Text

Semiannual Radioactive Effluent Release Rept,Jul-Dec 1989, Crystal River Unit 3. W/900227 Ltr
	ML20012B555
Person / Time
Site:	Crystal River
Issue date:	12/31/1989
From:	Boldt G, Johnson S; FLORIDA POWER CORP.
To:	; NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
	3F0290-17, 3F290-17, NUDOCS 9003150304
	Download: ML20012B555 (52)
	v • d • e

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. February 27, 1990 ,

3F0290-17 U.S. Nuclear Regulatory Commission $

Attention: Document Control Desk :

Washington, D.C. 20555 j

Subject:

Crystal River Unit 3 Docket No. 50-302 ;

Operating License No. DPR-72 l Semiannual Radioactive Effluent Release Report ]

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Dear Sir:

Pursuant to-Title ~10, Code of Federal Regulations, Part 50.36(a) (2) +

and Crystal River Unit 3 Technical Specification 6.9.1.5(d), Florida ;

Power Corporation hereby submits the Crystal River Unit 3 Semiannual j Radioactive Ef fluent Release Report for the period July 1,1989 through December 31, 1989. l sincerely, .

J Gary Boldt q Vice President i Nuclear Production )

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REF:wla xc: ' Regional Administrator, Region II i

Senior Resident-Inspector i

9003150304 891231 f} -

PDR ADOCK 05000302 R PDC %

POST OFFICE BOX 219 + CRYSTAL RIVER, FLORIDA 326290219 + (904) 7954486 A Florlds Progress Comparoy fhY-m .,

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SDEAlHAL RADIQPCCIVE !

EPMDENT REIEhSE j RERRP 1989 JUIX - IIIGNIER MIRIIR IONER 035GETION CRYS'mL RIVER - INIT 3 FACILITY OR!RATDG LICENSE NO. DRt-72 l D00GT NO. 50-302 mumar,1990 l.

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Approved by: M_^1 h. C n - ,.r, - sovi sita 9 Date: A 40

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Gam!MB EhGC o ININXXXTIN 1 o TMK7IAR DATA SLB00 RIES GMIEDUS EFMIENIS 2 LIQUID EFFILENIS 4 RM30sIE SHIINENIS 7 o TREND GUGHS GAnrTM RE2J!ASEE FISSIN AND ACTIVATIN PR0[11CIS 9 1RITEM 10 LISTID RE2J!ASES FISSIN AND ACTIVATIN PIODUCIS 11

'IRITEM 12 SDLID RAmesIE SPENT RESINS, EIC 13 l

130f GMPRESSIBLE 10sIE, EIC 14 o AM40AL DOSE SLBMhRIES INDIVI!1RL DOSES 15 POEULATI W DOSES 16 SITE BOWDMOf AIR DOSES 17

o. IMPUNED RE1EhSEB 18 o RN 90Wir!E TREEDENT SYS'I196 18 l o DNIR3MENIRL RADIOLOGIGL MEI7tRING PROGNM 18 o EFMIENT MMITER OPERABILITY 18 o CDCM AND PCP 19 i

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INTIKIIINEN t

'Ihis report is subnitted as required by Technical Specification 6.9.1.5.d to crystal River Facility Operating License No. DER-72. In acooth with 'Iwhnical Specifications, the following information must be included ;

in this report ;

i A sumary of the quantities of radioactive liquid and gammam effluents !

and solid wasta released fra the plant as outlined in Regulatory Guide l 1.21 (Rev.1,1974) with data sunearized cm a quarterly basis following !

the format of Appendix B thereof. !

. An annual sunmary of hourly meteorological data collected cwer the :

previous year. (In lieu of submittal this data is maintained on-site ;

and is available to the NRC upon request) '

An asseamment of radiation doses due to the radioactive liquid and gamecus affluents released fra the plant during the previous calardar "

year.

i As assessment of radiation doses fra radioactive liquid and gaseous effluents to MDGERS OF 'INE PUBLIC due to their activities inside the ,

SI'IE BOUNDMW. i An = =====arit of radiation doses to the hypothetical worst came individual frm reactor releases. '

l Ibr and type of solid waste shipped off site I i

! cantainer Voltans

! 'Ibtal curie Quantity (specified as measured or estimated) -

l Principal Radionuclides (specified as measured or estimated)

Type of Waste (e.g., spent resin, ==* tad dry waste) r Type of container (e.g., IRA, Type A, Type B)

Solidification Agent (e.g., cement)

A list and description of unplanned releases to unrestricted areas. 1 A description of any changes to the:

Prooses Control Program (PCP) off-site Dome calculation Marmal (OtxX)

Radioactive Westa Treatment Systems A listing of new Envircreantal Radiological Monitoring Fwp - done omlaulation locatica changes identitled by the land-use census.

l Information- rulatirg to affluent monitor instrumentation being inoperable for thirty or more days.

In addition to the required data, trend y@ of curies relaaaari per six

month period have been included in order to put the current data into historical perspective.

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'INER 1 MTIIEBrF ABE) WWE lEJEGEL IBED00ElhL NER5tt' - 1999 GhSIKXS EF7TDIBf!B - EK399GTCM OF ALL RELEASES Unit Quarter Quarter Est.7btal 3 4 Error %

A. Fission and Activation Camam

1. 7btal Release Ci 2.21E+03 1.41E+03 30

2. Average Release Rata for Period uCi/uec 2.78E+02 1.77E+02

3. INercent of Technical Specification Limit % 8.53E+00 2.98E400 s

B. Iodines

1. 7btal Iodine - 131 C1 1.60E-03 6.22E-05 30

2. Average Release Rate for Period uCi/sec 2.01E-04 7.82E-06

3. Percent of Tednical Specification Limit % 9.76E-00 2.60E-01

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C. Particulates 30 l -- 1. Particulatas with half-lives > 8 days Ci 7.70E-07 7.42E-07 2.. Average Release Rate for Period uCi/sec 9.69E-08 9.33E-08 1

3. Introent of 'Nchnical Specification Limit % 9.76E-00 2.60E-01

4. Gross Alpha anMaartivity Ci 6.96E-08 1.13E-07 D. Trititan

1. 7btal Hale:se Ci 1.97E+01 2.98E+00 30

2. Average Release Rate for Period uCi/sec 2.48E+00 3.75E-01

3. Percent of 'Mchnical Specification Limit % 9.76E-00 2.60E-01 2

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'5B[E 2 EFFIIRWfr AM) WWIE DESEUShL SEBEDWSElhL IWIR3tr - 1989 GMartN EFTIUENIS - GEKDE) IEVEL REtEhSES 03frDrrrs )rrF BA201 ) ODE Nuclides Relaaned Unit Quarter 3 Quarter 4 Quarter 3 Quarter 4

1. Fission games amtm-41 Ci krvaten-85 C1. 5.83E+00 6.54E+00 kryptzm-85m C1 4.66E+00 9.59E+00 5.79E-02 4.68E-01 krvotm -87 Ci 2.82E-01 1.28E+00 3.01E-02

)c:Votort-88 C1 1.04EMO 3.59E-02 3.37E-01 Menon-131m Ci 3.88E@0 4.02E+00 6.65E+00 Xenan-133 _Ci 1.55E+03 5.20E+02 5.18E+02 7.36E+02 xencm-133m Ci 2.06E+01 4.54E+00 8.26E+00 gencm-135 Ci 9.56E+01 1.06E+02 2.50E+00 1.05E+01 1:pnan-135m C1 1.22E-01 genen-138 Ci unidentified Ci

'Ibtal for Dariod C1 1.68E+03 6.37E+02 5.35E+02 7.69E+02

2. I0 dines iodinF131 C1 1.60E-03 6.22E-05 1.10E-06 lodine-133 C1 1.49E-07 l iodine-135 Ci L'Ibtal for 14 triad Cl 1.60E-03 6.22E-05 1.25E-06 l

