Text

2020 Radioactive Effluent Release Report for Nine Mile Point Units 1 and 2
	ML21127A147
Person / Time
Site:	Nine Mile Point, 07201036
Issue date:	04/30/2021
From:	Schuerman A; Exelon Generation Co
To:	; Office of Nuclear Material Safety and Safeguards, Office of Nuclear Reactor Regulation
References
	NMP1L3388
	Download: ML21127A147 (458)
	v • d • e

Exelon Generation 10 CFR 50.36a 10 CFR 72.44(d)(3)

Technical Specifications NMP1L3388 April 30, 2021 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001 Nine Mile Point Nuclear Station, Units 1 and 2 Renewed Facility Operating License Nos. DPR-63 and NPF-69 NRC Docket Nos. 50-220 and 50-410 Independent Spent Fuel Storage Installation (ISFSI)

ISFSI Docket No. 72-1036

Subject:

2020 Radioactive Effluent Release Report for Nine Mile Point Units 1 and 2 In accordance with 10 CFR 50.36a, and the Nine Mile Point Unit 1 (NMP1) and Nine Mile Point Unit 2 (NMP2) Technical Specifications, enclosed are the Radioactive Effluent Release Reports for NMP1 and NMP2 for the period of January through December 2020. This letter also satisfies the annual effluent reporting requirements for the ISFSI required by 10 CFR 72.44(d)(3).

The format _used for the effluent data is outlined in Appendix B of Regulatory Guide 1.21, Revision

1. During the reporting period, NMP1, NMP2, and the ISFSI did not exceed any 10 CFR 20, 10 CFR 50, 10 CFR 72, Technical Specification, or ODCM limits for gaseous or liquid effluents.

Should you have questions regarding the information in this submittal, please contact Mark Greer, Manager, Site Chemistry and Radwaste, at (315) 349-5226.

Sincerely,

/f o ZJ 9 am G Schuerman

~EJ/-t Plant Manager, Nine Mile Point Nuclear Station Exelon Generation Company, LLC

U /vf '5 5 2 Co AGS/KES N!cL NAfSS

2020 RERR April 30, 2021 Page 3

Enclosures:

(1) Nine Mile Point Nuclear Station, Unit 1 Radioactive Effluent Release Report, January - December 2020 (2) Nine Mile Point Nuclear Station, Unit 2 Radioactive Effluent Release Report, January - December 2020 Cc: NRC Regional Administrator, Region 1 NRC Project Manager NRC Resident Inspector R. Rolph, NRC

Enclosure 1 Nine Mile Point Nuclear Station, Unit 1 Radioactive Effluent Release Report, January - December 2020

NINE MILE POINT NUCLEAR STATION - UNIT 1 RADIOACTIVE EFFLUENT RELEASE REPORT January- December 2020

NINE MILE POINT NUCLEAR STATION - UNIT 1 RADIOACTIVE EFFLUENT RELEASE REPORT JANUARY - DECEMBER 2020 SUPPLEMENTAL INFORMATION Facility: Nine Mlle Point Unit 1 Licensee: Nine Mlle Point Nuclear Station, LLC

1. TECHNICAL SPECIFICATION UMITS/ODCM Limits A) FISSION AND ACTIVATION GASES

1. The dose rate limit of noble gases released in gaseous effluents from the site to areas at and beyond the srte boundary shall be less than or equal to 500 mrem/year to the total body and less than or equal to 3000 mrem/year to the skin 2 The air dose due to noble gases released in gaseous effluents from Nine Mile Point Unit 1 to areas at and beyond the site boundary shall be limited during any calendar quarter to less than or equal to 5 m111iroentgen for gamma radiation and less than or equal to 10 mrad for beta radiation, and dunng any calendar year to less than or equal to 10 mi Iii roentgen for gamma radiation and less than or equal to 20 mrad for beta radiation.

B&C) TRITIUM, IODINES AND PARTICULATES, HALF LIVES> 8 DAYS

1. The dose rate limit of lodme-131, lodine-133, Tritium and all radionuclides in particulate form with half-lives greater than eight days, released in gaseous, effluents from the site to areas at and beyond the site boundary shall be less than or equal to 1500 mrem/year to any organ.

2. The dose to a member of the public from lodine-131, lodine-133, Tritium and all radionuclides in particulate form with half-lives greater than eight days in gaseous effluents released from Nine Mile Point Unit 1 to areas at and beyond the site boundary shall be limited during any calendar quarter to less than or equal to 7.5 mrem to any organ, and dunng any calendar year to less'than or equal to 15 mrem to any organ.

D) LIQUID EFFLUENTS

1. The concentration of radioactive material released in liquid effluents to unrestricted areas shall be limited to ten times the concentrations specified in 10 CFR Part 20, Appendix B, Table 2, Column 2 for radionuclides other than dissolved or entrained noble gases. For dissolved or entrained noble gases, the concentration shall be limited to 2E-04 microcuries/ml total activity.

2. The dose or dose commitment to a member of the public from radioactive materials in liquid effluents released 'from Nine Mile Point Unit 1 to unrestricted areas shall be limrted during any calendar quarter to less than or equal to 1.5 mrem to the total body and to less than or equal to 5 mrem to any organ, and during any calendar year to less than or equal to 3 mrem to the total body and to less than or equal to 10 mrem to any organ.

2. MEASUREMENTS AND APPROXIMATIONS OF TOTAL RADIOACTIVITY Described below are the methods used to measure or approximate the total radioactivity and radionuclide composition in effluents.

A) FISSION AND ACTIVATION GASES Noble gas effluent activity is determined by on-line gross activity monitonng (calibrated against gamma isotopic analysis of a 4.0L Marinelli grab sample) of an 1sokinetic stack sample stream.

B) IODINES Iodine effluent activity Is determined by gamma spectroscopic analysis (at least weekly) of charcoal cartndges sampled from an Isokinet1c stack sample stream.

C) PARTICULATES Activity released from the main stack Is determined by gamma spectroscopic analysis (at least weekly) of particulate fitters sampled from an 1sokinetic sample stream and composite analysis of the filters for non-gamma emitters D) TRITIUM Tritium effluent activity is measured by liquid scintillation or gas proportional counting of monthly samples taken with an air sparging/water trap apparatus. Tritium effluent activity is measured during purge and weekly when fuel is offloaded until stable tritium release rates are demonstrated E) EMERGENCY CONDENSER VENT EFFLUENTS The effluent cune quantrties are estimated based on the isotopic d1stnbut1on in the Condensate Storage Tank water and the Emergency Condenser shell water. Actual 1SOtopic concentrations are found via gamma spectroscopy Initial release rates of Sr-89, Sr-90 and Fe-55 are estimated by applying scaling factors to release rates of gamma emitters and actual release rates are determined from post offsite analysis results. The activity of fission and activation gases released due to tube leaks is based on reactor steam leak rates using offgas isotopic analyses F) LIQUID EFFLUENTS Isotopic contents of liquid effluents are determined by isotopic analysis of a representative sample of each batch and composite analysis of non-gamma emitters. Trrtium aciJvity is estimated on the most recent analysis of the Condensate Storage Tank water. Initial release rates of Sr-89, Sr-90, -and Fe-55 are estimated by applying scaling factors to release rates of gamma emitters and actual release rates are determined from post offsrte analysis results.

G) SOLID EFFLUENTS Isotopic contents of waste shipments are determined by gamma spectroscopy analysis of a representative sample of each batch. Scaling factors established from primary composite sample analyses conducted off-site are applied, where appropriate, to find estimated concentration of non-gamma emitters For low activity trash shipments, curie content *is estimated by dose rate measurement and application of appropnate scaling factors H) C-1'4 The production of C-14 and the effluent dose consequences are estimates based on EPRI methodology provided in EPRI Report 1021106. Estimation of Carbon-14 in Nuclear Power Plant Gaseous Effluents, December 2010 and NUREG-0016, Calculation of Releases of Radioactive Materials in Gaseous and Liquid Effluents for Boilmg Water Reactors (BWR-GALE Code)

3. METEOROLOGICAL DATA Meteorological data Is an annual summary of hourly meteorological data collected over the previous year This annual summary may be either in the form of an hour-by-hour listing on magnetic tape of wind speed, wind direction, atmospheric stability, and precipitation (if measured), or in the form of Joint frequency distribution of wind speed, wind direction, and atmosphenc stability. In lieu of submIss1on with the Radiol6g1cal Effluent Release Report, the licensee Is exercising the option of r,etaining this summary of required meteorological data on site in a file that shall be provided to the NRC upon request. '

f

ATTACHMENT 1 Page 1 of 2 Unit 1 X Unlt2 Bm2!11ng Pi!:12!;!; J!D!,H!IY - Decem!22r 20~

Liquid Effluents:

ODCM Required MaXlmum Effluent Concentration (MEG)"' 10 x 10CFR20, Appendrx B Table 2, Column 2 There were no batch dJSCharges of liquid radwaste requlnng use of MEG to determine allowable release rate MEC for the Emergency Condenser Vent LlqUJd Discharge in the first quarter of 2020 is as follows Average MEC - µCl/ml (Qtr. 1) = I NO RELEASES Average MEC - µCl/ml (Qtr ~ = I NO RELEASES I Average MEC - µCt/ml (Qtr._zi m I NO RELEASES Average MEC - µCl/ml (Qtr ~ = I NO RELEASES I Average Energy (Fission and Activation gases - MeV):

Qrtr 1 ~y = NIA E13 = NIA Qrtr _l* ~y = NIA ~13 = NIA Qrtr J ~y = NIA E13 = NIA Qrtr. ~ Ey = NIA E13 = NIA Liquid: Raawaste ~

Number of Batch Releases 0 0.00 Total Tune Period for Batch Releases (hrs) 0 0.00 Maximum Time Period for a Batch Release (hrs) 0 0.00 Average Time Period for a Batch Release (hrs) 0 0.00 Minimum Time Penod for a Batch Release (hrs) 0 0.00 Total volume of water used to dilute 1fil ~ ~ 11b.

the liquid effluent dunng release penod (L) Radwaste NIA NIA NIA NIA EC Vent I NIA I NIA I NIA I NIA I Total volume of water available to 1.fil 2ml ~ 11.a dilute the liquid effluent dunng report penod (L) Radwaste 1.31E+11 1.33E+11 1.40E+11 1.~+11 I EC Vent NIA NIA NIA NIA I Gaseous (Emergency Conden&ef' Vent):

Nunber of Batch Releases 0 Total Time Penod for Batch Releases (hrs) 0.00 Maximum Time Penod for a Batch Release (hrs) 0.00 Average Time Penod for a Batch Release (hrs) 0.00 Minimum Time Period for a Batch Release (hrs) 0.00 Gaseous (Primary Containment Purge):

Number of Batch Releases 0 Total Time Period for Batch Releases (hrs) 0.00 M8X!mum Time Penod for a Batch Release (hrs) 0.00 Average Time Pertod for a Batch Release (hrs) 0.00 Mmlmum Time Penod for a Batch Release (hrs) 0.00

ATTACHMENT 1 Page 2 of2 Unit 1 X Unit 2 B~r!lag Eid2Sli ~!!nua!Y - Dec!i!Illber 20~2 AbnonnalReleases:

A. Liquids:

INumber of Releases 0

Total ActJvrty Released I NIA lei B. Gaseous:

INumber of Releases 0

Total ActJvrty Released I NIA lei I

ATTACHMENT 2 Page 1 of 1

) Unit 1 X Unit 2 , Bi122£11ne ei 2!i!= J!!n!.m!Y - Decem~[ 2oi2 GASEOUS EFFLUENTS - SUMMATION OF ALL RELEASES, ELEVATED AND GROUND LEVEL l1lQW1[1il: ~ Q181[W ~[!ii Qy1umr 4th Q~ner ~J.TQT~I,.

§RRQB, ~

r.;,,,.,,,.,_ /1 l I -... I -- I A t-1SS10n & " ~

tt tt 1 Total Release C1 5 OOE+01 2 Average Release Rate I

µCl/sec I I B lodloe§ (1)

I - I 1 Total lodme - 131 C1 1.18E-05 3 OOE+01

2. Average Release Rate for Penod µCi/sec I ... tt I 1.49E--06 I C Ps1rngylat~ (1) 1 Particulates wrth Half-hves>8 days C1 1.01E--04 1.28E--03 8.13E--04 2.55E-03 3 00E+01 2.

3 Average Release Rate for Penod Gross .AJpha RadIoacbvlty

µCJ/sec CJ -

1.30E--06 1.63E--04 1.02E--04

... 3.21E-04

... 2 50E+01 D. Illll!.im m 1 Total Release C1 1.50E+01 3.44E+OO 6.SOE+OO 5.42E+OO 5 OOE+01

2. Average Release Rate for Penod µCJ/sec 1.93E+OO 4.37E--01 7.30E--01 6.81E--01 E Percent Qf I~m Soe1<, L.!!!lll§ Flss/on and AcbvatJOn Gases Percent of Quarteny Gamma A1T' Dose

% O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO Limrt (5 mR)

Percent of Quarteny Beta Alr Dose Limit

% O.OOE+oo O.OOE+OO O.OOE+OO O.OOE+OO (10 mrad)

Percent of Annual Gamma Alr Dose Ltmit

% O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO to Date (10 mR)

Percent of Annual Beta Alr Dose Limit to

% O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO Date (20 mrad)

Percent of Vv11ole Body Dose Rate Lin-wt (500 mrem/yr)

% O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO Percent of Skin Dose Rate Limit (3000

% O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO mrem/yr)

Icttl!.!m, IQ!;lines, !Ind E11rt1cul;ites (:n'.111:!

half-Jives greater than adays}

Percent of Quarterty Dose Ltmlt (7 .5

% 1.0SE--02 1.07E--01 1.61 E--01 O.OOE+OO mrem)

Percent of Annual Dose Limit to Date (15

% 5.31E--03 5.57E--02 1.36E--01 1.36E--01 rnrem)

P6fC61lt of Organ Dose Limit (1500

% 2.1~-04 2.14E--03 3.20E--03 O.OOE+OO rnrem/yr (1) Concentrations less than the lower limit of detection of the counting system used are Indicated with a double astensk

ATTACHMENT 3 Page 1 of2 Unit 1 X Unit 2 R~ct.!ng e2aQS!: J~nUl!!Y

Dec2m~r 20~0 GASEOUS EFFLUENTS- ELEVATED RELEASE -

Continuous Mode (2)

Nuclldes Released M Q!!!!l!er 2ml Si!W!tlir ~Ill Q!!!!cter ~b QW!!Er Fission Gatts m c, ... .... ...

Argon-41 Krypton-BS CI Krypton-85m Krypton-87 CI CI c,

Krypto-n-88 c,

Xenon-127 c,

Xenon-131m c,

\

Xenon-133 c,

Xenon-133m Xenon-135 Cl .... .... ....

Xenon-135m Xenon-137 CI Cl ..... ...

Xenon-138 Ct lodlnu {1 l lodine-131 c, .... ... ... 1 18E-05 lodine-133 lodlne-135 c,

Cl 8 42E--05 6 97E--05 5 99E-05 2 32E-04 I

Partlculjrte8 (1l Stronbum-89 c, .... .... .... ...

c, Stronbum-90 Ces,um-134 c, .. .. .. ...

Ces,um-137 c, 5 70E-08 1 70E-05 2 60E-05 7 64E-05 Cobalt-60 c, 8 38E-05 8 19E-04 4 39E-04 9.08E-04 Cobalt-58 Cl c, ..

1.15E-05 5 44E-05 2.00E-04 7 53E-04 Manganese-54 Banum-140 c, .... ..

6 78E-05 1 17E-05 1.26E-04 c,

Lanthanum-140 c,

Niobtum-95 Cenum-141 Cl c, ..

Cenum-144 lron-59 c, .... ..

1-41E-05 1.51E-05 c,

Ces,um-136 Chromium-51 c, ..

..... 5.15E-05

.. 8 50E-05 Zmc-65 CI 9.90E-06 1 36E-04 lron-55 Molybdenum-99 CI c, ...

2 60E-04 1.26E-04 4 50E-04 Neodym,um-147 Cl **-

IritlYI!! C1l c, I 1 42E+01 I 246E+OO I 4 33E+O0 I 4 54E+OO I (1) Concentrations less than the lower ~mil of detection of the counting system used are mdlcated With a double astensk A la,ver hmrt of de1ectJon of 1 OOE-04 µCi/ml for reqLDred noble gases, 1 00E-11 µCi/ml for required particulates, 1.00E-12 µCl/ml for required lodmes, and 1 OOE--06 µCi/ml for Trrtlum as required by the ODCM, has been venfied (2) Contnbutlons from purges are included There were no other batch releases dunng the reporting period

ATTACHMENT 3 Page 2 of 2 Unit 1 X Unit 2 B1mort111g P~~: JA!JYill:Y - December ~QiQ GASEOUS EFFLUENTS- ELEVATED RELEASE Batch Mode (2)

Nuclldes Released l§1 Q!JAi:ter 2ml Q!arli![ ~a;I Q!J!!ru!r 4ttJ Qli!i!rm E!HIQ!!Gam(l) l Argon-41 Cr Krypton-85 Cr Kryptoo-85m Cr Krypton-87 Cr Kryptoo-88 Cr Xenon-127 Cr XeQon-131m Cl Xenon-133 Cr Xenon-133m Cr Xenon-135 Cr Xenon-135m Cr Xenon-137 Xenon-138 Cr Cr .. ..

Iodines {1l lodrne-131 Cr lodme-133 lodrne-135 Cr a I I .. I I I Particulates {1}

Stronbum--89 Cr ...

Strontium-90 Cr Cesrum-134

- Cr Cesrum-137 Cr Cobalt-60 Cr Cobalt-58 Cr Manganese-54 a a

Banum-140 Lanthanum-140 a Nloblum-95 Cr Cenum-141 Cl Cenum-144 Cr lron-59 Ceslum-136 Cr Cr Chromrum-51 Zmc--65 Cr Cr ..

lron-55 Molybdenum-99 Cr Cr ..

Neodymrum-147 Cr TrttllJ!!J {ll Cr I .. I .. I .. I .. I (1) Concentrations less than the lower hmrt of detectton of the counting system used are Indicated wrth a double asterisk A lower hmrt of detecbon of 1 OOE-04 µCi/ml for required noble gases, 1 OOE-11 µCi/ml for required particulates, 1 OOE-12 µCVml for required Iodines, and 1.00E-06 µCl/ml for Trlbum as required by the ODCM, has been venfted (2) Contnbutlons from purges, If any, are included There were no other ba1ch releases dunng the reporting period.

A TT AC HM ENT 4 Page 1 of 2 Unit 1 X Unit 2 B!i!Rf!r!IWI e!i!dQQi Jijll!,H!!Y - Decem~[ ~~

GASEOUS EFFLUENTS - GROUND LEVEL RELEASES Ground level releases are determmed in accOfdance with the Off-Site Dose Calculation Manual and Chemtstry procedures Continuous Mode Nuclldes Released 1stQ!:Hll:ti[ 2n!l Q!H!~r 3~ 91J!!rt&[ 4!!:l 9Y!!W![

Fission Gases 111 Argon-41 Krypton-85 C1 Ci ...

Krypton-85m CI Krypton--87 C1 Krypton-88 C1 Xenon-127 Xenon-131m CI C1 .... ....

Xenon-133 Cl Xenon-133m Xenon-135 Cl C1 Xenon-135m Cl Xenon-137 Xenon-138 CI CI .. .. ..

IQdinea (1) lodine-131 /' Cl I ..

lodlne-133 Cl lodine-135 Cl I ... I ... I I Particulate§ c11 Stronllurn-89 Cl .... ..... .... ....

stronbum-90 Ceslum-134 Ci Ci ees,um-137 Cobalt-60 Ci Cl Cobalt-58 Cl Manganese-54 Ci Banum-140 C1 Lanthanum-140 Cl Niobium-95 C1 Cenum-141 C1 Cenum-144 lron-59 CI Ci .... .... ....

Ceslum-136 Cl Chromium-51 CI Zinc-65 lron-55 CI CI .... .... ..

Molybdenum-99 Neodymium-147 CI CI

- Idtl!!!ll m Cl I 7 86E-01 I 9 BOE-01 I 1 47E+O0 I 8 79E-01 I (1) Concentrations less than the lower hmrt of detection of the counbng system used are 1ndIcated wrth a double asterisk

ATTACHMENT 4 Page 2 of 2 Unit 1 X Unit 2 Re~Q!l!ng P!lrl~: ~n!,!ar,: - December iQ 60 GASEOUS EFFLUENTS * $3ROUND LEVEL RELEASES Ground level releases are determmed in accordance wtth the Off-Site Dose Calculation Manual and Chemistry procedures Batch Mode Nuclldes Released j,:it Q!.JS(W[ 2nd Q!!!!!l~r ~rd Q!!!!!l~r ~tb Q!H!~r FlnlonGa&gs(jl Argon-41 Krypton-BS Cl CI Krypton-85m Krypton-87 Cl c, ...

Krypton-BB c, Xenon-127 Cl Xenon-131m Cl Xenon-133 Cl c,

Xenon-133m Xenon-135 Xenon-135m CI Ci Xenon-137 Xenon-138 Cl CI .. .. ... ..

lodjnes {1l lodine-131 C1 .... .... .... ...

lodme-133 lodine-135 C1 C1 I .. I .. I ... I .. I PartlcuJatn m strontl um-89 Ci strontiurn-90 Cl c, ... ... ....

Cesiurn-134 Cesiurn-137 Coball-80 Cl Ci Cobail-58 Manganese-54

' Cl Cl Bamin-14-0 Lanthanurn-14-0 C1 CI .... ....... ...... ....

N10bIurn-95 C1 Cenurn-141 Cenurn-144 C1 Cl lron-59 Ceslurn-136 C1 Cl .... .... -.. ....

Chrom,urn-51 Zinc-65 Cl Cl .... .... .... .... -

lron-55 Ci Molybdenurn-99 Neodymiurn-147 CI CI Trltl!.J!!! {1 l Cl I ... I .. I - I *: I (1) Concentrations less than the lcrwer limit of detecbon of the colllbng system used are indicated with a double astensk

ATTACHMENT 5 Page 1 of 2 Unit 1 X Unit 2 R!i!122nlag E!i!d2SJ: Ji!n!!~!)'. . ~!,!i!!!!~[ ,Q,Q LIQUID EFFLUENTS-SUMMATION OF ALL RELEASES (1)

J 1s Q!!i!tlir ,asi Q!!~!:l!lr ~lli Q!,!i!rter ~lb !d!di!!:1!H Est, I2lill ~rmr, ~

A F1ssmn & ActJvabon nooucm 1 Total Release (not including Tntlum, gases, alpha) a No Releases No Releases No Releases No Releases 5 OOE+01 2 Average diluted concentrabon dunng

µCJ/ml No Releases No Releases No Releases No Releases reporting period B...Ir!ll.!Jm.

1 Total release a No Releases No Releases No Releases No Releases 5 00E+01 2 Average diluted concentration dunng

µCVml No Releases No Releases No Releases No Releases the reporbng period C ~lll:§olved and Entr 2 1 ~ ~!ii~

1 Total release Cl No Releases No Releases No Releases No Releases 5 00E+01 2 Average diluted concentration dunng

µCJ/ml No Releases No Releases No Releases No Releases the reporting penod D Q(~ 81!2bfl B11!:11oact1V1!;y 1 Total release a INo Releases INo Releases INo Releases INo Releases I 5 OOE+01 E ~

1. Pnorto Dtlubon l..Jters No Releases No Releases No Releases No Releases 5 OOE+01 2 Volume of dilution water used dunng l..Jters No Releases No Releases No Releases No Releases 5 OOE+01 release per10d 3 Volume of dilution water available l..Jters 1.31E+11 1 33E+11 1 40E+11 1 34E+11 5 OOE+01 dunng reporting penod - Cooling Water F l:~~l QfI~ ~~ l.!mll§ Percent of Quarterly Vvtiole Body Dose.

% No Releases No Releases No Releases No Releases Umlt (1 5 mrem)

Percent of Annual Whole Body Dose

% No Releases No Releases No Releases No Releases l.Jmrt to Date (3 mrem)

Percent of Quarterly Orgwi Dose LImrt (5 nYem)

% No Releases No Releases No Releases No Releases Percent of Annual Organ Dose Limit to Date (10 mrem)

% No Releases No Releases No Releases No Releases Percent of 10CFR20 Concentration

% No Releases No Releases No Releases No Releases l.Jmrt Percent of DJSSOlved or Entrained Noble Gas Umrt (2 00E-04 µCI/mQ

% No Releases No Releases No Releases No Releases

\

(1) Concentrations less than the lower llmrt of detection of the counting system used are indicated wrth a double astensk.

ATTACHMENT 5 Page 2 of 2 Unit 1 X Unit 2 Be!!Q!l!Dg ~!!rlod, .li!DYICll - ~!i!!!~r 20i0 LIQUID EFFLUENTS RELEASED Batch Mode (1 ),(2)

Nuclldes Released jl!! Qlll!tli[ iml QYi!l:t!i! ~Ill QYllrn!r ~11:! Q!t!Bcter Nuclldes Released Strontium-89 CI No Releases No Releases No Releases No Releases Strontium-90 C1 No Releases No Releases No Releases No Releases Ceslum--134 C1 No Releases No Releases No Releases No Releases Ceslum-137 CJ No Releases No Releases No Releases No Releases lodine-131 Cl No Releases No Releases No Releases No Releases Cobalt-58 CJ No Releases No Releases No Releases No Releases Cobalt-60 C1 No Releases No Releases No Releases No Releases lron-59 CJ No Releases No Releases No Releases No Releases Zlnc-65 CJ No Releases No Releases No Releases No Releases Manganese-54 CJ No Releases No Releases No Releases No Releases Chromrnm-51 C1 No Releases No Releases No Releases No Releases Zlrcornum-95 C1 No Releases No Releases No Releases No Releases Nioblum-95 Cl No Releases No Releases No Releases No Releases Molybdenum-99 C1 No Releases No Releases No Releases No Releases Bartllll-140 CI No Releases No Releases No Releases No Releases Lanthanum-140 CJ No Releases No Releases No Releases No Releases Cenum-141 CI No Releases No Releases No Releases No Releases lodine-133 Cl No Releases No Releases No Releases No Releases lron-55 C1 No Releases No Releases No Releases No Releases Certum-144 CI No Releases No Releases No Releases No Releases Cesium-138 C1 No Releases No Releases No Releases No Releases Copper-64 C1 No Releases No Releases No Releases No Releases Manganese-56 t, No Releases No Releases No Releases No Releases Nickel-65 CI No Releases No Releases No Releases No Releases Sodium-24 Ci No Releases No Releases No Releases No Releases DISSOived Of Entralned Gases C1 I No Releases I No Releases I No Rele~s INo Releases I TntJum C1 I No Releases I No Releases I No Releases INo Releases I (1) No continuous mode release occurred dunng the report penod as Ind1cated by effluent sampling There were no Raclwaste Ba1ch Releases (2) Concentrabons less than the lower hmrt of detectJon of the counting system used have been venfied fOf sampled effluents A lower hmrt of detection of 5 00E-07 µC1/ml for required gamma emitting nucJIdes, 1 00E-05 µCi/ml for required dissolved and entrained noble gases and tnbum, 5 OOE-08 µCi/ml for ~-89/90, 1 00E-06 µCi/ml for 1-131 and Fe-55, and 1.00E-07 µCi/ml for gross alpha rad1oactlvrty, as ldentrfled In the ODCM, has been venfled. Concentrations less than the lower hmlt of detection of the counting system used are Ind1cated 'Mth a double astensk

ATTACHMENT 6 Page 1 of 4 Unlt1 X Unlt2 Bi!!Qrtlng e2a2!;!; Jg[!uaD£ - Q!!!.ember 2QZQ SOLID WASTE AND IRRADIATED FUEL SHIPMENTS A1. TYPE ~ ActMty: {11 (mi (Cl)

£1!1!! ~

A B C A B C a 1 Spent Resin (Dewatered) 1.49E+01 0 OOE+OO 0 OOE+OO 8 30E+OO 0 OOE+OO 0 OOE+OO a2 Filter Sludge ( O.OOE+OO 0 OOE+OO 0 OOE+OO 0 OOE+OO 0 OOE+OO 0 OOE+OO 83 Concentrated Waste 0 OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO 0 OOE+OO 0 OOE+OO Totals 1 49E+01 0 OOE+OO 0 OOE+OO 8 30E+OO O.OOE+OO 0 OOE+OO b 1 Dry Compressible Waste 5 78E+01 0 OOE+OO O.OOE+OO 2 21E-02 0 OOE+OO 0 OOE+OO b.2 Dry Non-Compressible Waste (Contaminated 0 OOE+OO O.OOE+OO 0 OOE+OO 0 OOE+OO 0 OOE+OO 0 OOE+OO Equipment)

Totals 5 78E+01 O.OOE+OO 0 OOE+OO 2.21E-02 O.OOE+OO 0 OOE+OOI C lrradl8.ted Components, 0 OOE+OO 0 OOE+OO 0 OOE+OO 0 OOE+OO 0 OOE+OO 0 OOE+OO Control Rods, etc d Other (to vendor for processing) d 1 Sewage Sludge 2 75E+01 O.OOE+OO 0 OOE+OO 3 01E+OO 0 OOE+OO 0 OOE+OO (1) The esbmated total error Is 5.0E+01%

ATTACHMENT 6 Page 2 of 4 Unit 1 X Unlt2 Reporting Period: January - December 2020 SOLID WASTE AND IRRADIATED FUEL SHIPMENTS A1. TYPE CQDii!IDi!: el!!.~lllli §Q!ld!fl~Qn ~!31 a 1 Spent Resin Poly Ll[l8f General Design None a 2 Filter Sludge Poly LJner Type B None b1 Dry Compressible Waste Seavan General Design None b2 Dry Non-Compressible Waste (contaminated NIA NIA NIA equipment) c Irradiated Components, NIA NIA NIA Control Rods d Other (To vendor for processing) d 1 Sewage Sludge Sack. General Design NIA

ATTACHMENT 6 Page 3 of 4 Unit 1 X Unit 2 B!QOrtlng et~i J!![!!,!B!Y - ~!!!!!!![ ~iQ SOLID WASTE AND IRRADIATED FUEL SHIPMENTS A2 ESTIMATE OF MAJOR NUCLIDE COMPOSITION (BY TYPE OF WASTE) a Spent Resins, Filter Sludges, Concentrated Waste

~ ~ ~

Fe-55 333% 2 76E-01 Co-60 92 91% 7 71E+O0 Cs-137 1 21% 1 01 E-01 b Dry Compressible Waste, Dry Noo-Compressible Waste (Contaminated Equipment)

~ ~ ~

Mn-54 440% 9 72E-04 Fe-55 16 97% 3 75E-03 Co-58 268% 5.93E-04 Co-60 6657% 1 47E-02 Zn-65 234% 517E-04 Cs-137 555% 1.23E-03 c Irradiated Components, Control Roos There were no shipments

~ -

~

NA NA I

\

d Other (To vendor for processing)

1 Sl.lllp Llner

~ ~ ~

H-3 529% 5 30E-01 Mn-54 767% 7 68E-01 -

Fe-55 18 34% 1 84E+O0 Co-58 1 58% 1.58E-01 Co-60 61 82% 619E+OO Zn-65 1 51% 1 51 E-01 Cs-137 238% 2 38E-01

ATTACHMENT 6 Paga 4 of 4 Unit 1 X Unit 2 Beoortlag Ei~i .!!!DYi!!:ll - Q~mbe[ ~oiQ SOLID WASTE AND IRRADIATED FUEL SHIPMENTS A3 SOLID WASTE DISPOSITION N!.Jmber of ~hiom!i!!J!§ M~ Qf I!1![]~oort51IlQ!J ~llrn!l!QD 1 Truck, highway Penna-Flx of Flonda 5 Truck, htghway Energy Solutions, Bear Creek 3 Truck, highway Energy Solutlons, Clive CWF B. IRRADIATED FUEL SHIPMENTS (Disposll.Jon)

~!J!Ill2s![ Qf ~l:!H11Il~I§ MQQ~ Qf Tra!J§QO[!a!!Qn De!l~!l§~Qn D SEWAGE WASTES SHIPPED TO A TREATMENT FACILITY FOR PROCESSING AND BURIAL There were no shipments of sewage sludge 'Mth detecl.Jble quantllles of plant-related nuci1des from NMP to the treatment faCJlrty during the reportmg penod.

ATTACHMENT 7 Rage 1 of 1 Unit 1 X Unit 2 - , UUM Period: *-----.... 2020

SUMMARY

OF CHANGES TO THE OFF-SITE DOSE CALCULATION MANUAL (ODCM)

The Unrt 1 Off-Srte Dose Calculation Manual (ODCM) was revised dunng the reporting period. The table below 1s a listing of changes.

REVISION 37

{

New/Amended Section Page# Description of Change Reason For Change I The Unit 1 ODCM was revised to NOTES FOR TABLED Under action (a), create (a)(2)(b) and move acbon include the optional DLCO action to I 3 1-10 3 6 14-2 sa)(2) to (a)(2)(b) ubhze stack rad1at1on monitor readings rather than take grab samples Under action (a), create (a)(2)(a) to read "Gross The Unit 1 ODCM was revised to NOTES FOR TABLE D activity 1s recorded every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when both Noble include the optional DLCO action to I 3 1-10 3 6 14-2 Gas Activity Radiation Monitor 1nd1cabons are ublize stack radiation monitor readings functional " rather than take grab samples NOTES FOR TABLE D Under acbon (1), create (1)(2) and move actJon (I) to 13 1-11 Formatbng 3.6.14-2 (cont'd) (1)(2)

The Unit) ODCM was revtsed to Under acbon (1), create (1)(1) to read "Gross acbvity NOTES FOR TABLED include the optional DLCO action to 131-11 1s recorded every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when both Noble Gas 3 6.14-2 (cont'd) utilize stack radiation monitor readings Act1v1ty Rad1at10n Monitor ind1cabons are functional."

rather than take grab samples Under acbon (g), append the statement "perform The Unit 1 ODCM was revised to NOTES FOR TABLE D one of the followmg AND restore the nonfuncbonal include the optional DLCO acbon to 131-11 3 6 14-2 (cont'd) channel(s) to funcbonal utilize stack radiation monitor readings status within 30 days O)" rather than take grab samples NOTES FOR TABLE D 13.1-11 Move acbon (g)(1) to (g)(3) formatting 3.6 14-2 (cont'd)

Change acbon (g)( 1) to read Wrth one or more channel nonfunctional due to alarm function only NOTES FOR TABLE D Added danfylng instruction for handling 131-11 every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months record gross acbvtty of both 3 6 14-2 (cont'd) readings channels as read on the meters in the control room and venfy within limits" Change action (g)(2) to read Take grab samples NOTES FOR TABLE D wrth1n 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter This had already existed 1n the ODCM 131-11 3 6 14-2 (cont'd) AND analyze samples for gross act1v1ty wrthin 24 It was moved for formattmg hours of sampling completion "

ATTACHMENT 8 Page 1 of 1 Unit 1 X Unit 2 Reporting Period: January - December 2020

SUMMARY

OF CHANGES TO THE PROCESS CONTROL PROGRAM (PCP)

There were no changes to the Process Control Program during this reporting penod

ATTACHMENT 9 Page 1 of 1 Unit 1 X Unit 2 Reporting Period: January - December 2020

SUMMARY

OF NON-FUNCTIONAL MONITORS Oates Monitor was Monitor Cause and Corrective Actions Non-Functional L1qu1d Radwaste Japuary 1, 2020 to These monitors were intentionally allowed to exceed their quarte~

Discharge Monitors December 31. 2020 functional tests and annual calibration frequency, as no discharges 11 and 12 are planned or expected. This condition is allowed as long as blank flanges are installed in the discharge line, precluding any unmonitored discharge. Blank ~anges are currently installed and no liquid waste discharges were performed during 2020. This non-functionality Is tracked in Equipment Status Log (ESL-Deficient-09 0029)

\

._J

ATTACHMENT 10 Page 1 of 4 Unit 1 _x__ Unit 2 _ _ Reporting PerJQCI: January - December 2020 DOSES TO MEMBERS OF THE PUBLIC DUE TO THEIR ACTIVITIES INSIDE THE SITE BOUNDARY Introduction An assessmenr of the radiation dose potentially received by a Member of the Public due to their activities inside the site boundary from Nine Mile Point Unit 1 (NMP1) liquid and gaseous effluents has been conducted for the period January through December 2020.

This assessment considers the maximum exposed individual and the various exposure pathways resulting from liquid and gaseous effluents to identify the maximum dose received by a Member of the Public during their activities within the site boundary.

Prior to September 11, 2001, the public had access to the Energy Information Center for purposes of observing the educational displays or for picnicking and associated actMties. Fishing also occurred near the shoreline adjacent to the Nine Mile Point (NMP) site. Fishing near the shoreline adjacent to the NMP site was the onsite activity that resulted in the potential maximum dose received by a Member of the Public. Following September 11, 2001 public access to the Energy Information Center has been restricted and fishing by Members of the Public at locations on site is also prohibited. Although fishing was not conducted during 2020, the annual dose to a hypothetical fisherman was still evaluated to provide continuity of data for the location.

Dose Pathways Dose pathways considered for this evaluation included direct radiation, inhalation and external ground (shoreline sediment or soil doses). other pathways, such as ingestion pathways, are not considered because they are either not applicable, insignificant, or are considered as part of the evaluation of the total dose to a member of the public located off-site. In addition, only releases from the NMP1 stack and emergency condenser vent were evaluated for the inhalation pathway. Dose due to aquatic pathways such as liquid effluents is not applicable since swimming is prohibited at the NMP site.

Dose to a hypothetical fisherman is received through the following pathways while standing on 'the shoreline fishing:

External ground pathway; this dose is received from plant related radionuclides detected in the shoreline sediment.

Inhalation pathway; this dose is received through inhalation of gaseous effluents released from the NMP1 Stack and Emergency Condenser Vent.

Direct radiation pathway; dose resulting from the operation of NMP1, Nine Mile Point Unit 2 (NMP2) and the James A. Fitzpatrick Nuclear Power Plant (JAFNPP) Facilities.

Methodologies for Detennlnlng Dose for Applicable Pathways External Ground (Shoreline Sediment) Pathway Dose from the external ground (shoreline sediment) is based on the methodology in the NMP1 Offsite Dose Calculation Manual (ODCM) as adapted from Regulatory Guide 1.109. For this evaluation it is assumed that the hypothetical maximum exposed individual fished from the shoreline at all times.

ATTACHMENT 10 Page 2 of 4 Unrt 1 _x_ Unrt 2 Reporting Per1odi January - December 2020 DOSES TO MEMBERS OF THE PUBLIC DUE TO THEIR ACTMTIES INSIDE THE SITE BOUNDARY The total dose received by the whole body and skin of the maximum exposed individual during 2020 was calculated using the following input parameters:

Usage Factor = 312 hours0.00361 days 0.0867 hours 5.15873e-4 weeks 1.18716e-4 months (fishing 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months per week, 39 weeks per year)

Density in grams per square meter= 40,000 Shore width factor= 0.3 Whole body and skin dose factor for each radionuclide= Regulatory Guide 1.109, Table E-6.

Fractional portion of the year = 1 (used average radionuclide concentration over total time peric No radionuclides were detected in sediment samples for 2020.

The total whole body and skin doses received by a hypothetical maximum exposed fishennan from the external ground pathway is presented in Table 1, Exposure Pathway Annual Dose.

Inhalation The inhalation dose pathway is evaluated by utilizing the inhalation equation in the NMP1 ODCM, as adapted from Regulatory Guide 1.109. The total whole body dose and organ dose received by the hypothetical maximum exposed fishennan during 2020 calculated using the following input parameters for gaseous effluents released from both the NMP1 Stack and Emergency Condenser Vent for the time period exposure is received:

NMP 1 Stack:

Variable 1 Flshennan XJQ (s/m3) 8.9E-06 Inhalation dose factor Table E-7, Regulatory Guide 1.109 Annual air intake (m 3/year) (adult) 8000 Fractional portion of the year 0.0356 H-3 (pCi/sec) 4.77E+05 C-14 (pCi/sec)2 2.80E+05 Mn-54 (pCi/sec) 8.66E+00 Cr-51 (pCL/sec) 5 74E+00 Fe-55 (pCi/sec) , 3.52E+01 Fe-59 (pCi/sec) 1.23E+00 Co-58 (pCi/sec) 4.24E+01 Co-60 (pCi/sec) 9 12E+01 /

Zn-65 (pCi/sec) 6.12E+O0 1-131 (pCi/sec) 4 98E-01 1-133 (pCi/sec) 1.52E+01 Cs-137 (pCL/sec) 5.02E+00

ATTACHMENT 10 Page 3 of 4 Unit 1 _x_ Unit 2 _ _ Reporting Period; January - December 2020 DOSES TO MEMBERS OF THE PUBLIC DUE TO THEIR ACTNITIES INSIDE THE SITE BOUNDARY-NMP1 Emergency Condenser Vent:

Variable 1 Fisherman X/Q (s/m3) 6 63E-06 Inhalation dose factor Table E-7, Regulatory Guide 1109 3 8000 Annual air intake (m /year) (adult)

Fractional portion of the year 0.0356 H-3 (pCi/sec) 1 40E+05 1

The maximum exposed fisherman is assumed to be present on site during the period of April through December at a rate of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months per week for 39 weeks per year equivalent to 312 hours0.00361 days 0.0867 hours 5.15873e-4 weeks 1.18716e-4 months for the year (fractional portion of the year = 0.0356). Therefore, the Average Sta~k and Emergency Condenser Vent flow rates and radionuclide concentrations used to determine the dose are represented by second, third and fourth quarter gaseous effluent flow and concentration values.

2 C-14 release rate determined from NUREG-0016, "Calculation of Releases of Radioactive Materials in Gaseous and Liquid Effluents for Boiling Water Reactors (BWR-GALE Code)," and EPRI Technical Report 1021106, "Estimation of Carbon-14 in Nuclear Power Plant Gaseous Effluents."

The total whole body dose and maximum organ dose received by the hypothetical maximum exposed fisherman is presented in Table 1, Exposure Pathway Annual Dose.

Direct Radiation Pathway The direct radiation pathway is evaluated in accordance with the methodology found in the NMP1 ODCM. This pathway considers four components: direct radiation from the generating facilities. direct radiation from any possible overhead plume, direct radiation from ground deposition and direct radiation from plume submersion. The direct radiation pathway is evaluated by the use of high sensitivity environmental Thermoluminescent Dosimeters (TLDs). Since fishing activities occur between April 1 and December 31, TLD data for the second, third, and fourth quarters of 2020 from TLDs placed in the general area where fishing once occurred were used to determine an average dose to the hypothetical maximum exposed fisherman from direct radiation. The following is a summary of the average dose rate and assumed time spent on site used to determine the total dose received.

Variable Flshennan Average Dose Rate (mRem/hr) 1.32E-03 Exposure time (hours) 312 Total doses received by the hypothetical m&ximum exposed fisherman from direct radiation is presented in Table 1, Exposure Pathway Annual Dose.

ATTACHMENT 10 Page 4 of 4 Unit 1 _x_ Unrt 2 Reporting Period: January - December 2020 DOSES TO MEMBERS OF THE PUBLIC DUE TO THEIR ACTIVITIES INl?IDE THE SITE BOUNDARY Dose Received By Hypothetical Maximum Exposed Member of the Public Inside the Site The following is a summary of the dose received by a hypothetical maximum exposed fisherman from liquid and gaseous effluents released from NMP1 during 2020:

TABLE 1 Exposure Pathway Annual Dose Fisherman Exposure Pathway Dose Type (mrem)

Whole Body 0.00E+00 External Ground Skin of Whole Body 0.00E+00 Whole Body 5.36E-04 Inhalation Maximum Organ Bone . 1.61 E-03 Thyroid 5.35E-04 Direct Radiation Whole Body 0 41 Based on these values the total annual dose received by a hypothetical maximum exposed Member of the Public inside the site boundary is as follows:

TABLE 2 Annual Dose Summary Flshennan -

Tota1Annua1Dosefor2020 (mrem)

Total Whole Body 4 12E-01 Skin of Whole Body 0.00E+00 Maximum Organ Bone : 1.61 E-03 Thyroid 5.35E-04

ATTACHMENT 11 Page 1 of 5 Unit 1 __x_ Unrt 2 _ _ Reporting Pedod; January - December 2020 DOSES TO MEMBERS OF THE PUBLIC DUE TO THEIR ACTIVITIES OUTSIDE THE SITE BOUNDARY Introduction An assessment of radiation doses potentially received by the likely most exposed Member of the Public located beyond the site boundary was conducted for the period January through December 2020 for comparison against the 40 CFR 190 annual dose-limits.

The intent of 40 CFR 190 requires that the effluents of Nine Mile Point Unit 1 (NMP1), as well as other nearby uranium fuel cycle facilities, be considered. In this case, the effluents of NMP1, Nine Mile Point Unit 2 (NMP2) and the James A. FitzPatrick Nuclear Power Plant (JAFNPP) facilities must be co"nsidered. -

40 CFR 190 requires the annual radiation dose received by Members of the Public in the general environment, as a result of plant operations, be limited to:

< 25 mRem whole body

<-25 mRem any organ (except thyroid)

< 75 mRem thyroid This evaluation compares doses resulting from liquid and gaseous effluents and direct radiation originating from the site as a result of the operation of the NMP1, NMP2 and JAFNPP nuclear facilities.

Dose Pathways Dose pathways considered for this evaluation included doses resulting from liquid effluents, gaseous effluents and direct radiation from all nuclear operating facilities located on the Nine Mile Point site.

Dose to the likely most exposed Member of the Public, outside the site boundary, is received through the following pathways:

Fish consumption pathway; this dose is received from plant radionuclides that have concentrated in consumed by a Member of the Public.

Vegetation consumption pathway; this dose is received from plant radionuclides that have vegetation that is consumed by a Member of the Public.

Shoreline Sediment; this dose is received as a result of an individual's exposure to plant in the shoreline sediment, which is used as a recreational area.

Deposition, Inhalation and Ingestion pathways resulting from gaseous effluents; this dose is exposure to gaseous effluents released from NMP1, NMP2 and JAFNPP operating facilities.

Direct Radiation pathway; radiation dose resulting from the operation of NMP1, NMP2 and JAFNPP (including the Independent Spent Fuel Storage-Installations (ISFSI)).

Methodologies for Determining Dose for Applicable Pathways Fish Consumption Dose received as a result of fish consumption is based on the methodology specified in the NMP1 Off-Site Dose Calculation Manual (ODCM) as adapted from Regulatory Guide 1.109. The dose for 2020 is calculated from actual analysis results of environmental fish samples taken near the site discharge points. For this evaluation it is assumed that the most likely exposed Member of the Public consumes fish taken near the site discharge points.

No radionuclides were detected in fish samples collected and analyzed during 2020; therefore, no dose was received by the whole body and organs of the likely most exposed Member of the Public during 2020.

ATTACHMENT 11 Page 2 of 5 Unrt 1 ___x_ Unrt 2 _ _ Reporting Period; January - December 2020 DOSES TO MEMBERS OF THE PUBLIC DUE TO THEIR ACTIVITIES OUTSIDE THE SITE BOUNDARY Vegetation Consumption ~ . .

NMP1 ODCM as adapted from Regulatory Guide 1.109. The dose for 2020 is calculated from actual analysis results of environmental vegetation samples taken near the most exposed Member of the Public.

No radionuclides were detected in vegetation samples collected ,and analyzed during 2020; therefore, no dose was received by the whole body and organs of the likely most exposed Member of the Public during 2020.

For estimating C-14, dose received as a result of vegetation consumption is based on the methodology specified in the NMP1 ODCM as adapted from Regulatory Guide 1.109. The estimated concentration of C-14 in vegetation is based on the estimated concentration of C-14 in plant gaseous effluents.

Shoreline Sediment Dose received from shoreline sediment is based on the methodology in the NMP1 ODCM as adapted from Regulatory Guide 1.109. For this evaluation it is assumed that the most likely exposed Member of the Public spends 67 hour7.75463e-4 days 0.0186 hours 1.107804e-4 weeks 2.54935e-5 months s/year along the shoreline for recreational purposes.

No radionuclides were detected in shoreline sediment samples collected and analyzed during 2020; therefore no dose was received by the whole body and organs of the likely most exposed Member of the Public during 2020.

Dose Pathways Resulting From Gaseous Effluents Dose received by the likely most exposed Member of the Public due to gaseous effluents is calculated in accordance with the methodology provided in the NMP1 ODCM, NMP2 ODCM, and the JAFNPP ODCM. These calculations consider deposition, inhalation and ingestion pathways. Actual meteorological data was used to calculate doses to the likely most exposed Member of the Public. The total sum of doses resulting from gaseous effluents from NMP1, NMP2 and JAFNPP during 2020

provides a total dose to the whole body and maximum organ dose for this pathway.

Carbon-14 Dose Pathways Resulting from Gaseous Effluents The Carbon-14 (C-14) effluent source tenns are used to estimate radiological doses from C-14 in site gaseous waste effluents. These estimates were generated in order to meet the NRC requirement to incorporate C-14 in nuclear power plant 2020 Annual Radiological Effluent Release Reports (ARERRs).

The C-14 production and effluent source tenn estimates were based on EPRI methodology provided in EPRI Report 1021106, Estimation of Carbon-14 in Nuclear Power Plant Gaseous Effluents, December 2010. The following methodology was used in estimating C-14 gaseous release actMty and dose components for the 2020 ARERR.

EPRI methodology for estimating C-14 production rates in Boiling Water Reactors (BWRs):

For BWRs, EPRI Report 1021106 summarized the distribution of C-14 in release pathways as follows:

gaseous 95% to 99%, liquid <0.5% and solid 1% to 5%. The report also states that ~95% of C-14 in BWR gaseous waste effluents exists in the carbon dioxide fonn, which contributes to population dose via photosynthesis uptake in the food consumption cycle.

ATTACHMENT 11 Page 3 of 5 Unrt 1 _x_ Unit 2 Reporting Period: January - December 2020 DOSES TO MEMBERS OF THE PUBLIC DUE TO THEIR ACTIVITIES OUTSIDE THE SITE BOUNDARY For NMP1 and NMP2, C-14 gaseous dose calculations in the site ARERR are made using the following assumptions for each unit: (1) continuous release of the estimated C-14 generated during power operation based on the number of Effective Full Power Days (EFPDs) for the period, (2) maximum C-14 activity from IH:erature values cited in EPRI Report 1021106, and (3) typical fraction as carbon dioxide for gaseous releases from literature values also cited in EPRI Report 1021106.

Equation 1 estimates the maximum annual production of C-14, PRMAX, for each BWR unH:.

PRMAX = 5.1

MWT / 1000 [Eq 1]

Where:

5.1 = BWR Normalized Production (CI/GWt-yr)

MWT = Megawatts Thermal (MWt) 1000 = Conversion Factor (Mwt to GWt)

Equation 2 estimates the C-14 activity released, A e- 14 , into the gaseous pathway during the time period for each BWR unit.

A e-14 = PR MAX

0.99

EFPD I 365, Ci (for time period) [Eq 2]

Where:

PRMAX = maximum annual production rate of C-14 0.99 = fraction of C-14 in BWR gaseous pathway releases (maximum literature value in EPRI Report 1021106; also Table 1)

EFPD = number of effective full power days for the unit during the time period; e.g., quarterly or yearly (Table 1) 365 = number of days in a typical year

ATTACHMENT 11 Page 4 of 5 Unrt 1 ___x_ Unit 2 _ _ Reporting Period; January - December 2020 DOSES TO MEMBERS OF THE PUBLIC DUE TO THEIR ACTIVITIES OUTSIDE THE SITE BOUNDARY Equation 3 estimates the y-14 activity released in carbon dioxide form, A e- 14, co2 , into the gaseous pathway during the time period for each BWR unit.

A C-14, CO2 = PR MAX

0.99

0.95

EFPD I 365, Ci (for time period) [Eq 3]

Where:

PR MAX = maximum annual production rate of C-14 0.99 = fraction of C-14 in BWR gaseous pathway releases (maximum literature value in EPRI Report 1021106; also Table 1) 0.95 = fraction of C-14 as carbon dioxide in BWR gaseous pathway releases (typical literature value in EPRI Report 1021106; also Table 1)

EFPD = number of effective full power days for the unit during the time period, e.g.*quarterly or yearly (Table 1) 365 = conversion factor, 365 days in a typical average year For each BWR unit, the 2020 estimated C-14 activity releases (total and carbon dioxide chemical form) are summarized in Table 1.

Table 1 2020 BWR Estimated C-14 Gaseous Releases Gaseous CO2 Fonn Max. Annual EFPD 2020 Total 2020 CO2 Release BWR Release Release Prod. Rate Operation Release (Eq 2) (Eq 3)

Fraction(*! Fractlon(bl (Eq 1) 364 879 EFPD NMP1 0 99 095 9 44 Ci/yr 9 34 CI 8 87 C1 (99 69%)

330 724 EFPD NMP2 0 99 095 (90 36%)

20 33 clfyr'-c) 18 24 C1 17.28 CI 327 83 EFPD JAFNPP 0.99 095 10 79 Ci/yr 963C1 915 C1 (89 57%)

(a) MaX1mum literature values from EPRI Report 1021106.

(b) Typical value from EPRI Report 1021106.

(c) NMP2 Reactor Power Rating inc~eased to 3988 Megawatts thennal.

As long as the core designs and power ratings are not significantly changed, the maximum annual production rates and annual total and carbon dioxide activity releases in Table 1 should be acceptable for use in estimating C-14 gaseous release activity and dose components for the ARERR.

ATTACHMENT 11 Page 5 of 5 Urnt 1 ___x_ Unit 2 _ _ Reporting Pedod: January - December 2020 DOSES TO MEMBERS OF THE PUBLIC DUE TO THEIR ACTIVmES OUTSIDE THE SITE BOUNDARY Direct Radiation Pathway Dose as a result of direct gamma radiation from the site, encompasses doses from direct "shine" from the generating facilities, direct radiation from any overhead gaseous plumes, plume submersion, and ground deposition. This total dose is measured by environmental TLDs. The critical location is based on the closest year-round residence from the generating facilities as well as the closest residence in the critical downwind sector in order to evaluate both direct radiation from the generating facilities and gaseous plumes as determined by the local meteorology During 2020, the closest residence and the critical downwind residence are at the same location.

Table 2 Dose Potentially Rec~ived by the Llkely Most Exposed Member of the Publlc Outside the Site Boundary During 2020 Exposure Pathway Dose Type Dose (mrem)

Fish and Vegetation Total Whole Body No Dose Consumption Total Maximum Organ No Dose Total Whole Body No Dose

-Shoreline Sediment Total Skin of Whole Body No Dose Total Whole Body 2.98E-03 Gaseous Effluents Thyroid 8.33E-03 (excluding C-14)

Maximum Organ Thyroid : 8.33E-03 Gaseous Effluent Total Whole Body 2.BSE-01 (C-14 only) Maximum Organ Bone

1 42E+OO Direct Radiation Total Whole Body 2 24E+O0 Based on these values the maximum total annual dose potentially received by the likely most exposed Member of the Public during 2020 is as follows:

Total Whole Body: 2.52E+OO Total Thyroid: 8.33E-03 Maximum Organ: Bone: 1.42E+00 40 CFR 190 Evaluation The maximum total doses presented in this attachment are the result of operations at the NMP1, NMP2 L.

and the JAFNPP facilities. The maximum organ dose (Bone: 1 42 mrem), maximum thyroid dose (8.33E-03 mrem) and the maximum whole body dose (2.52 mrem) are below the 40 CFR 190 criteria of 25 mrem per calendar year to the maximum exposed organ or the whole body, and below 75 mrem per calendar year to the thyroid.

ATTACHMENT 12 Page 1 of 1 Unrt 1 _a.__ Urnt2 - - B!!Q2r!lag EedQ~i i.!l!D!.!l!rl - D~~!2!!r iQ20 Well Identification # Samples Minimum Maximum

Positive Samples Number Collected Concentration (pCl/1) Concentration (pCl/1)

GMX-MW1* 1 0 <176 <176 MW-1 1 0 <180 <180 MW-5 3 0 <181 <196 MW-6 1 0 <172 <172 MW-7 1 0 <188 <188 MW-8 3 0 <179 <192 MW-91 3 0 <179 <195 MW-101 1 0 <177 <177 MW-11 1 0 <183 <183 MW-12 1 0 <178 <178 MW-13 1 0 <181 <181 MW-14* 1 0 <179 <179 MW-15 3 0 <178 <191 MW-16 1 0 <175 <175 MW-17 3 0 <179 <192 MW-18 3 0 <180 <193 MW-19 1 0 <182 <182 MW-20 1 0 <180 <180 MW-21 1 0 <179 <179 NMP2 MAT2,3 4 1 <183 <226 PZ-1 2 0 <174 <194 PZ-2 1 0 <178 <178 PZ-3 1 0 <177 <177 PZ-4 1 0 <177 <177 PZ-5 1 0 <175 <175 PZ-6 1 0 <184 <184 PZ-7 3 1 <181 242 PZ-8 3 0 <180 <195 PZ-9* 1 0 <178 <178 Notes: * - Control Location 1

- Sentinel well location 2

- NMP2 Groundwater Depression_Cone 3

- Samples collected from storm drain system which includes precipitation 4 nd

- No samples were collected during 2 due to Covid-19

ATTACHMENT 13 Page 1 of 1 Unit 1 X Unit 2 Reporting Period: January - December 2020 Off-Site Dose Calculation Manual <ODCM}

~

.,&#!?F:J"

.,a :saw Exelon Generation NINE MILE POINT UNIT 1 CY-NM-170-301 OFF-SITE DOSE CALCULATION MANUAL (ODCM)

REVISION37 Level 3 - Information Use Revision of this document requires PORC approval and changes are controlled by

'CY-AA-170-3100, Offsite Dose Calculation Manual Revisions

SUMMARY

OF REVISIONS Revision 3 7 (Effective January 2020)

PAGE DATE 1, 2, 5, 6, 8, 9, 11-13/15-18, 21, 24, 25, 36-44, 47-49, 52-81,86-116 February 1987 3, 4, 7, 10, 14, 19, 20, 22, 23, 26-35 December 1987 45,46,50,51,82-85 January1988

29 May 1988 (Reissue)

64, 77, 78 May 27, 1988 (Reissue) i, 19, 21, 22A, 22B, 124, 25, 26, 112 February 1990 i, ii, ill, 12-16, 18, 28-40, 45-47 52, 55, 59-89, 92, 93,97-129 June 1990 91-93, 95 June 1992 3,4,21,92,95a-c Februaryl993 10, 16-20 March 1993 5, 13, 18, 20, 25-30, 65, 79 June 1993 66, 69 December 1993 16, 69 June 1994 10,12 February1995 10, 18, 67, 69 December 1995 5, D-1 June 1996 5, D-1 June 1997 5, D-1 April 1999 D-1 December 1999 iv, 3, 6, 8, 9, 11, 13, 14, 27, 29, 65, 66, 69, 69a December 2001 Added Part I & Revised Part II - II 2-16, II 20-23, II 25, II 26, II 29, II 30 November 2002 iv, v, vii, viii, I 1.0-1 and 2, I 3.1-1, 7 to 9, 11, 14, 18 to 24, 26 and 27, I B 3.1-1, 3 to 7, I 6.0-2, 4, and 5, II 2, II 3, II 4, II 6, II 9 to 11, II 13 to 22, II 42, Figure D-8, Deleted Figures D-17, D-9, D-10 November 2002 CY-NM-170-301 Revision 37 January 2020

SUMMARY

OF REVISIONS (continued)

Revision 3 7 (Effective January 2020)

PAGE DATE x, I 1.0-1, I 3.1-22, I 3.1-38 and 39, I B 3.1-1, I 6.1-0 and 3, II 11, 12, 17, 18, 24 and 25 July 2003 I 3.1-7, I 3.1-8, I 3.1-9, I 3.1-10, I 3.1-11, I 3.1-121_1nd I B 3.1-1 February 2004 II 70, II 71, and II 73 December 2005 I 3.1-5, I 3.1-10, II 6, and II 24 May2006 I 1.0-1, I 3.1-1, I 3.1-2, I 3.1-3, I 3.1-4, I 3.1-5, I 3.1-7, I 3.1-8, I 3.1-9, I 3.1-10 I 3.1-11, I 3.1-12, I 3.1-23, I 3.1-27, I 3.1-28, I B 3.1-1, I B 3.1-8, I 6.0-4, II 2, II 5, II 11, II 17, and II 25 September 2006 viii, I 3.1-8, I 3.1-20, I 3.1-21, I 3.1-27, II 1, II 15, II 23, II 25, II 36- II 43, II 45, II 46, II 48 - II 51, II 53, II 55 - II 65, II 67, II 70, JI 71, II 78, II 83, II 92 February 2007 I 3.1-16, II 68 - II 71, II 73, D-1 October 2009 I 6.0-3 September 2010 I 3.1-9, I 3.1-12, II 5 February 2011 I 3.1-10, I 3.1-25, I B 3.1-7, II 10, II 28, II 30, II 32, II 69 June 2012 I 3.1-16, I B 3.1-2, II 7, II 74 December 2012 Coverpage, II 16, II 17, II 67, II 84-88 June2016 1,3.1-10, I 3.1-11, I B 3.1-1, II 19 May2017 I 3.1-10, I 3.1-11 January 2020 CY-NM-l 70-301 Revision 37 11 January 2020

ODCM - NINE MILE POINT UNIT 1 TABLE OF CONTENTS PAGE List of Figures .................................................................................................................................................... ix INTRODUCTION ...............................*............................................................................................................... X PART I - Radiological Effluent Controls SECTION 1.0: Definitions .............................................................................................':.. .......... I 1.0-0 SECTION 2.0: Not Used SECTIONS 3.0/4.0: Applicability ........................................................................................-.............. I 3.0-0 D 3/4.6.14 RADIOACTIVE EFFLUENT INSTRUMENTATION ................................ I 3.1-1 D 3/4.6.14.a. Liquid Effluent .................... ;............................................................................. I 3.1-1 D 3/4.6.14.b Gaseous Process and Effluent .................................................... ,..................... I 3.1-7 D 3/4.6.15 RADIOACTIVE EFFLUENTS ..................................................................... I 3.1-14 D 3/4.6.15.a.(1) Liquid Concentration ..................................................................................... I 3.1-14 D 3/4.6.15.a.(2) Liquid Dose...................................................................................................... I 3.1-15 D 3/4.6.15.b.(1) Gaseous Dose Rate ....................... .:. ................................................................. I 3.1-19 D 3/4.6.15.b.(2) Gaseous Air Dose ............................................................................................ I 3.1-20 D 3/4.6.15.b.(3) Gaseous Tritium, Iodines and Particulates ................................................... I 3.1-21 D 3/4.6.15.d Uranium Fuel Cycle ........................................................................................ I 3.1-24 D 3/4.6.16 RADIOACTIVE EFFLUENT TREATMENT SYSTEMS ......................... I 3.1-27 D 3/4.6.16.a Liquid ....................................... - .................................................................. I 3.1-27 D 3/4.6.16.b Gaseous ............................................................................................................ I 3.1-27 D 3/4.6.17 Not Used D 3/4.6.18 MARK I CONTAINMENT........................................................... ~ ............... I 3.1-29 D 3/4.6.19 LIQUID WASTE HOWUP TANKS ............................................................ I 3.1-30 D 3/4.6.20 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM ... I 3.1-31 D 3/4.6.21 INTERLABORATORY COMPARISON PROGRAM............................... I 3.1-41 D 3/4.6.22 LAND USE CENSUS ...................................................................................... I 3.1-42 CY-NM-170-301 Revision 37 lll January 2020

ODCM - NINE MILE POINT UNIT 1 TABLE OF CONTENTS (Cont)

PAGE BASES .. ---..**---...--..... - ............................. __ ............ - ................................... 1 B 3.1-0 B 3/4.6.14 BASES FOR RADIOACTIVE EFFLUENT INSTRUMENT ATION ............... I B 3.1-1 B 3/4.6.15 BASES FOR RADIOACTIVE EFFLUENTS-............ - ...................... - ...........1 B 3.1-2 Liquid Concentration ......................................................................................... 1 B 3.1-2 Liquid Dose .......................................................................................................... 1 B 3.1-3 Gaseous Dose Rate .............................................................................................. 1 B 3.1-4 Dose-Noble Gases ...................................................................................*-***-**I B 3.1-5 Dose-lodine-131, Iodine-133, Tritium, an~ Radionuclides in Particulate Form ..........................................................................................1 B 3.1-6 Total Dose-Uranium Fuel Cycle ................................... - .................................. 1 B 3.1-7 B 3/4.6.16 BASES FOR RADIOACTIVE EFFLUENT TREATMENT SYSTEMS ............ ! B 3.1-8 Liquid Radwaste Treatment System .............................-................................... 1 B 3.1-8 Gaseous Effluent Treatment Systems ................................ - ............................. 1 B 3.1-8 B 3/4.6.18 BASES FOR MARK I CONTAINMENT .............................................................. 1 B 3.1-9 B 3/4.6.19 BASES FOR LIQUID WASTE HOLDUP TANKS .............................................. I B 3.1-9 B 3/4.6.20 BASES FOR RADIOWGICAL ENVIRONMENTAL MONITORING PROGRAM ........................................................................... ! B 3.1-10 B 3/4.6.21 BASES FOR INTERLABORATORY COMPARISON PROGRAM ............... ! B 3.1-11 B 3/4.6.22 BASES FOR LAND USE CENSUS ......................................................................! B 3.1-12 SECTION 5.0 Not Used SECTION6.0 ADMINISTRATIVE CONTROLS ......................................................................... I 6.0-1 Reporting Requirements ....................................................................................... I 6.0-2 Special Reports ...................................................................................................... I 6.0-4 CY-NM-170-301 Revision 37 IV January 2020

ODCM - NINE MILE POINT UNIT 1 TABLE OF CONTENTS (Cont)

PAGE 1.0 LIQUID EFFLUENTS..................................................................................................................... II 2 1.1 Setpoint Determinations .............................................................................................. II 2 1.1.1 Basis .................................................. - ........................................................................ II 2 1.1.2 Service Water System Effluent Line Alarm Setpoint................................................ II 2 1.1.3 Liqnid Radwaste Effluent Line Alarm Setpoint ........................................................ II 3 1.1.4 Discussion ...................................................................................................................... II 5 1.1.4.1 Control of Liquid Effluent Batch Discharges ............................................................ II 5 1.1.4.2 Simultaneous Discharges of Radioactive Liquids ...................................................... II 5 1.1.4.3 Sample Representativeness ............. _._ ...................................................................... II 5 1.1.4.4 Liquid Radwaste System Operatio~ ........................................................................... II 6 1.1.4.5 Service Water System Contamination ........................................................................ 117 1.1.4.6 Reactor Building Perimeter Drain Discharges .......................................................... II 7 I

1.2 Liquid Effluent Concentration Calculation ............................................................... 117 1.3 Dose Determinations .................................................................................................... II 8 1.3.1 Maximum Dose Equivalent Pathway.......................................................................... II 8 1.3.2 Dose Projections - Determination of Need to Operate the Liquid Radwaste Treatment System ....... - ........ 1.................................................................................... II 11 2.0 GASEOUS EFFLUENTS .............................................................................................................. II 12 2.1 Setpoint Determinations ............................................................................................ II 12 2.1.1 Basis ............................................................................................................................. II 12 2.1.2 Stack Monitor Setpoints...... - .................................................................................... II 12 2.1.3 Recombiner Discharge (Off Gas) Monitor Setpoints .............................................. II 14 2.1.4 Emergency Condenser Vent Monitor_Setpoint ........................ _ ............................. II 15 CY-NM-170-301 Revision 37 V January 2020

ODCM - NINE MILE POINT UNIT 1 TABLE OF CONTENTS (Cont)

PAGE 2.1.5 Discussion .................................................................................................................... D 15 2.1.5.1 Stack Effluent Monitoring System Description ....................................................... D 15 2.1.5.2 Stack Sample Flow Path - RA GEMS Auxiliary Sample Point .............................. D 15 2.1.5.3 Stack Sample Flow Path - OGESMS ........................................................................ D 16 2.1.5.4 Sample Frequency/Sample Analysis ......................................................................... Il 16 2.1.5.5 1-133 and 1-135 Estimates .......................................................................................... D 16 2.1.5.6 Gaseous Radwaste Treatment System Operation ................................................... D 17 2.2 Dose and Dose Rate Determinations ......................................................................... D 17 2.2.1 Dose Rate ..................................................................................................................... D 18 2.2.1.1 Noble Gases ................................................................................................................. Il 19 2.2.1.2 Tritium, Iodines and Particulates ............................................................................. D 20 2.2.2 Dose .............................................................................................................................. Il 22 2.2.2.1 Noble Gas Air Dose .................................................................................................... II 22 2.2.2.2 Tritium, Iodines and Particulates ............................................................................. D 23 2.2.2.3 Accumulating Doses ................................................................................................... D 24 2.2.3 Dose Projections - Determination of Need to Operate Gaseous Radwaste Treatment System and Ventilation Exhaust Treatment System ............................ D 24 2.3 Critical Receptors ....................................................................................................... D 25 2.4 Refinement of Offsite Doses Resulting From Emergency Condenser Vent Releases .... _. .................................................................................... D 26

(

CY-NM-170-301 Revision 37 Vl January 2020

ODCM - NINE MILE POINT UNIT 1 TABLE OF CONTENTS (Cont)

PAGE 3.0 40 CFR 190 REQUIREMEN'fS .................................................................................................... II 27 3.1 Evaluation of Doses From Liquid Effluents ....................... ;. .................................... II 28 3.2 Evaluation of Doses From Gaseous Effluents .......................................................... II 29 3.3 Evaluation of Doses From Direct Radiation ............................................................ II 30 3.4 Doses to Members of the Public Within the Site Boundary .................................... II 30 4.0 ENVIRONMENTAL MONITORING PROGRAM ................................................................... II 33 4.1 Sampling Stations .................................................:--.................................................... II 33 4.2 lnterlaboratory Comparison Program ..................... ~ ............................. - ......... II 33 4.3 Capabilities for Thermolumlnescent Dosimeters Used for Environmental Measurements.. ................................................................................. II 34 Appendix A Liquid Dose Factor Derivation (A..,) ................................... ~ .................................... II 75 AppendixB Plume Shine Dose Factor Derivation (Bi and Vi) ..................................................... II 78 AppendixC Organ Dose Parameters for Iodine -131 & 133, Particulates and Tritium (Ri) ............ - ................................................................... II 82 AppendixD Diagrams of Radioactive Liquid and Gaseous Effluent Treatment Systems and Monitoring Systems ........................................................... II 92 CY-NM-170-301 Revision 37 vii January 2020

ODCM - NINE MILE POINT UNIT 1 LIST OF TABLES PART I - Radiological Effluent Controls PAGE D 3.6.14-1 Radioactive Liquid Effluent Monitoring Instrumentation ........................ I 3.1-3 D 4.6.14-1 Radioactive Liquid Effluent Monitoring Instrumentation -SR ............... I 3.1-5 D 3.6.14-2 Radioactive Gaseous Process and Effluent Monitoring Instrumentation ............................................................................................. I 3.1-8 D 4.6.14-2 Radioactive Gaseous Process and Effluent Monitoring Instrumentation - SR ................................................................................... I 3.1-12 D 4.6.15-1 Radioactiye Liquid Waste Sampling and Analysis Program - SR. ......... I 3.1-16 D 4.6.15-2 Radioactive Gaseous Waste Sampling and Analysis Program - SR ....... I 3.1-22 D 3.6.20-1 Operational Radiological Environmental Monitoring Program ............. I 3.1-34 D 4.6.20-1 Detection Capabilities for Environmental Sample Analysis Lower Limit of Detection LLD - SR .......................................................... I 3.1-38 D 6.9.3-1 Reporting Level for Radioactivity Concentration in Environmental Samples ..............................................................................:.. I 6.0-5 PART II - Calculational Methodologies Table 1-1 Average Energy Per Disintegration ................................................................ II 36 Tables 2-1 A1a1 Values for the NMP-1 Facility .................................................................. II 37 to 2-8 Table 3-1 Critical Receptor Dispersion Parameters for Ground Level and Elevated Releases .............................................................. Il 45 Table 3-2 Gamma Air and Whole Body Plume Shine Dose Factors for Noble Gases (Bi and V1) ................................................................ II 46 Table 3-3 Immersion Dose Factors for Noble Gases.. ..................................................... Il 47 Tables 3-4 to 3-22 Dose and Dose Rate Factors (Ri) ..................................................................... II 48 Table 3-23 Parameters for the Evaluation of Doses to Real Members of the Public from Gaseous and Liquid Effluents ............................................................................................................. II 67 Table 5.1 Nine Mile Point Nuclear Station Radiological Environmental Monitoring Program Sampling Locations .......................................... ~................................................................ II 68 CY-NM-170-301 Revision 37 Vlll January 2020

ODCM - NINE MILE POINT UNIT 1 LIST OF FIGURES PAGE Figure 5.1-1 Nine Mile Point On-Site Map_ ...................................................................... II 72 Figure 5.1-2 Nine Mile Point Offsite Map .. - ...................................................................... II 73 Figure 5.1.3-1 Site Boundaries .................. _ .. _ ...... _ .. _ ......... - ............................................. II 74 Figure D-0 Piping Instrument and Equipment Symbols ..................................... - ............ D-0 Figure D-1 Radioactive Waste Disposal _ ... - ..................... - ... - ............. - .... - ............. D-1 Figure D-2 Stm Packing, Exhauster, and Recombiner ............................... - ................ - .. D-2 Figure D-3 Reactor Building Vent System-....................................................................... D-3 Figure D-4 Waste Disposal Building Vent System ............................... _, ............................ D-4 Figure D-5 NMP-1 Stack .... _, .............. _ .. _ .......................................... _ .. _ ....................... D-5 Figure D-6 Offgas Building Vent System .................................................... ,........................ 0-6 Figure D-7 This Page/Figure Deleted Figure D-8 Stack Sample and Sample Return ...................................................... _._ ........ D-8 Figures D-9, D-10 These Pages/Figures Deleted FigureD-11 OGESMS Schematic............... - ..................................... - ............................... D-11 CY-NM-170-301 Revision 37 IX January 2020

INTRODUCTION The Offsite Dose Calculation Manual (ODCM) provides the methodology to be used for demonstrating compliance with 10 CFR 20, 10 CFR 50, and 40 CFR 190. The contents of the ODCM are based on Draft NUREG-0472, Revision 3, "Standard Radiological Effluent Technical Specifications for Pressurized Water Reactors," September 1982; Draft NUREG-0473, Revision 2, "Radiological Effluent Technical Specifications for BWR's", July 1979; NUREG 0133, "Preparation of Radiological Effluent Technical Specifications for Nuclear Power Plants,"

October 1978; the several Regulatory Guides referenced in these documents; and, communication with the NRC staff.

Should it be necessary to revise the ODCM, these revisions will be made in accordance with Technic,al Specifications.

The Offsite Dose Calculation Manual (ODCM) is a supporting document of the Technical Specifications Section 6.5.1, "Offsite Dose Cal~ulation Manual." The previous Limiting Conditions for Operation that were contained in the Radiological Effluent Technical Spocifications are now transferred to the ODCM as Radiological Eflluent Controls. The ODCM contains two parts: Radiological Effluent Controls Part I; ~d Calculational Methodologies, Part IL Radiological Effluent Controls, Part I, includes the following: (1) The Radioactive Effluent Controls and Radiological Environmental Monitoring Programs required by Technical Specifications 6.5.3, "Radioactive Effluent Controls Program" and 6.5.1, "Offsite Dose Calculation Manual", respectively, and (2) descriptions of the information that should be included in the Annual Radiological Environmental Operating and Radioactive Effluent Release Reports required by Technical Specifications 6.6.2, "Annual Radiological Environmental Operating Report and 6.6.3 "Radioactive Eflluent Release Report. Calculational Methodologies, Part II, describes methodology and parameters to be used in the calculation of liquid and gaseous effluent monitoring instrumentation alarm/trip setpoints and the calculation of offsite doses due to radioactive liquid and gaseous effluents. The ODCM also contains a list and graphical description of the specific sample locations for the radiological environmental monitoring program, and liquid and gaseous radwaste treatment system configurations.

CY-NM-170-301 Revision 37 X January 2020

PART I-RADIOLOGICAL EFFLUENT CONTROLS CY-NM-170-301 Revision 37 I January 2020

I I PART I - RADIOLOGICAL EFFLUENT CONTROLS Section 1.0 Definitions CY-NM-170-301 Revision 37 I 1.0-0 January 2020 \

1.0 DEFINITIONS DEFINITIONS 1.0 NOTE:

Technical Specfflcatlons defined tenns and the following additional defined tenns are applicable throughout these controls and bases.

Functional (Functionality)

Functionality is an attribute of Structures, Systems, or Components (SSCs) that is not controlled by Technical Specifications. An SSC shall be functional or have functionality when it is capable of performing its specified function as set forth in the Current Licensing Basis (CLB).

Functionality does not apply to specified safety functions, but does apply to the ability of non-Technical Specifications SSCs to perform specified support functions.

Gaseous Radwaste Treatment System A gaseous radwaste treatment system is any system designed and installed to reduce radioactive gaseous effluents by collecting main condenser offgas and providing for delay or holdup for the purpose of reducing the total radioactivity prior to release to the environment.

Member(s) of the Public Member(s) of the public shall include persons who are riot occupationally associated with the Nine Mile Point Nuclear Station. This category does not include employees of owners and operators of Nine Mile Point Nuclear Station and James A Fitzpatrick Nuclear Power Plant, their contractors or vendors who are occupationally associated with Nine Mile Point Unit 1. 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 Nine Mile Point Unit 1.

Milk Sampling Location A milk sampling location is that location where 10 or more head of milk animals are available for the collection of milk samples.

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DEFINITIONS 1.0 Offsite Dose Calculation Manual (ODCM)

The Offsite Dose Calculation Manual shall contain the current methodology and parameters used in the calculation of offsite doses due to radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring alann/trip setpoints, and in the conduct of the environmental radiological monitoring program. The ODCM shall also contain the radioactive effluent controls and radiological environmental monitoring activities, and descriptions of the infonnation that should be included in the Annual Radiological Environmental Operating and Radioactive Effluent Release Reports required by Technical Specifications 6.6.2, "Annual Radiological Environmental Operating Report" and 6.6.3, "Radioactive Effluent Release Report", and Controls D 6.9.1.d and D 6.9.1.e.

Purge - Purging Purge or purging Is 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. The purge is completed when the oxygen-concentration exceeds 19.5 percent. -

Site Boundary The site boundary shall be that line around the Nine Mile Point Nuclear Station beyond which the land is neither owned, leased, nor otherwise controlled by the owners and operators of Nine Mile Point Nuclear Station and James A. Fitzpatrick Nuclear Power Plant.

Source Check A source check shall be the qualitative assessment of channel response when the channel sensor is exposed to a source of increased radioactivity.

Un restricted Area The unrestricted area shall be any area at or beyond the site boundary access to which is not controlled by the owners and operators of Nine Mile Point Nuclear Station and James A. Fitzpatrick Nuclear Power Plant for purposes of protection of individuals from exposure to radiation and radioactive materials, or any area within the site boundary used for residential quarters or for industrial, commercial, institutional, and/or recreational purposes. That area outside the restricted area (10 CFR 20.1003) but within the site boundary will be controlled by the owner as required.,

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DEFINTIONS 1.0 Ventilation Exhaust Treatment System A ventilation exhaust treatment system is any system designed and installed to reduce gaseous radioiodine or radioactive material in particulate form in effluents by passing ventilation or vent exhaust gases through charcoal adsorbers and/or HEPA filters for the purpose of removing iodines or particulates from the gaseous exhaust stream prior to the release to the environment. Such a system is not considered to have any effect on noble gas effluents. Engineered Safety Feature (ESF) atmospheric cleanup systems are not considered to be ventilation exhaust treatment system components.

Venting Venting is 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 not provided or required during venting. Vent, used in system names, does not imply a venting process.

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PART I - RADIOLOGICAL EFFLUENT CONTROLS Sections 3.0/4.0 Applicability

\_

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3.0 CONTROLS APPLICABILITY 3.0/4.0 The Offsite Dose Calculation Manual (ODCM) Part I, Radiological Effluent Controls, is subject to Technical Specifications Section 3.0 requirements, as applicable.

4.0 SURVEILLANCE REQUIREMENTS The ODCM Part I, Radiological Effluent Controls, is subject to Technical Specjfications Section 4.0 requirements, as applicable.

r CY-NM-170-301 Revision 37 13.0-1 January 2020

RADIOACTIVE EFFLUENT INSTRUMENTATION - LIQUID D 3/4.6.14 CONTROLS SURVEILLANCE REQUIREMENT DLCO 3.6.14 RADIOACTIVE EFFLUENT INSTRUMENTATION DSR 4.6.14 RADIOACTIVE EFFLUENT INSTRUMENTATION Applicability: Applicability:

Applies to the functionality of plant instrumentation that Applies to the surveillance of instrumentation that monitors monitors plant effluents. plant effluents.

Objective: Objective:

To assure the functionality of instrumentation to monitor the To verify operation of monitoring instrumentation.

release of radioactive plant effluents.

Specification:

a. Liquid Effluent

a. Liquid Effluent Each radioactive liquid effluenfmonitoring The radioactive liquid effluent monitoring instru- instrumentation channel shall be demonstrated functional by performance of the sensor check, source check, mentation channels shown in Table D 3.6.14-1 shall instrument channel calibration and channel test be functlonal with their alarm set:polnts set to operations at the frequencies shown in Table D 4.6.14-1.

ensure that the llmlts of Control DLCO 3.6.15.a.1 are not exceeded. The alarm Records - Auditable records shall be maintained, in setpoints of these channels shall be determined and accordance with procedures in Part II, of all radioactive adjusted in accordance with the methodology and liquid effluent mof'!itoring instrumentation alarm parameters in Part II. setpoints. Setpoints and setpoint calculations shall be available for review to ensure that the limits of Control Wrth a radioactive liquid effluent monitoring DLCO 3.6.15.a.1 are met.

instrumentation channel alarm setpoint less conservative than a value which will ensure that the limits of DLCO 3.6.15.a.1 are met, immediately suspend the release of radioactive liquid effluents monitored by the affected channel, or declare the channel nonfunctional, or change the setpoint so it is acceptably conservative.

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RADIOACTIVE EFFLUENT INSTRUMENTATION - LIQUID D 3/4.6.14 CONTROLS SURVEILLANCE REQUIREMENT Wrth less than the minimum number of radioactive liquid effluent monitoring instrumentation channels functional, take the action shown in Table D 3.6.14-1~ Restore the instruments to functional status within 30 days, or outline in the next Radioactive Effluent Release Report the cause of the nonfunctionality and how the instruments were or will be restored to functional status.

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RADIOACTIVE EFFLUENT INSTRUMENTATION - LIQUID D 3/4.6.14 TABLED 3.6.14-1 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION Instrument Minimum Channels Functional Applicability

1. Gross Radioactivity Monitors(a)

A. Liquid Radwaste Effluent Line 1(c) At all timesCb>

1(d) At all times<1>

B. Service Water System Effluent Line

2. Flow Rate Measurement Devices 1(e) At all times A Liquid Radwaste Effluent Line-B. Discharge Canal ** -

3. Tank Level Indicating Devices(gl 1(f) At all times A Outside Liquid Radwaste Storage Tanks

-Pumps curves or rated capacity will be utilized to estimate flow.

