ML20153G109

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Semiannual Radioactive Effluent Release Rept for Jan-June 1988
ML20153G109
Person / Time
Site: Crystal River 
Issue date: 06/30/1988
From: Johnson S, Widell R
FLORIDA POWER CORP.
To:
NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
References
3FO888-16, NUDOCS 8809080041
Download: ML20153G109 (48)
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{{#Wiki_filter:.-. SRGANNLRL RADIQACPIVE DTUJENP NIEASE M 1988 i JANUARY - JUNE i t i l FIIRE PCHER 03GUUCICH CRYSTAL RIVER - LNIT 3 j FACIIIIY OPERATDC LIONSE No. DFR-72 DOCEEP Ho. 50-302 AUGJST, 1988 f t ( hl q nanagar, site Nucle e services onte: 4-22-7( 1 l l i. { 8809080041 880630 3 {DR ADOCK 05000302 i PDC ( 4

s o GNIENIS EhCiE o UflR XI) CIT W 1 o TARJIAR CATA SLM9AIES partM DTIUDfIS 2 LIQUID P37IDDfIS 4 i RADOSFE SHIINDfIS 7 o UUND GRAlHS t GAcrtM RELEASES FISSICH AND ACTIVATICH 190C0CIS 9 UUTILM 10 LIGJID REEASES FISSION AND ACTIVATICH H000CIS 11 'IRITILM 12 SOLID PADHASIT SPDir RESINS, EIC 13 IRY ONRGESIBIE WASIE, EIC 14 o INPIAltG:D REIEASES i 15 l o RADICW.TIVE WASIE ' IRE.N SYSIDG 15 D&DONDfIAL RADICIMICAL ENTICRDC HOGRAM o 15 o DTIUDff )ONTIUt OPDRBILITY 15 o CDCM AND PCI-16 i ~

INIHXIKTIW 'Ihis report is subnitted as iW by hchnical Specification 6.9.1.5.d to Crystal River Facility Operating License No. DER-72. In accordance with Technical Specifications, the following inforntion must be included in this report: A samary of the quantities of radioactive liquid ard gaseous effluents and solid waste released frca the plant as outlined in Regulatory Guide 1.21 (Rev.1,1974) with data sunnarized on a quarterly basis following the fomat of Appendix B thereof. For each type of solid wasta shipped off site: Container Volume ' Ictal Curie Quantity (specified as measured or estimated) Principal Radionuclides (specified as measured or estimated) Type of Wasta (e.g., spent resin, ocupacted dry waste) Type of container (e.g., ISA, Type A, Type B). Solidification Agent (e.g., cenant) A list and description of unplanned releases to unrestricted areas. A description of any charges to ther i Prreaam Control Program (ECP) Off-Site Dose Calculation Manual (OD34) l Radioactive Waste Treatmnt Systems A listing of new Environmntal Radiological Monitoring Program dose calculation location charges identified by the land-use census. Informtion relating to effluent monitor 1s being inoperable for thirty or more days. Information regartling meteorological data and environmental dose a=="nents will be included in the year-end semiantual report as required by hchnical 1 Specifications. In addition to the required data, trend graphs of m: ries relaa* rer six renth period have been included in order to mt the current data into historical perspective. With Itgarti to the 1987 year-ervi Semiannual report an error has been noted in Table IV-2 (Total curies): 1.69E+03 should be 1.69E+02. 1 e - r-m-y-~

1 TMEE1 EFTILENT AND NMFIE DISPOSAL SIIMIAltWAL REPCRT - 1988 i GpprrM EF7IUDfrS - SL999&ICN OF AIL REIEASES Unit Quarter Q.arter Est.'Ibtal 1 2 Error % A. Fissicn ani Activation Games 1. 'Ibtal Ralease Ci 2.22E+02 1.26E+02 110 2. Average Release Pata for Period uC1/sec 2.82E+01-1.60E+01 3. Pezcent of Waal Specificatien Limit 2.02E-01 1.14E-01 B. Ioiines 1. Total Iodine - 131 Ci 5.12E-05 1.73E-05 110 2. Average Release Rate for Period uC1/sec 6.52E-06 2.20E-06 3. Percent of Techni::al Specificaticm Limit

4. 55E-01 1.49E-01 C.

Particulates 1. Particulates with half-lives > 8 days Ci 1.35E-04 8.13E-05 100 2. Average Release Rats for Period uCi/sec 1.722-05 1.03E-05 3. Percent of h:hnical Specificaticn Limit 4.55E-01 1.49E-01 4. Gross Alpha Radioactivity Ci

  • f.38E-11

<IID D. Tritium 1. Total Ralease Ci 4.06E+00 5.32E-01 90 2. Average Release Rate for Period uC1/sec 5.16E-01 6.77E-02 3. Percent of Technical Specificatico Limit 4.55E-00 1.49E-01 1 l l 2

TABLE 2 EFTWDE AND WMrIT DISPOSAL SDEMOUAL REKRT - 1988 GASEQUS EFTWDirS - GIOUND I.EVEL REI?.ASES CarrINLOUS MXE BN1W MXE }Melides Released Unit Quartar 1 Quarter 2 Quarter 1 Quarter 2 1. Fission games aInon-41 Ci krvetm-85 Ci 8.02E-01 2.88E+00 kryptgri-85m Ci krveton-87 Ci krvotm -88 Ci x& se131m Ci 3.93E-01 7.9?E-01 xerdi-133 C1 1.88E42 8.79E+01 2.16E+01 2.65E+01 xerdr-133m Ci 6.72E-02 1.35E-01 1.59E-01 xeren-135 C1 1.11E+01 7.52E+00 2.11E-02 1.71E-02 xergr-135m C1 Xerdr-138 Ci unidentified Ci O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 Total for Paried C1 1.99E+02

9. 54 E+01
2. 30E401 3.04E+01 Iodince laiine-131 Cl 5.12E-05 1.73E-05 imiine-133 C1
1. 54 E-05 1.16E-05 icvH ne-135 Ci Tttal for Paried Ci 6.66E-05 2.89E-05 0.OOE+00 0.OOE+00 3.

Particulatas mnaaname-54 Ci c'nhalt-58 Ci iron-59 C1 ggNtit-60 Ci zine-65 Ci stiwit.ium-89 Ci 7.02E-08 2.46E-11 streii.i c -00 Ci rolvbderum-99 Ci te11urium-132 Ci cesium-134 C1 1.98E-05 5.75E-06 00sium-137 Ci 4.02E-05 1.06E-05 cesium-138 C1 2.96E-07 curium-141 Ci cerium-144 Ci l unidentified Ci 7.48E-05 6.49fE5 0.OOE+00 0.OOE+00 Total Ci 1.35E-04 8.15E-05 2.46E-11 0.OOE+00 3 L 1

TABLE 3 e DTIUDir AND Wp6'IE DISPOSAL SDEANNAUL REP:RT - 1988 IJQUID EFrIUDTIS - SNCH OF AIL RELEASES Unit Quarter Quarter Est.htal 1 2 Error % A. Fissica and Activatico Prc& acts 1. Total Release (not including tritium, games, al;ha) Ci 9.36E-02 7.02E-02 100 2. Average diluted concentratico daring period uCi/ml 6.08E-09 4.28E-09 3. Percent of applicable limit 4 2.3E+00 1.12E+00 B. Tritium 1. Total Release Ci 6.96E+01 1.48E+02 40 2. Average diluted concentration darirg period uCi/ml 4.52E-06 9.02E-06 3. Perrent of a plicable limit 1.51Z-01 3.01E-01 C. Dissolved and entrained gases 1. Total release C1 1.22E+00 1.62E+00 100 2. Average diluted ocmoentration &lring period uCi/ml 7.92E-08 9.88E-08 3. Percent of agplicable limit 3.96E-02 4.94E-02 D. Gross alpha radioactivity 1. %tal release Ci <IID <T m 100 E. Volume of Waste zulmased (prior to dilution) 1. Batch and Ccritinuous )bdes Liters 1.62E+07 9.20E+06 10 F. Volume of dilutice water usal during period 1. Batch and Continucas )b$es Liters 1.54E+10 1.64E+10 10 4 I

l TABLE 4 EFTIUfNT AND NPSIE DISP 06AL SDGANNUAL REPGC - 1988 TJQUID DTIUDfIS cmrINUCUS MXE BA20{ MXE Nuclides Released Unit Quarter 1 Quarter 2 Quarter 1 Quarter 2 sodium-24 Ci drmium-51 C1 5.79E-04 marmrmee-54 C1 1.7EE-03 7.89E-04 irtm-55 C1 4.57E-03 3.12F-03 2.29E-03

