AEP-NRC-2018-35, Annual Radioactive Effluent Release Report - Off-Site Dose Calculation Manual

Annual Radioactive Effluent Release Report - Off-Site Dose Calculation Manual

ML18123A350
Person / Time
Site:	Cook
Issue date:	04/30/2018
From:	American Electric Power, Indiana Michigan Power Co
To:	Office of Nuclear Reactor Regulation
References
AEP-NRC-2018-35 PMP-6010-0SD-001, Rev 25
Download: ML18123A350 (92)

Text

Doc No.: PMP-6010-0SD-001 Rev No.: 025

Title:

OFF-SITE DOSE CALCULATION MANUAL Alteration Cat.: Minor Revision CDl/50.59: NIA PORC Mtg. No.: NIA CARB Mtg. No.: NIA Admin Hold AR No.: N/A Superceding Proc(s): NIA Temp Proc Exp Date: NIA Temp Change Exp Date: NIA Temp Proc/Cbange End: N/A Effective Date: 7/22/2015 4:00:00 AM Approvals Name Review/Approval Type/Capacity Date j

/Wendzel, Regan  !!7 Approval Authority l/07/08/2015 07:49 I/

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====================

Signature Comments jApproved per Plant Manager, Sam Partin. II

!Approved, however, a few editorial comments need incorporated. The mark up was placed on 1

!Erik's desk. 11 1

0

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PMP-6010-0SD-001 Rev. 25 Page 1 of 89 POWER:

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OFF-SITE DOSE CALCULATION MANUAL Information I Erik Merchant Environmental Manager Environmental Writer Document Owner Cognizant Organization TABLE OF CONTENTS*

1 PURPOSE AND SCOPE ............................................................................. 4 2 DEF'INITIONS AND ABBREVIATIONS ........................................................ 4 3 DET~S *. ~ ...................................*............ "............................................. 6 3.1 Calculation of Off-Site Doses ................................................................ 6 3.1.1 Gaseous Effluent Releases .. .-....................................................... 6 3.1;2 Liquid Effluent Releases ..... .-................. :,........................ ~ ........ 12 3.2 Limits of Operation and Surveillances of the Effluent Release Points ............. 15 3.2.1 Radioactive Liquid Effluent Monitoring Instrumentation .... ............ 15 3.2.2 Radioactive Gaseous Effluent Monitoring Instrumentation ............... 16 3.2.3 Liquid Effluents ****************~*******************-~ ..-. ............................ 17

a. Concentration Excluding Releases via the Turbine Room Sump (TRS) Dischargy .........*.................................................... 17

b. Concentration of Releases from the TRS Discharge .................... 18

c. Dose ..... ; ........... , ............................................................ 19

d. Liquid Radwaste Treatment System ........ *............................... 19 3.2.4 Gasoous Effluents ...................................................*.............. 21

a. Dose Rate ............................................................ , ......... 21 *

b. Dose - Noble Gases .......................................................... 22

c. Do.se - Iodine-131, lodine-133, Tritium, and Radioactive Material in Particulate Form .......................:................. ; ............*....... 22 cl. Gaseous Radwaste Treatment. .............................................. 23 3.2.5 Radioactive Effluents -Total Dose ................................... ~ ............ 25 3.3 *caJculation of Alarm/Trip Setpoints **~**************************************************** 27 3.3.1 Liquid Monitors .................................................................... 27

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

b. Liquid Continuous Monitor Setpoint Methodology ..................... 29

. 3.3.2 Gaseous Monitors ..... 0 ********************* 0 ********.****************************** 31

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

_ 31 b: Waste Gas Storage Tanks ................................................... 35

c. Containment Purge and Exhaust System ................. : ............. *... 35

d. Steam Jet Air Ejee;tor System (SJAE) ..................................... 36

e. Gland Seal Condenser Exhaust ........................ : .................... 37

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PMP-6010-0SD-001 Rev. 25 Page2 of 89 FOWER'

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OFF-SITE DOSE CALCULATION MANUAL Information I Erik Merchant Environmental Manager Environmental Writer Document Owner Cognizant Organization 3.4 Radioactive Effluents Total Dose .~ ........................................ , ............... 37 3.5 Radiological Environmental Monitoring Program (REMP) .............*........... 37 3.5.1 Purpose of the REl\iP ............................................................. 37 3 .5 .2 Conduct of the REl\iP ............................................................ 38 3 .5 .3 .Anilual Land Use Census ........................................................*40 3.5.4 Interlaboratory Comparison Program ***.*.................*.*****.********.* 41 3.6 Meteorological Model .......................................................................... 42

3. 7 Reporting Requirements .......................... ~ .......................................... -42 3.7 .1 Annual Radiological Environmental Operating Report (AREOR) ..... 42 3.7 .2 Annual Radiological Effluent Release Report (ARERR) .................. 43 3.8 10 CFR 50.75 (g) Implementation *....................*................................... 45 3.9 Reporting/Management Review*****************************"!:*~**************************** 45 4 E'JN'AL CONDITIONS .................................................................................46 5 REFERENCES* *~*******************"****************~**************************************************46 SUPPLEMENTS Attachment 3.1 Dose Factors for Various Pathways .. , ................................... :... ,... Pages 49 - 52, Attachment 3 .2 Radioactive Liquid Effluent Monitoring Instruments ................... Pages 53 - 55 Attachment 3.3 Radioactive Liquid Effluent Monitoring Instrumentation Surveillance Requirements ............................................................ Pages 56 - 57

• Attachment 3.4 Radioactive Gaseous Effluent Monitoring Instrumentation .......... Pages 58 - 60 Attachment 3.5 Radioactive Gaseous Effluent Monitoring Instrumentation Surveillance Requirements ............................................................ Pages 61 - 62 Attachment 3 .6 Radioactive Liquid Waste Sampling and Analysis Program ....... '.Pages 63 - 64 Attachment 3.7 Radioactive Gaseous Waste Sampling and Analysis Program ..... Pages 65 - 66 Attachment 3.8 Multiple Release Point Factors for Release Points ................................ Page 67

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OFFMSITE DOSE CALCULATION MANUAL Information I Erik Merchant Environmental Manager Environmental Writer Document Owner Cognizant Organization .9 Liquid Effluent Release Systems ............................................................ Page 68 .10 Plant Liquid Effluent Parameters ........................................................... Page 69 .11 Volumetric Detection Efficiencies for Principle Gamma Emitting Radionuclides for Eberline Liquid Monitors ........................... Page 70 .12 Counting Efficiency Curves for R-19, and R-24 ........................... Pages 71 - 72 .13 Counting Efficiency Curve for R-20, and R-28 ..................................... Page 73 .14 Gaseous Effluent Release Systems ........................................................ Page 74 .15 Plant Gaseous Effluent Parameters ........................................................ Page 75 .16 10 Year Average of 1995-2004 Data ............................................ Pages 76 - 77 .17 Annual Evaluation of x/Q and D/Q Values For All Sectors ................. Page 78 .18 Dose Factors .................................................................................. Pages 79-80 .19 Radiological Environmental Monitoring Program Sample Stations, Sample Types, Sample Frequencies ............................... Pages 81 - 84 .20 Maximum Values for Lower Limits ofDetectionsA,B - REMP ..... Pages 85 - 86 .21 Reporting Levels for Radioactivity Concentrations in Environmental Samples .......................................................................... Page 87 .22 On-Site Monitoring Location - REMP ................................................... Page 88 .23 Off-Site Monitoring Locations -REMP ................................................. Page 89

Information I PMP-6010-0SD-001 I Rev. 25 I Page 4 of 89 OFF-SITE DOSE CALCULATION MANUAL 1 PURPOSE AND SCOPE

• 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 methodologies 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:

Sor 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 /> Dor 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 /> Wor weekly At least once per 7 days Mor 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.

SIU 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

Information I I Rev. 25 I Page 5 of 89 OFF-SITE DOSE CALCULATION MANUAL Member(s) of All per_sons who are not occupationally associated with the Public plant.-. boes 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(l; + C(2; + ... ~ 1 Lr1; L(2J Where; C(l) = Concentration of pt detected nuclide Cc2) = Concentration of 2nd detected nuclide L(l) = Reporting Level of 1st nuclide from Attachment 3.21, Reporting Levels for Radioactivity Concentrations in Environmental Samples.

Lc2) = Reporting Level of 2nd 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.

Information I PMP~6010-0SD-001 I Rev. 25 l Page 6 of 89 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

• MIDEX

• MIDEL

• MIDEG

• MIDEN

b. The subprogram used to enter and edit gaseous release data is called MDlEQ (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,D/J air=~ *I,[( M;or N;)*Q;

• 3.17E-8]

Where; Dy , Dp air = the gamma or beta air dose in mrad/yr to an individual receptor x IQ = the annual average or real time atmospheric dispersion factor over land, sec/m3 from Attachment 3.16, 10 Year Average of 1995-2004 Data Mi = the gamma air dose factor, mrad m 3 / yr µCi, from Attachment 3.18, Dose Factors Ni = the beta air dose factor, mrad m 3 /_ yr µCi, from Attachment 3.18, Dose Factors

Information I PMP-6010-0SD-001 I Rev. 25 I Page 7 of 89 OFF-SITE DOSE CALCULATION MANUAL

= the release rate of radionuclide, "i", in µCi/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).

,,,f.Q 2.03

• Al:

_

• X

• Lg T1

= Umg Where; 2

Lg === minimum of c,2 + H c or Lg= Ji er z Zg 2ff 8 x = distance downwind of the source, meters. This information is found in paramet~r 5 of MIDEX.

um 8

= wind speed for ground release, (meters/second) a-z8 = vertical dispersion coefficient for ground release, (meters),

(Reg. Guide 1.111 Fig.I)

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

(Containment Building = 49.4 meters)

Tr = terrain factor(= 1 for Cook Nuclear Plant) because we consider ~ our releases to be ground level (see parameter 5 inMIDEX).

2.03 = .J2..;.-:ti ..;.-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.

Information I PMP-6010-0SD-001 I Rev. 25 *I Page 8 of 89 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. G1,1ide 1. i09: *

1. Total Body Plume Pathway (Eq 10)

Dose(mrem/year)=3.17E-8*L(Q;* l7'Q*S1* DFB;)

Where; Sr = shielding factor that accounts for the dose reduction due to shlelding provided by residential structures. during occupancy (maximum exposed individual = 0. 7 per Table E-15 of Reg. Guide 1.109)

DFBi = the whole body dose factor from Table B-1 of Reg.

Guide 1.. 109, mrem - m 3 per µCi - yr. See Attachment 3.18,' Dose Factors.

Qi = the release rate of r~dionuclide "i", in µCi/yr

2. Skin Plume Pathway (Eq 11)

Dose(mrem/yr)=3.17E-8* s1 * ~*[~(Qt* 1.11* DF~)+ L(Q;

• DFS;)]

Where; 1.11 = conversion factor, tissue to air, mrem/mrad DF / = the gamma air dose factor for a uniform semi-infinite cloud of radionuclide ".i", in mrad m 3/µCi yr from Table B-1, Reg. Guide 1.109. See Attachment 3.18,.

Dose Factors,

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

Information I PMP-6010-0SD-001 I Rev. 25 I Page 9 of 89 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:

DIP (mrem/year) = 3. J7E -8

• L(R;

• W

• Q;)

Where; R; = the most restrictive dose factor for each identified radionuclide "i", in m2 mrem sec/ yr µ.Ci (for food and ground pathways) or mrem m 3 I yr µCi (for inhalation pathway), for the appropriate pathway For sectors with existing pathways within five miles of the site, use the values of R; for these real pathways, otherwise use pathways distance of five miles. See Attachment 3.1, Dose Factors for Various Pathways, for the maximum Ri values for the most controlling age group for select~d radionuclides. R; values were generated by computer code PARTS, see NUREG-0133, AppendixD.

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:

Wm x IQ for the inhalation pathway, in sec/m3

-OR-Wrg = DI Q for the food and ground pathways in 11m2 Q;c = 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 µ.Ci/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.

Information I PMP-6010-0SD-001 I Rev. 25 I Page 10 of 89 OFF-SITE DOSE CALCULATION MANUAL

1. 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. Carbon-14 (C-14) supplemental information

1. The quantity of C-14 released to the environment may be estimated by use of a C-14 source term scaling factor based on power generation (Ref. RG 1.21, Revision 2). A recent study recommends a source term scaling factor of approximately 9.0 to 9.8 Curies/GWe-yr for a Westinghouse Pressurized Water Reactor (Ref.

EPRI 1021106 "Estimation of Carbon-14 in Nuclear Plant Gaseous*.

Effluents" December 23, 2010). For this method, a scaling factor of 9.4 Curies/GWe-yr shal). be used.

2. C-14 releases from PWRs occur primarily as a mix of organic carbon (methane) and inorganic carbon (carbon dioxide). For this method, an average organic fraction of 80 % with the remaining 20 % being assumed as carbon dioxide shall be used.

3. Dose is calculated utilizing the methodology prescribed in RG 1.109 Appendix C, with the vegetation dose being the most predominant.

Adjustments for growing seasons, percentage of C-14 generated assumed released from the reactor coolant in gaseous form via batch releases, seasonal X I Q , and other industry methodologies being considered by the NRC may be applied as desired should their acceptance of these methods occur.

k. Steam Generator Blowdown System (Start Up Flash Tan1c 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= µCi* GPM

• time on flash tank (min)* 3. 785E -3 ml

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

3. The flow rate is determined from the blowdown valve position and the time on the start up tank, or using installed plant blowdown flow instrumentation. Chemistry Department performs the sampling and analysis of the samples.

Information I PMP-6010-0SD-001 I Rev. 25 I Page 11 of 89 OFF-SITE DOSE CALCULATION MANUAL

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 µ,Ci/g dose equivalent 1-131.

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

QY = Ci* IPF*. Rsgb Where; Qy = the release rate of 1-131 from the steam generator flash tank vent, in µ,Ci/sec Ci = the concentration (µ,Ci/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.

Information I PMP-6010-0SD-001 I Rev. 25 I Page 12 of 89 OFF-SITE DOSE CALCULATION MANUAL 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 MD lEB (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 p\lrposes to maintain uniformity with this defined pathway.

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

1. Potable Water (Eq 1)

Rap,-=1100* MP* F Uap *

• "Q D * * -;.,,,

• 2.23E-3 "7-' ; a,pje Where; Rapi = 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.

Information I PMP-6010-0SD-001 I Rev. 25 I Page13 of 89 OFF-SITE DOSE CALCULATION MANUAL 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 Qi = 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-11 through E-14 of Reg. Guide 1.109 and are located within the MIDAS code.

Ai = the radioactive decay constant for radionuclide "i", in hours- 1 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. (1:p = 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)

R apj

=1100* Uap Mp* F* 2.23E-3 ~ 1

• "'Q

• B*

• D aipJ*

'P

• e*).,tp Where, Bip = the equilibrium bioaccumulation factor for nuclide "i" in pathway "p", expressed as pCi LI 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 .

. 1:p = 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.

Information I PMP-6010-0SD-001 I Rev. 25 l Page 14 of 89 OFF-SITE DOSE CALCULATION MANUAL

3. Shoreline Deposits (Eq 3)

R ap;.= 110' 000

• Mp* F*

U *W ap * "'Q*

2.23E-3 ~ '

• T,

• Da,p;le 1

.. r -A;tp]* [1- -~lb]

l e 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 Daipi = the dose factor for standing on contaminated ground, in mrem m2 / hr pCi. The values are taken from table E-6 of Reg. Guide 1.109 and are located within the MIDAS code.

See Attachment 3.1, Dose Factors for Various Pathways.

ti, = the period of time for which sediment or soil is exposed to the contaminated water, l.31E+5 hours. Given in*

MIDEL as itein 6 of parameter 4.

ti, = the average transit time required for nuclides to reach the point of exposuret Ohours. Given as #28 of parameter 4 inMIDEL.

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

.aquati~ 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 DSlLI are printed in LDRPT (LP). These results are used in the monthly report of liquid releases.

Information PMP-6010-0SD-001 Rev. 25 - - - 'I- - ~Page 15 of 89

-""------II OFF-SITE DOSE CALCULATION MANUAL

g. In addition, the program DOSUM (DM) is used to search the results files of DSlLI to find the maximum liquid pathway individual ,doses. The highest exposures are then printed iri 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. G1;1ide 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 surveillan.ce interval.

3 .2 Limits of Operation and Surveillances <?f the Effluent Release Points

3. 2.1 Radioactive Liquid Effluent Monitoring Instrumentation

~- .The radioactive liquid effluent monitoring instrumentation channels_

shown in Attachment 3.2, Radioactive Liquid Effluent Monitoring Instruments, are operable *with their alann/trip setpoints .set to ensure that the limits of step 3 .2. 3a, Concentration Excluding Releases vfa the Turbine 1:loom Suml) (TRS) Discharge,.are not exceeded..

b. The applicabili1y of each ch~el is shown in Attachment ~.2, Radioactive Liquid Effluent Monitoring Instrum\;lnts:

c. With a radioactive liquid effluent monitoring instrumentation channel alatm/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 suryeillance interval, excluding the Wtial performance.

e. Determine the setpoints in accordance with the methodolbgy described in

.step .3.3'.1, Liquid Monitors. Record the setpoints.

Information I PMP-6010-0SD-001 I Rev. 25 I Page 16 of 89 OFF-SITE DOSE CALCULATION MANUAL

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 alann/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]

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 ensu~e that the limits of step 3.2.4a, Dose Rate, are nc;,t 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 alann/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.

/

Infonnation .. 1 PMP-6010-0SD-001 I Rey. 25 I Page 17 of 89 QFF.,SITE DOSE CALCULATION-MANUAL

d. With less than the minimum number of radioactive gaseous effluent monitoring instrumentation ~hannels 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 perforniance, 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 dOC\lIDent.

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, CltANNEL CALIBRATIO~, and CHANNEL OPERATIONAL TEST operations at the frequencies shown in Attachment 3 5, Radioactive Gaseous Effluent Monitoring .

Instrumentation Surveillance Requirements.

BASES ;_ GASEOUS The radiqactive gaseous effluent instrumentation is provided to monitor and control, as

• applicable, the ,releases of radioactive materials in gaseous effluents dqdng actual or potential releases. 1,'he 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 specjfied_ in Section 11. 3 of the Final Safety Analysis Report for the Donald C. Cook Nuclear Plant, 3.2.3 'Liquid Effluents

a. Concentratiqn Excluding Releases via the Turbine Room: SUmp (TRS)

Discharge

1. Limit the concentration of radioactive material relea/,ed via the Batch Release Tanks or Plant Continuous Releases (excluding only TRS discharge to the Absorption Pond) to uprestricted areas to the *

• concentrations in 10 CPR 20, Appendix 13, 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 µCi/ml total activity.

Information I PMP-6010-0SD-001 I Rev. 25 I Page 18 of 89 OFF-SITE DOSE CALCULATION MANUAL

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 µCi/ml total a_ctivity.

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. l 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.

Information I PMP-6010-0SD-001 I Rev. 25 I Page 19 of 89 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 :c::; 1.5 mrem/unit to the total body and to :c::; 5 mrem/unit to any organ, a.J._ld during any calendar year to :c::; 3 mrem/unit to the total body and to :c::; 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.l above, prepare and submit a Written Report, pursuant to 10 CPR 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, Special and 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 0.12 mrem (0.06 mrem/unit x 2 units) to the total body or 0.4 mrem (Ci.2 mrem/unit x 2 units) to any organ.

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

Information . I PMP-6010-0SD-001 I Rev. 25 I P~ge 20 of 89 OFF-SITE DOSE CALCULATION MANUAL e, Drai,nage of high conductivity water (Component Cooling Water and ice melt water containing sodium tetraborate) shall be evaluated to decide whether it sb.ould 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 bas on the radioactive wastewater dentjneralization 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:

( G)(Vi) + (Ca)(Va) = (C)(Vi)

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 c1* = the final concentration of the system after the addition Vt = . the final ~olume of the system after tb.e addition The intentis to keep the:

• WDS below 500 µmhos/cc.

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

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 CPR :Part 20; Appendix B, Table z. This limitation provides

, ad<litional assurance that the levels of radioactive materials ih bodies of water outside the

• site will not result in exposures *greater than 1) the-Section II.A design objectives of Appendix I, 10 CPR Part 50, to an individual and 2) the limits of 10 CPR 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.

Information I PMP-6010-0SD-001 I Rev. 25 I Page21 of 89 OFF-SITE DOSE CALCT)LATION MANUAL DOSE This specification is provided to implement the requirements of Sections IT.A, ill.A, and IV.A of Appendix I, 10 CFR Part 50. The dose limits implement the guides set forth in Section IT.A of ApP.endix 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 "a,s low as is reasonably achievable". Also, for fresh water sites with drinking water supplies which can be potentially affected by plant operations, there is reasonable assurance that the operation of the facility will not result in radionuclide concentrations in the finished drinking water that are in excess of the requirements .of 40 CFR 141. The dose calculations in the ODCM implement the requirements in Section ID.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 Amiual Doses to Man

• from Routine Releases of Reactor Effluents for tlie 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 e_ach reactor at the site. The liquid effluents from the shared system are proportj.oned among the units sharing the system.

LIQUID WASTE TREATMENT The operability of the liquid radwaste treatment system ensures that this system will be '

available for use whenever liquid effluents require treatment prior to release to the environment. The requirements that the appropriate portions of this system be used when specified provide assurance that the releases of radioactive materials in liquid effluents will be kept "as low as is reasonably achievable". This specification implements the requirements of 10 CPR 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 IT.D of Appendix Ito .10 CPR 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 IT.A of Appendix I, 10 CPR Part 50, for liquid effluents.

3.2.4 Gaseous Effluents

a. DoseRate

Information I PMP-6010-0SD-001 I *- Rev. 25 I Page22 of 89 OFF-SITE DOSE CALCULATION MANUAL

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 describe~ 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.

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 addre~sed 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 with half-lives greater than eight days in gaseous effluents

• released to unrestricted areas (site boundary) to the following:

a) During any calendar quarter to less than or equal to 7.5 mrem/unit to any organ b) During any calendar year to less than or equal to 15 mrem/unit to any organ.

Information I PMP-6010-0SD-001 I Rev. 25 I Page 23 of 89 OFF-SITE DOSE CALCULATION MANUAL

2. With the calculated dose from the release of radioiodines, radioactive materials in particulate form, or radionuclides other than noble gases in gaseous effluents exceeding any of the above limits, prepare and submit a Written Report, pursuant to 10 CFR 20.2203 and addressed in step 3.2.3c.2, within 30 days after learning of the event.

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

d. Gaseous Radwaste Treatment

1. The UFSAR (Updated Final Safety Analysis Report) states that radioactive waste gas should be held for 45 days of decay time.

2. 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 0.4 mrad (0.2 mrad/unit x 2 units) for gamma radiation and 0.8 mrad (0.4 mrad/unit x 2 units) 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.

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

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 CPR 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 11.C design objectives of appendix I to 10 CPR 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 dos~ 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 ra_te 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.

Information I PMP-6010-0SD-001 I Rev. 25 I Page24 of 89 OFF-SITE DOSE CALCULATION MANUAL 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 n.B, ill.A, and IV .A of Appendix. I, 10 CFR Part 50. The dose limits implem~nt the guides set forth in Section ILB of Appendix I.

The ACTIQN statements provide the required operating flexibility and at the same time implement the guides set forth in section IV .A of Appendix I to as so.re that the releases of radioactive material in gaseous effluents will be kept "as low as is reasonably achievable".

The Surveillance Requirements implement the requ_irements in Section III.A of Appendix I that conform with the guides of Appendix I to be shown by calculatiop_al 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 l, 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 l. July 1977. The ODCM equations provided for determining the air doses at the site boundary will be based upon the historical average atmospherical conditions. NUREG-0133 provides methods for dose calculations consistent with Regulatory Guides 1.109 and 1.111.

DOSE, RADIOIODINES, RADIOACTIVE MATERIAL 1N PARTICULATE FORM, ANp 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.

• Information PMP-6010-0SD-001 I Rev. 25 I Page25 of 89 OFF-SITE DOSE CALCULATION MANUAL The ACTION statements provide the required operating flexibility and at the saine 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 particu1ate form, and radionuclides, other than noble gases, are depenqent on the existing radionuclide pathwa,ys 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 Desigu Criterion Section 11.1 of the Final Safety Analysis Report for the Donald C. Cook Nuclear Plant, and design objective Section Il.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 Il.B and Il.C of Appendix. I, 10 CFR Part 50, for gaseous effluents.

3.2.5 Radioactive Effluents -Total Dose

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

b. With the calculated doses from the release of radioactive materials in liquid or gaseous effl~ents 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:

Information I PMP-6010-0SD-001 I Rev. 25 I Page 26 of 89 OFF-SITE DOSE CALCULATION MANUAL

• fuvestigate 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.ll(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 CPR 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]).

BASES -- TOTAL DOSE This specification is provided to meet the dose limitations of 40 CFR 190. The specification requires the preparati~n 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 coIJllilitment 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.

