ML080920393

Submittal of Annual Radioactive Effluent Release Report, January 1, 2007, Through December 31, 2007. Pages A3.0-1, Off-Site Dose Calculation Manual, PMP-6010-OSD-001, Revision 22 Through End

ML080920393
Person / Time
Site:	Cook
Issue date:	03/28/2008
From:	Indiana Michigan Power Co
To:	Office of Nuclear Reactor Regulation
References
AEP:NRC:8691-01, FOIA/PA-2010-0209 PMP-6010-OSD-001, Rev 22
Download: ML080920393 (95)

Text

-OFF-SITE DOSE CALCULATION MANUAL The Off-Site Dose Calculation Manual, PMP-6010-OSD-001, was revised during this reporting period. A copy of Revision 22 is included as part of the report. The reasons for the changes and the PORC approval are documented on the Review and Approval tracking form. 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 impact the accuracy or reliability of effluent, dose, or setpoint calculations.

A3.0-1

Document No.: PMP-6010-OSD-001 Revision No.: 022

Title:

OFF-SITE DOSE CALCULATION MANUAL Alteration Category: Minor Revision CDI/50.59 No(s): N/A PORC Mtg. No.: 4307 CR No.:

Superceding Procedure(s): N/A Temporary Procedure Expiration Date:

Temporary Change Expiration Date:

Temporary Procedure/Change Ending Activity: N/A Effective Date: 14-Jun-2007 12:00:00 AM Approvals Name Review/Approval Type/Capacity Date Wohlgamuth, Craig Cognizant Organization 16-May-2007 11:24:54 AM Hershberger, Robert Programmatic-Other 21-May-2007 08:53:49 AM Mutz, Brian Programmatic-Surveillance Section 21-May-2007 01:11:28 PM Beer, Joe Programmatic-ALARA 22-May-2007 03:48:17 PM Walker, David N Cross Discipline-BSP 24-May-2007 08:14:15 AM Turinetti, Dale Programmatic-Other 24-May-2007 12:40:16 PM Harner, Jon Technical Review 04-Jun-2007 10:39:00 AM Newmiller, Julie PORC I1-Jun-2007 12:10:49 PM Mottl, Teri Approval Authority 12-Jun-2007 08:41:46 AM Signature Comments PROCEDURE APPROVAL AUTHORIZED BY L. WEBER, PLANT MANAGER. TLM Document Number: PMP-6010-OSD-001 Revision: 022 Page 1 of I

ZAMERWIAM' POWER PMP-6010-OSD-001 Rev. 22 Page 1 of 91 OFF-SITE DOSE CALCULATION MANUAL Reference Doug Foster Environmental Manager Environmental Writer Document Owner Cognizant Organization TABLE OF CONTENTS 1 PURPOSEAND SCOPE ........................................................................... 4 2 DEFINITIONS AND ABBREVIATIONS ...................................................... 4 3 DETAILS ................................................................................................ 6 3.1 Calculation of Off-Site Doses ............................................................ 6 3.1.1 Gaseous Effluent Releases ...................................................... 6 3.1.2 Liquid Effluent Releases ..................................................... 11 3.2 Limits of Operation and Surveillances of the Effluent Release Points ............. 14 3.2.1 Radioactive Liquid Effluent Monitoring Instrumentation ................ 14 3.2.2 Radioactive Gaseous Effluent Monitoring Instrumentation .............. 16 3.2.3 Liquid Effluents .................................................................... 17

a. Concentration Excluding Releases via the Turbine Room Sump (TR S) D ischarge ............................. : ................................ 17

b. Concentration of Releases from the TRS Discharge .................... 17 c . D ose ............................................................................ 18

d. Liquid Radwaste Treatment System ..................................... 18 3.2.4 Gaseous Effluents .............................................................. 21

a. D ose Rate ................................................................... 21

b. Dose - Noble Gases ...................................................... 21

c. Dose - Iodine-131, Iodine-133, Tritium, and Radioactive Material in Particulate Form ........................................................... 22

d. Gaseous Radwaste Treatment ........................................... 22 3.2.5 Radioactive Effluents - Total Dose ............................................. 25 3.3 Calculation of Alarm/Trip Setpoints ...................................................... 26 3.3.1 Liquid Monitors .................................................................... 27

a. Liquid Batch Monitor Setpoint Methodology ......................... 27

b. Liquid Continuous Monitor Setpoint Methodology ..................... 28 3.3.2 Gaseous Monitors .............................................................. 30

a. Plant Unit Vent ............................................................ 30

b. Waste Gas Storage Tanks ................................................ 33

c. Containment Purge and Exhaust System .................................. 34

d. Steam Jet Air Ejector System (SJAE) ..................................... 35

e. Gland Seal Condenser Exhaust .......................................... 35

ZAMEDICAN POW .PMP-6010-OSD-001 Rev. 22 Page 2 of 91 OFF-SITE DOSE CALCULATION MANUAL Reference Doug Foster Environmental Manager Environmental Writer Document Owner Cognizant Organization 3.4 Radioactive Effluents Total Dose ....................................................... 36 3.5 Radiological Environmental Monitoring Program (REMP) ...................... 36 3.5.1 Purpose of the REMP ........................................................ 36 3.5.2 Conduct of the REMP ............................................................ 37 3.5.3 Annual Land Use Census ..................................................... 39 3.5.4 Interlaboratory Comparison Program ........................ 40 3.6 Meteorological Model ........................................... 40 3.7 Reporting Requirements ................................................................. 41 3.7.1 Annual Radiological Environmental Operating Report (AREOR) ..... 41 3.7.2 Annual Radiological Effluent Release Report (ARERR) .................. 42 3.8 10 CFR 50.75 (g) Implementation ................................................... 43 3.9 Reporting/Management Review ........ .......

... .............................. 44 4 FINAL CONDITIONS .......................................................................... 44 5 REFERENCES ..................................................................................... 44 SUPPLEMENTS .1 Dose Factors for Various Pathways .............................................. Pages 48 - 51 .2 Radioactive Liquid Effluent Monitoring Instruments ................... Pages 52 - 54 .3 Radioactive Liquid Effluent Monitoring Instrumentation Surveillance Requirem ents ............................................................ Pages 55 - 56 .4 Radioactive Gaseous Effluent Monitoring Instrumentation .......... Pages 57 - 59 .5 Radioactive Gaseous Effluent Monitoring Instrumentation Surveillance Requirem ents ............................................................ Pages 60 - 61 .6 Radioactive Liquid Waste Sampling and Analysis Program ........ Pages 62 - 63 .7 Radioactive Gaseous Waste Sampling and Analysis Program ..... Pages 64 - 65 .8 Multiple Release Point Factors for Release Points ................................ Page 66

• AMMRIAM*

POELRIC PMP-6010-OSD-001 Rev. 22 Page 3 of 91 OFF-SITE DOSE CALCULATION MANUAL Reference Doug Foster Environmental Manager Environmental Writer Document Owner Cognizant Organization .9 Liquid Effl uent Release Systems ........................................................... Page 67 .10 Plant Liquid Effl uent Param eters ........................................................... Page 68 .11 Volumetric Detection Efficiencies for Principle Gamma Emitting Radionuclides for Eberline Liquid Monitors ........................... Page 69 .12 Counting Efficiency Curves for R-19, and R-24 ........................... Pages 70 - 71 .13 Counting Efficiency Curve for R-20, and R-28 ..................................... Page 72 .14 Gaseous Effluent Release Systems ...................... Page 73 .15 Plant Gaseous Effl uent Parameters ............................... ..................... Page 74 .16 10 Year Average of 1995-2004 Data ................... ........................ Pages 75 - 76 .17 Annual Evaluation of x/Q and D/Q Values For All Sectors ................. Page 77 .18 D ose F actors .......................................................... ..................... Pages 78 - 79 . .19 Radiological Environmental Monitoring Program Sample Stations, Sample Types, Sample Frequencies ............................... Pages 80 - 83 .20 Maximum Values for Lower Limits of DetectionsAB - REMP ..... Pages 84 - 85 .21 Reporting Levels for Radioactivity Concentrations in Environmental Samples .................................. Page 86 .22 On-Site M onitoring Location - REM P .................................................. Page 87 .23 Off-Site Monitoring Locations - REMP ...................... Page 88 .24 Safety Evaluation By The Office Of Nuclear Reactor Regu lation ...................................................................................... P ages 89 - 9 1

Reference PMP-6010-OSD-001 Rev. 22 Page 4 of 91 OFF-SITE DOSE CALCULATION MANUAL 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.

The Off-Site Dose Calculation Manual (ODCM) is the top tier document for the Radiological Environmental Monitoring Program (REMP), 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.

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.

The ODCM provides flow diagrams detailing the treatment path and the major 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.

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.

2 DEFINITIONS AND ABBREVIATIONS Term: Meaning:

S or shiftly At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> D or daily At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 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.

S/U Prior to each reactor startup P Completed prior to each release B At least once per 24 months Sampling evolution Process of changing filters or obtaining grab samples

Reference PMP-6010-OSD-001 I Rev. 22 Page 5 of 91 OFF-SITE DOSE CALCULATION MANUAL Member(s) of All persons who are not occupationally associated with the Public 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 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.

Total Fractional Level (TFL) Total Fractional Level is defined as:

TFL -C(1) + C(2) . >.

L(j) L( 2)

Where; c() = Concentration of l.detected nuclide C(2) = Concentration of 2 "ddetected nuclide L() = Reporting Level of 1St nuclide from Attachment 3.21, Reporting Levels for Radioactivity Concentrations in Environmental Samples.

L(2) = Reporting Level of 2 nd nuclide from Attachment 3.21, Reporting Levels for Radioactivity Concentrations in Environmental Samples.

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.

Reference PMP-6010-OSD-001 t Rev. 22 Page 6 of 91 OFF-SITE DOSE CALCULATION MANUAL 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:

• MIDER 0 MIDEX

• MIDEL

• MIDEG

• MIDEN

b. The subprogram used to enter and edit gaseous release data is called MD1EQ (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, Df6 air = Z_*[(Mi or Nd)* Q

• 3.1 7E- 8]

Q Where; Dy Dp air = the gamma or beta air dose in mrad/yr to an individual receptor

-' / Q =the annual average or real time atmospheric dispersion factor over land, sec/m3 from Attachment 3.16, 10 Year Average of 199.5-2004 Data Mi = the gamma air dose factor, mrad in3 / yr ACi, from Attachment 3.18, Dose Factors Ni = the beta air dose factor, mrad m3 / yr MtCi, from Attachment 3.18, Dose Factors

Reference PMP-6010-OSD-001 I Rev. 22 Page 7 of 91 OFF-SITE DOSE CALCULATION MANUAL Qi the release rate of radionuclide, "i", in /Ci/yr.

Quantities are determined utilizing typical concentration times volumes equations that are documented in 12-THP-6010-RPP-601, Preparation of the Annual Radioactive Effluent Release Report.

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

d. The value for the ground average " /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).

=IQ 2.03 Tf 11 mg *X*Xg Where; F2

minimum of or -,F

3 Z3 x =distance downwind of the source, meters. This information is found in parameter 5 of MIDEX.

urn = wind speed for ground release, (meters/second) o7 = vertical dispersion coefficient for ground release, (meters),

(Reg. Guide 1.111 Fig. 1)

Hc = building height (meters) from parameter 28 of MIDER.

(Containment Building = 49.4 meters)

Tf = terrain factor (= 1 for Cook Nuclear Plant) because iI,we consider all our releases to be ground level (see parameter 5 in MIDEX).

2.03 = -2 + 0.393 radians(22.5')

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.

Reference PMP-6010-OSD-001 I Rev. 22 Page 8 of 91 OFF-SITE DOSE CALCULATION MANUAL

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

• Z (Q,

• X/Q

• Sf

• DFBi)

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 E-15 of Reg. Guide 1.109)

DFB = the whole body dose factor from Table B-I of Reg.

Guide 1.109, mrem - m3 per gCi - yr. See Attachment 3.18, Dose Factors.

Q= the release rate of radionuclide "i", in ICi/yr

2. Skin Plume Pathway (Eq 11)

Dose (mrem/yr)= 3.17E -8 * -* [f(Qi 1.1 *DFO)+ Z(Qi *DFS)]

Q Where; 1.11 = conversion factor, tissue to air, mrem/mrad DF il = the gamma air dose factor for a uniform semi-infinite cloud of radionuclide "i", in mrad m3/ItCi yr from Table B-1, Reg. Guide 1. 109. See Attachment 3.18, Dose Factors.

DFS = the beta skin dose factor for a semi-infinite cloud of radionuclide "i", in mrem m 3/1 iCi yr from Table B-I, Reg. Guide 1.109. See Attachment 3.18, Dose Factors.