  • 1t-58 C1 2.42F-02 4.43E-03 irtm-59 Ci tvhnit-60 Ci 1.95E-02 9.95E-03 Zinc-65 Ci rubiditm88 Ci s h ilium-89 C1 1.15E-04

_6.61E-04 1.91E-03 strailium-90 C1 2.78E-05 3.45E-05 streitiurr92 C1 1.78E-03 4.65E-03 n1&ium-95 Ci 6.45E-04 1.27E-04 zirc.adWS C1 1.71E-04 zirconium-97 C1 1.94E-04 5.94E-04 rolstiisi arr09 C1 6.94E-05 technetitr 09m C1 q 1.76E-04 4.37E-04 ruthenitan-103 Ci Ivthenium-106 Ci 3.00E-03 silver-110m Ci 5.10E-03 1.27E-02 icdine-131 Ci 1.05E-04 3.28E-05 lodine-132 Ci te11urlun-132 Ci iodine-133 Ci 6.24E-06 3.01E-05 cesitan-134 Ci 5.02E-05 1.06E-05 casitan-136 Ci cesium-137 Ci 2.30E-04 _1.79E-04 barium-133 C1 2.76E-04 lanthandm-140 Ci 5.85E-05 gerium-141 C1 cerium-144 C1 ynidentified Ci 9.92E-03 0.00E+00 1.94E-02 2.8?E-O? 'Ibtal for period (abcno) C1 1.46E-02 0.00E+O-O 7.90E-02 7.02E-02 5

TAM E 4 (O MPINUID) DTIUENT AND WMrIE DISICSAL SDURNUAL REPCRP - 1988 LIQUID DTUJINTS crwrINUCUS MXE BA1UI ECE Dissolved & Entrained Unit Quartar 1 Quarter 2 Quarter 1 Quarter 2 Gases amcn-41 C1 kn'otm-85 C1 2.94E-03 1.66E-02 krvetm -85m C1 2.17E-05 kn'ottn-88 Ci XEsr-131m C1 1.85E-02 2.68E-02 xersr-133 C1 1.1RF+00 1.56E+00 yersr-133m Ci 7.70E-03 1.15E-02 XEsir-135 C1 4.h5E-03 8.99E-03 xersr-135m Ci tritium C1 1.31 E-02 0.00E+00 6.96E+01 1.48E+02 i i 6 1

TMEZ 5 EF71 DINT AND WMTIE DISPCSAL SDUN#4UAL.6 - 1988 SOLID WMrIE AND IRRADIAIYD FUEL SKIMENIS A. SOLID WMTIE SHIPPED OF73rIE FtR RRIAL CR DISPOSAL (Nm irradiated fuel) First 1. Type of wasta Unit 6-scrith Est. 'Ibtal period Error, % Spent risins, filter slud; pas, evaporator m 2.92E+01 3 a. bottans, etc. Ci 8.332+02 20 b. Dry crapressible wasta, ccritandnated m 1.04E+02 3 equip, etc. Ci 9.44E+00 50 c. Irradiated WWJ, ocritrol rods, etc. d. Other (describe) Estimate

  • of major ruclide otapositicr1 (by type of waste) 2.

a. Co-58 5.05E+01 Ni-63 6.89E+00 Cs-137 1.33E+01 Mn-54 4.29E+00 CD-60 1.01E+01 Cr-51 2.22E+00 On-134 8.81E+00 b. G-137 2.67E+01 On-134 1.19E+01 Fe-55 2.02E+01 Ni-63 4.08E+00 CD-60 1.78E+01 0D-58 1.24E+01 L c. d. t 3. Solid Wasta Disposition Nunber of Shipments Mode of Transporaticri Destination 11 Transport Truck - Chase-Huclear sim, Inc. Exclusive Use Vehicle Barwell, SC B. IRRADIA'IID IVEL SHIRENIS (Disposition) Nurber of Shirrents )tde of Trargeration Destination Hone NA NA With the exc:!ptico of Secordary Resin shipnents all curie values and prirx:iple radionuclides are determinal by iniirect methods and are therefore estimates. Secondary Resin shipwnts are evaluated by direct (gama Wg) ard crii.e (applyirg scalirn factors) methods, with the results beirg a ocrbinatico of measured ard estimated value. 7

ThBIE 6 EFFIUD F AND WMTIE DISICSAL F2MIAl4 CAL REK5tr - 1988 SHIRENT SW99RY DNTE AND CCNIADER '!UIAL WMrIE CCNIAINER SOLIDIF SHIRENT f NOIDtE* CURIES 3RINCIPIE RADICNy'rm TYPE TYPE AGDff 1-22 H-3, C-14, Fe-55, Cs-134, 88-1 402.2 1.45E-04 On-137 SC ST** N.A. 1-28 Itt-54, Fe-55, Cb-58, Cb-60, 8 A-6 86 9 7.5 2.37E-01 Ni-63, 14)-95, Zr-95 CN SP N.A. 2-11 141-54, (b-58 Cb-60, Ni-63, 88-7 12 1 5.37E+02 On-134, On-147 SR ' Pipe B N.A. 2-16 Cr-51, W 54, Cb-58, C W, 88-8 120.3 2.91E+02 Ity-95, On-137 F 'Iype B N.A. 2-25 8597.5 )tt-54, }4-55, Cb-58, Cb-60, 88-9 1 9 11.6 2.73E-01 Ni-63, }ty-95, Cs-137 CN SP N.A. 3-10 8597.5 Wr-54, Fe-55, Cb-58, Cb-60, 88-11 1 9 11.6 2.92E-01 Ni-63, }t>-95, Os-137 CN SP N.A. 3-17 H-3, )tt-54, Cb-58, Cb-60, 88-12 193 1.61 Ni-63, Cs-134, Cs-137 SR St N.A. 4-15 Fe-55, Cb-58, Cb-60, Ni -63, 88-14 8647.5 4.38E-01 Cs-134, Os-137, N-241 CW Sr N.A. 4-31 H-3, Cr-51, Itt-54, Cb-58, Cb-60 88-15 194.1 2.97 Ni-63, Nb-95, Ag-110tn, Cs-134, OR Sr N.A. Ch-137 5-10 Fe-55, Cb-58, Cb-60, Ni-63 88-17 328.5 8.36 Os-134, Os-137, N-241 }M Sr N.A. 6-29 4996 141-54, Fe-55, Cb-58, Cb-60, 88-18 39120 2.82E-01 Ni-63, Sr-89, Nb-95, C2-134, Cs-137, N-241 }M ST N.A. ] f WASIE THE: SR - Spent Resin )M - Non-<trpacted Waste CE - Omtaminated Equiprnt SC - Secordary Resin CN - Cttpacted Waste IC - Irradiated Otmpctients F - Filters IB - Evaporator Battcrs o Container volume in cubic feet Sr - Strary; Tight 8