_3. Particulates marpaname-54 Ci '

g;iDlt-58 Ci imn-59 C1 cxhalt-60 Ci rinc-65 Ci stmntitan 49 C1 4.22E-07 5.55E-07 strontium-90 Ci 3.48E-07 1.87E-07 2.22E-11 -

mlvhdaram-99 Ci tellurium-132 Ci cesium-134 Ci cesi'.nn-137 - Ci cesitan-138 Cl unidentified Ci l Total for Period Ci 7.70E-07 7.42E-07 2.22E-11 3

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I N ABC DNfM EHEREAL SBENGEBE MR5tr - 1989 '

12(NID EFFIGBGB - IKBOWETCat W AIL MIEhSIB Unit Quarter Quarter Est.'Ibtal 3 4 Erzer %

A. Fissicri ard Activation Products '

1. Total Release (not including tritium, ,

gases, alpha) Ci 5.80E-02 1.51E-01 25

2. Average diluted cancert. ration during period uC1/ml 2.90E-08 6.61E-08

3. Percent of applicable limit % 1.50E-01 7.03E-01 B. Tritium

1. Total Release Ci 1.16E+02 1.36E+02 30

2. Average diluted canoantration durirug

! period uCi/ml 1.13E-05 7.98E-06 l 3. Percent of applicable limit % 3.77E-01 2.66E-01 C. Dissolved and entrained games

1. Total release Ci 1.29E+01 2.91E+01 25 l

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2. Average diluted concentration during period uCi/ml 1.24E-06 1.63E-06

3. Percent of applicable limit % 6.20E-01 8.15E-01 D. Gross alpha radioactivity I

1. 'Ibtal relemme Ci < IID 6.43E-05 30 i i

E. Volume of Wests released (prior to dilution) -

1. Batdt and 0:mtirmous Modes Litars 1.31E+07 3.37E+06 10 F. Volume of diluticri atar used during period I 1. Batd1 and Contirmous Modes Liters 2.23E+10 3.68E+10 10 i

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'IN E E 4 WFIIEBrf AID ISWFIE DISREAL MBED00E3hL 3 Mutt' - 1989 123 TID EFFIIEBGB !

CMTINDOUS MXE BMW MXE Nuclides Released Unit Quarter 3 Quarter 4 Quarter 3 Quarter 4

=M itan-24 Ci l chrrent tun-51 C1 2.62E-03 6.58E-03 *

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mangename-54 C1 1.53E-04 8.86E-04 irm-55 C1 1.04E-03 4.37E-03 8.17E-03 1.35E-02 mhalt-58 C1 2.72E-02 2.95E-02 iren-59 C1 1.1 W-04 2.37E-04 ,

mhalt-60 Ci 6.60E-03 1.41E-02 zire-65 Ci rubh44 tan-88 C1 1.64E-03 strentitan-89 C1 1.71E-04 2.59E-04 strertitan-90 Ci 9.69E-05 3.44E-05 4.16E-04 strmtitan-92 C1 1.21E-05 1.86E-QJ_,_,

niobitan-95 C1 3.58E-05 7.06E-04 >

zirecrsitan-95 C1 3.60E-04 zirecriitan-97 C1 2.38E-04 miolvh9arman-99 C1 2.84E-03 3.58E-04 -

&ar+triatitan-99m Ci 4.45E-03 1.84E-03 ruthartitan-103 Ci 3.57E-05 1.15E-04 ruthanitan-106 Ci 5.95E-04 silver-110in Ci 5.41E-03 iMiria-131 C1 1.50E-03 5.34E-03 iMi ria-133 C1 1.94E-05 1.19E-03 ta11uritam-132 Ci 8.51E-05 1.67E-05 iMiria-133 Ci >

<==itm>134 C1 2.31E-04 7.36E-04 rumitair-136 Ci

<==ltun-137 Ci 7.67E-04 1.48E-03 havitan-139 Ci 3.17E-04 lantharan>140 Ci 1.99E-03 1.38E-03 rweitan-141 Ci caritan-144 Ci unidentified Ci Ibtal for period Ci 1.04E-03 4.47E 03 5.70E-02 1.46E-01 5

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'DJEE 4 (03G'Dalm) arFmerr am went tusakaL seemeeamL msurr - inos LIGED EFFIGBHB O2fnNUEIS MXE IWIG MXE Dissolved & Entrained Unit (parter 3 Quarter 4 Quarter 3 Quarter 4 t'.n aan aruan-41 Ci krvoten-85 Ci 6.17E-03 2.51E-02 krvaten-85m C1 1.22E-04 2.20E-03 krvaten-88 C1 1.04E- L wanem-131m Ci 1.11E-01 1.98E-01 Mancm-133 C1 1.27E+01 2.84E+01 Manen-133m Ci 9.99E-02 3.00E-01 Manen-135 C1 2.73E-02 1.71E-01 xenem-135m Ci i

trititan Ci 4.04E-02 4.59E-02 1.15E+02 1.36E+02 l

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'INEE 5 '

EFFIIEDFF JWW WWrIE DDINGhL mermarar IER5tr - 1999 enrm Wurzg mm I3gggxmp putt, stIIDEBms A. SOLID WMTTE SHIPPED OFFSTIE FtR KRIAL OR DISPOSAL (Non irradiated fuel)

Secord

1. Type of wasta Unit 6-month Est. 'Ibtal !

w ied Error, t

a. Spent resins, filter sludges, avaporator m3 5.36E+01 bottans, etc. C1 2.38E+03 20

b. Dry ompressible waste, contaminated m3 9.87E+01 e@ lip, etc. Ci 3.14E+00 50

c. Irradiated mW uias, control .'

m3 r rods, etc. Ci

d. Other (describe) m3 l Ci i r

2. Estimate

of major nuclide ocuposition (by type of waste) !

a. On-137 3.97E+014 Ni-63 1.00E+01% ,

On-134 1.80E+01% Cc-58 3.17EMOt Cb-60 1.26E+01%

Fe-55 [

1.24E+01%

b. Fe-55 4.50E+01% NI-63 7.26E+00%

On-137 2.77E+01% C-14 2.30E+00%

00-60 1.42E+01%

l 3. Solid Waste Disposition Nunbar of Shipannts M:de of h imtation Destinaticri 12 Transport Truck - Chant-Nuclear Systems, Inc.

Ibeclusive Use Vehicle Barmell, SC i 3

1 #89-39 Sraa SEG**

Oak Ridge, 'IN B. INRADIATE) PUEL SKII9ENIS (Disposition)

Nunbar of Shipments Mode of h spurtation Destination 0 N/A N/A ,

With the e>compticm of Seoandary Rasin shipnants all curie values ard principle '

radiaruclides are determined by indirect methods and are therefore estimates.

Seoandary Rosin shipments are evaluated by direct (gansna wkW) and indirect (applying scaling factors) methods, with the results being a ocabination of naamnad and estimated value.

- Scientific Ecology Group rapacks secondary plant resin and irEry--N dry waste in order to reduce burial volume.

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EFFIIEBfr ADO Wdr!E IIIfEMAL SIBUAltERL NERP - 1989 fEIRENT R3094tY DME AND CINIAINER TUIAL WASTE O2fIAINER SOLIDIF.

SHIINDff # VOIIME* CIRIES IRINCIPIE 1%DIONOCLTDES T(PE TYPE AGDff 8-23 Cb-58, 00-60, Os-134, C2-137, 89-30 205.8 3.41 Ni-63, H-3, Fe-55, Sr-89 SR Type B la ;

8-25 00-58, 00-60, Cs-134, Os-137, 89-31 120.3 22.4 Ni-63, H-3, Fe-55, Sr-89 SR Type B !&

9-5 89-32 671.1 0.591 00-60, Cs-137, Ni-63, Fe-55 CM ST** la 9-21 00-58, 00-60, Cs-134, Os-137, 89-34 120.3 37.6 Sr-89, Fe-55, Ni-63, Mn-54 SR Type B IR 9-22 00-58, 00-60, Cs-134, Cs-137, 89-35 11-3, Ni-63, Fe-55 205.8 0.897 SR Type B la 10-31 89-39 1140 2.11 00-60, Cs-137, Ni-63, Fe-55 NW ST la 12-5 89-43 10 $ 100 50.6 00-58, Cb-60, N55, Cr-51 SR Type B la 12-8 Cb-60, Cs-134, Cs-137, Ni-63, 89-44 120.3 524. Fe-55 SR Type B la 12-12 89-46 120.3 973. Cb-58, 00-60, Cs-134, On-137 SR Type B IR 12-18 00-60, Os-137, Ni-63, Fe-55, 89-47 10 9 100 0.294 C-14 NW FF !R l

12-19 0o-60, Os-134, on-137, Ni-63, 89-48 120.3 574. Fe-55 SR Type B }&

12-20 C M O, On-137, Ni-63, Fe-55, 89-49 90 0 7.5 0.408 00-58, C-14 CM ST !&

l 12-21 00-60, Os-134, Os-137, Ni-63, 89-50 120.3 210. Fe-55 SR Type B !&

neeIE TYm: SR - Spent Resin Itf - Nan % Wasta G - Contaminated Equipment sc - Socordary Basin at - Otmpacted Haste IC - Irradiated Ozponents l

F - Filters 2 - Evaporator Bottans

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i

. a 1 1

- FIGURE 6 -

fd"..