Information I PMP-6010-0SD-001 I Rev.25 I Page 27 of 89 OFF-SITE DOSE CALCULATION MANUAL

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 CPR 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.

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 systeD,I design and operation can be found in the system descriptions. The setpoints are intended to keep releases within the limits of 10 CPR 20, Appendix B, Table 2, Column 2. Determine setpoints using either the batch or the cont4Iuous 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 CPR 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.

Information I PMP-6010-0SJ)-001 I Rev. 25 I Page.28 of 89 OFF-SITE, DOSE CALCULATION MANUAL.

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 Attacbment 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 i, Column 2. The equation to calculate the flow rate is from Addendum AAl of NUREG-0133:

[

}: C; ]*_f_-:;,F+f LIMIT; MRP Where; Ci = the concentration of nuclide "i" in µCi/ml LIM{Ti = the 10 CFR 20, Appendix B, Table 2, Column 2 limit of

.nuclide "i" in µCi/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 irroperation.

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 ~ay be rearranged to solve for the maximl,lill effluent release flow rate (f).

I

Information I PMP-6010-0SD-001 I Rev. 25 I Page 29 of 89 *

• OFF-SITE DOSE CALCULATION MANUAL

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. Tlle predicted v:;il.ue is generated by converting the effluent concentration for each gamma emitting radionuciide to counts per unit of time .as per Attachment 3 .11, Volumetric Detection Efficiencies for Principle Gamma Emitting Radionuclides for Eberline Liquid Monitors, or Attachment 3.12, Counting Efficiency Curves for R-19, and R-24. The sum of all the counts per unit of time is th~ predicted count rate. The predicted count rate can tlien 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 (E~W) ~ystems.

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

• R-19 ot 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, Appenqix B, Table 2.,

Column 2, entering the unrestricted area following dilution,

3. The monitors on steam generator blowdown and blowdown .*

treatment systems have trip functions associated With their setpoints.

Essential service water monitors are equipped with an alarm function only and monitor effluent in the event the Containment

• Spray Heat Exchangers are used.

4. The equation used to determine the setpoint for continuous monitors is from Addendum AAl ofNUREG-0133:

s<. C* E.r+*

=:JJ MRP*F*SF

. p- f Where; Sp = setpoint of monitor (cpm)

Information I PMP-6010-0SD-001 I Rev. ~5 I Page 30 of 89 OFF-SITE DOSE CALCULATION MANUAL C = 5E-7 µCi/ml, maximum effluent control limit from 10 CPR 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, 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:

I( C;

• Ejf ,) replaces C

• Eif.f I e, LIMIT, MRP = multiple release point factor. A factor such that when all the release points are operating at one time the limits of 10 CPR 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).

Information I PMP-6010-0SD-001 I Rev. 25 I P~e31 of89 OFF-SITE DOSE CALCULATION MANUAL 3.3.2 Gaseous Monitors For the purpose of implementing Step 3 .2.2, Radioactive Gaseous Effluent Monitoring Inst:rillhentation, 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 nor 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 paramete~(s).

Gaseous effluent monitor high al~ setpoints will routinely be established at a fraction of the maximum allowable setpoint (typically 10% of the setpoint) for ALARA purposes. Alert alarms will norn1ally be set to provide adequate indications of small changes in radiological conditions.

NOTE: IF the -setpoint calculation methodology changes or the as.sociated factors change for Unit Vent, Air Ejector and/or Gland Seal monitors, THEN initiate a review by"Emergency Planning to ensure that the requu;ements of 10 CFR 50.54 (q) are maintµned.

a. Plant Unit Vent

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 (Unifi), VRS-2505 (Unit.2)] to assure that appllcable alarms .and trip actions (isolation of gaseous release)

, will occur prior to exceeding .the limits in step 3.2.4,. Gaseous Eftluel).ts. The alarm setpoint values will be established using the following unit analysis equation:

• Where; Sp "'" the maximum setpoint of the monitor in µCi/cc for .

release point p, based on the mo~t limiting organ SF = an administrative operation safety factor, less than 1.0

Information I PMP-6010-0SD-001 I Rev. 25 I Page 32 of 89 OFF-SITE DOSE CALCULATION MANUAL 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, fp) from each release point p.

• Compute LQp (or Lfp) for all release points.

• Ratio Qp/LQp (or fp/Lfp) for each release point.

This ratio is the MRP for that specific release point

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

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 cfin for Unit 1 and

• 143,400 cfm for Unit 2.

Information I PMP-6010-0SD-001 I Rev. 25 I Page 33 of 89 OFF-SITE DOSE CALCULATION MANUAL DLi = dose rate limit to organ "j" in an unrestricted area (inrem/yr).

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

• Total Body ~ 500 ~em/year

• Skin ~ 3000 mrem/year

• • - Any Organ~ 1?00 mrem/year x IQ = The worst case annual average relative concentration in the applicable sector or area; in sec/m3 (see Attachment 3.16, 19 Year Average of 1995-2004 Data).

Wi = Weighted factor for* the radionuclide*:

Where, Ci = concentration of the most abundant radionuclide "i" Ck == total c;oncentration of all identified radionuclides in tbat release pathway. .For batch releases, this value may be set to 1- for conservatism.

DCFii = dose conversion factor used to relate radiation dose to organ "j "., from exposµre to radionuclide "i" in mrem m 3 / yr µCi, See

')

following equations.

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

For the whole body: DCF;j = Ki Where; Ki = whole body dose factor due to gail).II).a emissions for each identified noble gas radionuclide in mrem m.3 I yr µCL See Attachment 3.18, Dose Factors.

Information I PMP-6010-0SD-001 I Rev. 25 I Page 34 of 89 OFF-SITE DOSE CALCULATION MANUAL For the skin: DCFu =Li+ 1.lMi Where; Li = skin dose factor due to beta emissions for each identified noble gas radionuclide, in mrem m 3 / yr µCi. 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).

Mi = the air dose factor due to gamma emissions for each identified noble gas radionuclide in mrad m 3 I yr µCi. See Attachment 3.18, Dose Factors.

For the thyroid, via inhalation: DCFu = Pi Where; P; = the dose parameter, for radionuclides other than noble gas, for the inhalation pathway in mrem m 3 I yr µCi (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 cp.annel setpoint, SP, will be set such that the dose rate in unrestricted areas to the whole body, skill 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 eves 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.

Information

• I PMP-6010-0SD-001 I Rev. 25 I Pag~ 35 of-89 OFF-SITE DOSE CALCULATION MANUAL

7. At certain times, it may be desirable to increase the high alarm setpoint, if the vent flow rate is decreased. -1bis may be accomplished in one, of two ways.

Max.Cone( µCi/cc) *Max.Flowrate (cfin) =New.Maxcfm New M~Concentration ( µCi/cc)

-OR-Max.Conc ( µCi/cc)

• Max.Flowrate* (cfin)

NewMaxµCilcc New Max.Flowrate (cfin)

b. Waste Gas Storage Tanks L The gaseous effluents discharged from the Waste Gas System are monitored by the vent stack monitors VRS-1505 and VRS-2505.

2. fu 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 ~hannel (VRS-1°505 or VRS-2505). Therefore, for any gaseous release configuration,

• which includes nonnal operation and waste gas system gaseous discharges, the alarm setpoint of the plant vent radiation monitor will be recomputed to determine the most iimiting 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 apptove.d Operations' procedures. * *

3. It is nonnally prudent to allow 45 days of decay prior to releasing a Gas Decay Tank (GOT). There are extenuating, operattonaI

\ circumstances that may prevent this from occurring. Under these circumstances, such as high oxygen concentration creating a combustible atmosphere, it is prudent to waive the 45-day decay for safety's sake.

c. Containment J>u:rge and Exhaust System L The gaseous effluents discharged by the Containment Purge and Exhaust Systems, and Instrutnentation 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 Rat~.

fuformation I PMP-6010-0SD-001 I Rev. 25 I Page 36 of 89

_OFF:-SITE DOSE CALCULATION MANUAL

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 Conqrinment purge before release.

3. The Upper Containment area is monitored by normal range area gamma monitors (Tag No. VRS-1101/1201 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.

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 value will be established using the following unit analysis equation:

SF* MRP* DL*

s = J SJAE F P

• Xi~

.In *" (

Li W,

• DCFij )

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

Information I PMP-6010-0SD-001 I Rev. 25 I Page 37 of 89 OFF-SITE DOSE CALCULATION MANUAL

e. Gland Seal Condenser Exhaust

1. The gaseous effluents from the Gland Sea} Condell$er Exhaust discharged to the environment are continuously monitored by radiation monitor (fag 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:

Where; SGscE = the maximum setpoint, based on the most limiting o:rgan, in µCi/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 jn steps 3:2.3c (Dose), 3.2.4b (Dose - Noble Gases), and 3.2.4c (Dose - Iodine-131, Xodine-133, Tritiµ.m, and Radioactive*MateriaI in Particulate Form) of this /

procedure. Dose contribution from direct radiation exposur~ will be based on the results of the direct radiation monitoring devices locate.d at the REMP monitoring stations, and.reflects direct dose* bofu from the Dry Cask Storage Facility (ISFSI) licensed under Holtech International and both units of Cook.

See NUREG-0133, section 3.8.

• 3.5 Radiological Enviromilental Monitoring Program (REMP) 3 .5 .1 Purpose of the REMP

a. The purpose of the REMP is to:

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

• Monitor critical environmental exposure pathways,

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

Information I I Rev. 25 I Page 38 of 89 OFF-SITE DOSE CALCULATION MANUAL

• Assist with fulfilling the requirements of the Groundwater Protection fuitiative (GPI).

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 th~ scope of the REMP for the Donald C. Cook Nuclear Plant.

3.5.2 Conduct of the REMP [Ref. 5.2.lv]

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 ofDetectionsA,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 (ARBOR).

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.

fuformation I PMP-6010-0SD-001 I Rev. 25 I Page 39 of 89 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 J;iave 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.

• Document the notification using PMP-6090-PCP-100, Data Sheet 2, Part 4 Radioactive Liquid Spill Which May hnpact Groundwater.

• Forward a copy of the notification to the Environmental Department Manager.

Information I PMP-6010-0SD-001 I Rev. 25 I Page 40 of 89 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 moI).itoring 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 ofDetectionsA,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.lv]

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.

Information I PMP-6010-0SD-001 I Rev. 25 I Page 41 of 89 OFF-SITE DOSE CALCULATION MANUAL

b. In lieu of the garden census, broad leaf vegetation sampling of at least three different kinds of vegetation (if available) may be performed as close to the site boundary as possible (within 5 miles) in each of two different direction sectors 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, if required by the results of the census.

This census satisfies the requirements of Section IV.B.3 of Appendix I_ to 10 CPR 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 (26 kg/yr) of leafy vegetables assumed in Regulatory Guide 1.109 for consumption by 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 yield 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 ¥1 an Interlaboratory Comparison Program, for radioactive materials. Address program results and identified deficiencies in the ARBOR.

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

- - - - - - - - - - - - - - - - - - - - - - - - ~

Informatlon I P~-:6010.;0SD-001 _ -, Rev. 25 I Page 42 of 89 OFF-SITE. DOSE CALCULATION MANUAL BASES - JNl'ERLABORATORY COMPARISON PROGRAM The requirement for participation in an Interlaboratory Comparison Program is provided to ensur!;: independent checks oii 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 3.6.1 Tbree towers are used to determine the meteorological conditions .at Donald C.

Cook Nuclear Plant. One of the towers is located at the La;ke Michigan shoreline to determine the meteorological parameters associated with unmodified shoreline air. The data is accumulited by microprocessors at the tower sites and normally transferred to the central computer every 15 minutes. .

3.6.2 The central computer uses a meteorol_ogical 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 (ARBOR)

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. lj, TS 5.6.2]

b. Include in the ARBOR:

• Summaries, interpretations, and statistical evaluation of tlie 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 oµ the environment. .

• • The results of the land use censuses required by step 3.5.3, Annual Land Use Ce11.$US.

• 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.

Information I PMP-6010-0SD-001 I Rev. 25 I Page 43 of 89 OFF-SITE DOSE CALCULATION MANUAL

• Summarized and tabulated results of all radiological environmental samples taken during the reporting period. In the event that some results are not available for inclusion with the report, submit the report noting and explaining the reasons for the missing results.

Submit the missing data as soon as possible in a supplementary report.

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

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

• The results of participation in the Interlaboratory Comparison Program required by step 3.5.4, Interlaboratory Comparison Program.

• The results of non-REMP samples taken for informational purposes in support of non-program specific investigations, such as rainfall studies of tritium recapture for example.

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 prior to May pt of each year. [Ref 5.2. lj, 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 prior to May 1sr 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.

Information . I PMP-6010-0SD-001 I Rev. 25 I Page 44 of 89 OFF-SITE DOSE CALCULATION MANUAL

• 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 y~r.

• Include an assessment of the radiatfon 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 detennined by sampling frequency and measurement) for detennining the gaseous pathway doses.

• Inoperable radiation monitor periods exceeding 30 continuous days; explainLcauses of inope~ability and *actions taken to p:i;event reoccurrence.

d. Submit the ARERR [Ref. 5,2.lx] pri9r to May l51 of each year and .

include an assessment of radiation dosei; to the likely most exposed member of t,he publi~ from reactor releases and other nearby uranium fuel cycle sourc~s (including doses from primary eftluertt pathways and direct radiation) for the previous 12 consecutive months to s.how 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.l.

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

• Volume (cubic meters),

• Total curie quantity (specify whether detennined 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).

Information I I Rev. 25 I Page 45 of 89 OFF-SITE DOSE CALCULATION MANUAL

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 CPR 141, National Primary Drinking Water Regulations. Unit specific values are site values divided by two.

i. Include in the ARERR groundwater sample results taken that are in support of the Groundwater Protection Initiative (GPI) but are not part of the REMP.

3.8 10 CFR 50.75 (g) Jinplementation

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 item for 10 CPR 50. 75(g) 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 ~s procedure for potential impact on other related Department Procedures.

Information I PMP-6010-0SD-001 I Rev. 25 I Page 46 of 89 OFF-SITE DOSE CALCULATION MANUAL 3.9.5 Review this procedure during the first quarter of each year and update it if necessary. Review Attachment 3.16, IO Year Average *of 1995-2004 Data, and document using Attachment 3.17, Annual Evaluation of x/Qand D/Qvalues For All Sectors. The x IQ and DI 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:

5 .1.1 "Implementation of Programmatic Controls for Radiological Effluent Technical Sp~ifications 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 (ARBOR) Preparation And Submittal 5.1.4 PMP-6090-PCP-100, Spill Response- Oil, Polluting, Hazardous Materials, and Radioactjve Spills 5.2 Writing

References:

5.2.1 Source

References:

a. 10 CFR 20, *Standards for Protection Against Radtation

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

c. PMI-6010, Radiation Protection Plan

d. NUREG-0472

e. NUREG-1301

f. NUREG-0133

Information I PMP-6010-0SD-001 I Rev. 25 I Page 47 of 89 OFF-SITE DOSE CALCULATION MANUAL

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

h. Regulatory Guide 1.111

i. Regulatory Guide 1.113

j. Updated Final Safety Analysis Report (UFSAR)

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

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

m. NUREG-0017

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

ENGR 107-04 8112.1 Environs Rad Monitor System

o. HPPOS-223, Consideration of Measurement Uncertainty When Measuring Radiation Levels Approaching Regulatory Limits

p. 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

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

r. 40 CPR 190, Environmental Radiation Protection Standards for Nuclear Power Operations

s. NRC Commitment 6309 (N94083 dated 11/10/94)

t. NRC Commitment 1151

u. NRC Commitment 1217

v. NRC Commitment 3240

w. NRC Commitment 3850

x. NRC Commitmept 4859

y. NRC Commitment 6442

z. NRC Commitment 3768 aa. DIT-B-00277-00, HVAC Systems Design Flows bb. Regulatory Guide 1.21 cc. Regulatory Guide 4.1

Information I PMP-6010-0SD-001 I Rev. 25 I Page 48 of 89 OFF-SITE DOSE CALCULATION MANUAL dd. 1-2-V3-02-Calc #4, Unit Vent Sample Flow rate for isokinetic particulates and Iodine sampling ee. RPS Nl3.30-1996, Appendix. A Rationale for Methods of Determining Minimum Detectable Amount (MDA) and Minimum Testing Level (MDL ff. DIT-B-01971-00, Dose Factors for Radioactive Particulate Gaseous Effluents Associated with the Child by the Inhalation Pathway gg. DIT-B-01987-00, Ground Plane & Food Dose Factors Pi for Radioiodines and Radioactive Particulate Gaseous Effluents bh. NRC Commitment 1010 ii. NEI 07-07 Groundwater Protection Initiative jj. ANI 07-01 Potential for Unmonitored and Unplanned Off-Site Releases of Radioactive Material 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 eves Hold-Up Tank Directly to Unit Vent, May 1, 1992

• c. AEP Design Information Transmittal on Aux Building Ventilation Systems

d. PMP-4030.EIS.001, Event-Initiated Surveillance Testing

e. Environmental Position Paper, Fe Impact on Release Rates, approved 3/14/00

f. Environmental Position Paper, Methodology Change from Sampling Secondary System Gaseous Effluents for Power Changes Exceeding 15% within 1 hr to Responding to Gaseous Alert Alarms, approved 4/4/00

g. CR 02150078, RRS-1000 efficiency curve usage

b. Environmental Position Paper, Unit Vent Compensatory Sampling, approved 4/14/05

lnfQnnation PMP-6010-0SD-001 I Rev. 25 Page 49 of 89 OFF-SITE DOSE CALCULATION MANUAL Pages:

Attachment 3 .1 Dos~ Factors for Various Pathways 49-52 R.i Dose Factors PATHWAY Nuclide Ground Vegetable ).\feat CowJllilk Goat Milk Inhalation H-3 O.OE+OO 4.0E+o3 3.3E+02 2.4E+03 4.9E+03 l.3E+o3 C-14 O.OE+OO 3.5E+06 5.3E+05 3.2E+06 3.2E+o6 3.6E+o4 Cr-51 5.4E+06 l.1E+Q7 l.5E+06 6.9E+06 8.3E+o5 2.1E+o4 Mn-54 1.6E+09 9.4E+08 2.1E+07 2.9E+07 .*3.5E+06 2.0E+06 Fe-59 t2E+08 9.6E+o8 l.7E+09 3.1E+08 4.0E+o7 l.5E+o6 Co-58 4.4E+o8 6.0E+o8 2.9E+08 8.4E+07 l.OE+o7 1.3E+06 Co~60 2.5E+I0 3,2E+o9 1.dE+09 2.7E+o8 3.2E+o7 8.6E+o6 Zn-65 8.5E+08 2.7E+o9 95E+08 L6E+10 l.9E+o9 1.2E+d6 St-89 2,5E+04 3.5E+10 3.8E+o8 9.9E+09 2.lE+lO 2.4E+o6 Sr-90 .O.OE+oO l.4E+l2 9.6E+09 9.4E+10 2.0E+ll 1.1E+o8 Zr-95 2.9E+08* l.2E+o9 l.5E+09 9.3E+o5 l.1E+o5 2.7E+o6 Sb-124

• 6.9E+08 3.0E+09 4.4E+08 7.2E+o8 8.6E+o7 3.8E+o6 I-131 l.OE+o7 2.4E+10 2.5E+09 4.8E+I1 5.8E+ll 1.6E+o7 I-133 1.5E+o6 4.0E+o8 6.0E+Ol 4.4E+o9 :S.3E+09 3..8E+06 Cs~134

• 7.9E+o9 2.5E+10 l.1E+09 5.0E+lO 1.5E+ll 1.1E+06

• • • ,,.:::,., - . .. Cs-136 1.7E+08 Z..2E+o8 4.2E+07 5.1E+09 l.5E+I0 l.9E:+05 Cs-137 l.2E+10 2,5E+10

• l.OE+09 4.5E+I0 l.4E+ll 9.0E+05 Ba-140 2.3E+07 2.7E+o8 5.2E+07 2.1E+08 2.6E+o7. .. 2~0E+o6 Ce-141 1.5E+07 5.3E+o8. 3.0E+07 8.3E+07 l.OE+o7 6.1E+o5 Ce-144 7.9E+07 l.3E+10 3.6E+o8 7.3B+08 8.7E+o7 1.3E+07

• Uxµts for all ex<;:ept inhalation pathway are ~ 2 mr sec / yr µCi, .inhajatio~ pathway units are mr m3 l yr µCi.

Uap Values to be Used For the Maximum Exposed fudividual 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 1.6 21 Drinking water (L/yr) 330 510 510 730 Shoreline recreation (hr/yr) -- 14 67 12 3

Inhalation (tn. /yr) 1400 3700 8000 8000 Table E-5 of Reg. Guide 1.109.

Information PMP-6010-0SD-001 I Rev. 25 Page 50 of 89 OFF-SITE DOSE CALCULATION MANUAL Pages: .1 Dose Factors for Various Pathways 49-52 Bip Factors for Aquatic Foods pCil I kg pCi Element Fish Invertebrate H 9.0E-1 9.0E-1 C 4.6E3 9.1E3 Na l.OE2 2.0E2 p l.OE5 2.0E4 Cr 2.0E2 2.0E3 Mn 4.0E2 9.0E4 Fe l.OE2 3.2E3 Co 5.0El 2.0E2 Ni l.OE2 l.OE2 Cu 5.0El 4.0E2 Zn 2.0E3 l.OE4 Br 4.2E2 3.3E2 Rb 2.0E3 l.OE3 Sr 3.0El l.OE2 y 2.5El 1.0E3 Zr 3.3EO 6.7EO Nb 3.0E4 1.0E2 Mo 1.0El l.OEl Tc 1.5El 5.0EO Ru l.OEl 3.0E2 Rh l.OEl 3.0E2 Te 4.0E2 6.1E3 I l.5El 5.0EO Cs 2.0E3 l.OE3 Ba 4.0EO 2.0E2 La 2.5El l.OE3 Ce l.OEO l.OE3 Pr 2.5El 1.0E3 Nd 2.5El 1.0E3 w l.2E3 l.OEl Np 1.0El 4.0E2 Table A-1 ofReg.*Guide l.109.

Information PMP-6010-0SD-001 I Rev. 25 Page 51 of 89 OFF-SITE DOSE CALCULATION MANUAL Pages: .1 Dose Factors for Various Pathways 49-52 Drupj External Dose Factors for Standing on Contaminated Ground mrem m2 / hr pCi Radionuclide Total Body Skin H-3 0 0 C-14 0 0 Na-24 2.5E-8 2.9E-8 P-32 0 0 Cr-51 2.2E-10 2.6E-10 Mn-54 5.8E-9 6.8E-9 Mn-56 l.lE-8 1.3E-8 Fe-55 0 0 Fe-59 8.0E-9 9.4E-9 Co-58 7.0E-9 8.2E-9 Co-60 1.7E-8 2.0E-8 Ni-63 0 0 Ni-65 3.7E-9 4.3E-9 Cu-64 l.5E-9 l.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 l.4E-8 Br-85 0 0 Rb-86 6.3E-10 7.2E-10 Rb-88 3.SE-9 4.0E-9 Rb-89 l.5E-8 l.8E-8 Sr-89 5.6E-13 6.SE-13 Sr-91 7.lE-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 l.6E-9 l.9E-9 Y-93 5.7E-10 7.8E-10 Zr-95 5.0E-9 5.8E-9 Zr-97 5.SE-9 6.4E-9 Nb-95 5.lE-9 6.0E-9 Mo-99 l.9E-9 2.2E-9 Tc-99m 9.6E-10 l.lE-9 Tc-101 2.7E-9 3.0E-9 Ru-103 3.6E-9 4.2E-9 Ru-105 4.SE-9 *5.lE-9 Ru-106 l.SE-9 l.8E-9 Ag-llOm 1.8E-8 2.lE-8 Te-125m 3.SE-11 4.8E-ll

Information PMP-6010-0SD-001 I Rev. 25 Page 52 of 89 OFF-SITE DOSE CALCULATION MANUAL Pages:

I Attachment 3 .1 Dose Factors for Various Pathways 49-52 Radionuclide Total Body Skin Te-127m l.lE-12 l.3E-12 Te-127 1.0E-11 l.lE-11 Te-129ni 7.7E-10 9.0E-10 Te-129 7.lE-10 8.4E-10 Te-131m 8.4E-9 9 ..9E-9 Te-131 2.2E-9 2.6E-6 Te-132 l.7E-9 2.0E-9 1-130 1.4E-8 l.7E-8*

1-131 2.8E-9 3.4E-9 1-132 l.7E-8 2.0E-8 1-133 3.7E-9 4.SE-9 I-134 1.6E-8 l.9E-8 l-135 1.2E~8 l.4E-8 Cs-134 1.2E-8 1.4E-8 Cs-136 1.SE-8 l.7E-8 Cs-137 4.2E-9 4.9E-9 Cs-138 2.lE-8 2.4E-8 Ba-139 2.4E-9 2.7E-9 Ba-140 2.lE-9 2.4E-9 Ba-141 4.3E-9 4.9E-9 Ba-142 7.9E-9 9.0E-9 La-140 l.SE-8 1.7E-8 La-142 l.SE-8 l.8E-8 Ce-141 5.SE-10 6.2E-10 Ce-143 2.2E-9 2.SE-9 Ce-144 3.2E-10 3.7E-10 Pr-143 0 0 Pr-144 2.0E-10 2.3E-10 Nd-147 l.OE-9 1.2E-9 W-187 3.lE-9 3.6E-9 Np-239 9.SE-10 l.lE-9 Table E-6 of Reg. Guide 1.109.