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RADIOACTIVE EFFLUENT INSTRUMENTATION - LIQUID D 3/4.6.14 NOTES FOR TABLED 3.6.14-1 (a) Provide alarm, but do not provide automatic termination of release.

(b) An operator shall be present in the Radwaste Control Room at all times during a release.

(c) With the number of channels functional less than required by the minimum channels functional requirement, effluent releases may continue provided that prior to initiating a release:

1. At least two independent samples are analyzed in accordance with Specification DSR 4.6.15.a, and

2. At least two technically qualified members of the Facility Staff independently verify the release rate calculations and discharge line vaMng.

Otherwise suspend release of radioactive effluents via this pathway.

(d) Wrth the number of channels functional less than required by the minimum channels functional requirement, effluent releases via this pathway may continue provided that, at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , grab samples are collected and analyzed for gamma radioactivity at a lower limit of detection of at least 5x10-7 microcurie/ml. -

(e) During discharge, with the number of channels functional less than required by the minimum channels functional requirement, effluent releases via this pathway may continue provided the flow rate is estimated at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months during actual releases.

(f) With the number of channels functional less than requireq by the minimum channels functional requirement, liquid additions to this tank may continue provided the tank liquid level is estimated during liquid additions to the tank.

(g) Tanks included in this specification are those outdoor tanks that are not surrounded by liners, dikes or walls capable of holding the tank contents.

(h) Deleted.

(i) Monitoring will be conducted continuously by alternately sampling the reactor building and turbine building service water return lines for approximately 15-minute intervals.

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RADIOACTIVE EFFLUENT INSTRUMENTATION-LIQUID D 3/4.6.14 TABLED 4.6.14-1 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION Surveillance Requirement Instrument Sensor Check Source Check!fl Channel Test Channel Calibration Channel Test

1. Gross Beta or Gamma Radioactivity Monitors

a. Liquid Radwaste Effluent Line Once/day* Once/discharge* Once/3 months(e)* Oncelyear{bl*

Line Once/day Once/92 days Once/184 days<el Once/24 months(bl

b. Service Water Effluent Line

2. Flow Rate Measurement Devices Once/day(c) None None Once/24 months*

a. Liquid Radwaste Effluent Line None None None Once/year

b. Discharge Cana~d)

3. Tank Level Indicating Devices(e) Once/day** None Once/3 months Once/18 months

a. Outside Liquid Radwaste Storage Tanks

Required prior to removal of blank flange in discharge line and until blank flange is replaced.

- During liquid addition to the tank.

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RADIOACTIVE EFFLUENT INSTRUMENTATION - LIQUID D 3/4.6.14 NOTES FOR TABLE-D 4.6.14-1 (a) The channel test shall also demonstrate that control room alarm annunciation occurs if any of the following conditions exist:

1. Instrumentation indicates measured levels above the alarm setpoint.

2. Instrument indicates a downscale failure.

3. Instrument controls not set in operate mode.

(b) The channel calibration shall be performed using one or more reference standards certified by the National Institute of Standards and Technology (NIST), or using standards that are traceable to the NIST or using actual samples of liquid waste that have been analyzed on a system that has been calibrated with NIST traceable sources. These standards shall permit calibrating the system over its intended range of energy and measurement.

(c) Sensor check shall consist of verifying indication of flow during periods of release. Sensor check shall be made at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months on days on which continuous, periodic or batch releases are made.

(d) Pump performance curves or rated data may be used to estimate flow.

(e) Tanks included in this specification are those outdoor tanks that are not surrounded by liners, dikes or walls capable of holding the tank contents.

(f) Source check may consist of an installed check source, response to an external source, or (for liquid radwaste monitors) verification within 30 minutes of commencing discharge of monitor response to effluent.

CY-NM-170-301 Revision 37 I 3.1-6 January 2020

RADIOACTIVE EFFLUENT INSTRUMENTATION - GASEOUS D 3/4.6.14 CONTROLS SURVEILLANCE REQUIREMENT

b. Gaseous Process and Effluent b. Gaseous Process and Effluent The radioactive gaseous process and effluent Each radioactive gaseous process and monitoring Instrumentation channels shown in Table D effluent monitoring instrumentation channel 3.6.14-2 shall be functional. The Offgas process shall be demonstrated functional by monitor alarm setpoint shall be set to ensure that the performance of the sensor check, source limits of Technical Specification 3.6.15 are not check, instrument channel calibration and exceeded. The Effluent monitor alarm setpoints shall instrument channel test operations at the be set to ensure that the limits of Control DLCO , frequencies shown in Table D 4.6.14-2.

3.6.15.b.1 are not exceeded. The alarm setpoints of Auditable records shall be maintained of these channels shall be determined and adjusted in the calculations made, in accordance with accordance with the methodology and parameters in procedures in Part II, of radioactive Part II. gaseous process and effluent monitoring instrumentation alarm setpoints. Setpoints With a radioactive gaseous process and effluent and setpoint calculations shall be available monitoring instrumentation channel alarm setpoint less for review to ensure that the limits of conservative than required by the above specification, Control DLCO 3.6.15.b.1 are met.

immediately suspend the release of radloactive gaseous effluents monitored by the affected channel, or declare the channel nonfunctional, or change the setpoint so it Is acceptably conservative.

Wrth less than the minimum number of radioactive gaseous process and effluent monitoring instrumentation channels functional, take the action shown in Table D 3.6.14-2. Restore the instruments to functional status within 30 days or outline in the next Radioactive Effluent Release Report the cause of the nonfunctionality and how the instruments were or will be restored to functional status.

CY-NM-170-301 Revision 37 I 3.1-7 January 2020

RADIOACTIVE EFFLUENT INSTRUMENTATION -GASEOUS D 3/4.6.14 TABLED 3.6.14-2 RADIOACTIVE GASEOUS PROCESS AND EFFLUENT MONITORING INSTRUMENTATION

Minimum Instrument . Channels Functional Applicability Action

1. Stack Effluent Monitoring

a. Noble Gas Activity Monitors (1) High Range 2 * (a)

(2) Low Range 1 * (i)

b. Iodine Sampler Cartridge 1 * (b)

C. Particulate Sampler Filter 1 * (b)

d. Sampler Flow Rate Measuring Device 1 * (c)

e. Stacie Gas Flow Rate Measuring Device 1 * (c), (d)

2. Deleted

At all times.

- Note Deleted.

CY-NM-170-301 Revision 37 I 3.1-8 January 2020

RADIOACTNE EFFLUENT INSTRUMENTATION - GASEOUS D 3/4.6.14 TABLED 3.6.14-2 (cont'd)

RADIOACTIVE GASEOUS PROCESS AND EFFLUENT MONITORING INSTRUMENTATION Minimum Instrument Channels Functional Applicability Action

3. Condenser Air Ejector Process Monitor (Offgas System Recombiner Discharge)

a. Noble Gas Activity Monitor 2 (g)

b. Offgas System Flow Rate 1 (c)

Measuring Device

4. Emergency Condenser System Effluent (h)

a. Noble Gas Activity Monitor 1 per vent During operation of the main condenser air ejector During power operating conditions and whenever the reactor coolant temperature is greater than 212°F except for hydrostatic testing with the reactor not critical.

CY-NM-170-301 Revision 37 I 3.1-9 , January 2020

RADIOACTIVE EFFLUENT INSTRUMENTATION - GASEOUS D 3/4.6.14 NOTES FOR TABLED 3.6.14-2 (a) (1) With the number of channels functional 1 less than required by the minimum channels functional requirement, effluent releases via this pathway may continue provided:

'(a) The nonfunctional channel is placed in the tripped condition, OR (b) Vent and Purge valves are closed and administratively controlled, OR (c) Primary containment integrity is not required.

(2) With the number of channels functional 2 less than required by the minimum channels functional requirement, effluent releases via this pathway may continue provided:

(a) Gross.activity is recorded every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when both Noble Gas Activity Radiation Monitor indications are functional.

OR (b) Grab samples are taken once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and these samples are analyzed for gross activity within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . ,

(b) Wrth the number of channels functional less than required by the minimum channels functional requirements, effluent releases via this pathway may continue provided that samples are continuously collected with auxiliary sampling equipment starting within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months of discovery in accordance with the requirements of Table D 4.6.15-2. When OGESMS and the auxiliary sampling equipment are both non-functional.-

Attachment 3 of Chemistry procedure N1-CSP-V304 may be used for obtaining continuous particulate and iodine samples.

(c) Wrth the number of channels functional less than required by the minimum channels functional requirements, effluent releases via this pathway may continue provided the flow rate is estimated once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

(d) Stack gas flow rate may be estimated by exhaust fan operating configuration.

(e) Deleted (f) Deleted CY-NM-170-301 Revision 37 I 3.1-10 January 2020

RADIOACTIVE EFFLUENT INSTRUMENTATION - GASEOUS D 3/4.6.14 NOTES FOR TABLED 3.6.14-2 (cont'd)

(g) Wrth one or more channels nonfunctional, perfonn one of the following AND restate the nonfunctional channel(s) to functional status within 30 days U):

(1) With one or more channel nonfunctional due to alann function only, every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months record gross activity of both channels as read on the meters in the control room and verify within limits OR (2) Take grab samples within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter AND analyze samples for gross activity within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of sampling completion.

OR (3) Place the nonfunctional channel in the tripped condition within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months (h) Wrth the number of channels functional less than required by the minimum channels functional requirements, steam release via this pathway may commence or continue provided vent pipe radiation dose rates are monitored once per four hours.

(i) Wrth the number of channels functional less than required by the minimum channels functional requirement, effluent releases via this pathway may continue provided:

(1) Gross activity is recorded every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when both Noble Gas Activity Radiation Monitor indications are functional.

OR (2) Grab samples are taken once per 12_hours and these samples are analyzed for gross activity within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

U) If nonfunctional channel(s) not restored within specified time, explain why the inoperability was not corrected in a timely manner in the next Radioactive Effluent Release Report.

CY-NM-170-301 Revision 37 13.1-11 January 2020

RADIOACTIVE EFFLUENT INSTRUMENTATION - GASEOUS

. D 3/4.6.14 TABLED 4.6.14-2 RADIOACTIVE GASEOUS PROCESS AND EFFLUENT MONITORING INSTRUMENTATION Surveillance Requirements Instrument Sensor Check Source Check Channel Test Channel Calibration

1. Stack Effluent Monitoring System

a. Noble Gas Activity Monitors Once/day<al Once/92 days Once/184 days<gJ Once/24 monthsCbl (High Range and Low Range)

b. Iodine Sampler Cartridge None None None None

c. Particulate Sampler Filter None None None None

d. Sampler Flow Rate Measuring Device Once/day<al None None Once/24 months

e. Stack Gas Flow Rate Measuring Device

Once/day None None .Once/24 months

2. Deleted

3. Condenser Air Ejector Process Monitor (Offgas System Recombiner Discharge)

a. Noble Gas Activity Monitor Once/day(fl Once/92 days Once/24 months<c) Once/24 monthsCbl

b. Offgas System Flow Rate Measuring Once/day(fl None None Once/24 months Device

4. Emergency Condenser System Effluent Once/dayChl Once/92 days Once184 days<gl Once/24 monthsCb>

~ a. Noble Gas Activity Monitor CY-NM-170-301 Revision 37 I 3.1-12 January 2020

CY-NM-170-30 I Revision 37 I 3.1-13 January 2020

RADIOACTIVE EFFLUENT INSTRUMENTATION - GASEOUS D 3/4.6.14 NOTES FOR TABLED 4.6.14-2 (a) At all times.

(b) The channel calibration shall be performed using one or more of the reference standards certified by the National Institute of Standards and Technology (NIST), standards that are traceable to the NIST or using actual samples of gaseous effluent that have been analyzed on a system that has been calibrated with NIST traceable sources. These standards shall pennit calibrating the system over its intended range of energy and measurement.

(c) The channel function test shall demonstrate that control room alann annunciation occurs if either of the following conditions exist:

1) Instrument indicates measured levels above the Hi or Hi Hi alann setpoint.

2) Instrument indicates a downscale failure.

The channel function test shall also demonstrate that automatic isolation of this pathway occurs if either of the following conditions exist:

1) Instruments indicate two channels above Hi Hi alarm setpoint

2) Instruments indicate one channel above Hi Hi alarm setpoint and one channel downscale.

{d) Deleted (e) Deleted (f) During operation of the main condenser air ejector.

(g) The channel test shall produce upscale and downscale annunciation.

(h) During power operating conditions and whenever the reactor coolant temperature is greater than 212°F except for hydrostatic testing with the reactor not critical.

CY-NM-170-301 Revision 37 I 3.1-14 January 2020

RADIOACTIVE EFFLUENTS - LIQUID CONCENTRATION D 3/4.6.15 CONTROLS SURVEILLANCE REQUIREMENT DLCO 3.6.15 RADIOACTIVE EFFLUENTS DSR 4.6.15 RADIOACTIVE EFFLUENTS Applicability. Applicability:

Applies to the radioactive effluents from the station. Applies to the periodic test and recording requirements of the station process effluents.

Objective: Objective:

To assure that radioactive material is not released to the To ascertain that radioactive effluents from the station are environment in any uncontrolled manner and is within the limits within the allowable values of 10CFR20, Appendix B and of 10CFR20 and 10CFR50 Appendix I. 10CFR50, Appendix I.

Specification:

a. Liquid

a. Liquid (1) Concentration (1) Concentration The concentration of radioactive material released in liquid effluents to unrestricted areas shall be limited to ten times the concentrations specified in Radioactive liquid wastes shall be sampled and 10CFR Part 20, Appendix B, Table 2, Column 2 for analyzed according to the sampling and analysis radionuclides other than dissolved or entrained program of Table D 4.6.15-1.

noble gases. For dissolved or entrained noble gases, the concentration shall be limited to 2 x 10--4 The results of the radioactivity analyses shall be microcuries/ml total activity.

used in accordance with the methodology and parameters in Part II to assure that the' Should the concentration of radioactive material concentrations at the point of release are released in liquid effluents to unrestricted areas maintained within the limits of Control DLCO exceed the above limits, restore the concentration 3.6.15.a.(1 ).

to within the above limits immediately.

CY-NM-170-301 Revision 37 I 3.1-15 January 2020

RADIOACTIVE EFFLUENTS - LIQUID DOSE

- D 3/4.6.15 CONTROLS SURVEILLANCE REQUIREMENT (2) Dose (2) Dose The dose or dose commitment to a member of the Cumulative dose contributions from liquid public from radioactive materials in liquid effluents effluents for the current calendar quarter and the released, from each reactor unit, to unrestricted current calendar year shall be determined in areas (see Figures 5.1-1) shall be limited: accordance with the methodology and parameters in Part II monthly.

(a) During any calendar quarter to less than or equal to 1.5 mrems to the total body and to less than or equal to 5 mrems to any organ, and (b) During any calendar year to less than or equal to 3 mrems to the total body and to less than or equal to 10 mrems to any organ.

With the calculated dose from the release of radioactive materials in liquid effluents exceeding any of the above limits, prepare and submit to the Commission within 30 days, pursuant to Control D 6.9.3, a Special Report that identifies the cause(s) for exceeding the limit(s) and defines the corrective actions that have been taken to reduce the releases and the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with the above limits.

CY-NM-170-301 Revision 37 I 3.1-16 January 2020

RADIOACTIVE EFFLUENTS - LIQUID D 3/4.6.15 TABLE D 4.6.15-1 RADIOACTIVE LIQUID WASTE SAMPLING AND ANALYS!S PROGRAM Surveillance Requirement Sampling Frequency Minimum Lower Limit<a> of Detection Liquid Release Type Analysis Frequency Type of Activity Analysis (LLD) (µCl/ml)

A. Batch Waste(b> Tanks *

Each Batch Each Batch Principal Gamma<cJ Emitters 5 X 10-7 1-131 1 X 10-S Each Batch<dl Each Batch(d) Dissolved and Entrained Gases 1 X 10-5 (Gamma Emitters)

Monthly Composite<0 > H-3 1 X 10-5 Each Batch Gross Alpha 1 X 10-7

Quarterly Composite<0 > Sr-89, Sr-90 5 X 10-S Each Batch Fe 1 X 10-S B. Service Water System Once/month(f) Once/month(!) Principal Gamma<cl Emitters 5 X 10-7 Effluent 1-131 1 X 10-S Dissolved and Entrained Gases 1 X 10-15 H-3 1 X 10"6 Gross Alpha 1 X 10-7 Once/quarter(!) Once/q uarter(l) Sr-89, Sr-90 5x1o-8 Fe-55 1 X 10-S C. Reactor Building Perimeter Once/quarter Quarterly Composite<0 l Principal Gamma<c> Emitters 5 X 10-7 Dram 1-131 1 X 10-S H-3 1 X 1~

Gross Alpha 1 X 10-7 Sr-89, Sr-90 5 X 10-S Fe-55, Ni-63 1 X 10-S

Completed pnor to each release.

CY-NM-170-301 Revision 37 I 3.1-17 January 2020

RADIOACTIVE EFFLUENTS - LIQUID D 3/4.6.15 NOTES FOR TABLED 4.6.15-1 (a) The LLD is defined, for purposes of these specifications, as the smallest concentration of radioactive material in a sample that will yield a net count above system background that will be detected with 95 percent probability with only 5 percent probability of falsely concluding that a blank observation represents a "real" signal. For a particular measurement system which may include radiochemical separation:

LLD= 4.66 Sb E*V*2.22 x 106*Y*exp (-AAt)

Where:

LLD is the "a priori" lower limit of detection as defined above, as microcuries per unit mass or volume, Sb 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, as counts per disintegration, V is the sample size in units of mass or volume, 2.22 x 106 is the number of disintegrations per minute per microcurie, Y is the fractional radiochemical yield, when applicable, 1o is the radioactive decay constant for the particular radionuclide, and i1t for plant effluents is the elapsed time between the midpoint of sample collection and time of counting.

Typical values of E, V, Y and i1t should be used in the calculation.

It should be recognized that the LLD is defined as a before the fact limit representing the capability of a measurement system and not as an after the fact for a particular measurement.

CY-NM-170-301 Revision 37 13.1-18 January 2020

RADIOACTIVE EFFLUENTS - LIQUID D 3/4.6.15 NOTES FOR TABLED 4.6.15-1 (b) A batch release is the discharge of liquid wastes of a discrete volume. Prior to sampling for analyses, each batch shall be isolated and then thoroughly mixed to assure representative sampling.

(c) The principal gamma emitters for which the LLD specification applies exclusively are the following radionuclides: Mn-54, Fe-59, Co-58, C~0. 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 considered. Other gamma peaks that are identifiable, together with those of the above nuclides, shall also be analyzed and,, reported in the Radioactive Effluent Release Report.

(d) If more than one batch is released in a calendar month, only one batch need be sampled and analyzed during that month.

(e) A composite sample is one in which the quantity of liquid sampled is proportional t6 ijle quantity of liquid waste discharged and in which the method of sampling employed results in a specimen that is representative of the liquids released.

(f) If the alarm setpoint of the service water effluent monitor, as determined by the method presented in Part II, Is exceeded, the frequency of sampling shall be increased to daily until the condition no longer exists. Frequency of analysis shall be increased to daily for principal gamma emitters (including dissolved and entrained gases) and an incident composite for H-3, gross alpha, Sr-89, Sr-90 and Fe-55.

CY-NM-170-301 Revision 37 13.1-19 January 2020

RADIOACTIVE EFFLUENTS - GASEOUS DOSE RATE D 3/4.6.15 CONTROLS SURVEILLANCE REQUIREMENT

b. Gaseous b. Gaseous (1) Dose Rate (1) Dose Rate The dose rate due to radioactive materials The dose rate due to noble gases in gaseous released In gaseous effluents from the site to effluents shall be determined to be within the limits of areas at or beyond the site boundary shall be Control DLCO 3.6.15 in accordance with the limited to the following: methodology and parameters in Part II.

The dose rate due to iodine-131, iodine-133, tritium (a) For noble gases: Less than or equal and all radionuclides in particulate form with half lives to 500 mrems/year to the whole body greater than 8 days in gaseous effluents shall be and less than or equal to 3000 determined to be within the limits of Control DLCO mrems/year to the skin, and 3.6.15 in accordance with methodology and parameters in Part II by obtaining representative (b) For iodine-131, iodine-133, tritium and samples and performing analyses in accordance with all radionuclides in particulate form the sampling and analysis program specified in Table with half lives greater than 8 days: D 4.6.15-2.

Less than or equal to 1500 mrems/year to any organ.

Wrth the dose rate(s) exceeding the above limits, without delay restore the release rate to within the above limits(s).

CY-NM-170-301 Revision 37 I 3.1-20 January 2020

RADIOACTIVE EFFLUENTS - GASEOUS DOSE D 3/4.6.15 CONTROLS SURVEILLANCE REQUIREMENT (2) Air Dose (2) Air Dose The air dose due to noble gases released in gaseous Cumulative dose contributions for the current effluents, from each reactor unit, to areas beyond calendar quarter and current calendar year for the site boundary shall be limited to the following: noble gases shall be determined monthly in accordance with the methodology and parameters in Part II.

(a) During any calendar quarter: Less than or equal to 5 milliroentgen for gamma radiation and less than or equal to 10 mrads for beta radiation and, (b) During any calendar year: Less than or equal to 10 milliroentgen for gamma radiation and less than or equal to 20 mrads for beta radiation.

With the calculated air dose from radioactive noble gases in gaseous effluents exceeding any of the above limits, prepare and submit to the Commission within 30 days, pursuant to Control D 6.9.3, a Special Report that identifies the cause(s) for exceeding the limit(s) and defines the corrective actions that have been taken to reduce the releases and the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with the above limits.

CY-NM-170-301 Revision 37 I 3.1-21 January 2020

RADIOACTIVE EFFLUENTS - GASEOUS DOSE D 3/4.6.15 CONTROLS SURVEILLANCE REQUJREMENT (3) Tritium, Iodines and Particulates (3) Tritium, Iodines and Particulates The dose to a member of the public from lodine-131, Cumulative dose contributions for the current iodine-133, tritium and all radionuclides in particulate calendar quarter and current calendar year for form with half lives greater than 8 days In gaseous iodine-131, iodine-133, tritium and radionuclides effluents released, from each reactor unit, to areas in particulate form with half lives greater than 8 beyond the site boundary shall be limited to the days shall be determined monthly in accordance following: with the methodology and parameters in Part II.

(a) During any calendar quarter: Less than or equal to 7.5 mrems to any organ and, (b) ,. During any calendar year: Less than or equal to 15 mrems to any organ.

With the calculated dose from the release of iodine-131, iodine-133, tritium and radionuclides in particulate form with half lives greater than 8 days, in gaseous effluents exceeding any of the above limits, prepare and submit to the Commission within 30 days, pursuant to Control D 6.9.3, a Special Report that identifies the cause(s) for exceeding the limit and defines the corrective actions that have been taken to reduce the releases and the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with the above limits.

CY-NM-170-301 Revision 37 I 3.1-22 January 2020

RADIOACTIVE EFFLUENTS - GASEOUS D 3/4.6.15 TABLE D 4.6.15-2 RADIOACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM Surveillance Requirements Minimum Lower Limit<*l of Detection Gaseous Release Type Sampling Frequency Analysis Freq*uency Type of Activity Analysis (LLD) (µCl/ml)

A Containment Purge(bl Each Purge Prior to each release Principal Gamma Emitters<cJ 1 X 10-4 Grab Sample. Each Purge Priricipal Gamma Eniitters<cJ 1 x1o-4 H-3 1 X 10-6 B. Stack Once/Month(dl Once/Month(dl Principal Gamma Emitters<cJ 1 X 10-4 Once/Month(hl Once/Month H-3 1 X 10-6 C. Stack Continuous<eJ Once/Week<fl 1-131 1 X 10-12 Charcoal Sample Continuous<eJ OnceNVeek<fl Principal Gamma Emitters<cJ 1 X 10-11 Particulate Sample Continuous<eJ Once/Month Composite Gross alpha, Sr-89, Sr-90 1 X 10-11 Particulate Sample Continuous<eJ Noble Gas Monitor Noble Gases, Gross Gamma 1 X 1Q- 5 (g) or Principal Gamma Emitters<cJ CY-NM-170-301 Revision 37 I 3.1-23 January 2020

RADIOACTIVE EFFLUENTS - GASEOUS D 3/4.6.15 NOTES FOR TABLED 4.6.15-2 (a) The LLD is defined in notation (a) of Table D 4.6.15-1.

(b) Purge is defined in Section 1 0.

(c) The principal gamma emitters for which the LLD specification applies exclusively are the following radionuclides: Kr-87, Kr-88, Xe-133, Xe-135 and Xe-138 for gaseous emissions and Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, Cs-134, Cs-137, Ce-141, 1-131 and Ce-144 for particulate emissions.

This list does not mean that only these nuclides are to be considered. Other gamma peaks that are identifiable, together with those of the above nuclicles, shall also be analyzed and reported in the Radioactive Effluent Release Report pursuant to Technical Specification 6.6.3, "Radioactive Effluent Release Report", and Control D 6.9.1.

(d) Sampling and analysis shall also be perfonnecl following shutdown, startup or an increase on the recombiner discharge monitor of greater than 50 percent, factoring out increases due to changes in thennal power level or dilution flow; or when the stack release rate is in excess of 1000 µCi/second and steady-state gaseous release rate increases by 50 percent (e) The sample flow rate and the stack flow rate shall be known for the time period covered by each dose or dose rate calculation made in accordance with Controls DLCO 3.6.15.b.(1 ).(b) and DLCO 3.6.15.b.(3).

(f) When the release rate is in excess of 1000 µCi/sec and steady state gaseous release rate increases by 50 percent, the iodine and particulate collection device shall be removed and analyzed to detennine the changes in iodine-131 and particulate release rate. The analysis shall be done daily following each change until it is shown that a pattern exists which can be used to predict the release rate; after which it may revert to weekly sampling frequency. When samples collected tor 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months are analyzed, the corresponding LLD's may be increased by a factor of 10.

(g) When the continuous Noble Gas Monitor is nonfunctional the LLD for noble gas gamma analysis shall be 1 x 1 ~ µCi/cc.

(h) Tritium grab samples shall be taken weekly from the station ventilation exhaust (stack) when fuel is offloaded until stable tritium release levels can be demonstrated.

CY-NM-170-301 Revision 37 I 3.1-24 January 2020

RADIOACTIVE EFFLUENTS - MAIN CONDENSER, URANIUM FUEL CYCLE D 3/4.6.15 CONTROLS SURVEILLANCE REQUIREMENT

c. Deleted C. Main Condenser The radioactivity rate of noble gases at the recombiner discharge shall be continuously monitored in accordance with Table D 3.6.14-2.

d. Uranium Fuel Cycle d. Uranium Fuel Cycle The annual (calendar year) dose or dose Cumulative,dose contributions from liquid and commibnent to any member of the public due gaseous effluents shall be determined in to releases of radioactivity and to radiation accordance with Controls from uranium fuel cycle sources shall be DSR 4.6.15.a.(2), DSR 4.6.15.b.(2) and limited to less than or equal to 25 mrems to DSR 4.6.15.b.(3) and in accordance with the the whole body or any organ, except the methodology and parameters in Part II.

thyroid, which shall be limited to less than or equal to 75 mrems.

CY-NM-170-301 Revision 37 I 3.1-25 January 2020

RADIOACTIVE EFFLUENTS - URANIUM FUEL CYCLE D 3/4.6.15 CONTROLS SURVEILLANCE REQUIREMENT Wrth the calculated doses from the release of radioactive Cumulative dose contributions from direct radiation from the materials in liquid or gaseous effluents exceeding twice the reactor units, ISFSI, and from radwaste storage tanks shall be limits of Controls DLCO 3.6.15.a(2), DLCO 3.6.15.b(2) and determined in accordance with the methodology and parameters DLCO 3.6.15.b(3), calculations shall be made including direct in Part II. This requirement is applicable only under conditions radiation contributions from the reactor units, the Independent set forth in Control DLCO 3.6.15.d.

Spent Fuel Storage Installation (ISFSI), and from outside storage tanks to determine whether the above listed 40CFR190 limits have been exceeded. If such is the case, prepare and submit to the Commission within 30 days, pursuant to Control D 6.9.3, a Special Report that defines the corrective action to be taken to reduce subsequent releases to prevent recurrence of exceeding the above limits and includes the schedule for achieving conformance with the above limits. This Special Report, as defined in 10 CFR Part 20.2203(a)(4), shall include an analysis that estimates the radiation exposure (dose) to a member of the public from uranium fuel cycle sources, including all effluenf pathways and direct radiation, for the calendar year that includes the release(s) covered by this report.

CY-NM-170-301 Revision 37 I 3.1-26 January 2020

RADIOACTIVE EFFLUENTS - URANIUM FUEL CYLE

' D3/4.6.15 CONTROLS SURVEILLANCE REQUIREMENT It shall also describe levels of radiation and concentrations of radioactive material involved, and the cause of the exposure levels or concenJrations. If the estimated dose(s) exceeds the above limits, and if the release condition resulting in violation of 40CFR Part 190 has not already been corrected, the Special Report shall include a request for a variance in accordance with the provisions of 40CFR 190. Submittal of the report is considered a timely request and a variance is granted l!ntil staff action on the request is complete. '

CY-NM-170-301 Revision 37 I 3.1-27 January 2020

RADIOACTIVE EFFLUENT TREATMENT SYSTEMS D 3/4.6.16 CONTROLS SURVEILLANCE REQUIREMENT DLCO 3.6.16 RADIOACTIVE EFFLUENT TREATMENT SYSTEMS DSR 4.6.16 RADIOACTIVE EFFLUENT TREATMENT SYSTEMS Applicability: Applicability:

Applies to the operating status of the liquid- and gaseous Applies to the surveillance requirements for the liquid and effluent treatment systems. gaseous effluent treatment systems.

Objective: Objective:

To assure functionality of the liquid and gaseous effluent To verify functionality of the liquid and gaseous effluent treatment system. treatment system.

Specification: Specification:

)

a. Liquid a. Liquid The liquid radwaste treatment system shall be used to Doses due to liquid releases to unrestricted areas reduce the radioactive materials In liquid wastes prior shall be projected prior to the release of each batch of to their discharge when the projected dose due to the liquid radioactive waste in accordance with the liquid effluent, from each unit, to the Unrestricted methodology and parameters in Part II.

Areas would exceed 0.06 mrem to the total bcx:ly or 0.2 mrem to any organ for any batch. b. Gaseous (1) Doses due to gaseous releases to areas at or

b. Gaseous - beyond the site boundary shall be calculated in accordance with the methodology and (1) The Gaseous Radwaste Treatment System parameters In Part II.

shall be functional. The Gaseous Radwaste Treatment System shall be used to reduce radioactive materials in gaseous waste prior to their discharge as necessary to meet the requirements of Control DLCO 3.6.15.

CY-NM-170-301 Revision 37 13.1-28 January 2020

RADIOACTIVE EFFLUENT TREATMENT SYSTEMS D 3/4.6.16 CONTROLS With gaseous radwaste from the main condenser air ejector system being discliarged without treatment for more than 7 days, prepare and submit to the Commission within 30 days, pursuant to Control D _6.9.3, a Special Report that identifies the nonfunctional equipment and the reason for its nonfunctionality, actions taken to restore the nonfunctional equipment to functional status, and a summary description of those actions taken to prevent a recurrence.

(2) The Ventilation Exhaust Treatment System shall be functional (2) and appropriate portions of this system shall be used to reduce releases of radioactivity when the projected doses in NOTE:

31 days due to gaseous effluent releases, from each unit, to Only required to be met when the Ventilation areas beyond the site boundary would exceed 0.3 mrem to Exhaust Treatment System is not being fully utilized.

any organ of a member of the public.

Wrth radioactive gaseous waste being discharged without Project the doses from the iodine and particulate treatment and in excess of the above limit, complete a CR releases from each unit to areas beyond the Site evaluation of the degraded condition within 30 days that Boundary at least every 31 days.

identifies the nonfunctional equipment, the reason for the nonfunctionality, and plans and schedule to restore the equipment to functional status.

CY-NM-170-301 Revision 37 I 3.1-29 January 2020

MARK I CONTAINMENT D 3/4.6.18 CONTROLS SURVEILLANCE REQUIREMENT DLCO 3.6.18 MARK I CONTAINMENT DSR 4.6.18 MARK I CONTAINMENT Applicability: Applicability:

Applies to the venting/purging of the Mark I Applies to the surveillance requirement for venting Containment and purging of the Mark, I Containment when required to be vented/purged through the Emergency Ventilation System.

Objective: Objective:

To assure that the Mark I Containment is vented/purged so To verify that the Mark I Containment is vented through the that the limits of Controls DLCO 3.6.15.b(1) and Emergency Ventilation System when required.

DLCO 3.6.15. b(3) are met.

Specification: Specification:

The Mark I Containment drywell shall be vented/ purged The containment drywall shall be determined to be through the Emergency Ventilation System unless Controls aligned for venting/purging through the Emergency DLCO 3.6.15.b.(1) and DLCO 3.6.15.b.(3) can be met Ventilation System within four hours prior to start of without use of the Emergency Ventilation System. and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months during venting/purging of the drywall.

If these requirements are not satisfied, suspend all venting/purging of the drywall.

CY-NM-170-301 Revision 37 I 3.1-30 January 2020

LIQUID WASTE HOLDUP TANKS*

D 3/4.6.19 CONTROLS SURVEILLANCE REQUIREMENT DLCO 3.6.19 LIQUID WASTE HOLDUP TANKS* DSR 4.6.19 LIQUID WASTE HOLDUP TANKS Applicability: Applicability:

Applies to the quantity of radioactive material Applies to the surveillance requirements for outdoor that may be stored in an outdoor liquid waste liquid waste holdup tanks.

holdup tank.

Objective: Objective:

To assure that the quantity of radioactive material stored in To verify the quantity of radioactive material stored in an outdoor holdup tanks does not exceed a sp~fied level. outdoor liquid waste holdup tank.

Specification: _Specification:

The quantity of radioactive material contained in an outdoor The quantity of radioactive material contained in each of the liquid waste tank shall be limited to less than or equal to 1O tanks listed in Control DLCO 3.6.19 shall be determined to curies, excluding tritium and dissolved or entrained noble be within the limit of Control DLCO 3.6.19 by analyzing a gases. representative sample of the tank's contents at least weekly when radioactive materials are being added to the tank.

Wrth the quantity of radioactive material in any such tank exceeding the above limit, immediately suspend all additions of radioactive material to the tank. Within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months reduce the tank contents to within the limit and describe the events leading to this condition in the next Radioactive Effluent Release Report.

Tanks included in this Control are those outdoor tanks that are not surrounded by liners, dikes, or walls capable of holding the tank contents and that do not have tank overflows and surrounding area drains connected to the liquid radwaste treatment system. -

CY-NM-170-301 Revision 37 13.1-31 January 2020

RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM D 3/4.6.20 CONTROLS SURVEILLANCE REQUIREMENT DLCO 3.6.20 RADIOLOGICAL ENVIRONMENTAL MONITORING DSR 4.6.20 RADIOLOGICAL ENVIRONMENTAL PROGRAM MONITORING PROGRAM Applicability: Applicability:

Applies to radiologi~I samples of station Applies to the periodic sampling and monitoring environs. requirements of the radiological environmental monitoring program.

Objective: Objective:

To evaluate the effects of station operations and radioactive To ascertain what effect station operations and radioactive effluent releases on the environs and to verify the effluent releases have had upon the environment.

effectiveness of the controls on radioactive material sources.

Specification: Specification:

The radiological environmental monitoring program shall be The radiological environmental monitoring samples shall be conducted as specified in Table D 3.6.20-1. collected pursuant to Table D 3.6.20-1 from the specific locations given in the table and figure(s) in Part II and shall With the radiological environmental monitoring program not be analyzed pursuant to the requirements of Table D 3.6.20-being conducted as specified in Table D 3.6.20-1, prepare 1 and the detection capabilities required by Table D 4.6.20-and submit to the Commission, in the Annual Radiological 1.

Environmental Operating Report, a description of the reasons for not conducting the program as required and the plans for preventing a recurrence.

Deviations are permitted from the required sample schedule if samples are unobtainable due to hazardous conditions, seasonal unavailability, theft, uncooperative residents or to malfunction of automatic sampling equipment. In the event of the latter, every effort shall be made to complete corrective action prior to the end of the next sampling period.

CY-NM-170-301 Revision 37 I 3.1-32 January 2020

RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM D 3/4.6.20 CONTROLS SURVEILLANCE REQUIREMENT With the level of radioactivity (as the rE~sult of plant effluents), in an environmental sampling medium exceeding the reporting levels of Table D 6.9.3-1 when averaged over any calendar quarter, prepare and submit to the Commission within 30 days from the end of the affected calendar quarter a Special Report pursuant to Control D 6.9.3. The Special Report shall identify the cause(s) for exceeding the limit(s) and define the corrective action(s) to be taken to reduce radioactive effluents so that the potential annual dose to a member of the public is less than the calendar year limits of Controls DLCO 3.6.15.a.(2),

m!co 3.6.15 b.(2) and DLCO 3.6.15.b.(3). When more than one of the radionuclides in Table D 6.9.3 are detected in the sampling medium, this report shall be submitted if:

concentration(1) + concentration(2)+

limit level ( 1) limit level (2)

~1.0 When radionuclides other than those in Table D 6.9.3-1 are detected and are the result of plant effluents, this report shall be submitted if the potential annual dose to an individual is equal to or greater than the calendar year limits of Controls DLCO 3.6.15.a.(2), DLCO 3.6.15.b.(2) and DLCO 3.6.15.b.(3).

CY-NM-170-301 Revision 37 13.1-33 January 2020

RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM D 3/4.6.20 CONTROLS SURVEILLANCE REQUIREMENT This report is not required if the measured level of radioactMty was not the result of plant effluents; however, in such an event, the condition shall be reported and described in the Annual Radiological Environmental Operating Report.

Wrth milk or fruit and/or vegetables no longer available at one or more of the sample locations specified in Table D 3.6.20-1, identify locations for obtaining replacement samples and add them to the radiological environmental monitoring program within 30 days. The specific locations from which samples were unavailable may then be deleted from the monitoring program. Identify the cause of the unavailability of samples and identify the new location(s) for obtaining replacement samples in the next Radioactive Effluent Release Report and also include in the report a revised figure(s) and table for Part II reflecting the new location(s).

CY-NM-170-301 Revision 37 13.1-34 January 2020

RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM D 3/4.6.20 TABLE D 3.6.20-1 OPERATIONAL RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM Exposure Pathway Sampling and Collection Type of Analysis and Number of Samples<a> and Locations and/or Sample Frequency <*> Frequency Radioiodine & Samples from 5 locations: Continuous sampler operation Radioiodine Canisters analyze Particulates with sample collection weekly or once/week for 1-131.

1) 3 Samples from off-site locations in different as required by dust loading, sectors of the highest calculated site average D/Q whichever is more frequent (based on all site licensed reactors)

Particulate Samplers

2) 1 sample from the vicinity of an established year Gross beta radioactivity round community having the highest calculated following filter change, !bl site average D/Q (based on all site licensed composite (by location) for reactors) gamma isotopic analysis< 0 >

once per 3 months, (as a minimum)

3) 1 sample from a control location 10-17 miles distant and in a least prevalent wind direction<dl Direct Radiation<e> 32 stations with two or more dosimeters to be placed.as Once per 3 months Gamma dose once per 3 follows: an inner ring of stations in the general area of months the site boundary and an outer ring in the 4 to 5 mile range from the site with a station in each land based sector.* The balance of the stations should be placed in special interest areas such as population centers, nearby residences, schools and in 2 or 3 areas to serve as control stations.

At this distance, 8 wind rose\sectors are over Lake Ontario.

CY-NM-170-301 Revision 37 13.1-35 January 2020

RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM D 3/4.6.20 TABLE D 3.6.20-1 (Cont)

OPERATIONAL RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM ,

Exposure Pathway Sampllng and Collection Type of Analysis and Number of Samples(a) and Locations and/or Sample Frequency <11l Frequency WATERBORNE Surface(!) 1) 1 sample upstream Composite sample over 1 month Gamma isotopic analysis<c)

. period<gl once/month. Composite for once per 3 months tritium analysis.

2) 1 sample from the site's downstream cooling water intake Sediment from Shoreline 1 sample from a downstream area with existing or Twice per year Gamma isotopic analysis<c) potential recreational value INGESTION Milk 1) Samples from milk sampling locations in 3 Twice per month, April-December Gamma isotopic<c> and 1-131 locations within 3.5 miles distance having the (samples will be collected in analysis twice per month when highest calculated site average D/Q. If there are January-March if 1-131 is detected animals are on pasture (April-none, then 1 sample from milking animals in in November and December of December); once/month at each of 3 areas 3.5-5.0 miles distant having the ' the preceding year) other times (January-March) if highest calculated site average D/Q (based on required all site licensed reactors)

2) 1 sample from a milk sampling location at a control location (9-20 miles distant and in a least prevalent wind direction)<d)

CY-NM-170-301 Revision 37 13.1-36 January 2020

RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM D 3/4.6.20 TABLE D 3.6.20-1 (Cont)

OPERATIONAL RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM Exposure Pathway Sampling and Collection Type of Analysis and Number of SampleslaJ and Locations and/or Sample Frequency (al Frequency Fish 1) 1 sample each of two commercially or Twice per year Gamma isotopic analysis<c> on recreationally important species in the vicinity of a edible portions twice per year plant discharge a~ea.Chl

2) 1 sample each of the same species from an area at least 5 miles distant from the site. (dl Food Products 1) Samples of three dtfferent kinds of broad leaf Once per year during harvest Gamma isotopic<cJ analysis of vegetation (such as vegetables) grown nearest to season edible portions (isotopic to each of two different off-site locations of highest include 1-131 or a separate 1-calculated site average D/Q (based on all 131 analysis may be licensed site reactors). performed) once during the harvest season

2) One sample of each of the similar broad leaf vegetation grown at least 9.3-20 miles distant in a least prevalent wind direction.

CY-NM-170'..301 Revision 37 I 3.1-37 January 2020

RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM D 3/4.6.20 NOTES FOR TABLED 3.6.20-1 (a) It is recognized that, at times, it may not be possible or practical to obtain samples of the media of choice at the most desired location or time. In these instances, suitable alternative media and locations may be chosencfor the particular pathway in question and may be substituted. Actual locations (distance and directions) from the site shall be provided in the Annual Radiological Environmental Operating Report. Highest D/Q locations are based on historical meteorological data for all site licensed reactors.

(b) Particulate sample filters should be analyzed for gross beta 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or more after sampling to allow for radon and thoron daughter decay. If the gross beta activity in air is greater than 10 times a historical yearly mean of control samples, gamma isotopic analysis shall be performed on the individual samples.

(c) Gamma isotopic analysis means the identification and quantification of gamma emitting radionuclides that may be attributable to the effluents from the facility.

(d) The purpose of these samples is to obtain background information. If it is not practical to establish control locations in accordance with the distance and wind direction criteria, other sites, such as historical control locations which provide valid background data may be substiMed.

(e) One or more instruments, such as a pressurized ion chamber, for measuring and recording dose rate continuously may be used in place of, or in addition to, integrating dosimeters. For the purpose of this table, a thermoluminescent dosimeter may be considered to be one phosphor and two or more phosphors in a packet may be considered as two or more dosimeters. Film badges shall not be used for measuring direct radiation.

(f) The "upstream sample" should be taken at a distance beyond significant influence of the discharge. The "downstream sample" should be taken in an area beyond but near the mixing zone, if possible.

(g) Composite samples should be collected with equipment (or equwalent) which is capable of collecting an aliquot at time intervals which are very short (e.g. hourly) relative to the compositing period (e.g. monthly) in order to assure obtaining a representative sample.

(h) In the event commercial or recreational important species are not available as a result of three attempts, then other species may be utilized as available.

CY-NM-170-301 Revision 37 I 3.1-38 January 2020

RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM D 3/4.6.20 r TABLE D 4.6.20-1 DETECTION CAPABILITIES FOR ENVIRONMENTAL SAMPLE ANALYSISla.bl LOWER LIMIT OF DETECTION LLD(c)

Surveillance Requirement Water<c> Airborne Particulate or Fish Milk Food Products Sediment Analysis (pCi/1) Gases (pCilm3) (pCilkg, wet) (pCi/1) (pCllkg, wet) (pCilkg, dry) gross beta 4 0.01 H-3 2000*

Mn-54 15 130 Fe-59 30 260 Co-58, Co-60 15 130 Zn-65 30 260 Zr-95, Nb-95 15 1-131 1*

0.07 1 60 Cs-134 15 0.05 130 15 60 150 Cs-137 18 0.06 150 18 80 180 Ba/La-140 15 15

If no drinking water pathway exists, a value of 3000 pCl/liter may be used.

- If no drinking water pathway exists, a value of 15 pCl/liter may be used.

CY-NM-170-301 Revision 37 13.1-39 January 2020

RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM D 3/4.6.20 NOTES FOR TABLED 4.6.20-1 (a) This list does not mean that only these nuclides are to be considered. Other peaks that are identifiable, together with those of the above nuclides, shall also be analyzed and reported in the Annual Radiological Environmental Operating Report pursuant to Technical Specification 6.6.2, "Annual Radiological Environmental Operating Report", and Control D 6.9.1.d.

(b) Required detection capabilities for thermoluminescent dosimeters used for environmental measurements are given in ANSI N.545 (1975), Section 4.3. Allowable exceptions to ANSI N.545 (1975), Section 4.3 are contained in Part 11, Section 4.3.

(c) The LLD is defined, for purposes of these specifications, as the smallest concentration of radioactive material in a sample that will yield a net count, above system background, that will be detected with 95 percent probability with only 5 percent probability of falsely concluding that a blank observation represents a "real" signal.

For a particular measurement system, which may include radiochemical separation:

LLD= 4.66 Sb E*V*2.22*Y*exp (-Mt)

Where:

LLD is the "a priori" lower limit of detection as defined above, as picocuries per unit mass or volume, Sb 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, as counts per disintegration, V is the sample size in units of mass or volume, 2.22 is the number of disintegrations per minute per picocurie, Y is the fractional radiochemical yield, where applicable, 1c is the radioactive decay constant for the particular radionuclide, and Lit for environmental samples is the elapsed time between sample collection, or end of the sample collection period and time of counting.

Typical values of E, V, Y and Lit should be used in the calculation.

CY-NM-170-301 Revision 37 I 3.1-40 January 2020

RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM D 3/4.6.20 NOTES FOR TABLED 4.6.20-1 It should be recognized that the LLD is defined as a before the ~ct limit representing the capability of a measurement system an~ not as an after the fact limit for the particular measurement. Analyses shall be performed in such a manner that the stated LLDs will be achieved under routine conditions.

Occasionally, background fluctuations, unavoidable small sample sizes, the presence of interfering nuclides or other uncontrollable circumstances may render these LLDs unachievable. In such cases, the contributing factors shall be identified and described in the Annual Radiological Environmental Operatin~ Report pursuant to Technical Specification 6.6.2, "Annual Radiological Environmental Operating Report", and Control D 6.9.1.d.

CY-NM-170-301 Revision 37 I 3.1-41 January 2020

INTERLABORATORY COMPARISON PROGRAM D 3/4.6.21 CONTROLS SURVEILLANCE REQUIREMENT DLCO 3.6.21 INTERLABORATORY COMPARISON PROGRAM DSR 4.6.21 INTERLABORATORY COMPARISON PROGRAM Applicability: Applicability:

Applies to participation in an interlaboratory comparison Applies to testing the validity of measurements on program on environmental sample analysis. environmental sampl~.

Objective:

To ensure the accuracy of measurements of radioactive To verify the accuracy of measurements on radioactive material in environmental samples.

material in environmental samples.

Specification: Specification: j Analyses shall be perfonned on radioactive materials The lnterlaboratory Comparison Program shall be described supplied as part of an lnterlaboratory Comparison Program in Part II. A summary of the results obtained as part of the which has been approved by the Commission. Participation above required lnterlaboratory Comparison Program shall be in this program shall include media for which environmental included in the Annual Radiological Environmental Operating samples are routinely collected and for which Report. Participants in the EPA Cross Check Program may intercomparison samples are available. provide the EPA program code designation in lieu of providing results.

Wrth analyses not being perfonned as required above, report the corrective actions taken to prevent a recurrence to the Commission in the Annual Radiological Environmental Operating Report.

CY-NM-170-301 Revision 37 13.1-42 January 2020

LAND USE CENSUS D 3/4.6.22 CONTROLS SURVEILLANCE REQUIREMENT DLCO 3.6.22 LAND USE CENSUS DSR 4.6.22 LAND USE CENSUS Applicabiltty: Applicability:

Applies to the performance of a land use census in the Applies to assuring that current land use is known.

vicinity of the Nine Mile Point Nuclear Facility.

Objective:

To determine the utilization of land within a distance of three To verify the appropriateness of the environmental miles from the Facility.

surveillance program.

Specification: Specification:

A land use census shall be conducted and shall identify The land use census shall be conducted during the growing within a distance of three miles the location in each of the 16 season at least once per 12 months using that information meteorological sectors the nearest residence and within a that will provide the best results, such as conducting a door-distance of three miles the location in each of the 16 to-door survey, aerial survey or consulting local agriculture meteorological sectors of fill milk animals. In lieu of a authorities. The results of the land use census shall be garden census, specifications for vegetation sampling in included in the Annual Radiological Environmental Operating

Table D 3.6.20-1 shall be followed, including analysis of Report.

appropriate controls.

With a land use census identifying a milk animal location(s) that represents a calculated D/Q value greater than the D/Q value currently being used in Control DSR 4.6.15.b.(3), identify the new location(s) in the next Radioactive Effluent Release Report.

CY-NM-170-301 Revision 37 I 3.1-43 January 2020

LAND USE CENSUS D 3/4.6.22 CONTROLS SURVEILLANCE REQUIREMENT If the D/Q value at a new milk sampling location is -

significantly greater (50%) than the D/Q value at an existing milk sampling location, add the new location to the radiological environmental monitoring program within 30 days. The sampling location(s) excluding the control station location, having the lowest calculated D/Q may be deleted from this monitoring program after October 31 of the year in which this land use census was conducted. Pursuant to Control D 6.9.1.e identify the new location(s) in the next Radioactive Effluent Release Report and also include in the report a revised figure(s) and table for Part II reflecting the new location(s).

CY-NM-170-301 Revision 37 I 3.1-44 January 2020

PART I- RADIOLOGICAL EFFLUENT CONTROLS Bases CY-NM-170-301 Revision 37 I B 3.1-0 January 2020

RADIOACTIVE EFFLUENT INSTRUMENTATION, RADIOACTIVE EFFLUENTS-LIQUID CONCENTRATION B 3/4.6.14 BASES FOR DLCO 3.6.14 and DSR 4.6.14 RADIOACTIVE EFFLUENT INSTRUMENTATION The radioactive liquid and gaseous effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in liquid and gaseous effluents during actual or potential releases of liquid and gaseous effluents. The alann/trip setpoints for these instruments shall be calculated and adjusted in accordance with the methodology and parameters in Part II to ensure that the alann/trip will occur prior to exceeding the limits as described in Technical Specification 6.5.3, "Radioactive Effluent Controls Program".

Ttte alarm/trip setpoint for the Offgas process monitor is limited by Technical Specification 3.6.15. The Objective of that Specification is to assure radioactive material released is within the limits of 10CFR20 and 10CFRSO Appendix I. By doing so, total body exposure to an individual at the exclusion area boundary will not exceed a very small fraction of the limits of 10 CFR 100 in the event this effluent is discharged directly without treatment.

The Stack Effluent Monitors provide Effluent Monitoring (which requires a minimum of 1 Low Range and 1 High Range monitor) and Containment Purge and Vent Isolation (which requires 2 High Range monitors). When the Purge and Vent isolation capability is not required (Primary containment not required OR Purge and Vent valves shut and clearance applied), only 1 High Range Monitor and 1 Low Range monitor are required to satisfy the monitoring function.

The functionality and use of this instrumentation is consistent with the requirements of General Design Criteria 60, 63 and 64 of Appendix A to 10CFR Part 50. The purpose of tank level indicating devices is to assure the detection and control of leaks that if not controlled could potentially result in the transport of radioactive materials to unrestricted areas.

CY-NM-170-301 Revision 37 I B 3.1-1 January 2020

RADIOACTIVE EFFLUENT INSTRUMENTATION, RADIOACTIVE EFFLUENTS - LIQUID CONCENTRATION B 3/4.6.15 BASES FOR DLCO 3.6.16 AND DSR 4.6.16 RADIOACTIVE EFFLUENTS Liquid Concentration This control is provided to ensure that the concentration of radioactive materials released in liquid waste effluents to unrestricted areas will be less than ten times the concentration levels specified in 10CFR Part 20, Appendix B, Table 2, Column 2. This limitation provides additional assurance that the levels of radioactive materials in bodies of water in unrestricted areas will result in exposures within (1) the Section I I.A design objectives of Appendix I, 10CFR Part 50, to a member of the public and (2) the limits of 10 CFR 20.1301(e) to the population. The concentration limit for dissolved or entrained.noble gases is based upon the assumption thafXe-135 is the controlling radioisotope and its effluent concentration in air (submersion) was converted to an equivalent concentration in water using the methods described in International Commission on Radiological Protection (ICRP) Publication 2.

The required detection capabilities for radioactive materials in liquid waste samples are tabulated in terms of the lower limits of detection (LLDs).

Detailed discussion of the LLD and other detection limits can be found in HASL Procedures Manual, HASL-300 (revised annually), Currie, L. A.,

"Limits for Qualitative Detection and Quantitative Determination -Application to Radiochemistry," Anal. Chem. 40. 586-93 (1968), and Hartwell, J.

K., "Detection Limits for Radioanalytical Counting Techniques," Atlantic Richfield Hanford Company Report ARH-SA-215 (June 1975).

Because tritium was identified in the Reactor Building Perimeter Drains, (2012), a non-radioactive systern which discharges to the storm drains, the discharge from the Reactor Building Perimeter Drains will be sampled while the sump pumps are running and a composite sample analyzed for radioactMty. _The source of the radioactivity has been determined to be discharges from the Emergency Condenser Vents during surveillance testing and actual events. The radioactivity released, dose rates and doses have been, and will continue to be, determined for each release using existing Chemistry Surveillance Procedures and reported in the Radioactive Effluent Release Report (RERR). Because the radioactivity detected in the perimeter drain has already been accounted for in the liquid releases, discharges from the drains will be reported in the RERR as a separate item. Monitonng this pathway is to validate that the radioactivity released via this pathway is equal to or less than the quantity accounted for as released via the Emergency Condenser Vents.

CY-NM-170-301 Revision 37 1B3.1-2 January 2020

RADIOACTIVE EFFLUENTS - LIQUID DOSE B 3/4.6.15 BASES FOR DLCO 3.6.16 AND DSR 4.6.16 RADIOACTIVE EFFLUENTS Liquid Dose This control is provided to implement the requirements of Section II.A, Ill.A and IV.A of Appendix I, 10CFR Part 50. The controls expressed as quarter and annual limits are set at those values found in Section II.A. of Appendix I, in accordance with S~ction IV.A. The action statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive material in liquid effluents to unrestricted areas will be kept "as low as is reasonably achievable." There are no drinking water supplies that can be potentially affected by plant operations. The dose calculation methodology and parameters in Part II implement the requirements in Section Ill.A of Appendix I that conformance with the guides of Appendix I be shown by calculation procedures based on models and data, such that the actual exposure of a member of the public through appropriate pathways is unlikely to be substantially underestimated. The equations specified in Part II for calculating the doses due to the actual release rates of radioactive materials in liquid effluents are 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 10CFR 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.

CY-NM-170-301 Revision 37 IB3.l-3 January 2020

RADIOACTIVE EFFLUENTS - GASEOUS DOSE RATE B 3/4.6.15 BASES FOR DLCO 3.6.16 AND DSR 4.6.16 RADIOACTIVE EFFLUENTS Gaseous Dose Rate This control is provided to ensure that the dose at any time at or beyond the site boundary from gaseous effluents from all units on the site will be within the annual dose limits of 10CFR Part 20 to unrestricted aregs. The annual dose limits are the doses associated with the concentrations of 10CFR Part 20, Appendix B, Table 2, Column 1. These limits provide 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 within or outside the site boundary, to annual average concentrations exceeding the limits specified in Appendix B, Table 2 of 10CFR Part* 20 or as governed by 10 CFR 20.1302(c). For members of the public who may at times be within the site boundary, the occupancy of that member of the public will usually be sufficiently low to compensate for any increase in the atmospheric diffusion factor above that for the site boundary. The specified release rate limits restrict, at all times, the corresponding gamma and beta dose rates above background to a member of the public at or beyond the site boundary to less than or equal to 500 mrems/year to the total body or to less than or equal to 3000 mrems/year to the skin. These release rate limits also restrict, at all times, the corresponding thyroid dose rate above background to a child via the inhalation pathway to less than or equal to 1500 mrems/year.

The required detection capabilities for radioactive materials in gaseous waste samples are tabulated in terms of the lower limits of detection (LLDs).

Detailed discussion of the LLD and other detection limits can be found in HASL Procedures Manual, HASL-300 (revised annually), Currie, L.A.,

"Umits for Qualitative Detection and Quantitative Determination -Application to Radiochemistry," Anal. Chem. 40, 586-93 (1968), and Hartwell, J.

K., "Detection Limits for Radioanalytical Counting Techniques," Atlantic Richfield Hanford Company Report ARH-SA-215 (June 1975).

CY-NM-170-301 Revision 37 I B 3.1-4 January 2020

RADIOACTIVE EFFLUENTS - DOSE - NOBLE GASES B 3/4.6.15 BASES FOR DLCO 3.6.15 AND DSR 4.6.15 RADIOACTIVE EFFLUENTS Dose - Noble Gases This control is provided to implement the requirements of Sections 11.B, Ill.A and IV.A of Appendix I, 10CFR Part 50. The controls expressed as quarter and annual limits are set at those values found in Section 11.B of Appendix I in accordance with the guidance of Section IV.A The action statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV-A of Appendix I to assure that the releases of radioactive material in gaseous effluents to unrestricted areas will be kept "as low as is reasonably achievable." The Surveillance Requirement implements the requirements in Section Ill.A of Appendix I that conform with the guides of Appendix I be shown by calculational procedures based on models and data such that the actual exposure of a member of the public through appropriate pathways is unlikely to be substantially underestimated. The dose calculation methodology and parameters established in Part II for calculating the doses due to the actual release rates of radioactive noble gases in gaseous effluents are 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 10CFR Part 50, Appendix I, "Revision 1, October 1977 and Regulatory Guide 1.111, "Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water Cooled Reactors," Revision 1, July 1977.

The Offsite Dose Calculation Manual Part 11 equations provided to determine the air doses beyond the site boundary are based upon the historical average atmospheric conditions.

CY-NM-170-301 Revision 37 1B3.1-5 January 2020

RADIOACTIVE EFFLUENTS - DOSE - IODINE -131, IODINE -133, TRITIUM AND RADIONUCLIDES IN PARTICULATE FORM B 3/4.6.15 BASES FOR DLCO 3.6.15 AND DSR 4.6.15 RADIOACTIVE EFFLUENTS Dose- lodine-131, lodine-133, Tritium and Radionuclides in Particulate Form This control is provided to implement the requirements of Sections 11.C, Ill.A and IV.A of Appendix I, 10CFR Part 50. The controls expressed as quarter and annual limits are set at those values found in Section 11.C of Appendix I in accordance with the guidance of Section IV.A. The action statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive materials in gaseous effluents to unrestricted areas will be kept "as low as is reasonably achievable." The Part II calculational methods specified in the Surveillance Requirement implements the requirements in Section Ill.A of Appendix I that conformance with the guid~ of Appendix I be shown by calculational procedures based on models and data, such that the actual exposure of a member of the public through appropriate pathways is unlikely to be substantially underestimated. The Part II calculational methodology and parameters for calculating the doses due to the actual release rates of the subject materials are 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 10CFR 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 for determining the actual doses based upon the historical average atmospheric conditions. The release rate specifications for iodine-131, iodine-133, tritium and radionuclides in particulate form with half lives greater than 8 days are dependent upon the existing radionuclide pathways to man, in the areas beyond the site boundary. The pathways that were examined in the development of these calculations were: 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.

CY-NM-170-301 Revision 37 I B 3.1-6 January 2020

RADIOACTIVE EFFLUENTS-TOTAL DOSE-URANIUM FUEL CYCLE B 3/4.6.15 BASES FOR DLCO 3.6.15 AND DSR 4.6.15 RADIOACTIVE EFFLUENTS Total Dose - Uranium Fuel Cycle This control is provided to meet the dose limitations of 40CFR Part 190 that have been incorporated into 10CFR Part 20 by 46FR 18525. The control requires the preparation and submittal of a Special Report whenever the calculated doses from plant generated radioactive effluents and direct radiation exceed 25 mrems to the whole body or any organ, except the thyroid, which shall be limited to less than or equal to 75 mrems. 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 40CFR Part 190 if the individual reactors remain within twice the dose design objectives of Appendix I and if direct radiation doses from the reactor units, the on-site Independent Spent Fuel Storage Installation (ISFSI), and outside storage tanks are kept small. The Special Report will describe a course of action that should result in the limitation of the annual dose to a member of the public to within the 40CFR Part 190 limits. For the purposes of the Special Report, it may be assumed that the dose commitment to a member of the public from other uranium fuel cycle sources is negligible, with the exception that dose contribution from other nuclear fuel cycle facilities at the same site or within a radius of 8 km must be considered. If the dose to any member of the public is estimated to exceed the requirements of 40CFR Part 190, the Special Report with a request for variance (provided the release conditions resulting in violation of 40CFR Part 190 have not already been corrected), in accordance with the provisions of 40CFR Part 190.11 and 10 CFR Part 20.2203(a)(4) is considered to be a timely request and fulfills the requirements of 40CFR Part 190 until NRC staff action is completed. The variance only relates to the limits of 40CFR Part 190 and does not apply in any way to the other requirements for dose limitation of 10CFR Part 20, as addressed in Controls DLCO 3.6.15.a.(1) and DLCO 3.6.15.b.(1). An individual is not considered a member of the public durin-g any period in which he/she is engaged in carrying out any operation that is part of the nuclear fuel cycle.

CY-NM-170-301 Revision 37 I B 3.1-7 January 2020

RADIOACTIVE EFFLUENT TREATMENT SYSTEMS -

LIQUID AND GASEOUS B 3/4.6.16 BASES FOR DLCO 3.6.16 AND DSR 4.6.16 RADIOACTIVE EFFLUENT TREATMENT SYSTEMS Liquid Radwaste Treatment System The requirement that the appropriate portions of this system be used provides assurance that the releases of radioactive materials in liquid effluents will be kept "as low as is reasonably achievable." This control implements the requirements of 10 CFR Part 50.36a, General Design Criterion 60 of Appendix A to 10 CFR 50 and the design objective given in Section I1.D of Appendix I to 10 CFR Part 50. Projected doses are calculated on a batch rather than every 31 days due to the low frequency of releases.

Gaseous Effluent Treatment Systems The functionality of the Gaseous Radwaste Treatment System and the Ventilation Exhaust Treatment System ensures that the systems will be available for use whenever gaseous effluents require treatment prior to release to the environment. The requirement that 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 control implements the requirements of 10 CFR Part 50.36a, General Design Criterion 60 of Appendix A to 10 CFR 50 and the design objectives given in Section I1.D of Appendix I to 10 CFR Part 50. The control governing the use of appropriate portions of the Gaseous Radwaste Treatment System is based on time without treatment rather than dose, due to the wide variability in effluent with changing power conditions. Since the capability exists to operate within specification without use of the Gaseous Radwaste Treatment System, it is conceivable that due to unforeseen circumstances, limited operation without the system may be made sometime during the life of the plant. The control governing the use of appropriate portions of the Ventilation Exhaust Treatment System was specified as a suitable fraction of the dose design objectives set forth In I1.C:of Appendix I, 10CFR Part 50, for gaseous effluents.

CY-NM-170-301 Revision 37 I B 3.1-8 January 2020

MARK I CONTAINMENT, LIQUID HOLDUP TANKS B 3/4.6.18, B 3/4.6.19 BASES FOR DLCO 3.6.18 AND DSR 4.6.18 MARK I CONTAINMENT This control provides reasonable assurance that releases from drywell purging operations will not exceed the annual dose limits of 10CFR Part 20 for unrestricted areas.

BASES FOR DLCO 3.6.19 AND DSR 4.6.19 LIQUID HOLDUP TANKS This control applies to any outdoor tank that is not surrounded by liners, dikes or walls capable of holding the tank contents and that does not have tank overflows ~nd surrounding areas drains connected to the liquid radwaste treatment system.

Restricting the quantity of radioactive material contained in the specified tanks provides assurance that in the event of an uncontrolled release of the tanks' contents, the resulting concentrations would be less than ten times the concentrations of 10CFR Part 20, Appendix B, Table 2, Column 2, at the nearest potable water supply and the nearest surface water supply in an unrestricted area.

CY-NM-170-301 Revision 37 I B 3.1-9 January 2020

RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM B 3/4.6.20 BASES FOR DLCO 3.6.20 AND DSR 4.6.20 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM The radiological environmental monitoring program required by this control provides representative measurements of radiation and of radioactive materials in those exposure pathways and for those radionuclides that lead to the highest potential radiation exposures of members of the public resulting from the station operation. This monitoring program implementsSection IV.B.2 of Appendix I to 10CFR Part 50 and thereby supplements the radiological effluent monitoring program by verifying that the measurable concentrations of radioactive materials and levels of radiation are not higher than expected on the basis of the effluent measurements and the modeling of the environmental exposure pathways. Guidance for this monitoring program is provided by the Radiological Assessment Branch Technical Position on Environmental Monitoring. The initially specified monitoring program will be effective for at least the first three years of commercial operation. Following this period, program changes may be initiated based on operational experience.

The required detection capabilities for environmental sample analyses are tabulated in terms of the lower limits of detection (LLDs). The LLDs required by Table D 4.6.20-1 are considered optimum for routine environmental measurements in industrial laboratories. It should be recognized that the LLD is defined as a before the fact limit representing the capability of a measurement system and not as an after the fact limit for a particular measurement Detailed discussion of the LLD, and other detection limits, can be found in HASL Procedures Manual, HASL-300 (revised annually), Currie, L.A.,

"Limits for Qualitative Detection and Quantitative Determination - Application to Radiochemistry," Anal. Chem 40, 586-93 (1968) and Hartwell, J.K, "Detection Limits for Radioanalytical Counting Techniques," Atlantic Richfield Hanford Company Report ARH-SA-215 (June 1975).

CY-NM-170-301 Revision 37 I B 3.1-10 January 2020

INTERLABORATORY COMPARISON PROGRAM B 3/4.6.21 BASES FOR DLCO 3.6.21 AND DSR 4.6.21 INTERLABORATORY COMPARISON PROGRAM The requirement for parti_cipation in an approved lnterlaboratory Comparison Program is provided to ensure that independent checks on the precision and accuracy of the measurements of radioactive material in environmental sample matrices are performed as part of the quality assurance program for environmental monitoring for the purposes of Section IV. 8.2 of Appendix I to 10CFR Part 50.

CY-NM-170-301 Revision 37 IB3.1-11 January 2020

LAND USE CENSUS B 3/4.6.22 BASES FOR DLCO 3.6.22 AND DSR 4.6.22 LAND USE CENSUS This control is provided to ensure that changes in the use of areas at and beyond the site boundary are identified and that modifications to the radiological environmental monitoring program are made If required by the results of this census. The best survey infonnation such as from a door-to-door survey(s), from an aerial survey or from consulting with local agricultural authorities shall be used. This census satisfies the requirements of Section IV.B.3 of Appendix I to 10CFR Part 50.

In lieu of a garden census, the significance of the exposure via the garden pathway can be evaluated by the sampling of vegetation as specified in Table D 3.6.20-1.

A milk sampling location, as defined in Section 1, requires that at least 1O milking cows are present at a designated milk sample location. It has been found from past experience, and as a result of conferring with local farmers, that a minimum of 1O milking cows is necessary to guarantee an adequate supply of milk twice per month for analytical purposes. Locations with less than 1O milking cows are usually utilized for breeding purposes which eliminates a stable supply of milk for samples as a result of suckling calves and penods when the adult animals are dry.

CY-NM-170-301 Revision 37 I B 3.1-12 January 2020

PART I - RADIOLOGICAL EFFLUENT CONTROLS Section 6.0 Administrative Controls CY-NM-170-301 Revision 37 I 6.0-0 January 2020

Administrative Controls 6.0 6.0 ADMINISTRATIVE CONTROLS The ODCM Specifications are subject to Technical Specification Section 6.6.2, "Annual Radiological Environmental Operating Report," Section 6.6.3, Radioactive Effluent Release Report," Section 6.5.1, "Offsite Dose Calculation Manual (ODCM)," and Section 6.5.3,

Radioactive Effluent Controls Program."

CY-NM-170-301 Revision 37 I 6.0-1 January 2020

REPORTING REQUIREMENTS D 6.9.1.d D.6.9.1.e D6.9 Reporting Requirements D 6.9.1.d Annual*Radiological Environmental Operating Report The Annual Radiological Environmental Operating Report shall include a comparison with operational controls as appropriate, and with environmental surveillance reports from the previous 5 years, and an assessment of the observed impacts of the plant operation on the environment. The report shall also include the results of land use censuses required by Control DLCO 3.6.22.

The report shall also include the following: a summary description of the radiological environmental monitoring program; at least two legible maps.. covering all sampling locations keyed to a table giving distances and directions from the centerline of one reactor; the results of licensee participation in the lnterlaboratory Comparison Program, required by Control DLCO 3.6.21; discussion of all deviations from the sampling schedule of Table D 3.6.20-1; and discussion of all analyses in which the LLD required in Table D 4.6.20-1 was not achievable.

.. One map shall cover stations near the site boundary; a second shall include the more distant stations.

D 6.9.1.e Radioactive Effluent Release Report The Radioactive Effluent Release Report shall include a summary of the quantities of radioactive liquid-and gaseous effluents and solid waste releases from the unit as outlined in Regulatory Guide 1.21, "Measuring, Evaluating, and Reporting Radioactivity in Solid Wastes and Releases of Radioactive Materials in Liquid and Gaseous Effluents from Light-Water-Cooled Nuclear Power Plants", Revision 1, June 1974, with data summarized on a quarterly basis following the format of Appendix B thereof.

The Radioactive Effluent Release Report shall include an annual summary of hourly meteorological data collected over the previous year. This annual summary may be either in the form of an hour-by-hour listing on magnetic tape of wind speed, wind direction, atmospheric stability, and precipitation (if measured), or in the form of Joint frequency distributions of wind speed, wind direction, and atmospheric stability.* This same report shall include an assessment of the radiation doses from radioactive liquid and gaseous effluents to members of the public due to their activities inside the site boundary (Part II Figure 5.1.3-1) during the reporting period. All assumptions used in making these assessments, i.e., specific activity, exposure time and location, shall be included in these reports.

The assessment of radiation doses shall be performed in accordance with the methodology and parameters in Part II.

The Radioactive Effluent Release Report shall also include an assessment of radiation doses to the likely most exposed member of the public from reactor releases and other nearby uranium fuel cycle sources, including doses from primary effluent pathways and direct radiation, for the previous calendar year to show conformance with 40 CFR Part 190, Environmental Radiation Protection Standards for Nuclear Power Operation. Acceptable methods for calculating the doses from liquid and gaseous effluents are given in Part II.

In lieu of submission with the Radioactive Effluent Release Report, the licensee has the option of retaining this summary of required meteorological data on site in a file that shall be provided to the NRC upon request.

CY-NM-170-301 Revision 37 I 6.0-2 January 2020

REPORTING REQUIREMENTS D 6.9.1.e The Radioactive Effluent Release Reports shall include a list of unplanned releases from the site to unrestricted areas of radioactive materials in gaseous and liquid effluents made during the reporting period.

The Radioactive Effluent Release Reports shall include the following information for each class of solid waste (as defined by 10 CFR Part 61) shipped offsite during the report period.

a. Container volume,

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

c. Principal radionuclides (specify whether determined by measurement or estimate),

d. Source of waste and processing employed (e.g., dewatered spent resin, compacted dry waste, evaporator bottoms),

e. Type of container (e.g., LSA, Type A, Type B, Large Quantity), and,

f. Solidification agent or absorbent (e.g., cement)

The Radioactive Effluent Release Reports shall include any changes made during the reporting period to the Process Control Program (PCP) and to the Offsite Dose Calculation Manual (ODCM), as well as a listing of new locations for dose calculations and/or environmental monitoring identified by the land use census pursuant to Control DLCO 3.6.22.

Changes to the Process Control Program (PCP) shall be reported to the Commission in the Radioactive Effluent Release Report for the period in which the change(s) was made. This submittal shall contain:

a. Sufficiently detailed information to totally support the rationale for the change without benefit of additional or supplemental information;

b. A determination that the change did not reduce the overall conformance of the solidified wast!? product to existing criteria for solid wastes; and

c. Documentation of the fact that the change has been reviewed and found acceptable.

Changes to the Offsite Dose Calculation Manual (ODCM) shall be in accordance with Technical Specification 6.5.1, "Offsite Dose Calculation Manual (ODCM)".

CY-NM-170-301 Revision 37 I 6.0-3 January 2020

SPECIAL REPORTS D 6.9.3 D 6.9.3Special Reports Special reports shall be submitted in accordance with 10 CFR 50.4 to the Regional Office within the time period specified for each report.

These reports shall be submitted covering the activities identified below pursuant to the requirements of the applicable reference

a. .,._specification:

b.

C.

d. Not applicable to RETS e.

f.

g. +-------

h. Calculate Dose from Liquid Effluent in Excess of Limits, Control DLCO 3.6.15.a(2) (30 days from the end of the affected calendar quarter).

i. Calculate Air Dose from Noble Gases Effluent in Excess of Limits, Control DLCO 3.6.15.b(2) (30 days from the end of the affected calendar quarter).

j. Calculate Dose from 1-131, H-3 and Radioactive Particulates with half lives greater than eight days in Excess of Limits, Control DLCO 3.6.15.b(3)(b) (30 days from the end of the affected calendar quarter).

k. Caculated Doses from Uranium Fuel Cycle Source in Excess of Limits, Control DLCO 3.6.15.d (30 days from the end of the affected calendar year).

I. Nonfunctional Gaseous Radwaste Treatment System, Control DLCO 3.6.16.b (30 days from the end of the affected calendar year).

m. Environmental Radiological Reports. With the level of radioactivity (as the result of plant effluents) in an environmental sampling media exceeding the reporting level of Table D 6.9.3-1, when averaged over any calendar quarter, in lieu of a Licensee Event Report, prepare and submit to the Commission within thirty (30) days from the end of the calendar quarter a special report identifying the cause(s) for exceeding the limits, and define the corrective action to be taken.

CY-NM-170-301 Revision 37 I 6.0-4 January 2020

SPECIAL REPORTS D 6.9.3 Table D 6.9.3-1 REPORTING LEVEL FOR RADIOACTIVITY CONCENTRATIONS IN ENVIRONMENTAL SAMPLES L

REPORTING LEVELS Water Airborne Particulate or Fish Milk Food Products Analysis (pCi/1) Gases (pCl/m 3

) (pCi/kg, wet) (pCI/I) (pCi/kg, wet)

H-3 20,000*

Mn-54 1,000 30,000 Fe-59 400 10,000 Co-58 1,000 30,000 Co-60 300 10,000 Zn-65 300 20,000 Zr-95, Nb-95 400 1-131 2** 0.9 3 100 Cs-134 30 10.0 1,000 60 1,000

~

Cs-137 50 20.0 2,000 70 2,000 Ba/La-140 200 300

For drinking water samples. This is a 40 CFR 141 value. If no drinking water pathway exists, a value of 30,000 pCi/liter may be used.

If no drinking water pathway exists, a value of 20 pCi/liter may be used.

CY-NM-170-301 Revision 37 I 6.0-5 January 2020

(

PART II- CALCULATION METHODOLOGIES CY-NM-170-301 Revision 37 II 1 January 2020

1.0 LIQlJ!]) EFFLUENTS 1.1 Setpoint Determinations 1.1.1 Basis Monitor setpoints will be established such that the concentration ofradionuclides in the liquid effluent releases in the discharge canal shall be limited to ten times the concentrations specified in 10 CFR 20, Appendix B, Table 2, Column 2 for radionuclides other than dissolved or entrained noble gases. For dissolved or entrained noble gases, the concentration shall be limited to 2E-04 µCi/ml total activity. Setpoints for the Service Water System Eflluent Line will be calculated quarterly based on the radionuclides identified during the previous year's releases from the liquid radwaste system or the isotopes identified in the most recent radwaste release or other identified probable source. Setpoints for the Liquid Radwaste Effluent Line will be based on the radionuclides identified in each batch ofliquid waste prior to its release.

After release, the Liquid Radwaste monitor setpoint may remain as set, or revert back to a setpoint based on a previous Radioactive Eflluent Release Report, or install blank flange in the discharge line and declare nonfunctional in accordance with the ODCM Part I.

Since the Service Water System eflluent monitor and Liquid Radwaste eflluent monitor can only detect gamma radiation, the alarm setpoints are calculated by using the concentration of gamma emitting isotopes only ( or the corresponding Maximum Eflluent Concentration (MEC) values for the same isotopes, whichever are higher) in the L ,(µCi/ml)ry expression (Section 1.1.2, 1.1.3).

The Required Dilution Factor (RDF) is calculated using concentrations of all isotopes present (or the corresponding MEC values for the same isotopes, whichever are higher) including tritium and other non-gamma emitters to ensure that all radionuclides in the discharge canal do not exceed Technical Specifications Radioactive Effluent Controls Program limits.

1.1.2 Senice Water System Effluent Line Alarm Setpoint The detailed methods for establishing setpoints for the Service Water System Eflluent Line Monitor shall be contained in the Nine Mile Point Station Procedures. These methods shall be in accordance with the following:

The General Setpoint Equation is Setpoint < (Conservative Factor) (ConcentrationXADFXCF)

RDF From the above General Setpoint Equation the Hi and Alert alarms are calculated as follows:

Setpoint (Hi alarm)< _ T-1 (µCi I m/)1y (CF) TEDF I F sw 0 9 + background L, [(µCi I m/J,r I MEC1 ]

. (Al al ) "I:-; (µCi I m/)1y (CF) TEDF I F 9W b k Setpomt ert arm < 0 . 7 - - - - - ~ - - - - - - - + ac groun d

"'£.1 [(µCi I m/),7 I MEC1 ]

(µCi/ml)rr= concentration of gamma emitting isotope i in the sample, or the corresponding MEC of gamma emitting isotope i (MEC)i, whichev"er is higher (units= µCi/ml).

CY-NM-170-301 Revision 37 II 2 January 2020

1.1.2 Service Water System Effluent line Alarm Setpoint (Cont'd)

(µCi/ml),T = concentration of any radioactive isotope i in the sample including tritium and other non-gamma emitters or corresponding MEC of isotope i, MEC,, whichever is higher (units= µCi/ml).

TF = Tempering Fraction TDF Total Dilution Flow (units= gallons/minute).

TEDF = Total Effective Dilution Flow= TDF (1-TF) (units= gallons/minute)

Fsw Service Water Flow (units= gallons/minute).

CF Monitor calibration factor (units= net cpm/µCi/ml).

MEC, = Maximum Effluent Concentration, ten times the Effluent Concentration for radionuclide i as specified in 10 CPR 20, Appendix B, Table 2, Column 2 (units

µCi/ml).

Sample = Those nuclides present in the previous batch release from the liquid radwaste effluent system or those nuclides present in the last Radioactive Effluent Release Report (units=

µCi/ml) or those nuclides present in the service water system.**

(MEC),y same as MEC, but for gamma emitting nuclides only.

0.9 and 0.7= factors of conservatism to account for inaccuracies.

RDF Required Dilution Factor, L ,[(µCi/ml) IT /MEC,]. If MEC values are used in the

(µCi/ml),y, they must also be used in calculating RDF (numerator). RDF= FMEC (See Section 11-1.2).

ADP Actual Dilution Factor, TEDF/F"'

- For periods with known reactor water to Reactor Building Closed Loop Cooling (RBCLC) system leakage, RBCLC concentration may be prudently substituted for the above.

1.1.3 Liquid Radwaste Effluent Line Alarm Setpoint The detailed methods for establishing setpoints for the Liquid Radwaste Effluent Line Monitor shall be contained in the Nine Mile Point Station Procedures. These methods shall be in accordance with the following:

The General Setpoint Equation in Section 11-1.1.2 is used to develop the Hi-Hi and Hi alarm setpoints below:

1:, (µCi I ml)1 (CF) TEDF I Fre Setpoint (Hi-Hi alarm) < 0.9 . Y + background Li [(µCz I ml)tT I AfECi]

CY-NM-170-301 Revision 37 II 3 January 2020

1.1.3 Liquid Radwaste Effluent Line Alarm Setpoint (Cont'd)

L (µCi I ml) (CF) TEDF I Fre Setpoint (Hi alarm) < 0. 7 rr + background L1 [(µCi I ml)zT I MEC1 ]

(µCi/ml)r, = concentration of gamma emitting isotope i in the sample or the corresponding MEC of gamma emitting isotope i, (MEC)i whichever is higher.

(µCi/ml)iT = concentration of any radioactive isotope i in the sample including tritium and other non-gamma emitters or the corresponding MEC of isotope i, MEC1, whichever is higher. (units= µCi/ml).

TF = Tempering Fraction TDF = Total Dilution Flow (units= gallons/minute).

J TEDF Total Effective Dilution Flow= TDF (1-TF) (units= gallons/minute)

Fm = Radwaste Effluent Flow (units= gallons/minute).

CF = Monitor calibration factor (units= net cps/µCi/ml).

MEC, Maximum Effluent Concentration, ten times the Effluent Concentration for radionuclide i as specified in IO CFR 20, Appendix B, Table 2, Column 2, for those nuclides detected by spectral analysis of the contents of the radwaste tanks to be released. (units = µ0/ml)

(MEC)r, = same as MEC, but for gamma emitting nuclide only.

0.9 and 0.7 factors of conservatism to account for inaccuracies.

RDF = Required Dilution Factor, L 1[(µCi/ml) 1TIMEC,].

IfMEC values are used in the (µCi/ml)rr, they must also be used in calculating RDF (numerator).

ADF = Actual Dilution Factor= TEDF/Frc Notes: (a) IfTEDF/Frc = L 1[(µCi/ml) tTIMEC1](if ADF = RDF) the discharge could not be made, since the monitor would be continuously in alarm. To avoid this situation, Frc will be reduced (normally by a factor of 2) to allow setting the alarm point at a concentration higher than tank concentration.

This will also result in a discharge canal concentration at approximately 50%

Maximum Effluent Concentration.

(b) TF is tempering fraction (i.e., diversion of some fraction of discharge flow to the intake canal for the purpose of temperature control).