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7 SOLID RADWASTE !- i

+- DRY Cor RESSIBLE WASTE. !'~

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0.li 77 78 79 80 81 82 83 84 85 86 87 88 89 90 Year 14

'mBLE 7 l 3

1 AREERL IEEES '10 IMHVIIIRIS FEM AIL RE[EhSES [KEDt; 1989 Maole Bady 1)one Organ DJ8e (mrum) (arem) i ist Quartar Continuous Gaseous Effluents 2.37E-02 7.10E-02 i Batd1 Gaseous Effluents 1.49E-02 4.48E-02 '

contimous Ligaid Effluents 3.48E-04 8.55E-03 Batd1 Liglid Effluents 1.16E-03 1.90E-02 2nd..Quartar

Ccritinuous damar == Effluents 2.12E-02 6.68E-02 Batd1 namarum Effluents 5.63E-03 1.26E-02 !

Contimous Liglid Effluents 6.34E-05 1.56E43 Batd1 Ligild Effluents 7.04E-04 4.72E-03 3rd Quartar conti"Anous camar** Effluents 1.24E-01 8.22E-01 Batd1 namar*= Effluents 9.01E-02 2.51E-01

Continuous Ligild Effluents 3.46E-05 8.50E-04 Batd1 Ligaid Effluents 8.00E-04 6.98E-03 .

4th Quartar Continuous ca-= Effluents s 5.39E-02 1.39E-01 Batd1 namar== Effluents 2.44E-02 7.10E-02 Contimous Lig11d Effluents 1.65E-04 3.76E-03 4 Batd1 Lig11d Effluents 1.89E-03 3.54E-02 ANNUAL 7UELL............................. 3.63E-01 1.56E+00 NCTIE: Quarterly doses due to rJamme== releases are calculated using t.%

highest site boundary dispersicr1 factors (X/Q) and their w+.iirg depositien factors (D/Q).

Domes due to lig11d effluents are calculated using enly the dilution prt:wided by Units 1, 2, and 3 cooling water (i.e. Galf water dilution is not amruutted for) .

15

-. .- - . . ~ - - _ . . . .. .- .. - - ~

9 e

]

' maut a !

AREERL IIMDB 'ID 'IIE IGMJEICM FIEM AIL IEEEhSEB IIEEDE 1989 J

I b le Body Dome Oz1; pan Dog (Rem) (Rem) 1st Chaartar i

Carttirmacus ramar== Effluents 1.75E-02 5.48E-02 Bat 2 ca==r== Effluents 4.84E-03 1.84E-02 ,

Otmtirmous Liquid Effluents 2.932-02 7.22E-01 !

Batch Liquid Effluents 1.11E-01 1.75E+00 ;

g Continuous ramar== Effluents 4.52E-03 8.89E-03 Bate ramar== Erfluents 8.89E-04 2.79E-03 Continuous Liquid Effluents 5.35E-03 1.32E-01 Bat & Liquid Effluents 4.93E-02 4.18E-01 :

(

3rd Quartar l Continuous Famar== Effluents 2.98E-02 9.07E-02 l Bat & eamar== Effluents 2.88E-02 9.83E-02 Contismous Liquid Effluents 2.93E-03 7.18E-02 Batch Liquid Effluents 6.23E-02 4.68E-01 ,

4th Quartar

" contirmous ra-= Effluarts 1.25E-02 4.00E-02 Batch ramarum Effluents 4.00E-03 1.42E-02 Cbntinuous Liquid Effluents 1.40E-02 3.17E-01 Bat & Li@id Effluente 1.38E-01 3.72E+00 ANNUAL 'IUIL . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5.15E-01 7.93E+00 ll NL7FE: Population &ses are cum 11ative doses to the estimated 475,000 people l- residirq within 50 miles of the plant.

16

. L' [

'INER 9 l i

prat M1BEhRY AIR DCMIE E30EMtf PGt 1989 j e

i halemme Distance Beta Gatsna Quartar Itals Sector Miles inRads EBbdf ,

1 Batdt W 1.01 7.60E-02 2.37E-02 !

0:mtirmous N 0.83 9.32E-02 3.15E-02 l Quarterly 'Ibtal 1.69E-01 5.52E-02 2 Batdn N 0.83 1.67E-02 5.02E-03 Ocmtirmous W 1.01 3.15E-02 1.24E-02 ,

Quarterly 'Ibtal 4.82E-02 1.74E-02 3 Batdt SE 1.57 4.38E-01 1.48E-01 '

Q:ntirmous M 1.41 4.15E-01 1.56E-01 Quartarly 'Ibtal 8.53E-01 3.04E-01 4 Batd1 W 1.39 1.21E-01 4.15E-02 i Omtirmous W 1.01 1.77E-01 7.94E-02 Quartarly 'Ibtal 2.98E-01 1.21E-01 1

D MOUhL 'IUIAL 1.37E-00 4.98E-01 -

NCTIE: Quarterly domes are calculated using the highest site boundary dispersion factor (X/Q).

17

l

~ ..

c, . .. s .. : . : .

There were rm urplanned liqutid or rJamarna releases for the period of this report.

)

NEQhC1TVE WWFM 'MEMENf m5mRE l l

There were no sigrtificent chan; pas to the radioactive wasta treatment systans for the period of this t w t.. l l

\

EfVIMEGERemL NEMDGICRL MMPIGtIIC MKEEUsi ;

l 3

The June 1989 land-use census did not identify any new dose cala11ation locations.

i EFFHEREP IEMHtEt DEMEERWWI2EEf GBREELTrY i No effluent monitor was ir.ip-a.ble for a patiod of thirty days or more. ,

4 18 c__________________ . _ _ . . _ _ . _ _ _ _. . . _ _ _ . _ . . _ _ _ _ , _ . _ . . - _ , . , _ , _ , , . _ . _ . . . .

l

. ,+

. a .,

axM Als MP

'Ibers were several manges to the 000( (arvision 13 and Revis,icn 14), the descriptions of whim are provided below. 'Ihe revised pihipes have also been included. '1ha PCP was not manged.

GEM GIhlEEB - IRV. 13

)

HE I i

Introductico Revised to better reflect requirements for approval i 1,11 Table of Otmtants revised to reflect other changes

]

2, 11, 14 ' Revised for clarity arti to account for deletion of !

redundant does factor tables.

17, 18, 18a, 18b, Setpoint equations: deleted the administrative limit

18c, 19, 20, 21 whid had rdM setpoints to 10% of the Te&nical :

Spm::ification limits. Also deleted the term '

3.3 Bkg as it had a negligible influence on the setpoint, i

25 Deleted table because it was reckandant to Table 4.4-1 ,

en page 54. l 26 Clarified the use of does factor, Pi. !

27 Delstad table since it was redundant to Table 4.4-3 ;

on page 57. '

40a Changed purge run time requirement and added ,

clarification on the use of hatees when purge is shut down. ,

52, 53 Typos 55 Footnoted refersnoes; corrected reference '

60, 60s, 65, 65e, Typos arti ocupleted reference lists. .