W FIGURE 1 - T._ GASEGUS RELEASES FIS$10N AND ACTIVATION PRODUCTS e =+ 6 .w .e. .e ep= _.. - -e e. .~. 10,000 .. Inh N1 . h N-a 4 . 6 ., _ _ _%u Mp em e h@ e M eW - .= m -W Nhh6

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FIGURE 3 - .._1 LIQUID RELEASES ~ ~ ~ ~ - ~ ~ ~ ~ ~ ) _, _ ~ ^ ~ ~ FIS$10N AND ACTIVATION PRODUCTS ~ ~~ -~ -1 I l 1: (_-___ l l l* P -J i i . - ~... _, I p. l___ g i t y I ~. i~ i I t - t. g j 1.. _ :.. .. _. n l.._ - _ _. '. _ (. -- t. Curtes .11 ...e.. m. ee. w. s 9 b g 4 c.._ - 4 .__w._ _ - 1 t. i . ~.. _ _ -. __ _? f } } g ~ ~. _. - 9 ~C_*,.I_.'.*_'. ' ~.. l'. I,,. j ~ 1 l } l I r m . _.1 l { i t -4 h .01t I ,,,p..g _.m ,e eMi.e. ,~ _ .4... s.. _.. --. y-W en$ s i.-- pe awe e.g ,.gmw gw g p.e e =O.m. h. p _g .ee.+- .m.._-. -.e-_: ._e%.m. .m b 'W1' . 9 77~ 78-7f' 80 81 82 83 84 85 56 87 88~ 89 90 Year 11 ^ y

FIGURE 4 - LIQUID RELusts TRITIUM j i i 1,000. r . ~. - - - 1 ~. . - -. -.. + - ee,,---_ ...~._w-t w g 9 g m ...e ,e

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F M ?RE 5 - ~ ~ SOLIO RADIESTE 3 PENT RESINS. F'LTER 51.UDGES. EVAPORATOR 80TTONS. ETC. 3 q 5 4-3 a w: I4 '[i 3 I 8 f 1 I s r s .i i a k k 1 I k i 4 - _ _..I. -r. - : : ...

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IMPUMEID RE[EhSES 'Ihere wre no urplanned liquid relem for the pericd of this report. 'Ihere were two urplanned T-m ruim-as described below: D6IE DESGIPTIW 2-28-88 Steam relief due to reactor trip. Estim tad rulease amount: 5,J E-4 curies 4-1-88 Steam relief due to failure of limit switch durig valve (M3IVS) struking. Estimated release amount: 9.6 E-5 curien RADIQACTIVE 6 'Ihere were no si ficant charges to the radioactive vasta treatment systers for the period of report. 1 N RADIMOGImL KMritRIM3 ROGWE 'Ihe June 1988 lard-use census did not identify any new dose calculation locations. ] EFMIE!NF MMHGt GWWHLT1Y i l tb effluent monitor was incperable for a pericd of thirty days or core, i l 1 f 15

GEM AND KP 'Ibers were several changes to the 000( ard the PCP, the descriptiens Cf Weh are prwided below. 'Ihe twised pages have also been incitded. i M IElDQUITICtl l Introducticn Changed to iniloats s',1nual done calculations. l 6, 7, 8, 46, 47, 48 Deleted referince to noble gas acnitor IlD. 12, 13, 17, 18, 18a, 18b, AMed guidarce for establishing trip 18c, 19, 20, 21, 22, 23, 24 astpoints when affluent omountraticns are below monitor IID. 14 Corrweted grammar in latroductory sentance. 17, 18a, 18c, 19, 21, Changed intrtxhactory atatement to iniicata 22, 23, 24 that eff2.uent acnitor trips on otnoentraticn l enly. 80, 82 Corrected envim. atal station locations. i ER2HMS E IESGIITICH t Title Deleted non-requirwi signaturns atd mMai positicn for Manager, site Haclear servloes. i Mimi Arpenilx I to 'Inble of Ocntants 2 Incitded.krtlett PCP % rt Maber I 9 Clarifind Mirq Trugoancy in At.tdan 5.3.2.2. 14, 15 AMed drawinys of solidificaticn and dewataring systems. i 'Ibese PCP chvges WM effattive in Oaruary of this )wLr and were sutnitted in the 1987 year-eni report in February. 'Dochnical Specificaticos, h:urver, req 11re Winion of KP charges in the seniarnn1 Report for the period in shich the charges were rado. 0:nsecpently, the charnes are beirn sutnitted again. 16 1 )

INT 2000CT!0s The off-site Dose Calculation Manual (00CM) is provided to support implementation of the crystal River Unit 3 Radiological Iffluent Technical specifications. The 00cm contains calculational methods to be used in determining the dose to asabers of the public resulting from routine radioactive ef fluents released from Crystal River Unit 3. More accurate estination of doses is performed annually la preparatios of the year-end semiannual Radioactive Itfluent Release RepoJt. The 00CM also contains the methodology used to determine effluent aoattor alars/ trip setpoints which assure tAat releases of radioactive materiali remain within specified concestrations. The 000t will be controlled by the site Nuclear services Department and revisions shall be made with the approval of the Manager, site Nuclear services. In addition all revisions must se approved by the Plant Review Coenittee prior to implementation at Crystal River Omit 3. Ristorical documentation and distribution of the 00CM shall be the responsibility of the Nuclear Operations Records Manager per N00-05, Document Control Prograa, i l REV S ON 11 n

NUCLIDE ANALYSIS 1.2-1 REACTOR BUIta!NG PURGE EIRAOST NOCLIDE SAMPLE SOURCE LLD(b)(uci/ec) A. Principal Canna taitters (a) Mn-54 Fe-59 1x10* /1x10-11 1x10* /1x10-11 Co 54 Pre-release grab sample for Batch 1x10' /1x10*11 Co-60 Type release. Weekly Particulate 1:10* /1x10-11 In-65 Filter Analysis for continuous (c) 1:10* /1x10-11 Mo-19 type release. Cs-134 1x10" /1x10*11 Cs-137 1x10" /1x10*11 Ce-141 1:10* /1x10*11 Ce-144 1:10" /1x10*11 1:10* /1x10*11 ~ Kr-87 ) rrt-release grab sample for Batch 1:10*4 Kr 88 type release. Moble das monitor 1:10*4 i Xe 133 during batch and continuous releases ta10-4 Xe 133a Grab sample within 2-6 hr. following 1x10*4 Xe-135 startup, shutdown or > 15% RTP 1x10-4 Xe-130 changs in 1 kr. 1x10-4 3. Zodine 131 Pre-release greb sample for Batch RA/t a 10 12 , type release. Weekly charcoal filter and once per 24 hr for 7 days following startup shutdown or > 15% RTP change in 1 kr unless I-131 concentration at site boundary ( 10% 10 CFR 20 limit. C. Tritius Pre-release Grab sample and within 1x10-6 12 11 kr following flooding of refueling canal and once per 7 days while canal is flooded. D. Gross Alpha Monthly Particulate Filter Composite 1x10-11 E. St-49 Quarterly Particulate Filter Composite 1x10-11 T. Sr-90 Quarterly Particulate Filter Composite 1x10-11 (a) and setpoint calculations.Other identified Gaana Emitters not listed in this table shall (b) The first value refers to the LLD for pre-release grab sample; the second value refers to the LLO for weekly Particalate Tilter Analysis. (c) Reactor Building Purge is considered continueus af ter a miniaua of one Reactor Building volume has been released on a continuous basis (i.e., batch type!. first volume is a REV SIN 11 s

