averageD/Q land sectore I backgroand sampleofsimilr vegetation

.grown 15-25 miles distant in one of Broadleaf vegetation Monthly whan available Gamma Isotopic 1131 Monthly when available L

the less prevalent wind directions.

L Collect eomposite samples of Dinking and Surface waterat least daily. Analyze pLticulate sample filters for grossbeb activity24 ormose ho following filterremoval. This will allow for radon and thon daght decay. lfgossbcta activity in air or water is grceatr than 10 times the yearly men of contrl samples for any nedium perform gamm isotopic analysis on the individual samples. L If atleast thrvc indicator mlk samples and onebackgroundmlk sample cannotbe obtained, threeindicatorbrod lcafsamples wiDbe coUected at diffcrcntlocations. within ight miles ofthcplat inthe land sector wihthehighcstD/Q (refcrs to thc highest annual aveugeD/Q). Also, one backgrmnd broad leaf sample will be collected 15 to 25 miles fronm the plant in one of the less prevalent DIQ land sectors. L The three milk indicatorand two backgrnd farms will be determined by the Annual Land Use Census and those that ae willing to participate.

IFitis determined that the milk animals arefed stored feed, THEN monthly sampling is appropriate for that time period.

L V

L.~

Reference T PMP-6010-OSD-001 Rev. 19 Page 77 of84 OFF-SITE DOSE CALCULATION MANUAL Attachment3.20 l MatxnimiValuesforLowerLimitsofDetections^ -REMP l ages:

RC. .2I 1_ _ _ _ _ _ __ _ __ _ _ ___ ___

Radionuclides Food Product Water Milk Air Filter Fish Sediment pCi/kg, wet, pCi/I pCi/ pCi/m' pCi/kg, wet pCi/lkg, dliy Gross Beta 4 = 0.01 _ =

H-.3 2000 Ba-140 60 60 La 140 15 15 Cs-*134 60 15. 15 0.06 130 150 Cs *137 60 18 18 0.06 150 180 Zr-95 30 ,. . .

Nb-95 15 .. __.____. _

_N-54 15 -. 130 _

Fe*59 30 260 =

Zn-65 30 260 ___ =

Co-58 _ 15 = 130 ._ =

Co-60 15 130 _

1-131 60 1 1 0.07 Thi: Data is directly harn our plant-specific Tecludcal Specification.

Reference PMP-6010-OSD-001 Rev. 19 Page 78 of 84 - L OFF-SITE DOSE CALCULATION MANUAL Attachment3.20 Maxin limVauesforLbwerLimitsofDetectionsA -REMP l Pae7s NOTES A- he Lower Liiit of Detection (ID) is defined as the smallest concentradibi obfrdioactive naterial in a sample that will be detected with95% probability and 5% probability of falsely concluding that a blank observation represents a 'eal" signal.

For a particular measurement systan (which may include radiochemical separation), the LU) is given by the quation:

4.66a

S LLD =E*V

2.22

Ye(~

'acTr L1 ris the anpojn lower limit ofdetection as defined above (as pCi peramiit rns orvine). Perfarm analysis in such a mamer that the stated LLDs will be achieved under routine conditions. Occasionally background fluctuations, unavoidably small sample sizes, the presence of interfering radionuclides, or other mnontrollable circunstances may render these IlDs unadijevable. It should be futher clarified that the LU) represents the capability of a measurement system and not as an after the fact limit for a particilarmeasurmnt...

S is the standard deviation of the background counting rate or of the aunting rate of a blank sample as appropiate (as counts per minute).

B is te counting efficiency of the detection eqiipment as counts per transformation (that is, disintegration)

V is the sample size in appropriate mass or volume units {

2.22 is the conversion factor fonm picuries (pCi) to transformations (disintegrations) per minute Y is the frational radiochcrnical yield as appropriate X is the radioactive decay constant for the particular radionuclide L At is the elapsed time beteet the midpoint of sample collection (or end of sample collection period) and time of counting.

B. Identify and report other peaks which are measurable and identifiable, together with the radionudides listed in Attachment 3.20, Maximum Values for Lower Limits ofDetectionsA.B - REMe. L L

a A 2.71 value may be added to The equation to provide correction for deviations in the Poisson distobution at low count rates.

that is. 2.71 + 4.66 x S.

L.

Reference l PMP-6010-OSD-001 Rev. 19 Page 79 of 81 OFF-SITE DOSE CALCULATION MANUAL Attachment 321 lReporting Leyels for Radioactivity Concentrations l Page:

A in Environmental Samples 1 79 Ra Lionuclides Food Product Water Milk Air Filter Fish pCi/kg, wet pCi/l pCi/ pCi/m 3 pCi/kg, wet H-' . 20000 Ba-140 200 300 La-140 200 300 Cs 134 1000 30 60 10 1000 Cs-137 2000. 50 70 20 2000 Zr-95 400 Nb-95 . 400 kW .......