Reference PMP-6010-OSD-001 Rev. 22 Page 9 of 91 OFF-SITE DOSE CALCULATION MANUAL

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

The dose, Dip 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:

Dp (mrem/year)= 3.17E - 8

• Z(Ri

• W

• Q)

Where; R = the most restrictive dose factor for each identified radionuclide "i", in m 2 mrem sec / yr [Ci (for food and ground pathways) or mrem m3 / yr [Ci (for inhalation pathway), for the appropriate pathway For sectors with existing pathways within five miles 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 Pathways, for the maximum Ri values for the most controlling age group for selected radionuclides. R 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 further defined as:

Wi -- X / Q for the inhalation pathway, in sec/m 3

-OR-Wfg = D /Q for the food and ground pathways in 1/mn2 Qic = 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 uCi/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. 22 Page 10 of 91 OFF-SITE DOSE CALCULATION MANUAL 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 = Ci

• GPM

• time on flash tank (min)

• 3. 785E -3 ml Where; 3.785E-3 = conversion factor, ml Ci/ttCi gal.

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 ItCi/g dose equivalent 1-131.

Reference PMP-6010-OSD-001 ý Rev. 22 Page 11 of 91 OFF-SITE DOSE CALCULATION MANUAL

6. IF the specific activity of the secondary coolant system is less than 0.01 tCi/g dose equivalent 1-131, THEN the release rate must be determined once every six months. Use the following plant established equation:

Qv = Ci

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

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 DS1LI (LD) is used. It computes the individual dose for up to 5 receptors for 14 liquid 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.

Reference PMP-6010-OSD-001 I Rev. 22 Page 12 of 91 OFF-SITE DOSE CALCULATION MANUAL

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

1. Potable Water (Eq 1)

R,,j= 1100 l O0, U1,_ * "Q

• Dipj e-A'lp Mp*F*2.23E-3 Where; Rapj = 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 "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.

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, ft3 min / sec gal Q = 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-1 1 through E-14 of Reg. Guide 1.109 and are located within the MIDAS code.

X = the radioactive decay constant for radionuclide "i", in hours'

Reference PMP-6010-OSD-001 I Rev. 22 Page 13 of 91 OFF-SITE DOSE CALCULATION MANUAL tp = the average transit time required for nuclides to reach the point of exposure, 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. 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 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />)

2. Aquatic Foods (Eq 2)

Rapj =11O* Uap Mp*F*2.23E-3

• XI Q* Bip

• Daipj eh' Where, Bp- 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.

tp = the average transit time required for nuclides to reach the point of exposure, 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. 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 <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />)

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

3. Shoreline Deposits (Eq 3)

Rap =

10000 Uap*W - E i* T.* Daipj [e-AIp]* [1 -e _Yb]

Mp*F*2.23E-3 i 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.

T = the radioactive half-life of the nuclide, "i", in days Daipj = the dose factor for standing on contaminated ground, in mrem m2 / fir 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.

Reference T PMP-6010-OSD-001 Rev. 22 Page 14 of 91 OFF-SITE DOSE CALCULATION MANUAL tb = the period of time for which sediment or soil is exposed to the contaminated water, 1.31E+5 hours. Given in MIDEL as item 6 of parameter 4.

tp = the average transit time required for nuclides to reach the point of exposure, 0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />. Given as #28 of parameter 4 in MIDEL.

110,000 = conversion factor yr ft3 pCi / Ci sec m2 day, this accounts for proportionality constant in the sediment radioactivity model 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.

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

1. Irrigation rate = 0

2. Fraction of time on pasture = 0

3. Fraction of feed on pasture = 0

4. Shore width factor - 0.3 (from Reg. Guide 1.109, Table A-2)

f. The results of DS 1LI 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 of DS1LI to. find the maximum liquid pathway individual doses. The 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 Annual Radioactive Effluent Release Reports, required by Reg. Guide 1.21.

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

Reference PMP-6010-OSD-001 Rev. 22 Page 15 of 91 OFF-SITE DOSE CALCULATION MANUAL

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 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 (TRS) 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 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 system leak.

[Ref 5.2. lgg]

Reference PMP-6010-OSD-001 Rev. 22 Page 16 of 91 OFF-SITE DOSE CALCULATION MANUAL 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 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.

c. With a radioactive gaseous process or effluent monitoring instrumentation 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 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 Instrumentation, with a maximum allowable extension not to exceed 25 % 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 document.

e. Determine the setpoints in accordance with the methodology as described in step 3.3.2, Gaseous Monitors. Record the setpoints.

f. 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 Attachment 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.

Reference PMP-6010-OSD-001 Rev. 22 Page 17 of 91 OFF-SITE DOSE CALCULATION MANUAL 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 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 gCi/ml total activity.

2. With the concentration of radioactive material released from the site via the Batch Release Tanks or Plant Continuous Releases (other than 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.

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 2E-4 uCi/ml 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.

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.

Reference PMP-6010-OSD-001 Rev. 22 Page 18 of 91 OFF-SITE DOSE CALCULATION MANUAL

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 mrem/unit to the total body and to __5 mrem/unit to any organ, and during any calendar year to < 3 mrem/unit to the total body and to _<10 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 3.2.3c. 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 individuals to radiation and radioactive material, including, as appropriate:

a) Estimate of each individual's dose. This is to include the radiological impacts on finished drinking water supplies with 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 involved, c) Cause of elevated exposures, dose rates or concentrations,

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

These reports must be formatted in accordance with PMP-7030-001-002, Licensee Event Reports, Specialand Routine Reports, even though this is not an LER.

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/unit to the total body or 0.2 mrem/unit to any organ.

2. Project doses due to liquid releases to UNRESTRICTED AREAS at least once per 31 days, in accordance with this document.

Reference PMP-6010-OSD-001 Rev. 22 Page 19 of 91 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 NorthBoric 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 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:

(C)V)+ (Co)(WI) ='(CQ(V,)

Where; Ci = 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 Va = the volume of the water that is being added to the system Ct = the final concentration of the system after the addition Vt = the final volume of the system after the addition The intent is to keep the:

• WDS below 500 Vinmhos/cc.

• TRS below lE-5 jiC/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 of job order activities, and/or activities causing increased production of waste water.

Reference PMP-6010-OSD-001 I Rev. 22 Page 20 of 91 OFF-SITE DOSE CALCULATION MANUAL 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 not result in exposures greater than 1) the Section II.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 II.A, III.A, and IV.A of Appendix I, 10 CFR Part 50. The dose limits iniplement the guides set forth in Section II.A of Appendix I. 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 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 b' 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 requirements of 40 CFR 141. The dose calculations in the ODCM implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data such that the actual exposure of 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 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 1.113, "Estimating Aquatic Dispersion of Effluents from Accidental and Routine Reactor Releases for the Purpose of Implementing Appendix I", 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.

Reference PMP-6010-OSD-001 Rev., 22 Page 21 of 91 OFF-SITE DOSE CALCULATION MANUAL 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 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 II.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 II.A of Appendix I, 10 CFR 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 mrem/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. 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 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 _ 5 mrad/unit for gamma radiation and < 10 mrad/unit for beta radiation and during any calendar year, to

• 10 mrad/unit for gamma radiation and

• 20 mrad/unit for beta radiation.

Reference PMP-6010-OSD-001 I Rev. 22 Page 22 of 91 OFF-SITE DOSE CALCULATION MANUAL

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 event.

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

c. Dose - Iodine-131, 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 wvith 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 mrem/unit to any organ b) During any calendar year to less than- or equal to 15 mrem/unit to any organ.

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 mrad/unit 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 organ.

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

Reference T PMP-6010-OSD-001 ý Rev. 22 I Page 23 of 91 OFF-SITE DOSE CALCULATION MANUAL BASES -- GASEOUS EFFLUENTS 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 appropriately controlled. It provides operational flexibility for releasing gaseous effluents to satisfy the Section II.A and II.C 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 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 mrem/yr 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. 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.

DOSE, NOBLE GASES This specification is provided to implement the requirements of Sections II.B, III.A, and IV.A.

of Appendix I, 10 CFR Part 50. The dose limits implement the guides set forth in Section II.B of Appendix I.

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 radioactive material in gaseous effluents will be kept "as low as is reasonably achievable".

The 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 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 I", Revision 1, October 1977 and Regulatory Guide 1.111, "Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors",

Revision 1, July 1977. 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.

Reference T PMP-6010-OSD-001 I Rev. 22 1 Page 24 of 91 OFF-SITE DOSE CALCULATION MANUAL DOSE, RADIOIODINES, RADIOACTIVE MATERIAL IN PARTICULATE FORM, AND RADIONUCLIDES OTHER THAN NOBLE GASES This specification is provided to implement the requirements of Sections II.C, III.A, and IV.A of Appendix I, 10 CFR Part 50. The dose limits are the guides set forth in Section II.C of Appendix I.

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 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 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 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 CFR Part 50, Appendix I", Revision 1, October 1977 and Regulatory Guide 1.111, "Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors", Revision 1, July 1977. These equations also provide 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, and 4) deposition on the ground with subsequent exposure of man.

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 II.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 CFR Part 50, for gaseous effluents.

Reference PMP-6010-OSD-001 I Rev. 22. Page 25 of 91 OFF-SITE DOSE CALCULATION MANUAL J 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 twice the 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 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 CFR 190 and including the specified information of paragraph 190.1 (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 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 PMP-6010-OSD-001 Rev. 22 Page' 26 of 91 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 Alarm/Trip 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 (MRP) 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 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 Factors for Release Points.

The Site stance on instrument uncertainty is taken from HPPOS-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.

Reference PMP-6010-OSD-001 Rev. 22 Page 27 of 91 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, Column 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. This monitor is identified as RRS-1000. Steam Generator 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 sampling program. The sampling program determines the nuclides and concentrations of those nuclides prior to release. The 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 CFR 20, Appendix B, Table 2, Column 2. The equation to calculate the flow rate is from Addendum AA1 of NUREG-0133:

L LMTCi LIMTTi Mf _<F+f ARP Where; Ci = the concentration of nuclide "i" in .tCi/ml LIMIT = the 10 CFR 20, Appendix B, Table 2, Column 2 limit of nuclide "i" in p.Ci/ml 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 number of circulation pumps in operation.

Reference PMP-6010-OSD-001 Rev. 22 Page 28 of 91 OFF-SITE DOSE CALCULATION MANUAL 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.

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-24ý 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 (BDT), and essential service water (ESW) systems.

2. These Westinghouse monitors (R) are being replaced by Eberline monitors (DRS) and are identified as:

• R-19 or DRS 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.

Reference PMP-6010-OSD-001 I Rev. 22 Page 29 of 91 OFF-SITE DOSE CALCULATION MANUAL

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.

4. The equation used to determine the setpoint for continuous monitors is from Addendum AA1 of NUREG-0133:

SC*Eff *MRP*F*SF f

Where; Sp = setpoint of monitor (cpm)

C = 5E-7 pCi/ml, maximum effluent control limit from 10 CFR 20, Appendix B, Table 2, Column 2 of a known possible nuclide in effluent stream. (The 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 Sr90 is found. The concentration limit shall be adjusted appropriately.)

-OR-if a mixture is-to be specified, Zc, Ci LIMITi Eff = Efficiency, this information is located in Attachment 3.11, 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:

Y( Ci

• Effi) replaces C

• Eff Ci LIMITi

Reference PMP-6010-OSD-001 Rev. 22 Page 30 of 91 OFF-SITE DOSE CALCULATION MANUAL MRP = 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 (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 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).

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)..

Gaseous effluent monitor high alarm setpoints will routinely be established at a fraction of the maximum 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.

NOTE: IF the setpoint calculation methodology changes or the associated factors change for Unit Vent, Air Ejector and/or Gland Seal monitors, THEN initiate a review by Emergency Planning to ensure that the requirements of 10 CFR 50.54 (q) are maintained.

a. Plant Unit Vent

Reference PMP-6010-OSD-001 I Rev. 22 Page 31 of 91 OFF-SITE DOSE CALCULATION MANUAL 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:

Sp = SF

• MRP

• DLj Fp */Q*

r (Wi* DCF.)

Where; Sp = the maximum setpoint of the monitor in laCi/cc 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 arbitrary 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 XQp (or Xfp) for all release points.

0 Ratio Qp/EQp (or fp/>lfp) for each release point..

This ratio is the MRP for that specific release point 0 Repeat the above bullets for each of the site's eight gaseous release points.

Fp the maximum volumetric flow rate of release point "p",

at the time of the release, in cc/sec. The maximum Unit Vent flow rate, by design, is 186,600 cfm for Unit 1 and 143,400 cfm for Unit 2.

DLj = dose rate limit to organ "j" in an unrestricted area (mrem/yr).

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

" Total Body < 500 mrem/year

" Skin < 3000 mrem/year

" Any Organ< 1500 mrem/year

Reference PMP-6010-OSD-001 I Rev. 22 Page 32 of 91 OFF-SITE DOSE CALCULATION MANUAL

• / Q = The worst case annual average relative concentration in the applicable sector or area, in sec/m 3 (see Attachment 3.16, 10 Year Average of 1995-2004 Data).

Wi = weighted factor for the radionuclide:

-Ci ZC, Where, Ci = 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.

DCFij dose conversion factor used to relate radiation dose to organ "j", from exposure to radionuclide "i" in mrem in 3./ yr piCi. See following equations.