~ NUCLIDE ANALYS!$ 1.2o2 AUI!LIARY 90!LDING AND FUEL BANDLIF3 AAEA EXMAUST O NOCLIDE SAMPLE SOURCE LLD (b) (uci/al) A. Principal Gaasa Eastters (4) ~ Mn-54 Fe 59 1x10-4/1x10*11 Ix10-4/1x10* ' Co 58 Weekly Particulate Tilter Analysis. 1x10*4/1x10*11 Co-60 1x10*4/1x10*11 In 65 1x10*4/1x10-11 Mo-99 Cs-134 1x10-4/1x10*Il l Cs-137 1x10-4/1x10*11 Ce-141 1x10*4/1x10-11 ) Ce-144 1x10-4/1a10-11 1x10*4/1x10*11 t Kr-87 Monthly Grab Sample and x10*4 Kr-88 Continuous Noble Gas monitor. 1x10-4 Xe-133 Crab sample within 2-6 hr to11owing 1x10*4 Xe-133a startup, shutdown or > 15% RTP 1x10*4 Xe-135 change in 1 hr. Xe-138 1x10-4 1x10*4 3. Zodine 131 i Weekly Charcoal Filter analysis and once 1x10-12 per 24 hr for 7 days following startup i shutdown or > 15% ATP change in l 1 hr unless I-131 concentration at site boundary < 10% 10 CFR 20 limit. t C. Tritius Monthly Crab sample and within tx10-6 12-24 hr following flooding of refueling canal and once per'7 dayo t while cana), is flooded. D. Cross Alpha Monthly Particulate T11ter Composite 1x10 11 l E. St 89 Quarterly Particulate T11ter Composite 1x10*11 T. St-to Quarterly Particulate filter Composite 1x10*11 (a) other identif aed Gamma Emitters not listed in this table shall be ancluded in dose and setpoint calculations. (b) The first value refers to the LLD for pre-release grab sample; the second value refers to the LLD for weekly Particulate T11ter Analysts. 1 1 1 REVISION ll w g

NOCI,IDE AR&!YSIS 1.2 3 5487E CA8 DECAY TANit8 NUCLIDE SAM LE SOURCE !.LD(b)ImCi/a1) A. Principal Oaana taitters (a) Mn-54 Te-59 1x10-4/1x10*11 Co-58 1:10*4/1x10-11 Co 60 1x10-4/1x10*11 1:10*4/1x10*11 En-65 Pre-release crab' sample and Weekly 1:10*4/1x10-11 Mo-19 Particulate Filter $ ample free SM-12. 1:10*4/1x10*11 Cs-134 Cs-137 1x10*4/1x10*11 Co-141 ty10*4/1x10*11 Co-144 1:10*4/1x10-11 1:10*4/1x10*11 Er-87 Kr-88 1:10*4 1:10*4 te-133 Pre-release Orab sample. 1:10*4 Xe-133s Xe-135 1:10*4 Ie-138 1:10*4 1 10-4 5. todine 131 Weekly Charcoal Filter from RM-A2. 1 10-12 ta) other identified Gaana taitters not listed in this table shall be included in do and setpoint calculations. (b) The first value refers ta the LLD for pre-release grab sample the second value refers to the LLD for weekly particulate Tilter Analysis. ~ REVISIC N ll e n

3 Flow Rates (Variable - based on s;tpoint needs, nominal or sax- ) laus values listed belcS). 1) Reactor Building Purge Exhaust Duct = 50,000 cia = 2.4 x 107 cc/see 2) Auxiliary Building and Tuel Handling Area Exhaust Duct = 156,000 cia e 7.4 x 107 cc/sec 3) wasteGasDeca7TankReleaset,ine=50 cia max = 2.4 x 10 cc/sec X/Q = 2.5x10-6 sec/m3 for all vent releases. This value is the highest calculated annual average relative concentration for any area at or beyond the unrestricted ared boundary. In order for a gassaus release to be within the limits of specification 1.1-1, the Projected Dose Rate Ratio (PORR) must not exceed 1. The PDRA for each limit is calculated as fellows: PDRRyg = PDRTS / 500

  • (1.4)

PDRRSK PDRgg / 3000 (t.5) = l PDERogo = PDRoRc/ 1500 (1.6) PDRyg = Projected Dose Rate to the TOTAI, 8007 due to noble i gas eanissions. l PDRSK = Projected bose Rate to the SKIN due to noble gas emaissions. j PDRogos Projected Dose Rate to any organ due to inhalation of i

iodine, tritius and particulates trith half-lives I

greater than 8 days. i 500 = The allowable total body dose rate due to noble gas gamma emissions in area /yr. i 3000 ] = The allowable skin dose rate due to noble gas bota emissions in area /yr, i 1500 = The allowable organ dose rate in area /yr. i t t !! the concentratica of radionuclides to be released is less than the effluent monitor LLD set PDAR equal to 1. Equations 1.1

1. 2 a nd 1.3 are solved for each release type and release point currently releasing or awaiting release.

relationships 1.4, 1.5, and 1.6 are satisfied, the release can be made under the assumed flow rates. If one or more of the relationships 1.4, 1.5 and 1.6 are not satisfied, action aust be taken to reduce the radionuclide release rate prior to initiating a release (or to reduce the radionuclide release rate already in progress). REVISION 11

l \\ 'N The following acticns are available to reduce the release rates at release points. the three 1) Wanta das Decay Tanks a) Release valve may be throttled b) Tank contents may be diluted c) Release may be delayed for longer decay time, 21 asaetar nulldina Pures rakaunt Duet 1 a) Dilution flow may be opened to reduce purge rate while asistalain, the same flow rate. t 31 huvilimey muildina and Fuel namA11am Area rakaumt a) Reduce inlet air supply to areas la Aus111ary hildine to toduce radioactivity source rate to vent, b) Identify and isolate the sources of radioactive releases into ths Auxiliary Building. Effluent ihmitor LLD Determination l The Technical specification LLD equatica or the relationship given below any be 1 used to calculate a moattor LLD. 4.66 N uo. Slope S = Average moaltor background count rate la cya. L Slope = Slope of monitor calibration curve in cre/pci/a1 I i r REVISDN 11 .o. a

PRE-RELEASE CALCVLATION 1.3 2' LIQUID R&DWASTE RELIASE i I. inrnanocries a Prior to initiating a release of liquid radweste, it must be determined that the concentratica of radionuclides to be released, and the flow rates at which they will be released will aot lead to a release concentration greater than the limits of specification 1.1-2 at the point of discharge. Infanskrian mancIRED Atsults of appropriate Nuclide haalysis free Section 1.2 1 l CataEATIDE l Discharge C31 CI , Co C. Cy , Ce ,g '"1 4 p ~ IWCvi IWCg IWCo itPC. IWCy IWCs IEpCre (D + El / E ~ 1 wherei i Cyt The concentration of isotope 1 in the samma spectrum excluding = t-131 and dissolved or entraised noble gases. Cg Iodine 131 concentration. = Co !Jissolved or entrained noble 9as concentration. = Cy Tritius Concentration from most recent analysis. = C = Oross alpha concentration free most recent analysis. C = St-il, 90 concestration froe nest necent analysis. s Cy, Fe-55 concentration from most recent analysis. = } t = Efflueet Strema Flow Rate i 3 0 = Dilution stream Flow Rate (leuclear services seaweter flow only) { MPC = 10CTR20 Appendia 8. Table !!, Column 2 leasinua Feraissible Con-l contration by isotope. ItPC = 25 4 uC1/a1 total activity tor all dissolved or entrataed noble gases. 3 !! the calculated Discharge concentration is less than or equal to 1, the discharge may be initiated, !! the calculated i l discharge concentration is greater than 1, action must be taken to reduce the ef fluent concentration or ef fluent stress flew rate I prior to initiating discharge. a ~"~ REVISION ll a 1 I I = - - - - -