CY-NM:.170-301 Revision 37 II 4 January 2020

1.1.4 Discussion 1.1.4.1 Control of Liquid Effluent Batch Discharges At Nine Mile Point Unit 1 Liquid Radwaste Effluents are released only on a batch mode. To prevent the inadvertent release of any liquid radwaste effluents, radwaste discharge is mechanically isolated (blank flange installed or discharge valve chain-locked closed) following the completion of a batch release or series of batch releases.

This mechanical isolation remains in place and will only be removed prior to the next series of liquid radwaste discharges after all analyses required in station procedures and Table D 4.6.15-lA of Part I are performed and monitor setpoints have been properly adjusted.

, 1.1.4.2 Simultaneous Discharges of Radioactive Liquids If during the discharge of any liquid radwaste batch, there is an indication that the service water canal has become contaminated (through a service water monitor alarm or through a grab sample analysis in the event that the service water monitor is nonfunctional) the discharge shall be terminated immediately. The liquid radwaste discharge shall not be continued until the cause of the service water alarm (or high grab sample analysis result) has been determined and the appropriate corrective measures taken to ensure ten times the effluent concentrations specified in I 0CFR20, Appendix B, Table 2, Column 2 (Section D 3.6.15.a(l) of Part I) are not exceeded. In accordance with Liquid Waste procedures, controls are in place to preclude a simultaneous release of liquid radwaste batch tanks. In addition, an independent verification of the discharge valve line-up is performed prior to discharge to ensure that simultaneous discharges are prevented.

1.1.4.3 Sampling Representativeness This section covers Part I Table D 4.6.15-1 Note b concerning thoroughly mixing of each batch of liquid radwaste prior to sampling.

Liquid Radwaste Tanks scheduled for discharge at Nine Mile Point Unit I are isolated (i.e.

inlet valves marked up) and at least three tank volumes of entrained fluids are recirculated prior to sampling. Minimum recirculation time is calculated as follows:

Minimum Recirculation Time= 3.0(T/R)

Where:

3.0 = Plant established mixing factor, unitless T = Tank volume, gal R = Recirculation flow rate, gpm Additionally, the Hi Alarm setpoint of the Liquid Radwaste Effluent Radiation Monitor is

set at a value corresponding to not more than 70% of its calculated response to the grab sample or corresponding MEC values. Thus, this radiation monitor will alarm if the grab sample, or corresponding MEC value, is significantly lower in activity than any part of the tank contents being discharged.

CY-NM-170-301 Revision 37 IIS January 2020

1.1.4.4 Liquid Radwaste System Operation Part I Section DLCO 3.6.16.a requires that the liquid radwaste system shall be used to reduce the radioactive materials in liquid wastes prior to their discharge, as necessary, to meet the concentration and dose requirements of Section DLCO 3.6.15.

Utilization of the radwaste system will be based on the capability of the indicated components of each process system to process contents of the respective low conductivity and high conductivity collection tanks:

1) Low Conductivity (Equipment Drains): Radwaste Filter and Radwaste Demin.

(See Fig. D-1) or modular waste water technology ("THERMEX")

2) High Conductivity (Floor Drains): Waste Evaporator (See Fig. D-1) or modular waste water technology ("THERMEX") directly to the Waste Collector Tank or the Waste Sample Tanks.

Cumulative dose contributions from liquid effluents for the current calendar quarter and the current calendar year shall be determined as described in Section 11-1.3 of this manual prior to the release of each batch of liquid waste. This same dose projection of Section Il-1.3 will also be performed in the event that untreated liquid waste is discharged, to ensure that the dose limits of Part I DLCO 3.6.15.a(2) are not exceeded. (Thereby implementing the requirements of 10CFR50.36a, General Design Criteria 60 of Appendix A and the Design Objective given in Section II-D of Appendix I to 10 CFR50).

For the purpose of dose projection, the following assumptions shall be made with regard to concentrations of non-gamma emitting radionuclides subsequently analyzed:

a) [H-3] ::;;; H-3 Concentration found recent condensate storage tank analysis b) [Sr-89] ::;;; 4 x Cs-137 Concentration c) [Sr-90] :<,; 0.5 x Cs-137 Concentration d) [Fe-55] ::;;; 1 x Co-60 Concentration Assumed Scaling Factors used in b, c, and d above represent conservative estimates derived from analysis of historical data from process waste streams. Following receipt of H-3, Sr-89, Sr-90 and Fe-55 analysis information, dose estimates shall be revised using actual radionuclide concentrations and actual tank volumes discharged.

CY-NM-170-301 Revision 37 II 6 January 2020

1.1.4.5 Service Water System Contamination Service water is normally non-radioactive. If contamination is suspected, as indicated by a significant increase in service water effluent monitor response, grab samples will be obtained from the service water discharge lines and a gamma isotopic analysis meeting the LLD requirements of Part I Table D 4.6.15- I completed. If it is determined that an inadvertent radioactive discharge is occurring from the service water system, then:

a) A 10CFR 50.59 review shall be performed (ref. I&E Bulletin 80-10),

b) daily service water effluent samples shall be taken and analyzed for principal gamma emitters until the release is terminated, c) an incident composite shall be prepared for H-3, gross alpha, Sr-89, Sr-90 and Fe-55 analyses and, d) dose projections shall be performed in accordance with Section Il-1.3 of this manual (using estimated concentrations for H-3, Sr-89, Sr-90 and Fe-55 to be conservatively determined by supervision at the time of the incident).

Additionally, service water effluent monitor setpoints may be recalculated using the actual distribution of isotopes found from sample analysis.

When contamination is indicated by quantitative non-gamma emitter results, sample and analyze gamma and non-gamma emitters weekly.

1.1.4.6 Reactor Building Perimeter Drain Discharges The Reactor Building Perimeter Drain became contaminated to detectable levels with tritium, due to discharges from the Emergency Condenser (EC) Vents. This pathway has been evaluated by extensive sampling and analysis. It has been determined that any discharge to the environment via this pathway will be well within regulatory limits.

Discharges via this pathway wiU be monitored by composite sampling and laboratory analysis. The discharges will be tracked and reported annually in the ~dioactive Effluent Release Report (RERR). Because the radioactive discharge from the EC Vents are evaluated each time the ECs are actuated and the results reported in the RERR, the activity detected in the perimeter drain discharge will be reported separately from other liquid discharges.

1.2 Liquid Effluent Concentration Calculation This calculation documents compliance with Part I Section DLCO 3.6.15.a (1 ).

The concentration of radioactive material released in liquid effluents to unrestricted areas

( see Figure 5.1.3-1) shall be limited to ten times the effluent concentrations specified in 10CFR20, Appendix B, Table 2, Column 2, for radionuclides other than dissolved or entrained noble gases. For dissolved or entrained noble gases, the concentration shall be limited to 2 E-4 microcurie/milliliter (µCi/ml) total activity at the point of discharge. For dissolved and entrained noble gases, this limit may also be satisfied by using 2E-4 µCi/ml

' j as the MEC for each noble gas.

The concentration of radioactivity from Liquid Radwaste batch releases and, if applicable, Service Water System and emergency condenser start-up vent discharges are included in the CY-NM-170-301 Revision 37 II 7 January 2020

1.2 Liquid Effluent Concentration Calculation (Cont'd) calculation. The calculation is performed for a specific period of time. No credit taken for averaging. The limiting concentration is calculated as follows:

FMEC = L [(L .c~.) I (MEC, L .Fs)]

I Where:

FMEC The fraction of Maximum Effluent Concentration, the ratio at the point of discharge of the actual concentration to ten times the Effluent Concentration of 10 CFR 20, Appendix B, Table 2, Column 2 for radionuclides other than dissolved or entrained noble gases.

For noble gases, the concentration shall be limited to 2 E-4 microcurieJml total activity.

C,.=(µCi/ml),. The concentration of nuclide i in particular effluent stream s, µCi/ml.

= The flow rate of a particular effluent stream s, gpm.

MEC, = Maximum Effluent Concentration, ten times the Effluent Concentration of a specific nuclide i from 10CFR20, Appendix B, Table 2, Column 2 (noble gas limit is 2E-4 µCi/ml).

The total activity rate of nuclide i, in all effluent streams s.

L (Fs) 5

= The total flow rate of all effluent streams s, gpm (including those streams which do not contain radioactivity).

A value of less than one for FMEC is considered acceptable for compliance with Part I Section DLCO 3.6.15.a(l).

1.3 Dose Determinations 1.3.1 Maximum Dose Equivalent Pathway A dose assessment report was prepared for the Nine Mile Point Unit 1 facility by Charles T.

Main, Inc., of Boston, MA. This report presented the calculated dose equivalent rates to individuals as well as the population within a SO-mile radius of the facility based on the radionuclides released in liquid and gaseous effluents during the time periods of 1 July 1980 through 31 December 1980 and from January 1981 through 31 December 1981. The radwaste liquid releases are based on a canal discharge rate of 590 ft:3/sec which affects near field and far field dilution; therefore, this report is specific to this situation. Utilizing the effluent data contained in the Annual Radioactive Effluent Release Reports as source terms, dose equivalent rates were determined using the environmental pathway models specified in Regulatory Guides 1.109 and 1.111 as incorporated in the NRC computer codes LADTAP for liquid pathways, and XOQDOQ and GASPAR for gaseous effluent pathways. Dose equivalent rates were calculated for the total body as well as seven organs and/or tissues for the adult, teen, child, and infant age groups. From the standpoint of liquid effluents, the pathways evaluated included fish and drinking water ingestion, and external exposure to water and sediment CY-NM-170-301 Revision 37 January 2020

1.3.1 Maximum Dose Equivalent Pathway (Cont'd)

The majority of the dose for a radwaste liquid batch release was received via the fish pathway. However, to comply with Part I Specifications for dose projections, the drinking water and sediment pathways are included. Therefore, all doses due to liquid effluents are calculated monthly for the fish and drinking water ingestion pathways and the sediment external pathway from all detected nuclides in liquid effluents released to the unrestricted areas to_ each organ. The dose projection for liquid batch releases will also include discharges from the emergency condenser vent as applicable, for all pathways. Each age group dose factor, Aw, is given in Tables 2-1 to 2-8. To expedite time, the dose is calculated for a maxi~um individual instead of each age group. This maximum individual will be a composite of the highest dose factor of each age group for each organ, hence At.

The following expression from NUREG 0133, Section 4.3 is used to calculate dose:

Where:

D1 = The cumulative dose commitment to the total body or any organ, from the liquid effluents for the total time period (,HL), mrem.

LiTL The length of the L th time period over which c,i, and FL are averaged for all liquid releases, hours.

= The average concentration of radionuclide, i, in undiluted liquid effluents during time period t-,. TL from any liquid release, µCi/ml.

A11 = The site related ingestion dose commitment factor to the total body or any organ t for each identified principal gamma or beta emitter for a maximum individual, mrem/hr per µCi/ml.

FL = The near field average dilution factor for Cii, during any liquid effluent release.

Defined as the ratio of the maximum undiluted liquid waste flow during release to the average flow from the site discharge structure to unrestricted receiving waters, unitless.

Au.1 values for radwaste liquid batch releases at a discharge rate of 295 ft3/sec (one circulating water pump in operation) are presented in tables 2-1 to 2-4. Aw values for an emergency condenser vent release are presented in tables 2-5 to 2-8. The emergency condenser vent releases are assumed to travel to the perimeter drain system and released from the discharge structure at a rate of .33 ft3/sec. See Appendix A for the dose factor Atat derivation. To expedite time the dose is calculated to a maximum individual. This maximum individual is a composite of the highest dose factor Aw of each age group a for each organ t and each nuclide i. If a nuclide is detected for which a factor is not listed, then it will be calculated and included in a revision to the ODCM.

All doses calculated in this manner for each batch of liquid effluent will be summed for comparison with quarterly and annual limits, added to the doses accumulated from other releases in the quarter and year of interest. In all cases, the following relationships will hold:

CY-NM-170-301 Revision 37 II 9 January 2020

1.3.1 Maximum Dose Equivalent Pathway (Cont'd)

For a calendar quarter:

D1 :S: 1.5 mrem total body D1 :S: 5 mrem for any organ For the calendar year:

D1 :S: 3.0 mrem total body D1 :S: 10 mrem for any organ Where:

D1 = total dose received to the total body or any organ due to liquid effluent releases.

If these limits are exceeded, a special report will be submitted to the NRC identifying the cause and proposed corrective actions. In addition, if these limits are exceeded by a factor of two, calculations shall be made to determine if the dose limits contained in 40 CFR 190 have been exceeded. Dose limits, as contained in 40 CFR 190 are total body and organ doses of25 mrem per year and a thyroid dose of75 mrem per year.

These calculations will include doses as a result of liquid and gaseous pathways as well as doses from direct radiation. The liquid pathway analysis will only include the fish and sediment pathways since the drinking water pathway is insignificant. This pathway is only included in the station's effluent dose projections to comply with Part I Specifications.

Liquid, gaseous and direct radiation pathway doses will consider the James A. FitzPatrick and Nine Mile Point Unit 2 facilities as well as Nine Mile Point Unit 1 Nuclear Station and the on-site ISFSI facilities.

In the event the calculations demonstrate that the 40 CFR 190 dose limits, as defined above, have been exceeded, then a report shall be prepared and submitted to the Commission

'within 30 days as specified in Part I Section DLCO 3.6.15.d.

Section 3.0 of the ODCM contains more information concerning calculations for an evaluation of whether 40 CFR 190 limits have been exceeded.

CY-NM-170-301 Revision 37 II 10 January 2020

1.3.2 Dose Projections - Determinations of the Need to Operate the Liquid Radwaste Treatment System 1.3.2.1 Requirements DLCO 3.6.16.a requires that the liquid radwaste system be used to reduce the radioactive materials in liquid wastes prior to their discharge when the projected doses due to the liquid eflluent would exceed 0.06 rnrem to the total body or 0.2 mrem to any organ for the, batch.

This Control implements Technical Specification 6.5.3.fthat requires the Radioactive Effluent Controls Program to include limitations on the functional capability and use of the liquid effluent treatment system to ensure the appropriate portion of this system is used to reduce releases of radioactivity. This is required when the projected doses would exceed 0.06 mrem to the total body and 0.2 mrem to any organ. Since releases are performed much less frequently than once per month, doses are to be projected prior to each release and the above limits will be applied on a batch basis.

1.3.2.2 Methodology The dose projection for each batch is calculated in the same manner as cumulative dose calculations for the current calendar quarter and current calendar year. See II-1.1 .4.4 and II-1.3 .1. If the calculated dose is greater than 0.06 mrem to the total body or 0.2 mrem to any organ, the appropriate subsystems of the liquid rad waste system shall be used to reduce the radioactivity levels of the batch prior to release.

1.3.2.3 Continuous Liquid Release Dose Projections Each month that a continuous liquid release is in progress, or is anticipated, the expected dose to man can be accounted for or projected. Since a continuous release does not result from not operating a portion of the Liquid Radwaste System, projections are not required to determine or evaluate Radwaste System Functionality. Dose projections may be relevant to planning repairs, and in reporting intended actions. See 11-1.1.4.5.

CY-NM-170-301 Revision 37 II 11 January 2020

2.0 GASEOUS EFFLUENTS 2.1 Setpoint Determinations 2.1.1 Basis Stack gas monitor setpoints will be established such that the instantaneous release rate of radioactive materials in gaseous effluents does not exceed the 10 CPR 20 limits for annual release rate. The setpoi°its will be activated if the instantaneous dose rate at or beyond the (land) site boundary would exceed 500 mrem/yr to the whole body or 3000 mrem/yr to the skin from the continuous release of radioactive noble gas in the gaseous effluent.

The offgas (condenser air ejector activity) monitor setpoints provide assurance that the total body exposure to an individual at the exclusion area boundary does not exceed a small fraction of the dose guidelines of 10 CPR 100.

Emergency condenser vent monitor setpoints will be established such that the release rate for radioactive materials in gaseous effluents do not exceed the Technical Specification dose rate limits. Monitor setpoints for emergency condenser vent monitors are conservatively

fixed at 5 rnr/hr for reasons described in Section II-2.1.4 and therefore do not require periodic recalculations.

Monitor setpoints from continuous release points will be determined once per quarter under normal release rate conditions and will be based on the isotopic composition of the actual release in progress, or an offgas isotopic distribution or a more conservative default composition specified in the pertinent procedure. If the calculated setpoint is higher than the existing setpoint, it is not mandatory that the setpoint be changed.

Under abnormal site release rate conditions, monitor alarm setpoints from continuous release points will be recalculated and, if necessary, reset at more frequent intervals as deemed necessary by Chemistry Supervision. In particular, contributions from both JAF and NMP-2 and the Emergency Condenser Vents shall be assessed.

During outages and until steady state power operation is again realized, the last operating stack and off gas monitor alarm setpoints shall be used.

Since monitors respond to noble gases only, monitor alarm points are set to alarm prior to exceeding the corresponding whole body dose rates.

The skin dose rate limit is not used in setpoint calculations because it is never limiting.

2.1.2 Stack Monitor Setpoints The detailed methods for establishing setpoints shall be contained in the station procedures.

These methods shall apply the following general criteria:

(1) Rationale for Stack monitor settings is based on the general equation:.

release rate, actual = --=re=l=ease=-=--=-ra=t=e"--',m=ax.:=.cal=lo~w~a=b=l~e_

corresp. dose rate, actual corresp. dose rate, max. allowable L,Q, (Q)max L,Q, (V, + (SF)K, (XI Q),) 500mrem/ yr CY-NM-170-301 Revision 37 II 12 January 2020

2.1.2 Stack Monitor Setpoints (Cont'd)

Where:

Q, = release rate for each isotope i, µCi/sec.

V, gamma whole body dose factor in units of mrem/yr per µCi/sec. (See Table 3-2).

(Q)mu instantaneous release rate lirnit µCi/sec.

SF, K,, X/Q See Section 11-2.2.1.1.

(2) To ensure that Part I dose rate limits are not exceeded, the Hi Hi alanns on the stack monitors shall be set lower than or equal to (0.9) (Q)m.x. Hi alanns shall be set lower than or equal to (0.5)

'(Q)max.

(3) Based on the above conservatism, the dose contribution from JAF and NMP-2 can usually be ignored. During Emergency Classifications at JAF or NMP-2 due to airborne effluent, or after emergency condenser vent releases of significant proportions, the 500 mrem/yr value may be reduced accordingly.

(4) To convert monitor gross count rates to µCi/sec release rates, the following general formula shall be applied:

(Cm-B) K, = Q = µCi/sec, release rate Where:

Cm = monitor gross count rate in cps or cpm B = monitor background count rate K, = stack monitor efficiency factor with units of

µCi/sec-cps or µCi/sec-cpm (5) Monitor K. factors shall be determined using the general formula:

Where:

Q, = individual radionuclide stack effluent release rate as determined by isotopic analysis.

K. factors more conservative than those calculated by the above methodology may be assumed.

Alternatively, when stack release rates are near the lower limit of detection, the following general formula may be used to calculate K,:

1/K E (L, F, I

= - =--------------

k Yk Ek) (3.7E4 dis/sec per µC1)

f f Where:

f = stack flow in cc/sec. ,

E efficiency in units of cpm-cc/µCi or cps-cc/µCi ( cpm = counts per minute; cps = counts per second).

fa cpm-cc/bps or cps-cc/yps.

From energy calibration curve produced during NIST traceable primary gas calibration or transfer source calibration (bps = beta per second;yps = gammas per second).

CY-NM-170-301 Revision 37 II 13 January 2020

2.1.2 Stack Monitor Setpoints (Cont'd)

Yk = b/d (betas/disintegration) or y/d (gammas/disintegration).

Fi = Activity :fraction ofnuclide i in the mixture.

nuclide counter.

k discrete energy beta or gamma emitter per nuclide counter.

s seconds.

This monitor calibration method assumes a noble gas distribution typical of a recoil release mechanism. To ensure that the calculated efficiency is conservative, beta or gamma emissions whose energy is above the range of calibration of the detector are not included in the calculation.

2.1.3 Recombiner Discharge (Off Gas) Monitor Setpoints (1) The Hi-Hi alarm points shall activate with recombiner discharge rates equal to or less than 500,000 µCi/sec. This alarm point may be set equal to or less than 1 Ci/sec for a period of time not to exceed 60 days provided the offgas treatment system is in operation; According to Part I, Note (c) to Table D 4.6.14-2, the channel functional test of the condenser air ejector radioactivity monitor shall demonstrate that automatic isolation of this pathway occurs if either of the following conditions exist:

{

i) Instruments indicate two*channels above the Hi-Hi alarm setpoint, ii) Instruments indicate one channel above Hi-Hi alarm setpoint and one channel downscale.

This automatic isolation function is tested once per operating cycle in accordance with station procedures.

(2) The Hi alarm points shall be set to activate at equal to or less than five (5) times normal full power background.

If the monitor alarms at this setpoint, the offgas will be immediately sampled and analyzed, followed by an analysis of reactor coolant sample.

(3) To convert monitor mR/hr readings to µCi/sec, the formula below shall be applied:

(R)(KR) = QR. µCi/sec recombiner discharge release rate Where:

R mR/hr monitor indicator.

KR efficiency factor in units of µCi/sec/mR/hr determined prior to setting monitor alarm points.

(4) Monitor KR factors shall be determined using the general formula:

Where:

Q, = individual radionuclide recombiner discharge release rate as determined by isotopic analysis and flow rate monitor.

KR factors more conservative than those calculated by the above methodology may be assumed.

CY-NM-170-301 Revision 37 II 14 January 2020

2.1.3 Recombiner Discharge (Off Gas) Monitor Setpoints (Cont'd)

(5) The setpoints chosen provide assurance that the total body exposure to an individual at the exclusion area boundary will not exceed a very small fraction of the limits of 10CFR Part 100 in the event this effluent is inadvertently discharged directly to the environment without treatment (thereby implementing the requirements of General Design Criteria 60 and 64 of Appendix A to 10CFR Part 50). Additionally, these setpoints serve to limit buildup of fission product activity within the station systems which would result if high fuel leakage were to be permitted over extended periods.

2.1.4 Emergency Condenser Vent Monitor Setpoint The monitor setpoint was established by dalculation ("Emergency Condenser Vent Monitor Alarm Setpoint", January 13, 1986, NMPC File Code #16199). Assuming a hypothetical case with (1) reactor water iodine concentrations higher than the Technical Specification Limit, (2) reactor water noble gas concentrations higher than would be expected at Technical Specification iodine levels, and (3) leakage of reactor steam into the emergency condenser shell at 300% of rated flow (or 1.3 E6 lbs/hr), the calculation predicts an emergency condenser vent monitor response of20 mR/hr. Such a release would result in less than 10 CFR 20 dose rate values at the site boundary and beyond for typical emergency condenser cooldown periods.

Since a 20 mR/hr monitor response can, in theory, be achievable only when reactor water iodines are higher than permitted by Technical Specifications, a conservative monitor setpoint of 5 mr/hr has been adopted.

2.1.5 Discussion 2.1.5.1 Stack Effluent Monitoring System Description The NMP-1 St&.ck Effluent Monitoring System consisted of two subsystems; the Radioactive Gaseous Effluent Monitoring System (RAGEMS) and the Offgas Effluent Stack Monitoring System (OGESMS). The OGESMS shall be used to monitor station noble gas effluents and collect particulates and iodine samples in compliance with Part I requirements.,

The RAGEMS was designed to be promptly activated from the Main Control Room for use in high range monitoring during accident situations in compliance with NUREG 0737 criteria In accordance with a letter dated September 11, 2002 from the NRC to NMPNS, LLC, Nine Mile Point Nuclear Station Unit 1 - Use of the Offgas Effluent Stack Monitoring System to Meet Regulatory Guide 1.97, Revision 2 and NUREG-0737,"

OGESMS meets the objective and purpose ofNUREG-0737 and RG 1.97. The sample-line to RAGEMS will now be used as an additional auxiliary sample point 2.1.5.2 Stack Sample Flqw Path - RAGEMS Au:riliary Sample Point The effluent sample is obtained inside the stack at elevation 530' using an isokinetic probe with four orifices. The sample line then bends radially out and back into the stack; descends down the stack and out of the stack at approximately elevation 257'; runs horizontally

( enclosed in heat tracing) some 270' along the off gas tunnel; and enters Turbine Building 250' and Offgas Building 247'.

CY-NM-170-301 Revision 37 II 15 January 2020

2.1.5.3 Stack Sample Flow Path - OGESMS The OGESMS sample is obtained from the same stack sample probe as the RAGEMS Auxiliary Sample Point From the exit of the stack at elevation 257', the sample line runs east approximately 20' and then vertically approximately 8' to the OGESMS skid. In the OGESMS, sample flows thru a particulate/iodine cartridge housing and four noble gas scintillation detectors (i.e., 07 and 08 low range beta detectors and RN-03A and RN-03B high range gamma detectors). From OGESMS, the stack sample flows back into the stack at approximately elevation 257'.

All OGESMS detector outputs are monitored and recorded remotely in the Main Control Room. Alarming capabilities are provided to alert Operators of high release rate conditions prior to exceeding Part I Control DLCO 3.6.15.b (1 Xa) whole body dose rate limits.

Stack particulate and iodine samples are retrieved manually from the OGESMS and analyzed in the laboratory using gamma spectroscopy at :frequencies and LLDs specified in Part I Table D 4.6.15-2.

2.1.5A Sampling Frequency/Sample Analysis Radioactive gaseous wastes shall be sampled and analyzed in accordance with the sampling and analysis program specified in Part I Table D 4.6.15-2. Noble gas sample and analysis frequencies are increased during elevated release rate conditions. Noble gas sample and analysis are also performed following startup, shutdown and in conjunction with each drywell purge. Particulate samples are saved and analyzed for principal gamma emitters, gross alpha, Fe-55, Sr-89,_ Sr-90 at monthly intervals minimally, and in response to an increase in noble gas release rate. The latter three analyses are performed off-site from a composite sample.

Consistent with Part I Table D 4.6.15-2, stack effluent tritium is sampled monthly, during each drywell purge, and weekly when fuel is off loaded until stable release rates are demonstrated. Samples may be analyzed on-site or off-site.

Line loss correction factors are applied to all particulate and iodine results. Correction factors of2.0 and 1.5 are used for data obtained from RAGEMS Auxiliary Sample Point and OGESMS respectively. These correction factors are based on empirical data from sampling conducted at NMP-1 in 1985 (memo from J. Blasiak to RAGEMS File, 1/6/86, "Stack Sample Representativeness Study: RAGEMS versus In-Stack Auxiliary Probe Samples").

2.1.5.5 1-133 and 1-135 Estimates Monthly, the stack effluent shall be sampled for iodines over a 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period and the 1-135/1-131 and the 1-133/1-131 ratios calculated. These ratios shall be used to calculate 1-133, 1-135 release for longer acquisition samples collected during the month.

CY-NM-170-301 Revision 37 II 16 January 2020

2.1.5.5 1-133 and 1-135 Estimates (Cont'd)

Additionally, the 1-135/1-131 and 1-133/1-131 ratios should also be determined after a significant ch.ange in the ratio is suspected (eg, plant status changes from prolonged shutdown to power operation or fuel damage has occurred). 1-135 will be included in the Radioactive Effluent Release Report in accordance with Regulatory Guide 1.21 but it will not be included when totaling dose rate or dose.

2.1.5.6 Gaseous Radwaste Treatment System Operation Part I Control DLCO 3.6. 16.b requires that the gaseous radwaste treatment system shall be functional and shall be used to reduce radioactive materials in gaseous waste prior to their discharge as necessary to meet the requirements of Part I Control DLCO 3.6.15.b.

To ensure Part I Control DLCO 3.6.15.b limits are not exceeded, and to confirm proper radwaste treatment system operation as applicable, cumulative dose contributions for the current calendar quarter and current calendar year shall be determined monthly in accordance with section 2.2 of this manual. When actual results for the gross alpha, particulate, iodine, H-3, Sr-89, Sr-90 and Fe-55 concentrations are not available, dose contributions are calculated using concentration estimates. The doses are revised when actual results are obtained.

2.2 Dose and Dose Rate Determinations In accordance with Technical Specifications 6.5.3, "Radioactive Effluent Controls Program, and ODCM Part I Controls DSR 4.6.15.b.(l), DSR 4.6.15.b.(2), and DSR 4.6.15.b.(3) dose and dose rate determinations will be made monthly to determine:

( 1) Whole body dose rates and gamma air doses at the maximum X/Q land sector site boundary interface.

(2) Skin dose rates and beta air doses at the maximum X/Q land sector site boundary interface.

(3) The critical organ dose and dose rate at a critical receptor location beyond the site boundary.

Average meteorological data (ie, maximum five year annual average X/Q and D/Q values in the case of elevated releases or 1985 annual average X/Q and D/Q values, in the case of ground level releases) shall be utilized for dose and dose rate calculations. Where average meteorological data is assumed, dose and dose rates due to noble gases at locations beyond the site boundary will be lower than equivalent site boundary dose and dose rates. Therefore, under these conditions, calculations of noble gas dose and dose rates beyond the maximum X/Q land sector site boundary locations can be neglected.

CY-NM-170-301 Revision 37 II 17 January 2020

2.2 Dose an~ Dose Rate Determinations (Cont'd)

The frequency of dose rate calculations will be upgraded when elevated release rate conditions specified in subsequent sections II-2.2.1.1 and 11-2.2.1.2 are realized.

In accordance with Technical Specification 6.5.3.g, noble gas dose rate to the whole body and skin will be calculated at the site boundary. In accordance with Technical Specification 6.5.3.h, gamma and beta air doses may be calculated at a point beyond the site boundary.

To demonstrate compliance with Technical Specification 6.5.3, "R&dioactive Effluent Controls Program", critical organ doses and dose rates may be conservatively calculated by assuming the existence of a maximum individual. This individual is a composite of the highest dose factor of each age group, for each organ and total body, and each nuclide. It is assumed that all pathways are applicable and the highest XJQ and/or D/Q value for actual pathways as noted in Table 3-1 are in effect. The maximum individual's dose is equal to the same dose that person would receive if they were simultaneously subjected to the highest pathway dose at each critical receptor identified for each pathway. The pathways include grass-(cow and goat)-milk, grass-cow-meat, vegetation, ground plane and inhalation. To comply with Part I requirements the maximum individual dose rate will be calculated at this hypothetical critical residence.

If dose -or dose rates calculated, using the assumptions noted above, reach Part I limits, actual pathways will be evaluated, and dose/dose rates may be calculated at separate critical receptor locations and compared with applicable limits.

Emergency condenser vent release contributions to the monthly dose and dose rate determinations will be considered only when the emergency condenser return isolation valves have been opened for reactor cooldown, if Emergency Condenser tube leaks develop with or without the system's return isolation valve opened, or if significant activity is detected in the Emergency Condenser Shell.

Without tube leakage, dose contributions from emergency condenser vent releases are to be determined based on condensate storage tank and emergency condenser shell isotopic distributions.

When releases from the emergency condenser have occurred, dose rate and dose determinations shall be performed using methodology in II-2.2.1 and II-2.2.2. Furthermore, environmental sampling may also be initiated to refine any actual contribution to doses. See Section II-2.4.

2.2.1 Dose Rate Dose rates will be calculated monthly, at a minimum, or when the Hi-Hi stack monitor alarm setpoint is reached, to demonstrate that dose rates resulting from the release of noble gases, tritium, iodines, and particulates with half lives greater than 8 days are within the limits

\

specified in Technical Specifications Section 6.5.3, "Radioactive Effluent Controls Program".

These limits are:

Noble Gases Whole Body Dose Rate: 500 mrem/yr Skin Dose Rate: 3000 rnrem/yr Tritium, Iodines and Particulates Organ Dose Rate: 1500 mrem/yr CY-NM-170-301 Revision 37 II 18 January 2020

2.2.1.1 Noble Gases The following noble gas dose rate equation includes the contribution from the stack (s) elevated release and the emergency condenser vent (v) ground level release when applicable (See section 11-2.2).

For whole body dose rates (mrern/sec):

DR, (mrem/sec) = 3. l 7E-8 L ,[(V, Q.. + (SF) K, (X/Q).) Q,. + (SF)K, (X/Q}.Q,v]

For skin dose rates (mrern/sec):

DRrtil(mrern/sec)= 3.17E-8 L ,[(L,(X/Q). + 1.1 l(SF)(Bi + M(X/Q).))Q11 +

(L, + 1.11 (SF)M,)(X/Q)vQ,v]

Where:

= whole body gamma dose rate (mrern/sec).

DRr+il = skin dose rate from gamma and beta radiation (mrern/sec).

= the constant accounting for the gamma whole body dose rate from stack radiation for an elevated finite plume release for each identified noble gas nuclide, i. Listed on Table 3-2 in mrern/yr per µCi/sec.

= the constant accounting for the gamma whole body dose rate from immersion in the semi-infinite cloud for each identified noble gas nuclide, i. Listed in Table 3-3 in mrern/yr per µCi/m 3 (from Reg.

Guide 1.109) the release rate of isotope i from the stack(s) or emergency condenser vent(v); (µCi/sec)

SF = structural shielding factor (dimensionless). A shielding factor of 0.7 is discussed in Table E-15 of Regulatory Guide 1.109 Revision 1.

X/Q = the relative plume concentration (in units of sec/m3) at the land sector site boundary or beyond. Average meteorological data (Table 3-1) is used. "Elevated" X/Q values are used for stack releases (s = stack);

"Ground" X/Q values are used for Emergency Condenser Vent releases (v = vent).

= the constant accounting for the beta skin dose rate from immersion in the semi-infinite cloud for each identified noble gas nuclide, i. Listed in Table 3-3 in mrern/yr per µCi/m 3 (from Reg. Guide 1.109)

B, = the constant accounting for the gamma air radiation from the elevated Finite plume resulting from stack releases for each identified noble gas nuclide, i. Listed in Table 3-2 in mrad/yr per µCi/sec.

CY-NM-170-301 Revision 37 II 19 January 2020

2.2.1.1 Noble Gases (Cont'd)

M, = the constant accounting for the gamma air dose rate from immersion in the semi-infinite cloud for each identified noble gas nuclide, i. Listed in Table 3-3 in mrad/yr per µCVm 3 (from Reg. Guide 1.109)

See Appendix B for derivation ofB, and V,.

To ensure that the site noble gas dose rate limits are not exceeded, the following procedural actions are taken if the offsite dose rates from Unit 1 exceed 10% of the limits:

1) Notify Unit 1 SM (Shift Manager) and Supervisor Chemistry.

2) Notify Unit 2 SM and Supervisor Chemistry and request the Unit 2 contribution to offsite dose rate.

3) Notify SM of the James A Fitzpatrick Nuclear Plant and request the Fitzpatrick contribution to offsite dose rate.

4) Increase the frequency of performing noble gas dose and dose rate calculations, if necessary, to ensure Site (Nine Mile Point Units 1 and 2 and Fitzpatrick) limits are not exceeded.

Additionally, alarm setpoints are set at 90% of the dose rate lirnit to ensure that site limits are not exceeded. This alarm setpoint is adjusted if the noble gas dose rate from Unit 1 is greater than 10% of the limit 2.2.1.2 Tritium, Iodines and Particulates To ensure that the 1500 mrem/year site dose rate limit is not exceeded, offsite dose rates for tritium, iodine and particulates with half lives greater than 8 days shall be calculated monthly and when release rates (Q) exceed 0.34 µCi/sec using the following equation.

Da1: (mrem/sec) = 3.17E-8 L JL ,Rua1< [W, Q" + Wv Q,v))

Where:

De1c = Total dose rate to each organ k of an individual in age group a (mrem/sec).

WJ = dispersion parameter either X/Q (sec/m3) or D/Q (l/m2) depending on pathway and receptor location assumed. Average meteorological data is used (Table 3-1 ). "Elevated" WJ values are used for stack releases (s = stack); "Ground" WJ values are used for Emergency Condenser Vent releases (v = vent).

Q, = the release rate of isotope i, from the stack (s) or vent(v); (µCVsec).

CY-NM-170-301 Revision 37 II 20 January 2020

2.2.1.2 Tritium, Iodines and Particulates (Cont'd)

R.1a1c = the dose factor for each isotope i, pathway j, age group a, and organ k (Table 3-4, through 3-22; m2-mrem/yr 'per µCi/sec for all pathways except inhalation, mrem/yr per µCi/m 3* The R values contained in Tables 3-4 through 3-22 were calculated using the methodology defined in NUREG-0133 and parameters from Regulatory Guide 1.109, Revision l; as presented in Appendix C.

3. l 7E-8 = the inverse of the number of seconds in a year.

The use of the 0.34 µCi/sec release rate threshold to perform Unit 1 dose rate calculations is justified as follows: *

(a) The 1500 mrem/yr organ dose rate limit corresponds to a minimum release rate limit of 0.34 µCi/sec calculated using the equation:

Where:

1500 = site boundary dose rate limit (mrem/year).

(R.1W/)max the maximum curie-to-dose conversion factor equal to 4.34E3 mrem-sec/µCi-yr for Sr-90, child bone for the vegetation pathway at the critical residence receptor location beyond the site boundary for an elevated release.

(b) The use of0.34 µCi/sec release rate threshold and the 4.34E3 mrem-sec/µCi-yr curie-to-dose conversion factor is conservative since curie-to-dose conversion factors for other isotopes likely to be present are 1

significantly lower.

If the organ dose rate exceeds 5% of the annual limit, the following procedural actions will be taken:

1) Notify Unit 1 SM (Shift Manager) and Supervisor Chemistry.

2) Notify Unit 2 SM and Supervisor Chemistry and request the Unit 2 contribution to offsite dose rate.

3) Notify SM of James A. Fitzpatrick Nuclear Plant and request JAF's contribution to o:ffsite dose rate.*

4) Increase the frequency of performing dose and dose rate calculations if necessary to ensure site (Nine Mile Point Units 1 and 2 and Fitzpatrick) limits are not exceeded.

CY-NM-170-301 Revision 37 II 21 January 2020

2.2.2 Dose Calculations will be performed monthly at a minimum, to demonstrate that doses resulting from the release of noble gases, tritium, iodines, and particulates with half lives greater than 8 days are within the limits specified in 10 CFR 50, Appendix I.

These limits are:

Noble Gases 5 mR gamma/calendar quarter 10 mrad beta/calendar quarter 10 mR gamma/calendar year 20 mrad beta/calendar year Tritium, Iodines and Particulates 7.5 mrem to any organ/calendar quarter 15 mrem to any organ/calendar year 2.2.2.1 Noble Gas Air Dose The following Noble Gas air dose equation includes contributions from the stack (s) elevated release and the emergency condenser vent (v) ground level release when applicable (see section II-2.2):

For gamma radiation 1 (mrad):

Dr (mrad) = 3.17£-8 L ,[(B, + Mi(X/Q).) Q'" + Mi(X/Q)v Qrv] t For beta radiation (mrad):

D~ (mrad) = 3.17E-8 L ,Ni[(X/Q). Q 10 + (X/Q)v Qrv] t Where:

Dr = gamma air dose (mrad).

D~ = beta air dose (rnrad).

Note that the units for the gamma air dose are in mrad compared to the units for the limits are in mR The NRC recognizes that 1 mR=l mrad, for gamma radiation.

B, the constant accounting for the gamma air radiation from the elevated finite plume resulting from stack releases for each identified noble gas nuclide, i. Listed in Table 3-2 in mrad/yr per µCi/sec.

Ni the constant accounting for the beta air dose from immersion in the semi-infinite cloud for each identified noble gas nuclide, 'i.

Listed on Table 3-3 in mrad/yr per µCi/m 3 (from Reg. Guide 1.109).

the release rate of isotope i, from the stack (s) or vent (v);

(µCi/sec).

CY-NM-170-301 Revision 37 II 22 January 2020

2.2.2.1 Noble Gas Air Dose (Cont'd) 3.17E-8 the inverse of the number of seconds in a year.

the constant accounting for the gamma air dose from immersion in the semi-infinite cloud for each identified noble gas nuclide, i.

Listed on Table 3-3 in mrad/yr per µCi/m 3 (from Reg. Guide 1.109).

t total time during release period, sec.

All other parameters are as defined in section II-2.2.1.1.

2.2.2.2 Tritium, Iodines and Particulates To ensure that the 15 rnrem/yr facility dose limit is not exceeded, offsite doses for tritium, iodines, and particulates with half lives greater than 8 days shall be calculated monthly using the following equation:

Dak (mrem) = 3. l 7E-8 L i[L 1~ak [W. Qa + Wv QIV)) t Where:

total dose to each organ k of an individual in age group a(rnrem).

dispersion parameter either X/Q (sec/m3) or D/Q (l/m2) depending on pathway and receptor location assumed. Average meteorological data is used (Table 3-1 ). "Elevated" Wi values are used for stack releases (s = stack); "Ground" Wi values are used for Emergency Condenser Vent releases (v = vent).

Qa,Qrv the release rate of isotope i from ~tack(s) or vent (v); (µCi/sec).

the dose factor for each isotope i, pathway j, age group a, and organ k (Tables 3-4 through 3-7, mrem/yr per µCi/m 3; Tables 3-8 through 3-22, m 2-mrem/yr per µCi/sec). R values contained in Tables 3-4 through 3-22 were calculated using the methodology defined in NUREG-0133 and parameters from Regulatory Guide 1.109, Revision l; as presented in Appendix C.

3.17E-8 the inverse of the number of seconds in a year.

t total time during the release period, sec.

CY-NM-170-301 Revision 37 II 23 January 2020

2.2.2.3 Accumulating Doses Doses will be calculated monthly, at a minimum, for gamma air, beta air, and the critical organ for each age group. Dose estimates Jlill, also, be calculated monthly prior to receipt of any offsite or onsite analysis data i.e., strontium, tritium, and iron-55. Results will be summed for each calendar quarter and year.

The critical doses are based on the following:

noble gas plume air dose

- direct radiation from ground plane deposition

- inhalation dose

- cow milk ingestion dose goat milk ingestion dose

- cow meat ingestion dose

- vegetation (food crops) ingestion dose The quarterly and annual results shall be compared to the limits listed in paragraph II-22.2. If the limits are exceeded, special reports, as required by Part I Section D 6.9.3 shall be submitted.

2.2.3 Dose Projections - Determination of Need to Operate Gaseous Radwaste Treatment System and Ventilation Exhaust Treatment System 2.2.3.1 Requirement DLCO 3.6.16.b requires that the Gaseous Radwaste Treatment System be used to reduce the radioactive materials in gaseous waste prior to their discharge as necessary to meet the requirements ofDLCO 3.6:ls. DLCO 3.6.16.b(2) requires that the Ventilation Exhaust Treatment System be used to reduce releases of radioactivity when the projected doses in 31 days would exceed 0.3mrem to any organ. These Controls implement Technical Specification 6.5.3.f that requires the Radioactive Effluent Control Program to include limitations on the functional capability and use of the gaseous effluent treatment systems (Gaseous Radwaste Treatment System AND Ventilation Exhaust Treatment System) to ensure the appropriate portions of these systems are used to reduce releases of radioactivity.

The Gaseous Radwaste Treatment System is expected to be in service. For the Ventilation Exhaust Treatment System, use is required when the projected doses in a period of 31 days would exceed 2% of the guidelines for the annual dose or dose commitment, conforming to 10CFRS0, Appendix I, i.e., 3 mrem to any organ. When Treatment systems are not in use, doses are to be projected every 31 days.

The appropriate components, which affect iodine or particulate release, to be in use are:

Rad Waste Building FLT-204-24 FLT-204-25 FLT-204-69 FLT-204-70 RSSB FLT-204-147 CY-NM-170-301 Revision 37 II 24 January 2020

2.2.3.2 Methodology Due to system design and operating procedures the charcoal beds are always operated when the offgas system is in operation. Therefore, dose projection is not relevant to detennining need to operate.

If the Gaseous Radwaste Treatment System becomes nonfunctional for more than seven days a Special Report to the NRC is ~uired. This report will include appropriate dose assessments (cumulative and projected).

If Ventilation Exhaust Treatment System components become nonfunctional which prevent building effluent from being filtered, dose projections will be performed monthly using the methodology of Section 11-2.2.2.2. Assumptions for released activity will be added to historical routine stack emissions for calculating dose during the anticipated period-of component unavailability. The calculated projected doses for iodine and particulates will be compared to the DLCO 3.6.15.b limits and Technical Specifications Section 6.5.3.f limit, 0.3 mrem to any organ.

2.3 Critical Receptors.

In accordance with the provisions of 10 CFR 20 and 10 CFR 50, Appendix L the critical receptors have been identified and are contained in Table 3-1.

For elevated noble gas releases the critical receptor is the site boundary.

When 1985 average annual X/Q values are used for ground level noble gas releases, the critical receptor is the maximum X/Q land sector site boundary interface.

For tritium, iodines, and particulates with half lives greater than eight days, the critical pathways are grass-(cow and goat)-milk, grass-cow meat, vegetation, inhalation and direct radiation (ground plane) as a result of ground deposition.

The grass-( cow and goat)-milk, and grass-cow-meat pathways will be based on the greatest D/Q location. This location has been determined in conjunction with the land use census (Part I Control DLCO 3.6.22) and is subject to change. The vegetation (food crop) pathway is based on the greatest D/Q garden location from which samples are taken. This location may also be modified as a result of vegetation sampling surveys.

The inhalation and ground plane dose pathways will be calculated at the critical residence.

Because Part I states to calculate "at the site boundary or beyond", the doses and/or dose rates must be calculated for a maximum individual who is exposed to applicable pathways at the critical residence. The maximum individual is a composite of the highest dose factor of each age group, for each organ and total body, and each nuclide.

CY-NM-170-301 Revision 37 II 25 January 2020

2.4 Refinement ofOffsite Doses Resulting from Emergency Condenser Vent Releases The doses resulting from the operation of g-ie emergency condensers and calculated in \

accordance with 11-2.2.2 may be refined using data from actual environmental samples.

Ground deposition samples will be obtained from an area or areas of maximum projected deposition. These areas are anticipated to be at or near the site boundary and near projected plume centerline. Using the methodology found in Regulatory Guide 1.109, the dose will be calculated to the maximum exposed individual. This dose will then be compared to the dose calculated in accordance with 11-2.2.2. The comparison will result in an adjustment factor of less than or greater than one which will be used to adjust the other doses from other pathways. Other environmental samples may also be collected and the resultant calculated doses to the maximum exposed individual compared to the dose calculated per 11-2.2.2. Other environmental sample media may include milk, vegetation (such as garden broadleafvegetables), etc. The adjustment factors from these pathways may be applied to the doses calculated per II-2.2.2 on a pathway by pathway basis or several pathway adjustment factors may be averaged and used to adjust calculated doses.

Doses calculated from actual environmental sample media will be based on the methodology presented in Regulatory Guide 1.109. The regulatory guide equations may be slightly modified to account for short intervals of time (less than one year) or modified for simplicity purposes by deleting decay factors. Deletion of decay factors would yield more conservative results.

CY-NM-170-301 Revision 37 II 26 January 2020

3.0 40 CFR 190 REQUIREMENTS The "Uranium Fuel Cycle" is defined in 40 CPR Part 190.02 (b) as follows:

"Uranium fuel cycle means the operations of milling of uranium ore, chemical conversion of uranium, isotopic enrichment of uranium, fabrication of uranium fuel, generation of electricity by a light-water-cooled nuclear power plant using uranium fuel, and reprocessing of spent uranium fuel, to the extent that these directly support the production of electrical power for public use utilizing nuclear energy, but excludes mining operations, operations at waste disposal sites, transportation of any radioactive material in support of these operations, and the reuse of recovered non-uranium special nuclear and by-product materials from the cycle."

Control DLCO 3.6.15.d of Part I requires that when the calculated doses associated with the effluent releases exceed twice the applicable quarter or annual limits, the licensee shall evaluate the calendar year doses and, ifrequired, submit a Special Report to the NRC and limit subsequent releases such that the 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). This report is to demonstrate that radiation exposures to all real individuals from all uranium fuel cycle sources (including all liquid and gaseous effluent pathways and direct radiation) are less than the limits in 40 CPR Part 190.

If releases that result in doses exceeding the 40 CPR 190 limits have occurred, then a variance from the NRC to permit such releases will be requested and if possible, action will be taken to reduce subsequent releases.

The report to the NRC shall contain:

I) Identification of all uranium fuel cycle facilities or operations within 5 miles of the nuclear power reactor units at the site that contribute to the annual dose of the maximum exposed member of the public.

2) Identification of the maximum exposed member of the public and a determination of the total annual dose to this person from existing pathways and sources of radioactive effluents and direct radiation.

The total body and organ doses resulting from radioactive material in liquid effluents from Nine Mile Point Unit I will be summed with the maximum dos~s resulting from the releases of noble gases, radioiodines, and particulates for the other calendar quarters (as applicable) and from the calendar quarter in which twice the limit was exceeded. The direct dose components will be determined _by either calculation or actual measurement. Actual measurements will utilize environmental TLD dosimetry. Calculated measurements will utilize engineering calculations to determine a projected direct dose component. In the event calculations are used, the methodology will be detailed as required in Part I Section D 6.9.1.e.

CY-NM-170-301 Revision 37 II 27 January 2020

3.0 40 CFR 190 REQUIREMENTS (Cont'd)

The doses frob Nine Mile Point Unit 1 (including the on-site ISFSI) will be added to the doses to the maximum exposed individual that are contributed from other uranium fuel cycle operations within 5 miles of the site. Other uranium fuel cycle facilities within 5 miles of the Site include Nine Mile Point Nuclear Station Unit 2 and the James A.

Fitzpatrick Nuclear Power Plant (including its on-site ISFSI). Doses from other facilities will be calculated in accordance with each facilities' ODCM.

For the purpose of calculating doses, the results of the Radiological Environmental Monitoring Program may be included for providing more refined estimates of doses to a real maximum exposed individual. Estimated doses, as calculated from station effluents, may be replaced by doses calculated from actual environmental sample results. Reports will include all significant details of the dose determination if radiological sampling and analyses are used to determine if the dose limits of 40CFR190 are exceeded.

3.1 Evaluation of Doses From Liquid Effluents For the evaluation of doses to real members of the public from liquid effluents, the fish consumption and shoreline sediment ground dose will be considered. Since the doses from other aquatic pathways are insignificant, fish consumption and shoreline sediment are the only two pathways that will be considered. The dose associated with fish consumption may be calculated using eflluent data and Regulatory Guide 1.109 methodology or by calculating a dose to man based on actual fish sample analysis data Because of the nature of the receptor location and the extensive fishing in the area, the critical individual may be a teenager or an adult The dose associated with shoreline sediment is based on the assumption that the shoreline would be utilized as a ~reational area. This dose may be derived from liquid eflluent data and Regulatory Guide 1.109 methodology or from actual shoreline sediment sample analysis data.

Equations used to evaluate doses from actual fish and shoreline sediment samples are based on Regulatory Guide 1.109 methodology. Because of the sample medium type and the half-lives of the radionuclides historically observed, the decay corrected portions of the equations are deleted. This does not reduce the conservatism of the calculated doses but increases the simplicity from an evaluation point of view. Table 3-23 presents the parameters used for calculating doses from liquid effluents.

The dose from fish sample media is calculated as:

= L ,[C,r(U)(DBIJ>J) fJ (IE+3)

Where:

Rai,J = The total annual dose to organ j, of an individual of age group a, from nuclide i, via fish pathway p, in mrem per year.

Cr = The concentration of radionuclide i in fish samples in pCi/gram.

u The consumption rate offish in kg/yr.

1E+3 = Grams per kilogram.

CY-NM-170-301 Revision 37 II 28 January 2020

V 3.1 Evaluation of Doses From Liquid Effluents (Cont'd)

(Daq,J) = The ingestion dose factor for age group a, nuclide i, fish pathway p, and organj, (Reg. Guide 1.109, Table E-11) (mrem/pCi).

f The fractional portion of the year over which the dose is applicable.

The dose from shoreline sediment sample media is calculated as:

R,q,J = L ,[1/4 (U)(4E+4X0.3)(DIIJJ)J) fJ Where:

R,,pJ = The total annual dose to organ j, of an individual of age group a, from nuclide i, via the sediment pathway p, in mrem per year.

~ = The concentration of radionuclide i in shoreline sediment in pCi/gram.

u The usage factor, (hr/yr) (Reg. Guide 1.109).

4E+4 The product of the assumed density of shoreline sediment ( 40 kilogram per square meter to a depth of 2.5 cm) times the number of grams per kilogram.

0.3 = The shore width factor for a lake. (

D81JlJ The dose factor for age group a, nuclide i, sediment pathway s, and organj. (Reg. Guide 1.109, Table E-6Xmremlhr per pCi/m2).

f = The fractional portion of the year over which the dose is applicable.

3.2 Evaluation of Doses From Gaseous Effluents For the evaluation of doses to real memberS of the public from gaseous effluents, the pathways contained in section 11-2.2.2.3 of the OOCM will be considered. These include the deposition, inhalation cows milk, goats milk, meat, and food products (vegetation). However, any updated field data may be utilized that concerns locations of real individuals, real time meteorological data, location of critical receptors, etc. Data from the most recent census and sample location surveys should be utilized. Doses may also be calculated from actual environmental sample media, as available. Environmental sample media data such as TLD, air sample, milk sample and vegetable (food crop) sample data may be utilized in lieu of eflluent calculational data Doses to member of the public from the pathways contained in OOCM section II-2.2.2.3 as a

result of gaseous effluents will be calculated using the dose factors of Regulatory Guide 1.109 or_the methodology of the ODCM, as applicable. Doses calculated from environmental sample media will be based on the methodologies found in Regulatory Guide 1.109.

CY-NM-170-301 Revision 37 II 29 January 2020

(

3.3 Evaluation of Doses From Direct Radiation The dose contribution as a result of direct radiation shall be considered when evaluating whether the dose limitations of 40 CPR 190 have been exceeded.

Direct radiation doses as a result of the reactor, turbine and radwaste buildings, the Independent Spent Fuel Storage Installation (ISFSI), and outside radioactive storage tanks (as applicable) may be evaluated by engineering calculations or by evaluating environmental TLD results at critical receptor locations, site boundary or other special interest locations. For the evaluation of direct radiation doses utilizing environmental TLDs, the critical receptor in question, such as the critical residence, etc., will be compared to the control locations. The comparison involves the difference in environmental TLD results between the receptor location and the average control location result.

3.4 Doses to Members of-the Public Within the Site Boundary The Radioactive Effluent Release Report shall include an assessment of the radiation doses from radioactive liquid and gaseous effluents to members of the public due to their activities inside the site boundary as defined by Figure 5.1-1 of the Technical Specifications. A member of the public, as defined in Part I, would be represented by an individual who visits the site's Energy Center for the purpose of observing the educational displays or for picnicking and associated activities.

Fishing is a major recreational activity in the area and on the Site as a result of the salmonoid and trout populations in Lake Ontario. Fishermen have been observed fishing at the shoreline near the Energy Center from April through December in all weather conditions. Thus, fishing is the major activity performed by members of the public within the site boundary. Based on the nature of the fishermen and undocumented observations, it is conservatively assumed that the maximum exposed individual spends an average of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months per week fishing from the shoreline at a location between the Energy Center and the Unit 1 facility. This estimate is considered conservative but not necessarily excessive and accounts for occasions where individuals may fish more on weekends or on a few days in March of the year.

The pathways considered for the evaluation include the inhalation pathway, the ground dose pathway with the resultant whole body and skin dose and the direct radiation dose pathway with the associated whole body dose. The direct radiation dose pathway, in actuality, includes several pathways. These include: the direct radiation gamma dose to an individual from an overhead plume, a gamma submersion plume dose (as applicable), possible direct radiation dose from the facility (including the on-site ISFSI) and a ground plane dose ( deposition).

Because the location is in close proximity to the site, any beta plume submersion dose is felt to be insignificant Other pathways, such as the ingestion pathway, are not applicable since these doses are included under calculations for doses to members of the public outside of the site boundary.

In addition, pathways associated with water related recreational activities, other than fishing, are not applicable here. These 'include swimming, boating and wading which are prohibited at the facility.

CY-NM-I 70-301 Revision 37 II 30 January 2020

3.4 Doses to Members of the Public Within the Site Boundary (Cont'd)

The inhalation pathway is evaluated by identifying the applicable radionuclides (radioiodine, tritium and particulates) in the eflluent for the appropriate time period. The radionuclide concentrations are then multiplied by the appropriate X/Q value, inhalation dose factor, air intake rate, and the fractional portion of the year in question. Thus, the inhalation pathway is evaluated using the following equation adapted from Regulatory Guide 1.109. Table 3-23 presents the reference for the parameters used in the following equation.

NOTE: The following equation is adapted from equations C-3 and C-4 of Regulatory Guide 1.109. Since many of the factors are in units of pCi/m3, in3/sec., etc.,

and since the radionuclide decay expressions have been deleted because of the short distance to the receptor location, the equation presented here is not identical to the Regulatory Guide equations.

DJa = L 1 [(Ci)F (X/Q)(DF A}ia(BR)at]

Where:

DJa = The maximum dose from all nuclides to the organ j and age group (a) in mrem/yr.

= The average concentration in the stack release of nuclide i for the period in pCi/m3

F Unit 1 average stack flowrate in m 3/sec.

XIQ The plume dispersion parameter for a location approximately 0.50 miles west of NMP-1; the plume dispersion parameter is 8.9E-06 sec/m3 (stack) and was obtained from the C.T. Main five year average annual X/Q tables. The stack ( elevated) X/Q is conservative when based on 0.50 miles because of the close proximity of the stack and the receptor location.

(DFA}ia = The dose factor for nuclide i, organj, and age group a in mrem per pCi (Reg. Guide 1.109, Table E-7).

(BR)a = Annual air intake for individuals in age group a in m 3 per year (obtained from Table E-5 of Regulatory Guide 1.109).

t Fractional portion of the year for which radionuclide i was detected and for which a dose is to be calculated (in years).

CY-NM-170-301 Revision 37 II 31 January 2020

3.4 Doses to Members of the Public Within the Site Boundary (Cont'd)

The ground dose pathway (deposition) will be evaluated by obtaining at least one soil or shoreline sediment sample in the area where :fishing occurs. The dose will then becalculated using the sample results, the time period in question, and the methodology based on Regulatory Gurde 1.109 as presented in Section II-3.1. The resulw.nt dose may be adjusted for a background dose by subtracting the applicable off-site control soil or shoreline sediment sample radionuclide activities. In the event it is noted that fishing is not performed from the shoreline, but is instead performed in the water__(i.e., the use of waders), then the ground dose pathway (deposition) may not be evaluated.

The direct radiation gamma dose pathway includes any gamma doses from an overhead plume, potential submersion in the plume, possible direct radiation from the facility (including the on-site ISFSI) and ground plane dose ( deposition). This general path~ay will be evaluated by average environmental TLD readings. At least two environmental TLDs will be utilized at one location in the approximate area where :fishing occurs. The TLDs will be placed in the field on approximately the beginning of a calendar quarter and removed on approximately the end of the calendar quarter. For the purpo~es of this evaluation, TLD data from quarters 2, 3, and 4 will be utilized.

The average TLD readings will be adjusted by the average control TLD readings. This is accomplished by subtracting the average quarterly control TLD value from the average fishing location TLD value. The applicable quarterly control TLD values will be utilized after adjusting for the appropriate time period (as applicable). In the event ofloss or theft of the TLDs, results from a TLD or TLDs in a nearby area may be utilized.

CY-NM-170-301 Revision 37 II 32 January 2020

4.0 ENVIRONMENTAL MONITORING PROGRAM 4.1 Sampling Stations The current sampling locations are specified in Table 5-1 and Figures 5.1-1, 5 .1-2. The meteorological tower is shown in Figure 5.1-1. The location is shown as TLD location 17.

The Radiological Environmental Monitoring Program is a joint effort between the owners and operators of the Nine Mile Point Unit 1 and the James A. FitzPatrick Nuclear Power Plant Sampling locations are chosen on the basis of historical average dispersion or deposition parameters from both units. The environmental sampling location coordinates shown on Table 5-1 are based on the NMP-2 reactor centerline. '

The average dispersion and deposition 'parameters have been calculated for a 5 year period, 1978 through 1982. These average dispersion or deposition parameters for the site are used to compare results of the annual land use census.

If it is determined that sample locations required by Part I are unavailable or new locations are identified that yield a significantly higher (e.g. 50%) calculated D/Q value, actions will be taken as required by Controls DLCO 3.6.20 and DLCO 3.6.22, and the Radiological Environmental Monitoring program updated accordingly.

4.2 Interlaboratory Comparison Program Analyses shall be performed on samples containing known quantities of radioactive materials that are supplied as part of a Commission approved or: sponsored Interlaboratory Comparison Program, such as the EPA Crosscheck Program. Participation shall be only for those media, e.g., air, milk, water, etc., that are included in the Nine Mile Point Environmental Monitoring Program and for which crosscheck samples are available. An attempt will be made to obtain a QC sample to program sample ratio of 5% or better. The site identification symbol or the actual Quality Control sample results shall be reported in the Annual Radiological Environmental Operating Report so that the Commission staff may evaluate the results.

Specific sample media for which EPA Cross Check Program samples are available include the following:

- gross beta in air particulate filters

- gamma emitters in air particulate filters

- gamma emitters in milk

- gamma emitters in water

- tritium in water 131 in water CY-NM-170-301 Revision 37 II 33 January 2020

4.3 Capabilities for Thermoloninescent Dosimeters Used for Environmental Measurements Required detection capabilities for thermoluminescent dosimeters used for environmental measurements required by Table D 4.6.20-1, footnote b of Part I are based on ANSI Standard N545, section 4.3. TLDs are defined as phosphors packaged for field use. In regard to the detection capabilities for thermoluminescent dosimeters, only one determination is required to evaluate the above capabilities per type of TLD. Furthermore, the above capabilities may be determined by the vendor who supplies the TLDs. Required detection capabilities are as follows:

4.3.1 Uniformity shall be determined by giving TLDs from the same batch an exposure equal to that resulting from an exposure rate of 10 mR/hr during the field cycle. The responses obtained shall have a relative standard deviation of less than 7.5%. A total of at least 5 TLDs shall be evaluated.

4.3.2 Reproducibility shall be determined by giving TLDs repeated exposures equal to that resulting from an exposure rate of 10 uR/hr during the field cycle. The average of the relative standard deviations of the responses shall be less than 3.0°/c,. A total of at least 4 TLDs shall be evaluated.

4.3.3 Dependence of exposure interpretation on the length of a field cycle shall be examined by placing TLDs for a period equal to at least a field cycle and a period equal to half the same field cycle in an area where the exposure rate is known to be constant This test shall be conducted under approximate average winter temperatures and approximate average summer temperatures. For these tests, the ratio of the response obtained in the field cycle to twice that obtained for half the field cycle shall not be less than 0.85. At least 6 TLDs shall be evaluated.

4.3.4 Energy dependence shall be evaluated by the response ofTLDs to photons for several energies between approximately 30 keV and 3 MeV. The response shall not differ from that obtained with the calibration source by more than 25% for photons with energies greater than 80 keV and shall not be enhanced by more than a factor of two for photons with energies less than 80 keV. A total of at least 8 TLDs shall be evaluated.

4.3.5 The directional dependence of the TLD response shall be determined by comparing the response of the TLD exposed in the routine orientation with respect to the calibration source with the response obtained for different orientations. To accomplish this, the TLD shall be rotated through at least two perpendicular planes. The response averaged over all directions shall not differ from the response obtained in the standard calibration position by more than 100/o. A total of at least 4 TLDs shall be evaluated.

4.3.6 Light dependence shall be determined by placing TLDs in the field for a period equal to the field cycle under the four conditions found in ANSI N545, section 4.3.6. The results obtained for the unwrapped TLDs shall not differ from those obtained for the TLDs wrapped in aluminum foil by more than 10%. A total ofat least 4 TLDs shall be evaluated for each of the four conditions. '

CY-NM-170-301 Revision 37 II 34 January 2020

4~.7 Moisture dependence shall be determined by placing TLDs (that is, the phosphors packaged for field use) for a period equal to the field cycle in an area where the exposure rate is known to be cons1:imt The TLDs shall be exposed under two conditions: ('1) packaged in a thin, sealed plastic bag, and (2) packaged in a thin, sealed plastic bag with sufficient water to yield observable moisture throughout the field cycle. The TLD or phosphor, as appropriate, shall be dried before readout The response of the TLD exposed in the plastic bag containing water s ~ not differ from that exposed in the regular plastic bag by more than 100/o. A total of at least 4 TLDs shall be evaluated for each condition. ,, -

4.3.8 Self irradiation shall be determined by placing TLDs for a period equal to the field cycle in an area where the exposure rate is less than 10 uR/hr and the exposure during the field cycle is known. If necessary, corrections shall be applied for the dependence of exposure interpretation on the length of the field cycle (ANSI N545, section 4.3.3). The average exposure inferred from the responses of the TLDs shall not differ from the known exposure by more than an exposure equal to that resulting from an exposure rate of 10 uR/hr during the field cycle. A total of at least J TLDs shall be evaluated.

CY-NM-170-301 Revision 37 II 35 January 2020

TABLE 1-1 Average Energy Per Disintegration ISOTOPE E1mev/dis .IB!:!l J Etimev/dis<4> ~

Ar-41 1.294 (3) 0.464 (3)

Kr-83m 0.00248 (1) 0.0371 (1)

Kr-85 0.0022 (1) 0.250 (1)

Kr-85m 0.159 (1) 0.253 (1)

Kr-87 0.793 (1) 1.32 (1)

Kr-88 1.95 (1) 0.377 (1)

Kr-89 2.22 (2) 1.37 (2)

Kr-90 2.10 (2) 1.01 (2)

Xe-131m 0.0201 (1) 0.143 (1)

Xe-133 0.0454 (1) 0.135 (1)

Xe-133m 0.042 (1) 0.19 (1)

Xe-135 0.247 (1) 0.317 (1)

Xe-135m 0.432 (1) 0.095 (1)

Xe-137 0.194 (1) 1.64 (1)

Xe-138 1.18 (1) 0.611 (1)

(1) ORNL-4923, Radioactive Atoms - Supplement I, M.S, Martin, November 1973.

(2) NED0-12037, "Summary of Gamma and Beta Emitters and Intensity Data"; M.E. Meek, R.S.

Gilbert, January 1970. (The average ~nergy was computed from the maximum energy using the ICRP II equation, not the 1/3 value assumption used in this reference).

(3) NCRP Report No. 58, "A Handbook of Radioactivity Measurements Procedures"; 1978 (4) The average energy includes conversion electrons.

CY-NM-170-301 Revision 37 II 36 January 2020

TABLE2-1 Ai.1 VALUES - LIQUID*

RADWASTE TANK INFANT mrem - ml hr- µCi NU!:;LIDE BONE LIVER TBODY THYROID KIDNEY LUNG GI-TRACT H3 2.90E-1 2.90E-1 2.90E-1 2.90E-1 2.90E-1 2.90E-1 Cr 51 l.29E-2 839E-3 l.83E-3 l.63E-2 3.75E-l Cu64 l.13E-l 5.23E-2 1.91E-l 2.32 Mn54 l.87E+l 4.23 4.14 6.86 Fe 55 l.31E+l 8.44 2.26 4.13 1.07 Fe 59 2.84E+l 4.96E+l l.96E+l l.47E+l 2.37E+l Co 58 3.34 8.34 833 Co 60 l.02E+ 1 2.40E+l 2.42E+l Zn65 l.72E+l 5.91E+l 2.73E+l 2.87E+l 5.00E+l Sr 89 2.32E+3 6.66E+l 4.77E+l Sr 90 l.74E+4 4.43E+3 2.17E+2 Zr 95 l.91E-l 4.66E-2 3.30E-2 5.02E-2 2.32E+l Mn56 2.40E-4 4.15E-5 2.07E-4 2.18E-2 Mo99 2.34E+l 4.57 3.50E+l 7.71 Na24 2.37 2.37 2,37 237 2.37 2.37 2.37 I 131 3.03E+l 3.57E+l 1.57E+ 1 l.17E+4 4.17E+l 1.28 I 133 4.22 6.15 1.80 l.12E+3 723 1.04 Ni65 l.33E-3 l.51E-4 6.85E-5 l.15E-2 I 132 l.58E-4 3.21E-4 l.14E-4 l.S0E-2 3.58E-4 2.60E-4 Cs 134 3.54E+2. 6.60E+2 6.67E+l l.70E+2 6.97E+l 1.79 Cs 136 4.05E+l l.19E+2 4.45E+l 4.75E+l 9.71E+l 1.81 Cs 137 4.91E+2 5.75E+2 4.07E+l l.54E+2 624E+l 1.80 Ba 140, l.50E+2 l.S0E-1 7.74 3.57E-2 9.23E-2 3.69E+l Ce 141 7.21E-2 4.40E-2 5.l 7E-3 l.36E-2 2.27E+l Nb95 3.85E-2 l.59E-2 9.18E-3 l.14E-2 l.34E+l La 140 l.18E-2 4.67E-3 l.20E-3 5.48E+l Ce 144 2.79 1.14 l.57E-l 4.62E-l l.60E+2

Calculated in accordance with NUREG 0133, Section 4.3.1; and Regulatory Guide 1.109, Regulatory position C, Section 1.

CY-NM-170-301 Revision 37 II 37 January 2020

TABLE 2-2 Aw VALUES - LIQUID*

RADWASTE TANK CHILD mrem- ml hr - µCi NUCLIDE BQNE LIVER TBODY THYROID KIDNEY LUNG GI-TRACT H3 4.39E-1 4.39E-1 4.39E-1 4.39E-1 4.39E-1 4.39E-1 Cr 51 2.13E-2 2.13E-2 1.40 7.86E-1 2.30E-1 1.42 7.31E+l Cu64 2.SlE-6 2.70 1.63 2.51E-6 ' 6.52 2.51E-6 1.27E+2 Mn54 6.92 3.38E+3 9.06E+2 6.92 9.53E+2 6.92 2.84E+3 Fe 55 9.21E+2 4.88E+2 l.51E+2 2.76E+2 9.0SE+l Fe 59 ' l.30E+3 2.11E+3 l.05E+3 1.34 1.34 6.12E+2 2.19E+3 Co 58 1.89 7.46E+l 2.24E+2 1.89 1.89 1.89 4.26E+2 Co 60 l.12E+2 3.28E+2 7.48E+2 l.12E+2 l.12E+2 1.12E+2 l.31E+3 Zn65 2.15E+4 5.73E+4 3.56E+4 3.85 3.61E+4 3.85 1.01E+4 Sr 89 3.26E+4 l.ldE-4 9.32E+2 l.l0E-4 l.l0E-4 l.lOE-4 l.26E+3 Sr90 4.26E+5 l.08E+5 5.74E+3 Zr95 1.70 1.33 1.32 1.23 1.38 1.23 l.08E+2 Mn56 l.65E-1 3.73E-2 2.00E-1 2.39E+l Mo99 5.35E-3 9.57E+l 2.37EH 5.35E-3 2.04E+2 5.35E-3 7.91E+l Na24 l.52E+2 l.52E+2 l.52E+2 l.52E+2 l.52E+2 l.52E+2 l.52E+2

' I 131 2.09E+2 2.10E+2 l.19E+2 6.94E+4 3.45E+2 5.60E-2 l.87E+l I 133 3.39E+l 4.19E+l l.59E+l 7.78E+3 6.98E+l l.38E-4 l.69E+l Ni65 2.67E-1 2.51E-2 l.47E-2 3.08 I 132 6.13E-3 l.13E-2 5.18E-3 5.22E-I , l.72E-2 1.32E-2 Cs 134 3.68E+5 6.04E+5 l.27E+5 3.54E+l l.87E+5 6.72E+4 3.29E+3 Cs 136 3.52E+4 9.67E+4 6.26E+4 6.21E-1 5.15E+4 7.68E+3 3.40E+3 Cs 137 5.15E+5 4.93E+5 7.28E+4 5.37E+ 1 l.61E+5 5.78E+4 3.14E+3 Ba 140 3.61E+2 3.96E-1 2.1 lE+l 7.96E-2 l.82E-l 2.68E-l l.83E+2 Ce 141 I.SOE-I l.07E-l 6.99E-2 6.34E-2 8.24E-2 6.34E-2 5.40E+l Nb95 5.21E+2 2.03E+2 l.45E+2 6.39E-l l.91E+2 6.39E-l 3.75E+5 La 140 1.S0E-1 5.93E-2 2.68E-2 l.03E-2 l.03E-2 l.03E-2 l.36E+3 Ce 144 5.00 1.81 6.06E-1 3.58E-l 1.16 3.58E-1 3.80E+2

Calculated in accordance with NUREG 0133, Section 4.3.1; and Regulatory Guide 1.109, Regulatory position C, Section 1.

CY-NM-170-301 Revision 37 II 38 January 2020

TABLE 2-3 A1at VALVES - LIQUID*

RADWASTE TANK TEEN mrem- ml hr- Ci NUCLIDE BONE LIVER TBOQY THYROID KIDNEY LUNG GI-TRACT H3 3.28E-1 3.28E-1 3.28E-1 3.28E-1 3.28E-1 3.28E-1 Cr51 l.02E-1 l.02E-1 1.39 8.16E-1 3.84E-1 1.94 2.16E+2 Cu 64 l.20E-5 2.89 1.36 l.20E-5 7.32 1.20E-5 2.24E+2 Mn54 3.31E+l 4.34E+3 8.87E+2 3.31E+l l.32E+3 3.31E+l 8.86E+3 Fe 55 6.94E+2 4.92E+2 l.15E+2 3.12E+2 2.13E+2 Fe 59 l.07E+3 2.49E+3 9.64E+2 6.41 6.41 7.89E+2 5.87E+3 Co 58 9.03 9.82E+l 2.15E+2 9.03 9.03 9.03 l.24E+3 Co 60 536E+2 7.96E+2 l.12E+3 5.36E+2 5.36E+2 5.36E+2 3.93E+3 Zn 65 2.10E+4 7.28E+4 3.40E+4 l.84E+ 1 4.66E+4 l.84E+l 3.08E+4 Sr 89 2.44E+4 5.24E-4 6.98E+2 5.24E-4 5.24E-4 5.24E-4 2.90E+3 Sr 90 4.66E+5 l.15E+5 l.31E+4 Zr95 6.20 6.00 5.97 5.90 6.04 5.90 2.28E+2 Mn56 1.81E-1 3.22E-2 2.29E-1 l.19E+l Mo99 2.56E-2 9.22E+l 1.76E+l 2.56E-2 2.11E+2 2.56E-2 l.65E+2 Na24 l.39E+2 1.39E+2 l.39E+2 139E+2 l.39E+2 l.39E+2 l.39E+2 1131 l.55E+2 2.17E+2 l.16E+2 6.31E+4 3.73E+2 2.68E-1 4,30E+l I 133 2.53E+l 4.29E+l 131E+l 5.99E+3 7.52E+l 6.60E-4 3.25E+l Ni65 2.08E-1 2.66E-2 l.21E-2 1.44 I 132 4.90E-2 1.28E-2 4.60E-3 4.32E-1 2.02E-2 5.59E-3 Cs 134 3.05E+5 7.18E+5 3.33E+5 1.69E+2 2.28E+5 8.73E+4 9.10E+3 Cs 136 2.98E+4 l.17E+5 7.88E+4 2.97 6.38E+4 1.01E+4 9.44E+3 Cs 137 4.09E+5 5.44E+5 1.90E+5 2.57E+2 1.85E+5 7.21E+4 7.99E+3 Ba 140 2.35E+2 4.l0E-1 l.55E+l 3.81E-1 4.79E-1 5.75E-1 3.63E+2 Ce 141 3.46E-1 3.32E-1 3.07E-1 3.04E-1 3.l 7E-1 3.04E-1 8.16E+l Nb95 4.44E+2 2.48E+2 1.18E+2 3.06 2.40E+2 3.06 1.05E+6 La 140 J.57E-1 l.02E-1 6.35E-2 4.94E-2 4.94E-2 4.94E-2 3.05E+3 Ce 144 3.99 2.65 1.83 1.71 2.27 1.71 5.74E+2

Calculated in accordance with NUREG 0133, Section 4.3.1; and Regulatory Guide 1.109, Regulatory position C, Section 1.

CY-NM-170-301 Revision 37 II 39 January 2020

TABLE2-4 Aw VALUES - LIQUID"'

RADWASTE TANK ADULT mrem - ml hr- µCi NU~LIDE BONE LIVER TBQDY TI-IYRQID KIDNEY LUNG GI-TRACT H3 4.45E-1 4.45E-1 4.45E-1 4.45E-1 4.45E-1 4.45E-1 Cr51 l.82E-2 1.82E-2 1.27 7.64E-1 2.93E-1 1.67 3.14E+2 Cu64 2.75 1.29 6.94 2.35E+2 Mn54 5.94 4.38E+3 8.41E+2 5.94 l.31E+3 5.94 l.34E+4 Fe 55 6.64E+2 4.58E+2 l.07E+2 2.56E+2 2.63E+2 Fe 59 l.03E+3 2.43E+3 9.31E+2 1.15 1.15 6.79E+2 8.09E+3 Co 58 1.62 9.15E+l 2.03E+2 1.62 1.62 1.62 l.82E+3 Co 60 9.60E+l 3.57E+2 6.71E+2 9.60E+l 9.60E+l 9.60E+l 4.99E+3 Zn65 2.31E+4 7.36E+4 3.32E+4 3.30 4.92E+4 3.30 4.63E+4 Sr 89 2.25E+4 9.39E-5 6.45E+2 9.39E-5 9.39E-5 9.39E-5 3.60E+3 I

Sr 90 5.60E+5 l.37E+5 l.62E+4 Zr95 1.36 1.15 1.12 1.06 1.21 1.06 3.06E+2 Mn56 l.73E-1 3.07E-2 2.20E-1 5.52 Mo99 4.58E-3 8.70E+l l.66E+l 4.58E-3 l.97E+2 4.58E-3 2.02E+2 Na24 l.35E+2 l.35E+2 l.35E+2 l.35E+2 l.35E+2 l.35E+2 l.35E+2 I 131 l.45E+2 2.07E+2 l.19E+2 6.79E+4 3.55E+2 4.80E-2 5.47E+l I 133 2.35E+l 4.09E+l l.25E+l 6.02E+3 7.14E+l l.18E-4 3.68E+l Ni65 l.93E-1 2.51E-2 l.14E-2 6.36E-1 I 132 4.68E-3 l.25E-2 4.38E-3 4.38E-1 2.00E-2 2.35E-3 Cs 134 2.98E+5 -7.08E+5 5.79E+5 3.03E+l 2.29E+5 7.61E+4 l.24E+4 Cs 136 2.%E+4 1.17E+5 8.42E+4 5.32E-1 6.51E+4 8.93E+3 l.33E+4 Cs 137 3.82E+5 5.22E+5 3.42E+5 4.60E+l l.77E+5 5.90E+4 l.02E+4 Ba 140 2.24E+2 3.49E-1 l.47E+l 6.83E-2 l.64E-l 2.29E-1 4.61E+2

,Ce 141 9.53E-2 8.20E-2 5.75E-2 5.44E-2 6.72E-2 5.44E-2 l.06E+2 Nb95 4.39E+2 2.44E+2 l.32E+2 5.47E-1 2.41E+2 5.47E-1 l.48E+6 La 140 l.llE-1 6.03E-2 2.24E-2 8.84E-3 8.84E-3 8.84E-3 3.78E+3 Ce 144 2.48 1.22 4.24E-1 3.07E-1 8.47E-1 3.07E-l 7.3_7E+2

Calculated in accordance with NUREG 0133, Section 4.3.1; and Regulatory Guide 1.109, Regulatory position C, Section 1.

CY-NM-170-301 Revision 37 II 40 January 2020

TABLE2-5 Ai.1 VALUES - LIQUID*

EMERGENCY CONDENSER VENT INFANT mrem - ml hr- µCi NUCLIDE BONE LIVER TBODY THYRQID KIDNEY LUNQ QI-TRACT H3 7.43E-4 7.43E-4 7.43E-4 7.43E-4 7.43E-4 7.43E-4 Cr51 3.30E-5 2.15E-5 4.70E-6 4.18E-5 9.6IE-4 Cu64 2.89E-4 l.34E-4 4.89E-4 5.94E-3 Mn54 4.79E-2 l.08E-2 l.06E-2 l.76E-2 Fe 55 335E-2 2.16E-2 5.78E-3 l.06E-2 2.75E-3 Fe 59 7.29E-2 l.27E-I 5.02E-2 3.76E-2 6.08E-2 Co 58 8.58E-3 2.14E-2 2.14E-2 Co 60 2.60E-2 6.15E~2 6.19E-2 Zn65 4.42E-2 l.52E-I 6.99E-2 7.35E-2 l.28E-I Sr 89 5.95 l.71E-1 l.22E-I Sr 90 4.46E+ I l.14E+l 5.57E-l Zr 95 4.90E-4 l.l9E-4 8.47E-5 l.29E-4 5.95E-2 Mn56 6.l 7E-7 l.06E-7 5.30E-7 5.60E-5 Mo99 6.00E-2 l.l7E-2 8.97E-2 l.98E-2 Na24 6.07E-3 6.07E-3 6.07E-3 6.07E-3 6.07E-3 6.07E-3 6.07E-3 I 131 7.77E-2 9.16E-2 4.03E-2 3.0IE+l l.07E-1 3.27E-3 I 133 l.08E-2 l.58E-2 4.62E-3 2.87 l.85E-2 2.67E-3 Ni65 3.4IE-6 3.86E-7 l.76E-7 2.94E-5 I 132 4.05E-7 8.22E-7 2.93E-7 3.85E-5 9.17E-7 6.66E-7 Cs 134 9.08E-I 1.69 l.7IE-1 4.36E-1 l.79E-I 4.60E-3 Cs 136 l.04E-I 3.06E-I l.l4E-I l.22E-I 2.49E-2 4.64E-3 Cs 137 126 1.47 l.04E-I 3.95E-I l.60E-I 4.6IE-3 Ba 14-0 3.85E-I 3.85E-4 l.99E-2 9.15E~5 2.37E-4 9.47E-2

( Ce 141 l.85E-4 l.l3E-4 l.33E-5 3.48E-5 5.82E-2 Nb95 9.88E-5 4.07E-5 2.35E-5 2.92E-5 3.43E-2 La 140 3.03E-5 l.20E-5 3.08E-6 l.4IE-I Ce 144 7.16E-3 2.93E-3 4.02E-4 l.l9E-3 4.llE-1

Calculated in accordance with NUREG 0133, Section 4.3.1; and Regulatory Guide 1.109, Regulatory position C,Section I.