Mrn-54 . 1000 30000 Fe-59 400 10000 Zn-.65 300 20000 Co -58 1000 30000 Co.60 300 10000 I-131 100 2 3 0.90 _

Reference PMP-6010-OSD-001 I Rev. 19 Page 8"bof 84 OFF-SITE DOSE CALCULATION MANUAL L An . l . Page:

IAttachment :3 I ;On-Site Monitoring Location -RE I 0 ONS-SouM ONS-North L

L L

L LEGEND I

ONS-1 -ONS-6: Air Sampling Station T-1 -T-12 TLDSampling Station W-1 -W-14: REMP GroundwaterWells SWL- 2.3: Surface Water Sampling Stations SL-2 SL-3: Sediment Sampling Stations ONS-N & S: Fish sampling locatons SG-1. 2.4 and 5: Non REMP Information Wells

\^.;} I I ' .., '- . i I I Y; ., N> .

Laporte Background f~lk Farm R-WLvighos i ?

Reference PMP-6010-OSD-001 Rev. 19 -Page 82 of 84 OFF-SITE DOSE CALCULATION MANUAL Attachment324 Safety Evaluation By The Office Of Nuclear :.l,. Pages:

Reactor Regulation 82 - 84 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION '-

RELATED TO DISPOSAL OF SLIGHTLY CONTAMINATED SLUDGE INDIANA MICHIGAN POWER COMPANY DONALD C. COOK NUCLEAR PLANT, UNIT NOS. I AND 2 DOCKET NOS. 50-315 AND 50-316 (Ref. 5.2.1r]

(This is a 10 CPR 50.75 (g) item)

1. I1TROD.UCTION By lefers dated October 9, 1991, Odober 23, 1991,-Septrmbcr 3, 1993, and September 29, 1993. Indiana MHdigan Power Company (a&M) requested approval pursuant to 10 CFR2020(02 for the on-site disposal of licensed material not previously considered in.the Donald C. Cook Nuclear Plant Final Environmental Statement dated August 1973. Spefically, this request . :.

addresses actions taken in 1982 in vYhich approximately 942 cubic metrs of sightly cnntaminated sludge were rmoved from the turbine room sump absorptio pond and pumped to the upper parking lot located within the exclusion irea of the Donald C.

Cook Nuclear Plant The contaminated sludge was spread over an area of approximately 4.7 naas lhe sludge contained a total radionuclide inventory of 8.89 miflicuuies (mCi)oBfCcsium-l37, Cesium-136, Cesiun-134, Cobalt-60 and Iodine-131. ** . .. ,

hI its submitntal the lic Es ia:aisd specific infmmation requested in accordance with 10 CFR 202002(al prvided a detailed description of the licensed naterial, thoroughy analyzed and evaluated information pertinent to the impacts on the envirounent of die proposed disposal of licensed matmial and cornunitted to follow specific procedures to minuimiz the risk of Unexp le expositres.

L

2. DESCRIPTION OF WASTE The turbine room sump absorption pond is a collecdom place for water released from (he plants turbine room sump. The contamination was caused by a primaryto-secondary steam generator leak that entered the pond from the turbine buildig sump, a reognized release pathway. Sludge, consisting mainly of leaves and roots mixed with sand, built up in the pond. As a result, the licensee dredged the pond in 1982. The radioactive sludge removed by the dredging activities was pumped to a containment area located within the exdusion area Th total volume of 942 cubic meters of the radioactive sludge that was dredged from the bottom of the turbine room absorption pond was subsequently spread and made into a graveled road over the L

upper parking lot aea of approximately 4.7 acres.

he principal radionuclides idenified in the dredged material are listed below. L NUCLIDE TABLE 1 ACTVIMTY (mCi) ACTiVITY (mCi)

L (half-life) 1982 1991 6

3 Cs (13.2 d) 0.03 NA* L t34 Cs(2-1 y) 2.34 0.18 I37Cs (30.2 y) 6 0 Co (5.6 y) 5.59 0.90 4.57 0.27 L

311 (8.04 TOTAL:

d) 0.03 8.89 NA*

5.02 L

NA: not applicable due to decay

Reference I PMP-6010-OSD-001 l Rev. 19 ' l Page 83 of 84 OFF-SITE DOSE CALCULATION MANUAL L Atttachmet3.24 l Safety Evaluation By The Office Of Nuclear l Pages:

Reactor Regulation 82 - 84

3. RADIOLOGICAL IMPACTS The licensee in 1982 evaluated the following potential exposure pathways to members of the general public from the radionuclides in die sludge:

(1) external exposure caused by girwdshine from the disposal site; (2) internal exposure caused by inhalation ofre suspended radionudide;

-AND-(3) internal exposure from ingesting ground water.

The staff has reviewed the licensee's calculational methods and assumptions and finds that they are consistent with NUREG-I 101, "O~nsite Disposal of Radioactive Waste," Volumcs I and 2. November 1986 and February 1987. respectidely.

Ihe staff finds the assessment medkodol.gy a&ciable. Table 2 lists the doses calculated by the licensee for the maxinally exposed member of tie public based on a total activity of 8.9 nmCi disposed in that ytar.