Setpoint Calculation 1,4-1 N Reactor lu11 ding purge Exhaust Duct Monitor (RM All s" (latch Type Releas:s) i m 0DOCTIQ4 To11owing completion of 'the analyses required by Section 1.2-1 and determination of release rates and concentration limits in accordance with Section1.3-1, the monitor setpoint requires adjustment to ensure that alare and pathway isolation occur if nuclide concentration limits are exceeded. METEQD01aGY Reactor Building atmosphere is circulated through radiation noaltor.tM A6 (contain-ment atmosphere noble gas monitor) and the count rate is observed. The observed c unt 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 radt-ation and statistical counting variations, and the pressure difference between the detector chambers and exhaust vent. setpoint. The obtained value establishes the nazimum allowable to intitiating the release,The alara/ trip setpoint is adjusted to this or a more conservative value pr !! the concentration of radionuclides to be released is less than the offluent monitor LLO ' Met CPM' is obtained from the calibration curve by determining the C M which corresponds to 2.5t-2 wC1/al, cArzntAT10m I PDkk a AL {\\ 29. - V6 / (wC1/cc/CM) At +lig+3.3hg RM A1 Setpoint (CM) = E' U 'U I W 'I### 9 j where Net CM i The observed M-A6 count. r' ate, in cpa, less background, or = obtained from the calibration curve. = The vent fractions that portion of the total plant gaseous release associated with this vent and discharge type. to a number between 0 and 1. Value can be set f ractions of M.A1 and RM-A2 cannot esceed 1The summation o PORR The noble gas gassa emission Projected Dose Rate Ratto = j calculated in accordance with Section 1.3. actual projected dose rate This ratto is the j referenced in Section 1.3-1, relationsh1P 1.4. divided by the allowable 1 AL Administrative Limit = to reduce setroint to 10t of the allowable limit. AL = 10. [ V6 The actual gauge vacuun reading at RM 46 at the time of sampling. = V1 The actual or average gauge vacuus reading at RM*A1 during normal = 1 operation. ) -l (WC1/cc/CFM)A6

  • wC1/cc per cpa tor RM A6, derived from the calibration curve,This is based on an actual sample or l

\\ REVis CN 11 l

(WC1/cc/ cpm)At o WC1/cc per cpe f^r M Al. This is based cn an actual sample cr derived froe the calibration curve. Skg M-A1 background count rate in cpe. = 3.3 6 g = A statistical spread on the background count rate which represents a 19.95t contadence level on monitor counting. This factor is included to prevent inadvertent high/ trip alaras due to rances counts on the moattor. a I ] t l r t l 1 1 1 r 1 i l a 1 t e i l 1 l l l l REVISION 11 r

SETPo!NT CALCULATION 1.4 1A REACT 04 Mt!2!NG FORCE EIRAUtf 00CT NON! TOR (M-A1) (SPECIAL RELEAst,0d FUNCTIONAL TESTING OF Tst REACTOR DOILDIMC PORGE SYSTDtl l III1909CZllE t To11owing completion et the analyses required by section 1,2-1 and deteraanation of release rates and concentration limits in accordance with section 1,3 1, the monitor l i setpoint requires adjustment concentration limits are escoeded,to ensure that alara and pathway isolation occur af nuclide urTaoecleaf Auxiliary Building and fuel Hand 14ag Area atmosphere is continuously passed th! i radiation soalter M A2 and the count rate is observed. The ehserved count rate is i correlated to a corresponding count rate for M-A1, and factors are applied to accoun i backgrount radiation and statistical couating variations, and the pressure differenc i between the detector chambers and eahaust vent.The obtained value establishes the nazimum allowable setpoint. The alars/ trip setroint is ad)uated to this or a more conservative value ptior to i intitiating the release. If the concentration of radionuclides curve by deteralais, the CPM which corresponds to 2.5t 2 pC1/nl to be 5E4E34113E M.,,,et. int... --a,p29.9-V2p(wC1/cc/CPN) 1 PORA a C a,- ..,,.., s, At i Where: I Net CPM T'he observed RM A2 count rate, in cps, less background or = obtained froe the calthration curve. \\ VT = The vent traction that portion of the total plant gaseous j release associated with this vent and discharge type. set to a value fros 0 to s. VT can be fractions can not esceed 1. The sua of M-Al and M-A2 vent PDRA The noble gas games emission Project Dose Rate Ratio calculated = j in accordance with section 1.3. This ratio is the actual i l projected dose rate divided by the allowabis dose rate referenced in secties 1.3 1 relationekip 1.4. AL Administrative Limit to reduce setpoint to 10% of allowable = limit. AL = 10. ( { V2 The actual gauge vacuus, reading at M-M st the t.tse of = sampling. V1 The actual or average gauge vacuus reading at RM A1 during = A normal operation. i I l 1 I REVISKN 11 j i

(WCi/cc/ cpm)A2 o WCi/cc per eps for RM-A2. ' ~. \\ This is based en cn actual sample or derived from the enlibratica curve. (pCi/cc/ cpm)A1 = WCi/cc per cps for RM-Al. This is based on an actual sample or derived from the calibration curve. Skg RM-Al background count rate in cpe. l = 3.3 6 9 = A statistical spread on the background count rate which represents a 99.95% confidence level on monitoring counting. I This factor is inciaded to prevent inadvertent high/ tris alaras due to randos counts on the monitor. a l l ( I i i t 5 REVISION 11 J

) SITPOINT CALCULATION 1.4-IB REACTOR BOILDING PURGE EIIAOST DOCT IENt! TOR (RM-A1) ) (sPECIAL RELIASE FOLIMING IIAT OF REACTOR 90!LDING) urrmanocTias 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 should be adjusted to ensure that alara and pathway isolation occur if nuclide concentratien limits are exceeded. l MMOIM R actor Duilding atmosphere is circulated through a sampling apparatus. sample is analysed to determine the projected dose rate ratio (PORR). The Noble gas These values are combined with the monitor background, fro 3 the ca Met CPM is obtained limit to arrive at the monitor setpoint. The obtainedvent traction, and administrative allowable setpoint. value establishes the naziaua value prior to latitiating the release.The alara/ trip setpoint is adjusted to this or a acre conserva Shortly, after beginning the purge, new RM-A1 alara/ trip setpoints are determined methodology of Setpoint Calculation 1.4-2. e EMMAZIM RM-Al Setpoint (CPM) = ,'[ + 3kg + 3.3 N g where Net CPM A value derived from AM-A1 calibration curve. = VF = The vent fractions that portion of the total plant gaseous release associated with this vent and discharge type. set to a value froe o to 1 VF can be fractions can not exceed 1. The sum of RM-41 and RM-A2 vent l torr = i AL Administrative Limit to reduce setpoint to 10% of allowable = limit. AL = 10. Skg RM-At background count rate in cya. = 3.3 d kg A statiutical spread on the background count rate which = represents a 99.95% confidence level on monitoring counting. factor is included to prevent inadvertent high/ trip alaras due to This randos counts on the monitor. REVISKN ll

Setpoint Calculatien 1.4-2 N.,. Re:cter Building Purge Exhaust Duct Monitor (RM-A1) (Continuous Typs Releases) IETRQDOCTIoll i 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, monitor setpoint requires adjustment the occur if nuclide concentration limits are exceeded.to ensure that alara and pathway isolation METRODOIAGY Reactor Building atmosphere is passing through radiation monitor RM A1 during a continuous type release. Factors are applied to the observed count rate to account for background radiation and statistical counting variations. The obtained value establishes the maximum allowable setpoint. to this or a more conservative value weekly during continuous releases.The alara/ tr concentration of radionuclides to be released is less than the effluent monitor If the LLD ' Met CPM' is obtained from the calibration curve by determining the CPM which corresponds to 2.5E-2 pCi/al, cAIrnIATIce RM-A1 Setpoint (CPM) = Met CPN r VF PDRR x AL +Bkg+3.33[Bkg wheret Net CPM The observed RM-A1 count rate, in cpa, less background, = or obtained from the calibration curve. VF The vent fraction that portion of the total plant = gaseous release ass;ociated 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-At and RM-A2 cannot exceed 1. PDRR The noble gas gaasa emissAon Projected Dose Rate Ratio = calculated in accordance with Section 1.3. This ratio is the actual projected dose rate divideJ by the allowable dose rate referenced in Section 1.3-1, i relationsh'ip 1.4. i AL Administrative Limit = to reduce setpoint to 10% of the allowable limit. Admin. Limit = 10. Bkg RM-A1 background count rate in cpa. = 3.33[Bkg A statistical spread on the background count' rate which = represents a 99.95\\ confidence level on monitor counting. This factor is included to prevent inadver-tent inigh/ trip alaras due to randon counts on the son:- tor. REV S ON ll