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

For the whole body: DCF1 j = I Where; Ki = whole body dose factor due to gamma emissions for each identified noble gas radionuclide in mrem m3 / yr ýiCi. See Attachment 3.18, Dose Factors.

For the skin: DCFij = L + 1.1MI Where; Li skin dose factor due to beta emissions for each identified noble gas radionuclide, in mrem m3 / yr ýtCi. 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).

Reference PMP-6010-OSD-001 I Rev. 22 Page 33 of 91 OFF-SITE DOSE CALCULATION MANUAL Mi = the air dose factor due to gamma emissions for each identified noble gas radionuclide in mrad m3 / yr ptCi. See Attachment 3.18, Dose Factors.

For the thyroid, via inhalation: DCFij = Pi Where; Pi = the dose parameter, for radionuclides other than noble gas, for the inhalation pathway in mrem m3 / yr ptCi (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 mrem/yr, 3000 mrem/yr, and 1500 mrem/yr 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 ( AiCi/cc)

• Max Flowrate(cfm) New Max cin New Max Concentration(,aCi/cc)

-OR-Max Conc (,aCi/cc)* Max Flowrate(cfim) - New Max jiCi/cc New Max Flowrate(cfm)

b. Waste Gas Storage Tanks

Reference T PMP-6010-OSD-001 I Rev. 22 Page 34 of 91 OFF-SITE DOSE CALCULATION MANUAL

1. The gaseous effluents discharged from the Waste Gas System are monitored 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 alarm 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 System Hold 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 towaive the 45-day decay for safety's sake.

c. Containment Purge and Exhaust System

1. Thegaseous effluents discharged by the Containment Purge and Exhaust Systems and Instrumentation Room Purge and Exhaust System are monitored by the plant vent radiation monitor noble gas channels (VRS-1505 for 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 from the containment atmosphere is drawn through a closed, sealed system to 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 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-1 101/1201 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 PMP-6010-OSD-001

• Rev. 22 Page 35 of 91 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- 1300/2300 or VRS- 1101/2101) and one of the two Train B monitors (ERS-1400/2400 or VRS-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-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 maximum air ejector exhaust flow rate, (Attachment 3.15, Plant Gaseous Effluent Parameters). The alarm setpoint valuewill be established using the following unit analysis equation:

=SF *MRP* DLj SSJAE */ SF * *(w,*DCF)

Where; SSJAE = the maximum setpoint, based on the most limiting organ, in lICi/cc and where the other terms are as previously defined

e. Gland Seal Condenser Exhaust

1. The gaseous effluents from the Gland Seal Condenser Exhaust discharged to the environment are continuously monitored by 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 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:

Reference PMP-6010-OSD-001 Rev. 22 Page 36 of 91 OFF-SITE DOSE CALCULATION MANUAL SGSCE SF *MRP

• DLj Fp* X/Q* (Wi* DCFj)

Where; SGSCE = the maximum setpoint, based on the most limiting organ, in ptCi/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 REMP is to:

0 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 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 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 REMP for the Donald C. Cook Nuclear Plant.

Reference PMP-6010-OSD-001 I Rev. 22 Page 37 of 91 OFF-SITE DOSE CALCULATION MANUAL 3.5.2 Conduct of the REMP [Ref. 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 Frequencies, Attachment 3.20, Maximum Values for Lower Limits of DetectionsA,B - 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.

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).

3. 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.

Reference PMP-6010-OSD-001 Rev. 22 Page 38 of 91 OFF-SITE DOSE CALCULATION MANUAL NOTE: Only one report per event is required.

NOTE: Radioactivity from sources other than plant effluents do not require a Special Report.

4. IF any of the following conditions are identified:

• A radionuclide associated with plant effluents is detected in any REMP sample medium AND its concentration exceeded the limits specified in Attachment 3.21, Reporting Levels for Radioactivity Concentrations in Environmental Samples,

• More than one radionuclide associated with plant effluents is detected in any REMP sample medium AND the Total Fractional Level, when averaged over the calendar quarter, is greater than or equal to 1.

THEN complete the following steps, as applicable:

• Submit a Special Report to the Nuclear Regulatory Commission within 30 days.

• Submit a Special Report to designated state and local organizations for groundwater or surface water media which could. be used as drinking water.

• Evaluate the following items for inclusion in Special Reports:

1) Release conditions

2) Environmental factors

3) Corrective actions

4) Additional factors which may have contributed to the identified levels

5. WHEN submission of a Special Report to designated state and local organizations is required, THEN perform the following:

• Communicate event specific information to designated state and local organization personnel by the end of the next business day.

Reference PMP-6010-OSD-001 Rev. 22 Page 39 of 91 OFF-SITE DOSE CALCULATION MANUAL

6. IF a currently sampled milk farm location becomes unavailable, THEN conduct a special milk farm survey within 15 days.

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

b) IF the unavailable location was a background farm, THEN 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, THEN perform monthly vegetation sampling in lieu of milk sampling when vegetation is available.

BASES - RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM (REMP)

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 accordance with the requirements of Technical Specification 5.5.1 .c.

The detection capabilities, required by Attachment 3.20, Maximum Values for Lower Limits of DetectionsA,B - 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 LLDs 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 LLDs unachievable. In such cases, the contributing factors will be identified and described in the Annual Radiological Environmental Operating Report.

3.5.3 Annual Land Use Census [Ref. 5.2.lu]

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.

Reference PMP-6010-OSD-001 I Rev. 22 Page 40 of 91 OFF-SITE DOSE CALCULATION MANUAL

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.

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.

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 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 vegetables assumed in Regulatory Guide 1.109 for consumption of a child. To determine this minimum 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

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 for 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 program for environmental monitoring in order to demonstrate the results are reasonably valid.

3.6 Meteorological Model

Reference PMP-6010-OSD-001 Rev. 22 Page 41 of 91 OFF-SITE DOSE CALCULATION MANUAL 3.6.1 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 shoreline to determine the meteorological parameters associated with unmodified shoreline air. The data is accumulated by microprocessors at the tower sites and normally transferred to the central computer every 15 minutes.

3.6.2 The central computer uses a meteorological software program to provide atmospheric dispersion and deposition parameters. The meteorological model used is based on guidance provided in Reg. Guide 1.111 for routine releases. All calculations use the Gaussian plume model.

3.7 Reporting Requirements 3.7.1 Annual 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 15 of each year. [Ref 5.2.1j, TS 5.6.2]

b. Include in the AREOR:

• Summaries, interpretations, and statistical evaluation of the results of the radiological environmental surveillance activities for the reporting period.

• 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.

2

• 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.

Reference T PMP-6010-OSD-001 I Rev. 22 Page 42 of 91 OFF-SITE DOSE CALCULATION MANUAL 0 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.

3.7.2 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.

[Ref 5.2.1j, TS 5.6.3]

b. 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 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 an hour-by-hour listing of wind speed, wind direction, atmospheric stability, and precipitation (if measured) on magnetic tape, or in the form of joint 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.

Reference PMP-6010-OSD-001 1 Rev. 22 Page 43 of 91 OFF-SILTE DOSE CALCULATION MANUAL Inoperable radiation monitor periods exceeding 30 continuous days; explain causes of inoperability and actions taken to prevent reoccurrence.

d. Submit the ARERR [Ref. 5.2.1 w] 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 resin, compacted dry waste, evaporator bottoms),

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

-AND-

. 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 quarterly basis.

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

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 divided by two.

3.8 10 CFR 50.75 (g) Implementation

Reference PMP-6010-OSD-001 I Rev. 22 Page 44 of 91 OFF-SITE DOSE CALCULATION MANUAL 3.8.1 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.

3:8.2 These records shall include any known information or identification of involved nuclides, quantities, and concentrations.

3.8.3 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.

3.9 Reporting/Management Review 3.9.1 Incorporate any changes to this procedure in the ARERR.

3.9.2 Update this procedure when the Radiation Monitoring System, its instruments, or the specifications of instruments are changed.

3.9.3 Review or revise this procedure as appropriate based on the results of the land use census and REMP.

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

3:9.5 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 / Q and D / Q 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.

5 REFERENCES 5.1 Use

References:

Reference PMP-6010-OSD-001 I Rev. 22 Page 45 of 91 OFF-SITE DOSE CALCULATION MANUAL 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 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 CFR 50, Domestic Licensing of Production and Utilization Facilities

c. PMI-6010, Radiation Protection Plan

d. NUREG-0472

e. NUREG-0133

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

g. Regulatory Guide 1.111

h. Regulatory Guide 1.113

i. Final Safety Analysis Report (FSAR)

j. Technical Specifications 5.4.1.e, 5.5.1.c, 5.5.3, 5.6.2, and 5.6.3

k. Final Environmental Statement Donald. C. Cook Nuclear Plant, August 1973

1. 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 Limits

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

Reference PMP-6010-OSD-001 I Rev. 22 Page 46 of 91 OFF-SITE DOSE CALCULATION MANUAL

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

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 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, Dose Factors for Radioactive Particulate Gaseous Effluents Associated with the Child by the Inhalation Pathway ff. DIT-B-01987-00, Ground Plane & Food Dose Factors Pi for Radioiodines and Radioactive Particulate Gaseous Effluents gg. NRC Commitment 1010 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

c. AEP Design Information Transmittal on Aux Building Ventilation Systems

Reference PMP-6010-OSD-001 I Rev. 22 Page 47 of 91 OFF-SITE DOSE CALCULATION MANUAL

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

Reference PMP-6010-OSD-001 Rev. 22 Page 48 of 91 OFF-SITE DOSE CALCULATION MANUAL Pages:

Dose Factors for Various Pathways 48 - 51 Attachment 3.1 Ri Dose Factors PATHWAY Nuclide Ground Vegetable Meat Cow Milk Goat Milk Inhalation H-3 0.OE+00 4.OE+03 3.3E+02 2.4E+03 4.9E+03 1.3E+03 C-14 0.OE+00 3.5E+06 5.3E+05 3.2E+06 3.2E+06 3.6E+04 Cr-51 5.4E+06 1.1E+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 3.1E+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+06 Co-60 2.5E+10 3.2E+09 1.OE+09 2.7E+08 3.2E+07 8.6E+06 Zn-65 8.5E+08 2.7E+09 9.5E+08 1.6E+10 1.9E+09 1.2E+06 Sr-89 2.5E+04 3.5E+10 3.8E+08 9.9E+09 2.1E+10 2.4E+06 Sr-90 0.OE+00 1.4E+12 9.6E+09 9.4E+10 2.0E+1 1 1.1E+08 Zr-95 2.9E+08 1.2E+09 1.5E+09 9.3E+05 1.1E+05 2.7E+06 Sb-124 6.9E+08 3.OE+09 4.4E+08 7.2E+08 8.6E+07 3.8E+06 1-131 1.OE+07 2.4E+10 2.5E+09 4.8E+11 5.8E+11 1.6E+07 1-133 1.5E+06 4.OE+08 6.OE+01 4.4E+09 5.3E+09 3.8E+06 Cs-134 7.9E+09 2.5E+10 1.1E+09 5.OE+10 1.5E+l1 1.1E+06 Cs-136 1.7E+08 2.2E+08 4.2E+07 5.1E+09 1.5E+10 1.9E+05 Cs-137 1.2E+10 2.5E+10 1.OE+09 4.5E+10 1.4E+11 9.OE+05 Ba-140 2.3E+07 2.7E+08 5.2E+07 2.1E+08 2.6E+07 2.OE+06 Ce-141 1.5E+07 5.3E+08 3.OE+07 8.3E+07 1.0E+07 6.1E+05' Ce-144 7.9E+07 1.3E+10 3.6E+08 7.3E+08 8.7E+07 1.3E+07 3

Mr sec / yr pCi, inhalation pathway units are mr m yr PCi.

) except inhalation pathway are m2 Units for all Uap Values to be Used For the Maximum Exposed Individual Pathway Infant Child Teen Adult Fruits, vegetables and grain (kg/yr) -- 520 630 520 Leafy vegetables (kg/yr) -- 26 42 64 Milk (L/yr) 330 330 400 310 Meat and poultry (kg/yr) -- 41 65 110 Fish (kg/yr) -- 6.9 16 21 Drinking water (L/yr) 330 510 510 730 Shoreline recreation (hr/yr) -- 14 67 12 Inhalation (m3/yr) 1400 3700 8000 8000 Table E-5 of Reg. Guide 1.109.