69, 69a, 75-GEM 06 - MEV.14 18a Corrected typo '

80 REMP whul location, C47, In:wed to Orlando, Office of Radiatica cantzol k

19

- _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ <-----,-e -_ =2 m., -- ___ - _ _ _ _ -____,vm_,

4 .'

4, ,

d i

INTRODUCTION l The Off-Site Dose Calculation Manual (ODCM) is provided to support implementation of the Crystal River Unit 3 Radiological Effluent Technical Specifications. The-L:

n 00CM contains calculational methods to be used in determining the dose to members

- of the public resulting from routine radioactive effluents released from Crystal-River Unit 3. More accurate estimation of doses- is performed annually in ,

preparation of the year-end Semiannual Radioactive Effluent Release Report. The' L , ODCM also contains the methodology used to determine effluent monitor alarWtrip ,

I setpoints which assure that releases of radioactive materials remain within '

l specified concentrations.:

The 00CM will be controlled by.the Site Nuclear Services and revisions shall be ,

.made with the approval of the Manager, Site Nuclear Services. All revisions shall . be approved.' by- the _ Plant Review Committee prior to implementation at

'? Crystal River Unit 3.. Historical documentation and- distribution of the ODCM.

l shall be the responsibility of the Nuclear Operations Records Manager per N00-05, ;

L Document Control Program. 1

t. !

1: I i

s 5

g' f, '. .. <*

3. .

s

-3 TABLE OF CONTENTS 1

Section Eggs :

1 1.0 RADI0 ACTIVE EFFLUENTS NONITOR SETP0 INT SPECIFICATIONS 1 1.1 Effluent Monitor Setpoint Specifications 3 1.2 Nuclide Analyses 6 ,

1.3 Pre-Release Calculations 11 1.4 Setpoint Calculations 17

i. 2.0 RADI0 ACTIVE EFFLUENTS DOSE REDUCTION SPECIFICATIONS 28 2.1 Waste Reduction Specifications 30 2.2 Dose Projection Methodology 32 2.3 Total Dose Specification 9

34 3.0 RADI0 ACTIVE EFFLUENTS SANPLING SPECIFICATIONS 38 3.1-1 Liquid Releases (Batch) 40 3.l-2 Liquid Releases.(Continuous) 40 l

3.1-3 Gaseous Releases (Waste Gas Decay Tanks) 40 A 3.1-4 Gaseous Releases (RB & AB) 40 3.1-5 Reactor Bldg. with Personnel and Equipment Hatches Open .40a

- 4.0 RADI0 ACTIVE EFFLUENTS DOSE CALCULATION SPECIFICATIONS 41 4.1 Dose Specifications 43 4.2 Nuclide Analyses 46

.. 4.3- Dose Calculations L 51 L 4.4 Dose Factors 54 5.0 ENVIRONNENTAL NONITORING 79 6.0 ADNINISTRATIVE CONTROLS i

86

)

-i-1

1

-, .. l

, j 1

l LIST OF TABLES

]

Table

. Pane I RADI0 ACTIVE EFFLUENTS MONITOR SETP0INTS 2 II RADWASTE REDUCTION SYSTEN-DOSE PROJECTIONS 29 III GASE0US AND LIQUID EFFLUENT REPRESENTATIVE SAMPLING 39' ,

IV CUMULATIVE D0SE CALCULATIONS 42 4.4-1 Dose Factors for Exposure to a Semi-Infinite Cloud of Noble Gases $4 4.4-2 Inhalation Dose Factors - Infant -56 '

4.4-3 Inhalation Dose Factors - Child 57 4.4-4 Inhalation Dose Factors - Teen 58 4.4-5 Inhalation Dose Factors - Adult 59 4.4-6 Ingestion Dose Factors, Grass-Cow-Milk-Infant 61 4.4-7 Ingestion Dose Factors, Grass Cow-Milk-Child 62 4.4-8 Ingestion Dose Factors, Grass-Cow-Milk-Teen 63 4.4-9 Ingestion Dose Factors, Grass-Cow-Hilk-Adult 64 -

4.4-10' Ingestion, Dose Factors, Grass-Cow-Meat-Child 66 4.4-11 Ingestion Dose Factors, Grass-Cow-Meat-Teen 67 4.4-12 Ingestion-Dose Factors, Grass-Cow-Meat-Adult 68 4.4-13 Ingestion Dose Factors, Vegetation-Child 70 4.4-14_ Ingestion Dose Factors, Vegetation-Teen 71.

4.4-15 Ingestion Dose Factors, Vegetation-Adult 72

TABLE.I;- RADI0 ACTIVE EFFLUENT NONITOR SETPOINTS -- .

RELEASE TYPE- SETPOINT NUCLIDE ANAL. SETPOINT SETPOINT SPECIFICATION! CALCULATION MONITOR BATCH ~ CONT. ADJUSTMENT TYPE ** FREO.

'RM-Al X 1.1-1 1.2-1 P 1.3-1 '1.4-1

(Noble Gas)

RM-Al X 1.1-1 1.2-1 W l.3-1 1.4-2 (Noble Gas)

RM-A2- X* X -1.1-1 1.2-2 -l.3-1 W/P* 1.4-3 (Noble Gas) t j

RM-All X 1.1-1 1.2-3 1.3-1 P 1.4-4 't (Noble Gas)

RM-L2 X 1.1-2 1.2-4 1.3-2 P 1.4-5 .;

(Gasuna)

RM-L7 X X 1.1-2 1.2-5 W l.3-2 1.4-6 & l.4-7 (Ganuna)

RM-Al & !

RM-A2 N/A N/A 1.1-3 NA NA 1.3-3 NA (Iodine Channels)

This monitor is used in conjunction with -(or instead of) RM-All to monitor the release of the waste gas decay tanks. Nuclide analysis and setpoint' calculation must be perfonned for this monitor prior to waste gas decay tank release. At all other times,'it is a continuous source monitor and the setpoint is detennined weekly.

- For composited samples the results from the most recently. completed-analysis are used.

.l

_ __ -_ ~ - _ _ _ _ _ _ _ . . - .

.(

>S'_'

PRE-RELEASE CALCULATION 1.3-1 GASE0US RADWASTE RELEASE I. INTRODUCTION Prior to initiating a release of gaseous radwaste, it must be determined that the concentration of radionuclides to be released, and the flow rates at which they are released will not cause the dose rate limitations of Specification 1.1-1 to be exceeded.

!!. INFORMATION REQUIRED Results of appropriate Nuclide Analysis from Section 1.2

!!!. CALCULATIONS Noble Gas Gamma Emissions Dose Rate (Total Body)_= I (X/Q)KjQj mrem /yr. (1.1)

Noble Gas Beta Emissions Dose Rate (Skin) - I (X/Q)Qi(Lj + 1.1Mj) mrem /yr. (1.2)

Iodine 131. Tritium. Radioactive Particulates Dose Rate (I,T,P) = I (X/Q)PjQt mrem /yr. (1.3) where:

Kj = The total body dose factor due to gamma emissions for ench-identified noble gas radionuclide, in mrem /yr per pCi/mJ.

(See Table 4.4-1). l Lj =

The skin dose factor due to beta emissions for each identified noble gas radionuclide, in mrem /yr per pCi/m3 (See Table 4.4-1). g Mi

= .The air dose factor due to gamma emissions for each identified noble gas radionucide, in mrad /yr per pCi/m3 (unit conversion constant of 1.1 mrem / mrad converts air dose to skin dose). (See Table 4.4-1), j Pj = The dose parameter for radionuclides other than noble gases for the' inhalation pathway, in mrem /yr per pCi/m3,. (See Table 4.4-3).

l Qi

= The release. rate of radionuclides, i, in gaseous effluent from individual release sources, in pCi/sec (per unit, unless otherwise specified). Qj = Effluent stream nuclide concentration x flow rate.

x*

, L-PRE-RELEASE CALCULATION 1.3-2 LIQUID RADWASTE RELEASE I. INTRODUCTION Prior to initiating a release of liquid radwaste, it must be determined that the concentration of radionuclides to be released and the flow rates at which they will be released will not lead to a release concentration greater than the limits of specification 1.1-2 at the point of discharge.