Setpoint Calculcti:n 1.4-3 Auxiliary Building & Fu]1 Sandling Area Exhaust Monit:r (RM-A2 ) (Continuous Type Releases) Istmanecflaq Fulowing completion of 'the analyses required by Section 1.2-2 and determination of release rates and concentration limits in accordanca with Section 1.3-1, the monitor setpoint requires adjustment to assure that alars and pathway isolation occur if the nuclide concentration exceeds the determined limits, utTn000!aGY Auxiliary Building and Fuel Bandling Area atmosphere is continuously passing through radiation monitor M-AJ. Factors are applied to the observed count rate to account for background radiation and statistical counting variations. The obtained value establishew the maximum allowable setpoint. The alara/ 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 ' Met CPM' is obtained fron the calibration curve by determi'aing the cpm which corresponds to 8E-3 pC1/s1. cAunIATItst, RM-A2 Setpoints (CPM) Met CMI x VF + Skg + 3.3 h g = PDAR x AL where: Met CPM = The observed RM-A2 count rate, in cpa, 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-A1 and RM-A2 cannot exceed 1. PDRR i = 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. AL = Administrative Limit to reduce setpoint to 10\\ of the allowable limit. AL = 10. Skg = RM-A2 background count rate in cps. 3.3 h g = A statistical spread on the background count rate which represents a 99.95\\ confidence level on monitor counting. This fa: tor is included to prevent inadver-tent high/ trip alaras due to randos counts on the moni-tot. i I REVIS,0N 11 \\ m

I Em point Calculation 1.4-4 Weste Gas Decay Taak Monit:r (RM-A11) 1 (Batch Type A31eas:s) t ) t urrmanocrias i Following completion of the analyses required by Section 1.2-3 and deternnation of release rates and concentration limits in accordance with Section 1.3-1, the nonitor setpoint requires adjustment to assure that alara and pathway isolation occur if nuclide concentration limits are exceeded. IETEBOMGY Prior to initiating a Waste Gas Decay Tank release, its contents are drawn through radiation monitor RM-A11 and returned to the waste gas header. Factors are applied to the obeerved count rate to account for background radiation and statistical counting variations. The obtained value establishes the nazimum allowable setpoint. The alars/ trip setpoint is adjusted to this or a more conservative value prior to intitiatlag the release. If the concentration of radionuclides to be released is less than the affluent monitor LLD ' Wet CFM" is obtain calibration curve by determining the CPM which corresponds to 20 pC1/al. EMMAGE I RM-A11 Setpoint (CPM) = not cset u vy a to + Skg + 3.3 k g FORA E AL x F where: Net CPM i The observed RN-111 count rate, in cya, less background, or = obtained from the calibration curve. V7 The vent fraction; that portion of the total plant gaseous = release associated with this vent and discharge type. Value is equal to 0.5, PDRA The noble gas ganan 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. AL Administrative Limit to reduce setpoint to 10t of = allowable limit. AL = 10. 10 The naziaun pressure (psig) which RM-A11 detector chamber = should be subjected to. from the release line to the vent.This corresponds to a tiow of 15 CFM P Pressure (psig) in RM-111 at time of obtairling not CPM. = Skg PM-A11 background count rate in cya. =

3. 3Ng

= A statistical spread on the background count rate which represents a 99.95t confidence level on monitor counting. This tactor is included to prevent inadvertent high/ trip alaras due to randos counts on the monitor. ~ - REVIS ON 11 t

Setpoint Calculatica 1.4-5 4 Plant Discharg3 Line Monitor (M-L2) (Batch Type Roleases) i IETRODDCTICE l Following completion of the analyses required by Section 1.2-4 and determina-tion of release rates and concentration limits in accordance with section 1.3-2, the monitor setpoint requires adjustaent to ensure that alara and pathway isola-tion occur if nuclide concentration limits are exceeded, inrnoDotacY Evaporator Condensate Storage Tank or Laundry and Shower Susp Tank contents are circulated through radiation monitor M-L2 and returned to the auxiliary building sump to obtain the actual count rate at RM-L2 for the concentration contained in the tank for release. The observed count rate is adjusted for re-l lease flow, background and statistical counting variations, particular to this release flow path. The resulting value is used as the alara/ trip setpoint and M-L2 is adjusted to this or a more conservative value prior to initiating the release. If the concentration of radionuclides to be reinased is less than the effluent monitor LLD set 'ECi/MPCi' equal to 1 and determine ' Net CPM' fLos the calibration curve by locating the CPM which corresponds to 3E-7 pci/al. GIGI&ME. M-L2 setpoint (CPM) = Met c1PN z AF z (E + D) + Skg + 3.3 g (E C /MPC1) x E i where: i Net CPM The observed .".M-L2 count rate, in cpe, less back- = ground, or obtained from the calibration curve. AF Administration Factor to account for error in set- = point determination. AF = 0.8. 1 EC /MPci 1 = The ratio of the actual gamma emitting concentra-tions (excluding dissolved and entrained gases) of ,the tank contents to be released to the Marious Per-missible concentration (NPC) as listed in 10 CTR 20 Table II, Column 2 for unrestricted areas. E The release flow rate of waste to be discharged in = gallons per minute. A ma':inum flow rate of 100 gpa will be used for the Evaporator Condensate Storage Tanks and 40 gpa for the Laundry and Shower Sump Tanks. D The dilution flow from the Nuclear Services Sea = water systen in gallons per ainute. Bkg M-L2 background count rate in eps. = 3.3 k g = A statistical spread on the background count rate which represents a 99.95% confidence level on monitor counting. This factor is included to prevent inadver-tent high/ trip alares due to randon counts on the sonator. REVISION 11 i s a

~~. Setpoint Calculation 1.4-6 Turbine Building Basement Discharge Line leonitor (M-L7) (continuous Type Releases) HIEEEEEE. Tbe activity released through the Turbine Building Basement Discharge Line Itonitor M-L7 is analysed in accordance with section 1.2-5. The setpoint is a fixed concentration based on worst case nuclide released at the worst case rate as described in the Methodology Section below. The monitor setpoint is adjusted to ensure isolation of the release pathway if nuclide concentration limits are escoeded. m The alara/ trip setpoint determination is based on the worst case assumption that I-131 is the only nuclide being discharged. This assumption equates all counts on M-L7 to I-131 with an Mpc of 3 g 10-7 uci/al. I-131 has the most conservative Mpc of the nuclides available to this release path and ' visible

  • to M-L7.