Information PMP-6010-0SD-001 I Rev. 25 I Page 53 of 89 OFF-SITE DOSE CALCULATION MANUAL Pages:

Attachment 3.2 Radioactive Liquid Effluent Monitoring Instruments 53 - 55 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 3 System Effluent Line (R-20, R-28) train

3. Continuous Composite Sampler Flow Monitor

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

4. Flow Rate Measurement Devices

a. Liquid Radwaste Line (1) At times of release 4 (RFI-285)

b. Discharge Pipes* (1) At all times NA C. Steam Generator Blowdown (1) At times of release 4 Treatment Effluent (DFI-352)

d. Individual Stearn Generator sample flow ' (1) per At times of release 5 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 tenn 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 1 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 I 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.

Information PMP-6010-0SD-001 I Rev. 25 I Page 54 of 89 OFF-SITE DOSE CALCULATION MANUAL Pages:

Attachment 3 .2 Radioactive Liquid Effluent Monitoring Instruments 53- 55 A IF a.Ii RMS monitor is inoperable solely as the result of the loss of its control room alarm annunciation, TIIEN 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) a:t a limit of detection of at least 10-7 µCi/gram: *

1. At least once per shift when the specific activity of the secondary coolant is > 0.01 µCi/gram DOSE EQUIVALENT I-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 µCi/gram DOSE EQUIVALENT I-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 µCi/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 th(, 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.

Information PMP-6010-0SD-001 I Rev. 25 I Page 55 of 89 OFF-SITE DOSE CALCULATION MANUAL Pages:.

Attachment 3.2 Radioactive Liquid Effluent Monitoring Instruments 53 - 55 Action5 With the number of channels OPERABLE less than required by the Mmi.mum 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, TIIEN 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 des4"ed 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

Information Pl\IP-6010-0SD-001_ I Rev. 25 . Page 56 of 89 OFF-SITE DOSE CALCULATION MANUAL Radioactive Liquid Effluent Monitoring Pages:

Attachment 3)

Instrumentation Surveillance Requirements 56 -57 Instrum,ent CHANNEL SOURCE CHANNEL CHANNEL-CHECK CHECK CALJBRATION OPERATIONAL TEST

1. Gross Radioactivity Monitors Providing Automatic Release Termination

a. Liquid Radwaste D* p B(3) Q(S)

Efflm;nt Line (RRS-1001)

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

Blowdown Effluent Line C. Steam Generator *n* M B(3) Q(l)

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* NIA NIA NIA:.

Sump Effluent Line * ..

4. Flow Rate Measurement Devices

a. Liquid Radwaste D(4)* NIA B Q

-Effluent

b. Steam Generator D(4)* NIA NIA NIA Blowdown Treatment -

Line

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

lrtfonnation PMP-6010-0SD-001 l Rev. 25 Page57 of 89 OFF-SITE DOSE CALCULATION MANUAL Radioactive Liquid Effluent Monitoring Pages:

Attachment 3 .3 Instrumentation Surveillance Requirements 56.-- 57 TABLE NOTATION

1. Demonstrate with the CHANNEL OPERATIONAL TEST that automatic isolation of this pathway and control mom 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 downscii.le failure.*

4. Instrument control not set moperating mode.*

5. Loss of sample flow. *

2. Demonstrate with the CFJANNEL OPERATIONAL TEST that control room alarm annunciation occurs .if any of the following conditions exists: *

1. Instrument indicates measured levels above the a1ann setpoint.

2. Circuit failure.

3. Instrument indicates a downscale failure.

4. Instrumen,t controls not set in operating JIJ.Ode.

3. Perform the initial CHANNEL CALIBRATION using*one or more sourc~ with tr;aceability 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 m:;iy be used.

-4. Verify indication of flow during periods *of release with the CHANNEL CHECK. P~rform the CHANNEL CHECK at least once per :24 hours on days on which continuous, periodic *or batch releases are made.

5. De~onstrate with the CHANNEL OPERATIONAL TEST that automatic isolation of this p,athway 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 inc:lica,tes a downscale failure.**

4; Instrument control not.set in opera.ting mode.*

5. Loss ofs~ple flow,

• Instrument indicates, but does not provide for automatic isolation

• • Instrum(?nt indicates, but does not necessarily cause.automatic isolation. No credit is taken for the automatic isolatiop. on such occµrrences.

Operations currently performs the routine ch3Illlel checks and source checkS. Maintenance and Radiation Protection perform channel calibrations aJ,J.d channel operational tests. Chemistry performs the channel check on the continuous. composite sampler, Th~ responsibilities are subject to change without revision to this document.

Information PMP-6010-0SD-001 I Rev. 25 Page 58 of 89 OFF-SITE DOSE CALCULATION MANUAL Pages:

Attachment 3.4 Radioactive Gaseous Effluent Monitoring Instrumentation 58 -60 Instrument (Instrument #) Operable1 Minimum Action Channels Action

1. Condenser Evacuation System

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

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

OR (SFR-402 and l/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 and 1/2-MR-054) OR (VFR-315 and VFR-1510/2510)

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 (I) ****2,3 7 ERS-1305/1405 (ERS-2305/2405)

b. Containment Particulate Sampler Filter (I) **** 10 ERS-1301/1401 (ERS-2301/2401)

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

a. Noble Gas Activity (1) ****4 9 Alarm and Termination of Waste Gas Releases (VRS-1505/2505)

5. Gland Seal Exhaust

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

b. Flow Rate Monitor (SFR-201 and 1/2-MR- (1) **** 5

54) OR (SFR-201 and 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 associatea'with it like that associated with the two specific tank types listed here.

• • • • During releases via this pathway

. Information PMP-6010-0SD-001 I Rev. 25 Page 59 of 89

, OFF-SITE DOSE CALCULATION MANUAL Pages: -4 Radioactive Gaseous Effluent Moiµtoring Instrumentation

.58 - 60 TABLE NOTATIONS

1. IF an RMS monitor is inoperable solely as the result of the loss ofit's control room alarni annunciation, THEN one of the following actions is acceptable to satisfy the ODCM action statement compensatory surveillance reql).irement:

• l_ 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 ann.unciation,, THEN the only acceptable ac;tion is taking grab samples and conducting laboratory* ~yse~ as the reading is equivalent to a grab sample when the monitor is functional.

2. Consider releases as o~curring "via this pathway" under the following conditions:

• The Containment Purge System is in operation and Contawment Operability is applicable,

-OR- .

• The Containment Purge System is .in operation and the 'Clean-up' batch release of the Containment air volume has not been fully completed.

IF neither of the above are applicable AND .the unit is in Mode 5 or 6, THEN the contmnment purge system is acting as a ventilation system (an extension of the Auxiliary Bgildmg) and is covered by Item 2 of tliis Attachment. This is called 'Ventilation Mode'. 'Ventilate Mode' cannot be entered witho11t performing a Clean-up batch release.

-OR-

• A Containment Pressure Relief (CPR) is being performed.

Once the 'Clean-up' batch release has been completed and 'Ventilation' mode of Purge has commenced -

resultant return to 'Clean-up' mode can be made with no additional saµiplirig requirements or paperwor~ - so long as eit,her ERS-1305/2305 OR ERS~1405/2405 are operable. Containment particulate channels are not needed once the RCS has entered Mode 5 per Technical Specification 3.4.15.

3. For purge (including pressure relief) pm:poses only. Refe:r;ence T.S 3 .3.6, Contai.nment Purge Supply and Exhaust System Isolation Instrumentation and 3:4.15, RCS Leakage Detection lnstrQmentatlon for additional information. *

4. For waste gas releases only, see Item 2 (Unit Vent; Auxiliary Building Ventilation System) for !!-dditional 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 hours and provide a description of why the inoperability was not corrected in the next Annual Radiological Effluent Release Reporj:.

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.

Information

• PMP-6010-0SD-001 I Rev. 25 Pa2e 60 of 89 OFF-SITE DOSE CALCULATION MANUAL Pages:

Attachment 3 .4 Radioactive Gaseous Effluent Monitoring Instrumentation

. 58 - 60

7. With the number of channels OPERABLE less than required by the Minimun:i Channels OPERABLE requirements, imme(iiately suspend PURGING or VENTING (CPR) of radioactive ~ffluents via this pathway.

8. Willi. the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via the affected pathway may, continue f9r up to 30 days* provided samples required for weekly lodme & Particulates analysis are continuously collected with auxiliary sampling equl.pment as required in Attachment 3.7, Radioactive Gaseous Waste Sampling and Analysis Program. After 30 days, IF tp.e channels are not OPERABLE, THEN continue releases with sample collection by auxiliary sampling equipment and provide a description of why .the inoperabilit:y was not corrected in the next Annual Radiological Effluent Release Report. ..

Sampling evolutions are llOt an inJerruption 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 con~nts are analyzed anci,

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

After 14 days, IF the channels are not OPERABLE, THEN cop.tinue releases with sample collection by auxiliary sampling equipment and provide a description of why the inoperabilit:y was not corrected in the next Annual Radiological E~rient Release Report.

iO, Technical Specification 3-4.15, RCS Leakage Detection System Instrumentation.

Compensatory actions are.governed by PMP-4030-EIS-OOl, Event-lnitiat:(l Surveillance Testing.

Information PMP-6010-0SD-001 I Rev. 25 Page 61 of 89 OFF-SITE DOSE CALCULATION MANUAL Radioactive Gaseous Effluent Monitoring Pages:

Attachment 3.5 Instrumentation Surveillance Requirements 61 - 62 Instrument CHANNEL SOURCE I CHANNEL CHANNEL OPERATIONAL CHECK CHECK CALIBRATION TEST

1. Condenser Evacuation Alarm.Only System

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

(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(l)

(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* NIA 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 eves 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(l)

(SRA-1805/2805)

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

• At all times

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

Information PMP-6010-0SD-001 I Rev. 25 Page 62 .of 89 OFF-SITE DOSE CALCULATION MANUAL Radioactive Gaseous Effluent Monitoring Pages:

Attachment 3.5 Instrumentation Surveillance Requirements 61 -62 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.

Information PMP-6010-0SD-001 I Rev. 25 I Page 63 of89 OFF-SITE DOSE CALCULATION MANUAL Pages:

Attachment 3.6 Radioactive Liquid Waste Sampling and Analysis Program 63 -64

[Ref. 5.2.ltJ LIQUID SAMPLING MINIMUM TYPE OF LOWER RELEASE FREQUENCY ANALYSIS ACTIVITY LIMITOF TYPE FREQUENCY ANALYSIS DETECTION (LLD)

(µCi/ml) a A. Batch Waste p p Principal 5x10*1 Release Tanks c Each Batch Each Batch Gamma Emitters e I-131 lxl0*6 p p Dissolved and

- Entrained Gases Each Batch Each Batch (Gamma lx10*5 Emitters) p M lx1Q*5 H-3 Each Batch Compositeb Gross Alpha lx10*7 p Q Sr-89, Sr-90 5xl0*8 Each Batch Compositeb Fe-55 lxl0*6 B. Plant w Principal Continuous Daily Compositeb Gamma 5x10*7

Releases* d Emitterse I-131 lxl0*6 M M Dissolved and Grab Sample Entrained Gases 1x10*5 (Gamma Emitters)

M H-3 1x10*5 Daily Compositeb Gross Alpha 1x10*1

• Q Sr-89, Sr-90 5xl0*8 Daily Compositeb Fe-55 lxl0*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 Taruc tritium samples.

Information PMP-6010-0SD-001 I Rev. 25 I Page 64 of 89 OFF-SITE DOSE CALCULATION MANUAL Pages: .6 Radioactive Liquid Waste Sampling and Analysis Program 63 - 6A TABLE NOTATION

a. The lower limit of detection (LLD) is defined in Table Notation A. of Attachment 3.20, Maximum Values for Lower Limits ofDetectionsA,B _ 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 Tanlc 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.

Information PMP-6010-0SD-001 I Rev. 25 Page 65 of 89 OFF-SITE DOSE CALCULATION MANUAL R,adioactive Gaseous Waste Sampling and Pages:

Attachment 3.7 Analysis Program 65 -66 Gaseous Release Type Frequency Minimum Type of Lower Limit Analysis Activity of Detection Frequency Analysis (µCi/cc) a

a. Waste Gas Storage p p Principal Gamma Tanks and CVCS HUTs Each Tank Each Tank Emittersd 1 X 104 Grab Sample H-3 1 X 10-6

b. Containment Purge p p Principal Gamma Each Purge Each Purge Emitters d 1 X 104 Grab Sample CPR (vent)** Twice per Twice per Month Month H-3 1 X 10-6

c. Condenser Evacuation WorM M Principal Gamma System Grab Sample Particulate Sample Emittersd 1 X 10"11 Gland Seal Exhaust* ; 1 X 10-6 M H-3 wg Principle Gamma 1 X 104 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 0 wb 1-131 Vent* Iodine Adsorbing 1 X 10"12 Media Continuous c wb. Principal Gamma Particulate Sample Emittersd 1 X 10-ll Continuous 0 M Gross Alpha Composite Particulate 1 X 10"11 Sample w wh H-3 Grab Sample H-3 Sample 1 X 10-6 wgj Principle Gamma 1 X 104 Noble Gas Emitters d Continuous c Q Sr-89, Sr-90 Composite Particulate l X 10"11 Sample Continuous c Noble Gas Monitor Noble Gases 1 X 10-6

e. Incinerated Oil 0 p p Principal Gamma Each Batchr Each Batchr Emittersd 5 X 10*7

• 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.

Information PMP-6010-0SD-001 I Rev. 25 Page 66 of 89 OFF-SITE DOSE CALCULATION MANUAL Radioactive Gaseous Waste Sampling and Pages:

Attachment 3.7 Analysis Program 65 -66 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 !131 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 Il31 levels have returned to within the factor-of 3 of the pre-event 'normal' .[Ref. 5.2.lz]

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 or momentary interruptions to maintain sampling capability 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 andXe-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 contajner 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 IE-6 µC/cc. Attach the RMS daily averages in lieu of sampling. This is based on operating experience indicating no activity is detected in the Gland Seal Exhaust below this value. Compensatory sampling for out of service monitor is still required in the event 1805/2805 is inoperable.

j. Sampling and analysis shall also be perfonned following shutdown, startup or THERMAL POWER change exceeding 15% of RATED THERMAL POWER within a one hour period. This noble gas sampl~ 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.

Information PMP-6010-0SD-001 I Rev. 25 Page 67 of 89 OFF-SITE DOSE CALCULATION MANUAL Page:

Attachment 3.8 Multiple Release Point Factors for Release Points*

67 Liquid Factors Monitor Description Monitor Number MRP#

U 1 SO Blowdown 1R19/24, DRS 3100/3200* 0.35 U 2 SO 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 not respond 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 Unit2 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 I monitors.

1. Nominal Values a Two release points of2,754 cfin 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.

Information PMP-6010-0SD-001 I Rev. 25 Page 68 of 89 OFF-SITE DOSE CALCULATION MANUAL Page:

Attachment 3.9 Liquid Effluent Release Systems 68 RB.E:ASE SOURCES SYSTBulS PONfS DirtyWJSles.:

AoorDralns,, StatlonDfnin(Dirty) CJeansump Decontarr.inatton Sump Tan!< Tank RinsoS0lutlon9,,

Chemical Drain Tanl(,Etc. pump I a. H Wiste_~~:;~ Tank Clean WIS.es Equipment Drains, PumpSeal Leakolfs,.

Containment Fan Circulatlng pwnp Caoler LaundryHotSho\lo'er I l":'.'u 1/ihter C<JndenS1te,ac. I T:mks(2) ~ Discharge C\CS Boricacld Evaporator j BoriccAcidBiaporation ~ &aporatorCondensate MonitcrTanks Packagcs,North t Oeminaralizer I and South Steam StcamGeneralot Generator Bfowdovmand Slowdown Treatment Sy!tom (Potential)

I SaeenHouse I Circulating

\JI.bier Discharge Es:Entla1Servicc

\11.bterSystem (Potential)

Isample Point ClrculatingW:ater ln!akePipes 1!::!!.!!!.E.I r:----, I h Containment Spray l ~ a d f a t l o o g:i-----,rri_ HeatExchongers. I CirctJlating Vihtcr Dia::harge Turbine Room SumpUnil1and2 TurbineRoomSump 1--,--t I ~1 AowMctcr 1-~----~--------+-tl (Potential)

B.571'Sump Circulating V1,Mer Di!Charge

Information PMP-6010-0SD-001 I Rev. 25 Page 69 of 89 OFF-SITE DOSE CALCULATION MANUAL Page:

Attachment 3 .10 Plant Liquid Effluent Parameters 69 SYSTEM COMPONENTS CAPACITY FLOW RATE TANKS I PUMPS (EACH) (EACH)*

I Waste Disposal System

+ Chemical Drain Tanlc 1 1 600 GAL. 20 GPM

+ Laundry & Hot Shower Tanlcs 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. 580 GPM

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

+ Blowdown Treatment System 1 60 GPM III Essential Service Water System

+ Water Pumps 4 10,000 GPM

+ Containment Spray Heat 4 3,300 GPM Exchanger Outlet IV Circulating Water Pumps I Unit I 3 230,000 GPM Unit2 4 230,000GPM

• Nominal Values

1. 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.

Information PMP-6010-0SD-001 I Rev. 25 I Page 70 of 89 OFF-SITE DOSE CALCULATION MANUAL Volumetric Detection Efficiencies for Principle Gamma Page: .11 Emitting Radionuclides for Eberline Liquid Monitors 70 This includes the following monitors: RRS-1000, DRS 3100, DRS 3200, DRS 4100, and DRS 4200.

[Ref. 5.2.lq)

NUCLIDE EFFICIENCY (cpm/uCi/cc) 1-131 3.78 B7 Cs-137 3.00B7 Cs-134 7.93B7 Co-60 5.75 B7 Co-58 4.58 B7 Cr-51 3.60 B6 Mn-54 3.30B7 Zn-65 1.58 B7 Ag-llOM 9.93 E7 Ba-133 4.85 B7 Ba-140 l.92B7 Cd-109 9.58 E5 Ce-139 3.28 E7 Ce-141 1.92 B8 Ce-144 4.83 E6 Co-57 3.80E7 Cs-136 1.07B8 Fe-59 2.83 E7 Sb-124 5.93 E7 1-133 3.40 E7 I-134 7.23 E7 I-135 3.95 B7 Mo-99 8.68 E6 Na-24 4.45 E7 Nb-95 3.28 E7 Nb-97 3.50E7 Rb-89 5.00E7 Ru-103 3.48 E7 Ru-106 1.23 E7 Sb:-122 2.55 E7 Sb-125 3.15 E7 Sn-113 7.33 ES Sr-85 3.70E7 Sr-89 2.88 E3 Sr-92 3.67E7 Tc-99M 3.60 E7 Y-88 5.25 E7 Zr-95 3.38 E7 Zr-97 3.10E7 Kr-85 1.56 ES Kr-85M 3.53 E7 Kr-88 4.10E7 Xe-131M 8.15 E5 Xe-133 7.78E6 Xe-133M 5.75 E6 Xe-135 3.83E7

Information PMP-6010-0SD-001 I Rev. 25 Page 71 of 89 OFF-SITE DOSE CALCULATION MANUAL Pages: .12 Counting Efficiency Curves for R-19, and R-24 71 -72 Counting Efficiency Cunre for R-19 Efficiency Factor= 4.2 E6 cpm/uCi/ml (Based on empirical data taken during pre-<Jperational tcs ting with Cs-137)

'C C:

e J2 1.00E+04 +----------------------:;;,,'""'------------------!

il"

~

,g 1.00E+03 +-----------------'--------------------------4

a; n.

tJ rD 0 0 0 LL! +

0 UJ 0

~ ~

rnicrocuries/ml

Information PMP-6010-0SD-001 I Rev. 25 Page 72 of 89 OFF-SITE DOSE CALCULATION MANUAL Pages: .12 Counting Efficiency Curves for R-19, and R-24 71- 72 Counting Efficiency Curve for R-24 Efficiency Factor= 7.SE6 cpm/uCi/ml (Based oli empirical data taken during pre-operational testing with Mn-54)

"t, 1.00E+05 +--------------------------==-.-,::::.--------------1 C

e f 1.00E+04 -l:---------------------=_,,,.,.~-----------------1 B

Cl)

,8 1.00E+03 +-------------.....""'----------------------------l

ii Q.

(.) 1.00E+02 -l,---------=_,,,.,.~------------------------------1 CD It) v 0

(')

0 ~ 0 0 0 0 0 w

0 w

0 w

0 w

0 w

0 w

0 it0 q q q

_q

.... q

....q ....q mlcrocurles/ml

Information PMP-6010-0SD-001 I Rev. 25 I Page 73 of 89 OFF-SITE DOSE CALCULATION MANUAL Page: .13 Counting Efficiency Curve for R-20, and R-28 73 Counting Efficiency Curve for R-20 and R-28 Efficiency Factor= 4.3 E6 cpm/uCi/ml (Based on empirical data taken during pre"{)pcrational testing with Co-58)

N 9

UJ 9

UJ sl UJ 9

UJ ~

0 0 0 0 0 q q q q

~ ~

mlcrocurles/ml

Information PMP-6010-0SD-001 I Rev. 25 Page 74 of 89 OFF-SITE DOSE CALCULATION MANUAL

• Page:

Attachment 3 .14 Gaseous Effluent Release Systems 74 RB...EASE SOURCES SYSTBv1S R'.JINTS WltteGasDecay jvent Header

~a;s and C\.eS r-.

Moisture Se rator AU"t.. Building Vent ,---::--C'.'C'--, __ ..I Pm r*

Ventilation Filler Engineered Sar ety Feat uresVent ~

System Engineered ~ Pro HEPA >

SafetyFcaturcs VenlandPipo Filter Filter 1§ Enclosures s

J 0

I-E

'Li~~ ~  :>--------------

Fuel Hand[ing Ventilation 2 Q)

"" =: 1 Damper I I a::

Containment Purge and Relier sys1:em

. /~

arbon ~-----------~

Filter~

c::;;;;;;:J lnstrumerit Room Purge System HEPA CarbOn

~----~ ~

Filter Filter 1-.....J~,;---~~---;::=========:;-======~ Alrbornelo,o,flfConlnlnmenlRftdlation UpperConlaln1J1CJl\Are:i 1Jonl1or11, Theselr.ola\ete!\Salnnrn.t

~----~

~---------------='  :::S::om':p:::Ji=:n::gp::o::in:::l~L-H lnc1~:~~1::.: ~: n; :~~:: :dhl~h ahirm.

RadlD1lonND1lllor. Thi&

lscleteacon1alnm:.?nl puroe.ctnt.rdl"l,nnd lns11wnentrOG111

~-----------------------...'.:=========::....jo.du1v11l~o11hlch!l'J11r,n,,

Steam Generator Slowdown

~

~

. ~--------------------------------~

~

Trcatn1ent System ToTreatment Sy.stem no gaseousrelease Condenser Air EjectorSySl:enl I FlowData I r ------------, I Samo1incPolnt I _~

5

_'-'~"':"'c:c :c :s..Al-r__,l~~-1'~-*---------*--------------------,Q Gland Seal candcnser Exhaust ISS"-tea;;;;;:m;;;P;;;a;;;ck;;:in;;;g;l----------- ~

Eiehauster __

I semplloqPoJnt l~c,nl I~--r~-----~-,-----~

AowData I R a d i a t i o n ~ - - - - - - - - - - - - - - - - < atmo d~:. hero

Information PMP-6010-0SD-001 I Rev. 25 Page 75 of 89 OFF-SITE DOSE CALCULATION MANUAL Page:

Attachment 3.15 Plant Gaseous Effluent Parameters 75 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 FT3 @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 Umt NORMAL STEAM JET AIR 1&2 230 EJECTORS

• START UP STEAM JET AIR 1&2 3,600 EJECTORS Ill TURBINE SEALS SYSTEM 1 1,260 2 5,508 2 Release Points for Unit 2 START UP FLASH TANK VENT 1 1,536 2 1,536

+ Designates total flow for all fans.