Reference PMP-6010-OSD-001 Rev. 22 Page 49 of 91 OFF-SITE DOSE CALCULATION MANUAL Pages:

48 - 51 .1 Dose Factors for Various Pathways Bip Factors for Aquatic Foods pCi I / kg pCi Element Fish Invertebrate H 9.OE-1 9.OE-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.OE1 2.0E2 Ni 1.0E2 1.0E2 Cu 5.0E1 4.0E2 Zn 2.0E3 1.0E4 Br 4.2E2 3.3E2 Rb 2.0E3 1.0E3 Sr 3;.OE1 1.0E2 Y 2:5E1 1.0E3 Zr 3.3E0 6.7E0 Nb 3.0E4 1.0E2 Mo 1.OE1 1.OEl Tc 1.5El 5.0EO Ru 1.OEl 3.0E2 Rh 1.OE1 3.0E2 Te 4.0E2 6.1E3 I 1.5E1 5.OEO Cs 2.0E3 1.0E3 Ba 4.OEO 2.0E2 La 2.5E1, 1.0E3 Ce 1.OEO 1.0E3 Pr 2.5E1 1.0E3 Nd 2.5E1 L.0E3 W 1.2E3 1.OEl Np 1.OEl 4.0E2 Table A-I of Reg. Guide 1.109.

Reference PMP-6010-OSD-001 Rev. 22 Page 50 of 91 OFF-SITE DOSE CALCULATION MANUAL Dose Factors for Various Pathways Pages: .1 48 -51 Daipj External Dose Factors for Standing on Contaminated Ground mrem m 2 / 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 1.1E-8 1.3E-8 Fe-55 0 0 Fe-59 8.OE-9 9.4E-9 Co-58 7.0E-9 8.2E-9 Co-60 1.7E-8 2.OE-8 Ni-63 0 0 Ni-65 3.7E-9 4.3E-9 Cu-64 1.5E-9 1.7E-9 Zn-65 4.0E-9 4.6E-9 Zn-69 0 0 Br-83 6.4E-11 9.3E-11 Br-84 1.2E-8 1.4E-8 Br-85 0 0 Rb-86 6.3E-10 7.2E-10 Rb-88 3.5E-9 4.OE-9 Rb-89 1.5E-8 1.8E-8 Sr-89 5.6E-13 6.5E-13 Sr-91 7.1E-9 8.3E-9 Sr-92 9.0E-9 1.0E-8 Y-90 2.2E-12 2.6E-12 Y-91m 3.8E-9 4.4E-9 Y-91 2.4E-11 2.7E-11 Y-92 1.6E-9 1.9E-9 Y-93 5.7E-10 7.8E-10 Zr-95 5.OE-9 5.8E-9 Zr-97 5.5E-9 6.4E-9 Nb-95 5.1E-9 6.OE-9 Mo-99 1.9E-9 2.2E-9 Tc-99m 9.6E-10 1.1E-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-110m 1.8E-8 2.1E-8 Te-125m 3.5E-11 4.8E-1I

Reference PMP-6010-OSD-001 Rev. 22 Page 51 of 91 OFF-SITE DOSE CALCULATION MANUAL Pages: .1 Dose Factors for Various Pathways 48 - 51 Radionuclide Total Body Skin Te-127m 1.1E-12 1.3E-12 Te-127 1.OE-11 1.1E-11 Te-129m 7.7E-10 9.OE-10 Te-129 7.1E-10 8.4E-10 Te-131m 8.4E-9 9.9E-9 Te-131 2.2E-9 2.6E-6 Te-132 1.7E-9 2.OE-9 1-130 1.4E-8 1.7E-8 1-131 2.8E-9 3.4E-9 1-132 1.7E-8 2.OE-8 1-133 3.7E-9 4.5E-9 1-134 1.6E-8 1.9E-8 1-135 1.2E-8 1.4E-8 Cs-134 1.2E-8 1.4E-8 Cs-136 1.5E-8 1.7E-8 Cs-137 4.2E-9 4.9E-9 Cs-138 2.1E-8 2.4E-8 Ba-139 2.4E-9. 2.7E-9 Ba-140 2.1E-9 2.4E-9 Ba-141 4.3E-9 4.9E-9 Ba-142 7.9E-9 9.OE-9 La- 140 1.5E-8 1.7E-8 La- 142 1.5E-8 1.8E-8 Ce-141 5.5E-10 6.2E-10 Ce-143 2.2E-9 2.5E-9 Ce-144 3.2E-10 3.7E-10 Pr-143 0 0 Pr-144 2.OE-10 2.3E-10 Nd-147 1.OE-9 1.2E-9 W-187 3.1E-9 3.6E-9 Np-239 9.5E-10 1.1E-9 Table E-6 of Reg. Guide 1.109.

Reference PMP-6010-OSD-001 Rev. 22 Page 52 of 91 OFF-SITE DOSE CALCULATION MANUAL Radioactive Liquid Effluent Monitoring Instruments Page25 Attachment 3.2 52 - 54 ý INSTRUMENT Minimum Applicability Action Channels Operablea

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 3/4100 +)

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

2. Gross Radioactivity Monitors Not Providing Automatic Release Termination

a. Service Water (1) per At all times System Effluent Line (R-20, R-28) train

3. Continuous Composite Sampler Flow Monitor

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

4. Flow Rate Measurement Devices I

a. Liquid Radwaste Line (1) At times. of release (RFI-285)

b. Discharge Pipes* (1) At all times

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

d. Individual Steam Generator sample flow (1) per At times of release to Blowdown radiation monitors alarm generator (DFA-310, 320, 330 and 340)

• Pump curves and valve settings 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.

1. OPERABILITY of RRS-1001 includes OPERABILITY of sample flow switch RFS-1010, which is an attendant instrument as defined in Technical Specification section 1.1, under the term Operable - Operability. This item is also applicable for all Eberline liquid monitors (and their respective flow switches) listed here.

• • Since these monitors can be used for either batch or continuous release the appropriate action statement of I or 2 should apply (that is, Action 1 if a steam generator drain is being performed 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 samples while the sample system is lined up to discharge to the operating units blowdown system. In this case the steam generator drain samples can fulfill the sample requirement for Action 2 also. Action 2 would be exited when sampling was terminated.

+ Some Westinghouse (R) radiation monitors are being replaced by Eberline (DRS) monitors. Either monitor can fulfill the operability requirement. Ensure surveillances are current for operability of the instrumentation prior to using it to satisfy applicability requirement.

Reference PMP-6010-OSD-001 Rev. 22 Page 53 of 91 OFF-SITE DOSE CALCULATION MANUAL Radioactive Liquid Effluent Monitoring Instruments Pages:

Attachment 3.2 52 -54 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,

-OR-

2. Collect 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 control room annunciation, THEN 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.

TABLE NOTATION Action 1 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases may continue, provided that prior to initiating a release:

1. At least two independent samples are analyzed in accordance with Step 3.2.3a and;

2. At least two technically qualified members of the Facility Staff independently verify 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 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 jtCi/gram:

1. At least once per shift when the specific activity of the secondarycoolant is > 0.01 tCi/gram DOSE EQUIVALENT 1-131.

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

After 30 days, IF the channels are not OPERABLE, THEN continue releases with required grab samples provide a description of why the inoperability was not corrected in the next Annual Radiological Effluent release Report.

Action 3 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 that at least once 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 pCi/ml. Since the Westinghouse ESW 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 are only needed if the Containment Spray Heat Exchanger is in service. After 30 days, IF the channels are not OPERABLE, THEN continue releases with grab samples once per shift and provide a description of why the inoperability was not corrected in the next Annual Radiological Effluent release Report.

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 <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> during actual releases. After 30 days, IF the channels are not OPERABLE, THEN continue releases with grab samples once per shift and provide a description of why the inoperability was not corrected in the next Annual Radiological Effluent release Report.

Reference PMP-6010-OSD-001 Rev. 22 Page 54 of 91 OFF-SITE DOSE CALCULATION MANUAL Radioactive Liquid Effluent Monitoring Instruments Page25 Attachment 3.2 52 -54 Action 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 verified to.be within the required band at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> during actual releases. After 30 days, IF the channels are not OPERABLE, THEN continue releases with grab samples once per shift and provide a description of why the inoperability was not corrected in the next Annual Radiological Effluent release Report. IF the flow cannot be obtained within the desired band, THEN'declare the radiation monitor inoperable and enter the appropriate actions statement, Action 2.

Compensatory actions are governed by PMP-4030-EIS-001, Event-Initiated Surveillance Testing 2

Reference PMP-6010-OSD-001 Rev. 22 Page 55 of 91 OFF-SITE DOSE CALCULATION MANUAL Radioactive Liquid Effluent Monitoring Pages:

Instrumentation Surveillance Requirements 55 - 56 Instrument CHANNEL SOURCE CHANNEL CHANNEL CHECK CHECK CALIBRATION OPERATIONAL TEST

1. Gross Radioactivity Monitors Providing Automatic Release Termination

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

Effluent Line (RRS-1001)

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

Blowdown Effluent Line

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

Blowdown Treatment Effluent Line

2. Gross Radioactivity Monitors Not Providing Automatic Release Termination

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

System Effluent Line

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 B Q Effluent

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

• During releases via this pathway. This is applicable to all surveillances for the appropriate monitor.

2 Reference PMP-6010-OSD-001 Rev. 22 Page 56 of 91 OFF-SITE DOSE CALCULATION MANUAL Radioactive Liquid Effluent Monitoring Pages:

Instrumentation Surveillance Requirements 55 - 56 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 mode.*

5. Loss of sample flow. *

2. 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 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. For subsequent CHANNEL CALIBRATION, sources that have been related to the initial calibration may be used.

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 <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> on days on which continuous, periodic or batch releases are made.

5. 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 mode.*

5. Loss of sample flow.

• Instrument indicates, but does not provide for automatic isolation

• • Instrument indicates, but does not necessarily cause automatic isolation. No credit is taken for the automatic isolation on such occurrences.

Operations currently performs the routine channel checks and source checks. Maintenance and Radiation Protection perform channel calibrations and channel operational tests. Chemistry performs the channel check on the continuous composite sampler.

These responsibilities are subject to change without revision to this document.

Reference PMP-6010-OSD-001 Rev. 22 Page 57 of 91 OFF-SITE DOSE CALCULATION MANUAL Radioactive Gaseous Effluent Monitoring Instrumentation Pages:

Attachment 3.4 57-59 Instrument (Instrument #) Operable' Minimum Action Channels Action

1. Condenser Evacuation System

a. Noble Gas Activity (1) 1 T*6 Monitor (SRA-1905/2905)

b. Flow Rate Monitor (SFR-401, (1) **5 1/2-MR-054 and/or SRA- 1910/2910) OR (SFR-402 and 1/2-MR-054)

2. Unit Vent. Auxiliary Building Ventilation System

a. Noble Gas Activity (1) 6 Monitor (VRS-1505/2505)

b. Iodine Sampler (1) 8 Cartridge for VRA-1503/2503

c. Particulate Sampler Filter (1) 8 for VRA-1501/2501

d. Effluent System Flow Rate (1) *5 Measuring Device (VFR-315, MR-054 and/or VFR- 1510/25 10)

e. Sampler Flow Rate (1) 5 Measuring Device (VFS-1521/2521)

3. Containment Purge and Containment Pressure Relief (Vent) **

a. Containment Noble Gas Activity Monitor (1) ****2,3 7 ERS-13/1405 (ERS-23/2405)

b. Containment Particulate Sampler Filter (1) **** 10 ERS-13/1401 (ERS-23/2401)

4. Waste Gas Holdup System and CVCS HUT (Batch releases)**

a. Noble Gas Activity 9 Alarm and Termination (1) of Waste Gas Releases (VRS-1505/2505)

5. Gland Seal Exhaust

a. Noble Gas Activity Monitor (SRA-1805/2805)

b. Flow Rate Monitor (SFR-201, MR-054 or SFR-1810/2810)

• At all times

• • Containment Purge and other identified gaseous batch releases can be released utilizing the same double sampling compensatory action requirements of action 9 identified here even if there is no termination function associated with it like that associated with the two specific tank types listed here.

During releases via this pathway

Reference PMP-6010-OSD-001 Rev. 22 Page 58 of 91 OFF-SITE DOSE CALCULATION MANUAL Radioactive Gaseous Effluent Monitoring Instrumentation Pages:

Attachment 3.4 57 -59 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 statement,

-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 control room annunciation, THEN the only acceptable action is taking grab samples and conducting laboratory analyses as the reading is equivalent to a grab sample when the monitoris 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-

• 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 to completing both degas and depressurization of the RCS.

IF neither of the above are applicable, THEN the containment purge system is acting as a ventilation system (an extension of the Auxiliary Building) and is covered by Item 2 of this Attachment.

-OR-A Containment Pressure Relief (CPR) is being performed.

Once Purge (clean-up) has been completed and 'Ventilation' mode of Purge has commenced - resultant return to 'Clean-up' mode can be made with no additional sampling requirements or paperwork - so long as either ERS-1305/2305 OR ERS-1405/2405 are operable. Containment particulate channels are not needed since the RCS has been degassed and depressurized so leak detection is not an issue.

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

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 <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. After 30 days, IF the channels are not OPERABLE, THEN continue releases with estimation of the flow rate once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> 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 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 <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. After 30 days, IF the channels are not OPERABLE, THEN continue releases with grab samples once per shift and provide a description of why the inoperability was not corrected in the next Annual Radiological Effluent release Report.