The setpoint is based on assuring 1 MpC or less of I-131 in the i discharge canal and is determined by obtaining the eps on M-L7 calibration curve which corresponds to a concentration of 3 a 10-7 uci/mi and applying the flow dilution factor, background counts, and statistical counting variations. The resulting value is used as t.he alars/ trip setpoint and M-L7 is adjusted to this or a more conservative value to maintain control on release conditions. t catarIATIam CPN r (E + D) M-L7 SetPoint (CPM) = E + Skg + 3.3 h g where: CPM The counts per minute corresponding to 3 x 10-7 uci/a1 = (1 Mpc I-131) from the current M-L7 calibration curve. E = The maximum release flow rate of water able to be discharged in gallons per minute. D = The dilution flow from the Nuclear Services Sea Water system in gallons per minute. Skg The background count rrte at M-L7 in cpa. = 3.3TBkg= A statistical spread on the background count rate which represents a 99.95t confidence level on.nonitor counting. This factor is included to prevent inadvertent high/ trip alaras due to randos counts on the monitor. REVISION 11 1 ^

S:tpoint Calculatica 1.4-7 Turbine Building Basea:nt Discharge Lin? Moniter (RM-L7) (Batch Type Relets:s) IET10DOCTIQE To11owing completion of the analyses required by Section 1.2-4 and determination of release rates and concentration limits in accordance with Section 1.3-2, the monitor setpoint requires adjustment to ensure that alara and pathway isolation occur if nuclide concentration limits are exceeded. METEDDOIAGY Station Drain Tank (SDT-1) RM-L7 and returned to the sumpcontents are circulated through radiation monitor to obtain the actual count rate at RM-L7 for the concertration contained in the tank for release. The observed count rate is adjusted for release flow, background and statistical counting variations, particular to this release flow path. The resulting value is uscd as the alara/ trip setpoint and RM-L7 is adjusted to this or a acre conservative value prior to initiating the release. released is less than the effluent monitor LLD setIf the concentration of 'ECi/MPCi' equal to 1 ar4d determine ' Net CPM' from the calibration curve by locating the CPM which corresponds to 3E-7 pCi/al, cAImILaffas RM-L7 Setpoint (CPM) = +Skg+3.3hg (E C /MPC ) x E i i where: Net CPM The observed RM-L7 count rate, in cpa, less = background. AF Administration Factor to account for error in = setpoint determination. AF = 0.8. EC /MPCi i The ratio of the actual gaasa esitting concen- = trations (excluding dissolved and entrained gases) of the tank contents to be released to'the Maximum Permissible Concentration (MPC) as listed in 10 CTR 20, Table II, Column 2 for unrestricted areas. E The release flow rate of waste to be discharged in = gallons per minute. will be used. A saximum flow rate of 600 gpa D The dilution flow from the Nuclear Services Sea = Water system in gallons per minute. Bkg { RM-L7 background count rate in cps. l =

3. 3 Ng

= A statistical spread on the backgound count rate which represents a 99.95t confidence level on monitor counting. This factor is included to prevent inadvertent high/ trip alaras due to randon i counts on the monitor. REVISION ll. n l

l NOCLIDE ANALYSIS 4.2-1 M BOILDING FORGE IIEl40ST l O NOCLIDE SAMPLE ~00RCE l LJ(b)(uCi/a1) A. Principal Game' Esitters (4) Mn-54 Fe-59 1x10-4/1x10-11 1x10-4/1x10-11 Co-58 Batch release particulate filter 1xt0-4/1x10-11 Co-60 for Batch Releases. Weekly 1x10-4/1x10-11 In-65 Particulate Filter Analysis for 1x10-4/1x10-11 No-99 continuous (c) type release. 1x10-4/1x10-11 Cs-134 Cs-137 1x10-4/1x10-11 Ce-141 1x10-4/1x10-11 Co-144 1x10-4/1x10-11 1x10-4/1x10-11 Kr-87 Pre-release grab sample for Batch 1x10-4 Kr-48 type release. Weekly grab sample 4x10-4 Xe-133 for continuous type release. 1x10*4 Xe-133a Xe-135 1x10-4 Xe-138 1x10-4 1x10-4 9. Zodine 131 Batch release charcoal filter for NAf1x10-12 tatch Releases. Weekly charcoal filter for continuous releases. C. Tritium Pre-release Grab Sample. 1x10-6 D. Gross Alpha Monthly Particulate Filter Composite 1x10-11 E. St-89 Quarterly Particulate Filter Composite 1x10-11 F. Sr-90 Quarterly Particulate Filter Composite 1x10-11 l l ta) other identified Gamma Enitters not listed in this table shall be includ dose calculations. (b) The first value refers to the LLD for pre-release grab samples the second value refers to the LLD for weekly Particulate Filter Analysis. (c) Reactor Building Purge is considered continuous after ainlaus of one Reactor volume is a batch type). Building volumes have been released on a continuo REVISICN 11 48

NUCLIDE ANALYSIS 4.2-2 AUXILIARY BUILDING AND FUEL BANDLING AREA EXEADST NUCLIDE SAMPLE SOURCE LLD(b)( UC1/al) A. Principal Gamaa Esitters (a) Mn-54 Fe-59 1x10-4/1x10-11 1x10-4/1x10-11 Co-58 weekly Particulate Filter Analysis. 1x10-4/1x10-11 Co-60 1x10-4/1x10-11 In-65 1x10-4/1x10-11 Mo-99 Cs-134 1x10-4/1x10-11 Cs-137 1x10-4/1x10-11 Ce-141 1x10-4/1x10-11 Ce-144, 1x10-4/1x10-11 1x10-4/1x10-11 = Kr-87 Monthly Grab Sample. 1x10-4 i Kr-88 1x10-1 Xe-133 1x10-4 Xe-133a Xe-135 1x10-4 Xe-138 1x10-4 1x10-4 B. Iodine 131 weekly Charcoal Filter Analysis. 1x10-12 i C. Tritium Monthly Grab Sample. 1x10-6 D. Gross Alpha Monthly Particulate Filter Composite 1x10-11 E. Sr-89 Quarterly Particulate Filter Composite 1x10-11 F. Sr-90 Quarterly Particulate Filter Composite 1x10-11 { (a) Other identified Canaa Esitters not listed in this table shall be includ in dose calculations. (b) The first value refers to the LLD for pre-release grab sample; the second value refers to the LLD for weekly Particulate Filter Analysis. REV S10h 11 47 b ^

N NUCLIDE ANALYSIS 4.3-3 CASTE GAS DECAY TANKS NUCLIDE SAfrLE SOURCE Lu)(b)( uti/al) A. Principal Gamma F.altters (a) Mn-54, Fe-59, talo /1x10-11 Co-58 1:10 /1x10-11 Co-60 1:10 /1x10-11 1310-4/1x10-11 In-65 1x10-4/1x10-11 Mo-99 Weekly Particulate Filter sample (from RN-A2) 1x10-4/1x10-11 Cs-134 Cs-137 tato-4/1xto-11 Ce-141 1xto-4/tx10-11 1x10-4/1x10-11 ~ Ce-144 Kr-87 1x10-4 Kr-88 1x10-4 1x10-1 Xe-133 Pre-release Grab sample 1x10-4 Xe-133a Xe-135 1x10-4 Xe-138 1xto-4 1 10-4 ~ B. Iodine 131 Weekly Charcoal Filtar (from RM-12) 1x10-12 (a) other identified Gamma taitters not listed in this table shall be included dose and setpoint calculations. (b) The first value refers to the LLD for pre-release grab samples the second } value refers to the LLD for weekly Particulate Filter Analysis, i REVISIC \\1 11 4,. i ),

i 3 Te.ble 5.1-1 Enviroampatal Radiological flonitoring Stations Locations s?ATTas TMATTam _ DIRECTICII DISTANCE P% P =F Ps P "T fmi ) C04 State Park Old Can on River EME near road intersection 6.3 C07 Crystal River Public Water Plant Est 7.5 C09 Fort Island Gulf Beach 5 3.2 C10 Indian Waters Public Water supply Est 5.9 I C13 Mouth of Intake Canal WSW 3.4 C14H Nead of Discharge Canal NW O.1 C14M Midpoint of Discharge Canal W 1.2 Discharte Canal at Gulf of Mco C140 W 2.8 C18 Yankeetown City Well N 5.2 CIS NW Corner state' Roads 448 6 495 ENE 8.5 C29 Discharfe Arte N 2.0 C30 Intake Area WSW 3.6 C40 Near N.E. Site Boundary E near escavated pond & pump station 3.5 C41 Onsite meteorological tower SW 0.4 C46 North Pump Station E 0.4 C47 University of Florida, Gainesville NME $2 C48A Casite North of CR 4 & F ) M 0.8 C488 Onsite IWit of CR 4 & 5 NME 0.8 i 1 ) i REVISION 11 -.0 - ano' M n