II. INFORMATION REQUIRED Results of appropriate Nuclide Analysis from Section 1.2 III. CALCULATIONS Discharge C11 Co Ca CT Cs Ce F D+E Concentration - I + + + + + +

MPC7j 2E 4 MPCa- MPCy MPCs MPCre E where:

Cyj =

The concentration of isotope i, in the gamma spectrum-exluding dissolved or entrained noble gases.

CG

=

Total' dissolved or entrained noble gas concentration.

CT

=

Tritium Concentration from most recent analysis.

Ca

=

Gross alpha concentration from most recent analysis.

Cs =

Sr-89, 90 concentration from most recent analysis.

Cre =

Fe-55 concentration from most recent analysis.

E - Effluent Stream Flow Rate D =

Dilution Stream Flow Rate (Nuclear Services seawater flow only)

MPC< = 10CFR20 Appendix B, Table 'II, Column 2 Maximum Permissible . '

Concentration by isotope.

If the Calculated Discharge Concentration is less than or equal to 1, the discharge may be initiated. If the calculated discharge concentration is greater'than 1, action must be taken to reduce the effluent concentration or effluent stream flow rate prior to initiating discharge.

4

.. .c ,

. ,1 ;

Setpoint Calculation 1.4-1 Reactor Building Purge Exhaust Duct Monitor (RM-A1) <

(Batch Type Releases)

INTRODUCTION Following completion of the analyses required by Section 1.2-1 and determination of release rates and concentration-limits in accordance with Section 1.3-1, the monitor setpoint requires adjustment to ensure that alarm -

and pathway isolation occur if nuclide concentration limits are exceeded.

METHODOLOGY Reactor Building atmosphere is circulated through radiation monitor RM-A6 (containment atmosphere noble gas monitor) and the count rate is observed.

The observed count rate is correlated to a corresponding count rate for RM-Al (Reactor Building purge exhaust duct monitor), and factors are applied.to account.for background radiation, and the pressure difference between the detector chambers and exhaust vent. The obtained value establishes the l

- maximum allowable setpoint. The alarm / trip setpoint is adjusted to this or a more conservative value' prior to initiating the release. If the concentration of radionuclides.to be released is less than the effluent monitor LLD " Net CPM" is obtained from the calibration curve by determining the CPM which cor' responds to 2.5E-2 #Ci/ml.

JALCULATION Net CPM x VF /29.9 - VI) (pCi/cc/ CPM)A6 RM-Al Setpoint (CPM) =

+ Bkg PDRR (29.9-V6)(yCi/cc/ CPM)Al where:

Net' CPM =

The observed RM-A6 count rate, in cpm, less-background, or obtained from the calibration curve.

VF =

The vent fraction; that portion of the total plant gaseous release associated with this vent and discharge type.

Value can be set to a number between 0 and 1. The summation,of the vent fractions of RM-Al and RM-A2 cannot exceed 1.

PDRR = The noble gas gamma emission Projected Dose Rate Ratio calculated in accordance with Section 1.3. This ratio is the actual projected dose rate divided by the allowable dose rate referenced in Section 1.3-1, relationship 1.4.

V6 - The actual gauge vacuum reading at RM-A6 at the time of sampling.

VI -

The actual or average gauge vacuum reading at RM-Al during normal operation.

(pCi/cc/ CPM)A6 = #Ci/cc per cpm for RM-A6. This is based on an actual sample or derived from the calibration curve.

..... +

.o; ,..

, q ;.

1 .

~ ': .. .

- (gCi/cc/ CPM)Al' =

Ci/cc per cpm for RM-Al.. This is based on an actual 1., -

sample or derived from.the calibration curve.

Bk9 =

RM Al background count' rate'in cpm.

i l

t i

L L

I i

-I

.I 9 .

-e 1

i Setpoint Calculation 1.4-1A Reactor Building Purge Exhaust Duct Monitor (RM-A1) ,

(Special Release For Functional Testing of the ;

Reactor Building Purge System)

INTRODUCTION ;

Following completion of the analyses required by Section 1.2-1 and determination of release rates and concentration limits in accordance with Section l' 3-1, the monitor setpoint requires adjustment to ensure that alarm and pathway isolation occur if nuclide concentration limits are exceeded. ,

METHODOLOGY .

Auxiliary Building and Fuel Handling Area atmosphere-is continuously passed through radiation monitor RM-A2 and the count rate is observed. The observed count rate is correlated to a corresponding count rate for RM-A1, and factors are-applied to account for background radiation and the pressure difference between the detector chambers and exhaust vent. The obtained value establishes the maximum allowable setpoint. The alarm / trip setpoint is adjusted to this or a more conservative value prior to initiating.the release.

If the concentration of radionuclides to be released is less than the effluent monitor LLO " Net CPM" is obtained from the calibration curve by determining the CPM which corresponds to 2.5E-2 yCi/ml.

CALCULATION Net' CPM x VF [29.9 - Vlh (pCi/cc/ CPM)A2 RM-Al Setpoint (CPM) = 1 + Bkg PDRR (29.9 - V2) (pCi/cc/ CPM)Al where:

Net CPM =

The observed RM-A6 count rate, in cpm, less background, or obtained from the calibration curve.

VF =

The vent fraction; that portion of the total plant gaseous release associated with this vent and discharge type. VF can be set to a value from 0 and 1. The sum of RM-Al and RM-A2 vent fractions can not exceed 1.

PDRR = The noble gas gamma emission Projected Dose Rate Ratio calculated in accordance with Section 1.3. This ratio is the actual projected dose rate divided by the allowable dose rate referenced in Section 1.3-1, relationship 1.4.

V2 - The actual gauge vacuum reading at RM-A2 at the time of sampling.

V1 -

The actual or average gauge vacuum reading at RM-Al during normal operation.

- 18a -

v

Q ,M

,b , ro,.

c. ; ,

D F:. . ,, ,

9 : - . g l

'(#Ci/cc/ CPM)A2 "= pCi/cc per cpm for-RM-A2. This.is based on an actual 1 sample or-derived from the calibration curve.

N -(gCi/cc/ CPM)Al-- ' pC1/cc per cpm for RM-A1. This is based on an actual

. sample or derived from the calibration curve.

Bkg =

RM Al background count rate in cpm.

o

+

P t

r 5

I'

~

I r

l -l

,i I

t I

1:.r

- 18b -

g i ~

r.. ..*

'I. ;

setpoint calculation 1.4-1B Reactor Building Purge Exhaust Duct Monitor (RM-A1)

(Special Release Following ILRT of :

Reactor Building)

INTRODUCTION Following completion of the analyses required by Section 1,2-1 and determination of release rates and concentration limits in accordance with Section 1.3-1, the monitor setpoint requires adjustment to ensure that alarm and pathway isolation occur if nuclide concentration limits are exceeded.-

NETHODOLOGY Reactor Building atmosphere is circulated through a sampling apparatus. The Noble gas sample is analyzed to determine the projected dose rate ratio-(PDRR). Net CPM is obtained from the calibration curve by determining the CPM- -

which corresponds to 2.5E-2 yCi/ml. These values are combined with the monitor background and vent fraction, to arrive at the monitor setpoint. The obtained value establishes the maximum allowable setpoint. The alarm / trip setpoint is adjusted to this or a more conservative value prior to initiating the release.

Shortly, after beginning the purge, new RM-Al alarm / trip setpoints are determined using the methodology of Setpoint Calculation 1.4-2.

CALCULATION Net CPM x VF RM-Al Setpoint (CPM) = + Bkg PDRR where:

L Net CPM =

A value derived from RM-Al calibration curve.

VF =

The vent fraction; that portion of the total plant gaseous release associated with-this vent'and discharge type. VF t

L can be set to a value from 0 and 1. The sum of RM-Al and RM-A2 vent fractions can not exceed 1.

l PDRR = 1 l'

Bkg = RM-Al background count rate in epm.

I' A

I

- 18c -

r

< +- , , ,

e Setpoint Calculation 1.4-2 Reactor Building Purge Exhaust Duct Monitor (RM A1)

(Continuous Type Releases)

INTRODUCTION Following completion of the analyses required by Section 1.2-1 and determination of release rates and concentration limits in accordance with Section 1.3-1, the monitor setpoint requires adjustment to ensure that alarm and pathway isolation occur if-nuclide concentration limits are exceeded.