Reeence PMP-6010-OSD-001 I Rev. 22 Page 59 of 91 OFF-SITE DOSE CALCULATION MANUAL rAttachment 3.4 Radioactive Gaseous Effluent Monitoring Instrumentation Paes 57 -59

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.

8. With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via the affected pathway may continue for up to 30 days provided samples required for weekly analysis are continuously collected with auxiliary sampling equipment as required in Attachment 3.7, Radioactive Gaseous Waste Sampling and Analysis Program. After 30 days, IF the channels are not OPERABLE, THEN continue releases with sample collection by auxiliary sampling equipment 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 OPERABLE requirement, the contents of the tank(s) may be released to the environment for up to 14 days provided that prior to initiating the release:

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

b. At least two technically qualified members of the Facility Staff independently verify the release rate calculations and discharge valve lineups; otherwise, suspend release of radioactive effluents via this pathway.

'After 14 days, IF the channels are not OPERABLE, THEN continue releases with sample collection by auxiliary sampling equipment and provide a description of why the inoperability was not corrected in the next Annual Radiological Effluent Release Report

10. See Technical Specification 3.4.15, RCS Leakage Detection System Instrumentation.

Compensatory actions are governed by PMP-4030-EIS-001, Event-Initiated Surveillance Testing.

Reference PMP-6010-OSD-001 Rev. 22 Page 60 of 91 OFF-SITE DOSE CALCULATION MANUAL Radioactive Gaseous Effluent Monitoring Pages:

Instrumentation Surveillance Requirements 60 - 61 Instrument CHANNEL SOURCE CHANNEL CHANNEL OPERATIONAL CHECK CHECK CALIBRATION TEST

1. Condenser Evacuation Alarm Only System

a. Noble Gas Activity Monitor D** M B(2) Q(1)

(SRA-1905/2905)

b. System Effluent Flow Rate D** NA B Q (SFR-401, SFR-402, MR-054, SRA-1910/2910)

2. Auxiliary Building Unit Alarm Only Ventilation System

a. Noble Gas Activity Monitor D* M B(2) Q(1)

(VRS- 1505/2505)

b. Iodine Sampler W* NA NA' NA (For VRA-1503/2503)

c. Particulate Sampler W* NA NA NA (For VRA-1501/2501)

d. System' Effluent Flow Rate D* NA B Q Measurement Device (VFR-315, MR-054, VRS- 1510/2510)

e. Sampler Flow Rate D* N/A B Q Measuring Device (VFS-1521/2521)

3. Containment Purge System and Alarm and Trip Containment Pressure Relief

a. Containment Noble Gas S P B(2) Q Activity Monitor (ERS-13/1405 and ERS-23/2405)

b. Containment Particulate S NA B Q Sampler (ERS-13/1401 and ERS-23/2401)

4. Waste Gas Holdup System Alarm and Trip Including CVCS HUT

a. Noble Gas Activity Monitor P P B(2) Q(3)

Providing Alarm and Termination (VRS- 1505/2505)

5. Gland Seal Exhaust Alarm Only

a. Noble Gas Activity D** M B(2) Q(1)

(SRA- 1805/2805)

b. System Effluent Flow Rate D** NA B Q (SFR-201, MR-054, SRA-1810/2810)

• At all times

• • During releases via this pathway. This is applicable to all surveillances for the appropriate monitor.

Reference PMP-6010-OSD-001 Rev. 22 Page 61 of 91 OFF-SITE DOSE CALCULATION MANUAL Radioactive Gaseous Effluent Monitoring Pages:

Instrumentation Surveillance Requirements 60 - 61 TABLE NOTATIONS

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. 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 controls not set in operate mode.*

• Instrument indicates, but does not provide automatic isolation.

Operations currently performs the routine channel checks, and source checks. Maintenance and Radiation Protection perform channel calibrations and channel operational tests. These responsibilities are subject to change without revision to this document.

Reference PMP-6010-OSD-001 Rev. 22 Page 62 of 91 OFF-SITE DOSE CALCULATION MANUAL Radioactive Liquid Waste Sampling and Analysis Program P ages:

Attachment 3.6 Ln 62-63 J

[Ref. 5.2.1s]

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

(jICi/ml) a A. Batch Waste P P Principal 5x10-7 Release Tanks C Each Batch Each Batch Gamma Emitters 1-131 1x10-6 P P Dissolved and Entrained Gases Each Batch Each Batch (Gamma 1xl0 5 Emitters)

P M H-3 lx1O05 Each Batch Composite b Gross Alpha lxl0-7 P Q Sr-89, Sr-90 5x10-8 Each Batch Composite b Fe-55 1x10-6 B. Plant W Principal Continuous Daily Compositeb Gamma 5xlc1' Releases* d Emitters' 1-131 1x10-6 M M Dissolved and Grab Sample Entrained Gases Ix10 5 (Gamma Emitters)

M H-3 1xl0-5 Daily Composite b Gross Alpha 1X10-7 Q Sr-89, Sr-90 5x10 8-Daily Compositeb Fe-55 1x10 6

• During releases via this pathway This table provides the minimum requirements for the liquid sampling program. If additional sampling is performed then those sample results can be used to quantify releases in lieu of composite data for a more accurate quantification.

Examples of these samples are the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> secondary coolant activity and Monitor Tank tritium samples.

Reference PMP-6010-OSD-001 Rev. 22 Page 63 of 91 OFF-SITE DOSE CALCULATION MANUAL Radioactive Liquid Waste Sampling and Analysis Program Pages:

Attachment 3.6 62 -63 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 - 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 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 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.

Reference PMP-6010-OSD-001 Rev. 22 Page 64 of 91 OFF-SITE DOSE CALCULATION MANUAL Radioactive Gaseous Waste Sampling and Pages:

Attachment 37 Analysis Program 64 - 65 Gaseous Release Type Frequency Minimum Type of Lower Limit Analysis Activity of Detection Frequency Analysis (PCi/cc) a

a. Waste Gas Storage P P Principal Gamma Tanks and CVCS HUTs Each Tank Each Tank Emitters d 1X 10-4 Grab Sample H-3 1 X 10-6

b. Containment Purge P P Principal Gamma Each Purge Each Purge Emitters 1x 10-4 Grab Sample CPR (vent)** Twice per Twice per Month Month 6

H-3 1 X10-

c. Condenser Evacuation W or M M Principal Gamma System Grab Sample Particulate Sample Emitters' 1 x 10H 6

Gland Seal Exhaust* M H-3 1X10 Wg Principle Gamma 1x 10-4 Noble Gas Emitters d M 1-131 Iodine Adsorbing 1 x 10-12 Media Continuous Wg Noble Gases Noble Gas Monitor. 1 x 10-6

d. Auxiliary Building Unit Continuous C Wb 1-131 Vent* Iodine Adsorbing 1 x 10-2 Media Continuous C w b Principal Gamma Particulate Sample Emitters' 1 x 10`

Continuous C M Gross Alpha Composite Particulate 1 x 10H Sample W w h H-3 Grab Sample H-3 Sample 1 x, 10-6 W gi Principle Gamma 1 X1 0 -4 Noble Gas Emitters d Continuous C Q Sr-89, Sr-90 Composite Particulate 1 x 10.11 Sample Continuous C Noble Gas Monitor Noble Gases 1 x 10-6

e. Incinerated Oil e P P Principal Gamma Each Batch' Each Batch f Emitters d 5 x 10-'

• During releases via this pathway

• • Only a twice 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 performed then 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 or compensatory action sample results.

Reference PMP-6010-OSD-001 Rev. 22 Page 65 of 91 OFF-SITE DOSE CALCULATION MANUAL Attachment 37 Radioactive Gaseous Waste Sampling and Pages:

Analysis Program 64 - 65 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 DetectionsA'B - REMP

b. Change samples at least once per 7 days and complete analyses within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> after changing.

Perform analyses at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 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 <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 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 1131 concentration in the RCS has not increased more than a factor of 3; and (2) the noble gas monitor 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 'normal'. [Ref. 5.2.1 y]

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.

• d. The principal gamma emitters for which the LLD 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 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 samples are representative of container contents.

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

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

i. Grab sampling of the Gland Seal Exhaust pathway need not be performed if the RMS low range channel (SRA-1805/2805) readings are less than 1E-6 ftC/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 Put 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 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 PMP-6010-OSD-001 Rev. 22 Page 66 of 91 OFF-SITE DOSE CALCULATION MANUAL Multiple Release Point Factors for Release Points Page:

Attachment 3.8 66 Liquid Factors Monitor Description Monitor Number MRP #

U 1 SG Blowdown 1R19/24, DRS 3100/3200* 0.35 U 2 SG Blowdown 2R19/24, DRS 4100/4200* 0.35 U 1 & 2 Liquid Waste Discharge RRS-1000 0.30 Sources of radioactivity released from the Turbine Room Sump (TRS) typically originate from the secondary cycle which is already being monitored by instrumentation that utilizes multiple release point (MRP) factors. The MRP is an administrative value that is used to assist with maintaining releases ALARA. The TRS has no actual radiation monitor, but utilizes an automatic compositor for monitoring what has been released. The batch release path, through RRS-1000, is the predominant release path by several magnitudes. Tritium is the predominant radionuclide released from the site and the radiation monitors do notrespond to this low energy beta emitter. Based on this information and the large degree of conservatism built into the radiation monitor setpoint methodology it does not appear to warrant further reduction for the TRS release path since its source is predominantly the secondary cycle which is adequately covered by this factor.

Gaseous Factors Monitor Description Monitor Number Flow Rate (cfm) MRP #

Unit 1 Unit Vent VRS-1500 186,600 - 0.54 Gland Seal Vent SRA-1800 1,260 0.00363 Steam Jet Air Ejector SRA-1900 3,600 (b) 0.01 Start Up FT Vent 1,536 0.004 Total 192,996 Unit 2 Unit Vent VRS-2500 143,400 0.41 Gland Seal Vent SRA-2800 5,508 (a) 0.02 Steam Jet Air Ejector SRA-2900 3,600 (b) 0.01 Start Up FT Vent 1,536 0.004 Total 154,044

• Either R-19, 24, DRS 3/4100 or 3/4200 can be used for blowdown monitoring as the Eberline monitors (DRS) are replacing the Westinghouse (R) monitors.

1. Nominal Values a Two release points of 2,754 cfm each are totaled for this value.

b This is the total design maximum of the Start Up Air Ejectors. This is a conservative value for unit 1.

Reference PMP-6010-OSD-001 Rev. 22 Page 67 of 91 OFF-SITE DOSE CALCULATION MANUAL Liquid Effluent Release Systems Page: .9 67

(

Reference PMP-6010-OSD-001 Rev. 22 t Page 68 of 91 OFF-SITE DOSE CALCULATION MANUAL Plant Liquid Effluent Parameters Page:

Attachment 3.10 68 SYSTEM COMPONENTS CAPACITY FLOW RATE TANKS PUMPS (EACH) (EACH)*

I Waste Disposal System

+ Chemical Drain Tank 1 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 II Steam Generator Blowdown and Blowdown Treatment Systems

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

+ Normal Flash. Tank (Not Vented)

+ Blowdown Treatment System 1

1 525 GAL.

-1 1 100 GPM 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 Unit 1 3 230,000 GPM Unit 2 4 230,000 GPM

• Nominal Values The 580 gpm value is calculated from the Estimated Steam Generator Blowdown Flow vs. DRV Valve Position letter prepared by M. J. O'Keefe, dated 9/27/93. This is 830 gpm times the 70% that remains as liquid while the other 30% flashes to steam and exhausts out the flash tank vent.

Reference PMP-6010-OSD-001 Rev. 22 Page 69 of 91 OFF-SITE DOSE CALCULATION MANUAL Volumetric Detection Efficiencies for Principle Gamma Page: .11 Emitting Radionuclides for Eberline Liquid Monitors 69 This includes the following monitors: RRS-1000, DRS 3100, DRS 3200, DRS 4100, and DRS 4200.

[Ref. 5.2. lp]

NUCLIDE EFFICIENCY (cpm/1tCi/cc) 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 3.60 E6 Mn-54 3.30 E7 Zn-65 1.58 E7 Ag-1 10M 9.93 E7 Ba-133 4.85 E7 Ba-140 1.92 E7 Cd-109 9.58 E5 Ce-139 3.28 E7 Ce-141 1.92 E8 Ce-144 4.83 E6 Co-57 3.80 E7 Cs-136 1.07 E8 Fe-59 2.83 E7 Sb-124 5.93 E7 1-133 3.40 E7 1-134 *7.23 E7 1-135 3.95 E7 Mo-99 8.68 E6 Na-24 4.45 E7 Nb-95 3.28 E7 Nb-97 3.50 E7 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-113 7.33 E5 Sr-85 3.70 E7 Sr-89 2.88 E3 Sr-92 3.67 E7 Tc-99M 3.60 E7 Y-88 5.25 E7 Zr-95 3.38 E7 Zr-97 3.10 E7 Kr-85 1.56 E5 Kr-85M 3.53 E7 Kr-88 4.10 E7 Xe-131M 8.15 E5 Xe-133 7.78 E6 Xe-133M 5.75 E6 Xe-135 3.83 E7

Reference PMP-6010-OSD-001 Rev. 22 Page 70 of 91 OFF-SITE DOSE CALCULATION MANUAL .12 Counting Efficiency Curves for R-19, and R-24 Pages:

70-71 Counting Efficiency Curve for R-19 Efficiency Factor = 4.2 E6 cpm/uCi/ml (Based on empirical data taken during pre-operational testing with Cs-l 37) 1..E+07 1.00E+06 1.00E+05 2

x" 1.00E+04

.0 0

S1.OOE+03 C.)