Information PMP-6010-0SD-001 I Rev. 25 Page 76 of 89 OFF-SITE DOSE CALCULATION MANUAL Attachment 3.16 10 Year Average of 1995-2004 Data Pages: 76-77 x/Q GROUND AVERAGE (sec/m3)

DIRECTION DISTANCE (METERS)

(WINDFROM) 594 2416 4020 5630 7240 N 4.17E-06 4.82E-07 2.25E-07 l.33E-07 9.32E-08 N_NE 3.02E-06 3.64E-07 l.73E-07 l.04E-07 7.29E-08 NE 4.54E-06 5.31E-07 2.60E-07 l.59E-07 1.13E-07 ENE 7.16E-06 7.99E-07 4.04E-07 2.52E-07 l.80E-07 E l.04E-05 l.13E-06 5.82E-07 3.66E-07 2.63E-07 ESE 1.07E-05 l.18E-06 6.04E-07 3.78E-Q7 2.72E-07 SE 1.15E-05 l.24E-06 6.36E-07 4.00E-07 2.88E-07 SSE 1.3DE-05 l.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 l.72E-07 SW 4.12E-06 4.73E-07 2.28E-07 l.38E-07 9.73E-08 WSW 3.29E-06 3.65E-07 l.76E-07 1.06E-07 7.52E-08 w 3.63E-06 4.llE-07 1.96E-07 l.18E-07 8.31E-08 WNW 3.02E-06 3.43E-07 1.61E-07 9.59E-08 6.71E-08 NW 3.22E-06 3.61E-07 1.71E-07 1.02E-07 7 .. 16E-08 NNW 3.84E-06 4.29E-07 2.02E-07 l.20E-07 8.40E-08 DIRECTION DISTANCE (METERS)

(WINDFROM) 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 l.15E-08 7.41E-09 4.72E-09 ENE 9.30E-08 3.80E-08 l.91E-08 l.23E-08 7.90E-09 E l.37E-D7 5.65E-08 2.85E-08 1.83E-08 l.18E-08 ESE l.41B-07 5.81E-08 2.93E-08 1.88E-08 l.22E-08 SE l.50E-07 6.2DE-08 3.12E-08 2.0lE-08 l.30E-08 SSE l.71E-07 7.06E-08 3.56E-08 2.29E-08 l.48E-08 s 1.84E-07 7.49E-08 3.77E-08 2.43E-08 l.56E-08 SSW 8.86E-08 3.59E-08 l.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 l.51E-08 7 ..53E-09 4.83E-09 3.07E-09 w 4.17E-08 l.64E-08 8.13E-09 5.20E-09 3.28E-09 WNW 3.34E-08 l.29E-08 6.41E-09 4.lOE-09 2.57E-09 NW 3.57E-08 l.39E-08 6.89E-09 4.41E-09 2.77E-09 NNW 4.19E-08 3.35E-08 8.lOE-09 5.19E-09 3.27E-09 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 x/Q = 2.04E-05 sec/m3 in Sector H 2004

Information PMP-6010-0SD-001 I Rev. 25 Page 77 of 89 OFF-SITE DOSE CALCULATION MANUAL Attachment 3 .16 10 Year Average of 1995-2004 Data Pa,ges: 76-77 D/Q DEPOSITION (1/m2)

DIRECTION DISTANCE (METERS)

(WINDFROM) 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 l.29E-08 1.25E-09 5.67E-10 2.97E-10 1.90E-10 ENE l.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 l.85E-08 1.79E-09 8.12E-10 4.26E-10 2.72E-10 SE l.90E-08 1.83E-09 8.30E-10 4.36E-10 2.78E-10 SSE 2.40E-08 2.32E-09 1.0SE-09 5.52E-10 3.52p-10 s 3.68E-08 3.56E-09 1.61E.-09 8.46E-10 5.40E-10 SSW 2.30E-08 2.22E-09 1.0lE-09 5.28E-10 3.37E-10 SW 2.22E-08 2.lSE-09 9.74E-10 5.llE-10 3.26E-10 WSW 2.llE-08 2.04E-09 9.23E-10 4.84E-10 3.09E-10 w 2.00E-08 l.93E-09 8.74E-10 4.59E-10 2.93E-10 WNW 1.75E-08 1.69E-09 7.64E-10 4.0lE-10 2.56E-10 NW 1.58E-08 l.53E-09 6.94E-10 3.64E-10 2.32E-10 NNW 2.30E-08 2.22E-09 l.OlE-09 5.28E-10 3.37E-10 DIRECTION DISTANCE (METERS)

(WIND FROM) 12067 24135 40225 56315 80500

.N l.45E-10 4.72E-ll l.74E-ll 9.27E-12 4.65E-12 NNE 6.36E-ll 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-ll l.17E-ll 6.23E-12 3.13E-12 E 1.14E-10. 3.73E-11 1.37E-11 7.34E-12 3.68E-12 ESE l.13E-10 3.70E-11 1.36E-ll 7.26E-12 3.64E-12 SE 1.16E-10 3.78E-11 1.39E-11 7.42E-12 3.72E-12 SSE l.47E-10 4.79E-ll 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 L41E-10 4.59E-11 1.69E-ll 9.0lE-12 4.52E-12 SW l.36E-10 4.43E-ll 1.63E-11 8.71E-12 4.. 37E-12 WSW 1.29E-10 4.20E-11 1.SSE-11 8.26E-12 4.14E-12 w l.22E-10 3.98E-11 1.47E-11 7.82E-12 3.92E-12 WNW 1.07E-lb 3.48E-ll 1.28E-11 6.84E-12 3.43E-12 NW 9.70E-11 3.16E-ll l.16E-11 6.20E-12 3.llE-12 NNW l.41E-10 4.58E-ll 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 l/m2 in Sector A 2001

Information PMP'-6010-0SD-001 I Rev. 2S Page 78 of 89 OFF-SITE DOSE CALCULATION MANUAL Attachment 3 .17 Annual Evaluation of .x/Qand b/Q Values For Page:

All Sect9rs 78

1. Performed or received annual update of x/Q and D/Q values. Pro~ide a description of what has been received.

I Signature Date Environmental Department (print name, title)*

2. Worst ;c/Q and b/Q value and se.ctor determined. PMP-6010-0SD-001 has been updaJed, if necessrlfY, Provide ill evaluation.

I 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.

I Signature Date Environmental Department (print name, title)

4. Approved and verified_ by:_ .

I Signature Date Environmental Department (print name, title)

Information PMP-6010-0SD-001 I Rev. 25 Page 79 of 89 OFF-SITE DOSE CALCULATION MANUAL Pages:

. Attachment 3.18 Dose Factors 79- 80 DOSE FACTORS FOR NOBLE GASES AND DAUGHTERS*

TOTAL BODY SKIN DOSE GAMMA AIR BETA AIR DOSE FACTOR FACTOR DOSE FACTOR DOSE FACTOR Ki (DFB;) Li (DFSi) M; (DF71) Ni (DFl\)

mremm3 (mrem m3 (mradm3 (mradm3 RADIONUCLIDE per µCi yr) per µCi yr) per µCi yr) per µCi yr)

Kr-83m 7.56E-02 - - - l.93E+Ol 2.88E+02 Kr-85m l.17E+03 l.46E+03 l.23E+03 1.97E+03 Kr-85 l.61E+Ol l.34E+03 l.72E+Ol 1.95E+03 Kr-87 5.92E+03 9.73E+03 6.17E+03 l.03E+04 Kr-88 1.47E+04 2.37E+03 l.52E+04 2.93E+03

- Kr-89 1.66E+04 l.01E+04 l.73E+04 l.06E+04 Kr-90 l.56E+04 7.29E+03 l.63E+04 7.83E+03 Xe-131m 9.15E+Ol 4.76E+02 l.56E+02 l.11E+03 Xe-133m 2.51E+02 9.94E+02 3.27E+02 - l.48E+03 Xe-133 2.94E+02 3.06E+02 3.53E+02 l.05E+03 Xe-135m 3.12E+03 7.11E+02 3.36E+03 7.39E+02 Xe-135 l.81E+03 l.86E+03 l.92E+03 2.46E+03 Xe-137 l.42E+03 l.22E+04 l.51E+03 l.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 Reg. Guide 1.109, Table B-1.

Information PMP-6010-0SD-001 I Rev. 25 Page 80 of 89 OFF-SITE DOSE CALCULATION MANUAL Pages: .18 Dose Factors 79- 80 DOSE FACTORS FOR RADIOIODINES AND RADIOACTIVE PARTICULATE,

. JN GASEOUS EFFLUENTS FOR CIDLD* Ref. 5.2.leeandff P1 P1 INHALATION FOOD & GROUND PATHWAY PATHWAY (mremm3 (mremm2 sec RADIONUCLIDE per µCi yr) per µCi yr)

H-3 l.12E+03 l.57E+03 u P-32 2.60E+06 7.76E+10 Cr-51 l.70E+04 l.20E+07 Mn-54 l.58E+06 l.12E+09 Fe-59 l.27E+06 5.92E+08 Co-58 l.11E+06 5.97E+08 Co-60 7.07E+06 4.63E+09 Zn-65 9.95E+05 l.17E+10 Rb-86 l.98E+05 8.78E+09 Sr-89 2.16E+06 6.62E+09 Sr-90 l.01E+08 l.12E+ll 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 l.55E+08 Ru-106 l.43E+07 3.01E+08 Ag-llOrn 5.48E+06 l.99E+10 I-131 l.62E+07 4.34E+ll I-132 l.94E+05 l.78E+06 I-133 3.85E+06 3.95E+09 I-135 7.92E+05 l.22E+07 Cs-134 l.01E+06 4.00E+lO Cs-136 1.71E+05 3.00E+09 Cs-137 9.07E+05 3.34E+10 Ba-140 l.74E+D6 1.46E+08 Ce-141 5.44E+05 3.31E+07 Ce-144 l.20E+07 1.91E+08

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

1. The u~ts for both H3 factors are the same, mrem m' per µCi yr

Information PMP-6010-0SD-001 I Rev. 25 Page 81 of 89 OFF-SITE DOSE CALCULATION MANUAL Radiological Environmental Monitoring Program Pages:

Attachment 3.19 Sample Stations, Sample Types, Sample Frequencies 81- 84 e . . . w, 521

[Rf521 .u

.IV, 521]

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@ 18° from Plant Axis Airborne Particulate Weekly Gross Beta Weekly Gamma Isotopic Quart. Comp.

Airborne I-131 Weekly Radioiodine TLD Quarterly Direct Radiation Quarterly ONS-2 (T-2) 2338 ft@ 48° from Plant Axis Airborne Particulate Weekly Gross Beta Weekly Gamma Isotopic Quart. Comp.

Airborne I-131 Weekly Radioiodine TLD Quarterlv Direct Radiation Quarterly ONS-3 (T-3) 2407 ft @ 90° from Plant Axis Airborne Particulate Weekly Gross Beta Weekly Gamma Isotopic Quart. Comp.

Airborne I-131 Weekly Radioiodine TLD Quarterly Direct Radiation Quarterly ONS-4 (T-4) 1852 ft. @ 118° from Plant Axis Airborne Particulate Weekly Gross Beta Weekly Gamma Isotopic Quart. Comp.

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

Airborne I-131 Weekly Radioiodine TLD Quarterly Direct Radiation Quarterly T-7 2103 ft @ 36° from Plant Axis TLD Quarterly Direct Radiation Quarterly T-8 2208 ft @ 82° 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@ 11° from Plant Axis TLD Quarterly Direct Radiation Quarterly T-12 2292 ft@ 63° 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 N~w Buffalo, MI Gamma Isotopic Quart. Comp.

Airborne Radioiodine I-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 I-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 I-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 I-131 Weekly TLD Quarterly Direct Radiation Quarterly L

Information PMP-6010-0SD-001 I Rev. 25 Page 82 of 89 OFF-SITE DOSE CALCULATION MANUAL Radiological Environmental Monitoring Program Pages:

Attachment 3.19 Sample Stations, Sample Types, Sample Frequencies 81- 84 SAMPLE DESCRIPTION/ SAMPLE SAMPLE ANALYSIS ANALYSIS STATION LOCATION TYPE FREQUENCY TYPE FREQUENCY OFF-SITE DIRECT RADIATION (TLD) STATIONS OFf-1 4.5 miles NE, Pole #B294-44 TLD Quarterly Direct Radiation Quarterlv OFr-2 3.6 miles, NE, Stevensville TLD Quarterly Dire ct Radiation Quarterly Substation OFr-3 5.1 miles NE, Pole #B296-13 TLD Quarterly Direct Radiation Quarterly OFr-4 4.1 miles, E, Pole #B350-72 TLD Quarterly Direct Radiation Quarterly OFr-5 4.2 miles ESE, Pole #B387-32 TLD Quarterly Direct Radiation Quarterlv OFr-6 4.9 miles SE, Pole #B426-l TLD Quarterly Direct Radiation Quarterly OFf-7 2.5 miles S, Bridgman Substation TLD Quarterly Direct Radiation Quarterly OFr-8 4.0 miles S, Pole #B424-20 TLD Quarterly Direct Radiation Quarterly OFf-9 4.4 miles ESE, Pole #B369-214 TLD Quarterly Direct Radiation Quarterly OFr-10 3.8 miles S, Pole #B422-99 TLD Quarterly Direct Radiation Quarterlv OFr-11 3.8 miles S, Pole #B423-12 TLD Quarterly Direct Radiation Quarterly GROUNDWATER (WELL WATER) SAMPLE STATIONS W-1 1969 ft@ 11° from Plant Axis Groundwater Quarterly Gamma Isotopic Quarterly Tritium Quarterly W-2 2302 ft@ 63° from Plant Axis Groundwater Quarterly Gamma Isotopic Quarterly Tritium Quarterly W-3 3279 ft @ 107° from Plant Axis Groundwater Quarterly Gamma Isotopic Quarterly Tritium Quarterlv W-4 418 ft@ 301 ° 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@ 273° from Plant Axis Groundwater Quarterly Gamma Isotopic Quarterly Tritium Quarterly W-7 1895 ft @ 189° from Plant Axis Groundwater Quarterly Gamma Isotopic Quarterly Tritium Quarterlv W-8 1274 ft@ 54° from Plant Axis Groundwater Quarterly Gamma Isotopic Quarterly Tritium Quarterly W-9 1447 ft@ 22° from Plant Axis Groundwater Quarterly Gamma Isotopic Quarterly Tritium Quarterly W-10 4216 ft@ 129° from Plant Axis Groundwater Quarterly Gamma Isotopic Quarterly Tritium Quarterly W-11 3206 ft @ 153° from Plant Axis Groundwater Quarterly Gamma Isotopic Quarterly Tritium Quarterly W-12 2631 ft@ 162° from Plant Axis Groundwater Quarterly Gamma Isotopic Quarterly Tritium Quarterlv W-13 2152 ft@ 182° from Plant Axis Groundwater Quarterly Gamma Isotopic Quarterly Tritium Quarterlv W-14 1780 ft@ 164 ° from Plant Axis Groundwater Quarterly Gamma Isotopic Quarterly Tritium Quarterly W-15 725 ft @ 202° from Plant Axis Groundwater Quarterly Gamma Isotopic Quarterly NPDES well MW-12C Tritium Quarterly W-16 2200 ft @ 208° from Plant Axis Groundwater Quarterly Gamma Isotopic Quarterly NPDES well MW-20 Tritium Quarterly W-17 2200 ft @ 180° from Plant Axis Groundwater Quarterly Gamma Isotopic Quarterly NPDES well MW-21 Tritium Quarterly

Information PMP-6010-0SD-001 I Rev. 25 Page 83 of 89 OFF-SITE DOSE CALCULATION MANUAL Radiological Environmental Monitoring Program Pages:

Attachment 3.19 Sample Stations, Sample Types, Sample Frequencies 81- 84 DRINKING WATER STJ St. Joseph Public Intake Sta. Drinking water Once per calendar Gross Beta 14 day Comp.

9mi.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.

0.6mi. S Day Gamma Isotopic 14 dav Comp.

1-131 14dav Comp.

Tritium Quart.Como SAMPLE DESCRIPTION/ SAMPLE SAMPLE ANALYSIS ANALYSIS STATION LOCATION TYPE FREQUENCY TYPE FREQUENCY SURFACE WATER SWL-2 Plant Site Boundary - South Surface Water Once per calendar Gamma Isotopic Month. Comp.

- 500 ft. south of Plant Day Tritium Quart. Comp Centerline SWL-3 Plant Site Boundary - North Surface Water Once per calendar Gamma Isotooic Month. Comp.

- 500 ft. north of Plant Day Tritium Quart. Comp.

Centerline SEDIMENT SL-2 Plant Site Boundary - South Sediment Semi-Ann. Gamma Isotopic Semi-Annual

- 500 ft. south of Plant Centerline SL-3 Plant Site Boundary - North Sediment Semi-Ann. Gamma Isotopic Semi-Annual

- 500 ft. north of Plant Centerline INGESTION - MILK Indicator Farms Mille Once every 1-131 per sample 15 days Gamma Isotopic oer samole Mille Once every 1-131 per sample 15 days Gamma Isotooic oer samole Mille Once every I-131 per sample 15 days Gamma Isotopic oer samole INGESTION - MILK Background Farm*

I 1-131 I per sample I I Mille IOnceevery 15 days I Gamma Isotopic I 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 oortion 2/vear Gamma lsotooic oer 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 Michi"'m Fish - edible portion 2/year Gamma Isotopic per sample

Information PMP-6010-0SD-001 I Rev. 25 Page 84 of 89 OFF-SITE DOSE CALCULATION MANUAL Radiological Environmental Monitoring Program Pages:

Attachment 3.19 Sample Stations, Sample Types, Sample Frequencies 81- 84 INGESTION - FOOD PRODUCTS On Site ONS-G Nearest sample to Plant in the Grapes At time of Gamma Isotopic At time of highest D/Q land sector harvest harvest containing media.

ONS-V Broadleaf At time of Gamma Isotopic At time of vegetation harvest harvest Off Site OFS-G In a land sector containing Grapes At time of Gamma Isotopic At time of grapes, approximately 20 miles harvest Harvest from the plant, in one of the less prevalent D/Q land sectors OFS-V Broadleaf At time of Gamma Isotopic At time of vegetation harvest harvest INGESTION - BROADLEAF IN LIEU OF GARDEN CENSUS OR IN LIEU OF MILK (*)

3 samples of different kinds of broad leaf vegetation Broadleaf Monthly Gamma Isotopic Monthly collected at the site bonndary, within five vegetation when available Il31 when available miles of the plant, in each of 2 different sectors with the highest annual average D/Q containing media 1 background sample of similar vegetation Broad.leaf Monthly Gamma Isotopic Monthly grown 10-20 miles distant in one of vegetation when available Il31 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

~mes 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 broad leaf samples of different kinds will be collected in each of 2 different offsite locations, within five miles of the plant, with the highest D/Q (refers to the highest annual average ground D/QJ. Also, one background broad leaf sample of similar kinds will be collected 10 to 20 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 deterp:tlned that the milk animals are fed stored feed, THEN monthly sampling is appropriate for that time period.

Evaluate samples that identified positive plant effluent related radionuclides and determine if additional analysis are necessary to identify hard to detect radionuclides. The 10 CFR 61 scaling factor report should be consulted along with the radioactive material shipping program owner and the ODCM program owner to assist with this determination.

L

Information PMP-6010-0SD-001 I Rev. 25 I Page 85 of 89 OFF-SITE DOSE CALCULATION MANUAL P;:tges:

Attachment 3.20 Maximum Values for Lower Limits of DetectionsA,B - REMP 85 - 86

[Ref. 5.2.lw]

Radionuclides Food Product Water Milk Air Filter Fish Sediment pCi/kg, wet pCi/1 pCi/1 pCi/m3 pCi/kg, wet pCi/kg, dcy 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 I-131 60 1 l 0.07 This.Data is directly from 'our plant-specific Technical Specification.

i Information PMP-6010-0SD-001 l Rev. 25 I Page 86 of 89 OFF-SITE DOSE CALCULATION MANUAL Pages:

Attachment 3.20 Maximum Values for Lower Limits ofDetectionsA,B -REMP 85 - 86 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* /-J*M)

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

').. is the radioactive decay constant for the particular radionuclide

.M 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 ofDetectionsA,B _ REMP. *

• a 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.

Information PMP-6010-0SD-001 .I Rev. 25 Page 87 of 89 OFF-SITE DOSE CALCULATION MANUAL Reporting Levels for Radioactivity Concentrations Page:

Attachment 3.21 in Environmental Samples 87 Radionuclides Food Product Water Milk Air Filter Fish pCi/kg, wet pCi/1 pCi/1 pCi/m3 pCi/kg,. wet H-3 20000 Ba-140 200 300 La-140 200 300 Cs-134 1000 30 60 10 1000 Cs-137 2000 50 70 . 20 2000 Zr-95 400 Nb-95 400 Mn-54 1000 30000 Fe-59 400 10000 Zn-65 300 20000 Co-58 1000 30000 Co-60 300 10000 I-131 100 2 *3 0.90 JF any of the above concentration levels are l")xceeded THEN see guidance contained in step 3.5.2a. for additional "mformation.

Information PMP-6010-0SD-001 I Rev. 25 Page 88 of 89 OFF-SITE DOSE CALCULATION MANUAL Page: .22 On-Site Monitoring Location - REMP 88 ONS-North TLD T-6 TLD T-5 Air ONS- Air ONS-5 Surface Water 6 SWL-2 Sediment SL-2 Well W-16 LEGEND ONS-1-0NS-6: Air Sampling Station T 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 samolina locations

Information PMP-6010-0SD-001 I Rev. 25 Page 89 of 89 OFF-SITE DOSE CALCULATION MANUAL Page:

Attachment 3.23 Off-Site Monitoring Locations - REMP 89

~?-\.~-~,;;.,

Coloma Legend Offsite REMP Monitoring Locations Substation

.** .-:. :,.. ...,"':..-., :l Coloma Rd OFT OFT-11: TLD Locations Background AirfTLD Stations Drinking Water Locations Indicator Milk Farm Locations \

Background Milk Farm Locations **

...t Benton 20 Mile Radius OFS Offsite Fish locations ,:~ Harbor

.......,":" .. ... ,} ..

St Joseph Water Treatment Plant (STJ) Dowagiac Substation OFS-North Background AirffLD TLD OFT-3 (DOW)

TLD OFT-1 ColbvSt TLD OFT-2 TLD OFT-4 TLD OFT-9 TLD OFT-5

• • CS.*;

.,. ~ :** .

TLD OFT-7 TLD-OFT-10 TLD OFT-11 TLDOFT-8 TLDOFT-6

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L_

. REVISION

SUMMARY

Procedure No.: PMP-6010-0SD-001 Rev. No.: 25

Title:

OFF-SITE DOSE CALCULATION MANUAL Alteration Justification 10 CPR 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 ar.e made in accordance with Technical Specification 5. 5 .1 and implemented through 12-EA-6090-ENV-114, Effectiveness Review for ODCM/PCP Programs.

Step 2.1.2.j Added supplemental information Enhancement to provide clarity on the of the calculation of dose attributed to variables that would be utilized for carbon-14. calculating C-14 dose.

AR#2015-I439 Step 2.1.2.k.3 Added usage of installed Enhancement which reflects the installation blowdown flow instrumentation as an of modem technology allowing for more acceptable method for obtaining flows. accurate flow measurements of blowdown.

The data is obtainable on the Plant PPC computers. This change does not involve a change of procedure intent and reflects an improvement in measuring flows.

Step 2.4.1 Added specific mention of the Enhancement to ensure it is clearly dry cask storage facility (ISFSI) and the need understood that the direct radiation dose to to incorporate the dose from it with our dual public is a combination of that from our dual unit site data. unit site and the dry cask storage facility operated under a separate license.

AR#2015-1439 Step 2.7 .2 Revised verbiage throughout the Enhancement which clarifies the deadline to step to change wording from "by May 1" to provide the NRC the ARERR, based upon "prior to May l ". industry operating experience. This change does not involve a change of procedure intent.

AR#2015-1439 Office Infonnation for Fonn Tracking Only - Not Part of Fonn This is a free-form as called out in PMP-2010-PRC-002, Procedure Alteration, Review, and Approval. Page 1 of 3

REVISION

SUMMARY

Procedure No.: PMP-6010-0SD-001 Rev. No.: 25

Title:

OFF-SITE DOSE CALCULATION MANUAL Alteration Justification .19 Corrected distance The NUREG 1301 uses kilometers and an referenced and made clarification of the error was made where the unit conversion sector criteria where broadleaf sampling was from kilometers to miles did not convert the required, reflecting NUREG 1301 more actual number and only the unit. The value accurately. of 8 kilometers equals 5 miles. Adjusted the broadleaf distances to approximate miles equivalents of NUREG 1301 guidance given in kilometers.

The clarification on the broadleaf sampling involved an NRC Branch Technical Position having different verbiage than NUREG 1301. The previous revision reflected the verbiage from the Branch Technical Position. The change made here reflects the verbiage in NUREG 1301 and clearly states that samples from 2 different sectors with the highest D/Q are required. This does not involve a change of procedure intent and simply clarifies the process of broadleaf sampling.