TABLE 5.1-3 RING TLDs (5 NILE RING) IncanM DIRECTION p Hrfurr fmi,) C18 N 5.2 C03 NME 5.3 C04 NE 6.3 C74 ENE 5.5 C75 E 4.2 C76 ESE 5.4 C08 SE 3.5 C77 SSE 3.2 C09 S 3.2 C78 Vr3 4.1 C14G W 2.8 C01 NW 4.9 C79 NNW 5.0 l REV Sl0N 11 - a2 - W Y e

em3TR93 g ;m c%gqc#yY 4 94a 8 E %%b % 10 lid.YOid ;nhy kiiuEAR OPHi!Js0NS CRYST!.L RIVER UNIT 3 PROCESS CONTR0T. PROGRAM f ^" 2 Approved by: _ dfies/ Rad upt. Date: S f \\ Approved by: OA. ) Manager,SiteNucparServices i Date:, Al5IST )

t TABLE OF CONTENTS

1.0 INTRODUCTION

PAGE 1

2.0 REFERENCES

2 3.0 REGULATORY REQUIREMENTS. 3 4.0. ADMINISTRATIVF CONTROLS. 5 4.1 Responsibility / Revisions.. 5 4.2 Reporting. 5 4.3 Documentation. s 5 4.4 Definitions 5 5.0 SOLIDITICAT!?N........................... 6 5.1 System Description.. .6 5.2 Process Description. 6 5.3 Process Control Program Testing. 8 5.3.1 Process Testing. 8 5.3.2 Test Frequency. 9 5.3.3 Acceptance criteria. 9 5.3.4 Corrective Action.......... 9 5.3.5 Reportiny. 10 6.0 DEWATERING.. 6.1 Systen Description...................... 10 10 6.2 Process Description......... ............ 11 6.3 Dewatering Program Testing. 12 6.3.1 Sample Testing and Frequency. 13 6.3.2 Acceptance Criteria and Corrective Action.. 13 6.3.3 Reporting. . 1J 7.0 SHIPPING. 13 APPENDIXES I. Drawings Figure I-I: Solidificatic.1 System.. 14 Figure I-II: Secondary Dewatering System. . 15 Figure I-III: Primary Dewatering Systen. 15 REVISION 02.

l

2.0 REFERENCES

i 2.1. NUREG 0472, ' Radiological Effluent Technical Specifications for PWR's' 2,2 WP

101, "Packaging,
Storing, and Shipping of Radioactive Materials
  • 2.3 WP 102,
  • Radioactive Shipments Certificates of Compliance' 2.4 WP 201, "Processing with the Nuclear Waste Demineralizer System" 2.5 WP 202, ' Liner Dewatering Leg Construction" 2.6 WP 301, ' Radioactive Waste Solidification" 2.7 OP 413, "Waste Drumming Systen 2.8 OP 6018,
  • Secondary Resin Liner Dewatering' 2.9 SP 743, "Solidification Test Batch Verificat'on Progras' 2.10 Bartlett PCP, Report SA5889 2.11 CNSI Bead Resin Dewatering Procedure 2.12 10 CFR 61, "Licensing Requirements for Land Disposal of Radioactive Waste
  • 2.13 Crystal River Unit 3 Technical Specifications 2.14 South Carolina Department of Health and Environmental Control, Radioactive Materials License 4097, Amendment #34 REV SLN O2

t ' $.3.2 Test Frequ:ncy 5.3.2.1 Process Test Frequency A process test solidification shall be made prior to full scale solidification to determine ratios and additives as per Section 5.3.1.1. 5.3.2.2 Solidification Test Frequency The pCP shall be used to verify the solidification of at least one (1) repre-sentative test specimen from at least every tenth batch of each type of wet radioactive waste. 5.3.3 Acceptance criteria Sp 743, ' Solidification Test Batch Verification Program,' stipulates the activities and documentation necessary to verify acceptance of solidified waste. The solidified waste acceptance criteria is verified by: Visually inspecting for defects in the structure, a. b. Uniformity in color and density, No free-standing liquid (<0.5% of total waste volume). c. d. Free-standing monolith, e. Aftst 24 hours from solidification, the final cured product shall resist penetration when probed by hand with a spatula or firm object (>50 psi). If any portion of the specimen fails to pass the Acceptance criteria, the applicable actions of Section 5.3.4 aust be net. 5.3.4 Corrective Action If the initial test specimen from a batch of waste fails to verify solidi-a. fication, representative test specimens from each consecutive batch of the same type of wet waste shall be collected and tested until at least 3 consecutive initial test specimens demonstrate solidification. The process and/or additives shall be modified as required, as provided in Section 4.1, to assure solidification of subsequent batches of waste,

b. If any test specimen fails to verify solidification, the solidification of the batch under test shall be suspended until such time as additional test specimens can be obtained, alternate solidification parameters can be deterniaed in accordance with the Process Control Program, and a subsequent test verifies solidification.

Solidification of the batch may thtn be resumed using the alternative solidificStion parameters determined by the Process Control program,

c. With installed equipment incapable of meeting Technical Specification 3.7.13.4 or declared inoperable, restore the equipment to operate status or provide for contract capability to process wastes as necessary to satisfy all appl.icable transportation and disposal requirements.

l REV'SK1 02

o APPENDIX I e 1 Oils. Sludge, C Acids, etc. 1 - Solidification Dew i CONCENTM TED MSTE .CONCDITNATED BORIC STOME TMES ACID TMES teT-7A leT-73 LJ LJ LJ <J T+- -e O- -t I FIWRE l-1: SOLIDIFICATION SYSTDI i 1 u. l REVIS C N 02

~ o Resin ~1 Air / Water l Separator j o 47 s Floor Drsin FIENIE I-11: SECONDARY RE5!N DORTDl!M SYSTDI Assin to Vent Duct I T T Afr/ Water g Sluir.eable S*parater Dominere11rer l Needer i Spent Resin \\ Tank Outlet T l Floor Drain j FIENIE I-111: PRIl4UtY RESIN DOMTERIM SYSTDI i . Is.

J s .Oee.- o O. o' Power COR PO R AfIO N August 26, 1988 3F0888-16 M= ant Control Desk U.S. Nuclear Regulatory Cr.mnission Washington, DC 20555 SUBIECT: Crystal River Unit 3 Docket No. 50-302 Operating License No. DPR-72 Semianrmal Radioactive Effluent Release Report

Dear Sir:

Pursuant to Title 10, Code of Federal Regulations, Part 50.36(a)(2) and Crystal River Unit 3 hinical Specification 6.9.1.5(d), Florida Power Corporation hereby subnits the Crystal River Unit 3 Saniannual Radioactive Effluent Release Report for the period January 1, 1988 through June 30, 1988. If you have any questions concerning this matter please contact this office. Sincerely, Rolf b. Widell, Director i Nuclear Operations Site Stg. port RCW DUI xc: Regional Adrainistrator, Region II Senior Resident Inspector t Post Office Box 219

  • Crystal River, Florida 32029
  • Telephone (904) 7954802 l

g A Florida Progress Company}}

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