CY-NM-170-301 Revision 37 II 41 January 2020

TABLE2-6 Ai.1 VALUES - LIQUID"'

EMERGENCY CONDENSER VENT CHILD mrem- ml hr- µCi NU~LIDE BONE LIVER TBODY THYROID KIDNEY LUNQ GI-TRACT H3 l.44E-l l.44E-l l.44E-l l.44E-l l.44E-1 l.44E-1 Cr51 3.78E-5 3.78E-5 1.37 7.58E-l 2.07E-l 1.38 7.24E+l Cu 64 2.63 1.59 6.35 123E+2 Mn54 l.23E-2 3.36E+3 8.95E+2 l.23E-2 9.42E+2 l.23E-2 2.82E+3 Fe 55 9.04E+2 4.79E+2 l .49E+2 2.71E+2 8.88E+l Fe 59 l.28E+3 2.07E+3 l.03E+3 2.38E-3 2.38E-3 6.00E+2 2.15E+3 Co 58 3.36E-3 7.0IE+l 2.15E+2 3.36E-3 3.36E-3 3.36E-3 4.09E+2 Co 60 l.99E-l 2.08E+2 6.14E+2 l.99E-1 l.99E-l l.99E-1 l.15E+3 Zn 65 2.15E+4 5.73E+4 3.56E+4 6.84E-3 3.61E+4 6.84E-3 l.0IE+4 Sr 89 3.07E+4 8.78E+2 l.19E+3 Sr 90 4.0IE+5 l.02E+5 5.40E+3 Zr95 3.0IE-1 6.78E-2 6.06E-2 2.19E-3 9.61E-2 2.19E-3 6.84E+l Mn56 1.65E-1 3.73E-2 2.00E-1 2.39E+l Mo99 8.16E+l 2.02E+l l.74E+2 6.75E+l Na24 l.50E+2 l.50E+2 l.50E+2 l.50E+2 l.50E+2 l.50E+2 l.50E+2 I 131 l.86E+2 l.87E+2 l.06E+2 6.19E+4 3.08E+2 l.67E+ 1 I 133 3.08E+l 3.81E+I 1.44E+l 7.07E+3 6.35E+l l.53E+ 1 Ni65 2.66E-l 2.50E-2 l.46E-2 3.07 I 132 6.0IE-3 l.I0E-2 5.08E-3 5.12E-1 l.69E-2 l.30E-2 Cs 134 3.68E+5 6.04E+5 l.27E+5 6.29E-2 l.87E+5 6.71E+4 3.25E+3 Cs 136 3.51E+4 9.66E+4 6.25E+4 l.I0E-3 5.14E+4 7.67E+3 3.40E+3 Cs 137 5.14E+5 4.92E+5 7.27E+4 9.55E-2 l.60E+5 5.77E+4 3.08E+3 Ba 140 2.48E+2 2.l 7E-1 l.45E+l l.42E-4 7.09E-2 l.30E-l l.26E+2 Ce 141 3.08E-2 l.54E-2 2.39E-3 l.13E-4 6.83E-3 1.13E-4 l.91E+l Nb95 5.21E+2 2.03E+2 l.45E+2 l.14E-3 l.90E+2 l.14E-3 3.75E+5 La 140 l.31E-l 4.59E-2 l.55E-2 l.83E-5 l.83E-5 l.83E-5 l.28E+3 Ce 144 1.64 5. ISE-1 8.81E-2 6.36E-4 2.85E-l 6.36E-4 l.34E+2

Calculated in accordance with NUREG 0133, Section 4.3.1; and Regulatory Guide 1.109, Regulatory position C,Section I.

CY-NM-170-301 Revision 37 1142 January 2020

TABLE2-7 A,.1 VALUES - LIQUID*

EMERGENCY CONDENSER VENT TEEN mrem - ml '

hr- µCi NUCLIDE BONE LIVER TBODY THYROID KIDNEY LUNQ GI-TRACT H3 1.74E-l 1.74E-l 1.74E-l l.74E-l l.74E-l 1.74E-l Cr 51 l.81E-4 l.81E-4 1.28 7.12E-l 2.81E-l 1.83 2.15E+2 2.86 1.35 7.24 222E+2 1 Cu 64 Mn54 5.89E-2 4.29E+3 8.52E+2 5.89E-2l l.28E+3 5.89E-2 8.81E+3 Fe 55 6.89E+2 4.88E+2 1.14E+2 3.10E+2 2.11E+2 Fe 59 l.05E+3 2.46E+3 9.50E+2 1.14E-2 l.14E-2 7.76E+2 5.82E+3

\

Co 58 l.61E-2 8.78E+l 2.02E+2 l.61E-2 l.61E-2 l.61E-2 1.21E'+3 Co60 9.53E-l 2.57E+2 5.78E+2 9.53E-l 9.53E-l 9.53E-l 3.34E+3 Zn 65 2.10E+4 7.28E+4 3.39E+4 3.28E-2 4.66E+4 3.28E-2 3.08E+4 Sr 89 2.38E+4 6.81E+2 2.83E+3 Sr 90 4.54E+5 l.12E+5 l.27E+4 Zr95 2.56E-l 8.80E-2 6.38E-2 l.05E-2 l.24E-l l.05E-2 1.79E+2 Mn56 l.81E-l 322E-2 2.29E-l l.19E+l Mo99 - 8.57E+l l.63E+l l.96E+2 l.54E+2 Na24 l.38E+2 l.38E+2 l.38E+2 l.38E+2 l.38E+2 l.38E+2 l.38E+2 I 131 l.47E+2 2.06E+2 l.10E+2 6.00E+4 3.54E+2 4.77E-4 4.07E+l I 133 2.42E+l 4.llE+l l.25E+ 1 5.74E+3 7.21E+l 3.1 lE+l Ni65 2.08E-l 2.66E-2 l.21E-2 1.44 I 132 4.86E-3 1.27E-2 4.56E-3 429E-1 2.00E-2 5.54E-3 Cs 134 3.05E+5 7.18E+5 3.33E+5 3.0lE-1 2.28E+5 8.71E+4 8.93E+3 Cs 136 2.98E+4 l.17E+5 7.87E+4 5.28E-3 6.38E+4 l.01E+4 9.43E+3 Cs 137 4.09E+5 5.44E+5 l.89E+5 4.57E-1 1.85E+5 7.19E+4 7.73E+3 Ba 140 l.%E+2 2.47E-2 l.27E+l 6.77E-4 8.23E-2 l.62E-1 3.03E+2

.Ce 141 2.43E-2 1.64-E-2 2.36E-3 5.40E-4 8.02E-3 5.40E-4 4.54E+l Nb95 4.41E+2 2.45E+2 l.15E+2 5.43E-3 2.37E+2 5.43E-3 l.05E+6 La 140 1.05E-l 5.17E-2 l.38E-2 8.78E-5 8.78E-5 8.78E-5 2.96E+3 Ce 144 127 528E-l 7.12E-2 3.04E-3 3.17E-l 3.04E-3 3.19E+2

Calculated in accordance with NUREG 0133, Section 4.3.1;-and Regulatory Guide 1.109, Regulatory position C, Section 1.

CY-NM-170-301 Revision 37 II 43 January 2020

TABLE2-8 Ai.1 VALVES - LIQUID*

EMERGENCY CONDENSER VENT ADULT mrem - ml hr- Ci NU~LIDE BQNE LIVER TBODY THYROID KIDNEY LUNG QI-TRACT H3 2.27E-l 2.27E-l 2.27E-l 2.27E-l 2.27E-1 2.27E-1 Cr51 3.24E-5 3.24E-5 1.24 7.43E-1 2.74E-1 1.65 3.12E+2 Cu64 2.72 1.28 6.86 2.32E+2 Mn54 l.06E-2 4.37E+3 8.33E+2 l.06E-2 l.30E+3 1.06E-2 l.34E+4 Fe 55 6.58E+2 4.55E+2 l.06E+2 2.54E+2 2.6IE+2 Fe 59 1.02E+3 2.4IE+3 9.22E+2 2.04E-3 2.04E-3 6.72E+2 8.02E+3 Co 58 2.88E-3 8.83E+l l.98E+2 2.88E-3 2.88E-3 2.88E-3 l.79E+3 Co60 l.7IE-l 2.56E+2 5.65E+2 l.7IE-1 l.7IE-l l.71E-l 4.8IE+3 Zn65 2.3IE+4 7.36E+4 3.32E+4 5.87E-3 4.92E+4 5.87E-3 4.63E+4 Sr 89 2.18E+4 6.27E+2 3.50E+3 Sr 90 5.44E+5 l.34E+5 1.57E+4 Zr95 2.40E-1 7.8IE-2 5.35E-2 1.88E-3 l.22E-1 l.88E-3 2.42E+2 Mn56 l.73E-l 3.07E-2 2.20E-1 5.52 Mo99 8.04E+l l.53E+l l.82E+2 l.86E+2 Na24 l.34E+2 l.34E+2 l.34E+2 1.34E+2 l.34E+2 l.34E+2 l.34E+2 I 131 1.37E+2 l.96E+2 l.12E+2 6.43E+4 3.36E+2 5.17E+l I 133 2.25E+I 3.9IE+l l.19E+l 5.75E+3 6.82E+I 3.51E+l Ni65 l.93E-l 2.50E-2 l.14E-2 6.36E-l I 132 4.64E-3 l.24E-2 4.34E-3 4.34E-I l.98E-2 2.33E-3 Cs 134 2.98E+5 7.08E+5 5.79E+5 5.39E-2 2.29E+5 7.61E+4 l.24E+4 Cs 136 2.96E+4 1.17E+5 8.42E+4 9.46E-4 6.51E+4 8.92E+3 1.33E+4 Cs 137 3.82E+5 522E+5 3.42E+5 8.19E-2 l.77E+5 5.89E+4 l.01E+4 Ba 140 l.84E+2 2.32E-I l.2IE+l l.21E-4 7.88E-2 l.33E-I 3.79E+2 Ce 141 2.21E-2 l.50E-2 l.78E-3 9.67E-5 7.00E-3 9.67E-5 5.68E+l Nb95 4.38E+2 2.44E+2 1.3IE+2 9.73E-4 2.4IE+2 9.73E-4 l.48E+6 Lal40 9.90E-2 4.99E-2 l.32E-2 l.57E-5 1.57E-5 1.57E-5 3.66E+3 Ce 144 1.17 4.89E-1 6.33E-2 5.45E-4 2.90E-I 5.45E-4 3.95E+2

Calculated in accordance with NUREG 0133, Section 4.3.1; and Regulatory Guide 1.109, Regulatory position C, Section 1.

CY-NM-170-301 Revision 37 II 44 January 2020

TABLE3-1 Critical Receptor Dispersion Parameters" For Ground Level and Elevated Releases ELEVATED ELEVATED GROUNDe GRO~

LOCATION DIR MILES X/0 (sec/mJl D/0 (m-21 X/O(sec/mJl D/0 (m-21 Residences E (98°) 1.4 1.8 E-07b 5.2 E-09b 4.02 E-07 8.58 E-09 Dairy Cowsr SE (130°) 2.6 2.2 E-08c 7.0 E-lOC 6.00 E-08 1.64 E-09 MilkGoatsf SE (130°) 2.6 2.2 E-08c 7.0 E-lOC 6.00 E-08 1.64 E-09 Meat Animals ESE (115°) 1.8 5.1 E-08c 1.7 E-09° 1.16 E-07 3.54 E-09 Gardens E (97°) 1.8 1.0 E-07c 3.5 E-09° 2.53 E-07 5.55 E-09 Site Boundary ENE (67°) 0.4 2.4 E-06b,d 4.4 E-08c,d 6.63 E-06 6.35 E-08

_,r

a. These values will be used in dose calculations beginning in April 1986 but may be revised periodically to account for changes in locations offarms, gardens or critical residences.

b. Values based on 5 year annual meteorological data (C.T. Main, Rev. 2)

c. Values based on 5 year average grazing season meteorological data (C.T. Main Rev. 2)

d. Values are based on most restrictive X/Q land-based sector (ENE). (C.T. Main, Rev. 2)

e. Values are based on average annual meteorological data for the year 1985.

f. Conservative location based on past dairy cow and goat milk history.

CY-NM-170-301 Revision 37 II 45 January 2020

TABLE3-2 Gamma Air and Whole Body Plume Shine Dose Factors*

For Noble Gases Gamma Whole Gamma AirB. BodyV1 mrad!yr mrem/yr Nuclide uCl/sec µCi/sec 2.23E-6 Kr-85m l.75E-3 1.68E-3 Kr-87 l.02E-2 9.65E-3 Kr-88 2.23E-2 2.17E-2 Kr-89 2.50E-2 l.71E-2 Kr-83m 2.26E-6 Xe-133 2.91E-4 l.75E-4 Xe-133m 2.27E-4 l.87E-4 Xe-135 2.62E-3 2.50E-3 Xe-135m 5.20E-3 4.89E-3 Xe-137 2.30E-3 2.20E-3 Xe-138 l.54E-2 l.03E-2 Xe-131m l.74E-5 l.47E-6 Ar-41 l.64E-2 1.57E-2

Calculated in accordance with Regulatory Guide 1.109. (See Appendix B.)

CY-NM-170-301 Revision 37 II 46 January 2020

TABLE3-3 IMMERSION DOSE FACTORS FOR NOBLE GASES*

Nuclide K,(y-Body)** y(6--Skin)** Mi{y-Air)*** ~1(6--Air)***

Kr83m 7.56E-02 l.93El 2.88E2 Kr85m l.17E3 l.46E3 l.23E3 l.97E3 Kr85 l.61El l.34E3 l.72El l.95E3

~

Kr87 5.92E3 9.73E3 6.17E3 1.03E4 Kr88 l.47E4 237E3 1.52E4 2.93E3 Kr89 1.66E4 l.01E4 l.73E4 l.06E4 Kr90 1.56E4 729E3 l.63E4 7.83E3 Xe 131m 9.15El 4.76E2 1.56E2 1.11E3 Xe 133m 2.51E2 9.94E2 327E2 1.48E3 Xe 133 2.94E2 3.06E2 3.53E2 l.05E3 Xe 135m 3.12E3 7.l 1E2 3.36E3 7.39E2 Xe 135 l.81E3 1.86E3 l.92E3 2.46E3 Xe137 l.42E3 122E4 l.51E3 l.27E4 Xe 138 8.83E3 4.13E3 9.21E3 4.75E3 Ar41 8.84E3 2.69E3 9.30E3 3.28E3

From. Table B-1.Regulatory Guide 1.109 Rev. 1

- mrem/yr per µCi/m 3 *

mrad/yr per µCi/m 3

CY-NM-170-301 Revision 37 II 47 January 2020

TABLE3-4 DOSE AND DOSE RATE Ri VALUES- INHALATION - INFANT' mrem/yr

µCi/m 3 NUCLIDE BONE LIVER TBODY THYROID KIDNEY LUNG GI-LLI H3 6.47E2 6.47E2 6.47E2 6.47E2 6.47E2 6.47E2 C 14 2.65E4 5.31E3 531E3 5.31E3 5.31E3 5.31E3 5.31E3 Cr51 8.95El 5.75El l.32El l.28E4 3.57E2 Mn54 2.53E4 4.98E3 4.98E3 1.00E6 7.06E3 Fe 55 l.97E4 l.17E4 333E3 8.69E4 1.09E3 Fe 59 l.36E4 2.35E4 9.48E3 l.02E6 2.48E4 Co58 l.22E3 l.82E3 7.77E5 l.11E4 Co60 8.02E3 l.18E4 4.51E6 3.19E4 Zn65 l.93E4 6.26E4 3.l 1E4 3.25E4 6.47E5 5.14E4 Sr 89 3.98E5 l.14E4 2.03E6 6.40E4 Sr 90 4.09E7 2.59E6 l.12E7 l.31E5 Zr95 1.15E5 2.79E4 2.03E4 3.11E4 l.75E6 2.17E4 Nb95 l.57E4 6.43E3 3.78E3 4.72E3 4.79E5 l.27E4 Mo99 l.65E2 3.23El 2.65E2 l.35E5 4.87E4 I-131 3.79E4 4.44E4 l.96E4 1.48E7 5.18E4 l.06E3 I 133 l.32E4 l.92E4 5.60E3 3.56E6 2.24E4 2.16E3 Cs 134 3.%E5 7.03E5 7.45E4 l.90E5 7.97E4 l.33E3 Cs 137 5.49E5 6.12E5 4.55E4 l.72E5 7.13E4 1.33E3 Ba 140 5.60E4 5.60El 2.90E3 l.34El l.60E6 3.84E4 La 140 5.05E2 2.00E2 5.15El l.68E5 8.48E4 Ce 141 2.77E4 1.67E4 l.99E3 5.25E3 5.17E5 2.16E4 Ce144 3.19E6 l.21E6 l.76E5 5.38E5 9.84E6 1.48E5 Nd 147 7.94E3 8.13E3 5.00E2 3.15E3 3.22E5 3.12E4 1

This and following R, Tables Calculated in accordance with NUREG 0133, Section 5.3.1, except C 14 values in accordance with Regulatory Guide 1.109 Equation C-8.

J CY-NM-170-301 Revision 37 II 48 January 2020

TABLE3-5 DOSE AND DOSE RA TE Rt VALUES- INHALATION -CHILD mrem/yr

µCl/m 3 NUCLIDE BONE LIVER TBODY THYROID KIDNEY LUNG GI-LLI H3 l.12E3 l.12E3 l.12E3 l.12E3 1.12E3 l.12E3 C 14 3.59E4 6.73E3 6.73E3 6.73E3 6.73E3 6.73E3 6.73E3 Cr 51 1.54E2 8.55El 2.43El l.70E4 l.08E3 Mn54 4.29E4 9.51E3 l.00E4 l.58E6 229E4 Fe 55 4.74E4 2.52E4 7.77E3 l.11E5 2.87E3 Fe 59 2.07E4 3.34E4 l.67E4 1.27E6 7.07E4 Co 58 l.77E3 3.16E3 l.11E6 3.44E4 Co60 l.31E4 226E4 7.07E6 9.62E4 Zn65 4.26E4 l.13E5 7.03E4 7.14E4 9.95E5 l.63E4 Sr 89 5.99E5 1.72E4 2.16E6 l.67E5 Sr 90 l.01E8 6.44E6 1.48E7 3.43E5 Zr 95 l.90E5 4.18E4 3.70E4 5.96E4 223E6 6.11E4 Nb95 2.35E4 9.18E3 6.55E3 8.62E3 6.14E5 3.70E4 Mo99 1.72E2 4.26El 3.92E2 1.35E5 127E5 I 131 4.81E4 / 4.81E4 2.73E4 1.62E7 7.88E4 2.84E3 I 133 l.66E4 2.03E4 7.70E3 3.85E6 3.38E4 5.48E3 Cs 134 6.51E5 1.01E6 2.25E5 3.30E5 121E5 3.85E3 Cs 137 9.07E5 8.25E5 1.28E5 2.82E5 l.04E5 3.62E3 Ba 140 7.40E4 6.48El 4.33E3 2.llEI l.74E6 1.02E5 La 140 6.44E2 2.25E2 7.55El 1.83E5 2.26E5 Ce 141 3.92E4 1.95E4 2.90E3 8.55E3 5.44E5 5.66E4 Ce 144 6.77E6 2.12E6 3.61E5 l.17E6 l.20E7 3.89E5 Nd 147 1.08E4 8.73E3 6.81E2 4.81E3 3.28E5 821E4 CY-NM-170-301 Revision 37 II 49 January 2020

TABLE3-6 DOSE AND DOSE RATE Ri VALUES - INHALATION - TEEN mrem/yr

µCi/m 3 NUCLIDE BONE LIVER TBODY THYRQID KIDNEY LUNG GI-LLI H3 l.27E3 l.27E3 l.27E3 l.27E3 l.27E3 127£3 C 14 2.60£3 4.87E3 4.87£3 4.87E3 4.87E3 4.87E3 4.87E3 Cr 51 l.35E2 7.50EI 3.07EI 2.IOE4 3.00E3 Mn54 5.1IE4 8.40E3 l.27E4 1.98£6 6.68E4 Fe 55 3.34E4 2.38£4 5.54E3 l.24E5 6.39£3 Fe 59 l.59E4 3.70E4 l.43E4 1.53£6 l.78E5 Co 58 2.07E3 2.78E3 1.34£6 9.52E4 Co 60 1.5IE4 l.98E4 8.72E6 2.59E5 Zn65 3.86E4 l.34E5 6.24E4 8.64E4 1.24£6 4.66E4 Sr 89 4.34E5 l.25E4 2.42E6 3.7IE5 Sr 90 l.08E8 6.68E6 l.65E7 7.65E5 Zr95 l.46E5 4.58E4 3.15E4 6.74E4 2.69£6 l.49E5 Nb95 l.86E4 l.03E4 5.66£3 l.00E4 7.51E5 9.68E4 Mo99 l.69E2 3.22El 4.1IE2 l.54E5 2.69E5 I 131 3.54E4 4.9IE4 2.64E4 l.46E7 8.40E4 6.49E3 I 133 l.22E4 2.05E4 6.22£3 2.92E6 3.59E4 l.03E4 Cs 134 5.02£5 1.13£6 5.49E5 3.75E5 l.46E5 9.76£3 Cs 137 6.70E5 8.48E5 3.11E5 3.04E5 1.2IE5 8.48£3 Ba 140 5.47E4 6.70El 3.52£3 2.28El 2.03£6 2.29E5 La 140 4.79E2 2.36£2 6.26El 2.14E5 4.87E5 Ce 141 2.84E4 l:90E4 2.17£3 8.88E3 6.14E5 l.26E5 Ce 144 4.89£6 2.02£6 2.62E5 l.21E6 l.34E7 8.64E5 Nd 147 7.86£3 8.56£3 5.13E2 5.02£3 3.72E5 l.82E5 CY-NM-170-301 Revision 37 II 50 January 2020

TABLE3-7 DOSE AND DOSE RATE Ri VALUES - INHALATION - ADULT mrem/yr

µCi/m 3 NUCLIDE BONE LIVER TBQDY THYROID KIDNEY LUNG GI-LLI H3 l.26E3 l.26E3 1.26E3 1.26E3 1.26E3 l.26E3 C 14 1.82E4 3.41E3 3.41E3 3.41E3 3.41E3 3.41E3 3.41E3 Cr51 l.00E2 5.95El 2.28El 1.44E4 3.32E3 Mn54 3.96E4 6.30E3 9.84E3 1.40E6 7.74E4 Fe 55 2.46E4 1.70E4 3.94E3 7.21E4 6.03E3 Fe 59 l.18E5 2.78E4 1.06E4 l.02E6 l.88E5 Co 58 1.58E3 2.07E3 9.28E5 1.06E5 Co60 l.15E4 1.48E4 5.97E6 2.85E5 Zn65 3.24E4 1.03E5 4.66E4 6.90E4 8.64E5 5.34E4 Sr 89 3.04E5 8.72E3 l.40E6 3.50E5 Sr90 9.92E7 6.I0E6 9.60E6 7.22E5 Zr95 1_.07E5 3.44E4 2.33E4 5.42E4 1.77E6 l.50E5 Nb95 1.41E4 7.82E3 4.21E3 7.74E3 5.05E5 1.04E5 Mo99 1.21E2 2.30El 2.91E2 9.12E4 2.48E5 I 131 2.52E4 3.58E4 2.05E4 l.19E7 6.13E4 6.28E3 I 133 8.64E3 1.48E4 4.52E3 2.15E6 2.58E4 8.88E3 Cs 134 3.73E5 8.48E5 7.28E5 2.87E5 9.76E4 l.04E4 Cs 137 4.78E5 6.21E5 4.28E5 2.22E5 7.52E4 8.40E3 Ba 14-0 3.90E4 4.90El 2.57E3 l.67El l.27E6 2.18E5 La 140 3.44E2 1.74E2 4.58El l.36E5 4.58E5 Ce 141 1.99E4 l.35E4 l.53E3 6.26E3 3.62E5 1.20E5 Ce 144 3.43E6 1.43E6 l.84E5 8.48E5 7.78E6 8.16E5 Nd 147 5.27E3 6.10E3 3.65E2 3.56E3 2.21E5 l.73E5 CY-NM-170-301 Revision 37 II 51 January 2020

TABLE3-8 DOSE AND DOSE RATE Ri VALUES - GROUND PLANE ALL AGE GROUPS

,m.1 -mrem/yr

µCi/sec NUCLIDE TOTAL BODY SKIN H3 C 14 Cr51 4.65E6 5.50E6 Mn54 l.40E9 l.64E9 Fe 55 Fe 59 2.73E8 3.20E8 Co 58 3.80E8 4.45E8 Co 60 2.15EI0 2.53EI0 Zn65 7.46E8 8.57E8 Sr 89 2.16E4 2.5IE4 Sr 90 Zr 95 2.45E8 2.85E8 Nb95 l.36E8 l.61E8 Mo99 3.99E6 4.63E6 I 131 l.72E7 2.09E7 I 133 2.39E6 2.91E6 Cs 134 6.83E9 7.97E9 Cs 137 l.03EI0 120EIO Ba 140 2.05E7 2.35E7 La 140 l.92E7 2.18E7 Ce 141, l.37E7 l.54E7 Ce 144 6.96E7 8.07E7 Nd 147 8.46E6 1.0IE7 CY-NM-170-301 Revision 37 II 52 January 2020

TABLE3-9 DOSE AND DOSE RA TE Ri VALVES - COW MILK - INF ANT ml-mrem/yr

µCi/sec NUCLIDE BONE LIVER TBODY THYROID KIDNE:X LUNQ GILLI H3* 2.38E3 238E3 2.38E3 2.38E3 2.38E3 2.38E3 C 14* 3.23E6 6.89E5 6.89E5 6.89E5 6._89E5 6.89E5 6.89E5 Cr 51 8.35E4 5.45£4 l.19E4 l.06E5 2.43E6 Mn54 2.5IE7 5.68E6 5.56E6 9.21E6 Fe 55 8.43£7 5.44E7 l.45E7 2.66E7 6.91E6 Fe 59 l.22E8 2.13E8 8.38E7 6.29E7 l.02E8 Co 58 l.39E7 3.46E7 3.46E7 Co 60 5.90E7 l.39E8 1.40E8 Zn65 3.53E9 l.21E10 5.58E9 5.87E9 l.02E10 Sr 89 6.93E9 l.99E8 l.42E8 Sr 90 8.19E10 2.09EIO l.02E9 Zr 95 3.85E3 9.39E2 6.66E2 l.01E3 4.68E5 Nb95 421E5 l.64E5 l.17E5 l.54E5 3.03E8 Mo99 l.04E8 2.03E7 1.55E8 3.43E7 I 131 6.81E8 8.02E8 3.53E8 2.64El I 9.37E8 2.86E7 I 133 8.52E6 l.24E7 3.63E6 2.26E9 1.46E7 2.I0E6 Cs 134 2.41EI0 4.49EIO 4.54E9 l.16E10 4.74E9 l.22E8 Cs 137 3.47EIO 4.06EIO 2.88E9 l.09EI0 4.41E9 l.27E8 Ba 140 l.21E8 1.21E5 6.22E6 2.87E4 7.42E4 2.97E7 La 140 2.03EI 7.99 2.06 9.39E4 Ce 141 2.28E4 l.39E4 l.64E3 4.28E3 7.18E6 Ce 144 l.49E6 6.IOE5 8.34E4 2.46E5 8.54E7 Nd 147 4.43E2 4.55E2 2.79EI l.76E2 2.89E5

mrem/yr per µCi/m 3

CY-NM-170-301 rRevision 37 II 53 January 2020

TABLE3-10 DOSE AND DOSE RATE R. VALVES - COW MILK - CHILD m1-mrem/yr

µCi/sec NUCLIDE BONE LIVER TBODY THYROID KIDNEY LUNG GI-LLI H3* 1.57E3 l.57E3 1.57E3 l.57E3 l.57E3 l.57E3 C 14* 1.65E6 3.29E5 3.29E5 3.29E5 3.29E5 3.29E5 3.29E5 Cr 51 5.27E4 2.93E4 7.99E3 5.34E4 2.80E6 Mn54 l.35E7 3.59E6 3.78E6 l.13E7 Fe 55 6.97E7 3.07E7 l.15E7 2.09E7 6.85E6 Fe 59 6.52E7 l.06E8 5.26E7 3.06E7 l.10E8 Co58 6.94E6 2.13E7 4.05E7 Co60 2.89E7 8.52E7 l.60E8 Zn65 2.63E9 7.00E9 4.35E9 4.41E9 l.23E9 Sr 89 3.64E9 l.04E8 l.41E8 Sr 90 7.53E10 l.91E10 l.01E9 Zr 95 2.17E3 4.77E2 4.25E2 6.83E2 4.98E5 Nb95 l.86E5 l.03E4 5.69E4 l.00E5 4.42E8 Mo99 4.07E7 l.01E7 8.69E7 3.37E7 I 131 3.26E8 3.28E8 1.86£8 l.08El 1 5.39£8 2.92E7 I 133 4.04E6 4.99E6 l.89E6 9.27£8 8.32E6 2.01E6 Cs134 l.50El0 2.45E10 5.18E9 7.61E9 2.73E9 l.32E8 Cs 137 2.17E10 2.08E10 3.07E9 6.78E9 2.44E9 l.30E8 Ba 140 5.87E7 5.14E4 3.43E6 l.67E4 3.07E4 2.97E7 La 140 9.70 3.39 1.14 9.45E4 Ce 141 l.15E4 5.73E3 8.51E2 2.51E3 7.15E6 Ce 144 l.04E6 3.26E5 5.55E4 l.80E5 8.49E7 Nd 147 2.24E2 l.81E2 l.40El 9.94El 2.87E5

mrem/yr per µCi/m 3*

CY-NM-170-301 Revision 37 II 54 January 2020

TABLE 3-11 DOSE AND DOSE RA TE Ri VALUES - COW MILK - TEEN m2-mrem/yr

µCi/sec NUCLIDE BONE LIVER TBODY THYROID KIDNEY LUNG GI-LLI H3* 9.94£2 9.94E2 9.94E2 9.94£2 9.94E2 9.94E2 C 14* 6.70E5 l.34E5 l.34E5 l.34E5 l.34E5 l.35E5 l.34E5 Cr51 2.58E4 1.44E4 5.66E3 3.69E4 4.34E6 Mn54 9.01E6 l.79E6 2.69E6 l.85E7 Fe 55 2.78E7 l.97E7 4.59E6 125E7 8.52E6 Fe59 2.81E7 6.57E7 2.54E7 2.07E7 1.55£8 Co 58 4.55E6 1.05E7 6.27E7 Co 60 l.86E7 4.19E7 2.42£8 Zn65 1.34E9 4.65E9 2.17E9 2.97E9 L97E9 Sr 89 l.47E9 4.21E7 1.75£8 Sr90 4.45EIO 1.I0EIO l.25E9 Zr95 9.34E2 2.95E2 2.03E2 4.33E2 6.80E5 Nb95 l.86E5 l.03E5 5.69E4 l.O0E5 4.42£8 Mo99 2.24E7 4.27E6 5.12E7 4.01E7 I 131 1.34£8 1.88£8 1.01E8 5.49EI0 3.24E8 3.72E7 I 133 1.66E6 2.82E6 8.59E5 3.93E8 4.94E6 2.13E6 Cs 134 6.49E9 l.53EIO 7.08E9 4.85E9 l.85E9 l.90E8 Cs 137 9.02E9 l.20E10 4.18E9 4.08E9 l.59E9 l.71E8 Ba 140 2.43E7 2.98E4 l.57E6 l.01E4 2.00E4 3.75E7 La 140 4.05 1.99 5.30E-1 1.14E5 Ce 141 4.67E3 3.12E3 3.58E2 1.47E3 8.91E6 Ce 144 4.22E5 l.74E5 2.27E4 l.04E5 l.06E8 Nd 147 9.12El 9.91El 5.94EO 5.82El 3.58E5

mrem/yr per µCi/m 3

CY-NM-170-301 Revision 37 II 55 January 2020

TABLE3-12 DOSE AND DOSE RATE Ri VALVES - COW MILK - ADULT nr-mrem/yr

µCi/sec NUCLIDE BONE LIVER JBODY THYROID KIDNEY LUNG GI-LLI H3* 7.63E2 7.63E2 7.63E2 7.63E2 7.63E2 7.63E2 C 14* 3.63E5 7.26E4 7.26E4 726E4 7.26E4 7.26E4 7.26E4 Cr 51 l.48E4 8.85E3 3.26E3 l.96E4 3.72E6 Mn54 5.41E6 l.03E6 l.61E6 l.66E7 Fe 55 l.57E7 l.08E7 2.52E6 6.04E6 6.21E6 Fe59 l.61E7 3.79E7 1.45E7 l.06E7 l.26E8 Co 58 2.70E6 6.05E6 5.47E7 Co 60 1.I0E7 2.42E7 2.06E8 Zn65 8.71E8 2.77E9 l.25E9 l.85E9 l.75E9 Sr 89 7.99E8 2.29E7 l.28E8 Sr90 3.15E10 7.74E9 9.l 1E8 Zr95 5.34E2 l.71E2 l.16E2 2.69E2 5.43E5 Nb95 l.09E5 6.07E4 3.27E4 6.00E4 3.69E8 Mo99 l.24E7 2.36E6 2.81E7 2.87E7 I 131 7.41E7 l.06E8 - 6.08E7 3.47El0 l.82E8 2.80E7 I 133 9.09E5 l.58E6 4.82E5 2.32E8 2.76E6 l.42E6 Cs 134 3.74E9 8.89E9 7.27E9 2.88E9 9.55E8 l.56E8 Cs 137 4.97E9 6.80E9 4.46E9 2.31E9 7.68E8 l.32E8 Ba 140 l.35E7 l.69E4 8.83E5 5.75E3 9.69E3 2.77E7 La 140 2.26 1.14 3.0IE-1 8.35E4 Ce 141 2.54E3 l.72E3 l.95E2 7.99E2 6.58E6 Ce 144 2.29E5 9.58E4 l.23E4 5.68E4 7.74E7 Nd 147 4.74El 5.48El 3.28E0 3.20El 2.63E5

mrem/yr per µCi/m 3

CY-NM-170-301 Revision 37 II 56 January 2020

TABLE 3-13 DOSE AND DOSE RATE Rt VALUE~ - GOAT MILK - INFANT m2-mrem/yr

µCi/sec NUCLIDE BONE LIVER TBODY THYROID KIDNEY LUNG GI-LLI H3* 6.33E3 6.33E3 6.33E3 6.33E3 6.33E3 6.33E3 C 14* 3.23E6 6.89E5 6.89E5 6.89E5 6.89E5 6.89E5 6.89E5 Cr 51 l.00E4 6.56E3 1.43E3 1.28E4 2.93E5 Mn54 3.01E6 6.82E5 6.67E5 1.l 1E6 Fe 55 1.I0E6 7.08E5 l.89E5 3.46E5 8.98E4 Fe 59 l.59E6 2.78E6 1.09E6 8.21E5 1.33E6 Co 58 l.67E6 4.16E6 4.16E6 Co60 7.08E6 l.67E7 l.68E7 Zn65 4.24E8 1.45E9 6.70E8 7.04E8 l.23E9 Sr 89 l.48E10 4.24E8 3.04E8 Sr90 l.72El 1 4.38EIO 2.15E9 Zr95 4.66E2 1.13E2 8.04El 1.22E2 5.65E4 Nb95 9.42E4 3.88E4 2.24E4 2.78E4 3.27E7 Mo99 l.27E7 2.47E6 l.89E7 4.17E6 I 131 8.17E8 9.63E8 4.23E8 3.16El 1 l.12E9 3.44E7 I 133 l.02E7 1.49E7 4.36E6 2.71E9 l.75E7 2.52E6 Cs 134 7.23E10 1.35Ell 1.36EIO 3.47EIO l.42EIO 3.66E8 Cs 137 l.04El 1 l.22Ell 8.63E9 3.27El0 l.32E10 3.81E8 Ba 140 l.45E7 1.45E4 7.48E5 3.44E3 8.91E3 3.56E6 La 140 2.43E0 9.59E-1 2.47E-1 l.13E4 Ce 141 2.74E3 l.67E3 l.96E2 5.14E2 ' 8.62E5 Ce 144 l.79E5 7.32E4 1.00E4 2.96E4 1.03E7 Nd 147 5.32El 5.47El 3.35E0 2.1 IEl 3.46E4

mrem/yr per µCi/m 3

CY-NM-170-301 Revision 37 II 57 January 2020

TABLE 3-14 DOSE AND DOSE RATE R1 VALUES - GOAT MILK - CHILD m2-mrem/yr

µCi/sec NUCLIDE BONE LIVER TBODY THYROID KIDNEY LUNG GI-LLI H3* 4.17E3 4.17E3 4.17E3 4.17E3 4.17E3 4.17E3 C 14* 1.65E6 3.29E5 3.29E5 3.29E5 3.29E5 3.29E5 3.29E5 Cr51 6.34E3 3.52E3 9.62E2 6.43E3 336E5 Mn54 l.62E6 4.31E5 4.54E5 l.36E6 Fe 55 9.06E5 4.81E5 1.49E5 2.72E5 8.91E4 Fe 59 8.52E5 l.38E6 6.86E5 3.99E5 1.43E6 Co 58 8.35E5 2.56E6 4.87E6 Co 60 3.47E6 1.02E7 1.92E7 Zn65 3.15E8 8.40E8 5.23E8 5.29E8 l.48E8 Sr 89 7.77E9 2.22E8 3.01E8 Sr90 1.58Ell 4.0lEI0 2.13E9 Zr95 2.62E2 5.76El 5.13El 8.25El 6.01E4 Nb 95 5.05E4 1.96E4 l.40E4 -:- 1.85E4 3.63E7 Mo99 4.95E6 l.22E6\ l.06E7 4.09E6 I 131 3.91E8 3.94E8 2.24E8 l.30Ell 6.46E8 3.50E7 I 133 4.84E6 5.99E6 2.27E6 l.11E9 9.98E6 2.41E6 Cs 134 4.49El0 7.37El0 1.55El0 2.28El0 8.19E9 3.97E8 Cs 137 6.52E10 6.24El0 9.21E9 2.03El0 7.32E9 3.91E8 Ba 140 7.05E6 6.18E3 4.12E5 2.01E3 3.68E3 3.57E6 La 140 l.16EO 4.07E-1 l.37E-1 l.13E4 Ce 141 l.38E3 6.88E2 1.02E2 3.02E2 8.59E5 Ce 144 l.25E5 3.91E4 6.66E3 2.16E4 1.02E7 Nd 147 2.68El 2. l 7El 1.68EO l.19El 3.44E4

mrem/yr per µCi/m 3

CY-NM=l 70-301 Revision 37 II 58 January 2020

TABLE 3-15 DOSE AND DOSE RATE Ri VALUES - GOAT MILK - TEEN m2-mrem/yr

µCi/sec NUCLIDE BONE LIVER TBODY THYROID KIDNEY LUNG GI-LLI H 3* 2.64E3 2.64E3 2.64E3 2.64E3 2.64E3 2.64E3 C 14* 6.70E5 l.34E5 l.34E5 l.34E5 l.34E5 l.35E5 l.34E5 Cr 51 3.l 1E3 l.73E3 6.82E2 4.44E3 5.23E5 Mn54 l.08E6 2.15E5 3.23E5 2.22E6 Fe 55 3.61E5 2.56E5 5.97E4 l.62E5 l.IIE5 Fe 59 3.67E5 8.57E5 331E5 2.70E5 2.03E6 Co 58 5.46E5 l.26E6 7.53E6 Co60 2.23E6 5.03E6 2.91E7 Zn65 l.61E8 5.58E8 2.60E8 3.57E8 2.36E8 Sr 89 3.14E9 8.99E7 3.74E8 Sr 90 9.36EIO 2.3IE10 2.63E9 Zr95 l.13E2 3.56El 2.45El 5.23El 8.22E4 Nb95 2.23E4 l.24E4 6.82E3 l.20E4 5.30E7 Mo99 2.72E6 5.19E5 6.23E6 4.87E6 I 131 . l.61E8 2.26E8 l.21E8 6.59EIO 3.89E8 4.47E7 I 133 l.99E6 3.38E6 l.03E6 4.72E8 5.93E6 2.56E6 Csl34 l.95EIO 4.58EIO 2.13EIO l.46El0 5.56E9 5.70E8 Cs 137 2.71EIO 3.60EIO l.25EIO 1.23EIO 4.76E9 5.12E8 Ba 140 2.92E6 3.58E3 l.88E5 l.21E3 2.41E3 4.50E6 La 140 4.86E-l 2.39E-l 6.36E-2 l.37E4 Ce 141 5.60E2 3.74E2 4.30El I.76E2 l.07E6 Ce 144 5.06E4 2.09E4 2.72E3 l.25E4 l.27E7 Nd 147 l.09El l.19El 7.13E-l 6.99E0 4.29E4

mrem/yr per µCi/m 3

CY-NM-170-301 Revision 37 II 59 January 2020

TABLE 3-16 DOSE AND DOSE RATE Rt VALUES - GOAT MILK - ADULT

!!t-mrem!yr

µCl/sec NUCLIDE BONE LIVER TBODY THYROID KIDNEY LUNG GI-LLI H3* 2.03E3 2.03E3 2.03E3 2.03E3 2.03E3 2.03E3 C 14* 3.63E5 7.26E4 7.26E4 7.26E4 7.26E4 7.26E4 7.26E4 Cr51 l.78E3 l.06E3 3.92E2 2.36E3 4.48E5 Mn54 6.50E5 l.24E5 l.93E5 1.99E6 Fe 55 2.04E5 1.41E5 3.28E4 7.85E4 8.07E4 Fe 59 2.10E5 4.95E5 l.90E5 138E5 1.65E6 Co58 3.25E5 7.27E5 6.58E6 Co60 1.32E6 2.91E6 2.48E7 Zn65 l.05E8 3.33E8 l.51E8 2.23E8 2.10E8 Sr 89 l.70E9 4.89E7 2.73E8 Sr90 6.62E10 l.63E10 l.91E9 Zr95 6.45El 2.07El l.40El 3.25El 6.56E4 Nb95 l.31E4 7.29E3 3.92E3 7.21E3 4.42E7 Mo99 l.51E6 2.87E5 3.41E6 3.49E6 I 131 8.89E7 l.27E8 7.29E7 4.17E10 2.18E8 3.36E7 I 133 l.09E6 1.90E6 5.79E5 2.79E8 3.31E6 l.71E6 Cs 134 l.12E10 2.67E10 2.18E10 8.63E9 , 2.86E9 4.67E8 Cs 137 l.49E10 2.04E10 l.34E10 6.93E9 2.30E9 3.95E8 Ba 140 1.62E6 2.03E3 l.06E5 6.91E2 l.16E3 3.33E6 La 140 2.71E-1 l.36E-1 3.61E-2 l.00E4 Ce 141 3.06E2 2.07E2 2.34El 9.60El 7.90E5 Ce 144 2.75E4 l.15E4 1.48E3 6.82E3 9.30E6 Nd 147 5.69EO 6.57EO 3.93E-l 3.84E0 3.15E4

mrem/yr per µCi/m 3

CY-NM-170-301 Revision 37 II 60 January 2020

TABLE3-17 DOSE AND DOSE RATE Ri VALUES - COW MEAT - CHILD

!!12-mrem/yr

µCi/sec NUCLIDE BONE LIVER TBODY THYROID KIDNEY LUNG GI-LLI H3* '2.34E2 2.34E2 2.34E2 2.34E2 2.34E2 2.34E2 C 14* 5.29E5 l.06E5 l.06E5 l.06E5 l.06E5 l.06E5 l.06E5 Cr51 4.55E3 2.52E3 6.90E2 4.61E3 2.41E5 Mn54 5.15E6 l.37E6 l.44E6 4.32E6 Fe 55 2.89E8 1.53E8 4.74E7 8.66E7 2.84E7 Fe 59 2.04E8 3.30E8 l.65E8 9.58E7 3.44E8 Co 58 9.41E6 2.88E7 5.49E7 Co 60 4.64E7 l.37E8 2.57E8 Zn65 2.38E8 6.35E8 3.95E8 4.Q0E8 l.12E8 Sr 89 2.65E8 7.57E6 l.03E7 Sr 90 7.0IE9 l.78E9 9.44E7 Zr95 l.51E6 3.32E5 2.95E5 4.75E5 3.46E8 Nb95 4.I0E6 l.59E6 l.14E6 l.50E6 2.95E9 Mo99 5.42E4 l.34E4 l.16E5 4.48E4 I 131 4.15E6 4.18E6 2.37E6 l.38E9 6.86E6 3.72E5 I 133 9.38E-2 l.16E-l 4.39E-2 2.15El l.93E-l 4.67E-2 Cs 134 6.09E8 l.00E9 2.l 1E8 3.IOE8 l.11E8 5.39E6 Cs 137 8.99E8 8.60E8 l.27E8 2.80E8 l.01E8 5.39E6 Ba 140 2.20E7 l.93E4 l.28E6 6.27E3 l.15E4 l.l IE7 La 140 2.80E-2 9.78E-3 3.30E-3 2.73E2 Ce 141 l.17E4 5.82E3 8.64E2 2.55E3 7.26E6 Ce 1-44 l.48E6 4.65E5 7.91E4 2.57E5 l.21E8 Nd 147 5.93E3 4.80E3 3.72E2 2.64E3 7.61E6

mrem/yr per µCi/m 3*

CY-NM-170-301 Revision 37 II 61 January 2020

TABLE 3-18 DOSE AND DOSE RATE Rt VALUES - COW MEAT - TEEN

) nr-mrem/yr

µCl/sec NUCLIDE BONE LIVER TBODY THYROID KIDNEY LUNG GI-LLI H 3* l.94E2 l.94E2 l.94E2 l.94E2 l.94E2 l.94E2 C 14* 2.8IE5 5.62E4 5.62E4 5.62E4 5.62E4 5.62E4 5.62E4 Cr51 2.93E3 l.62E3 6.39E2 4.16E3 4.90E5 Mn54 4.50E6 8.93E5 l.34E6 924E6 Fe 55 l.50E8 l.07E8 2.49E7 6.77E7 4.62E7 Fe 59 l.15E8 2.69E8 l.04E8 8.47E7 6.36E8 Co 58 8.05E6 l.86E7 l.l 1E8 Co 60 3.90E7 8.80E7 5.09E8 Zn65 l.59E8 5.52E8 2.57E8 3.53E8 2.34E8 Sr 89 l.40E8 4.0IE6 l.67E7 Sr90 5.42E9 l.34E9 l.52E8 Zr95 8.50E5 2.68E5 l.84E5 3.94E5 6.19E8 Nb95 2.37E6 l.32E6 7.24E5 l.28E6 5.63E9 Mo99 3.90E4 7.43E3 8.92E4 6.98E4 I 131 2.24E6 3.13E6 l.68E6 9.15E8 5.40E6 6.20E5 I 133 5.05E-2 8.57E-2 2.6IE-2 l.20El l.50E-1 6.48E-2 Cs 134 3.46E8 8.13E8 3.77E8 2.58E8 9.87E7 1.0IE7 Cs 137 4.88E8 6.49E8 2.26E8 2.21E8 8.58E7 9.24E6 Ba 140 1.19E7 l.46E4 7.68E5 4.95E3 9.81E3 l.84E7 La 140 1.53E-2 7.51E-3 2.00E-3 4.3 IE2 Ce 141 6.19E3 4.14E3 4.75E2 l.95E3 1.18E7 Ce 144 7.87E5 3.26E5 4.23E4 l.94E5 l.98E8 Nd 147 3.16E3 3.44E3 2.06E2 2.02E3 l.24E7

mrem/yr per µCi/m 3

CY-NM-170-301 Revision 37 II 62 January 2020

TABLE 3-19 DOSE AND DOSE RATE Ri VALUES- COW MEAT-ADULT ml-mremfyr

µCl/sec NUCLIDE BONE LNER IBODY THYROID KIDNEY LUNG GI-LLI H3* 3.25E2 3.25E2 3.25E2 3.25E2 3.25E2 3.25E2 C 14* 3.33E5 6.~E4 6.66E4 6.66E4 6.66E4 6.66E4 6.66E4 Cr51 3.65E3 2.18E3 8.03E2 4.84E3 9.17E5 Mn54 5.90E6 l.13E6 l.76E6 l.81E7 Fe 55 l.85E8 l.28E8 2.98E7 7.14E7 7.34E7 Fe 59 l.44E8 339E8 130E8 9.46E7 l.13E9 Co58 l.04E7 2.34E7 2.12E8 Co60 5.03E7 l.11E8 9.45E8 Zn65 2.26E8 7.19E8 3.25E8 4.81E8 4.53E8 Sr 89 l.66E8 4.76E6 2.66E7 Sr 90 8.38E9 2.06E9 2.42E8 Zr95 l.06E6 3.40E5 2.30E5 5.34E5 l.08E9 Nb95 3.04E6 l.69E6 9.08E5 l.67E6 l.03E10 Mo99 4.71E4 8.97E3 l.07E5 l.09E5 I 131 2.69E6 3.85E6 2.21E6 126E9 6.61E6 1.02E6 I 133 6.04E-2 l.05E-1 3.20E-2 l.54El l.83E-1 9.44E-2 Cs 134 4.35E8 l.03E9 8.45E8 3.35E8 l.l 1E8 l.81E7 Cs 137 5.88E8 8.04E8 5.26E8 2.73E8 9.07E7 l.56E7 Ba 140 l.44E7 l.81E4 9.44E5 6.15E3 l.04E4 2.97E7 La 140 l.86E-2 9.37E-3 2.48E-3 6.88E2 Ce 141 7.38E3 4.99E3 5.66E2 2.32E3 l.91E7 Ce 144 9.33E5 3.90E5 5.01E4 2.31E5 3.16E8 Nd 147 3.59E3 4.15E3 2.48E2 2.42E3 l.99E7

mrem/yr per µCi/m 3

CY-NM-170-301 Revision 37 II 63 January 2020

TABLE3-20 DOSE AND DOSE RATE Ri VALUES-VEGETATION - CHILD

!!t-mrem/yr

µCl/sec NUCLIDE BONE LIVER TBODY THYROID

KIDNEY LUNG GI-LLI H3* 4.01E3 4.0IE3 4.0IE3 4.01E3 4.01E3 4.0IE3 C 14* 3.50E6 7.01E5 7.01E5 7.01E5 7.01E5 7.0IES 7.01E5 Cr51 l.l7E5 6.49E4 l.77E4 l.l8E5 6.20E6 Mn54 6.65E8 l.77E8 l.86E8 5.58E8 Fe 55 7.63E8 4.05E8 1.25E8 2.29E8 7.50E7 Fe 59 3.97E8 6.42E8 3.20E8 l.86E8 6.69E8 Co 58 6.45E7 l.97E8 3.76E8 Co60 3.78E8 l.12E9 2.IOE9 Zn65 8.12E8 2.16E9 135E9 l.36E9 3.80E8 Sr 89 3.59E10 l.03E9 l.39E9 Sr 90 l.24E12 3.15Ell l.67EI0 Zr95 3.86E6 8.50E5 7.56E5 l.22E6 8.86E8 Nb95 l.02E6 3.99E5 2.85E5 3.75E5 7.37E8 Mo99 7.70E6 l.91E6 l.65E7 6.37E6 I 131 7.16E7 7.20E7 4.09E7 2.38EI0 l.l8E8 6.41E6 I 133 l.69E6 2.09E6 7.92E5 3.89E8 3.49E6 8.44E5 Cs 134 l.60EIO 2.63EIO 5.55E9 8.15E9 2.93E9 l.42E8 Cs 137. 2.39E10 2.29EI0 3.38E9 7.46E9 2.68E9 l.43E8 Ba 140 2.77E8 2.43E5 l.62E7 7.90E4 1.45E5 l.40E8 La 140 3.25E3 1.13£3 3.83E2 3.16E7 Ce 141 6.56E5 3.27E5 4.85E4 l.43E5 4.08E8 Ce 144 l.27E8 3.98E7 6.78E6 2.21E7 l.04EI0 Nd 147 7.23E4 5.86E4 4.54E3 3.22E4 9.28E7

)

mrem/yr per µCi/m 3

CY-NM-170-301 Revision 37 II 64 January 2020

TABLE3-21 DOSE AND DOSE RATE Ri VALUES-VEGETATION -TEEN

!!!1-mrem/yr

µCi/sec NUCLIDE BONE LIVER TBODY THYROID KIDNEY LUNG GI-LLI H3* 2.59E3 2.59E3 2.59E3 2.59E3 2.59E3 2.59E3 C 14* l.45E6 2.91E5 2.9IE5 2.91E5 2.9IE5 2.9IE5- 2.9IE5 Cr 51 6.16E4 3.42E4 l.35E4 8.79E4 l.03E7 Mn54 4.54E8 9.0IE7 l.36E8 9.32E8 Fe 55 3.I0E8 2.20E8 5.13E7 l.40E8 9.53E7 Fe 59 l.79E8 4.18E8 l.61E8 l.32E8 9.89E8 Co 58 4.37E7 l.01E8 6.02E8 Co60 2.49E8 5.60E8 3.24E9 Zn65 4.24E8 l.47E9 6.86E8 9.41E8 6.23E8 Sr 89 l.5IEIO 4.33E8 l.80E9 Sr 90 7.5IEII l.85Ell 2.1 IEIO Zr95 l.72E6 5.44E5 3.74E5 7.99E5 l.26E9 Nb 95 4.80E5 2.66E5 l.46E5 2.58E5 l.14E9 Mo99 5.64E6 l.08E6 l.29E7 1.0IE7 I 131 3.85E7 5.39E7 2.89E7 l.57EI0 9.28E7 l.07E7 I 133 9.29E5 l.58E6 4.80E5 2.20E8 2.76E6 1.19E6 Cs 134 7.IOE9 l.67EI0 7.75E9 5.3IE9 2.03E9 2.08E8 Cs 137 I.0IEIO l.35EI0 4.69E9 4.59E9 l.78E9 l.92E8 Ba 140 l.38E8 l.69E5 8.91E6 5.74E4 l.14E5 2.l3E8 La 140 l.81E3 8.88E2 2.36E2 5.I0E7 Ce 141 2.83E5 l.89E5 2.17E4 8.89E4 5.40E8 Ce 144 5.27E7 2.18E7 2.83E6 l.30E7 l.33E10 Nd 147 3.66E4 3.98E4 2.38E3 2.34E4 l.44E8

mrem/yr per µCi/m 3 CY-NM-170-301 Revision 37 II 65 January 2020

TABLE3-22 DOSE AND DOSE RATE Ri VALUES - VEGETATION - ADULT m1-mrem/yr

µCi/sec NUCLIDE BONE LIVER T.BODY THYROID KIDNEY LUNG GI-LLI H3* '2.26E3 2.26E3 2.26E3 2.26E3 2.26E3 2.26E3 C 14* 8.97E5 1.79E5 1.79E5 1.79E5 1.79E5 1.79E5 1.79E5 Cr51 4.64E4 2.77E4 1.02E4 6.15E4 l.17E7 Mn54 3.13E8 5.97E7 9.31E7 9.58E8 Fe 55 2.00E8 1.38E8 3.22E7 7.69E7 7.91E7 Fe59 1.26E8 2.96E8 l.13E8 8.27E7 l.02E9 Co 58 3.08E7 6.90E7 6.24E8 Co 60 1.67E8 3.69E8 3.14E9 Zn65 3.17E8 1.01E9 4.56E8 6.75E8 6.36E8 Sr 89 9.96E9 2.86E8 1.60E9 Sr 90 6.05El 1 1.48Ell 1.75E10 Zr95 l.18E6 3.77E5 2.55E5 5.92E5 l.20E9 Nb95 3.55E5 l.98E5 1.06E5 1.95£5 l.20E9 Mo99 6.14£6 1.17£6 1.39E7 1.42£7 I 131 4.04E7 5.78E7 3.31E7 l.90E10 9.91E7 l.53E7 I 133 l.00E6 1.74£6 5.30E5 2.56E8 3.03£6 1.56£6 Cs 134 4.67E9 l.llEl0 9.08E9 3.59E9 l.19E9 l.94E8 Cs 137 6.36E9 8.70E9 5.70£9 2.95£9 9.81E8 l.68E8 Ba 140 1.29£8 l.61E5 8.42E6 5.49E4 9.25E4 2.65E8 La 140 l.98E3 9.97E2 2.63£2 7.32£7 Ce 141 1.97E5 1.33£5 l.51E4 6.19E4 5.09E8 Ce 144 3.29E7 1.38E7 1.77£6 8.16£6 1.1 lEl0 Nd 147 336E4 3.88E4 2.32E3 2.27£4 l.86E8

mrem/yr per µCi/m 3 CY-NM-170-301 Revision 37 II 66 January 2020

TABLE3-23 PARAMETERS FOR THE EVALUATION OF DOSES TO REAL MEMBERS OF THE PUBLIC FROM GASEOUS AND LIQUID EFFLUENTS Pathway Parameters Vaine Reference Fish U (kg/yr) - adult 21 Reg. Guide 1.109 Table E-5 Fish DaipJ (rnrem/pCi) Each Radionuclide Reg. Guide 1.109 Table E-11 Shoreline u (hr/yr)

- adult 67 Reg. Guide 1.109

- teen 67 Assumed to be same as Adult Shoreline DllljlJ Each Radionuclide Reg. Guide 1.109 (mrem/hr per pCi/m2) Table E-6 Inhalation DFAua Each Radionuclide Reg. Guide 1.109 Table E-7 CY-NM-170-301 Revision 37 II 67 January 2020

TABLE 5.1 NINE MILE POINT NUCLEAR STATION RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM SAMPLING WCATIONS

Map Collection Site Type of Sample Location (Env. Program No.) Location Radioiodine and 1 Nine Mile Point Road 1.8 mi@92°E Particulates (air) North (R-1)

Radioiodine and 2 Co. Rt. 29 & Lake Road (R-2) 1.1 mi@ 106° ESE Particulates (air)

Radioiodine and 3 Co. Rt 29 (R-3) 1.4 mi @ 134° SE Particulates (air)

Radioiodine and 4 Village of Lycoming, NY (R-4) 1.8 mi@ 145° SE Particulates (air)

Radioiodine and 5 Montario Point Road (R-5) 16.2 mi@42° NE Particulates (air)

Direct Radiation (TLD) 6 North Shoreline Area (75) 0.1 mi @354° N Direct Radiation (TLD) 7 North Shoreline Area (76) 0.1 mi @ 27° NNE Direct Radiation (TLD) 8 North Shoreline Area (77) 0.2 mi @37°NE Direct Radiation (TLD) 9 North Shoreline Area (23) 0.8 mi@ 74° ENE Direct Radiation (TLD) 10 JAF East Boundary (78) 1.0 mi@86°E Direct Radiation (TLD) 11 Rt 29 (79) 1.2 mi@ 121° ESE Direct Radiation (TLD) 12 Rt 29 (80) 1.5 mi @ 136° SE Direct Radiation (TLD) 13 Miner Road (81) 1.7 mi@ 160° SSE Direct Radiation (TLD) 14 Miner Road (82) 1.6 mi@ 180° s

_Direct Radiation (TLD) 15 Lakeview Road (83) 12 mi@203° SSW Direct Radiation (TLD) 16 Lakeview Road (84) 1.1 mi @ 225° SW Direct Radiation (TLD) 17 Site Meteorological Tower (7) 0.7 mi@ 244° WSW Direct Radiation (TLD) 18 Energy Information Center (18) 0.5 mi @ 266° W

Map = See Figures 5.1-1 and 5.1-2 CY-NM-170-301 Revision 37 II 68 'January 2020

TABLE 5.1 NINE MILE POINT NUCLEAR STATION RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM SAMPLING LOCATIONS

Map Collection Site Tl'.J!!l of S!!ml!le Location (Env. Proi:;ram No.) Location Direct Radiation (TLD) 19 North Shoreline (85) 0.2 mi @ 290° WNW Direct Radiation (TLD) 20 North Shoreline (86) 0.lmi@310°NW Direct Radiation (TLD) 21 North Shoreline (87) 0.1 mi@ 332° NNW Direct Radiation (TLD) 22 Hickory Grove (88) 4.5 mi@97°E Direct Radiation (TLD) 23 Leavitt Road (89) 4.3 mi@ 112° ESE Direct Radiation (TLD) 24 Rt. 104 (90) 4.2 mi @ 135° SE Direct Radiation (TLD) 25 Rt. SIA (91) 4.9 mi @ 157° SSE Direct Radiation (TLD) 26 Maiden Lane Road (92) 4.5 mi@ 183° s Direct Radiation (TLD) 27 Co. Rt. 53 (93) 4.4 mi @ 206° SSW Direct Radiation (TLD) 28 Co. Rt. 1 (94) 4.4 mi @ 224° SW Direct Radiation (TLD) 29 Lake Shoreline (95) 3.7 mi@239° WSW Direct Radiation (TLD) 30 Phoenix, NY Control (49) 19.7 mi@ 168° SSE Direct Radiation (TLD) 31 S. W. Oswego, Control (14) 12.5 mi @ 227° SW Direct Radiation (TLD) 32 Scriba, NY (96) 3.7 mi@ 199° s~w Direct Radiation (TLD) 33 Novelis, Rt IA (58) 3.0 mi@ 222° SW Direct Radiation (TLD) 34 Lycoming, NY (97) 1.8 mi@ 145° SE Direct Radiation (TLD) 35 New Haven, NY (56) 5.2 mi@ 124° SE Direct Radiation (TLD) 36 W. Boundary, Bible Camp (15) 0.9 mi @ 239° WSW Direct Radiation (TLD) 37 Lake Road (98) 12 mi@ 103° ESE Surface Water 38 OSS Inlet Canal (NA) 7.6 mi @236° SW Surface Water 39 JAFNPP Inlet Canal (NA) 0.5 mi@ 71° ENE (NA) = Not applicable

Map = See Figures 5.1-1 and 5.1-2 CY-NM-170-301 Revision 37 II 69 January 2020

TABLE 5.1 (Cont'd)

NINE MILE POINT NUCLEAR STATION RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM SAMPLING LOCATIONS

Map Collection Site Type of Sample Location {Env. Program No.) Location Shoreline Sediment 40 Sunset Bay Shoreline (NA) 1.2 mi@84°E Fish 41 NMP Site Discharge Area (NA) 0.3 mi@315°NW (and/or)

Fish 42 NMP Site Discharge Area (NA) 0.6 mi @55°NE Fish 43 Oswego Harbor Area (NA) 5.9 mi@ 237° WSW Milk 64 Milk Location #55 8.8 mi@97°E Milk(CR) 77 Milk Location 16.0 mi@ 190° S (Summerville)

Food Product 48 Produce Location #6 *

1.9 mi @ 143° SE (Bergenstock) (NA)

Food Product 49 Produce Location #1 *

1.6 mi@84°E (Culeton) (NA)

Food Product 50 Produce Location #2 *

l.9mi@l0l°E (Vitullo) (NA)

Food Product 51 Produce Location #5 *

1.5 mi @ 116° ESE (C.S. Parkhurst) (NA)

Food Product 52 Produce Location #3 *

1.5 mi@ 84°E (C. Narewski) (NA)

Map See Figures 5.1-1 and 5.1-2

- Food Product Samples need not necessarily be collected from all listed locations. Collected samples will be of the highest calculated site average D/Q.

(NA) Not applicable CR Control Result (location)

CY-NM-170-301 Revision 37 II 70 January 2020

TABLE 5.1 (Cont'd)

NINE MILE POINT NUCLEAR ST ATION RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM SAMPLING LOCATIONS

Map Collection Site Type of Sample Location {Env. Program No,) Location Food Product 53 Produce Location #4 *

1.7 mi @ 126° SE (P. Parkhurst) (NA)

Food Product (CR) 54 Produce Location #7*

15.1 mi @222° SW (Mc Millen) (NA)

Food Product (CR) 55 Produce Location #8 *

12.5 mi @227° SW (Denman) (NA)

Food Product 56 Produce Location #9*

1.6 mi@ 171° S (O'Connor) (NA)

Food Product 57 Produce Location #IO** 2.3 mi@ 124° SE (C. Lawton) (NA)

Food Product 58 Produce Location # 11 *

2.0 mi@ 112° ESE (C. R. Parkhurst) (NA)

Food Product 59 Produce Location #12*

2.0 mi @ 110° ESE (Barton) (NA)

Food Product (CR) 60 Produce Location # 13 *

15.4 mi @222° SW (Flack) (NA)

Food Product 61 Produce Location #14** 1.9 mi @97° E (Koeneke) (NA)

Food Product 62 Produce Location #15*

1.6 mi@ 139° SE (Whaley) (NA)

Food Product 63 Produce Location # 16 *

1.2 mi @ 209° SSW (Murray) (NA)

Food Product 67 Produce Location #17*

1.7 mi @98° E (Battles)

Food Product 68 Produce Location # 18 *

1.5 mi @84° E (Kronenbitter)

Food Product 69 Produce Location #119*

1.4 mi@ 132° SE (O'Connor)

Map See Figures 5.1-1 and 5.1-2

- Food Product Samples need not necessarily be collected from all listed locations. Collected samples will be of the highest calculated site average D/Q.

(NA) Not applicable CR = Control Result (location)

CY-NM-170-301 Revision 37 II 71 January 2020

Site Map 78' Lake Ontario-Hile Radius

-..J N

8W{L)

FIGURE 5.1-1 NINE MILE POINT ON-SITE MAP

.1 DENVIRONMENTAL SAMPLE I I LOCATIONS 1 0

Scale (Miles)

FIGURE 5.1-2 NINE MILE POINT N

OFF-SITE MAP D. ENVIRONMENTAL SAMPLE LOCATIONS Scale(Mlles)

! l 4 Lake Ontario

-..J w

NMPU1 Rx

--.J w

i:,:i ii!

i 7

n

._ --< FIGURE 5.1-2a N \

~

llJ ~ ~*

C:

Q :!:.

(l)

. .--.J'. .

NINE MILE POINT OFF-SITE MAP L-- NTY NOO o/::i L,.J 6,, ENVlRONMeITAL SAMPLE LOCATIONS \ SCALE (MILES)

Nwo 0--.J-

~ L--*3/i!.tl t-:.:.:*{*;:1 j

FIGURE 5. 1.3-1 Lake Ontario Site r11*u*1 Boundaries 1.tiuld ~ *

~

. -- L,Jacning Nate: NaGonll Grtd retains OM'IINhlp In certain 11'8n1rrt811an line and llfilchyard facilflee wlll'lln th& exc!Ulion . , . boundary.

.1 Aooe8B and 1.B198 ere oonlrallad ~NM,& Point Nuclaer I 1- -4 I

I I ,~I I I 0 1 statlCI\ LLC by lfill'b&iB d:.

Scale(t.lles)

APPENDIX A LIQUID DOSE FACTOR DERIVATION CY-NM-170-301 Revision 37 II 75 January 2020

Appendix A Liquid Effluent Dose Factor Derivation, Aw Arnt (mrem/hr per µCi/ml) which embodies the dose conversion factors, pathway transfer factors (e.g.,

bioaccumulation factors), pathway usage factors, and dilution factors for the points of pathway origin talces into account the dose from ingestion of fish and drinking water and the sediment. The total body and organ dose conversion factors for each radionuclide will be used from Table E-11 of Regulatory Guide 1.109. To expedite time, the dose is calculated for a maximum individual instead of each age group. The maximum individual dose factor is a composite of the highest dose factor A.at of each nuclide i age group a, and organ t, hence A1at. It should be noted that the fish ingestion pathway is the most significant pathway for dose from liquid effluents. The water consumption pathway is included for consistency, with NUREG 0133.

The equation for calculating dose contributions given in section 1.3 requires the use of the composite dose factor Art for each nuclide, i. The dose factor equation for a fresh water site is:

Where:

A.at = Is the dose factor for nuclide i, age group a, total body or organ t, for all appropriate pathways, (mrem/hr per µCi/ml).

Ko = Is the unit conversion factor, l.14ES=lE6pCi/µCi x 1E3 ml/kg-:- 8760 hr/yr.

Uw = Water consumption (I/yr); from Table E-5 of Reg. Guide 1.109.

Ur = Fish consumption (Kg/yr); from Table E-5 of Reg. Guide 1.109.

u. = Sediment Shoreline Usage (hr/yr); from Table E-5 of Reg. Guide 1.109.

(BF), = Bioaccumulation factor for nuclide, i, in fish, (pCi/kg per pCi/1 ),

from Table A-1 of Reg. Guide 1.109.

(DFL)m = Dose conversion factor for age, nuclide, i, group a, total body or organ t, (mrem/pCi); from Table E-11 of Reg. Guide 1.109.

(DPS), = Dose conversion factor for nuclide i and total body, from standing on contaminated ground (mem/hr per pCi/m2); from Table E-6 of Reg. Guide 1.109.

Dilution factor from the near field area within one-quarter mile of the release point to the potable water intalce for the adult water consumption. This is the Metropolitan Water Board, Onondaga County intalce structure located west of the City of Oswego;

( unitless ).

CY-NM-170-301 Revision 37 II 76 January 2020

Appendix A (Cont'd)

D. Dilution factor from the near field area within one quarter mile of the release point to the shoreline deposit (taken at the same point where we take environmental samples 1.5 miles; unitless).

69.3 conversion factor .693 x 100, 100 = Kc (L/kg-hr) *40*24 hr/day/.693 in Um 2-d, and Kc = transfer coefficient from water to sediment in Ukg per hour.

Average transit time required for each nuclide to reach the point of exposure for internal dose, it is the total time elapsed from release of the nuclides to either ingestion for water (w) and fish (f) or shoreline deposit (s), (hr).

Length of time the sediment is exposed to the contaminated water, nominally 15 yrs (approximate midpoint of facility operating life),

(hrs).

decay constant for nuclide i (hr" 1).

w Shore width factor (unitless) from Table A-2 of Reg. Guide 1.109.

Example Calculation For 1-131 Thyroid Dose Factor for an Adult from a Radwaste liquid effluents release:

(DFS), = 2.80E-9 mrem/hr per pCi/m2 (DFL)w = l.95E-3 mrem/pCi tpw = 30 hrs. (w = water)

BF, = 15 pCi/Kg per pCi/L tpr = 24 hrs. (f = fish)

Ur = 21 Kg/yr ti, = l.314E5 hrs. (5.48E3 days)

Dw = 40 unitless U., = 730 Uyr D, = 12 unitless Ko = 1.14E5 (~ilb!Ci){ml/kg)

u. = 12 hr/yr (hr/yr) w =0.3 Ai = 3.61E-3hr" 1 tps = 5.5 hrs (s = Shoreline Sediment)

These values will yield an Aa1 Factor of 6.79E4 mrem-ml per µCi-hr as listed in Table 2-4. It should be noted that only a limited number ofnuclides are listed on Tables 2-1 to 2-8. These are the most common nuclides encountered in effluents. If a nuclide is detected for which a factor is not listed, then it will be calculated and included in a revision to the ODCM.

In addition, not all dose factors are used for the dose calculations. A maximum individual is used, which is a composite of the maximum dose factor of each age group for each organ as reflected in the applicable chemistry procedures.

CY-NM-170-301 Revision 37 II 77 January 2020

APPENDIXB PLUME SHINE DOSE FACTOR DERIVATION CY-NM-170-301 Revision 37 II 78 January 2020

APPENDIXB For elevated releases the plume shine dose factors for gamma air (Bi) and whole body 0/1), are calculated using the finite plume model with an elevation above ground equal to the stack height To calculate the plume shine factor for gamma whole body doses, the gamma air dose factor is adjusted for the attenuation of tissue, and the ratio of mass absorption coefficients between tissue and air. The equations are as follows:

Gamma Air B, = ~ Ku.,E I, Where: Kl conversion factor (see Rev. below for actual value).

µ.. mass absorption coefficient (cm2/g; air for Bi, tissue for V1)

E = Energy of gamma ray per disintegration (Mev)

v. = average wind speed for each stability class (s), mis R = downwind distance (site boundary, m) 0 = sector width (radians) s = subscript for stability class I. = I function = L + kh for each stability class.

( unitless, see Regulatory Guide 1.109) k2 = Fraction of the attenuated energy that is actually absorbed in air (see Regulatory Guide 1.109, see below for equation)

Whole Body I

Where: ta tissue depth (g/cm2) shielding factor from structures (unitless) 1.11 = Ratio of mass absorption coefficients between tissue and air.

Where all other parameters are defined above.

1 K = conversion factor [3.7 El0 ~ ] 1.6 E-6 m;]

Cksec Mev= 0.46

[ 1293 ~ [ 100 m:g J g-rad Where: µ mass attenuation coefficient (cm2/g; air for B1, tissue for V1)

µ.. defined above CY-NM-170-301 Revision 37 II 79 January 2020

APPENDIX B (Cont'd)

There are seven stability classes, A thru F. The percentage of the year that each stability class occurs is taken from the U-2 FSAR. From this data, a plume shine dose factor is calculated for each stability class and each nuclide, multiplied by its respective :fraction and then summed.

The wind speeds corresponding to each stability class are, also, taken from the U-2 FSAR To confirm the accuracy of these values, an average of the 12 month wind speeds for 1985, 1986, 1987 and 1988 was compared to the average of the FSAR values. The average wind speed of the actual data is equal to 6.78 mis, which compared favorably to the FSAR average wind speed equal to 6.77 mis.

The average gamma energies were calculated using a weighted average of all gamma energies emitted from the nuclide. These energies were taken from the handbook "Radioactive Decay Data Tables", David C. Kocher.

The mass absorption (µ..) and attenuation (µ) coefficients were calculated by multiplying the mass absorption (J..l.a/µ) and mass attenuation (µ/p) coefficients given in the Radiation Health Handbook by the air density equal to 1.293 E-3 glee or the tissue density of 1 glee where applicable. The tissue depth is 5glcm2 for the whole body. '

The downwind distance is the site boundary.

CY-NM-170-301 Revision 37 II 80 January 2020

APPENDIX B (Cont'd)

SAMPLE CALCULATION Ex. Kr-89 F STABILilY CLASS ONLY - Gamma Alr

-DATA E = 2.22MeV k = !!:J!a = .871 K = 0.46

µa = 2.943 E-3m-1 µa VF = 5.55 m/sec

µ = 5.5064E-3m-1 R = 644m 0 = 0.39 crz = 19m ....... vertlcal plume spread taken from "Introduction to Nuclear Engineering", John R. LaMarsh

-I Function Ucrz = 0.06 Ii = 0.33 h = 0.45 I = Ii + kh = 0.33 + (0.871) (0.45) = 0.72 B,

~

= o.46lf_1-sec)

J CMev/eras C2,943E-3m- 1)(2.22Mev)C,72)

(7r1/2) (g/m 3) ~ (5.55 m/s) (.39) (644m)

(g-rad)

= 1.55(-6) rad/s (3600 s/hr) (24 hid) (365 d/y) (1E3mrad/rad) 0/s C1E6u0) a

= 2. 76(-2) mrad/yr

µ0/sec

-(.0253 an 2/g)(Sg/an 2)

= 1.11(.7) [2.76(-2) mrad/yr] [e ]

µ0/sec

= 1.89(-2) mrad/yr

µ0/sec NOTE: The above calculation is for the F stability class only. For Table 3-2 and procedure values, a c, weighted fraction of each stability class was used to determine the Bi and V1 values.

CY-NM-170-301 Revision 37 II 81 January 2020

APPENJ)IXC ORGAN DOSE PARAMETERS FOR IODINE 131 and 133, PARTICULATES AND TRITIUM

\

CY-NM-170-301 Revision 37 II 82 January 2020

APPENDIXC ORGAN OOSE PARAMETERS FOR IODINE-131 AND- 133, PARTICULATES AND TRITIUM This appendix contains the methodology which was used to calculate the organ dose factors for 1-131, 1-133, particulates, and tritium. The dose factor, R, was calculated using the methodology outlined in NUREG-0133. The radioiodine and particulate ODCM Part I (Control DLCO 3.6.15) is applicable to the location in the unrestricted area where the combination of existing pathways and receptor age groups indicates the maximum potential exposure occurs, i.e., the critical receptor. Washout was calculated and determined to be negligible. R values have been calculated for the adult, teen, child and infant age groups for all pathways. However, for dose compliance calculations, a maximum individual is assumed that is a composite of highest dose factor of each age group for each organ and pathway. The methodology used to calculate these values follows:

C.1 Inhalation Pathway R(I) K'(BR)a(DF A)ua where:

R(I) dose factor for each identified radionuclide i of the organ of interest (units= mrem/yr per µCi/m 3);

K' a constant of unit conversion, 1E6 pCi/µCi (BR)a Breathing rate of the receptor of age group a, (units= m3/yr);

The inhalation dose factor for nuclide i, organj and age group a, and organ t (units= mrem/pCi).

The breathing rates (BR)a for the various age groups, as given in Table E-5 of Regulatory Guide 1.109 Revision 1, are tabulated below.

Age Group (a) Breathing Rate (m3/yr)

Infant 1400 Child 3700 Teen 8000 Adult 8000 Inhalation dose factors (DF A)ua for the various age groups are given in Tables E-7 through E-10 of Regulatory Guide 1.109 Revision 1.

CY-NM-170-301 Revision 37 II 83 January 2020

APPENDIX C (Cont'd)

C.2 Ground Plane Pathway Where:

R.(G) = Dose factor for the ground plane pathway for each identified radionuclide i for the organ of interest (units= m2-mrern/yr per µCi/sec)

K' = A constant of unit conversion, 1E6 pCi/µCi K" = A constant of unit conversion, 8760 hr/year

~ = The radiological decay constant for radionuclide i, (units= sec- 1) t = The exposure time, sec, 4.73E8 sec (15 years)

(DFG), = The ground plane dose conversion factor for radionuclide i; (units= mrem/hr per pCi/m2)

SF The shielding factor (dimensionless)

A shielding factor of0.7 is discussed in Table E-15 of Regulatory Guide 1.109 Revision 1. A tabulation ofDFG1 values is presented in Table E-6 of Regulatory Guide 1.109 Revision 1.

CY-NM-170-301 Revision 37 II 84 January 2020

APPENDIX C (Cont'd)

C.3 Grass-(Cow or Goat}-Milk Pathway Where:

R,(C) Dose factor for the cow milk or goat milk pathway, for each identified radionuclide i for the organ of interest, (units= m2 -mrern/yr per µCi/sec)

K' A constant of unit conversion, 1E6 pCi/µCi Qr The cow's or goat's feed consumption rate, (units= kg/day-wet weight)

Uap The receptor's milk consumption rate for age group a, (units= liters/yr)

The agricultural productivity by unit area of pasture feed grass, (units= kg/m2)

The agricultural productivity by unit area of stored feed, ( units = kg/m2)

The stable element transfer coefficients, (units = pCi/liter per pCi/day) r Fraction of deposited activity retained on cow's feed grass (DFL)iat The ingestion dose factor for nuclide i, age group a, and total body or organ t (units= mrem/pCi)

The radiological decay constant for radionuclide i, (units=sec -1)

The decay constant for removal of activity on leaf and plant surfaces by weathering equal to 5.73E-7 sec -1 (corresponding to a 14 day half-life) tt = The transport time from pasture to cow or goat, to milk, to receptor, (units= sec)

The transport time from pasture, to harvest, to cow or goat, to milk, to receptor (units= sec)

Fraction of the year that the cow or goat is on pasture (dimensionless)

f. Fraction of the cow feed that is pasture grass while the cow is on pasture (dimensionless)

CY-NM-170-301 Revision 37 II 85 January 2020

APPENDIX C (Cont'd)

Milk cattle and goats are considered to be fed from two potential sources, pasture grass and stored feeds.

Following the- development in Regulatory Guide 1.109 Revision 1, the value off. is considered unity in lieu of site specific information. The value offp is 0.5 based on 6 month grazing period. This value for fp was obtained from the environmental group.

Table C-1 contains the appropriate values and their source in Regulatory Guide 1.109 Revision 1.

The concentration of tritium in milk is based on the airborne concentration rather than the deposition.

Therefore, the Rr(C) is based on X/Q:

Where:

Rr(C) Dose factor for the cow or goat milk pathway for tritium for the organ of interest, (units= mrem/yr per µCi/m 3)

K"' = A constant of unit conversion, 1E3 g/kg H Absolute humidity of the atmosphere, (units= g/m3) 0.75 The fraction of total feed that is water 0.5 The ratio of the specific activity of the feed grass water to the atmospheric water Other values are given previously. A site specific value of H equal to 6.14 g/m3. is used. This value was obtained from the environmental group using actual site data.

CY-NM-170-301 Revision 37 II 86 January 2020

APPENDIX C (Cont'd)

C.4 Grass-Cow-Meat Pathway R,(M) Dose factor for the meat ingestion pathway for radionuclide i for any organ of interest, (units= m2 -mrem/yr per µCi/sec)

Fr The stable element transfer coefficients, (units= pCi/kg per pCi/day)

The receptor's meat consumption rate for age group a, (units= kg/year)

The transport time from harvest, to cow, to receptor, (units= sec) tr The transport time from pasture, to cow, to receptor, (units= sec)

All other terms remain the same as defined for the milk pathway. Table C-2 contains the values which were used in calculating R,(M).

The concentration of tritium in meat is based on airborne concentration rather than deposition. Therefore, the Rr(M) is based on X/Q. \

Where:

Rr(M) = Dose factor for the meat ingestion pathway for tritium for any organ of interest, (units=

mrem/yr per µCi/m 3)

All other terms are defined above.

CY-NM-170-301 Revision 37 II 87 January 2020

APPENDIX C (Cont'd)

C.5 Vegetation Pathway The integrated concentration in vegetation consumed by man follows the expression developed for milk. Man is considered to consume two types of vegetation (fresh and stored) that differ only in the time period between harvest and consumption, therefore:

Where:

Dose factor for vegetable pathway for radionuclide i for the organ of interest, (units= m 2-mrem/yr per µCi/sec)

The consumption rate of fresh leafy vegetation by the receptor in age group a, (units= kg/yr)

The consumption rate of stored vegetation by the receptor in age group a (units= kg/yr)

The fraction of the annual intake of fresh leafy vegetation grown locally The fraction of the annual intake of stored vegetation grown locally ti, The average time between harvest ofleafy vegetation and its consumption, (units

= sec)

= The average time between harvest of stored vegetation and its consumption, (units = sec)

The vegetation areal P density, (units= kg/m 2)

All other factors have been defined previously.

Table C-3 presents the appropriate parameter values and their source in Regulatory Guide 1.109 Revision I.

In lieu of site-specific data, values for FL and F8 of, 1.0 and 0. 76, respectively, were used in the calculation. These values were obtained from Table E-15 of Regulatory Guide 1.109 Revision I.

The concentration of tritium in vegetation is based on the airborne concentration rather than the deposition. Therefore, the Rr(V) is based on X/Q:

Where:

Rr(V) dose factor for the vegetable pathway for tritium for any organ of interest, (units= mrem/yr per µCi/m 3).

All other terms are defined in preceeding sections.