AR#2015-1861 Office Infonnanon for Fonn Tracking Only - Not Part of Fonn This is a free-forni,as called out in PMP-2010-PRC-002, Procedure Alteration, Review, and Approval. Page 2 of 3

Doc No.: PMP-6010-0SD-001 Rev No.: 025

Title:

OFF-SITE DOSE CALCULATION MANUAL Alteration Cat.: Minor Revision CDl/50.59: NIA PORC Mtg. No.: NIA CARB Mtg. No.: NIA Admin Hold AR No.: N/A Superceding Proc(s): NIA Temp Proc Exp Date: NIA Temp Change Exp Date: NIA Temp Proc/Cbange End: N/A Effective Date: 7/22/2015 4:00:00 AM Approvals Name Review/Approval Type/Capacity Date j

/Wendzel, Regan  !!7 Approval Authority l/07/08/2015 07:49 I/

=IH=a=rn=er=,=Jo=n=====

ll.=5=M=a=n=ag=e=m=e=nt=&=e=v==ie=w=========:,I06/22/2015 15:09 i

,.=:z=or=d=el=l,=B=l=air====:!11;::::3=T=e=cbn==ic=a=lR=e=v=ie=w==========!!!06/17/20I5 13:46 i *

====================

Signature Comments jApproved per Plant Manager, Sam Partin. II

!Approved, however, a few editorial comments need incorporated. The mark up was placed on 1

!Erik's desk. 11 1

0

aJAMERKAH' ElEnJUC! ..

PMP-6010-0SD-001 Rev. 25 Page 1 of 89 POWER:

~.~~~~~

OFF-SITE DOSE CALCULATION MANUAL Information I Erik Merchant Environmental Manager Environmental Writer Document Owner Cognizant Organization TABLE OF CONTENTS*

1 PURPOSE AND SCOPE ............................................................................. 4 2 DEF'INITIONS AND ABBREVIATIONS ........................................................ 4 3 DET~S *. ~ ...................................*............ "............................................. 6 3.1 Calculation of Off-Site Doses ................................................................ 6 3.1.1 Gaseous Effluent Releases .. .-....................................................... 6 3.1;2 Liquid Effluent Releases ..... .-................. :,........................ ~ ........ 12 3.2 Limits of Operation and Surveillances of the Effluent Release Points ............. 15 3.2.1 Radioactive Liquid Effluent Monitoring Instrumentation .... ............ 15 3.2.2 Radioactive Gaseous Effluent Monitoring Instrumentation ............... 16 3.2.3 Liquid Effluents ****************~*******************-~ ..-. ............................ 17

a. Concentration Excluding Releases via the Turbine Room Sump (TRS) Dischargy .........*.................................................... 17

b. Concentration of Releases from the TRS Discharge .................... 18

c. Dose ..... ; ........... , ............................................................ 19

d. Liquid Radwaste Treatment System ........ *............................... 19 3.2.4 Gasoous Effluents ...................................................*.............. 21

a. Dose Rate ............................................................ , ......... 21 *

b. Dose - Noble Gases .......................................................... 22

c. Do.se - Iodine-131, lodine-133, Tritium, and Radioactive Material in Particulate Form .......................:................. ; ............*....... 22 cl. Gaseous Radwaste Treatment. .............................................. 23 3.2.5 Radioactive Effluents -Total Dose ................................... ~ ............ 25 3.3 *caJculation of Alarm/Trip Setpoints **~**************************************************** 27 3.3.1 Liquid Monitors .................................................................... 27

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

b. Liquid Continuous Monitor Setpoint Methodology ..................... 29

. 3.3.2 Gaseous Monitors ..... 0 ********************* 0 ********.****************************** 31

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

_ 31 b: Waste Gas Storage Tanks ................................................... 35

c. Containment Purge and Exhaust System ................. : ............. *... 35

d. Steam Jet Air Ejee;tor System (SJAE) ..................................... 36

e. Gland Seal Condenser Exhaust ........................ : .................... 37

. SAMIRICAN' l!'I.ECDIH:

PMP-6010-0SD-001 Rev. 25 Page2 of 89 FOWER'

""'-~.,,;,,,;;;.......

OFF-SITE DOSE CALCULATION MANUAL Information I Erik Merchant Environmental Manager Environmental Writer Document Owner Cognizant Organization 3.4 Radioactive Effluents Total Dose .~ ........................................ , ............... 37 3.5 Radiological Environmental Monitoring Program (REMP) .............*........... 37 3.5.1 Purpose of the REl\iP ............................................................. 37 3 .5 .2 Conduct of the REl\iP ............................................................ 38 3 .5 .3 .Anilual Land Use Census ........................................................*40 3.5.4 Interlaboratory Comparison Program ***.*.................*.*****.********.* 41 3.6 Meteorological Model .......................................................................... 42

3. 7 Reporting Requirements .......................... ~ .......................................... -42 3.7 .1 Annual Radiological Environmental Operating Report (AREOR) ..... 42 3.7 .2 Annual Radiological Effluent Release Report (ARERR) .................. 43 3.8 10 CFR 50.75 (g) Implementation *....................*................................... 45 3.9 Reporting/Management Review*****************************"!:*~**************************** 45 4 E'JN'AL CONDITIONS .................................................................................46 5 REFERENCES* *~*******************"****************~**************************************************46 SUPPLEMENTS Attachment 3.1 Dose Factors for Various Pathways .. , ................................... :... ,... Pages 49 - 52, Attachment 3 .2 Radioactive Liquid Effluent Monitoring Instruments ................... Pages 53 - 55 Attachment 3.3 Radioactive Liquid Effluent Monitoring Instrumentation Surveillance Requirements ............................................................ Pages 56 - 57

• Attachment 3.4 Radioactive Gaseous Effluent Monitoring Instrumentation .......... Pages 58 - 60 Attachment 3.5 Radioactive Gaseous Effluent Monitoring Instrumentation Surveillance Requirements ............................................................ Pages 61 - 62 Attachment 3 .6 Radioactive Liquid Waste Sampling and Analysis Program ....... '.Pages 63 - 64 Attachment 3.7 Radioactive Gaseous Waste Sampling and Analysis Program ..... Pages 65 - 66 Attachment 3.8 Multiple Release Point Factors for Release Points ................................ Page 67

fZD:t:=cAH' P:MP-6010-0SD-001 Rev. 25 Page 3 of 89 POWER A{lr..tmrr&2lo7T.J1tWr.Dtaf'..

OFFMSITE DOSE CALCULATION MANUAL Information I Erik Merchant Environmental Manager Environmental Writer Document Owner Cognizant Organization .9 Liquid Effluent Release Systems ............................................................ Page 68 .10 Plant Liquid Effluent Parameters ........................................................... Page 69 .11 Volumetric Detection Efficiencies for Principle Gamma Emitting Radionuclides for Eberline Liquid Monitors ........................... Page 70 .12 Counting Efficiency Curves for R-19, and R-24 ........................... Pages 71 - 72 .13 Counting Efficiency Curve for R-20, and R-28 ..................................... Page 73 .14 Gaseous Effluent Release Systems ........................................................ Page 74 .15 Plant Gaseous Effluent Parameters ........................................................ Page 75 .16 10 Year Average of 1995-2004 Data ............................................ Pages 76 - 77 .17 Annual Evaluation of x/Q and D/Q Values For All Sectors ................. Page 78 .18 Dose Factors .................................................................................. Pages 79-80 .19 Radiological Environmental Monitoring Program Sample Stations, Sample Types, Sample Frequencies ............................... Pages 81 - 84 .20 Maximum Values for Lower Limits ofDetectionsA,B - REMP ..... Pages 85 - 86 .21 Reporting Levels for Radioactivity Concentrations in Environmental Samples .......................................................................... Page 87 .22 On-Site Monitoring Location - REMP ................................................... Page 88 .23 Off-Site Monitoring Locations -REMP ................................................. Page 89

Information I PMP-6010-0SD-001 I Rev. 25 I Page 4 of 89 OFF-SITE DOSE CALCULATION MANUAL 1 PURPOSE AND SCOPE

• 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 methodologies 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:

Sor 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 /> Dor 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 /> Wor weekly At least once per 7 days Mor 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.

SIU 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

Information I I Rev. 25 I Page 5 of 89 OFF-SITE DOSE CALCULATION MANUAL Member(s) of All per_sons who are not occupationally associated with the Public plant.-. boes 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(l; + C(2; + ... ~ 1 Lr1; L(2J Where; C(l) = Concentration of pt detected nuclide Cc2) = Concentration of 2nd detected nuclide L(l) = Reporting Level of 1st nuclide from Attachment 3.21, Reporting Levels for Radioactivity Concentrations in Environmental Samples.

Lc2) = Reporting Level of 2nd 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.

Information I PMP~6010-0SD-001 I Rev. 25 l Page 6 of 89 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

• MIDEX

• MIDEL

• MIDEG

• MIDEN

b. The subprogram used to enter and edit gaseous release data is called MDlEQ (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,D/J air=~ *I,[( M;or N;)*Q;

• 3.17E-8]

Where; Dy , Dp air = the gamma or beta air dose in mrad/yr to an individual receptor x IQ = the annual average or real time atmospheric dispersion factor over land, sec/m3 from Attachment 3.16, 10 Year Average of 1995-2004 Data Mi = the gamma air dose factor, mrad m 3 / yr µCi, from Attachment 3.18, Dose Factors Ni = the beta air dose factor, mrad m 3 /_ yr µCi, from Attachment 3.18, Dose Factors

Information I PMP-6010-0SD-001 I Rev. 25 I Page 7 of 89 OFF-SITE DOSE CALCULATION MANUAL

= the release rate of radionuclide, "i", in µCi/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).

,,,f.Q 2.03

• Al:

_

• X

• Lg T1

= Umg Where; 2

Lg === minimum of c,2 + H c or Lg= Ji er z Zg 2ff 8 x = distance downwind of the source, meters. This information is found in paramet~r 5 of MIDEX.

um 8

= wind speed for ground release, (meters/second) a-z8 = vertical dispersion coefficient for ground release, (meters),

(Reg. Guide 1.111 Fig.I)

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

(Containment Building = 49.4 meters)

Tr = terrain factor(= 1 for Cook Nuclear Plant) because we consider ~ our releases to be ground level (see parameter 5 inMIDEX).

2.03 = .J2..;.-:ti ..;.-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.

Information I PMP-6010-0SD-001 I Rev. 25 *I Page 8 of 89 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. G1,1ide 1. i09: *

1. Total Body Plume Pathway (Eq 10)

Dose(mrem/year)=3.17E-8*L(Q;* l7'Q*S1* DFB;)

Where; Sr = shielding factor that accounts for the dose reduction due to shlelding provided by residential structures. during occupancy (maximum exposed individual = 0. 7 per Table E-15 of Reg. Guide 1.109)

DFBi = the whole body dose factor from Table B-1 of Reg.

Guide 1.. 109, mrem - m 3 per µCi - yr. See Attachment 3.18,' Dose Factors.

Qi = the release rate of r~dionuclide "i", in µCi/yr

2. Skin Plume Pathway (Eq 11)

Dose(mrem/yr)=3.17E-8* s1 * ~*[~(Qt* 1.11* DF~)+ L(Q;

• DFS;)]

Where; 1.11 = conversion factor, tissue to air, mrem/mrad DF / = the gamma air dose factor for a uniform semi-infinite cloud of radionuclide ".i", in mrad m 3/µCi yr from Table B-1, Reg. Guide 1.109. See Attachment 3.18,.

Dose Factors,

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

Information I PMP-6010-0SD-001 I Rev. 25 I Page 9 of 89 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:

DIP (mrem/year) = 3. J7E -8

• L(R;

• W

• Q;)

Where; R; = the most restrictive dose factor for each identified radionuclide "i", in m2 mrem sec/ yr µ.Ci (for food and ground pathways) or mrem m 3 I yr µCi (for inhalation pathway), for the appropriate pathway For sectors with existing pathways within five miles of the site, use the values of R; for these real pathways, otherwise use pathways distance of five miles. See Attachment 3.1, Dose Factors for Various Pathways, for the maximum Ri values for the most controlling age group for select~d radionuclides. R; values were generated by computer code PARTS, see NUREG-0133, AppendixD.

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:

Wm x IQ for the inhalation pathway, in sec/m3

-OR-Wrg = DI Q for the food and ground pathways in 11m2 Q;c = 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 µ.Ci/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.

Information I PMP-6010-0SD-001 I Rev. 25 I Page 10 of 89 OFF-SITE DOSE CALCULATION MANUAL

1. 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. Carbon-14 (C-14) supplemental information

1. The quantity of C-14 released to the environment may be estimated by use of a C-14 source term scaling factor based on power generation (Ref. RG 1.21, Revision 2). A recent study recommends a source term scaling factor of approximately 9.0 to 9.8 Curies/GWe-yr for a Westinghouse Pressurized Water Reactor (Ref.

EPRI 1021106 "Estimation of Carbon-14 in Nuclear Plant Gaseous*.

Effluents" December 23, 2010). For this method, a scaling factor of 9.4 Curies/GWe-yr shal). be used.

2. C-14 releases from PWRs occur primarily as a mix of organic carbon (methane) and inorganic carbon (carbon dioxide). For this method, an average organic fraction of 80 % with the remaining 20 % being assumed as carbon dioxide shall be used.

3. Dose is calculated utilizing the methodology prescribed in RG 1.109 Appendix C, with the vegetation dose being the most predominant.

Adjustments for growing seasons, percentage of C-14 generated assumed released from the reactor coolant in gaseous form via batch releases, seasonal X I Q , and other industry methodologies being considered by the NRC may be applied as desired should their acceptance of these methods occur.

k. Steam Generator Blowdown System (Start Up Flash Tan1c 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= µCi* GPM

• time on flash tank (min)* 3. 785E -3 ml

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

3. The flow rate is determined from the blowdown valve position and the time on the start up tank, or using installed plant blowdown flow instrumentation. Chemistry Department performs the sampling and analysis of the samples.

Information I PMP-6010-0SD-001 I Rev. 25 I Page 11 of 89 OFF-SITE DOSE CALCULATION MANUAL

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 µ,Ci/g dose equivalent 1-131.

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

QY = Ci* IPF*. Rsgb Where; Qy = the release rate of 1-131 from the steam generator flash tank vent, in µ,Ci/sec Ci = the concentration (µ,Ci/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.

Information I PMP-6010-0SD-001 I Rev. 25 I Page 12 of 89 OFF-SITE DOSE CALCULATION MANUAL 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 MD lEB (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 p\lrposes to maintain uniformity with this defined pathway.

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

1. Potable Water (Eq 1)

Rap,-=1100* MP* F Uap *

• "Q D * * -;.,,,

• 2.23E-3 "7-' ; a,pje Where; Rapi = 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.

Information I PMP-6010-0SD-001 I Rev. 25 I Page13 of 89 OFF-SITE DOSE CALCULATION MANUAL 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 Qi = 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-11 through E-14 of Reg. Guide 1.109 and are located within the MIDAS code.

Ai = the radioactive decay constant for radionuclide "i", in hours- 1 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. (1:p = 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)

R apj

=1100* Uap Mp* F* 2.23E-3 ~ 1

• "'Q

• B*

• D aipJ*

'P

• e*).,tp Where, Bip = the equilibrium bioaccumulation factor for nuclide "i" in pathway "p", expressed as pCi LI 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 .

. 1:p = 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.

Information I PMP-6010-0SD-001 I Rev. 25 l Page 14 of 89 OFF-SITE DOSE CALCULATION MANUAL

3. Shoreline Deposits (Eq 3)

R ap;.= 110' 000

• Mp* F*

U *W ap * "'Q*

2.23E-3 ~ '

• T,

• Da,p;le 1

.. r -A;tp]* [1- -~lb]

l e 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 Daipi = the dose factor for standing on contaminated ground, in mrem m2 / hr pCi. The values are taken from table E-6 of Reg. Guide 1.109 and are located within the MIDAS code.

See Attachment 3.1, Dose Factors for Various Pathways.

ti, = the period of time for which sediment or soil is exposed to the contaminated water, l.31E+5 hours. Given in*

MIDEL as itein 6 of parameter 4.

ti, = the average transit time required for nuclides to reach the point of exposuret Ohours. Given as #28 of parameter 4 inMIDEL.

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

.aquati~ 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 DSlLI are printed in LDRPT (LP). These results are used in the monthly report of liquid releases.

Information PMP-6010-0SD-001 Rev. 25 - - - 'I- - ~Page 15 of 89

-""------II OFF-SITE DOSE CALCULATION MANUAL

g. In addition, the program DOSUM (DM) is used to search the results files of DSlLI to find the maximum liquid pathway individual ,doses. The highest exposures are then printed iri 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. G1;1ide 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 surveillan.ce interval.

3 .2 Limits of Operation and Surveillances <?f the Effluent Release Points

3. 2.1 Radioactive Liquid Effluent Monitoring Instrumentation

~- .The radioactive liquid effluent monitoring instrumentation channels_

shown in Attachment 3.2, Radioactive Liquid Effluent Monitoring Instruments, are operable *with their alann/trip setpoints .set to ensure that the limits of step 3 .2. 3a, Concentration Excluding Releases vfa the Turbine 1:loom Suml) (TRS) Discharge,.are not exceeded..

b. The applicabili1y of each ch~el is shown in Attachment ~.2, Radioactive Liquid Effluent Monitoring Instrum\;lnts:

c. With a radioactive liquid effluent monitoring instrumentation channel alatm/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 suryeillance interval, excluding the Wtial performance.

e. Determine the setpoints in accordance with the methodolbgy described in

.step .3.3'.1, Liquid Monitors. Record the setpoints.

Information I PMP-6010-0SD-001 I Rev. 25 I Page 16 of 89 OFF-SITE DOSE CALCULATION MANUAL

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 alann/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]

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 ensu~e that the limits of step 3.2.4a, Dose Rate, are nc;,t 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 alann/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.

/

Infonnation .. 1 PMP-6010-0SD-001 I Rey. 25 I Page 17 of 89 QFF.,SITE DOSE CALCULATION-MANUAL

d. With less than the minimum number of radioactive gaseous effluent monitoring instrumentation ~hannels 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 perforniance, 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 dOC\lIDent.

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, CltANNEL CALIBRATIO~, and CHANNEL OPERATIONAL TEST operations at the frequencies shown in Attachment 3 5, Radioactive Gaseous Effluent Monitoring .

Instrumentation Surveillance Requirements.

BASES ;_ GASEOUS The radiqactive gaseous effluent instrumentation is provided to monitor and control, as

• applicable, the ,releases of radioactive materials in gaseous effluents dqdng actual or potential releases. 1,'he 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 specjfied_ in Section 11. 3 of the Final Safety Analysis Report for the Donald C. Cook Nuclear Plant, 3.2.3 'Liquid Effluents

a. Concentratiqn Excluding Releases via the Turbine Room: SUmp (TRS)

Discharge

1. Limit the concentration of radioactive material relea/,ed via the Batch Release Tanks or Plant Continuous Releases (excluding only TRS discharge to the Absorption Pond) to uprestricted areas to the *

• concentrations in 10 CPR 20, Appendix 13, 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 µCi/ml total activity.

Information I PMP-6010-0SD-001 I Rev. 25 I Page 18 of 89 OFF-SITE DOSE CALCULATION MANUAL

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 µCi/ml total a_ctivity.

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. l 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.

Information I PMP-6010-0SD-001 I Rev. 25 I Page 19 of 89 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 :c::; 1.5 mrem/unit to the total body and to :c::; 5 mrem/unit to any organ, a.J._ld during any calendar year to :c::; 3 mrem/unit to the total body and to :c::; 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.l above, prepare and submit a Written Report, pursuant to 10 CPR 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, Special and 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 0.12 mrem (0.06 mrem/unit x 2 units) to the total body or 0.4 mrem (Ci.2 mrem/unit x 2 units) to any organ.

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

Information . I PMP-6010-0SD-001 I Rev. 25 I P~ge 20 of 89 OFF-SITE DOSE CALCULATION MANUAL e, Drai,nage of high conductivity water (Component Cooling Water and ice melt water containing sodium tetraborate) shall be evaluated to decide whether it sb.ould 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 bas on the radioactive wastewater dentjneralization 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:

( G)(Vi) + (Ca)(Va) = (C)(Vi)

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 c1* = the final concentration of the system after the addition Vt = . the final ~olume of the system after tb.e addition The intentis to keep the:

• WDS below 500 µmhos/cc.

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

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 CPR :Part 20; Appendix B, Table z. This limitation provides

, ad<litional assurance that the levels of radioactive materials ih bodies of water outside the

• site will not result in exposures *greater than 1) the-Section II.A design objectives of Appendix I, 10 CPR Part 50, to an individual and 2) the limits of 10 CPR 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.

Information I PMP-6010-0SD-001 I Rev. 25 I Page21 of 89 OFF-SITE DOSE CALCT)LATION MANUAL DOSE This specification is provided to implement the requirements of Sections IT.A, ill.A, and IV.A of Appendix I, 10 CFR Part 50. The dose limits implement the guides set forth in Section IT.A of ApP.endix 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 "a,s low as is reasonably achievable". Also, for fresh water sites with drinking water supplies which can be potentially affected by plant operations, there is reasonable assurance that the operation of the facility will not result in radionuclide concentrations in the finished drinking water that are in excess of the requirements .of 40 CFR 141. The dose calculations in the ODCM implement the requirements in Section ID.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 Amiual Doses to Man

• from Routine Releases of Reactor Effluents for tlie 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 e_ach reactor at the site. The liquid effluents from the shared system are proportj.oned among the units sharing the system.

LIQUID WASTE TREATMENT The operability of the liquid radwaste treatment system ensures that this system will be '

available for use whenever liquid effluents require treatment prior to release to the environment. The requirements that the appropriate portions of this system be used when specified provide assurance that the releases of radioactive materials in liquid effluents will be kept "as low as is reasonably achievable". This specification implements the requirements of 10 CPR 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 IT.D of Appendix Ito .10 CPR 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 IT.A of Appendix I, 10 CPR Part 50, for liquid effluents.

3.2.4 Gaseous Effluents

a. DoseRate

Information I PMP-6010-0SD-001 I *- Rev. 25 I Page22 of 89 OFF-SITE DOSE CALCULATION MANUAL

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 describe~ 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.

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 addre~sed 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 with half-lives greater than eight days in gaseous effluents

• released to unrestricted areas (site boundary) to the following:

a) During any calendar quarter to less than or equal to 7.5 mrem/unit to any organ b) During any calendar year to less than or equal to 15 mrem/unit to any organ.

Information I PMP-6010-0SD-001 I Rev. 25 I Page 23 of 89 OFF-SITE DOSE CALCULATION MANUAL

2. With the calculated dose from the release of radioiodines, radioactive materials in particulate form, or radionuclides other than noble gases in gaseous effluents exceeding any of the above limits, prepare and submit a Written Report, pursuant to 10 CFR 20.2203 and addressed in step 3.2.3c.2, within 30 days after learning of the event.

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

d. Gaseous Radwaste Treatment

1. The UFSAR (Updated Final Safety Analysis Report) states that radioactive waste gas should be held for 45 days of decay time.

2. 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 0.4 mrad (0.2 mrad/unit x 2 units) for gamma radiation and 0.8 mrad (0.4 mrad/unit x 2 units) 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.

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

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 CPR 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 11.C design objectives of appendix I to 10 CPR 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 dos~ 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 ra_te 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.

Information I PMP-6010-0SD-001 I Rev. 25 I Page24 of 89 OFF-SITE DOSE CALCULATION MANUAL 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 n.B, ill.A, and IV .A of Appendix. I, 10 CFR Part 50. The dose limits implem~nt the guides set forth in Section ILB of Appendix I.

The ACTIQN statements provide the required operating flexibility and at the same time implement the guides set forth in section IV .A of Appendix I to as so.re that the releases of radioactive material in gaseous effluents will be kept "as low as is reasonably achievable".

The Surveillance Requirements implement the requ_irements in Section III.A of Appendix I that conform with the guides of Appendix I to be shown by calculatiop_al 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 l, 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 l. July 1977. The ODCM equations provided for determining the air doses at the site boundary will be based upon the historical average atmospherical conditions. NUREG-0133 provides methods for dose calculations consistent with Regulatory Guides 1.109 and 1.111.

DOSE, RADIOIODINES, RADIOACTIVE MATERIAL 1N PARTICULATE FORM, ANp 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.

• Information PMP-6010-0SD-001 I Rev. 25 I Page25 of 89 OFF-SITE DOSE CALCULATION MANUAL The ACTION statements provide the required operating flexibility and at the saine 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 particu1ate form, and radionuclides, other than noble gases, are depenqent on the existing radionuclide pathwa,ys 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 Desigu Criterion Section 11.1 of the Final Safety Analysis Report for the Donald C. Cook Nuclear Plant, and design objective Section Il.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 Il.B and Il.C of Appendix. I, 10 CFR Part 50, for gaseous effluents.

3.2.5 Radioactive Effluents -Total Dose

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

b. With the calculated doses from the release of radioactive materials in liquid or gaseous effl~ents 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:

Information I PMP-6010-0SD-001 I Rev. 25 I Page 26 of 89 OFF-SITE DOSE CALCULATION MANUAL

• fuvestigate 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.ll(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 CPR 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]).