1.OOE+02 1.00E+01 1.00E+O0 to U) -IT C4 0 Wi W iii Wj . Lii +

0D 0 CD D 0 0.

microcuries/mI

Reference PMP-6010-OSD-001 Rev. 22 Page 71 of 91 OFF-SITE DOSE CALCULATION MANUAL Counting Efficiency Curves for R-19, and R-24 Pages: .12 Counting Efficiency Curve for R-24 Efficiency Factor = 7.5E6 cpm/uCi/ml (Based on empirical data taken during pre-operational testingwth Mn-54) 1.00E+07 1.OOE+06 1.OOE+05 CM

.x 1.OOE+04 Cu 0 1.OOE+03 1.OOE+02 1.00E+01 1.00E+O00 (0 a)O(

9 9 C? 0C0 0 LU 0

CD LU 0 0UIJ L 0

0 0LUJ C

o.

C) 0 C+

C) mnicrocuries/mi

Reference PMP-6010-OSD-001 Rev. 22 Page 72 of 91 OFF-SITE DOSE CALCULATION MANUAL Counting Efficiency Curve for R-20, and R-28 Page: .13 Counting Efficiency Curve for R-20 and R-28 Efficiency Factor = 4.3 E6 cpm/uCi/mi (Based on empirical data taken during pre-operational testing with Co-58) 1.OOE+07 1.OOE+06 1.OOE+05 0

.. 1.00E+04

.0 1,0OE+03 1.00E+02 1.00E+01 1.00E+00. . . . * * ,, .

CD 0) 0 SN CD 9LLJ 9 LU. LU. 9 U!Jw +

microcuries/ml

Reference PMP-6010-OSD-001 Rev. 22 Page 73 of 91 OFF-SITE DOSE CALCULATION MANUAL Gaseous Effluent Release Systems Page:

Attachment 3.14 73 RELEASE SOURCES SYSTEMS POINTS WasteGas Decay Tanks and CVCS HUT MoistureSeparator s Decay Fswlation Tak Ha ater Valv Au. BuildingVent gPO From Areasand EngineeredSolely Damper FlowuresVent System adTa Syste seie FeaturesEldlnclosrs etanie cDFitr Fitr_______________________

Fuel Handling S ]

Ventilation A., 3. e, ning i Fitr nThso olte U Purge rsation Containment HEPA andRelietSystem Purg _ _ Vuloe FnIt RContinment F nm mdSt a etyf " aito oio >il]

Purge Syeltm Fiore 5iEhalar BIcteown rament Heaom ormal Fleeem Treamen Syste to Dan gaseousremosphee Ejecto Systemlt~

Samplin Pointme peif.aa Steam ~ Sea Packingat ý, st

Reference PMP-6010-OSD-001 Rev. 22 Page 74 of 91 OFF-SITE DOSE CALCULATION MANUAL Attachment 3.15 Plant Gaseous Effluent Parameters Page:

74 SYSTEM UNIT EXHAUST CAPACITY FLOW RATE (CFM)

I PLANT AUXILIARY BUILDING 1 186,600 max UNIT VENT 2 143,400 max WASTE GAS DECAY TANKS (8) 1 125 4082 FT 3 @100 psig AND CHEMICAL & VOLUME 28,741 ft3 max CONTROL SYSTEM HOLD UP @ 8#, 0 level TANKS (3)

+ AUXILIARY BUILDING 1 72,660 EXHAUST 2 59,400

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

+ FUEL HANDLING AREA VENT 1 30,000 SYSTEM CONTAINMENT PURGE SYSTEM 1 &2 32,000 CONTAINMENT PRESSURE 1& 2 1,000 RELIEF SYSTEM INSTRUMENT ROOM PURGE 1& 2 1,000 SYSTEM II CONDENSER AIR EJECTOR 2 Release Points SYSTEM One for Each Unit NORMAL STEAM JET AIR 1& 2 230 EJECTORS START UP STEAM JET AIR 1& 2 3,600 EJECTORS III TURBINE SEALS SYSTEM 1 1,260 2 5,508 2 Release Points for Unit 2 IV START UP FLASH TANK VENT 1 1,536

_ __2 1,536

+ Designates total flow for all fans.

Reference PMP-6010-OSD-001 Rev. 22 Page 75 of 91 OFF-SITE DOSE CALCULATION MANUAL Attachment 3.16 10 Year Average of 1995-2004 Data Pages: 75 - 76 X/IQ GROUND AVERAGE (sec/m 3 )

DIRECTION DISTANCE (METERS)

(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 1.59E-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 2.72E-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 1.41E-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 W 3.63E-06 4.11E-07 1.96E-07 1.18E-07 8.31E-08 WNW 3.02E-06 3.43E-07 1.61E-07 9.59E-08 6.71 E-08 NW 3.22E-06 3.61E-07 1.71E-07 1.02E-07 7.16E-08 NNW 3.84E-06 4.29E-07 2.02E-07 1.20E-07 8.40E-08 DIRECTION DISTANCE (METERS)

(WIND FROM) 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 4.56E-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 1.30E-08 SSE 1.71E-07 7.06E-08 3.56E-08 2.29E-08 1.48E-08 S 1.84E-07 7.49E-08 3.77E-08 2.43E-08 1.56E-08 SSW 8.86E-08 3.59E-08 1.80E-08 1.15E-08 7.39E-09 SW 4.93E-08 1.96E-08 9.77E-09 6.27E-09 :3.98E-09 WSW 3.80E-08 1.51E-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 WNW 3.34E-08 1.29E-08 6.41E-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 8.1OE-09 5.19E-09 3.27E-09 DIRECTION TO - SECTOR N A E E S =J W N NNE =B ESE = F SSW -K WNW = P NE =C SE = G SW =L NW =Q ENE = D SSE = H WSW = M NNW = R Worst Case I/Q = 2.04E-05 sec/m 3 in Sector H 2004

Reference PMP-6010-OSD-001 Rev. 22 Page 76 of 91 OFF-SITE DOSE CALCULATION MANUAL Attachment 3.16 10 Year Average of 1995-2004 Data Pages: 75 - 76 D/Q DEPOSITION (1/m2)

DIRECTION 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 1.29E-08 1.25E-09 5.67E-10 2.97E-10 1.90E-10 ENE 1.59E-08 1.54E-09 6.97E-10 3.66E-10 2.33E-10 E 1.87E-08 1.81E-09 8.20E-10 4.30E-10 2.75E-10 ESE 1.85E-08 1.79E-09 8.12E-10 4.26E-10 2.72E-10 SE 1.90E-08 1.83E-09 8.30E-10 4.36E-10 2.78E-10 SSE 2.40E-08 2.32E-09 1.05E-09 5.52E-10 3.52E-10 S 3.68E-08 3.56E-09 1.61E-09 8.46E-10 5.40E-10 SSW 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 2.11E-08 2.04E-09 9.23E-10 4.84E-10 3.09E-10 W 2.OOE-08 1.93E-09 8.74E-10 4.59E-10 2.93E-10 WNW 1.75E-08 1.69E-09 7.64E-10 4.01E-10 2.56E-10 NW 1.58E-08 1.53E-09 6.94E-10 3.64E-10 2.32E-10 NNW 2.30E-08 2.22E-09 1.01E-09 5.28E-10 3.37E-10 DIRECTION DISTANCE (METERS)

(WIND FROM) 12067 124135 1'40225 56315 ]80500 N 1.45E-10 4.72E-11 1.74E-11 9.27E-12 4.65E-12 NNE 6.36E-11 1.97E-11 7.24E-12 3.86E-12 1.94E-12 NE 8.07E-11 2.58E-11 9.51E-12 5.07E-12 2.54E-12 ENE 9.77E-11 3.17E-11 1.17E-11 6.23E-12 3.13E-12 E 1.14E-10 3.73E-11 1.37E-11 7.34E-12 3.68E-12 ESE 1.13E-10 3.70E-11 1.36E-11 7.26E-12 3.64E-12 SE 1.16E-10 3.78E-11 1.39E-11 7.42E-12 3.72E-12 SSE 1.47E-10 4.79E-11 1.76E-11 9.41E-12 4.72E-12 S 2.25E-10 7.34E-11 2.70E- 11 1.44E-11 7.23E-12 SSW 1.41E-10 4.59E-11 1.69E-11 ,9.01E-12 4.52E-12 SW 1.36E-10 4.43E-11 1.63E-11 8.71E-12 4.37E-12 WSW 1.29E-10 4.20E-11 1.55E-11 8.26E-12 4.14E-12 W 1.22E-10 3.98E-11 1.47E-11 7.82E-12 3.92E-12 WNW 1.07E-10 3.48E-11 1.28E-11 6.84E- 12 3.43E-12 NW 9.70E-11 3.16E-11 1.16E-11 6.20E-12 3.11E-12 NNW 1.41E-10 4.58E-11 1.69E-11 9.00E- 12 4.52E-12 DIRECTION TO - SECTOR N =A E =E S J W N NNE B ESE = F SSW = K WNW = P NE =C SE =G SW =L NW =Q ENE = D SSE = H WSW = M NNW = R Worst Case D/Q = 4.46E-08 1/M2 in Sector A 2001

Reference PMP-6010-OSD-001 - Rev. 22 Page 77 of 91 OFF-SITE DOSE CALCULATION MANUAL Annual Evaluation of x/Q and D/Q Values For Page:

Attachment 3.17 All Sectors 77

1. Performed 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)

2. Worst X/Q and D/Q value and sector determined. PMP-6010-OSD-001 has been updated, if necessary. Provide an evaluation.

/

Signature Date Environmental Department (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:

/

Signature Date Environmental Department (print name, title)

5. Copy to NS&A for information.

/

Signature Date Environmental Department (print name, title)

Reference PMP-6010-OSD-001 Rev. 22 Page 78 of 91 OFF-SITE DOSE CALCULATION MANUAL FPages: Dose Factors 78 - 79 Attachment 3.18 DOSE FACTORS FOR NOBLE GASES AND DAUGHTERS*

TOTAL BODY SKIN DOSE GAMMA AIR BETA AIR DOSE FACTOR FACTOR DOSE FACTOR DOSE FACTOR K (DFB) L (DFSi) Mi (DFi) Ni (DFi mrem m 3 (mrem m 3 (mrad m 3 (mrad m 3 RADIONUCLIDE per ýtCi yr) per ýtCi yr) per ýCi yr) per ýtCi 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 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 1.47E+04 2.37E+03 1.52E+04 2.93E+03 Kr-89 1.66E +04 1.01E+04 1.73E+04 1.06E+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 7.11E+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

• The listed dose factors are for radionuclides that may be detected in gaseous effluents, from keg. Guide 1.109, Table B-I.

Reference PMP-6010-OSD-001 Rev. 22 Page 79 of 91 OFF-SITE DOSE CALCULATION MANUAL Dose Factors Pages: .18 78-79 DOSE FACTORS FOR RADIOIODINES AND RADIOACTIVE PARTICULATE, IN GASEOUS EFFLUENTS FOR CHILD* Ref. 5.2.leeandff Pi Pi INHALATION FOOD & GROUND PATHWAY PATHWAY (mrem m 3 (mrem m2 sec per jtCi yr) per jCi yr)

H-3 1.12E+03 1.57E+03 H P-32 2.60E+06 7.76E+ 10 Cr-51 1.70E+04 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+1i0 Rb-86 1.98E+05 8.78E+09 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-110m 5.48E+06 1.99E+10 1-131 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.OOE+ 10 Cs-136 1.71E+05 3.OOE+09 Cs-137 9.07E+05 3.34E+ 10 Ba-140 1.74E+06 1.46E+08 Ce-141 5.44E+05 3.31E+07 Ce-144 1.20E+07 1.91E+08

• As Sr-90, Ru-106 and 1-131 analyses are performed, THEN use Pi given in P-32 for nonlisted radionuclides.

3 The units for both H3 factors are the same, mrem m per pCi yr

Reference PMP-6010-OSD-001 Rev. 22 Page 80 of 91 OFF-SITE DOSE CALCULATION MANUAL Attachment 3.19 Radiological Environmental Monitoring Program Pages:

Sample Stations, Sample Types, Sample Frequencies 80 - 83

[Ref. 5.2.1v, 5.2.1x, 5.2.1t]

SAMPLE DESCRIPTION/ SAMPLE SAMPLE ANALYSIS ANALYSIS STATION LOCATION TYPE FREQUENCY TYPE FREQUENCY ON-SITE AIRBORNE AND DIRECT RADIATION (TLD) STATIONS ONS-1 (T-1) 1945 ft @ 180 from Plant Axis AirborneParticulate Weekly Gross Beta Weekly Gamma Isotopic Quart. Comp.