CY-NM-170-301 Revision 37 II 88 January 2020

TABLEC-1 Parameters for Grass-(Cow or Goat)-Milk Pathways Reference Parameter Value (Reg. Guide 1.109 Rev. 1)

Qr (kg/day) 50 (cow) Table E-3 6 (goat) Table E-3 r 1.0 (radioiodines) Table E-15 0.2 (particulates) Table E-15

{DFL),Ja (mrem/pCi) Each radionuclide Tables E-11 to E-14 Fm (pCi/liter per pCVday) Each stable element Table E-1 (cow)

Table E-2 (goat) 2.0 Table E-15 0.7 Table E-15 tii (seconds) 7.78 x 106 (90 days) Table E-15 tr(seconds) 1.73 x 105 (2 days) Table E-15 Uap (liters/yr) 330 infant Table E-5 330 child Table E-5 400 teen Table E-5 310 adult Table E-5

\

CY-NM-170-301 Revision 37 II 89 January 2020

TABLEC-2 Parameters for the Grass-Cow-Meat Pathway Reference Parameter Value (Reg. Guide 1.109 Rev. 1) r 1.0 (radioiodines) Table E-15 0.2 (particulates) Table E-15 Fr(pCi!Kg per pCi/day) Each stable element Table E-1 Uap (Kg/yr) 0 infant Table E-5 41 child Table E-5 65 teen Table E-5 110 adult Table E-5 (DFLh]a (mrem/pCi) Each radionuclide Tables E-11 to E-14 r ,

0.7 Table E-15 2.0 Table E-15 tii (seconds) 7.78E6 (90 days) Table E-15 tr (seconds) 1.73E6 (20 days) Table E-15 Qr(kg/day) 50 Table E-3 CY-NM-170-301 Revision 37 II 90 January 2020

TABLEC-3 Parameters for the Vegetable Pathway Reference Parameter Value (Reg. Guide 1.109 Rev. l}

r (dimensionless) 1.0 (radioiodines) Table E-1 0.2 (particulates) Table E-1 (DFL),Ja (mrem/pCi) Each radionuclide Tables E-11 to E-14 UL)a (kg/yr) - infant 0 Table E-5

- child 26 Table E-5

- teen 42 Table E-5

- adult 64 Table E-5 U5)a (kg/yr) - infant 0 Table E-5

- child 520 Table E-5

-teen 630 Table E-5

- adult 520 .Table E-5 tL(seconds) 8.6E4 (1 day) Table E-15 ti, (seconds) 5.18E6 (60 days) Table E-15 2.0 Table E-15 CY-NM-170-301 Revision 37 II 91 January 2020

APPENDIXD DIAGRAMS OF RADIOACTIVE LIQUID AND GASEOUS EFFLUENT TREATMENT SYSTEMS AND MONITORING SYSTEMS CY-NM-170-301 Revision 37 II 92 January 2020

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This Page and Figure Deleted.

CY-NM-170-301 Revision 37 D-7 January 2020

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CY-NM-170-301 Revision 37 D-9 January 2020

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CY-NM-170-301 Revision 37 D-10 January 2020

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ATTACHMENT 14 Page 1 of 1 Unit 1 X Unit 2 Reporting Period; January - December 2020 Process Control Program (PCP)

There were no changes to the Process Control Program in 2020.

Enclosure 2 Nine Mile p-oint Nuclear Station, Unit 2 Radioactive Effluent Release Report, January - December 2020

NINE MILE POINT NUCLEAR STATION - UNIT 2 RADIOACTIVE EFFLUENT RELEASE REPORT January- December 2020

/

'NINE MILE POINT NUCLEAR STATION - UNIT 2 RADIOACTIVE EFFLUENT RELEASE REPORT JANUARY - DECEMBER 2020 SUPPLEMENTAL INFORMATION Facility: Nine Mlle Point Unit 2 Licensee: Nine Mile Point Nuclear Station, LLC

1. TECHNICAL SPECIFICATION/ODCM LIMITS A) FISSION AND ACTIVATION GASES 1 The dose rate limrt of noble gases released in gaseous effluents from the srte to areas at or beyond the srte boundary shall be less than or equal to 500 mrem/year to the whole body and less than or equal to 3000 mrem/year to the skin.

2 The air dose from noble gases released in gaseous effluents from Nine Mile Point Unrt 2 to areas at or beyond the site boundary shall be limited during any calendar quarter to less than or equal to 5 mrad for gamma radiation and less than or equal to 10 mrad for beta radiation, and during any calendar year to less than or equal to 10 mrad for gamma radiation and less than or equal to 20 mrad for beta radiation B&C) TRITIUM, IODINES AND PARTICULATES, HALF LIVES> 8 DAYS 1 The dose rate limit of lodme-131, lodme-133, TntIum and all radionuclides in particulate form wrth half-lives greater than eight days, released in gaseous effluents from the site to areas at or beyond the site boundary shall be less than or equal to 1500 mrem/year to any organ 2 The dose to a member of the public from lodine-131, lodine-133, Tritium and .all radionuclides m particulate form with half-lives greater than eight days in gaseous effluents released from Nine Mile Point Unit 2 to areas at or beyond the site boundary shall be limited dunng any calendar quarter to less than or equal to 7 5 mrem to any organ, and during any calendar year to less than or equal to 15 mrem to any organ D) LIQUID EFFLUENTS 1 Improved Technical Specifications (ITS) limit the concentration of radioactive matenal released in the l1qu1d effluents to unrestricted areas to ten times the concentrations specIf1ed in 10CFR20 1001-20 2402, Appendrx B, Table 2, Column 2 for radionuclides other than dissolved or entrained noble gases For dissolved or entrained noble gases, the concentration shall be lImrted to 2E-04 m1crocuries/ml total activity

2. The dose or dose commitment to a member of the public from radioactive matenals in liquid effluents released from Nine Mile Point Unit 2 to unrestricted areas shall be limited during any calendar quarter to less than or equal to 1.5 mrem to the whole body and to less than or equal to 5 mrem to any organ, and during any calendar year to less than or equal to 3 mrem to the whole body and to less than or equa(to 10 mrem to any organ

2. MEASUREMENTS AND APPROXIMATIONS OF TOTAL RADIOACTIVITY Described below are the methods used to measure or approximate the total radioactivity and rad1onuclIde composItIon in effluents A) FISSION AND ACTIVATION GASES Noble gas effluent activity Is detenrnned by an on-line scmt1llat1on detector (calibrated against gamma isotopic analysis of a 4 0L Mannelli grab sample) of an 1sokinetic sample stream.

B) IODINES Iodine effluent actIvIty is determined by gamma spectroscopic analysis (at least weekly) of charcoal cartridges sampled from an IsokinetIc sample stream C) PARTICULA TES Activrty released from the main stack and the combined Radwaste/Reactor Building vent is determined by gamma spectroscopic analysis (at least weekly) of particulate niters sampled from an 1sokinet1c sample stream and composite analysis of the filters for non-gamma emitters D) TRITIUM Tnt,um effluent act,vrty Is measured by liquid scintillation or gas proportional counting of monthly samples taken wrth an air sparging/water trap apparatus. '

E) LIQUID EFFLUENTS Isotopic contents of l1quId effluents are determined by IsotopIc analysis of a representative sample of each batch and composite analysis of non-gamma emrtters F) SOLID EFFLUENTS Isotopic contents of waste shipments are determined by gamma spectroscop y analyses of a representative sample of each batch Scaling factors established from primary composite sample analyses conducted off-site are applied, where appropriate, to find estimated concentration of non-gamma emrtters. For low activity trash shipments, curie content ,s estimated by dose rate measurement and appl1cat1on of appropriate scaling factors G) C-14 The production of C-14 and the effluent dose consequences are estimates based on EPRI methodology provided in EPRI Report 1021106, Estimation of Carbon-14 in Nuclear Power Plant Gaseous Effluents, December 2010 and NUREG-0016, Calculation of Releases of Radioactrve Materials in Gaseous and l.Jquld Effluents for Boiling Water Reactors (BWR-GALE Code)

3. METEOROLOGICAL DATA Meteorological data Is an annual summary of hourly meteorological data collected over the previous year This annual summary may be either in the form of an hour-by-hour listing on magnetic tape of wind speed, wind direction, atmospheric stab1lrty, and precIprtatIon (if measured), or in the form of joint frequency distribution of wind speed, wind direction, and atmosphenc stability In lieu of submIssIon wrth the Rad1olog1cal Effluent Release Report, the licensee Is exercising the option of retaining this summary of required meteorological data on srte m a file that shall be provided to the NRG upon request

Supplemental Information ATTACHMENT 1 Page 1 of 2

SUMMARY

DATA Unit 1 Unit 2 X R!!llQDlllQ Perl2g; .!l!Dl.llln'. - Q:i!,!!!I!b!ir 2020 Liquid Effluents:

ODCM Required Maximum Effluent Concentrallon (MEG) =10 x 10CFR20 1001 - 20 2402, AppendlX B, Table 2, Column 2 Average MEC - µCt/ml (Qtr 1) = I NO RELEASES Average MEC - µCr/ml (Qtr ~ = I NO RELEASES I Average MEC - µCr/ml (Qtr _g) = I NO RELEASES Average MEG - µCt/ml (Qtr 1) = I NO RELEASES I Average Energy (Fission and Activation ga88S - MEV):

Qrtr 1 ~y = N/A E~ = NIA Qrtr £ ~y = N/A . E~ = NIA Qrtr J ~y = N/A E~ = NIA Qrtr 5- Ey = NIA E~ = NIA Liquid:

Number of Batch Releases D Total Trme Perrod for Balch Releases (hrs) DD Maximum Trme Penod for a Batch Release (hrs) DD Averaoe Trme Penod for a Batch Release (hrs) DD Mrnrmum Trme Penod for a Batch Release DO Total volume of water used to drlute the lrqurd 1st 2nd 3rd 4th dunng the release penod (L) N/A NIA I NIA I NIA I Total volume of water available to dilute the hqurd 1st £Q.Q_ 3rd 4th effluent dunng the report penod (L) 1 12E+10 1 16E+10 I 1.24E+10 I 118E+10 I Gaseous (Emergency Condenser Vent) "Not applicable for Unrt 2" Number of Batch Releases N/A Total Trme Penod for Balch Releases (hrs) NIA Maximum Trme Penod for a Batch Release (hrs) NIA Average Time Penod for a Batch Release (hrs) NIA Mrnrmum Trme Penod for a Batch Release NIA Gaseous (Primary Containment Purge)

Number of Batch Releases 8 Total Trme Penod for Batch Releases (hrs) 170 6 MaxJmum Trrne Penod for a Batch Release (hrs) 944 Average Trme Penod for a Batch Release (hrs) 21 3 Mrnrmum Trme Penod for a Batch Release (hrs) DD I

Supplemental Information ATTACHMENT 1 Page 1 of 2

SUMMARY

DATA Unit 1 Unlt2 X Reoortlng eicl2sl; i:l!i!D!!IIO! - Decem~ ~~

Liquid Effluent5:

ODCM ReqUJred Maximum Effluent Concentration (MEG)= 10 x 10CFR20 1001 - 20 2402, Appendix B, Table 2. Column 2 Average MEG - µCJ/ml (Qtr. 1)

I NO RELEASES Average MEG - µCl/ml (Qtr ~ = I NO RELEASES I Average MEG - µCl/ml (Qtr g) = I NO RELEASES Average MEG - µCl/ml (Qtr £) = I NO RELEASES I Average Energy (Fission and Activation gases - MEV):

Qrtr 1 Ey = N/A E~ = N/A Qrtr i Ey = N/A E~ = NIA Qrtr ;}_ Ev = N/A E~ = NIA Qrtr. i Ev = NIA E~ = NIA Liquid:

NLrnber of Batch Releases 0 Total Time Pencxi for Batch Releases (hrs) 00 '

MaxJmum Time Pencxi for a Batch Release (hrs) 00 Average Time Penod for a Batch Release (hrs) 00 Minimum Time Pencxi for a Batch Release 00 Total volume of water used to dilute the hqutd 1st 2nd 3rd 4th dunng the release pencxi (L) NIA NIA I NIA I N/A I Total volume of water available to dilute the hql.Dd 1st 2nd 3rd 4th effluent dunng the report penod (L) 1 12E+10 1 16E+10 I 1 24E+10 I 1 18E+10 I Gaseous (Emergency Condenser Vent) "Not applicable for Unit 2

Number of Batch Releases N/A Total Time Penod for Batch Releases (hrs) NIA Maximum Tlme Penod for a Batch Release (hrs) NIA Average Time Penod for a Batch Release (hrs) N/A Minimum Tune Penod for a Batch Release NIA Gaseous (Primary Containment Purge)

Number of Batch Releases 8 Total Time Penod for Batch Releases (tvs) 170 6 MaxJmum Time Penod for a Batch Release (hrs) 944 Average Time Penod for a Batch Release (hrs) 21 3 Minimum Time Penod for a Batch Release (hrs) 0.0

Supplemental Information ATTACHMENT 1 Page 2 of 2

SUMMARY

DATA Unlt1 Unit 2 X B!!11:21l1ae E!![f~: J!!!lYilll'. - Decem!li[ 22~!!

Abnonnal Releases:

A. Liquids:

I Number of Releases 0 le, Total Acbvrty Released I NIA _J B. Gaseous:

I Number of Releases I 0 I ITotaJ ActMlv Released I N/A* lei r

Table 1A Gaseous Effluents - ATTACHMENT 2 Page 1 of 1 Summation of All Releases -

Elevated and Ground Level Unit 1 Unit 2 X R!i!oortlDQ ei!:!Qs!: Ji!Dln!D'.

Decem!!:!r .!!~

GASEOUS EFFLUENTS - SUMMATION OF ALL RELEASES, ELEVATED AND GROUND LEVEL 111 ans! Im ~ Est Total

~r ~r ~ ~ &1I2c, r.

A E~s!Q[] ~ Acb~t!on ~a~

1 Total Release CI 1 88E+01 0 OOE+OO 0 OOE+OO 0 00E+OO S 0OE+-01 2 Average Release Rate µCi/sec 2 39E+00 0 OOE+OO 0 OOE+OO 0 OOE+OO B~

1 Total Iodine -131 CI 1.14E-04 2 28E-06 0 OOE+00 0 OOE+OO 3 OOE+01

2. Average Release Rate for Period µCi/sec 1 4SE-OS 2 90E-07 0.OOE+OO 0 OOE+OO c Particulates 1 Parllculates v.,th Half-hves>8days Cl 4 52E-04 346E-04 2.04E--03 3 46E-04 3 0OE+-01

2. Average Release Rate for Period µCi/sec S 74E-OS 4 40E--OS 2 56E--04 4 36E--05

3. Gross Alpha Radioactivity CJ 0 OOE+O0 0 OOE+OO 0 00E+OO 0 00E+OO 2 50E+01 D ..Iolllim 1 Total Release CJ 3 02E+01 1 61E+01 2 50E+01 1 68E+01 5 OOE+01 2 Average Release Rate for Penod µC~sec 3 84E+O0 2 05E+OO 314E+OO 211E+OO E. P!)r~ l Qf Te1.b ~~ l.!rn1ts Eiss1Q[] and ActIVabon Gases Percent of Quarterly Gamma AJr Dose l..Jmlt

% 111E--02 0 OOE+00 0 OOE+OO 0 OOE+OO (5mR)

Percent of Quarterty Beta Air Dose Limit (10

% 5 1SE-04 0 00E+00 0 00E+OO 0 OOE+OO mrad)

Percent of Annual Gamma Air Dose l..Jmrt to

% 5.5SE--03 5.55E--03 5 55E--03 S.5SE--03 Date (10 mR)

Percent of Annual Beta Arr Dose Limit to Date (20 mrad)

% 2 S?E--04 2 S7E-04 2 57E-04 2 S?E--04 Percent of Whole Body Dose Rate Llmrt (500

% 4 27E-04 0 OOE+OO 0.OOE+00 0 OOE+00 mrem/yr)

Percent of 8km Dose Rate Limit (3000

% 8 69E--05 0OOE+OO O.OOE+00 0 OOE+OO mremlyr)

ITntium lnrlmes and Parbctil"tes <wrth half-llv<><, arPJ'lter than 8 rliivs\

Percent of Quarterly Dose l..Jmlt (7 5 mrem) % 4 39E--02 1.13E--02 6 09E--02 1.15E--02 Percent of Annual Dose Limit to Date (15

% 4 39E-02 5.49E-02 1 10E-01 1 21 E--01 mrem)

Percent of Organ Dose Limit (1500 mrem/yr % 7 40E--04 8 68E--OS 9 80E--OS 5 61 E--05

Table 1B Gaseous ATTACHMENT 3 Page 1 of 2 Effluents - Elevated Releases Unit 1 Unit2 X Reoortl!!9 P!!rlod: J!!nlli!!Y - December 2Q20 GASEOUS EFFLUENTS - ELEVATED RELEASE Continuous Mode (2)

Nuclldes Released 1§l91!!!rn!! ,ml Q!.li!tl!ll: 3!]! QUi:!!EC ~ttJ QU!![mr Etul9a Gases fll Argon-41 CI ..

Kryptoo-85 Cl a ....

Krypton-85m 6 03E-01 Krypton-87 Cl 5 97E-01 Krypton-88 C1 1 37E+OO Xenon-127 Xenon-131m C1 Cl .... I Xenon-133 CI 1 03E+01 Xenon-133m Cl Xenon-135 CI 5 52E+OO Xenon-135m Cl 4 18E-01 Xenon-137 Xenon-138 Cl CI ..

Iodines 111 lodine-131 CI 8 18E-05 2 28E-06 lodine-133 lodine-135 CI Cl I ..

528E-05 I I I .. I Partl!,!,!1~ (1l Chromium-51 a .... .... .... ....

a Manganese-54 lron-55 lron-59 Cl a ....

1 10E-07 1 12E-05 Cobalt-58 Cl Cobalt-60 Cl 8 40E-05 1.32E-04 1 49E-04 1.01E-04 Neodymium-147 CI Zinc-65 C1 Stronllum-89 CI 7 04E-08 1 12E-05 Stronllum-90 Nioblum-95 C1 Cl .... ......

3 47E-08 1 36E-06 Zircornum-95 C1 Molybdenum-99 Cl Ruthenmm-103 C1 Cesium-134 CI a

Cesium-136 Cesn.rn-137 Banum-140 a

C1 Lanthanum-140 Cl a

Cenum-141 Cenurn-144 a ..

Ii:l!!!.!m(:11 CI I 2 13E+01 I 1 08E+01 I 1 88E+01 I 1.27E+01 I (1) Concentratmns less than the lower hmrt of delec!Jon of the coun!Jng system used are Ind1caled with a double astensk A lower hmII of detecllon of 1.00E-04 µCl/ml for reqU1red noble gases, 1 OOE-11 µCi/ml for required particulates and gross alpha, 1 OOE-12 µCi/ml for required Iodines, 1 OOE-11

µCi/ml for Sr-89/90 and 1 OOE-06 µCl/ml for Trrt1um, as reqwed by the ODCM, has been venf1ed.

(2) Contnbu!Jons from purges are mciuded There were no other batch releases dunng the reporting penod

ATTACHMENT 3 Page 2 of 2 Unit 1 Unit 2 X Bi122ctlag e2[IQJ;!: Ja1rnarL - Decem!;!i[ 2222 GASEOUS EFFLUENTS - ELEVATED RELEASE Batch Mode (2)

Nuclldes Released 1st Q~!1ir ~n~ !dlli!tli[ 3f!! Quai::mr ~1b Q1H1!:1i[

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Xenon-138 C1 Cl ... .. ..

Iodines 11) lodine-131 Cl .. .. ... ..

lodine-133 lodine-135 Cl Cl I .. I .. I .. I .. I Pi!ctl,11li!1m! {j)

Chromium-51 ... .. ... ..

CI Manganese-54 C1 lron-55 Cl lron-59 C1 Cobalt-58 Cl Coball-60 C1 Neoclymiurn-147 Cl Zmc-65 Cl Strontium-89 Cl Stroruurn-90 Cl N1ob1urn-95 CI Zlrconiurn-95 Cl Molybdenum-99 Ruthenium-103 Cl C1 Ceslurn-134 Cl Ceslum-136 Cl CesJurn-137 Cl Banurn-140 CI Lanthanum-140 Cl Cenum-141 Cenurn-144 C1 CI Ir!~UI!! {1} Ci I .. I ... I .. I .. I (1) Concentrallons less than the lower hmrt of detection of the counllng system used are IndIcated wrth a double astensk A lower hmrt of detectJon of 1 OOE-04 µCi/mJ for required noble gases, 1 00E-11 µCi/ml for required partlCtiates and gross alpha, 1 OOE-12 µCi/ml for required lodrnes, 1 00E-11

µCi/ml for Sr-89/90 and 1 OOE-06 µCl/ml for Tnllum, as required by the ODCM, has been venfied (2) Contnbubons from purges are Induded There were no other batch releases dunng the reporllng penod.

Table 1C ATTACHMENT 4 Page 1 of 2 Unit 1 Unlt2 X Be~i:l!ml Pfld~; Janu!}rl - [lecmibet ~iQ GASEOUS EFFLUENTS - GROUND LEVEL RELEASES Continuous Mode (2)

Nuclides Released 1ll!l Ql!i!tw ~mlQ1,111~c 3rd Q!z!s!ru!r £ti QW!l1il:

Elsslon Gases {11 Argon-41 Krypton-85 Cl Cl Krypton-85m Cl Krypton-87 Kryplon-88 Cl Cl Xenon-127 Cl Xenon-131m Cl Xenon-133 Cl Xenon-133m Ci Xenon-135 Cl Xenon-135m Cl Xenon-137 Xenon-138 Cl Cl .. ...

lodl!!E!:!l lodine-131 Cl 3 25E-05 I .. I iodine-133 lodlne-135 Ci Cl I 1 16E-04 I I ...

Partlcy!!!l!! l:!l Chromium-51 Cl Manganese-54 Cl lron-55 Cl lron-59 Cl Cobalt-58 Cl Cobalt-60 Cl 3 68E-04 2 14E-04 1 86E-03 2 45E-04 Neodymiurn-147 Cl Zmc-65 Strontlum-89 Cl Cl .... .. ...

Strontlum-90 Cl N1ob1urn-95 Cl Zlrconlurn-95 Cl Molybdenurn-99 Cl Ru1hernurn-103 Cl Cesiurn-134 Cl Cesiurn-136 Cl CeslLHTI-137 Cl Banum-140 Cl Lanthanum-140 Cl Cenurn-141 Cenum-144 Cl Cl .. ..

Tt:!1!1:1mm Cl I 8 91E+00 I 5 28E+OO I 6,17E+OO I 4 09E+OO I (1) Concentrations less than the lower hmrt of detecbon of the counting system used are 1nd1cated wrth a double astensk A lower hlTl1I of detection of 1.00E-04 µCi/ml for required noble gases, 1 OOE-11 µCi/ml for required particulates and gross alpha, 1.00E-12 µCi/ml for required IodInes, 1 OOE -

11 µCi/ml for Sr-89/90 and 1 00E-00 µCl/ml to; Tnlium, as reqwred by the ODCM, has been venfled (2) There were no batch releases from this path dunng the reporbng penod

Table 1C ATTACHMENT 4 Page 2 of 2 Unit 1 Unit 2 X B~i;iortl!!Q E!i!!lod: Janua!Y - Dec!i!ffiber 202Q GASEOUS EFFLUENTS - GROUND LEVEL RELEASES Batch Mode Nuclldes Released ji! !2!!i!cwr ~as! Q!H!tl!!r ~mQY!!!:mr 4th au111:mr Fission Ar-41 Gases c1 l Cl .... ....- .... --

Kr-85 Ci Kr-85m Cl Kr-87 Kr-88 CI Cl Xe-127 Cl a ...

Xe-131m Xe-133 Cl Xe-133m Xa-135 Cl Cl .... , -.. .... ....

Xa-135m CI Xe-137 Xa-138 CJ C1 -

Iodines (1l 1-131 Cl .... ...... .... ....

1-132 1-133 Cl Cl I .. I I .. I .. I Eart1cu1ates ,11 Cr-51 Cl .... .... .... ..

Mn-54 Fa-55 Cl Ci .... .... ..

H Fe-59 Cl Co-58 Cl Co-80 Cl Nd-147 Cl Zn-65 CI Sr-89 C1 Sr-90 Cl Nb-95 CI Zr-95 C1 Mo-99 Ru-103 C1 Cl ..

Cs-134 Ci Cs-136 Cs-137 Cl Cl .... ....-

Ba-140 CI La-140 Ce-141 Ce-144 CI CI CJ .. .. -..

Trlt!IIll (1} Cl I .. I .. I .. I .. I (1) Coocentrabons less than the lower hmlt of detection of the cooobng system used are 1nchcated with a double ..

Table 2A ATTACHMENT 5 Page 1 of 2 Unit 1 Unit2 X Bi122!l!ng e1rl2~: JiDWlrl

Decem'2§[ 2Q.2 LIQUID EFFLUENTS-SUMMATION OF ALL RELEASES (1) 1~1 !d!!l!!lir ,llil9!H!!lir ~rn !dY11!lir ~1tl Qjgj[E[ ~!, Io~I ~!!Qr, ~

A F1SctIon & A.r:uvanon ProdL*Mct 1 Total Release (not mdudlng Tritium, Cl No Releases No Releases No Releases No Releases 5 OOE+01 gases, alpha) 2 Average diluted concentration during

µCJ/ml No Releases No Releases No Releases No Releases repor!Jng pertcx:I B ..I!!!ll!.!!J.

1 Total release Cl No Releases No Releases No Releases No Releases 5 OOE+01 2 Average diluted concentration dunng the reporting penod

µCJ/ml No Releases No Releases No Releases No Releases C DissQ!Y!ilQ s!DQ i;; t!:sll!l~ ~ii!~

1. Total release Cl No Releases No Releases No Releases No Releases 5 OOE+01 2 Average diluted concentrallon dunng the

µCt/ml No Releases No Releases No Releases No Releases reporting penod D Qr~ 6112!:!li! BIQIQ!i!!.llm

1. Total release Cl I No Releases INo Releases I N? Releases INo Releases I 5.00E+01 E.~

1 Prior to DI1ubon Lrters No Releases No Releases No Releases No Releases 5 OOE+01 2 VolLtTie of dilution water used dunng Lrters No Releases No Releases No Releases No Releases 5.00E+01 release penod 3 Volume of dilution water available dwmg Liters 1.12E+10 1 16E+10 1 24E+10 1 18E+10 5.00E+01 repor!Jng penod I

F i:!.l(!.§DI Qf I ~ Q~ ldm!!l!

Percent of Quarterly Wiole Body Dose Limit

% 0.OOE+OO 0 0OE+OO 0 00E+OO 0 OOE+00 (1.5 mrem)

Percent of Annual Whole Body Dose Limit to Date (3 mrem)

% 0.OOE+O0 0.0OE+OO 0.00E+OO 0 OOE+00 Percent of Quarterly Organ Dose Limit (5

% 0 00E+O0 0 00E+OO 0.OOE+OO 0 OOE+OO mrem)

Percent of Annual Organ Dose l..Jmrt to Date

% 0 00E+O0 0 00E+OO 0 OOE+OO 0 OOE+OO (10 mrem)

Percent of 10CFR20 Concentra!Jon l..Jmrt (2),

(3)

% 0 00E+OO 0 OOE+00 0 OOE+OO 0 00E+OO Percent of Dissolved or Entrained Noble Gas

% 0.OOE+OO 0 OOE+O0 0 00E+O0 0.00E+00 L1mrt (2.00E-04 µCJ/ml)

(1) Concentra!Jons less than the lower hm1t of detection of the counting system used are Ind1cated with a double astensk. A lower llmlt of detection of 5 00E-07 µCJ/ml for required gamma emI11Jng nucttdes, 1 OOE-05 µCi/ml for required dissolved and entrained noole gases and tntIum, 5 OOE-08

µCJ/ml for Sr-89190, 1 OOE-06 µCVml for 1-131 and Fe-55, arid 1 OOE-07 µCJ/ml for gross alpha radIoact1V1ty, as required by the Off-Site Dose Calcufaflon Manual (ODCM), has been venlled.

(2) The percent of 10CFR20 concentra!Jon hmrt Is based on the average concentra!Jon dunng the quarter (3) Improved Technical Speaficatlons nmrt the concentration of radl08c!Jve matertal released In the hqu!d effluents to unrestncted areas to ten !Jmes the concentraflons specified In 10CFR20 1001 - 20 2402 Append IX B, Table 2, Column 2 Maximum Effluent Concentrallons (MEC) numencally equal to ten times the 10CFR20 1001 - 20 2402 concentrations were adopted to evaluate hquId effluents.

Table 28 ATTACHMENT 5 Page 2 of 2 Unit 1 Unit 2 X B1oort1ng fitlfl!I: ~Dldllrt - 0ecem~er iQ2Q LIQUID EFFLUENTS RELEASED Batch Mode (1),(2)

Nuclldes Released l1!Qu11m1r im!Q1111rmr ~ Q!ds!rtir ~1119Y!!!ir Nucllcles Released Strontillll-89 c, No Releases No Releases No Releases No Releases stronbum-90 Ct No Releases No Releases No Releases No Releases Ceslum-134 Cl No Releases No Releases No Releases No Releases Cesium-137 C1 No Releases No Releases No Releases No Releases lodine-131 C1 No Releases No Releases No Releases No Releases Cobalt-58 C1 No Releases No Releases No Releases No Releases Coball-60 c, No Releases No Releases No Releases No Releases lron-59 c, No Releases No Releases No Releases No Releases Zlnc-65 Cl No Releases No Releases No Releases No Releases Manganese-54 Ci No Releases No Releases No Releases No Releases Chrom,um-51 Cl No Releases No Releases No Releases No Releases Zircornum-95 c, No Releases No Releases No Releases No Releases Niobrum-95 Cl No Releases No Releases No Releases No Releases Molybdenum-99 c, No Releases No Releases No Releases No Releases Technebum-99m Cl No Releases No Releases No Releases No Releases Banum-140 Cl No Releases No Releases No Releases No Releases Lanthanum-14-0 c, No Releases No Releases No Releases No Releases Cenum-141 Cl No Releases No Releases No Releases No Releases Tungsten-187 Cl No Releases No Releases No Releases No Releases Arsernc-76 c, No Releases No Releases No Releases No Releases lodine-133 Cl No Releases No Releases No Releases No Releases lron-55 Cl No Releases No Releases No Releases No Releases Nepturnum-239 Cl No Releases No Releases No Releases No Releases S,lver-11 Om Cl No Releases No Releases No Releases No Releases Gold-199 c, No Releases No Releases No Releases No Releases Cenum-144 Ci No Releases No Releases No Releases No Releases Ceslum-136 Cl No Releases No Releases No Releases No Releases Copper-64 Cl No Releases No Releases No Releases No Releases Dissolved or Entrained Gases C1 I No Releases I No Releases I No Releases I No Releases I Tnbum c, I No Releases I No Releases I No Releases I No Releases I (1) No continuous mode release occurred dunng the report penod as 1nd1cated by effluent sampling (2) Concentrations less than the IO'Wef firm of detecbon of the counting system used are indicated wrth a double astensk A lower nm,t of detecbon of 5 OOE-07 µC1/ml for required gamma emitting nuchdes. 1 00E-05 µCi/ml for required dissolved and entrained noble gases and tntlum, 5 OOE-08

µCi/ml for Sr-89/90, 1.00E-06 µCi/ml for 1-131 and Fe-55, and 1.00E--07 µCl/ml for gross alpha radioactivrty, as Identified in the ODCM, has been venfied.

Table 3 ATTACHMENT 6 Page 1 of 4 Unit 1 Unit 2 X B~rtl!!Q P!i!dQi;!: Jan1H!!Y - Q!!!d!filber ~~Q SOLID WASTE AND IRRADIATED FUEL SHIPMENTS A1. TYPE ~ ~t!Yln'.m (mi (Ci)

~ ~

A B C A B C a 1 Spent Resin (Dewatered) 3.06E+01 0 OOE+OO O.OOE+OO 5.86E+01 0 OOE+OO 0 OOE+OO a2 FIiter Sludge O.OOE+OO 0 OOE+OO OOOE+OO O.OOE+OO 0 OOE+OO 0 OOE+OO a.3 Concentrated Waste O.OOE+OO O.OOE+OO 0 OOE+OO O.OOE+OO 0 OOE+OO 0 OOE+OO Totals 3 06E+01 0 OOE+OO 0 OOE+OO 5 86E+01 0 OOE+OO O.OOE+OO b1 Dry Compressible Waste 5 25E+02 0 OOE+OO 0 OOE+OO 2 75E-01 0 OOE+OO 0 OOE+OO b2 Dry Non-Compressible Waste (Contaminated 0 OOE+OO 0 OOE+OO 0 OOE+OO O.OOE+OO 0 OOE+OO 0 OOE+OO Equipment)

Totals 5 25E+02 0 OOE+OO 0 OOE+OO 2 75E-01 0 OOE+OO 0 OOE+OO C Irradiated Components, O.OOE+OO 0 OOE+OO 0 OOE+OO 0 OOE+OO 0 OOE+OO O.OOE+OO Control Rods, etc.

d Other (to vendor for processmg) d.1 Otly waste 9 51E+01 O.OOE+OO O.OOE+OO 9.04E+OO O.OOE+OO 0 OOE+OO (1) The estimated total error 1s 5 OE+01%.

Table 3 ATTACHMENT 6 Page 2 of 4 Unit 1 Unit 2 X Reporting Period: January - December 2020 SOLID WASTE AND IRRADIATED FUEL SHIPMENTS A1. TYPE ~2amlam: Pi!!.~~ §Qll!,Jlfl~t!Q!l Aae!l!

a 1 Spent Resin (Dewatered) Poly Lmer General Design None a 2 Fitter Sludge Poly Lmer Type B None b 1 Dry Compressible Waste Seavan General Design None b.2 Dry Non-Compresslble Waste (contaminated NIA NIA Nia eqlllpment) c Irradiated Components, NIA NIA NIA Control Rods d Other (To vendor for processing)

Oil/Aqueous L.Jqy1d 55 gallon drums General Design None

Table 3 ATTACHMENT 6 Page 3 of 4 Unlt1 Unit 2 X B,~or!Jag eit!QS!: Jal!!d§!rY

D~rn~r 2020 SOLID WASTE AND IRRADIATED FUEL SHIPMENTS A2. ESTIMATE OF MAJOR NUCLIDE COMPOSITION (BY TYPE OF WASTE) :,

a Spent Resins, Filter Sludges, Concentrated Waste

~ ~ ~

Mn-54 2 52% 1 47E+OO Fe-55 19.55% 1 15E+01 Co-60 75 49"k 4.42E+01 Zn-65 1 21% 7 12E--01 b Dry Compressible Waste, Dry Non-Compressible Waste (Contaminated Equipment)

~ ~ ~

Mn-54 525% 1 44E--02 Fe-55 635% 1 75E-02 Co-60 8555% ' 2 35E--01 Zn-65 146% 4 02E--03 c Irradiated Components, Control Rods* There were no shipments

~ ~

NIA NIA d Other (To vendor for processing) ' /

Nucllde ~ ~

Mn-54 441% 3 99E--01 Fe-55 7.11% 6.42E--01 Co-60 8808% 7 96E+00

Table 3 ATTACHMENT 6 Page 4 of 4 Unlt1 Unit 2 X B~r!l!lil eerlod, Ja11Y~IY - Decembe[ ~~

SOLID WASTE AND IRRADIATED FUEL SHIPMENTS K3 SOLID WASTE DISPOSITION

~umllfil Qf ~tl!12flliints Mode Qf Tran~l.1Q ll11QD ~110slt!Q0 23 Truck,tvghway Bear Creek 7 Truck,hlghway Clrve CWF B IRRADIATED FUEL SHIPMENTS (01sposltlon)

~ymber Qf §h112!I!!:l!:!t!i! MQl;l~ Q[ I[iii!:m!21l!i!112 ~lJO!i!tlQ 0 NIA NIA D SEWAGE WASTES SHIPPED TO A TREATMENT FACILllY FOR PROCESSING AND BURIAL I

\

There are no shlprnents of sewage sludge wrth detecbble quantJbes of plant-related nucildes from NMP to the treatment faalrty dunng the reporting penod

ATTACHMENT 7 Page 1 of 1 Unit 1 Unit 2 X Be~!1!!!9 Eer1od: ~!!n!.!aDt'. - ~!fil!~r 20i2

SUMMARY

OF CHANGES TO THE OFF-SITE DOSE CALCULATION MANUAL (ODCM)

The Unit 2 Off-Site Dose Calculation Manual (ODCM) was revised during the reporting period. The table below is a listing of changes.

REVISION 37 New/Amended Page# Description of Change Reason For Change Section#

The change was made to both align with the Changed the analytical frequency for Unit 1 ODCM and NUREG 1302 analytical 13 2-2 Table D 3 2 1-1 alpha acbvity from 7 to 31 days on frequencies for alpha analysis on parbculate particulate filters filters

ATTACHMENT 8 Page 1 of 1 Unit 1 Unit 2 X Bi122!llWI ferlod: JiUl!.l!IY

December i!!20

SUMMARY

OF CHANGES TO THE PROCESS CONTROL PROGRAM (PCP)

No changes were made to the Process Control Program during 2020.

I I

~

ATTACHMENT 9 Page 1 of 1 Unit 1 Unit 2 X B!m2!l!ag Eicl2!!= :h!DIUI~ - 12~mt!i[ l:!!l:!!

SUMMARY

OF NON-FUNCTIONAL MONITORS Dates Monrtor was Monitor Cause and Corrective Actions Non-Functional 2LWS-CAB206, January 1, 2020 to No liquid waste discharges were performed during 2020, and therefore, these 2LWS-FT330 & December 31, 2020 monitors were not ~etumed to service. The discharge manual isolation valves, 2LWS-FT331, 2LWS-V420 and 2L\i\lS-V422, are locked closed durfng inoperable periods, Liquid Waste therefore, no inadvertent discharge can occur. Reference Equipment Status Discharge Monitor Log (ESL) 2010-0243.

ATTACHMENT 10 Page 1 of 4 Unit 1 Unlt2_X_ ReportJng Period: January - December 2020 DOSES TO MEMBERS OF THE PUBLIC DUE TO THEIR ACTIVITIES INSIDE THE SITE BOUNDARY Introduction An assessment of the radiation dose potentially received by a Member of the Public due to their activities inside the site boundary from Nine Mile Point Unit 2 (NMP2) liquid and gaseous effluents has been conducted for the period January through December 2020.

This assessment considers the maximum exposed individual and the various exposure pathways resulting from liquid and gaseous effluents to identify the maximum dose received by a Member of the Public during their activities within the site boundary.

Prior to September 11, 2001, the public had access to the Energy lnfom,ation Center for purposes of observing the educational displays or for picnicking and associated activities. Fishing also occurred near the shoreline adjacent to the Nine Mile Point (NMP) site. Fishing near the shoreline adjacent to the NMP site was the onsite actMty that resulted in the potential maximum dose received by a Member of the Public. Following September 11, 2001 public access to the Energy lnfomiation Center has been restricted and fishing by Members of the Public at locations on site is also prohibited. Although fishing was not conducted during 2020 the annual dose to a hypothetical fishemian was still evaluated to provide continuity of data for the location.

Dose Pathways Dose pathways considered for this evaluation included direct radiation, inhalation and external ground (shoreline sediment or soil doses). Other pathways, such as ingestion pathways, are not considered because they are either not applicable, insignificant, or are considered as part of the evaluation of the total dose to a member of the public located off-site. In addition, only releases from the NMP2 Stack and Radwaste/Reactor Building Vent were evaluated for the inhalation pathway. Dose due to aquatic pathways such as liquid effluents is not applicable since swimming is prohibited at the NMP site.

Dose to a hypothetical fishemian is received through the following pathways while standing on the shoreline fishing:

External ground pathway; this dose is received from plant related radionuclides detected in the shoreline sediment.

Inhalation pathway; this dose is received through inhalation of gaseous effluents released from the NMP2 Stack and Radwaste/Reactor Building Vent.

Direct radiation pathway; dose resulting from the operation of Nine Mile Point Unit 1 (NMP1),

NMP2 and the James A. Fitzpatrick Nuclear Power Plant (JAFNPP) Facilities.

Methodologies for Detennlnlng Dose for Applicable Pathways External Ground (Shore~ine SEK!lment) Pathway ,

Offsite Dose Calculation Manual (ODCM) as adapted from Regulatory Guide 1.109. For this evaluation it is assumed that the hypothetical maximum exposed individual fished from the shoreline at all times.

ATTACHMENT 10 Page 2 of 4 Unit 1 Unlt2_.x_ Reporting Period: January - December 2020 DOSES TO MEMBERS OF THE PUBLIC DUE TO THEIR ACTIVITIES INSIDE THE SITE BOUNDARY The total dose received by the whole body and skin of the maximum exposed individual during 2020 was calculated using the following input parameters:

Usage Factor= 312 hours0.00361 days 0.0867 hours 5.15873e-4 weeks 1.18716e-4 months (fishing 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months per week, 39 weeks per year)

Density i_n grams per square meter= 40,000 Shore width factor = 0.3 Whole body and skin dose factor for each radionuclide= Regulatory Guide 1.109, Table E-Fractional portion of the year = 1 (used average radionuclide concentration over total time p No radionuclides were detected in sediment samples for 2020.

The total whole body and skin doses received by a hypothetical maximum exposed fisherman from the external ground pathway is presented in Table 1, Exposure Pathway Annual Dose.

Inhalation Pathway The inhalation dose pathway is evaluated by utilizing the inhalation equation in the NMP2 ODCM, as adapted from Regulatory Guide 1.109. The total whole body dose and organ dose received by the hypothetical maximum exposed fisherman during 2020 calculated using the following input parameters for gaseous effluents released from both the NMP2 Stack and Radwaste/Reactor Building Vent for the time period exposure is received:

NMP2 Stack:

1 Variable Flshennan X/Q (slm3) 9.60E-07 Inhalation dose factor Table E-7, Regulatory Guide 1.109 Annual air intake (m 3/year) (adult) 8000 Fractional portion of the year 0.0356 H-3 (pCi/sec) 1.78E+06 Fe-55 (pCi/sec) 4.76E-01 Co-60 (pCi/sec) 1.61 E+01 Sr-89 (pCi/sec) 4.74E-01 Sr-90 (pCi/sec) 5.87E-02 1-131 (pCi/sec) 9.60E-02 C-14 (pCi/sec)2 5.46E+05 NMP2 Rae/waste/Reactor Building Vent:

Variable 1 Fishennan X/Q (s/m3) 2 80E-06 Inhalation dose factor Table E-7, Regulatory Guide 1.109 Annual air intake (m 3/year) (adult) 8000 Fractional portion of the year 0.0356 H-3 (pCi/sec) 6 53E+05 Co-60 (pCi/sec) 9.76E+01 I

ATTACHMENT 10 Page 3 of 4 Unit 1 Unit 2 ___x__ Reporting Period: January - December 2020 DOSES TO MEMBERS OF THE PUBLIC DUE TO THEIR ACTIVITIES INSIDE THE SITE BOUNDARY 1

The maximum exposed fisherman is assumed to be present on site during the period of April through December at a rate of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months per week for 39 weeks per year equivalent to 312 hours0.00361 days 0.0867 hours 5.15873e-4 weeks 1.18716e-4 months for the year (fra~ional portion of the year = 0.0356).

Therefore, the Average Stack and Radwaste/Reactor Building Vent flow rates and radionuclide concentrations used to determine the dose are represented by second, third and fourth quarter gaseous effluent flow and concentration values.

2* C-14 release rate determined from NUREG-0016, "Calculation of_ Releases of Radioactive Materials in Gaseous and Liquid Effluents for Boiling Water Reactors _

(BWR-GALE Code)," and EPRI Technical Report 1021106, "Estimation ofCarbon-14 in Nuclear Power Plant Gaseous Effluents."

The total whole body _dose and maximum organ dose received by the hypothetical maximum exposed fisherman is presented in Table 1, Exposure Pathway Annual Dose.

Direct Radiation Pathway The direct radiation pathway is evaluated in accordance with the methodology found in the NMP2 ODCM. This pathway considers four components: direct radiation from the generating facilities, direct radiation from any possible overhead plume, direct radiation from ground deposition and direct radiation from plume submersion. The direct radiation pathway is evaluated by the use of high sensitivity environmental Thermoluminescent Dosimeters (TLDs). Since fishing activities occur between April 1 and December 31, TLD data for the second, third, and fourth quarters of 2020 from TLDs placed in the general area where fishing once occurred were used to determine an average dose to the hypothetical maximum exposed fisherman from direct radiation. The following is a summary of the average dose rate and assumed time spent on site used to determine the total dose received:

Variable Flshennan Average Dose Rate (mRem/hr) 1.32E-03 Exposure time (hours) 312 Total Doses received by the hypothetical maximum exposed fisherman from direct radiation is presented in Table 1, Exposure Pathway Annual Dose.

ATTACHMENT 10 Page4 of 4 Unit 1 Unit 2 _x__ Reoortlaa PerJod; January - December 2020 DOSES TO MEMBERS OF THE PUBLIC DUE TO THEIR ACTIVITIES INSIDE THE SITE BOUNDARY Dose Received By A Hypothetical Maximum Exposed Member of the Public Inside the Site Boundary During 2020 The following is a summary of the dose received by a hypothetical m?)dmum exposed fisherman from liquid and gaseous effluents released from NMP2 during 2020:

TABLE 1 Exposure Pathway Annual Dose Flshennan Exposure Pathway Dose Type .. (mrem)

Whole Body 0.00E+00 External Ground Skin of Whole Body 0.00E+OO Whole Body 2.23E-04 Inhalation MaXJmum Organ Bone

3 39E-04 Thyroid 2.23E-04 Direct Radiation Whole Body 0 41 Based on these values, the total annual dose received by a hypothetical maximum exposed Member of the Public inside the site boundary is as follows:

TABLE2 Annual Dose Summary Flshennan Total Annual Dose for 2020 (mrem)

Total Whole Body 4.11E-01 Skin of Whole Body 0.00E+OO Maximum Organ Thyroid : 3 39E-04 Thyroid 2.23E-04

ATTACHMENT 11 Page 1 of 5 Unit 1 Unlt2_X_ Reporting Period: January - December 2020 DOSES TO MEMBERS OF THE PUBLIC DUE TO THEIR ACTIVmES OUTSIDE THE SITE BOUNDARY Introduction An assessment of radiation doses potentially received by the likely most exposed Member of the Public located beyond the site boundary was conducted for the period January through December 2020 for comparison against the 40 CFR 190 annual dose limits.

The intent of 40 CFR 190 requires that the effluents of Nine Mile Point Unit 1 (NMP1), as well as other nearby uranium fuel cycle facilities, be considered. In this case, the effluents of NMP1, Nine Mile Point Unit 2 (NMP2) and the James A. FitzPatrick Nuclear Power Plant (JAFNPP) facilities must be considered.

40 CFR 190 requires the annual radiation dose received by Members of the Public in the general environment, as a result of plant operations, be limited to:

< 25 mRem whole body

< 25 mRem any organ (except thyroid)

< 75 mRem thyroid This evaluation compares doses resulting from liquid and gaseous effluents and direct radiation originating from the site as a result of the operation of the NMP1, NMP2 and JAFNPP nuclear facilities.

Dose Pathways Dose pathways considered for this evaluation included doses resulting from liquid effluents, gaseous effluents and direct radiation from all nuclear operating facilities located on the Nine Mile Point site.

Dose to the likely most exposed Member of the Public, outside the site boundary, is received through the following pa~ways:

Fish consumption pathway; this dose is received from plant radionuclides that have concentrated in fish that is consum99 by a Member of the Public.

Vegetation consumption pathway; this dose is received from plant radionuclides that have concentrated in vegetation that is consumed by a Member of the Public.

Shoreline Sediment; this dose is received as a result of an individual's exposure to plant radionuclides in the shoreline sediment, which is used as a recreational area.

Deposition, Inhalation and Ingestion pathways resulting fron_, gaseous effluents; this dose is received through exposure to gaseous effluents released from NMP1, NMP2 and JAFNPP operating facilities.

Direct Radiation pathway; radiation dose resulting from the operation of NMP1, NMP2 and JAFNPP facilities (including the Independent Spent Fuel Storage Installations (ISFSI)).

Methodologles for Detennlning Dose for Applicable Pathways

.... Fish Consumption Dose received as a result of fish consumption is based on the methodology specified in the NMP1 Off-Site Dose Calculation Manual (ODCM) as adapt~d from Regulatory Guide 1: 109. The dose for 2020 is calculated from actual analysis results of environmental fish samples taken near the site discharge points. For this evaluation it is assumed that the most likely exposed Member of the Public consumes fish taken near the site discharge points. .

No radionuclides were detected in fish samples collected and analyzed during 2020; therefore, no dose was received by the whole body and organs of the likely most exposed Member of ttie Public during 2020.

r

ATTACHMENT 11 Page 2 of 5 Unit 1 Un It 2 __x__ Reporting Pertod: January - December 2020 DOSES TO MEMBERS OF THE PUBLIC DUE TO THEIR ACTNITIES OUTSIDE THE SITE BOUNDARY Vegetation Consumption Dose received as a result of vegetation consumption is based on the methodology specified in the NMP1 ODCM as adapted from Regulatory Guide 1.109. The dose for 2020 is calculated from actual analysis results of environmental vegetation samples taken near the most exposed Member of the Public.

No radionuclides were detected in vegetation samples collected and analyzed dunng 2020; therefore, no dose was received by the whole body and organs of the likely most exposed Member of the Public during 2020.

For estimating C-14, dose received as a result of vegetation consumption is based on tne methodology specified in the NMP1 ODCM as adapted from Regulatory Guide'1.109. The estimated concentration of C-14 in vegetation is based on the estimated concentration of C-14 in plant gaseous effluents.

Shorellne Sediment Dose received from shoreline sediment is based on the methodology in the NMP1 ODCM as adapted from Regulatory Guide 1.109. For this evaluation it is assumed that the most likely exposed Member of the Public spends 67 hour7.75463e-4 days 0.0186 hours 1.107804e-4 weeks 2.54935e-5 months s/year along the shoreline for recreational purposes.

No radionuclides were detected in shoreline sediment samples collected and analyzed during 2020; therefore no dose was received by the whole body and organs of the likely most exposed Member of the Public during 2020.

Dose Pathways Resultlng From Gaseous Effluents Dose received by the likely most exposed Member of the Public due to gaseous effluents Is calculated in accordance with the methodology provided in the NMP1 ODCM, NMP2 ODCM, and the JAFNPP ODCM. These calculations consider deposition, inhalation and ingestion pathways. Actual meteorological data was used to calculate doses to the likely most exposed Member of the Public. The total sum of doses resulting from gaseous effluents from NMP1, NMP2 and JAFNPP during 2020 provides a total dose to the whole body and maximum organ dose for this pathway.

Carbon-14 Dose Pathways Resulting from Gaseous Effluents The Carbon~14 (C-14) effluent source tem,s are used to estimate radiological doses from C-14 in site gaseous waste effluents. These estimates were generated in order to meet the NRC requirement to incorporate C-14 in nuclear power plant 2020 Annual Radiological Effluent Release Reports (ARERRs). The C-14 production and

,effluent $ource tern, estimates were based on EPRI methodology provided in EPRI Report 1021106, Estimation of Carbon-14 in Nuclear Power Plant Gaseous Effluents, December 2010. The following methodology was used in estimating C-14 gaseous release activity and dose components for the 2020 ARERR.

EPRI methodology for estimating C-14 production rates in Boiling Water Reactors (BWRs):

For BWRs, EPRI Report 1021106 summarized the distribution of C-14 in release pathways as follows. gaseous 95% to 99%, liquid <0.5% and solid 1% to 5%. The report also states that ~95% of C-14 in BWR gaseous waste effluents exists in the carbon dioxide form, which contributes to population dose via photosynthesis uptake in the 1

food consumption cycle.

ATTACHMENT 11 Page 3 of 5 Unit 1 Un1t2_x__ Reporting Pedod: January* December 2020 DOSES TO MEMBERS OF THE PUBLIC DUE TO THEIR ACTIVITIES OUTSIDE THE SITE BOUNDARY For NMP1 and NMP2, C-14 gaseous dose calculations in the site ARERR are made using the following assumptions for each unit: (1) continuous release of the estimated C-14 generated during power operation based on the number of Effective Full Power Days (EFPDs) for the period, (2) maximum C-14 activity from literature values cited in EPRI Report 1021106, and (3) typical fraction as carbon dioxide for gaseous releases from literature values also cited in EPRI Report 1021106.

Equation 1 estimates the maximum annual production of C-14, PRMAX, for each BWR unit.

PRMAX = 5.1

MWT / 1000 [Eq 1]

Where:

5.1 = BWR Normalized Production (Ci/GWt-yr)

MWT = Megawatts Thermal (MWt) 1000 = Conversion Factor (Mwt to GWt)

Equation 2 estimates the C-14 activity released, A e- 14 , into the gaseous pathway during the time period for each.

BWR unit.

A e-14 = PR MAX

0.99

EFPD I 365, Ci (for time period) [Eq 2]

Where:

PR MAX = maximum annual production rate of C-14 0.99 = fraction of C-14 in BWR gaseous pathway releases (maximum literature value in EPRI Report 1021106; also Table 1)

EFPD = number of effective full power days for the unit during the time period; e.g., quarterly or yearly (Table 1) 365 = number of days in a typical year

ATTACHMENT 11 Page 4 of 5 Unit 1 Un it 2 ___x__ Reporting Period: January* December 2020 DOSES 'fO MEMBERS OF THE PUBLIC DUE TO THEIR ACTIVITIES OUTSIDE THE SITE BOUNDARY Equation 3 estimates the C-14 activity released in carbon dioxide form, A e- 14, co2 , into the gaseous pathway during the time period for each BWR unit.

A C-14, CO2 = PR MAX

0. 99

0. 95

EFPD I 365, Ci (for time period) [Eq3]

Where:

PR MAX = maximum annual production rate of C-14 0.99 = fraction of C-14 in BWR gaseous pathway releases (maximum literature value in EPRI Report 1021106; also Table 1) 0.95 = fraction of C-14 as carbon dioxide in BWR gaseous pathway releases (typical literature value in EPRI Report 1021106; also Table 1)

EFPD = number of effective full power-days for the unit during the time period, e.g. quarterly or yearly (Table 1) 365 = conversion factor, 365 days in a typical average year For each BWR unit, the 2020 estimated C-14 activity releases (total and carbon dioxide chemical form) are summarized in Table 1.

Table 1 2020 BWR Estimated C-14 Gaseous Releases Gaseous CO2 Fonn Max. Annual 2020 Total EFPD 2020 CO2 Release BWR Release Release Prod. Rate Release Operation (Eq 3)

Fractlon 1~1 Fractlonlbl (Eq 1) (Eq 2) 364 879 EFPD NMP1 099 0.95 (99 69%) 9 44 Ci/yr 934 C1 8 87 c, 330.724 EFPD NMP2 0.99 0 95 (90 36%) 20 33 Ci/y~c) 18 24 Ci 17.28 c, 327.83 EFPD JAFNPP 0 99 0 95 (89 57%)

12 93 C1/yr 963 c, 915 Ci (a) Maximum literature values from EPRI Report 1021106.

(b) Typical value from EPRI Report 1021106 (c) NMP2 Reactor Power Rating increased to 3988 Megawatts thennal.

As long as the core designs and power ratings are not sign'ificantly changed, the maximum annual production rates and annual total and carbon dioxide activity releases in Table 1 should be acceptable for use in estimating C-14 gaseous release activity and dose components for the ARERR.

ATTACHMENT 11 Page 5 of 5 Unit 1 Unlt2_X_ Reporting Perfod: January- December 2020 DOSES TO MEMBERS OF THE PUBLIC DUE TO THEIR ACTIVITIES OUTSIDE THE SITE BOUNDARY Direct Radiation Pathway Dose as a result of direct gamma radiation from the site, encompasses doses from direct "shine" from the generating facilities, direct radiation from any overhead gaseous plumes, plume submersion, and ground deposition. This total dose is measured by environmental TLDs. The critical location is based on the closest year-round residence from the generating facilities as well as the closest residence in the critical downwind sector in order to evaluate both direct radiation from the generating facilities and gaseous plumes as detennined by the local meteorology. During 2020, the closest residence and the critical downwind residence are at the same location.

Table 2 Dose Potentially Received by the Likely Most Exposed Member of the Public Outside the Site Boundary During 2020 Exposure Pathway Dose Type Dose (mrem)

Fish and Vegetation Total Whole Body No Dose Consumption Total Maximum Organ No Dose Total Whole Body No Dose Shoreline Sediment Total Skin of Whole Body No Dose Total Whole Body 2.98E-03 Gaseous Effluents Thyroid 8.33E-03 (excluding C-14)

Maximum Organ Thyroid : 8.33E-03 Gaseous Effluent Total Whole Body 2.85E-01 (C-14 only) Maximum Organ Bone : 1.42E+O0 Direct Radiation Total Whole Body 2.24E+O0 Based on these values the maximum total annual dose potentially received by the likely most exposed Member of the Public during 2020 is as follows:

Total Whole Body: 2.52E+00 Total Thyroid: 8.33E-03 Maximum Organ: Bone : 1.42E+00 40 CFR 190 Evaluation The maximum total doses presented in this attachment are the result of operations at the NMP1, NMP2 and the JAFNPP facilities. The maximum organ dose (Bone: 1.42 mrem), maximum thyroid dose (8.33E-03 mrem) and the maximum whole body dose (2.52 mrem) are below the 40 CFR 190 criteria of 25 mrem per calendar year to the maximum exposed organ or the whole body, and below 75 mrem per calendar year to the thyroid.

ATTACHMENT 12 Page 1 of 1 Unit 1 Unlt2_x___

-- B!u2ort1ng Period: J!!!!l.li!~ - ~~bar 2Q2Q Well Identification #Samples # Positive Minimum Maximum Number Collected Samples Concentration (pCi/1) Concentration (pCi/1)

GMX-MW1* 1 0 <176 <176 MW-1 1 0 <180 <180 MW-5 3 0 <181 <196 MW-6 1 0 <172 <172 MW-7 1 0 <188 <188 MW-8 3 0 <179 <192 MW-91 3 0 <179 <195 MW-101 1 0 <177 <177 MW-11 1 0 <183 <183 MW-12 1 0 <178 <178 MW-13 1 0 <181 <181 MW-14* - 1 0 <179 <179 MW-15 3 0 <178 <191 MW-16 1 0 <175 <175 MW-17 3 0 <179 <192 MW-18 3 0 <180 <193 MW-19 1 0 <182 <182 MW-20 1 0 <180 <180 MW-21 1 0 <179 <179 NMP2 MAT2,3 4 1 <183 <226 PZ-1 2 0 <174 <194 PZ-2 1 0 <178 <178 PZ-3 1 0 <177 <177 PZ-4 1 0 <177 <177 PZ-5 1 0 <175 <175 PZ-6 1 0 <184 <184 PZ-7 3 1 <181 242 PZ-8 3 0 <180 <195 PZ-9* 1 0 <178 <178 Notes* * - Control Location 1

- Sentinel well location 2

- NMP2 Groundwater Depression Cone 3

- Samples collected from storm drain system which Includes precipitation 4 oo

- No samples were collected dunng 2 due to Covid-19

ATTACHMENT 13 Page 1 of 1 Unit 1 Unit 2 X Reporting Period: January - December 2020 Off-9ite Dose Calculation Manual <ODCM}

~

Exelon Generation NINE MILE POINT UNIT 2 N2-ODCM OFF-SITE DOSE CALCULATION MANUAL (ODCM)

REVISION 37 Level 3 - Information Use Revision of this document requires PORC approval and changes are controlled by CY-AA-170-3100, Offslte Dose Calculation Manual Revisions

SUMMARY

OF REVISIONS REVISION 37 ( Effective Jan 2020 )

PAGE DATE 13.3-13,14 August 2000 13.3-6 (

November 2000 14.0-1 November 2000 11 2-10,26,33-36,66,67,75,80 November 2000 ix, 11.0-1, 11.0-2, I B 3.3-2, 14.1-1 & 1a, 1111, 1115, 1129, 1163, II 107, II' 108 December 2001 13.3-9 December 2002 I 3.3-10 March 2003 I 3.3-7, I 3.3-12, and I 3.3-13 January 2004 II 63, II 64, and II 107 December 2005 113 and 114 May 2006 iv, 11.0-1, 13.1-7, 13.2-3, 13.2-10, 13.2-12, 13.3-1, 13.3-2, 13.3-3, I 3.3-7, I 3.3-8, I 3.3-9, I 3.3-10, I B 3.1-3, I B 3.2-5, I B 3.2-6, I B 3.3-1, I B 3.3-2, I 4.1-1a, 1110, 1113, 1120, and 1123 September 2006 1112, 1115, 1116 September 2007 II 16 September 2007 11.0-3, 11.0-4, 13.2-2, 13.2-3, 14.2-1, 14.2-2, 119, U 13, 1114, 1116, 11 20, 11 58, II 62-65, II 90, II 104, II 105, II 107, II 108, and II 109 December 2008 I 3.5-6, II 63, II 64, II 65, II 108 October 2009 I 3.3-13 February 2011 I 3.3-9, I 3.3-10, II 63 July 2011 I 3.1-5, I 3.2-5, I 3.2-8, I B 3.1-2, I B 3.2-2, I B 3.2-3, II 27, II 29, 1141 thru II 59 June 2012 I 3.3-11 January 2020 Unit 2 Revision 37 September 2020

TABLE OF CONTENTS List of Tables vii List of Figures ix Introduction X PART I - RADIOLOGICAL EFFLUENT CONTROLS SECTION 1.0 DEFINITIONS 11.0-0 SECTION 2.0 Not Used SECTION 3.0 APPLICABILITY 13.0-0 D 3.1 Radioactive Liquid Effluents I 3.1-1 D 3.1 1 Liquid Effluents Concentration I 3.1-1 D 31.2 Liquid Effluents Dose I 3.1-4 D 3.1.3 Liquid Radwaste Treatment System I 3.1-7 D 3.2 Radioactive Gaseous Effluents I 3.2-1 D 3.2.1 Gaseous Effluents Dose Rate I 3.2-1 D 3.2.2 Gaseous Effluents Noble Gas Dose I 3-2-4 D 3.2.3 Gaseous Effluents Dose:.... lodine-131, lodine-133, Tritium, and Radioactive Material in Particulate Form 13.2-7 D 3.2.4 Gaseous Radwaste Treatment System I 3.2-10 D 3.2.5 Ventilation Exhaust Treatment System I 3.2-12 D 3.2.6 Venting or Purging I 3.2-14 D3.3 Instrumentation I 3.3-1 D 3.3.1 Radioactive Liquid Effluent Monitoring Instrumentation I 3.3-1 D 3.3.2 Radioactive Gaseous Effluent Monitoring Instrumentation,,. 13.3-7 D 3.4 Radioactive Effluents Total Dose I 3 4-1 D 3.5 Radiological Environmental Monitoring I 3.5-1 D 3 5.1 Monitoring Program I 3.5-1 D..,3.5.2 Land Use Census I 3.5-13 D 3.5.3 lnterlaboratory Comparison Program I 3.5-16 BASES I B 3.1-0 B 3.1 Radioactive Liquid Effluents 183.1-1 B 3.1.1 Liquid Effluents Concentration I B 3.1-1 B 3.1.2 Liquid Effluents Dose I B 3.1-2 B 3.1.3 Liquid Radwaste Treatment System I B 3.1-3 TABLE OF CONTENTS (Cont)

PAGE B 3.2 Radioactive Gaseous Effluents I B 3.2-1 ii Unit 2 Revision 37 September 2020

B 3.2.1 Gaseous Effluents Dose Rate I B 3.2-1 B 3.2.2 Gaseous Effluents Noble Gas Dose I B 3-2-2 B 3.2.3 Gaseous Effluents Dose- lodine-131, lodine-133, Tritium, and Radioactive Material in Particulate Form I B 3.2-3 B 3.2.4 Gaseous Radwaste Treatment System I B 3.2-5 B 3.2.5 Ventilation Exhaust Treatment System I B 3.2-6 B 3.2.6 Venting or Purging I B 3.2-7 B 3.3 Instrumentation I B 3.3-1 B 3.3.1 Radioactive Liquid Effluent Monitoring Instrumentation I B 3.3-1 B 3.3.2 Radioactive Gaseous Effluent Monitoring Instrumentation I B 3.3-2 B 3.4 Radioactive Effluents Total Dose I B 3.4-1 B 3.5 Radiological Environmental Monitoring I B 3.5-1 B 3.5.1 Monitoring Program I B 3.5-1 B 3.5.2 Land Use Census I B 3.5-2 B 3.5.3 lnterlaboratory Comparison Program I B 3.5-3 SECTION 4.0 ADMINISTRATIVE CONTROLS 14.0-1 D4.1 Reporting Requirements I 4.1-1 D 4.1.1 Special Reports I 4.1-1 D4.2 Major Changes to Liquid, Gaseous and Solid Radwaste Treatment Systems 14.2-1

_j

(

Unit 2 Revision 37 September 2020

TABLE OF CONTENTS (Cont)

SECTION SUBJECT REF SECTION PAGE PART II - CALCULATIONAL METHODOLOGIES II 1 1.0 LIQUID EFFLUENTS II 2 1.1 Liquid Effluent Monitor Alarm Setpoints II 2 1.1.1 Basis 3.1.1 II 2 1.1.2 Setpoint Determination Methodology 3.3.1 II 2 1.1.2.1 Liquid Radwaste Effluent Radiation Alarm Setpoint II 2 1.1.2.2 Contaminated Dilution Water Radwaste Effluent Monitor Alarm Setpoint Calculations II 5 1.1.2.3 Service Water and Cooling Tower Slowdown Effluent Radiation Alarm Setpoint II 6 1.2 Liquid Effluent Concentration Calculation 3.1.1 II 7 DSR 3.1.1.2 1.3 Liquid Effluent Dose Calculation Methodology 3.1.2 II 8 DSR 3.1.2.1 1.4 Liquid Effluent Sampling Representativeness Table D 3.1.1-1 II 9 note b 1.5 Liquid Radwaste System FUNCTIONALITY 3.1.3 II 1O DSR 3.1.3.1 B 3.1.3 2.0 GASEOUS EFFLUENTS II 12

2. ~ Gaseous Effluent Monitor Alarm Setpoints II 12 2.1.1 Basis 3.2.1 II 12 2.1.2 Setpoint Determination Methodology Discussion 3.3.2 II 12 2.1.2.1 Stack Noble Gas Detector Alarm Setpoint Equation II 13 2.1.2.2 Vent Noble Gas Detector Alarm Setpomt Equation II 14 2.1.2.3 Offgas Pretreatment Noble Gas Detector Alarm Setpoint Equation II 15 2.2 Gaseous Effluent Dose Rate Calculation Methodology 3.2 1 II 16 2.2.1 X/Q and Wv - Dispersion Parameters for Dose Rate, Table D 3-23 ' II 16 2.2.2 \/Vhole Body Dose Rate Due to Noble Gases DLCO 3.2.1.a II 17 DSR 3.2.1.1 2.2.3 Skin Dose Rate Due to Noble Gases DLCO 3.2.1.a II 18 DSR 3.2.1.1 iv Unit 2 Revision 37 .

September 2020

TABLE OF CONTENTS (Cont)

SECTION SUBJECT REF SECTION PAGE 2.2.4 Organ Dose Rate Due to 1-131, 1-133, Tritium and DLCO 3.2.1.b Particulates with half-lives greater than 8 days DSR 3.2.1.2 1119 2.3 Gaseous Effluent Dose Calculation Methodology 3.2.2 1120 3.2.3 3.2.5 2.3.1 Ws and Wv - Dispersion Parameters For Dose, Table D 3-23 1120 2.3.2 Gamma Air Dose Due to Noble Gases 3.2.2 11 21 DSR 3.2.2.1 2.3.3 Beta Air Dose Due to Noble Gases 3.3.2 II 21 2.3.4 Organ Dose Due to 1-131, 1-133, Tritium and Particulates 3.2.3 with Half-Lives Greater than 8 Days 3.2.5 DSR 3.2.3.1 DSR 3.2.'5.1 21 2.4 1-133 and 1-135 Estimation 22 2.5 lsokinetic Sampling 22 2.6 Use of Concurrent Meteorological Data vs.- Historical Data 22 2.7 Gaseous Radwaste Treatment System Operation 3.2.4 22 2.8 Ventilation Exhaust Treatment System Operation 3.2.5 23 3.0 URANIUM FUEL CYCLE 3.4 1124 3.1 Evaluation of Doses From Liquid Effluents DSR 3.1.2.1 1125 3.2 Evaluation of Doses From Gaseous Effluents DSR 3.2.2.1 1126 3.3 Evaluation of Doses From Direct Radiation DSR 3.2.3.1 II 27 3.4 Doses to Members of the Public Within the Site Boundary 4.1 II 27 4.0 ENVIRONMENTAL MONITORING PROGRAM 35 1130 4.1 Sampling Stations 3.5.1 1130 DSR 3.5.1.1 4.2 lnterlaporatory Comparison Program DSR 3.5.3.2 1130 4.3 Capabilities for Thermoluminescent Dosimeters Used for Environmental Measurements II 31 j

V Unrt 2 Revision 37 September 2020

TABLE OF CONTENTS (Cont)

SECTION SUBJECT REF SECTION PAGE Appendix A Liquid Dose Factor Derivation II 66 Appendix B Plume Shine Dose Factor Derivation II 69 Appendix C Dose Parameters for Iodine 131 and 133, Particulates and Tritium II 73 Appendix D Diagrams of Liquid and Gaseous Radwaste Treatment Systems and Monitoring Systems II 83 Appendix E Nine Mile Point On-Site and Off-Site Maps 11106 vi Unit 2 Revision 37 September 2020

LIST OF TABLES PART I - RADIOLOGICAL EFFLUENT CONTROLS TABLE NO TITLE PAGE D 3.1.1-1 Radioactive Liquid Waste Sampling and Analysis I 3 1-2 D 3.2.1-1 Radioactive Gaseous Waste Sampling and Analysis 13.2-2 D 3.3.1-1 Radioactive Liquid Effluent Monitoring Instrumentation 13.3-6 D 3.3.2-1 Radioactive Gaseous Effluent Monitoring Instrumentation I 3.3-13 D 3.5.1-1 Radiological Environmental Monitoring Program 13.5-6 D 3.5.1-2 Reporting Levels for Radioactivity Concentrations in Environmental Samples I 3.5-10 D 3.5.1-3, Detection Capabilities for E_nvironmental Sample Analyses I 3.5-11 vii Unit2 Rev1s1on 37 September 2020

LIST OF TABLES (Cont)

PART II- CALCULA TIONAL METHODOLOGIES TABLE NO TITLE PAGE D 2-1 Liquid Effluent Detector Response 1133 D 2-2 thru D 2-5 Aiat. Values - Liquid Effluent Dose Factor 1134 D 3-1 Offgas Pretreatment Detector Response II 38 D 3-2 Finite Plume - Ground Level Dose II 39 Factors from an Elevated Release D 3-3 Immersion Dose Factors 1140 D 3-4 thru D 3-22 Dose And Dose Rate Factors, R, 1141 D 3-23 Dispersion Parameters at Controlling II 60 Locations, X/Q, Wv and Wa Values D 3-24 Parameters for the Evaluation of Doses to 1161 Real Members of the Public from Gaseous and Liquid Effluents D 51 Radiological Environmental Monitoring II 62 Program Sampling Locations viii Unit 2 Revision 37 September 2020 r.

LIST OF FIGURES FIGURE NO TITLE PAGE D 1.0-1 Site Area and Land Portion of Exclusion Area Boundaries

~

11.0-4 D 5.1-1 Nine Mile Point On-Site Map II 107 D 5.1-2 Nine Mile Point Off-Site Map (page 1 of 2) II 108 D 5.1-2a Nine Mile Point Off-Site Map (page 2 of 2) II 109 ix Unit2 Revision 37 September 2020

INTRODUCTION The OFFSITE DOSE CALCULATION MANUAL (ODCM) is a supporting document of the Technical Specifications Section 5.5.1. The previous Lir,niting Conditions for Operation that were contained in the Radiological Effluent Technical Specifications are now transferred to the ODCM as Radiological Effluent Controls. The ODCM contains two parts: Radiological Effluent Controls, Part I; and Calculational Methodologies, Part II. Radiological Effluent Controls, Part 1, includes the following: (1) The Radioactive Effluent Controls and Radiological Environmental Monitoring Programs required by Technical Specification 5.5.1 and (2) descriptions of the information that should be included in the Annual Radiological Environmental Operating and Radioactive Effluent Release Reports required by Technical Specifications 5.6.2 and 5.6.3. Calculational Methodologies, Part II, describes the methodology and parameters to be used in the calculation of liquid and gaseous effluent monitoring instrumentation alarm/trip setpoints and the calculation of offsite doses due to radioactive liquid and gaseous effluents. The ODCM also contains a list and graphical description of the specific sample locations for the radiological environmental monitoring program, and liquid and gaseous radwaste treatment system configurations.

The ODCM follows the methodology and models suggested by NUREG-0133 and Regulatory Guide 1.109, Revision 1. Simplifying assumptions have been applied in this manual where applicable to provide a more workable document for implementing the Radiological Effluent Control requirements; this simplified approach will result in a more conservative dose evaluation for determining compliance with regulatory requirements.

The ODCM will be maintained for use as a reference and training document of accepted methodologies and calculations. Changes to the calculation methods or parameters will be incorporated into the ODCM to assure that the ODCM represents the present methodology in all applicable areas. Any changes to the ODCM will be implemented in accordance with Section 5.5.1 of the Technical Specifications.

X Unit2 Revision 37 September 2020

PART I - RADIOLOGICAL EFFLUENT CONTROLS Unit2 Revision 37 I September 2020

Definitions 1.0 PART I - RADIOLOGICAL EFFLUENT CONTROLS SECTION 1.0 DEFINITIONS Unit2 Revision 37 I 1.0-0 September 2020

/

_J Definitions 1.0 1.0 DEFINITIONS

NOTE----- --------------------------------------------

Technical Specifications defined terms and the following additional defined terms appear in capitalized type and are applicable throughout these specifications and bases.

TERM DEFINITION FUNCTIONAL FUNCTIONALITY is an attribute of Structures, Systems, or Components (FUNCTIONALITY) (SSCs) that is not controlled by Technical Specifications. An SSC shall be functional or have functionality when it is capable of performing its specified function as set forth in the Current Licensing Basis (CLB).

FUNCTIONALITY does not apply to specified safety functions, but does apply to the ability of non-Technical Specifications SSCs to perform specified support functions.

GASEOUS A GASEOUS RADW ASTE TREATMENT SYSTEM shall be any system RADWASTE designed and installed to reduce radioactive gaseous effluents by collecting TREATMENT offgases from the main condenser evacuation system and providing for SYSTEM delay or holdup for the purpose of reducing the total radioactivity prior to release to the environment.

MEMBER(S) MEMBER(S) OF THE PUBLIC shall include all persons who are not OF THE PUBLIC occupationally associated with the Nine Mile Point Nuclear Station and James A. FitzPatrick Nuclear Power Plant. This category does not include employees of owners and operators of the Nine Mile Point Nuclear Station and James A. Fitzpatrick Nuclear Power Plant, their contractors or 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 Nine Mile Point Nuclear Station and James A. FitzPatrick Nuclear Power Plant.

MILK SAMPLING A MILK SAMPLING LOCATION is a location where 10 or more head of LOCATION milk animals are available for collection of milk samples.

(continued)

Unit2 Revision 37 I 1.0-1 September 2020

Definitions 1.0 1.0 DEFINITIONS (continued)

TERM DEFINITION OFFSITE DOSE The OFFSITE DOSE CALCULATION MANUAL (ODCM) shall contain CALCULATION the current methodology and parameters used in the calculation of offsite MANUAL doses that result from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring Alarm/frip Setpoints, and in the conduct of the environmental radiological monitoring program. The ODCM shall also contain: (1) the Radioactive Effluent Controls and Radiological Environmental Monitoring Program required by Specification 5.5.1 of Technical Specifications and, (2) descriptions of the information that should be included in the Annual Radiological Environmental Operating and Radioactive Effluent Release Reports required by Technical Specifications 5.6.2 and 5.6.3.

PURGE- PURGE and PURGING shall be the controlled process of discharging air PURGING or gas from a confinement to maintain temperature, pressure, concentration, or other operating condition, in such a manner that replacement air or gas is required to purify the confinement.

REPORTABLE A REPORTABLE EVENT shall be any of those conditions specified in EVENT 10 CFR 50.73.

SITE BOUNDARY The SITE BOUNDARY shall be that line around the Nine Mile Point Nuclear Station beyond which the land is pot owned, leas'ed or otherwise controlled by the owners and operators of Nine Mile Point Nuclear Station and James A. Fitzpatrick Nuclear Power Plant See Figure D 1.0-1.

SOURCE CHECK A SOURCE CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to a source of increased radioactivity.

UNRESTRICTED An UNRESTRICTED AREA shall be any area at or beyond the SITE AREA BOUNDARY, access to which is not controlled by the owners and operators of Nine Mile Point Nuclear Station and James A. Fitzpatrick Nuclear Power Plant for purposes of protection of individuals from exposure to radiation and radioactive materials, or any area within the SITE BOUNDARY used for residential quarters or for industrial, commercial, institutional, and/or recreational purposes.

(continued)

Unit2 Revision 37 I 1.0-2 September 2020

Definitions 1.0 1.0 DEFINITIONS (continued)

TERM DEFINITION VENTILATION A VENTILATION EXHAUST TREATMENT SYSTEM shall be any EXHAUST system designed and installed to reduce gaseous radioiodine or radioactive TREATMENT material in particulate form in effluents by passing ventilation or vent SYSTEM exhaust gases through charcoal adsorbers and/or HEPA filters for the purpose of removing iodines or particulates from the gaseous exhaust stream prior to the release to the environment (such a system is not considered to have any effect on noble gas effluents). Engineered safety features (ESF) atmospheric cleanup systems are not considered to be VENTILATION EXHAUST TREATtvfENT SYSTEM components.

VENTING VENTING shall be the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, concentration, or other operating condition, in such a manner that replacement air or gas is not provided or required during VENTING. Vent, used in system names, does not imply a VENTING process.

Unit2 Revision 37 I 1.0-3 September 2020

Definitions 1.0 FIGURED 1.0-1 ~PUnll2

Jl,f Liquid Discharge Lake Ontario 1

Site Area and Land Uqu!dD!adage Portion of Exclusion Area Boundartes IJbROld l

~

Enq LyconfaJ

.1 Note: National Grid retains ownership In certain transmission line and swltchyard facllltles within the exclusion area boundary.

0 1 Access and usage are controlled by Nine Mlle Point Nudear Scale (MIies)

Sta~on, LLC by agreement.

Unit2 Revision 37 I 1.0-4 September 2020

PART I- RADIOLOGICAL EFFLUENT CONTROLS SECTION 3.0 APPLICABILITY Unit2 Revision 37 I 3.0-0 September 2020

Applicability 3.0 3.0 APPLICABILITY The Offsite Dose Calculation Manual (ODCM) Specifications are contained in Section 3.0 of Part I. They contain operational requirements, Surveillance Requirements, and reporting requirements. Additionally, the Required Actions and associated Completion Times for degraded Conditions are specified. The format is consistent with the Technical Specifications (Appendix A to the NMP2 Operating License).

The rules of usage for the ODCM Specification are the same as those for the Technical Specifications. These rules are found in Technical Specifications Sections 1.2, "Logical Connectors," 1.3, "Completion Times," and 1.4, "Frequency."

The ODCM Specifications are subject to Technical Specifications Section 3.0, "Limiting Condition for Operation (LCO) Applicability and Surveillance Requirement (SR) Applicability,"

with the following exceptions:

1. LCO 3.0.6, regarding support/supported system ACTIONS is not applicable to ODCM Specifications.

2. LCO 3.0.7, regarding allowances to change specified Technical Specifications is not applicable to ODCM Specifications.

3. Section 3.0 requirements are not required when so stated in notes within individual specifications.

Unit2 Revision 37 13.0-1 September 2020

Liquid Effluents Concentration D 3.1.1 D 3.1 RADIOACTIVE LIQUID EFFLUENTS D 3.1.1 Liquid Effluents Concentration DLCO 3.1.1 The concentration of radioactive material released in liquid effluents to UNRESTRICTED AREAS (Figure D 1.0-1) shall be limited to:

a. Ten times the concentration s_pecified in 10 CFR Part 20, Appendix B, Table 2, Column 2 for rad1onuclides other than dissolved or entrained noble gases; and

b. 2 x 10-4 µCi/ml total activity concentration for dissolved or entrained noble gases.

APPLICABILITY: At all times.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Concentration of A. I Initiate action to restore Immediately radioactive material concentration to within limits.

released in liquid effluents to UNRESTRICTED AREAS exceeds limits.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY DSR 3.1.1.1 Perform radioactive liquid waste sampling and In accordance with activity analysis. Table D 3.1.1-1 DSR3.1.1.2 Verify the results of the DSR 3 .1.1.1 analyses to In accordance with assure that the concentrations at the point of release Table D 3.1.1-1 are maintained within the limits of DLCO 3.1.1.

Unit2 Revision 37 I 3.1-1 September 2020

Liquid Eflluents Concentration D3.l.l Table D 3.1.1-1 (Page 1 of2)

Radioactive Liquid Waste Sampling and Analysis SAMPLE LOWER LIMIT OF SAMPLE TYPE SAMPLE ANALYSIS SAMPLE DETECTION LIQUID RELEASE TYPE FREQUENCY FREQUENCY ANALYSIS (LLD) (a)

Batch Waste Rolease Grab Sample Each Batch (g) Each Batch (g) Pnnc1pal 5 X 10-1 µCt/ml Tanks (b) Gamma Enutters (c) a 2LWS-TK4A b 2LWS-TK4B 1-131 Ix JQ-6 µCt/ml C 2LWS-TK5A d 2LWS-TK5B Grab Sample One batch/31 31 days DISSOived and I X 10-5 µCt/ml days (g) Entrwned Gases (gamma emrtters)

Proportional Each batch (g) 31 days H-3 Ix 10-5 µCt/ml Composite of grab swnples Gross Alpha Ix 10-7 µCt/ml (d)

Proportional Each batch (g) 92 days Sr-89 5 x [()--8 µCt/ml Composite of grab samples (d)

Sr-90 5 x lo-" µCt/ml Fe-55 Ix 10-,; µCt/ml 2 Continuous Releases Grab Sample 31 days (e) 31 days (e) Pnnc1pal Sx 10-7 µCt/ml

'\_ Gamma II Service Water Emitters (c)

Effluent A b Service Water Grab Sample 31 days (e) 31 days(e) 1-131 I x I Q-6 µCt/ml Effluent B C Coolmg Tower B10 ....tlown Grab Sample 31 days (e) 31 days (e) DISSOived and I X 10-5 µCt/ml Entramed Gases (gamma ernrtters)

Grab Sample 31 days (e) 31 days (e) H-3 I x 10-5 µCt/ml Grab Sample 3 I days (e) 31 days (e) Gross Alpha . I x 10-1 µCt/ml Grab Sample 92 days (e) 92 days (e) Sr-89 5 x j()--8 µCt/ml Grab Sample 92 days (e) 92 days (e) Sr-90 5 x J()--8 µCt/ml Grab Sample 92 days (e) 92 days (e) Fe-55 I x I Q-6 µCt/ml 3 Contmuous Release Grab Sample 31 days (t) 31 days (t) Pnnc1pal 5 x 10-1 µCt/ml Gamma AUX1hmy Bmler Em1tters{c)

Pump Seal and Sample Cooling Grab Sample 92 days (t) 92 days (t) H-3 I X 10-5 µCt/ml Discharge (Somco Water)

Unit2 Revision 37 I 3.1-2 September 2020

Liquid Effluents Concentration 1

D 3.1.1 Table D 3.1.1-1 (Page 2 of2)

Radioactive Liquid Waste Sampling and Analysis

-(a) lbe LLD IS defined as the smallest concentration ofrad1oactrve matenal ma sample that will yield a net count, above system background, that will be detected with 95% probability With only 5% probability of falsely concludll18 that a blank observation represents a "real" signal For a particular measurement system, wluch may IIlClude radiochemical separation LLD (E) (V) (2.22xl0 6 ) (Y) e-ui where LLD The before-the-fact lower hrnrt of detection (µC1 per unJt mass or volwne ),

Si, The standard deviation of the background countmg rate or of the countmg rate of a blank sample as appropriate (counts per mmute),

E The counbng efficiency (=ts per d1smtegrat1on),

V The sample SIZO (units of mass or volume),

222x la6 The number of d1smtegratlons per mmute per µC1, y The fractional rad1ochem!Cal yield, when apphcable, The radioactive decay constant for the particular rad1onuchde (sec- 1), and M The elapsed tune between the m1dpomt of sample collection and the time of countmg (seconds)

Typical values ofE, V, Y, and Lit should be used m the calculation It should be recogm.zed that the LLD IS defined as a before-the-fact hmrt represenbng the capab1hty ofa measurement system and not as an after-the-fact hm1t for a particular measuremerrt (b) A batch release 1s the dISCharge of hqwd wastes of a dISCrete volume Pnor to sampling for analyses, each batch shall be ISOiated, and ~ thoroughly mIXed by the method descnbed m Part II,Section I 4 to assure representative samphng (c) The pnnc1pal gamma eillltters for which the LLD apphes mclude the folloWing radionuchdes Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, Cs-134, Cs-137, and Ce-141 Ce-144 shall also be measured, but with an LLD of5 x l(rµCi/ml Tlus lISt does not mean that only these nuchdes are to be considered Other gamma peaks that are 1dentlfiable, together With those of the above nuchdes, shall also be analyzed and reported m the Radioactive Effluent Release Report pursuant to Techrucal Specification 5 6 3 m the formatoutlmedmRG 121,AppendIXB,RevISIOn !,June 19,74 (d) A cornpos1te sample 1s one m which the quantity ofhquid sampled 1s proportional to the quantity ofhqwd waste discharged and m wluch the method ofsamplmg employed results ma specunen that IS representatrve of the hqwds released (e) lfthe alarm setpomt of the effluent morutor 1s exceeded, the frequency ofsamplmg shall be IIlCreased to daily until the condition no longer CXJSts Frequency of analysis shall be mcreased to daily for pnnc1pal gamma emrtters and an lllC1dent composite for H-3, gross alpha, Sr-89, Sr-90, and Fe-55 (f) lfthe alarm setpomt of Service Water Effluent Monitor A and/or BIS exceeded, the frequency ofsamplmg shall be increased to daily untII the conchtlon no longer CXIsts Frequency ofanalysIS shall be mcreased to daily for pnnc1pal gamma emitters and an lllCldent compoSJte for H-3, gross alpha, Sr-89, Sr-90, and Fe-55 (g) Complete pnor to each release Unit2 Revision 37 13.1-3 September 2020

Liquid Effluents Dose D 3.1.2 D 3.1 RADIOACTIVE LIQUID EFFLUENTS D 3.1.2 Liquid Effluents Dose DLCO 3.1.2 The dose or dose commitment to a MEMBER OF THE PUBLIC from radioactive materials released in liquid effluents from each unit to UNRESTRICTED AREAS (Figure D 1.0-1) shall be limited to:

a. :s; 1.5 mrem to the whole body an9 ::; 5 mrem to any organ during any calendar quarter; .and

b. ::; 3 mrem to the whole body and :s; 10 mrem to any organ during any calendar year.