METHODOLOGY Reactor Building atmosphere is passing through radiation monitor RM-Al during a_ continuous type release. Factors are applied to the observed count rate to account for background radiation and vent fraction. The obtained value establishes the maximum allowable setpoint. The alarm / trip setpoint is I adjusted to this or a more conservative value weekly during continuous releases. If the concentration of radionuclides to be released is less than the effluent monitor LLD " Net CPM" is obtained from the calibration curve by determining the CPM which corresponds to 2.5E-2 pCi/ml.

CALCULATION Net CPM x VF !

RM-Al Setpoint (CPM) = + 8kg PDRR where:

Net CPM -

The observed RM-Al count rate, in cpm, less background, or i obtained from the calibration curve.

VF =

The vent fraction; that portion of the total plant gaseous release associated with this vent and discharge type. .

Value can be set to a number between 0 and 1. The j summation of the vent fractions of RM-Al and RM-A2 cannot ,

exceed 1.

PDRR_ = The' noble gas gamma emission Projected Dose Rate Ratio calculated in accordance with Section 1.3. This ratio is the actual projected dose rate divided by the allowable.

dose rate referenced in Section 1.3-1, relationship 1.4.

Bkg = RM Al background count rate in cpm.

1

, =_______ _ __ _ -__-_-

7 .

-9 Setpoint Calculation 1,4-3 Auxiliary Building & Fuel Handling Area Exhaust Monitor (RM-A2)

.(ContinuousTypeReleases)- +

INTRODUCTION

\'

g Following completion of the analyses required by Section 1.2-2 and y determination of release rates and concentration limits in accordance with Section 1.3-1, the monitor setpoint requires adjustment to ensure that alarm

[ and pathway isolation occur if nuclide concentration limits are exceeded.

METHODOLOGY Auxiliary Building and Fuel Handling Area atmosphere is continuously passing through radiation monitor RM-A2. Factors are applied to the observed count rate to account for background radiation and vent fraction.. The obtained j value establishes the maximum allowable setpoint. The alarm / trip setpoint is adjusted to this or a more conservative value weekly during continuous releases. If the concentration of radionuclides to be released is less than the effluent monitor LLD " Net CPM" is obtained from the calibration curve by l-determining the CPM which corresponds to 8E-3 pCi/ml.

CALCULATION Net CPM x VF RM-A2 Setpoints (CPM) = + Bkg-PDRR-where:

Net CPM =

The observed RM-A2 count rate, in cpm,-less background, or obtained from the calibration curve.

g VF- -

The vent fraction; that portion of the total plant gaseous I

i release associated with'this vent and discharge type, Value can be set to a number between 0 and 1. The l

summation of the vent fractions of RM-Al and RM-A2 cannot exceed 1.

L' PDRR - The noble gas gamma emission Projected Dose Rate Ratio-calculated in accordance with Section 1.3. This ratio is the actual projected dose rate divided by the allowable -

dose rate referenced in Section 1.3-1, relationship 1.4.

8kg = RM-A2 background count rate in cpm.

,. . ;w Setpoint Calculation 1.4-4 Waste Gas Decay Tank Monitor (RM-All)

(Batch Type Releases) s INTRODUCTION L

L Following completion of the analyses required by Section 1.2-3 and determination of release rates and concentration limits in accordance with Section 1.3-1, the monitor setpoint requires adjustment to ensure that alarm and pathway isolation occur if nuclide concentration limits are exceeded.

METHODOLOGY l

Prior to initiating a Waste Gas Decay Tank release, its contents are-drawn through radiation monitor RM-All and returned to the waste gas header.

Factors are applied to the observed count rate to account for background radiation and vent fraction. The obtained value establishes the maximum-allowable setpoint. .The alarm / trip setpoint is adjusted to this or a more l=

conservative value weekly during continuous releases. If the concentration-of ,

radionuclides to be released is less than the effluent monitor LLD " Net CPM" is obtained from the calibration curve by determining the CPM which corresponds to 20 pC1/ml.

CALCULATION Net CPM x VF x 24.7 RM-All Setpoint (CPM) = . + Bkg PDRR x P where-Net CPM --

The observed RM-All count rate, in cpm, less background, or obtained from the calibration curve.

VF =

The vent fraction; that portion of the total plant gaseous release associated with this vent.and discharge type. ;

Value is equal to 0.5. i PDRR =- The noble gas gamma emission Projected Dose Rate Ratio calculated in accordance with Section 1.3. This ratio is

.the actual projected dose rate divided by the allowable dose rate referenced in Section.l.3-1, relationship 1.4.

24.7 - 'The maximum pressure.(psia) which RM-All detector chamber should be subjected to. This corresponds to a flow of 15 CFM-from the release line to the vent.

-P =

Pressure (psia) in RM-All at time of obtaining net CPM.

Bkg = RM-All background count rate in cpm.

x___-__ :

N .-

1

!- _ . . . . c-ts>

. .a s g

i.,

s h

s i R 's 4' ,

t j i

. 4'

... t u i i

1 I DELETED i 1

I J

1

'l 1

r I

'j e

4 J

I s

d f

4

(

_, 4

c --

s !

CALCULATION OF INHALATION ,

PATHWAY 00SE FACTOR (Pj)

I) .

Pi = K' (BR) DFAj mrem / year per uC1/m3 where:

K' -

A constant unit of conversion - 106 pCi/uci BR -

The Breathing Rate of the child age group - 3700 m3/ year *

.DFAj - The maximum organ inhalation dose factor for the child age >

group for the-ith radionuclide, in mrem /pci. The total body is considered as an organ in the-selection of DFA. -

NOTE: For the inhalation pathway Pj = Rj, so values of Pt may be taken from Table 4.4-3.

I r

1 i

H^

References:

1): NUREG-0133, Section 5.2.1.1 4

2)' Regulatory Guide 1.109, Table E-5, and Table E-9 it.

a g ;g -, ' =:evm

- ~ ngg x +

1 wp _

3.,

4 g-

.s q t

y s. gw - 3 1

v. % -

4 8 }

k e 1 8 h I

i l

.1 l

c ,. 4 t

{

}

a 1

.k t

1 4

cA f

h

,. t,

< 4 Wh:' ..\

l= DELETED r

x ;

f' j w .

'J P

-- 3

? ~-

e. $

h. -(

f

%. 1

}fA - '2

.k 4

s

%l

.st

..f

.?

n-.

gs

.5

't 4

l '1

.{:

l- l*

l l

l 4 p- .p 3

I '1

- 1a i f 1 Y IJ r3 l F l

l

.f

, 3

.}

t 9

. t i

.k i

i

..q l

{

e' Oc-

k. .

c-27 -

eo

}2 IkJ+NkFhhbhWy- s., , ;W% .A ,,a ,m , , , , ,_e m, ,- -- 3 ris -N

. _ - .- _. . - - . . - .. . _ =

4 !

v, ,.-

4, =

Representative Sampling Method No. 3.1-5 (Reactor Building With Personnel And Equipment Hatch Opened) .

j The following conditions must be satisfied prior to having the RB personnel l hatch and equipment hatches opened at the same time: ,

o The Reactor Building purge exhaust fans are operational and the make-up fans are shut down; o The initial purge ~must run for at least 2 days at 30,000 SCFM. <

o The Reactor Coolant System is degassed and depressurized; r o The Reactnr Building recirculation system is continuously monitored by RM-A6 (all channels); or air samples are taken daily on each elevation of the Reacto Building; -

o A parti',dlate sampler is installed and operating on the Reactor Building refue'ing floor; o Air samples are taken as jobs in the Reactor Building necessitates (i.e.,

jobs that risk increasing the particulate, iodine, or gaseous concentrations'of radionuclides in the Reactor Building);

NOTE If the purge exhaust fans must be shut down, then either the personnel hatch or equipment hatch openings must be closed (e.g., temporary door '

installed in the personnel hatch).