Airborne 1-131 Weekly Radioiodine TLD Quarterly Direct Radiation Quarterly ONS-2 (T-2) 2338 ft @ 480 from Plant Axis Airborne Particulate Weekly Gross Beta Weekly Gamma Isotopic Quart. Comp.

Airborne 1-131 Weekly Radioiodine TLD Quarterly Direct Radiation Quarterly ONS-3 (T-3) 2407 ft @ 900 from Plant Axis Airborne Particulate Weekly Gross Beta Weekly Gamma Isotopic Quart. Comp.

Airborne 1-131 Weekly Radioiodine TLD Quarterly Direct Radiation Quarterly ONS-4 (T-4) 1852 ft. @ 1180 from Plant Axis Airborne Particulate Weekly Gross Beta Weekly

-.Gamma Isotopic Quart. Comp.

Airborne 1-131 Weekly Radioiodine TLD Quarterly Direct Radiation Quarterly ONS-5 (T-5) 1895 ft @ 189' from Plant Axis Airborne Particulate Weekly Gross Beta Weekly Gamma Isotopic Quart. Comp.

Airborne 1-131 Weekly Radioiodine TLD Quarterly Direct Radiation Quarterly ONS-6 (j-6) 1917 ft @ 210' from Plant Axis Airborne Particulate Weekly Gross Beta Weekly Gamma Isotopic Quart. Comp.

Airborne 1-131 Weekly Radioiodine TLD Quarterly Direct Radiation Quarterly T-7 2103 ft @ 360 from Plant Axis TLD Quarterly Direct Radiation Quarterly T-8 2208 ft @ 820 from Plant Axis TLD Quarterly Direct Radiation Quarterly T-9 1368 ft @ 149' from Plant Axis TLD Quarterly Direct Radiation Quarterly T-10 1390 ft @ 127' from Plant Axis TLD Quarterly Direct Radiation Quarterly T-11 1969 ft @ 110 from PlantAxis TLD Quarterly Direct Radiation Quarterly T-12 2292 ft @ 630 from Plant Axis TLD Quarterly Direct Radiation Quarterly CONTROL AIRBORNE AND DIRECT RADIATION (TLD) STATIONS NBF 15.6 miles SSW Airborne Particulate Weekly Gross Beta Weekly New Buffalo, MI Gamma Isotopic Quart. Comp.

Airborne Radioiodine 1-131 Weekly TLD Quarterly Direct Radiation Quarterly SBN 26.2 miles SE Airborne Particulate Weekly Gross Beta Weekly South Bend, IN Gamma Isotopic Quart. Comp.

Airborne Radioiodine 1-131 Weekly TLD Quarterly Direct Radiation Quarterly DOW 24.3 miles ENE Airborne Particulate Weekly Gross Beta Weekly Dowagiac, MI Gamma Isotopic Quart. Comp.

Airborne Radioiodine 1-131 Weekly TLD Quarterly Direct Radiation Quarterly COL 18.9 miles NNE Airborne Particulate Weekly Gross Beta Weekly Coloma, MI Gamma Isotopic Quart. Comp.

Airborne Radioiodine 1-131 Weekly TLD Quarterly Direct Radiation Quarterly

Reference PMP-6010-OSD-001 Rev. 22 Page 81 of 91 OFF-SITE DOSE CALCULATION MANUAL Attachment 3.19 Radiological Environmental Monitoring Program Pages:

Sample Stations, Sample Types, Sample Frequencies 80 - 83 SAMPLE DESCRIPTION/ SAMPLE SAMPLE ANALYSIS ANALYSIS STATION LOCATION TYPE FREQUENCY TYPE FREQUENCY OFF-SITE DIRECT RADIATION (TLD) STATIONS OFT-1 4.5 miles NE, Pole #B294-44 TLD Quarterly Direct Radiation Quarterly OFT-2 3.6 miles, NE, Stevensville TLD Quarterly Direct Radiation Quarterly Substation OFT-3 5.1 miles NE, Pole #B296-13 TLD Quarterly Direct Radiation Quarterly OFT-4 4.1 miles, E, Pole #B350-72 TLD Quarterly Direct Radiation Quarterly OFT-5 4.2 miles ESE, Pole #B387-32 TLD Quarterly Direct Radiation Quarterly OFT-6 4.9 miles SE, Pole #B426-1 TLD Quarterly Direct Radiation Quarterly OFT-7 2.5 miles S, Bridgman Substation TLD Quarterly Direct Radiation Quarterly OFT-8 4.0 miles S, Pole #B424-20 TLD Quarterly Direct Radiation Quarterly OFT-9 4.4 miles ESE, Pole #B369-214 TLD Quarterly Direct Radiation Quarterly OFT-10 3.8 miles S, Pole #B422-99 TLD Quarterly Direct Radiation Quarterly OFT-11 3.8 miles S, Pole #1B423-12 TLD Quarterly Direct Radiation Quarterly GROUNDWATER (WELL WATER) SAMPLE STATIONS W:l 1969 ft @ 110 from Plant Axis Groundwater Quarterly Gamma Isotopic Quarterly Tritium Quarterly W-2 2302 ft @ 630 from Plant Axis Groundwater Quarterly Gamma Isotopic Quarterly Tritium Quarterly W-3 3279 ft @ 1070 from Plant Axis Groundwater Quarterly. Gamma Isotopic Quarterly Tritium Quarterly W-4 . 418 ft @ 3010 from Plant Axis Groundwater Quarterly Gamma Isotopic Quarterly Tritium Quarterly W-5. 404 ft @ 290' from Plant Axis Groundwater Quarterly Gamma Isotopic Quarterly Tritium Quarterly W-6 424 ft @ 2730 from Plant Axis Groundwater Quarterly Gamma Isotopic Quarterly Tritium Quarterly W-7 1895 ft @ 1890 from Plant Axis Groundwater Quarterly Gamma Isotopic Quarterly Tritium Quarterly W-8 1274 ft @ 540 from Plant Axis Groundwater Quarterly Gamma Isotopic Quarterly Tritium Quarterly W-9 1447 ft @ 220 from Plant Axis Groundwater Quarterly Gamma Isotopic Quarterly Tritium Quarterly W-10 4216 ft @ 1290 from Plant Axis Groundwater Quarterly Gamma Isotopic Quarterly Tritium Quarterly W-11 3206 ft @ 1530 from Plant Axis Groundwater Quarterly Gamma. Isotopic Quarterly Tritium Quarterly W-12 2631 ft @ 1620 from Plant Axis Groundwater Quarterly Gamma Isotopic Quarterly Tritium Quarterly W-13 2152 ft @ 1820 from Plant Axis Groundwater Quarterly Gamma Isotopic Quarterly Tritium Quarterly W-14 1780 ft @ 1640 from Plant Axis Groundwater Quarterly Gamma Isotopic Quarterly Tritium Quarterly W-15 725 ft @ 2020 from Plant Axis Groundwater Quarterly Gamma Isotopic Quarterly NPDES well MW-12C Tritium Quarterly W-16 2200 ft @ 2080 from Plant Axis Groundwater Quarterly Gamma Isotopic Quarterly NPDES well MW-20 Tritium Quarterly W-17 2200 ft @ 1800 from Plant Axis Groundwater Quarterly Gamma Isotopic Quarterly NPDES well MW-21 Tritium Quarterly f

Reference PMP-6010-OSD-001 Rev. 22 Page 82 of 91 OFF-SITE DOSE CALCULATION MANUAL Attachment 3.19 Radiological Environmental Monitoring Program Pages:

Sample Stations, Sample Types, Sample Frequencies 80 - 83 DRINKING WATER STJ St. Joseph Public Intake Sta. Drinking water Once per calendar Gross Beta 14 day Comp.

9 mi. NE Day Gamma Isotopic 14 day Comp.

1-131 14 day Comp.

Tritium Quart. Comp.

LTW Lake Twp. Public Intake Sta. Drinking water Once per calendar Gross Beta 14 day Comp.

0.6 mi. S Day Gamma Isotopic 14 day Comp.

1-131 14 day Comp.

Tritium Quart. Comp SEDIMENT SL-2

• Plant Site Boundary - South

- 500 ft. south of Plant Centerline SL-3 I.Plant Site Boundary - North 7-. 500 ft. north of Plant Centerline INGESTION - MILK Background Farm* I Milk Once every 1-131 per sample 15 days Gamma Isotopic per sample SAMPLE DESCRIPTION/ SAMPLE SAMPLE ANALYSIS ANALYSIS STATION LOCATION TYPE FREQUENCY TYPE FREQUENCY INGESTION - FISH ONS-N 0.3 mile N, Lake Michigan Fish - edible portion 2/year Gamma Isotopic per sample ONS-S 0.4 mile S, Lake Michigan Fish - edible portion " 2/year Gamma Isotopic per sample OFS-N 3.5 mile N, Lake Michigan Fish - edible portion 2/year Gamma Isotopic per sample OFS-S 5.0 mile S, Lake Michigan Fish - edible portion 2/year Gamma Isotopic per sample

Reference PMP-6010-OSD-001 Rev. 22 Page 83 of 91 OFF-SITE DOSE CALCULATION MANUAL Radiological Environmental Monitoring Program Pages:

Sample Stations, Sample Types, Sample Frequencies 80 - 83 INGESTION - BROADLEAF IN LIEU OF MILK 3 indicator samples of broad leaf vegetation Broadleaf Monthly Gamma Isotopic Monthly collected at different locations, within eight vegetation when available 1131- when available miles of the plant in the highest annual average D/Q land sector.

I background sample of similar vegetation Broadleaf Monthly Gamma Isotopic Monthly grown 15-25 miles distant in one of vegetation when available 1131 when available the less prevalent wind directions.

Collect composite samples of Drinking and Surface water at least daily. Analyze particulate sample filters for gross beta activity 24 or more hours following filter removal. This will allow for radon and thoron daughter decay. If gross beta activity in air or water is greater than 10 times the yearly mean of control samples for any medium, perform gamma isotopic analysis on the individual samples.

IF at least three indicator milk samples and one background milk sample cannot be obtained, THEN three indicator broad leaf samples will be collected at different locations, within eight miles of the plant, in the land sector with the highest D/Q (refers to the highest annual average D/Q). Also, one background broad leaf sample will be collected 15 to 25 miles from the plant in one of the less prevalent D/Q land sectors.

• The three milk indicator and one background farm will be determined by the Annual Land Use Census and those that are willing to participate. IF it is determined that the milk animals are fed stored feed, THEN monthly sampling is appropriate for that time period.

Reference PMP-6010-OSD-001 Rev. 22 Page 84 of 91 OFF-SITE DOSE CALCULATION MANUAL Maximum Values for Lower Limits of DetectionsA'B - REMP Pages:

Attachment 3.20 84-85

[Ref. 5.2. 1v]

Radionuclides Food Product Water Milk Air Filter Fish Sediment pCi/kg, wet pCi/! pCi/l pCi/m3 pCi/kg, 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 Zr-95 30 Nb-95 15 Mn-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 This Data is directly from our plant-specific Technical Specification.

• LLD for drinking water

Reference PMP-6010-OSD-001 Rev. 22 Page 85 of 91 OFF-SITE DOSE CALCULATION MANUAL Maximum Values for Lower Limits of DetectionsAB REMP Pages: .20 84 -8 NOTES A. The Lower Limit of Detection (LLD) is defined as the smallest concentration of radioactive material in a sample that will be detected with 95% probability and 5% probability of falsely concluding that a blank observation represents a "real" signal.

For a particular measurement system (which may include radiochemical separation), the LLD is given by the equation:

LLD- 4.66a *S E*V* 2.22*Y*e(-*A!)

Where LLD is the a priori lower limit of detection as defined above (as pCi per unit mass or volume). Perform analysis in such a manner that the stated LLDs will be achieved under routine conditions. Occasionally background fluctuations, unavoidably small sample sizes, the presence of interfering radionuclides, or other uncontrollable circumstances may render these LLDs unachievable. It should be further clarified that the LLD represents the capability of a measurement system and not as an after the fact limit for a particular measurement.

S is the standard deviation of the background counting rate or of the counting rate of a blank sample as appropriate (as counts per minute).

E is the counting efficiency of the detection equipment as counts per transformation (that is, disintegration)

V is the sample size in appropriate mass or volume units 2.22 is the conversion factor from picocuries (pCi) to transformations (disintegrations) per minute Y is the fractional radiochemical yield as appropriate k is the radioactive decay constant for the particular radionuclide At is the elapsed time between 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 radionuclides listed in Attachment 3.20, Maximum Values for Lower Limits of DetectionsA,B - REMP.

af A 2.71 value may be added to the equation to provide correction for deviations in the Poisson distribution at low count rates, that is, 2.71 + 4.66 x S.