APPLICABILITY: At all times.

ACTIONS

NOTES---------- *-------------------------------------------

1. LCO 3.0.3 is not applicable.

2. LCO 3.0.4 is not applicable.

CONDITION REQUIRED ACTION COMPLETION TIME A. Calculated dose to a A.1 Prepare and submit to the 30 days MEMBER OF THE NRC, pursuant to D 4.1.1, a

PUBLIC from the release Special Report that of radioactive materials in (1) Identifies the cause(s) for liquid effluents to exceeding the limit(s)

UNRESTRICTED AREAS and exceeds limits. (2) Defines the corrective actions that have been taken to reduce the .

relea&es and the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with DLCO 3.1.2.

( continued)

Unit2 Revision 37 13.1-4 September 2020

Liquid Effluents Dose D 3.1.2 v

A CTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B. Calculated dose to a B.1 Calculate the annual dose to a Immediately MEMBER OF THE MEMBER OF THE PUBLIC PUBLIC from the release which includes contributions of radioactive materials in from direct radiation from the liquid effluents exceeds 2 units (including outside times the limits. storage tanks, the on-site ISFSI, etc.).

AND Immediately B.2 Verify that the limits of DLCO 3.4 have not been exceeded.

C. Required Action B.2 and C.l Prepare and submit to the 30 days Associated Completion NRC, pursuant to D 4.1.1, a time not met Special Report, as defined in 10 CFR 20.2203 (a)(4), of Required Action A. l shall also include the foVowing:

(1) The corrective action(s) to be taken to prevent recurrence of exceeding the limits of DLCO 3.4 and the

- ~-, schedule for achieving conformance, (2) An analysis that estimates the dose to a MEMBER OF THE PUBLIC from uranium fuel cycle sources, including all effluent pathways and direct radiation, for the calendar year that includes the release(s), and (3) Describes the levels of radiation and concentrations of radioactive material involved and the cause of the exposure levels or concentrations.

Unit2 Revision 37 13.1-5 September 2020

Liquid Effluents Dose D 3.1.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY DSR 3.1.2.1 Detennine cumulative dose contributions from liquid 31 days

\

effluents for the current calendar quarter and the current calendar year.

Unit2 Revision 37 I 3.1-6 September 2020

Liquid Radwaste Treatment System D 3.1.3 D 3.1 RADIOACTIVE LIQUID EFFLUENTS D 3.1.3 Liquid Radwaste Treatment System DLCO 3.1.3 The liquid radwaste treatment system shall be FUNCTIONAL.

APPLICABILITY: At all times.

/

ACTIONS

*NOTES-----------------------------------------------------

1. LCO 3.0.3 is not applicable.

2. LCO 3.0.4 is not applicable.

CONDITION REQUIRED ACTION COMPLETION J TIME A. Radioactive liquid waste A.1 Prepare and submit to the 30 days being discharged without NRC, pursuant to D 4.1.1, a treatment. Special Report that includes:

(1) An explanation of why AND liquid radwaste was being discharged without

' Projected doses due to the treatment, identification of liquid effluent, from the any nonfunctional unit, to UNRESTRICTED equipment or subsystems, AREAS would exceed and the reason for the 0.06 mrem to the whole nonfunctionality, body or 0.2 mrem to any (2) Action(s) taken to restore organ in a 31 day period. the nonfunctional equipment to AND FUNCTIONAL status, and

, (3) Summary description of Any portion of the liquid action(s) taken to prevent a radwaste treatment system recurrence.

not in operation.

Unit2 Revision 37 13.1-7 September 2020

Liquid Radwaste Treatment System D 3.1.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY DSR 3.1.3.1 ------------------------NOTE----------------------------

Only required to be met when liquid radwaste treatment systems are not being fully utilized.

Project the doses due to liquid effluents from each 31 days unit to UNRESTRICTED AREAS.

Unit2 Revision 37 I 3.1-8 September 2020

Gaseous Effluents Dose Rate D 3.2.1 D 3.2 RADIOACTIVE GASEOUS EFFLUENTS D 3.2.1 Gaseous Effluents Dose Rate DLCO 3.2.1 The dose rate from radioactive materials released in gaseous effluents from the site to areas at or beyond the SITE BOUNDARY (Figure D 1.0-1) shall be limited to:

a. For noble gases, ~ 500 mrem/yr to the whole body and

~ 3000 mrem/yr to the skin and

b. For I-131, 1-133, H-3 and all radionuclides in particulate form with half-lives> 8 days, ~ 1500 mrem/yr to any organ.

APPLICABILITY: At all times.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. The dose rate( s) at or A.1 Restore the release rate to Immediately beyond the SITE within the limit.

BOUNDARY due to radioactive gaseous effluents exceeds limits.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY DSR 3.2.1.1 The dose rate from noble gases in gaseous effluents In accordance with shall be determined to be within the limits of DLCO Table D 3.2.1-1 3.2.1.a.

DSR 3.2.1.2 The dose rate from 1-131, 1-133, H-3 and all In accordance with Table D 3.2.1-1 radionuclides in particulate form with half-lives

> 8 days in gaseous effluents shall pe determined to be within the limits ofDLCO 3.2.1.b.

Unit2 Revision 37 I 3.2-1 September 2020

Gaseous Effluents Dose Rate D 3.2.1 Table D 3.2.1-1 (Page 1 of2)

Radioactive Gaseous Waste Sampling and Analysis SAMPLE LOWER LIMIT OF SAMPLE SAMPLE ANALYSIS SAMPLE DETECTION GASEOUS RELEASE TYPE FREQUENCY FREQUENCY ANALYSIS (LLD) (a)

TYPE I Contamment (b) Grab Sample Each Purge (h) Pnnc1pal Gamma Eillltters I x I o-4 µCt/ml (c)

Each Purge H-3 (oxide) Ix 10-6 µCt/ml Each Purge Pnnc1pal Ix 10"' µCt/ml Gamma Emitters (c) 2 Mwn Stack, Grab Sample 31 days (d) 31 days (d) Pnnc1pal 1 x I O"'µCt/ml Rad waste/Reactor Gamma Emitters BwldmgVent (c)

Grab Sample 31 days (e) 31 days (e) H-3 (oxide) Ix 10-6µCt/ml Charcoal Contmoous (f) 7 days (g) 1-131 Ix 10- 11 µCt/ml Sample Particulate Contmoous ( f) 7 days (g) Pnnc1pal Ix 10- 11 µCt/ml Sample Gamma EIIlltters (c) 31 days (g) Gross Alpha Ix 10- 11 µCt/ml Composrte Contmuous (f) 92 days Sr-89 Ix 10- 11 µCt/ml Particulate Sample Sr-90 Ix 10- 11 µCt/ml See the notes on the next page.

Unit2 Revision 37 I 3.2-2 September 2020

Gaseous Effluents Dose Rate D 3.2.1 Table D 3.2.1-1 (Page 2 of2)

Radioactive Gaseous Waste Sampling and Analysis (a) The LLD IS defined as the smallest concentration ofradioacuve matenal in a sample that WIii yield a net count, above system baclcgrmmd, that will be detected wrth 95% probability With only 5% probab11Ity of falsely concludmg that a blank observation represents a "real" signal For a paruct![,ar measurement system, wluch may include radiochenucal separation LLD (E) (V) (2.22xl0 6 ) (Y) e-Ut where LLD The before-the-fact lower limrt of detec!Ion (µC1 per umt mass or volume),

Si, 1be standard dev1abon of the backgrolllld eotmtmg rate or of the counting rate ofa blank sample as appropnate (counts per mmute ),

E The counting efficiency (counts per dISmtegra!lon),

V The sample StZC (uruts ofmas5 or volume),

222xla6 The number of d1sintegra!lons per mmute per µC1, y The fractional rad1ochermcal yield, when applicable, The rad1ooc!lve decay constant for the par!lcular radionuclide (sec*'), and The elapsed time between the m1dpomt of sample collec!ion and the !lme ofcoun!lng (seconds)

Typical values ofE, V, Y, and Af should be used m the calculatJon It should be recogruzed that the LLD 1s defined as a before-the-fact lumt represen!lng the capabil1ty of a measurement system and not as an after-the-fact limit for a par!lcular measurement (b) Sample and analysIS before PURGE IS used to determme pern11ss1ble PURGE rates Sample and analysIS dunng actual PURGE IS used for offs1te dose calculabons (c) The pnnc1pal gamma emrtters for which the LLD applies mclude the following rad1onuchdes Kr-87, Kr-88, Xe-133, Xe-133m, Xe-135, and Xe-138 m noble gas releases and Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo--99, 1-131, Cs-134, Cs-137, Ce-141 and Ce-144 m 1odme and partJculate releases 11us list does not mean -that only these nucl ides are to be considered Other gamma peaks that are identifiable, together with those of the above nuclides, shall also be analyzed and reported m the Radioactive Effluent Release Report pursuant to Techrucal Spec1ficatton 5 6 3 m the format outlmed in RG I 21, Appendix B, Rev1S1on I, June I 974 (d) If the main stack or reactor/radwaste bmlding radtanon monrtor IS not FUNCTIONAL, samplmg and analysis shall also be performed followmg shutdown, startup, or when there IS an alarm on the offgas pretreatment momtor (e) H-3 grab samples shall be taken once every 7 days from the reactor/radwaste ventilabon system when fuel IS offloaded until stable H-3 release levels can be demonstrated (f) lbe rabo of the sample flow rate to the sampled stream flow rate shall be known for the time penod covered by each dose or dose rate calculation made m accordance With DLCO 3 2 I b and DLCO 3 2 3 (g) When the release rate of the rnam stack or reactor/radwasto but Iding vent exceeds rts alarm setpomt, the 1odme and particulate device shall be removed and analyzed to determme the changes m 1ochne and particulate release rates The analysts shall be done once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months unttl the release no longer exceeds the alarm setpomt When samples collected for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months are analyzed, the correspondmg LLDs may be mcreased by a factor of I 0 (h) Complete pnor to each release Unit2 Revision 37 I 3.2-3 September 2020

Gaseous Effluents Noble Gas Dose D3.2.2 D3.2 RADIOACTIVE GASEOUS EFFLUENTS D 3.2.2 Gaseous Effluents Noble Gas Dose DLCO 3.2.2 The air dose from noble gases released in gaseous effluents from each unit to areas at or beyond the SITE BOUNDARY (Figure D 1.0-1) shall be limited to:

a. During any calendar quarter: s 5 mrad for gamma radiation and s 10 mrad for beta radiation and

b. During any calendar year: s 10 mrad for gamma radiation and s 20 mrad for beta radiation.

APPLICABILITY: At all times.

ACTIONS

NOTES- --------------------------------------------

1. LCO 3.0.3 is not applicable.

2. LCO 3.0.4 is not applicable.

CONDITION REQUIRED ACTION COMPLETION TIME A. The air dose at or beyond A.1 Prepare and submit to the 30 days the SITE BOUNDARY NRC, pursuant to D 4.1.1, a

~ue to noble gases released 7 m gaseous effluents Special Report that exceeds limits. (1) Identifies the cause(s) for exceeding the limit(s) and (2) Defines the corrective actions that have been taken to reduce the releases and the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with DLCO 3.2.2.

(continued)

Unit 2 Revision 37 I 3.2-4 September 2020

Gaseous Effluents Noble Gas Dose D 3.2.2 A CTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B. Calculated dose to a B.1 Calculate the annual dose to a Immediately MEMBER OF THE MEMBER OF THE PUBLIC PUBLIC from the release which includes contributions of radioactive materials in from direct radiation from the gaseous effiuents due to units (including outside noble gases exceeds 2 storage tanks, the on-site times the limits. ISFSI, etc.).

AND Immediately B.2 Verify that the limits of DLCO 3 .4 have not been exceeded.

C. Required Action B.2 and C.1 Special Report, as defined in 30 days Associated Completion 10 CPR 20.2203 (a)(4), of time not met. Required Action A. l shall also include the following:

(1) The corrective action(s) to be taken to prevent recurrence of exceeding the limits of DLCO 3.4 and the schedule for achieving conformance, '

(2) An analysis that estimates the dose to a MEMBER OF THE PUBLIC from uranium fuel cycle sources, including all effluent pathways and direct radiation, for the calendar year that includes the release(s), and (3) Describes the levels of radiation and concentrations of radioactive material involved and the cause of the exposure levels or -

concentrations.

Unit 2 Revision 37 I 3.2-5 September 2020

Gaseous Effluents Noble Gas Dose D 3.2.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY DSR 3.2.2.1 Determine cumulative dose contributions for the 31 days current calendar quarter and current calendar year.

Unit2 Revision 37 I 3.2-6 September 2020

Gaseous Effluents Dose -I-131, I-133, H-3 and Radioactive Material in Particulate Form D 3.2.3 D 3.2 RADIOACTIVE GASEOUS EFFLUENTS D 3.2.3 Gaseous Effluents Dose-I-131, I-133, H-3 and Radioactive Material in Particulate Form DLCO 3.2.3 The dose to a MEMBER OF THE PUBLIC from I-131, I-133, H-3, and all radioactive material in particulate form with half-lives > 8 days in gaseous effluents released, from each unit, to areas at or beyond the SITE BOUNDARY (Figure D 1.0-1) shall be limited to:

a. During any calendar quarter: :::; 7 .5 mrem to any organ and

b. During any calendar year: ~ 15 mrem to any organ.

APPLICABILITY: At all times.

ACTIONS

NOTES----------------------------------------------

1. LCO 3.0.3 is not applicable.

2. LCO 3.0.4 is not applicable.

CONDITION REQUIRED ACTION COMPLETION TIME A. The dose from 1-131, 1-133, A.I Prepare and submit to the NRC, 30 days H-3 and radioactive material pursuant to D 4.1.1, a Special in particulate form with half- Report that lives > 8 days released in (1) Identifies the cause(s) for gaseous effluents at or exceeding the Iimit(s) and beyond the SITE (2) Defines the corrective actions BOUNDARY exceeds limits. that have been taken to reduce the releases and the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with DLCO 3.2.3.

(continued)

Unit2 Revision 37 I 3.2-7 September 2020

Gaseous Effluents Dose-1-131, 1-133, H-3 and Radioactive Material in Particulate Form D 3.2.3 A CTIONS ( continued)

CONDITION REQUIRED ACTION COMPLETION TIME B.' Calculated dose to a B.l Calculate the annual dose to a Immediately MEMBER OF THE PUBLIC MEMBER OF THE PUBLIC from the release of which includes contributions radioactive materials in from direct radiation from the gaseous effluents exceeds 2 units (including outside storage times the limits. tanks, the on-site ISFSI, etc.).

AND B.2 Verify that the limits of DLCO Immediately 3 .4 have not been exceeded.

C. Required Action B.2 and C.l Special Report, as defined in 10 30 days Associated Completion time CFR 20.2203 (a)(4), of Required not met Action A. l shall also include the following:

(l)The corrective action(s) to be taken to prevent recurrence of exceeding the limits of DLCO 3 .4 and the schedule for achieving conformance, (2)An analysis that estimates the dose to a MEMBER OF THE PUBLIC from uranium fuel cycle sources, including all effluent pathways and direct radiation, for the calendar year that includes the release(s),

and (3)Describes the levels of radiation and concentrations of radioactive material involved and the cause of the exposure levels or concentrations.

Unit2 Revision 37 I 3.2-8 September 2020

Gaseous Effluents Dose - I-131, I-133, H-3 and Radioactive Material in Particulate Form D 3.2.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY DSR3.2.3.l Determine cumulative dose contributions for the 31 days current calendar quarter and current calendar year for I-131, I-133, H-3 and radioactive material in particulate form with half-lives > 8 days.

Unit2 Revision 37 I 3.2-9 September 2020

Gaseous Radwaste Treatment System D 3.2.4 D 3.2 RADIOACTIVE GASEOUS EFFLUENTS D 3.2.4 Gaseous Radwaste Treatment System DLCO 3.2.4 The GASEOUS RADWASTE TREATMENT SYSTEM shall be in operation.

APPLICABILITY: Whenever the main condenser air ejector system is in operation.

ACTIONS

NOTE----- ----------------------------------------

LCO 3.0.3 is not applicable.

CONDITION REQUIRED ACTION COMPLETION TIME A. The gaseous radwaste A.l Restore treatment of gaseous 7 days from the main condenser radwaste effluent.

air ejector system is being discharged without treatment.

B. Required Action and B.1 Prepare and submit to the NRC, 30 days associated Completion pursuant to D 4.1.1, a Special Time not met. Report that includes the following:

(1) Identification of any nonfunctional equipment or subsystems and the reason for the nonfunctionality, (2) Action(s) taken to restore the nonfunctional equipment to FUNCTIONAL status, and (3) Summary description of action(s) taken to prevent a recurrence.

Unit2 Revision 37 I 3.2-10 September 2020

Gaseous Radwaste Treatment System D 3.2.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY DSR 3.2.4.1 Check the readings of the relevant instruments to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ensure that the GASEOUS RADW ASTE TREATMENT SYSTEM is functioning. _ ,

Unit2 Revision 37 I 3.2-11 September 2020

Ventilation Exhaust Treatment System D 3.2.5 D 3.2 RADIOACTIVE GASEOUS EFFLUENTS D 3.2.5 Ventilation Exhaust Treatment System DLCO 3.2.5 The VENTILATION EXHAUST TREATMENT SYSTEM shall be FUNCTIONAL.

APPLICABILITY: At all times.

ACTIONS

--------------------------NOTES------------------ ------------------------

1. LCO 3.0.3 is not applicable.

2. LCO 3.0.4 is not applicable.

CONDITION REQUIRED ACTION COMPLETION TIME A. The radioactive gaseous A. I Prepare and submit to the 30 days waste is being discharged NRC, pursuant to D 4.1.1, a without treatment. Special Report that includes the following:

AND (I) Identification of any nonfunctional equipment or Projected doses in 31 days subsystems and the reason from iodine and particulate for the nonfunctionality, releases, from each unit, to (2) Action(s) taken to restore

,areas at or beyond the SITE the nonfunctional BOUNDARY (see Figure D equipment to 1.0-1) would exceed 0.3 FUNCTIONAL status, and mrem to any organ of a (3) Summary description of MEMBER OF THE action(s) taken to prevent a PUBLIC. recurrence.

Unit 2 Revision 3 I 3.2-12 September 2020

Ventilation Exhaust Treatment System D 3.2.5 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY DSR 3.2.5.1 ------------------------NOTE----------------------------

Only required to be met when the VENTILATION EXHAUST TREATMENT SYSTEM is not being fully utilized.

31 days Project the doses from iodine and particulate releases from each unit to areas at or beyond the SITE BOUNDARY.

Unit2 Revision 3 I 3.2-13 September 2020

Venting or Purging D 3.2.6 D 3.2 RADIOACTIVE GASEOUS EFFLUENTS D 3.2.6 Venting or Purging DLCO 3.2.6 VENTING or PURGING of the drywell and/or suppression chamber shall be through the standby gas treatment system.

APPLICABILITY: MODES 1, 2, and 3.

ACTIONS

NOTES-----------------------------------------------

1. LCO 3.0.3 is not applicable.

2. LCO 3.0.4 is not applicable.

CONDITION REQUIRED ACTION COMPLETION TIME A. VENTING or PURGING A.I Suspend all VENTING and Immediately of the drywell and/or PURGING of the drywell suppression chamber not and/or suppression chamber.

through the standby gas treatment system.

Unit2 Revision 37 I 3.2-14 September 2020

Venting or Purging D 3.2.6 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY DSR 3.2.6.1 The drywell and/or suppression chamber shall be Within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months determined to be aligned for VENTING or before start of PURGING through the standby gas treatment system. VENTING or PURGING 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter during VENTING or PURGING Unit2

\ Revision 37 I 3.2-15 September 2020

Radioactive Liquid Eflluent Monitoring Instrumentation D 3.3.1 D 3.3 INSTRUMENTATION D 3.3.1 Radioactive Liquid Effluent Monitoring Instrumentation DLCO 3.3.1 The radioactive liquid effluent monitoring instrumentation channels shown in Table D 3.3.1-1 shall be FUNCTIONAL with:

a. The minimum FUNCTIONAL channel(s) in service.

b. The alann/trip setpoints set to ensure that the limits ofDLCO 3.1.1 are not exceeded.

APPLICABILITY: According to Table D 3.3.1-1.

ACTIONS

NOTES---------------------------------------------

1. LCO 3.0.3 is not applicable.

2. Separate condition entry is allowed for each channel.

CONDITION REQUIRED ACTION COMPLETION TIME A. Liquid effluent monitoring A. I Suspend the release of Immediately instrumentation channel radioactive liquid effluents alarm/trip setpoint less monitored by the affected conservative than required. channel.

OR A.2 Declare the channel Immediately nonfunctional.

OR Immediately A.3 Change the setpoint so it is acceptably conservative.

(continued)

Unit2 Revision 37 I 3.3-1 September 2020

Radioactive Liquid Effluent Monitoring Instrumentation D 3.3.l A CTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B. One or more required B.1 Enter the Condition referenced Immediately channels nonfunctional. in Table D 3.3.1-1 for the channel.

AND B.2 Restore nonfunctional 30 days channel(s) to FUNCTIONAL status.

C. As required by Required C.l Analyze at least 2 independent Prior to initiating a Action B.1 and referenced samples in accordance with release in Table D 3.3.1-1. Table D 3.1.1-1.

AND C.2 ---------NOTE----------------

Verification Action will be

, performed by at least 2 separate technically qualified members of the facility staff.

Independently verify the Prior to initiating a release rate calculations and release discharge line valving.

D. As required by Required D.1 Collect and analyze grab 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Action B.1 and referenced samples for radioactivity at a in Table D 3.3.1-1. limit of detection of at least AND 5 x 10-7 µCi/ml.

Once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter (continued)

Unit 2 Revision 37 I 3.3-2 September 2020

Radioactive Liquid Effiuent Monitoring Instrumentation D 3.3.1 A CTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME E. As required by Required E.l --------------NOTE-----------

Action B.1 and referenced Pump performance curves in Table D 3.3.1-1. generated in place may be used to estimate flow.

Estimate the flow rate during 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months actual releases.

AND Once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months thereafter F. As required by Requireo F.l Estimate tank liquid level. Immediately Action B.1 and referenced in Table D 3.3.1-1. AND During liquid additions to the tank G. Required Action B.2 and G.l Explain in the next In accordance with associated Completion Radioactive Eftluent Release Radioactive Time not met. Report why the Effiuent Release nonfunctionality was not Report corrected in a timely manner.

H. Required Action and H.l Suspend liquid effiuent Immediately associated Completion releases monitored by the Time for Condition C, D, nonfunctional channel( s).

or E not met.

I. Required Action and I.I Suspend liquid additions to Immediately associated Completion Time the tank monitored by the for Condition F not met. nonfunctional channel(s).

Unit2 Revision 37 I 3.3-3 September 2020

Radioactive Liquid Effluent Monitoring Instrumentation D 3.3.1 SURVEILLANCE REQUIREMENTS

NOTE----------------------------------------------------

Refer to Table D 3.3.1-1 to determine which DSRs apply for each function.

SURVEILLANCE FREQUENCY DSR3.3.l.1 Perform CHANNEL CHECK. 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months DSR 3.3.1.2 Perform CHANNEL CHECK by verifying indication 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months on any of flow during periods of release. day on which continuous, periodic, or batch releases are made DSR 3.3.1.3 Perform SOURCE CHECK. Prior to release DSR 3.3.1.4 Perform SOURCE CHECK. 31 days DSR 3.3.1.5 Perform CHANNEL FUNCTIONAL TEST. The 31 days CHANNEL FUNCTIONAL TEST shall also demonstrate that automatic isolation of this pathway and control room alarm annunciation occurs if the instrument indicates measured levels above the alarm/trip setpoint; and control room alarm annunciation occurs for instrument indication levels measured above the alarm setpoint, circuit failure, instrument indicating a downscale failure, or I

instrument controls not set in operate mode.

DSR 3.3.1.6 Perform CHANNEL FUNCTIONAL TEST. 92 days (continued)

Unit2 Revision 37 I 3.3-4 September 2020

Radioactive Liquid Effluent,Monitoring Instrumentation D 3.3.l SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY DSR3.3.l.7 Perform CHANNEL FUNCTIONAL TEST. The 184 days CHANNEL FUNCTIONAL TEST shall also demonstrate control room alarm annunciation occurs for instrument indication levels measured above the alarm setpoint, circuit failure, instrument indicating a downscale failure, or instrument controls not set in operate mode.

\

DSR 3.3.1.8 Perform CHANNEL CALIBRATION. The initial 18 months CHANNEL CALIBRATION shall be performed using one or more of the reference standards certified by the National Institute of Standards and Technology (NIST), standards that are traceable to NIST standards, or using actual samples of liquid eftluents that have been analyzed on a system that has been calibrated with NIST traceable sources. These standards shall permit calibrating the system over its intended range of energy and measurement. For subsequent CHANNEL CALIBRATION, sources that have been related to the initial calibration may be used.

DSR 3.3.1.9 Perform CHANNEL CALIBRATION. 18 months Unit 2 Revision 37 I 3.3-5 September 2020

Radioactive Liquid Effluent Monitoring Instrumentation D 3.3.1 Table D 3.3.1-1 (pa$e 1 of 1)

Radioactive Liquid Effiuent Monitoring Instrumentation APPLICABILITY REQUIRED CONDITIONS OR OTHER CHANNELS REFERENCED SPECIFIED PER FROM REQUIRED SURVEILLANCE INSTRUMENT CONDITIONS INSTRUMENT ACTIONS I REQUIREMENTS Rad1oacttvrty Monrtors ProV1dmg Alarm and Automanc TennmatJ.on of Release L1q1I1d Radwaste Effluent (a) C DSR3 3 I I Lrne DSR3 3 I3 DSR3 3 15 DSR3 3 18 2 Rad1oact1V1ty Monrtors ProYldmg Alarm but not Pr0V1dmg Automatic Termlllll1lon of Release a Service Water Effluent (a) D DSR3 3 I I LmeA DSR3 3 14 DSR3 3 17 DSR3 3 I8 b Service Water Effluent (a) D DSR3 3 I I LmeB DSR3 3 14 DSR3 3 I7 DSR3 3 18 c Coolmg Tower (a) D DSR3 3 11 Blowdown Lme DSR3 3 14 DSR3 3 17 DSR3 3 18 3 Flow Rate Measurement Devices a L:timd Radwaste (a) E DSR3 3 12 E uent Lme DSR3 3 I 6 DSR3 3 1 9 b Service Water Effluent (a) E DSR3 3 12 LmeA DSR3 3 16 DSR3 3 I 9 C Service Water Effluent (a) E DSR3 3 12 LmeB DSR3 3 I 6 DSR3 3 1 9 d Coohng Tower (a) E DSR3 3 12 Blowdown Lme DSR3 3 1 6 DSR3 3 I 9 4 Tank Level Inchcatmg (b) F DSR3 3 1.1 Devices (c) DSR3 3 16 DSR3 3 19 (a) Dunng releases Via th!s pathway

~? Durmg h3u1d addition to the assoclllted tank Tanks IIlC uded m thIS DLCO are those outdoor tanks that are not surrounded~ lmers, dikes, or walls capable ofholdmg the tank contents and do not have tank overflows and surroundmg area drams connecte to the liquid radwaste treatment system, such as temporary tanks Unit2 Revision 37 I 3.3-6 September 2020

Radioactive Gaseous Effluent Monitoring Instrumentation D 3.3.2 D 3.3 INSTRUMENTATION D 3.3.2 Radioactive Gaseous Effluent Monitoring Instrumentation DLCO 3.3.2 The radioactive gaseous effluent monitoring instrumentation channels shown in Table D 3.3.2-1 shall be FUNCTIONAL with:

a. The minimum FUNCTIONAL channel(s) in service.

b. The alarm/trip setpoints of Offgas Noble Gas Activity Monitor set to ensure that the limit of Technical Specification LCO 3.7.4 is not exceeded.

C. The alarm/trip setpoints of Radwaste/Reactor Building Vent Effluent Noble Gas Activity Monitor and Main Stack Effluent Noble Gas Activity Monitor set to ensure that the limits of DLCO 3.2.1 are not exceeded.

APPLICABILITY: According to Table D 3.3.2-1.

ACTIONS

NOTES--------------------------------------------

1. LCO 3.0.3 is not applicable.

2. Separate condition entry is allowed for each channel.

CONDITION REQUIRED ACTION COMPLETION TIME A. Gaseous e,ffluent A.I Suspend the release of Immediately monitoring instrumentation radioactive gaseous effluents channel alann/trip setpoint monitored by the affected less conservative than channel.

required.

OR A.2 Declare the channel Immediately nonfunctional.

OR A.3 Change the setpoint so it is Immediately acceptably conservative.

(continued)

Unit 2 Revision 37 I 3.3-7 September 2020

Radioactive Gaseous Effluent Monitoring Instrumentation D3.3.2 A CTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B. One or more channels B.l Enter the Condition referenced Immediately nonfunctional. in Table D 3 .3 .2-1 for the channel.

AND B.2 Restore nonfunctional 30 days channel(s) to FUNCTIONAL status.

C. As required by Required C.l Place the nonfunctional 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Action B.l and referenced channel in the tripped in Table D 3.3.2-1. condition.

OR C.2.1 Take grab samples. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months AND Once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter

(

AND C.2.2 Analyze samples for gross 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months from time activity. of sampling completion (continued)

Unit2 Revision 37 I 3.3-8 September 2020

Radioactive Gaseous Effluent Monitoring Instrumentation D 3.3.2 A CTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME D. As required by Required D.l Estimate the flow rate for the 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Action B. l and referenced nmµunctional channel(s).

in Table D 3.3.2-1. AND Once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months thereafter E. As required by Required E.l Establish continuous 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months Action B. l and referenced sampling using auxiliary in Table D 3.3.2-1. sampling equtment as required in Ta le D 3.2.1-1.

F. As required by R~uired F.1.1 Take grab samples. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Action B.1 and re erenced in Table D 3.3.2-1. AND Once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter AND F.1.2 Analyze samples for gross 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months from time activity with a radioactivity of sampling limit of detection of at least completion 1 x 104 µCi/ml.

AND F.2.1 Restore the nonfunctional 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months channel(s) to FUNCTIONAL status.

OR ;,

F.2.2 Through a CR, determine: 14 days (1) The cause(s) of the nonfunctionality.

~

(2) The actions to be taken and the schedule for restoring the-system to FUNCTIONAL status.

(contmued)

Unit 2 Revision 37 I 3.3-9 September 2020

Radioactive Gaseous Effluent Monitoring Instrumentation D 3.3.2 A CTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME G. Required Action B.2 and G.l Explain in the next In accordance with associated Completion Radioactive Effluent Release Radioactive Time not met. Report why the Effluent Release nonfunctionality was not Report frequency corrected in a timely manner.

H. Required Action and H.1.1 Initiate action to implement 1 Hour associated Completion appropriate compensatory Time for Condition C, D, E actions. -

or F .1 not met.

AND 12 Hours H.1.2 Obtain Shift Manager approval of the compensatory actions and the plan for exiting this Condition H. 1 Unit2 Revision 37 I 3.3-10 September 2020

Radioactive Gaseous Effluent Monitoring Instrumentation D 3.3.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY DSR 3.3.2.1 Perform CHANNEL CHECK. 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months DSR 3.3.2.2 Perform CHANNEL CHECK. 7 days DSR3.3.2.3 Perform SOURCE CHECK. 31 days DSR 3.3.2.4 Perform CHANNEL FUNCTIONAL TEST. The 92 days CHANNEL FUNCTIONAL TEST shall also demonstrate the automatic isolation capability of this pathway and that control room alarm annunciation occurs 1f the instrument indicates measured levels above the alarm/tri1' setpoint (each channel will be tested independent y so as to not iNitiate isolation during operation); and control room alarm annunciation occurs for instrument indication levels measured above the alarm setpoint, circuit failure, instrument indicating a downscale failure, and instrument controls not set in operate mode.

DSR3.3.2.5 Perform CHANNEL FUNCTIONAL TEST. 92 days DSR 3.3.2.6 Perform CHANNEL FUNCTIONAL TEST. The 92 days CHANNEL FUNCTIONAL TEST shall also demonstrate control room alarm annunciation occurs for instrument indication levels measured above the alarm setint, circuit failure, instrument indicating a downsca e failure, and instrument controls not set m operate mode.

(continued)

Unit2 Revision 37 I 3.3-11 September 2020

Radioactive Gaseous Effluent Monitoring Instrumentation D 3.3.2 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY DSR 3.3.2.7 Perform CHANNEL CALIBRATION. The initial 24 months CHANNEL CALIBRATION shall be performed using one or more of the reference standards certified by the National Institute of Standards and Technology (NISn or using standards that have been obtained from suppliers that participate in measurement assurance activities with NIST, or using actual samples of gaseous effluents that have been analyzed on a system that has been calibrated with NIST traceable sources. These standards shall permit calibrating the system over its intended range of energy and measurement. For subsequent CHANNEL CALIBRATION, sources that have been related to the initial calibration may be used.

The CHANNEL CALIBRATION shall also demonstrate that automatic isolation of this pathway occurs when the instrument channels indicate measured levels above the Trip Setpoint.

DSR 3.3.2.8 Perform CHANNEL CALIBRATION. 18 months DSR3.3.2.9 Perform CHANNEL CALIBRATION. The initial 18 months CHANNEL CALIBRATION shall be performed using one or more of the reference standards certified by the National Institute of Standards and Technology (Nlsn or using standards that have been obtained from suppliers that participate in measurement assurance activities with NIST, or using actual samples of gaseous effluents that have been analyzed on a system that has been calibrated with NIST traceable sources. These standards shall permit calibrating the system over its intended range of energy anci measurement. For subsequent CHANNEL CALIBRATION, sources that have been related to the initial calibration may be used.

DSR 3.3.2.10 Perform CHANNEL CALIBRATION. 24 months Unit 2 Revision 37 I 3.3-12 September 2020

Radioactive Gaseous Effluent Monitoring Instrumentation D 3.3.2 Table D 3.3.2-1 (page 1 of 2)

Radioactive Gaseous Effluent Monitoring Instrumentation REQUIRED CONDITIONS APPLICABILITY OR CHANNELS REFERENCED OTIIER SPECIFIED PER FROM REQUIRED SURVEILLANCE INSTRUMENT CONDITIONS INSTRUMENT ACTIONB I REQUIREMENTS Offgas System II NobleG!ls (11) 2 C DSR3 3 2 1 Act1V1ty Monrtor DSR3 3 2 4

-Providing DSR3 3 2 7 Al11m1 and Automatic Temunanon of Release b System Flow- (11) D DSR3 3 2 1 Rate Measunng DSR3 3 2 5 Device DSR33210 2 Radwaste/Reactor Bwldrng Vent Effluent System II Noble Glls (b) F DSR3 3 2 I Activrty Monrtor DSR3 3 2 3 (c) DSR3 3 2 6 DSR3 3 2 9 b Iodme Sampler (b) E DSR3 3 2 2 C Part!culllte (b) E DSR3 3 2 2 Sampler d Flow-Rate (b) D DSR3 3 2 1 Morutor DSR3 3 2 5 DSR3 3 2 8 e Sample Flow- (b) D DSR3 3 2 I Rate Morutor DSR3 3 2 5 DSR3 3 2 8 (contmued)

(11) Dunng offgas system operation (b) At all tunes (c) Includes high range noble gas monrtonng C11pab1hty Unit 2 Revision 37 I 3.3-13 September 2020

Radioactive Gaseous Effluent Monitoring Instrumentation D 3.3.2 Table D 3.3.2-1 (page 2 of2)

Radioactive Gaseous Effluent Monitoring Instrumentation REQUIRED CONDITIONS APPLICABILITY OR CHANNELS REFERENCED OTHER SPECIFIED PER FROM REQUIRED SURVEILLANCE INSTRUMENT CONDI1lONS INSTRUMENT ACTION BI J REQUIREMENTS 3 Mam Stack Effluent a Noble Gas (b) F DSR 3 3 2 I Activity Morutor DSR3 3 2 3 (c) DSR3 3 26 DSR3 3 2 9 b Iodme Sampler (b) E DSR3 3 22 C Pllltlculate (b) E DSR3 3 22 Sampler d Flow-Rate (b) D DSR3 3 2 I Monrtor DSR3 3 2 5 DSR3 3 2 8 e Sample Flow- (b) D DSR3 3 2 I Rate Morutor DSR3 3 2 5 DSR3 3 2 8 (b) At all tunes (c) Includes !ugh range noble gas momtonng capability Unit 2 Revision 37 I 3.3-14 September 2020

.r Radioactive Effluents Total Dose D 3.4 D 3.4 RADIOACTIVE EFFLUENTS TOTAL DOSE D 3.4 Radioactive Effluents Total Dose DLCO 3.4 The annual (calendar year) dose or dose, commitment to any MEMBER OF THE PUBL1C due to releases of radioactivity and to radiation from uranium fuel cycle sources shall be limited to ~ 25 mrem to the whole body or any organ, except the thyroid, which shall be limited to ::;:; 75 mrem.

APPLICABILITY : At all times.

ACTIONS

\

-------------------- , - NOTES----------- -------------------- ----------------

1. LCO 3.0.3 is not applicable.

2. LCO 3.0.4 is not applicable.

CONDITION REQUIRED ACTION COMPLETION TIME A. Estimated dose or dose A.1 Verify the condition resulting in Immediately commitment due to direct doses exceeding these limits has radiation and the release of been corrected.

radioactive materials in liquid or gaseous effluents exceeds the limits.

B. Required Action and

)

B.1 ----------NOTE- --------

associated Completion Time This is the Special Report required not met. by D 3.1.2, D 3.2.2, or D 3.2.3-supplemented with the following.

/ -------------- ------

Submit a Special Report, 30 days pursuant to D 4.1. l, including a request for a variance in accordance with the provisions of 40 CFR 190. This submission is ,

considered a timely request, and a variance is granted until staff action on the request is complete.

Unit2 Revision 37 I 3.4-1 Septmber 2020

Radiological Environmental Monitoring Program D 3.5.1 D 3.5 RADIOLOGICAL ENVIRONMENTAL MONITORING D 3.5.1 Monitoring Program DLCO 3.5.1 The Radiolo~ical Environmental Monitoring Program shall be conducted as specified m Table D 3.5.1-1.

APPLICABILITY: At all times.

ACTIONS

NOTES-----------------------------------------

1. LCO 3.0.3 is not applicable.

2. LCO 3.0.4 is not applicable.

CONDITION REQUIRED ACTION COMPLETION TIME A. Radiological Environmental A.l Prepare and submit to the NRC In accordance with Monitoring Program not in the Annual Radiological the Annual conducted as specified in Environmental Operating Radiological Table D 3.5.1-1. Report, a description of the Environmental reasons for not conducting the Operating Report program as required and the frequency plans for preventing a recurrence.

B. Level of radioactivity in an B.1 --------NOTES--------

environmental sampling 1. Only applicable if the medium at a specified radioactivity/radionuclides are location exceeds the reporting the result of plant effluents.

levels of Table D 3.5.1-2 2. For radionuclides other than when averaged over any those in Table D 3.5.1-2, this calendar quarter. report shall indicate the methodology and parameters OR used to estimate the potential annual dose to a MEMBER OF THE PUBLIC.

( continued)

Unit2 Revision 37 13.5-1 September 2020

Radiological Environmental Monitoring Program D 3.5.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME More than one of the Prepare and submit to the NRC, 30 days radionuclides in Table pursuant to D 4.1.1, a Special D 3.5.1-2 are detected in the Report that environmental sampling (1) Identifies the cause(s) for medium and exceeding the limit(s) and (2) Defines the corrective actions Concentration 1 + to be taken to reduce reporting level 1 radioactive effluents so that the potential annual dose to a concentration 2 + ... :2: 1.0. MEMBER OF THE PUBLIC reporting level 2 is less than the calendar year limits ofD 3.1.2, D 3.2.2, or OR D 3.2.3.

Radionuclides other than OR those in Table D 3.5.1-2 are detected in an environmental B.2 -----NOTES-----

sampling medium at a 1.0nly applicable if the specified location which are radioactivity/radionuclides are the result of plant effluents not the result of plant effluents.

and the potential annual dose 2.For radionuclides other than to a MEMBER OF THE those in Table D 3.5.1-2, this PUBLIC from all report shall indicate the radionuclides is :2: the methodology and parameters calendar year limits of used to estimate the potential D 3.1.2, D 3.2.2 or D 3.2.3. annual dose to a MEMBER OF THE PUBLIC.

In accordance with Report and describe the condition th~ Annual in the Annual Radiological Radiological Environmental Operating Report Environmental Operating Report frequency (continued)

Unit 2 Revision 37 I 3.5-2 September 2020

Radiological Environmental Monitoring Program D 3.5.1 A CTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C. Mille or fresh leafy C.l Identiry specific locations for 30 days vegetation samples obtaimng recf.lacement unavailable from one or samples an add them to the more of the sample Rad1olosical Environmental locations required by Table Monitonng Program.

D 3.5.1-1.

AND C.2 Delete the specific locations 30 days from which samples were unavailable from the Radiolosical Environmental Monitonng Program.

' AND C.3 Pursuant to Technical In accordance with Specification 5.6.3, submit in the Radioactive the next Radioactive Effluent Effluent Release Release Report Report documentation for a chanJ,e in the ODCM reflecting e new location(s) with supporting information identin;infl the cause of the unavai ab1 ity of samples and justiftng the selection of the new ocation(s) for obtaining samples.

D. Environmental samples D.l Ensure all efforts are made to Prior to the end of required in Table D 3.5.1-1 complete corrective action(s). the next sampling are unobtainable due to 'period samE;ing equipment AND mal nctions.

D.2 Report all deviations from the In accordance with samplinft!chedule in the the Annual Annual diolo<Sical Radiological Environmental perating Environmental Report. Operating Report (continued)

Unit2 Revision 37 I 3.5-3 September 2020

Radiological Environmental Monitoring Program D 3.5.1 ACTIONS (continued)

I CONDITION REQUIRED ACTION COMPLETION TIME E. Samples required by Table E.1 Choose suitable alternative 30 days D 3.5.1-1 not obtained in media and locations for the the media of choice, at the pathway in question.

most desired location, or at the most desired time. AND E.2 Make appropriate 30 days substitutions in the Radiological Environmental Monitoring Program.

AND E.3 Submit in the next In accordance with Radioactive Effluent Release the Radioactive Report documentation for a Effluent Release change in the ODCM Report reflecting the new location(s) with supporting information identifying the cause of the unavailability of samples for that pathway and justifying the selection of the new location(s) for obtaining samples.

Unit2 Revision 37 I 3.5-4 September 2020

Radiological Environmental Monitoring Program D 3.5.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY DSR 3.5.1.1 Collect and analyze radiological environmental In accordance with monitoring samples pursuant to the requirements of Table D 3.5.1-1 Table D 3.5.1-1 and the detection capabilities required by Table D 3.5.1-3.

j Unit2 Revision 37 I 3.5-5 September 2020

Radiological Environmental Monitoring Profam D .5.1 Table D 3.5.1-1 ~age 1 of 4)

Radiological Environmenta Monitoring Program EXPOSURE NUMBEROF SAMPLING AND PATHWAY SAMPLE COLLECTION TYPE AND FREQUENCY AND/OR STATIONS SAMPLE FREQUENCY OF ANALYSIS SAMPLE LOCATIONS (a)

Duect 32 routme (1) An mner nng of stations, Once per 3 months Gamma dose once per!3 Radllltlon morutonng one meach months stations (b) meteorological sector m the general area of tho SITE BOUNDARY (2) An outer rmg of stallons, one rn each land base meteorological sector m the 4 to 5 nule (c) range from the s1to (3) lbe balance of the sta!tons should be placed rn special mterest areas such as popula!ton centers, nearby residences, schools, and m one or two areas to serve as control stattons (d) 2 Auborne 5 locations (I) 3 samples from offs1te Con!tnuous sampler Rad101odme caruster Rad101odme locattons close to the site operahon with Analyze weekly for I-13 I and boundary (withm I mtle) sample collec!ton Particulates m different sectors (e) weekly or more Parttculate sampler frequently tf (1) Analyze for gross beta (2) 1 sample from the vtctruty rad1oact1Y1ty ~ 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of an ostabhshed year- rcqwred by dust loading follo.wmg filter change round commumty (e) (f)

(2) Perform gamma ISOtop1c (3) I sample from a control analysIB on each sample loca!ton, at least 10 miles (g) m which gross beta dIStant and rn a least ac!Ivrty IB > IO tunes the prevalent Wind drrectton previous yearly mean of (d) control samples (3) Gamma 1sotop1c analysis of composite sample (g)

(by loca!Jon) once per 3 months 3 Waterborne a Surface 1 samplo Upstream (d) (h) Composite sample (1) Gamma 1sotop1c over a one month analysIB of each sample penod (1) (g) once per month 1 sample Site's downstream coolmg (2) H-3 analysts of each water mtake (h) composite sample and once per 3 months b Ground As requrred From one or two sources tf Grab sample once (1) Gamma 1sotop1c hkely to be affected (J) per 3 months analysIB of each sample (g) once per 3 months (2) H-3 analystS of each sample once per 3 months

( continued)

Unit2 Revision 37 I 3.5-6 September 2020

Radiological Environmental Monitoring Program 3.5.1 Table 3.5.1-1 (page 2 of 4)

Radiological Environmental Monitonng Program EXPOSURE PATHWAY SAMPLING AND AND/OR NUMBER.OF SAMPLE COLLECTION TYPE AND fREQUENCY SAMPLE SAMPLES LOCATIONS (a) FREQUENCY OF ANALYSIS 3 Waterborne (contmued)

C Dnnkmg 1 sample of each One to three of the nearest water Whon 1-13 1 ana.lySJS (1) I-131 Ellllliys1S on each supphes that could be affected 1s performed, a composite sample by its dlSCha.rge (k) composite sample when the dose over a two week calculated for the penod (1), othefW!se, consumption of the a composite sample water IS greater than 1 monthly rnrern/yr (I)

(2) Gross beta and gamma 1SOtop1c analyses of each composrte sample (g) monthly (3) H-3 Ellllliys1s of each composite sample once per 3 months d Sednnent 1 sample From a downstream area with Twice per year Gamma ISO!op1c analyslS of from existmg or potentJal recreatxmal each sample (g)

Sborelme value 4 Ingestion a Milk ( 1) 3 samples from In 3 locat10ns withrn 3 5 mLies Twice per month, (1) Gamma 1SOtop1c (g) and MILK (e) AprLI through 1~131 ana.lySJS of each SAMPLING December (m) sample twice per month LOCATIONS April through December (2) If there are In each of 3 areas 3 5-5 0 ITilles (2) Gamma 1sotop1c (g) and none, then I dlStant(e) 1-131 analysis of each sample from sample once per month MILK January through March SAMPLING ifreqwred LOCATIONS (3) I sample from a At a control locatJon 9-20 miles MILK dlS!a.nt and m a least prevalent SAMPLING wind drrectJon (d)

LOCATION b F!Sh ( 1) I sample each In the vicmrty of a plant Twice per year Gamma 1sotop1c analysis of of 2 discharge area each sample (g) on edible comrnerc1ally portJons twice per year or recreatJonally IIllportant species (n)

(2) 1 sample of the In areas oot m.fl uenced by same species stBtlon d1SChargo (d)

(continued)

Unit 2 Revision 37 13.5-7 September 2020

Radiological Environmental Monitoring Program D 3.5.1 Table D 3.5.1-1 (page 3 of 4)

Radiological Environmental Monitoring Program EXPOSURE PATIIWAY SAMPLING AND TYPE AND FREQUENCY AND/OR NUMBEROF SAMPLE COLLECTION OF ANALYSIS SAMPLE SAMPLES LOCATIONS (a) FREQUENCY 4 Ingestion (contmued)

C Food (1) I sample of Any area that IS 1IT1gated by At tune of harvest Gamma 1SOtop1c (g) and 1-131 Products each pnnc1pal water m wtuch hqwd plant (p) analysIS of each sample of class offood wastes have been dISCharged (o) edible port10ns products (2) Samples of 3 Grown nearest to each of2 Once per year dunng d!fferent lands drlferent offs1te locations (e) the harvest season-of broad leaf vegetabon (such as vegetables)

(3) I sample of Grown at least 9 3 n:ules d1Stant Once per year dunng each of the ma least prevalent wmd the harvest season sumlar broad drrect!on leaf vegelat!on Unit2 Revision 37 I 3.5-8 September 2020

Radiological Environmental Monitoring Program D 3.5.1 Table D 3.5.1-1 (page 4 of 4)

Radiological Environmental Monitoring Program (a) Specific parameters of dlStance and dll"CCt!on sector from the ccnterlme of ono reactor, and add1t1onal descnptlons where pertinent, shall be provided for each and every sample location m Table D 3 5 1-1 Refer to NUREG-0133, "Preparation of Rad1ological Effluent Technical Speclficat!ons for Nuclear Power Plants," October 1978, and to Radiological Assessment Branch Techrucal Posrtlon on Env1ronmental Morutormg. RevJs1on I, November 1979 Devwt10ns are perrmtted from the reqwred sampling schedule 1f spec!Illens are unobtamablo because of such circumstances as hazardous condrtJons, seasonal unavailability (wluch includes theft and uncooperative residents), or malfunction of automatic sampling eqwpment (b) One or moro mstrurnents, such as a pressunzed 10n chamber, for rneasunng and recording dose rate contmuously may be used 111 place of: or m addrtlon to rntegratmg doSIIlleters Each of the 32 routrne momtonng stations shall be equipped wrth 2 or more dos1D1eters or wrth I rnstnnnent for measunng and recordmg dose rate continuously For the purpose oftlus table, a therrnolurnmescent dOS1IT1eter (1LD) IS considered to be one phosphor, 2 or more phosphors ma packet are C011S1dered as 2 or more dosuneters Ftlrn badges shall not be used as dosrrneters for measunng drrect radiation (c) At tlus dLSt!lnce, 8 wmdrose sectors (W, WNW, NW, NNW, N, NNE, NE, and ENE) are over Lake Ontario (d) Tho purpose of these samples IS to obtain backgrotmd information Ifrt IS not practical to establ1Sh control locatJons m accordance with the distance and wind drrection cntena, other sites, wluch provtde vahd background data, may be substituted (e) Havmg the Jughest calculated annual srte average ground-level D/Q based on all srte licensed reactors (f) Airborne particulate sample filters shall be analyzed for gross beta actrvrty 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or more after sampling to allow for radon and thoron daughter decay (g) Gamma 1SOtop1c analys!S means the 1dentificabon and quantification of gamma ---ern1tting radlonucl!des that may be attnbutable to the effluents from the facility (h) The upstream samplo shall be taken at a distance beyond slgJllficant mfl uonce of the d1scharge The downstream sam pie shall be taken m an area beyond but near the mlXlng zone '

(1) In tlus program, representatrve composite sample aliquots shall be collected at tune mtervals that are very short (e g, hourly) relatJve to the composrtmg penod (e g, monthly) m order to assure obtmrung a representative sample (J) Groundwater samples shall be taken when tlus source 1S tapped for dnnlang or 1IT1gatlon purposes m areas where the hydrauhc gradient or recharge properties are smtable for contammatlon (k) Dnnkmg water samples shall be taken only when dnnkmg water 1S a close pathway (1) Analys1S for 1-131 may be accomphshed by Ge-L1 analyslS provided that the lower hm1t of detection (LLD) for 1-13 I m water samples found on Table D 3 5 1-2 can be met Doses shall be calculated for the I11llX1IllU!Il organ and age group (m) Samples will be collected January through March tfl-13 l 1S detected rn November and December of the precedmg year (n) In the ovent 2 commerc1ally or recreationally llllportant species are not available, after 3 attempts of collection, then 2 samples of one spec!CS or other species not necessanly commerc1ally or recreationally important may be uttlIZed (o) Apphcable only to maJor IITlgatlon projects within 9 miles of the srte rn the general downcurrent drrection (p) If harvest occurs more than once/year, sampling shall be performed dunng each dtsCrete harvest Ifharvost occurs contrnuously, samphng shall be taken monthly Attention should be prud to mcludmg samples of tuberous and root food products Unit2 Revision 37 I 3.5-9 September 2020

Radiological Environmental Monitoring Program D 3.5.1 Table D 3.5.1-2 (page 1 of 1)

Reporting Levels for Radioactivity in Environmental Samples AIRBORNE FOOD RADIONUCLIDE PARTIUCLATE OR FISH MILK PRODUCTS ANALYSIS WATER (pCI/L) GASES (pCJ/rn3) (pCI/kg, wet) (pCI/L) (pCJ/kg, wet)

H-3 20,000 (a)

Mn-54 1,000 30,000 Fe-59 400 10,000 Co-58 1,000 30,000 Co--00 300 10,000 Zn-65 300 20,000 Zr-95 400 Nb-95 400 1-131 2 (b) 09 3 100 Cs-134 30 10 1,000 60 1,000 Cs-137 50 20 2,000 70 2,000 Ba-140 200 300 La-140 200 300 (a) For dnnkmg water samples This 1s a 40 CFR 141 value lfno dnnlang water pathway eXISts, a value of30,000 pCI/L may be used (b) Ifno dnnkmg water pathway eXJSts, a value of20 pCI/L may be used Unit 2 Revision 37 I 3.5-10 September 2020

Radiological Environmental Monitoring Program D 3.5.1 Table D 3.5.1-3 (page 1 of 2)

Detection Capabilities for Environmental Sample Analysis Ca) Cb)

LOWER LIMIT OF DETECTION (LLD) c,J AIRBORNE PARTIUCLA TE OR FOOD RADIONUCLIDE WATER GASES (pCilm.3) FISH MILK PRODUCTS SEDIMENT ANALYSIS (pCu'L) (pCJ/kg, wet) (pCu'L) (pCI/kg, wet) (pCl/lcg, dry)

Gross Bota 4 0 01 H-3 2,000 (d)

Mn-54 15 130 Fe-59 30 260 Co-58 15 130 Co-60 15 130 Zlr65 30 260 Zr-95 15 Nb-95 15 I-131 1 <*> 0 07 60 Cs-134 15 0 05 130 15 60 150 Cs-137 18 006 150 18 80 180 Ba-140 15 15 La-140 15 15 See the notes on the next page Unit 2 Revision 37 I 3.5-11 September 2020

Radiological Environmental Monitoring Program D 3.5.1 Table 3.5.1-3 (page 2 of2)

Detection Capabilities for Environmental Sample Analysis (a) (b)

(a) 11us hst does not mean that only these nuchdes are to be considered Other peaks that are identifiable, together with those of the above nuchdes, shall also be analyzed and reported m the Annual Rad10logical Envrronrnental Operatmg Report (b) Requrred detectJon capab1lrtJes for thennolwrunescent dosrrneters used for environmental measurements are given m ANSI N-545, Sect.ton 4 3 1975 Allowable except10ns to ANSI N-545, Section 4 3 are contained m the ODCM (c) The LLD IS defined as the smallest concentranon ofrad1oac1:Jve matenal ma sample that wtll yield a net count, above system background. that will be detected with 95% probabtlrty with only 5% probab1hty of falsely concludmg that a blank observanon represents a "real" signal For a particular measurement system, wluch may mclude radtochem1cal separanon LLD (E) (V) (2.22) (Y) e-ui where LLD The before-the-fact lower hmrt of detection (pC1 per unrt mass or voltnne),

The standard dCVJatJon of the background counting rate or of the cotmtmg rate ofa blank sample as appropnate (counts per mmute),

E The countmg efficiency (counts per d1Smtegrailon),

V The sample stze (wuts of mass or volume),

222 The number of d1Smtegratlons per mmute per pC1, y The fractional radiochem1cal yield, when apphcable,

,_ The radl08Ctrve decay constant for the particular rad1onuchde (sec*'), and The elapsed tune between envrromnental collectlon or end of the sample collectlon penod, and the tune of countmg (seconds)

Typical values of E, V, Y, and "1t should be used m the calculatlon It should be recogruzed that the LLD 1s defined as a before-the-fact hm1t represerrtlng the capab1lrty of a measurement system and not as an after-the-fact hmrt for a particular measurement Analyses shall be performed m such a manner that the stated LLDs wtll be achieved under routme condrt!ons Occas10nally background fluctuations, unavoidable small sample stzeS, the presence of mterfenng nuchdes, or other uncontrollable circumstances may render these LLDs unachievable In such cases, the contnbutmg factors shall be 1dentrfied and descnbed m the Annual Radiological Envrronmental Operating Report (d) lfno dnnk:mg water pathway CXJsts, a value of3,000 pCt/L may be used (e) Ifno dnnkmg water pathway exists, a value of 15 pCt/L may be used Unit2 Revision 37 I 3.5-12 September 2020

Land Use Census D 3.5.2 D 3.5 RADIOLOGICAL ENVIRONMENTAL MONITORING D 3.5.2 Land Use Census DLCO 3.5.2 A land use census shall:

a Be conducted,

b. Identify within a distance of 5 miles the location in each of the 16 meteorologi,cal sectors of the nearest milk animal and the nearest residence, and the nearest garden (broad leaf vegetation sampling controlled by Table D 3.5.1-1, part 5.c may be performed in lieu of the garden census) of> 500 ft2 producing broad leaf vegetation, and

c. For elevated releases identify within a distance of 3 miles the locations in each of the 16 meteorological sectors of all milk animals and all gardens (broad leaf vegetation sampling controlled by Table D 3.5.1-1, part 5.c may be performed in lieu of the garden census)> 500 ft2 producing broad leaf vegetation.

APPLICABILITY: At all times.

ACTIONS

NOTES----------------------------------------------

1. LCO 3.0.3 is not applicable.

2. LCO 3.0.4 is not applicable.

CONDITION REQUIRED ACTION COMPLETION TIME A. Land use census identifies A.I Identify the new location(s) in In accordance with location(s) that yields a the next Radioactive Eftluent the Radioactive calculated dose, dose Release Report. Effiuent Release commitment, or D/Q value Report

> than the values currently being calculated in DSR 3.2.3.1.

(continued)

Unit2 Revision 37 I 3.5-13 September 2020

Land Use Census D3.5.2 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B. Land use census identifies B.1 Add the new location(s) to the 30 days location(s) that yields a Radiological Environmental calculated dose, dose Monitoring Program.

commitment, or D/Q value (via the same exposure pathway) 50% > than at a location from which B.2 Delete the sampling After October 31 of samples are currently being location(s), excluding the the year in which obtained in accordance control station location, the land use census with Table D 3.5.1-1. having the lowest calculated was conducted Jdose, dose commitment(s) or D/Q value, via the same exposure pathway, from the Radiological Environmental Monitoring Program.

AND B.3 Submit in the next In accordance with Radioactive Effluent Release the Radioactive Report documentation for a Effluent Release change in the ODCM Report incluqi_ng revised figure(s) and table(s) for* the ODCM reflecting the new location( s) with information supporting the change in sampling locations.

Unit2 Revision 37 I 3.5-14 September 2020

Land Use Census D 3.5.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY DSR 3.5.2.1 Conduct the land use census during the growing 366 days season using that information that will provide the best results, such as by a door-to-door survey, aerial survey, or by consulting local agriculture authorities.

DSR 3.5.2.2 Report the results of the land use census in the Annual In accordance with l Radiological Environmental Operating Report. the Annual Radiological Environmental Operating Report Unit2 Revision 37 I 3.5-15 September 2020

Interlaboratory Comparison Program D 3.5.3 D 3.5 RADIOLOGICAL ENVIRONMENTAL MONITORING D 3.5.3 Interlaboratory Comparison Program DLCO 3.5.3 The Interlaboratory Comparison Program shall be described in the ODCM.

Analyses shall be performed on all radioactive materials, supplied as part of an Interlaboratory Comparison Program that has been approved by the NRC, that correspond to samples required by Table D 3.5.1-1.

Participation in this program shall include media for which environmental samples are routinely collected and for which intercomparison samples are available.

APPLICABILITY: At all times.

ACTIONS

NOTES-----------------------------------------------

1. LCO 3.0.3 is not applicable.

2. LCO 3.0.4 is not applicable.

CONDITION REQUIRED ACTION COMPLETION TIME A.. Analyses not performed as A.1 Report the corrective actions In accordance with required. taken to prevent a recurrence the Annual to the NRC in the Annual Radiological Radiological Environmental Environmental Operating Report. Operating Report Unit2 Revision 37 I 3.5-16 September 2020

Interlaboratory Comparison Program D 3.5.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY DSR 3.5.3.1 Report a summary of the results obtained as part of In accordance with the Interlaboratory Comparison Program in the the Annual Annual Radiological Environmental Operating Radiological Report. Environmental Operating Report Unit2 Revision 37 I 3.5-17 September 2020

PART I - RADIOLOGICAL EFFLUENT CONTROLS BASES Unit2 Revision 37 I B 3.1-0 September 2020

Liquid Effluents Concentration B 3.1.1 B 3.1 RADIOACTIVE LIQUID EFFLUENTS B 3.1.1 Liquid Effluents Concentration BASES This is provided to ensure that the concentration of radioactive materials released in liquid waste effluents to UNRES1RICTED AREAS will be less than ten times the concentration levels specified in 10 CPR 20, Appendix B, Table 2, Column 2. This limitation provides additional assurance that the levels of radioactive materials in bodies of water in UNRES1RICTED AREAS will result in exposures within: (1) the Section II.A design objectives of Appendix I to 10 CPR 50, to a MEMBER OF THE PUBLIC and (2) the levels required by 10 CPR 20.1301(e) to the population. The concentration limit for dissolved or entrained noble gases is based upon the assumption that Xe-135 is the controlling radioisotope and its effluent concentration in air (submersion) was converted to an equivalent concentration in water using the methods described in International Commission on Radiological Protection (ICRP) Publication 2.

This applies to the release of radioactive materials in liquid effluents from all units at the site.

The required detection capabilities for radioactive materials in liquid waste samples are tabulated in terms of the lower limits of detection (LLDs). Detailed discussion of the LLD, and other detection limits can be found in L.A. Currie, "Lower Limit of Detection: Definition and

Elaboration of a Proposed Position for Radiological Effluent and Environmental Measurements,"

NUREG/CR-4007 (September 1984), and in the HASL Procedures Manual, HASL-300 (revised annually).

Unit2 Revision 37 I B 3.1-1 September 2020

Liquid Effluents Dose B 3.1.2 B 3.1 RADIOACTIVE LIQUID EFFLUENTS B 3.1.2 Liquid Effluents Dose BASES This is provided to implement the requirements of Sections II.A, ill.A, and IV.A of Appendix I to 10 CFR 50. This implements the guides set forth in Section II.A of Appendix I. The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive materials in liquid effluents to UNRESTRICTED AREAS will be kept as low as is reasonably achievable. Also, for fresh water sites with drinking water supplies that 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 potable drinking water that are in excess of the requirements of 40 CFR 141. For sites containing up to four 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 twice the dose design objectives of Appendix I, and if direct radiation doses from the units including the on-site Independent Spent Fuel Storage Installation (ISFSI), and the outside storage tanks, etc., are kept small. The Special Report will describe a course of action that should result in the limitation of the annual dose to al MEMBER OF THE PUBLIC to within the 40 CFR 190 limits. For the purposes of the Special Report, it may be assumed that the dose commitment to the MEMBERS 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. The dose calculation methodology and parameters implement the requirements in Section III.A of App__~ndix I that conformance with the guides of Appendix I be shown by Calculational procedures based on models and data, so that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated. The equations specified for calculating the doses that result from actual release rates of radioactive ipaterial in liquid effluents are consistent with the methodology provided in RG 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 R.G. 1.113, "Estimating Aqu~tic Dispersion of Effluents from Accidental and Routine Reactor Releases for the Purpose of Implementing Appendix I," April 1977. This applies to the release of radioactive materials in liquid effiuents from each unit at the site. For units with shared radwaste treatment systems, the liquid effluents from the shared system are to be proportioned among the units sharing that system.

Unit2 Revision 37 I B 3.1-2 September 2020

Liquid Radwaste Treatment System B 3.1.3 B 3.1 RADIOACTIVE LIQUID EFFLUENTS B 3.1.3 Liquid Radwaste Treatment System BASES The installed liquid radwaste treatment system shall be considered FUNCTIONAL by meeting DLCO 3.1.1 and DLCO 3.1.2. The FUNCTIONALITY of the liquid radwaste treatment system ensures that this system will be available for use whenever liquid effluents require treatment before release to the environment The requirement that the appropriate portions of this system be used when specified provides assurance that the releases of radioactive materials in liquid effluents will be kept as low as is reasonably achievable. This implements the requirements of 10 CFR 50.36a, GDC 60 of Appendix A to 10 CFR 50 and the design objective given in Section II.D of Appendix I to 10 CFR 50. The specified limits governing the use of appropriate portions of the liquid radwaste treatment system were specified as a suitable fraction of the dose design objectives set forth in Section II.A of Appendix I to 10 CFR 50 for liquid effluents. This applies to the release of radioactive materials in liquid effluents from each unit at the site. For units with shared radwaste treatment systems, the liquid effluents from the shared system are to be proportioned among the units sharing that system.

Unit2 Revision 37 IB3.l-3 September 2020

Gaseous Effluents Dose Rate B 3.2.1 B 3.2 RADIOACTIVE GASEOUS EFFLUENTS B 3.2.1 Gaseous Effluents Dose Rate BASES This is provided to ensure that the dose rate at any time at and beyond the SITE BOUNDARY from gaseous effluents from all units on the site will be within the annual dose limits of 10 CFR 20 to UNRESTRICTED AREAS.

The annual dose limits are the doses associated with the concentrations of 10 CFR 20, Appendix B, Table 2, Column 1. These limits provide 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 within or outside the SITE BOUNDARY, to annual average concentrations exceeding the limits specified in Appendix B, Table 2 of 10 CFR 20 or as governed by 10 CFR20.1302(c). For MEMBERS OF THE PUBLIC who may at times be within the SITE BOUNDARY, the occupancy of that MEMBER OF THE PUBLIC will usually be sufficiently low to compensate for any increase in the atmospheric diffusion factor above that for the SITE BOUNDARY. Examples of calculations for such MEMBERS OF THE PUBLIC, with the appropriate occupancy factors, shall be given in Part II. The specified release rate limits restrict, at all times, the corresponding gamma and beta dose rates above background to a MEMBER OF THE PUBLIC at or beyond the SITE BOUNDARY to less than or equal to 500 mrem/year to the whole body or to less than or equal to 3000 mrem/year to the skin. These release rate limits also restrict, at all times, the corresponding thyroid dose rate above background to a child via the inhalation pathway to less than or equal to 1500 mrem/year. This applies to the release of radioactive materials in gaseous effluents from all units at the site.

The required detection capabilities for radioactive materials in gaseous waste samples are tabulated in terms of the lower limits of detection (LLDs). Detailed discussion of the LLD, and other detection limits can be found in L. A. Currie, "Lower Limit of Detection: Definition and Elaboratioi;i of a Proposed Position for Radiological Effluent and Environments Measurements,"

NUREG/CR-4007 (September 1984), and in the HASL Procedures Manual, HASL-300 (revised annually).

Unit2 Revision 37 I B 3.2-1 September 2020

Gaseous Effluents Noble Gas Dose B 3.2.2 B 3.2 RADIOACTIVE GASEOUS EFFLUENTS B 3.2.2 Gaseous Effluents Noble Gas Dose BASES This is provided to implement the requirements of Section II.B, III.A, and IV.A of Appendix I to IO CFR 50. The DLCO implements the guides set forth in Section ll.B of Appendix I. The REQUIRED ACTIONS provide the required operating flexibility and, at the same time, implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive material in 'gaseous effluents to UNRESTRICTED AREAS will be kept as low as is reasonably achievable. The Surveillance Requirements implement the requirements in Section III.A of Appendix I that conformance with the guidelines of Appendix I be shown by calculational procedures based on models and data so that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated.

For sites containing up to four 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 twice the dose design objectives of Appendix I, and if direct radiation doses from the units including the on-site ISFSI, and the outside storage tanks, etc., are kept small. The Special Report will describe a course of action that should result in the limitation of the annual dose to a MEMBER OF THE PUBLIC to within the 40 CFR 190 limits. For the purposes of the Special Report, it may be assumed that the dose commitment to the 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. The dose calculation methodology and parameters for calculating the doses from the actual release rates of radioactive noble gases in gaseous effluents are consistent with the methodology provided in RG 1.109, "Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with IO CFR Part 50, Appendix I," Revision 1, October 1977, and RG 1.111, "Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water Cooled Reactors," Revision I," July 1977. The ODCM equations provided for determining the air doses at or beyond the SITE BOUNDARY are based upon real-time meteorological conditions or the historical average atmospheric condi~ions. This applies to the release of radioactive material in gaseous effluents from each unit at the site.

Unit 2 Revision 37 I B 3.2-2 September 2020

Gaseous Effluents Dose - Iodine-131, Iodine-133, Tritium, and Radioactive Material in Particulate Form B 3.2.3 B 3.2 RADIOACTIVE GASEOUS EFFLUENTS B 3.2.3 Gaseous Effluents Dose - Iodine-131, Iodine-133, Tritium, and Radioactive Material in Particulate Form BASES This is provided to implement the requirements of Sections II.C, III.A, and IV .A of Appendix I to 10 CFR 50. The DLCO implements the guides set forth in Section II.C of Appendix I. The REQUIRED ACTIONS provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive materials in gaseous effluents to UNRESTRICTED AREAS will be kept as low as is reasonably achievable. The calculational methods specified in the Surveillance Requirements implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data, so that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated. For sites containing up to four 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 twice the dose design objectives of Appendix I, and if direct radiation doses from the units including the on-site ISFSI, and the outside storage tanks, etc., are kept small. The Special Report will describe a course of action that should result in the limitation of the annual dose to a MEMBER OF THE PUBLIC to within the 40 CFR 190 limits.

For the purposes of the Special Report, it may be assumed that the dose commitment to the 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. The calculational methodology and parameters for calculating the doses from the actual release rates of the subject materials are consistent with the methodology provided in RG 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, 11 Revision 1, October 1977, and RG 1.111, "Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors, 11 Revision 1, July 1977. These equations also provide for determining the actual doses based upon the historical average atmospheric conditions. The release rate DLCO for iodine-131, iodine-133, tritium, and radioactive material in particulate form with half-lives greater than 8 days are dependent upon the existing radionuclide pathways to man, in the areas at or beyond the SITE BOUNDARY. The pathways that were examined in the development of these calculations were: (1) individual inhalation of airborne radioactive material, (2) deposition of radioactive material onto green leafy vegetation with subsequent consumption by man, (3)

Unit2 Revision 37 I B 3.2-3 September* 2020

Gaseous Effluents Dose -Iodine-131, Iodine-133, Tritium, and Radioactive Material in Particulate Form B 3.2.3 B 3.2.3 Gaseous Effluents Dose - Iodine-I 31, Iodine-I 33, Tritium, and Radioactive Material In Particulate Form (continued) deposition onto grassy areas where milk-producing animals and meat-producing animals graze (human consumption of the milk and meat is assumed), and (4) deposition on the ground with subsequent exposure to man. This applies to the release of radioactive materials in gaseous effluents from each unit at the site. For units with shared radwaste treatment systems, the gaseous effluents from the shared system are proportioned among the units sharing that system.

Unit2 Revision 37 I B 3.2-4 September 2020

Gaseous Radwaste Treatment System B 3.2.4 B 3.2 RADIOACTIVE GASEOUS EFFLUENTS B 3.2.4 Gaseous Radwaste Treatment System BASES The FUNCTIONALITY of the GASEOUS RADWASTE TREATMENT SYSTEM ensures that the system will be available for use whenever gaseous effluents require treatment before release to the environment The requirement that the appropriate portions of this system 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 implements the requirements of 1d CFR 50.36a, GDC 60 of Appendix A to 10 CFR 50, and the design objectives given in Section II.D of Appendix I to 10 CFR 50. Limits governing the use of appropriate portions of the system were specified as a suitable fraction of the dose design objectives set forth in Sections II.B and II.C of Appendix I to 10 CFR 50, for gaseous effluents. This applies to the release of radioactive materials in gaseous effluents from each unit at the site. For units with shared radwaste treatment systems, the gaseous effluents from the shared system are proportional among the units sharing that system.

Unit2 Revision 37 I B 3.2-5 September 2020

Ventilation Exhaust Treatment System B 3.2.5 B 3.2 RADIOACTIVE GASEOUS EFFLUENTS B 3.2.5 Ventilation Exhaust Treatment System BASES The FUNCTIONALITY of the VENTILATION EXHAUST TREATMENT SYSTEM ensures that the system will be available for use whenever gaseous effluents require treatment before release to the environment. The requirement that the appropriate portions of this system 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 implements the requirements of 10 CPR 50.36a, QDC 60 of Appendix A to 10 CFR 50, and the design objectives given in Section II.D of Appendix I to 10 CFR 50. Limits governing the use of appropriate portions of the system were specified as a suitable fraction of the dose design objectives set forth in Sections II.B and II.C of Appendix I to 10 CFR 50, for gaseous effluents. This applies to the release of radioactive materials in gaseous effluents from each unit at the site. For units with shared radwaste treatment systemt the gaseous effluents from the shared system are proportional among the units sharing that system.

The appropriate components, which affect iodine or particulate release, to be FUNCTIONAL are:

1) HEPA Filter - Radwaste Decon Area

2) HEPA Filter - Radwaste Equipment Area

3) HEPA Filter-Radwaste General Area Whenever one of these filters is not FUNCTIONAL, iodine and particulate dose projections will,:

be made for 31-day intervals starting with filter nonfunctionality, and continuing as long as the filter remains nonfunctional, in accordance with DSR 3 .2.5 .1.

Unit 2 Revision 37 I B 3.2-6 September 2020

Venting or Purging B 3.2.6 B 3.2 RADIOACTIVE GASEOUS EFFLUENTS B 3.2.6 Venting or Purging BASES This provides reasonable assurance that releases from drywell and/or suppression chamber purging operations will not exceed the annual dose limits of 10 CFR 20 for unrestricted areas.

Unit2 Revision 37 I B 3.2-7 September 2020

Radioactive Liquid Effluent Monitoring Instrumentation B 3.3.1 B 3.3 INSTRUMENTATION B 3.3.1 Radioactive Liquid Effluent Monitoring Instrumentation BASES 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 of liquid effluents. The alarm/trip setpoints for these instruments shall be calculated and adjusted in accordance with the methodology and parameters in Part II to ensure that the alarm/trip will occur before exceeding ten times the limits of 10 CFR 20. The FUNCTIONALITY and use of this instrumentation is consistent with the requirements of GDC 60, 63, and 64 of Appendix A to 10 CPR 50. The purpose of tank level indicating devices is to assure the detection and control of leaks that if not controlled could potentially result in the transport of radioactive materials to UNRESTRICTED AREAS.

Tanks included are those outdoor tanks that are not surrounded by liners, dikes, or walls capable of holding the tank contents and do not have tank overflows and surrounding area drains connected to the liquid radwaste treatment system, such as temporary tanks.

Unit2 Revision 37 I B 3.3-1 September 2020

Radioactive Gaseous Effluent Monitoring Instrumentation B 3.3.2 B 3.3 INSTRUMENTATION B 3.3.2 Radioactive Gaseous Effluent Monitoring Instrumentation BASES The radioactive gaseous effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in gaseous effluents during actual or potential releases of gaseous effluents. The alarm/trip setpoints for these instruments shall be calculated and adjusted in accordance with the methodology and parameters in Part II to ensure that the alarm/trip will occur before exceeding the limits of 10 CFR 20. Although the Offgas System Noble Gas Activity Monitor is listed in Table D 3.3.2-1, "Radioactive Gaseous Effluent Monitoring Instrumentation", these monitors are actually located upstream of the Main Stack noble gas activity monitor and are not effluent monitors. They were included in Table D 3.3.2-1 in accordance with NUREG-0473. As such, Offgas System Noble Gas Activity Monitor alarm and trip setpoints are not based on 10CFR20. The offgas system noble gas monitor alert setpoint is set at 1.5 times nominal full power background to assure compliance with ITS SR 3. 7.4.1 which requires offgas sampling be performed within four hours of a 500/4 increase in offgas monitoring readings, and to support MSLRM trip removal. The offgas system noble gas monitor trip setpoint is based on the 10CFRl 00 limits for the limiting design basis gaseous waste system accident which is the offgas system rupture. The range of the noble gas channels of the main stack and radwaste/reactor building vent effluent monitors is sufficiently large to envelope both normal and accident levels of noble gas activity. The capabilities of these instruments are consistent with the recommendations of Regulatory Guide 1.97, "Instrumentation for Light Water Cooled Nuclear Power Plants to Assess Plant Conditions During and Following an Accident," December 1980 and NUREG-0737, "Clarification of the TMI Action Plan Requirements," November 1980. This instrumentation also includes provisions for monitoring and controlling the concentrations of potentially explosive gas mixtures in the offgas system.

The FUNCTIONALITY and use of this instrumentation is consistent with the requirements of GDC 60, 63, and 64 of Appendix A to 10 CFR 50.

Unit 2 Revision 37 I B 3.3-2 September 2020

Radioactive Effluents Total Dose B 3.4 B 3.4 RADIOACTIVE EFFLUENTS TOTAL DOSE BASES This is provided to meet the dose limitations of 40 CFR 190 that have been incorporated into 10 CFR 20 by 46 FR 18525. This requires the preparation and submittal of a Special Report whenever the calculated doses from releases of radioactivity and from radiation from uranium fuel cycle sources exceed 25 mrem to the whole body or any organ, except the thyroid (which shall be limited to less than or equal to 75 mrem). 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 provisions of 40 CFR 190.11 and 10 CFR 20.405c, is considered to be a timely request and fulfills the requirements of 40 CFR 190 until NRC staff action is completed. The variance only relates to the limits of 40 CFR 190, and does not apply in any way to the other requirements for dose limitation of JO CFR 20, as addressed in 3.1.1 and 3.2.1. An individual is not considered a MEMBER OF THE PUBLIC during any period in which the individual is engaged in carrying out any operation that is part of the nuclear fuel cycle.

Unit2 Revision 37 I B 3.4-1 September 2020

Monitoring Program B 3.5.1 B 3.5 RADIOLOGICAL ENVIRONMENTAL MONITORING I B 3.5.1 Monitoring Program BASES The Radiological Environmental Monitoring Program provides representative measurements of radiation and of radioactive materials in those exposure pathways and for those radionuclides that lead to the highest potential radiation exposure of"MEMBERS OF THE PUBLIC resulting from the plant operation. This monitoring program implementsSection IV.B.2 of Appendix I to 10 CFR 50 and thereby supplements the Radiological Effluent Monitoring Program by verifying that the measurable concentrations of radioactive materials and levels of radiation are not higher than expected on the basis of the effluent measurements and the modeling of the environmental exposure pathways. Guidance for this monitoring program is provided by the Radiological Assessment Branch Technical Position on Environmental Monitoring, Revision 1, November 1979. Program changes may be initiated based on operational experience.

The required detection capabilities for environmental sample analyses are tabulated in terms of the lower limits of detection (LLDs). The LLDs required by Table D 3.5.1-3 are considered optimum for routine environmental measurements in industrial laboratories. It should be recognized that the LLD is defined as a before-the-fact limit representing the capability of a measurement system and not as an after-the-fact limit for a particular measurement.

Detailed discussion of the LLD, and other detection limits, can be found in L. A. Currie, "Lower Limit of Detection: Definition and Elaboration of a Proposed Position for Radiological Effluent and Environmental Measurements," NUREG/CR-4007 (September 1984),

and in the HASL Procedures Manual, HASL-300 (revised annually).

Unit2 Revision 37 I B 3.5-1 September 2020

Land Use Census B 3.5.2 B 3.5 RADIOLOGICAL ENVIRONMENTAL MONITORING B 3.5.2 Land Use Census BASES This is provided to ensure that changes in the use of areas at or beyond the SITE BOUNDARY are identified and that modifications to the Radiological Environmental Monitoring Program are made if required by the results of this census. The best information, such as from a door-to-door survey, from an aerial survey, or from consulting with local agricultural authorities, shall be used. This census satisfies the requirements of Section IV .B.3 of Appendix I to IO CFR 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/year) of leafy vegetables assumed in RG 1.109 for consumption by a child. To determine this minimum garden size, the following assumptions were made: (I) 20% of the garden was used for growing broad leaf vegetation (i.e., similar to lettuce and cabbage) and (2) the vegetation yield was 2 kg/m2

A MILK SAMPLING LOCATION, as defined in Section 1.0, requires that at least 10 milking cows are present at a designated milk sample location. It has been found from past experience, and as a result of conferring with local farmers, that a minimum of 10 milking cows is necessary to guarantee an adequate supply of milk twice a month for analytical purposes. Locations with fewer than IO milking cows are usually utilized for breeding purposes, eliminating a stable supply of milk for samples as a result of suckling calves and periods when the adult animals are dry. Elevated releases are defined in RG 1.111, Revision 1, July 1977.