E e >^ @TABLE 2 Paihway Whole Body Dose Received by Maximally Exposed Individual (mrem/year)

Gmnundshine 0.94 Inalation 0.94 Groundwater Ingestion . 0.73 Tot:al 2.61 On July 5, 1991. ihc lioense resampled the onsite disposal area to assure that no significant impa.ts and adverse cflets had occurred. A counting procedure based on the appropriate environmental low-level doses was used by the licensee; howevcr, no activity was detected during the re-samplingi. This is consistent with the original activity of the material and Ihe decay time The 1991 re-sampling process used by the licensee confims that tie environmental impact of the 1982 disposal was vey smalL The staff finds the licensee's methodology acceptable.

4. ENVIRONMENTALFINDING AND CONCLUSION The staff has evaluated dte cvironntal impact of the proposal to leave in place approximately 942 cubic meters of sligly contaminated sludge underneath the upper parkng lot on the Donald C. Cook Nudear Plnt site.

In 1982. the licensce evaluated the potential exposure to members ofthe general public from the radionuidides in the sludgc and calculated the potential dose to the maximally exposed member of the public, based on a total activity of 8.89 mCi disposed iii that years to be 2;61 mrem/yr. The staff has reviewed the licensens calculational methods and assumptions and fownd that they are consistent with NUREG-1101, OAsitc Disposal of Radioactive Waste, Volumes 1 and 2, November 1986 and Febmuary 1987. respectively. The staff finds the assessment methodology acceptable. For compariso the radiation from the naturally occurring radionuclides in soils and rocks plus cosmic radiation gives a person in Midhigan a whole-body dose rate of about 89 mrem per year outdoos Subsequent licensee sampling in 1991 identified no detectable activity. The staff evaluated the licensee's sampling and analysis methodology and finds it acceptable The results, of the 1991 re-sampling by the licensee.

confirm that the environmental impact of the 1982 disposal was very small.

Based on the above ffic staff finds that the potential environmental impacts of leaving the contaminated sludge in plaa: are insignificant With regard to the non-radiological impacts, the staffhas determined that leaving the soil in place represent; the -

least impact to the environment.

ft - Reference I PMP-6010-OSDO001 I Rev. 19 1- Page 84 of 84 OFF-SITE DOSE CALCULATION MANUAL Ali' nt 3.24 Safety Evaluation By The Office Of Nuclear Pages.-:,

It32 Reactor Regulation 82 - 84 S. CONCLUSION Based on the stffs review of the licaenses discussion, the stafffinds the licensee's proposal to retain the material in its present location as documented in this Safety Evaluation acceptable. Also, is Safety Evaluation shall be permanently mcorporated as an appendix to the licensee's Offsite Dose CalculationManual (ODCM), and anyfuture modifications shall be reported to NRC in accordancc with the applicable ODCM change protocoL a-I&M letter fromnF R-Fitzpatbick to thc NRC Document Control Desk, September 29, 1993 Therefore, thelicensec's proposal to consider the slightly contaminated sludge disposed by idention in place in the manner deseribed in time Donald C. Cook Nuclear Plant submittals date Odober 9, 1991, October 23, 1991, September 3, 1993, and September29, 1993, is acceptable.

Ihe guidelines used by the NRC staff for onsite disposil of licensed material and the staffs evaluation of how each guideline has been satisfied arc given in Table 3.

Pusantto 10 CFR51.32, theCommission has deuter-nned that grantirn ofthisapprovalwillhave no significantimpact onthe environment (October 31, 1994, 59 R 544Th: .

Principal Conuibutor.J. mins Date: November 10, 1994 L

TABLE 3 20.2002 GUIDELINE FOR ONSITE STAFF'S EVALUATION L DISPOSAL_

I. Due to thenature of the disposed material, recycling to the 1.

2.

The radioactive naterial should be disposed ofin such a manner that it is unlikely that the materil would be recycled.

Doses to the total bodyand anybody organ of a 2.

general public is not considered likely.

This guideline was addressed in Table 2. Although the L

maximally exposed individuals (a member of the general public or a non-oceupationally exposed worker) from the probable pathways of exposure to the disposed matenial should be less than 1 mremnlyar.

2.61 mrmra is greater than staffs guidelines, the stafffinds it acceptable due to 9 yrs decay following analysis and the cxpected lack of activity detedced in the 1991 survey. L

3. Doses to the totalbody and anybody organ of an inadvertent intruder fiom the probable pathways of exposure should be less thm S mremnyear.

3. Because ihe material willbe land-spread, the staff considers the maxinally exposed individual senario to also address the intruderscen aio. L

4. Doses to the total body and anybody organ of an 4. Even ifrecycling were to occur after release from regulatory individual from assumed recycling of the disposed material at the time the disposl site is released from regulatory control from all likely pathways of exposure should beless than 1 mra.

control the dose to a maximally exposed mcmber ofthe public is not expected to exceed I zmrenear, based on expaosu scenarios considered in this analysis. L 2 . p. Branagin, Jr. and F. J. Conge 'Dispod of Conlamninated Radioactive Wastes fn NuclearPower Pls,"presented at the Health Physics Societys Mid-Year Synmposium on Health Physics Consideation in 1econtauinationlDecomissioning L

Knoxville, Tennessee, Pebruary 1986, (CONF-860203)

L L.

L

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