Reference PMP-6010-OSD-001 Rev. 22 Page 86 of 91 OFF-SITE DOSE CALCULATION MANUAL Reporting Levels for Radioactivity Concentrations Page:

Attachment 3.21 in Environmental Samples 86 Radionuclides Food Product Water Milk Air Filter Fish pCi/kg, wet pCi/l pCi/l pCi/rn3 pCi/kg, wet H-3 20000 Ba-140 200 300 La-140 200 300 Cs-134 1000 30 60 10 y 1000 Cs-137 2000 50 70 20 2000 Zr-95 400 Nb-95 400 Mn-54 1000 30000 Fe-59 400 10000 Zn-65 300 20000 Co-58 1000 30000 Co-60 300 10000 i -i ________________ -- F- ------F- --

1-131 100 2 3 0.90 IF any of the above concentration levels are exceeded THEN see guidance contained in step 3,5.2a. for additional information.

Reference PMP-6010-OSD-001 Rev. 22 Page 87 of 91 OFF-SITE DOSE CALCULATION MANUAL On-Site Monitoring Location - REMP Page: .22 87 ONS-South ONS-North Surface Water TLD T-6 TLD T-5 SWL-3 Air ONS- Air ONS-5 Surface Water Well W-6 Well Se 6 SWL-2 r W_,eIi Sediment SL-3 Well L I Sediment SL-2 '. 2 W9 LEGEND ONS ONS-6: Air Sampling Station T-1 -T-12: TLD Sampling Station W-1 -W-17: REMP Groundwater Wells SWL- 2, 3: Surface Water Sampling Stations SL-2 SL-3: Sediment Sampling Stations ONS-N & S: Fish samolino locations

Reference PMP-6010-OSD-001 Rev. 22 Page 88 of 91 OFF-SITE DOSE CALCULATION MANUAL Off-Site Monitoring Locations - REMP Page: .23 88 Legend Offsite REMP Monitoring Locations OFT OFT-11: TLD Locations Background Air/TLD Stations Drinking Water Locations Indicator Milk Farm Locations Background Milk Farm Locations OFS Offsite Fish locations Kankakee Station

Background

Hwy 23 AirITLD (SBN)

Reference PMP-6010-OSD-001 Rev. 22 Page 89 of 91 OFF-SITE DOSE CALCULATION MANUAL Attachment 3.24 Safety Evaluation By The Office Of Nuclear Pages:

Reactor Regulation 89 - 91 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]

(This is a 10 CFR 50.75 (g) item)

1. INTRODUCTION By letters dated' October 9, 1991, October 23, 1991, September 3, 1993, and September 29, 1993, Indiana MichiganPower Company (I&M) requested approval pursuant to 10 CFR 20.2002 for the on-site disposal of licensed material not previously considered in the Donald C. Cook Nuclear Plant Final Environmental Statement dated August 1973. Specifically, this request addresses actions taken in 1982 in which approximately 942 cubic meters of slightly contaminated sludge were removed from the turbine room sump absorption pond and pumped to the upper parking lot located within the exclusion area of the Donald C. Cook Nuclear Plant. The contaminated sludge was spread over an area of approximately 4.7 acres. The sludge contained a total radionuclide inventory of 8.89 millicuries (mCi) of Cesium-137, Cesium-136, Cesium-134, Cobalt-60 and Iodine-131.

In its submittal, the licensee addressed specific information requested in accordance with 10 CFR 20.2002(a), provided a detailed description of the licensed material, thoroughly analyzed and evaluated information pertinent to the impacts on the environment of the proposed disposal of licensed material, and committed to follow specific procedures to minimize the risk of unexpected exposures.

2. DESCRIPTION OF WASTE The turbine room sump absorption pond is a collection place for water released from the plant's turbine room sump. The contamination was caused by a primary-t0-secondary steam generator leak that entered the pond from the turbine building sump, a recognized 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 exclusion area. The 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 upper parking lot area of approximately 4.7 acres.

The principal radionuclides identified in the dredged material are listed below.

TABLE 1 NUCLIDE ACTIVITY (mCi) ACTIVITY (mCi)

(half-life) 1982 1991 36 1 CS (13.2 d) 0.03 NA*

'34Cs (2.1 y) 2.34 0.18 137Cs (30.2 y) 5.59 4.57 60 Co (5.6 y) 0.90, 0.27 1311 (8.04 d) 0.03 NA*

TOTAL: 8.89 5.02

• NA: not applicable due to decay

Reference PMP-6010-OSD-001 Rev. 22 Page 90 of 91 OFF-SITE DOSE CALCULATION MANUAL Safety Evaluation By The Office Of Nuclear Pages:

Reactor Regulation 89-91

3. RADIOLOGICAL IMPACTS The licensee in 1982 evaluated the following potential exposure pathways to members of the general public from the radionuclides in the sludge:

(1) external exposure caused by groundshine from the disposal site; (2) internal exposure caused by inhalation of re suspended radionuclide;

-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-1101, "Onsite Disposal of Radioactive Waste," Volumes 1 and 2, November 1986 and February 1987, respectively. The staff finds the assessment methodology acceptable. Table 2 lists the doses calculated by the licensee for the maximally exposed member of the public based on a total activity of 8.89 mCi disposed in that year.

TABLE 2 Pathway Whole Body Dose Received by Maximally Exposed Individual (mrem/year)

Groundshine 0.94

.Inhalation 0.94 Groundwater Ingestion 0.73 Total 2.61 On July 5, 1991, the licensee re-sampled the onsite disposal area to assure that no significant impacts and adverse effects had occurred. A counting procedure based on the appropriate environmental low-level doses was used by the licensee; however, no activity was detected during the re-sampling'. This is consistent with the original activity of the material and the decay time. The 1991 re-sampling process used by the licensee confirms that the environmental impact of the 1982 disposal was very small. The staff finds the licensee's methodology acceptable.

4. ENVIRONMENTAL FINDING AND CONCLUSION The staff has evaluated the environmental impact of the proposal to leave in place approximately 942 cubic meters of slightly contaminated sludge underneath the upper parking lot on the Donald C. Cook Nuclear Plant site.

In 1982, the licensee evaluated the potential exposure to members of the general public from the radionuclides in the sludge and calculated the potential dose to the maximally exposed member of the public, based on a total activity of 8.89 mCi disposed in that year, to be 2.61 mrem/yr. The staff has reviewed the licensee's calculational methods and assumptions and found that they are consistent with NUREG-1 101, Onsite Disposal of Radioactive Waste, Volumes 1 and 2, November 1986 and February 1987, respectively. The staff finds the assessment methodology acceptable. For comparison, the radiation from the naturally occurring radionuclides in soils and rocks plus cosmic radiation gives a person in Michigan a whole-body dose rate of about 89 mrem per year outdoors. 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 the staff finds that the potential environmental impacts of leaving the contaminated sludge in place are insignificant. With regard to the non-radiological impacts, the staff has determined that leaving the soil in place represents the least impact to the environment.

Reference PMP-6010-OSD-001 Rev. 22 Page 91 of 91 OFF-SITE DOSE CALCULATION MANUAL Attachment 3.24 Safety Evaluation By The Office Of Nuclear Pages:

Reactor Regulation 89 - 91

5. CONCLUSION Based on the staff's review of the licensee's discussion, the staff finds the licensee's proposal to retain the material in its present location as documented in this Safety Evaluation acceptable. Also, this Safety Evaluation shall be permanently incorporated as an appendix to the licensee's Offsite Dose Calculation Manual (ODCM), and any future modifications shall be reported to NRC in accordance with the applicable ODCM change protocol.

I&M letter from E. E. Fitzpatrick to the NRC Document Control Desk, September 29, 1993 Therefore, the licensee's proposal to consider the slightly contaminated sludge disposed by retention in place in the manner described in the Donald C. Cook Nuclear Plant submittals date October 9, 1991, October 23, 1991, September 3, 1993, and September 29, 1993, is acceptable.

The guidelines used by the NRC staff for onsite disposal of licensed material and the staff's evaluation of how each guideline has been satisfied are given in Table 3.

Pursuant to 10 CFR 51.32, the Commission has determined that granting of this approval will have no significant impact on the environment (October 31, 1994, 59 FR 54477).

Principal Contributor: J. Minns Date: November 10, 1994 TABLE 3 20.2002 GUIDELINE FOR ONSITE STAFF'S EVALUATION 2

DISPOSAL 1, The radioactive material should be disposed of in such 1. Due to the nature of the disposed material, recycling to the a manner that it is unlikely that the material would be general public is not considered likely.

recycled.

2. Doses to the total body and any body organ of a 2. This guideline was addressed in Table 2. Although the maximally exposed individuals (a member of the 2.61 mrem/yr is greater than staff's guidelines, the staff general public or a non-occupationally exposed worker) finds it acceptable due to 9 yrs decay following analysis and from the probable pathways of exposure to the disposed the expected lack of activity detected in the 1991 survey.

material should be less than 1 mrern/year.

3. Doses to the total body and any body organ of an 3. Because the material will be land-spread, the staff considers inadvertent intruder from the probable pathways of the maximally exposed individual scenario to also address exposure should be less than 5 mrem/year. the intruder scenario.

4. Doses to the total body and any body organ of an 4. Even if recycling were to occur after release from regulatory individual from assumed recycling of the disposed control, the dose to a maximally exposed member of the material at the time the disposal site is released from public is not expected to exceed 1 mrem/year, based on regulatory control from all likely pathways of exposure exposure scenarios considered in this analysis.

should be less than 1 mrem.

, E: F. Branagan,. Jr. and F. J. Congel, "Disposal of Contaminated Radioactive Wastes from Nuclear Power Plants,"

presented at the Health Physics Society's Mid-Year Symposium on Health Physics Consideration in Decontamination/Decommissioning, Knoxville, Tennessee, February 1986, (CONF-860203).

REVISION

SUMMARY

Procedure No.: PMP-6010-OSD-001 Rev. No.: 22

Title:

OFF-SITE DOSE CALCULATION MANUAL Alteration Justification 10 CFR 50.59 is not applicable to this Per definition in Attachment 1 of PMP-2010-procedure revision. PRC-002. This is an administrative procedure governing the conduct of facility operations. Changes to this document are made in accordance with Technical Specification 5.5.1 and implemented through 12-EA-6090-ENV- 114, Effectiveness Review for ODCM/PCP Programs.

3.1. 1c, item Qi altered procedure number This corrects a typographical error as from 606 to 601 to correct typo. evidenced by the correct procedure title listed is for RPP-601. This meets editorial correction criteria (ECC) 1.

3.5.2a.6.c, added the clarifying words 'when This is clarifying information since vegetation is available' to the step. vegetation is not available year round and can only be collected when available. This is documented in AR 808291808291 This meets ECC p.

Deleted section 3.6 that discussed This-was deleted since it refers to 'data ornly' Mausoleum groundwater monitoring samples and is not part of the REMP and program. Renumbered remaining steps. was directed as part of AR 128023128023evaluation. The SG wells were also deleted from the map, Attachment 3.22. The locations had previously been removed from Attachment 3.19. This information currently resides in 12-THP-6010-RPP-634 for these wells. This meets ECC p.

Added clarifying Wording to

• footnotes in This was the result of the evaluation both Attachment 3.3 and 3.5 to denote that associated with AR 810352810352that concluded this applies to all of the surveillances to the that maintaining the blowdown treatment appropriate instrument, not just the channel radiation monitor in operational readiness check. was low value work and need not be maintained. Operations concurred with this deviation from the 'full deck' philosophy.

This meets ECC p.

Office Informationfor Form Tracking Only - Not Part of Form This is a free-form as called out in PMP-2010-PRC-002, Procedure Alteration, Review, and Approval. Page 1 of 2

REVISION

SUMMARY

Procedure No.: PMP-6010-OSD--(51 Rev. No.: 22

Title:

OFF-SITE DOSE CALCULATION MANUAL Alteration Justification Added groundwater (well water) sample Two additional wells, former NPDES wells stations W-16 and W-17 to Attachment 3.19 designated MW-20 and MW-21, were added and Attachment 3.22. to the REMP as part of the "D. C. Cook Nuclear Plant Groundwater Protection Project Charter" to more closely monitor groundwater flow characterized by the plant hydrology study. This change was done in response to AR 807643807643evaluation. .19 Reduced the spaces for NUREG-1301, Offsite Dose Calculation Ingestion - Milk Background Farm from two Manual Guidance- Standard Radiological to one and dropped farm to singular instead Effluent Controls for Pressurized Water of plural. Changed

• footnote from two Reactors, Table 3.12-1, REMP, only background farms to one also. Correctly requires one control (background) location.

formatted conditional if, then statement.

Reformatted map in Attachment 3.23 since it Map did not electronically transfer did not transfer well electronically during completely during the last revision. This revision 21. No changes were made to this meets ECC j and 1. It is the same map map. presented inrevision 20 of PMP-6010-OSD-001. No changes, were made during the past two revisions of this document. AR 808170808170documents this.

Office Informationfor Form Tracking Only - Not Part of Form This is a free-form as called out in PMP-2010-PRC-002, Procedure Alteration, Review, and Approval. Page 2 of 2