Unit2 Revision 37 I B 3.5-2 September 2020

Interlaboratory Comparison Program B 3.5.3 B 3.5 RADIOLOGICAL ENVIRONMENTAL MONITORING B 3.5.3 Interlaboratory Comparison Program BASES The requirement for participation in an approved Interlaboratory Comparison Program is provided to ensure that independent checks on the precision and accuracy of the measurements of radioactive materials in environmental sample matrices are performed as part of the quality assurance program for environmental monitoring in order to demonstrate that the results are valid for the purposes of Section IV.B.2 of Appendix I to 10 CFR 50.

Unit2 Revision 37 I B 3.5-3 September 2020

PART I - RADIOLOGICAL EFFLUENT CONTROLS SECTION 4.0 ADMINISTRATIVE CONTROLS Unit2 Revision 37 I 4.0-0 September 2020

Administrative Controls 4.0 4.0 ADMINISTRATIVE CONTROLS The ODCM Specifications are subject to Technical Specifications Section 5.5.4, "Radioactive Effluent Controls Program," Section 5.6.2, "Annual Radiological Environmental Operating ReJX)rt," Section 5.6.3, "Radioactive Effluent Release Report," and Section 5.5.1, "Offsite Dose Calculation Manual." r

\

Unit2 Revision 37 I 4.0-1 September 2020

Special Reports D 4.1.1 D 4.1.2 D 4.1.3 D 4.1 REPORTING REQUIREMENTS D 4.1.1 Special Reports Special Reports shall be submitted in accordance with 10 CFR 50.4 within the time period specified for each report.

D 4.1.2Annual Radiological Environmental Operating Reports In addition to the requirements of Technical Specification 5.6.2 the report shall also include the following:

A summary description of the Radiological Environmental Monitoring Program; at least two legible maps, one shall cover stations near the SITE BOUNDARY and the second shall include the more distant stations, covering all sample locations keyed to a table giving distances and directions from the centerline of one reactor; the results of license participation in the Interlaboratory Comparison Program, required by Control D 3.5.3; discussion of all deviations from the Sampling Schedule of Table D 3 .5 .1-1; and discussion of all analysis in which the LLD required by Table D 3.5.1-3 was not achievable.

D 4.1.3 Radioactive Eflluent Release Report The Radiological Eflluent Release Report described in Technical Specification section 5.6.3 shall include:

An annual summary of hourly meteorological data collected over the previous year. This annual summary may be either in the form of an hour-by-hour listing on magnetic tape of wind speed, wind direction, atmospheric stability, and precipitation (if measured), or in the form of joint frequency distribution of wind speed, wind direction, and atmospheric stability.

In lieu of submission with the Radiological Eflluent Release Report, the licensee has the option of retaining this summary of required meteorological data on site in a file that shall be provided to the NRC upon request. '

An assessment of radiation doses from the radioactive liquid and gaseous eflluents released from the unit during the previous year.

(Continued)

Unit2 Revision 37 I 4.1-1 September 2020

Special Reports D 4.1.3 D 4.1.3 Radioactive Effluent Release Report (continued)

As assessment of radiation doses from the radioactive liquid and gaseous effluents to MEMBERS OF THE PUBLIC from their activities inside the SITE ~OUNDARY during the report period. All assumptions used in making these assessments, i.e., specific activity, exposure time, and location shalLbe included in these reports. The assessment of radiation doses shall be performed in accordance with the methodology and parameters in Part IL

As assessment of doses to the likely most exposed MEMBER OF THE PUBLIC from reactor releases and other nearby uranium fuel cycle sources, including doses from primary effluent pathways and direct radiation, for the previous calendar year to show conformance with 40 CPR 190, '*'Environmental Radiation Protection Standards for Nuclear Power Operation."

Acceptable methods for calculating the dose contribution from liquid and gaseous effluents are given in Part II.

A list of unplanned releases from the site to UNRESTRICTED AREAS of radioactive materials in gaseous and liquid effluents made during the reporting period.

Any changes made during the reporting period to the PROCESS CONTROL PROGRAM and to the OFFSITE DOSE CALCULATION MANUAL (ODCM).

Any major changes to liquid, gaseous, or solid radwaste treatment systems pursuant to D 4.2.

A listing of new locations for dose calculations and/or environmental monitoring identified by the land use census pursuant to Control D 3.5.2.

An explanation of why the nonfunctionality of liquid or gaseous effluent monitoring instrumentation was not corrected within the time specified in Controls D 3.3.1 and D 3.3.2.

Description of events leading to liquid holdup tanks exceeding the limits of TRM 3.7.7.

Unit2 Revision 37 I 4.2-1 September 2020

Major Changes to Liquid, Gaseous, and Solid Radwaste Treatment System D4.2 D 4.2 MAJOR CHANGES TO LIQUID, GASEOUS, AND SOLID RADW ASTE TREATMENT SYSTEM

NOTE----- --------------------------------------------

Licensees may choose to submit this information as part of the annual USAR update.

Licensee-initiated major changes to the radwaste treatment systems (liquid, gaseous, and solid):

a. Shall be reported to the Commission in the Radioactive Effluent Release report for the period in which the evaluation was reviewed by the PORC. The discussion of each change shall contain:

1. A summary of the evaluation that led to the determination that the change could be made in accordance with 10 CFR 50.59.

2. Sufficient detailed information to totally support the reason for the change without benefit of additional or supplemental information;

3. A detailed description of the equipment, components, and processes involved and the interfaces with other plant systems;

4. An evaluation of the change, which shows the predicted releases of radioactive materials in liquid and gaseous effluents and/or quantity of solid waste that differ from those previously predicted in the license application and amendments thereto;

5. An evaluation of the change, which shows the expected maximum exposures to a MEMBER OF THE PUBLIC in the UNRESTRICTED AREA and to the general population that differ from those previously estimated in the license application and amendments thereto;

6. A comparison of the predicted releases of radioactive materials, in liquid and gaseous effluents and in solid waste, to the actual releases for the period that precedes the time when the change is to be made;

7. An estimate of the exposure to plant operating personnel as a result of the change; and (Continued)

Unit2 Revision 37 I 4.2-1 September 2020

Major Changes to Liquid, Gaseous, and Solid Radwaste Treatment System D 4.2 D 4.2 MAJOR CHANGES TO LIQUID, GASEOUS, AND SOLID RADWASTE TREATMENT SYSTEM (continued)

8. Documentation of the fact that the change was reviewed and found acceptable by the PORC.

b. Shall ~come effective upon review and acceptance by the PORC.

L Unit2 Revision 37 I 4.2-2 September 2020

PART II - CALCULATIONAL METHODOLOGIES I

1.0 LIQUID EFFLUENTS Service Water A and B, Cooling Tower Blowdown and the Liquid Radioactive Waste Discharges comprise the Radioactive Liquid Effluents at Unit 2. Presently there are no temporary outdoor tanks containing radioactive water capable of affecting the nearest ,

known or future water supply in an unrestricted area. NUREG 0133 and Regulatory Guide 1.109, Rev. 1were followed in-the development of this section.

1.1

Liquid Effluent Monitor Alarm Setpoints 1.1.1 Basis The concentration of radioactive material released in liquid effluents to UNRESTRICTED AREAS (see Figure D 1.0-1) shall be limited to ten times the concentrations specified in 10 CFR 20, Appendix B, Table 2, Column 2, for rad~onuclides other than dissolved or entrained noble gases. For dissolved or entrained noble gases, the concentration shall be limited to 2E-04 uCi/ml total activity.

1.1.2 Setpoint Determination Methodology 1.1.2.1 Liquid Radwaste Effluent Radiation Alarm Setpoint The Liquid Radioactive Waste System Tanks are pumped to the discharge tunnel which in tum flows directly to Lake Ontario. At the end of the discharge tunnel in Lake Ontario, a diffuser structure has been installed. Its purpose is to maintain surface water temperatures low enough to meet thermal pollution limits. However, it also assists in the near field dilution of any activity released. Service Water and the Cooling Tower Blowdown are also pumped to the discharge tunnel and will provide dilution. If the Service Water or the Cooling Tower Blowdown is found to b~ contaminated, then its activity will be accounted for when calculating the permissible radwaste eflluent flow for a Liquid Radwaste discharge. The Liquid Radwaste System Monitor provides alarm and automatic termination of release if radiation levels above its alarm setpoint are detected.

The radiation detector is a sodium iodide crystal. It is a scintillation device. The crystal is sensitive to-gamma and beta radiation. However, because of the metal walls of the sample chamber and the absorption characteristics of water, the monitor is not particularly sensitive to beta radiation. Actual detector response 3i ( CGi/CF i), cpm, has been evaluated by placing a sample of typical radioactive waste into the monitor and recording the gross count rate, cpm. A calibration ratio was developed by dividing the noted detector response, 31 ( CGi/ CF1) cpm, by total concentration of activity 3 i ( CGi), uCi/cc. The quantification of the gamma activity was completed with gamma spectrometry equipment whose calibration is traceable to NIST. This calibration ratio verified the manufacturer's prototype calibration, and any subsequent transfer calibrations performed. The current calibration factor (expressed as the reciprocal conversion factor, uCi/ml/cpm), will be used for subsequent setpoint calculations in the determination of detector response:

Where the factors are as defined above.

Unit 2 Revision 37 II 2 September 2020

The calculations of the required dilution factors (RDF) are performed as follows:

RDFy= IMECgammafraction= ~)CG/MECJ RDFTOTAL = I:MEC total fraction = L (C/MECJ RDFr is used to calculate the liquid radwaste effluent radiation monitor setpoint. This monitor is a gamma detector and has little or no response to non gamma emitters. Use of RDFr rather than RDFTOTAL, to determine the monitor setpoint prevents the condition where a tank with gamma concentrations near their LLD cannot be discharged due to spurious alarms received because the setpoint is close to the monitor background.

RDFrnTAL is used to determine the minimum dilution factor required to discharge the tank contents based on all activity, both gamma and non gamma, in the tank. This ensures that the concentrations of all radioactive materials released in liquid effluents will meet DLCO 3.1.1. Non gamma emitting nuclide activity, except tritium was initially estimated based on the expected ratios to quantified nuclides as listed in the USAR Table 11.2.5. Fe-55, Sr-89 and Sr-90 are 2.5, 0.25 and 0.02 times, respectively, the concentration of Co-60.

Currently, non gamma activity except tritium is estimated using the results from the latest analysis of composite samples.

Tritium*concentration is assumed to equal the latest concentration detected in the monthly tritium analysis of liquid radioactive waste tanks discharged.

Nominal flow rates of the Liquid Radioactive Waste System Tanks discharged is< 165 gpm while dilution flow from the Service Water Pumps, and Cooling Tower Blowdown cumulatively is typically over 10,200 gpm. Because of the large amount of dilution the alarm setpoint could be substantially greater than that which would correspond to the concentration actually in the tank. Potentially a discharge could continue even if the distribution of nuclides in the tank were substantially different from the grab sample obtained prior to discharge which was used to establish the detector alarm point. To avoid this possibility of Non representative Sampling" resulting in erroneous assumptions about the discharge of a tank, the tank is recirculated for a minimum of 2.5 tank volumes prior to sampling.

I This monitor's setpoint takes into account the dilution ofRadwaste Effluents provided by the Service Water and Cooling Tower Blowdown flows. Detector response for the nuclides to be discharged (cpm) is multiplied by the Actual Dilution Factor (dilution flow/waste stream flow) and divided by the Required Dilution Factor (total fraction of the effluent concentration in the waste stream). A safety factor is used to ensure that the limit is never exceeded. Service Water and Cooling Tower Blowdown are normally non-radioactive. If they are found to be contaminated prior to a Liquid Radwaste discharge then an alternative equation is used to take into account the contamination. If they become contaminated during a Radwaste discharge, then the discharge will be immediately terminated and the situation fully assessed.

Unit2 Revision 37 II 3 September 2020

~ .

Normal Radwaste Effluent Alarm Setpoint Calculation:

Alarm Setpoint ~ 0.8

TDF/PEF

TGC/CF

1/RDFy + Background.

Where:

Alarm Setpoint = The Radiation Detector Alarm Setpoint, cpm 0.8 = Safety Factor, unitless TDF Nonradioactive dilution flow rate, gpm. Service Water Flow (ranges from 30,000 to 58,000 gpm) +

Blowdown flow (typically 10,200 gpm) - Tempering PEP = The permissible Radwaste Effluent Flow rate, gpm, 165 gpm is the maximum value used in this equation TDF/PEF = An approximation to (TDF + PEF)/PEF, the Actual Dilution Factor in effect during a discharge.

TGC=LCG1 = Summation of all gamma emitting nuclides (which monitor will respond to)

CGi = Concentration of gamma emitting nuclide in Radwaste tank prior to dilution, µCi/ml CF = Monitor Conversion Factor, µCi/ml/cpm, determined at each calibration of the effluent monitor TGC/CF An approximation to 31(CGJCF1) using CF determined at each calibration of the effluent monitor RDFy = 31 (CG1/MEC1) = The total fraction of ten times the 10 CPR 20, Appendix B, Table 2, Column 2 limit that is in the Radwaste tank, unitless.

This is also known as the Required Dilution Factor Gamma (RDFy).

Background = Detector response when sample chamber is filled with nonradioactive water, cpm Ci = Concentration of isotope i in Radwaste tank prior to dilution,

µCi/ml (gamma+ non-gamma emitters)

CF1 = Detector response for isotope i, net µCi/ml/cpm See Table D 2-1 for a list of nominal values MECi = Maximum Effluent Concentration, ten times the limiting effluent concentration for isotope i from 10 CFR 20 Appendix B, Table 2, Column 2, µCi/ml The total detector response when exposed to the concentration of nuclides in the Radwaste tank, cpm '

The total fraction of ten times the 10 CFR 20, Appendix B, Table 2, Column 2 limit that is in the Radwaste tank, unitless.

This is also known as the Required Dilution Factor-Total and includes both the gamma and non-gamma emitters.

Tempering A diversion of some fraction of discharge flow to the intake canal for the purpose of temperature control, gpm.

Unit2 Revision 37 II4 September 2020

(

Permissible effluent flow, PEF, shall be calculated to determine that the maximum effluent concentration will not be exceeded in the discharge canal.

PEF = TDF (RDFTota!) 1.5 If Actual Dilution Factor is set equal to the Required Dilution Factor, then the alarm points required by the above equations correspond to a concentration of 80% of the Radwaste Tank concentration. No discharge could occur, since the monitor would be in alarm as soon as the discharge commenced. To avoid this situation, maximum allowable radwaste discharge flow is calculated using a multiple (usually 1.5 to 2) of the Required Dilution Factor, resulting in discharge canal concentration.of 2/3 to 1/2 of the maximum effluent concentration prior to alarm and termination of release. If no gamma emitters are detected in the Radwaste Tank samples, then the radiation monitor setpoint will be based on assuming gamma activity at the LLD of the most limiting nuclide from recent discharges. In performing the alarm calculation, the smaller of 165 gpm (the maximum possible flow) and PEF will be used.

To ensure the alarm setpoint is not exceeded, an alert alarm is provided. The alert alarm will be set in accordance with the equation above using a safety factor of 0.5 (or lower) instead of 0.8.

1.1.2.2 Contaminated Dilution Water Radwaste Effluent Monitor Alarm Setpoint Calculation:

The allowable discharge flow rate for a Radwaste tank, when one of the normal dilution streams (Service Water A, Service Water B, or Cooling Tower Blowdown) is*

contaminated, will be calculated by an iterative process. Using Radwaste tank concentrations with a total liquid effluent flow rate, the resulting fraction of the maximum effluent concentration in the discharge canal will be calculated.

FMEC = ~[F~s) Lt(Cis D MECi)]

Then the permissible radwaste effluent flow rate is given by:

PEF = Total Radwaste Effluent Flow FMEC The corresponding Alarm Setpoint will then be calculated using the following equation, with PEF limited as above.

Alarm Setpoint .:S 0.8 TGC/CF + Background FMEC Unit2 Revision 37 II 5 September 2020

Where:

Alarm Setpoint = The Radiation Detector Alarm Setpoint, cpm 0.8 = Safety Factor, Unitless An Effluent flow rate for stream s, gpm Concentration of isotope i in Radwaste tank prior to dilution, µCi/ml Concentration of isotope i in Effluent stream s including the Radwaste Effluent tank undiluted, µCi/ml CF = Average detector response for all isotopes in the waste stream, net ,

µCi/ml/cpm Maximum Effluent Concentration, ten times the effluent concentration limit for isotope i from 10CFR20 Appendix B, Table 2, Column 2, µCi/ml PEF = The permissible Radwaste Effluent Flow rate, gpm Background = Detector response when sample chamber is filled with nonradioactive water, cpm TGC/CF = The total detector response when exposed to the L1(CGJCF) concentration of nuclides in the Radwaste tank, cpm Ls[FsCis] = The total activity of nuclide i in all Effluent streams, µCi-gpm/ml Ls[Fs] = The total Liquid Effluent Flow rate, gpm (Service Water & CT Blowdown & Radwaste) 1.1.2.3 Service Water and Cooling Tower Blowdown Effluent Alarm Setpoint These monitor setpoints do not take any credit for dilution of each respective effluent stream. Detector response for the distribution of nuclides potentially discharged is divided by the total MEC fraction of the radionuclides potentially in the respective stream. A safety factor is used to ensure that the limit is never exceeded.

Service Water and Cooling Tower Blowdown are normally non-radioactive. If they are found to be contaminated by statistically significant increase in detector response then grab samples will be obtained and analysis meeting the LLD requirements of Table D 3.1.1-1 completed so that an estimate of offsite dose can be made and the situation fully assessed.

Service Water A and Band the Cooling Tower Blowdown are pumped to the discharge tunnel which in turn flows directly to Lake Ontario. Normal flow rates for each Service Water Pump is 10,000 gpm while that for the Cooling Tower Blowdown may be as much as 10,200 gpm. Credit is not taken for any dilution of these individual effluent streams.

The radiation detector is a sodium iodide crystal. It is a scintillation device. The crystal is sensitive to gamma and beta radiation. However, because of the metal walls in its sample chamber and the absorption characteristics of water, the monitor is not particularly sensitive to beta radiation.

Unit2 Revision 37 II 6 September 2020

Detector response l:1(CJCF1) has been evaluated by placing a diluted sample of Reactor Coolant (after a two hour decay) in a representative monitor and noting its gross count rate. Reactor Coolant was chosen because it represents the most,likely contaminant of Station Waters. '

A two hour decay was chosen by judgement of the staff of Nine Mile Point. Reactor Coolant with no decay contains a considerable amount of very energetic nuclides which would bias the detector response term high. However assuming a longer than 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months decay is not realistic as the most likely release mechanism is a leak through the Residual Heat Removal Heat Exchangers which would contain Reactor Coolant during shutdowns.

Service Water and Cooling Tower Blowdown Alarm Setpoint Equation:

Alarm Setpoint < 0.8 1/CF l:1 CJ[L(CJMECi)] + Background.

Where:

Alarm Setpoint = The Radiation Detector Al~ Setpoint, cpm 0.8 = Safety Factor, unitless Ci = Concentration of isotope i in potential contaminated stream,

µCi/ml CF1 = Detector response for isotope i, net µCi/ml/cpm See Table 2-1 for a list of nominal values

MECi = Maximum Effluent Concentration, ten times the effiuent concentration limit for isotope i from 10 CFR 20 Appendix B, Table 2, Column 2, µCi/ml Background Detector response when sample chamber is filled with rn;mradioactive water, cpm The total detector response when exposed to the concentration of nuclides in the potential contaminant, cpm The total fraction of ten times the 10CFR20, Appendix B, Table 2, Column 2 limit that is in the potential contaminated stream, unitless.

An approximation to L(CJCF1), determined at each calibration of the effluent monitor CF Monitor Conversion Factor, µCi/ml/cpm 1.2 Liquid Effluent Concentration Calculation This calculation documents compliance with Section D 3.1.1 of Part I:

Unit2 Revision 37 II 7 September 2020

As required by Technical Specification 5.5.4, "Radioactive Effiuent Controls Program," the concentration of radioactive material released in liquid effluents to UNRESTRICTED AREAS (see Figure D 1.0-1) shall be limited to ten times the concentrations specified in 10 CFR 20, Appendix B, Table 2, Column 2, for radionuclides other than dissolved or entrained noble gases. For dissolved or entrained noble gases, the concentration shall be limited to 2E-04 microcurie/ml total .activity.

The concentration of radioactivity from Liquid Radwaste, Service Water A and B and the Cooling Tower Blowdown are included in the calcul11tion. The calculation is performed for a specific period of time. No credit is taken for averaging. The limiting concentration is calculated as follows:

FMEC = ~[F~(Fs)L1(CsDMEC)]

(

Where: FMEC = The Fraction of Maximum Effluent Concentration, the ratio at the point of discharge of the actual concentfation to ten times the limiting concentration of 10 CFR 20, Appendix B, Table 2, Column 2, for radionuclides other than dissolved or entrained noble gases, unitless Cs = The concentration of nuclide i in a particular effluent stream ~' µCi/ml Fs = The flow rate of a particular effluent stream s, gpm MEC = Maximum Effluent Concentration, ten times the limiting Effluent Concentration of a specific nuclide i from 10CFR20, Appendix B, Table 2, Column 2 (for noble gases, the concentration shall be limited to 2E-4 microcurie/ml), µCi/ml Li(CdMEC) = The Maximum Effluent Concentration fraction of stream s prior to dilution by other streams

~(Fs) = The total flow rate of all effluent streams s, gpm A value of less than one for the MEC fraction is required for compliance.

1.3 Liquid Effluent Dose Calculation Methodology The dose or dose commitment to a MEMBER OF THE PUBLIC from radioactive materials in liquid effluents released, from each unit, to UNRESTRICTED AREAS (see Figure D 1.0-1) shall be limited:

a. During any calendar quarter to less than or equal to 1.5 mrem to the whole body and to less than or equal to 5 mrem to any organ, and

b. During any calendar year to less than or equal to 3 mrem to the whole body and to less than or equal to 10 mrem to any organ.

Unit 2 Revision 37 II 8 September 2020

Doses due to Liquid Effluents are calculated monthly for the fish and drinking water ingestion pathways and the external sediment exposure pathways from all detected nuclides in liquid effluents released to the unrestricted areas using the following expression from NUREG 0133, Section 4.3.

Where:

Dt = The cumulative dose commitment to the total body or any organ, t from the liquid effluents for the total time period L( TL), mrem The length of the L th time period over which CtL and FL are averaged for all liquid releases, hours The average concentration of radionuclide, i, in undiluted liquid effluents during time period ~ TL from any liquid release, µCi/ml The site related ingestion dose commitment factor for the maximum individual to the total body or any organ t for each identified principal gamma or beta emitter, rnrem/hr per µCi/ml. Table D 2-2.

The near field average dilution factor for Ci1 during any liquid effluent release. Defined as the ratio of the maximum undiluted liquid waste flow during release to the product of the average flow from the site discharge structure to unrestricted receiving waters times 5.9. (5.9 is the site specific applicable factor for the mixing effect of the discharge structure.) See the Nine Mile Point Unit 2 Environmental Report - Operating License Stage, Table 5.4-2 footnote 1.

These factors can be related to batch release parameters as follows:

PEF I (TDF x 5.9) (Terms defined in Section 1.1.2.1 and above)

[PEF (gpm) x ~TL (min) x 1.67E-2 (hr/min)] I [TDF (gpm) x 5.9]

[TV x 2.83E-3 (hours)]/ TDF For each batch, PEF (gpm) x ~TL (min)= Tank Volume. For each batch, a dose calculation common constant (~TLFL) is calculated to be used with the concentration of each nuclide and dose factor, A1, to calculate the dose to a receptor. Normally, the highest dose factor for any age group (adult, teen, child, infant) will be used for calculation, but specific age-group calculations to demonstrate compliance may be performed if required.

When actual results for the non-gamma emitter, (Gross Alpha, H-3, Sr-89, S.r-90 and Fe-55), concentrations are not available, dose contributions are calculated using concentration estimates. The doses are revised when actual results are obtained.

Unit2 Revision 37 II 9 September 2020

1.4 Liquid Effluent Sampling Representati,veness There are four tanks in the radwaste system designed to be discharged to the discharge canal. These tanks are labeled 4A, 4B, 5A, and 5B.

Liqui<;i Radwaste Tank SA.and 5B at Nine Mile Point Unit 2 contain a ,sparger spray ring which assists the mixing of the tank contents while it is being recirculated prior to 1sampling. This sparger effectively mixes the tank four times faster than simple recirculation.

Liquid Radwaste Tank 4A and 4B contain a mixing ring but no sparger. No credit is taken for the mixing effects of the ring. Normal recirculation flow is 150 gpm for tank 5A and 5B, 110 gpm for tank 4A and 4B while each tank contains up to 25,000 gallons although the entire contents are not discharged. To assure that the tanks are adequately mixed prior to sampling, it is a plant requirement that the tank be recirculated for the time required to pass 2.5 times the volume of the tank:

Recirculation Time = 2.5T/RM Where:

Recirculation Time Is the minimum time to recirculate the Tank, min 2.5 Is the plant requirement, unitless T Is the tank volume, gal R Is the recirculation flow rate, gpm.

M Is the factor that takes into account the mixing of the sparger, unitless, four for tank 5A and B, one for tank 4A andB.

Additionally, the Alef1: Alarm setpoint of,the Liquid Radwaste Effluent monitor is set at approximately 60% of the High alarm setpoint. This alarm will give indication of incomplete mixing with adequate margin before exceeding ten times the effluent concentration.

Service Water A and Band the Cooling Tower Blowdown are sampled from the radiation monitor on each respective stream. These monitors continuously withdraw a sample and pump it back to the effluent stream. The length of tubing between the continuously flowing sample and the sample spigot contains less than 200 ml which is adequately purged by requiring a purge of at least 1 liter when grabbing a sample.

1.5 Liquid Radwaste System FUNCTIONALITY The Liquid Radwaste Treatment System shall be FUNCTIONAL and used when projected doses due to liquid radwaste effluents would exceed 0.06 mrem to the whole body or 0.2 mrem to any organ in a 31-day period. Cumulative doses will be determined at least once per 31 days (as indicated in Section 1.3) and doses will also be projected if the radwaste treatment systems are not being fully utilized.

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The system collection tanks are processed as follows:

1) Low Conductivity (Waste Collector): Radwaste Filter and Radwaste Demineralizer or the Thermex System.

2) High Conductivity (Floor Drains): Regenerant Evaporator or the Thermex System.

3) Regenerant Waste: If resin regeneration is used at NMP-2; the waste will be processed through the regenerant evaporator or Thermex System.

The dose projection indicated above will be performed in accordance with the methodology of Section 1.3.

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2.0 GASEOUS EFFLUENTS The gaseous effluent release points are the stack and the combined Radwaste/Reactor Building vent. The stack effluent point includes Turbine Building ventilation, main condenser offgas (after charcoal bed holdup), and Standby Gas Treatment System exhaust NUREG 0133 and Regulatory Guide 1.109, Rev. 1 were followed in the development of this section.

2.1 Gaseous Effluent Monitor Alarm Setpoints 2.1.1 Basis The dose rate from radioactive materials released in gaseous effluents from the site to areas at or beyond the SITE BOUNDARY (see Figure D 1.0-1) shall be limited to the following in accordance with Technical Specification 5.5.4.g:

a. For noble gases: Less than or equal to 500 mrem/yr to the whole body and less than or equal to 3000 mrem/yr to the skin, and

b. For iodine-131, for iodine-I 33, for tritium, and for all radionuclides with half-lives greater than 8 days: Less than or equal to 1500 mrem/yr to any organ.

The radioactivity rate of noble gases measured at the recombiner effluent shall be less than or equal to 350,000 microcuries/second after decay of 30 minutes during offgas system operation in accordance with Technical Specification 3.7.4.

2.1.2 Setpoint Determination Methodology Discussion Nine Mile Point Unit 1 and the James A FitzPatrick nuclear plants occupy the same site as Nine Mile Point Unit 2. Because of the independence of these plants' safety systems, control rooms and operating staffs it is assumed that simultaneous accidents are not likely to occur at the different units. However, there are two release points at Unit 2. It is assumed that if an accident were to occur at Unit 2 that both release points could be involved.

The alarm setpoint for Gaseous Effluent Noble Gas Monitors are based on a dose rate limit of 500 rnRem/yr to the Whole Body. Since there are two release points at U:nit 2, the dose rate limit of 500 mRem/yr is divided equally for each release point, but may be apportioned otherwise, if required. These monitors are sensitive to only noble gases.

Because of this it is considered impractical to base their alarm setpoints on organ dose rates due to iodines or particulates. Additionally skin dose rate is never significantly greater than the whole body dose rate. Thus the factor R which is the basis for the alarm setpoint calculation is nominally taken as equal to 250 rnRem/yr. If there are significant releases from any gaseous release point on the site (>25 rnRem/yr) for an extended period of time then the setpoint will be recalculated with an appropriately smaller value forR.

The high alarm setpoint for the Offgas Noble Gas monitor is based on a limit of 350,000 uCi/sec after 30-minute delay. This is the release rate for which a USAR accident analysis was completed. At this rate the Offgas System charcoal beds will not contain Unit2 Revision 37 II 12 September 2020

enough activity so that their failure and subsequent release of activity will present a significant offsite dose assuming accident meteorology.

Initially, in accordance with Part I, Section D 3.3.2, the Radiation Monitoring System on the stack and vent will be calibrated with gas standards (traceable to NIST) in accordance with DSR 3.3.2.9. Subsequent calibrations may be performed with gas standards, or with related solid sources.

2.1.2.1 Stack Noble Gas Detector Alarm Setpoint Equation:

The stack at Nine Mile Point Unit 2 receives the Offgas after charcoal bed delay, Turbine Building Ventilation and the Standby Gas Treatment system exhaust. The Standby Gas Treatment System Exhausts the primary_containment during normal shutdowns and maintains a negative pressure on the Reactor Building to maintain secondary containment integrity. The Standby Gas Treatment will isolate on high radiation detected (by the SOTS monitor) during primary containment purges.*

The stack noble gas detector is a beta-gamma sensitive detector. It is able to accurately quantify the activity released in terms ofuCi ofXe-133 equivalent activity. A distribution of Noble Gases corresponding to offgas is chosen for the nominal alarm setpoint calculation. Offgas is chosen because it represents the most significant contaminant of gaseous activity in the plant. The release rate Qi, corresponds to offgas concentration expected with the plant design limit for fuel failure. The alarm setpoint may be recalculated if a significant release is encountered. In that case the actual distribution of noble gases will be used in the calculation.

The following calculation will be used for the initial Alarm Setpoint.

0.8R L(01)

Alarm Setpoint, µCi/sec ,:S L1(QiV1) 0.8 Safety Factor, un'itless R Allocation Factor. Normally, 250 mrem/yr; the value must be 500 mrem/yr or less depending upon the dose rate from other release points within the site such that the total dose rate corresponds to

< 500 mrem/yr

= The release rate of nuclide i, µCi/sec The constant for each identified noble gas nuclide accounting fo,r the whole body dose from the elevated finite plume listed on Table D 3-2, mrem/yr per µCi/sec

= The total release rate of noble gas nuclides in the stack effluent,

µCi/sec The total of the product of each isotope release rate times its respective whole body plume constant, mrem/yr The alert alarm is normally set at less than 10% of the high alarm.

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2.1.2.2 Vent Noble Gas Detector Alarm Setpoint Equation:

The vent contains the Reactor Building ventilation above and below the refuel floor and the Radwaste Building ventilation effluents. The Reactor Building Ventilation will isolate when radiation monitors detect high levels of radiation (these are separate monitors, not otherwise discussed in the ODCM). Nominal flow rate for the vent is 2.37E5 CFM.

This detector is a beta-gamma sensitive detector. It is able to accurately quantify the activity released in terms of uCi of Xe-133 equivalent activity. A distribution of Noble Gases corresponding to that expected with the design limit for fuel failure offgas is chosen for the nominal alarm setpoint calculation. Offgas is chosen because it represents the most significant contaminant of gaseous activity in the plant. The alarm setpoint may be recalculated if a significant release is encountered. In that case the actual distribution of noble gases will be used in the calculation.

0.8R Li(01)

Alarm Setpoint, uCi/sec < (X/Q)~ L1(Q1K1)

Where:

0.8 Safety Factor, unitless R = Allocation Factor. Normally, 250 mrem/yr; the value must be 500 mrem/yr or less depending upon the dose rate from other release points within the site such that the total rate corresponds to< 500 mrem/yr Qi = The release rate of nuclide i, µCi/sec (X/Q)v The highest annual average atmospheric dispersion coefficient at the site boundary as listed in the Final Environmental Statement, NUREG 1085, Table D-2, 2.0E-6 sec/m3

'fhe constant for each identified noble gas nuclide accounting for the whole body dose from the semi-infinite cloud, listed on Table D 3-3, mrem/yr per µCi/m 3 The total release rate of noble gas nuclides in the vent effluent, uCi/sec

= The total of the product of the each isotope release rate times its respective whole body immersion constant, mrem/yr per sec/m 3 The alert alarm is normally set at less than 10% of the high alarm.

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2.1.2.3 Offgas Pretreatment Noble Gas Detector Alarm Setpoint Equation:

The Offgas system has a radiation detector downstream of the recombiners and before the charcoal decay beds. The offgas, after decay, is exhausted to the main stack. The system will automatically isolate if its pretreatment radiation monitor detects levels of radiation above the high alarm setpoint.

The Radiation Detector contains a plastic scintillator disc. It is a beta scintillation detector. Calculation H21C-070, Offgas Radiation Monitor (2OFG-RE13A and 13B)

Alert and Alarm Setpoint Determination, assumes a distribution of offgas corresponding to that expected with the design limit for fuel failure to establish the setpoint. Calculation Table 5 tabulates Alarm/frip setpoints as a function of Offgas flow due to decay from the Recombiner discharge to the monitor/sampling location. The monitor design response values are confirmed during periodic calibration using a Transfer Standard source traceable to the primary calibration performed by the vendor.

Particulates and Iodines are not included in this calculation because this is a noble gas monitor.

To provide an alarm in the event of failure of the offgas system flow instrumentation, the low flow alarm setpoint will be set at or above 10 scfrn, (well below normal system flow) and the high flow alarm setpoint will be set at or below 110 scfm, which is well above exp*ected steady-state flow rates with a tight condenser.

To provide an alarm for changing conditions, the alert alarm will normally be set at less than or equal to 1.5 times nominal full power background (average, +3 standard deviations) to ensure that the Specific Activity Action required by ITS SR 3.7.4.1, are implemented in a timely fashion.

(ACTI(2.12 E-03) Li(CJCF1) + Background Alarm Setpoint, cpm ::: 0.8 F L1(Ci)

Where:

Alarm Setpoint = The alarm setpoint for the offgas pretreatment Noble Gas Detector, cpm

0.8 = Safety Factor, unitless ACT = The Technical Specification Limit for Offgas Pretreatment,

µCi/sec, equivalent to 350,000 µCi/sec after 30 minutes decay.

2.12E-03 = Unit conversion Factor, 60 sec/min I 28317 cc/CF Ci The concentration of nuclide, i, in the Offgas, µCi/cc CF1 The Detector response to nuclide i, µCi/cc/cpm; See Table D 3-1 for a list of nominal values F = The Offgas System Flow rate, CFM Unit2 Revision 37 II 15 September 2020

Background The detector response to something other than noble gases that will be released after 30 minutes decay. Includes purge background, response to activation gases, and response to fission gases with half-life less than 4 minutes, cpm The summation of the nuclide concentration divided by the corresponding detector response, net cpm The summation of the concentration of nuclides in offgas,

µCi/cc NOTE: Calculation H21 C-070 indicates Activity at the Monitor will range from 9.64 E+o5 µCi/sec to 1.22 E +o6 µCi/sec. A conservative Offgas Activity Release Rate of 900,000 µCi/sec applies to this location at all flow rates.

2.2 Gaseous Effluents Dose Rate Calculation Methodology Dose rates will be calculated monthly at a minimum to demonstrate that the release of noble gases, tritium, iodines, and particulates with half lives greater than 8 days are within the dose rate limits specified in 10CFR20. These limits are as follows:

The dose rate from radioactive materials released in gaseous effluents from the site to areas at or beyond the SITE BOUNDARY (see Figure D 1.0-1) shall be limited per Technical Specification 5.5.4.g to the following:

a. For noble gases: Less than or equal to 500 mrem/yr to the whole body and less than or equal to 3000 mrem/yr to the skin, and

b. For iodine-131, iodine-133, for tritium, and for all radionuclides in particulate form with half-lives greater than 8 days: Less than or equal to 1500 mrem/yr to any organ:

When actual results for the gross alpha, particulate, iodine, H-3, Sr-89, Sr-90 and Fe-55, concentrations are not available, dose contributions are calculated using concentration estimates. The doses are revised when actual results are obtained.

2.2.1 X/Q and Wv - Dispersion Parameters for Dose Rate, Table D 3-23 The dispersion parameters for the whole body and skin dose rate calculation correspond to the highest annual average dispersion parameters at or beyond the unrestricted area boundary. This is at the east site boundary. These values were obtained from the Nine Mile Point Unit 2 Final Environmental Statement, NUREG 1085 Table D-2 for the vent and stack. These were calculated using the methodology of Regulatory Guide 1.111, Rev. 1. The stack was modeled as an elevated release point because its height is more, than 2.5 times any adjacent building height. The vent was modeled as a ground level release because even though it is higher than any adjacent building it is not more than 2.5 times the height.

The NRC Final Environmental Statement values for the site boundary X/Q and D/Q terms were selected for use in calculating Effluent Monitor Alarm Points and compliance with 'Site Boundary Dose Rate specifications because they are conservative Unit 2 Revision 37 II 16 September 2020

when compared with the corresponding Nine Mile Point Environmental Report values.

In addition, the stack "intermittent release" X/Q was selected in lieu of the "continuous" value, since it is slightly larger, and also would allow not making a distinction between long term and short term releases.

The dispersion parameters for the organ dose calculations were obtained from the Environmental Report, Figures 7B-4 (stack) and 7B-8 (vent) by locating values corresponding to currently existing (1985) pathways. It should be noted that the most conservative pathways do not all exist at the same location. It is conservative to assume that a single individual would actually be at each of the receptor locations.

2.2.2 Whole Body Dose Rate Due to Noble Gases The ground level gamma radiation dose from a noble gas stack release (elevated),

referred to as plume shine, is calculated using the dose factors from Appendix B of this document. The ground level gamma radiation dose from a noble gas vent release accounts for the exposure from immersion in the semi-infinite cloud. The dispersion of the cloud from the point of release to the receptor at the east site boundary is factored into the plume shine dose factors for stack releases and through the use of X/Q in the equation for the immersion ground level dose rates for vent releases. The release rate is averaged over the period of concern. The factors are discussed in Appendix B.

Whole body dose rate (DR) due to noble gases:

Where:

DR Whole body dose rate (mrem/sec)

The constant accounting for the gamma whole body dose rate from the finite plume from the elevated stack releases for each identified noble gas nuclide, i. Listed on Table D 3-2, mrem/yr per µCi/sec The constant accounting for the gamma whole body dose rate from immersion in the semi-infinite cloud for each identified noble gas nuclide, i. Listed in Table D 3-3, mrem/yr per uCi/m 3 (From Reg.

Guide 1.109)

X/Qv The relative plume concentration at or beyond the land sector site X/Qs boundary. Average meteorological data is used. Elevated X/Q values are used for the stack releases (s=stack); ground X/Q values are used for the vent releases (v=vent). Listed on Table D 3-23 (sec/m 3)

= The release rate of each noble gas nuclide i, from the stack (s) or vent (v). Averaged over the time period of concern. (µCi/sec)

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3.l 7E-08 Conversion Factor; the inverse of the number of seconds in one year.

(yr/sec) 2.2.3 Skin Dose Rate Due to Noble Gases There are two types of radiation from noble gas releases that contribute to the skin dose rate: beta and gamma.

For stack releases this calculation takes into account the dose from beta radiation in a semi infinite cloud by using an immersion dose factor. Additionally, the dispersion of the released activity from the stack to the receptor is taken into account by use of the factor (X/Q). The gamma radiation dose from the elevated stack release is taken into account by the dose factors in Appendix B.

For vent releases the calculations also take into account the dose from the beta (tl ~) and gamma (Dy) radiation of the semi infinite cloud by using an immersion dose factor.

Dispersion is taken into account by use of the factor (X/Q).

The release rate is averaged over the period of concern.

Skin dose rate (DR) MI due to noble gases:

(DR)-, + ~

Where:

(DR) r+~ = Skin dose rate (mrem/sec)

The constant to account for the gamma and beta skin dose rates for each noble gas nuclide, i, from immersion in the semi-infinite cloud, mrem/yr per µCi/m3, listed on Table D 3-3 (from R.G. 1.109)

The constant to account for the air gamma dose rate for each noble gas nuclide, i, from immersion in the semi-infinite cloud, mrad/yr per

µCi/m 3, listed on Table D 3-3 (from R.G. 1.109) 1.11 Unit conversion constant, mrem/mrad

.7 Structural shielding factor, unitless The constant accounting for the air gamma dose rate from exposure to the overhead plume of elevated releases of each identified noble gas nuclide, i. Listed on Table D 3-2, mrad/yr per µCi/sec.

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(X/Q)s = The relative plume concentration at or beyond the land (X/Q)v sector site boundary. Average meteorological data is used. Elevated X/Q values are used for the stack releases (s=stack); ground X/Q values are used for the vent releases (v=vent). (sec/m 3) 3.17E-8 = Conversion Factor; the inverse of the numbet of seconds in a year; (yr/sec)

The release rate of each noble gas nuclide i, from the stack(s) or vent (v) averaged over the time period of concern, µCi/sec.

2.2.4 Organ Dose Rate Due to I-131, I-133, Tritium, and Particulates with l-Jalf-lives greater than 8 days.

The organ dose rate is calculated using the dose factors (Ri) from Appendix C. The factor R, takes into account the dose rate received from the ground plane, inhalation and ingestion pathways. Ws and Wv take into account the atmospheric dispersion from the release point to the location of the most conservative receptor for each of the respective pathways. The release rate is averaged over the period of concern.

Organ dose rates (DR)at due to iodine-131, iodine-133, tritium and all radionuclides in particulate form with half-lives greater than 8 days:

Where:

(DR)at = Organ dose rate (mrem/sec)

The factor that takes into account the dose from nuclide i through pathway j to an age group a, and individual organ t. Units for inhalation pathway, mrem/yr per µCi/m 3* Units for ground and ingestion pathways, m2-mrem/yr per uCi/sec. (See Tables D 3-4 through D 3-22). ,

Ws, Wv = Dispersion parameter either X/Q (sec/m 3) or D/Q (1/m2) depending on pathway and receptor location. Average meteorological data is used (Table D 3-23). Elevated Ws values are used for stack releases (s=stack); ground Wv values are used for vent releases (v=vent).

The release rates for nuclide i, from the stack (s) and vent (v) respectively, µCi/sec.

When the release rate exceeds 0.75 µCi/sec from the stack or vent, the dose rate assessment shall, also, include JAF and NMPl dose contributions. The use of the 0.75

µCi/sec release rate threshold is conservative because it is based on the dose conversion Unit2 Revision 37 II 19 September 2020

factor (Ri) for the Sr-90 child bone which is significantly higher than the dose factors for the other isotopes present in the stack or vent release.

2.3 Gaseous Effluent Dose Calculation Methodology Doses will be calculated monthly at a minimum to demonstrate that ddses resulting from the release of noble gases, tritium, iodines, and particulates with half lives greater than 8 days are within the limits specified in 10 CFR 50. These limits are as follows:

The air dose from noble gases released in gaseous effluents, from each unit, to areas at or beyond the SITE BOUNDARY (see Figure D 1.0-1) shall be limited to the following.

a. During any calendar quarter: Less than or equal to 5 mrad for gamma radiation and less than or equal to 10 mrad for beta radiation, and

b. During any calendar year: Less than or equal to 10 mrad for gamma radiation and less than or equal to 20 mrad for beta radiation.

The dose to a MEMBER OF THE PUBLIC from iodine-131, iodine-133, tritium, and all radioactive material in particulate form with half-lives greater than 8 days in gaseous effluents released, from each unit, to areas at or beyond the SITE BOUNDARY (see Figure D 1.0-1) shall be limited to the following:

a. During any calendar quarter: Less than or equal to 7 .5 mrem to any organ and,

b. During any calendar year: Less than or equal to 15 mrem to any organ.

The VENTILATION EXHAUST TREATMENT SYSTEM shall be FUNCTIONAL and appropriate portions of this system shall be used to reduce releases of radioactivity when the projected doses in 31 days from iodine and particulate releases, from each unit, to areas at or beyond the SITE BOUNDARY (see Figure D 1.0-1) would exceed 0.3 mrem to any organ of a MEMBER OF THE PUBLIC.

When actual results for the gross alpha, particulate, iodine, H-3, Sr-89, Sr-90 and Fe-55, concentrations are not available, dose contributions are calculated using concentration estimates. The doses are revised when actual results are obtained.

2.3.1 Wv and Ws - Dispersion Parameters for Dose, Table D 3-23 The dispersion parameters for dose calculations were obtained chiefly from the Nine Mile Point Unit 2 Environmental Report Appendix 7B. These were calculated using the methodology of Regulatory Guide 1.111 and NUREG 0324. The stack was modeled as an elevated release point because height is more than 2.5 times the height of any adjacent building. The vent was modeled as a combined elevated/ground level release because the vent's height is not more than 2.5 times the height of any adjacent building. Average meteorology over the appropriate time period was used. Dispersion parameters not available from the ER were obtained from C.T. Main Data report dated November, 1985, or the FES.

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2.3.2 Gamma Air Dose Due to Noble Gases Gamma air dose from the stack or vent noble gas releases is calculated monthly. The gamma air dose equation is similar to the gamma dose rate equation except the receptor is air instead of the whole body or skin of whole body. Therefore, the stack noble gas releases use the finite plume air dose factors, and the vent noble gas releases use semi-infinite cloud immersion dose factors. The factor X/Q takes into account the dispersion of vent releases to the most conservative location. The release activity is totaled over the period of concern.

The finite plume factor is discussed in Appendix B.

Gamma air dose due to noble gases:

Dy The gamma air dose for the period of concern, mrad t = The duration of the dose period of concern, sec Where all other parameters have been previously defined.

2.3.3 Beta Air Dose Due to Noble Gases The beta air dose from the stack or vent noble gas releases is calculated using the semi-infinite cloud immersion dose factor in beta radiation. The factor X/Q takes into account the dispersion of releases to the most conservative location.

Beta air dose due to noble gases:

Dti Beta air dose (mrad) for the period of concern N, The constant accounting for the beta air dose from immersion in the semi-infinite cloud for each identified noble gas nuclide, i. Listed on Table D 3-3, mrad/yr per uCi/m 3 * (From Reg. Guide 1.109).

t The duration of the dose period of concern, sec Where all other parameters have been previously defined.

2.3.4 Organ Dose Due to 1-131, 1-133, Tritium and Particulates with Half-Lives Greater than 8 Days.

The organ dose is based on the same equation as the dose rate equation except the dose is compared to the 10CFR50 dose limits. The factor Ri takes into account the dose received from the ground plane, inhalation, food (cow milk, cow meat and vegetation) pathways. Ws and Wv take into account the atmospheric dispersion from the release point to the location of the most conservative receptor for each of the respective pathways. The release is totaled over the period of concern. The Ri factors are discussed in Appendix C.

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Organ dose Det due to iodine-131, iodine-133, tritium and radionuclides in particulate form with half-lives greater than 8 days.

Dat = 3.l 7E-8 LJ [ Li Ri1at [Ws QIS + Wv Qrv)) x t Where:

Det = Dose to the critical organ t, for age group a, mrem t = The duration of the dose period of concern, sec Where all other parameters have been previously defined in Section 2.2.4.

2.4 I-133 and I-135 Estimation Stack and vent effluent iodine cartridges are analyzed to a sensitivity of at least 1E-12 uCi/cc. If detected in excess of the LLD, the I-131 and I-133 analysis results will be reported directly from each cartridge analyzed. Periodically, (usually quarterly but-on a monthly frequency if effluent iodines are routinely detected) a short-duration (12 to 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ) effluent sample is collected and analyzed to establish an I-135/I-131 ratio and an I-133/I-131 ratio, if each activity exceeds LLD. The short-duration ratio is used to confirm the routinely measured I-133 values. The short-duration I-135/I-131 ratio (if determined) is used with the I-131 release to estimate the I-135 release. The short-duration I-133/I-131 ratio may be used with the I-131 release to estimate the I-133 release if the directly measured I-133 release appears non-conservative.

2.5 Isokinetic Sampling Sampling systems for the stack and vent effluent releases are designed to maintain isokinetic sample flow at normal ventilation flow rates. During periods of reduced ventilation flow, sample flow may be maintained at a minimum flow rate (above the calculated isokinetic rate) in order to minimize sample line losses due to particulate deposition at low velocity.

2.6 Use of Concurrent Meteorological Data vs. Historical Data It is the intent to use dispersion parameters based on historical meteorological data to set alarm points and to determine or predict dose and dose rates in the environment due to gaseous effluents. If effluent levels approach limiting values, meteorological conditions concurrent with the time of release may be used to determine gaseous pathway doses.

2.7 Gaseous Radwaste Treatment System Operation Part I, Section D 3.2.4 requires the GASEOUS RADWASTE TREATMENT SYSTEM to be in operation whenever the main condenser air ejector system is in operation. The system may be operated for short periods with the charcoal beds bypassed to facilitate transients.

The components of the system which normally should operate to treat offgas are the Preheater, Recombiner, Condenser, Dryer, Charcoal Adsorbers, HEPA Filter, and Vacuum Pump. (See Appendix D, Offgas System).

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2.8 Ventilation Exhaust Treatment System Operation Part I, Section D 3.2.5 requires the VENTILATION EXHAUST TREATMENT SYSTEM to be FUNCTIONAL when projected doses in 31 days due to iodine and particulate releases would exceed 0.3 mrem to any organ of a member of the public. The appropriate

-components, which affect iodine or particulate release, to be FUNCTIONAL are:

1) HEPA Filter- Radwaste Decon Area

2) HEPA Filter - Radwaste Equipment Area

3) 'HEPA Filter - Radwaste General Area Whenever one of these filters is not FUNCTIONAL, iodine and particulate dose projections will be made for 31-day intervals starting with filter nonfunctionality, and continuing as long :

as the filter remains nonfunctional, in accordance with DSR 3.2.5.1. Predicted release rates will be used, along with the methodology of Section 2.3.4. (See Appendix D, Gaseous Radiation Monitoring.)

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3.0 URANIUM FUEL CYCLE The "Uranium Fuel Cycle" is defined in 40 CFR Part 190.02 (b) as follows:

"Uranium fuel cycle means the operations of milling of uranium ore chemical conversion of uranium, isotopic enrichment of uranium, fabrication of uranium fuel, generation of electricity by a light-water-cooled nuclear power plant using uranium fuel, and reprocessing of spent uranium fuel, to the extent that these directly support the production of electrical power for public use utilizing nuclear energy, but excludes mining operations, operations at waste disposal sites, transportation of any radioactive material in support of these operations, and the reuse of recovered non-uranium special nuclear and by-product materials from the cycle."

Sections D 3.1.2, D 3.2.2, and D 3.2.3 of Part I requires that when the calculated doses associated with the eftluent releases exceed twice the applicable quarter or annual limits, the licensee shall evaluate the calendar year doses and, if required, submit a Special Report to the NRC and limit subsequent releases such that the 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). This report is to demonstrate that radiation exposures to all real individuals from all uranium fuel cycle sources (including all liquid and gaseous eftluent pathways and direct radiation) are less than the limits in 40 CFR Part 190. If releases that result in doses exceeding the 40 CFR 190 limits have occurred, then a variance from the NRC to permit such releases will be requested and if possible, action will be taken to reduce subsequent releases.

The report to the NRG shall contain:

1) Identification of all uranium fuel cycle facilities or operations within 5 miles of the nuclear power reactor units at the site, that contribute to the annual dose of the maximum exposed member of the public.

2) Identification of the maximum exposed member of the public and a determination of the total annual dose to this person from all existing pathways and sources of radioactive eftluents and direct radiation.

The total body and organ doses resulting from radioactive material in liquid eftluents from Nine Mile Point Unit 2 will be summed with the doses resulting from the releases of noble gases, radioiodines, and particulates. The direct dose components will also be determined by either calculation or actual measurement. Actual measurements will utilize environmental TLD dosimetry. Calculated measurements will utilize engineering calculations to determine a projected direct dose component. In the event calculations are used, the methodology will be detailed as required by Technical Specification 5.6.3. The doses from Nine Mile Point Unit 2 will be added to the doses to the maximum exposed individual that are contributed from other uranium fuel cycle operations within 5 miles of the site.

For the purpose of calculating doses, the results of the Environmental Monitoring Program may be included to provide more refined estimates of doses to a real maximum exposed Unit2 Revision 37 II 24 September 2020

individual. Estimated doses, as calculated from station effluents, may be replaced by doses calculated from actual environmental sample results.

3.1 Evaluation of Doses From Liquid Eftluents For the evaluation of doses to real members of the public from liquid effluents, the fish consumption and shoreline sediment ground dose will be considered. Since the doses from other aquatic pathways are insignificant, fish consumption and shoreline sediment are the only two pathways that will be considered. The dose associated with fish consumption may be calculated using eftluent data and Regulatory Guide 1.109 methodology or by calculating a dose to man based on actual fish sample analysis data. Because of the nature of the receptor location and the extensive fishing in the area, the critical individual may be a teenager or an adult. The dose associated with shoreline sediment is based on the as,sumption that the shoreline would be utilized as a recreational area. This dose may be derived from liquid effluent data and Regulatory Guide 1.109 methodology or from actual shoreline sediment sample analysis data.

Equations used to evaluate fish and shoreline sediment samples are based on Regulatory Guide 1.109 methodology. Because of the sample medium type and the half-lives of the radionuclides historically observed, the decay corrected portions of the equations are deleted.

This does not reduce the conservatism of the calculated doses but increases the simplicity from an evaluation point of view. Table D 3-24 presents the parameters used for calculating doses from liquid effluents.

The dose from fish sample media is calculated as:

RapJ = L1 [Cir(U)(DarpJ) f] (1E+3)

Where:

RapJ The total annual dose to organ j, of an individual of age group a, from nuclide i, via fish pathway p, in mrem per year; ex. if calculating to the adult whole body, then RapJ = Rwb and DarpJ = D1WB Cir The concentration of radionuclide i in fish samples in pCi/gram u The consumption rate of fish IE+3 Grams per kilogram

= The ingestion dose factor for age group a, nuclide i, fish pathway p, and organj, (Reg. Guide 1.109, Table E-11) (mrem/pCi). ex. w4en calculating to the adult whole body DaipJ = D1ws f The fractional portion of the year over which the dose is applicable Unit2 Revision 37 II 25 September 2020

The dJse from shorelirn~ sediment sample media is calculated as:

RapJ = L1 [Cis (U)(4E+4J(0.3)(Da1pJ) f]

Where:

RapJ = The total annual dose to organ j, of an individual of age group a, from

_nuclide i, via the sediment pathway p, in mrem per year; ex. if calculating to the adult whole body, then RapJ = RWB and Da!pJ = D1WB

(

= The concentration of radionuclide i in shoreline sediment in pCi/gram u The usage factor, (hr/yr) (Reg. Guide 1.109) 4E+4 The product of the assumed density of shoreline sediment (40 kilogram per square meter to a depth of 2.5 cm) times the number of grams per kilogram 0.3 The shore width factor for a lake Da!pJ The dose factor for age group a, nuclide i, sediment pathway s, and organ

j. (Reg. Guide 1.109, Table E-6) (mrem/hr per pCi/m2); ex. when calculating to the adult whole body Da!pJ = D1WB f The fractional portion of the year over which the dose is applicable NOTE: Because of the nature of the receptor location and the extensive fishing in the area, the critical individual may be a teenager or an adult.

3.2 Evaluation of Doses From Gaseous Effluents For the evaluation of doses to real members of the public from gaseous effluents, the pathways contained in section 2 of the calculational methodologies section will be considered and include ground deposition, inhalation, cows milk, goats milk, meat, and food products (vegetation). However, any updated field data may be utilized that concerns locations of real individuals, real time meteorological data, location of critical receptors, etc.

Data from the most recent census and sample location surveys should be utilized. Doses may also be calculated from actual environmental sample media, as available. Environmental sample media data such as TLD, air sample, milk sample and vegetable (food crop) sample data may be utilized in lieu of effluent calculational data.

Doses to members of the public from the pathways considered in section 2 as a result of gaseous effluents will be calculated using the methodology of Regulatory Guide 1.109 or the methodology of the ODCM, as applicable. Doses calculated from environmental sample media will be based on methodologies found in Regulatory Guide 1.109.

Unit2 Revision 37 II 26 September 2020

3.3 Evaluation of Doses From Direct Radiation The dose contribution as a result of direct radiation shall be considered when evaluating whether the dose limitations of 40 CFR 190 have been exceeded. Direct radiation doses as a result of the reactor, turbine and radwaste buildings', the Independent Spent Fuel Storage Installation (ISFSI), and outside radioactive storage tanks (as applicable) may be evaluated by engineering calculations or by evaluating environmental TLD results at critical receptor locations, site boundary or other special interest locations. For the evaluation of direct radiation doses utilizing environmental TLDs, the critical receptor in question, such as the critical residence, etc., will be compared to the control locations.

The comparison involves the difference in environmental TLD results between the receptor location and the average control location result.

3.4 Doses to Members of the Public Within the Site Boundary The Radioactive Effiuent Release Report shall include an assessment of the radiation doses from radioactive liquid and gaseous eflluents to members of the public due to their activities inside the site boundary as defined by Figure D 1.0-1. A member of the public, would be represented by an individual who visits the sites' Energy Center for the purpose of observing the educational displays or for picnicking and associated activities.

Fishing is a major recreational activity in the area and on the Site as a result of the salmon and trout populations in Lake Ontario. Fishermen have been observed fishing at the shoreline near the Energy Center froin April through December in all weather conditions.

Thus, fishing is the major activity performed by members of the public within the site boundary. Based on the nature of the fishermen and undocumented observations, it is conservatively assumed that the maximum exposed individual spends an average of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months per week fishing from the shoreline at a location,between the Energy Center and the Unit 1 facility. This estimate is considered conservative but not necessarily excessive and accounts for occasions where individuals may fish more on weekends or on a few days in March of the year.

The pathways considered for the evaluation include the inhalation pathway with the resultant lung dose, the ground dose pathway with the resultant whole body and skin dose and the

'direct radiation dose pathway with the associated total body dose. The direct radiation dose pathway, in actuality, includes several pathways. These include: the direct radiation gamma dose to an individual from an overhead plume, a gamma submersion plume dose, possible direct radiation dose from the facility (including the ISFSI) and a ground plane dose (deposition). Because the location is in close proximity to the site, any beta plume submersion dose is felt to be insignificant.

Other pathways, such as the ingestion pathway, are not applicable. In addition, pathways associated with water related recreational activities, other than fishing, are not applicable here. These include swimming, boating and wading which are prohibited at the facility.

Unit2 Revision 37 II 27 September 2020

The inhalation pathway is evaluated by identifying the applicable radionuclides (radioiodine, tritium and particulates) in the effiuent for the appropriate time period. The radionuclide concentrations are then multiplied by the appropriate X/Q value, inhalation dose factor, air intake rate, and the fractional portion of the year in question. Thus, the inhalation pathway is evaluated using the following equation adapted from Regulatory Guide 1.109. Table D 3-24 presents the reference for the parameters used in the following equation.

NOTE:The following equation is adapted from equations C-3 and C-4 of Regulatory Guide 1.109. Since many of the factors are in units of pCi/m3, m 3/sec., etc., and since the radionuclide decay expressions have been deleted because of the short distance to the receptor location, the equation presented here is not identical to the Regulatory Guide equations.

= Lt [(Ci)F (X/Q)(DFA)11a(BR)at]

Where:

The maximum dose from all nuclides to the organ j and age group (a) in mrem/yr; ex. if calculating to the adult lung, then DJa = DL and DF A11a = DFAIL

= The average concentration in the stack or vent release of nuclide i for the period in pCi/m 3

F Unit 2 average stack or vent flowrate in m 3/sec.

XIQ The plume dispersion parameter for a location approximately 0.50 miles west of NMP-2 (The plume dispersion parameters are

, 9.6E-07 (stack) and 2.8E-06 (vent) and were obtained from the C.T. Main five year average annual X/Q tables. The vent X/Q (ground level) is ten times the listed 0.50 mile X/Q because the vent is approximately 0.3 miles from the receptor location. The stack (elevated) X/Q is conservative when based on 0.50 miles because of the close proximity of the stack and the receptor location.

the dose factor for nuclide i, organj, and age group a in mrem per pCi (Reg. Guide 1.109, Table E-7); ex. if calculating to the adult lung the DF A11a = DF A1L (BR)a annual air intake for individuals in age group a in M 3 per year (obtained from Table E-5 of Regulatory Guide 1.109).

t fractional portion of the year for which radionuclide i was detected and for which a dose is to be calculated (in years).

Unit2 Revision 37 II 28 September 2020

The ground dose pathway (deposition) will be evaluated by obtaining at least one soil or shoreline sediment sample in the area where fishing occurs. The dose will then be calculated using the sample results, the time period in question, and the methodology based on Regulatory Guide 1.109 as presented in Section 3.1. The resultant dose may be adjusted for a background dose by subtracting the applicable off-site control soil or shoreline sediment, sample radionuclide activities. In the event it is noted that fishing is not performed from the shoreline but is instead performed in the water (i.e., the use of waders), then the ground dose pathway (deposition) will not be evaluated.

The direct radiation gamma dose pathway includes any gamma doses from an overhead plume, submersion in the plume, possible radiation from the facility (including the ISFSI) and ground plane dose (deposition). This general pathway will be evaluated by average environmental TLD readings. At least two environmental TLDs will be used at one location in the approximate area where fishing occurs. The TLDs will be placed in the field on approximately the beginning of each calendar quarter and removed approximately at the end of each calendar quarter (quarter 2, 3, and 4).

The average TLD readings will be adjusted by the average control TLD readings. This is accomplished by subtracting the average quarterly control TLD value from the average fishing location TLD value. The applicable quarterly control TLD values will be used after adjusting for the appropriate time period (as applicable). In the event of loss or theft of the TLDs, results from a TLD or TLDs in a nearby area may be utilized.

Unit2 Revision 37 II 29 September 2020

4.0 ENVIRONMENTAL MONITORING PROGRAM 4.1 Sampling Stations The current sampling locations are specified in Table D 5-1 and Figures D 5.1-1 and D 5.1-2. The meteorological tower location is shown on Figure D 5.1-1 and is located where TLD location #17 is identified. The Environmental Monitoring Program is a joint effort between the owners and operators of the Nine Mile Point Units 1 and 2 and the James A. FitzPatrick Nuclear Power Plants. Sampling locations are chosen on the basis of historical average dispersion or deposition parameters from both units. The environmental sampling location_ coordinates shown on Table D 5-1 are based on the NMP-2 reactor centerline.

The average dispersion and deposition parameters for the three units have been calculated for a 5 year period, 1978 through 1982. Average dispersion or deposition parameters for the site are calculated using the 1978 through 1982 data and are used to compare the results of the annual land use census. If it is detem,ined that sample locations required by Control D 3.5.1 are unavailable or new locations* are identified that yield a significantly higher (i.e., 50%) calculated D/Q value, actions will be taken as required by Controls D 3.5.1 and D 3.5.2 and_the Radiological Environmental Monitoring Program updated accordingly.

4.2 lnterlaboratory Comparison Program Analyses shall be perfom,ed on samples containing known quantities of radioactive materials that are supplied as part of a Commission approved or sponsored lnterlaboratory Comparison Program, such as the EPA Crosscheck Program.

Participation shall be only for those media, e.g., air, milk, water, etc., that are included in the Nine Mile Point Environmental Monitoring Program and for which cross check samples are available. An attempt will be made to obtain a QC sample to program sample ratio of 5% or better. The Quality Control sample results shall be reported in the Annual Radiological Environmental Operating Report so that the Commission staff may evaluate the results.

Specific sample media for which EPA Cross Check Program samples are available include the following:

gross beta in air particulate filters

gamma emitters in air particulate filters

gamma emitters in milk

gamma emitters in water

tritium in water

1-131 in water Unit2 Revision 37 II 30 September 2020

4.3 Capabilities for Thermoluminescent Dosimeters Used for Environmental Measurements Required detection capabilities for thermoluminescent dosimeters used for environmental measurements required by the Technical Specifications are based on ANSI Standard N545, section 4.3. TLDs are defined as phosphors packaged for field use. In regard to the detection capabilities for thermoluminescent dosimeters, only one determination is required to evaluate the above capabilities per type of TLD. Furthermore, the above capabiliti.es may be determined by the vendor who supplies the TLDs. Required detection capabilities are as follows.

4.3.1 Uniformity shall be determined by giving TLDs from the same batch an exposure equal to that resulting from an exposure rate of 10 uR/hr during the field cycle. The responses obtained shall have a relative standard deviation of less than 7.5%. A total of at least 5 TLDs shall be evaluated.

4.3.2 Reproducibility shall be determined by giving TLDs repeated exposures equal to that resulting from an exposure rate of 10 uR/hr during the field cycle. The average of the relative stand~rd deviations of the responses shall be less than 3.0%. A total of at least 4 TLDs shall be evaluated.

4.3.3 Dependence of exposure interpretation on the length of a field cycle shall be examined by placing TLDs for a period equal to at least a field cycle and a period equal to half the same field cycle in an area where the exposure rate is known to be constant. This test shall be conducted under approximate average winter temperatures and approximate average summer temperatures. For these tests, the ratio of the response obtained in the field cycle to twice that obtained for half the field cycle shall not be less than 0.85. At least 6 TLDs shall be evaluated.

4.3.4 Energy dependence shall be evaluated by the response of TLDs to photons for several energies between approximately 30 keV and 3 MeV. The response shall not differ from that obtained with the calibration source by more than 25% for photons with energies greater than 80 keV and shall not be enhanced by more than a factor of two for photons with energies less than 80 keV. A total of at least 8 TLDs shall be evaluated.

4.3.5 The directional dependence of the TLD response shall be determined by comparing the response of the TLD exposed in the routine orientation with respect to the calibration source with the response obtained for different orientations. To accomplish this, the TLD shall be rotated through at least two perpendicular planes.

The response averaged over all directions shall not differ from the response obtained in the standard calibration position by more than 10%. A total of at least 4 TLDs shall be evaluated.

Unit2 Revision 37 II 31 September 2020

4.3.6 Light dependence shall be determined by placing TLDs in the field for a period equal to the field cycle under the four conditions found in ANSI N545, section 4.3.6. The results obtained for the unwrapped TLDs shall not differ from those obtained for the TLDs wrapped in aluminum foil by more than 10%. A total of at least 4 TLDs shall be evaluated for each of the four conditions.

4.3.7 Moisture dependence shall be determined by placing TLDs (that is, the phosphors packaged for field use) for a period equal to the field cycle in an area where the exposure rate is known to be constant. The TLDs shall be exposed under two conditions: (1) packaged in a thin, sealed plastic bag, and (2) packaged in a thin, sealed plastic bag with sufficient water to yield observable moisture throughout the field cycle. The TLD or phosphor, as appropriate, shall be dried before readout.

The response of the TLD exposed in the plastic bag containing water shall not differ from that exposed in the regular plastic bag by more than 10%. A total of at least 4 TLDs shall be evaluated for each condition.

4.3.8 Self irradiation shall be determined by placing TLDs for a period equal to the field cycle in an area where the exposure rate is less than 10 µR/hr and the exposure during the field cycle is known. If necessary, corrections shall be applied for the dependence of exposure interpretation on the length of the field cycle (ANSI N545, section 4.3.3). The average exposure inferred from the responses of the TLDs shall not differ from the known exposure by more than an exposure equal to that resulting from an exposure rate of 10 uR/hr during the field cycle. A total of at least 3 TLDs shall be evaluated.

Unit2 Revision 37 II 32 September 2020

TABLED 2-1 LIQUID EFFLUENT DETECTORS RESPONSES*

NUCLIDE (CPM/µCi/ml x 108)

Sr 89 0.78E-04 Sr 91 1.22 Sr92 0.817 Y91 2.47 Y92 0.205 Zr95 0.835 Nb95 0.85 Mo99, 0.232 Tc99m 0.232 Te 132 1.12 Ba 140 0.499 Ce 144 0.103 Br 84 1.12 I 131 1.01 I 132 2.63 I 133 0.967 I 134 2.32 I 135 1.17 Cs 134 1.97 Cs 136 2.89 Cs 137 0.732 Cs 138 1.45 Mn54 0.842 Mn56 1.2 Fe 59 0.863 Co 58 1.14 Co 60 1.65

,

Values from SWEC purchase specification NMP2-P281F.

Unit 2 Revision 37 II 33 September 2020

TABLED2-2 A1at VALUES - LIQUID 1 ADULT mrem- ml hr- µCi NUCLIDE TBODY GI-TRACT BONE LIVER KIDNEY THYROID LUNG H3 3.67E-1 3.67E-1 3.67E-1 3.67E-1 3.67E-1 3.67E-1 Cr 51 1.26 3.13£2 l.18E-2 l.18E-2 2.86E-1 7.56E-1 1.66 Cu64 1.28 2.33£2 2.73 6.89 Mn54 8.38£2 1.34£4 3.98 4.38£3 l.31E3 3.98 3.98 Fe 55 1.07£2 2.62£2 6.62£2 4.57£2 2.55£2 Fe 59 9.28E2 8.06£3 1.03£3 2.42£3 7.53E-1 7.53E-1 6.76£2 Co57 5.43El 5.36£2 2.1 lEl Co58 2.01£2 1.81£3 1.07 9.04El 1.07 1.07 1.07 Co 60 6.36£2 4.93£3 6.47El 3.24E2 6.47El 6.47El 6.47El Zn 65 3.32£4 4.63£4 2.31£4 7.35E4 4.92E4 2.21 2.21 Sr 89 6.38£2 3.57E3 2.22E4 6.18E-5 6.18E-5 6.18E-5 6.18E-5 Sr 90 l.36E5 1.60£4 5.55£5 Sr 92 l.44E-2 6.61 3.34E-1 Zr95 7.59E-1 2.83£2 9.77E-1 7.88E-1 8.39E-1 6.99E-1 6.99E-1 Mn56 3.07E-2 5.52 l.73E-1 2.20E-1 Mo99 l.60El l.95E2 l.97E-3 8.42El l.91E2 l.97E-3 l.97E-3 Na24 1.34E2 1.34E2 1.34E2 l.34E2 1.34E2 1.34E2 1.34£2 I 131 1.16E2 5.36El 1.42E2 2.03£2 3.48£2 6.65E4 2.77E-2 1132 4.34E-3 2.33E-3 4.64E-3 l.24E-2 1.98E-2 4.34E-1 I 133 l.22El 3.59El 2.30El 3.99El 6.97El 5.87E3 I 135 l.32EO 3.79EO l.28EO 3.36EO 5.39EO 2.22E2 Ni 65 l.14E-2 6.35E-1 l .93E-1 2.50E-2 Cs 134 5.79E5 l.24E4 2.98£5 7.08£5 2.29E5 2.04El 7.61£4 Cs 136 8.42E4 1.33£4 2.96E4 1.17£5 6.51E4 3.28E-1 8.92E3 Cs 137 3.42E5 l.01E4 3.82E5 5.22E5 1.77£5 3.lOEl 5.89E4 Ba 140 l.37El 4.30£2 2.09£2 3.04E-1 1.3 lE-1 4.17E-2 l.92E-1 Ce 141 3.79E-2 8.81El 6.93E-2 5.83E-2 4.60E-2 3.53E-2 3.53E-2 Nb95m l.51El l.44E6 3.53El 2.74El 2.70El Nb95 1.31£2 1.48E6 4.38E2 2.44E2 2.41£2 3.56E-1 3.56E-1 La 140 l.62E-2 3.72£3 l.03E-1 5.36E-2 2.83E-3 2.83E-3 2.83E-3 Ce 144 3.03E-1 6.15E2 2.02 9.66E-1 6.57E-1 2.06E-1 2.06E-1 Tc99m 2.05E-2 9.54£-01 5.71E-4 l.61E-3 2.45E-2 7.90E-4 Np239 l.8E-3 4.47£2 2.28E-2 2.78E-3 7.40E-3 5.95E-4 5.95E-4 Te 132 1.18£3 5.97£4 1.95£3 l.26E3 l.22E4 1.39£3 2.66E-3 Zr97 5.08E-4 3.39E2 5.44E-3 l.lOE-3 1.66E-3 7.1 lE-6 7.1 lE-6 W 187 4.31El 4.04£4 1.48E2 l.23E2 4.43E-5 4.43E-5 4.43E-5 Ag 110m l.09El 3.94E2 1.14El 1.13El 1.22El l.04El 1.04El Sb 124 4.72El 3.36E2 l.07E3 4.33El 4.3 lEl 4.31El 5.12El Zn69m 5.40El 3.60E4 2.46E2 5.90£2 3.57E2 6.90E-2 6.90E-2 Au 199 3.95 7.33£2 l.26E-1 4.67 l.79El l.26E-1 1.26E-1 As 76 5.94 l.24E4 l.60E-1 6.19 1.16El 1.60E-1 l.60E-1 Calculated in accordance with NUREG O133, Section 4.3.1, and Regulatory Guide I.I 09, Regulatory position C,Section I.

Unit2 Revision 37 II 34 September 2020

TABLED2-3 A1at VALVES - LIQUID1 TEEN mrem-ml hr- µCi NUCLIDE TBODY GI-TRACT BONE LIVER KIDNEY THYROID LUNG H3 2.73E-1 2.73E-1 2.73E-1 2.73E-1 2.73E-1 2.73E-1 Cr51 1.35 2.16E2 6.56E-2 6.56E-2 3.47E-1 7.79E-1 1.90 Cu64 1.35 2.23E2 2.87 7.27 Mn54 8.75E2 8.84E3 2.22El 4.32E3 1.31E3 2.22El 2.22El Fe 55 1.15E2 2.13E2 6.93E2 4.91E2 3.11E2 Fe 59 9.59E2 5.85E3 1.06E3 2.48E3 4.20 -4.20 7.84E2 Co 57 1.44E2 4.08E2 2.19El Co 58 2.10E2 1.23E3 5.98 9.47El 5.98 5.98 5.98 Co60 9.44E2 3.73E3 3.61E2 6.20E2 3.61E2 3.61E2 3.61E2 r

Zn65 3.40E4 3.08E4 2.10E4 7.28E4 4.66E4 1.24El 1.24El Sr 89 6.92E2 2.88E3 2.42E4 3.45E-4 3.45E-4 3.45E-4 3.45E-4 Sr 90 1.14E5 1.30E4 4.62E5 Sr92 1.54E-2 9.19El 3.61E-1 Zr95 ( 3.96 2.10E2 4.19 3.99 4.03 3.90 3.90 Mn56 3.22E-2 1.19El 1.81E-1 2.29E-l Mo99 1.71El 1.60E2 1.l0E-2 8.95El 2.05E2 1.l0E-2 1.l0E-2 Na24 1.38E2 1.38£2 l.38E2 1.38E2 l.38E2 1.38E2 1.38E2 I 131 1.14E2 4.21El 1.52E2 2.12E2 3.66E2 6.19E4 1.55E-1 I 132 4.56E-3 5.54£-3 4.86E-3 1.27E-2 2.00E-2 4.29E-1 I 133 l.28El 3.17El 2.47El 4.19El 7.35El 5.85E3 1.02E-4 I 135 1.76E0 3.84E0

1.34EO 3.46EO 5.47E0 2.23E2 Ni 65 1.21E-2 1.44 2.08£-1 2.66E-2 Cs 134 3.33E5 9.05E3 3.05E5 7.18E5 2.28E5 1.14E2 8.72E4 Cs 136 7.87E4 9.44E3 2.98E4 l.17E5 6.38E4 1.83 1.01E4 Cs 137 1.90E5 7.91E3 4.09E5 5.44E5 l.85E5 1.73E2 7.21E4 Ba 140 1.44El 3.40E2 2.21E2 5.03E-1 3.25E-1 2.33E-1 4.15E-1 Ce 141 2.00E-1 6.85El 2.33E-1 2.21E-1 2.08E-1 1.97E-1 l.97E-1 Nb95m l.69El 1.14E6 3.87El 2.99El 2.96El Nb95 1.17E2 l.05E6 4.43E2 2.47E2 2.39E2 1.99 1.99 La 140 2.97E-2 3.01E3 l.22E-1 6.82E-2 l.58E-2 1.58E-2 1.58E-2 Ce 144 1.25 4.83E2 3.07 1.94 1.62 1.15 1.15 Tc99m 2.1 lE-2 1.07 5.84E-4 l.63E-3 2.43E-2 9.04E-4 Np239 4.63E-3 3.78E2 2.82E-2 5.67E-3 l.07E-2 3.32E-3 3.32E-3 Te 132 1.23E3 4.13E4 2.06E3 1.30E3 l.25E4 l.37E3 l.48E-2 Zr 97 5.68E-4 3.l 1E2 5.84E-3 1.19E-3 l.78E-3 3.97E-5 3.97E-5 W 187 4.55El 3.52E4 l.59E2 l.30E2 2.47E-4 2.47E-4 2.47E-4 Ag 110m 5.85El 3.17E2 5.89El 5.88El 5.97El 5.79El 5.79El Sb 124 2.45E2 4.53E2 2.51E2 2.41E2 2.41E2 2.41E2 2.50E2 Zn69m 5.76El 3.43E4 2.65E2 6.24E2 3.79E2 3.85E-1 3.85E-1 Au 199 4.85 5.78E2 7.04E-1 5.60 2.0lEl 7.04E-1 7.04E-1 As76 7.18 1.06E4 8.92E-1 7.40 /

l.33El 8.92E-1 8.92E-1 1

Calculated m accordance with NUREG 0133, Section 4.3.1; and Regulatory Guide 1.109, Regulatory position C, Section 1 Unit2 Revision 37 II 35 September 2020

TABLED2-4 Aut VALUES - LIQUID 1 CHILD mrem- ml hr- µCi NUCLIQE TBODY GI-TRACT BONE LIVER KIDNEY THYROID LUNG H3 3.34E-1 3.34E-1 3.34E-1 3.34E-1 3.34E-1 3.34E-1 Cr51 1.39 7.29El l.37E-2 l.37E-2 2.22E-1 7.76E-1 1.41 Cu64 1.60 l.25E2 2.65 '6.41 Mn54 9.02E2 2.83E3 4.65 3.37E3 9.49E2 4.65 4.65 Fe 55 l.50E2 8.99El 9.15E2 4.85E2 2.74E2 Fe 59 1.04E3 2.18E3 l.29E3 2.09E3 8.78E-1 8.78E-1 6.08E2 Co 57 6.24El l.62E2 2.00El -'-

Co 58 2.21E2 4.20E2 1.25 7.30El 1.25 1.25 1.25 Co60 7.03E2 l.25E3 7.55El 2.88E2 7.55El 7.55El 7.55El Zn65 3.56E4 l.01E4 2.15E4 5.73E4 3.61E4 2.58 2.58 Sr 89 9.13E2 l.24E3 3.20E4 Sr90 l.06E5 5.62E3 4.17E5 \

Sr92 l.85E-2 8.73 4.61E-1 Zr95 8.95E-1 9.36El 1.22 9.04E-1 9.43E-1 8.15E-1 8.15E-1 Mn56 3.73E-2 2.39El l.65E-1 2.00E-1 Mo99 2.22El 7.42El 2.30E-3 8.98El l.92E2 2.30E-3 2.30E-3 Na24 1.51E2 l.51E2 1.51E2 l.51E2 l.51E2 l.51E2 l.51E2 I 131 l.14E2 1.80El 2.00E2 2.01E2 3.31E2 6.66E4 3.23E-2 I 132 5.0SE-3 l.30E-2 6.0lE-3 l.lOE-2 l.69E-2 5.13E-1 I 133 l.51El l.60El 3.22El 3.98El 6.64El 7.40E3 I 135 l.53EO 2.30EO 1.68EO 3.02EO 4.63EO 2.67E2 Ni65 1.46E-2 3.07 2.66E-1 2.51E-2 Cs 134 l.27E5 3.28E3 3.68E5 6.04E5 l.87E5 2.38El 6.72E4 Cs 136 6.26E4 3.40E3 3.52E4 9.67E4 5.15E4 3.82E-1 7.68E3 Cs 137 7.28E4 3.12E3 5.15E5 4.93E5 l.61E5 3.62El 5.78E4 Ba 140 l.87El l.62E2 3.19E2 3.28E-1 1.40E-1 4.87E-2 2.15E-1 Ce 141 4.61E-2 4.14El l .08E-1 7.43E-2 5.57E-2 4.12E-2 4.12E-2 Nb95m 2.14El 5.28E5 4.99El 2.92El 2.68El Nb95 1.45E2 3.75E5 5.21E2 2.03E2 l.91E2 4.16E-1 4.16E-1 La 140 l.93E-2 l.33E3 1.39E-1 5.09E-2 3.30E-3 3.30E-3 3.30E-3 Ce 144 4.3 lE-1 2.92E2 3.81 1.36 8.61E-1 2.40E-1 2.40E-1 Tc99m 2.29E-2 - 7.87E-1 7.05E-4 l .38E-3 2.0IE-2 7.02E-4 Np239 2.40E-3 1.79E2 1 3.44E-2 3.12E-3 7.70E-3 6.94E-4 6.94E-4 Te 132 l.38E3 l.15E4 2.57E3 1.14E3 l.06E4 l.66E3 3.lOE-3 Zr 97 6.99E-4 l.77E2 8.1 lE-3 1.18E-3 l.69E-3 8.29E-6 8.29E-6 W 187 5.37El l.68E4 2.02E2 1.20E2 5.16E-5 5.16E-5 5.16E-5 Ag 110m 1.29El 1.24E2 l .35El l.30El l.39El 1.21El l.21El Sb 124 5.69El l.68E2 6.92El 5.06El 5.03El 5.04El 6.08El Zn69m 6.80El l.87E4 3.37E2 5.75E2 3.34E2 8.05E-2 8.05E-2 Au 199 5.58 2.75E2 l.47E-1 5.02 1.80El l.47E-1 l.47E-1 As 76 8.31 5.47E3 l.86E-1 6.58 1.15El l.86E-1 l.86E-1 1

Calculated in accordance with NUREG 0133, Section 4.3.1; and Regulatory Guide 1.109, Regulatory position C, Section 1.