BASES -- TOTAL DOSE This specification is provided to meet the dose limitations of 40 CFR 190. The specification requires the preparati~n 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 coIJllilitment 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.

Information I PMP-6010-0SD-001 I Rev.25 I Page 27 of 89 OFF-SITE DOSE CALCULATION MANUAL

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 CPR 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.

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 systeD,I design and operation can be found in the system descriptions. The setpoints are intended to keep releases within the limits of 10 CPR 20, Appendix B, Table 2, Column 2. Determine setpoints using either the batch or the cont4Iuous 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 CPR 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.

Information I PMP-6010-0SJ)-001 I Rev. 25 I Page.28 of 89 OFF-SITE, DOSE CALCULATION MANUAL.

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 Attacbment 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 i, Column 2. The equation to calculate the flow rate is from Addendum AAl of NUREG-0133:

[

}: C; ]*_f_-:;,F+f LIMIT; MRP Where; Ci = the concentration of nuclide "i" in µCi/ml LIM{Ti = the 10 CFR 20, Appendix B, Table 2, Column 2 limit of

.nuclide "i" in µCi/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 irroperation.

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 ~ay be rearranged to solve for the maximl,lill effluent release flow rate (f).

I

Information I PMP-6010-0SD-001 I Rev. 25 I Page 29 of 89 *

• OFF-SITE DOSE CALCULATION MANUAL

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. Tlle predicted v:;il.ue is generated by converting the effluent concentration for each gamma emitting radionuciide to counts per unit of time .as per Attachment 3 .11, Volumetric Detection Efficiencies for Principle Gamma Emitting Radionuclides for Eberline Liquid Monitors, or Attachment 3.12, Counting Efficiency Curves for R-19, and R-24. The sum of all the counts per unit of time is th~ predicted count rate. The predicted count rate can tlien 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 (E~W) ~ystems.

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

• R-19 ot 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, Appenqix B, Table 2.,

Column 2, entering the unrestricted area following dilution,

3. The monitors on steam generator blowdown and blowdown .*

treatment systems have trip functions associated With their setpoints.

Essential service water monitors are equipped with an alarm function only and monitor effluent in the event the Containment

• Spray Heat Exchangers are used.

4. The equation used to determine the setpoint for continuous monitors is from Addendum AAl ofNUREG-0133:

s<. C* E.r+*

=:JJ MRP*F*SF

. p- f Where; Sp = setpoint of monitor (cpm)

Information I PMP-6010-0SD-001 I Rev. ~5 I Page 30 of 89 OFF-SITE DOSE CALCULATION MANUAL C = 5E-7 µCi/ml, maximum effluent control limit from 10 CPR 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, 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:

I( C;

• Ejf ,) replaces C

• Eif.f I e, LIMIT, MRP = multiple release point factor. A factor such that when all the release points are operating at one time the limits of 10 CPR 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).

Information I PMP-6010-0SD-001 I Rev. 25 I P~e31 of89 OFF-SITE DOSE CALCULATION MANUAL 3.3.2 Gaseous Monitors For the purpose of implementing Step 3 .2.2, Radioactive Gaseous Effluent Monitoring Inst:rillhentation, 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 nor 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 paramete~(s).

Gaseous effluent monitor high al~ setpoints will routinely be established at a fraction of the maximum allowable setpoint (typically 10% of the setpoint) for ALARA purposes. Alert alarms will norn1ally be set to provide adequate indications of small changes in radiological conditions.

NOTE: IF the -setpoint calculation methodology changes or the as.sociated factors change for Unit Vent, Air Ejector and/or Gland Seal monitors, THEN initiate a review by"Emergency Planning to ensure that the requu;ements of 10 CFR 50.54 (q) are maintµned.

a. Plant Unit Vent

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 (Unifi), VRS-2505 (Unit.2)] to assure that appllcable alarms .and trip actions (isolation of gaseous release)

, will occur prior to exceeding .the limits in step 3.2.4,. Gaseous Eftluel).ts. The alarm setpoint values will be established using the following unit analysis equation:

• Where; Sp "'" the maximum setpoint of the monitor in µCi/cc for .

release point p, based on the mo~t limiting organ SF = an administrative operation safety factor, less than 1.0

Information I PMP-6010-0SD-001 I Rev. 25 I Page 32 of 89 OFF-SITE DOSE CALCULATION MANUAL 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, fp) from each release point p.

• Compute LQp (or Lfp) for all release points.

• Ratio Qp/LQp (or fp/Lfp) for each release point.

This ratio is the MRP for that specific release point

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

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 cfin for Unit 1 and

• 143,400 cfm for Unit 2.

Information I PMP-6010-0SD-001 I Rev. 25 I Page 33 of 89 OFF-SITE DOSE CALCULATION MANUAL DLi = dose rate limit to organ "j" in an unrestricted area (inrem/yr).

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

• Total Body ~ 500 ~em/year

• Skin ~ 3000 mrem/year

• • - Any Organ~ 1?00 mrem/year x IQ = The worst case annual average relative concentration in the applicable sector or area; in sec/m3 (see Attachment 3.16, 19 Year Average of 1995-2004 Data).

Wi = Weighted factor for* the radionuclide*:

Where, Ci = concentration of the most abundant radionuclide "i" Ck == total c;oncentration of all identified radionuclides in tbat release pathway. .For batch releases, this value may be set to 1- for conservatism.

DCFii = dose conversion factor used to relate radiation dose to organ "j "., from exposµre to radionuclide "i" in mrem m 3 / yr µCi, See

')

following equations.

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

For the whole body: DCF;j = Ki Where; Ki = whole body dose factor due to gail).II).a emissions for each identified noble gas radionuclide in mrem m.3 I yr µCL See Attachment 3.18, Dose Factors.

Information I PMP-6010-0SD-001 I Rev. 25 I Page 34 of 89 OFF-SITE DOSE CALCULATION MANUAL For the skin: DCFu =Li+ 1.lMi Where; Li = skin dose factor due to beta emissions for each identified noble gas radionuclide, in mrem m 3 / yr µCi. 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).

Mi = the air dose factor due to gamma emissions for each identified noble gas radionuclide in mrad m 3 I yr µCi. See Attachment 3.18, Dose Factors.

For the thyroid, via inhalation: DCFu = Pi Where; P; = the dose parameter, for radionuclides other than noble gas, for the inhalation pathway in mrem m 3 I yr µCi (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 cp.annel setpoint, SP, will be set such that the dose rate in unrestricted areas to the whole body, skill 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 eves 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.

Information

• I PMP-6010-0SD-001 I Rev. 25 I Pag~ 35 of-89 OFF-SITE DOSE CALCULATION MANUAL

7. At certain times, it may be desirable to increase the high alarm setpoint, if the vent flow rate is decreased. -1bis may be accomplished in one, of two ways.

Max.Cone( µCi/cc) *Max.Flowrate (cfin) =New.Maxcfm New M~Concentration ( µCi/cc)

-OR-Max.Conc ( µCi/cc)

• Max.Flowrate* (cfin)

NewMaxµCilcc New Max.Flowrate (cfin)

b. Waste Gas Storage Tanks L The gaseous effluents discharged from the Waste Gas System are monitored by the vent stack monitors VRS-1505 and VRS-2505.

2. fu 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 ~hannel (VRS-1°505 or VRS-2505). Therefore, for any gaseous release configuration,

• which includes nonnal operation and waste gas system gaseous discharges, the alarm setpoint of the plant vent radiation monitor will be recomputed to determine the most iimiting 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 apptove.d Operations' procedures. * *

3. It is nonnally prudent to allow 45 days of decay prior to releasing a Gas Decay Tank (GOT). There are extenuating, operattonaI

\ circumstances that may prevent this from occurring. Under these circumstances, such as high oxygen concentration creating a combustible atmosphere, it is prudent to waive the 45-day decay for safety's sake.

c. Containment J>u:rge and Exhaust System L The gaseous effluents discharged by the Containment Purge and Exhaust Systems, and Instrutnentation 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 Rat~.

fuformation I PMP-6010-0SD-001 I Rev. 25 I Page 36 of 89

_OFF:-SITE DOSE CALCULATION MANUAL

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 Conqrinment purge before release.

3. The Upper Containment area is monitored by normal range area gamma monitors (Tag No. VRS-1101/1201 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.

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 value will be established using the following unit analysis equation:

SF* MRP* DL*

s = J SJAE F P

• Xi~

.In *" (

Li W,

• DCFij )

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

Information I PMP-6010-0SD-001 I Rev. 25 I Page 37 of 89 OFF-SITE DOSE CALCULATION MANUAL

e. Gland Seal Condenser Exhaust

1. The gaseous effluents from the Gland Sea} Condell$er Exhaust discharged to the environment are continuously monitored by radiation monitor (fag 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:

Where; SGscE = the maximum setpoint, based on the most limiting o:rgan, in µCi/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 jn steps 3:2.3c (Dose), 3.2.4b (Dose - Noble Gases), and 3.2.4c (Dose - Iodine-131, Xodine-133, Tritiµ.m, and Radioactive*MateriaI in Particulate Form) of this /

procedure. Dose contribution from direct radiation exposur~ will be based on the results of the direct radiation monitoring devices locate.d at the REMP monitoring stations, and.reflects direct dose* bofu from the Dry Cask Storage Facility (ISFSI) licensed under Holtech International and both units of Cook.

See NUREG-0133, section 3.8.

• 3.5 Radiological Enviromilental Monitoring Program (REMP) 3 .5 .1 Purpose of the REMP

a. The purpose of the REMP is to:

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

• Monitor critical environmental exposure pathways,

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

Information I I Rev. 25 I Page 38 of 89 OFF-SITE DOSE CALCULATION MANUAL

• Assist with fulfilling the requirements of the Groundwater Protection fuitiative (GPI).

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 th~ scope of the REMP for the Donald C. Cook Nuclear Plant.

3.5.2 Conduct of the REMP [Ref. 5.2.lv]

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 ofDetectionsA,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 (ARBOR).

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.

fuformation I PMP-6010-0SD-001 I Rev. 25 I Page 39 of 89 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 J;iave 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.

• Document the notification using PMP-6090-PCP-100, Data Sheet 2, Part 4 Radioactive Liquid Spill Which May hnpact Groundwater.

• Forward a copy of the notification to the Environmental Department Manager.

Information I PMP-6010-0SD-001 I Rev. 25 I Page 40 of 89 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 moI).itoring 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 ofDetectionsA,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.lv]

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.

Information I PMP-6010-0SD-001 I Rev. 25 I Page 41 of 89 OFF-SITE DOSE CALCULATION MANUAL

b. In lieu of the garden census, broad leaf vegetation sampling of at least three different kinds of vegetation (if available) may be performed as close to the site boundary as possible (within 5 miles) in each of two different direction sectors 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, if required by the results of the census.

This census satisfies the requirements of Section IV.B.3 of Appendix I_ to 10 CPR 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 (26 kg/yr) of leafy vegetables assumed in Regulatory Guide 1.109 for consumption by 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 yield 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 ¥1 an Interlaboratory Comparison Program, for radioactive materials. Address program results and identified deficiencies in the ARBOR.

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

- - - - - - - - - - - - - - - - - - - - - - - - ~

Informatlon I P~-:6010.;0SD-001 _ -, Rev. 25 I Page 42 of 89 OFF-SITE. DOSE CALCULATION MANUAL BASES - JNl'ERLABORATORY COMPARISON PROGRAM The requirement for participation in an Interlaboratory Comparison Program is provided to ensur!;: independent checks oii 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 3.6.1 Tbree towers are used to determine the meteorological conditions .at Donald C.

Cook Nuclear Plant. One of the towers is located at the La;ke Michigan shoreline to determine the meteorological parameters associated with unmodified shoreline air. The data is accumulited by microprocessors at the tower sites and normally transferred to the central computer every 15 minutes. .

3.6.2 The central computer uses a meteorol_ogical 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 (ARBOR)

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. lj, TS 5.6.2]

b. Include in the ARBOR:

• Summaries, interpretations, and statistical evaluation of tlie 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 oµ the environment. .

• • The results of the land use censuses required by step 3.5.3, Annual Land Use Ce11.$US.

• 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.

Information I PMP-6010-0SD-001 I Rev. 25 I Page 43 of 89 OFF-SITE DOSE CALCULATION MANUAL

• Summarized and tabulated results of all radiological environmental samples taken during the reporting period. In the event that some results are not available for inclusion with the report, submit the report noting and explaining the reasons for the missing results.

Submit the missing data as soon as possible in a supplementary report.

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

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

• The results of participation in the Interlaboratory Comparison Program required by step 3.5.4, Interlaboratory Comparison Program.

• The results of non-REMP samples taken for informational purposes in support of non-program specific investigations, such as rainfall studies of tritium recapture for example.

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 prior to May pt of each year. [Ref 5.2. lj, 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 prior to May 1sr 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.

Information . I PMP-6010-0SD-001 I Rev. 25 I Page 44 of 89 OFF-SITE DOSE CALCULATION MANUAL

• 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 y~r.

• Include an assessment of the radiatfon 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 detennined by sampling frequency and measurement) for detennining the gaseous pathway doses.

• Inoperable radiation monitor periods exceeding 30 continuous days; explainLcauses of inope~ability and *actions taken to p:i;event reoccurrence.

d. Submit the ARERR [Ref. 5,2.lx] pri9r to May l51 of each year and .

include an assessment of radiation dosei; to the likely most exposed member of t,he publi~ from reactor releases and other nearby uranium fuel cycle sourc~s (including doses from primary eftluertt pathways and direct radiation) for the previous 12 consecutive months to s.how 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.l.

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

• Volume (cubic meters),

• Total curie quantity (specify whether detennined 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).

Information I I Rev. 25 I Page 45 of 89 OFF-SITE DOSE CALCULATION MANUAL

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 CPR 141, National Primary Drinking Water Regulations. Unit specific values are site values divided by two.

i. Include in the ARERR groundwater sample results taken that are in support of the Groundwater Protection Initiative (GPI) but are not part of the REMP.

3.8 10 CFR 50.75 (g) Jinplementation

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 item for 10 CPR 50. 75(g) 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 ~s procedure for potential impact on other related Department Procedures.

Information I PMP-6010-0SD-001 I Rev. 25 I Page 46 of 89 OFF-SITE DOSE CALCULATION MANUAL 3.9.5 Review this procedure during the first quarter of each year and update it if necessary. Review Attachment 3.16, IO Year Average *of 1995-2004 Data, and document using Attachment 3.17, Annual Evaluation of x/Qand D/Qvalues For All Sectors. The x IQ and DI 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:

5 .1.1 "Implementation of Programmatic Controls for Radiological Effluent Technical Sp~ifications 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 (ARBOR) Preparation And Submittal 5.1.4 PMP-6090-PCP-100, Spill Response- Oil, Polluting, Hazardous Materials, and Radioactjve Spills 5.2 Writing

References:

5.2.1 Source

References:

a. 10 CFR 20, *Standards for Protection Against Radtation

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

c. PMI-6010, Radiation Protection Plan

d. NUREG-0472

e. NUREG-1301

f. NUREG-0133

Information I PMP-6010-0SD-001 I Rev. 25 I Page 47 of 89 OFF-SITE DOSE CALCULATION MANUAL

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

h. Regulatory Guide 1.111

i. Regulatory Guide 1.113

j. Updated Final Safety Analysis Report (UFSAR)

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

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

m. NUREG-0017

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

ENGR 107-04 8112.1 Environs Rad Monitor System

o. HPPOS-223, Consideration of Measurement Uncertainty When Measuring Radiation Levels Approaching Regulatory Limits

p. 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

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

r. 40 CPR 190, Environmental Radiation Protection Standards for Nuclear Power Operations

s. NRC Commitment 6309 (N94083 dated 11/10/94)

t. NRC Commitment 1151

u. NRC Commitment 1217

v. NRC Commitment 3240

w. NRC Commitment 3850

x. NRC Commitmept 4859

y. NRC Commitment 6442

z. NRC Commitment 3768 aa. DIT-B-00277-00, HVAC Systems Design Flows bb. Regulatory Guide 1.21 cc. Regulatory Guide 4.1

Information I PMP-6010-0SD-001 I Rev. 25 I Page 48 of 89 OFF-SITE DOSE CALCULATION MANUAL dd. 1-2-V3-02-Calc #4, Unit Vent Sample Flow rate for isokinetic particulates and Iodine sampling ee. RPS Nl3.30-1996, Appendix. A Rationale for Methods of Determining Minimum Detectable Amount (MDA) and Minimum Testing Level (MDL ff. DIT-B-01971-00, Dose Factors for Radioactive Particulate Gaseous Effluents Associated with the Child by the Inhalation Pathway gg. DIT-B-01987-00, Ground Plane & Food Dose Factors Pi for Radioiodines and Radioactive Particulate Gaseous Effluents bh. NRC Commitment 1010 ii. NEI 07-07 Groundwater Protection Initiative jj. ANI 07-01 Potential for Unmonitored and Unplanned Off-Site Releases of Radioactive Material 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 eves Hold-Up Tank Directly to Unit Vent, May 1, 1992

• c. AEP Design Information Transmittal on Aux Building Ventilation Systems

d. PMP-4030.EIS.001, Event-Initiated Surveillance Testing

e. Environmental Position Paper, Fe Impact on Release Rates, approved 3/14/00

f. Environmental Position Paper, Methodology Change from Sampling Secondary System Gaseous Effluents for Power Changes Exceeding 15% within 1 hr to Responding to Gaseous Alert Alarms, approved 4/4/00

g. CR 02150078, RRS-1000 efficiency curve usage

b. Environmental Position Paper, Unit Vent Compensatory Sampling, approved 4/14/05

lnfQnnation PMP-6010-0SD-001 I Rev. 25 Page 49 of 89 OFF-SITE DOSE CALCULATION MANUAL Pages:

Attachment 3 .1 Dos~ Factors for Various Pathways 49-52 R.i Dose Factors PATHWAY Nuclide Ground Vegetable ).\feat CowJllilk Goat Milk Inhalation H-3 O.OE+OO 4.0E+o3 3.3E+02 2.4E+03 4.9E+03 l.3E+o3 C-14 O.OE+OO 3.5E+06 5.3E+05 3.2E+06 3.2E+o6 3.6E+o4 Cr-51 5.4E+06 l.1E+Q7 l.5E+06 6.9E+06 8.3E+o5 2.1E+o4 Mn-54 1.6E+09 9.4E+08 2.1E+07 2.9E+07 .*3.5E+06 2.0E+06 Fe-59 t2E+08 9.6E+o8 l.7E+09 3.1E+08 4.0E+o7 l.5E+o6 Co-58 4.4E+o8 6.0E+o8 2.9E+08 8.4E+07 l.OE+o7 1.3E+06 Co~60 2.5E+I0 3,2E+o9 1.dE+09 2.7E+o8 3.2E+o7 8.6E+o6 Zn-65 8.5E+08 2.7E+o9 95E+08 L6E+10 l.9E+o9 1.2E+d6 St-89 2,5E+04 3.5E+10 3.8E+o8 9.9E+09 2.lE+lO 2.4E+o6 Sr-90 .O.OE+oO l.4E+l2 9.6E+09 9.4E+10 2.0E+ll 1.1E+o8 Zr-95 2.9E+08* l.2E+o9 l.5E+09 9.3E+o5 l.1E+o5 2.7E+o6 Sb-124

• 6.9E+08 3.0E+09 4.4E+08 7.2E+o8 8.6E+o7 3.8E+o6 I-131 l.OE+o7 2.4E+10 2.5E+09 4.8E+I1 5.8E+ll 1.6E+o7 I-133 1.5E+o6 4.0E+o8 6.0E+Ol 4.4E+o9 :S.3E+09 3..8E+06 Cs~134

• 7.9E+o9 2.5E+10 l.1E+09 5.0E+lO 1.5E+ll 1.1E+06

• • • ,,.:::,., - . .. Cs-136 1.7E+08 Z..2E+o8 4.2E+07 5.1E+09 l.5E+I0 l.9E:+05 Cs-137 l.2E+10 2,5E+10

• l.OE+09 4.5E+I0 l.4E+ll 9.0E+05 Ba-140 2.3E+07 2.7E+o8 5.2E+07 2.1E+08 2.6E+o7. .. 2~0E+o6 Ce-141 1.5E+07 5.3E+o8. 3.0E+07 8.3E+07 l.OE+o7 6.1E+o5 Ce-144 7.9E+07 l.3E+10 3.6E+o8 7.3B+08 8.7E+o7 1.3E+07

• Uxµts for all ex<;:ept inhalation pathway are ~ 2 mr sec / yr µCi, .inhajatio~ pathway units are mr m3 l yr µCi.

Uap Values to be Used For the Maximum Exposed fudividual 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 1.6 21 Drinking water (L/yr) 330 510 510 730 Shoreline recreation (hr/yr) -- 14 67 12 3

Inhalation (tn. /yr) 1400 3700 8000 8000 Table E-5 of Reg. Guide 1.109.

Information PMP-6010-0SD-001 I Rev. 25 Page 50 of 89 OFF-SITE DOSE CALCULATION MANUAL Pages: .1 Dose Factors for Various Pathways 49-52 Bip Factors for Aquatic Foods pCil I kg pCi Element Fish Invertebrate H 9.0E-1 9.0E-1 C 4.6E3 9.1E3 Na l.OE2 2.0E2 p l.OE5 2.0E4 Cr 2.0E2 2.0E3 Mn 4.0E2 9.0E4 Fe l.OE2 3.2E3 Co 5.0El 2.0E2 Ni l.OE2 l.OE2 Cu 5.0El 4.0E2 Zn 2.0E3 l.OE4 Br 4.2E2 3.3E2 Rb 2.0E3 l.OE3 Sr 3.0El l.OE2 y 2.5El 1.0E3 Zr 3.3EO 6.7EO Nb 3.0E4 1.0E2 Mo 1.0El l.OEl Tc 1.5El 5.0EO Ru l.OEl 3.0E2 Rh l.OEl 3.0E2 Te 4.0E2 6.1E3 I l.5El 5.0EO Cs 2.0E3 l.OE3 Ba 4.0EO 2.0E2 La 2.5El l.OE3 Ce l.OEO l.OE3 Pr 2.5El 1.0E3 Nd 2.5El 1.0E3 w l.2E3 l.OEl Np 1.0El 4.0E2 Table A-1 ofReg.*Guide l.109.

Information PMP-6010-0SD-001 I Rev. 25 Page 51 of 89 OFF-SITE DOSE CALCULATION MANUAL Pages: .1 Dose Factors for Various Pathways 49-52 Drupj External Dose Factors for Standing on Contaminated Ground mrem m2 / hr pCi Radionuclide Total Body Skin H-3 0 0 C-14 0 0 Na-24 2.5E-8 2.9E-8 P-32 0 0 Cr-51 2.2E-10 2.6E-10 Mn-54 5.8E-9 6.8E-9 Mn-56 l.lE-8 1.3E-8 Fe-55 0 0 Fe-59 8.0E-9 9.4E-9 Co-58 7.0E-9 8.2E-9 Co-60 1.7E-8 2.0E-8 Ni-63 0 0 Ni-65 3.7E-9 4.3E-9 Cu-64 l.5E-9 l.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 l.4E-8 Br-85 0 0 Rb-86 6.3E-10 7.2E-10 Rb-88 3.SE-9 4.0E-9 Rb-89 l.5E-8 l.8E-8 Sr-89 5.6E-13 6.SE-13 Sr-91 7.lE-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 l.6E-9 l.9E-9 Y-93 5.7E-10 7.8E-10 Zr-95 5.0E-9 5.8E-9 Zr-97 5.SE-9 6.4E-9 Nb-95 5.lE-9 6.0E-9 Mo-99 l.9E-9 2.2E-9 Tc-99m 9.6E-10 l.lE-9 Tc-101 2.7E-9 3.0E-9 Ru-103 3.6E-9 4.2E-9 Ru-105 4.SE-9 *5.lE-9 Ru-106 l.SE-9 l.8E-9 Ag-llOm 1.8E-8 2.lE-8 Te-125m 3.SE-11 4.8E-ll

Information PMP-6010-0SD-001 I Rev. 25 Page 52 of 89 OFF-SITE DOSE CALCULATION MANUAL Pages:

I Attachment 3 .1 Dose Factors for Various Pathways 49-52 Radionuclide Total Body Skin Te-127m l.lE-12 l.3E-12 Te-127 1.0E-11 l.lE-11 Te-129ni 7.7E-10 9.0E-10 Te-129 7.lE-10 8.4E-10 Te-131m 8.4E-9 9 ..9E-9 Te-131 2.2E-9 2.6E-6 Te-132 l.7E-9 2.0E-9 1-130 1.4E-8 l.7E-8*

1-131 2.8E-9 3.4E-9 1-132 l.7E-8 2.0E-8 1-133 3.7E-9 4.SE-9 I-134 1.6E-8 l.9E-8 l-135 1.2E~8 l.4E-8 Cs-134 1.2E-8 1.4E-8 Cs-136 1.SE-8 l.7E-8 Cs-137 4.2E-9 4.9E-9 Cs-138 2.lE-8 2.4E-8 Ba-139 2.4E-9 2.7E-9 Ba-140 2.lE-9 2.4E-9 Ba-141 4.3E-9 4.9E-9 Ba-142 7.9E-9 9.0E-9 La-140 l.SE-8 1.7E-8 La-142 l.SE-8 l.8E-8 Ce-141 5.SE-10 6.2E-10 Ce-143 2.2E-9 2.SE-9 Ce-144 3.2E-10 3.7E-10 Pr-143 0 0 Pr-144 2.0E-10 2.3E-10 Nd-147 l.OE-9 1.2E-9 W-187 3.lE-9 3.6E-9 Np-239 9.SE-10 l.lE-9 Table E-6 of Reg. Guide 1.109.