Representative Sampling Method No. 3.1-6 (Reactor Building During Integrated Leak Rate Test)

DuetobuildIngoverpressure,prepurgesamplescannotbetakenfromRM-A6. -

Representative gas, iodine, particulate and tritium samples may be obtained from the Intermediate Building containment sampling apparatus or the Post-Accident Sampling System.

Reference:

Telecon-FPC (Dan Green, Dan Wilder) to NRC (Charles Willis) dated 03/15/85 at 0930;

Subject:

Personnel and Equipment Hatch Openings.

- 40a-

4 ;

i . .. )

O ,:

.. t DOSE CALCULATION 4.3-2' I (RADIOI0 DINES & PARTICULATES)-

The dose to an individual at or beyond the SITE BOUNDARY due to Iodine-131, Tritium and radioactive particulates with half lives of greater than 8 days is calculated as follows:

0 = 3.17 x 10-8 I WRjQt .

where:

0 -

The radiation dose to an individual at or beyond _the UNRESTRICTED AREA B0UNDARY, in mrem.

Rj =

The dose factor for each identified radionuclide m2(mrem / year) per uCi/sec or mrem / year per uCi/m$.1, in W =

X/Qforinhalationpatgway,25x106sec/m3atthesite boundary and 7.5 x 10- sec/m$ at the critical receptor.

W =

D/Qforfoodandgroundplanepgm2atthecriticalthway, the site boundary and 5.7 x 10- 1.9 x 10-8 receptor.

t Qi. -

Total yCi of isotope i released during the calendar quarter or calendar year, as appropriate.

3.17 x 10-8 =

The number of years in one second, yr/sec. :

l l^

1

Reference:

NUREG 0133, Section 5.3.1 FSAR, Table 2-20 l

l

,y 1 m r .. . L._*- ,

in ,

k DOSE CALCULATION 4.3-3 '

(LIQUID EFFLUENTS)

The dose or dose commitment to a MEMBER OF THE'PUBLIC from radioactive -

materials in liquid effluents released to UNRESTRICTED AREAS is calculated as '

follows: ,

y<cq D-I Aj7 ItCk ik Fk i k ;

i where: ' '

=

D The cumulative dose commitment to the total-body or any organ, 7, from the liquid effluents for the total time period I tk in mrem. l '

tk -

The length of the kth time period over which C ik 15 averaged for all _ liquid releases, in hours.

L Cik The average concentration of radionuclide, i, in undiluted liquid effluent during time period'tk from any liquid release, in pCi/ml.

s Ari =

The site related injestion dose commitment factor to the .

total body or any organ for each identified principal l-gamma and beta emitter as shown in Table 4.4-17 of this L

manual, in mrem /hr per #Ci/ml.

l-Fk Waste release flowrate + (Waste flow rate + Dilution flow rate)

Dilution flowrate is. equal to the total circulating water flow and/or Nuclear Services seawater flow available t during the release.

1 i-

'k

References:

1)_NUREG~0133,Section4.3.

2) Telecon/ Meeting Summary with C. Willis (USNRC) dated 01/16/85 regarding F-k l f: e'., .e b J

CALCULATION.0F INHALATION .*

PATHWAY DOSE FACTOR-(Rj) s

.~

Rj = K' (BR) DFAj mrem / year per uCi/m3 where: ,

K' - A constant unit of conversion - 106 pCi/uCi -

BR - The Breathing Rate of the represented age group:

1400 m3/yr - infant .

3700 m3/yr - child 8000 m /yr - teen

'8000 m3/yr - adult .'

DFAj - The maximum organ inhalation dose factor for the '

represented age: group for the ith radionuclide, in mrem /pC1.

l

References:

. 1) NUREG-0133, Section 5.3.1.1

? :2) Regulatory Guide 1.109, Table E-5, and Tables E-7 through E-10 L

s l- t i

=.

oc "

' t. : . ;'

Calculation-of Ingestion Dose Factor - !

Grass-Cow-Milk Pathway

. 5 R

c Qp(Vap) .

Fm(r)(DFLj)a

. fp fs (1 - f pfs) e-A tth i [ D/Q ] = K'

+ e'Aiy t I Aj + Aw Yp Ys l where: Unit = m2 . mrem /yr per pCi/sec Reference Table R.G.' l .109 i K' =

A constant of unit conversion,106 pCi/ Ci.

1

=

QF The cow's consumption' rate, 50 kg/ day (wet weight) E-3 U

ap

=

The receptor's milk consumption rate for age (a), E-5 in liters, yr Infant & Child - 330 Teen - 400 Adult - 310 '

Yp -

The agricultural productivity by unit area of pasture E-15 feed grass 0.7 kg/m2 1.

l Ys The agricultural productivity of unit area of E-15 stored feed 2.0 kg/m2 Fm The stable element transfer coefficients, in days /kg. E-1 s r =

Fraction of deposited activity retained on cow's E-15 feed grass 1.0 radiciodine 0.2 particulates I tr - Transport time from pasture to regeptor, in sec. E-15 l 1.73x100 sec (2 days)'

1 th =

' Transport time from crop field to receDtor, in sec. E-15 ,

7.78x100 sec. (90 days)

-(DFLj)a - The maximum organ-ingestion dose factor for the ith E-Il to

, radionuclide for the receptor in-age group (a), - E-14

-in mrem /pci

l Aj =' - The decay constant for the ith radionuclide. in sec -1

=

Aw The plantdecay constant surfaces for removal by weathering 5.73ofxactivjty 10- see on eaf and E-15 (corresponding to a 14 day half-life).

fp -. Fraction of the year that the cow is:on pasture ----

(dimensionless)=1*.

f3 - Fraction of the cow feed that is pasture grass ----

while the cow is on pasture (dimensionless) = 1*.

Milk cattle are considered to be fed from two potential sources, pasture p grass and stored feeds.

I-

= ,

r

3

. 4 . . , * <-

'* . Note: :The above' equation-does agl apply to_the' concentration of tritium in .r meat. A-separate equation is provided in NUREG 0133, section 5.3.1.4 to determine-Tritiur.value. '

t C

Reference:

The equation for R (D/Q) was taken from NUREG -0133'Section- '

5.3.1.3 i i

a k

V 5

l-

1. s

' )[

e t

- 60a -

- - + .

+ s Calculation of. Ingestion Dose Factor Grass-Cow-Meat Pathway >

M QF (Uap)

R [ D/Q) = K' Fr(r)(OFLj)a fp fs

+

(1 -~ pf fs) e d tih 1 I Aj + Aw _ Yp e'A tif

. . Ys s-where:- Unit = m2-: mrem /yr per pCi/sec Reference Table R.G. 1.109 K' =

A constant of unit conversion, 106 pCi/ C1.

.i

=

QF The cow's consumption rate, 50 kg/ day (wet weight) E-3

.Vap - The receptor's meat consumption rate for age (a), E.5 in kg/yr Infant - 0 2 Child - 41 Teen - 65 Adult - 110 Y

p = .The agricultural productivity by unit area of pasture E-15 feed grass 0.7 kg/m2 p -Ys

=

The agricultural productivity of unit area of E- H stored feed 2.0 kg/m2 o,

=

Fr The ' stable element transfer coefficients, in days /kg. E-1 l-r -

Fraction of deposited activity retained on cow's E-15 feed grass 1.0 radiciodine 0.2 particulates tr = Transport time from pasture to receptor, in sec. E-15 1.73x106 see 1

(20 days) i t.-

h Transport time from crop field to receptor, in sec. E-15 7.78x100 sec.. l s

(90 days)

(DFLj)a = The maximum organ ingestion dose factor for the ith E-Il to i+

" radionuclide for the receptor in age group (a), E-14 in arem/pCi Aj =

The decay constant for the ith radionuclide, in sec " ----

L Aw

=

The decay constant for removal of activity on E-15 I i.

L leaf and plant surfaces by weathering, 5.73 x 10-7 sec -1 (corresponding to a 14 day half-life),

fp =

Fraction of the year that the con is oc pasture ----

(dimensionless) = 1*.

f3 -

Fraction of the cow feed that is pasture grass ----

while the cow is on pasture (dimensionless) = 1.

t z

3 3 .. ,

.w. : as e _

~

i

[_ $ "; . '. i > #

e , ,, )

I br 7b' - -

Milk cattleJare considered to be fed from-two potential sources, pasture . .

grass and stored feeds.