Unit 2 Revision 37 II 36 September 2020

TABLED2-5 Aiat VALUES - LIQUID 1 INFANT mrem- ml hr - µCi NUCLIDE TBODY GI-TRACT BONE LIVER KIDNEY THYROID LUNG H3 1.87E-1 1.87E-1 l.87E-1 1.87E-1 1.87E-1 l.87E-1 Cr 51 8.21E-3 2.39E-1 1.17E-3 5.36E-3 l.04E-2 Cu64 l.96E-2 8.70E-1 4.24E-2 7.17E-2 Mn54 2.73 4A-2 1.20El 2.67 Fe 55 1.45 6.91E-1 8.42 5.44 2.66 Fe 59 1.25El 1.52El 1.82El 3.18El 9.41 Co 57 l.13E0 2.37E0 6.95El Co 58 5.36 5.36 2.15 Co 60 l.55El l.56El 6.55 Zn65 l.76El 3.22El 1.1 lEl 3.81El l.85El Sr89 4.27El 3.06El 1.49E3 Sr90 2.86E3 1.40E2 l.12E4 Sr92 1.56E-5 4.54E-3 4.21E-4 Zr95 2.12E-2 l.49El l.23E-1 2.99E-2 3.23E-2 Mn56 l.81E-6 9.56E-4 l.05E-5 9.05E-6 Mo99 2.65 4.48 1.36El 2.03El Na24 9.61E-1 9.61E-1 9.61E-1 9.61E-1 9.61E-1 9.61E-1' 9.61E-1 I 131 9.78 7.94E-1 1.89El 2.22El 2.60El 7.31E3 I 132 3.43E-6 7.80E-6 4.75E-6 9.63E-6 1.07E-5 4.52E-4 1133 8.26E-1 4.77E-1 1.94 2.82 3.31 5.13E2 I 135 2.38E2 2.36E2 3.29E2 6.54E2 7.28E2 5.86E0 Ni 65 2.96E-6 4.96E-4 5.75E-5 6.51E-6 Cs 134 4.30El 1.16 2.28E2 4.26E2 1.10E2 4.50El Cs 136 2.81El 1.14 2.56El 7.53El 3.00El 6.13 Cs 137 2.63El 1.16 3.17E2 3.71E2 9.95El 4.03El Ba140 4.88 2.33El 9.48El 9.48E-2 2.25E-2 5.82E-2 Ce 141 3.3 lE-3 1.45El 4.61E-2 2.81E-2 8.67E-3 Nb95m l.02E3 l.20El 2.39E3 l.73E3 1.10E3 Nb95 5.87E-3 8.57 2.47E-2 l.02E-2 7.28E-3 La 140 6.52E-4 2.98El 6.43E-3 2.53E-3 Ce 144 1.0lE-1 1.03E2 1.80 7.37E-1 2.98E-1 Tc99m 3.17E-4 7.14E-3 1.19E-5 2.46E-5 2.64E-4 1.28E-5 Np239 2.08E-4 l.06El 4.12E-3 3.68E-4 7.34E-4 Te 132 4.08 l.62El 8.83 4.37 2.74El 6.46 Zr97 l.38E-4 1.92El 1.76E-3 3.02E-4 3.04E-4 W 187 4.13E-2 7.02 1.72E-1 1.19E-1 Ag 110m 2.91E-1 2.28El 6.02E-1 4.39E-1 6.28E-1 Sb 124 3.95 3.93El l.27El 1.87E-1 3.38E-2 7.98 Zn69m 2.30E-2 3.50 1.24E-1 2.52E-1 1.02E-1 Au 199 2.23E-1 5.38 2.48E-1 6.26E-1 1

As 76 8.67E-2 2.85El 8.46E-2 1.03E-1 1Calculated in accordana; with NUREG O133, Section 4.3.1; and Regulatory Guide 1.109, Regulatory position C, Section 1.

Unit2 Revision 37 II 37 September 2020

TABLED3-l OFFGASPRETREATMENT*

DETECTOR RESPONSE NUCLIDE NET CPM/µCi/cc Kr83m Kr85 4.28E+03 Kr85m 3.85E+o3 Kr87 6.68E+o3 Kr88 3.97E+03 Kr89 6.48E+03 Xe 131m Xe 133 1.69E+o3 Xe 133m Xe 135 4.91E+03 Xe 135m Xe 137 6.89E+03 )

Xe 138 5.51E+o3

Values from calculation H21C-070 Unit2 Revision 37 II 38 September 2020

TABLED3-2 PLUME SHINE P ARAMETERS 1 NUCLIDE B1 mrad/yr V1 mrem/vr

µCi/sec µCi/sec Kr83m 9.0lE-7 Kr85 6.92E-7 Kr85m 5.09E-4 4.91E-4 Kr 87 2.72E-3 2.57E-3 Kr88 7.23E-3 7.04E-3 Kr 89 l .15E-2 1.13E-2 Kf 90 6.57E-3 4.49E-3 Xe 131m 7.76E-6 Xe 133 7.46E-5 6.42E-5 Xe 133m 4.79E-5 3.95E-5 Xe 135 7.82E-4 7.44E-4 Xe 135m 1.45E-3 l.37E-3 Xe 137 6.25E-4 5.98E-4 Xe 138 4.46E-3 4.26E-3 Xe-127 l.96E-3 l.31E-3 Ar41 5.00E-3 4.79E-3 B1 and V1 are calculated for critical site boundary location; 1.6km in the easterly direction. See Appendix B. Those values that show a dotted line were negligible

' because of high energy absorption coefficients.

Unit2 Revision 37 II 39 September 2020

TABLED3-3 IMMERSION DOSE FACTORS 1 Nuclide I(i{y -Body)2 Lt{BD-Skin)2 Mi{y -Air) 3 'Nj(DB-Air}3 Kr83m 7.56E-02 l.93El 2.88E2

~85m l.17E3 1.46E3 1.23E3 l.97E3 Kr85 l.61El l.34E3 l.72El 1.95E3 Kr87 5.92E3 9.73E3 6.17E3 l.03E4 Kr 88 l.47E4 2.37E3 1.52E4 2.93E3 Kr 89 l.66E4 l.01E4 l.73E4 l.06E4 Kr90 1.56E4 7.29E3 l.63E4 7.83E3 Xe 131m 9.15El 4.76E2 l.56E2 l.11E3 Xe 133m 2.51E2 9.94E2 3.27E2 1.48E3 Xe 133 2.94E2 3.06E2 3.53E2 l.05E3 Xe 135m 3.12E3 7.11E2 3.36E3 7.39E2 Xe 135 l.81E3 l.86E3 l.92E3 2.46E3 Xe 137 l.42E3 1.22E4 l.51E3 l.27E4 Xe 138 8.83E3 4.13E3 9.21E3 4.75E3 Ar41 8.84E3 2.69E3 9.30E3 3.28E3

/

1 From, Table B-1.Regulatory Guide 1.109 Rev. 1 2mrem/yr per µCi/m 3 . l..._

3 mrad/yr per µCi/m 3 .

Unit 2 Revision 37 II 40 September 2020

TABLED3-4 DOSE AND DOSE RATE Rt VALVES - INHALATION - INFANT 1 mrem/vr

µCi/m 3 NUCLIDE BONE LIVER T.BODY THYROID KIDNEY LUNG GI-LLI H3 6.47E2 6.47E2 6.47E2 6.47E2 6.47E2 6.47E2 C 14 2.65E4 5.31E3 5.31E3 5.31E3 5.31E3 5.31E3 5.31E3 Cr 51 8.95El 5.75El l.32El l.28E4 3.57E2 Mn54 2.53E4 4.98E3 4.98E3 l.00E6 7.06E3 Fe 55 l.97E4 l.17E4 3.33E3 8.69E4 l.09E3 Fe 59 1.36E4 2.35E4 9.48E3 l.02E6 2.48E4 Co 58 1.22E3 l.82E3 7.77E5 l.11E4 Co60 8.02E3 l.18E4 4.51E6 3.19E4 Zn65 l.93E4 6.26E4 3.11E4 3.25E4 6.47E5 5.14E4 Sr 89 3.98E5 l.14E4 2.03E6 6.40E4 Sr 90 4.09E7 2.59E6 l.12E7 1.31E5 Zr95 l.15E5 2.79E4 2.03E4 3.11E4 l.75E6 2.17E4 Nb95 l.57E4 6.43E3 3.78E3 4.72E3 4.79E5 l.27E4 Mo99 1.65E2 3.23El 2.65E2 l.35E5 4.87E4 Ru 103 2.02E3 6.79E2 4.24E3 5.52E5 l.61E4 Ag 110m 9.99E3 7.22E3 5.00E3 l.09E4 3.67E6 3.30E4 1-131 3.79E4 4.44E4 l.96E4 1.48E7 5.18E4 l.06E3 I 133 1.32E4 l.92E4 5.60E3 3.56E6 2.24E4 2.16E3 Cs 134 3.96E5 7.03E5 7.45E4 l.90E5 7.97E4 l.33E3 Cs 137 5.49E5 6.12E5 4.55E4 l.72E5 7.13E4 l.33E3 Ba 140 5.60E4 5.60El 2.90E3 l.34El l.60E6 - 3.84E4 La 140 5.05E2 2.00E2 5.15El 1.68E5 8.48E4 Ce 141 2.77E4 l.67E4 1.99E3 5.25E3 5.l 7E5 2.16E4 Ce 144 3.19E6 l.21E6 l.76E5 5.38E5 9.84E6 1.48E5 Nd 147 7.94E3 8.13E3 5.00E2 3.15E3 3.22E5 3.12E4

)

1 This and following R1 Tables Calculated in accordance with NUREG 0133, Section 5.3.1, except C 14 values in accordance with Regulatory Guide 1.109 Equation C-8.

Unit2 Revision 37 II 41 September 2020

\

TABLED3-5 DOSE AND DOSE RATE Rt V ALOES - INHALATION - CHILD mrem/yr

µCi/m 3 NUCLIDE BONE LIVER T.BODY lflRMioib KIDNEY tirNd GI-LLI H3 l.12E3 1.12E3 l.12E3 l.12E3 1.12E3 l.12E3 C 14 , 3.59E4 6.73E3 6.73E3 6.73E3 6.73E3 6.73E3 6.73E3 Cr 51 1.54E2 8.55El 2.43El 1.70E4 1.08E3 Mn54 4.29E4 9.51E3 l.00E4 1.58E6 2.29E4 Fe 55 4.74E4 2.52E4 7.77E3 1.11E5 2.87E3 Fe 59 2.07E4 3.34E4 1.67E4 1.27E6 7.07E4 Co 58 l.77E3 3.16E3 l.11E6 3.44E4 Co60 l.31E4 2.26E4 7.07E6 9.62E4 Zn 65 4.26E4 1.13E5 7.03E4 7.14E4 9.95E5 l.63E4 Sr 89 5.99E5 1.72E4 2.16E6 l.67E5 Sr 90 l.01E8 6.44E6 1.48E7 3.43E5 Zr95 1.90E5 4.18E4 3.70E4 5.96E4 2.23E6 6.11E4 I

Nb95 2.35E4 9.18E3 6.55E3 8.62E3 6.14E5 3.70E4 Mo99 l.72E2 4.26El 3.92E2 1.35E5 1.27E5

- Ru 103 2.79E3 l.07E3 7.03E3 6.62E5 4.48E4 Ag 110m 1.69E4 l.14E4 9.14E3 2.12E4 5.48E6 l.00E5 I 131 4.81E4 4.81E4 2.73E4 l.62E7 7.88E4 2.84E3 I 133 l.66E4 2.03E4 7.70E3 3.85E6 3.38E4 5.48E3 Cs 134 6.51E5 1.01E6 2.25E5 3.30E5 1.21E5 3.85E3 Cs 137 9.07E5 8.25E5 1.28E5 2.82E5 1.04E5 3.62E3 Ba 140 7.40E4 6.48El 4.33E3 2.1 lEl 1.74E6 l.02E5 La 140 6.44E2 2.25E2 7.55El 1.83E5 2.26E5 Ce 141 3.92£4 l.95E4 2.90£3 8.55E3 5.44E5 5.66E4 Ce 144 6.77E6 2.12E6 3.61E5 1.17£6 1.20E7 3.89E5 Nd 147 1.08E4 8.73E3 6.81E2 4.81E3 3.28E5 8.21E4 Unit2 Revision 37 II 42 September 2020

TABLED3-6 DOSE AND DOSE RATE Ri VALUES - INHALATION - TEEN mrem/yr

µCi/m 3 NUCLIDE BONE LIVER T.BODY THYROID KIDNEY LUNG GI-LLI H3 1.27E3 1.27E3 1.27E3 1.27E3 1.27E3 1.27E3 C 14 2.60E4 4.87E3 4.87E3 4.87E3 4.87E3 4.87E3 4.87E3 Cr 51 1.35E2 7.50El 3.07El 2.10E4 3.00E3 Mn54 5.11E4 8.40E3 l.27E4 1.98E6 6.68E4 Fe 55 3.34E4 2.38E4 5.54E3 1.24E5 6.39E3 Fe 59 1.59E4 3.70E4 1.43E4 1.53E6 1.78E5 Co 58 2.07E3 2.78E3 1.34E6 9.52E4 Co60 1.51E4 l.98E4 8.72E6 2.59E5 Zn65 3.86E4 1.34E5 6.24E4 8.64E4 l.24E6 4.66E4 Sr 89 4.34E5 1.25E4 2.42E6 3.71E5 Sr 90 1.08E8 6.68E6 l.65E7 7.65E5 Zr 95 1.46E5 4.58E4 3.15E4 6.74E4 2.69E6 1.49E5 Nb95 l.86E4 l.03E4 5.66E3 l.00E4 7.51E5 9.68E4 Mo99 l.69E2 3.22El 4.11E2 1.54E5 2.69E5 Ru 103 2.10E3 8.96E2 7.43E3 7.83E5 1.09E5 Ag 110m 1.38E4 1.31E4 7.99E3 2.50E4 6.75E6 2.73E5 I 131 3.54E4 4.91E4 2.64E4 1.46E7 8.40E4 6.49E3 I 133 1.22E4 2.05E4 6.22E3 ! 2.92E6 3.59E4 1.03E4 Cs 134 5.02E5 1.13E6 5.49E5 3.75E5 1.46E5 9.76E3 Cs 137 6.70E5 8.48E5 3.11E5 3.04E5 1.21E5 8.48E3 Ba 140 5.47E4 6.70El 3.52E3 2.28El 2.03E6 2.29E5 La 140 4.79E2 2.36E2 6.26El 2.14E5 4.87E5 Ce 141 2.84E4 l.90E4 2.17E3 8.88E3 6.14E5 l.26E5 Ce 144 4.89E6 2.02E6 2.62E5 1.21E6 l.34E7 8.64E5 Nd 147 7.86E3 8.56E3 5.13E2 5.02E3 3.72E5 1.82E5 Unit2 Revision 37 II 43 September 2020

TABLED3-7 DOSE AND DOSE RATE Rt VALUES - INHALATION - ADULT mremll:'.r

µCi/m 3 NUCLIDE BONE LIVER -T.BODY THYROID KIDNEY LUNG GI-LLI H3 1.26E3 l.26E3 1.26E3 1.26E3 l.26E3 1.26E3 C 14 l.82E4 3.41E3 3.41E3 3.41E3 3.41E3 3.41E3 3.41E3 Cr 51 l.00E2 5.95El 2.28El l.44E4 3.32E3 Mn54 3.96E4 6.30E3 9.84E3 1.40E6 7.74E4 Fe 55 2.46E4 l.70E4 3.94E3 7.21E4 6.03E3 Fe 59 l.18E4 2.78E4 l.06E4 l.02E6 1.88E5 Co 58 1.58E3 2.07E3 9.28E5 l.06E5 Co 60 1.15E4 1.48E4 5.97E6 2.85E5 Zn65 3.24E4 1.03E5

4.66E4 6.90E4 8.64E5 5.34E4 Sr 89 3.04E5 8.72E3 1.40E6 3.50E5 Sr 90 9.92E7 6.10E6 9.60E6 7.22E5 Zr95 l.07E5 3.44E4 2.33E4 5.42E4 1.77E6 1.50E5 Nb95 1.41E4 7.82E3 4.21E3 7.74E3 5.05E5 1.04E5 Mo99 1.21E2 2.30El 2.91E2 9.12E4 2.48E5 Ru 103 1.53E3 6.58E2 5.83E3 5.05E5 1.10E5 Ag 110m 1.38E4 1.31E4 7.99E3 2.50E4 6.75E6 2.73E5 I 131 2.52E4 3.58E4 2.05E4 1.19E7 6.13E4 6.28E3 I 133 8.64E3 1.48E4 4.52E3 2.15E6 2.58E4 8.88E3 Cs 134 3.73E5 8.48E5 7.28E5 2.87E5 9.76E4 1.04E4 Cs 137 4.78E5 6.21E5 4.28E5 2.22E5 7.52E4 8.40E3 Ba140 3.90E4 4.90El 2.57E3 l.67El 1.27E6 2.18E5 La 140 3.44E2 1.74E2 4.58El l.36E5 4.58E5 Ce 141 l.99E4 l.35E4 l.53E3 6.26E3 3.62E5 1.20E5 Ce 144 3.43E6 l.43E6 l.84E5 8.48E5 7.78E6 8.16E5 Nd 147 5.27E3 6.10E3 3.65E2 3.56E3 2.21E5 l.73E5 Ag 110m l.38E4 l.31E4 7.99E3 2.50E4 6.75E6 2.73E5 Unit2 Revision 37 II 44 September 2020

TABLED3-8 DOSE AND DOSE RATE Ri VALUES- GROUND PLANE ALL AGE GROUPS m2 -mrem/yr

µCi/sec NUCLIDE TOTAL BODY SKIN H3 C 14 Cr 51 4.65E6 5.50E6 Mn54 1.40E9 l.64E9 Fe 55 Fe 59 2.73E8 3.20E8 Co 58 3.80E8 4.45E8 Co 60 2.15E10 2.53E10 Zn65 7.46E8 8.57E8 Sr 89 2.16E4 2.51E4 Sr 90 Zr95 2.45E8 2.85E8 Nb95 l.36E8 l.61E8 Mo99 3.99E6 4.63E6 Ru 103 1.08E8 1.26E8 Ag 110m 3.44E9 4.01E9 I 131 l.72E7 2.09E7 I 133 2.39E6 2.91E6 Cs134 6.83E9 7.97E9 Cs137 l.03E10 l.20E10 Ba 140 2.05E7 2.35E7 La 140 l.92E7 2.18E7 Ce 141 l.37E7 l.54E7 Ce 144 6.96E7 8.07E7*

Nd 147 8.46E6 l.01E7 Unit2 Revision 37 II 45 September 2020

TABLED3-9 DOSE AND DOSE RATE Ri VALUES - COW MILK - INFANT (

m 2-mrem/yr

µCi/sec NU~[fflE BO~ 'LIVER T.BODY THYR~ID KIDNE" LUNG GI-LLI H 3* 2.38E3 2.38E3 2.38E3 2.38E3 2.38E3 2.38E3 C 14* 3.23E6 6.89E5 6.89E5 6.89E5 6.89E5 6.89E5 6.89E5 Cr 51 8.35E4 5.45E4 l.19E4 l.06E5 2.43E6 Mn54 2.51E7 5.68E6 5.56E6 9.21E6 Fe 55 8.43E7 5.44E7 1.45E7 2.66E7 6.91E6 Fe 59 l .22E8 2.13E8 8.38E7 6.29E7 l.02E8 Co 58 l.39E7 3.46E7 3.46E7 Co 60 5.90E7 l.39E8 1.40E8 Zn 65 3.53E9 - l.21E10 5.58E9 5.87E9 l.02E10 Sr 89 6.93E9 1.99E8 1.42E8 Sr 90 8.19E10 2.09E10 l.02E9 Zr95I 3.85E3 9.39E2 6.66E2 l.01E3 4.68E5 Nb95 4.21E5 l.64E5 l.17E5 l.54E5 3.03E8 Mo99 l.04E8 2.03E7 l.55E8 3.43E7 Ru 103 4.65E3 l.55E3 9.67E3 5.65E4 Ag 110in 2.46E8 l.79E8 l.19E8 2.56E8 9.29E9 I 131 6.81E8 8.02E8 3.53E8 2.64El l 9.37E8 2.86E7 I 133 8.52E6 l.24E7 3.63E6 2.26E9 1.46E7 2.10E6 Cs 134 2.41E10 4.49E10 4.54E9 l.16E10 4.74E9 l.22E8 Cs 137 3.47E10 4.06E10 2.88E9 l.09E10 4.41E9 l.27E8 Ba 140 l.21E8 l.21E5 6.22E6 2.87E4 7.42E4 2.97E7 La 140 2.03El 7.99 2.06 9.39E4 Ce 141 2.28E4 l.39E4 l.64E3 4.28E3 7.18E6 Ce 144 1.49E6 6.10E5 8.34E4 2.46E5 8.54E7 Nd 147 4.43E2 4.55E2 2.79El l.76E2 2.89E5

mrem/yr per µCi/m 3*

Unit2 Revision 37 II 46 September 2020

TABLED 3-10 DOSE AND DOSE RATE Rt VALUES - COW MILK- CHILD m2-mrem/yr

µCi/sec NUCLIDE BONE LIVER T.BODY THYROID KIDNEY LUNG GI-LLI H3* 1.57£3 1.57E3 l.57E3 1.57£3 1.57£3 1.57£3 C 14* l.65E6 3.29E5 3.29E5 3.29£5 3.29E5 3.29£5 3.29£5 Cr 51 5.27£4 2.93E4 7.99£3 5.34£4 2.80£6 Mn54 l.35E7 3.59E6 3.78£6 1.13£7 Fe 55 6.97E7 3.07£7 l.15E7 2.09£7 6.85£6 Fe 59 6.52E7 l.06E8 5.26E7 3.06£7 1.10£8 Co 58 6.94£6 2.13E7 4.05£7 Co 60 2.89E7 8.52£7 1.60£8 Zn65 2.63E9 7.00E9 4.35E9 4.41£9 1.23£9 Sr 89 3.64E9 1.04£8 1.41£8 Sr 90 7.53E10 1.91£10 1.01£9 Zr 95 2.17£3 4.77£2 4.25E2 6.83£2 4.98£5 Nb95 l.86E5 1.03£4 5.69E4 l.00E5 4.42£8 Mo99 4.07E7 l.01E7 8.69£7 3.37£7 Ru 103 2.29E3 8.82E2 5.78£3 5.93£4 Ag 110m 1.33E8 8.97£7 7.17£7 1.67£8 1.07£10 I 131 3.26£8 3.28£8 1.86£8 l.08El 1 5.39£8 2.92£7 I 133 4.04E6 4.99£6 l.89E6 9.27£8 8.32E6 2.01£6 Cs 134 1.50£10 2.45£10 5.18E9 7.61£9 2.73£9 1.32£8 Cs 137 2.17£10 2.08£10 3.07E9 6.78E9 2.44£9 1.30£8 Ba 140 5.87E7 5.14£4 3.43E6 1.67£4 3.07£4 2.97£7 La 140 9.70 3.39 1.14 9.45£4 Ce 141 1.15£4 5.73£3 8.51£2 2.51£3 7.15£6 Ce 144 1.04£6 3.26£5 5.55E4 1.80£5 8.49£7 Nd 147 2.24E2 l.81E2 l.40El 9.94El 2.87£5

mrem/yr per µCi/m 3.

Unit2 Revision 37 II 47 September 2020

TABLED 3-11 DOSE AND DOSE RATE Rt VALVES - COW MILK - TEEN m2 -mrem/yr

µCi/sec NUCLIDE BONE LIVER T.BODY THYROID KIDNEY LUNG GI-LLI H3* 9.94E2 9.94E2 9.94E2 9.94E2 9.94E2 9.94E2 C 14* 6.70E5 l.34E5 l.34E5 l.34E5 l.34E5 l.35E5 l.34E5 Cr 51 2.58E4 1.44E4 5.66E3 3.69E4 4.34E6 Mn54 9.01E6 l.79E6 2.69E6 l.85E7 Fe 55 2.78E7 l.97E7 4.59E6 l.25E7 8.52E6 Fe 59 2.81E7 6.57E7 2.54E7 2.07E7 l.55E8 Co 58 4.55E6 l.05E7 6.27E7 Co 60 1.86E7 4.19E7 2.42E8 Zn65 l.34E9 4.65E9 2.17E9 2.97E9 l.97E9 Sr 89 l.47E9 4.21E7 l.75E8 Sr 90 4.45E10 1.l0ElO l.25E9 Zr 95 9.34E2 2.95E2 2.03E2 4.33E2 6.80E5 Nb95 l.86E5 l.03E5 5.69E4 l.00E5 4.42E8 Mo99 2.24E7 4.27E6 5.12E7 4.01E7 Ru 103 9.70E2 4.15E2 3.42E3 8.10E4 Ag 110m 6.13E7 5.80E7 3.53E7 l.11E8 l.63El0 I 131 l.34E8 l.88E8 l.01E8 5.49E10 3.24E8 3.72E7 I 133

~

l.66E6 2.82E6 8.59E5 3.93E8 4.94E6 2.l3E6 Cs 134 6.49E9 l.53E10 7.08E9 4.85E9 l.85E9 1.90E8 Cs 137 9.02E9 l.20E10 4.18E9 4.08E9 1.59E9 l.71E8 Ba 140 2.43E7 2.98E4 l.57E6 l.01E4 2.00E4 3.75E7 La 140 4.05 1.99 5.30E-1 l.14E5 Ce 141 4.67E3 3.12E3 3.58E2 l.47E3 8.91E6 Ce 144 4.22E5 l.74E5 2.27E4 1.04E5 l.06E8 Nd 147 9.12El 9.91El 5.94E0 5.82El 3.58E5

mrem/yr per µCi/m 3

Unit 2 Revision 37 II 48 September 2020

TABLED3-12 DOSE AND DOSE RATE Ri VALUES - COW MILK - ADULT m 2-mrem/yr

µCi/sec NUCLIDE BONE LIVER T.BODY THYROID KIDNEY LUNG GI-LLI H3° 7.63E2 7.63E2 7.63E2 7.63E2 7.63E2 7.63E2 C 14° 3.63E5 7.26E4 7.26E4 7.26E4 7.26E4 7.26E4 7.26E4 Cr 51 1.48E4 8.85E3 3.26E3 l.96E4 3.72E6 Mn54 5.41E6 l.03E6 l.61E6 l.66E7 Fe 55 1.57E7 1.08E7 2.52E6 6.04E6 6.21E6 Fe 59 1.61E7 3.79E7 1.45E7 1.06E7 l.26E8 Co 58 2.70E6 6.05E6 5.47E7 Co 60 1.10E7 2.42E7 2.06E8 Zn 65 8.71E8 2.77E9 l.25E9 l.85E9 l.75E9 Sr 89 7.99E8 2.29E7 l.28E8 Sr 90 3.15E10 7.74E9 9.11E8 Zr 95 5.34E2 l.71E2 l.16E2 2.69E2 5.43E5 Nb95 l.09E5 6.07E4 3.27E4 6.00E4 3.69E8 Mo99 l.24E7 2.36E6 2.81E7 2.87E7 Ru 103 5.45E2 2.35E2 2.08E3 6.34E4 Ag 110m 3.71E7 3.43E7 2.04E7 6.74E7 l.40El0 I 131 7.41E7 l.06E8 6.08E7 3.47El 0 l.82E8 2.80E7 I 133 9.09E5 1.58E6 4.82E5 2.32E8 2.76E6 1.42E6 Cs 134 3.74E9 8.89E9 7.27E9 2.88E9 9.55E8 1.56E8 Cs 137 4.97E9 6.80E9 4.46E9 2.31E9 7.68E8 l.32E8 Ba 140 l.35E7 l.69E4 8.83E5 5.75E3 9.69E3 2.77E7 La 140 2.26 1.14 3.0lE-1 8.35E4 Ce 141 2.54E3 l.72E3 l.95E2 7.99E2 6.58E6 Ce 144 2.29E5 9.58E4 l.23E4 5.68E4 7.74E7 Nd 147 4.74El 5.48El 3.28E0 3.20El 2.63E5

rnrem/yr per µCi/m 3*

Unit2 Revision 37 II 49 September 2020

TABLED3-13 DOSE AND DOSE RATE Ri VALVES - GOAT MILK - INFANT m 2-mrem/yr

µCi/sec NUCLIDE BONE LIVER T.BODY THYROID KIDNEY LUNG GI-LLI H3* 6.33E3 6.33E3 6.33E3 6.33E3 6.33E3 6.33E3 C 14* 3.23E6 6.89E5 6.89E5 6.89E5 6.89E5 6.89E5 6.89E5 Cr 51 l.00E4 6.56E3 1.43E3 1.28E4 2.93E5 Mn54 3.01E6 6.82E5 6.67E5 l.11E6 Fe 55 l.IOE6 7.08E5 l.89E5 3.46E5 8.98E4 Fe 59 1.59E6 2.78E6 l.09E6 8.21E5 l.33E6 Co 58 l.67E6 4.16E6 4.16E6 Co 60 7.08E6 l.67E7 1.68E7 Zn 65 4.24E8 1.45E9 6.70E8 7.04E8 1.23E9 Sr 89 1.48E10 4.24E8 3.04E8

\ ..

Sr 90 1.72Ell 4.38E10 2.15E9 Zr95 4.66E2 1.13E2 8.04El l.22E2 5.65E4 Nb95 9.42E4 3.88E4 2.24E4 2.78E4 3.27E7 Mo99 1.27E7 2.47E6 l.89E7 4.17E6 Ru 103 5.57E2 l.86E2 1.16E3 6.78E3 Ag 110m 2.95E7 2.15E7 1.43E7 3.07E7 l.11E9 I 131 8.17E8 9.63E8 4.23E8 3.16El l l.12E9 3.44E7 I 133 l.02E7 1.49E7 4.36E6 2.71E9 1.75E7 2.52E6 Cs 134 7.23E10 l.35El 1 1.36EIO 3.47E10 l.42E10 3.66E8 Cs 137 1.04El 1 1.22El 1 8.63E9 3.27E10 l.32E10 3.81E8 Ba 140 1.45E7 1.45E4 7.48E5 3.44E3 8.91E3 3.56E6 La 140 2.430 9.59E-1 2.47E-1 l.13E4 Ce 141 2.74E3 l.67E3 1.96E2 5.14E2 8.62E5 Ce 144 l.79E5 7.32E4 1.00E4 2.96E4 1.03E7 Nd 147 5.32El 5.47El 3.35E0 2.llEl 3.46E4

mrem/yr per µCi/m 3*

Unit 2 Revision 37 II 50 September 2020

TABLED3-14 DOSE AND DOSE RATE Rt VALUES - GOAT MILK - CHILD m.2-mrem/yr

µCi/sec NUCLIDE BONE LIVER T.BODY THYROID KIDNEY LUNG GI-LLI H 3* 4.17E3 4.l7E3 4.17E3 4.17E3 4.17E3 4.17E3 C 14* l.65E6 3.29E5 3.29E5 3.29E5 3.29E5 3.29E5 3.29E5 Cr 51 6.34E3 3.52E3 9.62E2 6.43E3 3.36E5 Mn54 l.62E6 4.31E5 4.54E5 l.36E6 Fe 55 9.06E5 4.81E5 l.49E5 2.72E5 8.91E4 Fe 59 8.52E5 l.38E6 6.86E5 3.99E5 l.43E6 Co 58 8.35E5 2.56E6 4.87E6 Co 60 3.47E6 l.02E7 l.92E7 Zn 65 3.15E8 8.40E8 5.23E8 5.29E8 l.48E8 Sr 89 7.77E9 2.22E8 3.01E8 Sr 90 l.58El 1 4.0lEl0 2.13E9 Zr 95 2.62E2 5.76El 5.13El 8.25El 6.01E4 Nb95 5.05E4 l.96E4 1.40E4 l.85E4 3.63E7 Mo99 4.95E6 1.22E6 l.06E7 4.09E6 Ru 103 2.75E2 l.06E2 6.93E2 7.12E3 Ag 110m 1.60E7 l.08E7 8.60E6 2.00E7 l.28E9 I 131 3.91E8 3.94E8 2.24E8 1.30Ell 6.46E8 3.50E7 I 133 4.84E6 5.99E6 2.27E6 1. l 1E9 9.98E6 2.41E6 Cs 134 4.49El0 7.37El0 1.55E10 2.28El0 8.19E9 3.97E8 Cs 137 6.52E10 6.24El0 9.21E9 2.03El0 7.32E9 3.91E8 Ba 140 7.05E6 6.18E3 4.12E5 2.01E3 3.68E3 3.57E6 La 140 1.16 4.07E-1 l.37E-l l.13E4 Ce 141 l.38E3 6.88E2 l.02E2 3.02E2 8.59E5 Ce 144 l.25E5 3.91E4 6.66E3 2.16E4 l.02E7 Nd 147 2.68El 2.17El l.68E0 1.19.El 3.44E4

mrem/yr per µCi/m 3*

Unit2 Revision 37 II 51 September 2020

TABLED3-15 DOSE AND DOSE RATE Ri VALUES - GOAT MILK - TEEN m2-mrem/yr

µCi/sec NUCLIDE BONE LIVER T.BODY THYROID KIDNEY LUNG GI-LLI H3* 2.64E3. 2.64E3 2.64E3 2.64E3 2.64E3 2.64E3 C 14* 6.70E5 1.34E5 1.34E5 1.34E5 1.34E5 l.35E5 1.34E5 Cr 51 3.11E3 l.73E3 6.82E2 4.44E3 5.23E5 Mn54 1.08E6 2.15E5 3.23E5 2.22E6 Fe 55 3.61E5 2.56E5 5.97E4 1.62E5 1.11E5 Fe 59 3.67E5 8.57E5 3.31E5 -- ( 2.70E5 2.03E6 Co 58 5.46E5 l.26E6 7.53E6 Co 60 i.23E6 5.03E6 2.91E7 Zn65 l.61E8 5.58E8 2.60E8 3.57E8 2.36E8 Sr 89 3.14E9 8.99E7 3.74E8 Sr 90 9.36E10 2.31E10 2.63E9 Zr 95 l.13E2 3.56El 2.45El 5.23El 8.22E4 Nb95 2.23E4 1.24E4 6.82E3 l.20E4 5.30E7 Mo99 2.72E6 5.19E5 6.23E6 4.87E6 Ru 103 1.16E2 4.98El 4.10E2 9.72E3

,:\g 110m 7.36E6 6.96E6 4.24E6 l.33E7 l.96E9 I 131 l.61E8 2.26E8 l.21E8 6.59EI0 3.89E8 4.47E7 I 133 l.99E6 3.38E6 1.03E6 4.72E8 5.93E6 2.56E6 Cs 134 l.95E10 4.58E10 2.13E10 1.46E10 5.56E9 5.70E8 Cs 137 2.71E10 3.60E10 1.25E10 1.23El0 4.76E9 5.12E8 Ba 140 2.92E6 3.58E3 l.88E5 1.21E3 2.41E3 4.50E6 La 140 4.86E-1 2.39E-1 6.36E-2 1.37E4 Ce 141 5.60E2 3.74E2 4.30El 1.76E2 l.07E6 Ce 144 5.06E4 2.09E4 2.72E3 1.25E4 1.27E7 Nd 147 1.09El l.19El 7.13E-1 6.99E0 4.29E4

mrern/yr per µCi/m 3*

Unit2 Revision 37 II 52 September 2020

TABLED3-16 DOSE AND DOSE RATE Rt VALUES- GOAT MILK-ADULT m2-mrem/yr

µCi/sec NUCLIDE BONE LIVER T.BODY THYROID KIDNEY LUNG GI-LLI H3* 2.03E3 2.03E3 2.03E3 2.03E3 2.03E3 2.03E3 C 14* 3.63E5 7.26E4 7.26E4 7.26E4 7.26E4 7.26E4 7.26E4 Cr 51 l.78E3 1.06E3 3.92E2 2.36E3 4.48E5 Mn54 6.50E5 l.24E5 l.93E5 l.99E6 Fe 55 2.04E5 l.41E5 3.28E4 7.85E4 8.07E4 Fe 59 2.10E5 4.95E5 l.90E5 1.38E5 l.65E6 Co 58 3.25E5 7.27E5 6.58E6 Co 60 l.32E6 2.91E6 2.48E7 Zn 65 1.05E8 3.33E8 l.51E8 2.23E8 2.10E8 Sr 89 1.70E9 4.89E7 2.73E8 Sr 90 6.62E10 l.63E10 1.91E9 Zr 95 6.45El 2.07El l.40El 3.25El 6.56E4 Nb95 1.31E4 7.29E3 3.92E3 7.21E3 4.42E7 Mo99 l.51E6 2.87E5 3.41E6 3.49E6 Ru 103 6.55El 2.82El 2.50E2 7.64E3 Ag 110m 4.45E6 4.12E6 2.45E6 8.09E6 l.68E9 I 131 8.89E7 1.27E8 7.29E7 4.17E10 2.18E8 3.36E7 I 133 l.09E6 l.90E6 5.79E5 2.79E8 3.31E6 1.71E6 Csl34 1.12E10 2.67E10 2.18E10 8.63E9 2.86E9 4.67E8 Cs 137 1.49E10 2.04E10 l.34E10 6.93E9 2.30E9 3.95E8 Ba 140 l.62E6 2.03E3 l.06E5 6.91E2 l.16E3 3.33E6 La 140 2.71E-1 l.36E-1 3.61E-2 l.00E4 Ce 141 3.06E2 2.07E2 2.34El 9.60El 7.90E5 Ce 144 2.75E4 l.15E4 1.48E3 6.82E3 9.30E6 Nd 147 5.69E0 6.57E0 3.93E-1 3.84E0 3.15E4

mrem/yr per µCi/m 3 .

Unit2 Revision 37 II 53 September 2020

TABLED3-17 DOSE AND DOSE RATE Ri VALUES - COW MEAT - CHILD m 2-mrem/yr

µCi/sec NUCLIDE BONE LIVER T.BODY THYROID KIDNEY LUNG GI-LLI H3* 2.34E2 2.34E2 2.34E2 2.34E2 2.34E2 2.34E2 C 14* 5.29E5 l.06E5 l.06E5 l.06E5 l.06E5 l.06E5 l.06E5 Cr 51 4.55E3 2.52E3 6.90E2 4.61E3 2:41E5 Mn54 5.15E6 1.37E6 l.44E6 4.32E6 Fe 55 2.89E8 1.53E8 4.74E7 8.66E7 2.84E7 Fe 59 2.04E8 3.30E8 1.65E8 9.58E7 3.44E8 Co 58 9.41E6 2.88E7 5.49E7 Co 60 4.64E7 l.37E8 2.57E8 Zn65 2.38E8 6.35E8 3.95E8 4.00E8 l.12E8 Sr 89 2.65E8 7.57E6 l.03E7 Sr 90 7.01E9 l.78E9 9.44E7 Zr95 l.51E6 3.32E5 - 2.95E5 4.75E5 3.46E8 Nb95 4.10E6 l.59E6 l.14E6 l.50E6 2.95E9 Mo99 5.42E4 1.34E4 l.16E5 4.48E4 Ru 103 8.30E7 3.19E7 2.09E8 2.15E9 Ag 110m 5.62E6 3.79E6 3.03E6 7.05E6 4.52E8 I 131 4.15E6 4.18E6 2.37E6 l.38E9 6.86E6 3.72E5 I 133 9.38E-2 l.16E-1 4.39E-2 2.15El l.93E-1 4.67E-2 Cs 134 6.09E8 l.00E9 2.11E8 3.10E8 l.11E8 5.39E6 Cs 137 8.99E8 8.60E8 l.27E8 2.80E8 l.01E8 5.39E6 Ba 140 2.20E7 l.93E4 1.28E6 6.27E3 l.15E4 l.11E7 La 140 2.80E-2 9.78E-3 3.30E-3 2.73E2 Ce 141 l.17E4 5.82E3 8.64-E2 2.55E3 7.26E6 Ce 144 1.48E6 4.65E5 7.91E4 2.57E5 l.21E8 Nd 147 5.93E3 4.80E3 3.72E2 2.64-E3 7.61E6

mrern/yr per µCi/m 3 .

Unit2 Revision 37 II 54 September 2020

TABLED3-18 DOSE AND DOSE RATE Ri VALUES - COW :MEAT -TEEN m 2-mrem/yr

µCi/sec NUCLIDE BONE LMR T.BODY THYROID KIDNEY tirNd GI-LLI H 3* l.94E2 l.94E2 1.94E2 l.94E2 l.94E2 l.94E2 C 14* 2.81E5 5.62E4 5.62E4 5.62E4 5.62E4 5.62E4 5.62E4 Cr 51 2.93E3 l.62E3 6.39E2 4.16E3 4.90E5 Mn54 4.50E6 8.93E5 1.34E6 9.24E6 Fe 55 l.50E8 l.07E8 2.49E7 6.77E7 4.62E7 Fe 59 l.15E8 2.69E8 l.04E8 8.47E7 6.36E8 Co 58 8.05E6 1.86E7 1.11E8 Co 60 3.90E7 8.80E7 5.09E8 Zn 65 l.59E8 5.52E8 2.57E8 3.53E8 2.34E8 Sr 89 1.40E8 4.01E6 1.67E7 Sr 90 5.42E9 1.34E9 l.52E8 Zr95 8.50E5 2.68E5 l.84E5 3.94E5 6.19E8 Nb95 2.37E6 l.32E6 7.24E5 1.28E6 5.63E9 Mo99 3.90E4 7.43E3 8.92E4 6.98E4 Ru 103 4.59E7 l.96E7 1.62E8 3.84E9 Ag 110m 3.39E6 3.20E6 1.95E7 6.13E6 9.01E8 I 131 2.24E6 3.13E6 1.68E6 9.15E8 5.40E6 6.20E5 I 133 5.05E-2 8.57E-2 2.61E-2 1.20El 1.50E-1 6.48E-2 Cs 134 3.46E8 8.13E8 3.77E8 2.58E8 9.87E7 l.01E7 Cs 137 4.88E8 6.49E8 2.26E8 2.21E8 8.58E7 9.24E6 Ba140 l.19E7 l.46E4 7.68E5 4.95E3 9.81EJ 1.84E7 La 140 l.53E-2 7.51E-3 2.00E-3 4.31E2 Ce 141 6.19E3 4.14E3 4.75E2 l.95E3 l.18E7 Ce 144 7.87E5 3.26E5 4.23E4 1.94E5 1.98E8 Nd 147 3.16E3 3.44E3 2.06E2 2.02E3 l.24E7

mrem/yr per µCi/m 3

Unit2 Revision 37 II 55 September 2020

TABLED3-19 DOSE AND DOSE RATE Rt VALUES - COW MEAT - ADULT m 2 -mrem/yr

µCi/sec NUCLIDE BONE LIVER *T.BODY THYROID KIDNEY LUNG GI-LLI H3* 3.25E2 3.25E2 3.25E2 3.25E2 3.25E2 3.25E2 C 14* 3.33E5 6.66E4 6.66E4 6.66E4 6.66E4 6.66E4 6.66E4 Cr 51 3.65E3 2.18E3 8.03E2 4.84E3 9.17E5 Mn54 5.90E6 1.13E6 l.76E6 l.81E7 Fe 55 l.85E8 1.28E8 2.98E7 7.14E7 7.34E7 Fe 59 1.44E8 3.39E8 1.30E8 9.46E7 1.13E9 Co 58 l.04E7 2.34E7 2.12E8 Co 60 5.03E7 l.11E8 9.45E8 Zn 65 2.26E8 7.19E8 3.25E8 4.81E8 4.53E8 Sr 89 1.66E8 4.76E6 2.66E7 Sr 90 8.38E9 2.06E9 2.42E8 Zr95 l.06E6 3.40E5 2.30E5 5.34E5 1.08E9 Nb95 3.04E6 1.69E6 9.08E5 l.67E6 1.03El0 Mo99 4.71E4 8.97E3 1.07E5 1.09E5 Ru 103 5.64E7 2.43E7 2.15E8 6.58E9 Ag 110m 4.48E6 4.14E6 2.46E6 8.13E6 l.69E9 I 131 2.69E6 3.85E6 2.21E6 1.26E9 6.61E6 l.02E6 I 133 6.04E-2 l.05E-1 3.20E-2 l.54El l.83E-1 9.44E-2 Cs 134 4.35E8 l.03E9 8.45E8 3.35E8 l.11E8 1.81E7 Cs 137 5.88E8 8.04E8 5.26E8 2.73E8 9.07E7 l.56E7 Ba 140 l.44E7 1.81E4 9.44E5 6.15E3 l.04E4 2.97E7 La 140 1.86E-2 9.37E-3 2.48E-3 6.88E2 Ce 141 7.38E3 4.99E3 5.66E2 2.32E3 l.91E7 Ce 144 9.33E5 3.90E5 5.01E4 2.31E5 3.16E8 Nd 147 3.59E3 4.15E3 2.48E2 2.42E3 1.99E7

mrem/yr per µCi/m 3*

Unit2 Revision 37 II 56 September 2020

TABLE ]) 3-20 DOSE AND DOSE RATE Ri VALUES- VEGETATION -CIDLD m 2-mrem/yr

µCi/sec NUCLIDE BONE LIVER T.BODY THYROID KIDNEY LUNG GI-LLI H3* 4.01E3 4.01E3 4.01E3 4.01E3 4.01E3 4.01E3 C 14* 3.50E6 7.01E5 7.01E5 7.01E5 7.01E5 7.01E5 7.01E5 Cr 51 1.17E5 6.49E4 l.77E4 l.18E5 6.20E6 Mn54 6.65E8 l.77E8 l.86E8 5.58E8 Fe 55 7.63E8 4.05E8 1.25E8 2.29E8 7.50E7 Fe 59 3.97E8 6.42E8 3.20E8 1.86E8 6.69E8 Co 58 6.45E7 l.97E8 3.76E8 Co 60 3.78E8 l.12E9 2.I0E9 Zn65 8.12E8 2.16E9 l.35E9 1.36E9 3.80E8 Sr 89 3.59EI0 1.03E9 l.39E9 Sr 90 1.24£12 3.15El 1 1.67£10 Zr95 3.86E6 8.50E5 7.56E5 l.22E6 8.86E8 Nb95 l.02E6 3.99E5 2.85E5 3.75E5 7.37E8 Mo99 7.70E6 l.91E6 l.65E7 6.37E6 Ru 103 l.53E7 5.90E6 3.86E7 3.97E8 Ag 110m 3.21E7 2.17E7 l.73E7 4.04E7 2.58E9 I 131 7.16E7 7.20E7 4.09E7 2.38£10 l.18E8 6.41E6 I 133 l.69E6 2.09E6 7.92E5 3.89£8, 3.49E6 8.44E5 Cs 134 l.60EI0 2.63E10 5.55E9 8.15E9 2.93E9 l.42E8 Cs 137 2.39El0 2.29El0 3.38E9 7.46E9 2.68E9 l.43E8 Ba 140 2.77E8 2.43E5 l.62E7 7.90E4 1.45E5 l.40E8 La 140 3.25E3 1..13E3 3.83E2 3.16E7 Ce 141 6.56E5 3.27E5 4.85E4 l.43E5 4.08E8 Ce 144 l.27E8 3.98E7 6.78E6 2.21E7 1.04£10 Nd 147 7.23E4 5.86E4 4.54E3 3.22E4 9.28E7

rnrem/yr per µCi/m 3

Unit2 Revision 37 II 57 September 2020

TABLED3-21 DOSE AND DOSE RATE HJ VALUES- VEGETATION -TEEN m2-mrem/yr

µCi/sec NUCLIDE BONE LIVER T.BODY THYROID KIDNEY LUNG GI-LLI H3* 2.59E3 2.59E3 2.59E3 2.59E3 2.59E3 2.59E3 C 14* 1.45E6 2.91E5 2.91E5 2.91E5 2.91E5 2.91E5 2.91E5 Cr 51 6.16E4 3.42E4 l.35E4 8.79E4 l.03E7 Mn54 4.54E8 9.01E7 1.36E8 9.32E8 Fe 55 3.10E8 2.20E8 5.13E7 1.40E8 9.53E7 Fe 59 l.79E8 4.18E8 l.61E8 1.32E8 9.89E8 Co 58 4.37E7 1.01E8 6.02E8 Co 60 2.49E8 5.60E8 3.24E9 Zn65 4.24E8 l.47E9 6.86E8 9.41E8 6.23E8 Sr 89 l.51E10 4.33E8 1.80E9 Sr 90 7.51El 1 l.85El 1 2.1 lEl0 Zr95 l.72E6 5.44E5 3.74E5 7.99E5 l.26E9 Nb95 4.80E5 2.66E5 l.46E5 2.58E5 l.14E9 Mo99 5.64E6 l.08E6 1.29E7 l.01E7 Ru 103 6.82E6 2.92E6 2.40E7 5.70E8 Ag 110m l.51E7 l.43E7 8.72E6 2.74E7 4.03E9 I 131 3.85E7 5.39E7 2.89E7 l.57E10 9.28E7 l.07E7 I 133 9.29E5 l.58E6 4.80E5 2.20E8 2.76E6 l.19E6 Cs 134 7.10E9 l.67E10 7.75E9 5.31E9 2.03E9 2.08E8 Cs 137 l.0lEl0 l.35E10 4.69E9 4.59E9 l.78E9 1.92E8 Ba 140 l.38E8 l.69E5 8.91E6 5.74E4 l.14E5 2.13E8 La 140 l.81E3 8.88E2 2.36E2 5.10E7 Ce 141 2.83E5 l.89E5 2.17E4 8.89E4 5.40E8 Ce 144 5.27E7 2.18E7 2.83E6 1.30E7 l.33El0 Nd 147 3.66E4 3.98E4 2.38E3 2.34E4 1.44E8 I

mrem/yr per µCi/m 3 Unit2 Revision 37 II 58 September 2020

TABLED3-22 DOSE AND DOSE RATE RiVALUES'-VEGETATION-ADULT m.2-mrem/yr

µCi/sec NUCLIDE BONE LIVER T.BODY THYROID KIDNEY LUNG GI-LLI H3* 2.26E3 2.26E3 2.26E3 2.26E3 2.26E3 2.26E3 C 14* 8.97E5 1.79E5 l.79E5 l.79E5 l.79E5 l.79E5 l.79E5 Cr 51 4.64E4 2.77E4 l.02E4 6.15E4 l.17E7 Mn54 3.13E8 5.97E7 9.31E7 9.58E8

/ Fe 55 2.00E8 l.38E8 3.22E7 7.69E7 7.91E7 Fe 59 1.26E8 2.'96E8 l.13E8 8.27E7 l.02E9 Co 58 3.08E7 6.90E7 6.24E8 Co 60 l.67E8 3.69E8 3.14E9 Zn65 3.17E8 l.01E9 4.56E8 6.75E8 6.36E8 Sr 89 9.96E9 2.86E8 l.60E9 Sr 90 6.05El 1 l.48El 1 l.75E10 Zr95 l'.18E6 3.77E5 2.55E5 5.92E5 1.20E9 Nb95 I 3.55E5 l.98E5 1.06E5 l.95E5 l.20E9 Mo99 6.14E6 l.17E6 1.39E7, 1.42E7 Ru 103 4.77E6 2.06E6 1.82E7 5.57E8 Ag 110m l.05E7 9.75E6 5.79E6 1.92E7 3.98E9 I 131 4.04E7 5.78E7 3.31E7 l.90E10 9.91E7 1.53E7 I 133 l.00E6 l.74E6 5.30E5 2.56E8 3.03E6 l.56E6 Cs 134 4.67E9 1.1 lEl0 9.08E9 3.59E9 l.19E9 1.94E8 Cs 137 6.36E9 8.70E9 5.70E9 2.95E9 9.81E8 l.68E8 Ba 140 1.29E8 l.61E5 8.42E6 5.49E4 9.25E4 2.65E8 La 140 l.98E3 9.97E2 2.63E2 7.32E7 Ce 141 l.97E5 l.33E5 l.51E4 6.19E4 5.09E8 Ce 144 3.29E7 l.38E7 l.77E6 *8.16E6 l.llEl0 Nd 147 3.36E4 3.88E4 2.32E3 2.27E4 l.86E8

mrern/yr per µCi/m 3 Unit2 Revision 37 II 59 September 2020

TABLED 3-23 DISPERSION PARAMETERS AT CONTROLLING LOCATIONS 1 X/Q,Wv and We VALUES VENT DIRECTION DISTANCE (ml X/Q (sec/m 3 ) D/Q (m*2)

Site Boundary2 E 1,600 2.00 E-6 2.1 0E-9 Inhalation and Ground E (104°) 1,800 1.42E-7 2.90E-9 Plane t

Cow Milk ESE (130°) 4,300 4.11 E-8 4.73E-10 Goat Milk3 SE (140°) 4,800 3.56E-08 5.32E-10 Meat Animal E (114°) 2,600 1.17E-7 1.86E-9 Vegetation E (96°) 2,900 1.04E-7 1.50E-9 STACK Site Boundary2 E 1,600 4.50E-8 6.00E-9 Inhalation and Ground E (109°) 1,700 8.48E-9 1.34E-9 Plane Cow Milk ESE (135°) 4,200 1.05E-8 3.64E-10 Goat Milk3 SE (140°) 4,800 2.90E-08 5.71 E-10 Meat Animal E (114°) 2,500 1.13E-8 1.15E-9

)

Vegetation E (96°) 2,800 1.38E-8 9.42E-10 NOTE: Inhalation and Ground Plane are annual average values. Others are grazing season only.

1 X/Q and D/Q values from NMP-2 ER-OLS.

2 X/Q and D/Q from NMP-2 FES, NUREG-1085, May 1985, Table D-2.

3 X/Q and D/Q from C.T. Main Data Report dated November 1985.

Unit2 Revision 37 II 60 September 2020

TABLED 3-24 PARAMETERS FOR THE EVALUATION OF DOSES TO REAL MEMBERS OF THE PUBLIC FROM GASEOUS AND LIQUID EFFLUENTS Pathway Parameter Value Reference Fish U (kg/yr) - adult 21 Reg. Guide 1.109 Table E-5 Fish DaipJ (mrem/pCi) Each Radionuclide Reg. Guide 1.109 Table E-11 Shoreline U (hr/yr)

- adult 67 Reg. Guide 1.109

- teen 67 Assumed to be Same as Adult Shoreline Ow~ Each Radionuclide Reg. Guide 1.109 (mrem/hrperpCum2) Table E-6 Inhalation DFAja Each Radionuclide Reg. Guide 1.109 Table E-7 Unit2 Revision 37 II 61 September 2020

TABLED 5.1 NINE MILE POINT NUCLEAR STATION RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM SAMPLING LOCATIONS

Map Collectlon Site TyQ!! of S!,!mE!le Location {§DY, ~rogrnm ~o.} Location Radioiod1ne and 1 Nine Mile Point Road North 1.8 m1@92°E Particulates (air) (R-1)

Radioiodine and 2 County Route 29 & Lake Road 1.1 mi@ 106" ESE Particulates (air) (R-2)

Rad101odine and 3 County Route 29 14m1@134°SE Particulates (air) (R-3)

Rad101odme and 4 Village of Lycoming, NY 1 8 mI @ 145' SE Particulates (air) (R-4)

Rad101odme and 5 Montano Point Road 16 2 mI @42' NE Particulates (air) (R-5)

Direct Radiation (TLD) 6 North Shoreline Area 0.1 mI @354" N (75)

Direct Radiation (TLD) 7 North Shoreline Area 01 m1@2T NNE (76)

Direct RadiatJon (TLD) 8 North Shoreline Area 0 2 m1@37" NE (77)

Direct Radiatlon (TLD) 9 North Shoreline Area 0 8 m1@74° ENE (23)

Direct Radiation (TLD) 10 JAF East Boundary 1.0 m1@86° E (78)

Direct Rad1at1on (TLD) 11 Route 29 1 2 mi@ 121" ESE (79)

Direct Rad1auon (TLD) 12 Route 29 15mi@136°SE (80)

Direct Radiation (TLD) 13 Miner Road 1 7 mi@ 160" SSE (81)

Direct Radiation (TLD) 14 Miner Road 1 6 m1@ 180" S (82)

Direct Radiation (TLD) 15 Lakeview Road 12m1@203°SSW (83)

Direct Radiabon (TLD) 16 LakeVJew Road 1.1 mi @225' SW (84)

Direct Radiation (TLD) 17 Srte Meteorological Tower 0 7 mi @ 244° WSW (7)

Direct Radiation (TLD) 18 Energy Information Center 05m1@266°W (18)

Direct Rad,abon (TLD) 19 North Shoreline 0.2 mI @ 290" WNW (85)

Unit2 Revision 37 II 62 September 2020

TABLE D 5.1 (Cont'd)

NINE MILE POINT NUCLEAR STATION RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM SAMPLING LOCATIONS

Map Collection Site Type of Sample Location CEnv. Program No,) Location Direct Radiation (TLD) 20 North Shoreline 0.1 mI @310° NW (86)

Direct Rad1at1on (TLD) 21 North Shoreline 0 1 mI @332" NNW (87)

Direct Radiation (TLD) 22 Hickory Grove 45ml@9T E (88)

Direct Radiation (TLD) 23 Leavitt Road 4.3 m1@ 112° ESE (89)

Direct Radiation (TLD) 24 Route 104 4.2 ml@ 135° SE (90)

Direct Radiation (TLD) 25 Route 51A 4 9 mI @ 15T SSE (91)

Direct Radiation (TLD) 26 Maiden Lane Road 4 5 m1@ 183" S (92)

Direct Radiabon (TLD) 27 County Route 53 4 4 mI @ 206° SSW (93)

Direct Radiation (TLD) 28 County Route 1 4.4 mI @ 224° SW (94)

Direct Rad1abon (TLD) 29 Lake Shoreline 3 7 m1@239°WSW (95)

Direct Radiabon (TLD) 30 Phoenix, NY Control 19 7 mI@ 168° SSE (49)

Direct Radiation (TLD) 31 S W Oswego, Control 125 m1@22T SW (14)

Direct Radiation (TLD) 32 Scnba, NY 3 7 mi @ 199° SSW (96)

Direct Rad1at1on (TLD) 33 Novells, Route 1A 3 0 mI @ 222° SW (58)

Direct Radiation (TLD) 34 Lycoming, NY 1 8 mI @ 145° SE (97)

Direct RadIabon (TLD) 35 New Haven, NY 5 2 m1@ 124° SE (56)

Direct Radiation (TLD) 36 W Boundary, Bible Camp 0 9 mI @ 239° WSW (15)

Direct Radiation (TLD) 37 Lake Road 1.2 m1@ 103° ESE (98)

Unit2 Revision 37 II 63 September 2020

TABLE D 5.1 (Cont'd)

NINE MILE POINT NUCLEAR STATION RADIOLOGICAL ENVIRONMENiJ' AL MONITORING PROGRAM SAMPLING LOCATIONS

Map Collectlon Site T)l~ Qf Sam12!e Location {Env. Program No.} Location Surface Water 38 ass Inlet Canal 7 6 mi @ 236° SW (NA)

Surface Water 39 JAFNPP Inlet Canal 0 5 mi @ 71 ° ENE (NA)

Shoreline Sediment 40 Sunset Bay Shoreline 1.2 m1@84° E (NA)

Fish 41 NMP Site Discharge Area 0.3 mi@ 315° NW (NA)

(and/or)

Fish 42 NMP Site Discharge Area 0.6 mi @ 55° NE (NA)

Fish 43 Oswego Harbor Area 59 m1@237WSW (NA)

Milk 64 Milk Location #55 8.8 mi@97 E Milk (CR) 77 Milk Location 16.0 mi@ 190° S (Summerville)

Food Product 48 Produce Location #6** 1 9 m1@ 143° SE (Bergenstock) (NA)

Food Product 49 Produce Location #1,... 1.6 mi@84° E (Culeton) (NA)

Food Product 50 Produce Location #2** 1.9 mi@ 101° E (Vrtullo) (NA)

Food Product 51 Produce Location #5** 1.5 mi@ 116°ESE (C.S. Parkhurst) (NA)

Food Product 52 Produce Location #3** 1 5 m1@84° E (C. Narewski) (NA)

Map = See Figures D 5.1-1 and D 5.1-2.

- = Food Product Samples need not necessarily be collected from all listed locations. Collected samples will be of the highest calculated site average D/Q.

(NA) = Not applicable.

CR = Control Result (location)

Unit2 Revision 37 II 64 September 2020

TABLE D 5.1 (Cont'd)

NINE MILE POINT NUCLEAR STATION RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM SAMPLING LOCATIONS

Map Collectlon Site Type of Sample Location (Env. Program No.) Location Food Product 53 Produce Location #4..,, 1.7 mi@ 126° SE (P Parkhurst) (NA)

Food Product (CR) -) 54 Produce Location #7..,, 15 1 mI @ 222° SW (Mc Millen) (NA)

Food Product (CR) 55 Produce Location #8** 125mi@22TSW (Denman) (NA)

Food Product 56 Produce Location #9..,, 1.6 mi@ 171° S (O'Connor) (NA)

Food Product 57 Produce Location #1 0** 2.3 m1@ 124° SE (C. Lawton) (NA)

Food Product 58 Produce Location #11- 2.0 m1@ 112° ESE (C. R. Parkhurst) (NA)

Food Product 59 Produce Location #12- 2.0 mi@ 110° ESE (Barton) (NA)

Food Product (CR) 60 Produce Location #13** 15.4 mi @ 222° SW (Flack) (NA)

Food Product 61 Produce Location #14** 1.9 m1@9T E (Koeneke) (NA)

Food Product 62 Produce Location #15 ... 1.6 mi@ 139° SE (Whaley) (NA)

Food Product 63 Produce Location #16** 1.2 mi @ 209° SSW (Murray) (NA)

Food Product 67 Produce Location #17** 1.7 mi@ 98° E (Battles) (NA)

Food Product 68 Product Location #18"" 1.5 mi@84° E (Kronenbitter)

Food Product 69 Product Location #119** 1.4 mi@ 132°_SE (O'Connor)

Map = See Figures D 5.1-1 and D 51-2

- = Food Product Samples need not necessarily be collected from all listed locations. Collected samples will be of the highest calculated site average D/Q.

(NA) = Not applicable.

CR = Control Result (location).

Unit2 Revision 37 II 65 September 2020

APPENDIX A LIQUID DOSE FACTOR DERIVATION Unit2 Revision 37 II 66 September 2020

Appendix A Liquid Effluent Dose Factor Derivation, Aat Aiat (mrem/hr per uCi/ml) which embodies the dose conversion factors, pathway transfer factors (e.g., bioaccumulation factors), pathway usage factors, and dilution factors for the points of pathway origin takes into account the dose from ingestion of fish and drinking water and the sediment. The total body and organ dose conversion factors for each radionuclide will be used from Table E-11 of Regulatory Guide 1.109. To expedite time, the dose is calculated for a maximum individual instead of each age group. The maximum individual dose factor is a composite of the highest dose factor A1at of each nuclide i age group a, and organ t, hence A1at. It should be noted that the fish ingestion pathway is the most significant pathway for dose from liquid effluents. The water consumption pathway is included for consistency with NUREG 0133.

The equation for calculating dose contributions given in section 1.3 requires the use of the composite dose factor Art for each nuclide, i. The dose factor equation for a fresh water site is:

A,at=Ko J Uwe -Ji, t""+UJBF,e-,l,tPI DFLtat+ 69.3UsWe -Ji, tP' (1-e-,l,tb)DFS, ]

((

~ ~~

Where:

A1at = Is the dose factor for nuclide i, age group a, total body or organ t, for all appropriate pathw~ys, (mrem/hr per uCi/ml)

Ko = Is the unit conversion factor, 1.14E5=1 E6pCi/uCi x 1 E3 ml/liter-:- 8760 hr/yr Uw = Water consumption (liters/yr); from Table E-5 of Reg. Guide 1.109 Ut = Fish consumption (kg/yr); from Table E-5 of Reg. Guide-1.109 Us = Sediment Shoreline Usage (hr/yr); from Table E-5 of Reg. Guide 1.109 I BF, = Bioaccumulation factor for nuclide, i, in fish, (pCi/kg per pCi/liter),

from Table A-1 of Reg. Guide 1.109 DFLiat = Dose conversion factor for age, nuclide, i, group a, total body or organ t, (mrem/pCi); from Table E-11 of Reg. Guide 1.109 DFSi= Dose conversion factor for nuclide i and total body, from standing on contaminated ground (mrem/hr per pCi/m 2 ); from Table E-6 of Reg. Guide 1.109 Unit2 Revision 37 II 67 September 2020

Appendix A (Cont'd)

Dw = Dilution factor from the near field area within one-quarter mile of the release point to the potable water intake for the adult water consumption. This is the Metropolitan Water Board, Onondaga County intake structure located west of the City of Oswego.

(Unitless)

Ds = Dilution factor from the near field area within one quarter mile of the release point to the shoreline deposit (taken at the same point where we take environmental samples 1.5 miles; unitless) 69.3 = conversion factor .693 x 100, 100 = Kc (liters/kg-hr)*40 kg/m 2*24 hr/day/.693 in liters/m 2-d, and Kc= transfer coefficient from water to sediment in liters/kg per hour.

= Average transit time required for each nuclide to reach the point of exposure for internal dose, it is the total time elapsed from release of the nuclides to either ingestion for water (w) and fish (f) or shoreline deposit (s), (hr)

= Length of time the sediment is exposed to the contaminated water, nominally 15 yrs (approximate midpoint,of facility operating life), (hrs).

= decay constant for nuclide i (hr1)

= Shore width factor (unitless) from Table A-2 of Reg. Guide 1.109 Example Calculation For 1-131 Thyroid Dose Factor for an Adult from a Radwaste liquid effluents release:

r (DFS)1 = 2.80E-9 mrem/hr per pC~m 2 (DFL),at = 1.95E-3 mrem/pCi tpw = 40 hrs. (w = water)

BFi = 15 pCi/kg per pCi/liter 4)f = 24 hrs. (f = fish)

Ut = 21 kg/yr tb = 1.314E5 hr (5.48E3 days)

Dw = 62 unitless Uw = 730 liters/yr Ds = 17.8 unitless Ko = 1.14E5 {gCi/uCi)(ml/kg}

Us = 12 hr/yr (hr/yr) w = 0.3 Ai = 3.61E-3hr 1 tps = 7.3 hrs (s=Shoreline Sediment)

These values will yield an Aat Factor of 6.65E4 mrem-ml per uCi-hr as listed in Table D 2-2.

It should be noted that only a limited number of nuclides are listed on Tables D 2-2 to D 2-5.

These are the most common nuclides encountered in effluents. If a nuclide is detected for which a factor is not listed, then it will be calculated and included in a revision to the ODCM.

In addition, not all dose factors are used for the dose calculations. A maximum individual is used, which is a composite of the maximum dose factor of each age group for each organ as reflected in the applicable chemistry procedures.

(

Unit2 Revision 37 II 68 September 2020

APPENDIX B PLUME SHINE DOSE FACTOR DERIVATION Unit2 Revision 37 II 69 September 2020

Appendix B For elevated releases the plume shine dose factors for gamma air (B,) and whole body 0/1), are calculated using the finite plume model with an elevation above ground equal to the stack height. To calculate the plume shine factor for gamma whole body doses, the gamma air dose factor is adjusted for the attenuation of tissue, and the ratio of mass absorption coefficients between tissue and air. The equations are as follows:

Gamma Air B1 = L KuaE Is Where: K1 = conversion factor (see s R8Vs below for actual value).

= mass absorption coefficient (cm 2/g; air for B1, tissue for V,)

E = Energy of gamma ray per disintegration (Mev)

Vs = average wind speed for each stability class (s), mis

,)

R = downwind distance (site boundary, m) 8 = sector width (radians) s = subscript for stability class Is = I function = 11 + klz for each stability class. (unitless, see Regulatory Guide 1.109) k2 = Fraction of the attenuated energy that is actually absorbed in air (see Regulatory Guide 1.109, see below for equation)

Whole Body

- µatd V, = 1 .11 SFBie Where: ti = tissue depth (g/cm 2 )

SF = shielding factor from structures (unitless) 1.11 = Ratio of mass absorption coefficients between tissue and air.

Where

. all other parameters are defined above.

)

Unit2 Revision 37 II 70 September 2020

(

Appendix B (Cont'd) 1K =conversion factor = 3.7 E10 dis 1.6 E-6 erg Ci-sec Mev = .46 1293 g 100 erg m3 g-rad Where: µ = mass attenuation coefficient (cm 2lg; air for B1, tissue for V1)

= defined above There are seven stability classes, A thru G. The percentage of the year that each stability class is taken from the U-2 USAR. From this data, a plume shine dose factor is calculated for each stability class and each nuclide, multiplied by its respective fraction and then summed.

The wind speeds corresponding to each stability class are, also, taken from the Unit 2 USAR. To confirm the accuracy of these values, an average of the 12 month wind speeds for 1985, 1986, 1987 and 1988 was compared to the average of the USAR values. The average wind speed of the actual data is equal to 6. 78 mis, which compared favorably to the USAR average wind speed equal to 6. 77 mis.

The average gamma energies were calculated using a weighted average of all gamma energies emitted from the nuclide. These energies were taken from the handbook "Radioactive Decay Data Tables", David C. Kocher.

The mass absorption (µa) and attenuation (µ) coefficients were calculated by multiplying the mass absorption (µalp) and mass attenuation (µIp) coefficients given in the Radiation Health Handbook by the air density equal to 1.293 E-3 glee or the tissue density of 1 glee where applicable. The tissue depth is 5glcm 2 for the whole body.

The downwind distance is the site boundary.

SAMPLE CALCULATION Ex. Kr-89 F STABILITY CLASS ONLY - Gamma Air

-DATA E = 2.22MeV k = .b!.:.b!a = .871 K = .46

µa = 2.943 E-3m-1 µa VF = 5.55 mlsec

µ = 5.5064E-3m- 1 R = 1600m 8 = .39 crz = 19m vertical plume spread taken from "Introduction to Nuclear Engineering", John R. LaMarsh Unit2 Revision 37 II 71 September 2020

Appendix B (Cont'd)

-I Function

' Uoz = .11 11 = .3 12 = .4 I = 11 +kb= .3 + (.871) (.4) = .65 dis.

B1 = 0.46 [Ci-sec)(Mev/ergs] (2.943E-3m-1}(2.22Mev}(.65)

(TT 112 (g/m 3) (fil9.§) (5.55 mis) (.39) (1600m)

(g-rad)

= 3.18(-7) rad/s (3600 s/hr} (24 hid) (365 d/y) (1 E3mrad/rad)

Ci/s (1 E6µCi)

Ci

= 1.00(-2) mrad/yr

µCi/sec -(.0253 cm2/g)(5g/cm 2)

V, = 1.11 (.7) [(1 E-2)mrad/yr] [e l

µCi/sec

= 6.85(-3) mrad/yr

µCi/sec Note: The above calculation is for the F stability class only. For Table D 3-2 and procedure values, a weighted fraction of each stability class was used to determine the B1 and V1 values.

Unit2 Revision 37 II 72 September 2020

(

APPENDIX C DOSE PARAMETERS FOR IODINE 131 and 133, PARTICULATES AND TRITIUM J

Unit2 Revision 37 II 73 September 2020 ,

Appendix C DOSE PARAMETERS FOR IODINE-131 AND -133, PARTICULATES AND TRITIUM This appendix contains the methodology which was used to calculate the organ dose factors for 1-131, 1-133, particulates, and tritium. The dose factor, Ri, was calculated using the methodology outlined in NUREG-0133. The radioiodine and particulate DLCO 3.2.1 is applicable to the location in the unrestricted area where the combination of existing pathways and receptor age groups indicates the maximum potential exposure occurs, i.e., the critical receptor. Washout was calculated and determined to be negligible. R1 values have been calculated for the adult, teen, child and infant age groups for all pathways. However, for dose compliance calculations, a maximum individual is assumed that is a composite of highest dose factor of each age group for each organ and pathway. The methodology used to calculate these values follows:

C.1 Inhalation Pathway

= K'(BR)a(DFA}iia where:

dose factor for each identified radionuclide i of the organ of interest (units = mrem/yr per uCi/m 3);

K' = a constant of unit conversion, 1E6 pCi/µCi (BR)a = Breathing rate of the receptor of age group a, (units = m3/yr);

(DFA)IJB = The inhalation dose factor for nuclide i, organ j and age group a, and organ j (units = mrem/pCi).

The breathing rates (BR)a for the various age groups, as given in Table E-5 of Regulatory Guide 1.109 Revision 1, are tabulated below.

Age Group (a) Breathing Rate (m 3/yr)

Infant 1400 Child 3700 Teen 8000 Adult 8000 Inhalation dose factors (DFA)ua for the various age groups are given in Tables E-7 through E-10 of Regulatory Guide 1.109 Revision 1.

Unit2 Revision 37 II 74 September 2020

Appendix C (Cont'd)

C.2 Ground Plane Pathway

-M K'K"(SF)(DFG), (1-e )

Ai Where:

= Dose factor for the ground plane pathway for each identified radionuclide i for the organ of interest (units = m2 -mrem/yr per

µCi/sec)

K' = A constant of unit conversion, 1E6 pCi/µCi K = A constant of unit conversion, 8760 .hr/year Ai = The radiological decay constant for radionuclide i, (units = sec- 1) t = The exposure time, sec, 4. 73E8 sec (15 years)

(DFG), = The ground plane dose conversion factor for radionuclide i; (units

= mrem/hr per pCi/m 2)

SF = The shielding factor (dimensionless)

A shielding factor of 0. 7 is discussed in Table E-15 of Regulatory Guide 1.109 Revision 1. A tabulation of DFG, values is presented in Table E-6 of Regulatory Guide 1 .109 Revision 1.

Unit2 Revision 37 II 75 September 2020

Appendix C (Cont'd)

C.3 ' Grass-(Cow or Goat}-Milk Pathway Where:

= Dose factor for the cow milk or goat milk pathway, for each identified radionuclide i for the organ of interest, (units = m2-mrem/yr per µCi/sec)

K' = A constant of unit conversion, 1E6 pCi/µCi Ot = The cow's or goat's feed consumption rate, (units = kg/day-wet weight)

= The receptor's milk consumption rate for age group a, (units = liters/yr)

The agricultural productivity by unit area of pasture feed grass, (units

kg/m 2)

Ys = The agricultural productivity by unit area of stored feed, (units = kg/m2)

Fm = The stable element transfer coefficients, (units= pCi/liter per pCi/day) r = Fraction of deposited activity retained on cow's feed grass (DFL)iai = The ingestion dose factor for nuclide i, age group a, and total body or organ j (units = mrem/pCi)

,J = The radiological decay constant for radionuclide i, (units=sec- 1)

')..W = The decay constant for removal of activity on leaf and plant surfaces by weathering equal to 5.73E-7 sec-1 (corresponding to a 14 day half-life)

The transport time from pasture to cow or goat, to milk, to receptor, (units

sec)

= The transport time from pasture, to harvest, tb cow or goat, to milk, to receptor (units = sec)

= Fraction of the year that the cow or goat is on pasture (dimensionless) fs = Fraction of the cow feed that is pasture grass while the cow is on pasture (dimensionless)

Unit2 Revision 37 n 76 September 2020

Appendix C (Cqnt'd)

Milk cattle and goats are considere~ to be fed from two potential sources, pasture grass and stored feeds. Following the development in Regulatory Guide 1.109 Revision 1, the value offs is considered unity in lieu of site specific information. The value of fp is 0.5 based on 6 month grazing period. This value for fp was obtained from the environmental group.

Table C-1 contains the appropriate values and their source in Regulatory Guide 1.109 Revision 1.

The concentration of tritium in milk is based on the airborne concentration rather than the deposition. Therefore, the Rr(C) is based on X/Q:

Rr(C) = K'K'" FmOtllap(DFL)iat 0.75(0.5/H)

Where:

Rr(C) = Dose factor for the cow or goat milk pathway for tritium for the organ of interest, (units = mrern/yr per µCi/m 3)

K'" = A constant of unit conversion, 1E3 g/kg H = Absolute humidity of the atmosphere, (units = g/m3) 0.75 = The fraction of total feed that is water 0.5 = The ratio of the specific activity of the feed grass water to the atmospheric water Other values are given previously. A site specific value of H equal to 6.14 g/m 3 is used. This value was obtained from the environmental group using actual site data.

Unit2

. , Revision 37 II 77 September 2020

Appendix C (Cont'd)

C.4 Grass-Cow-Meat Pathway R,(M)= Dose factor for the {Tleat ingestion pathway for radionuclide i for any organ of interest, (units = m2-mrem/yr per µCi/sec)

Ft = The stable element transfer coefficients, (units = pCi/kg per pCi/day}

Uap = The receptor's meat consumption rate for age group a, (units = kg/year) bi = The transport time from harvest, to cow, to receptor, (units = sec) tt = The transport time from pasture, to cow, to receptor, (units = sec)

All other terms remain the same as defined for the milk pathway. Table C-2 contains the values which were used in calculating R,(M).

The concentration of tritium in meat is based on'airbome concentration rather than deposition. Therefore, the Rr(M) is based on X/Q.

Rr(M) = K'K"'FtOtUap(DFL),at [0.75(0.5/H)]

Where:

Rr(M) = Dose factor for the meat ingestion pathway for tritium for any organ of interest, (units = mrem/yr per µCi/m 3)

( ,

All other terms are defined above.

C.5 Vegetation Pathway The integrated concentration in vegetation consumed by man follows the expression developed for milk., Man is considered to consume two types of vegetation (fresh and stored) that differ only in the time period between harvest and consumption, therefore:

-ML R,M = K' r (DFL)iai [ LJLaFt.e Yv('AJ + 'Aw)

Unit2 Revision 37 II 78 September 2020

Appendix C (Cont'd)

Where:

R,M = Dose factor for vegetable pathway for radionuclide i for the organ of interest, (units = m2-mrem/yr per µCi/sec)

K' = A constant of unit conversion, 1E6 pCi/µCi LJLa = The consumption rate of fresh leafy vegetation by the receptor in age group a, (units = kg/yr)

U8 a = The consumption rate of stored vegetation by the receptor in age group a (units = kg/yr)

FL = The fraction of the annual intake of fresh leafy vegetation grown locally F9 = The fraction of the annua1 intake of stored vegetation grown locally tL = The average time between-harvest of leafy vegetation and its consumption, (units = sec) b, = The average time between harvest of stored vegetation and its consumption, (units = sec)

Yv = The vegetation areal P density, (units = kg/m 2)

All other factors have been defined previously.

Table C-3 presents the appropriate parameter values and their source in Regulatory 1

Guide 1 .109 Revision 1.

In lieu of site-specific data, values for FL and F9 of, 1.0 and 0.76, respectively, were used in the calculation. These values were obtained from Table E-15 of Regulatory Guide 1.109 Revision 1 .

The concentration of tritium in vegetation is' based on the airborne concentration rather than the deposition. Therefore, the RrM is based on X/Q:

RrM = K'K'" [Ula fL + U5 a fg](DFL),aJ 0.75(0.5/H)

Where:

RrM = dose factor for the vegetable pathway for tritium for any organ of interest, (units = mrem/yr per µCi/m 3).

All other terms are defined in preceding sections.

Unit2 Revision 37 II 79 September 2020

TABLE C-1 Parameters for Grass - (Cow or Goat) - Milk Pathways Reference Parameter (Reg. Guide 1.109 Rev. 1)

Ot (kg/day) 50 (cow) Table E-3 6 (goat) Table E-3 r 1.0 (radioiodines) Table E-15 0.2 (particulates) Table E-15 (DFL)1Ja (mrem/pCi) Each radionuclide Tables E-11 to E-14 Fm (pCi/liter per pCi/day) Each stable element Table E-1 (cow)

Table E-2 (goat)

Ys (kg/m2) 2.0 Table E-15

(_

Yp (kg/m2) 0.7 Table E-15 hi (seconds) 7.78 x 106 (90 days) Table E-15 tt (seconds)

1. 73 x 105 (2 days) Table E-15 Uap (liters/yr) 330 infant Table E-5 330 child Table E-5 400 teen Table E-5 310 adult Table E-5 Unit 2 Revision 37 II 80 September 2020

TABLE C-2 Parameters for the Grass-Cow-Meat Pathway Reference Parameter Value (Reg. Guide 1.109 Rev. 1) r 1.0 (radioiodines) Table E-15 0.2 (particulates) Table E-15 Ft (pCi/kg per pCi/day) Each stable element Table E-1 Uap (kg/yr) 0 infant Table E-5 41 child Table E-5 65 teen Table E-5 110adult Table E-5 (DFL)1Ja (mrem/pCi) Each radionuclide Tables E-11 to E-14 I

Yp (kg/m 2) 0.7 Table E-15 Ys (kg/m2) 2.0 Table E-15 hi (seconds) 7.78E6 (90 days) Table E-15 tt (seconds) 1. 73E6 (20 days) Table E-15 Ot (kg/day) 50 Table E-3 Unit2 Revision 37 II 81 September 2020

TABLE C-3 Parameters for the Vegetable Pathway Reference Parameter Value (Reg. Guide 1.109 Rev. 1) r (dimensionless) 1.0 (radioiodines) Table E-1 0.2 (particulates) Table E-1 (DFL)ua (mrem/pCi) Each radionuclide Tables E-11 to E-14 LJL)a (kg/yr) - infant 0 Table E-5

- child 26 Table E-5

- teen 42 Table E-5

- adult 64 Table E-5 U5 )a (kg/yr) - infant 0 Table E-5

- child 520 .Table E-5

-teen 630 Table E-5

- adult 520 Table E-5 tL (seconds) 8.64E4 (1 day) Table E-15 tti (seconds) 5.18E6 (60 days) Table E-15 y V (kg/m-2) 2.0 Table E-15 Unit2 Revision 37 II 82 September 2020

APPENDIX D DIAGRAMS OF LIQUID AND GASEOUS TREATMENT SYSTEMS AND MONITORING SYSTEMS Unit2 Revision 37 II 83 September 2020

Liquid Radwaste Treatment System Diagrams Unit2 Revision 37 II 84 September 2020

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APPENDIX E NINE MILE POINT ON-SITE AND OFF-SITE MAPS Unit2 Revision 37 II 106 September 2020
Site Map Lake Ontario FIGURED 5.1-1 NINE MILE POINTON-SITE MAP
,1 D_ ENVIRONMENTAL SAMPLE I.
LOCATIONS 0 1 Scale(Mlles)
Unit2 Revision 37 II 107 September 2020
FIGURE D 5.1-2 (1 of 2)
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Unit2 Revision 37 II 108 September 2020
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ATTACHMENT 14 Page 1 of 1 Unit 1 Unit 2 X Reoortlna PeoQd: January - December 2020 Process Control Program <PCP}
There were no changes to the Process Control Program in 2019.

NMP1L3388, 2020 Radioactive Effluent Release Report for Nine Mile Point Units 1 and 2

Contents

Text

Subject:

Enclosures:

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

The agricultural productivity by unit area of pasture feed grass, (units

The transport time from pasture to cow or goat, to milk, to receptor, (units

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NMP1L3388, 2020 Radioactive Effluent Release Report for Nine Mile Point Units 1 and 2

Text

Subject:

Enclosures:

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

The agricultural productivity by unit area of pasture feed grass, (units

The transport time from pasture to cow or goat, to milk, to receptor, (units

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