Information PMP-6010-0SD-001 I Rev. 25 I Page 53 of 89 OFF-SITE DOSE CALCULATION MANUAL Pages:

Attachment 3.2 Radioactive Liquid Effluent Monitoring Instruments 53 - 55 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 3 System Effluent Line (R-20, R-28) train

3. Continuous Composite Sampler Flow Monitor

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

4. Flow Rate Measurement Devices

a. Liquid Radwaste Line (1) At times of release 4 (RFI-285)

b. Discharge Pipes* (1) At all times NA C. Steam Generator Blowdown (1) At times of release 4 Treatment Effluent (DFI-352)

d. Individual Stearn Generator sample flow ' (1) per At times of release 5 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 tenn 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 1 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 I 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.

Information PMP-6010-0SD-001 I Rev. 25 I Page 54 of 89 OFF-SITE DOSE CALCULATION MANUAL Pages:

Attachment 3 .2 Radioactive Liquid Effluent Monitoring Instruments 53- 55 A IF a.Ii RMS monitor is inoperable solely as the result of the loss of its control room alarm annunciation, TIIEN 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) a:t a limit of detection of at least 10-7 µCi/gram: *

1. At least once per shift when the specific activity of the secondary coolant is > 0.01 µCi/gram DOSE EQUIVALENT I-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 µCi/gram DOSE EQUIVALENT I-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 µCi/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 th(, 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.

Information PMP-6010-0SD-001 I Rev. 25 I Page 55 of 89 OFF-SITE DOSE CALCULATION MANUAL Pages:.

Attachment 3.2 Radioactive Liquid Effluent Monitoring Instruments 53 - 55 Action5 With the number of channels OPERABLE less than required by the Mmi.mum 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, TIIEN 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 des4"ed 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

Information Pl\IP-6010-0SD-001_ I Rev. 25 . Page 56 of 89 OFF-SITE DOSE CALCULATION MANUAL Radioactive Liquid Effluent Monitoring Pages:

Attachment 3)

Instrumentation Surveillance Requirements 56 -57 Instrum,ent CHANNEL SOURCE CHANNEL CHANNEL-CHECK CHECK CALJBRATION OPERATIONAL TEST

1. Gross Radioactivity Monitors Providing Automatic Release Termination

a. Liquid Radwaste D* p B(3) Q(S)

Efflm;nt Line (RRS-1001)

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

Blowdown Effluent Line C. Steam Generator *n* M B(3) Q(l)

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* NIA NIA NIA:.

Sump Effluent Line * ..

4. Flow Rate Measurement Devices

a. Liquid Radwaste D(4)* NIA B Q

-Effluent

b. Steam Generator D(4)* NIA NIA NIA Blowdown Treatment -

Line

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

lrtfonnation PMP-6010-0SD-001 l Rev. 25 Page57 of 89 OFF-SITE DOSE CALCULATION MANUAL Radioactive Liquid Effluent Monitoring Pages:

Attachment 3 .3 Instrumentation Surveillance Requirements 56.-- 57 TABLE NOTATION

1. Demonstrate with the CHANNEL OPERATIONAL TEST that automatic isolation of this pathway and control mom 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 downscii.le failure.*

4. Instrument control not set moperating mode.*

5. Loss of sample flow. *

2. Demonstrate with the CFJANNEL OPERATIONAL TEST that control room alarm annunciation occurs .if any of the following conditions exists: *

1. Instrument indicates measured levels above the a1ann setpoint.

2. Circuit failure.

3. Instrument indicates a downscale failure.

4. Instrumen,t controls not set in operating JIJ.Ode.

3. Perform the initial CHANNEL CALIBRATION using*one or more sourc~ with tr;aceability 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 m:;iy be used.

-4. Verify indication of flow during periods *of release with the CHANNEL CHECK. P~rform the CHANNEL CHECK at least once per :24 hours on days on which continuous, periodic *or batch releases are made.

5. De~onstrate with the CHANNEL OPERATIONAL TEST that automatic isolation of this p,athway 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 inc:lica,tes a downscale failure.**

4; Instrument control not.set in opera.ting mode.*

5. Loss ofs~ple flow,

• Instrument indicates, but does not provide for automatic isolation

• • Instrum(?nt indicates, but does not necessarily cause.automatic isolation. No credit is taken for the automatic isolatiop. on such occµrrences.

Operations currently performs the routine ch3Illlel checks and source checkS. Maintenance and Radiation Protection perform channel calibrations aJ,J.d channel operational tests. Chemistry performs the channel check on the continuous. composite sampler, Th~ responsibilities are subject to change without revision to this document.

Information PMP-6010-0SD-001 I Rev. 25 Page 58 of 89 OFF-SITE DOSE CALCULATION MANUAL Pages:

Attachment 3.4 Radioactive Gaseous Effluent Monitoring Instrumentation 58 -60 Instrument (Instrument #) Operable1 Minimum Action Channels Action

1. Condenser Evacuation System

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

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

OR (SFR-402 and l/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 and 1/2-MR-054) OR (VFR-315 and VFR-1510/2510)

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 (I) ****2,3 7 ERS-1305/1405 (ERS-2305/2405)

b. Containment Particulate Sampler Filter (I) **** 10 ERS-1301/1401 (ERS-2301/2401)

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

a. Noble Gas Activity (1) ****4 9 Alarm and Termination of Waste Gas Releases (VRS-1505/2505)

5. Gland Seal Exhaust

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

b. Flow Rate Monitor (SFR-201 and 1/2-MR- (1) **** 5

54) OR (SFR-201 and 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 associatea'with it like that associated with the two specific tank types listed here.

• • • • During releases via this pathway

. Information PMP-6010-0SD-001 I Rev. 25 Page 59 of 89

, OFF-SITE DOSE CALCULATION MANUAL Pages: -4 Radioactive Gaseous Effluent Moiµtoring Instrumentation

.58 - 60 TABLE NOTATIONS

1. IF an RMS monitor is inoperable solely as the result of the loss ofit's control room alarni annunciation, THEN one of the following actions is acceptable to satisfy the ODCM action statement compensatory surveillance reql).irement:

• l_ 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 ann.unciation,, THEN the only acceptable ac;tion is taking grab samples and conducting laboratory* ~yse~ as the reading is equivalent to a grab sample when the monitor is functional.

2. Consider releases as o~curring "via this pathway" under the following conditions:

• The Containment Purge System is in operation and Contawment Operability is applicable,

-OR- .

• The Containment Purge System is .in operation and the 'Clean-up' batch release of the Containment air volume has not been fully completed.

IF neither of the above are applicable AND .the unit is in Mode 5 or 6, THEN the contmnment purge system is acting as a ventilation system (an extension of the Auxiliary Bgildmg) and is covered by Item 2 of tliis Attachment. This is called 'Ventilation Mode'. 'Ventilate Mode' cannot be entered witho11t performing a Clean-up batch release.

-OR-

• A Containment Pressure Relief (CPR) is being performed.

Once the 'Clean-up' batch release has been completed and 'Ventilation' mode of Purge has commenced -

resultant return to 'Clean-up' mode can be made with no additional saµiplirig requirements or paperwor~ - so long as eit,her ERS-1305/2305 OR ERS~1405/2405 are operable. Containment particulate channels are not needed once the RCS has entered Mode 5 per Technical Specification 3.4.15.

3. For purge (including pressure relief) pm:poses only. Refe:r;ence T.S 3 .3.6, Contai.nment Purge Supply and Exhaust System Isolation Instrumentation and 3:4.15, RCS Leakage Detection lnstrQmentatlon for additional information. *

4. For waste gas releases only, see Item 2 (Unit Vent; Auxiliary Building Ventilation System) for !!-dditional 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 hours and provide a description of why the inoperability was not corrected in the next Annual Radiological Effluent Release Reporj:.

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.

Information

• PMP-6010-0SD-001 I Rev. 25 Pa2e 60 of 89 OFF-SITE DOSE CALCULATION MANUAL Pages:

Attachment 3 .4 Radioactive Gaseous Effluent Monitoring Instrumentation

. 58 - 60

7. With the number of channels OPERABLE less than required by the Minimun:i Channels OPERABLE requirements, imme(iiately suspend PURGING or VENTING (CPR) of radioactive ~ffluents via this pathway.

8. Willi. the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via the affected pathway may, continue f9r up to 30 days* provided samples required for weekly lodme & Particulates analysis are continuously collected with auxiliary sampling equl.pment as required in Attachment 3.7, Radioactive Gaseous Waste Sampling and Analysis Program. After 30 days, IF tp.e channels are not OPERABLE, THEN continue releases with sample collection by auxiliary sampling equipment and provide a description of why .the inoperabilit:y was not corrected in the next Annual Radiological Effluent Release Report. ..

Sampling evolutions are llOt an inJerruption 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 con~nts are analyzed anci,

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

After 14 days, IF the channels are not OPERABLE, THEN cop.tinue releases with sample collection by auxiliary sampling equipment and provide a description of why the inoperabilit:y was not corrected in the next Annual Radiological E~rient Release Report.

iO, Technical Specification 3-4.15, RCS Leakage Detection System Instrumentation.

Compensatory actions are.governed by PMP-4030-EIS-OOl, Event-lnitiat:(l Surveillance Testing.

Information PMP-6010-0SD-001 I Rev. 25 Page 61 of 89 OFF-SITE DOSE CALCULATION MANUAL Radioactive Gaseous Effluent Monitoring Pages:

Attachment 3.5 Instrumentation Surveillance Requirements 61 - 62 Instrument CHANNEL SOURCE I CHANNEL CHANNEL OPERATIONAL CHECK CHECK CALIBRATION TEST

1. Condenser Evacuation Alarm.Only System

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

(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(l)

(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* NIA 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 eves 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(l)

(SRA-1805/2805)

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

• At all times

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

Information PMP-6010-0SD-001 I Rev. 25 Page 62 .of 89 OFF-SITE DOSE CALCULATION MANUAL Radioactive Gaseous Effluent Monitoring Pages:

Attachment 3.5 Instrumentation Surveillance Requirements 61 -62 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.

Information PMP-6010-0SD-001 I Rev. 25 I Page 63 of89 OFF-SITE DOSE CALCULATION MANUAL Pages:

Attachment 3.6 Radioactive Liquid Waste Sampling and Analysis Program 63 -64

[Ref. 5.2.ltJ LIQUID SAMPLING MINIMUM TYPE OF LOWER RELEASE FREQUENCY ANALYSIS ACTIVITY LIMITOF TYPE FREQUENCY ANALYSIS DETECTION (LLD)

(µCi/ml) a A. Batch Waste p p Principal 5x10*1 Release Tanks c Each Batch Each Batch Gamma Emitters e I-131 lxl0*6 p p Dissolved and

- Entrained Gases Each Batch Each Batch (Gamma lx10*5 Emitters) p M lx1Q*5 H-3 Each Batch Compositeb Gross Alpha lx10*7 p Q Sr-89, Sr-90 5xl0*8 Each Batch Compositeb Fe-55 lxl0*6 B. Plant w Principal Continuous Daily Compositeb Gamma 5x10*7

Releases* d Emitterse I-131 lxl0*6 M M Dissolved and Grab Sample Entrained Gases 1x10*5 (Gamma Emitters)

M H-3 1x10*5 Daily Compositeb Gross Alpha 1x10*1

• Q Sr-89, Sr-90 5xl0*8 Daily Compositeb Fe-55 lxl0*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 Taruc tritium samples.

Information PMP-6010-0SD-001 I Rev. 25 I Page 64 of 89 OFF-SITE DOSE CALCULATION MANUAL Pages: .6 Radioactive Liquid Waste Sampling and Analysis Program 63 - 6A TABLE NOTATION

a. The lower limit of detection (LLD) is defined in Table Notation A. of Attachment 3.20, Maximum Values for Lower Limits ofDetectionsA,B _ 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 Tanlc 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.

Information PMP-6010-0SD-001 I Rev. 25 Page 65 of 89 OFF-SITE DOSE CALCULATION MANUAL R,adioactive Gaseous Waste Sampling and Pages:

Attachment 3.7 Analysis Program 65 -66 Gaseous Release Type Frequency Minimum Type of Lower Limit Analysis Activity of Detection Frequency Analysis (µCi/cc) a

a. Waste Gas Storage p p Principal Gamma Tanks and CVCS HUTs Each Tank Each Tank Emittersd 1 X 104 Grab Sample H-3 1 X 10-6

b. Containment Purge p p Principal Gamma Each Purge Each Purge Emitters d 1 X 104 Grab Sample CPR (vent)** Twice per Twice per Month Month H-3 1 X 10-6

c. Condenser Evacuation WorM M Principal Gamma System Grab Sample Particulate Sample Emittersd 1 X 10"11 Gland Seal Exhaust* ; 1 X 10-6 M H-3 wg Principle Gamma 1 X 104 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 0 wb 1-131 Vent* Iodine Adsorbing 1 X 10"12 Media Continuous c wb. Principal Gamma Particulate Sample Emittersd 1 X 10-ll Continuous 0 M Gross Alpha Composite Particulate 1 X 10"11 Sample w wh H-3 Grab Sample H-3 Sample 1 X 10-6 wgj Principle Gamma 1 X 104 Noble Gas Emitters d Continuous c Q Sr-89, Sr-90 Composite Particulate l X 10"11 Sample Continuous c Noble Gas Monitor Noble Gases 1 X 10-6

e. Incinerated Oil 0 p p Principal Gamma Each Batchr Each Batchr Emittersd 5 X 10*7

• 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.

Information PMP-6010-0SD-001 I Rev. 25 Page 66 of 89 OFF-SITE DOSE CALCULATION MANUAL Radioactive Gaseous Waste Sampling and Pages:

Attachment 3.7 Analysis Program 65 -66 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 !131 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 Il31 levels have returned to within the factor-of 3 of the pre-event 'normal' .[Ref. 5.2.lz]

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 or momentary interruptions to maintain sampling capability 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 andXe-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 contajner 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 IE-6 µC/cc. Attach the RMS daily averages in lieu of sampling. This is based on operating experience indicating no activity is detected in the Gland Seal Exhaust below this value. Compensatory sampling for out of service monitor is still required in the event 1805/2805 is inoperable.

j. Sampling and analysis shall also be perfonned following shutdown, startup or THERMAL POWER change exceeding 15% of RATED THERMAL POWER within a one hour period. This noble gas sampl~ 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.

Information PMP-6010-0SD-001 I Rev. 25 Page 67 of 89 OFF-SITE DOSE CALCULATION MANUAL Page:

Attachment 3.8 Multiple Release Point Factors for Release Points*

67 Liquid Factors Monitor Description Monitor Number MRP#

U 1 SO Blowdown 1R19/24, DRS 3100/3200* 0.35 U 2 SO 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 not respond 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 Unit2 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 I monitors.

1. Nominal Values a Two release points of2,754 cfin 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.

Information PMP-6010-0SD-001 I Rev. 25 Page 68 of 89 OFF-SITE DOSE CALCULATION MANUAL Page:

Attachment 3.9 Liquid Effluent Release Systems 68 RB.E:ASE SOURCES SYSTBulS PONfS DirtyWJSles.:

AoorDralns,, StatlonDfnin(Dirty) CJeansump Decontarr.inatton Sump Tan!< Tank RinsoS0lutlon9,,

Chemical Drain Tanl(,Etc. pump I a. H Wiste_~~:;~ Tank Clean WIS.es Equipment Drains, PumpSeal Leakolfs,.

Containment Fan Circulatlng pwnp Caoler LaundryHotSho\lo'er I l":'.'u 1/ihter C<JndenS1te,ac. I T:mks(2) ~ Discharge C\CS Boricacld Evaporator j BoriccAcidBiaporation ~ &aporatorCondensate MonitcrTanks Packagcs,North t Oeminaralizer I and South Steam StcamGeneralot Generator Bfowdovmand Slowdown Treatment Sy!tom (Potential)

I SaeenHouse I Circulating

\JI.bier Discharge Es:Entla1Servicc

\11.bterSystem (Potential)

Isample Point ClrculatingW:ater ln!akePipes 1!::!!.!!!.E.I r:----, I h Containment Spray l ~ a d f a t l o o g:i-----,rri_ HeatExchongers. I CirctJlating Vihtcr Dia::harge Turbine Room SumpUnil1and2 TurbineRoomSump 1--,--t I ~1 AowMctcr 1-~----~--------+-tl (Potential)

B.571'Sump Circulating V1,Mer Di!Charge

Information PMP-6010-0SD-001 I Rev. 25 Page 69 of 89 OFF-SITE DOSE CALCULATION MANUAL Page:

Attachment 3 .10 Plant Liquid Effluent Parameters 69 SYSTEM COMPONENTS CAPACITY FLOW RATE TANKS I PUMPS (EACH) (EACH)*

I Waste Disposal System

+ Chemical Drain Tanlc 1 1 600 GAL. 20 GPM

+ Laundry & Hot Shower Tanlcs 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. 580 GPM

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

+ Blowdown Treatment System 1 60 GPM III Essential Service Water System

+ Water Pumps 4 10,000 GPM

+ Containment Spray Heat 4 3,300 GPM Exchanger Outlet IV Circulating Water Pumps I Unit I 3 230,000 GPM Unit2 4 230,000GPM

• Nominal Values

1. 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.

Information PMP-6010-0SD-001 I Rev. 25 I Page 70 of 89 OFF-SITE DOSE CALCULATION MANUAL Volumetric Detection Efficiencies for Principle Gamma Page: .11 Emitting Radionuclides for Eberline Liquid Monitors 70 This includes the following monitors: RRS-1000, DRS 3100, DRS 3200, DRS 4100, and DRS 4200.

[Ref. 5.2.lq)

NUCLIDE EFFICIENCY (cpm/uCi/cc) 1-131 3.78 B7 Cs-137 3.00B7 Cs-134 7.93B7 Co-60 5.75 B7 Co-58 4.58 B7 Cr-51 3.60 B6 Mn-54 3.30B7 Zn-65 1.58 B7 Ag-llOM 9.93 E7 Ba-133 4.85 B7 Ba-140 l.92B7 Cd-109 9.58 E5 Ce-139 3.28 E7 Ce-141 1.92 B8 Ce-144 4.83 E6 Co-57 3.80E7 Cs-136 1.07B8 Fe-59 2.83 E7 Sb-124 5.93 E7 1-133 3.40 E7 I-134 7.23 E7 I-135 3.95 B7 Mo-99 8.68 E6 Na-24 4.45 E7 Nb-95 3.28 E7 Nb-97 3.50E7 Rb-89 5.00E7 Ru-103 3.48 E7 Ru-106 1.23 E7 Sb:-122 2.55 E7 Sb-125 3.15 E7 Sn-113 7.33 ES Sr-85 3.70E7 Sr-89 2.88 E3 Sr-92 3.67E7 Tc-99M 3.60 E7 Y-88 5.25 E7 Zr-95 3.38 E7 Zr-97 3.10E7 Kr-85 1.56 ES Kr-85M 3.53 E7 Kr-88 4.10E7 Xe-131M 8.15 E5 Xe-133 7.78E6 Xe-133M 5.75 E6 Xe-135 3.83E7

Information PMP-6010-0SD-001 I Rev. 25 Page 71 of 89 OFF-SITE DOSE CALCULATION MANUAL Pages: .12 Counting Efficiency Curves for R-19, and R-24 71 -72 Counting Efficiency Cunre for R-19 Efficiency Factor= 4.2 E6 cpm/uCi/ml (Based on empirical data taken during pre-<Jperational tcs ting with Cs-137)

'C C:

e J2 1.00E+04 +----------------------:;;,,'""'------------------!

il"

~

,g 1.00E+03 +-----------------'--------------------------4

a; n.

tJ rD 0 0 0 LL! +

0 UJ 0

~ ~

rnicrocuries/ml

Information PMP-6010-0SD-001 I Rev. 25 Page 72 of 89 OFF-SITE DOSE CALCULATION MANUAL Pages: .12 Counting Efficiency Curves for R-19, and R-24 71- 72 Counting Efficiency Curve for R-24 Efficiency Factor= 7.SE6 cpm/uCi/ml (Based oli empirical data taken during pre-operational testing with Mn-54)

"t, 1.00E+05 +--------------------------==-.-,::::.--------------1 C

e f 1.00E+04 -l:---------------------=_,,,.,.~-----------------1 B

Cl)

,8 1.00E+03 +-------------.....""'----------------------------l

ii Q.

(.) 1.00E+02 -l,---------=_,,,.,.~------------------------------1 CD It) v 0

(')

0 ~ 0 0 0 0 0 w

0 w

0 w

0 w

0 w

0 w

0 it0 q q q

_q

.... q

....q ....q mlcrocurles/ml

Information PMP-6010-0SD-001 I Rev. 25 I Page 73 of 89 OFF-SITE DOSE CALCULATION MANUAL Page: .13 Counting Efficiency Curve for R-20, and R-28 73 Counting Efficiency Curve for R-20 and R-28 Efficiency Factor= 4.3 E6 cpm/uCi/ml (Based on empirical data taken during pre"{)pcrational testing with Co-58)

N 9

UJ 9

UJ sl UJ 9

UJ ~

0 0 0 0 0 q q q q

~ ~

mlcrocurles/ml

Information PMP-6010-0SD-001 I Rev. 25 Page 74 of 89 OFF-SITE DOSE CALCULATION MANUAL

• Page:

Attachment 3 .14 Gaseous Effluent Release Systems 74 RB...EASE SOURCES SYSTBv1S R'.JINTS WltteGasDecay jvent Header

~a;s and C\.eS r-.

Moisture Se rator AU"t.. Building Vent ,---::--C'.'C'--, __ ..I Pm r*

Ventilation Filler Engineered Sar ety Feat uresVent ~

System Engineered ~ Pro HEPA >

SafetyFcaturcs VenlandPipo Filter Filter 1§ Enclosures s

J 0

I-E

'Li~~ ~  :>--------------

Fuel Hand[ing Ventilation 2 Q)

"" =: 1 Damper I I a::

Containment Purge and Relier sys1:em

. /~

arbon ~-----------~

Filter~

c::;;;;;;:J lnstrumerit Room Purge System HEPA CarbOn

~----~ ~

Filter Filter 1-.....J~,;---~~---;::=========:;-======~ Alrbornelo,o,flfConlnlnmenlRftdlation UpperConlaln1J1CJl\Are:i 1Jonl1or11, Theselr.ola\ete!\Salnnrn.t

~----~

~---------------='  :::S::om':p:::Ji=:n::gp::o::in:::l~L-H lnc1~:~~1::.: ~: n; :~~:: :dhl~h ahirm.

RadlD1lonND1lllor. Thi&

lscleteacon1alnm:.?nl puroe.ctnt.rdl"l,nnd lns11wnentrOG111

~-----------------------...'.:=========::....jo.du1v11l~o11hlch!l'J11r,n,,

Steam Generator Slowdown

~

~

. ~--------------------------------~

~

Trcatn1ent System ToTreatment Sy.stem no gaseousrelease Condenser Air EjectorSySl:enl I FlowData I r ------------, I Samo1incPolnt I _~

5

_'-'~"':"'c:c :c :s..Al-r__,l~~-1'~-*---------*--------------------,Q Gland Seal candcnser Exhaust ISS"-tea;;;;;:m;;;P;;;a;;;ck;;:in;;;g;l----------- ~

Eiehauster __

I semplloqPoJnt l~c,nl I~--r~-----~-,-----~

AowData I R a d i a t i o n ~ - - - - - - - - - - - - - - - - < atmo d~:. hero

Information PMP-6010-0SD-001 I Rev. 25 Page 75 of 89 OFF-SITE DOSE CALCULATION MANUAL Page:

Attachment 3.15 Plant Gaseous Effluent Parameters 75 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 FT3 @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 Umt NORMAL STEAM JET AIR 1&2 230 EJECTORS

• START UP STEAM JET AIR 1&2 3,600 EJECTORS Ill TURBINE SEALS SYSTEM 1 1,260 2 5,508 2 Release Points for Unit 2 START UP FLASH TANK VENT 1 1,536 2 1,536

+ Designates total flow for all fans.