Following the. development in ' Regulatory Guide 1.109, i? the values of-fp and fs will be considered unity, in lieu of site-specific i information provided in the annual land census report by the licensee.

Note: ' The above equation does nat apply to the concentration of tritium in tc <

- meat; A separate-equation is provided in NUREG 0133, section 523.1.4:

, _to determine Tritium value. -

f .

i

${ _t r

i t.

A t t 1

P A

y

-f m

Reference:

-The equation deriving R _(D/Q) t was.taken frcm NU.".EG 0133, Section 5.3.1.4.

tr;in NUREG 0133 is equivalent to ts in R.G.-1.109 Table E-15.

I 4

e - 65a--

e . ..e N '

= .

A Calculation of Ingestion Dose Factor

, Vegetation Pathway

)

v r L S R1 [ 0/0 ] - K' -

(DFLj)a Ua fL e-A tih+ Ua fg e'A tih

,(v (11 + Aw)

~

$' where: Units m2 . rarem/yr per uti/sec'. Reference Table. R.G. 1.iO9 K' - A constant of unit conversion, 106 pCi/gC1.

3 L h

L Ut - The consumotion rate of fresh leafy vegetation by the E-5 receptor in age group (a), in kg/yr.

Infant - 0 Child - 26

" Teen - 42 Adult - 64 hj_ Ua S

c The consumption rate of stored vegetation by the E-5 L_ receptor in age group (a), in kg/yr

=

Infant - 0 Child - 520 Teen - 630 Adult - 520 (DFLj)a - The maximum organ ingesting dose factor for the ith E-11 to E-14 rudionuclide for the receptor in age group (a),

in mrem /pCi.

ft -

The fraction of the annual intake of fresh leafy E-15 l

vegetation grown locally. (default 1.0) fg - The fraction of the annual intake of stored vegetation E-15 !

grown locally. (default 0.76) tl - The average time between harvest of leafy vegetation E-15 l i

and its consumption, 8.6 x 109, seconds, (1 day) th

The average time between harvest of stored vegetation E-15 and its consu'uption, 5.18 x 100 seccads, (60 days)

Yy - The vegetation a real density, 2.0 kg/m2 E-15 r - Fraction of deposited activity retained on the E-15 e'

vegetation l.0 radiciodine 0.2 particulates

=

Ai - The decay constant for the ith radionuclide, in sec 'I --

Aw -

E-15 The decay constant fcr remval of activi}y on lgaf and plant scrfaces by weathering, 5.73 x 10- sec ~

(corresponding to a 14 day half-life).

mJ - _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ . _ _ - - - - , - - - - - - - - - - - - -

4 4:

, 4 _- ...o n -, - . . _ . . ,

I Note: The above equation does DQ1 apply to the concentrations of tritium in vegetation. A separate equation is provided in NUREG 0133,-section-5.3.1.5 to. determine tritium values.

v

Reference:

-The equation deriving R, (D/Q). was taken from NUREG 0133, Section 5.3.1.5.

m-k a

,E +

r-I s.. 69a -

q r +,. <

..1 ,as '

u ' 4 !- p'

' ' -/ L j ;

6 ,

+

CALCULATION OF LIQUID EFFLUENT ADULT INGESTION DOSE FACTOR $ _'

Aj7 - 1.14E5 (21BFj + 5 Bij) DFj

[

Air =

Composite dose parameter for the total body or critical organ. l of an adult for nuclide,1, for all appropriate pathways, ;

mrem /hr per #1/mi 1.14E5 =

units-conversion factor 106 pci/pci x 10 3ml/kg + 8760 hr/yr

[

BFj =

. Bioaccumulation factor for nuclide,1, in fish, pCi/kg per pCi/L, from Table A-1 of Regulatory Guide 1.109 (Rev.1) or Table A-8 of Regulatory Guide 1.109 (original draft).

BIj -

Bioaccumulation factor for nuclide, i, in invertebrates, pCi/kg per pCi/L, from Table A-1 of Regulatory Guide 1.109 (Rev.1) or Table A-8 of Regulatory Guide 1.109 (original draft).

DFj. -

Dose conversion factor for nuclide, i', for. adults in c pre-selected organ, 7. in mrem /pci, from Table E-11 or. ;

Regulatory Guide 1.109 (Rev.1) or Table A-3~ of Regulatory l l

Guide. l.109 (original draft).

t 1

l l

t n

a l

I:

Reference:

The equation for Saltwater sites from NUREG 0133, Section 4.3.1, where Uw/Dw = 0 since no drinking water pathway exists.

l e

l:

1s '. . s I s . ;

. .y Setpoint Calculation 1.4 1A !

Reactor Building Purge Exhaust Duct Monitor (RM A1) !

($pecial Release For Functional Testing of the ReactorBuildingPurgeSystem) ;

INTR 000CT105 Following completion of the analyses required by Section 1.2-l iand determination of release rates and concentration limits in accordance with Section 1.3 1, the monitor setpoint requires adjustment to ensure that alarm and pathway isolation occur if nuclide concentration limits are exceeded. :

METHOD 0L0tY Auxiliary Building and Fuel Handling Area atmosphere is continuously passed through radiation monitor RM A2 and the count rate is observed. The observed count rate is correlated to a corresponding count rate for RM A1, and factors are applied to account for background radiation and the pressure difference between the detector chambers and exhaust vent. The obtained value

' establishes the maximum allowable setpoint. The alarm / trip setpoint is adjusted to this or a more conservative value prior to initiating the release, if the concentration of radionuclides to be released is less than the effluent monitor LLO

Net CPM" is obtained from the calibration curve by determining the CPM which corresponds to 2.5E 2 #C1/ml. '

CALCULATION '

Net CPM x VF [ 29.y - V1h (sci /cc/ CPM)A2 l RM Al Setpoint (CPM) . + Bkg -

PDRR ( 29.9 - V2) (pti/cc/ CPM)Al where:

l '

Net CPM =

The observed RM A2 count rate, in cps, less background, or ,

obtained from the calibration curve.

l VF =

The vent fraction; that portion of the total plant gaseous release associated witn this vent and discharge type. VF can be set to a value from 0 and 1. The sum of RM Al and RM A2 vent fractions car, not exceed 1.

PORR = The noble gas gamma emission Projected Dose Rate Ratio calculated in accordance with Section 1.3. This ratio is the actual pro.iected dnse rate divided by the allowable .

dose rate referencei in Section 1.3-1, relationship 1.4.

'V2 -

The actual gauge vacuum reading at RM-A2 at the time of sampling.

VI =

The actual or average gauge vacuum reading at RM Al during '

normal operatien.

18a -

. -- .. .. . . . . - = - - - _ - -

=

i Table 5.1-1 i

Environmental Radiological Monitoring !

Stations Locations '

DIRECTION DISTANCE STATION LOCATION !

FROM PLANT FROM PLANT (mi) l l

C04 State Park Old Dam on River '

ENE 6.3 near road intersection C07 Crystal River Public Water Plant ESE 7.5 C09 Fort Island Gulf 8each S 3.1 C10 Indian Waters Public Water Supply ESE 5.9 Cl3 Mouth of Intake Canal wsw 3,4 f

Cl4H Head of Discharge Canal NW 0.1 i Cl4M Midpoint of Discharge Canal W l.2 '

Cl4G Discharge Canal at Gulf of Mexico W 2.8 i C18 Yankeetown City Well N 5.2 >

C19 NW Corner State Roads 488 & 495 ENE 8.5 C29 Discharge Area N 2.0 C30 Intake Area WSW 3.6 C40 Near N.E. Site Boundary E 3.5 near exca.vated pond & pump station '

C41 Onsite Meteorological tower SW 0.4 C46 North Pump Station N 0.4 C47 Office of Radiation Control, Orlando ESE 67 C48A Onsite North of CR 4 & 5 N 0.8

C488 Onsite NNE of CR 4 & 5 L

NNE 0.8 l-l l

l

_ ___ _ _ o NA -- ---- - - ----

ML20012B555

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