Information PMP-6010-0SD-001 I Rev. 25 Page 76 of 89 OFF-SITE DOSE CALCULATION MANUAL Attachment 3.16 10 Year Average of 1995-2004 Data Pages: 76-77 x/Q GROUND AVERAGE (sec/m3)

DIRECTION DISTANCE (METERS)

(WINDFROM) 594 2416 4020 5630 7240 N 4.17E-06 4.82E-07 2.25E-07 l.33E-07 9.32E-08 N_NE 3.02E-06 3.64E-07 l.73E-07 l.04E-07 7.29E-08 NE 4.54E-06 5.31E-07 2.60E-07 l.59E-07 1.13E-07 ENE 7.16E-06 7.99E-07 4.04E-07 2.52E-07 l.80E-07 E l.04E-05 l.13E-06 5.82E-07 3.66E-07 2.63E-07 ESE 1.07E-05 l.18E-06 6.04E-07 3.78E-Q7 2.72E-07 SE 1.15E-05 l.24E-06 6.36E-07 4.00E-07 2.88E-07 SSE 1.3DE-05 l.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 l.72E-07 SW 4.12E-06 4.73E-07 2.28E-07 l.38E-07 9.73E-08 WSW 3.29E-06 3.65E-07 l.76E-07 1.06E-07 7.52E-08 w 3.63E-06 4.llE-07 1.96E-07 l.18E-07 8.31E-08 WNW 3.02E-06 3.43E-07 1.61E-07 9.59E-08 6.71E-08 NW 3.22E-06 3.61E-07 1.71E-07 1.02E-07 7 .. 16E-08 NNW 3.84E-06 4.29E-07 2.02E-07 l.20E-07 8.40E-08 DIRECTION DISTANCE (METERS)

(WINDFROM) 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 l.15E-08 7.41E-09 4.72E-09 ENE 9.30E-08 3.80E-08 l.91E-08 l.23E-08 7.90E-09 E l.37E-D7 5.65E-08 2.85E-08 1.83E-08 l.18E-08 ESE l.41B-07 5.81E-08 2.93E-08 1.88E-08 l.22E-08 SE l.50E-07 6.2DE-08 3.12E-08 2.0lE-08 l.30E-08 SSE l.71E-07 7.06E-08 3.56E-08 2.29E-08 l.48E-08 s 1.84E-07 7.49E-08 3.77E-08 2.43E-08 l.56E-08 SSW 8.86E-08 3.59E-08 l.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 l.51E-08 7 ..53E-09 4.83E-09 3.07E-09 w 4.17E-08 l.64E-08 8.13E-09 5.20E-09 3.28E-09 WNW 3.34E-08 l.29E-08 6.41E-09 4.lOE-09 2.57E-09 NW 3.57E-08 l.39E-08 6.89E-09 4.41E-09 2.77E-09 NNW 4.19E-08 3.35E-08 8.lOE-09 5.19E-09 3.27E-09 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 x/Q = 2.04E-05 sec/m3 in Sector H 2004

Information PMP-6010-0SD-001 I Rev. 25 Page 77 of 89 OFF-SITE DOSE CALCULATION MANUAL Attachment 3 .16 10 Year Average of 1995-2004 Data Pa,ges: 76-77 D/Q DEPOSITION (1/m2)

DIRECTION DISTANCE (METERS)

(WINDFROM) 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 l.29E-08 1.25E-09 5.67E-10 2.97E-10 1.90E-10 ENE l.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 l.85E-08 1.79E-09 8.12E-10 4.26E-10 2.72E-10 SE l.90E-08 1.83E-09 8.30E-10 4.36E-10 2.78E-10 SSE 2.40E-08 2.32E-09 1.0SE-09 5.52E-10 3.52p-10 s 3.68E-08 3.56E-09 1.61E.-09 8.46E-10 5.40E-10 SSW 2.30E-08 2.22E-09 1.0lE-09 5.28E-10 3.37E-10 SW 2.22E-08 2.lSE-09 9.74E-10 5.llE-10 3.26E-10 WSW 2.llE-08 2.04E-09 9.23E-10 4.84E-10 3.09E-10 w 2.00E-08 l.93E-09 8.74E-10 4.59E-10 2.93E-10 WNW 1.75E-08 1.69E-09 7.64E-10 4.0lE-10 2.56E-10 NW 1.58E-08 l.53E-09 6.94E-10 3.64E-10 2.32E-10 NNW 2.30E-08 2.22E-09 l.OlE-09 5.28E-10 3.37E-10 DIRECTION DISTANCE (METERS)

(WIND FROM) 12067 24135 40225 56315 80500

.N l.45E-10 4.72E-ll l.74E-ll 9.27E-12 4.65E-12 NNE 6.36E-ll 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-ll l.17E-ll 6.23E-12 3.13E-12 E 1.14E-10. 3.73E-11 1.37E-11 7.34E-12 3.68E-12 ESE l.13E-10 3.70E-11 1.36E-ll 7.26E-12 3.64E-12 SE 1.16E-10 3.78E-11 1.39E-11 7.42E-12 3.72E-12 SSE l.47E-10 4.79E-ll 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 L41E-10 4.59E-11 1.69E-ll 9.0lE-12 4.52E-12 SW l.36E-10 4.43E-ll 1.63E-11 8.71E-12 4.. 37E-12 WSW 1.29E-10 4.20E-11 1.SSE-11 8.26E-12 4.14E-12 w l.22E-10 3.98E-11 1.47E-11 7.82E-12 3.92E-12 WNW 1.07E-lb 3.48E-ll 1.28E-11 6.84E-12 3.43E-12 NW 9.70E-11 3.16E-ll l.16E-11 6.20E-12 3.llE-12 NNW l.41E-10 4.58E-ll 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 l/m2 in Sector A 2001

Information PMP'-6010-0SD-001 I Rev. 2S Page 78 of 89 OFF-SITE DOSE CALCULATION MANUAL Attachment 3 .17 Annual Evaluation of .x/Qand b/Q Values For Page:

All Sect9rs 78

1. Performed or received annual update of x/Q and D/Q values. Pro~ide a description of what has been received.

I Signature Date Environmental Department (print name, title)*

2. Worst ;c/Q and b/Q value and se.ctor determined. PMP-6010-0SD-001 has been updaJed, if necessrlfY, Provide ill evaluation.

I 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.

I Signature Date Environmental Department (print name, title)

4. Approved and verified_ by:_ .

I Signature Date Environmental Department (print name, title)

Information PMP-6010-0SD-001 I Rev. 25 Page 79 of 89 OFF-SITE DOSE CALCULATION MANUAL Pages:

. Attachment 3.18 Dose Factors 79- 80 DOSE FACTORS FOR NOBLE GASES AND DAUGHTERS*

TOTAL BODY SKIN DOSE GAMMA AIR BETA AIR DOSE FACTOR FACTOR DOSE FACTOR DOSE FACTOR Ki (DFB;) Li (DFSi) M; (DF71) Ni (DFl\)

mremm3 (mrem m3 (mradm3 (mradm3 RADIONUCLIDE per µCi yr) per µCi yr) per µCi yr) per µCi yr)

Kr-83m 7.56E-02 - - - l.93E+Ol 2.88E+02 Kr-85m l.17E+03 l.46E+03 l.23E+03 1.97E+03 Kr-85 l.61E+Ol l.34E+03 l.72E+Ol 1.95E+03 Kr-87 5.92E+03 9.73E+03 6.17E+03 l.03E+04 Kr-88 1.47E+04 2.37E+03 l.52E+04 2.93E+03

- Kr-89 1.66E+04 l.01E+04 l.73E+04 l.06E+04 Kr-90 l.56E+04 7.29E+03 l.63E+04 7.83E+03 Xe-131m 9.15E+Ol 4.76E+02 l.56E+02 l.11E+03 Xe-133m 2.51E+02 9.94E+02 3.27E+02 - l.48E+03 Xe-133 2.94E+02 3.06E+02 3.53E+02 l.05E+03 Xe-135m 3.12E+03 7.11E+02 3.36E+03 7.39E+02 Xe-135 l.81E+03 l.86E+03 l.92E+03 2.46E+03 Xe-137 l.42E+03 l.22E+04 l.51E+03 l.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 Reg. Guide 1.109, Table B-1.

Information PMP-6010-0SD-001 I Rev. 25 Page 80 of 89 OFF-SITE DOSE CALCULATION MANUAL Pages: .18 Dose Factors 79- 80 DOSE FACTORS FOR RADIOIODINES AND RADIOACTIVE PARTICULATE,

. JN GASEOUS EFFLUENTS FOR CIDLD* Ref. 5.2.leeandff P1 P1 INHALATION FOOD & GROUND PATHWAY PATHWAY (mremm3 (mremm2 sec RADIONUCLIDE per µCi yr) per µCi yr)

H-3 l.12E+03 l.57E+03 u P-32 2.60E+06 7.76E+10 Cr-51 l.70E+04 l.20E+07 Mn-54 l.58E+06 l.12E+09 Fe-59 l.27E+06 5.92E+08 Co-58 l.11E+06 5.97E+08 Co-60 7.07E+06 4.63E+09 Zn-65 9.95E+05 l.17E+10 Rb-86 l.98E+05 8.78E+09 Sr-89 2.16E+06 6.62E+09 Sr-90 l.01E+08 l.12E+ll 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 l.55E+08 Ru-106 l.43E+07 3.01E+08 Ag-llOrn 5.48E+06 l.99E+10 I-131 l.62E+07 4.34E+ll I-132 l.94E+05 l.78E+06 I-133 3.85E+06 3.95E+09 I-135 7.92E+05 l.22E+07 Cs-134 l.01E+06 4.00E+lO Cs-136 1.71E+05 3.00E+09 Cs-137 9.07E+05 3.34E+10 Ba-140 l.74E+D6 1.46E+08 Ce-141 5.44E+05 3.31E+07 Ce-144 l.20E+07 1.91E+08

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

1. The u~ts for both H3 factors are the same, mrem m' per µCi yr

Information PMP-6010-0SD-001 I Rev. 25 Page 81 of 89 OFF-SITE DOSE CALCULATION MANUAL Radiological Environmental Monitoring Program Pages:

Attachment 3.19 Sample Stations, Sample Types, Sample Frequencies 81- 84 e . . . w, 521

[Rf521 .u

.IV, 521]

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@ 18° from Plant Axis Airborne Particulate Weekly Gross Beta Weekly Gamma Isotopic Quart. Comp.

Airborne I-131 Weekly Radioiodine TLD Quarterly Direct Radiation Quarterly ONS-2 (T-2) 2338 ft@ 48° from Plant Axis Airborne Particulate Weekly Gross Beta Weekly Gamma Isotopic Quart. Comp.

Airborne I-131 Weekly Radioiodine TLD Quarterlv Direct Radiation Quarterly ONS-3 (T-3) 2407 ft @ 90° from Plant Axis Airborne Particulate Weekly Gross Beta Weekly Gamma Isotopic Quart. Comp.

Airborne I-131 Weekly Radioiodine TLD Quarterly Direct Radiation Quarterly ONS-4 (T-4) 1852 ft. @ 118° from Plant Axis Airborne Particulate Weekly Gross Beta Weekly Gamma Isotopic Quart. Comp.

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

Airborne I-131 Weekly Radioiodine TLD Quarterly Direct Radiation Quarterly T-7 2103 ft @ 36° from Plant Axis TLD Quarterly Direct Radiation Quarterly T-8 2208 ft @ 82° 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@ 11° from Plant Axis TLD Quarterly Direct Radiation Quarterly T-12 2292 ft@ 63° 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 N~w Buffalo, MI Gamma Isotopic Quart. Comp.

Airborne Radioiodine I-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 I-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 I-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 I-131 Weekly TLD Quarterly Direct Radiation Quarterly L

Information PMP-6010-0SD-001 I Rev. 25 Page 82 of 89 OFF-SITE DOSE CALCULATION MANUAL Radiological Environmental Monitoring Program Pages:

Attachment 3.19 Sample Stations, Sample Types, Sample Frequencies 81- 84 SAMPLE DESCRIPTION/ SAMPLE SAMPLE ANALYSIS ANALYSIS STATION LOCATION TYPE FREQUENCY TYPE FREQUENCY OFF-SITE DIRECT RADIATION (TLD) STATIONS OFf-1 4.5 miles NE, Pole #B294-44 TLD Quarterly Direct Radiation Quarterlv OFr-2 3.6 miles, NE, Stevensville TLD Quarterly Dire ct Radiation Quarterly Substation OFr-3 5.1 miles NE, Pole #B296-13 TLD Quarterly Direct Radiation Quarterly OFr-4 4.1 miles, E, Pole #B350-72 TLD Quarterly Direct Radiation Quarterly OFr-5 4.2 miles ESE, Pole #B387-32 TLD Quarterly Direct Radiation Quarterlv OFr-6 4.9 miles SE, Pole #B426-l TLD Quarterly Direct Radiation Quarterly OFf-7 2.5 miles S, Bridgman Substation TLD Quarterly Direct Radiation Quarterly OFr-8 4.0 miles S, Pole #B424-20 TLD Quarterly Direct Radiation Quarterly OFf-9 4.4 miles ESE, Pole #B369-214 TLD Quarterly Direct Radiation Quarterly OFr-10 3.8 miles S, Pole #B422-99 TLD Quarterly Direct Radiation Quarterlv OFr-11 3.8 miles S, Pole #B423-12 TLD Quarterly Direct Radiation Quarterly GROUNDWATER (WELL WATER) SAMPLE STATIONS W-1 1969 ft@ 11° from Plant Axis Groundwater Quarterly Gamma Isotopic Quarterly Tritium Quarterly W-2 2302 ft@ 63° from Plant Axis Groundwater Quarterly Gamma Isotopic Quarterly Tritium Quarterly W-3 3279 ft @ 107° from Plant Axis Groundwater Quarterly Gamma Isotopic Quarterly Tritium Quarterlv W-4 418 ft@ 301 ° 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@ 273° from Plant Axis Groundwater Quarterly Gamma Isotopic Quarterly Tritium Quarterly W-7 1895 ft @ 189° from Plant Axis Groundwater Quarterly Gamma Isotopic Quarterly Tritium Quarterlv W-8 1274 ft@ 54° from Plant Axis Groundwater Quarterly Gamma Isotopic Quarterly Tritium Quarterly W-9 1447 ft@ 22° from Plant Axis Groundwater Quarterly Gamma Isotopic Quarterly Tritium Quarterly W-10 4216 ft@ 129° from Plant Axis Groundwater Quarterly Gamma Isotopic Quarterly Tritium Quarterly W-11 3206 ft @ 153° from Plant Axis Groundwater Quarterly Gamma Isotopic Quarterly Tritium Quarterly W-12 2631 ft@ 162° from Plant Axis Groundwater Quarterly Gamma Isotopic Quarterly Tritium Quarterlv W-13 2152 ft@ 182° from Plant Axis Groundwater Quarterly Gamma Isotopic Quarterly Tritium Quarterlv W-14 1780 ft@ 164 ° from Plant Axis Groundwater Quarterly Gamma Isotopic Quarterly Tritium Quarterly W-15 725 ft @ 202° from Plant Axis Groundwater Quarterly Gamma Isotopic Quarterly NPDES well MW-12C Tritium Quarterly W-16 2200 ft @ 208° from Plant Axis Groundwater Quarterly Gamma Isotopic Quarterly NPDES well MW-20 Tritium Quarterly W-17 2200 ft @ 180° from Plant Axis Groundwater Quarterly Gamma Isotopic Quarterly NPDES well MW-21 Tritium Quarterly

Information PMP-6010-0SD-001 I Rev. 25 Page 83 of 89 OFF-SITE DOSE CALCULATION MANUAL Radiological Environmental Monitoring Program Pages:

Attachment 3.19 Sample Stations, Sample Types, Sample Frequencies 81- 84 DRINKING WATER STJ St. Joseph Public Intake Sta. Drinking water Once per calendar Gross Beta 14 day Comp.

9mi.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.

0.6mi. S Day Gamma Isotopic 14 dav Comp.

1-131 14dav Comp.

Tritium Quart.Como SAMPLE DESCRIPTION/ SAMPLE SAMPLE ANALYSIS ANALYSIS STATION LOCATION TYPE FREQUENCY TYPE FREQUENCY SURFACE WATER SWL-2 Plant Site Boundary - South Surface Water Once per calendar Gamma Isotopic Month. Comp.

- 500 ft. south of Plant Day Tritium Quart. Comp Centerline SWL-3 Plant Site Boundary - North Surface Water Once per calendar Gamma Isotooic Month. Comp.

- 500 ft. north of Plant Day Tritium Quart. Comp.

Centerline SEDIMENT SL-2 Plant Site Boundary - South Sediment Semi-Ann. Gamma Isotopic Semi-Annual

- 500 ft. south of Plant Centerline SL-3 Plant Site Boundary - North Sediment Semi-Ann. Gamma Isotopic Semi-Annual

- 500 ft. north of Plant Centerline INGESTION - MILK Indicator Farms Mille Once every 1-131 per sample 15 days Gamma Isotopic oer samole Mille Once every 1-131 per sample 15 days Gamma Isotooic oer samole Mille Once every I-131 per sample 15 days Gamma Isotopic oer samole INGESTION - MILK Background Farm*

I 1-131 I per sample I I Mille IOnceevery 15 days I Gamma Isotopic I 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 oortion 2/vear Gamma lsotooic oer 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 Michi"'m Fish - edible portion 2/year Gamma Isotopic per sample

Information PMP-6010-0SD-001 I Rev. 25 Page 84 of 89 OFF-SITE DOSE CALCULATION MANUAL Radiological Environmental Monitoring Program Pages:

Attachment 3.19 Sample Stations, Sample Types, Sample Frequencies 81- 84 INGESTION - FOOD PRODUCTS On Site ONS-G Nearest sample to Plant in the Grapes At time of Gamma Isotopic At time of highest D/Q land sector harvest harvest containing media.

ONS-V Broadleaf At time of Gamma Isotopic At time of vegetation harvest harvest Off Site OFS-G In a land sector containing Grapes At time of Gamma Isotopic At time of grapes, approximately 20 miles harvest Harvest from the plant, in one of the less prevalent D/Q land sectors OFS-V Broadleaf At time of Gamma Isotopic At time of vegetation harvest harvest INGESTION - BROADLEAF IN LIEU OF GARDEN CENSUS OR IN LIEU OF MILK (*)

3 samples of different kinds of broad leaf vegetation Broadleaf Monthly Gamma Isotopic Monthly collected at the site bonndary, within five vegetation when available Il31 when available miles of the plant, in each of 2 different sectors with the highest annual average D/Q containing media 1 background sample of similar vegetation Broad.leaf Monthly Gamma Isotopic Monthly grown 10-20 miles distant in one of vegetation when available Il31 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

~mes 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 broad leaf samples of different kinds will be collected in each of 2 different offsite locations, within five miles of the plant, with the highest D/Q (refers to the highest annual average ground D/QJ. Also, one background broad leaf sample of similar kinds will be collected 10 to 20 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 deterp:tlned that the milk animals are fed stored feed, THEN monthly sampling is appropriate for that time period.

Evaluate samples that identified positive plant effluent related radionuclides and determine if additional analysis are necessary to identify hard to detect radionuclides. The 10 CFR 61 scaling factor report should be consulted along with the radioactive material shipping program owner and the ODCM program owner to assist with this determination.

L

Information PMP-6010-0SD-001 I Rev. 25 I Page 85 of 89 OFF-SITE DOSE CALCULATION MANUAL P;:tges:

Attachment 3.20 Maximum Values for Lower Limits of DetectionsA,B - REMP 85 - 86

[Ref. 5.2.lw]

Radionuclides Food Product Water Milk Air Filter Fish Sediment pCi/kg, wet pCi/1 pCi/1 pCi/m3 pCi/kg, wet pCi/kg, dcy 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 I-131 60 1 l 0.07 This.Data is directly from 'our plant-specific Technical Specification.

i Information PMP-6010-0SD-001 l Rev. 25 I Page 86 of 89 OFF-SITE DOSE CALCULATION MANUAL Pages:

Attachment 3.20 Maximum Values for Lower Limits ofDetectionsA,B -REMP 85 - 86 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* /-J*M)

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

').. is the radioactive decay constant for the particular radionuclide

.M 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 ofDetectionsA,B _ REMP. *

• a 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.

Information PMP-6010-0SD-001 .I Rev. 25 Page 87 of 89 OFF-SITE DOSE CALCULATION MANUAL Reporting Levels for Radioactivity Concentrations Page:

Attachment 3.21 in Environmental Samples 87 Radionuclides Food Product Water Milk Air Filter Fish pCi/kg, wet pCi/1 pCi/1 pCi/m3 pCi/kg,. wet H-3 20000 Ba-140 200 300 La-140 200 300 Cs-134 1000 30 60 10 1000 Cs-137 2000 50 70 . 20 2000 Zr-95 400 Nb-95 400 Mn-54 1000 30000 Fe-59 400 10000 Zn-65 300 20000 Co-58 1000 30000 Co-60 300 10000 I-131 100 2 *3 0.90 JF any of the above concentration levels are l")xceeded THEN see guidance contained in step 3.5.2a. for additional "mformation.

Information PMP-6010-0SD-001 I Rev. 25 Page 88 of 89 OFF-SITE DOSE CALCULATION MANUAL Page: .22 On-Site Monitoring Location - REMP 88 ONS-North TLD T-6 TLD T-5 Air ONS- Air ONS-5 Surface Water 6 SWL-2 Sediment SL-2 Well W-16 LEGEND ONS-1-0NS-6: Air Sampling Station T 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 samolina locations

Information PMP-6010-0SD-001 I Rev. 25 Page 89 of 89 OFF-SITE DOSE CALCULATION MANUAL Page:

Attachment 3.23 Off-Site Monitoring Locations - REMP 89

~?-\.~-~,;;.,

Coloma Legend Offsite REMP Monitoring Locations Substation

.** .-:. :,.. ...,"':..-., :l Coloma Rd OFT OFT-11: TLD Locations Background AirfTLD Stations Drinking Water Locations Indicator Milk Farm Locations \

Background Milk Farm Locations **

...t Benton 20 Mile Radius OFS Offsite Fish locations ,:~ Harbor

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St Joseph Water Treatment Plant (STJ) Dowagiac Substation OFS-North Background AirffLD TLD OFT-3 (DOW)

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. REVISION

SUMMARY

Procedure No.: PMP-6010-0SD-001 Rev. No.: 25

Title:

OFF-SITE DOSE CALCULATION MANUAL Alteration Justification 10 CPR 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 ar.e made in accordance with Technical Specification 5. 5 .1 and implemented through 12-EA-6090-ENV-114, Effectiveness Review for ODCM/PCP Programs.

Step 2.1.2.j Added supplemental information Enhancement to provide clarity on the of the calculation of dose attributed to variables that would be utilized for carbon-14. calculating C-14 dose.

AR#2015-I439 Step 2.1.2.k.3 Added usage of installed Enhancement which reflects the installation blowdown flow instrumentation as an of modem technology allowing for more acceptable method for obtaining flows. accurate flow measurements of blowdown.

The data is obtainable on the Plant PPC computers. This change does not involve a change of procedure intent and reflects an improvement in measuring flows.

Step 2.4.1 Added specific mention of the Enhancement to ensure it is clearly dry cask storage facility (ISFSI) and the need understood that the direct radiation dose to to incorporate the dose from it with our dual public is a combination of that from our dual unit site data. unit site and the dry cask storage facility operated under a separate license.

AR#2015-1439 Step 2.7 .2 Revised verbiage throughout the Enhancement which clarifies the deadline to step to change wording from "by May 1" to provide the NRC the ARERR, based upon "prior to May l ". industry operating experience. This change does not involve a change of procedure intent.

AR#2015-1439 Office Infonnation for Fonn Tracking Only - Not Part of Fonn This is a free-form as called out in PMP-2010-PRC-002, Procedure Alteration, Review, and Approval. Page 1 of 3

REVISION

SUMMARY

Procedure No.: PMP-6010-0SD-001 Rev. No.: 25

Title:

OFF-SITE DOSE CALCULATION MANUAL Alteration Justification .19 Corrected distance The NUREG 1301 uses kilometers and an referenced and made clarification of the error was made where the unit conversion sector criteria where broadleaf sampling was from kilometers to miles did not convert the required, reflecting NUREG 1301 more actual number and only the unit. The value accurately. of 8 kilometers equals 5 miles. Adjusted the broadleaf distances to approximate miles equivalents of NUREG 1301 guidance given in kilometers.

The clarification on the broadleaf sampling involved an NRC Branch Technical Position having different verbiage than NUREG 1301. The previous revision reflected the verbiage from the Branch Technical Position. The change made here reflects the verbiage in NUREG 1301 and clearly states that samples from 2 different sectors with the highest D/Q are required. This does not involve a change of procedure intent and simply clarifies the process of broadleaf sampling.

AR#2015-1861 Office Infonnanon for Fonn Tracking Only - Not Part of Fonn This is a free-forni,as called out in PMP-2010-PRC-002, Procedure Alteration, Review, and Approval. Page 2 of 3